National Library of Energy BETA

Sample records for non-hlw wds llw

  1. LLW Forum meeting report

    SciTech Connect (OSTI)

    NONE

    1996-08-01

    This report summarizes the Low-Level Radioactive Waste Forum (LLW Forum) meeting on May 29 through May 31, 1996.The LLW Forum is an association of state and compact representatives, appointed by governors and compact commissions, established to facilitate state and compact implementation of the Low-Level Waste Policy Act of 1980 and the Low-Level Radioactive Waste Policy Amendments Act of 1985 and to promote the objectives of low-level radioactive waste regional compacts. The LLW forum provides an opportunity for state and compact officials to share information with one another and to exchange views with officials of federal agencies and other interested parties.

  2. Quantitative WDS analysis using electron probe microanalyzer

    SciTech Connect (OSTI)

    Ul-Hamid, Anwar [Research Institute, King Fahd University of Petroleum and Minerals, P.O. Box 1073, Dhahran 31261 (Saudi Arabia)]. E-mail: anwar@kfupm.edu.sa; Tawancy, Hani M. [Research Institute, King Fahd University of Petroleum and Minerals, P.O. Box 1073, Dhahran 31261 (Saudi Arabia); Mohammed, Abdul-Rashid I. [Research Institute, King Fahd University of Petroleum and Minerals, P.O. Box 1073, Dhahran 31261 (Saudi Arabia); Al-Jaroudi, Said S. [Saudi Aramco, P.O. Box 65, Tanajib 31311 (Saudi Arabia); Abbas, Nureddin M. [Research Institute, King Fahd University of Petroleum and Minerals, P.O. Box 1073, Dhahran 31261 (Saudi Arabia)

    2006-04-15

    In this paper, the procedure for conducting quantitative elemental analysis by ZAF correction method using wavelength dispersive X-ray spectroscopy (WDS) in an electron probe microanalyzer (EPMA) is elaborated. Analysis of a thermal barrier coating (TBC) system formed on a Ni-based single crystal superalloy is presented as an example to illustrate the analysis of samples consisting of a large number of major and minor elements. The analysis was performed by known standards and measured peak-to-background intensity ratios. The procedure for using separate set of acquisition conditions for major and minor element analysis is explained and its importance is stressed.

  3. Dusty WDs in the WISE all sky survey ? SDSS

    SciTech Connect (OSTI)

    Barber, Sara D.; Kilic, Mukremin; Gianninas, A. [Homer L. Dodge Department of Physics and Astronomy, University of Oklahoma, 440 W. Brooks St., Norman, OK 73019 (United States); Brown, Warren R., E-mail: barber@nhn.ou.edu [Smithsonian Astrophysical Observatory, 60 Garden Street, Cambridge, MA 02138 (United States)

    2014-05-10

    A recent cross-correlation between the Sloan Digital Sky Survey (SDSS) Data Release 7 White Dwarf Catalog with the Wide-Field Infrared Survey Explorer (WISE) all-sky photometry at 3.4, 4.6, 12, and 22 ?m performed by Debes et al. resulted in the discovery of 52 candidate dusty white dwarfs (WDs). However, the 6'' WISE beam allows for the possibility that many of the excesses exhibited by these WDs may be due to contamination from a nearby source. We present MMT+SAO Wide-Field InfraRed Camera J- and H-band imaging observations (0.''5-1.''5 point spread function) of 16 of these candidate dusty WDs and confirm that four have spectral energy distributions (SEDs) consistent with a dusty disk and are not accompanied by a nearby source contaminant. The remaining 12 WDs have contaminated WISE photometry and SEDs inconsistent with a dusty disk when the contaminating sources are not included in the photometry measurements. We find the frequency of disks around single WDs in the WISE ? SDSS sample to be 2.6%-4.1%. One of the four new dusty WDs has a mass of 1.04 M {sub ?} (progenitor mass 5.4 M {sub ?}) and its discovery offers the first confirmation that massive WDs (and their massive progenitor stars) host planetary systems.

  4. Waste simulant development for evaluation of LLW melter system technology

    SciTech Connect (OSTI)

    Shade, J.W.

    1994-05-25

    This document describes the LLW simulant compositions, basis for the simulants, and recipes for preparing nonradioactive simulants for LLW melter tests.

  5. LLW Notes, Volume 12, Number 3

    SciTech Connect (OSTI)

    Norris, C.; Brown, H. [eds.; Colsant, J.; Lovinger, T.; Scheele, L.; Shaker, M.A.

    1997-03-01

    Contents include articles entitled: California DHS sues US Interior Department to compel land transfer; LLW Forum holds winter meeting; LLW Forum waste information working group meets; LLW Forum regulatory issues discussion group meets; Envirocare investigation transferred to feds; Host state TCC meets in Laughlin, Nevada; BLM to require new permit for California site testing; Federal agencies and committees; Pena sworn in as Energy Secretary, Grumbly departs DOE; U.S. Supreme Court tackles property rights issues; GAO to study DOI`s actions; Congress scrutinizes FY `98 budget requests; and Senate committee passes high-level waste bill: Clinton threatens to veto.

  6. LLW notes, Vol. 11, No. 2

    SciTech Connect (OSTI)

    1996-03-01

    `LLW Notes` is distributed by Afton Associates, Inc. to Low-Level Radioactive Waste Forum Participants and other state, and compact officials identified by those Participants to receive LLW Notes. The Low-Level Radioactive Waste Forum (LLW Forum) is an association of state and compact representatives, appointed by governors and compact commissions, established to facilitate state and compact implementation of the Low-Level Radioactive Waste Policy Act of 1980 and the Low-Level Radioactive Waste Policy amendments Act of 1985 and to promote the objectives of low-level radioactive waste regional compacts. The LLW Forum provides an opportunity for state and compact officials to share information with one another and to exchange views with officials of federal agencies and other interested parties.

  7. LLW notes. Volume 11, No.8

    SciTech Connect (OSTI)

    NONE

    1996-12-01

    `LLW Notes` is distributed by Afton Associates, Inc. to Low-Level Radioactive Waste Forum Participants and other state, and compact officials identified by those Participants to receive `LLW Notes`. The Low-Level Radioactive Waste Forum (LLW Forum) is an association of state and compact representatives, appointed by governors and compact commissions, established to facilitate state and compact implementation of the Low-Level Radioactive Waste Policy Act of 1980 and the Low-Level Radioactive Waste Policy Amendments Act of 1985 and to promote the objectives of low-level radioactive waste regional compacts. The LLW Forum provides an opportunity for state and compact officials to share information with one another and to exchange views with officials of federal agencies and other interested parties.

  8. He-Accreting WDs: AM CVn stars with WD Donors

    E-Print Network [OSTI]

    Piersanti, Luciano; Tornambe', Amedeo

    2015-01-01

    We study the physical and evolutionary properties of the "WD family" of AM CVn stars by computing realistic models of IDD systems. We evaluate self-consistently both the mass transfer rate from the donor, as determined by GW emission and interaction with the binary companion, and the thermal response of the accretor to mass deposition. We find that, after the onset of mass transfer, all the considered systems undergo a strong non-dynamical He-flash. However, due to the compactness of these systems, the expanding accretors fill their Roche lobe very soon, thus preventing the efficient heating of the external layers of the accreted CO WDs. Moreover, due to the loss of matter from the systems, the orbital separations enlarge and mass transfer comes to a halt. The further evolution depends on the value of \\mdot\\, after the donors fill again their lobe. On one hand, if the accretion rate, as determined by the actual value of (M_don,M_acc), is high enough, the accretors experience several He-flashes of decreasing s...

  9. WDS-Based Layer 2 Routing for Wireless Mesh Networks [Extended Abstract

    E-Print Network [OSTI]

    Chuah, Chen-Nee

    - struct flow-based MAC tables at each wireless router. For an incoming packet, the router will matchWDS-Based Layer 2 Routing for Wireless Mesh Networks [Extended Abstract] Dhruv Gupta, Jason Le-communication Networks]: Network Architecture and Design--wireless communication General Terms design

  10. LLW Notes, vol. 9, no. 1. February/March 1994

    SciTech Connect (OSTI)

    NONE

    1994-03-01

    LLW Notes is published ten times each year and is distributed to Low- Level Radioactive Waste Forum Participants and other state and compact officials identified by those Participants to receive LLW Notes. The Low-Level Radioactive Waste Forum is an association of representatives of states and compacts established to facilitate state and compact commission implementation of the Low-Level Radioactive Waste Policy Amendments Act of 1985 and to promote the objectives of low-level radioactive waste regional compacts. The Forum provides an opportunity for states and compacts to share information with one another and to exchange views with officials of federal agencies.

  11. WRAP low level waste (LLW) glovebox acceptance test report

    SciTech Connect (OSTI)

    Leist, K.J.

    1998-02-17

    In June 28, 1997, the Low Level Waste (LLW) glovebox was tested using glovebox acceptance test procedure 13031A-85. The primary focus of the glovebox acceptance test was to examine control system interlocks, display menus, alarms, and operator messages. Limited mechanical testing involving the drum ports, hoists, drum lifter, compacted drum lifter, drum tipper, transfer car, conveyors, lidder/delidder device and the supercompactor were also conducted. As of November 24, 1997, 2 of the 131 test exceptions that affect the LLW glovebox remain open. These items will be tracked and closed via the WRAP Master Test Exception Database. As part of Test Exception resolution/closure the responsible individual closing the Test Exception performs a retest of the affected item(s) to ensure the identified deficiency is corrected, and, or to test items not previously available to support testing. Test Exceptions are provided as appendices to this report.

  12. LLW Forum meeting report, May 7--9, 1997

    SciTech Connect (OSTI)

    Norris, C.; Brown, H.; Lovinger, T.; Scheele, L.; Shaker, M.A.

    1997-12-31

    The Low-Level Radioactive Waste Forum met in Chicago, Illinois, on may 7--9, 1997. Twenty-three Forum Participants, Alternate Forum Participants, and meeting designees representing 20 compacts and states participated. A report on the meeting is given under the following subtitles: New developments in states and compacts; Upgrading an existing disposal facility; Revisions to DOE Order 5820 re DOE waste management; Conference of radiation control program directors: Recent and upcoming activities; National Conference of State Legislatures` (NCSL) low-level radioactive waste working group: Recent and upcoming activities; Executive session; LLW forum business session; Public involvement and risk communication: Success at West Valley, New York; DOE low-level waste management program; impact of the International Atomic Energy Agency`s convention on waste; Panel discussion: The environmental justice concept--Past, present and future; New technologies for processing and disposal of LLRW; High-level and low-level radioactive waste: A dialogue on parallels and intersections; Draft agreement re uniform application of manifesting procedures; Regulatory issues focus; LLW forum October 1997 agenda planning; Resolutions; LLW forum regulatory issues discussion group meets; and Attendance.

  13. LLW Forum meeting report, October 20--22, 1997

    SciTech Connect (OSTI)

    Norris, C.; Brown, H.; Lovinger, T.; Scheele, L.; Shaker, M.A.

    1997-12-31

    The Low-Level Radioactive Waste Forum met in Annapolis, Maryland, on October 20--22, 1997. Twenty-six Forum Participants, Alternate Forum Participants, and meeting designees representing 22 compacts and states participated. A report on the meeting is given under the following subtitles: New developments in states and compacts; Discussion with NRC Commissioner McGaffigan; Regulatory issues session; Executive session; LLW forum business session; DOE low-level waste management program; Transportation of radioactive waste; Environmental equity: Title VI; Congressional studies on Ward Valley Site; Implementation of DOE`s strategy for waste management; Relicensing Envirocare; Draft agreement for uniform application of manifesting procedures; CRCPD report; Panel: Future of low-level radioactive waste management; Agenda planning: February 1998; Resolutions; and Attendance.

  14. LLW Forum meeting report, April 18--19, 1991

    SciTech Connect (OSTI)

    NONE

    1991-12-31

    The Low-Level Radioactive Waste Forum is an association of representatives of states and compacts established to facilitate state and compact commission implementation of the Low-Level Radioactive Waste Policy Act of 1980 and the Low-Level Radioactive Waste Policy Amendments Act of 1985 and to promote the objectives of low-level radioactive waste regional compacts. The Forum provides an opportunity for states and compacts to share information with one another and to exchange views with officials of federal agencies. LLW Forum participants include representatives from regional compacts, designated host states, unaffiliated states, and states with currently- operating low-level radioactive waste facilities. This quarterly meeting was held on April 18-19, 1991.

  15. LLW Forum meeting report, April 25--27, 1994

    SciTech Connect (OSTI)

    NONE

    1994-12-31

    The Low-Level radioactive Waste Forum is an association of representatives of states and compacts established to facilitate state and compact commission implementation of the Low-Level Radioactive Waste Policy Act of 1980 and the Low-Level Radioactive Waste Policy Amendments Act of 1985 and to promote the objectives of low-level radioactive waste regional compacts. The Forum provides an opportunity for states and compacts to share information with one another and to exchange views with officials of federal agencies. LLW Forum participants include representatives from regional compacts, designated host states, unaffiliated states, and states with currently-operating low-level radioactive waste facilities. This quarterly meeting was held April 25-27, 1994 and activities during the first quarter of 1994 are detailed..

  16. LLW (Low-Level Waste) Notes, Volume 13, Number 1, February 1998

    SciTech Connect (OSTI)

    NONE

    1998-02-01

    LLW Notes is a newsletter distributed to Low-Level Radioactive Waste Forum Participants and other state and compact officials. The LLW Forum provides an opportunity for state and compact officials to share information with one another and to exchange views with officials of federal agencies and other interested parties. This issue focuses on the following topics: DOI approves Ward Valley permit application; Project evidentiary hearings begin in Texas; and Summary judgment motions in California breach of contract action.

  17. Preliminary low-level waste feed definition guidance - LLW pretreatment interface

    SciTech Connect (OSTI)

    Shade, J.W.; Connor, J.M.; Hendrickson, D.W.; Powell, W.J.; Watrous, R.A.

    1995-02-01

    The document describes limits for key constituents in the LLW feed, and the bases for these limits. The potential variability in the stream is then estimated and compared to the limits. Approaches for accomodating uncertainty in feed inventory, processing strategies, and process design (melter and disposal system) are discussed. Finally, regulatory constraints are briefly addressed.

  18. Greater-than-Class C low-level radioactive waste characterization. Appendix E-2: Mixed GTCC LLW assessment

    SciTech Connect (OSTI)

    Kirner, N.P. [Ebasco Environmental, Idaho Falls, ID (United States)

    1994-09-01

    Mixed greater-than-Class C low-level radioactive waste (mixed GTCC LLW) is waste that combines two characteristics: it is radioactive, and it is hazardous. This report uses information compiled from Greater-Than-Class C Low-Level Radioactive Waste Characterization: Estimated Volumes, Radionuclide Activities, and Other Characteristics (DOE/LLW 1 14, Revision 1), and applies it to the question of how much and what types of mixed GTCC LLW are generated and are likely to require disposal in facilities jointly regulated by the DOE and the NRC. The report describes how to classify a RCRA hazardous waste, and then applies that classification process to the 41 GTCC LLW waste types identified in the DOE/LLW-114 (Revision 1). Of the 41 GTCC LLW categories identified, only six were identified in this study as potentially requiring regulation as hazardous waste under RCRA. These wastes can be combined into the following three groups: fuel-in decontamination resins, organic liquids, and process waste consisting of lead scrap/shielding from a sealed source manufacturer. For the base case, no mixed GTCC LLW is expected from nuclear utilities or sealed source licensees, whereas only 177 ml of mixed GTCC LLW are expected to be produced by other generators through the year 2035. This relatively small volume represents approximately 40% of the base case estimate for GTCC wastes from other generators. For these other generators, volume estimates for mixed GTCC LLW ranged from less than 1 m{sup 3} to 187 m{sup 3}, depending on assumptions and treatments applied to the wastes.

  19. Test Plan: Phase 1, Hanford LLW melter tests, GTS Duratek, Inc.

    SciTech Connect (OSTI)

    Eaton, W.C.

    1995-06-14

    This document provides a test plan for the conduct of vitrification testing by a vendor in support of the Hanford Tank Waste Remediation System (TWRS) Low-Level Waste (LLW) Vitrification Program. The vendor providing this test plan and conducting the work detailed within it [one of seven selected for glass melter testing under Purchase Order MMI-SVV-384215] is GTS Duratek, Inc., Columbia, Maryland. The GTS Duratek project manager for this work is J. Ruller. This test plan is for Phase I activities described in the above Purchase Order. Test conduct includes melting of glass with Hanford LLW Double-Shell Slurry Feed waste simulant in a DuraMelter{trademark} vitrification system.

  20. WRAP low level waste restricted waste management (LLW RWM) glovebox acceptance test report

    SciTech Connect (OSTI)

    Leist, K.J.

    1997-11-24

    On April 22, 1997, the Low Level Waste Restricted Waste Management (LLW RWM) glovebox was tested using acceptance test procedure 13027A-87. Mr. Robert L. Warmenhoven served as test director, Mr. Kendrick Leist acted as test operator and test witness, and Michael Lane provided miscellaneous software support. The primary focus of the glovebox acceptance test was to examine glovebox control system interlocks, operator Interface Unit (OIU) menus, alarms, and messages. Basic drum port and lift table control sequences were demonstrated. OIU menus, messages, and alarm sequences were examined, with few exceptions noted. Barcode testing was bypassed, due to the lack of installed equipment as well as the switch from basic reliance on fixed bar code readers to the enhanced use of portable bar code readers. Bar code testing was completed during performance of the LLW RWM OTP. Mechanical and control deficiencies were documented as Test Exceptions during performance of this Acceptance Test. These items are attached as Appendix A to this report.

  1. WDS.cdr

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home RoomPreservationBio-Inspired SolarAbout /Two0 -UsingHeat &DOE-RL Contracts/ProcurementsWaste

  2. Proven concepts for LLW-treatment of large components for free release and recycling

    SciTech Connect (OSTI)

    Bergstroem, Lena; Lindstrom, Anders; Lindberg, Maria; Wirendal, Bo; Lorenzen, Joachim [Studsvik RadWaste AB, SE-611 82 Nykoeping (Sweden)

    2007-07-01

    This paper describes Studsvik's technical concept of LLW-treatment of large, retired components from nuclear installations in operation or in decommissioning. Many turbines, heat exchangers and other LLW components have been treated in Studsvik during the last 20 years. This also includes development of techniques and tools, especially our latest experience gained under the pilot project for treatment of one full size PWR steam generator from Ringhals NPP, Sweden. The ambition of this pilot project was to minimize the waste volumes for disposal and to maximize the material recycling. Another objective, respecting ALARA, was the successful minimization of the dose exposure to the personnel. The treatment concept for large, retired components comprises the whole sequence of preparations from road and sea transports and the management of the metallic LLW by segmentation, decontamination and sorting using specially devised tools and shielded treatment cell, to the decision criteria for recycling of the metals, radiological analyses and conditioning of the residual waste into the final packages suitable for customer-related disposal. For e.g. turbine rotors with their huge number of blades the crucial moments are segmentation techniques, thus cold segmentation is a preferred method to keep focus on minimization of volumes for secondary waste. Also a variety of decontamination techniques using blasting cabinet or blasting tumbling machines keeps secondary waste production to a minimum. The technical challenge of the treatment of more complicated components like steam generators also begins with the segmentation. A first step is the separation of the steam dome in order to dock the rest of the steam generator to a specially built treatment cell. Thereafter, the decontamination of the tube bundle is performed using a remotely controlled manipulator. After decontamination is concluded the cutting of the tubes as well as of the shell is performed in the same cell with remotely controlled tools. Some of the sections of steam dome shell or turbine shafts can be cleared directly for unconditional reuse without melting after decontamination and sampling program. Experience shows that the amount of material possible for clearance for unconditional use is between 95 - 97 % for conventional metallic scrap. For components like turbines, heat exchangers or steam generators the recycling ratio can vary to about 80 - 85% of the initial weight. (authors)

  3. A Comprehensive Solution for Managing TRU and LLW From Generation to Final Disposition - 13205

    SciTech Connect (OSTI)

    Tozer, Justin C.; Sanchez, Edwina G.; Dorries, Alison M.

    2013-07-01

    A LANL multi-disciplinary team faced the challenge of building and delivering a waste information system capable of managing radioactive, hazardous, and industrial waste from cradle to grave. The result is the Waste Compliance and Tracking System (WCATS) a flexible, adaptive system that has allowed LANL to consolidate its legacy applications into one system, and leverage the advantages of managing all waste types within a single scalable enterprise application. Key functionality required for robust waste operations, include: waste characterization, waste identification, transportation, inventory management, waste processing, and disposal. In order to maintain data quality, field operations such as waste identification, surveillance checklists, wall-to-wall inventory assessments, waste transfers, shipment pickup and receipt, and simple consolidation operations are captured by the operator or technician using mobile computers. Work flow is managed via end-user defined work paths, to ensure that unit operations are performed in the correct order. Regulatory compliance reports and algorithms are provided to support typical U.S. EPA, DOT, NRC, and DOE requirements, including the EPA hazardous waste manifest, NRC LLW manifest, DOE nuclear material at risk, RCRA TSDF inventory rules, and so forth. The WCATS application has allowed LANL to migrate and consolidate its disparate legacy applications. The design and implementation is generalized so that facility owners can customize the user interface, setup facilities and unit operations (i.e., treatment, storage, disposal, characterization, and administrative), define inventory compliance rules, and establish custom work flow requirements. (authors)

  4. ASSESSING EXPOSURE TO THE PUBLIC FROM LOW LEVEL RADIOACTIVE WASTE (LLW) TRANSPORTATION TO THE NEVADA TEST SITE.

    SciTech Connect (OSTI)

    Miller, J.J.; Campbell, S.; Church, B.W.; Shafer, D. S.; Gillespie, D.; Sedano, S.; Cebe, J.J.

    2003-02-27

    The United States (U.S.) Department of Energy (DOE) Nevada Test Site (NTS) is one of two regional sites where low-level radioactive waste (LLW) from approved DOE and U.S. DOD generators across the United States is disposed. In federal fiscal year (FY) 2002, over 57,000 cubic meters of waste was transported to and disposed at the NTS. DOE and U.S. Department of Transportation (DOT) regulations ensure that radiation exposure from truck shipments to members of the public is negligible. Nevertheless, particularly in rural communities along transportation routes in Utah and Nevada, there is perceived risk from members of the public about incremental exposure from LLW trucks, especially when ''Main Street'' and the LLW transportation route are the same. To better quantify the exposure to gamma radiation, a stationary monitoring array of four pressurized ion chambers (PICs) have been set up in a pullout just before LLW trucks reach the entrance to the NTS. The PICs are positioned at a distance of one meter from the sides of the truck trailer and at a height appropriate for the design of the trucks that will be used in FY2003 to haul LLW to the NTS. The use of four PICs (two on each side of the truck) is to minimize and to correct for non-uniformity where radiation levels from waste packages vary from side to side, and from front to back in the truck trailer. The PIC array is being calibrated by collecting readings from each PIC exposed to a known 137Cs source that was positioned at different locations on a flatbed stationed in the PIC array, along with taking secondary readings from other known sources. Continuous data collection using the PICs, with and without a truck in the array, is being used to develop background readings. In addition, acoustic sensors are positioned on each side of the PIC array to record when a large object (presumably a truck) enters the array. In FY2003, PIC surveys from as many incoming LLW trucks as possible will be made and survey data recorded automatically by dataloggers that will be periodically downloaded. Solar panels provide power for the batteries to run both the dataloggers and PICs. Truck drivers have been asked to park their truck within the PIC array for only the time it takes to complete an information log before moving on to one of two Radioactive Waste Management Sites (RWMS) on the NTS. On the log, the truck drivers record their shipment identification number, the time of day, where the waste originated, and information on the route they used to reach the NTS. This data will facilitate comparison of PIC readings with waste manifests and other waste disposal operations data collected at the RWMSs. Gamma radiation measurements collected from the PICs will be analyzed using standard health physics and statistical methods for comparison to DOT standards, but with the added benefit of obtaining an improved understanding of the variability of readings that can occur in the near vicinity of a LLW truck. The data collected will be combined with measurements of street width and other information about transportation routes through towns to develop realistic dose scenarios for citizens in Nevada and Utah towns.

  5. Caustic Recycling Pilot Unit to Separate Sodium from LLW at Hanford Site - 12279

    SciTech Connect (OSTI)

    Pendleton, Justin; Bhavaraju, Sai; Priday, George; Desai, Aditya; Duffey, Kean; Balagopal, Shekar [Ceramatec Inc., Salt Lake City, UT 84119 (United States)

    2012-07-01

    As part of the Department of Energy (DOE) sponsored Advanced Remediation Technologies initiative, a scheme was developed to combine Continuous Sludge Leaching (CSL), Near-Tank Cesium Removal (NTCR), and Caustic Recycling Unit (CRU) using Ceramatec technology, into a single system known as the Pilot Near-Tank Treatment System (PNTTS). The Cesium (Cs) decontaminated effluent from the NTCR process will be sent to the caustic recycle process for recovery of the caustic which will be reused in another cycle of caustic leaching in the CSL process. Such an integrated mobile technology demonstration will give DOE the option to insert this process for sodium management at various sites in Hanford, and will minimize the addition of further sodium into the waste tanks. This allows for recycling of the caustic used to remove aluminum during sludge washing as a pretreatment step in the vitrification of radioactive waste which will decrease the Low Level Waste (LLW) volume by as much as 39%. The CRU pilot process was designed to recycle sodium in the form of pure sodium hydroxide. The basis for the design of the 1/4 scale pilot caustic recycling unit was to demonstrate the efficient operation of a larger scale system to recycle caustic from the NTCR effluent stream from the Parsons process. The CRU was designed to process 0.28 liter/minute of NTCR effluent, and generate 10 M concentration of 'usable' sodium hydroxide. The proposed process operates at 40 deg. C to provide additional aluminum solubility and then recover the sodium hydroxide to the point where the aluminum is saturated at 40 deg. C. A system was developed to safely separate and vent the gases generated during operation of the CRU with the production of 10 M sodium hydroxide. Caustic was produced at a rate between 1.9 to 9.3 kg/hr. The CRU was located inside an ISO container to allow for moving of the unit close to tank locations to process the LLW stream. Actual tests were conducted with the NTCR effluent simulant from the Parsons process in the CRU. The modular CRU is easily scalable as a standalone system for caustic recycling, or for NTTS integration or for use as an In-Tank Treatment System to process sodium bearing waste to meet LLW processing needs at the Hanford site. The standalone pilot operation of the CRU to recycle sodium from NTCR effluent places the technology demonstration at TRL level 6. Multiple operations were performed with the CRU to process up to 500 gallons of the NTCR effluent and demonstrate an efficient separation of up to 70 % of the sodium without solids precipitation while producing 10 M caustic. Batch mode operation was conducted to study the effects of chemistry variation, establish the processing rate, and optimize the process operating conditions to recycle caustic from the NTCR effluent. The performance of the CRU was monitored by tracking the density parameter to control the concentration of caustic produced. Different levels of sodium were separated in tests from the effluent at a fixed operating current density and temperature. The voltage of the modules remained stable during the unit operation which demonstrated steady operation to separate sodium from the NTCR effluent. The sodium transfer current efficiency was measured in testing based on the concentration of caustic produced. Measurements showed a current efficiency of 99.8% for sodium transfer from the NTCR effluent to make sodium hydroxide. The sodium and hydroxide contents of the anolyte (NTCR feed) and catholyte (caustic product) were measured before and after each batch test. In two separate batch tests, samples were taken at different levels of sodium separation and analyzed to determine the stability of the NTCR effluent after sodium separation. The stability characteristics and changes in physical and chemical properties of the NTCR effluent chemistry after separation of sodium hydroxide as a function of storage time were evaluated. Parameters such as level of precipitated alumina, total alkalinity, analysis of Al, Na, K, Cs, Fe, OH, nitrate, nitrite, total dissolved and

  6. Integration of US Department of Energy contractor installations for the purpose of optimizing treatment, storage, and disposal of low-level radioactive waste (LLW)

    SciTech Connect (OSTI)

    Lucas, M.; Gnoose, J.; Coony, M.; Martin, E.; Piscitella, R.

    1998-02-01

    The US Department of Energy (DOE) manages a multibillion dollar environmental management (EM) program. In June 1996, the Assistant Secretary of Energy for EM issued a memorandum with guidance and a vision for a ten year planning process for the EM Program. The purpose of this process, which became known as the Accelerated Cleanup: Focus on 2006, is to make step changes within the DOE complex regarding the approach for making meaningful environmental cleanup progress. To augment the process, Assistant Secretary requested the site contractors to engage in an effort to identify and evaluate integration alternatives for EM waste stream treatment, storage, and disposal (TSD) that would parallel the 2006 Plan. In October 1996, ten DOE contractor installations began the task of identifying alternative opportunities for low level radioactive waste (LLW). Cost effective, efficient solutions were necessary to meet all requirements associated with storing, characterizing, treating, packaging, transporting, and disposing of LLW while protecting the workers` health and safety, and minimizing impacts to the environment. To develop these solutions, a systems engineering approach was used to establish the baseline requirements, to develop alternatives, and to evaluate the alternatives. Key assumptions were that unique disposal capabilities exist within the DOE that must be maintained; private sector disposal capability for some LLW may not continue to exist into the foreseeable future; and decisions made by the LLW Team must be made on a system or complex wide basis to fully realize the potential cost and schedule benefits. This integration effort promoted more accurate waste volume estimates and forecasts; enhanced recognition of existing treatment, storage, and disposal capabilities and capacities; and improved identification of cost savings across the complex.

  7. Disposal of LLW and ILW in Germany - Characterisation and Documentation of Waste Packages with Respect to the Change of Requirements

    SciTech Connect (OSTI)

    Bandt, G.; Spicher, G.; Steyer, St.; Brennecke, P.

    2008-07-01

    Since the 1998 termination of LLW and ILW emplacement in the Morsleben repository (ERAM), Germany, the treatment, conditioning and documentation of radioactive waste products and packages have been continued on the basis of the waste acceptance requirements as of 1995, prepared for the Konrad repository near Salzgitter in Lower Saxony, Germany. The resulting waste products and packages are stored in interim storage facilities. Due to the Konrad license issued in 2002 the waste acceptance requirements have to be completed by additional requirements imposed by the licensing authority, e. g. for the declaration of chemical waste package constituents. Therefore, documentation of waste products and packages which are checked by independent experts and are in parts approved by the responsible authority (Office for Radiation Protection, BfS) up to now will have to be checked again for fulfilling the final waste acceptance requirements prior to disposal. In order to simplify these additional checks, databases are used to ensure an easy access to all known facts about the waste packages. A short balance of the existing waste products and packages which are already checked and partly approved by BfS as well as an overview on the established databases ensuring a fast access to the known facts about the conditioning processes is presented. (authors)

  8. Management of nuclear materials and non-HLW | Department of Energy

    Broader source: Energy.gov (indexed) [DOE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirleyEnergy A plug-inPPLforLDRDEnergyTurbineProcessesEnergyofEnergyThe 2020consolidation

  9. A Novel and Cost Effective Approach to the Decommissioning and Decontamination of Legacy Glove Boxes - Minimizing TRU Waste and Maximizing LLW Waste - 13634

    SciTech Connect (OSTI)

    Pancake, Daniel; Rock, Cynthia M.; Creed, Richard [Argonne National Laboratory, 9700 South Cass Avenue, Lemont, IL 60439 (United States)] [Argonne National Laboratory, 9700 South Cass Avenue, Lemont, IL 60439 (United States); Donohoue, Tom; Martin, E. Ray; Mason, John A. [ANTECH Corporation 9050 Marshall Court, Westminster, CO, 80031 (United States)] [ANTECH Corporation 9050 Marshall Court, Westminster, CO, 80031 (United States); Norton, Christopher J.; Crosby, Daniel [Environmental Alternatives, Inc., 149 Emerald Street, Suite R, Keene, NH 03431 (United States)] [Environmental Alternatives, Inc., 149 Emerald Street, Suite R, Keene, NH 03431 (United States); Nachtman, Thomas J. [InstaCote, Inc., 160 C. Lavoy Road, Erie, MI, 48133 (United States)] [InstaCote, Inc., 160 C. Lavoy Road, Erie, MI, 48133 (United States)

    2013-07-01

    This paper describes the process of decommissioning two gloveboxes at the Argonne National Laboratory (ANL) that were employed for work with plutonium and other radioactive materials. The decommissioning process involved an initial phase of clearing tools and materials from the glove boxes and disconnecting them from the laboratory infrastructure. The removed materials, assessed as Transuranic (TRU) waste, were packaged into 55 gallon (200 litre) drums and prepared for ultimate disposal at the Waste Isolation Pilot Plant (WIPP) at Carlsbad New Mexico. The boxes were then sampled to determine the radioactive contents by means of smears that were counted with alpha and beta detectors to determine the residual surface contamination, especially in terms of alpha particle emitters that are an indicator of TRU activity. Paint chip samples were also collected and sent for laboratory analysis in order to ascertain the radioactive contamination contributing to the TRU activity as a fixed contamination. The investigations predicted that it may be feasible to reduce the residual surface contamination and render the glovebox structure low level waste (LLW) for disposal. In order to reduce the TRU activity a comprehensive decontamination process was initiated using chemical compounds that are particularly effective for lifting and dissolving radionuclides that adhere to the inner surfaces of the gloveboxes. The result of the decontamination process was a reduction in the TRU surface activity on the inner surfaces of the gloveboxes by four orders of magnitude in terms of disintegrations per unit area (DPA). The next phase of the process involved a comprehensive assay of the gloveboxes using a combination of passive neutron and gamma ray scintillation detectors and a shielded and collimated high purity Germanium (HPGe) gamma ray detector. The HPGe detector was used to obtain gamma ray spectra for a variety of measurement positions within the glovebox. The spectra were used to determine the TRU content of the boxes by assessing the activity of Am-241 (59 keV) and Pu-241 (414 keV). Using the data generated it was possible for qualified subject matter experts (SME) to assess that the gloveboxes could be consigned for disposition as LLW and not as TRU. Once this determination was assessed and accepted the gloveboxes were prepared for final disposition to the Nevada National Security Site (NNSS) - formerly the Nevada Test Site (NTS). This preparation involved fixing any remaining radioactive contamination within the gloveboxes by filling them with a foam compound, prior to transportation. Once the remaining contamination was fixed the gloveboxes were removed from the laboratory and prepared for transported by road to NNSS. This successful glovebox decontamination and decommissioning process illustrates the means by which TRU waste generation has been minimized, LLW generation has been maximized, and risk has been effectively managed. The process minimizes the volume of TRU waste and reduced the decommissioning time with significant cost savings as the result. (authors)

  10. LLW Notes supplement, Volume 12, Number 6

    SciTech Connect (OSTI)

    NONE

    1997-07-01

    Contents include articles related to environmental justice concerns and Title VI, entitled as follows: Civil Rights Act of 1964; Exec order on environmental justice; Applicability to states; Philosophical differences -- Environmental justice and Title VI; Ambiguities in existing Title VI guidance; Clarification of existing Title VI guidance; Federal financial assistance; Administrative complaints vs. lawsuits; Effect and disparate impact; Termination, suspension or refusal to grant federal financial assistance; DOJ guidance defines environmental justice; NEJAC meets, adopts far-reaching resolution re siting; Indigenous Peoples Resolution No. 23; and States meet, support environmental justice concept and express concerns about federal approach and composition of NEJAC.

  11. LLW Notes, Volume 12, Number 8

    SciTech Connect (OSTI)

    Norris, C.; Brown, H. [eds.; Gedden, R.; Lovinger, T.; Scheele, L.; Shaker, M.A.

    1997-12-31

    Contents include articles entitled: Chem-Nuclear documents new plan for Barnwell; Nebraska releases technical analysis of LLRW facility; Southeast Compact suspends funding for NC facility development; NC governor and Southeast Compact differ on proposed MOU; Midwest Compact to return export fees; State legislators` group revises radioactive waste policy; Internal documents discuss administration`s policy on Ward Valley; BLM issues EA for Ward Valley testing; California DHS, NRC criticize DOI`s testing protocols; Army removes training mines from Ward Valley site; The 1997 gubernatorial elections and a look ahead to 1998; Court throws out case challenging Pennsylvania`s siting law; DOE files notice of appeal in WCS suit; Central Compact moves to dismiss ``Veto`` authority suit; Congress exempts NAS from FACA; Judge sets schedule for Ward Valley case; Court won`t order DOE to accept spent fuel by deadline; NRC chairman expresses concern re CERCLA reauthorization; Senators question EPA`s guidance on remediation; EPA issues guidance, criticizes NRC decommissioning rule; Members of Congress clarify FUSRAP transfer; HLW legislation passes House by wide margin; Takings legislation passes House; Energy and water bill signed into law; and Senate confirms 5 of 6 DOE appointees.

  12. LLW Notes, Volume 12, Number 6

    SciTech Connect (OSTI)

    Norris, C.; Brown, H.; Gedden, R.; Lovinger, T.; Scheele, L.; Shaker, M.A.

    1997-07-01

    Contents include articles entitled: GAO concludes most Ward Valley SEIS issues previously addressed; Midwest compact halts facility development; Texas publishes proposal to issue WCS radioactive materials license; Central Compact issues export authorizations over NE`s objection; Nebraska governor to host LLRW summit; California regulators reassured re US ecology facility in WA; Southeast Compact augments funding for North Carolina; State and compact calendar of events; IAEA Director General to UN: reexamine nuclear power; DOI convenes meetings on Ward Valley Title VI complaint; California BLM: Tribes fully represented and consulted; MW, NE, and SW file amici curiae briefs in Ward Valley suit; Court denies state`s motion for protective order; WCS files suit against Envirocare and others; States attack DOE`s claim re lack of authority to store spent fuel; House committee passes Texas legislation; Ward Valley land transfer bill introduced in Senate; Senate committee holds hearing on Ward Valley legislation and related GAO report; NRDC threatens to sue DOE re Envirocare; NRC chair criticizes Deputy Interior Secretary`s use of Ward Valley fact sheet; Utility consortium submits license application for storage on Goshute land to NRC; Envirocare cited for SNM violation; EPA begins audit; and EPA rejects Title VI claim re Texas site.

  13. LLW Notes, Volume 12, Number 7

    SciTech Connect (OSTI)

    Norris, C.; Brown, H.; Gedden, R.; Lovinger, T.; Scheele, L.; Shaker, M.A.

    1997-09-01

    Contents include articles entitled: House votes 309 to 107 to approve Texas compact; Nebraska governor hosts LLRW meeting; Southeast Compact considers funding proposal; Chem-Nuclear explores options re SC revenue requirements; Legislation sets revenue requirements for Barnwell; TCC meets: Supports CA request for technical assistance; DOE approves part of California`s technical assistance request; State legislators discuss LLRW management for OH, IL, NC; Washington governor re Potential New Hanford Role; Federal court enjoins DOE from excluding WCS on new disposal; Appellate court in favor of DOE in surcharge rebates dispute; Hearing set for October in Ward Valley case; court rejects federal motion to dismiss Ward Valley suit; NE sues commission re veto over export authorizations; US Supreme Court dismisses line-item veto challenge; Department of Interior Inspector General investigation requested; USEC privatization plan approved; DOD finalizes LLRW disposal charter; Clinton nominates six DOE appointees; Congress moves FUSRAP to Army Corps of Engineers; Schaefer named interim director of USGS: Nichols leaves EPA: NRC Commissioner Rogers` term expires; NRC: CA ``Well-Quantified`` to license Ward Valley facility; EPA objects to state permit for Louisiana facility; Petitions submitted to EPA oppose Shintech permits; ECOS draft recommendations re Enviro programs; Legislation introduced to prohibit spent fuel shipments to the Goshutes; and HLW legislation ready for floor action.

  14. LLW Notes, Volume 12, Number 4

    SciTech Connect (OSTI)

    Norris, C.; Brown, H.; Gedden, R.; Lovinger, T.; Scheele, L.; Shaker, M.A.

    1997-04-01

    Contents include articles entitled: Texas Authority`s funding pending before conference committee: Auditor`s report favors authority; Revisions likely for Illinois siting law; Midwest Compact votes on Ohio fundings: Less approved than requested; Walter Sturgeon named executive director of North Carolina authority; New forum participant for Massachusetts; CRCPD holds fifth workshop for LLRW regulators; DOD generators hold annual meeting; State legislators` LLRW working group meets; NRC Chairman Jackson responds to proposal to amend the Policy Act; US Ecology uses to recover costs and lost profits and/or to compel Ward Valley land transfer; New suit against Envirocare and others alleges unlawful business practices; Federal court finds line-item veto unconstitutional; States/utilities seek to escrow nuclear waste payments; High-level waste bill passes Senate; NRC releases decommissioning rule; EPA Region VI re La Paz Agreement; EPA, NRC debate NRC`s decommissioning rule: No progress re approaches to risk harmonization; and Mousseau heads DOE`s national low-level waste management program.

  15. LLW Notes, Volume 12, Number 1

    SciTech Connect (OSTI)

    Norris, C.; Brown, H.; Colsant, J.; Lovinger, T.; Scheele, L.; Shaker, M.A.

    1997-01-01

    Contents include articles entitled: Suit against Envirocare sparks investigations: Formal petition filed with NRC; Group alleges misconduct by USGS re Beatty study; EPA rescinds NESHAPs subpart 1; Northwest Compact executive director changes jobs; New forum participant for the state of New Jersey; and Director of North Carolina division of radiation control retires.

  16. LLW Notes, Volume 12, Number 2

    SciTech Connect (OSTI)

    Norris, C.; Brown, H.; Colsant, J.; Lovinger, T.; Scheele, L.; Shaker, M.A.

    1997-02-01

    Contents include the following articles: National Environmental Justice Advisory Council considers Ward Valley resolution; NGA urges Congressional and Presidential support for low-level radioactive waste compacts and transfer of federal land in Ward Valley; RFP issued for SEIS on Ward Valley land transfer; Illinois siting criteria finalized; Consideration of tribal concerns during Ward Valley siting process; State legislators` LLRW working group meets in D.C.; Upcoming state and compact events; Court calendar; Texas compact legislation introduced in Congress; Superfund reform is a priority for 105th Congress; High-level waste bill gets off to an early start; Fort Mojave petition NEJAC for Ward Valley resolution; EPA withdraws cleanup rule from OMB; Board ruling raises doubts about proposed Louisiana enrichment facility; DOE recommends external regulation by NRC; and Supplement--Background on environmental justice.

  17. Characterization and Pre-treatment of LLW in Turkey - 12572

    SciTech Connect (OSTI)

    Osmanlioglu, Ahmet Erdal

    2012-07-01

    Pre-treatment of radioactive waste is the first step in waste management program that occurs after waste generation from various applications in Turkey. Pre-treatment and characterization practices are carried out in Radioactive Waste Management Unit (RWMU) at Cekmece Nuclear Research and Training Center (CNRTC) in Istanbul. This facility has been assigned to take all low-level radioactive wastes generated by nuclear applications in Turkey. The wastes are generated from research and nuclear applications mainly in medicine, biology, agriculture, quality control in metal processing and construction industries. These wastes are classified as low- level radioactive wastes. Pre-treatment practices cover several steps. In this paper, main steps of pre-treatment and characterization are presented. Basically these are; collection, segregation, chemical adjustment, size reduction and decontamination operations. (author)

  18. Greater-than-Class C Low-Level Radioactive Waste (GTCC LLW) ...

    Energy Savers [EERE]

    Pilot Plant in Carlsbad, N.M. On February 17, 2011, DOE issued the Draft Environmental Impact Statement (EIS) for the Disposal of Greater-Than-Class C (GTCC) Low-Level Radioactive...

  19. LLW Processing and Operational Experience using a Plasma ARC Centrifugal Treatment (PACT{sup TM}) System

    SciTech Connect (OSTI)

    Shuey, M.W.; Ottmer, P.P.

    2006-07-01

    After several years of development, a commercially available high-temperature treatment system has been developed, licensed, and installed that treats heterogeneous low-level radioactive waste. High temperature plasma processing, unique torch design and operating features make it feasible to achieve a volume reduced, permanent, high integrity waste form while eliminating the personnel exposure and costs associated with conventional sorting, characterizing and handling. The Plasma Arc Centrifugal Treatment system or PACT{sup TM} manufactured by Retech Systems LLC is a licensed thermal plasma system that processes and consolidates low level radioactive wastes. The first PACT{sup TM} thermal plasma system to be licensed was at ZWILAG (Zwischenlager Wuerenlingen AG, Switzerland) in May 2004, and the second is utilized by the Japan Atomic Power Company (JAPC) in Tsuruga, Japan in March 2005. ZWILAG uses a drum feeder that processes the 200-liter drums from storage horizontally and pours the molten slag into molds. The drums contain organic and inorganic wastes (mixed waste), and by processing the drums directly lowers exposure to processing personnel. ZWILAG production data mid-2004 through mid-June 2005 has fed 9.4 E+10 Bq of mixed waste and stabilized 8.5 E+10 Bq in slag with a mean activity of 2.1 E+09 Bq/drum. The operational experience demonstrated by ZWILAG and JAPC has been a testament to the success of thermal plasma and their unique status has proven the real benefits of using the PACT{sup TM} system. (authors)

  20. PNL vitrification technology development project glass formulation strategy for LLW vitrification

    SciTech Connect (OSTI)

    Kim, D.; Hrma, P.R.; Westsik, J.H. Jr.

    1996-03-01

    This Glass Formulation Strategy describes development approaches to optimize glass compositions for Hanford`s low-level waste vitrification between now and the projected low-level waste facility start-up in 2005. The objectives of the glass formulation task are to develop optimized glass compositions with satisfactory long-term durability, acceptable processing characteristics, adequate flexibility to handle waste variations, maximize waste loading to practical limits, and to develop methodology to respond to further waste variations.

  1. Greater-than-Class C Low-Level Radioactive Waste (GTCC LLW) | Department of

    Broader source: Energy.gov (indexed) [DOE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirleyEnergy A plug-inPPLforLDRD Report11, SolarMat 4"Gloria B.CharlesEnergyasEnergy

  2. INNOVATIVE DISPOSAL PRACTICES AT THE NEVADA TEST SITE TO MEET ITS LLW GENERATORS FUTURE DISPOSAL NEEDS

    National Nuclear Security Administration (NNSA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefield Municipal GasAdministration Medal01 Sandia4) AugustA. -71- Particulate: Columns 59 and R

  3. Central Characterization Program (CCP) Contact-Handled (CH) TRU Waste Certification and Waste Information System/Waste Data System (WWIS/WDS) Data Entry

    Broader source: Energy.gov [DOE]

    Supporting Technical Document for the Radiological Release Accident Investigation Report (Phase II Report)

  4. Office of the Assistant General Counsel for Civilian Nuclear...

    Broader source: Energy.gov (indexed) [DOE]

    of Nuclear Materials and Non-HLW Nuclear Fuel Cycle Energy Research and Development Non-Proliferation Nuclear Regulatory Commission Regulatory and Licensing Matters Nuclear...

  5. RH-LLW Disposal Facility Project CD-2/3 to Design/Build Proposal Reconciliation Report

    SciTech Connect (OSTI)

    Annette L. Schafer

    2012-06-01

    A reconciliation plan was developed and implemented to address potential gaps and responses to gaps between the design/build vendor proposals and the Critical Decision-2/3 approval request package for the Remote-Handled Low Level Waste Disposal Facility Project. The plan and results of the plan implementation included development of a reconciliation team comprised of subject matter experts from Battelle Energy Alliance and the Department of Energy Idaho Operations Office, identification of reconciliation questions, reconciliation by the team, identification of unresolved/remaining issues, and identification of follow-up actions and subsequent approvals of responses. The plan addressed the potential for gaps to exist in the following areas: Department of Energy Order 435.1, Radioactive Waste Management, requirements, including the performance assessment, composite analysis, monitoring plan, performance assessment/composite analysis maintenance plan, and closure plan Environmental assessment supporting the National Environmental Policy Act Nuclear safety Safeguards and security Emplacement operations Requirements for commissioning General project implementation. The reconciliation plan and results of the plan implementation are provided in a business-sensitive project file. This report provides the reconciliation plan and non-business sensitive summary responses to identified gaps.

  6. Waste Determination and Section 3116 of the 2005 National Defense...

    Office of Environmental Management (EM)

    y * Both paths provide a methodology to treat and manage waste incidental to reprocessing as non-HLW * Section 3116 can only be applied in the states of South Carolina and...

  7. DOE Railcar Fleet Asset Planning & Lessons Learned

    Office of Environmental Management (EM)

    Campaigns - LLW -Fernald, Mound, Savannah River *In Progress - LLW -Savannah River, Brookhaven, Moab *ForeCast - LLW -Portsmouth, Paducah, D&D, DUF6 2 *Lessons Learned F ld Cl P j...

  8. Physical, Chemical and Structural Evolution of Zeolite-Containing Waste Forms Produced from Metakaolinite and Calcined Sodium Bearing Waste (HLW and/or LLW)

    SciTech Connect (OSTI)

    Grutzeck, Michael W.

    2005-06-27

    Zeolites are extremely versatile. They can adsorb liquids and gases and serve as cation exchange media. They occur in nature as well cemented deposits. The ancient Romans used blocks of zeolitized tuff as a building material. Using zeolites for the management of radioactive waste is not a new idea, but a process by which the zeolites can be made to act as a cementing agent is. Zeolitic materials are relatively easy to synthesize from a wide range of both natural and man-made substances. The process under study is derived from a well known method in which metakaolin (an impure thermally dehydroxylated kaolinite heated to {approx}700 C containing traces of quartz and mica) is mixed with sodium hydroxide (NaOH) and reacted in slurry form (for a day or two) at mildly elevated temperatures. The zeolites form as finely divided powders containing micrometer ({micro}m) sized crystals. However, if the process is changed slightly and only just enough concentrated sodium hydroxide solution is added to the metakaolinite to make a thick crumbly paste and then the paste is compacted and cured under mild hydrothermal conditions (60-200 C), the mixture will form a hard ceramic-like material containing distinct crystalline tectosilicate minerals (zeolites and feldspathoids) imbedded in an X-ray amorphous hydrated sodium aluminosilicate matrix. Due to its lack of porosity and vitreous appearance we have chosen to call this composite a ''hydroceramic''.

  9. Tank farms compacted low-level waste

    SciTech Connect (OSTI)

    Hetzer, D.C.

    1997-08-01

    This report describes the process of Low-Level Waste (LLW) volume reduction by compaction. Also included is the data used for characterization of LLW destined for compaction. Scaling factors (ratios) are formed based on data contained in this report.

  10. Tank farms compacted low level waste

    SciTech Connect (OSTI)

    Waters, M.S., Westinghouse Hanford

    1996-07-01

    This report describes the process of Low Level Waste (LLW) volume reduction by compaction. Also included is the data used for characterization of LLW destined for compaction. Scaling factors (ratios) are formed based on data contained in this report.

  11. EIS-0243: Nevada Test Site and Off-Site Locations in the State of Nevada

    Broader source: Energy.gov [DOE]

    This EIS evaluates the potential environmental impacts of the management of low-level waste (LLW) at all sites and continue, to the extent practicable, disposal of on- site LLW at the Idaho...

  12. 70 DA WHITE DWARFS IDENTIFIED IN LAMOST PILOT SURVEY

    SciTech Connect (OSTI)

    Zhao, J. K.; Luo, A. L.; Zhao, G. [Key Laboratory of Optical Astronomy, National Astronomical Observatories, Chinese Academy of Sciences, Beijing 100012 (China); Oswalt, T. D., E-mail: zjk@bao.ac.cn, E-mail: gzhao@bao.ac.cn, E-mail: lal@bao.ac.cn, E-mail: toswalt@fit.edu [Physics and Space Science Department, Florida Institute of Technology, Melbourne, FL 32901 (United States)

    2013-06-01

    We present a spectroscopically identified catalog of 70 DA white dwarfs (WDs) from the LAMOST pilot survey. Thirty-five are found to be new identifications after cross-correlation with the Eisenstein et al. and Villanova catalogs. The effective temperature and gravity of these WDs are estimated by Balmer lines fitting. Most of them are hot WDs. The cooling times and masses of these WDs are estimated by interpolation in theoretical evolution tracks. The peak of the mass distribution is found to be {approx}0.6 M {sub Sun }, which is consistent with prior work in the literature. The distances of these WDs are estimated using the method of synthetic spectral distances. All of these WDs are found to be in the Galactic disk from our analysis of space motions. Our sample supports the expectation that WDs with high mass are concentrated near the plane of the Galactic disk.

  13. The late emission of thermonuclear supernovae

    E-Print Network [OSTI]

    Pilar Ruiz-Lapuente

    1996-04-16

    The subject of late-time emission of Type Ia supernovae and its implications for the understanding of the explosions of C+O WDs is reviewed.

  14. The role of organic complexants and microparticulates in the facilitated transport of radionuclides

    SciTech Connect (OSTI)

    Schilk, A.J.; Robertson, D.E.; Abel, K.H.; Thomas, C.W.

    1996-12-01

    This progress report describes the results of ongoing radiological and geochemical investigations of the mechanisms of radionuclide transport in groundwater at two low-level waste (LLW) disposal sites within the waste management area of the Chalk River Laboratories (CRL), Ontario, Canada. These sites, the Chemical Pit liquid disposal facility and the Waste Management Area C solid LLW disposal site, have provided valuable 30- to 40-year-old field locations for characterizing the migration of radionuclides and evaluating a number of recent site performance objectives for LLW disposal facilities. This information will aid the NRC and other federal, state, and local regulators, as well as LLW disposal site developers and waste generators, in maximizing the effectiveness of existing or projected LLW disposal facilities for isolating radionuclides from the general public and thereby improving the health and safety aspects of LLW disposal.

  15. Estimating heel retrieval costs for underground storage tank waste at Hanford. Draft

    SciTech Connect (OSTI)

    DeMuth, S.

    1996-08-26

    Approximately 100 million gallons ({approx}400,000 m{sup 3}) of existing U.S. Department of Energy (DOE) owned radioactive waste stored in underground tanks can not be disposed of as low-level waste (LLW). The current plan for disposal of UST waste which can not be disposed of as LLW is immobilization as glass and permanent storage in an underground repository. Disposal of LLW generally can be done sub-surface at the point of origin. Consequently, LLW is significantly less expensive to dispose of than that requiring an underground repository. Due to the lower cost for LLW disposal, it is advantageous to separate the 100 million gallons of waste into a small volume of high-level waste (HLW) and a large volume of LLW.

  16. Low-level radioactive waste disposal technologies used outside the United States

    SciTech Connect (OSTI)

    Templeton, K.J.; Mitchell, S.J.; Molton, P.M.; Leigh, I.W.

    1994-01-01

    Low-level radioactive waste (LLW) disposal technologies are an integral part of the waste management process. In the United States, commercial LLW disposal is the responsibility of the State or groups of States (compact regions). The United States defines LLW as all radioactive waste that is not classified as spent nuclear fuel, high- level radioactive waste, transuranic waste, or by-product material as defined in Section II(e)(2) of the Atomic Energy Act. LLW may contain some long-lived components in very low concentrations. Countries outside the United States, however, may define LLW differently and may use different disposal technologies. This paper outlines the LLW disposal technologies that are planned or being used in Canada, China, Finland, France, Germany, Japan, Sweden, Taiwan, and the United Kingdom (UK).

  17. Review of private sector and Department of Energy treatment, storage, and disposal capabilities for low-level and mixed low-level waste

    SciTech Connect (OSTI)

    Willson, R.A.; Ball, L.W.; Mousseau, J.D.; Piper, R.B.

    1996-03-01

    Private sector capacity for treatment, storage, and disposal (TSD) of various categories of radioactive waste has been researched and reviewed for the Idaho National Engineering Laboratory (INEL) by Lockheed Idaho Technologies Company, the primary contractor for the INEL. The purpose of this document is to provide assistance to the INEL and other US Department of Energy (DOE) sites in determining if private sector capabilities exist for those waste streams that currently cannot be handled either on site or within the DOE complex. The survey of private sector vendors was limited to vendors currently capable of, or expected within the next five years to be able to perform one or more of the following services: low-level waste (LLW) volume reduction, storage, or disposal; mixed LLW treatment, storage, or disposal; alpha-contaminated mixed LLW treatment; LLW decontamination for recycling, reclamation, or reuse; laundering of radioactively-contaminated laundry and/or respirators; mixed LLW treatability studies; mixed LLW treatment technology development. Section 2.0 of this report will identify the approach used to modify vendor information from previous revisions of this report. It will also illustrate the methodology used to identify any additional companies. Section 3.0 will identify, by service, specific vendor capabilities and capacities. Because this document will be used to identify private sector vendors that may be able to handle DOE LLW and mixed LLW streams, it was decided that current DOE capabilities should also be identified. This would encourage cooperation between DOE sites and the various states and, in some instances, may result in a more cost-effective alternative to privatization. The DOE complex has approximately 35 sites that generate the majority of both LLW and mixed LLW. Section 4.0 will identify these sites by Operations Office, and their associated LLW and mixed LLW TSD units.

  18. Microsoft PowerPoint - Camper, ORNL-TN CAB-04-2010-final, via...

    Office of Environmental Management (EM)

    to Clarifying Regulatory Approach * Commission Policy Decision 6 LLW Disposal Volume Projections 7 EPRI, 2006 8 "May you live in interesting times." Chi P b 9 Chinese Proverb...

  19. Remote-Handled Low-Level Waste (RHLLW) Disposal Project Code of Record

    SciTech Connect (OSTI)

    S.L. Austad, P.E.; L.E. Guillen, P.E.; C. W. McKnight, P.E.; D. S. Ferguson, P.E.

    2010-10-01

    The Remote-Handled Low-Level Waste Disposal Project addresses an anticipated shortfall in remote-handled LLW disposal capability following cessation of operations at the existing facility, which will continue until it is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of fiscal year 2015). Development of a new onsite disposal facility, the highest ranked alternative, will provide necessary remote handled LLW disposal capability and will ensure continuity of operations that generate remote-handled LLW. This report documents the Code of Record for design of a new LLW disposal capability.

  20. Attendee Introduction

    Office of Environmental Management (EM)

    do not exhibit a hazardous characteristic. * Powdered U-hydrides or tritides are LLW. Palladium hydride, or tritide, is not AEA exempt. Hydrogen isotopes are tightly held and...

  1. Comparison of low-level waste disposal programs of DOE and selected international countries

    SciTech Connect (OSTI)

    Meagher, B.G. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States); Cole, L.T. [Cole and Associates (United States)

    1996-06-01

    The purpose of this report is to examine and compare the approaches and practices of selected countries for disposal of low-level radioactive waste (LLW) with those of the US Department of Energy (DOE). The report addresses the programs for disposing of wastes into engineered LLW disposal facilities and is not intended to address in-situ options and practices associated with environmental restoration activities or the management of mill tailings and mixed LLW. The countries chosen for comparison are France, Sweden, Canada, and the United Kingdom. The countries were selected as typical examples of the LLW programs which have evolved under differing technical constraints, regulatory requirements, and political/social systems. France was the first country to demonstrate use of engineered structure-type disposal facilities. The UK has been actively disposing of LLW since 1959. Sweden has been disposing of LLW since 1983 in an intermediate-depth disposal facility rather than a near-surface disposal facility. To date, Canada has been storing its LLW but will soon begin operation of Canada`s first demonstration LLW disposal facility.

  2. Nevada Test Site Waste Acceptance Criteria (NTSWAC)

    SciTech Connect (OSTI)

    NNSA /NSO Waste Management Project

    2008-06-01

    This document establishes the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office, Nevada Test Site Waste Acceptance Criteria (NTSWAC). The NTSWAC provides the requirements, terms, and conditions under which the Nevada Test Site will accept low-level radioactive (LLW) and LLW Mixed Waste (MW) for disposal.

  3. Commercial disposal options for Idaho National Engineering Laboratory low-level radioactive waste

    SciTech Connect (OSTI)

    Porter, C.L.; Widmayer, D.A.

    1995-09-01

    The Idaho National Engineering Laboratory (INEL) is a Department of Energy (DOE)-owned, contractor-operated site. Significant quantities of low-level radioactive waste (LLW) have been generated and disposed of onsite at the Radioactive Waste Management Complex (RWMC). The INEL expects to continue generating LLW while performing its mission and as aging facilities are decommissioned. An on-going Performance Assessment process for the RWMC underscores the potential for reduced or limited LLW disposal capacity at the existing onsite facility. In order to properly manage the anticipated amount of LLW, the INEL is investigating various disposal options. These options include building a new facility, disposing the LLW at other DOE sites, using commercial disposal facilities, or seeking a combination of options. This evaluation reports on the feasibility of using commercial disposal facilities.

  4. Supplemental information related to risk assessment for the off-site transportation of low-level waste for the U.S. Department of Energy waste management programmatic environmental impact statement

    SciTech Connect (OSTI)

    Monette, F.A.; Biwer, B.M.; LePoire, D.J.; Chen, S.Y. [Argonne National Lab., IL (United States). Environmental Assessment Div.

    1996-12-01

    This report presents supplemental information to support the human health risk assessment conducted for the transportation of low-level waste (LLW) in support of the US Department of Energy Waste Management Programmatic Environmental Impact Statement (WM PEIS). Detailed descriptions of the transportation health risk assessment method and results of the assessment are presented in Appendix E of the WM PEIS and are not repeated in this report. This report presents additional information that is not presented in Appendix E but that was needed to conduct the transportation risk assessment for Waste Management (WM) LLW. Included are definition of the LLW alternatives considered in the WM PEIS, data related to the inventory and to the physical and radiological characteristics of WM LLW, an overview of the risk assessment method, and detailed results of the assessment for each WM LLW alternative considered.

  5. Disposal of low-level and mixed low-level radioactive waste during 1990

    SciTech Connect (OSTI)

    Not Available

    1993-08-01

    Isotopic inventories and other data are presented for low-level radioactive waste (LLW) and mixed LLW disposed (and occasionally stored) during calendar year 1990 at commercial disposal facilities and Department of Energy (DOE) sites. Detailed isotopic information is presented for the three commercial disposal facilities located near Barnwell, SC, Richland, WA, and Beatty, NV. Less information is presented for the Envirocare disposal facility located near Clive, UT, and for LLW stored during 1990 at the West Valley site. DOE disposal information is included for the Savannah River Site (including the saltstone facility), Nevada Test Site, Los Alamos National Laboratory, Idaho National Engineering Laboratory, Hanford Site, Y-12 Site, and Oak Ridge National Laboratory. Summary information is presented about stored DOE LLW. Suggestions are made about improving LLW disposal data.

  6. Remote-Handled Low-Level Waste Disposal Project Code of Record

    SciTech Connect (OSTI)

    S.L. Austad, P.E.; L.E. Guillen, P.E.; C. W. McKnight, P.E.; D. S. Ferguson, P.E.

    2012-06-01

    The Remote-Handled Low-Level Waste (LLW) Disposal Project addresses an anticipated shortfall in remote-handled LLW disposal capability following cessation of operations at the existing facility, which will continue until it is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). Development of a new onsite disposal facility will provide necessary remote-handled LLW disposal capability and will ensure continuity of operations that generate remote-handled LLW. This report documents the Code of Record for design of a new LLW disposal capability. The report is owned by the Design Authority, who can authorize revisions and exceptions. This report will be retained for the lifetime of the facility.

  7. Remote-Handled Low-Level Waste Disposal Project Code of Record

    SciTech Connect (OSTI)

    S.L. Austad, P.E.; L.E. Guillen, P.E.; C. W. McKnight, P.E.; D. S. Ferguson, P.E.

    2011-01-01

    The Remote-Handled Low-Level Waste (LLW) Disposal Project addresses an anticipated shortfall in remote-handled LLW disposal capability following cessation of operations at the existing facility, which will continue until it is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). Development of a new onsite disposal facility, the highest ranked alternative, will provide necessary remote-handled LLW disposal capability and will ensure continuity of operations that generate remote-handled LLW. This report documents the Code of Record for design of a new LLW disposal capability. The report is owned by the Design Authority, who can authorize revisions and exceptions. This report will be retained for the lifetime of the facility.

  8. Remote-Handled Low-Level Waste Disposal Project Code of Record

    SciTech Connect (OSTI)

    Austad, S. L.; Guillen, L. E.; McKnight, C. W.; Ferguson, D. S.

    2015-04-01

    The Remote-Handled Low-Level Waste (LLW) Disposal Project addresses an anticipated shortfall in remote-handled LLW disposal capability following cessation of operations at the existing facility, which will continue until it is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). Development of a new onsite disposal facility will provide necessary remote-handled LLW disposal capability and will ensure continuity of operations that generate remote-handled LLW. This report documents the Code of Record for design of a new LLW disposal capability. The report is owned by the Design Authority, who can authorize revisions and exceptions. This report will be retained for the lifetime of the facility.

  9. Remote-Handled Low-Level Waste Disposal Project Code of Record

    SciTech Connect (OSTI)

    S.L. Austad, P.E.; L.E. Guillen, P.E.; C. W. McKnight, P.E.; D. S. Ferguson, P.E.

    2014-06-01

    The Remote-Handled Low-Level Waste (LLW) Disposal Project addresses an anticipated shortfall in remote-handled LLW disposal capability following cessation of operations at the existing facility, which will continue until it is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). Development of a new onsite disposal facility will provide necessary remote-handled LLW disposal capability and will ensure continuity of operations that generate remote-handled LLW. This report documents the Code of Record for design of a new LLW disposal capability. The report is owned by the Design Authority, who can authorize revisions and exceptions. This report will be retained for the lifetime of the facility.

  10. Remote-Handled Low-Level Waste Disposal Project Code of Record

    SciTech Connect (OSTI)

    S.L. Austad, P.E.; L.E. Guillen, P.E.; C. W. McKnight, P.E.; D. S. Ferguson, P.E.

    2012-04-01

    The Remote-Handled Low-Level Waste (LLW) Disposal Project addresses an anticipated shortfall in remote-handled LLW disposal capability following cessation of operations at the existing facility, which will continue until it is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). Development of a new onsite disposal facility will provide necessary remote-handled LLW disposal capability and will ensure continuity of operations that generate remote-handled LLW. This report documents the Code of Record for design of a new LLW disposal capability. The report is owned by the Design Authority, who can authorize revisions and exceptions. This report will be retained for the lifetime of the facility.

  11. Remote-Handled Low-Level Waste Disposal Project Code of Record

    SciTech Connect (OSTI)

    S.L. Austad, P.E.; L.E. Guillen, P.E.; C. W. McKnight, P.E.; D. S. Ferguson, P.E.

    2011-04-01

    The Remote-Handled Low-Level Waste (LLW) Disposal Project addresses an anticipated shortfall in remote-handled LLW disposal capability following cessation of operations at the existing facility, which will continue until it is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). Development of a new onsite disposal facility, the highest ranked alternative, will provide necessary remote-handled LLW disposal capability and will ensure continuity of operations that generate remote-handled LLW. This report documents the Code of Record for design of a new LLW disposal capability. The report is owned by the Design Authority, who can authorize revisions and exceptions. This report will be retained for the lifetime of the facility.

  12. The white dwarfs within 25 pc of the Sun: Kinematics and spectroscopic subtypes

    SciTech Connect (OSTI)

    Sion, Edward M.; McCook, George P.; Wasatonic, Richard; Myszka, Janine; Holberg, J. B.; Oswalt, Terry D. E-mail: george.mccook@villanova.edu E-mail: janine.myszka@villanova.edu E-mail: toswalt@fit.edu

    2014-06-01

    We present the fractional distribution of spectroscopic subtypes, range and distribution of surface temperatures, and kinematical properties of the white dwarfs (WDs) within 25 pc of the Sun. There is no convincing evidence of halo WDs in the total 25 pc sample of 224 WDs. There is also little to suggest the presence of genuine thick disk subcomponent members within 25 pc. It appears that the entire 25 pc sample likely belongs to the thin disk. We also find no significant kinematic differences with respect to spectroscopic subtypes. The total DA to non-DA ratio of the 25 pc sample is 1.8, a manifestation of deepening envelope convection, which transforms DA stars with sufficiently thin H surface layers into non-DAs. We compare this ratio with the results of other studies. We find that at least 11% of the WDs within 25 pc of the Sun (the DAZ and DZ stars) have photospheric metals that likely originate from accretion of circumstellar material (debris disks) around them. If this interpretation is correct, then it suggests the possibility that a similar percentage have planets, asteroid-like bodies, or debris disks orbiting them. Our volume-limited sample reveals a pileup of DC WDs at the well-known cutoff in DQ WDs at T {sub eff} ? 6000 K. Mindful of small number statistics, we speculate on its possible evolutionary significance. We find that the incidence of magnetic WDs in the 25 pc sample is at least 8% in our volume-limited sample, dominated by cool WDs. We derive approximate formation rates of DB and DQ degenerates and present a preliminary test of the evolutionary scenario that all cooling DB stars become DQ WDs via helium convective dredge-up with the diffusion tail of carbon extending upward from their cores.

  13. Life-Cycle Cost and Risk Analysis of Alternative Configurations for Shipping Low-Level Radioactive Waste to the Nevada Test Site

    SciTech Connect (OSTI)

    PM Daling; SB Ross; BM Biwer

    1999-12-17

    The Nevada Test Site (NTS) is a major receiver of low-level radioactive waste (LLW) for disposal. Currently, all LLW received at NTS is shipped by truck. The trucks use highway routes to NTS that pass through the Las Vegas Valley and over Hoover Dam, which is a concern of local stakeholder groups in the State of Nevada. Rail service offers the opportunity to reduce transportation risks and costs, according to the Waste Management Programmatic Environmental Impact Statement (WM-PEIS). However, NTS and some DOE LLW generator sites are not served with direct rail service so intermodal transport is under consideration. Intermodal transport involves transport via two modes, in this case truck and rail, from the generator sites to NTS. LLW shipping containers would be transferred between trucks and railcars at intermodal transfer points near the LLW generator sites, NTS, or both. An Environmental Assessment (EA)for Intermodal Transportation of Low-Level Radioactive Waste to the Nevada Test Site (referred to as the NTSIntermodal -M) has been prepared to determine whether there are environmental impacts to alterations to the current truck routing or use of intermodal facilities within the State of Nevada. However, an analysis of the potential impacts outside the State of Nevada are not addressed in the NTS Intermodal EA. This study examines the rest of the transportation network between LLW generator sites and the NTS and evaluates the costs, risks, and feasibility of integrating intermodal shipments into the LLW transportation system. This study evaluates alternative transportation system configurations for NTS approved and potential generators based on complex-wide LLW load information. Technical judgments relative to the availability of DOE LLW generators to ship from their sites by rail were developed. Public and worker risk and life-cycle cost components are quantified. The study identifies and evaluates alternative scenarios that increase the use of rail (intermodal where needed) to transport LLW from generator sites to NTS.

  14. Project Execution Plan for the Remote Handled Low-Level Waste Disposal Project

    SciTech Connect (OSTI)

    Danny Anderson

    2014-07-01

    As part of ongoing cleanup activities at the Idaho National Laboratory (INL), closure of the Radioactive Waste Management Complex (RWMC) is proceeding under the Comprehensive Environmental Response, Compensation, and Liability Act (42 USC 9601 et seq. 1980). INL-generated radioactive waste has been disposed of at RWMC since 1952. The Subsurface Disposal Area (SDA) at RWMC accepted the bulk of INLs contact and remote-handled low-level waste (LLW) for disposal. Disposal of contact-handled LLW and remote-handled LLW ion-exchange resins from the Advanced Test Reactor in the open pit of the SDA ceased September 30, 2008. Disposal of remote-handled LLW in concrete disposal vaults at RWMC will continue until the facility is full or until it must be closed in preparation for final remediation of the SDA (approximately at the end of fiscal year FY 2017). The continuing nuclear mission of INL, associated ongoing and planned operations, and Naval spent fuel activities at the Naval Reactors Facility (NRF) require continued capability to appropriately dispose of contact and remote handled LLW. A programmatic analysis of disposal alternatives for contact and remote-handled LLW generated at INL was conducted by the INL contractor in Fiscal Year 2006; subsequent evaluations were completed in Fiscal Year 2007. The result of these analyses was a recommendation to the Department of Energy (DOE) that all contact-handled LLW generated after September 30, 2008, be disposed offsite, and that DOE proceed with a capital project to establish replacement remote-handled LLW disposal capability. An analysis of the alternatives for providing replacement remote-handled LLW disposal capability has been performed to support Critical Decision-1. The highest ranked alternative to provide this required capability has been determined to be the development of a new onsite remote-handled LLW disposal facility to replace the existing remote-handled LLW disposal vaults at the SDA. Several offsite DOE and commercial disposal options exist for contact-handled LLW; however, offsite disposal options are either not currently available (i.e., commercial disposal facilities), practical, or cost-effective for all remote-handled LLW streams generated at INL. Offsite disposal of all INL and tenant-generated remote-handled waste is further complicated by issues associated with transporting highly radioactive waste in commerce; and infrastructure and processing changes at the generating facilities, specifically NRF, that would be required to support offsite disposal. The INL Remote-Handled LLW Disposal Project will develop a new remote handled LLW disposal facility to meet mission-critical, remote-handled LLW disposal needs. A formal DOE decision to proceed with the project has been made in accordance with the requirements of National Environmental Policy Act (42 USC 4321 et seq.). Remote-handled LLW is generated from nuclear programs conducted at INL, including spent nuclear fuel handling and operations at NRF and operations at the Advanced Test Reactor. Remote-handled LLW also will be generated by new INL programs and from segregation and treatment (as necessary) of remote handled scrap and waste currently stored in the Radioactive Scrap and Waste Facility at the Materials and Fuels Complex.

  15. Methods for verifying compliance with low-level radioactive waste acceptance criteria

    SciTech Connect (OSTI)

    NONE

    1993-09-01

    This report summarizes the methods that are currently employed and those that can be used to verify compliance with low-level radioactive waste (LLW) disposal facility waste acceptance criteria (WAC). This report presents the applicable regulations representing the Federal, State, and site-specific criteria for accepting LLW. Typical LLW generators are summarized, along with descriptions of their waste streams and final waste forms. General procedures and methods used by the LLW generators to verify compliance with the disposal facility WAC are presented. The report was written to provide an understanding of how a regulator could verify compliance with a LLW disposal facility`s WAC. A comprehensive study of the methodology used to verify waste generator compliance with the disposal facility WAC is presented in this report. The study involved compiling the relevant regulations to define the WAC, reviewing regulatory agency inspection programs, and summarizing waste verification technology and equipment. The results of the study indicate that waste generators conduct verification programs that include packaging, classification, characterization, and stabilization elements. The current LLW disposal facilities perform waste verification steps on incoming shipments. A model inspection and verification program, which includes an emphasis on the generator`s waste application documentation of their waste verification program, is recommended. The disposal facility verification procedures primarily involve the use of portable radiological survey instrumentation. The actual verification of generator compliance to the LLW disposal facility WAC is performed through a combination of incoming shipment checks and generator site audits.

  16. Preliminary Hazard Analysis for the Remote-Handled Low-Level Waste Disposal Facility

    SciTech Connect (OSTI)

    Lisa Harvego; Mike Lehto

    2010-02-01

    The need for remote handled low level waste (LLW) disposal capability has been identified. A new onsite, remote-handled LLW disposal facility has been identified as the highest ranked alternative for providing continued, uninterrupted remote-handled LLW disposal capability for remote-handled LLW that is generated as part of the nuclear mission of the Idaho National Laboratory and from spent nuclear fuel processing activities at the Naval Reactors Facility. Historically, this type of waste has been disposed of at the Radioactive Waste Management Complex. Disposal of remote-handled LLW in concrete disposal vaults at the Radioactive Waste Management Complex will continue until the facility is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). This document supports the conceptual design for the proposed remote-handled LLW disposal facility by providing an initial nuclear facility hazard categorization and by identifying potential hazards for processes associated with onsite handling and disposal of remote-handled LLW.

  17. Preliminary Hazard Analysis for the Remote-Handled Low-Level Waste Disposal Project

    SciTech Connect (OSTI)

    Lisa Harvego; Mike Lehto

    2010-10-01

    The need for remote handled low level waste (LLW) disposal capability has been identified. A new onsite, remote-handled LLW disposal facility has been identified as the highest ranked alternative for providing continued, uninterrupted remote-handled LLW disposal capability for remote-handled LLW that is generated as part of the nuclear mission of the Idaho National Laboratory and from spent nuclear fuel processing activities at the Naval Reactors Facility. Historically, this type of waste has been disposed of at the Radioactive Waste Management Complex. Disposal of remote-handled LLW in concrete disposal vaults at the Radioactive Waste Management Complex will continue until the facility is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). This document supports the conceptual design for the proposed remote-handled LLW disposal facility by providing an initial nuclear facility hazard categorization and by identifying potential hazards for processes associated with onsite handling and disposal of remote-handled LLW.

  18. Preliminary Hazard Analysis for the Remote-Handled Low-Level Waste Disposal Facility

    SciTech Connect (OSTI)

    Lisa Harvego; Mike Lehto

    2010-05-01

    The need for remote handled low level waste (LLW) disposal capability has been identified. A new onsite, remote-handled LLW disposal facility has been identified as the highest ranked alternative for providing continued, uninterrupted remote-handled LLW disposal capability for remote-handled LLW that is generated as part of the nuclear mission of the Idaho National Laboratory and from spent nuclear fuel processing activities at the Naval Reactors Facility. Historically, this type of waste has been disposed of at the Radioactive Waste Management Complex. Disposal of remote-handled LLW in concrete disposal vaults at the Radioactive Waste Management Complex will continue until the facility is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). This document supports the conceptual design for the proposed remote-handled LLW disposal facility by providing an initial nuclear facility hazard categorization and by identifying potential hazards for processes associated with onsite handling and disposal of remote-handled LLW.

  19. Risk Analysis and Adaptive Response Planning for Water Distribution Systems Contamination Emergency Management

    E-Print Network [OSTI]

    Rasekh, Amin

    2012-10-19

    of terrorism, several aspects of emergency management for WDSs remain at an undeveloped stage. A set of methods is developed to analyze the risk and consequences of WDS contamination events and develop emergency response support tools. Monte Carlo...

  20. Dynamical Tides in Compact White Dwarf Binaries: Influence of Rotation

    E-Print Network [OSTI]

    Fuller, Jim

    2014-01-01

    Tidal interactions play an important role in the evolution and ultimate fate of compact white dwarf (WD) binaries. Not only do tides affect the pre-merger state (such as temperature and rotation rate) of the WDs, but they may also determine which systems merge and which undergo stable mass transfer. In this paper, we attempt to quantify the effects of rotation on tidal angular momentum transport in binary stars, with specific calculations applied to WD stellar models. We incorporate the effect of rotation using the traditional approximation, in which the dynamically excited gravity waves within the WDs are transformed into gravito-inertial Hough waves. The Coriolis force has only a minor effect on prograde gravity waves, and previous results predicting the tidal spin-up and heating of inspiraling WDs are not significantly modified. However, rotation strongly alters retrograde gravity waves and inertial waves, with important consequences for the tidal spin-down of accreting WDs. We identify new dynamical tidal...

  1. Imprint of modified Einstein's gravity on white dwarfs: Unifying type Ia supernovae

    E-Print Network [OSTI]

    Das, Upasana

    2015-01-01

    We establish the importance of modified Einstein's gravity (MG) in white dwarfs (WDs) for the first time in the literature. We show that MG leads to significantly sub- and super-Chandrasekhar limiting mass WDs, depending on a single model parameter. However, conventional WDs on approaching Chandrasekhar's limit are expected to trigger type Ia supernovae (SNeIa), a key to unravel the evolutionary history of the universe. Nevertheless, observations of several peculiar, under- and over-luminous SNeIa argue for the limiting mass widely different from Chandrasekhar's limit. Explosions of MG induced sub- and super-Chandrasekhar limiting mass WDs explain under- and over-luminous SNeIa respectively, thus unifying these two apparently disjoint sub-classes. Our discovery questions both the global validity of Einstein's gravity and the uniqueness of Chandrasekhar's limit.

  2. A Risk-based Optimization Modeling Framework for Mitigating Fire Events for Water and Fire Response Infrastructures

    E-Print Network [OSTI]

    Kanta, Lufthansa Rahman

    2011-02-22

    ) minimizing the cost of mitigation. Third, a stochastic modeling approach is developed to assess urban fire risk for the coupled water distribution and fire response systems that includes probabilistic expressions for building ignition, WDS failure, and wind...

  3. Greater-than-Class C low-level radioactive waste characterization: Estimated volumes, radionuclide activities, and other characteristics. Revision 1

    SciTech Connect (OSTI)

    Not Available

    1994-09-01

    The Department of Energy`s (DOE`s) planning for the disposal of greater-than-Class C low-level radioactive waste (GTCC LLW) requires characterization of the waste. This report estimates volumes, radionuclide activities, and waste forms of GTCC LLW to the year 2035. It groups the waste into four categories, representative of the type of generator or holder of the waste: Nuclear Utilities, Sealed Sources, DOE-Held, and Other Generator. GTCC LLW includes activated metals (activation hardware from reactor operation and decommissioning), process wastes (i.e., resins, filters, etc.), sealed sources, and other wastes routinely generated by users of radioactive material. Estimates reflect the possible effect that packaging and concentration averaging may have on the total volume of GTCC LLW. Possible GTCC mixed LLW is also addressed. Nuclear utilities will probably generate the largest future volume of GTCC LLW with 65--83% of the total volume. The other generators will generate 17--23% of the waste volume, while GTCC sealed sources are expected to contribute 1--12%. A legal review of DOE`s obligations indicates that the current DOE-Held wastes described in this report will not require management as GTCC LLW because of the contractual circumstances under which they were accepted for storage. This report concludes that the volume of GTCC LLW should not pose a significant management problem from a scientific or technical standpoint. The projected volume is small enough to indicate that a dedicated GTCC LLW disposal facility may not be justified. Instead, co-disposal with other waste types is being considered as an option.

  4. Melter system technology testing for Hanford Site low-level tankwaste vitrification

    SciTech Connect (OSTI)

    Wilson, C.N.

    1996-05-03

    Following revisions to the Tri-Party Agreement for Hanford Site cleanup, which specified vitrification for Complete melter feasibility and system operability immobilization of the low-level waste (LLW) tests, select reference melter(s), and establish reference derived from retrieval and pretreatment of the radioactive LLW glass formulation that meets complete systems defense wastes stored in 177 underground tanks, commercial requirements (June 1996). Available melter technologies were tested during 1994 to 1995 as part of a multiphase program to select reference Submit conceptual design and initiate definitive design technologies for the new LLW vitrification mission.

  5. Low-level waste inventory, characteristics, generation, and facility assessment for treatment, storage, and disposal alternatives considered in the US Department of Energy waste management programmatic environmental impact statement

    SciTech Connect (OSTI)

    Goyette, M.L.; Dolak, D.A.

    1996-12-01

    This report provides technical support information for use in analyzing environmental impacts associated with U.S. Department of Energy (DOE) low-level radioactive waste (LLW) management alternatives in the Waste-Management (WM) Programmatic Environmental Impact Statement (PEIS). Waste loads treated and disposed of for each of the LLW alternatives considered in the DOE WM PEIS are presented. Waste loads are presented for DOE Waste Management (WM) wastes, which are generated from routine operations. Radioactivity concentrations and waste quantities for treatment and disposal under the different LLW alternatives are described for WM waste. 76 refs., 14 figs., 42 tabs.

  6. Sulfur polymer cement as a low-level waste glass matrix encapsulant. Part 1: Thermal processing

    SciTech Connect (OSTI)

    Sliva, P.; Peng, Y.B.; Bunnell, L.R.; Peeler, D.K.; Feng, X.; Martin, P.; Turner, P.J. [Pacific Northwest National Lab., Richland, WA (United States)

    1996-08-01

    Sulfur polymer cement (SPC) is a candidate material to encapsulate low-level waste (LLW) glass. Molten SPC will be poured into a LLW glass cullet-filled canister, surrounding the glass to act as an additional barrier to groundwater intrusion. This paper covers the first part of a study performed at Pacific Northwest National Laboratory concerned with the fundamental aspects of embedding LLW glass in SPC. Part one is a study of the SPC itself. Variations in SPC properties are discussed, especially in relation to long-term stability and controlling crystallization in a cooling canister.

  7. Environmental Assessment Idaho National Engineering Laboratory, low-level and mixed waste processing

    SciTech Connect (OSTI)

    Not Available

    1994-06-01

    The Department of Energy (DOE) has prepared an environmental assessment (EA), DOE/EA-0843, for the Idaho National Engineering Laboratory (INEL) low-level and mixed waste processing. The original proposed action, as reviewed in this EA, was (1) to incinerate INEL`s mixed low-level waste (MLLW) at the Waste Experimental Reduction Facility (WERF); (2) reduce the volume of INEL generated low-level waste (LLW) through sizing, compaction, and stabilization at the WERF; and (3) to ship INEL LLW to a commercial incinerator for supplemental LLW volume reduction.

  8. Super-Eddington wind scenario for the progenitors of type Ia supernovae: Accreting He-rich matter onto white dwarfs

    E-Print Network [OSTI]

    Wang, Bo; Ma, Xin; Liu, Dongdong; Cui, Xiao; Han, Zhanwen

    2015-01-01

    Supernovae of type Ia (SNe Ia) are believed to be thermonuclear explosions of carbon-oxygen white dwarfs (CO WDs). However, the mass accretion process onto CO WDs is still not completely understood. In this paper, we study the accretion of He-rich matter onto CO WDs and explore a scenario in which a strong wind forms on the surface of the WD if the total luminosity exceeds the Eddington limit. Using a stellar evolution code called modules for experiments in stellar astrophysics (MESA), we simulated the He accretion process onto CO WDs for WDs with masses of 0.6-1.35Msun and various accretion rates of 10^{-8}-10^{-5}Msun/yr. If the contribution of the total luminosity is included when determining the Eddington accretion rate, then a super-Eddington wind could be triggered at relatively lower accretion rates than those of previous studies based on steady-state models. The super-Eddington wind can prevent the WDs with high accretion rates from evolving into red-giant-like He stars. We found that the contribution...

  9. Low-level radioactive waste management: transitioning to off-site disposal at Los Alamos National Laboratory

    SciTech Connect (OSTI)

    Dorries, Alison M

    2010-11-09

    Facing the closure of nearly all on-site management and disposal capability for low-level radioactive waste (LLW), Los Alamos National Laboratory (LANL) is making ready to ship the majority of LLW off-site. In order to ship off-site, waste must meet the Treatment, Storage, and Disposal Facility's (TSDF) Waste Acceptance Criteria (WAC). In preparation, LANL's waste management organization must ensure LANL waste generators characterize and package waste compliantly and waste characterization documentation is complete and accurate. Key challenges that must be addressed to successfully make the shift to off-site disposal of LLW include improving the detail, accuracy, and quality of process knowledge (PK) and acceptable knowledge (AK) documentation, training waste generators and waste management staff on the higher standard of data quality and expectations, improved WAC compliance for off-site facilities, and enhanced quality assurance throughout the process. Certification of LANL generators will allow direct off-site shipping of LLW from their facilities.

  10. Biological and Physical Assessment of Streams in Northern California: Evaluating the Effects of Global Change and Human Disturbance

    E-Print Network [OSTI]

    Lawrence, Justin Earl

    2011-01-01

    as large woody debris (LWD), whereas living large wood washave established the importance of LWD on pool formation in2) whether it was dead (LWD) or living (LLW); 3) the length

  11. Potential co-disposal of greater-than-class C low-level radioactive waste with Department of Energy special case waste - greater-than-class C low-level waste management program

    SciTech Connect (OSTI)

    Allred, W.E.

    1994-09-01

    This document evaluates the feasibility of co-disposing of greater-than-Class C low-level radioactive waste (GTCC LLW) with U.S. Department of Energy (DOE) special case waste (SCW). This document: (1) Discusses and evaluates key issues concerning co-disposal of GTCC LLW with SCW. This includes examining these issues in terms of regulatory concerns, technical feasibility, and economics; (2) Examines advantages and disadvantages of such co-disposal; and (3) Makes recommendations. Research and analysis of the issues presented in this report indicate that it would be technically and economically feasible to co-dispose of GTCC LLW with DOE SCW. However, a dilemma will likely arise in the current division of regulatory responsibilities between the U.S. Nuclear Regulatory Commission and DOE (i.e., current requirement for disposal of GTCC LLW in a facility licensed by the Nuclear Regulatory Commission). DOE SCW is currently not subject to this licensing requirement.

  12. Summary report. Low-level radioactive waste management activities in the states and compacts. Volume 4, No. 2

    SciTech Connect (OSTI)

    NONE

    1996-08-01

    `Low-Level Radioactive Waste Management Activities in the States and Compacts` is a supplement to `LLW Notes` and is distributed periodically by Afton Associates, Inc. to state, compact and federal officials that receive `LLW Notes`. The Low-Level Radioactive Waste Forum (LLW Forum) is an association of state and compact representatives, appointed by governors and compact commissions, established to facilitate state and compact implementation of the Low- Level Radioactive Waste Policy Act of 1980 and the Low-Level Radioactive Waste Policy Amendments Act of 1985 and to promote the objectives of low-level radioactive waste regional compacts. The LLW Forum provides an opportunity for state and compact officials to share information with one another and to exchange views with officials of federal agencies and other interested parties.

  13. Summary report, low-level radioactive waste management activities in the states and compacts. Vol. 4. No. 1

    SciTech Connect (OSTI)

    NONE

    1996-01-01

    `Low-Level Radioactive Waste Management Activities in the States and Compacts` is a supplement to `LLW Notes` and is distributed periodically by Afton Associates, Inc. to state, compact and federal officials that receive `LLW Notes`. The Low-Level Radioactive Waste Forum (LLW Forum) is an association of state and compact representatives, appointed by governors and compact commissions, established to facilitate state and compact implementation of the Low- Level Radioactive Waste Policy Act of 1980 and the Low-Level Radioactive Waste Policy Amendments Act of 1985 and to promote the objectives of low-level radioactive waste regional compacts. The LLW Forum provides an opportunity for state and compact officials to share information with one another and to exchange views with officials of federal agencies and other interested parties.

  14. EA-0843: Idaho National Engineering Laboratory Low-Level and Mixed Waste Processing, Idaho Falls, Idaho

    Broader source: Energy.gov [DOE]

    This EA evaluates the environmental impacts of a proposal to (1) reduce the volume of the U.S. Department of Energy's Idaho National Engineering Laboratory's (INEL) generated low-level waste (LLW)...

  15. Greater-than-Class C low-level radioactive waste characterization. Appendix E-4: Packaging factors for greater-than-Class C low-level radioactive waste

    SciTech Connect (OSTI)

    Quinn, G.; Grant, P.; Winberg, M.; Williams, K.

    1994-09-01

    This report estimates packaging factors for several waste types that are potential greater-than-Class C (GTCC) low-level radioactive waste (LLW). The packaging factor is defined as the volume of a GTCC LLW disposal container divided by the as-generated or ``unpackaged`` volume of the waste loaded into the disposal container. Packaging factors reflect any processes that reduce or increase an original unpackaged volume of GTCC LLW, the volume inside a waste container not occupied by the waste, and the volume of the waste container itself. Three values are developed that represent (a) the base case or most likely value for a packaging factor, (b) a high case packaging factor that corresponds to the largest anticipated disposal volume of waste, and (c) a low case packaging factor for the smallest volume expected. GTCC LLW is placed in three categories for evaluation in this report: activated metals, sealed sources, and all other waste.

  16. DOE to Weigh Alternatives for Greater Than Class C Low-Level...

    Broader source: Energy.gov (indexed) [DOE]

    where to safely dispose of GTCC LLW that is currently stored at commercial nuclear power plants and other generator sites across the country. The Energy Policy Act of 2005...

  17. United States Office of Radiation and EPA 402-R-00-007 Environmental Protection Indoor Air August 2000

    E-Print Network [OSTI]

    -level radioactive waste (LLW). Since the mainframe versions of the PRESTO-EPA-CPG and PRESTO-EPA-POP models were to radioactivity through atmospheric, groundwater, and surface water transport pathways; through drinking water

  18. Radioactive Materials Product Stewardship

    E-Print Network [OSTI]

    Radioactive Materials Product Stewardship ABackground Report for the National Dialogue...................................................................................................26 Low Level Waste (LLW) Disposal Regulations on Radioactive Materials Product Stewardship Prepared by the: Product Stewardship Institute University

  19. Evaluation of Exothermic Reactions from Bulk-Vitrification Melter Feeds Containing Cellulose

    SciTech Connect (OSTI)

    Scheele, Randall D.; McNamara, Bruce K.; Bagaasen, Larry M.; Bos, Stanley J.; Kozelisky, Anne E.; Berry, Pam

    2007-06-25

    PNNL has demonstrated that cellulose effectively reduces the amount of molten ionic salt during Bulk Vitrification of simulated Hanford Low Level Waste (LLW). To address concerns about the potential reactivity of cellulose-LLW, PNNL used thermogravimetric analysis, differential thermal analysis, and accelerating rate calorimetry to determine in these preliminary studies that these mixtures will support a self-sustaining reaction if heated to 110C at adiabatic conditions. Additional testing is recommended.

  20. Greater-than-Class C low-level radioactive waste characterization: Estimated volumes, radionuclide activities, and other characteristics

    SciTech Connect (OSTI)

    Hulse, R.A.

    1991-08-01

    Planning for storage or disposal of greater-than-Class C low-level radioactive waste (GTCC LLW) requires characterization of that waste to estimate volumes, radionuclide activities, and waste forms. Data from existing literature, disposal records, and original research were used to estimate the characteristics and project volumes and radionuclide activities to the year 2035. GTCC LLW is categorized as: nuclear utilities waste, sealed sources waste, DOE-held potential GTCC LLW; and, other generator waste. It has been determined that the largest volume of those wastes, approximately 57%, is generated by nuclear power plants. The Other Generator waste category contributes approximately 10% of the total GTCC LLW volume projected to the year 2035. Waste held by the Department of Energy, which is potential GTCC LLW, accounts for nearly 33% of all waste projected to the year 2035; however, no disposal determination has been made for that waste. Sealed sources are less than 0.2% of the total projected volume of GTCC LLW.

  1. Final environmental assessment for off-site transportation of low-level waste from four California sites under the management of the U.S. Department of Energy Oakland Operations Office

    SciTech Connect (OSTI)

    NONE

    1997-10-01

    The Department of Energy Oakland Operations Office (DOE/OAK) manages sites within California that generate Low Level Waste (LLW) in the course or routine site operations. It is the preference of the DOE to dispose of LLW at federally owned and DOE-operated disposal facilities; however, in some circumstances DOE Headquarters has determined that disposal at commercial facilities is appropriate, as long as the facility meets all regulatory requirements for the acceptance and disposal of LLW, including the passage of a DOE audit to determine the adequacy of the disposal site. The DOE would like to ship LLW from four DOE/OAK sites in California which generate LLW, to NRC-licensed commercial nuclear waste disposal facilities such as Envirocare in Clive, Utah and Chem Nuclear in Barnwell, South Carolina. Transportation impacts for shipment of LLW and MLLW from DOE Oakland sites to other DOE sites was included in the impacts identified in the Department`s Waste Management Programmatic Environmental Impact Statement (WM-PEIS), published in May, 1997, and determined to be low. The low impacts for shipment to commercial sites identified herein is consistent with the WM-PEIS results.

  2. Conceptual Safety Design Report for the Remote Handled Low-Level Waste Disposal Facility

    SciTech Connect (OSTI)

    Boyd D. Christensen

    2010-02-01

    A new onsite, remote-handled LLW disposal facility has been identified as the highest ranked alternative for providing continued, uninterrupted remote-handled LLW disposal for remote-handled LLW from the Idaho National Laboratory and for spent nuclear fuel processing activities at the Naval Reactors Facility. Historically, this type of waste has been disposed of at the Radioactive Waste Management Complex. Disposal of remote-handled LLW in concrete disposal vaults at the Radioactive Waste Management Complex will continue until the facility is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). This conceptual safety design report supports the design of a proposed onsite remote-handled LLW disposal facility by providing an initial nuclear facility hazard categorization, by identifying potential hazards for processes associated with onsite handling and disposal of remote-handled LLW, by evaluating consequences of postulated accidents, and by discussing the need for safety features that will become part of the facility design.

  3. Conceptual Safety Design Report for the Remote Handled Low-Level Waste Disposal Facility

    SciTech Connect (OSTI)

    Boyd D. Christensen

    2010-05-01

    A new onsite, remote-handled LLW disposal facility has been identified as the highest ranked alternative for providing continued, uninterrupted remote-handled LLW disposal for remote-handled LLW from the Idaho National Laboratory and for spent nuclear fuel processing activities at the Naval Reactors Facility. Historically, this type of waste has been disposed of at the Radioactive Waste Management Complex. Disposal of remote-handled LLW in concrete disposal vaults at the Radioactive Waste Management Complex will continue until the facility is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). This conceptual safety design report supports the design of a proposed onsite remote-handled LLW disposal facility by providing an initial nuclear facility hazard categorization, by identifying potential hazards for processes associated with onsite handling and disposal of remote-handled LLW, by evaluating consequences of postulated accidents, and by discussing the need for safety features that will become part of the facility design.

  4. 3D Model Atmospheres for Extremely Low-Mass White Dwarfs

    E-Print Network [OSTI]

    Tremblay, P -E; Kilic, M; Ludwig, H -G; Steffen, M; Freytag, B; Hermes, J J

    2015-01-01

    We present an extended grid of mean three-dimensional (3D) spectra for low-mass, pure-hydrogen atmosphere DA white dwarfs (WDs). We use CO5BOLD radiation-hydrodynamics 3D simulations covering Teff = 6000-11,500 K and logg = 5-6.5 (cgs units) to derive analytical functions to convert spectroscopically determined 1D temperatures and surface gravities to 3D atmospheric parameters. Along with the previously published 3D models, the 1D to 3D corrections are now available for essentially all known convective DA WDs (i.e., logg = 5-9). For low-mass WDs, the correction in temperature is relatively small (a few per cent at the most), but the surface gravities measured from the 3D models are lower by as much as 0.35 dex. We revisit the spectroscopic analysis of the extremely low-mass (ELM) WDs, and demonstrate that the 3D models largely resolve the discrepancies seen in the radius and mass measurements for relatively cool ELM WDs in eclipsing double WD and WD + milli-second pulsar binary systems. We also use the 3D cor...

  5. Design and operational considerations of United States commercial near-surface low-level radioactive waste disposal facilities

    SciTech Connect (OSTI)

    Birk, S.M.

    1997-10-01

    In accordance with the Low-Level Radioactive Waste Policy Amendments Act of 1985, states are responsible for providing for disposal of commercially generated low-level radioactive waste (LLW) within their borders. LLW in the US is defined as all radioactive waste that is not classified as spent nuclear fuel, high-level radioactive waste, transuranic waste, or by-product material resulting from the extraction of uranium from ore. Commercial waste includes LLW generated by hospitals, universities, industry, pharmaceutical companies, and power utilities. LLW generated by the country`s defense operations is the responsibility of the Federal government and its agency, the Department of Energy. The commercial LLRW disposal sites discussed in this report are located near: Sheffield, Illinois (closed); Maxey Flats, Kentucky (closed); Beatty, Nevada (closed); West Valley, New York (closed); Barnwell, South Carolina (operating); Richland, Washington (operating); Ward Valley, California, (proposed); Sierra Blanca, Texas (proposed); Wake County, North Carolina (proposed); and Boyd County, Nebraska (proposed). While some comparisons between the sites described in this report are appropriate, this must be done with caution. In addition to differences in climate and geology between sites, LLW facilities in the past were not designed and operated to today`s standards. This report summarizes each site`s design and operational considerations for near-surface disposal of low-level radioactive waste. The report includes: a description of waste characteristics; design and operational features; post closure measures and plans; cost and duration of site characterization, construction, and operation; recent related R and D activities for LLW treatment and disposal; and the status of the LLW system in the US.

  6. A search for planetary eclipses of white dwarfs in the Pan-STARRS1 medium-deep fields

    SciTech Connect (OSTI)

    Fulton, B. J.; Tonry, J. L.; Flewelling, H.; Burgett, W. S.; Chambers, K. C.; Hodapp, K. W.; Huber, M. E.; Kaiser, N.; Wainscoat, R. J.; Waters, C.

    2014-12-01

    We present a search for eclipses of ?1700 white dwarfs (WDs) in the Pan-STARRS1 medium-deep fields. Candidate eclipse events are selected by identifying low outliers in over 4.3 million light curve measurements. We find no short-duration eclipses consistent with being caused by a planetary size companion. This large data set enables us to place strong constraints on the close-in planet occurrence rates around WDs for planets as small as 2 R {sub ?}. Our results indicate that gas giant planets orbiting just outside the Roche limit are rare, occurring around less than 0.5% of WDs. Habitable-zone super-Earths and hot super-Earths are less abundant than similar classes of planets around main-sequence stars. These constraints provide important insight into the ultimate fate of the large population of exoplanets orbiting main-sequence stars.

  7. TIDAL NOVAE IN COMPACT BINARY WHITE DWARFS

    SciTech Connect (OSTI)

    Fuller, Jim; Lai Dong [Department of Astronomy, Cornell University, Ithaca, NY 14850 (United States)

    2012-09-01

    Compact binary white dwarfs (WDs) undergoing orbital decay due to gravitational radiation can experience significant tidal heating prior to merger. In these WDs, the dominant tidal effect involves the excitation of outgoing gravity waves in the inner stellar envelope and the dissipation of these waves in the outer envelope. As the binary orbit decays, the WDs are synchronized from outside in (with the envelope synchronized first, followed by the core). We examine the deposition of tidal heat in the envelope of a carbon-oxygen WD and study how such tidal heating affects the structure and evolution of the WD. We show that significant tidal heating can occur in the star's degenerate hydrogen layer. This layer heats up faster than it cools, triggering runaway nuclear fusion. Such 'tidal novae' may occur in all WD binaries containing a CO WD, at orbital periods between 5 minutes and 20 minutes, and precede the final merger by 10{sup 5}-10{sup 6} years.

  8. Contribution of White Dwarfs to Cluster Masses

    E-Print Network [OSTI]

    Ted von Hippel

    1998-02-14

    I present a literature search through 31 July 1997 of white dwarfs (WDs) in open and globular clusters. There are 36 single WDs and 5 WDs in binaries known among 13 open clusters, and 340 single WDs and 11 WDs in binaries known among 11 globular clusters. From these data I have calculated WD mass fractions for four open clusters (the Pleiades, NGC 2168, NGC 3532, and the Hyades) and one globular cluster (NGC 6121). I develop a simple model of cluster evolution that incorporates stellar evolution but not dynamical evolution to interpret the WD mass fractions. I augment the results of my simple model with N-body simulations incorporating stellar evolution (Terlevich 1987; de la Feunte Marcos 1996; Vesperini & Heggie 1997). I find that even though these clusters undergo moderate to strong kinematical evolution the WD mass fraction is relatively insensitive to kinematical evolution. By comparing the cluster mass functions to that of the Galactic disk, and incorporating plausibility arguments for the mass function of the Galactic halo, I estimate the WD mass fraction in these two populations. I assume the Galactic disk is ~10 Gyrs old (Winget et al. 1987; Liebert, Dahn, & Monet 1988; Oswalt et al. 1996) and that the Galactic halo is ~12 Gyrs old (Reid 1997b; Gratton et al. 1997; Chaboyer et al. 1998), although the WD mass fraction is insensitive to age in this range. I find that the Galactic halo should contain 8 to 9% (alpha = -2.35) or perhaps as much as 15 to 17% (alpha = -2.0) of its stellar mass in the form of WDs. The Galactic disk WD mass fraction should be 6 to 7% (alpha = -2.35), consistent with the empirical estimates of 3 to 7% (Liebert, Dahn, & Monet 1988; Oswalt et al. 1996). (abridged)

  9. Application of Probabilistic Performance Assessment Modeling for Optimization of Maintenance Studies for Low-Level Radioactive Waste Disposal Sites at the Nevada Test Site

    SciTech Connect (OSTI)

    Crowe, B.; Yucel, V.; Rawlinson, S.; Black, P.; Carilli, J.; DiSanza, F.

    2002-02-25

    The U.S. Department of Energy (DOE), National Nuclear Security Administration of the Nevada Operations Office (NNSA/NV) operates and maintains two active facilities on the Nevada Test Site (NTS) that dispose defense-generated low-level radioactive waste (LLW), mixed radioactive waste, and ''classified waste'' in shallow trenches and pits. The operation and maintenance of the LLW disposal sites are self-regulated by the DOE under DOE Order 435.1. This Order requires formal review of a performance assessment (PA) and composite analysis (CA; assessment of all interacting radiological sources) for each LLW disposal system followed by an active maintenance program that extends through and beyond the site closure program. The Nevada disposal facilities continue to receive NTS-generated LLW and defense-generated LLW from across the DOE complex. The PA/CAs for the sites have been conditionally approved and the facilities are now under a formal maintenance program that requires testing of conceptual models, quantifying and attempting to reduce uncertainty, and implementing confirmatory and long-term background monitoring, all leading to eventual closure of the disposal sites. To streamline and reduce the cost of the maintenance program, the NNSA/NV is converting the deterministic PA/CAs to probabilistic models using GoldSim, a probabilistic simulation computer code. The output of probabilistic models will provide expanded information supporting long-term decision objectives of the NTS disposal sites.

  10. Secondary Low-Level Waste Treatment Strategy Analysis

    SciTech Connect (OSTI)

    D.M. LaRue

    1999-05-25

    The objective of this analysis is to identify and review potential options for processing and disposing of the secondary low-level waste (LLW) that will be generated through operation of the Monitored Geologic Repository (MGR). An estimate of annual secondary LLW is generated utilizing the mechanism established in ''Secondary Waste Treatment Analysis'' (Reference 8.1) and ''Secondary Low-Level Waste Generation Rate Analysis'' (Reference 8.5). The secondary LLW quantities are based on the spent fuel and high-level waste (HLW) arrival schedule as defined in the ''Controlled Design Assumptions Document'' (CDA) (Reference 8.6). This analysis presents estimates of the quantities of LLW in its various forms. A review of applicable laws, codes, and standards is discussed, and a synopsis of those applicable laws, codes, and standards and their impacts on potential processing and disposal options is presented. The analysis identifies viable processing/disposal options in light of the existing laws, codes, and standards, and then evaluates these options in regard to: (1) Process and equipment requirements; (2) LLW disposal volumes; and (3) Facility requirements.

  11. Preliminary PCT data on glass formulations developed for Hanford Site low-level wastes

    SciTech Connect (OSTI)

    Feng, X.; Schweiger, M.J.; Hrma, P.R.; Palmer, S.E.; Smith, D.E.; Kim, D.; Gong, M.; Westsik, J.H. Jr.

    1995-09-01

    Tank wastes stored at the Hanford Site are to be separated into high-level and low-level waste (LLW) fractions and vitrified for disposal. The high content of Na{sub 2}O in the LLW, averaging about 80 wt% on an oxide basis, necessitates the development of durable high-sodium glasses. Pacific Northwest Laboratory (PNL) is providing glass formulations for the LLW vitrification program. The most important considerations for acceptable LLW waste glass compositions are the following: (1) the capability to incorporate high sodium content from LLW; (2) satisfactory long-term durability; and (3) proper processability, such as desired viscosity at melting temperature. To develop durable high-sodium glasses in a reasonably short time and to supply data needed for modeling the glass`s long-term performance, several short-term test methods, such as 7-day PCT (Product Consistency Test), 28-day Materials Characterization Center MCC-1 test, single-pass flow through test, and vapor-hydration test have been used. A long-term static test (up to one year using PCT) is also being performed for selected glasses. This data report includes only the PCT data available at the time of the publication.

  12. Development of low-level radioactive waste disposal capacity in the United States - progress or stalemate?

    SciTech Connect (OSTI)

    Devgun, J.S. [Argonne National Lab., IL (United States); Larson, G.S. [Midwest Low-Level Radioactive Waste Commission, St. Paul, MN (United States)

    1995-12-31

    It has been fifteen years since responsibility for the disposal of commercially generated low-level radioactive waste (LLW) was shifted to the states by the United States Congress through the Low-Level Radioactive Waste Policy Act of 1980 (LLRWPA). In December 1985, Congress revisited the issue and enacted the Low-Level Radioactive Waste Policy Amendments Act of 1985 (LLRWPAA). No new disposal sites have opened yet, however, and it is now evident that disposal facility development is more complex, time-consuming, and controversial than originally anticipated. For a nation with a large nuclear power industry, the lack of availability of LLW disposal capacity coupled with a similar lack of high-level radioactive waste disposal capacity could adversely affect the future viability of the nuclear energy option. The U.S. nuclear power industry, with 109 operating reactors, generates about half of the LLW shipped to commercial disposal sites and faces dwindling access to waste disposal sites and escalating waste management costs. The other producers of LLW - industries, government (except the defense related research and production waste), academic institutions, and medical institutions that account for the remaining half of the commercial LLW - face the same storage and cost uncertainties. This paper will summarize the current status of U.S. low-level radioactive waste generation and the status of new disposal facility development efforts by the states. The paper will also examine the factors that have contributed to delays, the most frequently suggested alternatives, and the likelihood of change.

  13. Gas generation from low-level radioactive waste: Concerns for disposal

    SciTech Connect (OSTI)

    Siskind, B.

    1992-01-01

    The Advisory Committee on Nuclear Waste (ACNW) has urged the Nuclear Regulatory Commission (NRC) to reexamine the topic of hydrogen gas generation from low-level radioactive waste (LLW) in closed spaces to ensure that the slow buildup of hydrogen from water-bearing wastes in sealed containers does not become a problem for long-term safe disposal. Brookhaven National Laboratory (BNL) has prepared a report, summarized in this paper, for the NRC to respond to these concerns. The paper discusses the range of values for G(H{sub 2}) reported for materials of relevance to LLW disposal; most of these values are in the range of 0.1 to 0.6. Most studies of radiolytic hydrogen generation indicate a leveling off of pressurization, probably because of chemical kinetics involving, in many cases, the radiolysis of water within the waste. Even if no leveling off occurs, realistic gas leakage rates (indicating poor closure by gaskets on drums and liners) will result in adequate relief of pressure for radiolytic gas generation from the majority of commercial sector LLW packages. Biodegradative gas generation, however, could pose a pressurization hazard even at realistic gas leakage rates. Recommendations include passive vents on LLW containers (as already specified for high integrity containers) and upper limits to the G values and/or the specific activity of the LLW.

  14. Gas generation from low-level radioactive waste: Concerns for disposal

    SciTech Connect (OSTI)

    Siskind, B.

    1992-04-01

    The Advisory Committee on Nuclear Waste (ACNW) has urged the Nuclear Regulatory Commission (NRC) to reexamine the topic of hydrogen gas generation from low-level radioactive waste (LLW) in closed spaces to ensure that the slow buildup of hydrogen from water-bearing wastes in sealed containers does not become a problem for long-term safe disposal. Brookhaven National Laboratory (BNL) has prepared a report, summarized in this paper, for the NRC to respond to these concerns. The paper discusses the range of values for G(H{sub 2}) reported for materials of relevance to LLW disposal; most of these values are in the range of 0.1 to 0.6. Most studies of radiolytic hydrogen generation indicate a leveling off of pressurization, probably because of chemical kinetics involving, in many cases, the radiolysis of water within the waste. Even if no leveling off occurs, realistic gas leakage rates (indicating poor closure by gaskets on drums and liners) will result in adequate relief of pressure for radiolytic gas generation from the majority of commercial sector LLW packages. Biodegradative gas generation, however, could pose a pressurization hazard even at realistic gas leakage rates. Recommendations include passive vents on LLW containers (as already specified for high integrity containers) and upper limits to the G values and/or the specific activity of the LLW.

  15. Hanford low-level vitrification melter testing -- Master list of data submittals

    SciTech Connect (OSTI)

    Hendrickson, D.W.

    1995-03-15

    The Westinghouse Hanford Company (WHC) is conducting a two-phased effort to evaluate melter system technologies for vitrification of liquid low-level radioactive waste (LLW) streams. The evaluation effort includes demonstration testing of selected glass melter technologies and technical reports regarding the applicability of the glass melter technologies to the vitrification of Hanford LLW tank waste. The scope of this document is to identify and list vendor document submittals in technology demonstration support of the Hanford Low-Level Waste Vitrification melter testing program. The scope of this document is limited to those documents responsive to the Statement of Work, accepted and issued by the LLW Vitrification Program. The purpose of such a list is to maintain configuration control of vendor supplied data and to enable ready access to, and application of, vendor supplied data in the evaluation of melter technologies for the vitrification of Hanford low-level tank wastes.

  16. Low-level waste vitrification phase 1 vendor test sample analysis data

    SciTech Connect (OSTI)

    Mast, E.S.

    1995-10-04

    A multi-phase program was initiated in 1994 to test commercially available melter technologies for the vitrification of the low-level waste (LLW) stream from defense wastes stored in underground tanks at the Hanford Site in southeastern Washington State. Phase 1 of the melter demonstration tests was performed in vendor test facilities using simulated LLW and was completed during FY-1995. Test samples taken during Phase 1 testing were analyzed by independent laboratories who reported the analyses results to Westinghouse Hanford Company for integration and evaluation. The reported analytical results were integrated into an electronic data base using Microsoft Excel*5.0. This report documents this data base as of the end of FY-1995, and is supplemental to the Phase 1 LLW melter testing summary report, WHC-SD-WM-ER-498, revision 0.

  17. Evaluation of melter technologies for vitrification of Hanford site low-level tank waste - phase 1 testing summary report

    SciTech Connect (OSTI)

    Wilson, C.N., Westinghouse Hanford

    1996-06-27

    Following negotiation of the fourth amendment to the Tri- Party Agreement for Hanford Site cleanup, commercially available melter technologies were tested during 1994 and 1995 for vitrification of the low-level waste (LLW) stream to be derived from retrieval and pretreatment of the radioactive defense wastes stored in 177 underground tanks. Seven vendors were selected for Phase 1 testing to demonstrate vitrification of a high-sodium content liquid LLW simulant. The tested melter technologies included four Joule-heated melters, a carbon electrode melter, a combustion melter, and a plasma melter. Various dry and slurry melter feed preparation processes also were tested. The technologies and Phase 1 testing results were evaluated and a preliminary technology down-selection completed. This report describes the Phase 1 LLW melter vendor testing and the tested technologies, and summarizes the testing results and the preliminary technology recommendations.

  18. Low-Level waste phase 1 melter testing off gas and mass balance evaluation

    SciTech Connect (OSTI)

    Wilson, C.N.

    1996-06-28

    Commercially available melter technologies were tested during 1994-95 as part of a multiphase program to test candidate technologies for vitrification of the low-level waste (LLW) stream to be derived from retrieval and pretreatment of Hanford Site tank wastes. Seven vendors were selected for Phase 1 testing to demonstrate vitrification of a high sodium content liquid LLW simulant. The tested melter technologies included four Joule-heated melters, a carbon electrode melter, a combustion melter, and a plasma melter. Various dry and slurry melter feed preparation processes were also tested. Various feed material samples, product glass samples, and process offgas streams were characterized to provide data for evaluation of process decontamination factors and material mass balances for each vitrification technology. This report describes the melter mass balance evaluations and results for six of the Phase 1 LLW melter vendor demonstration tests.

  19. Greater-than-Class C low-level radioactive waste characterization. Appendix A-2: Timing of greater-than-Class C low-level radioactive waste from nuclear power plants

    SciTech Connect (OSTI)

    Steinke, W.F.

    1994-09-01

    Planning for the storage or disposal of greater-than-Class C low-level radioactive waste (GTCC LLW) requires characterization of that waste. Timing, or the date the waste will require storage or disposal, is an integral aspect of that planning. The majority of GTCC LLW is generated by nuclear power plants, and the length of time a reactor remains operational directly affects the amount of GTCC waste expected from that reactor. This report uses data from existing literature to develop high, base, and low case estimates for the number of plants expected to experience (a) early shutdown, (b) 40-year operation, or (c) life extension to 60-year operation. The discussion includes possible effects of advanced light water reactor technology on future GTCC LLW generation. However, the main focus of this study is timing for shutdown of current technology reactors that are under construction or operating.

  20. Evaluation of solid-based separation materials for the pretreatment of radioactive wastes

    SciTech Connect (OSTI)

    Lumetta, G.J.; Wagner, M.J.; Wester, D.W.; Morrey, J.R.

    1993-05-01

    Separation science will play an important role in pretreating nuclear wastes stored at various US Department of Energy Sites. The application of separation processes offers potential economic and environmental benefits with regards to remediating these sites. For example, at the Hanford Site, the sizeable volume of radioactive wastes stored in underground tanks could be partitioned into a small volume of high-level waste (HLW) and a relatively large volume of low-level waste (LLW). After waste separation, only the smaller volume of HLW would require costly vitrification and geologic disposal. Furthermore, the quality of the remaining LLW form (e.g., grout) would be improved due to the lower inventory of radionuclides present in the LLW stream. This report investigates extraction chromatography as a possible separation process for Hanford wastes.

  1. Exploratory study of complexant concentrate waste processing

    SciTech Connect (OSTI)

    Lumetta, G.J.; Bray, L.A.; Kurath, D.E.; Morrey, J.R.; Swanson, J.L.; Wester, D.W.

    1993-02-01

    The purpose of this exploratory study, conducted by Pacific Northwest Laboratory for Westinghouse Hanford Company, was to determine the effect of applying advanced chemical separations technologies to the processing and disposal of high-level wastes (HLW) stored in underground tanks. The major goals of this study were to determine (1) if the wastes can be partitioned into a small volume of HLW plus a large volume of low-level waste (LLW), and (2) if the activity in the LLW can be lowered enough to meet NRC Class LLW criteria. This report presents the results obtained in a brief scouting study of various processes for separating radionuclides from Hanford complexant concentrate (CC) waste.

  2. Evaluation of melter system technologies for vitrification of high-sodium content low-level radioactive liquid wastes

    SciTech Connect (OSTI)

    Wilson, C.N.

    1994-03-21

    Westinghouse Hanford Company (WHC) is conducting a two-phased demonstration testing and evaluation of candidate melter system technologies for vitrification of Hanford Site low-level tank wastes. The testing is to be performed by melter equipment and vitrification technology commercial suppliers. This Statement of Work is for Phases 1 and 2 of the demonstration testing program. The primary objective of the demonstration testing is to identify the best available melter system technology for the Hanford Site LLW vitrification facility. Data obtained also will support various WHC engineering studies and conceptual design of the LLW vitrification facility. Multiple technologies will be selected for demonstration and evaluation. Testing will be conducted using non-radioactive LLW simulants in Seller-specified pilot/testing facilities.

  3. Alternative generation and analysis for phase I privatization transfer system needs

    SciTech Connect (OSTI)

    Galbraith, J.D.

    1996-09-10

    This decision document provides input for the Phase I Privatization waste staging plans for the High-Level Waste (HLW)and Low-Level Waste (LLW) Disposal Programs. This AGA report evaluates what infrastructure upgrades to existing 200 East waste transfer systems are necessary for delivery of HLW and LLW streams to the Phase I Privatization vendor. The AGA identifies the transfer routing alternatives for supernatant waste transfers from the 241-AN, 241-AW, and 241-AP Tank Farms to the 241-AP-102 tank and/or the 241-AP-104 tank. These two tanks have been targeted as the initial LLW feed staging tanks. In addition,this report addresses the transfer of slurry waste from the 241-AY and 241-AZ Tank Farms to the Phase I Privatization vendor`s facilities for HLW immobilization.

  4. Low-level tank waste simulant data base

    SciTech Connect (OSTI)

    Lokken, R.O.

    1996-04-01

    The majority of defense wastes generated from reprocessing spent N- Reactor fuel at Hanford are stored in underground Double-shell Tanks (DST) and in older Single-Shell Tanks (SST) in the form of liquids, slurries, sludges, and salt cakes. The tank waste remediation System (TWRS) Program has the responsibility of safely managing and immobilizing these tank wastes for disposal. This report discusses three principle topics: the need for and basis for selecting target or reference LLW simulants, tanks waste analyses and simulants that have been defined, developed, and used for the GDP and activities in support of preparing and characterizing simulants for the current LLW vitrification project. The procedures and the data that were generated to characterized the LLW vitrification simulants were reported and are presented in this report. The final section of this report addresses the applicability of the data to the current program and presents recommendations for additional data needs including characterization and simulant compositional variability studies.

  5. Draft low level waste technical summary

    SciTech Connect (OSTI)

    Powell, W.J.; Benar, C.J.; Certa, P.J.; Eiholzer, C.R.; Kruger, A.A.; Norman, E.C.; Mitchell, D.E.; Penwell, D.E.; Reidel, S.P.; Shade, J.W.

    1995-09-01

    The purpose of this document is to present an outline of the Hanford Site Low-Level Waste (LLW) disposal program, what it has accomplished, what is being done, and where the program is headed. This document may be used to provide background information to personnel new to the LLW management/disposal field and to those individuals needing more information or background on an area in LLW for which they are not familiar. This document should be appropriate for outside groups that may want to learn about the program without immediately becoming immersed in the details. This document is not a program or systems engineering baseline report, and personnel should refer to more current baseline documentation for critical information.

  6. Thermal denitration and mineralization of waste constituents

    SciTech Connect (OSTI)

    Nenni, J.A.; Boardman, R.D.

    1997-08-01

    In order to produce a quality grout from LLW using hydraulic cements, proper conditioning of the waste is essential for complete cement curing. Several technologies were investigated as options for conditions. Since the LLW is dilute, removal of all, or most, of the water will significantly reduce the final waste volume. Neutralization of the LLW is also desirable since acidic liquids to not allow cement to cure properly. The nitrate compounds are very soluble and easily leached from solid waste forms; therefore, denitration is desirable. Thermal and chemical denitration technologies have the advantages of water removal, neutralization, and denitration. The inclusion of additives during thermal treatment were investigated as a method of forming insoluable waste conditions.

  7. Catalog of documents produced by the Greater-Than-Class C Low-Level Waste Management Program

    SciTech Connect (OSTI)

    Winberg, M.R.

    1995-03-01

    This catalog provides a ready reference for documents prepared by the Greater-Than-Class C Low-Level Waste (GTCC LLW) Management Program. The GTCC LLW Management Program is part of the National Low-Level Waste Management Program (NLLWMP). The NLLWMP is sponsored by the US Department of Energy (DOE) and is responsible for assisting the DOE in meeting its obligations under Public Law 99-240, The Low-Level Radioactive Waste Policy Amendments Act of 1985. This law assigns DOE the responsibility of ensuring the safe disposal of GTCC LLW in a facility licensed by the Nuclear Regulatory Commission (NRC). The NLLWMP is managed at the Idaho National Engineering Laboratory (INEL).

  8. GTS Duratek, phase I Hanford low-level waste melter tests: Final report

    SciTech Connect (OSTI)

    Eaton, W.C.

    1995-10-26

    A multiphase program was initiated in 1994 to test commercially available melter technologies for the vitrification of the low-level waste (LLW) stream from defense waste stored in underground tanks at the Hanford Site in southeastern Washington State. Phase 1 of the melter demonstration tests using simulated LLW was completed during fiscal year 1995. This document is the final report on testing performed by GTS Duratek Inc. in Columbia, Maryland. GTS Duratek (one of the seven vendors selected) was chosen to demonstrate Joule heated melter technology under WHC subcontract number MMI-SVV-384215. The report contains description of the tests, observations, test data and some analysis of the data as it pertains to application of this technology for LLW vitrification. The document also contains summaries of the melter offgas reports issued as separate documents for the 100 kg melter (WHC-SD-WM-VI-028) and for the 1000 kg melter (WHC-SD-WM-VI-029).

  9. Environmental assessment for the construction, operation, and decommissioning of the Waste Segregation Facility at the Savannah River Site

    SciTech Connect (OSTI)

    1998-01-01

    This Environmental Assessment (EA) has been prepared by the Department of Energy (DOE) to assess the potential environmental impacts associated with the construction, operation and decontamination and decommissioning (D&D) of the Waste Segregation Facility (WSF) for the sorting, shredding, and compaction of low-level radioactive waste (LLW) at the Savannah River Site (SRS) located near Aiken, South Carolina. The LLW to be processed consists of two waste streams: legacy waste which is currently stored in E-Area Vaults of SRS and new waste generated from continuing operations. The proposed action is to construct, operate, and D&D a facility to process low-activity job-control and equipment waste for volume reduction. The LLW would be processed to make more efficient use of low-level waste disposal capacity (E-Area Vaults) or to meet the waste acceptance criteria for treatment at the Consolidated Incineration Facility (CIF) at SRS.

  10. Department of Energy treatment capabilities for greater-than-Class C low-level radioactive waste

    SciTech Connect (OSTI)

    Morrell, D.K.; Fischer, D.K.

    1995-01-01

    This report provides brief profiles for 26 low-level and high-level waste treatment capabilities available at the Idaho National Engineering Laboratory (INEL), Lawrence Livermore National Laboratory (LLNL), Los Alamos National Laboratory (LANL), Oak Ridge National Laboratory (ORNL), Pacific Northwest Laboratory (PNL), Rocky Flats Plant (RFP), Savannah River Site (SRS), and West Valley Demonstration Plant (WVDP). Six of the treatments have potential use for greater-than-Class C low-level waste (GTCC LLW). They include: (a) the glass ceramic process and (b) the Waste Experimental Reduction Facility incinerator at INEL; (c) the Super Compaction and Repackaging Facility and (d) microwave melting solidification at RFP; (e) the vitrification plant at SRS; and (f) the vitrification plant at WVDP. No individual treatment has the capability to treat all GTCC LLW streams. It is recommended that complete physical and chemical characterizations be performed for each GTCC waste stream, to permit using multiple treatments for GTCC LLW.

  11. Microbial degradation of low-level radioactive waste. Final report

    SciTech Connect (OSTI)

    Rogers, R.D.; Hamilton, M.A.; Veeh, R.H.; McConnell, J.W. Jr

    1996-06-01

    The Nuclear Regulatory Commission stipulates in 10 CFR 61 that disposed low-level radioactive waste (LLW) be stabilized. To provide guidance to disposal vendors and nuclear station waste generators for implementing those requirements, the NRC developed the Technical Position on Waste Form, Revision 1. That document details a specified set of recommended testing procedures and criteria, including several tests for determining the biodegradation properties of waste forms. Information has been presented by a number of researchers, which indicated that those tests may be inappropriate for examining microbial degradation of cement-solidified LLW. Cement has been widely used to solidify LLW; however, the resulting waste forms are sometimes susceptible to failure due to the actions of waste constituents, stress, and environment. The purpose of this research program was to develop modified microbial degradation test procedures that would be more appropriate than the existing procedures for evaluation of the effects of microbiologically influenced chemical attack on cement-solidified LLW. The procedures that have been developed in this work are presented and discussed. Groups of microorganisms indigenous to LLW disposal sites were employed that can metabolically convert organic and inorganic substrates into organic and mineral acids. Such acids aggressively react with cement and can ultimately lead to structural failure. Results on the application of mechanisms inherent in microbially influenced degradation of cement-based material are the focus of this final report. Data-validated evidence of the potential for microbially influenced deterioration of cement-solidified LLW and subsequent release of radionuclides developed during this study are presented.

  12. Conceptual design report for handling Fort St. Vrain fuel element components

    SciTech Connect (OSTI)

    Gavalya, R.A.

    1993-09-01

    This report presents conceptual designs for containment of high-level wastes (HLW) and low-level wastes (LLW) that will result from disassembly of fuel elements from the High Temperature Gas-Cooled Reactor at the Fort St. Vrain nuclear power plant in Platteville, Colorado. Hexagonal fuel elements will enter the disassembly area as a HLW and exit as either as HLW or LLW. The HLW will consist of spent fuel compacts that have been removed from the hexagonal graphite block. Graphite dust and graphite particles produced during the disassembly process will also be routed to the container that will hold the HLW spent fuel compacts. The LLW will consist of the emptied graphite block. Three alternatives have been introduced for interim storage of the HLW containers after the spent fuel has been loaded. The three alternatives are: (a) store containers where fuel elements are currently being stored, (b) construct a new dry storage facility, and (c) employ Multi-Purpose Canisters (currently in conceptual design stage). Containment of the LLW graphite block will depend on several factors: (a) LLW classification, (b) radiation levels, and (c) volume-reducing technique (if used). Packaging may range from cardboard boxes for incinerable wastes to 55-ton cask inserts for remote-handled wastes. Before final designs for the containment of the HLW and LLW can be developed, several issues need to be addressed: (a) packing factor for fuel compacts in HLW container, (b) storage/disposal of loaded HLW containers, (c) characterization of the emptied graphite blocks, and (d) which technique for volume-reduction purposes (if any) will be used.

  13. An Update On Waste Control Specialists' 2004 License Application For Safe Disposal Of Class A, B, and C Low-Level Radioactive Waste In Texas

    SciTech Connect (OSTI)

    Baltzer, R.; Eriksson, L. [Waste Control Specialists LLC, Three Lincoln Centre, Dallas, Texas (United States)

    2008-07-01

    On December 10, 2007, Waste Control Specialists LLC (WCS) received notification that the Executive Director of the Texas Commission on Environmental Quality (TCEQ) had prepared an interim draft license and made a preliminary decision that it met all statutory and regulatory requirements for safe disposal of low-level radioactive waste (LLW) at the WCS' site in Texas. Pursuant to this interim draft license, WCS will be authorized to dispose Class A, B, and C LLW in two enhanced near-surface landfills at WCS' 5.4-square-kilometer (1,338-acre) treatment, storage, and disposal (TSD) site in Andrews County, Texas (Fig. 1). One landfill will be dedicated to LLW generated within the member/party states of the Texas Compact (Texas and Vermont), while the other will be dedicated to LLW generated by the federal government. The calculated annual peak dose to the maximally exposed member of the general public, i.e., an adjacent resident, from any of the proposed LLW-disposal landfills occurs approximately 36,400 years after closure and is 0.034 milli-sievert (mSv) (3.4 milli-rem (mrem)), which is less than 14 percent of the applicable regulatory limit of 25 mSv (25 mrem). The draft license will be published in February 2008, which will be followed by 12 months of public hearings, and three months for preparation of the final license. Based on this schedule, the final license is due in May 2009. When opened, the WCS site will achieve a national milestone; it will be the first new Compact LLW-disposal site in the USA to open under the Low-Level Radioactive Waste Policy Act of 1980, as amended in 1985. (authors)

  14. Identifying industrial best practices for the waste minimization of low-level radioactive materials

    SciTech Connect (OSTI)

    Levin, V.

    1996-04-01

    In US DOE, changing circumstances are affecting the management and disposal of solid, low-level radioactive waste (LLW). From 1977 to 1991, the nuclear power industry achieved major reductions in solid waste disposal, and DOE is interested in applying those practices to reduce solid waste at DOE facilities. Project focus was to identify and document commercial nuclear industry best practices for radiological control programs supporting routine operations, outages, and decontamination and decommissioning activities. The project team (DOE facility and nuclear power industry representatives) defined a Work Control Process Model, collected nuclear power industry Best Practices, and made recommendations to minimize LLW at DOE facilities.

  15. Chloride removal from vitrification offgas

    SciTech Connect (OSTI)

    Slaathaug, E.J. [Westinghouse Hanford Co., Richland, WA (United States)

    1995-06-01

    This study identified and investigated techniques of selectively purging chlorides from the low-level waste (LLW) vitrification process with the purge stream acceptable for burial on the Hanford Site. Chlorides will be present in high concentration in several individual feeds to the LLW Vitrification Plant. The chlorides are highly volatile in combustion type melters and are readily absorbed by wet scrubbing of the melter offgas. The Tank Waste Remediation System (TWRS) process flow sheets show that the resulting chloride rich scrub solution is recycled back to the melter. The chlorides must be purged from the recycle loop to prevent the buildup of excessively high chloride concentrations.

  16. RECOUNTING HISTORY THROUGH RADIOASSAY

    SciTech Connect (OSTI)

    JASEN, W.G.

    2005-12-21

    This paper describes a proposed method for using historical documentation to identify unknown wastes resulting from retrieving suspect transuranic (TRU) waste. Identification is accomplished by a historical review of radionuclides identified by radioassay, along with the project controls used to ensure an accurate segregation of TRU from low-level waste (LLW). This paper presents an historical perspective on the identification of radionuclides at the Hanford Site from various waste generators of suspect TRU waste with an emphasis on the Data Quality Objectives (DQO's) and project controls used to ensure the waste is properly classified as TRU or LLW.

  17. Vectra GSI, Inc. low-level waste melter testing Phase 1 test report

    SciTech Connect (OSTI)

    Stegen, G.E.; Wilson, C.N.

    1996-02-21

    A multiphase program was initiated in 1994 to test commercially available melter technologies for the vitrification of the low-level waste (LLW) stream from defense wastes stored in underground tanks at the Hanford Site in southeastern Washington State. Vectra GSI, Inc. was one of seven vendors selected for Phase 1 of the melter demonstration tests using simulated LLW that were completed during fiscal year 1995. The attached report prepared by Vectra GSI, Inc. describes results of melter testing using slurry feed and dried feeds. Results of feed drying and prereaction tests using a fluid bed calciner and rotary dryer also are described.

  18. Type Ia supernovae from exploding oxygen-neon white dwarfs

    E-Print Network [OSTI]

    Marquardt, Kai S; Ruiter, Ashley J; Seitenzahl, Ivo R; Ohlmann, Sebastian T; Kromer, Markus; Pakmor, Ruediger; Roepke, Friedrich K

    2015-01-01

    The progenitor problem of Type Ia supernovae (SNe Ia) is still unsolved. Most of these events are thought to be explosions of carbon-oxygen (CO) white dwarfs (WDs), but for many of the explosion scenarios, particularly those involving the externally triggered detonation of a sub-Chandrasekhar mass WD (sub-M Ch WD), there is also a possibility of having an oxygen-neon (ONe) WD as progenitor. We simulate detonations of ONe WDs and calculate synthetic observables from these models. The results are compared with detonations in CO WDs of similar mass and observational data of SNe Ia. We perform hydrodynamic explosion simulations of detonations in initially hydrostatic ONe WDs for a range of masses below the Chandrasekhar mass (M Ch), followed by detailed nucleosynthetic postprocessing with a 384-isotope nuclear reaction network. The results are used to calculate synthetic spectra and light curves, which are then compared with observations of SNe Ia. We also perform binary evolution calculations to determine the nu...

  19. THE WIRED SURVEY. IV. NEW DUST DISKS FROM THE McCOOK and SION WHITE DWARF CATALOG

    SciTech Connect (OSTI)

    Hoard, D. W.; Wachter, Stefanie; Debes, John H.; Leisawitz, David T.; Cohen, Martin

    2013-06-10

    We have compiled photometric data from the Wide-field Infrared Survey Explorer All Sky Survey and other archival sources for the more than 2200 objects in the original McCook and Sion Catalog of Spectroscopically Identified White Dwarfs. We applied color-selection criteria to identify 28 targets whose infrared spectral energy distributions depart from the expectation for the white dwarf (WD) photosphere alone. Seven of these are previously known WDs with circumstellar dust disks, five are known central stars of planetary nebulae, and six were excluded for being known binaries or having possible contamination of their infrared photometry. We fit WD models to the spectral energy distributions of the remaining ten targets, and find seven new candidates with infrared excess suggesting the presence of a circumstellar dust disk. We compare the model dust disk properties for these new candidates with a comprehensive compilation of previously published parameters for known WDs with dust disks. It is possible that the current census of WDs with dust disks that produce an excess detectable at K-band and shorter wavelengths is close to complete for the entire sample of known WDs to the detection limits of existing near-IR all-sky surveys. The WD dust disk candidates now being found using longer wavelength infrared data are drawn from a previously underrepresented region of parameter space, in which the dust disks are overall cooler, narrower in radial extent, and/or contain fewer emitting grains.

  20. Mobile Sensor Networks for Leak and Backflow Detection in Water Distribution Systems

    E-Print Network [OSTI]

    Shihada, Basem

    Mobile Sensor Networks for Leak and Backflow Detection in Water Distribution Systems M. Agumbe detection are essential aspects of Water Distribution System (WDS) monitoring. Most existing solutions for leak detection in water distribution systems focus on the placement of expensive static sensors located

  1. POPULATION STUDIES OF CATACLYSMIC VARIABLES F. A. RINGWALD

    E-Print Network [OSTI]

    Ringwald, Frederick A.

    . The Palomar-Green catalog includes 1715 objects of all types (QSOs, sdBs, WDs) in its complete sample & Engel 1994). This color index also de nes inclusion in the Palomar-Green (PG) survey (Green, Schmidt & Liebert 1986), and a preliminary list of PG CVs was given by Green et al. (1982). My Ph.D. thesis

  2. REVISED DRAFT 2 7/16/01 Testing the Performance of Uniform Price and Discriminative Auctions*

    E-Print Network [OSTI]

    REVISED DRAFT 2 7/16/01 Testing the Performance of Uniform Price and Discriminative Auctions* T. D and Management 301 Warren Hall Cornell University Ithaca, NY 14853 WDS3@cornell.edu R. J. Thomas Professor School the relative performance of different types of auctions for electricity markets. The experiments involved

  3. THE BIRTH RATE OF SNe Ia FROM HYBRID CONe WHITE DWARFS

    SciTech Connect (OSTI)

    Meng, Xiangcun [Yunnan Observatories, Chinese Academy of Sciences, Kunming 650011 (China); Podsiadlowski, Philipp, E-mail: xiangcunmeng@ynao.ac.cn [Department of Astronomy, Oxford University, Oxford OX1 3RH (United Kingdom)

    2014-07-10

    Considering the uncertainties of the C-burning rate (CBR) and the treatment of convective boundaries, Chen et al. found that there is a regime where it is possible to form hybrid CONe white dwarfs (WDs), i.e., ONe WDs with carbon-rich cores. As these hybrid WDs can be as massive as 1.30 M {sub ?}, not much mass needs to be accreted for these objects to reach the Chandrasekhar limit and to explode as Type Ia supernovae (SNe Ia). We have investigated their contribution to the overall SN Ia birth rate and found that such SNe Ia tend to be relatively young with typical time delays between 0.1 and 1 Gyr, where some may be as young as 30 Myr. SNe Ia from hybrid CONe WDs may contribute several percent to all SNe Ia, depending on the common-envelope ejection efficiency and the CBR. We suggest that these SNe Ia may produce part of the 2002cx-like SN Ia class.

  4. Nutrition, antigenicity and serological characteristics of different strains of Moraxella bovis

    E-Print Network [OSTI]

    Chowdhury, T. I. M. Fazlayrabbi

    1963-01-01

    . of distilled water in a 1000 ml. flask snd heated to di. ssalve the medium according to the direction of the manufacturer. The non-enriched agar wds made, sterilised in sn autoclave for 15 minutes at 15 pounds of pressure and poured into Nolle culture...

  5. An Initial Survey of White Dwarfs in the Sloan Digital Sky Survey

    E-Print Network [OSTI]

    H. C. Harris; J. Liebert; S. J. Kleinman; A. Nitta; S. F. Anderson; G. R. Knapp; J. Krzesinski; G. Schmidt; M. A. Strauss; D. Vanden Berk; D. Eisenstein; S. Hawley; B. Margon; J. A. Munn; N. M. Silvestri; A. Smith; P. Szkody; M. J. Collinge; C. C. Dahn; X. Fan; P. B. Hall; D. P. Schneider; J. Brinkmann; S. Burles; J. E. Gunn; G. S. Hennessy; R. Hindsley; Z. Ivezic; S. Kent; D. Q. Lamb; R. H. Lupton; R. C. Nichol; J. R. Pier; D. J. Schlegel; M. SubbaRao; A. Uomoto; B. Yanny; D. G. York

    2003-05-19

    An initial assessment is made of white dwarf and hot subdwarf stars observed in the Sloan Digital Sky Survey. In a small area of sky (190 square degrees), observed much like the full survey will be, 269 white dwarfs and 56 hot subdwarfs are identified spectroscopically where only 44 white dwarfs and 5 hot subdwarfs were known previously. Most are ordinary DA (hydrogen atmosphere) and DB (helium) types. In addition, in the full survey to date, a number of WDs have been found with uncommon spectral types. Among these are blue DQ stars displaying lines of atomic carbon; red DQ stars showing molecular bands of C_2 with a wide variety of strengths; DZ stars where Ca and occasionally Mg, Na, and/or Fe lines are detected; and magnetic WDs with a wide range of magnetic field strengths in DA, DB, DQ, and (probably) DZ spectral types. Photometry alone allows identification of stars hotter than 12000 K, and the density of these stars for 15WDs and the fraction of WDs rises to ~90% at g=20. The remainder are hot sdB and sdO stars.

  6. THE WIRED SURVEY. II. INFRARED EXCESSES IN THE SDSS DR7 WHITE DWARF CATALOG

    SciTech Connect (OSTI)

    Debes, John H.; Leisawitz, David T.; Hoard, D. W.; Wachter, Stefanie; Cohen, Martin

    2011-12-01

    With the launch of the Wide-field Infrared Survey Explorer (WISE), a new era of detecting planetary debris and brown dwarfs (BDs) around white dwarfs (WDs) has begun with the WISE InfraRed Excesses around Degenerates (WIRED) Survey. The WIRED Survey is sensitive to substellar objects and dusty debris around WDs out to distances exceeding 100 pc, well beyond the completeness level of local WDs. In this paper, we present a cross-correlation of the preliminary Sloan Digital Sky Survey (SDSS) Data Release 7 (DR7) WD catalog between the WISE, Two-Micron All Sky Survey (2MASS), UKIRT Infrared Deep Sky Survey (UKIDSS), and SDSS DR7 photometric catalogs. From {approx}18,000 input targets, there are WISE detections comprising 344 'naked' WDs (detection of the WD photosphere only), 1020 candidate WD+M dwarf binaries, 42 candidate WD+BD systems, 52 candidate WD+dust disk systems, and 69 targets with indeterminate infrared excess. We classified all of the detected targets through spectral energy distribution model fitting of the merged optical, near-IR, and WISE photometry. Some of these detections could be the result of contaminating sources within the large ( Almost-Equal-To 6'') WISE point-spread function; we make a preliminary estimate for the rates of contamination for our WD+BD and WD+disk candidates and provide notes for each target of interest. Each candidate presented here should be confirmed with higher angular resolution infrared imaging or infrared spectroscopy. We also present an overview of the observational characteristics of the detected WDs in the WISE photometric bands, including the relative frequencies of candidate WD+M, WD+BD, and WD+disk systems.

  7. CLEARANCE ISSUES FOR ADVANCED FUSION POWER PLANTS L. El-Guebaly, D. Henderson, A. Abdou, P. Wilson, and the ARIES Team

    E-Print Network [OSTI]

    in the nuclear industry. The primary options for managing the waste of the ARIES power plants2 include near waste management, it is essential to assess the implication of the clearance option on the waste-surface disposal as Class A or Class C Low-Level Waste (LLW), recycling and reuse in nuclear facilities

  8. Preliminary Deep Water Results in Single-Beacon One-Way-Travel-Time Acoustic Navigation for Underwater Vehicles

    E-Print Network [OSTI]

    Eustice, Ryan

    Preliminary Deep Water Results in Single-Beacon One-Way-Travel-Time Acoustic Navigation results from the first deep-water evaluation of this method using data collected from an autonomous of Mechanical Engineering Johns Hopkins University, Baltimore, MD 21218 email: {swebster,llw}@jhu.edu Department

  9. Engineering report of plasma vitrification of Hanford tank wastes

    SciTech Connect (OSTI)

    Hendrickson, D.W.

    1995-05-12

    This document provides an analysis of vendor-derived testing and technology applicability to full scale glass production from Hanford tank wastes using plasma vitrification. The subject vendor testing and concept was applied in support of the Hanford LLW Vitrification Program, Tank Waste Remediation System.

  10. Design Challenges and Activation Concerns for ARIES Vacuum Vessel

    E-Print Network [OSTI]

    composition specifically tailored for LLW) 3Cr-3WV bainitic steels 8-9%Cr reduced activation ferritic-martensitic is unacceptable for VV due to complex heat treatment requirement Selection of austenitic stainless steel (such-based steel). #12;7 Seven Steels Selected for Further Analysis The seven steels: F82H ferritic steel (having

  11. Fission Product Impact Reduction via Protracted In-core Retention in Very High Temperature Reactor (VHTR) Transmutation Scenarios

    E-Print Network [OSTI]

    Alajo, Ayodeji Babatunde

    2011-08-08

    ?.???.......................................................... 8 I.F Outline and Strategy ??....................................................................... 9 II FISSION PRODUCT VECTOR, SOURCE AND TREATMENT ????.. 12 II.A LWR Fission Product Inventories ??????..??????....... 13 II... Library JNDC Japanese Nuclear Data Committee KAERI Korea Atomic Energy Research Institute viii LANL Los Alamos National Laboratory LLFP Long Lived Fission Product LLW Low Level Waste LWR Light Water Reactor MCNP Monte Carlo N ? Particle MCNPX...

  12. National Environmental Policy Act Compliance Strategy for the Remote-Handled Low-level Waste Disposal Facility

    SciTech Connect (OSTI)

    Peggy Hinman

    2010-10-01

    The U.S. Department of Energy (DOE) needs to have disposal capability for remote-handled low level waste (LLW) generated at the Idaho National Laboratory (INL) at the time the existing disposal facility is full or must be closed in preparation for final remediation of the INL Subsurface Disposal Area in approximately the year 2017.

  13. EUROGRAPHICS 2012/ A. Fusiello, M. Wimmer Poster Multiview-Consistent Color Sampling for Alpha Matting

    E-Print Network [OSTI]

    Eisert, Peter

    formulate a new cost function penalizing deviations of foreground color estimates between associated pixel was supported by the European FP7 project REACT (288369) for this sampling process to be efficiently carried out to the closed-form global matting method proposed in [LLW08]. 2. Constructing a 3D trimap In a multiview setup

  14. Taiwan industrial cooperation program technology transfer for low-level radioactive waste final disposal - phase I.

    SciTech Connect (OSTI)

    Knowlton, Robert G.; Cochran, John Russell; Arnold, Bill Walter; Jow, Hong-Nian; Mattie, Patrick D.; Schelling, Frank Joseph Jr.

    2007-01-01

    Sandia National Laboratories and the Institute of Nuclear Energy Research, Taiwan have collaborated in a technology transfer program related to low-level radioactive waste (LLW) disposal in Taiwan. Phase I of this program included regulatory analysis of LLW final disposal, development of LLW disposal performance assessment capabilities, and preliminary performance assessments of two potential disposal sites. Performance objectives were based on regulations in Taiwan and comparisons to those in the United States. Probabilistic performance assessment models were constructed based on limited site data using software including GoldSim, BLT-MS, FEHM, and HELP. These software codes provided the probabilistic framework, container degradation, waste-form leaching, groundwater flow, radionuclide transport, and cover infiltration simulation capabilities in the performance assessment. Preliminary performance assessment analyses were conducted for a near-surface disposal system and a mined cavern disposal system at two representative sites in Taiwan. Results of example calculations indicate peak simulated concentrations to a receptor within a few hundred years of LLW disposal, primarily from highly soluble, non-sorbing radionuclides.

  15. Siting Study for the Remote-Handled Low-Level Waste Disposal Project

    SciTech Connect (OSTI)

    Lisa Harvego; Joan Connolly; Lance Peterson; Brennon Orr; Bob Starr

    2010-10-01

    The U.S. Department of Energy has identified a mission need for continued disposal capacity for remote-handled low-level waste (LLW) generated at the Idaho National Laboratory (INL). An alternatives analysis that was conducted to evaluate strategies to achieve this mission need identified two broad options for disposal of INL generated remote-handled LLW: (1) offsite disposal and (2) onsite disposal. The purpose of this study is to identify candidate sites or locations within INL boundaries for the alternative of an onsite remote handled LLW disposal facility and recommend the highest-ranked locations for consideration in the National Environmental Policy Act process. The study implements an evaluation based on consideration of five key elements: (1) regulations, (2) key assumptions, (3) conceptual design, (4) facility performance, and (5) previous INL siting study criteria, and uses a five-step process to identify, screen, evaluate, score, and rank 34 separate sites located across INL. The result of the evaluation is identification of two recommended alternative locations for siting an onsite remote-handled LLW disposal facility. The two alternative locations that best meet the evaluation criteria are (1) near the Advanced Test Reactor Complex and (2) west of the Idaho Comprehensive Environmental Response, Compensation, and Liability Act Disposal Facility.

  16. Melter system technology testing for Hanford Site low-level tank waste vitrification

    SciTech Connect (OSTI)

    Wilson, C.N. [Westinghouse Hanford Company, Richland, WA (United States)

    1996-12-31

    Following revisions to the Tri-Party Agreement for Hanford Site cleanup, which specified vitrification for immobilization of the low-level waste (LLW) stream to be derived from retrieval and pretreatment of the radioactive defense wastes stored in 177 underground tanks, commercially available melter technologies were tested during 1994 to 1995 as part of a multiphase program to select reference technologies for the new LLW vitrification mission. Seven vendors were selected for Phase 1 testing to demonstrate vitrification of a high-sodium content liquid LLW simulant. The tested melter technologies included four Joule-heated melters, a carbon electrode melter, a combustion melter, and a plasma melter. Various dry and slurry melter feed preparation processes were also tested. The technologies and Phase 1 testing results were evaluated and a preliminary technology down-selection and recommendations for Phase 2 testing completed. This paper describes the Phase 1 LLW melter vendor testing program and the tested technologies, and summarizes the testing results and the preliminary technology recommendations.

  17. Vendor glass durability study during evaluation of melter system technologies for vitrification of Hanford low-level wastes

    SciTech Connect (OSTI)

    Feng, X.; Kim, D.; Schweiger, M.J. [Pacific Northwest Lab., Richland, WA (United States)] [and others

    1995-12-31

    The low level radioactive wastes (LLW) separated from the single-shell tanks and double-shell tanks at the Hanford Site will be immobilized into glass. A melter system technology testing, and evaluation program is being conducted to identify the demonstration, best overall melter system technology available to vitrify the Hanford LLW streams. The melter technologies being demonstrated use a variety of heating methods to melt the glass, including plasma torch, fossil-fuel-fired cyclone burner, carbon arc and joule-heating. The Phase I testing is a {open_quotes}proof of principle{close_quotes} test to demonstrate that a melter system technology can process a simulated highly alkaline, high nitrate/nitrite content LLW feed and produce a glass product of consistent quality. Target waste oxide loading of LLW simulant was specified to be about 25 wt%. Pacific Northwest Laboratory (PNL) is providing glass formulation support for this program. The five candidate vendor glasses at 20 wt% Na{sub 2}O level provided by PNL are alumino-borosilicate and aluminosilicate glasses with melting points around 1300{degrees}. Glasses adopted by vendors were tested at PNL to verify the required properties. The testing included durability evaluation through product consistency test, MCC-1 tests, and flow through tests and viscosity measurements.

  18. Accelerated Leach Test(s) Program: Annual report

    SciTech Connect (OSTI)

    Dougherty, D.R.; Pietrzak, R.F.; Fuhrmann, M.; Colombo, P.

    1986-09-01

    A computerized data base of LLW leaching data has been developed. Long-term tests on portland cement, bitumen and vinyl ester-styrene (VES) polymer waste forms containing simulated wastes are underway which are designed to identify and evaluate factors that accelerate leaching without changing the mechanisms.

  19. 2nd Quarter Transportation Report FY 2014

    SciTech Connect (OSTI)

    Gregory, L.

    2014-07-30

    This report satisfies the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Field Office (NNSA/NFO) commitment to prepare a quarterly summary report of radioactive waste shipments to the Nevada National Security Site (NNSS) Radioactive Waste Management Complex (RWMC) at Area 5. There were no shipments sent for offsite treatment and returned to the NNSS this quarter. This report summarizes the second quarter of fiscal year (FY) 2014 low-level radioactive waste (LLW) and mixed low-level radioactive waste (MLLW) shipments. This report also includes annual summaries for FY 2014 in Tables 4 and 5. Tabular summaries are provided which include the following: Sources of and carriers for LLW and MLLW shipments to and from the NNSS; Number and external volume of LLW and MLLW shipments; Highway routes used by carriers; and Incident/accident data applicable to LLW and MLLW shipments. In this report shipments are accounted for upon arrival at the NNSS, while disposal volumes are accounted for upon waste burial. The disposal volumes presented in this report do not include minor volumes of non-radioactive materials that were approved for disposal. Volume reports showing cubic feet (ft3) generated using the Low-Level Waste Information System may vary slightly due to differing rounding conventions.

  20. Caustic leaching of high-level radioactive tank sludge: A critical literature review

    SciTech Connect (OSTI)

    McGinnis, C.P.; Welch, T.D.; Hunt, R.D.

    1998-08-01

    The Department of Energy (DOE) must treat and safely dispose of its radioactive tank contents, which can be separated into high-level waste (HLW) and low-level waste (LLW) fractions. Since the unit costs of treatment and disposal are much higher for HLW than for LLW, technologies to reduce the amount of HLW are being developed. A key process currently being studied to reduce the volume of HLW sludges is called enhanced sludge washing (ESW). This process removes, by water washes, soluble constituents such as sodium salts, and the washed sludge is then leached with 2--3 M NaOH at 60--100 C to remove nonradioactive metals such as aluminum. The remaining solids are considered to be HLW while the solutions are LLW after radionuclides such as {sup 137}Cs have been removed. Results of bench-scale tests have shown that the ESW will probably remove the required amounts of inert constituents. While both experimental and theoretical results have shown that leaching efficiency increases as the time and temperature of the leach are increased, increases in the caustic concentration above 2--3 M will only marginally improve the leach factors. However, these tests were not designed to validate the assumption that the caustic used in the ESW process will generate only a small increase (10 Mkg) in the amount of LLW; instead the test conditions were selected to maximize leaching in a short period and used more water and caustic than is planned during full-scale operations. Even though calculations indicate that the estimate for the amount of LLW generated by the ESW process appears to be reasonable, a detailed study of the amount of LLW from the ESW process is still required. If the LLW analysis indicates that sodium management is critical, then a more comprehensive evaluation of the clean salt process or caustic recycle would be needed. Finally, experimental and theoretical studies have clearly demonstrated the need for the control of solids formation during and after leaching.

  1. Caustic leaching of high-level radioactive tank sludge: A critical literature review

    SciTech Connect (OSTI)

    McGinnis, C.P.; Welch, T.D.; Hunt, R.D.

    1997-12-31

    The Department of Energy (DOE) must treat and safely dispose of its radioactive tank contents, which can be separated into high-level waste (HLW) and low-level waste (LLW) fractions. Since the unit costs of treatment and disposal are much higher for HLW than for LLW, technologies to reduce the amount of HLW are being developed. A key process currently being studied to reduce the volume of HLW sludges is called enhanced sludge washing (ESW). This process removes, by water washes, soluble constituents such as sodium salts, and the washed sludge is then leached with 2--3 M NaOH at 60--100 C to remove nonradioactive metals such as aluminum. The remaining solids are considered to be HLW while the solutions are LLW after radionuclides such as {sup 137}Cs have been removed. Results of bench-scale tests have shown that the ESW will probably remove the required amounts of inert constituents. While both experimental and theoretical results have shown that leaching efficiency increases as the time and temperature of the leach are increased, increases in the caustic concentration above 2--3 M will only marginally improve the leach factors. However, these tests were not designed to validate the assumption that the caustic used in the ESW process will generate only a small increase (10 Mkg) in the amount of LLW; instead, the test conditions were selected to maximize leaching in a short period and used more water and caustic than is planned during full-scale operations. Even though calculations indicate that the estimate for the amount of LLW generated by the ESW process appears to be reasonable, a detailed study of the amount of LLW from the ESW process is still required. If the LLW analysis indicates that sodium management is critical, then a more comprehensive evaluation of the clean salt process or caustic recycle would be needed. Finally, experimental and theoretical studies have clearly demonstrated the need for the control of solids formation during and after leaching.

  2. Separation, Concentration, and Immobilization of Technetium and Iodine from Alkaline Supernate Waste

    SciTech Connect (OSTI)

    James Harvey; Michael Gula

    1998-12-07

    Development of remediation technologies for the characterization, retrieval, treatment, concentration, and final disposal of radioactive and chemical tank waste stored within the Department of Energy (DOE) complex represents an enormous scientific and technological challenge. A combined total of over 90 million gallons of high-level waste (HLW) and low-level waste (LLW) are stored in 335 underground storage tanks at four different DOE sites. Roughly 98% of this waste is highly alkaline in nature and contains high concentrations of nitrate and nitrite salts along with lesser concentrations of other salts. The primary waste forms are sludge, saltcake, and liquid supernatant with the bulk of the radioactivity contained in the sludge, making it the largest source of HLW. The saltcake (liquid waste with most of the water removed) and liquid supernatant consist mainly of sodium nitrate and sodium hydroxide salts. The main radioactive constituent in the alkaline supernatant is cesium-137, but strontium-90, technetium-99, and transuranic nuclides are also present in varying concentrations. Reduction of the radioactivity below Nuclear Regulatory Commission (NRC) limits would allow the bulk of the waste to be disposed of as LLW. Because of the long half-life of technetium-99 (2.1 x 10 5 y) and the mobility of the pertechnetate ion (TcO 4 - ) in the environment, it is expected that technetium will have to be removed from the Hanford wastes prior to disposal as LLW. Also, for some of the wastes, some level of technetium removal will be required to meet LLW criteria for radioactive content. Therefore, DOE has identified a need to develop technologies for the separation and concentration of technetium-99 from LLW streams. Eichrom has responded to this DOE-identified need by demonstrating a complete flowsheet for the separation, concentration, and immobilization of technetium (and iodine) from alkaline supernatant waste.

  3. Clean option: An alternative strategy for Hanford Tank Waste Remediation. Volume 2, Detailed description of first example flowsheet

    SciTech Connect (OSTI)

    Swanson, J.L.

    1993-09-01

    Disposal of high-level tank wastes at the Hanford Site is currently envisioned to divide the waste between two principal waste forms: glass for the high-level waste (HLW) and grout for the low-level waste (LLW). The draft flow diagram shown in Figure 1.1 was developed as part of the current planning process for the Tank Waste Remediation System (TWRS), which is evaluating options for tank cleanup. The TWRS has been established by the US Department of Energy (DOE) to safely manage the Hanford tank wastes. It includes tank safety and waste disposal issues, as well as the waste pretreatment and waste minimization issues that are involved in the ``clean option`` discussed in this report. This report describes the results of a study led by Pacific Northwest Laboratory to determine if a more aggressive separations scheme could be devised which could mitigate concerns over the quantity of the HLW and the toxicity of the LLW produced by the reference system. This aggressive scheme, which would meet NRC Class A restrictions (10 CFR 61), would fit within the overall concept depicted in Figure 1.1; it would perform additional and/or modified operations in the areas identified as interim storage, pretreatment, and LLW concentration. Additional benefits of this scheme might result from using HLW and LLW disposal forms other than glass and grout, but such departures from the reference case are not included at this time. The evaluation of this aggressive separations scheme addressed institutional issues such as: radioactivity remaining in the Hanford Site LLW grout, volume of HLW glass that must be shipped offsite, and disposition of appropriate waste constituents to nonwaste forms.

  4. Environmental assessment for the treatment of Class A low-level radioactive waste and mixed low-level waste generated by the West Valley Demonstration Project

    SciTech Connect (OSTI)

    NONE

    1995-11-01

    The U.S. Department of Energy (DOE) is currently evaluating low-level radioactive waste management alternatives at the West Valley Demonstration Project (WVDP) located on the Western New York Nuclear Service Center (WNYNSC) near West Valley, New York. The WVDP`s mission is to vitrify high-level radioactive waste resulting from commercial fuel reprocessing operations that took place at the WNYNSC from 1966 to 1972. During the process of high-level waste vitrification, low-level radioactive waste (LLW) and mixed low-level waste (MILLW) will result and must be properly managed. It is estimated that the WVDP`s LLW storage facilities will be filled to capacity in 1996. In order to provide sufficient safe storage of LLW until disposal options become available and partially fulfill requirements under the Federal Facilities Compliance Act (FFCA), the DOE is proposing to use U.S. Nuclear Regulatory Commission-licensed and permitted commercial facilities in Oak Ridge, Tennessee; Clive, Utah; and Houston, Texas to treat (volume-reduce) a limited amount of Class A LLW and MLLW generated from the WVDP. Alternatives for ultimate disposal of the West Valley LLW are currently being evaluated in an environmental impact statement. This proposed action is for a limited quantity of waste, over a limited period of time, and for treatment only; this proposal does not include disposal. The proposed action consists of sorting, repacking, and loading waste at the WVDP; transporting the waste for commercial treatment; and returning the residual waste to the WVDP for interim storage. For the purposes of this assessment, environmental impacts were quantified for a five-year operating period (1996 - 2001). Alternatives to the proposed action include no action, construction of additional on-site storage facilities, construction of a treatment facility at the WVDP comparable to commercial treatment, and off-site disposal at a commercial or DOE facility.

  5. Dark matter burners

    E-Print Network [OSTI]

    Moskalenko, I V; Moskalenko, Igor V.; Wai, Lawrence L.

    2007-01-01

    We show that a star orbiting close enough to an adiabatically grown supermassive black hole (SMBH) can capture weakly interacting massive particles (WIMPs) at an extremely high rate. The stellar luminosity due to annihilation of captured WIMPs in the stellar core may be comparable to or even exceed the luminosity of the star due to thermonuclear burning. The model thus predicts the existence of unusual stars, essentially WIMP burners, in the vicinity of a SMBH. We find that the most efficient WIMP burners are stars with degenerate electron cores, e.g. white dwarfs (WDs); such WDs may have a very high surface temperature. If found, such stars would provide evidence for the existence of particle dark matter and can possibly be used to establish its density profile. On the other hand, the lack of such unusual stars may provide constraints on the WIMP density near the SMBH, as well as the WIMP-nucleus scattering and pair annihilation cross-sections.

  6. Dark matter burners

    E-Print Network [OSTI]

    Igor V. Moskalenko; Lawrence L. Wai

    2007-02-24

    We show that a star orbiting close enough to an adiabatically grown supermassive black hole (SMBH) can capture weakly interacting massive particles (WIMPs) at an extremely high rate. The stellar luminosity due to annihilation of captured WIMPs in the stellar core may be comparable to or even exceed the luminosity of the star due to thermonuclear burning. The model thus predicts the existence of unusual stars, essentially WIMP burners, in the vicinity of a SMBH. We find that the most efficient WIMP burners are stars with degenerate electron cores, e.g. white dwarfs (WDs); such WDs may have a very high surface temperature. If found, such stars would provide evidence for the existence of particle dark matter and can possibly be used to establish its density profile. On the other hand, the lack of such unusual stars may provide constraints on the WIMP density near the SMBH, as well as the WIMP-nucleus scattering and pair annihilation cross-sections.

  7. Shortest recurrence periods of novae

    SciTech Connect (OSTI)

    Kato, Mariko [Department of Astronomy, Keio University, Hiyoshi, Yokohama 223-8521 (Japan); Saio, Hideyuki [Astronomical Institute, Graduate School of Science, Tohoku University, Sendai 980-8578 (Japan); Hachisu, Izumi [Department of Earth Science and Astronomy, College of Arts and Sciences, The University of Tokyo, 3-8-1 Komaba, Meguro-ku, Tokyo 153-8902 (Japan); Nomoto, Ken'ichi, E-mail: mariko@educ.cc.keio.ac.jp [Kavli Institute for the Physics and Mathematics of the Universe (WPI), The University of Tokyo, 5-1-5 Kashiwanoha, Kashiwa, Chiba 277-8583 (Japan)

    2014-10-01

    Stimulated by the recent discovery of the 1 yr recurrence period nova M31N 2008-12a, we examined the shortest recurrence periods of hydrogen shell flashes on mass-accreting white dwarfs (WDs). We discuss the mechanism that yields a finite minimum recurrence period for a given WD mass. Calculating the unstable flashes for various WD masses and mass accretion rates, we identified a shortest recurrence period of about two months for a non-rotating 1.38 M {sub ?} WD with a mass accretion rate of 3.6 10{sup 7} M {sub ?} yr{sup 1}. A 1 yr recurrence period is realized for very massive (? 1.3 M {sub ?}) WDs with very high accretion rates (? 1.5 10{sup 7} M {sub ?} yr{sup 1}). We revised our stability limit of hydrogen shell burning, which will be useful for binary evolution calculations toward Type Ia supernovae.

  8. QUARK-NOVAE IN LOW-MASS X-RAY BINARIES. II. APPLICATION TO G87-7 AND TO GRB 110328A

    SciTech Connect (OSTI)

    Ouyed, Rachid; Staff, Jan; Jaikumar, Prashanth

    2011-12-20

    We propose a simple model explaining two outstanding astrophysical problems related to compact objects: (1) that of stars such as G87-7 (alias EG 50) that constitute a class of relatively low-mass white dwarfs (WDs) which nevertheless fall away from the C/O composition and (2) that of GRB 110328A/Swift J164449.3+57345 which showed spectacularly long-lived strong X-ray flaring, posing a challenge to standard gamma-ray burst models. We argue that both these observations may have an explanation within the unified framework of a quark-nova (QN) occurring in a low-mass X-ray binary (LMXB; neutron star (NS)-WD). For LMXBs, where the binary separation is sufficiently tight, ejecta from the exploding NS triggers nuclear burning in the WD on impact, possibly leading to Fe-rich composition compact WDs with mass 0.43 M{sub Sun} < M{sub WD} < 0.72 M{sub Sun }, reminiscent of G87-7. Our results rely on the assumption, which ultimately needs to be tested by hydrodynamic and nucleosynthesis simulations, that under certain circumstances the WD can avoid the thermonuclear runaway. For heavier WDs (i.e., M{sub WD} > 0.72 M{sub Sun }) experiencing the QN shock, degeneracy will not be lifted when carbon burning begins, and a sub-Chandrasekhar Type Ia supernova may result in our model. Under slightly different conditions and for pure He WDs (i.e., M{sub WD} < 0.43 M{sub Sun }), the WD is ablated and its ashes raining down on the quark star (QS) leads to accretion-driven X-ray luminosity with energetics and duration reminiscent of GRB 110328A. We predict additional flaring activity toward the end of the accretion phase if the QS turns into a black hole.

  9. X-shooter Science Verification Proposal A special co-moving white dwarfmain sequence pair

    E-Print Network [OSTI]

    Liske, Jochen

    and HD 122750 form a co-moving proper motion pair (PM in mas/yr: RA, Dec, from UCAC-2): WD 1401-147 (-170: The white dwarf (WD) WD 1401-147 and the main sequence (MS) star HD 122750 form a co-moving proper motion laboratory to study the complex physics of pulsating WDs in a quantitative manner. Scientific Case: WD 1401-147

  10. A&A manuscript no. (will be inserted by hand later)

    E-Print Network [OSTI]

    ?ur Radioastronomie, Auf dem H?ugel 69, D­53121 Bonn, Germany (driebe@speckle.mpifr­bonn.mpg.de; bloecker@speckle.mpifr­bonn.mpg calculated a grid of evolutionary models for white dwarfs with helium cores (He­WDs) and investigated the occurrence of hydrogen­shell flashes due to unstable hy­ drogen burning via CNO cycling. Our calculations

  11. Zeeman tomography of magnetic white dwarfs IV. The complex field structure of the polars EF Eri, BL Hyi, and CP Tuc

    E-Print Network [OSTI]

    K. Beuermann; F. Euchner; K. Reinsch; S. Jordan; B. T. Gaensicke

    2006-10-26

    The magnetic fields of the accreting white dwarfs (WDs) in magnetic cataclysmic variables (mCVs) determine the accretion geometries, the emission properties, and the secular evolution of these objects. We determine the structure of the surface magnetic fields of the WDs primaries in magnetic CVs using Zeeman tomography. Our study is based on orbital-phase resolved optical flux and circular polarization spectra of the polars EF Eri, BL Hyi, and CP Tuc obtained with FORS1 at the ESO VLT. An evolutionary algorithm is used to synthesize best fits to these spectra from an extensive database of pre-computed Zeeman spectra. The general approach has been described in previous papers of this series. The results achieved with simple geometries as centered or offset dipoles are not satisfactory. Significantly improved fits are obtained for multipole expansions that are truncated at degree l(max)=3 or 5 and include all tesseral and sectoral components with 0CP Tuc and the ranges of field strength covered are similar for the dipole and multipole models, but only the latter provide access to accreting matter at the right locations on the WD. The results suggest that the field geometries of the WDs in short-period mCVs are quite complex with strong contributions from multipoles higher than the dipole in spite of a typical age of the WDs in CVs in excess of 1 Gyr. It is feasible to derive the surface field structure of an accreting WD from phase-resolved low-state circular spectropolarimetry of sufficiently high signal-to-noise ratio. The fact that independent information is available on the strength and direction of the field in the accretion spot from high-state observations helps in unraveling the global field structure.

  12. Deflagrations in hybrid CONe white dwarfs: a route to explain the faint Type Iax supernova 2008ha

    E-Print Network [OSTI]

    Kromer, M; Pakmor, R; Ruiter, A J; Hillebrandt, W; Marquardt, K S; Roepke, F K; Seitenzahl, I R; Sim, S A; Taubenberger, S

    2015-01-01

    Stellar evolution models predict the existence of hybrid white dwarfs (WDs) with a carbon-oxygen core surrounded by an oxygen-neon mantle. Being born with masses ~1.1 Msun, hybrid WDs in a binary system may easily approach the Chandrasekhar mass (MCh) by accretion and give rise to a thermonuclear explosion. Here, we investigate an off-centre deflagration in a near-MCh hybrid WD under the assumption that nuclear burning only occurs in carbon-rich material. Performing hydrodynamics simulations of the explosion and detailed nucleosynthesis post-processing calculations, we find that only 0.014 Msun of material is ejected while the remainder of the mass stays bound. The ejecta consist predominantly of iron-group elements, O, C, Si and S. We also calculate synthetic observables for our model and find reasonable agreement with the faint Type Iax SN 2008ha. This shows for the first time that deflagrations in near-MCh WDs can in principle explain the observed diversity of Type Iax supernovae. Leaving behind a near-MCh...

  13. Population synthesis of accreting white dwarfs: II. X-ray and UV emission

    E-Print Network [OSTI]

    Chen, Hai-Liang; Yungelson, L R; Gilfanov, M; Han, Zhanwen

    2015-01-01

    Accreting white dwarfs (WDs) with non-degenerate companions are expected to emit in soft X-rays and the UV, if accreted H-rich material burns stably. They are an important component of the unresolved emission of elliptical galaxies, and their combined ionizing luminosity may significantly influence the optical line emission from warm ISM. In an earlier paper we modeled populations of accreting WDs, first generating WD with main-sequence, Hertzsprung gap and red giant companions with the population synthesis code \\textsc{BSE}, and then following their evolution with a grid of evolutionary tracks computed with \\textsc{MESA}. Now we use these results to estimate the soft X-ray (0.3-0.7keV), H- and He II-ionizing luminosities of nuclear burning WDs and the number of super-soft X-ray sources for galaxies with different star formation histories. For the starburst case, these quantities peak at $\\sim 1$ Gyr and decline by $\\sim 1-3$ orders of magnitude by the age of 10 Gyr. For stellar ages of $\\sim$~10 Gyr, predict...

  14. MULTIPLICITY OF NOVA ENVELOPE SOLUTIONS AND OCCURRENCE OF OPTICALLY THICK WINDS

    SciTech Connect (OSTI)

    Kato, Mariko; Hachisu, Izumi E-mail: hachisu@ea.c.u-tokyo.ac.jp

    2009-07-10

    We revisit the occurrence condition of optically thick winds reported by Kato in 1985 and Kato and Hachisu in 1989 who mathematically examined nova envelope solutions with an old opacity and found that optically thick winds are accelerated only in massive white dwarfs (WDs) of {approx}>0.9 M{sub sun}. With the OPAL opacity we find that the optically thick wind occurs for {approx}>0.6 M{sub sun} WDs and that the occurrence of winds depends not only on the WD mass but also on the ignition mass. When the ignition mass is larger than a critical value, winds are suppressed by a density-inversion layer. Such a static solution can be realized in WDs of mass {approx}0.6-0.7 M{sub sun}. We propose that sequences consisting only of static solutions correspond to slow evolutions in symbiotic novae like PU Vul because PU Vul shows no indication of strong winds in a long-lasted flat peak followed by a very slow decline in its light curve.

  15. Assessing Potential Exposure from Truck Transport of Low-level Radioactive Waste to the Nevada Test Site

    SciTech Connect (OSTI)

    J. Miller; D. Shafer; K. Gray; B. Church; S. Campbell; B. Holz

    2005-08-01

    Since 1980, over 651,558 m{sup 3} (23,000,000 ft{sup 3}) of low-level radioactive waste (LLW) have been disposed of at the Nevada Test Site (NTS) by shallow land burial. Since 1988, the majority of this waste has been generated at other United States (U.S.) Department of Energy (DOE) and Department of Defense (DoD) sites and facilities in the U.S. Between fiscal year (FY) 2002 and the publication date, the volumes of LLW being shipped by truck to the NTS increased sharply with the accelerated closure of DOE Environmental Management (EM) Program sites (DOE, 2002). The NTS is located 105 km (65 mi) northwest of Las Vegas, Nevada, in the U.S. There continue to be public concerns over the safety of LLW shipments to the NTS. They can be broadly divided into two categories: (1) the risk of accidents involving trucks traveling on public highways; and (2) whether residents along transportation routes receive cumulative exposure from individual LLW shipments that pose a long-term health risk. The DOE and U.S. Department of Transportation (DOT) regulations ensure that radiation exposure from truck shipments to members of the public is negligible. Nevertheless, particularly in rural communities along transportation routes in Utah and Nevada, there is a perceived risk from members of the public about cumulative exposure, particularly when ''Main Street'' and the routes being used by LLW trucks are one in the same. To provide an objective assessment of gamma radiation exposure to members of the public from LLW transport by truck, the Desert Research Institute (DRI) and the DOE, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) established a stationary and automated array of four pressurized ion chambers (PICs) in a vehicle pullout for LLW trucks to pass through just outside the entrance to the NTS. The PICs were positioned at a distance of 1.0 m (3.3 ft) from the sides of the truck trailer and at a height of 1.5 m (5.0 ft) to simulate conditions that a member of the public (Turner, 1995) might experience if a truck were to pass while the person was on the side of the road, or if a truck were to come to a stop at a stoplight in one of the smaller towns along the transportation routes. The 1.0-m (3.3-ft) distance also allowed for comparison with gamma readings of trucks taken with portable, hand-held instruments at the two LLW disposal sites at the NTS: the Area 5 Radioactive Waste Management Complex (RWMC) and the Area 3 Radioactive Waste Management Site (RWMS). The purpose in automating the system was to provide the most objective and consistent measurement and calculation of radiation exposure from the trucks possible. The array was set up in November 2002 and equipment was tested and calibrated over the next two months. Data collection on trucks began on February 13, 2003, and continued to the end of December 2003. In all, external gamma readings were collected from 1,012 of the 2,260 trucks that delivered LLW to the NTS during this period. Because DOE could not contractually require waste generators to participate in the study, the database is biased toward voluntary participants; however, data were collected from the 10 generators that represented 92 percent of the LLW shipments to the NTS during the study period, with another eight generators accounting for the balance of the shipments. Because of the voluntary nature of the participation, the identity of the waste generators is not used in the report. Previous studies on potential exposure to the public from transporting LLW to the NTS either relied on calculated exposures (Davis et al., 2002) or was based on a small population of trucks (e.g., 88) where a relatively high-background value of 50 microRoentgens per hour (R/h) (background value measured at the LLW disposal sites) were subtracted from the gross reading of the truck trailer as measured by portable, handheld instruments (Gertz, 2001). The dataset that resulted from the DRI study is the largest collection of measurements of LLW trucks in transit of which the authors are aware.

  16. Assessing Potential Exposure from Truck Transport of Low-level Radioactive Waste to the Nevada Test Site

    SciTech Connect (OSTI)

    Miller, J; Shafer, D; Gray, K; Church, B; Campbell, S; Holtz, B.

    2005-08-15

    Since 1980, over 651,558 m{sup 3} (23,000,000 ft{sup 3}) of low-level radioactive waste (LLW) have been disposed of at the Nevada Test Site (NTS) by shallow land burial. Since 1988, the majority of this waste has been generated at other United States (U.S.) Department of Energy (DOE) and Department of Defense (DoD) sites and facilities in the U.S. Between fiscal year (FY) 2002 and the publication date, the volumes of LLW being shipped by truck to the NTS increased sharply with the accelerated closure of DOE Environmental Management (EM) Program sites (DOE, 2002). The NTS is located 105 km (65 mi) northwest of Las Vegas, Nevada, in the U.S. There continue to be public concerns over the safety of LLW shipments to the NTS. They can be broadly divided into two categories: (1) the risk of accidents involving trucks traveling on public highways; and (2) whether residents along transportation routes receive cumulative exposure from individual LLW shipments that pose a long-term health risk. The DOE and U.S. Department of Transportation (DOT) regulations ensure that radiation exposure from truck shipments to members of the public is negligible. Nevertheless, particularly in rural communities along transportation routes in Utah and Nevada, there is a perceived risk from members of the public about cumulative exposure, particularly when ''Main Street'' and the routes being used by LLW trucks are one in the same. To provide an objective assessment of gamma radiation exposure to members of the public from LLW transport by truck, the Desert Research Institute (DRI) and the DOE, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) established a stationary and automated array of four pressurized ion chambers (PICs) in a vehicle pullout for LLW trucks to pass through just outside the entrance to the NTS. The PICs were positioned at a distance of 1.0 m (3.3 ft) from the sides of the truck trailer and at a height of 1.5 m (5.0 ft) to simulate conditions that a member of the public (Turner, 1995) might experience if a truck were to pass while the person was on the side of the road, or if a truck were to come to a stop at a stoplight in one of the smaller towns along the transportation routes. The 1.0-m (3.3-ft) distance also allowed for comparison with gamma readings of trucks taken with portable, hand-held instruments at the two LLW disposal sites at the NTS: the Area 5 Radioactive Waste Management Complex (RWMC) and the Area 3 Radioactive Waste Management Site (RWMS). The purpose in automating the system was to provide the most objective and consistent measurement and calculation of radiation exposure from the trucks possible. The array was set up in November 2002 and equipment was tested and calibrated over the next two months. Data collection on trucks began on February 13, 2003, and continued to the end of December 2003. In all, external gamma readings were collected from 1,012 of the 2,260 trucks that delivered LLW to the NTS during this period. Because DOE could not contractually require waste generators to participate in the study, the database is biased toward voluntary participants; however, data were collected from the 10 generators that represented 92 percent of the LLW shipments to the NTS during the study period, with another eight generators accounting for the balance of the shipments. Because of the voluntary nature of the participation, the identity of the waste generators is not used in the report. Previous studies on potential exposure to the public from transporting LLW to the NTS either relied on calculated exposures (Davis et al., 2002) or was based on a small population of trucks (e.g., 88) where a relatively high-background value of 50 microRoentgens per hour ({micro}R/h) (background value measured at the LLW disposal sites) were subtracted from the gross reading of the truck trailer as measured by portable, handheld instruments (Gertz, 2001). The dataset that resulted from the DRI study is the largest collection of measurements of LLW trucks in transit of which the authors are aware.

  17. The accretion of solar material onto white dwarfs: No mixing with core material implies that the mass of the white dwarf is increasing

    SciTech Connect (OSTI)

    Starrfield, Sumner

    2014-04-15

    Cataclysmic Variables (CVs) are close binary star systems with one component a white dwarf (WD) and the other a larger cooler star that fills its Roche Lobe. The cooler star is losing mass through the inner Lagrangian point of the binary and some unknown fraction of this material is accreted by the WD. One consequence of the WDs accreting material, is the possibility that they are growing in mass and will eventually reach the Chandrasekhar Limit. This evolution could result in a Supernova Ia (SN Ia) explosion and is designated the Single Degenerate Progenitor (SD) scenario. This paper is concerned with the SD scenario for SN Ia progenitors. One problem with the single degenerate scenario is that it is generally assumed that the accreting material mixes with WD core material at some time during the accretion phase of evolution and, since the typical WD has a carbon-oxygen CO core, the mixing results in large amounts of carbon and oxygen being brought up into the accreted layers. The presence of enriched carbon causes enhanced nuclear fusion and a Classical Nova explosion. Both observations and theoretical studies of these explosions imply that more mass is ejected than is accreted. Thus, the WD in a Classical Nova system is losing mass and cannot be a SN Ia progenitor. However, the composition in the nuclear burning region is important and, in new calculations reported here, the consequences to the WD of no mixing of accreted material with core material have been investigated so that the material involved in the explosion has only a Solar composition. WDs with a large range in initial masses and mass accretion rates have been evolved. I find that once sufficient material has been accreted, nuclear burning occurs in all evolutionary sequences and continues until a thermonuclear runaway (TNR) occurs and the WD either ejects a small amount of material or its radius grows to about 10{sup 12} cm and the evolution is ended. In all cases where mass ejection occurs, the mass of the ejecta is far less than the mass of the accreted material. Therefore, all the WDs are growing in mass. It is also found that the accretion time to explosion can be sufficiently short for a 1.0M{sub ?} WD that recurrent novae can occur on a low mass WD. This mass is lower than typically assumed for the WDs in recurrent nova systems. Finally, the predicted surface temperatures when the WD is near the peak of the explosion imply that only the most massive WDs will be significant X-ray emitters at this time.

  18. Evaluation of Department of Energy-Held Potential Greater-Than-Class C Low-Level Radioactive Waste. Revision 1

    SciTech Connect (OSTI)

    NONE

    1994-09-01

    A number of commercial facilities have generated potential greater-than-Class C low-level radioactive waste (GTCC LLW), and, through contractual arrangements with the US Department of Energy (DOE) or for health and safety reasons, DOE is storing the waste. This report presents the results of an assessment conducted by the GTCC LLW Management Program to consider specific circumstances under which DOE accepted the waste, and to determine whether disposal in a facility licensed by the US Nuclear Regulatory Commission, or by DOE in a nonlicensed facility, is appropriate. Input from EG&G Idaho, Inc., and DOE Idaho Operations Office legal departments concerning the disposal requirements of this waste were the basis for the decision process used in this report.

  19. Test Plan: Phase 1 demonstration of 3-phase electric arc melting furnace technology for vitrifying high-sodium content low-level radioactive liquid wastes

    SciTech Connect (OSTI)

    Eaton, W.C.

    1995-05-31

    This document provides a test plan for the conduct of electric arc vitrification testing by a vendor in support of the Hanford Tank Waste Remediation System (TWRS) Low-Level Waste (LLW) Vitrification Program. The vendor providing this test plan and conducting the work detailed within it [one of seven selected for glass melter testing under Purchase Order MMI-SVV-384216] is the US Bureau of Mines, Department of the Interior, Albany Research Center, Albany, Oregon. This test plan is for Phase I activities described in the above Purchase Order. Test conduct includes feed preparation activities and melting of glass with Hanford LLW Double-Shell Slurry Feed waste simulant in a 3-phase electric arc (carbon electrode) furnace.

  20. Test plan for evaluation of plasma melter technology for vitrification of high-sodium content low-level radioactive liquid wastes

    SciTech Connect (OSTI)

    McLaughlin, D.F.; Lahoda, E.J.; Gass, W.R.; D`Amico, N.

    1994-10-20

    This document provides a test plan for the conduct of plasma arc vitrification testing by a vendor in support of the Hanford Tank Waste Remediation System (TWRS) Low-Level Waste (LLW) Vitrification Program. The vendor providing this test plan and conducting the work detailed within it [one of seven selected for glass melter testing under Purchase Order MMI-SVV-384212] is the Westinghouse Science and Technology Center (WSTC) in Pittsburgh, PA. WSTC authors of the test plan are D. F. McLaughlin, E. J. Lahoda, W. R. Gass, and N. D`Amico. The WSTC Program Manager for this test is D. F. McLaughlin. This test plan is for Phase I activities described in the above Purchase Order. Test conduct includes melting of glass frit with Hanford LLW Double-Shell Slurry Feed waste simulant in a plasma arc fired furnace.

  1. Initial Activation Assessment for ARIES Compact Stellarator Power Plant

    SciTech Connect (OSTI)

    El-Guebaly, L.; Wilson, P.; Paige, D. [University of Wisconsin, Fusion Technology Institute (United States)] (and others)

    2005-04-15

    As the safety assessment frequently requires knowledge of the activation parameters, we estimated the highest possible activity, decay heat, and waste disposal rating on the time scale after shutdown for the compact stellarator power plant ARIES-CS. We selected two widely different systems employing SiC/SiC composites and low-activation ferritic steel (FS) as structural materials. Our results show that components of both systems qualify as Class C low-level waste (LLW) at the end of a 100 y storage period following the decommissioning of the plant. The SiC blanket, vacuum vessel, and magnet offer very low waste disposal rating to the extent that a Class A LLW seems achievable for these components. On this last point, we discussed the split between the Class A and Class C wastes, emphasizing our motivation to lower the level of ARIES-CS radioactive waste.

  2. Assessment of Preferred Depleted Uranium Disposal Forms

    SciTech Connect (OSTI)

    Croff, A.G.; Hightower, J.R.; Lee, D.W.; Michaels, G.E.; Ranek, N.L.; Trabalka, J.R.

    2000-06-01

    The Department of Energy (DOE) is in the process of converting about 700,000 metric tons (MT) of depleted uranium hexafluoride (DUF6) containing 475,000 MT of depleted uranium (DU) to a stable form more suitable for long-term storage or disposal. Potential conversion forms include the tetrafluoride (DUF4), oxide (DUO2 or DU3O8), or metal. If worthwhile beneficial uses cannot be found for the DU product form, it will be sent to an appropriate site for disposal. The DU products are considered to be low-level waste (LLW) under both DOE orders and Nuclear Regulatory Commission (NRC) regulations. The objective of this study was to assess the acceptability of the potential DU conversion products at potential LLW disposal sites to provide a basis for DOE decisions on the preferred DU product form and a path forward that will ensure reliable and efficient disposal.

  3. Performance assessment for the disposal of low-level waste in the 200 West Area Burial Grounds

    SciTech Connect (OSTI)

    Wood, M.I.; Khaleel, R.; Rittmann, P.D.; Lu, A.H.; Finfrock, S.H.; DeLorenzo, T.H. [Westinghouse Hanford Co., Richland, WA (United States); Serne, R.J.; Cantrell, K.J. [Pacific Northwest Lab., Richland, WA (United States)

    1995-06-01

    This document reports the findings of a performance assessment (PA) analysis for the disposal of solid low-level radioactive waste (LLW) in the 200 West Area Low-Level Waste Burial Grounds (LLBG) in the northwest corner of the 200 West Area of the Hanford Site. This PA analysis is required by US Department of Energy (DOE) Order 5820.2A (DOE 1988a) to demonstrate that a given disposal practice is in compliance with a set of performance objectives quantified in the order. These performance objectives are applicable to the disposal of DOE-generated LLW at any DOE-operated site after the finalization of the order in September 1988. At the Hanford Site, DOE, Richland Operations Office (RL) has issued a site-specific supplement to DOE Order 5820.2A, DOE-RL 5820.2A (DOE 1993), which provides additiona I ce objectives that must be satisfied.

  4. Numerical evaluation of monofil and subtle-layered evapotranspiration (ET) landfill caps

    SciTech Connect (OSTI)

    Wilson, G.V.; Henley, M.; Valceschini, R.

    1998-01-01

    The US Department of Energy/Nevada Operations Office (DOE/NV) has identified the need to design a low-level waste (LLW) closure cap for the arid conditions at the Nevada Test Site (NTS). As a result of concerns for subsidence impacting the cover, DOE/NV redesigned the LLW cover from one containing a `hard` infiltration barrier that would likely fail, to a `soft` (ET) cover that is sufficiently deep to accommodate the hydrologic problems of subsidence. An ET cover is one that does not contain hydrologic barrier layers but relies on soil-water retention and sufficient thickness to store water until evapotranspiration (ET) can remove the moisture. Subtle layering within an ET cap using the native soil could be environmentally beneficial and cost effective.

  5. Mixed waste characterization, treatment & disposal focus area

    SciTech Connect (OSTI)

    NONE

    1996-08-01

    The mission of the Mixed Waste Characterization, Treatment, and Disposal Focus Area (referred to as the Mixed Waste Focus Area or MWFA) is to provide treatment systems capable of treating DOE`s mixed waste in partnership with users, and with continual participation of stakeholders, tribal governments, and regulators. The MWFA deals with the problem of eliminating mixed waste from current and future storage in the DOE complex. Mixed waste is waste that contains both hazardous chemical components, subject to the requirements of the Resource Conservation and Recovery Act (RCRA), and radioactive components, subject to the requirements of the Atomic Energy Act. The radioactive components include transuranic (TRU) and low-level waste (LLW). TRU waste primarily comes from the reprocessing of spent fuel and the use of plutonium in the fabrication of nuclear weapons. LLW includes radioactive waste other than uranium mill tailings, TRU, and high-level waste, including spent fuel.

  6. Vitrification of Polyvinyl Chloride Waste from Korean Nuclear Power Plants

    SciTech Connect (OSTI)

    Sheng, Jiawei [Kyoto University (Japan); Choi, Kwansik [Nuclear Environment Technology Institute (Korea, Republic of); Yang, Kyung-Hwa [Nuclear Environment Technology Institute (Korea, Republic of); Lee, Myung-Chan [Nuclear Environment Technology Institute (Korea, Republic of); Song, Myung-Jae [Nuclear Environment Technology Institute (Korea, Republic of)

    2000-02-15

    Vitrification is considered as an economical and safe treatment technology for low-level radioactive waste (LLW) generated from nuclear power plants (NPPs). Korea is in the process of preparing for its first ever vitrification plant to handle LLW from its NPPs. Polyvinyl chloride (PVC) has the largest volume of dry active wastes and is the main waste stream to treat. Glass formulation development for PVC waste is the focus of study. The minimum additive waste stabilization approach has been utilized in vitrification. It was found that glasses can incorporate a high content of PVC ash (up to 50 wt%), which results in a large volume reduction. A glass frit, KEP-A, was developed to vitrify PVC waste after the optimization of waste loading, melt viscosity, melting temperature, and chemical durability. The KEP-A could satisfactorily vitrify PVC with a waste loading of 30 to 50 wt%. The PVC-frit was tolerant of variations in waste composition.

  7. GTS Duratek, Phase I Hanford low-level waste melter tests: 100-kg melter offgas report

    SciTech Connect (OSTI)

    Eaton, W.C. [Westinghouse Hanford Co., Richland, WA (United States)] [Westinghouse Hanford Co., Richland, WA (United States)

    1995-11-01

    A multiphase program was initiated in 1994 to test commercially available melter technologies for the vitrification of the low-level waste (LLW) stream from defense wastes stored in underground tanks at the Hanford Site in southeastern Washington State. Phase 1 of the melter demonstration tests using simulated LLW was completed during fiscal year 1995. This document is the 100-kg melter offgas report on testing performed by GTS Duratek, Inc., in Columbia, Maryland. GTS Duratek (one of the seven vendors selected) was chosen to demonstrate Joule heated melter technology under WHC subcontract number MMI-SVV-384215. The document contains the complete offgas report on the 100-kg melter as prepared by Parsons Engineering Science, Inc. A summary of this report is also contained in the GTS Duratek, Phase I Hanford Low-Level Waste Melter Tests: Final Report (WHC-SD-WM-VI-027).

  8. Performance test report for the 1000 kg melter

    SciTech Connect (OSTI)

    Eaton, W.C.

    1995-11-01

    A multiphase program was initiated in 1994 to test commercially available melter technologies for the vitrification of the low-level waste (LLW) stream from defense wastes stored in underground tanks at the Hanford Site in southeastern Washington State. Phase 1 of the melter demonstration tests using simulated LLW was completed during fiscal year 1995. This document is the 100 kg melter offgas report on testing performed by GTS Duratek Inc., in Columbia, Maryland. GTS Duratek (one of the seven vendors selected) was chosen to demonstrate Joule heated melter technology under WHC subcontract number MMI-SVV- 384215. The document contains the complete offgas report on the 100 kg melter as prepared by Parsons Engineering Science, Inc. A summary of this report is also contained in the ``GTS Duratek, Phase 1 Hanford Low-Level Waste Melter Tests: Final Report`` (WHC-SD-VI-027).

  9. Wittgenstein on Practice and the Myth of the Giving

    E-Print Network [OSTI]

    Hurley, Susan

    1995-01-01

    WITTGENSTEIN ON PRACTICE AND THE MYTH OF THE GIVING by S. L. HURLEY The Lindley Lecture The University of Kansas 1995 'l'lw E. II. Line! It, \\ ltnwl iall.tcttllt:'ih ip h md "~'' ntahli-.hed in I ~I l l in mcmon ol I t tH'-.t I I. I... indh-\\. ( ham l'llnr of llw l' n iwr:.il\\ of K

  10. Overview of Nevada Test Site Radioactive and Mixed Waste Disposal Operations

    SciTech Connect (OSTI)

    J.T. Carilli; S.K. Krenzien; R.G. Geisinger; S.J. Gordon; B. Quinn

    2009-03-01

    The U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office Environmental Management Program is responsible for carrying out the disposal of on-site and off-site generated low-level radioactive waste (LLW) and low-level radioactive mixed waste (MW) at the Nevada Test Site (NTS). Core elements of this mission are ensuring safe and cost-effective disposal while protecting workers, the public, and the environment. This paper focuses on the impacts of new policies, processes, and opportunities at the NTS related to LLW and MW. Covered topics include: the first year of direct funding for NTS waste disposal operations; zero tolerance policy for non-compliant packages; the suspension of mixed waste disposal; waste acceptance changes; DOE Consolidated Audit Program (DOECAP) auditing; the 92-Acre Area closure plan; new eligibility requirements for generators; and operational successes with unusual waste streams.

  11. Siting study for a consolidated waste capability at Los Alamos National Laboratory

    SciTech Connect (OSTI)

    Booth, Steven Richard

    2011-01-26

    Decision analysis was used to rank alternative sites for a new Consolidated Waste Capability (CWC) to replace current hazardous solid waste operations (hazardous/chemical, mixed lowlevel, transuranic, and low-level waste) at Los Alamos National Laboratory's TA-54 Area G. An original list of 21 site alternatives was pre-screened to ten sites that were assessed using the analytical hierarchy process with five top-level criteria and fifteen sub-criteria. Three passes of the analysis were required to assess different site scenarios: 1) a fully consolidated CWC with both transfer/storage and LL W disposal in one location (45 acre minimum), 2) CWC transfer/storage only (12 acre minimum), and 3) LLW disposal only (33 acre minimum). The top site choice for all three options is TA-63/52/46; the second choice is TA-18/36. TA-54 East, Zone 4 also deserves consideration as a LLW disposal site.

  12. Estimation of natural ground water recharge for the performance assessment of a low-level waste disposal facility at the Hanford Site

    SciTech Connect (OSTI)

    Rockhold, M.L.; Fayer, M.J.; Kincaid, C.T.; Gee, G.W.

    1995-03-01

    In 1994, the Pacific Northwest Laboratory (PNL) initiated the Recharge Task, under the PNL Vitrification Technology Development (PVTD) project, to assist Westinghouse Hanford Company (WHC) in designing and assessing the performance of a low-level waste (LLW) disposal facility for the US Department of Energy (DOE). The Recharge Task was established to address the issue of ground water recharge in and around the LLW facility and throughout the Hanford Site as it affects the unconfined aquifer under the facility. The objectives of this report are to summarize the current knowledge of natural ground water recharge at the Hanford Site and to outline the work that must be completed in order to provide defensible estimates of recharge for use in the performance assessment of this LLW disposal facility. Recharge studies at the Hanford Site indicate that recharge rates are highly variable, ranging from nearly zero to greater than 100 mm/yr depending on precipitation, vegetative cover, and soil types. Coarse-textured soils without plants yielded the greatest recharge. Finer-textured soils, with or without plants, yielded the least. Lysimeters provided accurate, short-term measurements of recharge as well as water-balance data for the soil-atmosphere interface and root zone. Tracers provided estimates of longer-term average recharge rates in undisturbed settings. Numerical models demonstrated the sensitivity of recharge rates to different processes and forecast recharge rates for different conditions. All of these tools (lysimetry, tracers, and numerical models) are considered vital to the development of defensible estimates of natural ground water recharge rates for the performance assessment of a LLW disposal facility at the Hanford Site.

  13. The Diversity of Objections to Inequality

    E-Print Network [OSTI]

    Scanlon, T. M.

    1996-01-01

    them may seem to involve unacceptable costs not only in economic efficiency and the quality of the products of a culture but also in individual fulfillment. One thing individuals naturally and reasonably want is to develop their talents... Professor ofPhilosophy, Ulli vcrsity of Pittsburgh. 1990. 'Justice and the Good Life." By Ronald Dworkin, Professor of jurisprudence, Oxford University, New York University L'lw School. 1991. "Equality or Pliority?" By Derek Parfit, Professor...

  14. Coupled Environmental Processes in the Mojave Desert and Implications for ET Covers as Stable Landforms

    SciTech Connect (OSTI)

    D. Shafer; M. Y oung; S. Zitzer; E. McDonald; T. Caldwell

    2006-01-18

    Monolayer evapotranspiration (ET) covers are the baseline method for closure of disposal sites for low-level radioactive waste (LLW), mixed LLW, and transuranic (TRU) waste at the Nevada Test Site (NTS). The regulatory timeline is typically 1,000 years for LLW and 10,000 years for TRU waste. Covers for such waste have different technical considerations than those with shorter timelines because they are subject to environmental change for longer periods of time, and because the environmental processes are often coupled. To evaluate these changes, four analog sites (approximately 30, 1,000 to 2,000, 7,000 to 12,500, and 125,000 years in age) on the NTS were analyzed to address the early post-institutional control period (the youngest site), the 1,000-year compliance period for disposal of LLW, and the 10,000-year period for TRU waste. Tests included soil texture, structure, and morphology; surface soil infiltration and hydraulic conductivity; vegetation and faunal surveys; and literature reviews. Separate measurements were made in plant undercanopy and intercanopy areas. The results showed a progressive increase in silt and clay content of surface soils with age. Changes in soil texture and structure led to a fivefold decline in saturated hydraulic conductivity in intercanopy areas, but no change in undercanopies, which were subject to bioturbation. These changes may have been responsible for the reduction in total plant cover, most dramatically in intercanopy areas, primarily because more precipitation either runs off the site or is held nearer to the surface where plant roots are less common. The results suggest that covers may evolve over longer timeframes to stable landforms that minimize the need for active maintenance.

  15. Immobilization and Waste Form Product Acceptance for Low Level and TRU Waste Forms

    SciTech Connect (OSTI)

    Holtzscheiter, E.W. [Westinghouse Savannah River Company, AIKEN, SC (United States); Harbour, J.R.

    1998-05-01

    The Tanks Focus Area is supporting technology development in immobilization of both High Level (HLW) and Low Level (LLW) radioactive wastes. The HLW process development at Hanford and Idaho is patterned closely after that of the Savannah River (Defense Waste Processing Facility) and West Valley Sites (West Valley Demonstration Project). However, the development and options open to addressing Low Level Waste are diverse and often site specific. To start, it is important to understand the breadth of Low Level Wastes categories.

  16. Linking RESRAD-OFFSITE and HYDROGEOCHEM Model for Performance Assessment of Low-Level Radioactive Waste Disposal Facility - 13429

    SciTech Connect (OSTI)

    Lin, Wen-Sheng [Hydrotech Research Institute, National Taiwan University, Taiwan (China)] [Hydrotech Research Institute, National Taiwan University, Taiwan (China); Yu, Charley; Cheng, Jing-Jy; Kamboj, Sunita; Gnanapragasam, Emmanuel [Argonne National Laboratory, Argonne, IL 60439 (United States)] [Argonne National Laboratory, Argonne, IL 60439 (United States); Liu, Chen-Wuing [Department of Bioenvironmental Systems Engineering, National Taiwan University, Taiwan (China)] [Department of Bioenvironmental Systems Engineering, National Taiwan University, Taiwan (China); Li, Ming-Hsu [Institute of Hydrological and Oceanic Sciences, National Central University, Taiwan (China)] [Institute of Hydrological and Oceanic Sciences, National Central University, Taiwan (China)

    2013-07-01

    Performance assessments are crucial steps for the long-term radiological safety requirements of low-level waste (LLW) disposal facility. How much concentration of radionuclides released from the near-field to biosphere and what radiation exposure levels of an individual can influence on the satisfactory performance of the LLW disposal facility and safety disposal environment. Performance assessment methodology for the radioactive waste disposal consists of the reactive transport modeling of safety-concerned radionuclides released from the near-field to the far-field, and the potential exposure pathways and the movements of radionuclides through the geosphere, biosphere and man of which the accompanying dose. Therefore, the integration of hydrogeochemical transport model and dose assessment code, HYDROGEOCHEM code and RESRAD family of codes is imperative. The RESRAD family of codes such as RESRAD-OFFSITE computer code can evaluate the radiological dose and excess cancer risk to an individual who is exposed while located within or outside the area of initial (primary) contamination. The HYDROGEOCHEM is a 3-D numerical model of fluid flow, thermal, hydrologic transport, and biogeochemical kinetic and equilibrium reactions in saturated and unsaturated media. The HYDROGEOCHEM model can also simulate the crucial geochemical mechanism, such as the effect of redox processes on the adsorption/desorption, hydrogeochemical influences on concrete degradation, adsorption/desorption of radionuclides (i.e., surface complexation model) between solid and liquid phase in geochemically dynamic environments. To investigate the safety assessment of LLW disposal facility, linking RESRAD-OFFSITE and HYDROGEOCHEM model can provide detailed tools of confidence in the protectiveness of the human health and environmental impact for safety assessment of LLW disposal facility. (authors)

  17. Performance assessment for continuing and future operations at solid waste storage area 6

    SciTech Connect (OSTI)

    1997-09-01

    This revised performance assessment (PA) for the continued disposal operations at Solid Waste Storage Area (SWSA) 6 on the Oak Ridge Reservation (ORR) has been prepared to demonstrate compliance with the performance objectives for low-level radioactive waste (LLW) disposal contained in the US Department of Energy (DOE) Order 5820.2A. This revised PA considers disposal operations conducted from September 26, 1988, through the projects lifetime of the disposal facility.

  18. Justification Of The Use Of Boreholes For Disposal Of Sealed Radiological Sources

    SciTech Connect (OSTI)

    Zarling, John [Los Alamos National Laboratory; Johnson, Peter [Los Alamos National Laboratory

    2008-01-01

    Soon there will be only 14 states in two compacts that are able to dispose of Low Level Waste (LLW): the Northwest and Rocky Mountain compact with disposal options in Richland, Washington, and the Atlantic compact with disposal options in Barnwell, South Carolina. How do states not in one of the two compacts dispose of their LLW? The Off-Site Source Recovery Project can take possession and dispose of some of the unwanted transuranic sources at the Waste Isolation Pilot Plant (WIPP). However, there will be no path forward for states outside of the two compacts for disposal of their non-transuranic LLW. A solution that has been much discussed, debated and researched, but has not been put into wide scale practice, is the borehole disposal concept. It is the author's position that companies that drill and explore for oil have been disposing of sources in borehole-like structures for years. It should be noted that these companies are not purposely disposing of these sources, but the sources are irretrievable and must be abandoned. Additionally, there are Nuclear Regulatory Commission (NRC) regulations that must be followed to seal the well that contains the lost and abandoned source. According to the NRC Event Notification Reports database, there were a minimum of 29 reports of lost and abandoned sources in oil wells between December 1999 and October 2006. The sources were lost at depths between 2,018-18,887 feet, or 600-5,750 meters. The companies that are performing explorations with the aid of sealed radiological sources must follow regulation 10 CFR Part 39. Subsection 15 outlines the procedures that must be followed if sources are determined to be irretrievable and abandoned in place. If the NRC allows and has regulations in place for oil companies, why can't states and/or companies be allowed to dispose of LLW in a similar fashion?

  19. Waste Receiving and Processing Facility Module 1 Data Management System Software Requirements Specification

    SciTech Connect (OSTI)

    Brann, E.C. II

    1994-09-09

    This document provides the software requirements for Waste Receiving and Processing (WRAP) Module 1 Data Management System (DMS). The DMS is one of the plant computer systems for the new WRAP 1 facility (Project W-026). The DMS will collect, store and report data required to certify the low level waste (LLW) and transuranic (TRU) waste items processed at WRAP 1 as acceptable for shipment, storage, or disposal.

  20. Radiated waste and irradiated fuel management in western Europe

    SciTech Connect (OSTI)

    NONE

    1989-04-01

    A number of countries in Western Europe, many of which reprocess spent nuclear fuel, have or are developing storage and permanent disposal facilities for their radioactive waste. Low-Level Waste (LLW), Intermediate Level Waste (ILW) and Medium-Level Waste (MLW), Transuranic Waste (TRU), and High-Level Waste (HLW) each have unique characteristics and thus specific disposal requirements. How eight Western European countries are managing and planning for the safe and efficient disposal of nuclear waste is summarized by country.

  1. Operating Experience and Lessons Learned in the Use of Soft-Sided Packaging for Transportation and Disposal of Low Activity Radioactive Waste

    SciTech Connect (OSTI)

    Kapoor, A.; Gordon, S.; Goldston, W.

    2013-07-08

    This paper describes the operating experience and lessons learned at U.S. Department of Energy (DOE) sites as a result of an evaluation of potential trailer contamination and soft-sided packaging integrity issues related to the disposal of low-level and mixed low-level (LLW/MLLW) radioactive waste shipments. Nearly 4.3 million cubic meters of LLW/MLLW will have been generated and disposed of during fiscal year (FY) 2010 to FY 2015either at commercial disposal sites or disposal sites owned by DOE. The LLW/MLLW is packaged in several different types of regulatory compliant packaging and transported via highway or rail to disposal sites safely and efficiently in accordance with federal, state, and local regulations and DOE orders. In 1999, DOE supported the development of LLW containers that are more volumetrically efficient, more cost effective, and easier to use as compared to metal or wooden containers that existed at that time. The DOE Idaho National Engineering and Environmental Laboratory (INEEL), working in conjunction with the plastic industry, tested several types of soft-sided waste packaging systems that meet U.S. Department of Transportation requirements for transport of low specific activity and surface contaminated objects. Since then, soft-sided packaging of various capacities have been used successfully by the decontamination and decommissioning (D&D) projects to package, transport, and dispose D&D wastes throughout the DOE complex. The joint team of experts assembled by the Energy Facility Contractors Group from DOE waste generating sites, DOE and commercial waste disposal facilities, and soft-sided packaging suppliers conducted the review of soft-sided packaging operations and transportation of these packages to the disposal sites. As a result of this evaluation, the team developed several recommendations and best practices to prevent or minimize the recurrences of equipment contamination issues and proper use of soft-sided packaging for transport and disposal of waste.

  2. RESULTS FOR THE THIRD QUARTER 2011 TANK 50 WAC SLURRY SAMPLE: CHEMICAL AND RADIONUCLIDE CONTAMINANT RESULTS

    SciTech Connect (OSTI)

    Reigel, M.

    2011-10-20

    The Saltstone Facility is designed and permitted to immobilize and dispose of low-level radioactive and hazardous liquid waste (salt solution) remaining from the processing of radioactive material at the Savannah River Site. Low-level waste (LLW) streams from the Effluent Treatment Project (ETP), H-Canyon, and the decontaminated salt solution product from the Actinide Removal Process/Modular Caustic Side Solvent Extraction (CSSX) Unit (ARP/MCU) process are stored in Tank 50 until the LLW can be transferred to the Saltstone Facility for treatment and disposal. The LLW must meet the specified waste acceptance criteria (WAC) before it is processed into saltstone. The specific chemical and radionuclide contaminants and their respective WAC limits are in the current Saltstone WAC. Waste Solidification Engineering (WSE) requested that Savannah River National Laboratory (SRNL) perform quarterly analysis on saltstone samples. The concentrations of chemical and radionuclide contaminants are measured to ensure the saltstone produced during each quarter is in compliance with the current WAC. This report documents the concentrations of chemical and radionuclide contaminants for the 2011 Third Quarter samples collected from Tank 50 on July 7, 2011 and discusses those results in further detail than the previously issued results report.

  3. RESULTS FOR THE FOURTH QUARTER 2011 TANK 50 WAC SLURRY SAMPLE: CHEMICAL AND RADIONUCLIDE CONTAMINANT RESULTS

    SciTech Connect (OSTI)

    Bannochie, C.

    2012-01-31

    The Saltstone Facility is designed and permitted to immobilize and dispose of low-level radioactive and hazardous liquid waste (salt solution) remaining from the processing of radioactive material at the Savannah River Site. Low-level waste (LLW) streams from the Effluent Treatment Project (ETP), H-Canyon, and the decontaminated salt solution product from the Actinide Removal Process/Modular Caustic Side Solvent Extraction (CSSX) Unit (ARP/MCU) process are stored in Tank 50 until the LLW can be transferred to the Saltstone Facility for treatment and disposal. The LLW must meet the specified waste acceptance criteria (WAC) before it is processed into saltstone. The specific chemical and radionuclide contaminants and their respective WAC limits are in the current Saltstone WAC. Waste Solidification Engineering (WSE) requested that Savannah River National Laboratory (SRNL) perform quarterly analysis on saltstone samples. The concentrations of chemical and radionuclide contaminants are measured to ensure the saltstone produced during each quarter is in compliance with the current WAC. This report documents the concentrations of chemical and radionuclide contaminants for the 2011 Fourth Quarter samples collected from Tank 50 on October 12, 2011 and discusses those results in further detail than the previously issued results report.

  4. Letter report: Pre-conceptual design study for a pilot-scale Non-Radioactive Low-Level Waste Vitrification Facility

    SciTech Connect (OSTI)

    Thompson, R.A.; Morrissey, M.F.

    1996-03-01

    This report presents a pre-conceptual design study for a Non-Radioactive Low-Level Waste, Pilot-Scale Vitrification System. This pilot plant would support the development of a full-scale LLW Vitrification Facility and would ensure that the full-scale facility can meet its programmatic objectives. Use of the pilot facility will allow verification of process flowsheets, provide data for ensuring product quality, assist in scaling to full scale, and support full-scale start-up. The facility will vitrify simulated non-radioactive LLW in a manner functionally prototypic to the full-scale facility. This pre-conceptual design study does not fully define the LLW Pilot-Scale Vitrification System; rather, it estimates the funding required to build such a facility. This study includes identifying all equipment necessary. to prepare feed, deliver it into the melter, convert the feed to glass, prepare emissions for atmospheric release, and discharge and handle the glass. The conceived pilot facility includes support services and a structure to contain process equipment.

  5. Standardization of DOE Disposal Facilities Waste Acceptance Processes

    SciTech Connect (OSTI)

    Shrader, T. A.; Macbeth, P. J.

    2002-02-26

    On February 25, 2000, the U.S. Department of Energy (DOE) issued the Record of Decision (ROD) for the Waste Management Programmatic Environmental Impact Statement (WM PEIS) for low-level and mixed low-level wastes (LLW/ MLLW) treatment and disposal. The ROD designated the disposal sites at Hanford and the Nevada Test Site (NTS) to dispose of LLW/MLLW from sites without their own disposal facilities. DOE's Richland Operations Office (RL) and the National Nuclear Security Administration's Nevada Operations Office (NV) have been charged with effectively implementing the ROD. To accomplish this task NV and RL, assisted by their operating contractors Bechtel Nevada (BN), Fluor Hanford (FH), and Bechtel Hanford (BH) assembled a task team to systematically map out and evaluate the current waste acceptance processes and develop an integrated, standardized process for the acceptance of LLW/MLLW. A structured, systematic, analytical process using the Six Sigma system identified dispos al process improvements and quantified the associated efficiency gains to guide changes to be implemented. The review concluded that a unified and integrated Hanford/NTS Waste Acceptance Process would be a benefit to the DOE Complex, particularly the waste generators. The Six Sigma review developed quantitative metrics to address waste acceptance process efficiency improvements, and provides an initial look at development of comparable waste disposal cost models between the two disposal sites to allow quantification of the proposed improvements.

  6. Equity of commercial low-level radioactive waste disposal fees. Report to Congress

    SciTech Connect (OSTI)

    NONE

    1998-02-01

    In the Report accompanying the Fiscal Year 1997 Senate Energy and Water Development Appropriations Bill, the Senate Appropriations Committee directed the Department of Energy (DOE) to prepare a study of the costs of operating a low-level radioactive waste (LLW) disposal facility such as the one at Barnwell, South Carolina, and to determine whether LLW generators are paying equitable disposal fees. The disposal costs of four facilities are reviewed in this report, two operating facilities and two planned facilities. The operating facilities are located at Barnwell, South Carolina, and Richland, Washington. They are operated by Chem-Nuclear, LLC, (Chem-Nuclear), and US Ecology, Inc., (US Ecology), respectively. The planned facilities are expected to be built at Ward Valley, California, and Sierra Blanca, Texas. They will be operated by US Ecology and the State of Texas, respectively. This report found that disposal fees vary significantly among facilities for a variety of reasons. However, the information suggests that at each disposal facility, LLW generators pay equitable disposal fees.

  7. WRAP Module 1 waste characterization plan

    SciTech Connect (OSTI)

    Mayancsik, B.A.

    1995-01-23

    The purpose of this document is to present the characterization methodology for waste generated, processed, or otherwise the responsibility of the Waste Receiving and Processing (WRAP) Module 1 facility. The scope of this document includes all solid low level waste (LLW), transuranic (TRU), mixed waste (MW), and dangerous waste. This document is not meant to be all-inclusive of the waste processed or generated within WRAP Module 1, but to present a methodology for characterization. As other streams are identified, the method of characterization will be consistent with the other streams identified in this plan. The WRAP Module 1 facility is located in the 200 West Area of the Hanford Site. The facility`s function is two-fold. The first is to verify/characterize, treat and repackage contact handled (CH) waste currently in retrievable storage in the LLW Burial Grounds, Hanford Central Waste Complex, and the Transuranic Storage and Assay Facility (TRUSAF). The second is to verify newly generated CH TRU waste and LLW, including MW. The WRAP Module 1 facility provides NDE and NDA of the waste for both drums and boxes. The NDE is used to identify the physical contents of the waste containers to support waste characterization and processing, verification, or certification. The NDA results determine the radioactive content and distribution of the waste.

  8. Underground storage tank integrated demonstration: Evaluation of pretreatment options for Hanford tank wastes

    SciTech Connect (OSTI)

    Lumetta, G.J.; Wagner, M.J.; Colton, N.G.; Jones, E.O.

    1993-06-01

    Separation science plays a central role inn the pretreatment and disposal of nuclear wastes. The potential benefits of applying chemical separations in the pretreatment of the radioactive wastes stored at the various US Department of Energy sites cover both economic and environmental incentives. This is especially true at the Hanford Site, where the huge volume (>60 Mgal) of radioactive wastes stored in underground tanks could be partitioned into a very small volume of high-level waste (HLW) and a relatively large volume of low-level waste (LLW). The cost associated with vitrifying and disposing of just the HLW fraction in a geologic repository would be much less than those associated with vitrifying and disposing of all the wastes directly. Futhermore, the quality of the LLW form (e.g., grout) would be improved due to the lower inventory of radionuclides present in the LLW stream. In this report, we present the results of an evaluation of the pretreatment options for sludge taken from two different single-shell tanks at the Hanford Site-Tanks 241-B-110 and 241-U-110 (referred to as B-110 and U-110, respectively). The pretreatment options examined for these wastes included (1) leaching of transuranic (TRU) elements from the sludge, and (2) dissolution of the sludge followed by extraction of TRUs and {sup 90}Sr. In addition, the TRU leaching approach was examined for a third tank waste type, neutralized cladding removal waste.

  9. U.S. Bureau of Mines, Phase 1 Hanford low-level waste melter tests. Final report

    SciTech Connect (OSTI)

    Eaton, W.C. [Westinghouse Hanford Co., Richland, WA (United States); Oden, L.L.; O`Connor, W.K. [Bureau of Mines, Albany, OR (United States). Albany Research Center

    1995-11-01

    A multiphase program was initiated in 1994 to test commercially available melter technologies for the vitrification of the low-level waste (LLW) stream from defense wastes stored in underground tanks at the Hanford Site in southeastern Washington State. Phase 1 of the melter demonstration tests using simulated LLW was completed during fiscal year 1995. This document is the melter offgas report on testing performed by the U.S. Department of the Interior, Bureau of Mines, Albany Research Center in Albany, Oregon. The Bureau of Mines (one of the seven vendors selected) was chosen to demonstrate carbon electrode melter technology (also called carbon arc or electric arc) under WHC Subcontract number MMI-SVV-384216. The report contains description of the tests, observation, test data and some analysis of the data as it pertains to application of this technology for LLW vitrification. Testing consisted of melter feed preparation and three melter tests, the first of which was to fulfill the requirements of the statement of work (WHC-SD-EM-RD-044), and the second and third were to address issues identified during the first test. The document also contains summaries of the melter offgas report issued as a separate document U.S. Bureau of Mines, Phase 1 Hanford Low-Level Waste Melter Tests: Melter Offgas Report (WHC-SD-WM-VI-032).

  10. Alternative methods for dispoal of low-level radioactive wastes. Task 1. Description of methods and assessment of criteria. [Alternative methods are belowground vaults, aboveground vaults; earth mounded concrete bunkers, mined cavities, augered holes

    SciTech Connect (OSTI)

    Bennett, R.D.; Miller, W.O.; Warriner, J.B.; Malone, P.G.; McAneny, C.C.

    1984-04-01

    The study reported herein contains the results of Task 1 of a four-task study entitled Criteria for Evaluating Engineered Facilities. The overall objective of this study is to ensure that the criteria needed to evaluate five alternative low-level radioactive waste (LLW) disposal methods are available to the Nuclear Regulatory Commission (NRC) and the Agreement States. The alternative methods considered are belowground vaults, aboveground vaults, earth mounded concrete bunkers, mined cavities, and augered holes. Each of these alternatives is either being used by other countries for low-level radioactive waste (LLW) disposal or is being considered by other countries or US agencies. In this report the performance requirements are listed, each alternative is described, the experience gained with its use is discussed, and the performance capabilities of each method are addressed. Next, the existing 10 CFR Part 61 Subpart D criteria with respect to paragraphs 61.50 through 61.53, pertaining to site suitability, design, operations and closure, and monitoring are assessed for applicability to evaluation of each alternative. Preliminary conclusions and recommendations are offered on each method's suitability as an LLW disposal alternative, the applicability of the criteria, and the need for supplemental or modified criteria.

  11. Hydrologic evaluation methodology for estimating water movement through the unsaturated zone at commercial low-level radioactive waste disposal sites

    SciTech Connect (OSTI)

    Meyer, P.D.; Rockhold, M.L.; Nichols, W.E.; Gee, G.W. [Pacific Northwest Lab., Richland, WA (United States)

    1996-01-01

    This report identifies key technical issues related to hydrologic assessment of water flow in the unsaturated zone at low-level radioactive waste (LLW) disposal facilities. In addition, a methodology for incorporating these issues in the performance assessment of proposed LLW disposal facilities is identified and evaluated. The issues discussed fall into four areas: estimating the water balance at a site (i.e., infiltration, runoff, water storage, evapotranspiration, and recharge); analyzing the hydrologic performance of engineered components of a facility; evaluating the application of models to the prediction of facility performance; and estimating the uncertainty in predicted facility performance. To illustrate the application of the methodology, two examples are presented. The first example is of a below ground vault located in a humid environment. The second example looks at a shallow land burial facility located in an arid environment. The examples utilize actual site-specific data and realistic facility designs. The two examples illustrate the issues unique to humid and arid sites as well as the issues common to all LLW sites. Strategies for addressing the analytical difficulties arising in any complex hydrologic evaluation of the unsaturated zone are demonstrated.

  12. Performance of a feasibility study for remediation of WAG 6 at Oak Ridge National Laboratory

    SciTech Connect (OSTI)

    Kubarewicz, J.; Pfeffer, J. [CH2M Hill, Oak Ridge, TN (United States); Garland, S.B. II [Oak Ridge National Lab., TN (United States); Riddle, S.P. [USDOE Oak Ridge Field Office, TN (United States); Branscom, K.S. [Radian Corp., Oak Ridge, TN (United States)

    1992-10-01

    This paper describes the process of preparing a feasibility study (FS) for remediation of a low-level radioactive waste (LLW) disposal site at Oak Ridge National Laboratory (ORNL). ORNL conducts research and development and is one of three DOE-owned facilities on the Oak Ridge Reservation (ORR). Waste Area Grouping (WAG) 6 is located in Melton Valley, approximately 2 miles southwest of the plant in Roane County, Tennessee. WAG 6 includes Solid Waste Storage Area (SWSA) 6, which is still used for shallow land burial of LLW and nonradioactive materials and was the primary focus of the FS. SWSA 6 covers 68 acres, 19 of which contain wastes such as low-level radioactive liquids, solids, sludges, asbestos, and biological and associated laboratory wastes. During the first 15 years of operation, the site also received chemical wastes, but since 1986, it has been used only for LLW. Until 1986, wastes were placed in unlined trenches and auger holes, but since then, wastes have been disposed in greater confinement disposal silos, lined pipe wells and auger holes, and above-ground tumulus units. A list of the sitewide alternatives initially developed for WAG 6 remediation is presented. The alternatives combined capping, structural stabilization (dynamic compaction/grouting), waste consolidation, and groundwater collection/treatment components. In situ vitrification was also considered for areas with significant long-life source inventories.

  13. SIPS: A small modular process unit for the in-tank pretreatment of high-level wastes

    SciTech Connect (OSTI)

    Reich, M.; Powell, J.; Barletta, R. [Brookhaven National Lab., Upton, NY (United States)

    1996-12-31

    As a result of the U.S. weapons production program, there are now hundreds of large tanks containing highly radioactive wastes. Safe disposal of these wastes requires their processing and separations into a small volume of highly radioactive waste (HLW) and a much larger volume of low-level waste (LLW). The HLW waste would then be vitrified and transported to a geologic repository. To date, the principal approach proposed for the separation envisions a large, centralized process facility. The small in-tank processing system (SIPS) is a proposed new, small modular concept for the in-tank processing and separation of wastes into HLW and LLW output streams suitable for vitrification. Instead of pumping the retrieved tank wastes as a solid/liquid slurry over long distances to a centralized process facility, SIPS would employ a small process module, typically {approximately}1 m in diameter and 4 m long, which would be inserted into the tank. Over a period of {approx} 6 months, the module would process the solid/liquid materials in the tank, producing separated liquid HLW and liquid LLW output streams that are pumped away in two small-diameter ({approx}3-cm outside diameter) pipes. The SIPS module would be serviced by five auxiliary small pipes - a water feed pipe, a water feed pipe containing micron-size ferromagnetic particles, a nitric acid ({approx}3 M) feed pipe, and input/out pipes to hydraulically load/unload ion exchange beads.

  14. Chemical behavior of P,S,Cl,F, and Cr in borosilicate nuclear waste glasses

    SciTech Connect (OSTI)

    Feng, X.; Freeman, C.J.; Luey, J.; Li, Hong; Schweiger, M.J.; Gong, M. [Pacific Northwest National Lab., Richland, WA (United States)

    1996-12-31

    Vitrification is one technology being considered for the immobilization of the tank low-level nuclear waste (LLW) at Hanford site of Washington. The LLW contains approximately 80 wt% Na{sub 2}O and borosilicate glasses are being evaluated as the final waste form. Some of these tank wastes also have significant concentrations of one or more of the minor components such as chloride, fluoride, phosphate, sulfate, and chromium oxide. These minor components could cause potential problems in the vitrification of LLW, which include component volatility and the formation of segregated phases on the melt surface and in the final product. Volatility of halides (Cl{sup -}, F{sup -}) and sulfate (SO{sub 4}{sup 2-}) can accelerate corrosion of melter off-gas equipment and electrodes. Additionally, phase segregation on the melt surface can adversely affect the waste melting rate and can have adverse effects on glass durability. This paper describes a study on borosilicate glass with minor components added to the waste glass. Results of the retention behaviors of the components are discussed.

  15. In-tank pretreatment of high-level tank wastes: The SIPS system

    SciTech Connect (OSTI)

    Reich, M.; Powell, J.; Barletta, R.

    1996-03-01

    A new approach, termed SIPS (Small In-Tank Processing System), that enables the in-tank processing and separation of high-level tank wastes into high-level waste (HLW) and low-level waste (LLW) streams that are suitable for vitrification, is described. Presently proposed pretreatment systems, such as enhanced sludge washing (ESW) and TRUEX, require that the high-level tank wastes be retrieved and pumped to a large, centralized processing facility, where the various waste components are separated into a relatively small, radioactively concentrated stream (HLW), and a relatively large, predominantly non-radioactive stream (LLW). In SIPS, a small process module, typically on the order of 1 meter in diameter and 4 meters in length, is inserted into a tank. During a period of approximately six months, it processes the solid/liquid materials in the tank, separating them into liquid HLW and liquid LLW output streams that are pumped away in two small diameter (typically 3 cm o.d.) pipes. The SIPS concept appears attractive for pretreating high level wastes, since it would: (1) process waste in-situ in the tanks, (2) be cheaper and more reliable than a larger centralized facility, (3) be quickly demonstrable at full scale, (4) have less technical risk, (5) avoid having to transfer unstable slurries for long distances, and (6) be simple to decommission and dispose of. Further investigation of the SIPS concept appears desirable, including experimental testing and development of subscale demonstration units.

  16. Report for Westinghouse Hanford Company: Makeup procedures and characterization data for modified DSSF and modified remaining inventory simulated tank waste

    SciTech Connect (OSTI)

    Lokken, R.O.

    1996-03-01

    The majority of defense wastes generated from reprocessing spent reactor fuel at Hanford are stored in underground Double-Shell Tanks (DST) and in older Single-Shell Tanks (SST). The Tank Waste Remediation System (TWRS) Program has the responsibility of safely managing and immobilizing these tank wastes for disposal. A reference process flowsheet is being developed that includes waste retrieval, pretreatment, and vitrification. Melter technologies for vitrifying low-level tank wastes are being evaluated by Westinghouse Hanford Company. Chemical simulants are being used in the technology testing. For the first phase of low-level waste (LLW) vitrification simulant development, two waste stream compositions were investigated. The first waste simulant was based on the analyses of six tanks of double-shell slurry feed (DSSF) waste and on the projected composition of the wastes exiting the pretreatment operations. A simulant normalized to 6 M sodium was based on the anticipated chemical concentrations after ion exchange and initial separations. The same simulant concentrated to 10 M sodium would represent a waste that had been concentrated by evaporation to reduce the overall volume. The second LLW simulant, referred to as the remaining inventory (RI), included wastes not included in the DSSF tanks and the projected LLW fraction of single-shell tank wastes.

  17. Preliminary assessment of candidate immobilization technologies for retrieved single-shell tank wastes

    SciTech Connect (OSTI)

    Wiemers, K.D.; Mendel, J.E.; Kruger, A.A.; Bunnell, L.R.; Mellinger, G.B.

    1992-01-01

    This report describes the initial work that has been performed to select technologies for immobilization of wastes that may be retrieved from Hanford single-shell tanks (SSTs). Two classes of waste will require immobilization. One is the combined high-level waste/transuranic (HLW/TRU) fraction, the other the low-level waste (LLW) fraction. A number of potential immobilization technologies are identified for each class of waste. Immobilization technologies were initially selected based on a number of considerations, including (1) the waste loading that could likely be achieved within the constraint of producing acceptable waste forms, (2) process flexibility (primarily compatibility with anticipated waste variability), (3) process complexity, and (4) state of development. Immobilization technologies selected for further development include the following: for HLW/TRU waste -- borosilicate glass, lead-iron phosphate glass, glass-calcine composites, glass-ceramics, and cement based forms; for non-denitrated LLW -- grout, laxtex-modified concrete, and polyethylene; and for denitrated LLW -- silicate glass, phosphate glass, and clay calcination or tailored ceramic in various matrices.

  18. Hanford low-level waste process chemistry testing data package

    SciTech Connect (OSTI)

    Smith, H.D.; Tracey, E.M.; Darab, J.G.; Smith, P.A.

    1996-03-01

    Recently, the Tri-Party Agreement (TPA) among the State of Washington Department of Ecology, U.S. Department of Energy (DOE) and the US Environmental Protection Agency (EPA) for the cleanup of the Hanford Site was renegotiated. The revised agreement specifies vitrification as the encapsulation technology for low level waste (LLW). A demonstration, testing, and evaluation program underway at Westinghouse Hanford Company to identify the best overall melter-system technology available for vitrification of Hanford Site LLW to meet the TPA milestones. Phase I is a {open_quotes}proof of principle{close_quotes} test to demonstrate that a melter system can process a simulated highly alkaline, high nitrate/nitrite content aqueous LLW feed into a glass product of consistent quality. Seven melter vendors were selected for the Phase I evaluation: joule-heated melters from GTS Duratek, Incorporated (GDI); Envitco, Incorporated (EVI); Penberthy Electomelt, Incorporated (PEI); and Vectra Technologies, Incorporated (VTI); a gas-fired cyclone burner from Babcock & Wilcox (BCW); a plasma torch-fired, cupola furnace from Westinghouse Science and Technology Center (WSTC); and an electric arc furnace with top-entering vertical carbon electrodes from the U.S. Bureau of Mines (USBM).

  19. Distribution coefficient values describing iodine, neptunium, selenium, technetium, and uranium sorption to Hanford sediments. Supplement 1

    SciTech Connect (OSTI)

    Kaplan, D.I.; Seme, R.J.

    1995-03-01

    Burial of vitrified low-level waste (LLW) in the vadose zone of the Hanford Site is being considered as a long-term disposal option. Regulations dealing with LLW disposal require that performance assessment (PA) analyses be conducted. Preliminary modeling efforts for the Hanford Site LLW PA were conducted to evaluate the potential health risk of a number of radionuclides, including Ac, Am, C, Ce, Cm, Co, Cs, Eu, 1, Nb, Ni, Np, Pa, Pb, Pu, Ra, Ru, Se, Sn, Sr, Tc, Th, U, and Zr (Piepho et al. 1994). The radionuclides, {sup 129}I, {sup 237}Np, {sup 79}Se, {sup 99}Tc, and {sup 234,235,238}U, were identified as posing the greatest potential health hazard. It was also determined that the outcome of these simulations were very sensitive to the parameter describing the extent to which radionuclides sorbed to the subsurface matrix, described as a distribution coefficient (K{sub d}). The distribution coefficient is a ratio of the radionuclide concentration associated with the solid phase to that in the liquid phase. The literature-derived K{sub d} values used in these simulations were conservative, i.e., lowest values within the range of reasonable values used to provide an estimate of the maximum health threat. Thus, these preliminary modeling results reflect a conservative estimate rather than a best estimate of what is likely to occur. The potential problem with providing only a conservative estimate is that it may mislead us into directing resources to resolve nonexisting problems.

  20. Selected radionuclides important to low-level radioactive waste management

    SciTech Connect (OSTI)

    1996-11-01

    The purpose of this document is to provide information to state representatives and developers of low level radioactive waste (LLW) management facilities about the radiological, chemical, and physical characteristics of selected radionuclides and their behavior in the environment. Extensive surveys of available literature provided information for this report. Certain radionuclides may contribute significantly to the dose estimated during a radiological performance assessment analysis of an LLW disposal facility. Among these are the radionuclides listed in Title 10 of the Code of Federal Regulations Part 61.55, Tables 1 and 2 (including alpha emitting transuranics with half-lives greater than 5 years). This report discusses these radionuclides and other radionuclides that may be significant during a radiological performance assessment analysis of an LLW disposal facility. This report not only includes essential information on each radionuclide, but also incorporates waste and disposal information on the radionuclide, and behavior of the radionuclide in the environment and in the human body. Radionuclides addressed in this document include technetium-99, carbon-14, iodine-129, tritium, cesium-137, strontium-90, nickel-59, plutonium-241, nickel-63, niobium-94, cobalt-60, curium -42, americium-241, uranium-238, and neptunium-237.

  1. PUREX low-level waste radionuclide characterization

    SciTech Connect (OSTI)

    Ellis, M.W.; LeBaron, G.J.

    1995-01-16

    The PUREX low-level waste (LLW) radionuclide characterization document describes the methodology for the characterization of solid LLW and solid low-level mixed waste (MW) with the respect to radiological characteristics. This document only serves as an overview of the PUREX radionuclide characterization methodology and provides specific examples for how the radionuclide distribution is derived. It would be impractical to provide all background information in this document. If further clarification and background information is required, consult the PUREX Regulatory Compliance group files. This document applies to only that waste generated in or is the responsibility of the PUREX facilities. The US Department of Energy (DOE) establishes the requirements for radioactive solid waste in DOE Order 5820.2A Radioactive Waste Management. Chapters 2 and 3 from DOE Order 5820.2A requires that generators of solid wastes in the LLW categories and the radioactive mixed waste subcategories: (1) identify the major radionuclides in each solid waste matrix and (2) determine the radionuclide concentrations and waste classes of their solid wastes. In addition, the Order also requires each generator to carry out a compliance program that ensures the proper certification of the solid waste generated.

  2. Implementation plan for WRAP Module 1 operational readiness review

    SciTech Connect (OSTI)

    Irons, L.G.

    1994-11-04

    The Waste Receiving and Processing Module 1 (WRAP 1) will be used to receive, sample, treat, and ship contact-handled (CH) transuranic (TRU), low-level waste (LLW), and low-level mixed waste (LLMW) to storage and disposal sites both on the Hanford site and off-site. The primary mission of WRAP 1 is to characterize and certify CH waste in 55-gallon and 85-gallon drums; and its secondary function is to certify CH waste standard waste boxes (SWB) and boxes of similar size for disposal. The WRAP 1 will provide the capability for examination (including x-ray, visual, and contents sampling), limited treatment, repackaging, and certification of CH suspect-TRU waste in 55-gallon drums retrieved from storage, as well as newly generated CH LLW and CH TRU waste drums. The WRAP 1 will also provide examination (X-ray and visual only) and certification of CH LLW and CH TRU waste in small boxes. The decision to perform an Operational Readiness Review (ORR) was made in accordance with WHC-CM-5-34, Solid Waste Disposal Operations Administration, Section 1.4, Operational Readiness Activities. The ORR will ensure plant and equipment readiness, management and personnel readiness, and management programs readiness for the initial startup of the facility. This implementation plan is provided for defining the conduct of the WHC ORR.

  3. Waste management facilities cost information for transportation of radioactive and hazardous materials

    SciTech Connect (OSTI)

    Feizollahi, F.; Shropshire, D.; Burton, D.

    1995-06-01

    This report contains cost information on the U.S. Department of Energy (DOE) Complex waste streams that will be addressed by DOE in the programmatic environmental impact statement (PEIS) project. It describes the results of the task commissioned by DOE to develop cost information for transportation of radioactive and hazardous waste. It contains transportation costs for most types of DOE waste streams: low-level waste (LLW), mixed low-level waste (MLLW), alpha LLW and alpha MLLW, Greater-Than-Class C (GTCC) LLW and DOE equivalent waste, transuranic (TRU) waste, spent nuclear fuel (SNF), and hazardous waste. Unit rates for transportation of contact-handled (<200 mrem/hr contact dose) and remote-handled (>200 mrem/hr contact dose) radioactive waste are estimated. Land transportation of radioactive and hazardous waste is subject to regulations promulgated by DOE, the U.S. Department of Transportation (DOT), the U.S. Nuclear Regulatory Commission (NRC), and state and local agencies. The cost estimates in this report assume compliance with applicable regulations.

  4. Waste-Incidental-to-Reprocessing Evaluation for the West Valley Demonstration Project Vitrification Melter - 12167

    SciTech Connect (OSTI)

    McNeil, Jim; Kurasch, David; Sullivan, Dan; Crandall, Thomas

    2012-07-01

    The Department of Energy (DOE) has determined that the vitrification melter used in the West Valley Demonstration Project can be disposed of as low-level waste (LLW) after completion of a waste-incidental-to-reprocessing evaluation performed in accordance with the evaluation process of DOE Manual 435.1-1, Radioactive Waste Management Manual. The vitrification melter - which consists of a ceramic lined, electrically heated box structure - was operated for more than 5 years melting and fusing high-level waste (HLW) slurry and glass formers and pouring the molten glass into 275 stainless steel canisters. Prior to shutdown, the melter was decontaminated by processing low-activity decontamination flush solutions and by extracting molten glass from the melter cavity. Because it could not be completely emptied, residual radioactivity conservatively estimated at approximately 170 TBq (4,600 Ci) remained in the vitrification melter. To establish whether the melter was incidental to reprocessing, DOE prepared an evaluation to demonstrate that the vitrification melter: (1) had been processed to remove key radionuclides to the maximum extent technically and economically practical; (2) would be managed to meet safety requirements comparable to the performance objectives for LLW established by the Nuclear Regulatory Commission (NRC); and (3) would be managed by DOE in accordance with DOE's requirements for LLW after it had been incorporated in a solid physical form with radionuclide concentrations that do not exceed the NRC concentration limits for Class C LLW. DOE consulted with the NRC on the draft evaluation and gave other stakeholders an opportunity to submit comments before the determination was made. The NRC submitted a request for additional information in connection with staff review of the draft evaluation; DOE provided the additional information and made improvements to the evaluation, which was issued in January 2012. DOE considered the NRC Technical Evaluation Report as well as comments received from other stakeholders prior to making its determination that the vitrification melter is not HLW, does not require permanent isolation in a geologic repository, and can be disposed of as LLW. (authors)

  5. Technical area status report for waste destruction and stabilization

    SciTech Connect (OSTI)

    Dalton, J.D.; Harris, T.L.; DeWitt, L.M.

    1993-08-01

    The Office of Environmental Restoration and Waste Management (EM) was established by the Department of Energy (DOE) to direct and coordinate waste management and site remediation programs/activities throughout the DOE complex. In order to successfully achieve the goal of properly managing waste and the cleanup of the DOE sites, the EM was divided into five organizations: the Office of Planning and Resource Management (EM-10); the Office of Environmental Quality Assurance and Resource Management (EM-20); the Office of Waste Operations (EM-30); the Office of Environmental Restoration (EM-40); and the Office of Technology and Development (EM-50). The mission of the Office of Technology Development (OTD) is to develop treatment technologies for DOE`s operational and environmental restoration wastes where current treatment technologies are inadequate or not available. The Mixed Waste Integrated Program (MWIP) was created by OTD to assist in the development of treatment technologies for the DOE mixed low-level wastes (MLLW). The MWIP has established five Technical Support Groups (TSGs) whose purpose is to identify, evaluate, and develop treatment technologies within five general technical areas representing waste treatment functions from initial waste handling through generation of final waste forms. These TSGs are: (1) Front-End Waste Handling, (2) Physical/Chemical Treatment, (3) Waste Destruction and Stabilization, (4) Second-Stage Destruction and Offgas Treatment, and (5) Final Waste Forms. This report describes the functions of the Waste Destruction and Stabilization (WDS) group. Specifically, the following items are discussed: DOE waste stream identification; summary of previous efforts; summary of WDS treatment technologies; currently funded WDS activities; and recommendations for future activities.

  6. Dusty disks around central stars of planetary nebulae

    SciTech Connect (OSTI)

    Clayton, Geoffrey C. [Department of Physics and Astronomy, Louisiana State University, Baton Rouge, LA 70803 (United States); De Marco, Orsola [Department of Physics and Astronomy, Macquarie University, Sydney, NSW 2109 (Australia); Nordhaus, Jason [Center for Computational Relativity and Gravitation, and National Technical Institute for the Deaf, Rochester Institute of Technology, Rochester, NY 14623 (United States); Green, Joel [Department of Astronomy, The University of Texas, 1 University Station, C1400, Austin, TX 78712-0259 (United States); Rauch, Thomas; Werner, Klaus [Institute for Astronomy and Astrophysics, Kepler Center for Astro and Particle Physics, Eberhard Karls University, Sand 1, D-72076 Tbingen (Germany); Chu, You-Hua, E-mail: gclayton@fenway.phys.lsu.edu, E-mail: orsola@science.mq.edu.au, E-mail: nordhaus@astro.rit.edu, E-mail: joel@astro.as.utexas.edu, E-mail: rauch@astro.uni-tuebingen.de, E-mail: werner@astro.uni-tuebingen.de, E-mail: chu@astro.uiuc.edu [Department of Astronomy, University of Illinois at Urbana-Champaign, 1002 West Green Street, Urbana, IL 61801 (United States)

    2014-06-01

    Only a few percent of cool, old white dwarfs (WDs) have infrared excesses interpreted as originating in small hot disks due to the infall and destruction of single asteroids that come within the star's Roche limit. Infrared excesses at 24 ?m were also found to derive from the immediate vicinity of younger, hot WDs, most of which are still central stars of planetary nebulae (CSPNe). The incidence of CSPNe with this excess is 18%. The Helix CSPN, with a 24 ?m excess, has been suggested to have a disk formed from collisions of Kuiper belt-like objects (KBOs). In this paper, we have analyzed an additional sample of CSPNe to look for similar infrared excesses. These CSPNe are all members of the PG 1159 class and were chosen because their immediate progenitors are known to often have dusty environments consistent with large dusty disks. We find that, overall, PG 1159 stars do not present such disks more often than other CSPNe, although the statistics (five objects) are poor. We then consider the entire sample of CSPNe with infrared excesses and compare it to the infrared properties of old WDs, as well as cooler post-asymptotic giant branch (AGB) stars. We conclude with the suggestion that the infrared properties of CSPNe more plausibly derive from AGB-formed disks rather than disks formed via the collision of KBOs, although the latter scenario cannot be ruled out. Finally, there seems to be an association between CSPNe with a 24 ?m excess and confirmed or possible binarity of the central star.

  7. Head-on collisions of binary white dwarf-neutron stars: Simulations in full general relativity

    SciTech Connect (OSTI)

    Paschalidis, Vasileios; Etienne, Zachariah; Liu, Yuk Tung; Shapiro, Stuart L.

    2011-03-15

    We simulate head-on collisions from rest at large separation of binary white dwarf-neutron stars (WDNSs) in full general relativity. Our study serves as a prelude to our analysis of the circular binary WDNS problem. We focus on compact binaries whose total mass exceeds the maximum mass that a cold-degenerate star can support, and our goal is to determine the fate of such systems. A fully general relativistic hydrodynamic computation of a realistic WDNS head-on collision is prohibitive due to the large range of dynamical time scales and length scales involved. For this reason, we construct an equation of state (EOS) which captures the main physical features of neutron stars (NSs) while, at the same time, scales down the size of white dwarfs (WDs). We call these scaled-down WD models 'pseudo-WDs (pWDs)'. Using pWDs, we can study these systems via a sequence of simulations where the size of the pWD gradually increases toward the realistic case. We perform two sets of simulations; One set studies the effects of the NS mass on the final outcome, when the pWD is kept fixed. The other set studies the effect of the pWD compaction on the final outcome, when the pWD mass and the NS are kept fixed. All simulations show that after the collision, 14%-18% of the initial total rest mass escapes to infinity. All remnant masses still exceed the maximum rest mass that our cold EOS can support (1.92M{sub {center_dot}}), but no case leads to prompt collapse to a black hole. This outcome arises because the final configurations are hot. All cases settle into spherical, quasiequilibrium configurations consisting of a cold NS core surrounded by a hot mantle, resembling Thorne-Zytkow objects. Extrapolating our results to realistic WD compactions, we predict that the likely outcome of a head-on collision of a realistic, massive WDNS system will be the formation of a quasiequilibrium Thorne-Zytkow-like object.

  8. Permitting plan for the immobilized low-activity waste project

    SciTech Connect (OSTI)

    Deffenbaugh, M.L.

    1997-09-04

    This document addresses the environmental permitting requirements for the transportation and interim storage of the Immobilized Low-Activity Waste (ILAW) produced during Phase 1 of the Hanford Site privatization effort. Tri-Party Agreement (TPA) Milestone M-90 establishes a new major milestone, and associated interim milestones and target dates, governing acquisition and/or modification of facilities necessary for: (1) interim storage and disposal of Tank Waste Remediation Systems (TWRS) immobilized low-activity tank waste (ILAW) and (2) interim storage of TWRS immobilized HLW (IHLW) and other canistered high-level waste forms. Low-activity waste (LAW), low-level waste (LLW), and high-level waste (HLW) are defined by the TWRS, Hanford Site, Richland, Washington, Final Environmental Impact Statement (EIS) DOE/EIS-0189, August 1996 (TWRS, Final EIS). By definition, HLW requires permanent isolation in a deep geologic repository. Also by definition, LAW is ``the waste that remains after separating from high-level waste as much of the radioactivity as is practicable that when solidified may be disposed of as LLW in a near-surface facility according to the NRC regulations.`` It is planned to store/dispose of (ILAW) inside four empty vaults of the five that were originally constructed for the Group Program. Additional disposal facilities will be constructed to accommodate immobilized LLW packages produced after the Grout Vaults are filled. The specifications for performance of the low-activity vitrified waste form have been established with strong consideration of risk to the public. The specifications for glass waste form performance are being closely coordinated with analysis of risk. RL has pursued discussions with the NRC for a determination of the classification of the Hanford Site`s low-activity tank waste fraction. There is no known RL action to change law with respect to onsite disposal of waste.

  9. The Future Through the Past: The Use of Analog Sites for Design Criteria and Long Term Performance Assessment of Evapotranspiration Landfill Covers

    SciTech Connect (OSTI)

    Shafer, D. S.; Miller, J. J.; Young, M. H.; Edwards, S. C.; Rawlinson, S. E.

    2002-02-26

    There is growing support for using evapotranspiration (ET) covers for closure of low-level waste (LLW) and other types of waste disposal sites, particularly in the lower latitude arid regions of the western United States. At the Nevada Test Site (NTS), monolayer ET covers are the baseline technology for closure of LLW and mixed LLW cells. To better predict the long-term performance of monolayer ET covers, as well as to identify design criteria that will potentially improve their performance, the properties of, and processes occurring on, analog sites for ET covers on the NTS are being studied. The project is funded through the Subsurface Contaminants Focus Area of the U.S. Department of Energy. Four analog sites on the NTS have been selected to predict performance of ET covers over a 1,000-year compliance period. Two sites are relatively recently disturbed (within the last 50 years) and have been selected to evaluate processes and changes on ET covers for the early period after active cover maintenance is discontinued. Two other sites, late to mid-Holocene in age, are intended as analogs for the end of the compliance period (1,000 years or more); both surfaces are abandoned alluvial/colluvial deposits. The history of the early post-institutional control analog sites are being evaluated by an archaeologist to help determine when the sites were last disturbed or modified, and the mode of disturbance to help set baseline conditions. Similar to other ''landforms,'' ET covers will evolve over time because of pedogenic, biotic, and climatic processes. Properties of analog sites that could affect ET water balance performance will be evaluated to help understand ET cover performance over time.

  10. Lessons Learned from Raw Treatment in the Slovak Republic - Minimization for Final Disposal

    SciTech Connect (OSTI)

    Hanusik, V.; Hladky, E.; Krajc, T.; Pekar, A.; Stubna, M.; Urbanec, M. [Milan Zatkulak, VUJE, a.s., Trnava (Slovakia); Ehn, L.; Kover, M.; Remias, V.; Slezak, M. [JAVYS, a.s., Bohunice (Slovakia)

    2008-07-01

    This paper is referring about the utilization of technologies for the treatment and conditioning of low and intermediate level RAW from operation and decommissioning of nuclear facilities in Slovakia. This experience represents more than 116 reactor years of NPP operation, mainly of NPPs equipped with VVER 440 reactors, 30 years of decommissioning activities, 27 years of development and operation of technologies for the treatment and conditioning of RAW and 7 years of LLW and ILW final repository operation. These technologies are located in two localities: Jaslovske Bohunice and Mochovce. The complex treatment and conditioning center (cementation, bituminization, incineration, vitrification, fragmentation and compacting) for almost all types of radioactive waste is located in Jaslovske Bohunice NPP site. The treatment and conditioning center for liquid radioactive waste (cementation and bituminization) and the surface type repository for LLW and ILW final disposal are located in Mochovce area. The treated waste forms are disposed to repository in cubical Fiber Reinforced Concrete (FRC) containers. The experience from the phase of technology development and the phase of technology modifications for various types of RAW, the experience from long term operation of technologies and the experience from transportation of original and packed wastes are described in this paper. The method of optimally combined technology utilization in order to maximize the radionuclide inventory at the same time with respect of disposal safety limitations of repository is described, too. The significant RAW volume reduction for final disposal was achieved through mediation of the combination of treatment and conditioning technologies. The disposal of treated RAW in cubic FRC containers allowed the optimal utilization of volume and radiological capacity of LLW and ILW repository in Mochovce and the fulfillment of determined safety requirements at the same time. (authors)

  11. Experimental data and analysis to support the design of an ion-exchange process for the treatment of Hanford tank waste supernatant liquids

    SciTech Connect (OSTI)

    Kurath, D.E.; Bray, L.A.; Brooks, K.P.; Brown, G.N.; Bryan, S.A.; Carlson, C.D.; Carson, K.J.; DesChane, J.R.; Elovich, R.J.; Kim, A.Y.

    1994-12-01

    Hanford`s 177 underground storage tanks contain a mixture of sludge, salt cake, and alkaline supernatant liquids. Disposal options for these wastes are high-level waste (HLW) glass for disposal in a repository or low-level waste (LLW) glass for onsite disposal. Systems-engineering studies show that economic and environmental considerations preclude disposal of these wastes without further treatment. Difficulties inherent in transportation and disposal of relatively large volumes of HLW make it impossible to vitrify all of the tank waste as HLW. Potential environmental impacts make direct disposal of all of the tank waste as LLW glass unacceptable. Although the pretreatment and disposal requirements are still being defined, most pretreatment scenarios include retrieval of the aqueous liquids, dissolution of the salt cakes, and washing of the sludges to remove soluble components. Most of the cesium is expected to be in the aqueous liquids, which are the focus of this report on cesium removal by ion exchange. The main objectives of the ion-exchange process are removing cesium from the bulk of the tank waste (i.e., decontamination) and concentrating the separated cesium for vitrification. Because exact requirements for removal of {sup 137}Cs have not yet been defined, a range of removal requirements will be considered. This study addresses requirements to achieve {sup 137}Cs levels in LLW glass between (1) the Nuclear Regulatory Commission (NRC) Class C (10 CFR 61) limit of 4600 Ci/m{sup 3} and (2) 1/10th of the NRC Class A limit of 1 Ci/m{sup 3} i.e., 0.1/m{sup 3}. The required degrees of separation of cesium from other waste components is a complex function involving interactions between the design of the vitrification process, waste form considerations, and other HLW stream components that are to be vitrified.

  12. Proposal for Construction/Demonstration/Implementation of A Material Handling System

    SciTech Connect (OSTI)

    Jim Jnatt

    2001-08-24

    Vortec Corporation, the United States Enrichment Corporation (USEC) and DOE/Paducah propose to complete the technology demonstration and the implementation of the Material Handling System developed under Contract Number DE-AC21-92MC29120. The demonstration testing and operational implementation will be done at the Paducah Gaseous Diffusion Plant. The scope of work, schedule and cost for the activities are included in this proposal. A description of the facility to be constructed and tested is provided in Exhibit 1, attached. The USEC proposal for implementation at Paducah is presented in Exhibit 2, and the commitment letters from the site are included in Exhibit 3. Under our agreements with USEC, Bechtel Jacobs Corporation and DOE/Paducah, Vortec will be responsible for the construction of the demonstration facility as documented in the engineering design package submitted under Phase 4 of this contract on August 9, 2001. USEC will have responsibility for the demonstration testing and commercial implementation of the plant. The demonstration testing and initial commercial implementation of the technology will be achieved by means of a USEC work authorization task with the Bechtel Jacobs Corporation. The initial processing activities will include the processing of approximately 4,250 drums of LLW. Subsequent processing of LLW and TSCA/LLW will be done under a separate contract or work authorization task. To meet the schedule for commercial implementation, it is important that the execution of the Phase 4 project option for construction of the demonstration system be executed as soon as possible. The schedule we have presented herein assumes initiation of the construction phase by the end of September 2001. Vortec proposes to complete construction of the demonstration test system for an estimated cost of $3,254,422. This price is based on the design submitted to DOE/NETL under the Phase 4 engineering design deliverable (9 august 2001). The cost is subject to the assumptions and conditions identified in Section 6 of this proposal.

  13. Operational and Regulatory Performance of Waste Crate Assay Systems at RFETS

    SciTech Connect (OSTI)

    Clapham, M. J.; Franco, J.; Simpson, A.; Santo, J.; Menlove, H. O.; Durel, F. M.

    2003-02-27

    As Rocky Flats Environmental Technology Site (RFETS) approaches its closure target of 2006 emphasis for Non-Destructive Assay (NDA) has shifted from small waste package assay systems towards larger systems that are designed to accommodate Standard Waste Boxes (SWB) and larger Low Level Waste (LLW) containers. To this end, Kaiser Hill, with the support of BNFL Instruments, Inc. (BII) and Los Alamos National Laboratory (LANL), has recently deployed two new crate assay systems. These systems provide the capacity to meet the assay requirements associated with the Deactivation and Decommissioning (D&D) at RFETS. The Super High Efficiency Neutron Coincidence Counting System (SuperHENC) was designed and fabricated as a collaborative effort between RFETS, LANL and BII. The purpose of this counter is to provide a WIPP certified assay capability for SWBs with a sensitivity that allows for TRU/LLW sorting. The SuperHENC has been in operation since early 2001. The BII Multi-Purpose Crate Counter (MPCC) is based on the Imaging Passive Active Neutron (IPAN) technology. This counter was designed to provide diverse capacity for WIPP certified assay of SWBs and to provide assay capability for larger LLW crates that are generated at RFETS. The MPCC has been in operation since early 2002. In order to meet the requirement for measurement of the WIPP tracked radionuclides, both systems incorporate a BII Gamma Energy Analysis sub-system. The unique Energy Times Attenuation (ETA) method is used to provide isotopic mass fractions for diverse waste streams. These systems were the first, and at this time the only, waste crate assay systems that have achieved WIPP certification. This represents a significant achievement given that the performance criteria applied to the measurements of large crates is identical to the criteria for 55-gallon (208 liter) drums. They are now both fully operational at RFETS and continue to successfully support the site closure mission.

  14. Closure Strategy for a Waste Disposal Facility with Multiple Waste Types and Regulatory Drivers at the Nevada Test Site

    SciTech Connect (OSTI)

    L. Desotell; D. Wieland; V. Yucel; G. Shott; J. Wrapp

    2008-03-01

    The U.S. Department of Energy, National Security Administration Nevada Site Office (NNSA/NSO) is planning to close the 92-Acre Area of the Area 5 Radioactive Waste Management Site (RWMS) at the Nevada Test Site (NTS), which is about 65 miles northwest of Las Vegas, Nevada. Closure planning for this facility must take into account the regulatory requirements for a diversity of waste streams, disposal and storage configurations, disposal history, and site conditions. This paper provides a brief background of the Area 5 RWMS, identifies key closure issues, and presents the closure strategy. Disposals have been made in 25 shallow excavated pits and trenches and 13 Greater Confinement Disposal (GCD) boreholes at the 92-Acre Area since 1961. The pits and trenches have been used to dispose unclassified low-level waste (LLW), low-level mixed waste (LLMW), and asbestiform waste, and to store classified low-level and low-level mixed materials. The GCD boreholes are intermediate-depth disposal units about 10 feet (ft) in diameter and 120 ft deep. Classified and unclassified high-specific activity LLW, transuranic (TRU), and mixed TRU are disposed in the GCD boreholes. TRU waste was also disposed inadvertently in trench T-04C. Except for three disposal units that are active, all pits and trenches are operationally covered with 8-ft thick alluvium. The 92-Acre Area also includes a Mixed Waste Disposal Unit (MWDU) operating under Resource Conservation and Recovery Act (RCRA) Interim Status, and an asbestiform waste unit operating under a state of Nevada Solid Waste Disposal Site Permit. A single final closure cover is envisioned over the 92-Acre Area. The cover is the evapotranspirative-type cover that has been successfully employed at the NTS. Closure, post-closure care, and monitoring must meet the requirements of the following regulations: U.S. Department of Energy Order 435.1, Title 40 Code of Federal Regulations (CFR) Part 191, Title 40 CFR Part 265, Nevada Administrative Code (NAC) 444.743, RCRA requirements as incorporated into NAC 444.8632, and the Federal Facility Agreement and Consent Order (FFACO). A grouping of waste disposal units according to waste type, location, and similarity in regulatory requirements identified six closure units: LLW Unit, Corrective Action Unit (CAU) 111 under FFACO, Asbestiform LLW Unit, Pit 3 MWDU, TRU GCD Borehole Unit, and TRU Trench Unit. The closure schedule of all units is tied to the closure schedule of the Pit 3 MWDU under RCRA.

  15. Low-level waste management program and interim waste operations technologies

    SciTech Connect (OSTI)

    Mezga, L.J.

    1983-01-01

    The Department of Energy currently supports an integrated technology development and transfer program aimed at ensuring that the technology necessary for the safe management and disposal of LLW by the commercial and defense sectors is available. The program focuses on five technical areas: (1) corrective measures technology, (2) improved shallow land burial technology, (3) greater confinement disposal technology, (4) model development and validation, and (5) treatment methods for problem wastes. The results of activities in these areas are reported in the open literature and the Proceedings of the LLWMP Annual Participants Information Meeting.

  16. Surficial Geology and Landscape Development in Northern Frenchman Flat, Interim Summary and Soil Data

    SciTech Connect (OSTI)

    Raytheon Services Nevada Environmental Restoration and Waste Management Division

    1995-09-01

    This report summarizes geologic studies by Raytheon Services Nevada near the Area 5 Radioactive Waste Management Site at the Nevada Test Site. These studies are part of a program to satisfy data needs of (1) the Greater Confinement Disposal (GCD) Program Performance Assessment (PA), (2) the low-level waste (LLW) PA, and (3) the Resource Conservation and Recovery Act (RCRA) permit application. The geologic studies were integrated into a single program that worked toward a landscape evolution model of northern Frenchman Flat, with more detailed geologic studies of particular topics as needed. Only the Holocene tectonism and surficial geology components of the landscape model are presented in this report.

  17. Entombment: It is Time to Reconsider this Technology

    SciTech Connect (OSTI)

    Birk, Sandra Margaret; Hanson, Robert Gail; Vernon, Donald Keith

    2000-09-01

    It is time to reconsider entombment of nuclear reactors and other facilities. Decommissioning worker exposure and safety, transportation, cost, potential loss of LLW disposal capacity, and need for strong technical basis are shared drivers for the renewed interest in developing the entombment D&D option. Entombment relies on retarding the release of radionuclides for a very long period, a number of factors must be considered prior to selection and implementation of entombment. A technical basis for addressing and evaluating these factors with associated stakeholder acceptance of the technology is needed before entombment becomes an accepted D&D option.

  18. Low-level waste program technical strategy

    SciTech Connect (OSTI)

    Bledsoe, K.W.

    1994-10-01

    The Low-Level Waste Technical Strategy document describes the mechanisms which the Low-Level Waste Program Office plans to implement to achieve its mission. The mission is to manage the receipt, immobilization, packaging, storage/disposal and RCRA closure (of the site) of the low-level Hanford waste (pretreated tank wastes) in an environmentally sound, safe and cost-effective manner. The primary objective of the TWRS Low-level waste Program office is to vitrify the LLW fraction of the tank waste and dispose of it onsite.

  19. Cotton Breeding of the Bureau of Plant Industry U.S. Department of Agriculture and the Texas Experiment Station.

    E-Print Network [OSTI]

    Bennett, R. L. (Robert Love)

    1905-01-01

    and first fruit limbs are hivh up from ground. l'lw ~,larrt llas a colnpaet appesmnre b;lt that IS clue to the many primary'limbs. Plants of this type or structure are late, fru~t slo~\\~ly and are not ada~ted to boll weevil conditions. demand... that will have a greater per cent of lint t I to 33, which is now the general yield, is greatly desired and would gely increase the profits of the grower at no cost whatever. Corn- ,re a cotton that yields 33* per cent of lint or a 500-pound bale he small g...

  20. Civic and Cosmopolitan Justice

    E-Print Network [OSTI]

    O'Neill, Onora

    2000-01-01

    ." The next lecture was given in 1959 by Professor Evereu C. H ughes, and has been published by the Cniversity of Kansas School of L'lw as part of his book Strulmt.s' Culture and PerspeclitJfS: l.Actures on Medical and CeneralEducatimr. The selection... premise asserting the claims of that authority; and dear enough what it costs-the relativisation of conclusions to that arbitrary premise. It is easy to agree that reasoning is limited as soon as it is be holden to any civilly constituted authority...

  1. Technical information report: Plasma melter operation, reliability, and maintenance analysis

    SciTech Connect (OSTI)

    Hendrickson, D.W. [ed.

    1995-03-14

    This document provides a technical report of operability, reliability, and maintenance of a plasma melter for low-level waste vitrification, in support of the Hanford Tank Waste Remediation System (TWRS) Low-Level Waste (LLW) Vitrification Program. A process description is provided that minimizes maintenance and downtime and includes material and energy balances, equipment sizes and arrangement, startup/operation/maintence/shutdown cycle descriptions, and basis for scale-up to a 200 metric ton/day production facility. Operational requirements are provided including utilities, feeds, labor, and maintenance. Equipment reliability estimates and maintenance requirements are provided which includes a list of failure modes, responses, and consequences.

  2. The effect of vitrification technology on waste loading

    SciTech Connect (OSTI)

    Hrma, P.R.; Smith, P.A.

    1994-08-01

    Radioactive wastes on the Hanford Site are going to be permanently disposed of by incorporation into a durable glass. These wastes will be separated into low and high-level portions, and then vitrified. The low-level waste (LLW) is water soluble. Its vitrifiable part (other than off-gas) contains approximately 80 wt% Na{sub 2}O, the rest being Al{sub 2}O{sub 3}, P{sub 2}O{sub 5}, K{sub 2}O, and minor components. The challenge is to formulate durable LLW glasses with as high Na{sub 2}O content as possible by optimizing the additions of SiO{sub 2}, Al{sub 2}O{sub 3}, B{sub 2}O{sub 3}, CaO, and ZrO{sub 2}. This task will not be simple, considering the non-linear and interactive nature of glass properties as a function of composition. Once developed, the LLW glass, being similar in composition to commercial glasses, is unlikely to cause major processing problems, such as crystallization or molten salt segregation. For example, inexpensive LLW glass can be produced in a high-capacity Joule-heated melter with a cold cap to minimize volatilization. The high-level waste (HLW) consists of water-insoluble sludge (Fe{sub 2}O{sub 3}, Al{sub 2}O{sub 3}, ZrO{sub 2}, Cr{sub 2}O{sub 3}, NiO, and others) and a substantial water-soluble residue (Na{sub 2}O). Most of the water-insoluble components are refractory; i.e., their melting points are above the glass melting temperature. With regard to product acceptability, the maximum loading of Hanford HLW in the glass is limited by product durability, not by radiolytic heat generation. However, this maximum may not be achievable because of technological constraints imposed by melter feed rheology, frit properties, and glass melter limits. These restrictions are discussed in this paper. 38 refs.

  3. Improvement to low-level radioactive-waste vitrification processes. Master's thesis

    SciTech Connect (OSTI)

    Horton, W.S.

    1986-05-01

    Low-level radioactive waste vitrification (LLWV) is a technically feasible and cost-competitive alternative to the traditional immobilization options, i.e., cementation or bituminization. This thesis analyzes cementation, bituminization and vitrification, reviews the impact of the low-level Waste-stream composition on the vitrification process, then proposes and discusses several techniques to control the volatile radionuclides in a Process Improved LLWV system (PILLWV). The techniques that control the volatile radionuclides include chemical precipitation, electrodialysis, and ion exchange. Ion exchange is preferred. A comparison of the technical specifications, of the regulatory compliance, and of the cost considerations shows the PILLWV to be the superior LLW immobilization option.

  4. Nevada Test Site Waste Acceptance Criteria

    SciTech Connect (OSTI)

    U. S. Department of Energy, National Nuclear Security Administration Nevada Site Office

    2005-10-01

    This document establishes the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (NNSA/NSO) waste acceptance criteria (WAC). The WAC provides the requirements, terms, and conditions under which the Nevada Test Site (NTS) will accept low-level radioactive (LLW) and mixed waste (MW) for disposal. It includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the NTS Area 3 and Area 5 Radioactive Waste Management Complex (RWMC) for storage or disposal.

  5. NEVADA TEST SITE WASTE ACCEPTANCE CRITERIA, JUNE 2006

    SciTech Connect (OSTI)

    U.S. DEPARTMENT OF ENERGY, NATIONAL NUCLEAR SECURITY ADMINISTRATION NEVADA SITE OFFICE

    2006-06-01

    This document establishes the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (NNSA/NSO) waste acceptance criteria (WAC). The WAC provides the requirements, terms, and conditions under which the Nevada Test Site (NTS) will accept low-level radioactive (LLW) and mixed waste (MW) for disposal. It includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the NTS Area 3 and Area 5 Radioactive Waste Management Complex (RWMC) for storage or disposal.

  6. Preliminary Project Execution Plan for the Remote-Handled Low-Level Waste Disposal Project

    SciTech Connect (OSTI)

    David Duncan

    2011-05-01

    This preliminary project execution plan (PEP) defines U.S. Department of Energy (DOE) project objectives, roles and responsibilities of project participants, project organization, and controls to effectively manage acquisition of capital funds for construction of a proposed remote-handled low-level waste (LLW) disposal facility at the Idaho National Laboratory (INL). The plan addresses the policies, requirements, and critical decision (CD) responsibilities identified in DOE Order 413.3B, 'Program and Project Management for the Acquisition of Capital Assets.' This plan is intended to be a 'living document' that will be periodically updated as the project progresses through the CD process to construction and turnover for operation.

  7. AM Canum Venaticorum Progenitors with Helium Star Donors and the resultant Explosions

    E-Print Network [OSTI]

    Brooks, Jared; Marchant, Pablo; Paxton, Bill

    2015-01-01

    We explore the outcome of mass transfer via Roche lobe overflow (RLOF) of $M_{\\rm He}\\lesssim0.51 M_\\odot$ pure helium burning stars in close binaries with white dwarfs (WDs). The evolution is driven by the loss of angular momentum through gravitational wave radiation (GWR), and both stars are modeled using Modules for Experiments in Stellar Astrophysics (MESA). The donors have masses of $M_{\\rm He}=0.35, 0.4, \\&\\ 0.51M_\\odot$ and accrete onto WDs of mass $M_{\\rm WD}$ from $0.6M_\\odot$ to $1.26M_\\odot$. The initial orbital periods ($P_{\\rm{orb}}$) span 20 to 80 minutes. For all cases, the accretion rate onto the WD is below the stable helium burning range, leading to accumulation of helium followed by unstable ignition. The mass of the convective core in the donors is small enough so that the WD accretes enough helium-rich matter to undergo a thermonuclear runaway in the helium shell before any carbon-oxygen enriched matter is transferred. The mass of the accumulated helium shell depends on $M_{\\rm WD}$ a...

  8. The cosmic rate of supernovae and the range of stars ending as Type Ia SNe

    E-Print Network [OSTI]

    P. Ruiz-Lapuente; R. Canal

    2000-09-20

    The present cosmic rate of Type Ia supernovae (SNeIa) suggests that about 6% of all stars in binary systems with primaries in the initial mass range $3-9\\ M\\sun$ end up as SNeIa. If that is confirmed, the unavoidable conclusion is that SNeIa can only be explained by the single degenerate scenario. At most 1% of stars in binary systems in the above range end up as CO + CO WD pairs, with total mass equal to or larger than the Chandrasekhar mass. Given that the number of mergers from pairs of CO + He WDs that reach the Chandrasekhar mass is even lower, the conclusion strongly favors binaries containing just one CO WD as the progenitors of SNeIa, since the SNeIa production efficiency (relative to the instantaneous star formation rate) predicted for double degenerate (DD) pairs lies more than $3\\sigma$ below the observational data, and the DD scenario can be rejected at more than 99% confidence level. Only if the SFR measurements from $z\\sim 0.1$ to $z\\sim 0.5$ are being underestimated by a factor of 6 while SNeIa rates are not, can we escape the above conclusion. We evaluate the numbers and characteristics of double WD systems with different chemical compositions (CO and He WDs) that should form and compare them with the observations, in order to check our predictions. Our conclusions appear robust after that test.

  9. Near-UV absorption in very cool DA white dwarfs

    SciTech Connect (OSTI)

    Saumon, D.; Holberg, J. B.; Kowalski, P. M. E-mail: holberg@argus.lpl.arizona.edu

    2014-07-20

    The atmospheres of very cool, hydrogen-rich white dwarfs (WDs) (T{sub eff} < 6000 K) are challenging to model because of the increased complexity of the equation of state, chemical equilibrium, and opacity sources in a low-temperature, weakly ionized dense gas. In particular, many models that assume relatively simple models for the broadening of atomic levels and mostly ideal gas physics overestimate the flux in the blue part of their spectra. A solution to this problem that has met with some success is that additional opacity at short wavelengths comes for the extreme broadening of the Lyman ? line of atomic H by collisions primarily with H{sub 2}. For the purpose of validating this model more rigorously, we acquired Hubble Space Telescope STIS spectra of eight very cool WDs (five DA and three DC stars). Combined with their known parallaxes, BVRIJHK, and Spitzer IRAC photometry, we analyze their entire spectral energy distribution (from 0.24 to 9.3 ?m) with a large grid of model atmospheres and synthetic spectra. We find that the red wing of the Lyman ? line reproduces the rapidly decreasing near-UV flux of these very cool stars very well. We determine better constrained values of T{sub eff} and gravity as well as upper limits to the helium abundance in their atmospheres.

  10. The Evolution of Compact Binary Star Systems

    E-Print Network [OSTI]

    Konstantin Postnov; Lev Yungelson

    2014-03-21

    We review the formation and evolution of compact binary stars consisting of white dwarfs (WDs), neutron stars (NSs), and black holes (BHs). Mergings of compact binary stars are expected to be the most important sources for the forthcoming gravitational-wave (GW) astronomy. In the first part of the review, we discuss observational manifestations of close binary stars with NS and/or black components and their merger rate, crucial points in the formation and evolution of compact stars in binary systems, including the treatment of the natal kicks which NSs and BHs acquire during the core collapse of massive stars and the common envelope phase of binary evolution, which are most relevant to the merging rates of NS-NS, NS-BH and BH-BH binaries. The second part of the review is devoted mainly to formation and evolution of binary WDs and their observational manifestations, including their role as progenitors of cosmologically important thermonuclear SN Ia. We also consider AM CVn-stars which are thought to be the best verification binary GW sources for future low-frequency GW space interferometers.

  11. THE HYBRID CONe WD + He STAR SCENARIO FOR THE PROGENITORS OF TYPE Ia SUPERNOVAE

    SciTech Connect (OSTI)

    Wang, B.; Meng, X.; Liu, D.-D.; Han, Z. [Yunnan Observatories, Chinese Academy of Sciences, Kunming 650011 (China); Liu, Z.-W., E-mail: wangbo@ynao.ac.cn [Argelander-Institut fr Astronomie, Auf dem Hgel 71, D-53121, Bonn (Germany)

    2014-10-20

    Hybrid CONe white dwarfs (WDs) have been suggested to be possible progenitors of type Ia supernovae (SNe Ia). In this Letter, we systematically studied the hybrid CONe WD + He star scenario for the progenitors of SNe Ia, in which a hybrid CONe WD increases its mass to the Chandrasekhar mass limit by accreting He-rich material from a non-degenerate He star. We obtained the SN Ia birthrates and delay times for this scenario using to a series of detailed binary population synthesis simulations. The SN Ia birthrates for this scenario are ?0.033-0.539 10{sup 3} yr{sup 1}, which roughly accounts for 1%-18% of all SNe Ia. The estimated delay times are ?28Myr-178Myr, which makes these the youngest SNe Ia predicted by any progenitor model so far. We suggest that SNe Ia from this scenario may provide an alternative explanation for type Iax SNe. We also presented some properties of the donors at the point when the WDs reach the Chandrasekhar mass. These properties may be a good starting point for investigating the surviving companions of SNe Ia and for constraining the progenitor scenario studied in this work.

  12. Analysis of HST/COS spectra of the bare C-O stellar core H1504+65 and a high-velocity twin in the Galactic halo

    E-Print Network [OSTI]

    Werner, K

    2015-01-01

    H1504+65 is an extremely hot white dwarf (effective temperature Teff = 200,000 K) with a carbon-oxygen dominated atmosphere devoid of hydrogen and helium. This atmospheric composition was hitherto unique among hot white dwarfs (WDs), and it could be related to recently detected cooler WDs with C or O dominated spectra. The origin of the H and He deficiency in H1504+65 is unclear. To further assess this problem, we performed ultraviolet spectroscopy with the Cosmic Origins Spectrograph (COS) aboard the Hubble Space Telescope (HST). In accordance with previous far-ultraviolet spectroscopy performed with the Far Ultraviolet Spectroscopic Explorer, the most prominent lines stem from C IV, O V-VI, and Ne VI-VIII. Archival HST/COS spectra are utilized to prove that the supersoft X-ray source RX J0439.8-6809 is, considering the exotic composition, a twin of H1504+65 that is even hotter (Teff = 250,000 K). In contrast to earlier claims, we find that the star is not located in the Large Magellanic Cloud but a foregrou...

  13. Solid-polymer-electrolyte tritiated water electrolyzer for Water Detritiation System

    SciTech Connect (OSTI)

    Iwai, Y.; Yamanishi, T. [Tritium Engineering Group, JAEA, Tokai, Naka, Ibaraki, 319-1195 (Japan); Hiroki, A.; Yagi, T.; Tamada, M. [Environmental Polymer Group, JAEA, Watanuki, Takasaki, Gunma, 370-1292 (Japan)

    2008-07-15

    A solid-polymer-electrolyte (SPE) water electrolyzer for high-level tritiated water was designed for the Water Detritiation System (WDS). Polymeric materials were selected from a main viewpoint of radiation durability to keep their functions beyond ITER-WDS requirement (530 kGy). Our selection was Pt + Ir applied Nafion{sup R} N117 ion exchange membrane, VITON{sup R} O-ring seal and polyimide insulator. A {gamma}-ray irradiation test of the SPE cell demonstrated the durability of the cell against 530 kGy. The electrolyzer is designed to handle around 9 TBq/kg of high-level tritiated water. The detritiation of the polymeric materials is thus a critical problem for the maintenance or for the disposal of the electrolyzer. As for the Nafion membrane, most of tritiated water in the membrane was rapidly removed by such as vacuum dehydration. It was difficult, by contrast, to remove bound tritiated water in the membrane. An effective method to remove tritiated water in the bound water is to promote an isotope exchange. (authors)

  14. Final Evolution and Delayed Explosions of Spinning White Dwarfs in Single Degenerate Models for Type Ia Supernovae

    E-Print Network [OSTI]

    Benvenuto, Omar G; Kitamura, Hikaru; Hachisu, Izumi

    2015-01-01

    We study the occurrence of delayed SNe~Ia in the single degenerate (SD) scenario. We assume that a massive carbon-oxygen (CO) white dwarf (WD) accretes matter coming from a companion star, making it to spin at the critical rate. We assume uniform rotation due to magnetic field coupling. The carbon ignition mass for non-rotating WDs is M_{ig}^{NR} \\approx 1.38 M_{\\odot}; while for the case of uniformly rotating WDs it is a few percent larger (M_{ig}^{R} \\approx 1.43 M_{\\odot}). When accretion rate decreases, the WD begins to lose angular momentum, shrinks, and spins up; however, it does not overflow its critical rotation rate, avoiding mass shedding. Thus, angular momentum losses can lead the CO WD interior to compression and carbon ignition, which would induce an SN~Ia. The delay, largely due to the angular momentum losses timescale, may be large enough to allow the companion star to evolve to a He WD, becoming undetectable at the moment of explosion. This scenario supports the occurrence of delayed SNe~Ia if...

  15. Nevada National Security Site Waste Acceptance Criteria

    SciTech Connect (OSTI)

    none,

    2013-06-01

    This document establishes the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Field Office (NNSA/NFO), Nevada National Security Site Waste Acceptance Criteria (NNSSWAC). The NNSSWAC provides the requirements, terms, and conditions under which the Nevada National Security Site (NNSS) will accept the following: DOE hazardous and non-hazardous non-radioactive classified waste DOE low-level radioactive waste (LLW) DOE mixed low-level waste (MLLW) U.S. Department of Defense (DOD) classified waste The LLW and MLLW listed above may also be classified waste. Classified waste is the only waste accepted for disposal that may be non-radioactive and shall be required to meet the waste acceptance criteria for radioactive waste as specified in this document. Classified waste may be sent to the NNSS as classified matter. Section 3.1.18 provides the requirements that must be met for permanent burial of classified matter. The NNSA/NFO and support contractors are available to assist the generator in understanding or interpreting this document. For assistance, please call the NNSA/NFO Environmental Management Operations (EMO) at (702) 295-7063, and the call will be directed to the appropriate contact.

  16. Closure Strategy for a Waste Disposal Facility with Multiple Waste Types and Regulatory Drivers at the Nevada Test Site

    SciTech Connect (OSTI)

    D. Wieland, V. Yucel, L. Desotell, G. Shott, J. Wrapp

    2008-04-01

    The U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) plans to close the waste and classified material storage cells in the southeast quadrant of the Area 5 Radioactive Waste Management Site (RWMS), informally known as the '92-Acre Area', by 2011. The 25 shallow trenches and pits and the 13 Greater Confinement Disposal (GCD) borings contain various waste streams including low-level waste (LLW), low-level mixed waste (LLMW), transuranic (TRU), mixed transuranic (MTRU), and high specific activity LLW. The cells are managed under several regulatory and permit programs by the U.S. Department of Energy (DOE) and the Nevada Division of Environmental Protection (NDEP). Although the specific closure requirements for each cell vary, 37 closely spaced cells will be closed under a single integrated monolayer evapotranspirative (ET) final cover. One cell will be closed under a separate cover concurrently. The site setting and climate constrain transport pathways and are factors in the technical approach to closure and performance assessment. Successful implementation of the integrated closure plan requires excellent communication and coordination between NNSA/NSO and the regulators.

  17. Environmental assessment for the Radioactive and Mixed Waste Management Facility: Sandia National Laboratories/New Mexico

    SciTech Connect (OSTI)

    Not Available

    1993-06-01

    The Department of Energy (DOE) has prepared an environmental assessment (EA) (DOE/EA-0466) under the National Environmental Policy Act (NEPA) of 1969 for the proposed completion of construction and subsequent operation of a central Radioactive and Mixed Waste Management Facility (RMWMF), in the southeastern portion of Technical Area III at Sandia National Laboratory, Albuquerque (SNLA). The RMWMF is designed to receive, store, characterize, conduct limited bench-scale treatment of, repackage, and certify low-level waste (LLW) and mixed waste (MW) (as necessary) for shipment to an offsite disposal or treatment facility. The RMWMF was partially constructed in 1989. Due to changing regulatory requirements, planned facility upgrades would be undertaken as part of the proposed action. These upgrades would include paving of road surfaces and work areas, installation of pumping equipment and lines for surface impoundment, and design and construction of air locks and truck decontamination and water treatment systems. The proposed action also includes an adjacent corrosive and reactive metals storage area, and associated roads and paving. LLW and MW generated at SNLA would be transported from the technical areas to the RMWMF in containers approved by the Department of Transportation. The RMWMF would not handle nonradioactive hazardous waste. Based on the analysis in the EA, the proposed completion of construction and operation of the RMWMF does not constitute a major Federal action significantly affecting the quality of the human environment within the meaning of NEPA. Therefore, preparation of an environmental impact statement for the proposed action is not required.

  18. Volatility and entrainment of feed components and product glass characteristics during pilot-scale vitrification of simulated Hanford site low-level waste

    SciTech Connect (OSTI)

    Shade, J.W.

    1996-05-03

    Commercially available melter technologies were tested for application to vitrification of Hanford site low-level waste (LLW). Testing was conducted at vendor facilities using a non-radioactive LLW simulant. Technologies tested included four Joule-heated melter types, a carbon electrode melter, a cyclone combustion melter, and a plasma torch-fired melter. A variety of samples were collected during the vendor tests and analyzed to provide data to support evaluation of the technologies. This paper describes the evaluation of melter feed component volatility and entrainment losses and product glass samples produced during the vendor tests. All vendors produced glasses that met minimum leach criteria established for the test glass formulations, although in many cases the waste oxide loading was less than intended. Entrainment was much lower in Joule-heated systems than in the combustion or plasma torch-fired systems. Volatility of alkali metals, halogens, B, Mo, and P were severe for non-Joule-heated systems. While losses of sulfur were significant for all systems, the volatility of other components was greatly reduced for some configurations of Joule-heated melters. Data on approaches to reduce NO{sub x} generation, resulting from high nitrate and nitrite content in the double-shell slurry feed, are also presented.

  19. Integration of health physics, safety and operational processes for management and disposition of recycled uranium wastes at the Fernald Environmental Management Project (FEMP)

    SciTech Connect (OSTI)

    Barber, James; Buckley, James

    2003-02-23

    Fluor Fernald, Inc. (Fluor Fernald), the contractor for the U. S. Department of Energy (DOE) Fernald Environmental Management Project (FEMP), recently submitted a new baseline plan for achieving site closure by the end of calendar year 2006. This plan was submitted at DOE's request, as the FEMP was selected as one of the sites for their accelerated closure initiative. In accordance with the accelerated baseline, the FEMP Waste Management Project (WMP) is actively evaluating innovative processes for the management and disposition of low-level uranium, fissile material, and thorium, all of which have been classified as waste. These activities are being conducted by the Low Level Waste (LLW) and Uranium Waste Disposition (UWD) projects. Alternatives associated with operational processing of individual waste streams, each of which poses potentially unique health physics, industrial hygiene and industrial hazards, are being evaluated for determination of the most cost effective and safe met hod for handling and disposition. Low-level Mixed Waste (LLMW) projects are not addressed in this paper. This paper summarizes historical uranium recycling programs and resultant trace quantity contamination of uranium waste streams with radionuclides, other than uranium. The presentation then describes how waste characterization data is reviewed for radiological and/or chemical hazards and exposure mitigation techniques, in conjunction with proposed operations for handling and disposition. The final part of the presentation consists of an overview of recent operations within LLW and UWD project dispositions, which have been safely completed, and a description of several current operations.

  20. Dancing with the regulations - Part Deux

    SciTech Connect (OSTI)

    Nitschke, R.L.

    1995-12-31

    The disposal of low-level radioactive waste (LLW) in the United States has long been subjected to two very similar regulations depending upon the location. Disposal sites located on Department of Energy (DOE) Reservations are subject to DOE Order 5820.2A {open_quotes}Radioactive Waste Management,{close_quotes} while disposal sites located elsewhere are subject to the Nuclear Regulatory Commission regulation 10 CFR 61 {open_quotes}Licensing Requirements for Land Disposal of Radioactive Waste.{close_quotes} While life was not necessarily good, there was only one sheet of music to dance to. Recently a new player, named CERCLA (Comprehensive Environmental Response, Compensation, and Liability Act), has ridden into those DOE towns, and for those whose disposal facilities lie within or adjacent to Superfund sites, she has brought along a different drum to dance to. This paper discusses the differences and similarities between the different dance partners and their associated musical scores (i.e., the performance assessment (PA) required by the DOE order and the baseline risk assessment (BRA) required by CERCLA). The paper then provides a brief discussion on the latest dancer to cut in: the Defense Nuclear Facilities Safety Board (DNFSB). This discussion should help to alleviate the confusion while dancing on the LLW disposal regulatory ballroom floor.

  1. Evaluation of Groundwater Impacts to Support the National Environmental Policy Act Environmental Assessment for the INL Remote-Handled Low-Level Waste Disposal Project

    SciTech Connect (OSTI)

    Annette Schafer; Arthur S. Rood; A. Jeffrey Sondrup

    2010-08-01

    The groundwater impacts have been analyzed for the proposed RH-LLW disposal facility. A four-step analysis approach was documented and applied. This assessment compared the predicted groundwater ingestion dose to the more restrictive of either the 25 mrem/yr all pathway dose performance objective, or the maximum contaminant limit performance objective. The results of this analysis indicate that the groundwater impacts for either proposed facility location are expected to be less than the performance objectives. The analysis was prepared to support the NEPA-EA for the top two ranking of the proposed RH-LLW sites. As such, site-specific conditions were incorporated for each set of results generated. These site-specific conditions were included to account for the transport of radionuclides through the vadose zone and through the aquifer at each site. Site-specific parameters included the thickness of vadose zone sediments and basalts, moisture characteristics of the sediments, and aquifer velocity. Sorption parameters (Kd) were assumed to be very conservative values used in Track II analysis of CERCLA sites at INL. Infiltration was also conservatively assumed to represent higher rates corresponding to disturbed soil conditions. The results of this analysis indicate that the groundwater impacts for either proposed facility location are expected to be less than the performance objectives.

  2. Colloid suspension stability and transport through unsaturated porous media

    SciTech Connect (OSTI)

    McGraw, M.A.; Kaplan, D.I.

    1997-04-01

    Contaminant transport is traditionally modeled in a two-phase system: a mobile aqueous phase and an immobile solid phase. Over the last 15 years, there has been an increasing awareness of a third, mobile solid phase. This mobile solid phase, or mobile colloids, are organic or inorganic submicron-sized particles that move with groundwater flow. When colloids are present, the net effect on radionuclide transport is that radionuclides can move faster through the system. It is not known whether mobile colloids exist in the subsurface environment of the Hanford Site. Furthermore, it is not known if mobile colloids would likely exist in a plume emanating from a Low Level Waste (LLW) disposal site. No attempt was made in this study to ascertain whether colloids would form. Instead, experiments and calculations were conducted to evaluate the likelihood that colloids, if formed, would remain in suspension and move through saturated and unsaturated sediments. The objectives of this study were to evaluate three aspects of colloid-facilitated transport of radionuclides as they specifically relate to the LLW Performance Assessment. These objectives were: (1) determine if the chemical conditions likely to exist in the near and far field of the proposed disposal site are prone to induce flocculation (settling of colloids from suspension) or dispersion of naturally occurring Hanford colloids, (2) identify the important mechanisms likely involved in the removal of colloids from a Hanford sediment, and (3) determine if colloids can move through unsaturated porous media.

  3. Greater-than-Class C low-level radioactive waste shipping package/container identification and requirements study. National Low-Level Waste Management Program

    SciTech Connect (OSTI)

    Tyacke, M.

    1993-08-01

    This report identifies a variety of shipping packages (also referred to as casks) and waste containers currently available or being developed that could be used for greater-than-Class C (GTCC) low-level waste (LLW). Since GTCC LLW varies greatly in size, shape, and activity levels, the casks and waste containers that could be used range in size from small, to accommodate a single sealed radiation source, to very large-capacity casks/canisters used to transport or dry-store highly radioactive spent fuel. In some cases, the waste containers may serve directly as shipping packages, while in other cases, the containers would need to be placed in a transport cask. For the purpose of this report, it is assumed that the generator is responsible for transporting the waste to a Department of Energy (DOE) storage, treatment, or disposal facility. Unless DOE establishes specific acceptance criteria, the receiving facility would need the capability to accept any of the casks and waste containers identified in this report. In identifying potential casks and waste containers, no consideration was given to their adequacy relative to handling, storage, treatment, and disposal. Those considerations must be addressed separately as the capabilities of the receiving facility and the handling requirements and operations are better understood.

  4. Vitrification treatment options for disposal of greater-than-Class-C low-level waste in a deep geologic repository

    SciTech Connect (OSTI)

    Fullmer, K.S.; Fish, L.W.; Fischer, D.K.

    1994-11-01

    The Department of Energy (DOE), in keeping with their responsibility under Public Law 99-240, the Low-Level Radioactive Waste Policy Amendments Act of 1985, is investigating several disposal options for greater-than-Class C low-level waste (GTCC LLW), including emplacement in a deep geologic repository. At the present time vitrification, namely borosilicate glass, is the standard waste form assumed for high-level waste accepted into the Civilian Radioactive Waste Management System. This report supports DOE`s investigation of the deep geologic disposal option by comparing the vitrification treatments that are able to convert those GTCC LLWs that are inherently migratory into stable waste forms acceptable for disposal in a deep geologic repository. Eight vitrification treatments that utilize glass, glass ceramic, or basalt waste form matrices are identified. Six of these are discussed in detail, stating the advantages and limitations of each relative to their ability to immobilize GTCC LLW. The report concludes that the waste form most likely to provide the best composite of performance characteristics for GTCC process waste is Iron Enriched Basalt 4 (IEB4).

  5. Volatility and entrainment of feed components and product glass characteristics during pilot-scale vitrification of simulated Hanford Site low-level waste

    SciTech Connect (OSTI)

    Whyatt, G.A. [Pacific Northwest National Lab., Richland, WA (United States); Shade, J.W.; Stegen, G.E. [Westinghouse Hanford Co., Richland, WA (United States)

    1996-12-31

    Commercially available melter technologies were tested for application to vitrification of Hanford Site low-level waste (LLW). Testing was conducted at vendor facilities using a non-radioactive LLW stimulant. Technologies tested included four Joule-heated melter types, a carbon electrode melter, a cyclone combustion melter, and a plasma torch-fired melter. A variety of samples were collected during the vendor tests and analyzed to provide data to support evaluation of the technologies. This paper describes the evaluation of melter feed component volatility and entrainment losses and product glass samples produced during the vendor tests. All vendors produced glasses that met minimum leach criteria established for the test glass formulations, although in many cases the waste oxide loading was less than intended. Entrainment was much lower in Joule-heated systems than in the combustion or plasma torch-fired systems. Volatility of alkali metals, halogens, B, Mo, and P were severe for non-Joule-heated systems. While losses of sulfur were significant for all systems, the volatility of other components was greatly reduced for some configurations of Joule-heated melters. Data on approaches to reduce NO{sub x} generation, resulting from high nitrate and nitrite content in the double-shell slurry feed, are also presented.

  6. A West Valley Demonstration Project Milestone - Achieving Certification to Ship Waste to the Nevada Test Site

    SciTech Connect (OSTI)

    Jackson, J. P.; Pastor, R. S.

    2002-02-28

    The West Valley Demonstration Project (WVDP) has successfully pretreated and vitrified nearly all of the 600,000 gallons of liquid high-level radioactive waste that was generated at the site of the only commercial nuclear fuel reprocessing plant to have operated in the United States. Low-level waste (LLW) generated during the course of the cleanup effort now requires disposal. Currently the WVDP only ships Class A LLW for off-site disposal. It has been shipping Class A wastes to Envirocare of Utah, Inc. since 1997. However, the WVDP may also have a future need to ship Class B and Class C waste, which Envirocare is not currently authorized to accept. The Nevada Test Site (NTS), a U.S. Department of Energy (DOE) facility, can accept all three waste classifications. The WVDP set a goal to receive certification to begin shipping Class A wastes to NTS by 2001. Formal certification/approval was granted by the DOE Nevada Operations Office on July 12, 2001. This paper discusses how the WVDP contractor, West Valley Nuclear Services Company (WVNSCO), completed the activities required to achieve NTS certification in 2001 to ship waste to its facility. The information and lessons learned provided are significant because the WVDP is the only new generator receiving certification based on an NTS audit in January 2001 that resulted in no findings and only two observations--a rating that is unparalleled in the DOE Complex.

  7. Innovative vitrification for soil remediation

    SciTech Connect (OSTI)

    Jetta, N.W.; Patten, J.S.; Hnat, J.G. [Vortec Corp., Collegeville, PA (United States)] [and others

    1996-03-01

    The objective of this DOE demonstration program is to validate the performance and operation of the Vortec Cyclone Melting System (CMS{trademark}) for the processing of LLW contaminated soils found at DOE sites. This DOE vitrification demonstration project has successfully progressed through the first two phases. Phase 1 consisted of pilot scale testing with surrogate wastes and the conceptual design of a process plant operating at a generic DOE site. The objective of Phase 2, which is scheduled to be completed the end of FY 95, is to develop a definitive process plant design for the treatment of wastes at a specific DOE facility. During Phase 2, a site specific design was developed for the processing of LLW soils and muds containing TSCA organics and RCRA metal contaminants. Phase 3 will consist of a full scale demonstration at the DOE gaseous diffusion plant located in Paducah, KY. Several DOE sites were evaluated for potential application of the technology. Paducah was selected for the demonstration program because of their urgent waste remediation needs as well as their strong management and cost sharing financial support for the project.

  8. Evaluation of Phase II glass formulations for vitrification of Hanford Site low-level waste

    SciTech Connect (OSTI)

    Feng, X.; Hrma, P.R.; Schweiger, M.J. [and others

    1996-03-01

    A vendor glass formulation study was carried out at Pacific Northwest Laboratory (PNL), supporting the Phase I and Phase II melter vendor testing activities for Westinghouse Hanford Company. This study is built upon the LLW glass optimization effort that will be described in a separate report. For Phase I vendor melter testing, six glass formulations were developed at PNL and additional were developed by Phase I vendors. All the doses were characterized in terms of viscosity and chemical durability by the 7-day Product Consistency Test. Twelve Phase II glass formulations (see Tables 3.5 and 3.6) were developed to accommodate 2.5 wt% P{sub 2}O{sub 5} and 1.0 wt% S0{sub 3} without significant processing problems. These levels of P{sub 2}O{sub 5} and SO{sub 3} are expected to be the highest possible concentrations from Hanford Site LLW streams at 25 wt% waste loading in glass. The Phase H compositions formulated were 6 to 23 times more durable than the environmental assessment (EA) glass. They melt within the temperature range of 1160{degrees} to 1410{degrees}C to suit different melting technologies. The composition types include boron-free for volatilization sensitive melters; boron-containing glasses for coId-cap melters; Zr-containing, glasses for enhanced Iong-term durability; and Fe-containing glasses for reducing melting temperature and melt volatility while maintaining chemical durability.

  9. Interim data quality objectives for waste pretreatment and vitrification. Revision 1

    SciTech Connect (OSTI)

    Kupfer, M.J.; Conner, J.M.; Kirkbride, R.A.; Mobley, J.R.

    1994-09-15

    The Tank Waste Remediation System (TWRS) is responsible for storing, processing, and immobilizing the Hanford Site tank wastes. Characterization information on the tank wastes is needed so that safety concerns can be addressed, and retrieval, pretreatment, and immobilization processes can be designed, permitted, and implemented. This document describes the near-term tank waste sampling and characterization needs of the Pretreatment, High-Level Waste (HLW) Disposal, and Low-Level Waste (LLW) Disposal Programs to support the TWRS disposal mission. The final DQO (Data Quality Objective) will define specific waste tanks to be sampled, sample timing requirements, an appropriate analytical scheme, and a list of required analytes. This interim DQO, however, focuses primarily on the required analytes since the tanks to be sampled in FY 1994 and early FY 1995 are being driven most heavily by other considerations, particularly safety. The major objective of this Interim DQO is to provide guidance for tank waste characterization requirements for samples taken before completion of the final DQO. The characterization data needs defined herein will support the final DQO to help perform the following: Support the TWRS technical strategy by identification of the chemical and physical composition of the waste in the tanks and Guide development efforts to define waste pretreatment processes, which will in turn define HLW and LLW feed to vitrification processes.

  10. Tank Waste Remediation System tank waste pretreatment and vitrification process development testing requirements assessment

    SciTech Connect (OSTI)

    Howden, G.F.

    1994-10-24

    A multi-faceted study was initiated in November 1993 to provide assurance that needed testing capabilities, facilities, and support infrastructure (sampling systems, casks, transportation systems, permits, etc.) would be available when needed for process and equipment development to support pretreatment and vitrification facility design and construction schedules. This first major report provides a snapshot of the known testing needs for pretreatment, low-level waste (LLW) and high-level waste (HLW) vitrification, and documents the results of a series of preliminary studies and workshops to define the issues needing resolution by cold or hot testing. Identified in this report are more than 140 Hanford Site tank waste pretreatment and LLW/HLW vitrification technology issues that can only be resolved by testing. The report also broadly characterizes the level of testing needed to resolve each issue. A second report will provide a strategy(ies) for ensuring timely test capability. Later reports will assess the capabilities of existing facilities to support needed testing and will recommend siting of the tests together with needed facility and infrastructure upgrades or additions.

  11. Innovative vitrification for soil remediation

    SciTech Connect (OSTI)

    Jetta, N.W.; Patten, J.S.; Hnat, J.G. [Vortec Corp., Collegeville, PA (United States)

    1995-10-01

    The objective of this DOE demonstration program is to validate the performance and operation of the Vortec Cyclone Melting System (CMS{trademark}) for the processing of LLW contaminated soils found at DOE sites. This DOE vitrification demonstration project has successfully progressed through the first two phases. Phase I consisted of pilot scale testing with surrogate wastes and the conceptual design of a process plant operating at a generic DOE site. The objective of Phase 2, which is scheduled to be completed the end of FY 95, is to develop a definitive process plant design for the treatment of wastes at a specific DOE facility. During Phase 2, a site specific design was developed for the processing of LLW soils and muds containing TSCA organics and RCRA metal contaminants. Phase 3 will consist of a full scale demonstration at the DOE gaseous diffusion plant located in Paducah, KY. Several DOE sites were evaluated for potential application of the technology. Paducah was selected for the demonstration program because of their urgent waste remediation needs as well as their strong management and cost sharing financial support for the project.

  12. U.S. Bureau of Mines, phase I Hanford low-level waste melter tests: Melter offgas report

    SciTech Connect (OSTI)

    Eaton, W.C.

    1995-10-27

    A multiphase program was initiated in 1994 to test commercially available melter technologies for the vitrification of the low-level waste (LLW) stream from defense wastes stored in underground tanks at the Hanford Site in southeastern Washington State. Phase 1 of the melter demonstration tests using simulated LLW was completed during fiscal year 1995. This document is the melter offgas report on testing performed by the U.S. Department of the Interior, Bureau of Mines, Albany Research Center in Albany, Oregon. The Bureau of Mines (one of the seven vendors selected) was chosen to demonstrate carbon electrode melter technology (also called carbon arc or electric arc) under WHC subcontract number MMI-SVV-384216. The document contains the complete offgas report for the first 24-hour melter test (WHC-1) as prepared by Entropy Inc. A summary of this report is also contained in the``U.S. Bureau of Mines, Phase 1 Hanford Low-Level Waste Melter Tests: Final Report`` (WHC-SD-WM-VI-030).

  13. Low-Level Waste Regulation: Putting Principles Into Practice - 13297 - The Richard S. Hodes, M.D., Honor Lecture Award

    SciTech Connect (OSTI)

    Kennedy, James E.

    2013-07-01

    In carrying out its mission to ensure the safe use of radioactive materials for beneficial civilian purposes while protecting people and the environment, the U.S. Nuclear Regulatory Commission (NRC) adheres to its Principles of Good Regulation. The Principles-Independence, Openness, Efficiency, Clarity, and Reliability-apply to the agency as a whole in its decision-making and to the individual conduct of NRC employees. This paper describes the application of the Principles in a real-life staff activity, a guidance document used in the NRC's low-level radioactive waste (LLW) program, the Concentration Averaging and Encapsulation Branch Technical Position (CA BTP). The staff's process to revise the document, as well as the final content of the document, were influenced by following the Principles. For example, consistent with the Openness Principle, the staff conducted a number of outreach activities and received many comments on three drafts of the document. Stakeholder comments affected the final staff positions in some cases. The revised CA BTP, once implemented, is expected to improve management and disposal of LLW in the United States. Its positions have an improved nexus to health and safety; are more performance-based than previously, thus providing licensees with options for how they achieve the required outcome of protecting an inadvertent human intruder into a disposal facility; and provide for disposal of more sealed radioactive sources, which are a potential threat to national security. (author)

  14. INNOVATIVE FOSSIL FUEL FIRED VITRIFICATION TECHNOLOGY FOR SOIL REMEDIATION

    SciTech Connect (OSTI)

    J. Hnat; L.M. Bartone; M. Pineda

    2001-07-13

    This Summary Report summarizes the progress of Phases 3, 3A and 4 of a waste technology Demonstration Project sponsored under a DOE Environmental Management Research and Development Program and administered by the U.S. Department of Energy National Energy Technology Laboratory-Morgantown (DOE-NETL) for an ''Innovative Fossil Fuel Fired Vitrification Technology for Soil Remediation''. The Summary Reports for Phases 1 and 2 of the Program were previously submitted to DOE. The total scope of Phase 3 was to have included the design, construction and demonstration of Vortec's integrated waste pretreatment and vitrification process for the treatment of low level waste (LLW), TSCA/LLW and mixed low-level waste (MLLW). Due to funding limitations and delays in the project resulting from a law suit filed by an environmental activist and the extended time for DOE to complete an Environmental Assessment for the project, the scope of the project was reduced to completing the design, construction and testing of the front end of the process which consists of the Material Handling and Waste Conditioning (MH/C) Subsystem of the vitrification plant. Activities completed under Phases 3A and 4 addressed completion of the engineering, design and documentation of the Material Handling and Conditioning System such that final procurement of the remaining process assemblies can be completed and construction of a Limited Demonstration Project be initiated in the event DOE elects to proceed with the construction and demonstration testing of the MH/C Subsystem.

  15. Extensive separations (CLEAN) processing strategy compared to TRUEX strategy and sludge wash ion exchange

    SciTech Connect (OSTI)

    Knutson, B.J.; Jansen, G.; Zimmerman, B.D.; Seeman, S.E. [Westinghouse Hanford Co., Richland, WA (United States); Lauerhass, L.; Hoza, M. [Pacific Northwest Lab., Richland, WA (United States)

    1994-08-01

    Numerous pretreatment flowsheets have been proposed for processing the radioactive wastes in Hanford`s 177 underground storage tanks. The CLEAN Option is examined along with two other flowsheet alternatives to quantify the trade-off of greater capital equipment and operating costs for aggressive separations with the reduced waste disposal costs and decreased environmental/health risks. The effect on the volume of HLW glass product and radiotoxicity of the LLW glass or grout product is predicted with current assumptions about waste characteristics and separations processes using a mass balance model. The prediction is made on three principal processing options: washing of tank wastes with removal of cesium and technetium from the supernatant, with washed solids routed directly to the glass (referred to as the Sludge Wash C processing strategy); the previous steps plus dissolution of the solids and removal of transuranic (TRU) elements, uranium, and strontium using solvent extraction processes (referred to as the Transuranic Extraction Option C (TRUEX-C) processing strategy); and an aggressive yet feasible processing strategy for separating the waste components to meet several main goals or objectives (referred to as the CLEAN Option processing strategy), such as the LLW is required to meet the US Nuclear Regulatory Commission Class A limits; concentrations of technetium, iodine, and uranium are reduced as low as reasonably achievable; and HLW will be contained within 1,000 borosilicate glass canisters that meet current Hanford Waste Vitrification Plant glass specifications.

  16. Hazard Classification of the Remote Handled Low-Level Waste Disposal Facility

    SciTech Connect (OSTI)

    Boyd D. Christensen

    2012-05-01

    The Battelle Energy Alliance (BEA) at the Idaho National Laboratory (INL) is constructing a new facility to replace remote-handled low-level radioactive waste disposal capability for INL and Naval Reactors Facility operations. Current disposal capability at the Radioactive Waste Management Complex (RWMC) will continue until the facility is full or closed for remediation (estimated at approximately fiscal year 2015). Development of a new onsite disposal facility is the highest ranked alternative and will provide RH-LLW disposal capability and will ensure continuity of operations that generate RH-LLW for the foreseeable future. As a part of establishing a safety basis for facility operations, the facility will be categorized according to DOE-STD-1027-92. This classification is important in determining the scope of analyses performed in the safety basis and will also dictate operational requirements of the completed facility. This paper discusses the issues affecting hazard classification in this nuclear facility and impacts of the final hazard categorization.

  17. Integrated data base for 1993: US spent fuel and radioactive waste inventories, projections, and characteristics. Revision 9

    SciTech Connect (OSTI)

    Klein, J.A.; Storch, S.N.; Ashline, R.C.

    1994-03-01

    The Integrated Data Base (IDB) Program has compiled historic data on inventories and characteristics of both commercial and DOE spent fuel; also, commercial and U.S. government-owned radioactive wastes through December 31, 1992. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest U.S. Department of Energy/Energy Information Administration (DOE/EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste (HLW), transuranic (TRU), waste, low-level waste (LLW), commercial uranium mill tailings, environmental restoration wastes, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) LLW. For most of these categories, current and projected inventories are given through the calendar-year (CY) 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal.

  18. Advanced Fuel Cycle Cost Basis

    SciTech Connect (OSTI)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2008-03-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  19. Advanced Fuel Cycle Cost Basis

    SciTech Connect (OSTI)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert

    2007-04-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 26 cost modules24 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, and high-level waste.

  20. Advanced Fuel Cycle Cost Basis

    SciTech Connect (OSTI)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2009-12-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  1. 55-Gallon Drum Attenuation Corrections for Waste Assay Measurements

    SciTech Connect (OSTI)

    Casella, V.R.

    2002-04-03

    The present study shows how the percent attenuation for low-level waste (LLW), carbon-steel 55-gallon drums (44 and 46 mil) and for transuranic (TRU) DOT Type 7A 55-gallon drums (approximately 61 mil) changes with gamma energy from 60 keV to 1400 keV. Attenuation for these drums is in the range of 5 to 15 percent at energies from 400 to 1400 keV and from 15 to 35 percent at energies from 120 to 400 keV. At 60 keV, these drums attenuate 70-80 percent of the gamma rays. Correction factors were determined in order to correct for gamma attenuation of a TRU drum if a calibration is performed with a LLW drum. These correction factors increase the activities of the TRU drum by from 10 percent to 2 percent in the energy range of 165 to 1400 keV, with an increase of about 50 percent at 60 keV. Correction factors for TRU drums and for analyses without a drum were used to adjust the percent yield for frequently measured gamma rays, so that the assay libraries could be modified to provide the drum attenuation corrections.

  2. Conversion of transuranic waste to low level waste by decontamination: a site specific update

    SciTech Connect (OSTI)

    Allen, R.P.; Hazelton, R.F.

    1985-09-01

    As a followup to an FY-1984 cost/benefit study, a program was conducted in FY-1985 to transfer to the relevant DOE sites the information and technology for the direct conversion of transuranic (TRU) waste to low-level waste (LLW) by decontamination. As part of this work, the economic evaluation of the various TRUW volume reduction and conversion options was updated and expanded to include site-specific factors. The results show, for the assumptions used, that size reduction, size reduction followed by decontamination, or in situ decontamination are cost effective compared with the no-processing option. The technology transfer activities included site presentations and discussions with operations and waste management personnel to identify application opportunities and site-specific considerations and constraints that could affect the implementation of TRU waste conversion principles. These discussions disclosed definite potential for the beneficial application of these principles at most of the sites, but also confirmed the existence of site-specific factors ranging from space limitations to LLW disposal restrictions that could preclude particular applications or diminish expected benefits. 8 refs., 2 figs., 4 tabs.

  3. Treatability study of absorbent polymer waste form for mixed waste treatment

    SciTech Connect (OSTI)

    Herrmann, S. D.; Lehto, M. A.; Stewart, N. A.; Croft, A. D.; Kern, P. W.

    2000-02-10

    A treatability study was performed to develop and characterize an absorbent polymer waste form for application to low level (LLW) and mixed low level (MLLW) aqueous wastes at Argonne National Laboratory-West (ANL-W). In this study absorbent polymers proved effective at immobilizing aqueous liquid wastes in order to meet Land Disposal Restrictions for subsurface waste disposal. Treatment of aqueous waste with absorbent polymers provides an alternative to liquid waste solidification via high-shear mixing with clays and cements. Significant advantages of absorbent polymer use over clays and cements include ease of operations and waste volume minimization. Absorbent polymers do not require high-shear mixing as do clays and cements. Granulated absorbent polymer is poured into aqueous solutions and forms a gel which passes the paint filter test as a non-liquid. Pouring versus mixing of a solidification agent not only eliminates the need for a mixing station, but also lessens exposure to personnel and the potential for spread of contamination from treatment of radioactive wastes. Waste minimization is achieved as significantly less mass addition and volume increase is required of and results from absorbent polymer use than that of clays and cements. Operational ease and waste minimization translate into overall cost savings for LLW and MLLW treatment.

  4. Evaluation of Groundwater Impacts to Support the National Environmental Policy Act Environmental Assessment for the INL Remote-Handled Low-Level Waste Disposal Project

    SciTech Connect (OSTI)

    Annette Schafer; Arthur S. Rood; A. Jeffrey Sondrup

    2011-08-01

    The groundwater impacts have been analyzed for the proposed RH-LLW disposal facility. A four-step analysis approach was documented and applied. This assessment compared the predicted groundwater ingestion dose to the more restrictive of either the 25 mrem/yr all pathway dose performance objective, or the maximum contaminant limit performance objective. The results of this analysis indicate that the groundwater impacts for either proposed facility location are expected to be less than the performance objectives. The analysis was prepared to support the NEPA-EA for the top two ranking of the proposed RH-LLW sites. As such, site-specific conditions were incorporated for each set of results generated. These site-specific conditions were included to account for the transport of radionuclides through the vadose zone and through the aquifer at each site. Site-specific parameters included the thickness of vadose zone sediments and basalts, moisture characteristics of the sediments, and aquifer velocity. Sorption parameters (Kd) were assumed to be very conservative values used in Track II analysis of CERCLA sites at INL. Infiltration was also conservatively assumed to represent higher rates corresponding to disturbed soil conditions. The results of this analysis indicate that the groundwater impacts for either proposed facility location are expected to be less than the performance objectives.

  5. Evaluation of Groundwater Impacts to Support the National Environmental Policy Act Environmental Assessment for the INL Remote-Handled Low-Level Waste Disposal Project

    SciTech Connect (OSTI)

    Annette Schafer; Arthur S. Rood; A. Jeffrey Sondrup

    2011-12-01

    The groundwater impacts have been analyzed for the proposed RH-LLW disposal facility. A four-step analysis approach was documented and applied. This assessment compared the predicted groundwater ingestion dose to the more restrictive of either the 25 mrem/yr all pathway dose performance objective, or the maximum contaminant limit performance objective. The results of this analysis indicate that the groundwater impacts for either proposed facility location are expected to be less than the performance objectives. The analysis was prepared to support the NEPA-EA for the top two ranking of the proposed RH-LLW sites. As such, site-specific conditions were incorporated for each set of results generated. These site-specific conditions were included to account for the transport of radionuclides through the vadose zone and through the aquifer at each site. Site-specific parameters included the thickness of vadose zone sediments and basalts, moisture characteristics of the sediments, and aquifer velocity. Sorption parameters (Kd) were assumed to be very conservative values used in Track II analysis of CERCLA sites at INL. Infiltration was also conservatively assumed to represent higher rates corresponding to disturbed soil conditions. The results of this analysis indicate that the groundwater impacts for either proposed facility location are expected to be less than the performance objectives.

  6. Sublimation-induced orbital perturbations of extrasolar active asteroids and comets: application to white dwarf systems

    E-Print Network [OSTI]

    Veras, Dimitri; Gaensicke, Boris T

    2015-01-01

    The metal budgets in some white dwarf (WD) atmospheres reveal that volatile-rich circumstellar bodies must both exist in extrasolar systems and survive the giant branch phases of stellar evolution. The resulting behaviour of these active asteroids or comets which orbit WDs is not well-understood, but may be be strongly influenced by sublimation due to stellar radiation. Here we develop a model, generally applicable to any extrasolar system with a main sequence or WD star, that traces sublimation-induced orbital element changes in approximately km-sized extrasolar minor planets and comets traveling within hundreds of au. We derive evolution equations on orbital timescales and for arbitrarily steep power-law sublimation dependencies on distance, and place our model in a Solar system context. We also demonstrate the importance of coupling sublimation and general relativity, and the orbital consequences of outgassing in arbitrary directions. We prove that nongravitational accelerations alone cannot result in orbi...

  7. Present State of Electron Backscatter Diffraction and Prospective Developments

    SciTech Connect (OSTI)

    Schwarzer, R A; Field, D P; Adams, B L; Kumar, M; Schwartz, A J

    2008-10-24

    Electron backscatter diffraction (EBSD), when employed as an additional characterization technique to a scanning electron microscope (SEM), enables individual grain orientations, local texture, point-to-point orientation correlations, and phase identification and distributions to be determined routinely on the surfaces of bulk polycrystals. The application has experienced rapid acceptance in metallurgical, materials, and geophysical laboratories within the past decade (Schwartz et al. 2000) due to the wide availability of SEMs, the ease of sample preparation from the bulk, the high speed of data acquisition, and the access to complementary information about the microstructure on a submicron scale. From the same specimen area, surface structure and morphology of the microstructure are characterized in great detail by the relief and orientation contrast in secondary and backscatter electron images, element distributions are accessed by energy dispersive spectroscopy (EDS), wavelength dispersive spectroscopy (WDS), or cathodoluminescence analysis, and the orientations of single grains and phases can now be determined, as a complement, by EBSD.

  8. Relationship between pyrite formation and organic sulfur content of coal as revealed by electron microscopy

    SciTech Connect (OSTI)

    Raymond, R. Jr.; Hagan, R.C.

    1982-01-01

    There are a large number of questions concerning the mode of occurrence of organic sulfur in peat, and what, if anything, alters its occurrence during and after coalification. The formation of pyrite during periods of peatification and coalification has been hypothesized to have a great effect on the organic sulfur content of organic material surrounding the pyrite. Measurement of organic sulfur contents at different distances from pyrite particles would serve as direct experimental proof for or against this pypothesis. A combination of in situ energy dispersive spectrometer (EDS) line profiles, EDS x-ray maps, and WDS analyses across pyrite/coal interfaces in a variety of coals shows unequivocally that formation of pyrite does not alter the organic sulfur contents of the surrounding coal macerals.

  9. ELEMENTAL ABUNDANCES IN THE EJECTA OF OLD CLASSICAL NOVAE FROM LATE-EPOCH SPITZER SPECTRA

    SciTech Connect (OSTI)

    Helton, L. Andrew; Vacca, William D.; Gehrz, Robert D.; Woodward, Charles E.; Shenoy, Dinesh P.; Wagner, R. Mark; Evans, Aneurin; Krautter, Joachim; Schwarz, Greg J.; Starrfield, Sumner

    2012-08-10

    We present Spitzer Space Telescope mid-infrared IRS spectra, supplemented by ground-based optical observations, of the classical novae V1974 Cyg, V382 Vel, and V1494 Aql more than 11, 8, and 4 years after outburst, respectively. The spectra are dominated by forbidden emission from neon and oxygen, though in some cases, there are weak signatures of magnesium, sulfur, and argon. We investigate the geometry and distribution of the late time ejecta by examination of the emission line profiles. Using nebular analysis in the low-density regime, we estimate lower limits on the abundances in these novae. In V1974 Cyg and V382 Vel, our observations confirm the abundance estimates presented by other authors and support the claims that these eruptions occurred on ONe white dwarfs (WDs). We report the first detection of neon emission in V1494 Aql and show that the system most likely contains a CO WD.

  10. Bolometric light curves of supernovae and post-explosion magnetic fields

    E-Print Network [OSTI]

    P. Ruiz-Lapuente; H. Spruit

    1997-11-20

    The various effects leading to diversity in the bolometric light curves of supernovae are examined: nucleosynthesis, kinematic differences, ejected mass, degree of mixing, and configuration and intensity of the magnetic field are discussed. In Type Ia supernovae, a departure in the bolometric light curve from the full-trapping decline of $^{56}$Co can occur within the two and a half years after the explosion, depending on the evolutionary path followed by the WD during the accretion phase. If convection has developed in the WD core during the presupernova evolution, starting several thousand years before the explosion, a tangled magnetic field close to the equipartition value should have grown in the WD. Such an intense magnetic field would confine positrons where they originate from the $^{56}$Co decays, and preclude a strong departure from the full-trapping decline, as the supernova expands. This situation is expected to occur in C+O Chandrasekhar WDs as opposed to edge-lit detonated sub-Chandrasekhar WDs. If the pre-explosion magnetic field of the WD is less intense than 10$^{5-8}$G, a lack of confinement of the positrons emitted in the $^{56}$Co decay and a departure from full-trapping decline would occur. The time at which it takes place can provide estimates of the original magnetic field of the WD, its configuration, and also of the mass of the supernova ejecta. In SN 1991bg, the bolometric light curve suggests absence of a significant tangled magnetic field (intensity lower than $10^{3}$ G). Chandrasekhar-mass models do not reproduce the bolometric light curve of this supernova. For SN 1972E, on the contrary, there is evidence for a tangled configuration of the magnetic field and its light curve is well reproduced by a Chandrasekhar WD explosion.

  11. Type Ia Supernova Scenarios and the Hubble Sequence

    E-Print Network [OSTI]

    P. Ruiz-Lapuente; A. Burkert; R. Canal

    1995-05-19

    The dependence of the Type Ia supernova (SN Ia) rate on galaxy type is examined for three currently proposed scenarios: merging of a Chandrasekhar--mass CO white dwarf (WD) with a CO WD companion, explosion of a sub--Chandrasekhar mass CO WD induced by accretion of material from a He star companion, and explosion of a sub--Chandrasekhar CO WD in a symbiotic system. The variation of the SNe Ia rate and explosion characteristics with time is derived, and its correlation with parent population age and galaxy redshift is discussed. Among current scenarios, CO + He star systems should be absent from E galaxies. Explosion of CO WDs in symbiotic systems could account for the SNe Ia rate in these galaxies. The same might be true for the CO + CO WD scenario, depending on the value of the common envelope parameter. A testable prediction of the sub--Chandrasekhar WD model is that the average brightness and kinetic energy of the SN Ia events should increase with redshift for a given Hubble type. Also for this scenario, going along the Hubble sequence from E to Sc galaxies SNe Ia events should be brighter on average and should show larger mean velocities of the ejecta. The observational correlations strongly suggest that the characteristics of the SNe Ia explosion are linked to parent population age. The scenario in which WDs with masses below the Chandrasekhar mass explode appears the most promising one to explain the observed variation of the SN Ia rate with galaxy type together with the luminosity--expansion velocity trend.

  12. UNDERSTANDING THE EVOLUTION OF CLOSE BINARY SYSTEMS WITH RADIO PULSARS

    SciTech Connect (OSTI)

    Benvenuto, O. G.; De Vito, M. A.

    2014-05-01

    We calculate the evolution of close binary systems (CBSs) formed by a neutron star (behaving as a radio pulsar) and a normal donor star, which evolve either to a helium white dwarf (HeWD) or to ultra-short orbital period systems. We consider X-ray irradiation feedback and evaporation due to radio pulsar irradiation. We show that irradiation feedback leads to cyclic mass transfer episodes, allowing CBSs to be observed in between episodes as binary radio pulsars under conditions in which standard, non-irradiated models predict the occurrence of a low-mass X-ray binary. This behavior accounts for the existence of a family of eclipsing binary systems known as redbacks. We predict that redback companions should almost fill their Roche lobe, as observed in PSR J1723-2837. This state is also possible for systems evolving with larger orbital periods. Therefore, binary radio pulsars with companion star masses usually interpreted as larger than expected to produce HeWDs may also result in such quasi-Roche lobe overflow states, rather than hosting a carbon-oxygen WD. We found that CBSs with initial orbital periods of P{sub i} < 1day evolve into redbacks. Some of them produce low-mass HeWDs, and a subgroup with shorter P{sub i} becomes black widows (BWs). Thus, BWs descend from redbacks, although not all redbacks evolve into BWs. There is mounting observational evidence favoring BW pulsars to be very massive (? 2 M {sub ?}). As they should be redback descendants, redback pulsars should also be very massive, since most of the mass is transferred before this stage.

  13. Superconducting Open-Gradient Magnetic Separation for the Pretreatment of Radioactive or Mixed Waste Vitrification Feeds

    SciTech Connect (OSTI)

    Nunez', L.; Kaminsky', M.D.,; Crawford, C.; Ritter, J.A.

    1999-12-31

    An open-gradient magnetic separation (OGMS) process is being considered to separate deleterious elements from radioactive and mixed waste streams prior to vitrification or stabilization. By physically segregating solid wastes and slurries based on the magnetic properties of the solid constituents, this potentially low-cost process may serve the U.S. Department of Energy (DOE) by reducing the large quantities of glass produced from defense-related high-level waste (HLW). Furthermore, the separation of deleterious elements from low-level waste (LLW) also can reduce the total quantity of waste produced in LLW immobilization activities. Many HLW 'and LLW waste' streams at both Hanford and the Savannah River Site (SRS) include constituents deleterious to the durability of borosilicate glass and the melter many of the constituents also possess paramagnetism. For example, Fe, Cr, Ni, and other transition metals may limit the waste loading and affect the durability of the glass by forming spine1 phases at the high operating temperature used in vitrification. Some magnetic spine1 phases observed in glass formation are magnetite (Fe,O,), chromite (FeCrO,), and others [(Fe, Ni, Mg, Zn, Mn)(Al, Fe, Ti, Cr)O,] as described elsewhere [Bates-1994, Wronkiewicz-1994] Stable spine1 phases can cause segregation between the glass and the crystalline phases. As a consequence of the difference in density, the spine1 phases tend to accumulate at the bottom of the glass melter, which decreases the conductivity and melter lifetime [Sproull-1993]. Crystallization also can affect glass durability [Jantzen-1985, Turcotte- 1979, Buechele-1990] by changing the chemical composition of the matrix glass surrounding the crystals or causing stress at the glass/crystal interface. These are some of the effects that can increase leaching [Jantzen-1985]. A SRS glass that was partially crystallized to contain 10% vol. crystals composed of spinels, nepheline, and acmite phases showed minimal changes in short term leachability [Jantzen-1985, Hench-1982]. However, Jantzen et k > al. found that leaching increased preferentially at grain boundary interfaces [Jantzen-1985]. For a SRL 165 glass crystallized up to 30% vol., leachability measured by normalized boron release increased by a factor of three compared to the uncrystallized glass [Kelly-1975, Plodinec-1979]. In general, the magnitude of the crystallization effect depends highly on glass composition and cooling rate.

  14. RESULTS FOR THE SECOND QUARTER 2011 TANK 50 WAC SLURRY SAMPLE: CHEMICAL AND RADIONUCLIDE CONTAMINANT RESULTS

    SciTech Connect (OSTI)

    Eibling, R.

    2011-08-25

    The Saltstone Facility is designed and permitted to immobilize and dispose of low-level radioactive and hazardous liquid waste (salt solution) remaining from the processing of radioactive material at the Savannah River Site. Low-level waste (LLW) streams from the Effluent Treatment Project (ETP), H-Canyon, the DDA (Deliquification, Dissolution, and Adjustment) process, and the decontaminated salt solution product from the Actinide Removal Process/Modular Caustic Side Solvent Extraction (CSSX) Unit (ARP/MCU) process are stored in Tank 50 until the LLW can be transferred to the Saltstone Facility for treatment and disposal. The LLW must meet the specified waste acceptance criteria (WAC) before it is processed into saltstone. The specific chemical and radionuclide contaminants and their respective WAC limits are listed in the current Saltstone WAC. Waste Solidification Engineering (WSE) requested that Savannah River National Laboratory (SRNL) perform quarterly analysis on saltstone samples. The concentrations of chemical and radionuclide contaminants are measured to ensure the saltstone produced during each quarter is in compliance with the current WAC. This report documents the concentrations of chemical and radionuclide contaminants for the 2011 Second Quarter samples collected from Tank 50 on April 4, 2011 and discusses those results in further detail than the previously issued results report. The following conclusions are drawn from the analytical results provided in this report: (1) The concentrations of the reported chemical and radioactive contaminants were less than their respective WAC targets or limits unless noted in this section. (2) The reported detection limit for {sup 59}Ni is above the requested limit from Reference 2 but below the established limit in Reference 3. (3) The reported detection limit for {sup 94}Nb is above the requested limit from Reference 2; however, it is below the established limits in Reference 3. (4) The reported concentration of {sup 242m}Am is above the target in Listed in Attachment 8.4 of the Saltstone WAC. (5) {sup 247}Cm and {sup 249}Cf are above the requested limits from Reference 2. However, they are below the limits established in Reference 3. (6) The reported detection limit for Norpar 13{sup 5} is greater than the limit from Table 4 and Attachment 8.2 of the WAC. (7) The reported detection limit for Isopar L is greater than the limit from Table 3 of the WAC. (8) Isopar L and Norpar 13 have limited solubility in aqueous solutions making it difficult to obtain consistent and reliable sub-samples. The values reported in this memo are the concentrations in the sub-sample as detected by the instrument; however, the results may not accurately represent the concentrations of the analytes in Tank 50.

  15. Lessons Learned in the Design and Use of IP1 / IP2 Flexible Packaging - 13621

    SciTech Connect (OSTI)

    Sanchez, Mike; Reeves, Wendall; Smart, Bill

    2013-07-01

    For many years in the USA, Low Level Radioactive Waste (LLW), contaminated soils and construction debris, have been transported, interim stored, and disposed of, using IP1 / IP2 metal containers. The performance of these containers has been more than adequate, with few safety occurrences. The containers are used under the regulatory oversight of the US Department of Transportation (DOT), 49 Code of Federal Regulations (CFR). In the late 90's the introduction of flexible packaging for the transport, storage, and disposal of low level contaminated soils and construction debris was introduced. The development of flexible packaging came out of a need for a more cost effective package, for the large volumes of waste generated by the decommissioning of many of the US Department of Energy (DOE) legacy sites across the US. Flexible packaging had to be designed to handle a wide array of waste streams, including soil, gravel, construction debris, and fine particulate dust migration. The design also had to meet all of the IP1 requirements under 49CFR 173.410, and be robust enough to pass the IP2 testing 49 CFR 173.465 required for many LLW shipments. Tens of thousands of flexible packages have been safely deployed and used across the US nuclear industry as well as for hazardous non-radioactive applications, with no recorded release of radioactive materials. To ensure that flexible packages are designed properly, the manufacturer must use lessons learned over the years, and the tests performed to provide evidence that these packages are suitable for transporting low level radioactive wastes. The design and testing of flexible packaging for LLW, VLLW and other hazardous waste streams must be as strict and stringent as the design and testing of metal containers. The design should take into consideration the materials being loaded into the package, and should incorporate the right materials, and manufacturing methods, to provide a quality, safe product. Flexible packaging can be shown to meet the criteria for safe and fit for purpose packaging, by meeting the US DOT regulations, and the IAEA Standards for IP-1 and IP-2 including leak tightness. (authors)

  16. Testing of a portable ultrahigh pressure water decontamination system (UHPWDS)

    SciTech Connect (OSTI)

    Lovell, A.; Dahlby, J.

    1996-02-01

    This report describes the tests done with a portable ultrahigh pressure water decontamination system (UHPWDS) on highly radioactively contaminated surfaces. A small unit was purchased, modified, and used for in-situ decontamination to change the waste level of the contaminated box from transuranic (TRU) waste to low- level waste (LLW). Low-level waste is less costly by as much as a factor of five or more if compared with TRU waste when handling, storage, and disposal are considered. The portable unit we tested is commercially available and requires minimal utilities for operation. We describe the UHPWDS unit itself, a procedure for its use, the results of the testing we did, and conclusions including positive and negative aspects of the UHPWDS.

  17. Composite analysis E-area vaults and saltstone disposal facilities

    SciTech Connect (OSTI)

    Cook, J.R.

    1997-09-01

    This report documents the Composite Analysis (CA) performed on the two active Savannah River Site (SRS) low-level radioactive waste (LLW) disposal facilities. The facilities are the Z-Area Saltstone Disposal Facility and the E-Area Vaults (EAV) Disposal Facility. The analysis calculated potential releases to the environment from all sources of residual radioactive material expected to remain in the General Separations Area (GSA). The GSA is the central part of SRS and contains all of the waste disposal facilities, chemical separations facilities and associated high-level waste storage facilities as well as numerous other sources of radioactive material. The analysis considered 114 potential sources of radioactive material containing 115 radionuclides. The results of the CA clearly indicate that continued disposal of low-level waste in the saltstone and EAV facilities, consistent with their respective radiological performance assessments, will have no adverse impact on future members of the public.

  18. Disposal concepts and characteristics of existing and potential low-waste repositories - 9076

    SciTech Connect (OSTI)

    Johnson, Peter J [Los Alamos National Laboratory; Zarling, John C [Los Alamos National Laboratory

    2009-01-01

    The closure of the Barnwell low-level waste (LLW) disposal facility to non-Atlantic Compact users poses significant problems for organizations seeking to remove waste material from public circulation. Beta-gamma sources such as {sup 137}Cs and {sup 90}Sr in particular create problems because in 36 states no path forward exists for disposal. Furthermore, several other countries are considering disposition of sealed sources in a variety of facilities. Like much of the United States, many of these countries currently have no means of disposal. Consequently, there is a greater tendency for sources to be misplaced or stored in insufficient facilities, resulting in an increased likelihood of unwitting exposure of nearby people to radioactive materials. This paper provides an overview of the various disposal concepts that have been employed or attempted in the United States. From these concepts, a general overview of characteristics necessary for long-term disposal is synthesized.

  19. A DAW lockout scheme

    SciTech Connect (OSTI)

    Landeche, D.A. [Entergy Operations, Inc., Killona, LA (United States)

    1996-10-01

    In it`s document TR-105834, Project 24414 dated November 1995 and titled {open_quotes}Zero Plastics and the Radiologically Protected Area Low Level Waste Lockout Program{close_quotes}, EPRI describes DAW lockout as a {open_quotes}very advanced{close_quotes} and {open_quotes}perhaps the most powerful of all approaches to LLW minimization.{close_quotes} The report discusses implementation of a lockout program at Kewaunee, LaSalle and Zion stations but states that {open_quotes}no quantifiable results{close_quotes} were available at time of report printing. Waterford 3 Nuclear Station began implementing a limited lockout scheme in 1993. This paper presents Waterford`s experience and results achieved through the lockout approach.

  20. Performance assessment for a hypothetical low-level waste disposal facility

    SciTech Connect (OSTI)

    Smith, C.S.; Rohe, M.J.; Ritter, P.D.

    1997-01-01

    Disposing of low-level waste (LLW) is a concern for many states throughout the United States. A common disposal method is below-grade concrete vaults. Performance assessment analyses make predictions of contaminant release, transport, ingestion, inhalation, or other routes of exposure, and the resulting doses for various disposal methods such as the below-grade concrete vaults. Numerous assumptions are required to simplify the processes associated with the disposal facility to make predictions feasible. In general, these assumptions are made conservatively so as to underestimate the performance of the facility. The objective of this report is to describe the methodology used in conducting a performance assessment for a hypothetical waste facility located in the northeastern United States using real data as much as possible. This report consists of the following: (a) a description of the disposal facility and site, (b) methods used to analyze performance of the facility, (c) the results of the analysis, and (d) the conclusions of this study.

  1. Evaluating off-site disposal of low-level waste at LANL-9498

    SciTech Connect (OSTI)

    Hargis, Kenneth M [Los Alamos National Laboratory; French, Sean B [Los Alamos National Laboratory; Boyance, Julien A [NORTH WIND, INC.

    2009-01-01

    Los Alamos National Laboratory generates a wide range of waste types, including solid low-level radioactive waste (LL W), in conducting its national security mission and other science and technology activities. Although most ofLANL's LLW has been disposed on-site, limitations on expansion, stakeholder concerns, and the potential for significant volumes from environmental remediation and decontamination and demolition (D&D) have led LANL to evaluate the feasibility of increasing off-site disposal. It appears that most of the LL W generated at LANL would meet the Waste Acceptance Criteria at the Nevada Test Site or the available commercial LL W disposal site. Some waste is considered to be problematic to transport to off-site disposal even though it could meet the off-site Waste Acceptance Criteria. Cost estimates for off-site disposal are being evaluated for comparison to estimated costs under the current plans for continued on-site disposal.

  2. Volatility literature of chlorine, iodine, cesium, strontium, technetium, and rhenium; technetium and rhenium volatility testing

    SciTech Connect (OSTI)

    Langowski, M.H.; Darab, J.G.; Smith, P.A.

    1996-03-01

    A literature review pertaining to the volatilization of Sr, Cs, Tc (and its surrogate Re), Cl, I and other related species during the vitrification of Hanford Low Level Waste (LLW) streams has been performed and the relevant information summarized. For many of these species, the chemistry which occurs in solution prior to the waste stream entering the melter is important in dictating their loss at higher temperatures. In addition, the interactive effects between the species being lost was found to be important. A review of the chemistries of Tc and Re was also performed. It was suggested that Re would indeed act as an excellent surrogate for Tc in non-radioactive materials testing. Experimental results on Tc and Re loss from sodium aluminoborosilicate melts of temperatures ranging from 900--1350{degrees}C performed at PNL are reported and confirm that Re behaves in a nearly identical manner to that of technetium.

  3. Preliminary design requirements document for Project W-378, low-level waste vitrification plant

    SciTech Connect (OSTI)

    Swanson, L.M.

    1995-03-31

    The scope of this preliminary Design Requirements Document (DRD) is to identify and define the functions, with associated requirements, which must be performed to accomplish vitrification and disposal of the pretreated low-level waste (LLW) fraction of the Hanford Site tank waste. This document sets forth function requirements, performance requirements and design constraints necessary to begin conceptual design for the Low-Level Waste Vitrification Plant (LLWVP). System and physical interfaces between the LLWVP Project and the Tank Waste Remediation System (TWRS) are identified. The constraints, performance requirements, and transfer of information and data across a technical interface will be documented in an Interface Control Document. The design requirements provided in this document will be augmented by additional detailed design data to be documented by the project.

  4. An empirical modeling approach to high sodium glass durability

    SciTech Connect (OSTI)

    Shine, E.P.; Sadler, A.L.K. [Westinghouse Savannah River Company, Aiken, SC (United States)

    1996-12-31

    Empirical mixture models have been developed for chemical durability of high sodium borosilicate glass. The response of boron to a seven-day Product Consistency Test (PCT) was chosen as the measure of durability. The objective of the model development was to support the proposed vitrification of Hanford low-level waste (LLW), the bulk of which is primarily sodium oxide. A full first-order model and a second order model were developed from a database of high-sodium borosilicate glasses. First-order models proved to be satisfactory in a qualitative sense, but root mean squared errors were fairly large for quantitative predictive purposes. The results imply that mechanistic models relating durability to composition should include higher order compositional interactions; a second-order model yielded much improved statistics. The modeling results also suggest that calcium, which is considered a network modifier yet is also regarded as a glass {open_quotes}stiffener{close_quotes}, may improve durability.

  5. Bench-scale studies with mercury contaminated SRS soil

    SciTech Connect (OSTI)

    Cicero, C.A.

    1996-05-08

    The Savannah River Technology Center (SRTC) has been charactered by the Department of Enregy (DOE) - Office of Technology Development (OTD) to investigate vitrification technology for the treatment of Low Level Mixed Wastes (LLMW). In fiscal year 1995, LLW streams containing mercury and organics were targeted. This report will present the results of studies with mercury contaminated waste. In order to successfully apply vitrification technology to LLMW, the types and quantities of glass forming additives necessary for producing homogeneous glasses from the wastes had to be determined, and the treatment for the mercury portion had to also be determined. The selected additives had to ensure that a durable and leach resistant waste form was produced, while the mercury treatment had to ensure that hazardous amounts of mercury were not released into the environment.

  6. Pretreatment of neutralized cladding removal waste sludge: Status Report

    SciTech Connect (OSTI)

    Lumetta, G J; Swanson, J L

    1993-03-01

    This report describes the status of process development for pretreating Hanford neutralized cladding removal waste (NCRW) sludge, of which [approximately] 3.3 [times] 10[sup 6] L is stored in Tanks 103-AW and 105-AW at the Hanford Site. The initial baseline process chosen for pretreating NCRW sludge is to dissolve the sludge in nitric acid and extract the -transuranic (MU) elements from the dissolved sludge solution with octyl(phenyl)-N,N-diisobutylcarbamoyl methyl phosphine oxide (CNWO). This process converts the NCRW sludge into a relatively large volume of low-level waste (LLW) to be disposed of as grout, leaving only a small volume of high-level waste (HLW) requiring vitrification in the Hanford Waste Vitrification Plant (HWVP).

  7. Final Environmental Assessment and Finding of No Significant Impact: Waste Disposition Activities at the Paducah Site Paducah, Kentucky

    SciTech Connect (OSTI)

    N /A

    2002-11-05

    The U.S. Department of Energy (DOE) has completed an environmental assessment (DOE/EA-1339), which is incorporated herein by reference, for proposed disposition of polychlorinated biphenyl (PCB) wastes, low-level radioactive waste (LLW), mixed low-level radioactive waste (MLLW), and transuranic (TRU) waste from the Paducah Gaseous Diffusion Plant Site (Paducah Site) in Paducah, Kentucky. All of the wastes would be transported for disposal at various locations in the United States. Based on the results of the impact analysis reported in the EA, DOE has determined that the proposed action is not a major federal action that would significantly affect the quality of the human environment with in the context of the National Environmental Policy Act of 1969 (NEPA). Therefore, preparation of an environmental impact statement is not necessary, and DOE is issuing this Finding of No Significant Impact (FONSI).

  8. Low-Level Waste Forum notes and summary reports for 1994. Volume 9, Number 3, May-June 1994

    SciTech Connect (OSTI)

    NONE

    1994-06-01

    This issue includes the following articles: Vermont ratifies Texas compact; Pennsylvania study on rates of decay for classes of low-level radioactive waste; South Carolina legislature adjourns without extending access to Barnwell for out-of-region generators; Southeast Compact Commission authorizes payments for facility development, also votes on petitions, access contracts; storage of low-level radioactive waste at Rancho Seco removed from consideration; plutonium estimates for Ward Valley, California; judgment issued in Ward Valley lawsuits; Central Midwest Commission questions court`s jurisdiction over surcharge rebates litigation; Supreme Court decides commerce clause case involving solid waste; parties voluntarily dismiss Envirocare case; appellate court affirms dismissal of suit against Central Commission; LLW Forum mixed waste working group meets; US EPA Office of Radiation and Indoor Air rulemakings; EPA issues draft radiation site cleanup regulation; EPA extends mixed waste enforcement moratorium; and NRC denies petition to amend low-level radioactive waste classification regulations.

  9. Low-level waste certification plan for the WSCF Laboratory Complex

    SciTech Connect (OSTI)

    Morrison, J.A.

    1994-09-19

    The solid, low-level waste certification plan for the Waste Sampling and Characterization Facility (WSCF) describes the organization and methodology for the certification of the solid low-level waste (LLW) that is transferred to the Hanford Site 200 Areas Storage and Disposal Facilities. This plan incorporates the applicable elements of waste reduction, including up-front minimization, and end product treatment to reduce the volume or toxicity of the waste. The plan also includes segregation of different waste types. This low-level waste certification plan applies only to waste generated in, or is the responsibility of the WSCF Laboratory Complex. The WSCF Laboratory Complex supports technical activities performed at the Hanford Site. Wet Chemical and radiochemical analyses are performed to support site operations, including environmental and effluent monitoring, chemical processing, RCRA and CERCLA analysis, and waste management activities. Environmental and effluent samples include liquid effluents, ground and surface waters, soils, animals, vegetation, and air filters.

  10. Safety Evaluation for Packaging for onsite Transfer of plutonium recycle test reactor ion exchange columns

    SciTech Connect (OSTI)

    Smith, R.J.

    1995-09-11

    The purpose of this Safety Evaluation for Packaging (SEP) is to authorize the use of three U.S. Department of Transportation (DOT) 7A, Type A metal boxes (Capital Industries Part No. S 0600-0600-1080- 0104) to package 12 Plutonium Recycle Test Reactor (PRTR) ion exchange columns as low-level waste (LLW). The packages will be transferred from the 309 Building in the 300 Area to low level waste burial in the 200 West Area. Revision 1 of WHC-SD-TP-SEP-035 (per ECN No. 621467) documents that the boxes containing ion exchange columns and grout will maintain the payload under normal conditions of transport if transferred without the box lids

  11. Annual Transportation Report for Radioactive Waste Shipments to and from the Nevada Test Site, Fiscal Year 2009

    SciTech Connect (OSTI)

    U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office

    2010-02-01

    In February 1997, the U.S. Department of Energy (DOE), Nevada Operations Office (now known as the Nevada Site Office) issued the Mitigation Action Plan which addressed potential impacts described in the Final Environmental Impact Statement for the Nevada Test Site and Off-Site Locations in the State of Nevada (DOE/EIS 0243). The DOE, Nevada Operations Office committed to several actions, including the preparation of an annual report, which summarizes waste shipments to and from the Nevada Test Site (NTS) Radioactive Waste Management Site (RWMS) at Area 5 and Area 3. Since 2006, the Area 3 RWMS has been in cold stand-by. This document satisfies requirements regarding low-level radioactive waste (LLW) and mixed low-level radioactive waste (MLLW) transported to and from the NTS during FY 2009. In addition, this document provides shipment, volume, and route information on transuranic (TRU) waste shipped from the NTS to the Idaho National Laboratory, near Idaho Falls, Idaho.

  12. Groundwater monitoring in the Savannah River Plant Low Level Waste Burial Ground

    SciTech Connect (OSTI)

    Carlton, W.H.

    1983-12-31

    This document describes chemical mechanisms that may affect trace-level radionuclide migration through acidic sandy clay soils in a humid environment, and summarizes the extensive chemical and radiochemical analyses of the groundwater directly below the SRP Low-Level Waste (LLW) Burial Ground (643-G). Anomalies were identified in the chemistry of individual wells which appear to be related to small amounts of fission product activity that have reached the water table. The chemical properties which were statistically related to trace level transport of Cs-137 and Sr-90 were iron, potassium, sodium and calcium. Concentrations on the order of 100 ppM appear sufficient to affect nuclide migration. Several complexation mechanisms for plutonium migration were investigated.

  13. Low Level and Transuranic Waste Segregation and Low Level Waste Characterization at the 200 Area of the Hanford Site - 12424

    SciTech Connect (OSTI)

    Donohoue, Tom; Martin, E. Ray; Mason, John A. [ANTECH Corporation 9050 Marshall Court, Westminster, CO, 80031 (United States); Blackford, Ty; Estes, Michael; Jasen, William [CH2M Hill Plateau Remediation Company, 2420 Stevens Drive, Richland, WA, 99352 (United States); Cahill, Michael [Fluor Federal Services, 1200 Jadwin Avenue, Richland, WA, 99352 (United States)

    2012-07-01

    This paper describes the waste measurement and waste characterization activities carried out by ANTECH Corporation (ANTECH) and CH2M Hill Plateau Remediation Company (CHPRC) at the 200 Area of the Hanford Site under Contracts No. 22394 and No. 40245 for the US Department of Energy (DOE). These include Low Level Waste (LLW) and Transuranic (TRU) Waste segregation and LLW characterization for both 55-gallon (200-litre) drums with gross weight up to 454 kg and 85-gallon over-pack drums. In order to achieve efficient and effective waste drum segregation and assay, ANTECH deployed an automated Gamma Mobile Assay Laboratory (G-MAL) at the trench face in both 200 Area West and East. The unit consists of a modified 40 foot ISO shipping container with an automatic flow through roller conveyor system with internal drum weigh scale, four measurement and drum rotation positions, and four high efficiency high purity Germanium (HPGe) detectors with both detector and shadow shields. The unit performs multiple far-field measurements and is able to segregate drums at levels well below 100 nCi/g. The system is sufficiently sensitive that drums, which are classified as LLW, are characterized at measurement levels that meet the Environmental Restoration Disposal Facility (ERDF) Waste Acceptance Criteria (WAC). With measurement times of between 20 and 30 minutes the unit can classify and characterize over 40 drums in an 8-hour shift. The system is well characterized with documented calibrations, lower limits of detection (LLD) and total measurement uncertainty. The calibrations are confirmed and verified using nationally traceable standards in keeping with the CHPRC measurement requirements. The performance of the system has been confirmed and validated throughout the measurement process by independent CHPRC personnel using traceable standards. All of the measurement and maintenance work has been conducted during the period under a Quality Assurance Plan (QAP) compliant with the applicable criteria of NQA-1 (2000). This includes not only the calibrations and measurements but also the data analysis activities of the ANTECH Subject Matter Experts (SME) and ANTECH support and maintenance activities as well as the activities of CHPRC staff who recover, transport and load waste drums and disposition measured and characterized drums. The overall processes of drum recovery and analysis are described in the paper. Specific spectral data is presented which illustrates the segregation, sentencing and assay process for different types of drums with different radionuclide profiles. The process of identifying and quantifying a wide range of non-TRU radionuclide isotopes is explained and illustrated with spectral examples. The difficulties associated with the measurement of drums with a high gamma ray background, usually arising from high levels of Cs-137 are considered. These drums, which would normally be declared indeterminate and treated as TRU, are addressed under contract No. 40245 by the deployment of the ANTECH Neutron Mobile Assay Laboratory. This is an Active-Passive neutron assay system housed in a modified ISO shipping container. The unit is designed for the measurement and assay of both drums and crates (including B-25 boxes and SWB containers) and will quantify the content of both plutonium and uranium. The neutron system has been employed to perform further evaluation on indeterminate drums to classify them to either LLW or TRU. The experiences of both gamma ray and neutron system operation in different conditions are described; as are the issues of throughput, drum handling and system maintenance. All of these are considered in the overriding context of safe drum handling and safe assay system operation. (authors)

  14. PUBLIC AND REGULATORY ACCEPTANCE OF BLENDING OF RADIOACTIVE WASTE VS DILUTION

    SciTech Connect (OSTI)

    Goldston, W.

    2010-11-30

    On April 21, 2009, the Energy Facilities Contractors Group (EFCOG) Waste Management Working Group (WMWG) provided a recommendation to the Department of Energy's Environmental Management program (DOE-EM) concerning supplemental guidance on blending methodologies to use to classify waste forms to determine if the waste form meets the definition of Transuranic (TRU) Waste or can be classified as Low-Level Waste (LLW). The guidance provides specific examples and methods to allow DOE and its Contractors to properly classify waste forms while reducing the generation of TRU wastes. TRU wastes are much more expensive to characterize at the generator's facilities, ship, and then dispose at the Waste Isolation Pilot Plant (WIPP) than Low-Level Radioactive Waste's disposal. Also the reduction of handling and packaging of LLW is inherently less hazardous to the nuclear workforce. Therefore, it is important to perform the characterization properly, but in a manner that minimizes the generation of TRU wastes if at all possible. In fact, the generation of additional volumes of radioactive wastes under the ARRA programs, this recommendation should improve the cost effective implementation of DOE requirements while properly protecting human health and the environment. This paper will describe how the message of appropriate, less expensive, less hazardous blending of radioactive waste is the 'right' thing to do in many cases, but can be confused with inappropriate 'dilution' that is frowned upon by regulators and stakeholders in the public. A proposal will be made in this paper on how to communicate this very complex and confusing technical issue to regulatory bodies and interested stakeholders to gain understanding and approval of the concept. The results of application of the proposed communication method and attempt to change the regulatory requirements in this area will be discussed including efforts by DOE and the NRC on this very complex subject.

  15. Innovative vitrification for soil remediation

    SciTech Connect (OSTI)

    Jetta, N.W.; Patten, J.S.; Hart, J.G.

    1995-12-01

    The objective of this DOE demonstration program is to validate the performance and operation of the Vortec Cyclone Melting System (CMS{trademark}) for the processing of LLW contaminated soils found at DOE sites. This DOE vitrification demonstration project has successfully progressed through the first two phases. Phase 1 consisted of pilot scale testing with surrogate wastes and the conceptual design of a process plant operating at a generic DOE site. The objective of Phase 2, which is scheduled to be completed the end of FY 95, is to develop a definitive process plant design for the treatment of wastes at a specific DOE facility. During Phase 2, a site specific design was developed for the processing of LLW soils and muds containing TSCA organics and RCRA metal contaminants. Phase 3 will consist of a full scale demonstration at the DOE gaseous diffusion plant located in Paducah, KY. Several DOE sites were evaluated for potential application of the technology. Paducah was selected for the demonstration program because of their urgent waste remediation needs as well as their strong management and cost sharing financial support for the project. During Phase 2, the basic nitrification process design was modified to meet the specific needs of the new waste streams available at Paducah. The system design developed for Paducah has significantly enhanced the processing capabilities of the Vortec vitrification process. The overall system design now includes the capability to shred entire drums and drum packs containing mud, concrete, plastics and PCB`s as well as bulk waste materials. This enhanced processing capability will substantially expand the total DOE waste remediation applications of the technology.

  16. Volatilization of heavy metals and radionuclides from soil heated in an induction ``cold`` crucible melter

    SciTech Connect (OSTI)

    Aloy, A.S.; Belov, V.Z.; Trofimenko, A.S. [Khlopin Radium Inst., St. Petersburg (Russian Federation); Dmitriev, S.A.; Stefanovsky, S.V. [SIA Radon, Moscow (Russian Federation); Gombert, D.; Knecht, D.A. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States)

    1997-12-31

    The behavior of heavy metals and radionuclides during high-temperature treatment is very important for the design and operational capabilities of the off-gas treatment system, as well as for a better understanding of the nature and forms of the secondary waste. In Russia, a process for high-temperature melting in an induction heated cold crucible system is being studied for vitrification of Low Level Waste (LLW) flyash and SYNROC production with simulated high level waste (HLW). This work was done as part of a Department of Energy (DOE) funded research project for thermal treatment of mixed low level waste (LLW). Soil spiked with heavy metals (Cd, Pb) and radionuclides (Cs-137, U-239, Pu-239) was used as a waste surrogate. The soil was melted in an experimental lab-scale system that consisted of a high-frequency generator (1.76 MHz, 60 kW), a cold crucible melter (300 mm high and 90 mm in diameter), a shield box, and an off-gas system. The process temperature was 1,350--1,400 C. Graphite and silicon carbide were used as sacrificial conductive materials to start heating and initial melting of the soil batch. The off-gas system was designed in such a manner that after each experiment, it can be disconnected to collect and analyze all deposits to determine the mass balance. The off-gases were also sampled during an experiment to analyze for hydrogen, NO{sub x}, carbon dioxide, carbon monoxide and chlorine formation. This paper describes distribution and mass balance of metals and radionuclides in various parts of the off-gas system. The leach rate of the solidified blocks identified by the PCT method is also reported.

  17. The incorporation of P, S, Cr, F, Cl, I, Mn, Ti, U, and Bi into simulated nuclear waste glasses: Literature study

    SciTech Connect (OSTI)

    Langowski, M.H.

    1996-02-01

    Waste currently stored on the Hanford Reservation in underground tanks will be into High Level Waste (HLW) and Low Level Waste (LLW). The HLW melter will high-level and transuranic wastes to a vitrified form for disposal in a geological repository. The LLW melter will vitrify the low-level waste which is mainly a sodium solution. Characterization of the tank wastes is still in progress, and the pretreatment processes are still under development Apart from tank-to-tank variations, the feed delivered to the HLW melter will be subject to process control variability which consists of blending and pretreating the waste. The challenge is then to develop glass formulation models which can produce durable and processable glass compositions for all potential vitrification feed compositions and processing conditions. The work under HLW glass formulation is to study and model glass and melt pro functions of glass composition and temperature. The properties of interest include viscosity, electrical conductivity, liquidus temperature, crystallization, immiscibility durability. It is these properties that determine the glass processability and ac waste glass. Apart from composition, some properties, such as viscosity are affected by temperature. The processing temperature may vary from 1050{degrees}C to 1550{degrees}C dependent upon the melter type. The glass will also experience a temperature profile upon cooling. The purpose of this letter report is to assess the expected vitrification feed compositions for critical components with the greatest potential impact on waste loading for double shell tank (DST) and single shell tank (SST) wastes. The basis for critical component selection is identified along with the planned approach for evaluation. The proposed experimental work is a crucial part of model development and verification.

  18. ALTERNATIVES ANALYSIS FOR SELECTING ET #3 SITE

    SciTech Connect (OSTI)

    Collard, L.; Hamm, L.

    2012-02-13

    Engineered trenches (ETs) are considered to be a cost-effective method for disposing Low Level Waste (LLW). Based on waste forecasts from waste generators, the last engineered trench in operation (ET No.2) is anticipated to close in FY14, requiring development of a new ET. Solid Waste requested that SRNL develop an assessment report that reviews four disposal options for this new ET (ET No.3) and determine which option would provide the 'best' Performance Assessment (PA) disposal limits for LLW (Appendix A). Those four options (see option footprint locations in Figure 1-1) are: (1) Disposal at grade on TRU Pads 7-13 where soil would be mounded over waste packages; (2) Excavation at a slightly modified SLIT No.13 location - near the Used Equipment Storage Area; (3) Excavation at a modified SLIT No.12 location - near the 643-26E Naval Reactor Component Disposal Area; and (4) Excavation east of TRU Pad No.26 that replaces northeast portions of four slit trench (ST) disposal units in the eastern set of STs. The assessment consisted of both quantitative and qualitative analyses. The quantitative analysis captured key aspects that were readily quantifiable and had predictable impacts on limits and doses. A simplified modeling strategy stemming from current Special Analysis (SA) practices was employed. Both inventory capacity for a specific nuclide (a quasi-inventory limit) and overall performance for specified inventory mixtures (doses resulting from historical inventories) were considered. The qualitative analysis evaluated other key aspects based on engineering judgment in the form of pros and cons.

  19. Enhancing RESRAD-OFFSITE for Low Level Waste Disposal Facility Performance Assessment

    Broader source: Energy.gov [DOE]

    Abstract: The RESRAD-OFFSITE code was developed to evaluate the radiological dose and excess cancer risk to an individual who is exposed while located within or outside the area of initial (primary) contamination. The primary contamination, which is the source of all releases modeled by the code, is assumed to be a layer of soil. The code considers the release of contamination from the source to the atmosphere, to surface runoff, and to groundwater. The radionuclide leaching was modeled as a first order (without transport) release using radionuclide distribution coefficient and infiltration rate calculated from water balance (precipitation, surface runoff, evapotranspiration, etc.). Recently, a new source term model was added the RESRAD-OFFSITE code so that it can be applied to the evaluation of Low Level Waste (LLW) disposal facility performance assessment. This new improved source term model include (1) first order with transport, (2) equilibrium desorption (rinse) release, and (3) uniform release (constant dissolution). With these new source release options, it is possible to simulate both uncontainerized (soil) contamination and containerized (waste drums) contamination. A delay time in the source release was also added to the code. This allows modeling the LLW container degradation as a function of time. The RESRAD-OFFSITE code also allows linking to other codes using improved flux and concentration input options. Additional source release model such as diffusion release may be added later. In addition, radionuclide database with 1252 radionuclides (ICRP 107) and the corresponding dose coefficients (DCFPAK 3.02) and the Department of Energys new gender- and age-averaged Reference Person dose coefficients (DOE-STD-1196-2011) which is based on the US census data will be added to the next version of RESRAD-OFFSITE code

  20. Decommissioning of the secondary containment of the steam generating heavy water reactor at UKAEA-Winfrith

    SciTech Connect (OSTI)

    Miller, Keith; Cornell, Rowland; Parkinson, Steve; McIntyre, Kevin; Staples, Andy

    2007-07-01

    Available in abstract form only. Full text of publication follows: The Winfrith SGHWR was a prototype nuclear power plant operated for 23 years by the United Kingdom Atomic Energy Authority (UKAEA) until 1990 when it was shut down permanently. The current Stage 1 decommissioning contract is part of a multi-stage strategy. It involves the removal of all the ancillary plant and equipment in the secondary containment and non-containment areas ahead of a series of contracts for the decommissioning of the primary containment, the reactor core and demolition of the building and all remaining facilities. As an outcome of a competitive tending process, the Stage 1 decommissioning contract was awarded to NUKEM with operations commencing in April 2005. The decommissioning processes involved with these plant items will be described with some emphasis of the establishment of multiple work-fronts for the production, recovery, treatment and disposal of mainly tritium-contaminated waste arising from its contact with the direct cycle reactor coolant. The means of size reduction of a variety of large, heavy and complex items of plant made from a range of materials will also be described with some emphasis on the control of fumes during hot cutting operations and establishing effective containments within a larger secondary containment structure. Disposal of these wastes in a timely and cost-effective manner is a major challenge facing the decommissioning team and has required the development of a highly efficient means of packing the resultant materials into mainly one-third height ISO containers for disposal as LLW. Details of the quantities of LLW and exempt wastes handled during this process will be given with a commentary about the difficulty in segregating these two waste streams efficiently. (authors)

  1. Landfill stabilization focus area: Technology summary

    SciTech Connect (OSTI)

    1995-06-01

    Landfills within the DOE Complex as of 1990 are estimated to contain 3 million cubic meters of buried waste. The DOE facilities where the waste is predominantly located are at Hanford, the Savannah River Site (SRS), the Idaho National Engineering Laboratory (INEL), the Los Alamos National Laboratory (LANL), the Oak Ridge Reservation (ORR), the Nevada Test Site (NTS), and the Rocky Flats Plant (RFP). Landfills include buried waste, whether on pads or in trenches, sumps, ponds, pits, cribs, heaps and piles, auger holes, caissons, and sanitary landfills. Approximately half of all DOE buried waste was disposed of before 1970. Disposal regulations at that time permitted the commingling of various types of waste (i.e., transuranic, low-level radioactive, hazardous). As a result, much of the buried waste throughout the DOE Complex is presently believed to be contaminated with both hazardous and radioactive materials. DOE buried waste typically includes transuranic-contaminated radioactive waste (TRU), low-level radioactive waste (LLW), hazardous waste per 40 CFR 26 1, greater-than-class-C waste per CFR 61 55 (GTCC), mixed TRU waste, and mixed LLW. The mission of the Landfill Stabilization Focus Area is to develop, demonstrate, and deliver safer,more cost-effective and efficient technologies which satisfy DOE site needs for the remediation and management of landfills. The LSFA is structured into five technology areas to meet the landfill remediation and management needs across the DOE complex. These technology areas are: assessment, retrieval, treatment, containment, and stabilization. Technical tasks in each of these areas are reviewed.

  2. INNOVATIVE FOSSIL FUEL FIRED VITRIFICATION TECHNOLOGY FOR SOIL REMEDIATION

    SciTech Connect (OSTI)

    J. Hnat; L.M. Bartone; M. Pineda

    2001-10-31

    This Final Report summarizes the progress of Phases 3,3A and 4 of a waste technology Demonstration Project sponsored under a DOE Environmental Management Research and Development Program and administered by the U.S. Department of Energy National Energy Technology Laboratory-Morgantown (DOE-NETL) for an ''Innovative Fossil Fuel Fired Vitrification Technology for Soil Remediation''. The Summary Reports for Phases 1 and 2 of the Program were previously submitted to DOE. The total scope of Phase 3 was to have included the design, construction and demonstration of Vortec's integrated waste pretreatment and vitrification process for the treatment of low level waste (LLW), TSCA/LLW and mixed low-level waste (MLLW). Due to funding limitations and delays in the project resulting from a law suit filed by an environmental activist and the extended time for DOE to complete an Environmental Assessment for the project, the scope of the project was reduced to completing the design, construction and testing of the front end of the process which consists of the Material Handling and Waste Conditioning (MH/C) Subsystem of the vitrification plant. Activities completed under Phases 3A and 4 addressed completion of the engineering, design and documentation of the MH/C System such that final procurement of the remaining process assemblies can be completed and construction of a Limited Demonstration Project be initiated in the event DOE elects to proceed with the construction and demonstration testing of the MH/C Subsystem. Because of USEPA policies and regulations that do not require treatment of low level or low-level/PCB contaminated wastes, DOE terminated the project because there is no purported need for this technology.

  3. A simulation of the transport and fate of radon-222 derived from thorium-230 low-level waste in the near-surface zone of the Radioactive Waste Management Site in Area 5 of the Nevada Test Site

    SciTech Connect (OSTI)

    Lindstrom, F.T.; Cawlfield, D.E.; Donahue, M.E.; Emer, D.F.; Shott, G.J.

    1993-12-01

    US Department of Energy (DOE) Order 5820.2A (DOE, 1988) requires performance assessments on all new and existing low-level radioactive waste (LLW) disposal sites. An integral part of performance assessment is estimating the fluxes of radioactive gases such as radon-220 and radon-222. Data needs pointed out by mathematical models drive site characterization. They provide a logical means of performing the required flux estimations. Thorium-230 waste, consisting largely of thorium hydroxide and thorium oxides, has been approved for disposal in shallow trenches and pits at the LLW Radioactive Waste Management Site in Area 5 of the Nevada Test Site. A sophisticated gas transport model, CASCADR8 (Lindstrom et al., 1992b), was used to simulate the transport and fate of radon-222 from its source of origin, nine feet below a closure cap of native soil, through the dry alluvial earth, to its point of release into the atmosphere. CASCADR8 is an M-chain gas-phase radionuclide transport and fate model. It has been tailored to the site-specific needs of the dry desert environment of southern Nevada. It is based on the mass balance principle for each radionuclide and uses gas-phase diffusion as well as barometric pressure-induced advection as its main modes of transport. CASCADR8 uses both reversible and irreversible sorption kinetic rules as well as the usual classical Bateman (1910) M-chain decay rules for its kinetic processes. Worst case radon-222 gas-phase concentrations, as well as surface fluxes, were estimated over 40 days. The maximum flux was then used in an exposure assessment model to estimate the total annual dose equivalent received by a person residing in a standard 2500-square-foot house with 10-foot walls. Results are described.

  4. Closure Report for Corrective Action Unit 166: Storage Yards and Contaminated Materials, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    NSTec Environmental Restoration

    2009-08-01

    Corrective Action Unit (CAU) 166 is identified in the Federal Facility Agreement and Consent Order (FFACO) as 'Storage Yards and Contaminated Materials' and consists of the following seven Corrective Action Sites (CASs), located in Areas 2, 3, 5, and 18 of the Nevada Test Site: CAS 02-42-01, Condo Release Storage Yd - North; CAS 02-42-02, Condo Release Storage Yd - South; CAS 02-99-10, D-38 Storage Area; CAS 03-42-01, Conditional Release Storage Yard; CAS 05-19-02, Contaminated Soil and Drum; CAS 18-01-01, Aboveground Storage Tank; and CAS 18-99-03, Wax Piles/Oil Stain. Closure activities were conducted from March to July 2009 according to the FF ACO (1996, as amended February 2008) and the Corrective Action Plan for CAU 166 (U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office, 2007b). The corrective action alternatives included No Further Action and Clean Closure. Closure activities are summarized. CAU 166, Storage Yards and Contaminated Materials, consists of seven CASs in Areas 2, 3, 5, and 18 of the NTS. The closure alternatives included No Further Action and Clean Closure. This CR provides a summary of completed closure activities, documentation of waste disposal, and confirmation that remediation goals were met. The following site closure activities were performed at CAU 166 as documented in this CR: (1) At CAS 02-99-10, D-38 Storage Area, approximately 40 gal of lead shot were removed and are currently pending treatment and disposal as MW, and approximately 50 small pieces of DU were removed and disposed as LLW. (2) At CAS 03-42-01, Conditional Release Storage Yard, approximately 7.5 yd{sup 3} of soil impacted with lead and Am-241 were removed and disposed as LLW. As a BMP, approximately 22 ft{sup 3} of asbestos tile were removed from a portable building and disposed as ALLW, approximately 55 gal of oil were drained from accumulators and are currently pending disposal as HW, the portable building was removed and disposed as LLW, and accumulators, gas cylinders, and associated debris were removed and are currently pending treatment and disposal as MW. (3) At CAS 05-19-02, Contaminated Soil and Drum, as a BMP, an empty drum was removed and disposed as sanitary waste. (4) At CAS 18-01-01, Aboveground Storage Tank, approximately 165 gal of lead-impacted liquid were removed and are currently pending disposal as HW, and approximately 10 gal of lead shot and 6 yd{sup 3} of wax embedded with lead shot were removed and are currently pending treatment and disposal as MW. As a BMP, approximately 0.5 yd{sup 3} of wax were removed and disposed as hydrocarbon waste, approximately 55 gal of liquid were removed and disposed as sanitary waste, and two metal containers were grouted in place. (5) At CAS 18-99-03, Wax Piles/Oil Stain, no further action was required; however, as a BMP, approximately l.5 yd{sup 3} of wax were removed and disposed as hydrocarbon waste, and one metal container was grouted in place.

  5. Application for a Permit to Operate a Class III Solid Waste Disposal Site at the Nevada National Security Site Area 5 Asbestiform Low-Level Solid Waste Disposal Site

    SciTech Connect (OSTI)

    NSTec Environmental Programs

    2010-10-04

    The Nevada National Security Site (NNSS) is located approximately 105 km (65 mi) northwest of Las Vegas, Nevada. The U.S. Department of Energy National Nuclear Security Administration Nevada Site Office (NNSA/NSO) is the federal lands management authority for the NNSS and National Security Technologies, LLC (NSTec) is the Management and Operations contractor. Access on and off the NNSS is tightly controlled, restricted, and guarded on a 24-hour basis. The NNSS is posted with signs along its entire perimeter. NSTec is the operator of all solid waste disposal sites on the NNSS. The Area 5 Radioactive Waste Management Site (RWMS) is the location of the permitted facility for the Solid Waste Disposal Site (SWDS). The Area 5 RWMS is located near the eastern edge of the NNSS (Figure 1), approximately 26 km (16 mi) north of Mercury, Nevada. The Area 5 RWMS is used for the disposal of low-level waste (LLW) and mixed low-level waste. Many areas surrounding the RWMS have been used in conducting nuclear tests. The site will be used for the disposal of regulated Asbestiform Low-Level Waste (ALLW), small quantities of low-level radioactive hydrocarbon-burdened (LLHB) media and debris, LLW, LLW that contains Polychlorinated Biphenyl (PCB) Bulk Product Waste greater than 50 ppm that leaches at a rate of less than 10 micrograms of PCB per liter of water, and small quantities of LLHB demolition and construction waste (hereafter called permissible waste). Waste containing free liquids, or waste that is regulated as hazardous waste under the Resource Conservation and Recovery Act (RCRA) or state-of-generation hazardous waste regulations, will not be accepted for disposal at the site. Waste regulated under the Toxic Substances Control Act (TSCA) that will be accepted at the disposal site is regulated asbestos-containing materials (RACM) and PCB Bulk Product Waste greater than 50 ppm that leaches at a rate of less than 10 micrograms of PCB per liter of water. The term asbestiform is used throughout this document to describe RACM. The disposal site will be used as a depository of permissible waste generated both on site and off site. All generators designated by NNSA/NSO will be eligible to dispose regulated ALLW at the Asbestiform Low-Level Waste Disposal Site in accordance with the DOE/NV-325, Nevada National Security Site Waste Acceptance Criteria (NNSSWAC, current revision). Approval will be given by NNSA/NSO to generators that have successfully demonstrated through process knowledge (PK) and/or sampling and analysis that the waste is low-level, contains asbestiform material, or contains PCB Bulk Product Waste greater than 50 ppm that leaches at a rate of less than 10 micrograms of PCB per liter of water, or small quantities of LLHB demolition and construction waste and does not contain prohibited waste materials. Each waste stream will be approved through the Radioactive Waste Acceptance Program (RWAP), which ensures that the waste meets acceptance requirements outlined in the NNSSWAC.

  6. DECONTAMINATION OF ZIRCALOY SPENT FUEL CLADDING HULLS

    SciTech Connect (OSTI)

    Rudisill, T; John Mickalonis, J

    2006-09-27

    The reprocessing of commercial spent nuclear fuel (SNF) generates a Zircaloy cladding hull waste which requires disposal as a high level waste in the geologic repository. The hulls are primarily contaminated with fission products and actinides from the fuel. During fuel irradiation, these contaminants are deposited in a thin layer of zirconium oxide (ZrO{sub 2}) which forms on the cladding surface at the elevated temperatures present in a nuclear reactor. Therefore, if the hulls are treated to remove the ZrO{sub 2} layer, a majority of the contamination will be removed and the hulls could potentially meet acceptance criteria for disposal as a low level waste (LLW). Discard of the hulls as a LLW would result in significant savings due to the high costs associated with geologic disposal. To assess the feasibility of decontaminating spent fuel cladding hulls, two treatment processes developed for dissolving fuels containing zirconium (Zr) metal or alloys were evaluated. Small-scale dissolution experiments were performed using the ZIRFLEX process which employs a boiling ammonium fluoride (NH{sub 4}F)/ammonium nitrate (NH{sub 4}NO{sub 3}) solution to dissolve Zr or Zircaloy cladding and a hydrofluoric acid (HF) process developed for complete dissolution of Zr-containing fuels. The feasibility experiments were performed using Zircaloy-4 metal coupons which were electrochemically oxidized to produce a thin ZrO{sub 2} layer on the surface. Once the oxide layer was in place, the ease of removing the layer using methods based on the two processes was evaluated. The ZIRFLEX and HF dissolution processes were both successful in removing a 0.2 mm (thick) oxide layer from Zircaloy-4 coupons. Although the ZIRFLEX process was effective in removing the oxide layer, two potential shortcomings were identified. The formation of ammonium hexafluorozirconate ((NH{sub 4}){sub 2}ZrF{sub 6}) on the metal surface prior to dissolution in the bulk solution could hinder the decontamination process by obstructing the removal of contamination. The thermal decomposition of this material is also undesirable if the cladding hulls are melted for volume reduction or to produce waste forms. Handling and disposal of the corrosive off-gas stream and ZrO{sub 2}-containing dross must be addressed. The stability of Zr{sup 4+} in the NHF{sub 4}/NH{sub 4}NO{sub 3} solution is also a concern. Precipitation of ammonium zirconium fluorides upon cooling of the dissolving solution was observed in the feasibility experiments. Precipitation of the solids was attributed to the high fluoride to Zr ratios used in the experiments. The solubility of Zr{sup 4+} in NH{sub 4}F solutions decreases as the free fluoride concentration increases. The removal of the ZrO{sub 2} layer from Zircaloy-4 coupons with HF showed a strong dependence on both the concentration and temperature. Very rapid dissolution of the oxide layer and significant amounts of metal was observed in experiments using HF concentrations {ge} 2.5 M. Treatment of the coupons using HF concentrations {le} 1.0 M was very effective in removing the oxide layer. The most effective conditions resulted in dissolution rates which were less than approximately 2 mg/cm{sup 2}-min. With dissolution rates in this range, uniform removal of the oxide layer was obtained and a minimal amount of Zircaloy metal was dissolved. Future HF dissolution studies should focus on the decontamination of actual spent fuel cladding hulls to determine if the treated hulls meet criteria for disposal as a LLW.

  7. RESULTS FOR THE FOURTH QUARTER TANK 50 WAC SLURRY SAMPLE: CHEMICAL AND RADIONUCLIDE CONTAMINANT RESULTS

    SciTech Connect (OSTI)

    Reigel, M.; Bibler, N.

    2010-01-27

    The Saltstone Facility is designed and permitted to immobilize and dispose of low-level radioactive and hazardous liquid waste (salt solution) remaining from the processing of radioactive material at the Savannah River Site. Low-level waste (LLW) streams from the Effluent Treatment Project (ETP), H-Canyon, the DDA (Deliquification, Dissolution, and Adjustment) process, and the decontaminated salt solution product from the Actinide Removal Process/Modular Caustic Side Solvent Extraction (CSSX) Unit (ARP/MCU) process are stored in Tank 50 until the LLW can be transferred to the Saltstone Facility for treatment and disposal. The LLW must meet the specified waste acceptance criteria (WAC) before it is processed into saltstone. The specific chemical and radionuclide contaminants and their respective WAC limits are listed in the current Saltstone WAC. SRS Liquid Waste Operations (LWO) requested that Savannah River National Laboratory (SRNL) perform quarterly analysis on saltstone samples. The concentrations of chemical and radionuclide contaminants are measured to ensure the saltstone produced during each quarter is in compliance with the current WAC. This report documents the concentrations of chemical and radionuclide contaminants for the 2009 Fourth Quarter samples collected from Tank 50 on October 2, 2009 and discusses those results in further detail than the previously issued results report. This report details the chemical and radionuclide contaminant results for the characterization of the 2009 Fourth Quarter sampling of Tank 50 for the Saltstone Waste Acceptance Criteria (WAC). Information from this characterization will be used by Liquid Waste Operations (LWO) to support the transfer of low-level aqueous waste from Tank 50 to the Salt Feed Tank in the Saltstone Facility in Z-Area, where the waste will be immobilized. This information is also used to update the Tank 50 Waste Characterization System. The following conclusions are drawn from the analytical results provided in this report: (1) The concentrations of the reported chemical and radioactive contaminants were less than their respective WAC targets or limits unless noted in this section. (2) The reported detection limit for Isopar L is greater than the limit from Table 3 of the WAC. (3) The reported detection limits for {sup 59}Ni and {sup 94}Nb are above the requested limits from Reference 4. However, they are each below the limits established in Reference 3. (4) The reported detection limit for isopropanol is greater than the requested limit from Table 4 of the WAC. (5) The reported detection limits for {sup 247}Cm and {sup 249}Cf are above the requested limits from Reference 4. However, they are below the limits established in Reference 3. (6) Isopar L and Norpar 13 have limited solubility in aqueous solutions making it difficult to obtain consistent and reliable sub-samples. The values reported in this memo are the concentrations in the sub-sample as detected by the GC/MS; however, the results may not accurately represent the concentrations of the analytes in Tank 50.

  8. Transuranic (TRU) Waste Phase I Retrieval Plan

    SciTech Connect (OSTI)

    MCDONALD, K.M.

    2000-09-28

    From 1970 to 1987, TRU and suspect TRU wastes at Hanford were placed in the SWBG. At the time of placement in the SWBG these wastes were not regulated under existing Resource Conservation and Recovery Act (RCRA) regulations, since they were generated and disposed of prior to the effective date of RCRA at the Hanford Site (1987). From the standpoint of DOE Order 5820.2A1, the TRU wastes are considered retrievably stored, and current plans are to retrieve these wastes for shipment to WIPP for disposal. This plan provides a strategy for the Phase I retrieval that meets the intent of TPA milestone M-91 and Project W-113, and incorporates the lessons learned during TRU retrieval campaigns at Hanford, LANL, and SRS. As in the original Project W-113 plans, the current plan calls for examination of approximately 10,000 suspect-TRU drums located in the 218-W-4C burial ground followed by the retrieval of those drums verified to contain TRU waste. Unlike the older plan, however, this plan proposes an open-air retrieval scenario similar to those used for TRU drum retrieval at LANL and SRS. Phase I retrieval consists of the activities associated with the assessment of approximately 10,000 55-gallon drums of suspect TRU-waste in burial ground 218-W-4C and the retrieval of those drums verified to contain TRU waste. Four of the trenches in 218-W-4C (Trenches 1, 4, 20, and 29) are prime candidates for Phase I retrieval because they contain large numbers of suspect TRU drums, stacked from 2 to 5 drums high, on an asphalt pad. In fact, three of the trenches (Trenches 1 , 20, and 29) contain waste that has not been covered with soil, and about 1500 drums can be retrieved without excavation. The other three trenches in 218-W-4C (Trenches 7, 19, and 24) are not candidates for Phase I retrieval because they contain significant numbers of boxes. Drums will be retrieved from the four candidate trenches, checked for structural integrity, overpacked, if necessary, and assayed at the burial ground. A mobile assay system will be used to determine if the drum is LLW (Le., contains <100 nCi/g). LLW will remain disposed of in the 218-W-4C Burial Ground. TRU waste will be retrieved and staged in the burial ground until it can be shipped to the CWC. The TRU drums will be stored at the CWC until they can be moved to WRAP. The WRAP facility will prepare the waste for shipment to WIPP for final disposal. For planning purposes, approximately 50% of the 10,000 drums have been estimated to contain LLW.

  9. Evaluation of Low-Level Waste Disposal Receipt Data for Los Alamos National Laboratory Technical Area 54, Area G Disposal Facility - Fiscal Year 2011

    SciTech Connect (OSTI)

    French, Sean B. [Los Alamos National Laboratory; Shuman, Robert [WPS: WASTE PROJECTS AND SERVICES

    2012-04-17

    The Los Alamos National Laboratory (LANL or the Laboratory) generates radioactive waste as a result of various activities. Operational or institutional waste is generated from a wide variety of research and development activities including nuclear weapons development, energy production, and medical research. Environmental restoration (ER), and decontamination and decommissioning (D and D) waste is generated as contaminated sites and facilities at LANL undergo cleanup or remediation. The majority of this waste is low-level radioactive waste (LLW) and is disposed of at the Technical Area 54 (TA-54), Area G disposal facility. U.S. Department of Energy (DOE) Order 435.1 (DOE, 2001) requires that radioactive waste be managed in a manner that protects public health and safety, and the environment. To comply with this order, DOE field sites must prepare and maintain site-specific radiological performance assessments for LLW disposal facilities that accept waste after September 26, 1988. Furthermore, sites are required to conduct composite analyses that account for the cumulative impacts of all waste that has been (or will be) disposed of at the facilities and other sources of radioactive material that may interact with the facilities. Revision 4 of the Area G performance assessment and composite analysis was issued in 2008 (LANL, 2008). These analyses estimate rates of radionuclide release from the waste disposed of at the facility, simulate the movement of radionuclides through the environment, and project potential radiation doses to humans for several on-site and off-site exposure scenarios. The assessments are based on existing site and disposal facility data and on assumptions about future rates and methods of waste disposal. The accuracy of the performance assessment and composite analysis depends upon the validity of the data used and assumptions made in conducting the analyses. If changes in these data and assumptions are significant, they may invalidate or call into question certain aspects of the analyses. For example, if the volumes and activities of waste disposed of during the remainder of the disposal facility's lifetime differ significantly from those projected, the doses projected by the analyses may no longer apply. DOE field sites are required to implement a performance assessment and composite analysis maintenance program. The purpose of this program is to ensure the continued applicability of the analyses through incremental improvement of the level of understanding of the disposal site and facility. Site personnel are required to conduct field and experimental work to reduce the uncertainty in the data and models used in the assessments. Furthermore, they are required to conduct periodic reviews of waste receipts, comparing them to projected waste disposal rates. The radiological inventory for Area G was updated in conjunction with Revision 4 of the performance assessment and composite analysis (Shuman, 2008). That effort used disposal records and other sources of information to estimate the quantities of radioactive waste that have been disposed of at Area G from 1959, the year the facility started receiving waste on a routine basis, through 2007. It also estimated the quantities of LLW that will require disposal from 2008 through 2044, the year in which it is assumed that disposal operations at Area G will cease. This report documents the fourth review of Area G disposal receipts since the inventory was updated and examines information for waste placed in the ground during fiscal years (FY) 2008 through 2011. The primary objective of the disposal receipt review is to ensure that the future waste inventory projections developed for the performance assessment and composite analysis are consistent with the actual types and quantities of waste being disposed of at Area G. Toward this end, the disposal data that are the subject of this review are used to update the future waste inventory projections for the disposal facility. These projections are compared to the future inventory projections that were develope

  10. Closure Report for Corrective Action Unit 139: Waste Disposal Sites, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    NSTec Environmental Restoration

    2009-07-31

    Corrective Action Unit (CAU) 139 is identified in the Federal Facility Agreement and Consent Order (FFACO) as 'Waste Disposal Sites' and consists of the following seven Corrective Action Sites (CASs), located in Areas 3, 4, 6, and 9 of the Nevada Test Site: CAS 03-35-01, Burn Pit; CAS 04-08-02, Waste Disposal Site; CAS 04-99-01, Contaminated Surface Debris; CAS 06-19-02, Waste Disposal Site/Burn Pit; CAS 06-19-03, Waste Disposal Trenches; CAS 09-23-01, Area 9 Gravel Gertie; and CAS 09-34-01, Underground Detection Station. Closure activities were conducted from December 2008 to April 2009 according to the FFACO (1996, as amended February 2008) and the Corrective Action Plan for CAU 139 (U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office, 2007b). The corrective action alternatives included No Further Action, Clean Closure, and Closure in Place with Administrative Controls. Closure activities are summarized. CAU 139, 'Waste Disposal Sites,' consists of seven CASs in Areas 3, 4, 6, and 9 of the NTS. The closure alternatives included No Further Action, Clean Closure, and Closure in Place with Administrative Controls. This CR provides a summary of completed closure activities, documentation of waste disposal, and confirmation that remediation goals were met. The following site closure activities were performed at CAU 139 as documented in this CR: (1) At CAS 03-35-01, Burn Pit, soil and debris were removed and disposed as LLW, and debris was removed and disposed as sanitary waste. (2) At CAS 04-08-02, Waste Disposal Site, an administrative UR was implemented. No postings or post-closure monitoring are required. (3) At CAS 04-99-01, Contaminated Surface Debris, soil and debris were removed and disposed as LLW, and debris was removed and disposed as sanitary waste. (4) At CAS 06-19-02, Waste Disposal Site/Burn Pit, no work was performed. (5) At CAS 06-19-03, Waste Disposal Trenches, a native soil cover was installed, and a UR was implemented. (6) At CAS 09-23-01, Area 9 Gravel Gertie, a UR was implemented. (7) At CAS 09-34-01, Underground Detection Station, no work was performed.

  11. ANALYSIS OF THE SALT FEED TANK CORE SAMPLE

    SciTech Connect (OSTI)

    Reigel, M.; Cheng, W.

    2012-01-26

    The Saltstone Production Facility (SPF) immobilizes and disposes of low-level radioactive and hazardous liquid waste (salt solution) remaining from the processing of radioactive material at the Savannah River Site (SRS). Low-level waste (LLW) streams from processes at SRS are stored in Tank 50 until the LLW can be transferred to the SPF for treatment and disposal. The Salt Feed Tank (SFT) at the Saltstone Production Facility (SPF) holds approximately 6500 gallons of low level waste from Tank 50 as well as drain water returned from the Saltstone Disposal Facility (SDF) vaults. Over the past several years, Saltstone Engineering has noted the accumulation of solids in the SFT. The solids are causing issues with pump performance, agitator performance, density/level monitoring, as well as taking up volume in the tank. The tank has been sounded at the same location multiple times to determine the level of the solids. The readings have been 12, 25 and 15 inches. The SFT is 8.5 feet high and 12 feet in diameter, therefore the solids account for approximately 10 % of the tank volume. Saltstone Engineering has unsuccessfully attempted to obtain scrape samples of the solids for analysis. As a result, Savannah River National Laboratory (SRNL) was tasked with developing a soft core sampler to obtain a sample of the solids and to analyze the core sample to aid in determining a path forward for removing the solids from the SFT. The source of the material in the SFT is the drain water return system where excess liquid from the Saltstone disposal vaults is pumped back to the SFT for reprocessing. It has been shown that fresh grout from the vault enter the drain water system piping. Once these grout solids return to the SFT, they settle in the tank, set up, and can't be reprocessed, causing buildup in the tank over time. The composition of the material indicates that it is potentially toxic for chromium and mercury and the primary radionuclide is cesium-137. Qualitative measurements show that the material is not cohesive and will break apart with some force.

  12. Neutron Screening Measurements of 110 gallon drums at T Plant

    SciTech Connect (OSTI)

    Mozhayev, Andrey V.; Hilliard, James R.; Berg, Randal K.

    2011-01-14

    The Pacific Northwest National Laboratory (PNNL) Nondestructive Assay (NDA) Service Center was contracted to develop and demonstrate a simple and inexpensive method of assaying 110 gallon drums at the Hanford Sites T-Plant. The drums contained pucks of crushed old drums used for storage of transuranic (TRU) waste. The drums were to be assayed to determine if they meet the criteria for TRU or Low Level Waste (LLW). Because of the dense matrix (crushed steel drums) gamma measurement techniques were excluded and a mobile, configurable neutron system, consisting of four sequentially connected slab detectors was chosen to be used for this application. An optimum measurement configuration was determined through multiple test measurements with californium source. Based on these measurements the initial calibration of the system was performed applying the isotopic composition for aged weapon-grade plutonium. A series of background and blank puck drum measurements allowed estimating detection limits for both total (singles) and coincidence (doubles) counting techniques. It was found that even conservative estimates for minimum detection concentration using singles count rate were lower than the essential threshold of 100 nCi/g. Whereas the detection limit of coincidence counting appeared to be about as twice as high of the threshold. A series of measurements intended to verify the technique and revise the initial calibration obtained were performed at the Waste Receiving and Processing (WRAP) facility with plutonium standards. Standards with a total mass of 0.3 g of plutonium (which is estimated to be equivalent of 100 nCi/g for net waste weight of 300 kg) loaded in the test puck drum were clearly detected. The following measurements of higher plutonium loadings verified the calibration factors obtained in the initial exercise. The revised and established calibration factors were also confirmed within established uncertainties by additional measurements of plutonium standards in various locations in the test drum. Due to necessity to dispense the blank test drum an alternative method of baseline determination was established during field measurements. Count rates of ambient background were corrected by the differences between observed background and blank test drum count rates which were previously determined over a series of measurements. Only 31 drums out of 352 counted during the intensive measurement campaign at T-Plant were determined to be Suspect TRU. 25 of these drums were re-measured at the WRAP facility using the SuperHENC. Of the 25 drums measured, 21 were confirmed to be TRU and the remaining four LLW.

  13. Repackaging Rocky Flats Legacy Transuranic Waste

    SciTech Connect (OSTI)

    McTaggart, Jerri Lynne

    2008-01-15

    Repackaging legacy Transuranic (TRU), Transuranic Mixed (TRM), Low Level Waste (LLW), and Low Level Mixed (LLM) waste requires good characterization skills and the ability to adapt to less than ideal conditions. Repackaging legacy waste in a facility that is not undergoing Decontamination and Decommission (D and D) is optimum. However, repackaging any waste in a D and D facility, under cold and dark conditions, can be difficult. Cold and dark conditions are when the heating and air conditioning are no longer in service and the lighting consists of strands of lights hung throughout each of the rooms. Working under these conditions adds an additional level of stress and danger that must be addressed. The use of glovebags was very useful at Rocky Flats during the D and D of many buildings. Glovebags can be adapted for many different types of wastes and unusual conditions. Repackaging of legacy TRU waste, in a D and D facility, can be accomplished safely and cost effectively with the use of glovebags. In conclusion: the use of glovebags to repackage legacy TRU, TRM, LLW, or LLM waste was done safely and cost effectively at Rocky Flats. The cost of using glovebags was minimal. Glovebags are easily adaptable to whatever the waste configuration is. The use of glovebags, for repackaging of Legacy waste, allows D and D efforts to stay on schedule and on task. Without the use of glovebags, additional gloveboxes would have been required at Rocky Flats. Larger items, such as the HEPA filters, would have required the construction of a new large item repackaging glovebox. Repackaging in glovebags allows the freedom to either locate the glovebag by the waste or locate the glovebag in a place that least impacts D and D efforts. The use of glovebags allowed numerous configurations of waste to be repackaged without the use of gloveboxes. During the D and D of the Rocky Flats facility, which was in a cold and dark stage, D and D work was not impacted by the repackaging activity. Glovebags work well in facilities that are in the process of D and D or still in full operations because glovebags are very safe and cost effective.

  14. Assessing Potential Exposure from Truck Transport of Low-level Radioactive Waste to the Nevada Test Site

    SciTech Connect (OSTI)

    J. Miller; D. Shafer; K. Gray; B. Church; S.Campbell; B. Holz

    2005-08-15

    This study has shown that, based upon measurements from industry standard radiation detection instruments, such as the RS model RSS-131 PICs in a controlled configuration, a person may be exposed to gamma radiation above background when in close proximity to some LLW trucks. However, in approximately half (47.7 percent) the population of trucks measured in this study, a person would receive no exposure above background at a distance of 1.0 m (3.3 ft) away from a LLW truck. An additional 206 trucks had net exposures greater than zero, but equal to or less than 1 {micro}R/h. Finally, nearly 80 percent of the population of trucks (802 of 1,012) had net exposures less than or equal to 10 {micro}R/h. Although there are no shipping or exposure standards at 1.0 m (3.3 ft) distance, one relevant point of comparison is the DOT shipping standard of 10 mrem/h at 2.0 m (6.6 ft) distance. Assuming a one-to-one correspondence between Roentgens and Rems, then 903 trucks (89.2 percent of the trucks measured) were no greater than one percent of the DOT standard at 1.0 m (3.3 ft). Had the distance at which the trucks been measured increased to 2.0 m (6.6 ft), the net exposure would be even less because of the increase in distance between the truck and the receptor. However, based on the empirical data from this study, the rate of decrease may be slower than for either a point or line source as was done for previous studies (Gertz, 2001; Davis et al., 2001). The highest net exposure value at 1.0 m (3.3 ft) distance, 11.9 mR/h, came from the only truck with a value greater than 10 mR/h at 1.0 m (3.3 ft) distance.

  15. Unvented Drum Handling Plan

    SciTech Connect (OSTI)

    MCDONALD, K.M.

    2000-08-01

    This drum-handling plan proposes a method to deal with unvented transuranic drums encountered during retrieval of drums. Finding unvented drums during retrieval activities was expected, as identified in the Transuranic (TRU) Phase I Retrieval Plan (HNF-4781). However, significant numbers of unvented drums were not expected until excavation of buried drums began. This plan represents accelerated planning for management of unvented drums. A plan is proposed that manages unvented drums differently based on three categories. The first category of drums is any that visually appear to be pressurized. These will be vented immediately, using either the Hanford Fire Department Hazardous Materials (Haz. Mat.) team, if such are encountered before the facilities' capabilities are established, or using internal capabilities, once established. To date, no drums have been retrieved that showed signs of pressurization. The second category consists of drums that contain a minimal amount of Pu isotopes. This minimal amount is typically less than 1 gram of Pu, but may be waste-stream dependent. Drums in this category are assayed to determine if they are low-level waste (LLW). LLW drums are typically disposed of without venting. Any unvented drums that assay as TRU will be staged for a future venting campaign, using appropriate safety precautions in their handling. The third category of drums is those for which records show larger amounts of Pu isotopes (typically greater than or equal to 1 gram of Pu). These are assumed to be TRU and are not assayed at this point, but are staged for a future venting campaign. Any of these drums that do not have a visible venting device will be staged awaiting venting, and will be managed under appropriate controls, including covering the drums to protect from direct solar exposure, minimizing of container movement, and placement of a barrier to restrict vehicle access. There are a number of equipment options available to perform the venting. The preferred option is to use equipment provided by a commercial vendor during the first few years of retrieval and venting. This is based on a number of reasons. First, retrieval funding is uncertain. Using a commercial vendor will allow DOE-RL to avoid the investment and maintenance costs if retrieval is not funded. Second, when funding can be identified, retrieval will likely be performed with minimal initial throughput and intermittent operations. Again, costs can be saved by using contracted vendor services only as needed, rather than supporting Hanford equipment full time. When full-scale retrieval begins and the number of drums requiring venting increases significantly, then use of the Hanford container venting system (CVS) should be considered.

  16. Regulations, Policies and Strategies for LLRW Management in Bangladesh - 12368

    SciTech Connect (OSTI)

    Mollah, A.S.

    2012-07-01

    Low level radioactive waste (LLW) is generated from various nuclear applications in Bangladesh. The major sources of radioactive waste in the country are at present: (a) the 3 MW TRIGA Mark-II research reactor; (b) the radioisotope production facility; (c) the medical, industrial and research facilities that use radionuclides; and (d) the industrial facility for processing monazite sands. Radioactive waste needs to be safely managed because it is potentially hazardous to human health and the environment. According to Nuclear Safety and Radiation Control Act-93, the Bangladesh Atomic Energy Commission (BAEC) is the governmental body responsible for the receipt and final disposal of radioactive wastes in the whole country. Waste management policy has become an important environmental, social, and economical issue for LLW in Bangladesh. Policy and strategies will serve as a basic guide for radioactive waste management in Bangladesh. The waste generator is responsible for on-site collection, conditioning and temporary storage of the waste arising from his practice. The Central Waste Processing and Storage Unit (CWPSU) of BAEC is the designated national facility with the requisite facility for the treatment, conditioning and storage of radioactive waste until a final disposal facility is established and becomes operational. The Regulatory Authority is responsible for the enforcement of compliance with provisions of the waste management regulation and other relevant requirements by the waste generator and the CWPSU. The objective of this paper is to present, in a concise form, basic information about the radioactive waste management infrastructure, regulations, policies and strategies including the total inventory of low level radioactive waste in the country. For improvement and strengthening in terms of operational capability, safety and security of RW including spent radioactive sources and overall security of the facility (CWPSF), the facility is expected to serve waste management need in the country and, in the course of time, the facility may be turned into a regional level training centre. It is essential for safe conduction and culture of research and application in nuclear science and technology maintaining the relevant safety of man and environment and future generations to come. (authors)

  17. Decommissioning of the secondary containment of the steam generating heavy water reactor at UKAEA Winfrith

    SciTech Connect (OSTI)

    Miller, K.D.; Cornell, R.M.; Parkinson, S.J.; McIntyre, K.; Staples, A.

    2007-07-01

    The Winfrith SGHWR was a prototype nuclear power plant operated for 23 years by the United Kingdom Atomic Energy Authority (UKAEA) until 1990 when it was shut down permanently. The current Stage 1 decommissioning contract is part of a multi-stage strategy. It involves the removal of all the ancillary plant and equipment in the secondary containment and non-containment areas ahead of a series of contracts for the decommissioning of the primary containment, the reactor core and demolition of the building and ail remaining facilities. As an outcome of a competitive tending process, the Stage 1 decommissioning contract was awarded to NUKEM with operations commencing in April 2005. The decommissioning processes involved with these plant items will be described with some emphasis of the establishment of multiple work-fronts for the production, recovery, treatment and disposal of mainly tritium-contaminated waste arising from its contact with the direct cycle reactor coolant. The means of size reduction of a variety of large, heavy and complex items of plant made from a range of materials will also be described with some emphasis on the control of fumes during hot cutting operations and establishing effective containments within a larger secondary containment structure. Disposal of these wastes in a timely and cost-effective manner is a major challenge facing the decommissioning team and has required the development of a highly efficient means of packing the resultant materials into mainly one-third height IS0 containers for disposal as LLW. Details of the quantities of LLW and exempt wastes handled during this process will be given with a commentary about the difficulty in segregating these two waste streams efficiently. The paper sets out to demonstrate the considerable progress that has been made with these challenging decommissioning operations at the SGHWR plant and to highlight some of the techniques and processes that have contributed to the overall success of the process. The overall management and control of safety during ail aspects of this challenging contract are major features of the paper, greatly assisted by the adoption from the outset of a non-adversarial team working approach between client and contractor. This has greatly assisted in developing safe and cost-effective solutions to a number of problems that have arisen during the programme, demonstrating the worth of adopting this co-operative approach for mutual benefit. (authors)

  18. Evolution Of USDOE Performance Assessments Over 20 Years

    SciTech Connect (OSTI)

    Seitz, Roger R. [Savannah River Site (SRS), Aiken, SC (United States); Suttora, Linda C. [U.S. Department of Energy, Office of Site Restoration, Germantown, MD (United States)

    2013-02-26

    Performance assessments (PAs) have been used for many years for the analysis of post-closure hazards associated with a radioactive waste disposal facility and to provide a reasonable expectation of the ability of the site and facility design to meet objectives for the protection of members of the public and the environment. The use of PA to support decision-making for LLW disposal facilities has been mandated in United States Department of Energy (USDOE) directives governing radioactive waste management since 1988 (currently DOE Order 435.1, Radioactive Waste Management). Prior to that time, PAs were also used in a less formal role. Over the past 20+ years, the USDOE approach to conduct, review and apply PAs has evolved into an efficient, rigorous and mature process that includes specific requirements for continuous improvement and independent reviews. The PA process has evolved through refinement of a graded and iterative approach designed to help focus efforts on those aspects of the problem expected to have the greatest influence on the decision being made. Many of the evolutionary changes to the PA process are linked to the refinement of the PA maintenance concept that has proven to be an important element of USDOE PA requirements in the context of supporting decision-making for safe disposal of LLW. The PA maintenance concept represents the evolution of the graded and iterative philosophy and has helped to drive the evolution of PAs from a deterministic compliance calculation into a systematic approach that helps to focus on critical aspects of the disposal system in a manner designed to provide a more informed basis for decision-making throughout the life of a disposal facility (e.g., monitoring, research and testing, waste acceptance criteria, design improvements, data collection, model refinements). A significant evolution in PA modeling has been associated with improved use of uncertainty and sensitivity analysis techniques to support efficient implementation of the graded and iterative approach. Rather than attempt to exactly predict the migration of radionuclides in a disposal unit, the best PAs have evolved into tools that provide a range of results to guide decision-makers in planning the most efficient, cost effective, and safe disposal of radionuclides.

  19. Expert System for Building TRU Waste Payloads - 13554

    SciTech Connect (OSTI)

    Bruemmer, Heather; Slater, Bryant

    2013-07-01

    The process for grouping TRU waste drums into payloads for shipment to the Waste Isolation Pilot Plant (WIPP) for disposal is a very complex process. Transportation and regulatory requirements must be met, along with striving for the goals of shipment efficiency: maximize the number of waste drums in a shipment and minimize the use of empty drums which take up precious underground storage space. The restrictions on payloads range from weight restrictions, to limitations on flammable gas in the headspace, to minimum TRU alpha activity concentration requirements. The Overpack and Payload Assistant Tool (OPAT) has been developed as a mixed-initiative intelligent system within the WIPP Waste Data System (WDS) to guide the construction of multiple acceptable payloads. OPAT saves the user time while at the same time maximizes the efficiency of shipments for the given drum population. The tool provides the user with the flexibility to tune critical factors that guide OPAT's operation based on real-time feedback concerning the results of the execution. This feedback complements the user's external knowledge of the drum population (such as location of drums, known challenges, internal shipment goals). This work demonstrates how software can be utilized to complement the unique domain knowledge of the users. The mixed-initiative approach combines the insight and intuition of the human expert with the proficiency of automated computational algorithms. The result is the ability to thoroughly and efficiently explore the search space of possible solutions and derive the best waste management decision. (authors)

  20. Death by Dynamics: Planetoid-Induced Explosions on White Dwarfs

    E-Print Network [OSTI]

    Di Stefano, Rosanne; Guillochon, James; Steiner, James F

    2015-01-01

    At intervals as short as ten thousand years, each white dwarf (WD) passes within a solar radius of a planetoid, i.e., a comet, asteroid, or planet. Gravitational tidal forces tear the planetoid apart; its metal-rich debris falls onto the WD, enriching the atmosphere. A third of WDs exhibit atmospheric "pollution". For roughly every hundred planetoid disruptions, a planetoid collides with a WD. We simulate a small number of collisions, in which "death-by-dynamics" refers to the fate of the planetoid. We also compute the energies and likely durations of a broad sample of collision events, and identify detection strategies at optical and X-ray wavelengths. Collisions with the most massive planetoids can be detected in external galaxies. Some may trigger nuclear burning. If one in $\\sim 10^7-10^8$ of WD-planetoid collisions creates the conditions needed for a Type Ia supernova (SN~Ia), "death-by-dynamics" would also refer to the fate of the WD, and could provide a novel channel for the production of SN~Ia. We con...

  1. Illuminating Massive Black Holes With White Dwarfs: Orbital Dynamics and High Energy Transients from Tidal Interactions

    E-Print Network [OSTI]

    MacLeod, Morgan; Ramirez-Ruiz, Enrico; Guillochon, James; Samsing, Johan

    2014-01-01

    White dwarfs (WDs) can be tidally disrupted only by massive black holes (MBHs) with masses less than approximately $10^5 M_\\odot$. These tidal interactions feed material to the MBH well above its Eddington limit, with the potential to launch a relativistic jet. The corresponding beamed emission is a promising signpost to an otherwise quiescent MBH of relatively low mass. We show that the mass transfer history, and thus the lightcurve, are quite different when the disruptive orbit is parabolic, eccentric, or circular. The mass lost each orbit exponentiates in the eccentric-orbit case leading to the destruction of the WD after several tens of orbits and making it difficult to produce a Swift J1644+57-like lightcurve via this channel. We then examine the stellar dynamics of clusters surrounding these MBHs to show that single-passage WD disruptions are substantially more common than repeating encounters in eccentric orbits. The $10^{49}$ erg s$^{-1}$ peak luminosity of these events makes them visible to cosmologi...

  2. UV-Bright Stellar Populations and Their Evolutionary Implications in the Collapsed-Core Cluster M15

    E-Print Network [OSTI]

    Haurberg, Nathalie C; Cohn, Haldan N; Lugger, Phyllis M; Anderson, Jay; Cool, Adrienne M; Serenelli, Aldo; 10.1088/0004-637X/722/1/158

    2010-01-01

    We performed deep photometry of the central region of Galactic globular cluster M15 from archival Hubble Space Telescope data taken on the High Resolution Channel and Solar Blind Channel of the Advanced Camera for Surveys. Our data set consists of images in far-UV (FUV$_{140}$; F140LP), near-UV (NUV$_{220}$; F220W), and blue (B$_{435}$; F435W) filters. The addition of an optical filter complements previous UV work on M15 by providing an additional constraint on the UV-bright stellar populations. Using color-magnitude diagrams (CMDs) we identified several populations that arise from non-canonical evolution including candidate blue stragglers, extreme horizontal branch stars, blue hook stars (BHks), cataclysmic variables (CVs), and helium-core white dwarfs (He WDs). Due to preliminary identification of several He WD and BHk candidates, we add M15 as a cluster containing a He WD sequence and suggest it be included among clusters with a BHk population. We also investigated a subset of CV candidates that appear in...

  3. Hydrodynamical evolution of merging carbon-oxygen white dwarfs: their pre-supernova structure and observational counterparts

    E-Print Network [OSTI]

    Tanikawa, Ataru; Sato, Yushi; Nomoto, Ken'ichi; Maeda, Keiichi; Hachisu, Izumi

    2015-01-01

    We perform smoothed particle hydrodynamics (SPH) simulations for merging binary carbon-oxygen (CO) white dwarfs (WDs) with masses of $1.1$ and $1.0$ $M_\\odot$, until the merger remnant reaches a dynamically steady state. Using these results, we assess whether the binary could induce a thermonuclear explosion, and whether the explosion could be observed as a type Ia supernova (SN Ia). We investigate three explosion mechanisms: a helium-ignition following the dynamical merger (`helium-ignited violent merger model'), a carbon-ignition (`carbon-ignited violent merger model'), and an explosion following the formation of the Chandrasekhar mass WD (`Chandrasekhar mass model'). An explosion of the helium-ignited violent merger model is possible, while we predict that the resulting SN ejecta are highly asymmetric since its companion star is fully intact at the time of the explosion. The carbon-ignited violent merger model can also lead to an explosion. However, the envelope of the exploding WD spreads out to $\\sim 0.1...

  4. Thermonuclear detonations ensuing white dwarf mergers

    E-Print Network [OSTI]

    Dan, Marius; Brggen, Marcus; Ramirez-Ruiz, Enrico; Rosswog, Stephan

    2015-01-01

    The merger of two white dwarfs (WDs) has for many years not been considered as the favoured model for the progenitor system of type Ia supernovae (SNe Ia). But recent years have seen a change of opinion as a number of studies, both observational and theoretical, have concluded that they should contribute significantly to the observed type Ia supernova rate. In this paper, we study the ignition and propagation of detonation through post-merger remnants and we follow the resulting nucleosynthesis up to the point where a homologous expansion is reached. In our study we cover the entire range of WD masses and compositions. For the emergence of a detonation we study several setups, guided by both merger remnants from our own simulations and by results taken from the literature. We carefully compare the nucleosynthetic yields of successful explosions with SN Ia observations. Only three of our models are consistent with all the imposed constraints and potentially lead to a standard type Ia event. The first one, a $0...

  5. Optical Thermonuclear Transients From Tidal Compression of White Dwarfs as Tracers of the Low End of the Massive Black Hole Mass Function

    E-Print Network [OSTI]

    MacLeod, Morgan; Ramirez-Ruiz, Enrico; Kasen, Daniel; Rosswog, Stephan

    2015-01-01

    In this paper, we model the observable signatures of tidal disruptions of white dwarf (WD) stars by massive black holes (MBHs) of moderate mass, $\\approx 10^3 - 10^5 M_\\odot$. When the WD passes deep enough within the MBH's tidal field, these signatures include thermonuclear transients from burning during maximum compression. We combine a hydrodynamic simulation that includes nuclear burning of the disruption of a $0.6 M_\\odot$ C/O WD with a Monte Carlo radiative transfer calculation to synthesize the properties of a representative transient. The transient's emission emerges in the optical, with lightcurves and spectra reminiscent of type I SNe. The properties are strongly viewing-angle dependent, and key spectral signatures are $\\approx 10,000$ km s$^{-1}$ Doppler shifts due to the orbital motion of the unbound ejecta. Disruptions of He WDs likely produce large quantities of intermediate-mass elements, offering a possible production mechanism for Ca-rich transients. Accompanying multiwavelength transients ar...

  6. Supernovae from direct collisions of white dwarfs and the role of helium shell ignition

    E-Print Network [OSTI]

    Papish, Oded

    2015-01-01

    Models for supernovae (SNe) arising from thermonuclear explosions of white dwarfs (WDs) have been extensively studied over the last few decades, mostly focusing on the single degenerate (accretion of material of a WD) and double degenerate (WD-WD merger) scenarios. In recent years it was suggested that WD-WD direct collisions provide an additional channel for such explosions. Here we extend the studies of such explosions, and explore the role of Helium-shells in affecting the thermonuclear explosions. We study both the impact of low-mass helium ($\\sim0.01$ M$_{\\odot})$ shells, as well as high mass shells ($\\ge0.1$ M$_{\\odot}$). We find that detonation of the massive helium layers precede the detonation of the WD Carbon-Oxygen (CO) bulk during the collision and can change the explosive evolution and outcomes for the cases of high mass He-shells. In particular, the He-shell detonation propagates on the WD surface and inefficiently burns material prior to the CO detonation that later follows in the central parts...

  7. Thermonuclear Supernovae: Probing Magnetic Fields by Late-Time IR Line Profiles

    E-Print Network [OSTI]

    Penney, R

    2014-01-01

    We study the imprint of magnetic fields B on late-time IR line profiles and light curves of Type Ia Supernovae. As a benchmark, we use the explosion of a Chandrasekhar mass M_{Ch White Dwarf (WD) and, specifically, a delayed detonation model. We assume WDs with initial magnetic surface fields between 1 and 1E9G. We discuss large-scale dipole and small-scale magnetic fields. We find that the [Fe II] line at 1.644 mu can be used to analyze the overall chemical and density structure of the exploding WD up to day 200 without considering B. Subsequently, positron transport and magnetic field effects become important. By day 500, the profile becomes sensitive to the morphology of B and directional dependent for dipole fields. Small or no directional dependence of the spectra is found for small-scale B. After about 200 days, persistent broad-line, flat-topped or stumpy profiles require high density burning which is the signature of a WD close to M_Ch. Narrow peaked profiles are a signature of chemical mixing or sub-...

  8. Low-Metallicity Inhibition of Type Ia Supernovae and Galactic and Cosmic Chemical Evolution

    E-Print Network [OSTI]

    Chiaki Kobayashi; Takuji Tsujimoto; Ken'ich Nomoto; Izumi Hachisu; Mariko Kato

    1998-06-25

    We introduce a metallicity dependence of Type Ia supernova (SN Ia) rate into the Galactic and cosmic chemical evolution models. In our SN Ia progenitor scenario, the accreting white dwarf (WD) blows a strong wind to reach the Chandrasekhar mass limit. If the iron abundance of the progenitors is as low as [Fe/H] 1-2, SNe Ia can be found only in the environments where the timescale of metal enrichment is sufficiently short as in starburst galaxies and ellipticals. The low-metallicity inhibition of SNe Ia can shed new light on the following issues: 1) The limited metallicity range of the SN Ia progenitors would imply that ``evolution effects'' are relatively small for the use of high redshift SNe Ia to determine the cosmological parameters. 2) WDs of halo populations are poor producers of SNe Ia, so that the WD contribution to the halo mass is not constrained from the iron abundance in the halo. 3) The abundance patterns of globular clusters and field stars in the Galactic halo lack of SN Ia signatures in spite of their age difference of several Gyrs, which can be explained by the low-metallicity inhibition of SNe Ia. 4) It could also explain why the SN Ia contamination is not seen in the damped Ly\\alpha systems for over a wide range of redshift.

  9. Radionuclide Retention in Concrete Wasteforms

    SciTech Connect (OSTI)

    Wellman, Dawn M.; Jansik, Danielle P.; Golovich, Elizabeth C.; Cordova, Elsa A.

    2012-09-24

    Assessing long-term performance of Category 3 waste cement grouts for radionuclide encasement requires knowledge of the radionuclide-cement interactions and mechanisms of retention (i.e., sorption or precipitation); the mechanism of contaminant release; the significance of contaminant release pathways; how wasteform performance is affected by the full range of environmental conditions within the disposal facility; the process of wasteform aging under conditions that are representative of processes occurring in response to changing environmental conditions within the disposal facility; the effect of wasteform aging on chemical, physical, and radiological properties; and the associated impact on contaminant release. This knowledge will enable accurate prediction of radionuclide fate when the wasteforms come in contact with groundwater. Data collected throughout the course of this work will be used to quantify the efficacy of concrete wasteforms, similar to those used in the disposal of LLW and MLLW, for the immobilization of key radionuclides (i.e., uranium, technetium, and iodine). Data collected will also be used to quantify the physical and chemical properties of the concrete affecting radionuclide retention.

  10. Identification of technical problems encountered in the shallow land burial of low-level radioactive wastes

    SciTech Connect (OSTI)

    Jacobs, D.G.; Epler, J.S.; Rose, R.R.

    1980-03-01

    A review of problems encountered in the shallow land burial of low-level radioactive wastes has been made in support of the technical aspects of the National Low-Level Waste (LLW) Management Research and Development Program being administered by the Low-Level Waste Management Program Office, Oak Ridge National Laboratory. The operating histories of burial sites at six major DOE and five commercial facilities in the US have been examined and several major problems identified. The problems experienced st the sites have been grouped into general categories dealing with site development, waste characterization, operation, and performance evaluation. Based on this grouping of the problem, a number of major technical issues have been identified which should be incorporated into program plans for further research and development. For each technical issue a discussion is presented relating the issue to a particular problem, identifying some recent or current related research, and suggesting further work necessary for resolving the issue. Major technical issues which have been identified include the need for improved water management, further understanding of the effect of chemical and physical parameters on radionuclide migration, more comprehensive waste records, improved programs for performance monitoring and evaluation, development of better predictive capabilities, evaluation of space utilization, and improved management control.

  11. Retrofit of an Engineered Glove-port to a Los Alamos National Laboratory's Plutonium Facility Glovebox

    SciTech Connect (OSTI)

    Rael, P.E.D.; Cournoyer, M.E.Ph.D.; Chunglo, S.D.; Vigil, T.J.; Schreiber, P.E.S.

    2008-07-01

    At the Los Alamos National Laboratory's Plutonium Facility (TA-55), various isotopes of plutonium along with other actinides are routinely handled such that the spread of radiological contamination and excursions of contaminants into the operator's breathing zone are prevented through the use of a variety of gloveboxes (the glovebox coupled with adequate negativity providing primary confinement). The current technique for changing glovebox gloves are the weakest part of this engineering control. 1300 pairs of gloves are replaced each year at TA-55, generating approximately 500 m{sup 3}/yr of transuranic (TRU) waste and Low Level Waste (LLW) waste that represents an annual disposal cost of about 4 million dollars. By retrofitting the LANL 8'' glove-port ring, a modern 'Push-Through' technology is utilized. This 'Push-Through' technology allows relatively fast glove changes to be done by operators with much less training and experience and without breaching containment. A dramatic reduction in waste is realized; exposure of the worker to residual contamination reduced, and the number of breaches due to installation issues is eliminated. In the following presentation, the evolution of the 'Push- Through' technology, the features of the glove-port retrofit, and waste savings are discussed. (author)

  12. Evaluation of plasma melter technology for verification of high-sodium content low-level radioactive liquid wastes: Demonstration test No. 4 preliminary test report

    SciTech Connect (OSTI)

    McLaughlin, D.F.; Gass, W.R.; Dighe, S.V.; D`Amico, N.; Swensrud, R.L.; Darr, M.F.

    1995-01-10

    This document provides a preliminary report of plasma arc vitrification testing by a vendor in support of the Hanford Tank Waste Remediation System Low-Level Waste (LLW) Vitrification Program. Phase I test conduct included 26 hours (24 hours steady state) of melting of simulated high-sodium low-level radioactive liquid waste. Average processing rate was 4.9 kg/min (peak rate 6.2 kg/min), producing 7330 kg glass product. Free-flowing glass pour point was 1250 C, and power input averaged 1530 kW(e), for a total energy consumption of 19,800 kJ/kg glass. Restart capability was demonstrated following a 40-min outage involving the scrubber liquor heat exchanger, and glass production was continued for another 2 hours. Some volatility losses were apparent, probably in the form of sodium borates. Roughly 275 samples were collected and forwarded for analysis. Sufficient process data were collected for heat/material balances. Recommendations for future work include lower boron contents and improved tuyere design/operation.

  13. DEISCODES. For Environmental Impact Statements

    SciTech Connect (OSTI)

    Widmayer, D.A. [U.S. NRC, Office of Material Safety and Safegaurds, Washington, D.C., (United States)

    1983-01-01

    DEISCODES, the Draft Environmental Impact Statement CODES are five separate FORTRAN codes used to perform the analysis in the Draft Environmental Impact Statement written to support 10 CFR 61, Licensing Requirements for Land Disposal of Radioactive Waste. The five codes are named OPTIONS, GRWATER, INTRUDE, INVERSW, and INVERSI. These codes calculate impact measures associated with the management of Low-Level radioactive Waste (LLW). Three phases of waste management are considered: waste processing, transportation, and disposal, utilizing (1) information on waste characteristics, (2) data and assumptions on disposal technologies and (3) impact calculational methodologies presented in NUREG/CR-1759 and NUREG-0782. The INTRUDE code determines the radiological impacts resulting from potential inadvertent human intrusion into a selected disposal facility containing processed waste as a function of time after disposal. GRWATER calculates the individual exposures resulting from use of contaminated water drawn from various human access locations such as a well that may become contaminated as a result of potential groundwater migration or radionuclides. The OPTIONS code calculates the waste volume-averaged inadvertent intruder impacts, impacts resulting from exposed waste scenarios, as well as those resulting from operational accidents, and those associated with short term consideration such as waste processing and transportation impacts, disposal costs, energy use, land use, etc. INVERSI, calculates the limiting concentrations in waste to meet a specific dose criterion for a disposal facility. INVERSW, calculates disposal facility radionuclide concentrations and inventories to meet specific allowable dose criteria for groundwater migration for the facility design and regionally representative environmental characteristics.

  14. Proceedings of the eighth annual DOE low-level waste management forum: Executive summary, opening plenary session, closing plenary session, attendees

    SciTech Connect (OSTI)

    Not Available

    1987-02-01

    The Eighth Annual DOE (Department of Energy) Low-Level Waste Management Forum was held in September 1986, in Denver, Colorado, to provide a forum for exchange of information on low-level radioactive waste (LLW) management activities, requirements, and plans. The one hundred ninety attendees included representatives from the DOE Nuclear Energy and Defense Low-Level Waste Management Programs, DOE Operations Offices and their contractors; representatives from the US Nuclear Regulatory Commission (NRC), US Environmental Protection Agency (EPA), US Geological Survey, and their contractors; representatives of states and regions responsible for development of new commercial low-level waste disposal facilities; representatives of utilities, private contractors, disposal facility operators, and other parties concerned with low-level waste management issues. Plenary sessions were held at the beginning and conclusion of the meeting, while eight concurrent topical sessions were held during the intervening two days. The meeting was organized by topical areas to allow for information exchange and discussion on current and future low-level radioactive waste management challenges. Session chairmen presented summaries of the discussions and conclusions resulting from their respective sessions. Selected papers in this volume have been processed for inclusion in the Energy Data Base.

  15. Closure Plan for the Area 5 Radioactive Waste Management Site at the Nevada Test Site

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2008-09-01

    The Area 5 Radioactive Waste Management Site (RMWS) at the Nevada Test Site (NTS) is managed and operated by National Security Technologies, LLC (NSTec), for the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (NNSA/NSO). This document is the first update of the preliminary closure plan for the Area 5 RWMS at the NTS that was presented in the Integrated Closure and Monitoring Plan (DOE, 2005a). The major updates to the plan include a new closure schedule, updated closure inventory, updated site and facility characterization data, the Title II engineering cover design, and the closure process for the 92-Acre Area of the RWMS. The format and content of this site-specific plan follows the Format and Content Guide for U.S. Department of Energy Low-Level Waste Disposal Facility Closure Plans (DOE, 1999a). This interim closure plan meets closure and post-closure monitoring requirements of the order DOE O 435.1, manual DOE M 435.1-1, Title 40 Code of Federal Regulations (CFR) Part 191, 40 CFR 265, Nevada Administrative Code (NAC) 444.743, and Resource Conservation and Recovery Act (RCRA) requirements as incorporated into NAC 444.8632. The Area 5 RWMS accepts primarily packaged low-level waste (LLW), low-level mixed waste (LLMW), and asbestiform low-level waste (ALLW) for disposal in excavated disposal cells.

  16. Mixed Waste Management Options: 1995 Update. National Low-Level Waste Management Program

    SciTech Connect (OSTI)

    Kirner, N.; Kelly, J.; Faison, G.; Johnson, D.

    1995-05-01

    In the original mixed Waste Management Options (DOE/LLW-134) issued in December 1991, the question was posed, ``Can mixed waste be managed out of existence?`` That study found that most, but not all, of the Nation`s mixed waste can theoretically be managed out of existence. Four years later, the Nation is still faced with a lack of disposal options for commercially generated mixed waste. However, since publication of the original Mixed Waste Management Options report in 1991, limited disposal capacity and new technologies to treat mixed waste have become available. A more detailed estimate of the Nation`s mixed waste also became available when the US Environmental Protection Agency (EPA) and the US Nuclear Regulatory Commission (NRC) published their comprehensive assessment, titled National Profile on Commercially Generated Low-Level Radioactive Mixed Waste (National Profile). These advancements in our knowledge about mixed waste inventories and generation, coupled with greater treatment and disposal options, lead to a more applied question posed for this updated report: ``Which mixed waste has no treatment option?`` Beyond estimating the volume of mixed waste requiring jointly regulated disposal, this report also provides a general background on the Atomic Energy Act (AEA) and the Resource Conservation and Recovery Act (RCRA). It also presents a methodical approach for generators to use when deciding how to manage their mixed waste. The volume of mixed waste that may require land disposal in a jointly regulated facility each year was estimated through the application of this methodology.

  17. Nevada National Security Site Waste Acceptance Criteria

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2012-02-28

    This document establishes the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (NNSA/NSO), Nevada National Security Site Waste Acceptance Criteria (NNSSWAC). The NNSSWAC provides the requirements, terms, and conditions under which the Nevada National Security Site (NNSS) will accept DOE non-radioactive classified waste, DOE non-radioactive hazardous classified waste, DOE low-level radioactive waste (LLW), DOE mixed low-level waste (MLLW), and U.S. Department of Defense (DOD) classified waste for permanent disposal. Classified waste is the only waste accepted for disposal that may be non-radioactive and will be required to meet the waste acceptance criteria for radioactive waste as specified in this document. The NNSA/NSO and support contractors are available to assist you in understanding or interpreting this document. For assistance, please call the NNSA/NSO Waste Management Project (WMP) at (702) 295-7063, and your call will be directed to the appropriate contact.

  18. Composite Analysis for the Area 5 Radioactive Waste Management Site at the Nevada Test Site, Nye County, Nevada

    SciTech Connect (OSTI)

    V. Yucel

    2001-09-01

    This report summarizes the results of a Composite Analysis (CA) for the Area 5 Radioactive Waste Management Site (RWMS). The Area 5 RWMS is a US Department of Energy (DOE)-operated low-level radioactive waste (LLW) management site located in northern Frenchman Flat on the Nevada Test Site (NTS). The Area 5 RWMS has disposed of low-level radioactive waste in shallow unlined pits and trenches since 1960. Transuranic waste (TRU) and high-specific activity waste was disposed in Greater Confinement Disposal (GCD) boreholes from 1983 to 1989. The purpose of this CA is to determine if continuing operation of the Area 5 RWMS poses an acceptable or unacceptable risk to the public considering the total waste inventory and all other interacting sources of radioactive material in the vicinity. Continuing operation of the Area 5 RWMS will be considered acceptable if the total effective dose equivalent (TEDE) is less than 100 mrem in a year. If the TEDE exceeds 30 mrem in a year, a cost-benefit options analysis must be performed to determine if cost-effective management options exist to reduce the dose further. If the TEDE is found to be less than 30 mrem in a year, an analysis may be performed if warranted to determine if doses are as low as reasonably achievable (ALARA).

  19. Concrete Property and Radionuclide Migration Tests

    SciTech Connect (OSTI)

    Wellman, Dawn M.; Mattigod, Shas V.; Powers, Laura; Parker, Kent E.; Clayton, Libby N.; Wood, Marcus I.

    2008-10-01

    The Waste Management Project provides safe, compliant, and cost-effective waste management services for the Hanford Site and the DOE Complex. Part of theses services includes safe disposal of LLW and MLLW at the Hanford Low-Level Waste Burial Grounds (LLBG) in accordance with the requirements listed in DOE Order 435.1, Radioactive Waste Management. To partially satisfy these requirements, a Performance Assessment (PA) analyses were completed and approved. DOE Order 435.1 also requires that continuing data collection be conducted to enhance confidence in the critical assumptions used in these analyses to characterize the operational features of the disposal facility that are relied upon to satisfy the performance objectives identified in the Order. One critical assumption is that concrete will frequently be used as waste form or container material to control and minimize the release of radionuclide constituents in waste into the surrounding environment. Data was collected to (1) quantify radionuclide migration through concrete materials similar to those used to encapsulate waste in the LLBG, (2) measure the properties of the concrete materials, especially those likely to influence radionuclide migration, and (3) quantify the stability of U-bearing solid phases of limited solubility in concrete.

  20. Maintenance Plan for the Performance Assessments and Composite Analyses for the Area 3 and Area 5 Radioactive Waste Management Sites at the NTS

    SciTech Connect (OSTI)

    Vefa Yucel

    2007-01-03

    U.S. Department of Energy (DOE) Manual M 435.1-1 requires that performance assessments (PAs) and composite analyses (CAs) for low-level waste (LLW) disposal facilities be maintained by the field offices. This plan describes the activities performed to maintain the PA and the CA for the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) at the Nevada Test Site (NTS). This plan supersedes the Maintenance Plan for the Performance Assessments and Composite Analyses for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada Test Site (DOE/NV/11718--491-REV 1, dated September 2002). The plan is based on U.S. Department of Energy (DOE) Order 435.1 (DOE, 1999a), DOE Manual M 435.1-1 (DOE, 1999b), the DOE M 435.1-1 Implementation Guide DOE G 435.1-1 (DOE, 1999c), and the Maintenance Guide for PAs and CAs (DOE, 1999d). The plan includes a current update on PA/CA documentation, a revised schedule, and a section on Quality Assurance.

  1. SOLID WASTE INTEGRATED FORECAST TECHNICAL (SWIFT) REPORT FY2005 THRU FY2035 2005.0 VOLUME 2

    SciTech Connect (OSTI)

    BARCOT, R.A.

    2005-08-17

    This report provides up-to-date life cycle information about the radioactive solid waste expected to be managed by Hanford's Waste Management (WM) Project from onsite and offsite generators. It includes: (1) an overview of Hanford-wide solid waste to be managed by the WM Project; (2) multi-level and waste class-specific estimates; (3) background information on waste sources; and (4) comparisons to previous forecasts and other national data sources. The focus of this report is low-level waste (LLW), mixed low-level waste (MLLW), and transuranic waste, both non-mixed and mixed (TRU(M)). Some details on hazardous waste are also provided, however, this information is not considered comprehensive. This report includes data requested in December, 2004 with updates through March 31,2005. The data represent a life cycle forecast covering all reported activities from FY2005 through the end of each program's life cycle and are an update of the previous FY2004.1 data version.

  2. Regulatory Closure Options for the Residue in the Hanford Site Single-Shell Tanks

    SciTech Connect (OSTI)

    Cochran, J.R. Shyr, L.J.

    1998-10-05

    Liquid, mixed, high-level radioactive waste (HLW) has been stored in 149 single-shell tanks (SSTS) located in tank farms on the U.S. Department of Energy's (DOE's) Hanford Site. The DOE is developing technologies to retrieve as much remaining HLW as technically possible prior to physically closing the tank farms. In support of the Hanford Tanks Initiative, Sandia National Laboratories has addressed the requirements for the regulatory closure of the radioactive component of any SST residue that may remain after physical closure. There is significant uncertainty about the end state of each of the 149 SSTS; that is, the nature and amount of wastes remaining in the SSTS after retrieval is uncertain. As a means of proceeding in the face of these uncertainties, this report links possible end-states with associated closure options. Requirements for disposal of HLW and low-level radioactive waste (LLW) are reviewed in detail. Incidental waste, which is radioactive waste produced incidental to the further processing of HLW, is then discussed. If the low activity waste (LAW) fraction from the further processing of HLW is determined to be incidental waste, then DOE can dispose of that incidental waste onsite without a license from the U.S. Nuclear Regulatory Commissions (NRC). The NRC has proposed three Incidental Waste Criteria for determining if a LAW fraction is incidental waste. One of the three Criteria is that the LAW fraction should not exceed the NRC's Class C limits.

  3. Waste Receiving and Processing Facility Module 1: Volume 1, Preliminary Design report

    SciTech Connect (OSTI)

    Not Available

    1992-03-01

    The Preliminary Design Report (Title 1) for the Waste Receiving and Processing (WRAP) Module 1 provides a comprehensive narrative description of the proposed facility and process systems, the basis for each of the systems design, and the engineering assessments that were performed to support the technical basis of the Title 1 design. The primary mission of the WRAP 1 Facility is to characterize and certify contact-handled (CH) waste in 55-gallon drums for disposal. Its secondary function is to certify CH waste in Standard Waste Boxes (SWBs) for disposal. The preferred plan consist of retrieving the waste and repackaging as necessary in the Waste Receiving and Processing (WRAP) facility to certify TRU waste for shipment to the Waste Isolation Pilot Plant (WIPP) in New Mexico. WIPP is a research and development facility designed to demonstrate the safe and environmentally acceptable disposal of TRU waste from National Defense programs. Retrieved waste found to be Low-Level Waste (LLW) after examination in the WRAP facility will be disposed of on the Hanford site in the low-level waste burial ground. The Hanford Site TRU waste will be shipped to the WIPP for disposal between 1999 and 2013.

  4. Integrated Disposal Facility FY2011 Glass Testing Summary Report

    SciTech Connect (OSTI)

    Pierce, Eric M.; Bacon, Diana H.; Kerisit, Sebastien N.; Windisch, Charles F.; Cantrell, Kirk J.; Valenta, Michelle M.; Burton, Sarah D.; Westsik, Joseph H.

    2011-09-29

    Pacific Northwest National Laboratory was contracted by Washington River Protection Solutions, LLC to provide the technical basis for estimating radionuclide release from the engineered portion of the disposal facility (e.g., source term). Vitrifying the low-activity waste at Hanford is expected to generate over 1.6 x 10{sup 5} m{sup 3} of glass (Certa and Wells 2010). The volume of immobilized low-activity waste (ILAW) at Hanford is the largest in the DOE complex and is one of the largest inventories (approximately 8.9 x 10{sup 14} Bq total activity) of long-lived radionuclides, principally {sup 99}Tc (t{sub 1/2} = 2.1 x 10{sup 5}), planned for disposal in a low-level waste (LLW) facility. Before the ILAW can be disposed, DOE must conduct a performance assessment (PA) for the Integrated Disposal Facility (IDF) that describes the long-term impacts of the disposal facility on public health and environmental resources. As part of the ILAW glass testing program PNNL is implementing a strategy, consisting of experimentation and modeling, in order to provide the technical basis for estimating radionuclide release from the glass waste form in support of future IDF PAs. The purpose of this report is to summarize the progress made in fiscal year (FY) 2011 toward implementing the strategy with the goal of developing an understanding of the long-term corrosion behavior of low-activity waste glasses.

  5. Standardization of DOE Disposal Facilities Waste Acceptance Process

    SciTech Connect (OSTI)

    SHRADER, T.; MACBETH, P.

    2002-01-01

    On February 25, 2000, the US. Department of Energy (DOE) issued the Record of Decision (ROD) for the Waste Management Programmatic Environmental Impact Statement (WM PEIS) for low-level and mixed low-level wastes (LLW/ MLLW) treatment and disposal. The ROD designated the disposal sites at Hanford and the Nevada Test Site (NTS) to dispose of LLWMLLW from sites without their own disposal facilities. DOE's Richland Operations Office (RL) and the National Nuclear Security Administration's Nevada Operations Office (NV) have been charged with effectively implementing the ROD. To accomplish this task NV and RL, assisted by their operating contractors Bechtel Nevada (BN), Fluor Hanford (FH), and Bechtel Hanford (BH) assembled a task team to systematically map out and evaluate the current waste acceptance processes and develop an integrated, standardized process for the acceptance of LLWMLLW. A structured, systematic, analytical process using the Six Sigma system identified disposal process improvements and quantified the associated efficiency gains to guide changes to be implemented. The review concluded that a unified and integrated Hanford/NTS Waste Acceptance Process would be a benefit to the DOE Complex, particularly the waste generators. The Six Sigma review developed quantitative metrics to address waste acceptance process efficiency improvements, and provides an initial look at development of comparable waste disposal cost models between the two disposal sites to allow quantification of the proposed improvements.

  6. Pretreatment status report on the identification and evaluation of alternative processes. Milestone Report No. C064

    SciTech Connect (OSTI)

    Sutherland, D.G.; Brothers, A.J.; Beary, M.M.; Nicholson, G.A.

    1993-09-01

    The purpose of this report is to support the development and demonstration of a pretreatment system that will (1) destroy organic materials and ferrocyanide in tank wastes so that the wastes can be stored safely, (2) separate the high-activity and low-activity fractions, (3) remove radionuclides and remove or destroy hazardous chemicals in LLW as necessary to meet waste form feed requirements, (4) support development and demonstration of vitrification technology by providing representative feeds to the bench-scale glass melter, (5) support full-scale HLW vitrification operations, including near-term operation, by providing feed that meets specifications, and (6) design and develop pretreatment processes that accomplish the above objectives and ensure compliance with environmental regulations. This report is a presentation of candidate technologies for pretreatment of Hanford Site tank waste. Included are descriptions of studies by the Pacific Northwest Laboratory of Battelle Memorial Institute; Science Applications International Corporation, an independent consultant; BNFL, Inc. representing British technologies; Numatec, representing French technologies; and brief accounts of other relevant activities.

  7. Site characterization and monitoring data from Area 5 Pilot Wells, Nevada Test Site, Nye County, Nevada

    SciTech Connect (OSTI)

    NONE

    1994-02-01

    The Special Projects Section (SPS) of Reynolds Electrical & Engineering Co., Inc. (REECO) is responsible for characterizing the subsurface geology and hydrology of the Area 5 Radioactive Waste Management Site (RWMS) at the Nevada Test Site (NTS) for the US Department of Energy, Nevada Operations Office (DOE/NV), Environmental Restoration and Waste Management Division, Waste Operations Branch. The three Pilot Wells that comprise the Pilot Well Project are an important part of the Area 5 Site Characterization Program designed to determine the suitability of the Area 5 RWMS for disposal of low-level waste (LLW), mixed waste (MW), and transuranic waste (TRU). The primary purpose of the Pilot Well Project is two-fold: first, to characterize important water quality and hydrologic properties of the uppermost aquifer; and second, to characterize the lithologic, stratigraphic, and hydrologic conditions which influence infiltration, redistribution, and percolation, and chemical transport through the thick vadose zone in the vicinity of the Area 5 RWMS. This report describes Pilot Well drilling and coring, geophysical logging, instrumentation and stemming, laboratory testing, and in situ testing and monitoring activities.

  8. Code requirements for concrete repository and processing facilities

    SciTech Connect (OSTI)

    Hookham, C.J. [Black & Veatch, Ann Arbor, MI (United States); Palaniswamy, R. [Bechtel Savannah River, Inc., North Augusta, SC (United States)

    1993-04-01

    The design and construction of facilities and structures for the processing and safe long-term storage of low- and high-level radioactive wastes will likely employ structural concrete. This concrete will be used for many purposes including structural support, shielding, and environmental protection. At the present time, there are no design costs, standards or guidelines for repositories, waste containers, or processing facilities. Recently, the design and construction guidelines contained in American Concrete Institute (ACI), Code Requirements for Nuclear Safety Related Concrete Structures (ACI 349), have been cited for low-level waste (LLW) repositories. Conceptual design of various high-level (HLW) repository surface structures have also cited the ACI 349 Code. However, the present Code was developed for nuclear power generating facilities and its application to radioactive waste repositories was not intended. For low and medium level radioactive wastes, concrete has a greater role and use in processing facilities, engineered barriers, and repository structures. Because of varied uses and performance/safety requirements this review of the current ACI 349 Code document was required to accommodate these special classes of structures.

  9. Integrating Volume Reduction and Packaging Alternatives to Achieve Cost Savings for Low Level Waste Disposal at the Rocky Flats Environmental Technology Site

    SciTech Connect (OSTI)

    Church, A.; Gordon, J.; Montrose, J. K.

    2002-02-26

    In order to reduce costs and achieve schedules for Closure of the Rocky Flats Environmental Technology Site (RFETS), the Waste Requirements Group has implemented a number of cost saving initiatives aimed at integrating waste volume reduction with the selection of compliant waste packaging methods for the disposal of RFETS low level radioactive waste (LLW). Waste Guidance Inventory and Shipping Forecasts indicate that over 200,000 m3 of low level waste will be shipped offsite between FY2002 and FY2006. Current projections indicate that the majority of this waste will be shipped offsite in an estimated 40,000 55-gallon drums, 10,000 metal and plywood boxes, and 5000 cargo containers. Currently, the projected cost for packaging, shipment, and disposal adds up to $80 million. With these waste volume and cost projections, the need for more efficient and cost effective packaging and transportation options were apparent in order to reduce costs and achieve future Site packaging a nd transportation needs. This paper presents some of the cost saving initiatives being implemented for waste packaging at the Rocky Flats Environmental Technology Site (the Site). There are many options for either volume reduction or alternative packaging. Each building and/or project may indicate different preferences and/or combinations of options.

  10. THE PRODUCT CONSISTENCY TEST HOW AND WHY IT WAS DEVELOPED

    SciTech Connect (OSTI)

    Jantzen, C; Ned Bibler, N

    2008-12-15

    The Product Consistency Test (PCT), American Society for Testing Materials (ASTM) Standard C1285, is currently used world wide for testing glass and glass-ceramic waste forms for high level waste (HLW), low level waste (LLW), and hazardous wastes. Development of the PCT was initiated in 1986 because HLW glass waste forms required extensive characterization before actual production began and required continued characterization during production ({ge}25 years). Non-radioactive startup was in 1994 and radioactive startup was in 1996. The PCT underwent extensive development from 1986-1994 and became an ASTM consensus standard in 1994. During the extensive laboratory testing and inter- and intra-laboratory round robins using non-radioactive and radioactive glasses, the PCT was shown to be very reproducible, to yield reliable results rapidly, to distinguish between glasses of different durability and homogeneity, and to easily be performed in shielded cell facilities with radioactive samples. In 1997, the scope was broadened to include hazardous and mixed (radioactive and hazardous) waste glasses. In 2002, the scope was broadened to include glass-ceramic waste forms which are currently being recommended for second generation nuclear wastes yet to be generated in the nuclear renaissance. Since the PCT has proven useful for glass-ceramics with up to 75% ceramic component and has been used to evaluate Pu ceramic waste forms, the use of this test for other ceramic/mineral waste forms such as geopolymers, hydroceramics, and fluidized bed steam reformer mineralized product is under investigation.

  11. Minor component study for simulated high-level nuclear waste glasses (Draft)

    SciTech Connect (OSTI)

    Li, H.; Langowskim, M.H.; Hrma, P.R.; Schweiger, M.J.; Vienna, J.D.; Smith, D.E.

    1996-02-01

    Hanford Site single-shell tank (SSI) and double-shell tank (DSI) wastes are planned to be separated into low activity (or low-level waste, LLW) and high activity (or high-level waste, HLW) fractions, and to be vitrified for disposal. Formulation of HLW glass must comply with glass processibility and durability requirements, including constraints on melt viscosity, electrical conductivity, liquidus temperature, tendency for phase segregation on the molten glass surface, and chemical durability of the final waste form. A wide variety of HLW compositions are expected to be vitrified. In addition these wastes will likely vary in composition from current estimates. High concentrations of certain troublesome components, such as sulfate, phosphate, and chrome, raise concerns about their potential hinderance to the waste vitrification process. For example, phosphate segregation in the cold cap (the layer of feed on top of the glass melt) in a Joule-heated melter may inhibit the melting process (Bunnell, 1988). This has been reported during a pilot-scale ceramic melter run, PSCM-19, (Perez, 1985). Molten salt segregation of either sulfate or chromate is also hazardous to the waste vitrification process. Excessive (Cr, Fe, Mn, Ni) spinel crystal formation in molten glass can also be detrimental to melter operation.

  12. Preliminary time-phased TWRS process model results

    SciTech Connect (OSTI)

    Orme, R.M.

    1995-03-24

    This report documents the first phase of efforts to model the retrieval and processing of Hanford tank waste within the constraints of an assumed tank farm configuration. This time-phased approach simulates a first try at a retrieval sequence, the batching of waste through retrieval facilities, the batching of retrieved waste through enhanced sludge washing, the batching of liquids through pretreatment and low-level waste (LLW) vitrification, and the batching of pretreated solids through high-level waste (HLW) vitrification. The results reflect the outcome of an assumed retrieval sequence that has not been tailored with respect to accepted measures of performance. The batch data, composition variability, and final waste volume projects in this report should be regarded as tentative. Nevertheless, the results provide interesting insights into time-phased processing of the tank waste. Inspection of the composition variability, for example, suggests modifications to the retrieval sequence that will further improve the uniformity of feed to the vitrification facilities. This model will be a valuable tool for evaluating suggested retrieval sequences and establishing a time-phased processing baseline. An official recommendation on tank retrieval sequence will be made in September, 1995.

  13. Optimization of microwave heating in an existing cubicle cavity by incorporating additional wave guide and control components

    SciTech Connect (OSTI)

    Erle, R.R.; Eschen, V.G.; Sprenger, G.S.

    1995-04-01

    The use of microwave energy to thermally treat Low Level (LLW), Transuranic (TRU), and mixed waste has been under development at the Rocky Flats Environmental Technology Site (Site) since 1986. During that time, the technology has progressed from bench-scale tests, through pilot-scale tests, and finally to a full-scale demonstration unit. Experimental operations have been conducted on a variety of non-radioactive surrogates and actual radioactive waste forms. Through these studies and development efforts, the Microwave Vitrification Engineering Team (MVET) at Rocky Flats has successfully proven the application of microwave energy for waste treatment operations. In the microwave solidification process, microwave energy is used to heat a mixture of waste and glass frit to produce a vitrified product that meets all the current acceptance criteria at the final disposal sites. All of the development to date has utilized a multi-mode microwave system to provide the energy to treat the materials. Currently, evaluations are underway on modifications to the full-scale demonstration system that provide a single-mode operation as a possible method to optimize the system. This poster presentation describes the modifications made to allow the single-mode operation.

  14. Waste management system alternatives for treatment of wastes from spent fuel reprocessing

    SciTech Connect (OSTI)

    McKee, R.W.; Swanson, J.L.; Daling, P.M.; Clark, L.L.; Craig, R.A.; Nesbitt, J.F.; McCarthy, D.; Franklin, A.L.; Hazelton, R.F.; Lundgren, R.A.

    1986-09-01

    This study was performed to help identify a preferred TRU waste treatment alternative for reprocessing wastes with respect to waste form performance in a geologic repository, near-term waste management system risks, and minimum waste management system costs. The results were intended for use in developing TRU waste acceptance requirements that may be needed to meet regulatory requirements for disposal of TRU wastes in a geologic repository. The waste management system components included in this analysis are waste treatment and packaging, transportation, and disposal. The major features of the TRU waste treatment alternatives examined here include: (1) packaging (as-produced) without treatment (PWOT); (2) compaction of hulls and other compactable wastes; (3) incineration of combustibles with cementation of the ash plus compaction of hulls and filters; (4) melting of hulls and failed equipment plus incineration of combustibles with vitrification of the ash along with the HLW; (5a) decontamination of hulls and failed equipment to produce LLW plus incineration and incorporation of ash and other inert wastes into HLW glass; and (5b) variation of this fifth treatment alternative in which the incineration ash is incorporated into a separate TRU waste glass. The six alternative processing system concepts provide progressively increasing levels of TRU waste consolidation and TRU waste form integrity. Vitrification of HLW and intermediate-level liquid wastes (ILLW) was assumed in all cases.

  15. LLNL Site plan for a MOX fuel lead assembly mission in support of surplus plutonium disposition

    SciTech Connect (OSTI)

    Bronson, M.C.

    1997-10-01

    The principal facilities that LLNL would use to support a MOX Fuel Lead Assembly Mission are Building 332 and Building 334. Both of these buildings are within the security boundary known as the LLNL Superblock. Building 332 is the LLNL Plutonium Facility. As an operational plutonium facility, it has all the infrastructure and support services required for plutonium operations. The LLNL Plutonium Facility routinely handles kilogram quantities of plutonium and uranium. Currently, the building is limited to a plutonium inventory of 700 kilograms and a uranium inventory of 300 kilograms. Process rooms (excluding the vaults) are limited to an inventory of 20 kilograms per room. Ongoing operations include: receiving SSTS, material receipt, storage, metal machining and casting, welding, metal-to-oxide conversion, purification, molten salt operations, chlorination, oxide calcination, cold pressing and sintering, vitrification, encapsulation, chemical analysis, metallography and microprobe analysis, waste material processing, material accountability measurements, packaging, and material shipping. Building 334 is the Hardened Engineering Test Building. This building supports environmental and radiation measurements on encapsulated plutonium and uranium components. Other existing facilities that would be used to support a MOX Fuel Lead Assembly Mission include Building 335 for hardware receiving and storage and TRU and LLW waste storage and shipping facilities, and Building 331 or Building 241 for storage of depleted uranium.

  16. Processing mixed-waste compressed-gas cylinders at the Oak Ridge Reservation

    SciTech Connect (OSTI)

    Morris, M.I.; Conley, T.B.; Osborne-Lee, I.W.

    1998-05-01

    Until recently, several thousand kilograms of compressed gases were stored at the Oak Ridge Reservation (ORR), in Oak Ridge, Tennessee, because these cylinders could not be taken off-site in their state of configuration for disposal. Restrictions on the storage of old compressed-gas cylinders compelled the Waste Management Organization of Lockheed Martin Energy Systems, Inc. (LMES) to dispose of these materials. Furthermore, a milestone in the ORR Site Treatment Plan required repackaging and shipment off-site of 21 cylinders by September 30, 1997. A pilot project, coordinated by the Chemical Technology Division (CTD) at the Oak Ridge National Laboratory (ORNL), was undertaken to evaluate and recontainerize or neutralize these cylinders, which are mixed waste, to meet that milestone. Because the radiological component was considered to be confined to the exterior of the cylinder, the contents (once removed from the cylinder) could be handled as hazardous waste, and the cylinder could be handled as low-level waste (LLW). This pilot project to process 21 cylinders was important because of its potential impact. The successful completion of the project provides a newly demonstrated technology which can now be used to process the thousands of additional cylinders in inventory across the DOE complex. In this paper, many of the various aspects of implementing this project, including hurdles encountered and the lessons learned in overcoming them, are reported.

  17. Waste minimization measures associated with the analysis of {sup 137}Cs in coconut milk collected from the Marshall Islands

    SciTech Connect (OSTI)

    Hamilton, T.; Jones, H.; Wong, K.; Robinson, W.

    1998-05-01

    The Marshall Islands Environmental Characterization and Dose Assessment Program has recently implemented waste minimization measures to reduce low level radioactive (LLW) and low level mixed (LLWMIXED) waste streams at the Lawrence Livermore National Laboratory (LLNL). Several thousand environmental samples are collected annually from former US nuclear test sites in the Marshall Islands, and returned to LLNL for processing and radiometric analysis. In the past, we analyzed coconut milk directly by gamma-spectrometry after adding formaldehyde (as preservative) and sealing the fluid in metal cans. This procedure was not only tedious and time consuming but generated storage and waste disposal problems. We have now reduced the number of coconut milk samples required for analysis from 1500 per year to approximately 250, and developed a new analytical procedure which essentially eliminates the associated mixed radioactive waste stream. Coconut milk samples are mixed with a few grams of ammonium-molydophosphate (AMP) which quantitatively scavenges the target radionuclide cesium 137 in an ion-exchange process. The AMP is then separated from the mixture and sealed in a plastic container. The bulk sample material can be disposed of as a non- radioactive non-hazardous waste, and the relatively small amount of AMP conveniently counted by gamma-spectrometry, packaged and stored for future use.

  18. Analysis of accident sequences and source terms at treatment and storage facilities for waste generated by US Department of Energy waste management operations

    SciTech Connect (OSTI)

    Mueller, C.; Nabelssi, B.; Roglans-Ribas, J.; Folga, S.; Policastro, A.; Freeman, W.; Jackson, R.; Mishima, J.; Turner, S.

    1996-12-01

    This report documents the methodology, computational framework, and results of facility accident analyses performed for the US Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies assessed, and the resultant radiological and chemical source terms evaluated. A personal-computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for the calculation of human health risk impacts. The WM PEIS addresses management of five waste streams in the DOE complex: low-level waste (LLW), hazardous waste (HW), high-level waste (HLW), low-level mixed waste (LLMW), and transuranic waste (TRUW). Currently projected waste generation rates, storage inventories, and treatment process throughputs have been calculated for each of the waste streams. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated, and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. Key assumptions in the development of the source terms are identified. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also discuss specific accident analysis data and guidance used or consulted in this report.

  19. Documented Safety Analysis for the Waste Storage Facilities March 2010

    SciTech Connect (OSTI)

    Laycak, D T

    2010-03-05

    This Documented Safety Analysis (DSA) for the Waste Storage Facilities was developed in accordance with 10 CFR 830, Subpart B, 'Safety Basis Requirements,' and utilizes the methodology outlined in DOE-STD-3009-94, Change Notice 3. The Waste Storage Facilities consist of Area 625 (A625) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area portion of the DWTF complex. These two areas are combined into a single DSA, as their functions as storage for radioactive and hazardous waste are essentially identical. The B695 Segment of DWTF is addressed under a separate DSA. This DSA provides a description of the Waste Storage Facilities and the operations conducted therein; identification of hazards; analyses of the hazards, including inventories, bounding releases, consequences, and conclusions; and programmatic elements that describe the current capacity for safe operations. The mission of the Waste Storage Facilities is to safely handle, store, and treat hazardous waste, transuranic (TRU) waste, low-level waste (LLW), mixed waste, combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL (as well as small amounts from other DOE facilities).

  20. Nonradioactive air emissions notice of construction for the Waste Receiving And Processing facility

    SciTech Connect (OSTI)

    Not Available

    1993-02-01

    The mission of the Waste Receiving And Processing (WRAP) Module 1 facility (also referred to as WRAP 1) is to examine assay, characterize, treat, and repackage solid radioactive and mixed waste to enable permanent disposal of the wastes in accordance with all applicable regulations. WRAP 1 will contain equipment and facilities necessary for non-destructive examination (NDE) of wastes and to perform a non-destructive examination assay (NDA) of the total radionuclide content of the wastes, without opening the outer container (e.g., 55-gal drum). WRAP 1 will also be equipped to open drums which do not meet waste acceptance and shipping criteria, and to perform limited physical treatment of the wastes to ensure that storage, shipping, and disposal criteria are met. The solid wastes to be handled in the WRAP 1 facility include low level waste (LLW), transuranic (TRU) waste, and transuranic and low level mixed wastes (LLMW). The WRAP 1 facility will only accept contact handler (CH) waste containers. A Best Available Control Technology for Toxics (TBACT) assessment has been completed for the WRAP 1 facility (WHC 1993). Because toxic emissions from the WRAP 1 facility are sufficiently low and do not pose any health or safety concerns to the public, no controls for volatile organic compounds (VOCs), and installation of HEPA filters for particulates satisfy TBACT for the facility.

  1. A modeling study of the effect of depth of burial of depleted uranium and thorium on radon gas flux at a dry desert alluvial soil radioactive waste management site (RWMS)

    SciTech Connect (OSTI)

    Lindstrom, F.T.; Cawlfield, D.E.; Emer, D.F.; Shott, G.J.

    1993-08-01

    An integral part of designing low-level waste (LLW) disposal pits and their associated closure covers in very dry desert alluvium is the use of a radon gas transport and fate model. Radon-222 has the potential to be a real heath hazard. The production of radon-222 results from the radioactive decay (a particle emission) of radium-226 in the uranium-235 and 238 Bateman chains. It is also produced in the thorium-230 series. Both long lived radionuclides have been proposed for disposal in the shallow land burial pits in Area 5 RWMS compound of Nevada Test Site (NTS). The constructed physics based model includes diffusion and barometric pressure-induced advection of an M-chain of radionuclides. The usual Bateman decay mechanics are included for each radionuclide. Both linear reversible and linear irreversible first order sorption kinetics are assumed for each radionuclide. This report presents the details of using the noble gas transport model, CASCADR9, in an engineering design study mode. Given data on the low-level waste stream, which constitutes the ultimate source of radon-222 in the RWMS, CASCADR9 is used to generate the surface flux (pCi/cm{sup 2}-sec) of radon-222 under the realistic atmospheric and alluvial soil conditions found in the RWMS at Area 5, of the NTS. Specifically, this study examines the surface flux of radon-222 as a function of the depth of burial below the land surface.

  2. Low level tank waste disposal study

    SciTech Connect (OSTI)

    Mullally, J.A.

    1994-09-29

    Westinghouse Hanford Company (WHC) contracted a team consisting of Los Alamos Technical Associates (LATA), British Nuclear Fuel Laboratories (BNFL), Southwest Research Institute (SwRI), and TRW through the Tank Waste Remediation System (TWRS) Technical Support Contract to conduct a study on several areas concerning vitrification and disposal of low-level-waste (LLW). The purpose of the study was to investigate how several parameters could be specified to achieve full compliance with regulations. The most restrictive regulation governing this disposal activity is the National Primary Drinking Water Act which sets the limits of exposure to 4 mrem per year for a person drinking two liters of ground water daily. To fully comply, this constraint would be met independently of the passage of time. In addition, another key factor in the investigation was the capability to retrieve the disposed waste during the first 50 years as specified in Department of Energy (DOE) Order 5820.2A. The objective of the project was to develop a strategy for effective long-term disposal of the low-level waste at the Hanford site.

  3. Cost savings associated with landfilling wastes containing very low levels of uranium

    SciTech Connect (OSTI)

    Boggs, C.J. [Argonne National Lab., Germantown, MD (United States); Shaddoan, W.T. [Lockheed Martin Energy Systems, Paducah, KY (United States)

    1996-03-01

    The Paducah Gaseous Diffusion Plant (PGDP) has operated captive landfills (both residential and construction/demolition debris) in accordance with the Commonwealth of Kentucky regulations since the early 1980s. Typical waste streams allowed in these landfills include nonhazardous industrial and municipal solid waste (such as paper, plastic, cardboard, cafeteria waste, clothing, wood, asbestos, fly ash, metals, and construction debris). In July 1992, the U.S. Environmental Protection Agency issued new requirements for the disposal of sanitary wastes in a {open_quotes}contained landfill.{close_quotes} These requirements were promulgated in the 401 Kentucky Administrative Record Chapters 47 and 48 that became effective 30 June 1995. The requirements for a new contained landfill include a synthetic liner made of high-density polyethylene in addition to the traditional 1-meter (3-foot) clay liner and a leachate collection system. A new landfill at Paducah would accept waste streams similar to those that have been accepted in the past. The permit for the previously existing landfills did not include radioactivity limits; instead, these levels were administratively controlled. Typically, if radioactivity was detected above background levels, the waste was classified as low-level waste (LLW), which would be sent off-site for disposal.

  4. Performance assessment for the class L-II disposal facility

    SciTech Connect (OSTI)

    1997-03-01

    This draft radiological performance assessment (PA) for the proposed Class L-II Disposal Facility (CIIDF) on the Oak Ridge Reservation (ORR) has been prepared to demonstrate compliance with the requirements of the US Department of Energy Order 5820.2A. This PA considers the disposal of low-level radioactive wastes (LLW) over the operating life of the facility and the long-term performance of the facility in providing protection to public health and the environment. The performance objectives contained in the order require that the facility be managed to accomplish the following: (1) Protect public health and safety in accordance with standards specified in environmental health orders and other DOE orders. (2) Ensure that external exposure to the waste and concentrations of radioactive material that may be released into surface water, groundwater, soil, plants, and animals results in an effective dose equivalent (EDE) that does not exceed 25 mrem/year to a member of the public. Releases to the atmosphere shall meet the requirements of 40 CFR Pt. 61. Reasonable effort should be made to maintain releases of radioactivity in effluents to the general environment as low as reasonably achievable. (1) Ensure that the committed EDEs received by individual who inadvertently may intrude into the facility after the loss of active institutional control (100 years) will not exceed 100 mrem/year for continuous exposure of 500 mrem for a single acute exposure. (4) Protect groundwater resources, consistent with federal, state, and local requirements.

  5. Proceedings of the US Department of Energy Office of Environmental Restoration and Waste Management

    SciTech Connect (OSTI)

    Not Available

    1990-09-01

    The fifth of a series of waste minimization (WMIN)/reduction workshops (Waste Reduction Workshop V) was held at the Little Tree Inn in Idaho Falls, Idaho, on July 24--26, 1990. The workshops are held under the auspices of the US Department of Energy's (DOE's) Office of Environmental Restoration and Waste Management (EM). The purpose of this workshop was to provide a forum for sharing site activities in WMIN/reduction planning. Topics covered were management commitment, organizational structure, goal setting, reporting requirements, data bases and tracking systems, pollution prevention, awareness and incentives, information exchange, process waste assessment (PWA) implementation, and recycling internal and external. The workshops assist DOE waste-generating sites in implementing WMIN/reduction programs, plans, and activities, thus providing for optimal waste reduction within the DOE complex. All wastes are considered within this discipline: liquid, solid, and airborne, within the categories of high-level waste (HLW), transuranic waste (TRU), low-level waste (LLW), hazardous waste, and mixed waste.

  6. Fast neutron incineration in the energy amplifier as alternative to geologic storage the case of Spain

    E-Print Network [OSTI]

    Rubbia, Carlo; Kadi, Y; Rubio, Juan Antonio

    1997-01-01

    In previous reports [1][2] we have presented the conceptual design of a fast neutron driven sub-critical device (Energy Amplifier) designed both for energy amplification (production) and for the incineration of unwanted waste from Nuclear Light Water Reactors (LWR). The latter scheme is here applied to the specific case of Spain, where 9 large LWRs are presently in operation. It is shown that a cluster of 5 EAs is a very effective and realistic solution to the elimination (in 37 years) of the present and foreseen (till 2029) LWR-Waste stockpiles of Spain, but with major improvements over Geologic Storage, since: (1) only a Low Level Waste (LLW) surface repository of reasonable size is ultimately required; (2) the large amount of energy stored in the trans-Uranics is recovered, amounting for each of the 37 years of incineration to a saving of about 8% of the present primary energy demand of Spain (100 MTep/y); (3) the slightly enriched (1.1%) Uranium, unburned by LWRs, can be recovered for further us...

  7. Operational Strategies for Low-Level Radioactive Waste Disposal Site in Egypt - 13513

    SciTech Connect (OSTI)

    Mohamed, Yasser T. [Hot Laboratories and Waste Management Center, Atomic Energy Authority, 3 Ahmed El-Zomor St., El-Zohour District, Naser City, 11787, Cairo (Egypt)] [Hot Laboratories and Waste Management Center, Atomic Energy Authority, 3 Ahmed El-Zomor St., El-Zohour District, Naser City, 11787, Cairo (Egypt)

    2013-07-01

    The ultimate aims of treatment and conditioning is to prepare waste for disposal by ensuring that the waste will meet the waste acceptance criteria of a disposal facility. Hence the purpose of low-level waste disposal is to isolate the waste from both people and the environment. The radioactive particles in low-level waste emit the same types of radiation that everyone receives from nature. Most low-level waste fades away to natural background levels of radioactivity in months or years. Virtually all of it diminishes to natural levels in less than 300 years. In Egypt, The Hot Laboratories and Waste Management Center has been established since 1983, as a waste management facility for LLW and ILW and the disposal site licensed for preoperational in 2005. The site accepts the low level waste generated on site and off site and unwanted radioactive sealed sources with half-life less than 30 years for disposal and all types of sources for interim storage prior to the final disposal. Operational requirements at the low-level (LLRW) disposal site are listed in the National Center for Nuclear Safety and Radiation Control NCNSRC guidelines. Additional procedures are listed in the Low-Level Radioactive Waste Disposal Facility Standards Manual. The following describes the current operations at the LLRW disposal site. (authors)

  8. Glass optimization for vitrification of Hanford Site low-level tank waste

    SciTech Connect (OSTI)

    Feng, X.; Hrma, P.R.; Westsik, J.H. Jr.

    1996-03-01

    The radioactive defense wastes stored in 177 underground single-shell tanks (SST) and double-shell tanks (DST) at the Hanford Site will be separated into low-level and high-level fractions. One technology activity underway at PNNL is the development of glass formulations for the immobilization of the low-level tank wastes. A glass formulation strategy has been developed that describes development approaches to optimize glass compositions prior to the projected LLW vitrification facility start-up in 2005. Implementation of this strategy requires testing of glass formulations spanning a number of waste loadings, compositions, and additives over the range of expected waste compositions. The resulting glasses will then be characterized and compared to processing and performance specifications yet to be developed. This report documents the glass formulation work conducted at PNL in fiscal years 1994 and 1995 including glass formulation optimization, minor component impacts evaluation, Phase 1 and Phase 2 melter vendor glass development, liquidus temperature and crystallization kinetics determination. This report also summarizes relevant work at PNNL on high-iron glasses for Hanford tank wastes conducted through the Mixed Waste Integrated Program and work at Savannah River Technology Center to optimize glass formulations using a Plackett-Burnam experimental design.

  9. Documented Safety Analysis for the Waste Storage Facilities

    SciTech Connect (OSTI)

    Laycak, D

    2008-06-16

    This documented safety analysis (DSA) for the Waste Storage Facilities was developed in accordance with 10 CFR 830, Subpart B, 'Safety Basis Requirements', and utilizes the methodology outlined in DOE-STD-3009-94, Change Notice 3. The Waste Storage Facilities consist of Area 625 (A625) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area portion of the DWTF complex. These two areas are combined into a single DSA, as their functions as storage for radioactive and hazardous waste are essentially identical. The B695 Segment of DWTF is addressed under a separate DSA. This DSA provides a description of the Waste Storage Facilities and the operations conducted therein; identification of hazards; analyses of the hazards, including inventories, bounding releases, consequences, and conclusions; and programmatic elements that describe the current capacity for safe operations. The mission of the Waste Storage Facilities is to safely handle, store, and treat hazardous waste, transuranic (TRU) waste, low-level waste (LLW), mixed waste, combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL (as well as small amounts from other DOE facilities).

  10. An assessment and evaluation for recycle/reuse of contaminated process and metallurgical equipment at the DOE Rocky Flats Plant Site -- Building 865. Final report

    SciTech Connect (OSTI)

    Not Available

    1993-08-01

    An economic analysis of the potential advantages of alternatives for recycling and reusing equipment now stored in Building 865 at the Rocky Flats Plant (RFP) in Colorado has been conducted. The inventory considered in this analysis consists primarily of metallurgical and process equipment used before January 1992, during development and production of nuclear weapons components at the site. The economic analysis consists of a thorough building inventory and cost comparisons for four equipment dispositions alternatives. The first is a baseline option of disposal at a Low Level Waste (LLW) landfill. The three alternatives investigated are metal recycling, reuse with the government sector, and release for unrestricted use. This report provides item-by-item estimates of value, disposal cost, and decontamination cost. The economic evaluation methods documented here, the simple cost comparisons presented, and the data provided as a supplement, should provide a foundation for D&D decisions for Building 865, as well as for similar D&D tasks at RFP and at other sites.

  11. Reduced waste generation technical work plan

    SciTech Connect (OSTI)

    Not Available

    1987-05-01

    The United States Department of Energy has established policies for avoiding plutonium losses to the waste streams and minimizing the generation of wastes produced at its nuclear facilities. This policy is evidenced in DOE Order 5820.2, which states Technical and administrative controls shall be directed towards reducing the gross volume of TRU waste generated and the amount of radioactivity in such waste.'' To comply with the DOE directive, the Defense Transuranic Waste Program (DTWP) supports and provides funding for specific research and development tasks at the various DOE sites to reduce the generation of waste. This document has been prepared to give an overview of current and past Reduced Waste Generation task activities which are to be based on technical and cost/benefit factors. The document is updated annually, or as needed, to reflect the status of program direction. Reduced Waste Generation (RWG) tasks encompass a wide range of goals which are basically oriented toward (1) avoiding the generation of waste, (2) changing processes or operations to reduce waste, (3) converting TRU waste into LLW by sorting or decontamination, and (4) reducing volumes through operations such as incineration or compaction.

  12. Reduced waste generation, FY 1986

    SciTech Connect (OSTI)

    Not Available

    1986-02-01

    The United States Department of Energy is committed to the principles of minimizing the quantity and transuranic content of its transuranium (TRU) waste being generated at its nuclear facilities. The reasons are to reduce costs associated with waste handling and disposal, and also to reduce radiation exposure to workers and risk for radionuclide release to man and the environment. The purpose of this document is to provide the USDOE with a plan of research and development tasks for waste minimization, and is prepared so as to provide the maximum impact on volumes based on cost/benefit factors. The document is to be updated annually or as needed to reflect current and future tasks. The Reduced Waste Generation (RWG) tasks encompass a wide range of activities with the principal goals of (1) preventing the generation of waste and (2) converting TRU waste into low-level wastes (LLW) by sorting or decontamination. Concepts for reducing the volume such as in incineration and compaction are considered within the discipline of Reduced Waste Generation, but are considered as somewhat developed technology with only a need for implementation. 33 refs.

  13. Low-level radioactive-waste compacts. Status report as of July 1982

    SciTech Connect (OSTI)

    Not Available

    1982-07-01

    The Low-Level Radioactive Waste Policy Act (P.L. 96-573), enacted in December 1980, established as federal policy that states take responsibility for providing disposal capacity for low-level radioactive waste (LLW) generated within their borders, except for defense waste and Federal R and D. At the request of Senator James A. McClure, Chairman of the Senate Committee on Energy and Natural Resources, DOE has documented the progress of states individually and collectively in fulfilling their responsibilities under the Public Law. Regionalization through formation of low-level waste compacts has been the primary vehicle by which many states are assuming this responsibility. To date seven low-level waste compacts have been drafted and six have been enacted by state legislatures or ratified by a governor. As indicated by national progress to date, DOE considers the task of compacting achievable by the January 1, 1986, exclusionary date set in law, although several states and NRC questioned this.

  14. On-Site Decontamination System for Liquid Low Level Radioactive Waste - 13010

    SciTech Connect (OSTI)

    OSMANLIOGLU, Ahmet Erdal [Cekmece Nuclear Research and Training Center, Kucukcekmece Istanbul (Turkey)] [Cekmece Nuclear Research and Training Center, Kucukcekmece Istanbul (Turkey)

    2013-07-01

    This study is based on an evaluation of purification methods for liquid low-level radioactive waste (LLLW) by using natural zeolite. Generally the volume of liquid low-level waste is relatively large and the specific activity is rather low when compared to other radioactive waste types. In this study, a pilot scale column was used with natural zeolite as an ion exchanger media. Decontamination and minimization of LLLW especially at the generation site decrease operational cost in waste management operations. Portable pilot scale column was constructed for decontamination of LLW on site. Effect of temperature on the radionuclide adsorption of the zeolite was determined to optimize the waste solution temperature for the plant scale operations. In addition, effect of pH on the radionuclide uptake of the zeolite column was determined to optimize the waste solution pH for the plant scale operations. The advantages of this method used for the processing of LLLW are discussed in this paper. (authors)

  15. Waste characterization for the F/H Effluent Treatment Facility in support of waste certification

    SciTech Connect (OSTI)

    Brown, D.F.

    1994-10-17

    The Waste Acceptance Criteria (WAC) procedures define the rules concerning packages of solid Low Level Waste (LLW) that are sent to the E-area vaults (EAV). The WACs tabulate the quantities of 22 radionuclides that require manifesting in waste packages destined for each type of vault. These quantities are called the Package Administrative Criteria (PAC). If a waste package exceeds the PAC for any radionuclide in a given vault, then specific permission is needed to send to that vault. To avoid reporting insignificant quantities of the 22 listed radionuclides, the WAC defines the Minimum Reportable Quantity (MRQ) of each radionuclide as 1/1000th of the PAC. If a waste package contains less than the MRQ of a particular radionuclide, then the package`s manifest will list that radionuclide as zero. At least one radionuclide has to be reported, even if all are below the MRQ. The WAC requires that the waste no be ``hazardous`` as defined by SCDHEC/EPA regulations and also lists several miscellaneous physical/chemical requirements for the packages. This report evaluates the solid wastes generated within the F/H Effluent Treatment Facility (ETF) for potential impacts on waste certification.

  16. Assessment of commercially available ion exchange materials for cesium removal from highly alkaline wastes

    SciTech Connect (OSTI)

    Brooks, K.P.; Kim, A.Y.; Kurath, D.E.

    1996-04-01

    Approximately 61 million gallons of nuclear waste generated in plutonium production, radionuclide removal campaigns, and research and development activities is stored on the Department of Energy`s Hanford Site, near Richland, Washington. Although the pretreatment process and disposal requirements are still being defined, most pretreatment scenarios include removal of cesium from the aqueous streams. In many cases, after cesium is removed, the dissolved salt cakes and supernates can be disposed of as LLW. Ion exchange has been a leading candidate for this separation. Ion exchange systems have the advantage of simplicity of equipment and operation and provide many theoretical stages in a small space. The organic ion exchange material Duolite{trademark} CS-100 has been selected as the baseline exchanger for conceptual design of the Initial Pretreatment Module (IPM). Use of CS-100 was chosen because it is considered a conservative, technologically feasible approach. During FY 96, final resin down-selection will occur for IPM Title 1 design. Alternate ion exchange materials for cesium exchange will be considered at that time. The purpose of this report is to conduct a search for commercially available ion exchange materials which could potentially replace CS-100. This report will provide where possible a comparison of these resin in their ability to remove low concentrations of cesium from highly alkaline solutions. Materials which show promise can be studied further, while less encouraging resins can be eliminated from consideration.

  17. The Hubble constant from $^{56}$Co-powered Nebular Candles

    E-Print Network [OSTI]

    Pilar Ruiz-Lapuente

    1996-04-10

    Type Ia supernovae (SNe Ia), produced by the thermonuclear explosion of white dwarf (WD) stars, are used here to derive extragalactic distances and an estimate of the Hubble constant from their emission signatures at late phases ({\\it Nebular SNe Ia Method}, NSM). The method, first developed in Ruiz--Lapuente \\& Lucy (1992), makes use here of an improved modeling of the forbidden line emission at late phases. Hydrodynamic models of the explosion of WDs of different masses, both sub--Chandrasekhar and Chandrasekhar, provide the basis for comparison with the observations. It is shown that it is possible to probe the overall density structure of the ejecta and the mass of the exploding WD by the effect that the electron density profile has in shaping the forbidden line emission of the iron ions, and that a robust diagnostic of the mass of the exploding WD can be obtained. Cosmic distance scale can thus be related to basic diagnostics of excitation of iron lines. Once the most adequate model is selected, comparison of the predicted line emission at these phases with the observed spectra gives an internal estimate of both the reddening and the distance to the SNe Ia. The results presented here favor denser models than those corresponding to sub--Chandrasekhar explosions. From a sample of seven SNe Ia in Leo, Virgo, Fornax and beyond, a value of the Hubble constant $H_{0} = 68 \\ \\pm 6\\ (stat) \\pm 7\\ (syst)\\ km\\ s^{-1}\\ Mpc^{-1}$ is derived. The depth of the Virgo cluster is found to be large, ranging from 13 to 23 Mpc at least. If NGC 4526 traces well the core of the Virgo Cluster, then the latter is located at $16\\pm 2 \\ Mpc$. The galaxy NGC 3267 in Leo appears to be located at 9.8 $\\pm$ 1.5 Mpc.

  18. Constraints on shallow {sup 56}Ni from the early light curves of type Ia supernovae

    SciTech Connect (OSTI)

    Piro, Anthony L. [Theoretical Astrophysics, California Institute of Technology, 1200 E California Boulevard, M/C 350-17, Pasadena, CA 91125 (United States); Nakar, Ehud, E-mail: piro@caltech.edu [Raymond and Beverly Sackler School of Physics and Astronomy, Tel Aviv University, Tel Aviv 69978 (Israel)

    2014-03-20

    Ongoing transient surveys are presenting an unprecedented account of the rising light curves of Type Ia supernovae (SNe Ia). This early emission probes the shallowest layers of the exploding white dwarf (WD), which can provide constraints on the progenitor star and the properties of the explosive burning. We use semianalytic models of radioactively powered rising light curves to analyze these observations. As we have summarized in previous work, the main limiting factor in determining the surface distribution of {sup 56}Ni is the lack of an unambiguously identified time of explosion, as would be provided by detection of shock breakout or shock-heated cooling. Without this the SN may in principle exhibit a 'dark phase' for a few hours to days, where the only emission is from shock-heated cooling that is too dim to be detected. We show that by assuming a theoretically motivated time-dependent velocity evolution, the explosion time can be better constrained, albeit with potential systematic uncertainties. This technique is used to infer the surface {sup 56}Ni distributions of three recent SNe Ia that were caught especially early in their rise. In all three we find fairly similar {sup 56}Ni distributions. Observations of SN 2011fe and SN 2012cg probe shallower depths than SN 2009ig, and in these two cases {sup 56}Ni is present merely ?10{sup 2} M {sub ?} from the WDs' surfaces. The uncertainty in this result is up to an order of magnitude given the difficulty of precisely constraining the explosion time. We also use our conclusions about the explosion times to reassess radius constraints for the progenitor of SN 2011fe, as well as discuss the roughly t {sup 2} power law that is inferred for many observed rising light curves.

  19. Early hydrodynamic evolution of a stellar collision

    SciTech Connect (OSTI)

    Kushnir, Doron; Katz, Boaz [Institute for Advanced Study, Einstein Drive, Princeton, NJ 08540 (United States)

    2014-04-20

    The early phase of the hydrodynamic evolution following the collision of two stars is analyzed. Two strong shocks propagate from the contact surface and move toward the center of each star at a velocity that is a small fraction of the velocity of the approaching stars. The shocked region near the contact surface has a planar symmetry and a uniform pressure. The density vanishes at the (Lagrangian) surface of contact, and the speed of sound diverges there. The temperature, however, reaches a finite value, since as the density vanishes, the finite pressure is radiation dominated. For carbon-oxygen white dwarf (CO WD) collisions, this temperature is too low for any appreciable nuclear burning shortly after the collision, which allows for a significant fraction of the mass to be highly compressed to the density required for efficient {sup 56}Ni production in the detonation wave that follows. This property is crucial for the viability of collisions of typical CO WD as progenitors of type Ia supernovae, since otherwise only massive (>0.9 M {sub ?}) CO WDs would have led to such explosions (as required by all other progenitor models). The divergence of the speed of sound limits numerical studies of stellar collisions, as it makes convergence tests exceedingly expensive unless dedicated schemes are used. We provide a new one-dimensional Lagrangian numerical scheme to achieve this. A self-similar planar solution is derived for zero-impact parameter collisions between two identical stars, under some simplifying assumptions (including a power-law density profile), which is the planar version of previous piston problems that were studied in cylindrical and spherical symmetries.

  20. THE SOLAR NEIGHBORHOOD. XXV. DISCOVERY OF NEW PROPER MOTION STARS WITH 0.''40 yr{sup -1} > {mu} {>=} 0.''18 yr{sup -1} BETWEEN DECLINATIONS -47{sup 0} AND 00{sup 0}

    SciTech Connect (OSTI)

    Boyd, Mark R.; Winters, Jennifer G.; Henry, Todd J.; Jao, Wei-Chun; Finch, Charlie T.; Subasavage, John P.; Hambly, Nigel C. E-mail: winters@chara.gsu.edu E-mail: jao@chara.gsu.edu E-mail: jsubasavage@ctio.noao.edu

    2011-07-15

    We present 2817 new southern proper motion systems with 0.''40 yr{sup -1} > {mu} {>=} 0.''18 yr{sup -1} and declination between -47{sup 0} and 00{sup 0}. This is a continuation of the SuperCOSMOS-RECONS (SCR) proper motion searches of the southern sky. We use the same photometric relations as previous searches to provide distance estimates based on the assumption that the objects are single main-sequence stars. We find 79 new red dwarf systems predicted to be within 25 pc, including a few new components of previously known systems. Two systems-SCR 1731-2452 at 9.5 pc and SCR 1746-3214 at 9.9 pc-are anticipated to be within 10 pc. We also find 23 new white dwarf (WD) candidates with distance estimates of 15-66 pc, as well as 360 new red subdwarf candidates. With this search, we complete the SCR sweep of the southern sky for stars with {mu} {>=} 0.''18 yr{sup -1} and R{sub 59F} {<=} 16.5, resulting in a total of 5042 objects in 4724 previously unreported proper motion systems. Here we provide selected comprehensive lists from our SCR proper motion search to date, including 152 red dwarf systems estimated to be within 25 pc (9 within 10 pc), 46 WDs (10 within 25 pc), and 598 subdwarf candidates. The results of this search suggest that there are more nearby systems to be found at fainter magnitudes and lower proper motion limits than those probed so far.

  1. The FBXW7 {beta}-form is suppressed in human glioma cells

    SciTech Connect (OSTI)

    Gu, Zhaodi; Inomata, Kenichi; Ishizawa, Kota; Horii, Akira . E-mail: horii@mail.tains.tohoku.ac.jp

    2007-03-23

    FBXW7 (F-box and WD40 domain protein 7) is an F-box protein with 7 tandem WDs (tryptophan-aspartic acid) that functions as a phosphoepitope-specific substrate recognition component of SCF (Skp1-Cul1-F-box protein) ubiquitin ligases and catalyzes the ubiquitination of proteins promoting cell proliferation, such as CCNE1, MYC, AURKA, NOTCH1, and JUN, which are frequently activated in a wide range of human cancers. FBXW7 is a candidate tumor suppressor, and mutations have been reported in some human tumors. In this study, we analyzed 84 human tumor cell lines in search for genetic alterations of FBXW7, as well as mRNA and protein expressional changes, and compared them with expression levels of the CCNE1, MYC, and AURKA proteins. We found a novel nonsense mutation in a colon cancer cell line SCC and confirmed the missense mutations in SKOV3, an ovarian cancer cell line, and LoVo, a colon cancer cell line. Moreover, suppressed expression of FBXW7 accompanied by activation of the target proteins were observed in ovarian, colon, endometrial, gastric, and prostate cancers. It is notable that highly suppressed mRNA expression of the FBXW7 {beta}-form was found in all the human glioma cell lines analyzed; enhanced expressions of CCNE1, MYC, and AURKA were observed in these cells. Our present results imply that FBXW7 plays a pivotal role in many tissues by controlling the amount of cell cycle promoter proteins and that dysfunction of this protein is one of the essential steps in carcinogenesis in multiple organs.

  2. Type Ia Supernovae: Progenitors and Evolution with Redshift

    E-Print Network [OSTI]

    Ken'ichi Nomoto; Hideyuki Umeda; Chiaki Kobayashi; Izumi Hachisu; Mariko Kato; Takuji Tsujimoto

    2000-03-09

    Relatively uniform light curves and spectral evolution of Type Ia supernovae (SNe Ia) have led to the use of SNe Ia as a ``standard candle'' to determine cosmological parameters. Whether a statistically significant value of the cosmological constant can be obtained depends on whether the peak luminosities of SNe Ia are sufficiently free from the effects of cosmic and galactic evolutions. Here we first review the single degenerate scenario for the Chandrasekhar mass white dwarf (WD) models of SNe Ia. We identify the progenitor's evolution and population with two channels: (1) the WD+RG (red-giant) and (2) the WD+MS (near main-sequence He-rich star) channels. In these channels, the strong wind from accreting WDs plays a key role, which yields important age and metallicity effects on the evolution. We then address the questions whether the nature of SNe Ia depends systematically on environmental properties such as metallicity and age of the progenitor system and whether significant evolutionary effects exist. We suggest that the variation of the carbon mass fraction $X$(C) in the C+O WD (or the variation of the initial WD mass) causes the diversity of the brightness of SNe Ia. This model can explain the observed dependences of SNe Ia brightness on the galaxy types and the distance from the galactic center. Finally, applying the metallicity effect on the evolution of SN Ia progenitors, we make a prediction of the cosmic supernova rate history as a composite of the supernova rates in different types of galaxies.

  3. OPERATIONAL AND COMPOSITIONAL FACTORS THAT AFFECT THE PERFORMANCE PROPERTIES OF ARP/MCU SALTSTONE GROUT

    SciTech Connect (OSTI)

    Reigel, M.; Edwards, T.; Pickenheim, B.

    2012-02-15

    The Saltstone Production Facility (SPF) receives low level waste (LLW) salt solution from Tank 50H for treatment and disposal. Tank 50H receives transfers from the Effluent Treatment Project (ETP), the H-Canyon General Purpose Evaporator, and the Actinide Removal Process/Modular Caustic Side Solvent Extraction Unit (ARP/MCU) Decontaminated Salt Solution Hold Tank (DSS-HT). At the SPF, the LLW is mixed with premix (a cementitious mixture of portland cement (PC), blast furnace slag (BFS) and Class F fly ash (FA)) in a Readco mixer to produce fresh (uncured) saltstone that is transferred to the Saltstone Disposal Facility (SDF) vaults. The saltstone formulation (mix design) must produce a grout waste form that meets both placement and performance properties. In previous simulated saltstone studies, multiple compositional factors were identified that drive the performance properties of saltstone made from the projected ARP/MCU salt solution. This composition was selected as salt solution simulant since ARP/MCU is the primary influent into Tank 50H. The primary performance property investigated was hydraulic conductivity since it is a variable input property to the saltstone Performance Assessment (PA) transport model. In addition, the porosity, also referred to as void structure, is another variable that impacts the PA response. In addition, Young's modulus and cured density are other performance properties analyzed in this report; however they are indicators of the performance of saltstone and not direct inputs into the PA. The data from previous studies showed that the largest impact on the performance properties of saltstone was due to curing temperature, followed by aluminate concentration in the salt solution, water to premix ratio and premix composition. However, due to the scope of the previous studies, only a few mixes were cured and analyzed at higher temperatures. The samples cured at 60 C had an increased hydraulic conductivity of approximately 600 times that of the sample cured at room temperature. The hydration reactions initiated during the mixing of the premix and salt solution continue during the curing period in the vaults to produce the hardened waste form product. The heat generated from exothermic hydration reactions results in a temperature increase in the vaults that depends on the composition of the decontaminated salt solution being dispositioned, the grout formulation (mix design) and the pour frequency and volume. This heat generation is a contributing factor to the temperature increase in the vaults that leads to an increased cure temperature for the grout. This report will further investigate the impact of curing temperature on saltstone performance properties (hydraulic conductivity, Young's modulus, porosity, etc.) over a range of aluminate concentration, water to premix (w/p) ratio and weight percent fly ash in the premix processed at the SPF. The three curing temperatures selected for this study were chosen to provide data at fixed cure temperatures that represent measured temperatures in the SDF vaults. This does not represent the conditions in the vault where the temperature of the saltstone is continually changing with time. For example, it may take several days for the saltstone to reach 60 C at a given elevation. Previous results demonstrated that the rates at which a selected curing temperature is reached affect the performance properties. The approach taken in this task, a rapid increase to the curing temperature, may be conservative with respect to decreased performance. Nevertheless, the data will provide a basis from which to determine the impact of curing temperature on saltstone performance as a function of key variables. A statistical evaluation of the results for these mixes will be performed to provide the range, and associated uncertainties, of hydraulic conductivity and other properties over this factor space.

  4. An ideal sealed source life-cycle

    SciTech Connect (OSTI)

    Tompkins, Joseph Andrew [Los Alamos National Laboratory

    2009-01-01

    In the last 40 years, barriers to compliant and timely disposition of radioactive sealed sources have become apparent. The story starts with the explosive growth of nuclear gauging technologies in the 1960s. Dozens of companies in the US manufactured sources and many more created nuclear solutions to industrial gauging problems. Today they do not yet know how many Cat 1, 2, or 3 sources there are in the US. There are, at minimum, tens of thousands of sources, perhaps hundreds of thousands of sources. Affordable transportation solutions to consolidate all of these sources and disposition pathways for these sources do not exist. The root problem seems to be a lack of necessary regulatory framework that has allowed all of these problems to accumulate with no national plan for solving the problem. In the 1960s, Pu-238 displaced Pu-239 for most neutron and alpha source applications. In the 1970s, the availability of inexpensive Am-241 resulted in a proliferation of low energy gamma sources used in nuclear gauging, well logging, pacemakers, and X-ray fluorescence applications for example. In the 1980s, rapid expansion of worldwide petroleum exploration resulted in the expansion of Am-241 sources into international locations. Improvements of technology and regulation resulted in a change in isotopic distribution as Am-241 made Pu-239 and Pu-238 obsolete. Many early nuclear gauge technologies have been made obsolete as they were replaced by non-nuclear technoogies. With uncertainties in source end of life disposition and increased requirements for sealed source security, nuclear gauging technology is the last choice for modern process engineering gauging solutions. Over the same period, much was learned about licensing LLW disposition facilities as evident by the closure of early disposition facilities like Maxey Flats. The current difficulties in sealed source disposition start with adoption of the NLLW policy act of 1985, which created the state LLW compact system they we have today. This regulation created a new regulatory framework seen as promising at the time. However, now they recognize that, despite the good intentions, the NIJWP/85 has not solved any source disposition problems. The answer to these sealed source disposition problems is to adopt a philosophy to correct these regulatory issues, determine an interim solution, execute that solution until there is a minimal backlog of sources to deal with, and then let the mechanisms they have created solve this problem into the foreseeable future. The primary philosophical tenet of the ideal sealed source life cycle follows. You do not allow the creation (or importation) of any source whose use cannot be justified, which cannot be affordably shipped, or that does not have a well-delinated and affordable disposition pathway. The path forward dictates that we fix the problem by embracing the Ideal Source Life cycle. In figure 1, we can see some of the elements of the ideal source life cycle. The life cycle is broken down into four portions, manufacture, use, consolidation, and disposition. These four arbitrary elements allow them to focus on the ideal life cycle phases that every source should go through between manufacture and final disposition. As we examine the various phases of the sealed source life cycle, they pick specific examples and explore the adoption of the ideal life cycle model.

  5. Revision of the Branch Technical Position on Concentration Averaging and Encapsulation - 12510

    SciTech Connect (OSTI)

    Heath, Maurice; Kennedy, James E.; Ridge, Christianne; Lowman, Donald [U.S. NRC, Washington, DC, 20555-0001 (United States); Cochran, John [Sandia National Laboratory (United States)

    2012-07-01

    The U.S. Nuclear Regulatory Commission (NRC) regulation governing low-level waste (LLW) disposal, 'Licensing Requirements for Land Disposal of Radioactive Waste', 10 CFR Part 61, establishes a waste classification system based on the concentration of specific radionuclides contained in the waste. The regulation also states, at 10 CFR 61.55(a)(8), that, 'the concentration of a radionuclide (in waste) may be averaged over the volume of the waste, or weight of the waste if the units are expressed as nanocuries per gram'. The NRC's Branch Technical Position on Concentration Averaging and Encapsulation provides guidance on averaging radionuclide concentrations in waste under 10 CFR 61.55(a)(8) when classifying waste for disposal. In 2007, the NRC staff proposed to revise the Branch Technical Position on Concentration Averaging and Encapsulation. The Branch Technical Position on Concentration Averaging and Encapsulation is an NRC guidance document for averaging and classifying wastes under 10 CFR 61. The Branch Technical Position on Concentration Averaging and Encapsulation is used by nuclear power plants (NPPs) licensees and sealed source users, among others. In addition, three of the four U.S. LLW disposal facility operators are required to honor the Branch Technical Position on Concentration Averaging and Encapsulation as a licensing condition. In 2010, the Commission directed the staff to develop guidance regarding large scale blending of similar homogenous waste types, as described in SECY-10-0043 as part of its Branch Technical Position on Concentration Averaging and Encapsulation revision. The Commission is improving the regulatory approach used in the Branch Technical Position on Concentration Averaging and Encapsulation by moving towards a making it more risk-informed and performance-based approach, which is more consistent with the agency's regulatory policies. Among the improvements to the Branch Technical Position on Concentration Averaging and Encapsulation are more risk-informed limits for the sizes of sealed sources for safe disposal. Using more realistic intruder exposure scenarios, the suggested limits for Class B and C waste disposal of sealed sources, particularly Cs-137 and Co-60, have been increased. These suggested changes, and others in the Branch Technical Position on Concentration Averaging and Encapsulation, if adopted by Agreement States, have the potential to eliminate numerous orphan sources (i.e., sources that currently have no disposal pathway) that are now being stored. Permanent disposal of these sources, rather than temporary storage, will help reduce safety and security risks. The revised Branch Technical Position on Concentration Averaging and Encapsulation has an alternative approach section which provides flexibility to generators and processors, while also ensuring that intruder protection will be maintained. Alternative approaches provide flexibility by allowing for consideration of likelihood of intrusion, the possibility of averaging over larger volumes and allowing for disposal of large activity sources. The revision has improved the organization of the Branch Technical Position on Concentration Averaging and Encapsulation, improved its clarity, better documented the bases for positions, and made the positions more risk informed while also maintaining protection for intruder as required by 10 CFR Part 61. (authors)

  6. Closeout of the Melton Valley Completion Project at the Oak Ridge National Laboratory

    SciTech Connect (OSTI)

    Bonilla, R.; Johnson, Ch. [Bechtel Jacobs Company, LLC, Oak Ridge, TN (United States); Skinner, R. [U.S. DOE, Oak Ridge Operations Office, Oak Ridge, TN (United States); Adams, V. [U.S. DOE, Office of Groundwater and Soil Remediation, Washington, DC (United States)

    2008-07-01

    The U.S. Department of Energy DOE Order 413.3A (Program and Project Management for the Acquisition of Capital Assets) identifies major milestones in a project life cycle that require approval upon achievement, including Critical Decision-4 (CD-4), the project completion milestone. A CD-4 document is required for all DOE projects in accordance with DOE Order 413.3A, Program and Project Management for Acquisition of Capital Assets. A conditional CD-4 report was prepared for the Melton Valley Completion Project (MVCP) in order to document the completion of the remedial action in the Melton Valley watershed. Approval of the MVCP CD-4 was 'conditional' pending final resolution of a small quantity of unexpected pyrophoric material, with no current disposition pathway, remaining in one of the waste trenches that were to be remediated as part of the completion project. This paper will provide an overview of the MVCP remediation work and the process successfully used to demonstrate closeout of a major CERCLA project in accordance with DOE requirements. In summary: The MV ROD was planned to be completed in 12 years; however, under the ACP the remedial activities were completed 6 years ahead of schedule (September 2006). Highlights of the remedial actions include: - Completion of remedial action activities at 219 release sites identified in the MV ROD. - Construction of 58 hectares of multilayer caps for SWSA 4, SWSA 5, SWSA 6, Pits 2, 3 and 4 and Trenches 5, 6, and 7; - Complete excavation, retrieval and over-pack of 204 casks, 8 boxes, and 530 m{sup 3} of loose waste from the TRU waste retrieval project; - Complete demolition and disposition of approx. 557 m{sup 2} of various buildings such as HRE ancillary facilities, NHF, Liquid LLW pumping stations, 7841 Scrap Yard, misc. storage buildings, and well P and A; - Removal of LLW sludges and stabilization of T-1, T-2, and HFIR tanks; - In situ grouting/stabilization of approx. 12 km of inactive waste pipelines; - Excavation and disposition of approx. 38,000 m{sup 3} of soil from HFIR impoundments, HRE cryogenic pond, IHP and contaminated soil from miscellaneous leak sites; - Completion of in-situ grouting of LLLW matrix at Trenches 5 and 7. The CD-4 document to demonstrate attainment of the project completion milestone was prepared to document close-out activities. Approval of the MVCP CD-4 was obtained but deemed 'conditional' pending final resolution of a small quantity of unexpected pyrophoric material, with no current disposition pathway, remaining in Trench 13 of the 22 Trench Area in SWSA 5N. DOE is currently evaluating options for the final disposition of this material and has committed to resolution of the issue by the end of Fiscal Year 2008. At that time, all aspects of the Melton Valley Completion Project will be complete. (authors)

  7. Radiation Stability of Benzyl Tributyl Ammonium Chloride Towards Technetium-99 Extraction

    SciTech Connect (OSTI)

    Jared Horkley; Audrey Roman; Keri Campbell; Ana Nunez; Amparo Espartero

    2013-02-01

    A closed nuclear fuel cycle combining new separation technologies along with generation III and generation IV reactors is a promising way to achieve a sustainable energy supply. But it is important to keep in mind that future recycling processes of used nuclear fuel (UNF) must minimize wastes, improve partitioning process, and integrate waste considerations into processes. New separation processes are being developed worldwide to complement the actual industrialized PUREX process which selectively separates U(VI) and Pu(IV) from the raffinate. As an example, low nitric acid concentration in the aqueous phase of a UREX based process will co-extract U(VI) and Tc(VII) by tri-n-butyl phosphate (TBP). Technetium (Tc-99) is recognized to be one of the most abundant, long-lived radiotoxic isotopes in UNF (half-life, t1/2 = 2.13 105 years), and as such, it is targeted in UNF separation strategies for isolation and encapsulation in solid waste forms for final disposal in a nuclear waste repository. Immobilization of Tc-99 by a durable solid waste form is a challenge, and its fate in new advanced technology processes is of importance. It is essential to be able to quantify and locate 1) its occurrence in any new developed flow sheets, 2) its chemical form in the individual phases of a process, 3) its potential quantitative transfer in any waste streams, and consequently, 4) its quantitative separation for either potential transmutation to Ru-100 or isolation and encapsulation in solid waste forms for ultimate disposal. Furthermore, as a result of an U(VI)-Tc(VII) co-extraction in a UREX-based process, Tc(VII) could be found in low level waste (LLW) streams. There is a need for the development of new extraction systems that would selectively extract Tc-99 from LLW streams and concentrate it for feed into high level waste (HLW) for either Tc-99 immobilization in metallic waste forms (Tc-Zr alloys), and/or borosilicate-based waste glass. Studies have been launched to investigate the suitability of new macrocompounds such as crown-ethers, aza-crown ethers, and resorcinarenes for the selective extraction of Tc-99 from nitric acid solutions. The selectivity of the ligand is important in evaluating potential separation processes and also the radiation stability of the molecule is essential for minimization of waste and radiolysis products. In this paper, we are reporting the extraction of TcO4- by benzyltributyl ammonium chloride (BTBA). Experimental efforts were focused on determining the best extraction conditions by varying the ligands matrix conditions and concentration, as well as varying the organic phase composition (i.e., diluent variation). Furthermore, the ligand has been investigated for radiation stability. The ?-irradiation was performed on the neat organic phases containing the ligand at different absorbed doses to a maximum of 200 kGy using external Co-60 source. Post-irradiation solvent extraction measurements will be discussed.

  8. On-site Destruction of Radioactive Oily Wastes Using Adsorption Coupled with Electrochemical Regeneration - 12221

    SciTech Connect (OSTI)

    Brown, N.W.; Wickenden, D.A.; Roberts, E.P.L.

    2012-07-01

    Arvia{sup R}, working with Magnox Ltd, has developed the technology of adsorption coupled with electrochemical regeneration for the degradation of orphan radioactive oil wastes. The process results in the complete destruction of the organic phase where the radioactivity is transferred to liquid and solid secondary wastes that can then be processed using existing authorised on-site waste-treatment facilities.. Following on from successful laboratory and pilot scale trials, a full scale, site based demonstrator unit was commissioned at the Magnox Trawsfynydd decommissioning site to destroy 10 l of LLW and ILW radioactive oils. Over 99% of the emulsified oil was removed and destroyed with the majority of activity (80 - 90%) being transferred to the aqueous phase. Secondary wastes were disposed of via existing routes with the majority being disposed of via the sites active effluent treatment plant. The regeneration energy required to destroy a litre of oil was 42.5 kWh/l oil. This on-site treatment approach eliminates the risks and cost associated with transporting the active waste oils off site for incineration or other treatment. The Arvia{sup R} process of adsorption coupled with electrochemical regeneration has successfully demonstrated the removal and destruction of LLW and ILW radioactive oils on a nuclear site. Over 99.9% of the emulsified oil was removed, with the majority of the radioactive species transferred to the aqueous, supernate, phase (typically 80 - 90 %). The exception to this is Cs-137 which appears to be more evenly distributed, with 43% associated with the liquid phase and 33 % with the Nyex, the remainder associated with the electrode bed. The situation with Plutonium may be similar, but this requires confirmation, hence further work is underway to understand the full nature of the electrode bed radioactive burden and its distribution within the body of the electrodes. - Tritium gaseous discharges were negligible; hence no off-gas treatment before direct discharge to atmosphere is necessary. All secondary wastes were suitable for disposal using existing disposal routes, with the majority of the activity being successfully discharged as active water via the site active drains. - Oil destruction was achieved at a rate of 28.2 ml/hr using a regeneration energy of 42.5 kWh/l oil. The treatment of different active and non-active oils was achieved using the same operating parameters, providing strong evidence that the process is robust and will treat a wide range of oils, organic wastes and additives. - Currently the design of a plant capable of processing 1000 ml/hr is being established in discussion with Magnox Ltd. The plant will run automatically with little operator attention and so process between 5-8 m{sup 3} of ILW oil per annum. (authors)

  9. Closure Strategy Nevada Test Site Area 5 Radioactive Waste Management Site

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2007-03-01

    This paper presents an overview of the strategy for closure of part of the Area 5 Radioactive Waste Management Site (RWMS) at the Nevada Test Site (NTS), which is about 65 miles northwest of Las Vegas, Nevada (Figure 1). The Area 5 RWMS is in the northern part of Frenchman Flat, approximately 14 miles north of Mercury. The Area 5 RWMS encompasses 732 acres subdivided into quadrants, and is bounded by a 1,000-foot (ft)-wide buffer zone. The northwest and southwest quadrants have not been developed. The northeast and southeast quadrants have been used for disposal of unclassified low-level radioactive waste (LLW) and indefinite storage of classified materials. This paper focuses on closure of the 38 waste disposal and classified material storage units within the southeast quadrant of the Area 5 RWMS, called the ''92-Acre Area''. The U.S Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (NNSA/NSO) is currently planning to close the 92-Acre Area by 2011. Closure planning for this site must take into account the regulatory requirements for a diversity of waste streams, disposal and storage configurations, disposal history, and site conditions. For ease of discussion, the 92-Acre Area has been subdivided into six closure units defined by waste type, location, and similarity in regulatory requirements. Each of the closure units contains one or more waste disposal units; waste disposal units are also called waste disposal cells. The paper provides a brief background of the Area 5 RWMS, identifies key closure issues for the 92-Acre Area, recommends actions to address the issues, and provides the National Security Technologies, LLC (NSTec), schedule for closure.

  10. Near-Field Hydrology Data Package for the Integrated Disposal Facility 2005 Performance Assessment

    SciTech Connect (OSTI)

    Meyer, Philip D.; Saripalli, Prasad; Freedman, Vicky L.

    2004-06-25

    CH2MHill Hanford Group, Inc. (CHG) is designing and assessing the performance of an Integrated Disposal Facility (IDF) to receive immobilized low-activity waste (ILAW), Low-Level and Mixed Low-Level Wastes (LLW/MLLW), and the Waste Treatment Plant (WTP) melters used to vitrify the ILAW. The IDF Performance Assessment (PA) assesses the performance of the disposal facility to provide a reasonable expectation that the disposal of the waste is protective of the general public, groundwater resources, air resources, surface water resources, and inadvertent intruders. The PA requires prediction of contaminant migration from the facilities, which is expected to occur primarily via the movement of water through the facilities and the consequent transport of dissolved contaminants in the pore water of the vadose zone. Pacific Northwest National Laboratory (PNNL) assists CHG in its performance assessment activities. One of PNNLs tasks is to provide estimates of the physical, hydraulic, and transport properties of the materials comprising the disposal facilities and the disturbed region around them. These materials are referred to as the near-field materials. Their properties are expressed as parameters of constitutive models used in simulations of subsurface flow and transport. In addition to the best-estimate parameter values, information on uncertainty in the parameter values and estimates of the changes in parameter values over time are required to complete the PA. These parameter estimates and information were previously presented in a report prepared for the 2001 ILAW PA. This report updates the parameter estimates for the 2005 IDF PA using additional information and data collected since publication of the earlier report.

  11. Integrated Closure and Monitoring Plan for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada Test Site

    SciTech Connect (OSTI)

    S. E. Rawlinson

    2001-09-01

    Bechtel Nevada (BN) manages two low-level Radioactive Waste Management Sites (RWMSs) (one site is in Area 3 and the other is in Area 5) at the Nevada Test Site (NTS) for the U.S. Department of Energy's (DOE's) National Nuclear Security Administration Nevada Operations Office (NNSA/NV). The current DOE Order governing management of radioactive waste is 435.1. Associated with DOE Order 435.1 is a Manual (DOE M 435.1-1) and Guidance (DOE G 435.1-1). The Manual and Guidance specify that preliminary closure and monitoring plans for a low-level waste (LLW) management facility be developed and initially submitted with the Performance Assessment (PA) and Composite Analysis (CA) for that facility. The Manual and Guidance, and the Disposal Authorization Statement (DAS) issued for the Area 3 RWMS further specify that the preliminary closure and monitoring plans be updated within one year following issuance of a DAS. This Integrated Closure and Monitoring Plan (ICMP) fulfills both requirements. Additional updates will be conducted every third year hereafter. This document is an integrated plan for closing and monitoring both RWMSs, and is based on guidance issued in 1999 by the DOE for developing closure plans. The plan does not follow the format suggested by the DOE guidance in order to better accommodate differences between the two RWMSs, especially in terms of operations and site characteristics. The modification reduces redundancy and provides a smoother progression of the discussion. The closure and monitoring plans were integrated because much of the information that would be included in individual plans is the same, and integration provides efficient presentation and program management. The ICMP identifies the regulatory requirements, describes the disposal sites and the physical environment where they are located, and defines the approach and schedule for both closing and monitoring the sites.

  12. The Swedish Program has Entered the Site Selection Phase

    SciTech Connect (OSTI)

    Nygards, P.; Hedman, T.; Eng, T.; Olsson, O.

    2003-02-25

    Facilities for intermediate storage of spent fuel and HLW and for final disposal of ILW and LLW together with a system for sea transportation have been in operation in Sweden for more then 15 years. To complete the ''back end system'' the remaining parts are to build facilities for encapsulation and final storage of spent fuel and HLW. The Swedish reference method for final disposal of spent fuel, KBS-3, is to encapsulate the fuel elements in copper canisters and dispose them in a deep geological repository. The Swedish program up to 2001 was focused on the establishment of general acceptance of the reference method for final storage and SKB's selection of candidate sites for a deep geological repository. In the end of year 2000 SKB presented a report as a base for a Government decision about the siting process. This report gave the background for the selection of three candidate sites. It also presented the program for geological surveys of the candidate sites as well as the background for the choice of the method for final disposal of spent nuclear fuel and HLW. In the end of 2001 the Swedish government endorsed the plan for the site selection phase and stated that the KBS-3 design of the repository shall be used as the planning base for the work. Permissions were also granted for the fieldwork from the municipalities of Forsmark and Oskarshamn where the candidate sites are located. Site investigations on these two sites started during 2002. The technical development and demonstration of the KBS 3-method is ongoing at the Dspv Hard Rock Laboratory and the Canister Laboratory. The goal for the coming five years period is to select the site for the repository and apply for licenses to construct and operate the facilities for encapsulation and final storage of spent fuel. The encapsulation plant and the repository are planned to be in operation around year 2015.

  13. Plutonium Equivalent Inventory for Belowground Radioactive Waste at the Los Alamos National Laboratory Technical Area 54, Area G Disposal Facility - Fiscal Year 2011

    SciTech Connect (OSTI)

    French, Sean B.; Shuman, Rob

    2012-04-18

    The Los Alamos National Laboratory (LANL) generates radioactive waste as a result of various activities. Many aspects of the management of this waste are conducted at Technical Area 54 (TA-54); Area G plays a key role in these management activities as the Laboratory's only disposal facility for low-level radioactive waste (LLW). Furthermore, Area G serves as a staging area for transuranic (TRU) waste that will be shipped to the Waste Isolation Pilot Plant for disposal. A portion of this TRU waste is retrievably stored in pits, trenches, and shafts. The radioactive waste disposed of or stored at Area G poses potential short- and long-term risks to workers at the disposal facility and to members of the public. These risks are directly proportional to the radionuclide inventories in the waste. The Area G performance assessment and composite analysis (LANL, 2008a) project long-term risks to members of the public; short-term risks to workers and members of the public, such as those posed by accidents, are addressed by the Area G Documented Safety Analysis (LANL, 2011a). The Documented Safety Analysis uses an inventory expressed in terms of plutonium-equivalent curies, referred to as the PE-Ci inventory, to estimate these risks. The Technical Safety Requirements for Technical Area 54, Area G (LANL, 2011b) establishes a belowground radioactive material limit that ensures the cumulative projected inventory authorized for the Area G site is not exceeded. The total belowground radioactive waste inventory limit established for Area G is 110,000 PE-Ci. The PE-Ci inventory is updated annually; this report presents the inventory prepared for 2011. The approach used to estimate the inventory is described in Section 2. The results of the analysis are presented in Section 3.

  14. Thirty-Year Solid Waste Generation Maximum and Minimum Forecast for SRS

    SciTech Connect (OSTI)

    Thomas, L.C.

    1994-10-01

    This report is the third phase (Phase III) of the Thirty-Year Solid Waste Generation Forecast for Facilities at the Savannah River Site (SRS). Phase I of the forecast, Thirty-Year Solid Waste Generation Forecast for Facilities at SRS, forecasts the yearly quantities of low-level waste (LLW), hazardous waste, mixed waste, and transuranic (TRU) wastes generated over the next 30 years by operations, decontamination and decommissioning and environmental restoration (ER) activities at the Savannah River Site. The Phase II report, Thirty-Year Solid Waste Generation Forecast by Treatability Group (U), provides a 30-year forecast by waste treatability group for operations, decontamination and decommissioning, and ER activities. In addition, a 30-year forecast by waste stream has been provided for operations in Appendix A of the Phase II report. The solid wastes stored or generated at SRS must be treated and disposed of in accordance with federal, state, and local laws and regulations. To evaluate, select, and justify the use of promising treatment technologies and to evaluate the potential impact to the environment, the generic waste categories described in the Phase I report were divided into smaller classifications with similar physical, chemical, and radiological characteristics. These smaller classifications, defined within the Phase II report as treatability groups, can then be used in the Waste Management Environmental Impact Statement process to evaluate treatment options. The waste generation forecasts in the Phase II report includes existing waste inventories. Existing waste inventories, which include waste streams from continuing operations and stored wastes from discontinued operations, were not included in the Phase I report. Maximum and minimum forecasts serve as upper and lower boundaries for waste generation. This report provides the maximum and minimum forecast by waste treatability group for operation, decontamination and decommissioning, and ER activities.

  15. Radioactive Waste Conditioning, Immobilisation, And Encapsulation Processes And Technologies: Overview And Advances (Chapter 7)

    SciTech Connect (OSTI)

    Jantzen, Carol M.; Lee, William E.; Ojovan, Michael I.

    2012-10-19

    The main immobilization technologies that are available commercially and have been demonstrated to be viable are cementation, bituminization, and vitrification. Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. Most of the nations that have generated HLW are immobilizing in either alkali borosilicate glass or alkali aluminophosphate glass. The exact compositions of nuclear waste glasses are tailored for easy preparation and melting, avoidance of glass-in-glass phase separation, avoidance of uncontrolled crystallization, and acceptable chemical durability, e.g., leach resistance. Glass has also been used to stabilize a variety of low level wastes (LLW) and mixed (radioactive and hazardous) low level wastes (MLLW) from other sources such as fuel rod cladding/decladding processes, chemical separations, radioactive sources, radioactive mill tailings, contaminated soils, medical research applications, and other commercial processes. The sources of radioactive waste generation are captured in other chapters in this book regarding the individual practices in various countries (legacy wastes, currently generated wastes, and future waste generation). Future waste generation is primarily driven by interest in sources of clean energy and this has led to an increased interest in advanced nuclear power production. The development of advanced wasteforms is a necessary component of the new nuclear power plant (NPP) flowsheets. Therefore, advanced nuclear wasteforms are being designed for robust disposal strategies. A brief summary is given of existing and advanced wasteforms: glass, glass-ceramics, glass composite materials (GCMs), and crystalline ceramic (mineral) wasteforms that chemically incorporate radionuclides and hazardous species atomically in their structure. Cementitious, geopolymer, bitumen, and other encapsulant wasteforms and composites that atomically bond and encapsulate wastes are also discussed. The various processing technologies are cross-referenced to the various types of wasteforms since often a particular type of wasteform can be made by a variety of different processing technologies.

  16. Dry-out and low temperature calcination of DST/SST waste blend high temperature melter feed

    SciTech Connect (OSTI)

    Smith, H.D.; Tracey, E.M.

    1996-02-01

    The FY1994 DST/SST blend was prepared in accordance with the DST/SST blend feed specification. The laboratory preparation steps and observations were compared with an existing experience base to verify the acceptability of the feed specification for simulant make-up. The most significant test results included a variety of features. Ferrocyanide breaks down to NH{sub 3} plus formate, during the low-temperature calcining phase of the tests. Ferrocyanide displayed no redox reactivity with the nitrates and nitrites contained in the slurry in the absence of sugar. Sugar displays a redox reaction with the nitrates and nitrites in the blend similar to the redox. reaction observed in the LLW feed simulant. Boiling of a free flowing slurry occurs at temperatures below about 120{degrees}C. When about 45% of the total water loss has occurred, the feed slurry congeals and continues to lose water, shrinking and developing shrinkage cracks. Water stops coming off between 350{degrees}C and 400{degrees}C. Slurry shear strength and viscosity strongly increase as the weight percent solids increases from 20 wt% to 45 wt%. The 45 wt% solids corresponds to approximately a 40 % water loss. The principle beat sensitivity for this material is the exothermic reaction which is activated when the temperature exceeds about 250{degrees}C. The breakdown of ferrocyanide to ammonia and formate under strongly basic conditions may begin at temperatures less than 100{degrees}C, but the rate increased strongly with increasing temperature and appeared to be completed in the time of our tests. Differential thermal analysis (DTA) results on feed slurry without and with ferrocyanide showed only endothermic behavior. This is consistent with the dry out and low temperature calcine studies which did not indicate any exothermic behavior for the feed slurry with and without ferrocyanide.

  17. Evaluation of Alternatives for Hanford 327 Building Hot Cell Removal and Transport

    SciTech Connect (OSTI)

    Stevens, Ray W.; Jasen, William G.

    2003-02-27

    The Department of Energy (DOE) Hanford site 327 Building, built in 1953, played a key role in reactor material and fuel research programs. The facility includes nine shielded hot cells, a fuel storage basin, dry sample storage, and a large inerted hot (SERF) cell. In 1996, the 327 Building was transferred from Pacific Northwest National Laboratory (PNNL) to Fluor Hanford, Inc., to begin the transition from the mission of irradiated fuel examination to stabilization and deactivation. In 2001, a multi-contractor team conducted a review of the concept of intact (one piece) removal, packaging, and disposal of the 327 hot cells. This paper focuses on challenges related to preparing the 327 Building hot cells for intact one-piece disposal as Low Level Waste (LLW) at the Hanford Site. These challenges, described in this paper, are threefold and include: Sampling and characterization of the cells for low level waste designation; Packaging of the cells for transportation and waste disposal; Transportation from the facility to the disposal site. The primary technical challenges in one-piece removal, packaging, and disposal of the hot cells involve the techniques required to characterize, remove, handle, package and transport a large (approximately up to 12-feet long and 8-feet high) contaminated object that weighs 35 to 160 tons. Specific characterization results associated with two hot cells, G and H cells will be reported. A review of the activities and plans to stabilize and deactivate the 327 Building provides insight into the technical challenges faced by this project and identifies a potential opportunity to modify the baseline strategy by removing the hot cells in one piece instead of decontaminating and dismantling the cells.

  18. Intermediate depth burial of classified transuranic wastes in arid alluvium

    SciTech Connect (OSTI)

    Cochran, J.R. [Sandia National Labs., Albuquerque, NM (United States). Environmental Risk and Decision Analysis Dept.; Crowe, B.M. [Los Alamos National Lab., NM (United States). Geologic Integration Group; Di Sanza, F. [Dept. of Energy, Las Vegas, NV (United States). Nevada Operations Office

    1999-04-01

    Intermediate depth disposal operations were conducted by the US Department of Energy (DOE) at the DOE`s Nevada Test Site (NTS) from 1984 through 1989. These operations emplaced high-specific activity low-level wastes (LLW) and limited quantities of classified transuranic (TRU) wastes in 37 m (120-ft) deep, Greater Confinement Disposal (GCD) boreholes. The GCD boreholes are 3 m (10 ft) in diameter and founded in a thick sequence of arid alluvium. The bottom 15 m (50 ft) of each borehole was used for waste emplacement and the upper 21 m (70 ft) was backfilled with native alluvium. The bottom of each GCD borehole is almost 200 m (650 ft) above the water table. The GCD boreholes are located in one of the most arid portions of the US, with an average precipitation of 13 cm (5 inches) per year. The limited precipitation, coupled with generally warm temperatures and low humidities results in a hydrologic system dominated by evapotranspiration. The US Environmental Protection Agency`s (EPA`s) 40 CFR 191 defines the requirements for protection of human health from disposed TRU wastes. This EPA standard sets a number of requirements, including probabilistic limits on the cumulative releases of radionuclides to the accessible environment for 10,000 years. The DOE Nevada Operations Office (DOE/NV) has contracted with Sandia National Laboratories (Sandia) to conduct a performance assessment (PA) to determine if the TRU wastes emplaced in the GCD boreholes complies with the EPA`s 40 CFR 191 requirements. This paper describes DOE`s actions undertaken to evaluate whether the TRU wastes in the GCD boreholes will, or will not, endanger human health. Based on preliminary modeling, the TRU wastes in the GCD boreholes meet the EPA`s requirements, and are, therefore, protective of human health.

  19. Transuranic (TRU) Waste Phase I Retrieval Plan

    SciTech Connect (OSTI)

    MCDONALD, K.M.

    1999-08-27

    Phase I retrieval of post-1970 TRU wastes from burial ground 218-W-4C can be done in a safe, efficient, and cost-effective manner. Initiating TRU retrieval by retrieving uncovered drums from Trenches 1, 20, and 29, will allow retrieval to begin under the current SWBG safety authorization basis. The retrieval of buried drums from Trenches 1, 4, 20, and 29, which will require excavation, will commence once the uncovered drum are retrieved. This phased approach allows safety analysis for drum venting and drum module excavation to be completed and approved before the excavation proceeds. In addition, the lessons learned and the operational experience gained from the retrieval of uncovered drums can be applied to the more complicated retrieval of the buried drums. Precedents that have been set at SRS and LANL to perform retrieval without a trench cover, in the open air, should be followed. Open-air retrieval will result in significant cost savings over the original plans for Phase I retrieval (Project W-113). Based on LANL and SRS experience, open-air retrieval will have no adverse impacts to the environment or to the health and safety of workers or the public. Assaying the waste in the SWBG using a mobile assay system, will result in additional cost savings. It is expected that up to 50% of the suspect-TRU wastes will assay as LLW, allowing those waste to remain disposed of in the SWBG. Further processing, with its associated costs, will only occur to the portion of the waste that is verified to be TRU. Retrieval should be done, to the extent possible, under the current SWBG safety authorization basis as a normal part of SWBG operations. The use of existing personnel and existing procedures should be optimized. By working retrieval campaigns, typically during the slow months, it is easier to coordinate the availability of necessary operations personnel, and it is easier to coordinate the availability of a mobile assay vendor.

  20. Savannah River Site waste vitrification projects initiated throughout the United States: Disposal and recycle options

    SciTech Connect (OSTI)

    Jantzen, C.M.

    2000-04-10

    A vitrification process was developed and successfully implemented by the US Department of Energy's (DOE) Savannah River Site (SRS) and at the West Valley Nuclear Services (WVNS) to convert high-level liquid nuclear wastes (HLLW) to a solid borosilicate glass for safe long term geologic disposal. Over the last decade, SRS has successfully completed two additional vitrification projects to safely dispose of mixed low level wastes (MLLW) (radioactive and hazardous) at the SRS and at the Oak Ridge Reservation (ORR). The SRS, in conjunction with other laboratories, has also demonstrated that vitrification can be used to dispose of a wide variety of MLLW and low-level wastes (LLW) at the SRS, at ORR, at the Los Alamos National Laboratory (LANL), at Rocky Flats (RF), at the Fernald Environmental Management Project (FEMP), and at the Hanford Waste Vitrification Project (HWVP). The SRS, in conjunction with the Electric Power Research Institute and the National Atomic Energy Commission of Argentina (CNEA), have demonstrated that vitrification can also be used to safely dispose of ion-exchange (IEX) resins and sludges from commercial nuclear reactors. In addition, the SRS has successfully demonstrated that numerous wastes declared hazardous by the US Environmental Protection Agency (EPA) can be vitrified, e.g. mining industry wastes, contaminated harbor sludges, asbestos containing material (ACM), Pb-paint on army tanks and bridges. Once these EPA hazardous wastes are vitrified, the waste glass is rendered non-hazardous allowing these materials to be recycled as glassphalt (glass impregnated asphalt for roads and runways), roofing shingles, glasscrete (glass used as aggregate in concrete), or other uses. Glass is also being used as a medium to transport SRS americium (Am) and curium (Cm) to the Oak Ridge Reservation (ORR) for recycle in the ORR medical source program and use in smoke detectors at an estimated value of $1.5 billion to the general public.

  1. Pilot-scale Tests to Vitrify Korean Low-Level Wastes

    SciTech Connect (OSTI)

    Choi, K.; Kim, C.-W.; Park, J. K.; Shin, S. W.; Song, M.-J.; Brunelot, P.; Flament, T.

    2002-02-26

    Korea is under preparation of its first commercial vitrification plant to handle LLW from her Nuclear Power Plants (NPPs). The waste streams include three categories: combustible Dry Active Wastes (DAW), borate concentrates, and spent resin. The combustible DAW in this research contains vinyl bag, paper, and protective cloth and rubber shoe. The loaded resin was used to simulate spent resin from NPPs. As a part of this project, Nuclear Environment Technology Institute (NETEC) has tested an operation mode utilizing its pilot-scale plant and the mixed waste surrogates of resin and DAW. It has also proved, with continuous operation for more than 100 hours, the consistency and operability of the plant including cold crucible melter and its off-gas treatment equipment. Resin and combustible DAW were simultaneously fed into the glass bath with periodic addition of various glass frits as additives, so that it achieved a volume reduction factor larger than 70. By adding various glass frits, this paper discusses about maintaining the viscosity and electrical conductivity of glass bath within their operable ranges, but not about obtaining a durable glass product. The operating mode starts with a batch of glass where a titanium ring is buried. When the induced power ignites the ring, the joule heat melts the surrounding glass frit along with the oxidation heat of titanium. As soon as the molten bath is prepared, in the first stage of the mode, the wastes consisting of loaded resin and combustible DAW are fed with no or minimum addition of glass frits. Then, in the second stage, the bath composition is kept as constant as possible. This operation was successful in terms of maintaining the glass bath under operable condition and produced homogeneous glass. This operation mode could be adapted in commercial stage.

  2. Vitrification of NAC process residue

    SciTech Connect (OSTI)

    Merrill, R.A.; Whittington, K.F.; Peters, R.D. [Pacific Northwest Lab., Richland, WA (United States)

    1995-12-31

    Vitrification tests have been performed with simulated waste compositions formulated to represent the residue which would be obtained from the treatment of low-level, nitrate wastes from Hanford and Oak Ridge by the nitrate to ammonia and ceramic (NAC) process. The tests were designed to demonstrate the feasibility of vitrifying NAC residue and to quantify the impact of the NAC process on the volume of vitrified waste. The residue from NAC treatment of low-level nitrate wastes consists primarily of oxides of aluminum and sodium. High alumina glasses were formulated to maximize the waste loading of the NAC product. Transparent glasses with up to 35 wt% alumina, and even higher contents in opaque glasses, were obtained at melting temperatures of 1,200 C to 1,400 C. A modified TCLP leach test showed the high alumina glasses to have good chemical durability, leaching significantly less than either the ARM-1 or the DWPF-EA high-level waste reference glasses. A significant increase in the final waste volume would be a major result of the NAC process on LLW vitrification. For Hanford wastes, NAC-treatment of nitrate wastes followed by vitrification of the residue will increase the final volume of vitrified waste by 50% to 90%; for Melton Valley waste from Oak Ridge, the increase in final glass volume will be 260% to 280%. The increase in volume is relative to direct vitrification of the waste in a 20 wt% Na{sub 2}O glass formulation. The increase in waste volume directly affects not only disposal costs, but also operating and/or capital costs. Larger plant size, longer operating time, and additional energy and additive costs are direct results of increases in waste volume. Such increases may be balanced by beneficial impacts on the vitrification process; however, those effects are outside the scope of this report.

  3. Vitrification of NAC process residue

    SciTech Connect (OSTI)

    Merrill, R.A.; Whittington, K.F.; Peters, R.D.

    1995-09-01

    Vitrification tests have been performed with simulated waste compositions formulated to represent the residue which would be obtained from the treatment of low-level, nitrate wastes from Hanford and Oak Ridge by the nitrate to ammonia and ceramic (NAC) process. The tests were designed to demonstrate the feasibility of vitrifying NAC residue and to quantify the impact of the NAC process on the volume of vitrified waste. The residue from NAC treatment of low-level nitrate wastes consists primarily of oxides of aluminum and sodium. High alumina glasses were formulated to maximize the waste loading of the NAC product. Transparent glasses with up to 35 wt% alumina, and even higher contents in opaque glasses, were obtained at melting temperatures of 1200{degrees}C to 1400{degrees}C. A modified TCLP leach test showed the high alumina glasses to have good chemical durability, leaching significantly less than either the ARM-1 or the DWPF-EA high-level waste reference glasses. A significant increase in the final waste volume would be a major result of the NAC process on LLW vitrification. For Hanford wastes, NAC-treatment of nitrate wastes followed by vitrification of the residue will increase the final volume of vitrified waste by 50% to 90%; for Melton Valley waste from Oak Ridge, the increase in final glass volume will be 260% to 280%. The increase in volume is relative to direct vitrification of the waste in a 20 wt% Na{sub 2}O glass formulation. The increase in waste volume directly affects not only disposal costs, but also operating and/or capital costs. Larger plant size, longer operating time, and additional energy and additive costs are direct results of increases in waste volume. Such increases may be balanced by beneficial impacts on the vitrification process; however, those effects are outside the scope of this report.

  4. Process system evaluation-consolidated letters. Volume 1. Alternatives for the off-gas treatment system for the low-level waste vitrification process

    SciTech Connect (OSTI)

    Peurrung, L.M.; Deforest, T.J; Richards, J.R.

    1996-03-01

    This report provides an evaluation of alternatives for treating off-gas from the low-level waste (LLW) melter. The study used expertise obtained from the commercial nonradioactive off-gas treatment industry. It was assumed that contact maintenance is possible, although the subsequent risk to maintenance personnel was qualitatively considered in selecting equipment. Some adaptations to the alternatives described may be required, depending on the extent of contact maintenance that can be achieved. This evaluation identified key issues for the off-gas system design. To provide background information, technology reviews were assembled for various classifications of off-gas treatment equipment, including off-gas cooling, particulate control, acid gas control, mist elimination, NO{sub x} reduction, and SO{sub 2} removal. An order-of-magnitude cost estimate for one of the off-gas systems considered is provided using both the off-gas characteristics associated with the Joule-heated and combustion-fired melters. The key issues identified and a description of the preferred off-gas system options are provided below. Five candidate treatment systems were evaluated. All of the systems are appropriate for the different melting/feed preparations currently being considered. The lowest technical risk is achieved using option 1, which is similar to designs for high-level waste (HLW) vitrification in the Hanford Waste Vitrification Project (HWVP) and the West Valley. Demonstration Project. Option 1 uses a film cooler, submerged bed scrubber (SBS), and high-efficiency mist eliminator (HEME) prior to NO{sub x} reduction and high-efficiency particulate air (HEPA) filtration. However, several advantages were identified for option 2, which uses high-temperature filtration. Based on the evaluation, option 2 was identified as the preferred alternative. The characteristics of this option are described below.

  5. Summary of national and international fuel cycle and radioactive waste management programs, 1984

    SciTech Connect (OSTI)

    Harmon, K.M.; Lakey, L.T.; Leigh, I.W.

    1984-07-01

    Worldwide activities related to nuclear fuel cycle and radioactive waste management programs are summarized. Several trends have developed in waste management strategy: All countries having to dispose of reprocessing wastes plan on conversion of the high-level waste (HLW) stream to a borosilicate glass and eventual emplacement of the glass logs, suitably packaged, in a deep geologic repository. Countries that must deal with plutonium-contaminated waste emphasize pluonium recovery, volume reduction and fixation in cement or bitumen in their treatment plans and expect to use deep geologic repositories for final disposal. Commercially available, classical engineering processing are being used worldwide to treat and immobilize low- and intermediate-level wastes (LLW, ILW); disposal to surface structures, shallow-land burial and deep-underground repositories, such as played-out mines, is being done widely with no obvious technical problems. Many countries have established extensive programs to prepare for construction and operation of geologic repositories. Geologic media being studied fall into three main classes: argillites (clay or shale); crystalline rock (granite, basalt, gneiss or gabbro); and evaporates (salt formations). Most nations plan to allow 30 years or longer between discharge of fuel from the reactor and emplacement of HLW or spent fuel is a repository to permit thermal and radioactive decay. Most repository designs are based on the mined-gallery concept, placing waste or spent fuel packages into shallow holes in the floor of the gallery. Many countries have established extensive and costly programs of site evaluation, repository development and safety assessment. Two other waste management problems are the subject of major R and D programs in several countries: stabilization of uranium mill tailing piles; and immobilization or disposal of contaminated nuclear facilities, namely reactors, fuel cycle plants and R and D laboratories.

  6. International fuel cycle and waste management technology exchange activities sponsored by the United States Department of Energy: FY 1982 evaluation report

    SciTech Connect (OSTI)

    Lakey, L.T.; Harmon, K.M.

    1983-02-01

    In FY 1982, DOE and DOE contractor personnel attended 40 international symposia and conferences on fuel reprocessing and waste management subjects. The treatment of high-level waste was the topic most often covered in the visits, with geologic disposal and general waste management also being covered in numerous visits. Topics discussed less frequently inlcude TRU/LLW treatment, airborne waste treatment, D and D, spent fuel handling, and transportation. The benefits accuring to the US from technology exchange activities with other countries are both tangible, e.g., design of equipment, and intangible, e.g., improved foreign relations. New concepts initiated in other countries, particularly those with sizable nuclear programs, are beginning to appear in US efforts in growing numbers. The spent fuel dry storage concept originating in the FRG is being considered at numerous sites. Similarly, the German handling and draining concepts for the joule-heated ceramic melter used to vitrify wastes are being incorporated in US designs. Other foreigh technologies applicable in the US include the slagging incinerator (Belgium), the SYNROC waste form (Australia), the decontamination experience gained in decommissioning the Eurochemic reprocessing plant (Belgium), the engineered surface storage of low- and intermediate-level waste (Belgium, FRG, France), the air-cooled storage of vitrified high-level waste (France, UK), waste packaging (Canada, FRG, Sweden), disposal in salt (FRG), disposal in granite (Canada, Sweden), and sea dumping (UK, Belgium, The Netherlands, Switzerland). These technologies did not necessarily originated or have been tried in the US but for various reasons are now being applied and extended in other countries. This growing nuclear technological base in other countires reduces the number of technology avenues the US need follow to develop a solid nuclear power program.

  7. Nuclear carrier business volume projections, 1980-2000

    SciTech Connect (OSTI)

    Lebo, R.G.; McKeown, M.S.; Rhyne, W.R.

    1980-05-01

    The expected number of shipments of commodities in the nuclear fuel cycle are projected for the years 1980 thru 2000. Projections are made for: yellowcake (U/sub 3/O/sub 8/); natural, enriched and reprocessed uranium hexafluoride (UF/sub 6/); uranium dioxide powder (UO/sub 2/); plutonium dioxide powder (PuO/sub 2/); fresh UO/sub 2/ and mixed oxide (MOX) fuel; spent UO/sub 2/ fuel; low-level waste (LLW); transuranic (TRU) waste; high-activity TRU waste; high-level waste (HLW), and cladding hulls. Projections are also made for non-fuel cycle commodities such as defense TRU wastes and institutional wastes, since they also are shipped by the commercial transportation industry. Projections of waste shipments from LWRs are based on the continuation of current volume reduction and solidification techniques now used by the utility industry. Projections are also made based on a 5% per year reduction in LWR waste volume shipped which is assumed to occur as a result of increased implementation of currently available volume reduction systems. This assumption results in a net 64% decrease in the total waste shipped by the year 2000. LWR waste shipment projections, and essentially all other projections for fuel cycle commodities covered in this report, are normalized to BWR and PWR generating capacity projections set forth by the Department of Energy (DOE) in their low-growth projection of April, 1979. Therefore these commodity shipment projections may be altered to comply with future changes in generating capacity projections. Projected shipments of waste from the reprocessing of spent UO/sub 2/ fuel are based on waste generation rates proposed by Nuclear Fuels Services, Allied-General Nuclear Services, Exxon Nuclear, and the DOE. Reprocessing is assumed to begin again in 1990, with mixed oxide fresh fuel available for shipment by 1991.

  8. Radioactive air emissions notice of construction for the Waste Receiving And Processing facility

    SciTech Connect (OSTI)

    Not Available

    1993-02-01

    The mission of the Waste Receiving And Processing (WRAP) Module 1 facility (also referred to as WRAP 1) includes: examining, assaying, characterizing, treating, and repackaging solid radioactive and mixed waste to enable permanent disposal of the wastes in accordance with all applicable regulations. The solid wastes to be handled in the WRAP 1 facility include low-level waste (LLW), transuranic (TRU) waste, TRU mixed wastes, and low-level mixed wastes (LLMW). Airborne releases from the WRAP 1 facility will be primarily in particulate forms (99.999 percent of total unabated emissions). The release of two volatilized radionuclides, tritium and carbon-14 will contribute less than 0.001 percent of the total unabated emissions. Table 2-1 lists the radionuclides which are anticipated to be emitted from WRAP 1 exhaust stack. The Clean Air Assessment Package 1988 (CAP-88) computer code (WHC 1991) was used to calculate effective dose equivalent (EDE) from WRAP 1 to the maximally exposed offsite individual (MEI), and thus demonstrate compliance with WAC 246-247. Table 4-1 shows the dose factors derived from the CAP-88 modeling and the EDE for each radionuclide. The source term (i.e., emissions after abatement in curies per year) are multiplied by the dose factors to obtain the EDE. The total projected EDE from controlled airborne radiological emissions to the offsite MEI is 1.31E-03 mrem/year. The dose attributable to radiological emissions from WRAP 1 will, then, constitute 0.013 percent of the WAC 246-247 EDE regulatory limit of 10 mrem/year to the offsite MEI.

  9. Transition from Consultation to Monitoring-NRC's Increasingly Focused Review of Factors Important to F-Area Tank Farm Facility Performance - 13153

    SciTech Connect (OSTI)

    Barr, Cynthia; Grossman, Christopher; Alexander, George; Parks, Leah; Fuhrmann, Mark; Shaffner, James; McKenney, Christepher [U.S. NRC, Rockville, MD (United States)] [U.S. NRC, Rockville, MD (United States); Pabalan, Roberto; Pickett, David [Center for Nuclear Waste Regulatory Analyses, Southwest Research Institute, San Antonio, TX (United States)] [Center for Nuclear Waste Regulatory Analyses, Southwest Research Institute, San Antonio, TX (United States); Dinwiddie, Cynthia [Southwest Research Institute, San Antonio, TX (United States)] [Southwest Research Institute, San Antonio, TX (United States)

    2013-07-01

    In consultation with the NRC, DOE issued a waste determination for the F-Area Tank Farm (FTF) facility in March 2012. The FTF consists of 22 underground tanks, each 2.8 to 4.9 million liters in capacity, used to store liquid high-level waste generated as a result of spent fuel reprocessing. The waste determination concluded stabilized waste residuals and associated tanks and auxiliary components at the time of closure are not high-level and can be disposed of as LLW. Prior to issuance of the final waste determination, during the consultation phase, NRC staff reviewed and provided comments on DOE's revision 0 and revision 1 FTF PAs that supported the waste determination and produced a technical evaluation report documenting the results of its multi-year review in October 2011. Following issuance of the waste determination, NRC began to monitor DOE disposal actions to assess compliance with the performance objectives in 10 CFR Part 61, Subpart C. To facilitate its monitoring responsibilities, NRC developed a plan to monitor DOE disposal actions. NRC staff was challenged in developing a focused monitoring plan to ensure limited resources are spent in the most cost-effective manner practical. To address this challenge, NRC prioritized monitoring areas and factors in terms of risk significance and timing. This prioritization was informed by NRC staff's review of DOE's PA documentation, independent probabilistic modeling conducted by NRC staff, and NRC-sponsored research conducted by the Center for Nuclear Waste Regulatory Analyses in San Antonio, TX. (authors)

  10. Disposal options for burner ash from spent graphite fuel. Final study report November 1993

    SciTech Connect (OSTI)

    Pinto, A.P.

    1994-08-01

    Three major disposal alternatives are being considered for Fort St. Vrain Reactor (FSVR) and Peach Bottom Reactor (PBR) spent fuels: direct disposal of packaged, intact spent fuel elements; (2) removal of compacts to separate fuel into high-level waste (HLW) and low-level waste (LLW); and (3) physical/chemical processing to reduce waste volumes and produce stable waste forms. For the third alternative, combustion of fuel matrix graphite and fuel particle carbon coatings is a preferred technique for head-end processing as well as for volume reduction and chemical pretreatment prior to final fixation, packaging, and disposal of radioactive residuals (fissile and fertile materials together with fission and activation products) in a final repository. This report presents the results of a scoping study of alternate means for processing and/or disposal of fissile-bearing particles and ash remaining after combustion of FSVR and PBR spent graphite fuels. Candidate spent fuel ash (SFA) waste forms in decreasing order of estimated technical feasibility include glass-ceramics (GCs), polycrystalline ceramic assemblages (PCAs), and homogeneous amorphous glass. Candidate SFA waste form production processes in increasing order of estimated effort and cost for implementation are: low-density GCs via fuel grinding and simultaneous combustion and waste form production in a slagging cyclone combustor (SCC); glass or low-density GCs via fluidized bed SFA production followed by conventional melting of SFA and frit; PCAs via fluidized bed SFA production followed by hot isostatic pressing (HIPing) of SFA/frit mixtures; and high-density GCs via fluidized bed SFA production followed by HIPing of Calcine/Frit/SFA mixtures.

  11. Closure Plan for the Area 3 Radioactive Waste Management Site at the Nevada Test Site

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2007-09-01

    The Area 3 Radioactive Waste Management Site (RMWS) at the Nevada Test Site (NTS) is managed and operated by National Security Technologies, LLC (NSTec) for the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office (NNSA/NSO). This document is the first update of the interim closure plan for the Area 3 RWMS, which was presented in the Integrated Closure and Monitoring Plan (ICMP) (DOE, 2005). The format and content of this plan follows the Format and Content Guide for U.S. Department of Energy Low-Level Waste Disposal Facility Closure Plans (DOE, 1999a). The major updates to the plan include a new closure date, updated closure inventory, the new institutional control policy, and the Title II engineering cover design. The plan identifies the assumptions and regulatory requirements, describes the disposal sites and the physical environment in which they are located, presents the design of the closure cover, and defines the approach and schedule for both closing and monitoring the site. The Area 3 RWMS accepts low-level waste (LLW) from across the DOE Complex in compliance with the NTS Waste Acceptance Criteria (NNSA/NSO, 2006). The Area 3 RWMS accepts both packaged and unpackaged unclassified bulk LLW for disposal in subsidence craters that resulted from deep underground tests of nuclear devices in the early 1960s. The Area 3 RWMS covers 48 hectares (119 acres) and comprises seven subsidence craters--U-3ax, U-3bl, U-3ah, U-3at, U-3bh, U-3az, and U-3bg. The area between craters U-3ax and U-3bl was excavated to form one large disposal unit (U-3ax/bl); the area between craters U-3ah and U-3at was also excavated to form another large disposal unit (U-3ah/at). Waste unit U-3ax/bl is closed; waste units U-3ah/at and U-3bh are active; and the remaining craters, although currently undeveloped, are available for disposal of waste if required. This plan specifically addresses the closure of the U-3ah/at and the U-3bh LLW units. A final closure cover has been placed on unit U-3ax/bl (Corrective Action Unit 110) at the Area 3 RWMS. Monolayer-evapotranspirative closure cover designs for the U-3ah/at and U-3bh units are provided in this plan. The current-design closure cover thickness is 3 meters (10 feet). The final design cover will have an optimized cover thickness, which is expected to be less than 3 m (10 ft). Although waste operations at the Area 3 RWMS have ceased at the end of June 2006, disposal capacity is available for future disposals at the U-3ah/at and U-3bh units. The Area 3 RWMS is expected to start closure activities in fiscal year 2025, which include the development of final performance assessment and composite analysis documents, closure plan, closure cover design for construction, cover construction, and initiation of the post-closure care and monitoring activities. Current monitoring at the Area 3 RWMS includes monitoring the cover of the closed mixed waste unit U-3ax/bl as required by the Nevada Department of Environmental Protection, and others required under federal regulations and DOE orders. Monitoring data, collected via sensors and analysis of samples, are needed to evaluate radiation doses to the general public, for performance assessment maintenance, to demonstrate regulatory compliance, and to evaluate the actual performance of the RWMSs. Monitoring provides data to ensure the integrity and performance of waste disposal units. The monitoring program is designed to forewarn management and regulators of any failure and need for mitigating actions. The plan describes the program for monitoring direct radiation, air, vadose zone, biota, groundwater, meteorology, and subsidence. The requirements of post-closure cover maintenance and monitoring will be determined in the final closure plan.

  12. Application for a Permit to Operate a Class III Solid Waste Disposal Site at the Nevada Test Site Area 5 Asbestiform Low-Level Solid Waste Disposal Site

    SciTech Connect (OSTI)

    NSTec Environmental Programs

    2010-09-14

    The NTS solid waste disposal sites must be permitted by the state of Nevada Solid Waste Management Authority (SWMA). The SWMA for the NTS is the Nevada Division of Environmental Protection, Bureau of Federal Facilities (NDEP/BFF). The U.S. Department of Energy's National Nuclear Security Administration Nevada Site Office (NNSA/NSO) as land manager (owner), and National Security Technologies (NSTec), as operator, will store, collect, process, and dispose all solid waste by means that do not create a health hazard, a public nuisance, or cause impairment of the environment. NTS disposal sites will not be included in the Nye County Solid Waste Management Plan. The NTS is located approximately 105 kilometers (km) (65 miles [mi]) northwest of Las Vegas, Nevada (Figure 1). The U.S. Department of Energy (DOE) is the federal lands management authority for the NTS, and NSTec is the Management and Operations contractor. Access on and off the NTS is tightly controlled, restricted, and guarded on a 24-hour basis. The NTS has signs posted along its entire perimeter. NSTec is the operator of all solid waste disposal sites on the NTS. The Area 5 RWMS is the location of the permitted facility for the Solid Waste Disposal Site (SWDS). The Area 5 RWMS is located near the eastern edge of the NTS (Figure 2), approximately 26 km (16 mi) north of Mercury, Nevada. The Area 5 RWMS is used for the disposal of low-level waste (LLW) and mixed low-level waste. Many areas surrounding the RWMS have been used in conducting nuclear tests. A Notice of Intent to operate the disposal site as a Class III site was submitted to the state of Nevada on January 28, 1994, and was acknowledged as being received in a letter to the NNSA/NSO on August 30, 1994. Interim approval to operate a Class III SWDS for regulated asbestiform low-level waste (ALLW) was authorized on August 12, 1996 (in letter from Paul Liebendorfer to Runore Wycoff), with operations to be conducted in accordance with the ''Management Plan for the Disposal of Low-Level Waste with Regulated Asbestos Waste.'' A requirement of the authorization was that on or before October 9, 1999, a permit was required to be issued. Because of NDEP and NNSA/NSO review cycles, the final permit was issued on April 5, 2000, for the operation of the Area 5 Low-Level Waste Disposal Site, utilizing Pit 7 (P07) as the designated disposal cell. The original permit applied only to Pit 7, with a total design capacity of 5,831 cubic yards (yd{sup 3}) (157,437 cubic feet [ft{sup 3}]). NNSA/NSO is expanding the SWDS to include the adjacent Upper Cell of Pit 6 (P06), with an additional capacity of 28,037 yd{sup 3} (756,999 ft{sup 3}) (Figure 3). The proposed total capacity of ALLW in Pit 7 and P06 will be approximately 33,870 yd{sup 3} (0.9 million ft{sup 3}). The site will be used for the disposal of regulated ALLW, small quantities of low-level radioactive hydrocarbon-burdened (LLHB) media and debris, LLW, LLW that contains PCB Bulk Product Waste greater than 50 ppm that leaches at a rate of less than 10 micrograms of PCB per liter of water, and small quantities of LLHB demolition and construction waste (hereafter called permissible waste). Waste containing free liquids, or waste that is regulated as hazardous waste under the Resource Conservation and Recovery Act (RCRA) or state-of-generation hazardous waste regulations, will not be accepted for disposal at the site. The only waste regulated under the Toxic Substances Control Act (TSCA) that will be accepted at the disposal site is regulated asbestos-containing materials (RACM). The term asbestiform is used throughout this document to describe this waste. Other TSCA waste (i.e., polychlorinated biphenyls [PCBs]) will not be accepted for disposal at the SWDS. The disposal site will be used as a depository of permissible waste generated both on site and off site. All generators designated by NNSA/NSO will be eligible to dispose regulated ALLW at the Asbestiform Low-Level Waste Disposal Site in accordance with the U.S. Department of Energy, Nevada Operations Office (DOE/NV) 325

  13. Generalized charge-screening in relativistic ThomasFermi model

    SciTech Connect (OSTI)

    Akbari-Moghanjoughi, M.

    2014-10-15

    In this paper, we study the charge shielding within the relativistic Thomas-Fermi model for a wide range of electron number-densities and the atomic-number of screened ions. A generalized energy-density relation is obtained using the force-balance equation and taking into account the Chandrasekhar's relativistic electron degeneracy pressure. By numerically solving a second-order nonlinear differential equation, the Thomas-Fermi screening length is investigated, and the results are compared for three distinct regimes of the solid-density, warm-dense-matter, and white-dwarfs (WDs). It is revealed that our nonlinear screening theory is compatible with the exponentially decaying Thomas-Fermi-type shielding predicted by the linear response theory. Moreover, the variation of relative Thomas-Fermi screening length shows that extremely dense quantum electron fluids are relatively poor charge shielders. Calculation of the total number of screening electrons around a nucleus shows that there is a position of maximum number of screening localized electrons around the screened nucleus, which moves closer to the point-like nucleus by increase in the plasma number density but is unaffected due to increase in the atomic-number value. It is discovered that the total number of screening electrons, (N{sub s}?r{sub TF}{sup 3}/r{sub d}{sup 3} where r{sub TF} and r{sub d} are the Thomas-Fermi and interparticle distance, respectively) has a distinct limit for extremely dense plasmas such as WD-cores and neutron star crusts, which is unique for all given values of the atomic-number. This is equal to saying that in an ultrarelativistic degeneracy limit of electron-ion plasma, the screening length couples with the system dimensionality and the plasma becomes spherically self-similar. Current analysis can provide useful information on the effects of relativistic correction to the charge screening for a wide range of plasma density, such as the inertial-confined plasmas and compact stellar objects.

  14. Phase Transformations in Cast Duplex Stainless Steels

    SciTech Connect (OSTI)

    Yoon-Jun Kim

    2004-12-19

    Duplex stainless steels (DSS) constitute both ferrite and austenite as a matrix. Such a microstructure confers a high corrosion resistance with favorable mechanical properties. However, intermetallic phases such as {sigma} and {chi} can also form during casting or high-temperature processing and can degrade the properties of the DSS. This research was initiated to develop time-temperature-transformation (TTT) and continuous-cooling-transformation (CCT) diagrams of two types of cast duplex stainless steels, CD3MN (Fe-22Cr-5Ni-Mo-N) and CD3MWCuN (Fe-25Cr-7Ni-Mo-W-Cu-N), in order to understand the time and temperature ranges for intermetallic phase formation. The alloys were heat treated isothermally or under controlled cooling conditions and then characterized using conventional metallographic methods that included tint etching, and also using electron microscopy (SEM, TEM) and wavelength dispersive spectroscopy (WDS). The kinetics of intermetallic-phase ({sigma} + {chi}) formation were analyzed using the Johnson-Mehl-Avrami (MA) equation in the case of isothermal transformations and a modified form of this equation in the case of continuous cooling transformations. The rate of intermetallic-phase formation was found to be much faster in CD3MWCuN than CD3MN due mainly to differences in the major alloying contents such as Cr, Ni and Mo. To examine in more detail the effects of these elements of the phase stabilities; a series of eight steel castings was designed with the Cr, Ni and Mo contents systematically varied with respect to the nominal composition of CD3MN. The effects of varying the contents of alloying additions on the formation of intermetallic phases were also studied computationally using the commercial thermodynamic software package, Thermo-Calc. In general, {sigma} was stabilized with increasing Cr addition and {chi} by increasing Mo addition. However, a delicate balance among Ni and other minor elements such as N and Si also exists. Phase equilibria in DSS can be affected by local composition fluctuations in the cast alloy. This may cause discrepancy between thermodynamic prediction and experimental observation.

  15. An accreting white dwarf near the Chandrasekhar limit in the Andromeda galaxy

    SciTech Connect (OSTI)

    Tang, Sumin; Bildsten, Lars; Wolf, William M.; Li, K. L.; Kong, Albert K. H.; Cao, Yi; Kulkarni, Shrinivas R.; Perley, Daniel A.; Prince, Thomas A.; Cenko, S. Bradley; De Cia, Annalisa; Kasliwal, Mansi M.; Laher, Russ R.; Surace, Jason; Nugent, Peter E.

    2014-05-01

    The intermediate Palomar Transient Factory (iPTF) detection of the most recent outburst of the recurrent nova (RN) system RX J0045.4+4154 in the Andromeda galaxy has enabled the unprecedented study of a massive (M > 1.3 M {sub ?}) accreting white dwarf (WD). We detected this nova as part of the near-daily iPTF monitoring of M31 to a depth of R ? 21 mag and triggered optical photometry, spectroscopy and soft X-ray monitoring of the outburst. Peaking at an absolute magnitude of M{sub R} = 6.6 mag, and with a decay time of 1 mag per day, it is a faint and very fast nova. It shows optical emission lines of He/N and expansion velocities of 1900-2600 km s{sup 1} 1-4 days after the optical peak. The Swift monitoring of the X-ray evolution revealed a supersoft source (SSS) with kT {sub eff} ? 90-110 eV that appeared within 5 days after the optical peak, and lasted only 12 days. Most remarkably, this is not the first event from this system, rather it is an RN with a time between outbursts of approximately 1 yr, the shortest known. Recurrent X-ray emission from this binary was detected by ROSAT in 1992 and 1993, and the source was well characterized as a M > 1.3 M {sub ?} WD SSS. Based on the observed recurrence time between different outbursts, the duration and effective temperature of the SS phase, MESA models of accreting WDs allow us to constrain the accretion rate to M-dot >1.710{sup ?7} M{sub ?} yr{sup ?1} and WD mass >1.30 M {sub ?}. If the WD keeps 30% of the accreted material, it will take less than a Myr to reach core densities high enough for carbon ignition (if made of C/O) or electron capture (if made of O/Ne) to end the binary evolution.

  16. LANL OPERATING EXPERIENCE WITH THE WAND AND HERCULES PROTOTYPE SYSTEMS

    SciTech Connect (OSTI)

    K. M. GRUETZMACHER; C. L. FOXX; S. C. MYERS

    2000-09-01

    The Waste Assay for Nonradioactive Disposal (WAND) and the High Efficiency Radiation Counters for Ultimate Low Emission Sensitivity (HERCULES) prototype systems have been operating at Los Alamos National Laboratory's (LANL's) Solid Waste Operation's (SWO'S) non-destructive assay (NDA) building since 1997 and 1998, respectively. These systems are the cornerstone of the verification program for low-density Green is Clean (GIC) waste at the Laboratory. GIC waste includes all non-regulated waste generated in radiological controlled areas (RCAS) that has been actively segregated as clean (i.e., nonradioactive) through the use of waste generator acceptable knowledge (AK). The use of this methodology alters LANL's past practice of disposing of all room trash generated in nuclear facilities in radioactive waste landfills. Waste that is verified clean can be disposed of at the Los Alamos County Landfill. It is estimated that 50-90% of the low-density room trash from radioactive material handling areas at Los Alamos might be free of contamination. This approach avoids the high cost of disposal of clean waste at a radioactive waste landfill. It also reduces consumption of precious space in the radioactive waste landfill where disposal of this waste provides no benefit to the public or the environment. Preserving low level waste (LLW) disposal capacity for truly radioactive waste is critical in this era when expanding existing radioactive waste landfills or permitting new ones is resisted by regulators and stakeholders. This paper describes the operating experience with the WAND and HERCULES since they began operation at SWO. Waste for verification by the WAND system has been limited so far to waste from the Plutonium Facility and the Solid Waste Operations Facility. A total of461 ft3 (13.1 m3) of low-density shredded waste and paper have been verified clean by the WAND system. The HERCULES system has been used to verify waste from four Laboratory facilities. These are the Solid Waste Operations Facility, the TA-48 Chemistry Facility, the Shops Facility, and the Environmental Facility. A total of 3150 ft3 (89.3 m3) of low-density waste has been verified clean by the HERCULES system.

  17. Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

    SciTech Connect (OSTI)

    Shropshire, D.E.; Herring, J.S.

    2004-10-03

    The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrain all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim storage, packaging, transportation, waste forms, waste treatment, decontamination and decommissioning issues; and low-level waste (LLW) and high-level waste (HLW) disposal.

  18. FY 1996 solid waste integrated life-cycle forecast characteristics summary. Volumes 1 and 2

    SciTech Connect (OSTI)

    Templeton, K.J.

    1996-05-23

    For the past six years, a waste volume forecast has been collected annually from onsite and offsite generators that currently ship or are planning to ship solid waste to the Westinghouse Hanford Company`s Central Waste Complex (CWC). This document provides a description of the physical waste forms, hazardous waste constituents, and radionuclides of the waste expected to be shipped to the CWC from 1996 through the remaining life cycle of the Hanford Site (assumed to extend to 2070). In previous years, forecast data has been reported for a 30-year time period; however, the life-cycle approach was adopted this year to maintain consistency with FY 1996 Multi-Year Program Plans. This document is a companion report to two previous reports: the more detailed report on waste volumes, WHC-EP-0900, FY1996 Solid Waste Integrated Life-Cycle Forecast Volume Summary and the report on expected containers, WHC-EP-0903, FY1996 Solid Waste Integrated Life-Cycle Forecast Container Summary. All three documents are based on data gathered during the FY 1995 data call and verified as of January, 1996. These documents are intended to be used in conjunction with other solid waste planning documents as references for short and long-term planning of the WHC Solid Waste Disposal Division`s treatment, storage, and disposal activities over the next several decades. This document focuses on two main characteristics: the physical waste forms and hazardous waste constituents of low-level mixed waste (LLMW) and transuranic waste (both non-mixed and mixed) (TRU(M)). The major generators for each waste category and waste characteristic are also discussed. The characteristics of low-level waste (LLW) are described in Appendix A. In addition, information on radionuclides present in the waste is provided in Appendix B. The FY 1996 forecast data indicate that about 100,900 cubic meters of LLMW and TRU(M) waste is expected to be received at the CWC over the remaining life cycle of the site. Based on ranges provided by the waste generators, this baseline volume could fluctuate between a minimum of about 59,720 cubic meters and a maximum of about 152,170 cubic meters. The range is primarily due to uncertainties associated with the Tank Waste Remediation System (TWRS) program, including uncertainties regarding retrieval of long-length equipment, scheduling, and tank retrieval technologies.

  19. Technology for Treatment of Liquid Radioactive Waste Generated during Uranium and Plutonium Chemical and Metallurgical Manufacturing in FSUE PO Mayak - 13616

    SciTech Connect (OSTI)

    Adamovich, D. [SUE MosSIA Radon, 2/14 7th Rostovsky lane, Moscow, 119121 (Russian Federation)] [SUE MosSIA Radon, 2/14 7th Rostovsky lane, Moscow, 119121 (Russian Federation); Batorshin, G.; Logunov, M.; Musalnikov, A. [FSUE 'PO Mayak', 31 av. Lenin, Ozyorsk, Chelyabinsk region, 456780 (Russian Federation)] [FSUE 'PO Mayak', 31 av. Lenin, Ozyorsk, Chelyabinsk region, 456780 (Russian Federation)

    2013-07-01

    Created technological scheme for treatment of liquid radioactive waste generated while uranium and plutonium chemical and metallurgical manufacturing consists of: - Liquid radioactive waste (LRW) purification from radionuclides and its transfer into category of manufacturing waste; - Concentration of suspensions containing alpha-nuclides and their further conversion to safe dry state (calcinate) and moving to long controlled storage. The following technologies are implemented in LRW treatment complex: - Settling and filtering technology for treatment of liquid intermediate-level waste (ILW) with volume about 1500m{sup 3}/year and alpha-activity from 10{sup 6} to 10{sup 8} Bq/dm{sup 3} - Membrane and sorption technology for processing of low-level waste (LLW) of radioactive drain waters with volume about 150 000 m{sup 3}/year and alpha-activity from 10{sup 3} to 10{sup 4} Bq/dm{sup 3}. Settling and filtering technology includes two stages of ILW immobilization accompanied with primary settling of radionuclides on transition metal hydroxides with the following flushing and drying of the pulp generated; secondary deep after settling of radionuclides on transition metal hydroxides with the following solid phase concentration by the method of tangential flow ultrafiltration. Besides, the installation capacity on permeate is not less than 3 m{sup 3}/h. Concentrates generated are sent to calcination on microwave drying (MW drying) unit. Membrane and sorption technology includes processing of averaged sewage flux by the method of tangential flow ultrafiltration with total capacity of installations on permeate not less than 18 m{sup 3}/h and sorption extraction of uranium from permeate on anionite. According to radionuclide contamination level purified solution refers to general industrial waste. Concentrates generated during suspension filtering are evaporated in rotary film evaporator (RFE) in order to remove excess water, thereafter they are dried on infrared heating facility. Solid concentrate produced is sent for long controlled storage. Complex of the procedures carried out makes it possible to solve problems on treatment of LRW generated while uranium and plutonium chemical and metallurgical manufacturing in Federal State Unitary Enterprise (FSUE) Mayak and cease its discharge into open water reservoirs. (authors)

  20. Integrated Disposal Facility FY2010 Glass Testing Summary Report

    SciTech Connect (OSTI)

    Pierce, Eric M.; Bacon, Diana H.; Kerisit, Sebastien N.; Windisch, Charles F.; Cantrell, Kirk J.; Valenta, Michelle M.; Burton, Sarah D.; Serne, R Jeffrey; Mattigod, Shas V.

    2010-09-30

    Pacific Northwest National Laboratory was contracted by Washington River Protection Solutions, LLC to provide the technical basis for estimating radionuclide release from the engineered portion of the disposal facility (e.g., source term). Vitrifying the low-activity waste at Hanford is expected to generate over 1.6 105 m3 of glass (Puigh 1999). The volume of immobilized low-activity waste (ILAW) at Hanford is the largest in the DOE complex and is one of the largest inventories (approximately 0.89 1018 Bq total activity) of long-lived radionuclides, principally 99Tc (t1/2 = 2.1 105), planned for disposal in a low-level waste (LLW) facility. Before the ILAW can be disposed, DOE must conduct a performance assessement (PA) for the Integrated Disposal Facility (IDF) that describes the long-term impacts of the disposal facility on public health and environmental resources. As part of the ILAW glass testing program PNNL is implementing a strategy, consisting of experimentation and modeling, in order to provide the technical basis for estimating radionuclide release from the glass waste form in support of future IDF PAs. The purpose of this report is to summarize the progress made in fiscal year (FY) 2010 toward implementing the strategy with the goal of developing an understanding of the long-term corrosion behavior of low-activity waste glasses. The emphasis in FY2010 was the completing an evaluation of the most sensitive kinetic rate law parameters used to predict glass weathering, documented in Bacon and Pierce (2010), and transitioning from the use of the Subsurface Transport Over Reactive Multi-phases to Subsurface Transport Over Multiple Phases computer code for near-field calculations. The FY2010 activities also consisted of developing a Monte Carlo and Geochemical Modeling framework that links glass composition to alteration phase formation by 1) determining the structure of unreacted and reacted glasses for use as input information into Monte Carlo calculations, 2) compiling the solution data and alteration phases identified from accelerated weathering tests conducted with ILAW glass by PNNL and Viteous State Laboratory/Catholic University of America as well as other literature sources for use in geochemical modeling calculations, and 3) conducting several initial calculations on glasses that contain the four major components of ILAW-Al2O3, B2O3, Na2O, and SiO2.

  1. Revised Analyses of Decommissioning Reference Non-Fuel-Cycle Facilities

    SciTech Connect (OSTI)

    MC Bierschbach; DR Haffner; KJ Schneider; SM Short

    2002-12-01

    Cost information is developed for the conceptual decommissioning of non-fuel-cycle nuclear facilities that represent a significant decommissioning task in terms of decontamination and disposal activities. This study is a re-evaluation of the original study (NUREG/CR-1754 and NUREG/CR-1754, Addendum 1). The reference facilities examined in this study are the same as in the original study and include: a laboratory for the manufacture of {sup 3}H-labeled compounds; a laboratory for the manufacture of {sup 14}C-labeled compounds; a laboratory for the manufacture of {sup 123}I-labeled compounds; a laboratory for the manufacture of {sup 137}Cs sealed sources; a laboratory for the manufacture of {sup 241}Am sealed sources; and an institutional user laboratory. In addition to the laboratories, three reference sites that require some decommissioning effort were also examined. These sites are: (1) a site with a contaminated drain line and hold-up tank; (2) a site with a contaminated ground surface; and (3) a tailings pile containing uranium and thorium residues. Decommissioning of these reference facilities and sites can be accomplished using techniques and equipment that are in common industrial use. Essentially the same technology assumed in the original study is used in this study. For the reference laboratory-type facilities, the study approach is to first evaluate the decommissioning of individual components (e.g., fume hoods, glove boxes, and building surfaces) that are common to many laboratory facilities. The information obtained from analyzing the individual components of each facility are then used to determine the cost, manpower requirements and dose information for the decommissioning of the entire facility. DECON, the objective of the 1988 Rulemaking for materials facilities, is the decommissioning alternative evaluated for the reference laboratories because it results in the release of the facility for restricted or unrestricted use as soon as possible. For a facility, DECON requires that contaminated components either be: (1) decontaminated to restricted or unrestricted release levels or (2) packaged and shipped to an authorized disposal site. This study considers unrestricted release only. The new decommissioning criteria of July 1997 are too recent for this study to include a cost analysis of the restricted release option, which is now allowed under these new criteria. The costs of decommissioning facility components are generally estimated to be in the range of $140 to $27,000, depending on the type of component, the type and amount of radioactive contamination, the remediation options chosen, and the quantity of radioactive waste generated from decommissioning operations. Estimated costs for decommissioning the example laboratories range from $130,000 to $205,000, assuming aggressive low-level waste (LLW) volume reduction. If only minimal LLW volume reduction is employed, decommissioning costs range from $150,000 to $270,000 for these laboratories. On the basis of estimated decommissioning costs for facility components, the costs of decommissioning typical non-fuel-cycle laboratory facilities are estimated to range from about $25,000 for the decommissioning of a small room containing one or two fume hoods to more than $1 million for the decommissioning of an industrial plant containing several laboratories in which radiochemicals and sealed radioactive sources are prepared. For the reference sites of this study, the basic decommissioning alternatives are: (1) site stabilization followed by long-term care and (2) removal of the waste or contaminated soil to an authorized disposal site. Cost estimates made for decommissioning three reference sites range from about $130,000 for the removal of a contaminated drain line and hold-up tank to more than $23 million for the removal of a tailings pile that contains radioactive residue from ore-processing operations in which tin slag is processed for the recovery of rare metals. Total occupational radiation doses generally range from 0.00007 person-rem to 13 person-rem for

  2. Special Analysis: Disposal Plan for Pit 38 at Technical Area 54, Area G

    SciTech Connect (OSTI)

    French, Sean B. [Los Alamos National Laboratory; Shuman, Rob [URS Coporation

    2012-06-26

    Los Alamos National Laboratory (LANL) generates radioactive waste as a result of various activities. Operational waste is generated from a wide variety of research and development activities including nuclear weapons development, energy production, and medical research; environmental restoration (ER), and decontamination and decommissioning (D&D) waste is generated as contaminated sites and facilities at LANL undergo cleanup or remediation. The majority of this waste is low-level radioactive waste (LLW) and is disposed of at the Technical Area 54 (TA-54), Area G disposal facility. U.S. Department of Energy (DOE) Order 435.1 (DOE, 2001) requires that radioactive waste be managed in a manner that protects public health and safety, and the environment. To comply with this order, DOE field sites must prepare site-specific radiological performance assessments for LLW disposal facilities that accept waste after September 26, 1988. Furthermore, sites are required to conduct composite analyses that account for the cumulative impacts of all waste that has been (or will be) disposed of at the facilities and other sources of radioactive material that may interact with the facilities. Revision 4 of the Area G performance assessment and composite analysis was issued in 2008 (LANL, 2008). These analyses estimate rates of radionuclide release from the waste disposed of at the facility, simulate the movement of radionuclides through the environment, and project potential radiation doses to humans for several on- and off-site exposure scenarios. The assessments are based on existing site and disposal facility data, and on assumptions about future rates and methods of waste disposal. The Area G disposal facility consists of Material Disposal Area (MDA) G and the Zone 4 expansion area. To date, disposal operations have been confined to MDA G and are scheduled to continue in that region until MDA G undergoes final closure at the end of 2013. Given its impending closure, efforts have been made to utilize the remaining disposal capacity within MDA G to the greatest extent possible. One approach for doing this has been to dispose of low-activity waste from cleanup operations at LANL in the headspace of selected disposal pits. Waste acceptance criteria (WAC) for the material placed in the headspace of pits 15, 37, and 38 have been developed (LANL, 2010) and the impacts of placing waste in the headspace of these units has been evaluated (LANL, 2012a). The efforts to maximize disposal efficiency have taken on renewed importance because of the disposal demands placed on MDA G by the large volumes of waste that are being generated at LANL by cleanup efforts. For example, large quantities of waste were recently generated by the retrieval of waste formerly disposed of at TA-21, MDA B. A portion of this material has been disposed of in the headspace of pit 38 in compliance with the WAC developed for that disposal strategy; a large amount of waste has also been sent to off-site facilities for disposal. Nevertheless, large quantities of MDA B waste remain that require disposal. An extension of pit 38 was proposed to provide the disposal capacity that will be needed to dispose of institutional waste and MDA B waste through 2013. A special analysis was prepared to evaluate the impacts of the pit extension (LANL, 2012b). The analysis concluded that the disposal unit could be extended with modest increases in the exposures projected for the Area G performance assessment and composite analysis, as long as limits were placed on the radionuclide concentrations in the waste that is placed in the headspace of the pit. Based, in part, on the results of the special analysis, the extension of pit 38 was approved and excavation of the additional disposal capacity was started in May 2012. The special analysis presented here uses performance modeling to identify a disposal plan for the placement of waste in pit 38. The modeling uses a refined design of the disposal unit and updated radionuclide inventories to identify a disposal configuration that promotes efficie

  3. Thermonuclear supernovae: probing magnetic fields by positrons and late-time IR line profiles

    SciTech Connect (OSTI)

    Penney, R.; Hoeflich, P., E-mail: phoeflich77@gmail.com, E-mail: rpenney@g.clemson.edu [Department of Physics, Florida State University, Tallahassee, FL 32305 (United States)

    2014-11-01

    We show the importance of ? and positron transport for the formation of late-time spectra in Type Ia supernovae (SNe Ia). The goal is to study the imprint of magnetic fields (B) on late-time IR line profiles, particularly the [Fe II] feature at 1.644 ?m, which becomes prominent two to three months after the explosion. As a benchmark, we use the explosion of a Chandrasekhar mass (M {sub Ch}) white dwarf (WD) and, specifically, a delayed detonation model that can reproduce the light curves and spectra for a Branch-normal SN Ia. We assume WDs with initial magnetic surface fields between 1 and 10{sup 9} G. We discuss large-scale dipole and small-scale magnetic fields. We show that positron transport effects must be taken into account for the interpretation of emission features starting at about one to two years after maximum light, depending on the size of B. The [Fe II] line profile and its evolution with time can be understood in terms of the overall energy input by radioactive decay and the transition from a ?-ray to a positron-dominated regime. We find that the [Fe II] line at 1.644 ?m can be used to analyze the overall chemical and density structure of the exploding WD up to day 200 without considering B. At later times, positron transport and magnetic field effects become important. After about day 300, the line profile allows one to probe the size of the B-field. The profile becomes sensitive to the morphology of B at about day 500. In the presence of a large-scale dipole field, a broad line is produced in M {sub Ch} mass explosions that may appear flat-topped or rounded depending on the inclination at which the SN is observed. Small or no directional dependence of the spectra is found for small-scale B. We note that narrow-line profiles require central {sup 56}Ni as shown in our previous studies. Persistent broad-line, flat-topped profiles require high-density burning, which is the signature of a WD close to M {sub Ch}. Good time coverage is required to separate the effects of optical depth, the size and morphology of B, and the aspect angle of the observer. The spectra require a resolution of about 500 km s{sup 1} and a signal-to-noise ratio of about 20%. Two other strong near-IR spectral features at about 1.5 and 1.8 ?m are used to demonstrate the importance of line blending, which may invalidate a kinematic interpretation of emission lines. Flat-topped line profiles between 300 and 400 days have been observed and reported in the literature. They lend support for M {sub Ch} mass explosions in at least some cases and require magnetic fields equal to or in excess of 10{sup 6} G. We briefly discuss the effects of the size and morphology of B on light curves, as well as limitations. We argue that line profiles are a more direct measurement of B than light curves because they measure both the distribution of {sup 56}Ni and the redistribution of the energy input by positrons rather than the total energy input. Finally, we discuss possible mechanisms for the formation of high B-fields and the limitations of our analysis.

  4. Energy Saving Melting and Revert Reduction Technology (Energy SMARRT): Development of CCT Diagrams

    SciTech Connect (OSTI)

    L. Scott Chumbley

    2011-08-20

    One of the most energy intensive industries in the U.S. today is in the melting and casting of steel alloys for use in our advanced technological society. While the majority of steel castings involve low or mild carbon steel for common construction materials, highly-alloyed steels constitute a critical component of many industries due to their excellent properties. However, as the amount of alloying additions increases, the problems associated with casting these materials also increases, resulting in a large waste of energy due to inefficiency and a lack of basic information concerning these often complicated alloy systems. Superaustenitic stainless steels constitute a group of Fe-based alloys that are compositionally balanced to have a purely austenitic matrix and exhibit favorable pitting and crevice corrosion resistant properties and mechanical strength. However, intermetallic precipitates such as sigma (???¯??????³) and Laves can form during casting or exposure to high-temperature processing, which degrade the corrosion and mechanical properties of the material. Knowledge of the times and temperatures at which these detrimental phases form is imperative if a company is to efficiently produce castings of high quality in the minimum amount of time, using the lowest amount of energy possible, while producing the least amount of material waste. Anecdotal evidence from company representatives revealed that large castings frequently had to be scrapped due to either lower than expected corrosion resistance or extremely low fracture toughness. It was suspected that these poor corrosion and / or mechanical properties were directly related to the type, amount, and location of various intermetallic phases that formed during either the cooling cycle of the castings or subsequent heat treatments. However, no reliable data existed concerning either the time-temperature-transformation (TTT) diagrams or the continuous-cooling-transformation (CCT) diagrams of the super-austenitics. The goal of this study was to accurately characterize the solid-solid phase transformations seen in cast superaustenitic stainless steels. Heat treatments were performed to understand the time and temperature ranges for intermetallic phase formations in alloys CN3MN and CK3McuN. Microstructures were characterized using scanning electron microscopy (SEM), transmission electron microscopy (TEM), and energy and wavelength dispersive spectroscopy (EDS, WDS). In this way TTT and CCT diagrams could be developed for the matrix of samples chosen. As this study consisted of basic research into the development of TTT and CCT diagrams as an aid to the US steel casting industry, there is no formal commercialization plan associated with this task other than presentations and publications via the Steel Founders Society of America to their members. The author is confident that the data contained in this report can be used by steel foundries to refine their casting procedures in such a way as to reduce the amount of waste produced and energy wasted by significantly reducing or eliminating the need for remelting or recasting of material due to unwanted, premature intermetallic formation. This development of high alloy steel CCT diagrams was predicted to result in an average energy savings of 0.05 trillion BTU???¢????????s/year over a 10 year period (with full funding). With 65% of the proposed funding, current (2011) annual energy saving estimates, based on initial dissemination to the casting industry in 2011and market penetration of 97% by 2020, is 0.14 trillion BTU???¢????????s/year. The reduction of scrap and improvement in casting yield will also result in a reduction of environmental emissions associated with the melting and pouring of the steel. The average annual estimate of CO2 reduction per year through 2020 is 0.003 Million Metri

  5. CLEAN CAST STEEL TECHNOLOGY: DETERMINATION OF TRANSFORMATION DIAGRAMS FOR DUPLEX STAINLESS STEEL.

    SciTech Connect (OSTI)

    Chumbley. L., S.

    2005-09-18

    Duplex stainless steels (DSS) constitute both ferrite and austenite as a matrix. Such a microstructure confers a high corrosion resistance with favorable mechanical properties. However, intermetallic phases such as sigma (???????????????¯??????????????????????????????³) and chi (???????????????¯??????????????????????????????£) can also form during casting or high-temperature processing and can degrade the properties of the DSS. This research was initiated to develop time-temperature-transformation (TTT) and continuous-cooling- transformation (CCT) diagrams of two types of cast duplex stainless steels, CD3MN (Fe 22Cr-5Ni-Mo-N) and CD3MWCuN (Fe-25Cr-7Ni-Mo-W-Cu-N), in order to understand the time and temperature ranges for intermetallic phase formation. The alloys were heat treated isothermally or under controlled cooling conditions and then characterized using conventional metallographic methods that included tint etching, and also using electron microscopy (SEM, TEM) and wavelength dispersive spectroscopy (WDS). The kinetics of intermetallic-phase (???????????????¯??????????????????????????????³ + ???????????????¯??????????????????????????????£) formation were analyzed using the Johnson-Mehl-Avrami (JMA) equation in the case of isothermal transformations and a modified form of this equation in the case of continuous cooling transformations, The rate of intermetallic-phase formation was found to be much faster in CD3MWCuN than CD3MN due mainly to differences in the major alloying contents such as Cr, Ni and Mo. To examine in more detail the effects of these elements of the phase stabilities, a series of eight steel castings was designed with the Cr, Ni and Mo contents systematically varied with respect to the nominal composition of CD3MN. The effects of varying the contents of alloying additions on the formation of intermetallic phases were also studied computationally using the commercial thermodynamic software package, Thermo-Calc. In general, ???????????????¯??????????????????????????????³ was stabilized with increasing Cr addition and by increasing Mo addition. However, a delicate balance among Ni and other minor elements such as N and Si also exists. Phase equilibria in DSS can be affected by local composition fluctuations in the cast alloy. This may cause discrepancy between thermodynamic prediction and experimental observation.

  6. Transuranic Waste Processing Center (TWPC) Legacy Tank RH-TRU Sludge Processing and Compliance Strategy - 13255

    SciTech Connect (OSTI)

    Rogers, Ben C.; Heacker, Fred K.; Shannon, Christopher [Wastren Advantage, Inc., Transuranic Waste Processing Center, 100 WIPP Road, Lenoir City, Tennessee 37771 (United States)] [Wastren Advantage, Inc., Transuranic Waste Processing Center, 100 WIPP Road, Lenoir City, Tennessee 37771 (United States); and others

    2013-07-01

    The U.S. Department of Energy (DOE) needs to safely and efficiently treat its 'legacy' transuranic (TRU) waste and mixed low-level waste (LLW) from past research and defense activities at the Oak Ridge National Laboratory (ORNL) so that the waste is prepared for safe and secure disposal. The TWPC operates an Environmental Management (EM) waste processing facility on the Oak Ridge Reservation (ORR). The TWPC is classified as a Hazard Category 2, non-reactor nuclear facility. This facility receives, treats, and packages low-level waste and TRU waste stored at various facilities on the ORR for eventual off-site disposal at various DOE sites and commercial facilities. The Remote Handled TRU Waste Sludge held in the Melton Valley Storage Tanks (MVSTs) was produced as a result of the collection, treatment, and storage of liquid radioactive waste originating from the ORNL radiochemical processing and radioisotope production programs. The MVSTs contain most of the associated waste from the Gunite and Associated Tanks (GAAT) in the ORNL's Tank Farms in Bethel Valley and the sludge (SL) and associated waste from the Old Hydro-fracture Facility tanks and other Federal Facility Agreement (FFA) tanks. The SL Processing Facility Build-outs (SL-PFB) Project is integral to the EM cleanup mission at ORNL and is being accelerated by DOE to meet updated regulatory commitments in the Site Treatment Plan. To meet these commitments a Baseline (BL) Change Proposal (BCP) is being submitted to provide continued spending authority as the project re-initiation extends across fiscal year 2012 (FY2012) into fiscal year 2013. Future waste from the ORNL Building 3019 U-233 Disposition project, in the form of U-233 dissolved in nitric acid and water, down-blended with depleted uranyl nitrate solution is also expected to be transferred to the 7856 MVST Annex Facility (formally the Capacity Increase Project (CIP) Tanks) for co-processing with the SL. The SL-PFB project will construct and install the necessary integrated systems to process the accumulated MVST Facilities SL inventory at the TWPC thus enabling safe and effective disposal of the waste. This BCP does not include work to support current MVST Facility Surveillance and Maintenance programs or the ORNL Building 3019 U-233 Disposition project, since they are not currently part of the TWPC prime contract. The purpose of the environmental compliance strategy is to identify the environmental permits and other required regulatory documents necessary for the construction and operation of the SL- PFB at the TWPC, Oak Ridge, TN. The permits and other regulatory documents identified are necessary to comply with the environmental laws and regulations of DOE Orders, and other requirements documented in the SL-PFB, Safety Design Strategy (SDS), SL-A-AD-002, R0 draft, and the Systems, Function and Requirements Document (SFRD), SL-X-AD-002, R1 draft. This compliance strategy is considered a 'living strategy' and it is anticipated that it will be revised as design progresses and more detail is known. The design basis on which this environmental permitting and compliance strategy is based is the Wastren Advantage, Inc., (WAI), TWPC, SL-PFB (WAI-BL-B.01.06) baseline. (authors)

  7. Disposition of Radioisotope Thermoelectric Generators Currently Located at the Oak Ridge National Laboratory - 12232

    SciTech Connect (OSTI)

    Glenn, J.; Patterson, J.; DeRoos, K.; Patterson, J.E.; Mitchell, K.G.

    2012-07-01

    Under the American Recovery and Reinvestment Act (ARRA), the U.S. Department of Energy (DOE) awarded SEC Federal Services Corporation (SEC) a 34-building demolition and disposal (D and D) project at the Oak Ridge National Laboratory (ORNL) that included the disposition of six Strontium (Sr-90) powered Radioisotope Thermoelectric Generators (RTGs) stored outside of ORNL Building 3517. Disposition of the RTGs is very complex both in terms of complying with disposal facility waste acceptance criteria (WAC) and U.S. Department of Transportation (DOT) requirements for packaging and transportation in commerce. Two of the RTGs contain elemental mercury which requires them to be Land Disposal Restrictions (LDR) compliant prior to disposal. In addition, all of the RTGs exceed the Class C waste concentration limits under Nuclear Regulatory Commission (NRC) Waste Classification Guidelines. In order to meet the LDR requirements and Nevada National Security Site (NNSS) WAC, a site specific treatability variance for mercury was submitted to the U.S. Environmental Protection Agency (EPA) to allow macro-encapsulation to be an acceptable treatment standard for elemental mercury. By identifying and confirming the design configuration of the mercury containing RTGs, the SEC team proved that the current configuration met the macro-encapsulation standard of 40 Code of Federal Regulations (CFR) 268.45. The SEC Team also worked with NNSS to demonstrate that all radioisotope considerations are compliant with the NNSS low-level waste (LLW) disposal facility performance assessment and WAC. Lastly, the SEC team determined that the GE2000 Type B cask met the necessary size, weight, and thermal loading requirements for five of the six RTGs. The sixth RTG (BUP-500) required a one-time DOT shipment exemption request due to the RTG's large size. The DOT exemption justification for the BUP-500 relies on the inherent robust construction and material make-up of the BUP- 500 RTG. DOE-ORO, SEC, and the entire SEC RTG team are nearing the conclusion of the Sr-90 RTG disposition challenge - a legacy now 50 years in the making. Over 600,000 Ci of Sr-90 waste await disposal and its removal from ORNL will mark an historical moment in the clean-up of the cold-war legacy in the ORNL central industrial area. Elimination (i.e., removal) of the RTGs will reduce security risks at ORNL and disposal will permanently eliminate security risks. The RTGs will eventually decay to benign levels within a reasonable timeframe relative to radiological risks posed by long-lived isotopes. The safety authorization basis at ORNL Building 3517 will be reduced enabling greater operational flexibility in future clean-out and D and D campaigns. Upon disposition the Department of Energy will realize reduced direct and indirect surveillance and maintenance costs that can be reapplied to accelerated and enhanced clean-up of the Oak Ridge Reservation. At present, waste profiles for the RTGs are developed and under review by NNSS RWAP staff and approval authorities. Disposition schedule is driven by the availability of compliant shipping casks necessary to safely transport the RTGs from ORNL to NNSS. The first disposal of the RCA RTG is expected in April 2012 and the remaining RTGs disposed in 2012 and 2013. (authors)

  8. Geochemical Data Package for the 2005 Hanford Integrated Disposal Facility Performance Assessment

    SciTech Connect (OSTI)

    Krupka, Kenneth M.; Serne, R JEFFREY.; Kaplan, D I.

    2004-09-30

    CH2M HILL Hanford Group, Inc. (CH2M HILL) is designing and assessing the performance of an integrated disposal facility (IDF) to receive low-level waste (LLW), mixed low-level waste (MLLW), immobilized low-activity waste (ILAW), and failed or decommissioned melters. The CH2M HILL project to assess the performance of this disposal facility is the Hanford IDF Performance Assessment (PA) activity. The goal of the Hanford IDF PA activity is to provide a reasonable expectation that the disposal of the waste is protective of the general public, groundwater resources, air resources, surface-water resources, and inadvertent intruders. Achieving this goal will require prediction of contaminant migration from the facilities. This migration is expected to occur primarily via the movement of water through the facilities, and the consequent transport of dissolved contaminants in the vadose zone to groundwater where contaminants may be re-introduced to receptors via drinking water wells or mixing in the Columbia River. Pacific Northwest National Laboratory (PNNL) assists CH2M HILL in their performance assessment activities. One of the PNNL tasks is to provide estimates of the geochemical properties of the materials comprising the IDF, the disturbed region around the facility, and the physically undisturbed sediments below the facility (including the vadose zone sediments and the aquifer sediments in the upper unconfined aquifer). The geochemical properties are expressed as parameters that quantify the adsorption of contaminants and the solubility constraints that might apply for those contaminants that may exceed solubility constraints. The common parameters used to quantify adsorption and solubility are the distribution coefficient (Kd) and the thermodynamic solubility product (Ksp), respectively. In this data package, we approximate the solubility of contaminants using a more simplified construct, called the solution concentration limit, a constant value. The Kd values and solution concentration limits for each contaminant are direct inputs to subsurface flow and transport codes used to predict the performance of the IDF system. In addition to the best-estimate Kd values, a reasonable conservative value and a range are provided. The data package does not list estimates for the range in solubility limits or their uncertainty. However, the data package does provide different values for both the Kd values and solution concentration limits for different spatial zones in the IDF system and does supply time-varying Kd values for the cement solidified waste. The Kd values and solution concentration limits presented for each contaminant were previously presented in a report prepared by Kaplan and Serne (2000) for the 2001 ILAW PA, and have been updated to include applicable data from investigations completed since the issuance of that report and improvements in our understanding of the geochemistry specific to Hanford. A discussion is also included of the evolution of the Kd values recommended from the original 1999 ILAW PA through the 2001 ILAW and 2003 Supplement PAs to the current values to be used for the 2005 IDF PA for the key contaminants of concern: Cr(VI), nitrate, 129I, 79Se, 99Tc, and U(VI). This discussion provides the rationale for why certain Kd have changed with time.

  9. FY2010 ANNUAL REVIEW E-AREA LOW-LEVEL WASTE FACILITY PERFORMANCE ASSESSMENT AND COMPOSITE ANALYSIS

    SciTech Connect (OSTI)

    Butcher, T.; Swingle, R.; Crapse, K.; Millings, M.; Sink, D.

    2011-01-01

    The E-Area Low-Level Waste Facility (ELLWF) consists of a number of disposal units described in the Performance Assessment (PA)(WSRC, 2008b) and Composite Analysis (CA)(WSRC, 1997; WSRC, 1999): Low-Activity Waste (LAW) Vault, Intermediate Level (IL) Vault, Trenches (Slit Trenches [STs], Engineered Trenches [ETs], and Component-in-Grout [CIG] Trenches), and Naval Reactor Component Disposal Areas (NRCDAs). This annual review evaluates the adequacy of the approved 2008 ELLWF PA along with the Special Analyses (SAs) approved since the PA was issued. The review also verifies that the Fiscal Year (FY) 2010 low-level waste (LLW) disposal operations were conducted within the bounds of the PA/SA baseline, the Savannah River Site (SRS) CA, and the Department of Energy (DOE) Disposal Authorization Statement (DAS). Important factors considered in this review include waste receipts, results from monitoring and research and development (R&D) programs, and the adequacy of controls derived from the PA/SA baseline. Sections 1.0 and 2.0 of this review are a summary of the adequacy of the PA/SA and CA, respectively. An evaluation of the FY2010 waste receipts and the resultant impact on the ELLWF is summarized in Section 3.1. The results of the monitoring program, R&D program, and other relevant factors are found in Section 3.2, 3.3 and 3.4, respectively. Section 4.0 contains the CA annual determination similarly organized. SRS low-level waste management is regulated under DOE Order 435.1 (DOE, 1999a) and is authorized under a DAS as a federal permit. The original DAS was issued by the DOE-Headquarters (DOE-HQ) on September 28, 1999 (DOE, 1999b) for the operation of the ELLWF and the Saltstone Disposal Facility (SDF). The 1999 DAS remains in effect for the regulation of the SDF. Those portions of that DAS applicable to the ELLWF were superseded by revision 1 of the DAS on July 15, 2008 (DOE, 2008b). The 2008 PA and DAS were officially implemented by the facility on October 31, 2008 and are the authorization documents for this FY2010 Annual Review. Department of Energy Headquarters approval of the 2008 DAS was subject to numerous conditions specified in the document. Two of those conditions are to update the ELLWF closure plan and monitoring plan to align with the conceptual model analyzed in the PA. Both of these conditions were met with the issuance of the PA Monitoring Plan (Millings, 2009a) and the Closure Plan (Phifer et al, 2009a). The PA Monitoring Plan was approved by DOE on July 22, 2009 and the Closure Plan was approved by DOE on May 21, 2009. Both will be updated as needed to remain consistent with the PA. The DAS also specifies that the maintenance plan include activities to resolve each of the secondary issues identified in the DOEHQ review of the 2008 PA that were not completely addressed either with supplemental material provided to the review team or in final revisions to the PA. These outstanding issues were originally documented in the 2008 update of the PA/CA Maintenance Plan (WSRC, 2008a) and in subsequent PA/CA Maintenance Plans (most recently SRNS, 2010a) as required and are actively being worked.

  10. Transuranic (TRU) Waste Repackaging at the Nevada Test Site

    SciTech Connect (OSTI)

    E.F. Di Sanza; G. Pyles; J. Ciucci; P. Arnold

    2009-03-01

    This paper describes the activities required to modify a facility and the process of characterizing, repackaging, and preparing for shipment the Nevada Test Sites (NTS) legacy transuranic (TRU) waste in 58 oversize boxes (OSB). The waste, generated at other U.S. Department of Energy (DOE) sites and shipped to the NTS between 1974 and 1990, requires size-reduction for off-site shipment and disposal. The waste processing approach was tailored to reduce the volume of TRU waste by employing decontamination and non-destructive assay. As a result, the low-level waste (LLW) generated by this process was packaged, with minimal size reduction, in large sea-land containers for disposal at the NTS Area 5 Radioactive Waste Management Complex (RWMC). The remaining TRU waste was repackaged and sent to the Idaho National Laboratory Consolidation Site for additional characterization in preparation for disposal at the Waste Isolation Pilot Plant (WIPP), near Carlsbad, New Mexico. The DOE National Nuclear Security Administration Nevada Site Office and the NTS Management and Operating (M&O) contractor, NSTec, successfully partnered to modify and upgrade an existing facility, the Visual Examination and Repackaging Building (VERB). The VERB modifications, including a new ventilation system and modified containment structure, required an approved Preliminary Documented Safety Analysis prior to project procurement and construction. Upgrade of the VERB from a radiological facility to a Hazard Category 3 Nuclear Facility required new rigor in the design and construction areas and was executed on an aggressive schedule. The facility Documented Safety Analysis required that OSBs be vented prior to introduction into the VERB. Box venting was safely completed after developing and implementing two types of custom venting systems for the heavy gauge box construction. A remotely operated punching process was used on boxes with wall thickness of up to 3.05 mm (0.120 in) to insert aluminum bronze filters and sample ports to prevent sparking during penetration. A remotely operated cold-drilling process with self-drilling, self-tapping titanium coated spark-resistant filters was used for boxes with wall thickness of up to 6.35 mm (0.25 in). The box headspace was sampled for the presence of flammable gases. To further accelerate the project schedule, an innovative treatment process was used. Several of the OSBs were re-assayed and determined to be mixed low-level waste (MLLW) which allowed treatment, followed by disposal in the Mixed Waste Disposal Unit at the NTS Area 5 Radioactive Waste Management Complex (RWMC). The MLLW boxes were certified using real-time radiography and overpacked into custom-built polyethylene-lined macroencapsulation containers. The polyethylene-lined lid was welded to the poly-lined box using automatically controlled resistance heating through embedded wiring in the lid. The work was performed under the existing Documented Safety Analysis since plastic welding is accomplished at low temperature and does not introduce the risks of other macroencapsulation processes, such as welding stainless steel containers. The macroencapsulation process for MLLW not only accelerated the schedule by reducing the number of boxes requiring size reduction, but it also resulted in significantly improved safety with as low as reasonable achievable levels of exposure to workers plus reduced cost by eliminating the need to perform repackaging in the VERB.

  11. Estimating Residual Solids Volume In Underground Storage Tanks

    SciTech Connect (OSTI)

    Clark, Jason L.; Worthy, S. Jason; Martin, Bruce A.; Tihey, John R.

    2014-01-08

    The Savannah River Site liquid waste system consists of multiple facilities to safely receive and store legacy radioactive waste, treat, and permanently dispose waste. The large underground storage tanks and associated equipment, known as the 'tank farms', include a complex interconnected transfer system which includes underground transfer pipelines and ancillary equipment to direct the flow of waste. The waste in the tanks is present in three forms: supernatant, sludge, and salt. The supernatant is a multi-component aqueous mixture, while sludge is a gel-like substance which consists of insoluble solids and entrapped supernatant. The waste from these tanks is retrieved and treated as sludge or salt. The high level (radioactive) fraction of the waste is vitrified into a glass waste form, while the low-level waste is immobilized in a cementitious grout waste form called saltstone. Once the waste is retrieved and processed, the tanks are closed via removing the bulk of the waste, chemical cleaning, heel removal, stabilizing remaining residuals with tailored grout formulations and severing/sealing external penetrations. The comprehensive liquid waste disposition system, currently managed by Savannah River Remediation, consists of 1) safe storage and retrieval of the waste as it is prepared for permanent disposition; (2) definition of the waste processing techniques utilized to separate the high-level waste fraction/low-level waste fraction; (3) disposition of LLW in saltstone; (4) disposition of the HLW in glass; and (5) closure state of the facilities, including tanks. This paper focuses on determining the effectiveness of waste removal campaigns through monitoring the volume of residual solids in the waste tanks. Volume estimates of the residual solids are performed by creating a map of the residual solids on the waste tank bottom using video and still digital images. The map is then used to calculate the volume of solids remaining in the waste tank. The ability to accurately determine a volume is a function of the quantity and quality of the waste tank images. Currently, mapping is performed remotely with closed circuit video cameras and still photograph cameras due to the hazardous environment. There are two methods that can be used to create a solids volume map. These methods are: liquid transfer mapping / post transfer mapping and final residual solids mapping. The task is performed during a transfer because the liquid level (which is a known value determined by a level measurement device) is used as a landmark to indicate solids accumulation heights. The post transfer method is primarily utilized after the majority of waste has been removed. This method relies on video and still digital images of the waste tank after the liquid transfer is complete to obtain the relative height of solids across a waste tank in relation to known and usable landmarks within the waste tank (cooling coils, column base plates, etc.). In order to accurately monitor solids over time across various cleaning campaigns, and provide a technical basis to support final waste tank closure, a consistent methodology for volume determination has been developed and implemented at SRS.

  12. Closed Fuel Cycle Waste Treatment Strategy

    SciTech Connect (OSTI)

    Vienna, J. D.; Collins, E. D.; Crum, J. V.; Ebert, W. L.; Frank, S. M.; Garn, T. G.; Gombert, D.; Jones, R.; Jubin, R. T.; Maio, V. C.; Marra, J. C.; Matyas, J.; Nenoff, T. M.; Riley, B. J.; Sevigny, G. J.; Soelberg, N. R.; Strachan, D. M.; Thallapally, P. K.; Westsik, J. H.

    2015-02-01

    This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) (Gombert et al. 2008) as a general strategy and associated Waste Treatment Baseline Study (WTBS) (Gombert et al. 2007). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significant additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to a representative full recycle reprocessing flowsheet. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. These applications are occasionally discussed where the data are more readily available. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include; Tritium in low-water cement in high integrity containers (HICs); Iodine-129: As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form with encapsulated nano-sized AgI crystals; Carbon-14 immobilized as a CaCO3 in a cement waste form; Krypton-85 stored as a compressed gas; An aqueous reprocessing high-level waste (HLW) raffinate waste immobilized by the vitrification process; An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel either included in the borosilicate HLW glass or immobilized in the form of a metal alloy or titanate ceramics; Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware super-compacted for disposal or purified for reuse (or disposal as low-level waste, LLW) of Zr by reactive gas separations; Electrochemical process salt HLW incorporated into a glass bonded Sodalite waste form; and Electrochemical process UDS and SS cladding hulls melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from a representative flowsheet are reported. In addition to the above listed primary waste streams, a range of secondary process wastes are generated by aqueous reprocessing of LWR fuel, metal SFR fuel fabrication, and electrochemical reprocessing of SFR fuel. These secondary wastes have been summarized and volumes estimated by type and classification. The important waste management data gaps and research needs have been summarized for each primary waste stream and selected waste process.

  13. Chooz A, First Pressurized Water Reactor to be Dismantled in France - 13445

    SciTech Connect (OSTI)

    Boucau, Joseph [Westinghouse Electric Company, 43 rue de l'Industrie, Nivelles (Belgium)] [Westinghouse Electric Company, 43 rue de l'Industrie, Nivelles (Belgium); Mirabella, C. [Westinghouse Electric France, Orsay (France)] [Westinghouse Electric France, Orsay (France); Nilsson, Lennart [Westinghouse Electric Sweden, Vaesteraas (Sweden)] [Westinghouse Electric Sweden, Vaesteraas (Sweden); Kreitman, Paul J. [Westinghouse Electric Company, Lake Bluff, IL 60048 (United States)] [Westinghouse Electric Company, Lake Bluff, IL 60048 (United States); Obert, Estelle [EDF - DPI - CIDEN, Lyon (France)] [EDF - DPI - CIDEN, Lyon (France)

    2013-07-01

    Nine commercial nuclear power plants have been permanently shut down in France to date, of which the Chooz A plant underwent an extensive decommissioning and dismantling program. Chooz Nuclear Power Station is located in the municipality of Chooz, Ardennes region, in the northeast part of France. Chooz B1 and B2 are 1,500 megawatt electric (MWe) pressurized water reactors (PWRs) currently in operation. Chooz A, a 305 MWe PWR implanted in two caves within a hill, began operations in 1967 and closed in 1991, and will now become the first PWR in France to be fully dismantled. EDF CIDEN (Engineering Center for Dismantling and Environment) has awarded Westinghouse a contract for the dismantling of its Chooz A reactor vessel (RV). The project began in January 2010. Westinghouse is leading the project in a consortium with Nuvia France. The project scope includes overall project management, conditioning of the reactor vessel (RV) head, RV and RV internals segmentation, reactor nozzle cutting for lifting the RV out of the pit and seal it afterwards, dismantling of the RV thermal insulation, ALARA (As Low As Reasonably Achievable) forecast to ensure acceptable doses for the personnel, complementary vacuum cleaner to catch the chips during the segmentation work, needs and facilities, waste characterization and packaging, civil work modifications, licensing documentation. The RV and RV internals will be segmented based on the mechanical cutting technology that Westinghouse applied successfully for more than 13 years. The segmentation activities cover the cutting and packaging plan, tooling design and qualification, personnel training and site implementation. Since Chooz A is located inside two caves, the project will involve waste transportation from the reactor cave through long galleries to the waste buffer area. The project will end after the entire dismantling work is completed, and the waste storage is outside the caves and ready to be shipped either to the ANDRA (French National Radioactive Waste Management Agency) waste disposal facilities - (for low-level waste [LLW] and very low-level waste [VLLW], which are considered short lived) - or to the EDF Interim Storage Facility planned to be built on another site - (for low- and intermediate-level waste [LILW], which is considered long lived). The project has started with a detailed conceptual study that determines the step-by-step approach for dismantling the reactor and eventually supplying the packed containers ready for final disposal. All technical reports must be verified and approved by EDF and the French Nuclear Safety Authority before receiving the authorization to start the site work. The detailed conceptual study has been completed to date and equipment design and manufacturing is ongoing. This paper will present the conceptual design of the reactor internals segmentation and packaging process that will be implemented at Chooz A, including the planning, methodology, equipment, waste management, and packaging strategy. (authors)

  14. Assessment of Potential Flood Events and Impacts at INL's Proposed Remote-Handled Low-Level Waste Disposal Facility Sites

    SciTech Connect (OSTI)

    A. Jeff Sondrup; Annette L. Schafter

    2010-09-01

    Rates, depths, erosion potential, increased subsurface transport rates, and annual exceedance probability for potential flooding scenarios have been evaluated for the on-site alternatives of Idaho National Laboratorys proposed remote handled low-level waste disposal facility. The on-site disposal facility is being evaluated in anticipation of the closure of the Radioactive Waste Management Complex at the INL. An assessment of flood impacts are required to meet the Department of Energys Low-Level Waste requirements (DOE-O 435.1), its natural phenomena hazards assessment criteria (DOE-STD-1023-95), and the Radioactive Waste Management Manual (DOE M 435.1-1) guidance in addition to being required by the National Environmental Policy Act (NEPA) environmental assessment (EA). Potential sources of water evaluated include those arising from (1) local precipitation events, (2) precipitation events occurring off of the INL (off-site precipitation), and (3) increased flows in the Big Lost River in the event of a Mackay Dam failure. On-site precipitation events include potential snow-melt and rainfall. Extreme rainfall events were evaluated for the potential to create local erosion, particularly of the barrier placed over the disposal facility. Off-site precipitation carried onto the INL by the Big Lost River channel was evaluated for overland migration of water away from the river channel. Off-site precipitation sources evaluated were those occurring in the drainage basin above Mackay Reservoir. In the worst-case scenarios, precipitation occurring above Mackay Dam could exceed the dams capacity, leading to overtopping, and eventually complete dam failure. Mackay Dam could also fail during a seismic event or as a result of mechanical piping. Some of the water released during dam failure, and contributing precipitation, has the potential of being carried onto the INL in the Big Lost River channel. Resulting overland flows from these flood sources were evaluated for their erosion potential, ability to overflow the proposed disposal facility, and for their ability to increase migration of contaminants from the facility. The assessment of available literature suggests that the likelihood of detrimental flood water impacting the proposed RH-LLW facility is extremely low. The annual exceedance probability associated with uncontrolled flows in the Big Lost River impacting either of the proposed sites is 1x10-5, with return interval (RI) of 10,000yrs. The most probable dam failure scenario has an annual exceedance probability of 6.3x10-6 (1.6x105 yr RI). In any of the scenarios generating possible on-site water, the duration is expected to be quite short, water depths are not expected to exceed 0.5 m, and the erosion potential can easily be mitigated by emplacement of a berm (operational period), and an engineered cover (post closure period). Subsurface mobilization of radionuclides was evaluated for a very conservative flooding scenario resulting in 50 cm deep, 30.5 day on-site water. The annual exceedance probability for which is much smaller than 3.6x10-7 (2.8x106 yr RI). For the purposes of illustration, the facility was assumed to flood every 500 years. The periodically recurring flood waters were predicted to marginally increase peak radionuclide fluxes into the aquifer by at most by a factor of three for non-sorbing radionuclides, and to have limited impact on peak radionuclide fluxes into the aquifer for contaminants that do sorb.

  15. SPECIAL ANALYSIS AIR PATHWAY MODELING OF E-AREA LOW-LEVEL WASTE FACILITY

    SciTech Connect (OSTI)

    Hiergesell, R.; Taylor, G.

    2011-08-30

    This Special Analysis (SA) was initiated to address a concern expressed by the Department of Energy's Low Level Waste Disposal Facility Federal Review Group (LFRG) Review Team during their review of the 2008 E-Area Performance Assessment (PA) (WSRC, 2008). Their concern was the potential for overlapping of atmospheric plumes, emanating from the soil surface above SRS LLW disposal facilities within the E-Area, to contribute to the dose received by a member of the public during the Institutional Control (IC) period. The implication of this concern was that the dose to the maximally-exposed individual (MEI) located at the SRS boundary might be underestimated during this time interval. To address this concern a re-analysis of the atmospheric pathway releases from E-Area was required. In the process of developing a new atmospheric release model (ARM) capable of addressing the LFRG plume overlap concern, it became obvious that new and better atmospheric pathway disposal limits should be developed for each of the E-Area disposal facilities using the new ARM. The scope of the SA was therefore expanded to include the generation of these new limits. The initial work conducted in this SA was to develop a new ARM using the GoldSim{reg_sign} program (GTG, 2009). The model simulates the subsurface vapor diffusion of volatile radionuclides as they release from E-Area disposal facility waste zones and migrate to the land surface. In the process of this work, many new features, including several new physical and chemical transport mechanisms, were incorporated into the model. One of the most important improvements was to incorporate a mechanism to partition volatile contaminants across the water-air interface within the partially saturated pore space of the engineered and natural materials through which vapor phase transport occurs. A second mechanism that was equally important was to incorporate a maximum concentration of 1.9E-07 Ci/m{sup 3} of {sup 14}CO{sub 2} in the air-filled pores of cementitious materials. The ARM also combines the individual transport models constructed for each E-Area disposal facility into a single model, and was ultimately used to analyze the LFRG concern regarding the potential for atmospheric plume overlap at the SRS boundary during the IC period. To evaluate the plume overlap issue, a conservative approach was adopted whereby the MEI at the SRS boundary was exposed to the releases from all E-Area disposal facilities simultaneously. This is equivalent to a 100% overlap of all atmospheric plumes emanating from E-Area. Should the dose received from this level of atmospheric plume overlap still fall below the permissible exposure level of 10 mrem/yr, then the LFRG concern would be alleviated. The structuring of the ARM enables this evaluation to be easily performed. During the IC period, the peak of the 'total plume overlap dose' was computed to be 1.9E-05 mrem/yr, which is five orders of magnitude lower than the 10 mrem/yr PA performance objective for the atmospheric release pathway. The main conclusion of this study is that for atmospheric releases from the E-Area disposal facilities, plume overlap does not cause the total dose to the MEI at the SRS boundary during the IC to exceed the Performance Assessment (PA) performance objective. Additionally, the potential for plume overlap was assessed in the post-Institutional Control period. Atmospheric plume overlap is less likely to occur during this period but conceivably could occur if the prevailing wind direction shifted so as to pass directly over all EArea disposal facilities and transport airborne radionuclides to the MEI at the 100 m point of compliance (POC). This concern was also demonstrated of little concern, as the maximum plume overlap dose was found to be 1.45E+00 mrem/yr (or {approx}15% of the performance measure) during this period and under these unlikely conditions.