National Library of Energy BETA

Sample records for vitrified hlw quantities

  1. EM Waste Acceptance Product Specification (WAPS) for Vitrified...

    Office of Environmental Management (EM)

    EM Waste Acceptance Product Specification (WAPS) for Vitrified High-Level Waste Forms Presentation to the HLW Corporate Board July 24, 2008 By Tony KlukKen Picha 2 Background * ...

  2. Vitrified waste option study report

    SciTech Connect (OSTI)

    Lopez, D.A.; Kimmitt, R.R.

    1998-02-01

    A {open_quotes}Settlement Agreement{close_quotes} between the Department of Energy and the State of Idaho mandates that all radioactive high-level waste (HLW) now stored at the Idaho Chemical Processing Plant (ICPP) will be treated so that it is ready to be moved out of Idaho for disposal by a target date of 2035. This report investigates vitrification treatment of all ICPP calcine, including the existing and future HLW calcine resulting from calcining liquid Sodium-Bearing Waste (SBW). Currently, the SBW is stored in the tank farm at the ICPP. Vitrification of these wastes is an acceptable treatment method for complying with the Settlement Agreement. This method involves vitrifying the calcined waste and casting the vitrified mass into stainless steel canisters that will be ready to be moved out of the Idaho for disposal by 2035. These canisters will be stored at the Idaho National Engineering and Environmental Laboratory (INEEL) until they are sent to a HLW national repository. The operating period for vitrification treatment will be from 2013 through 2032; all HLW will be treated and in storage by the end of 2032.

  3. Vitrified underground structures

    DOE Patents [OSTI]

    Murphy, Mark T. (Kennewick, WA); Buelt, James L. (Richland, WA); Stottlemyre, James A. (Richland, WA); Tixier, Jr., John S. (Richland, WA)

    1992-01-01

    A method of making vitrified underground structures in which 1) the vitrification process is started underground, and 2) a thickness dimension is controlled to produce substantially planar vertical and horizontal vitrified underground structures. Structures may be placed around a contaminated waste site to isolate the site or may be used as aquifer dikes.

  4. TWRS HLW interim storage facility search and evaluation

    SciTech Connect (OSTI)

    Calmus, R.B., Westinghouse Hanford

    1996-05-16

    The purpose of this study was to identify and provide an evaluation of interim storage facilities and potential facility locations for the vitrified high-level waste (HLW) from the Phase I demonstration plant and Phase II production plant. In addition, interim storage facilities for solidified separated radionuclides (Cesium and Technetium) generated during pretreatment of Phase I Low-Level Waste Vitrification Plant feed was evaluated.

  5. HLW Glass Waste Loadings

    Office of Environmental Management (EM)

    HLW Glass Waste Loadings Ian L. Pegg Vitreous State Laboratory The Catholic University of ... (JHCM) technology Factors affecting waste loadings Waste loading requirements ...

  6. PAIRWISE BLENDING OF HIGH LEVEL WASTE (HLW)

    SciTech Connect (OSTI)

    CERTA, P.J.

    2006-02-22

    The primary objective of this study is to demonstrate a mission scenario that uses pairwise and incidental blending of high level waste (HLW) to reduce the total mass of HLW glass. Secondary objectives include understanding how recent refinements to the tank waste inventory and solubility assumptions affect the mass of HLW glass and how logistical constraints may affect the efficacy of HLW blending.

  7. The NUMO Strategy for HLW and TRU Waste Disposal

    SciTech Connect (OSTI)

    Kitayama, K.; Oda, Y. [Nuclear Waste Management Organization of Japan (NUMO), Tokyo (Japan)

    2008-07-01

    Shortly after the Nuclear Waste Management Organization of Japan (NUMO) was established, we initiated an open call to all municipalities, requesting volunteers to host a repository for vitrified HLW. The first volunteer applied for a preliminary literature survey phase last January but, unfortunately, it withdrew the application in April. This failure provided an invaluable lesson for both the relevant authorities and NUMO; subsequently the Atomic Energy Commission of Japan and associated organizations are examining a support plan to back up NUMO's open solicitation. On another front, a recent amendment of 'The Specified Radioactive Waste Final Disposal Act' also allocates specific 'TRU' waste for deep geological disposal, requiring a demonstration of safety to a similar level as that for HLW. To implement the radioactive waste disposal project, NUMO has developed a methodology appropriate to our specific boundary conditions - the NUMO Structured Approach. This takes into account, in particular, our need to balance competing goals, such as operational safety, post-closure safety and cost, during repository tailoring to specific sites. The most important challenge for NUMO is, however, to attract volunteers. We believe that our open and structured R and D program is critical to demonstrate technical competence which, in turn, enhances the credibility of our various public relations activities. (authors)

  8. HLW Glass Studies: Development of Crystal-Tolerant HLW Glasses

    SciTech Connect (OSTI)

    Matyas, Josef; Huckleberry, Adam R.; Rodriguez, Carmen P.; Lang, Jesse B.; Owen, Antionette T.; Kruger, Albert A.

    2012-04-02

    In our study, a series of lab-scale crucible tests were performed on designed glasses of different compositions to further investigate and simulate the effect of Cr, Ni, Fe, Al, Li, and RuO2 on the accumulation rate of spinel crystals in the glass discharge riser of the HLW melter. The experimental data were used to expand the compositional region covered by an empirical model developed previously (Maty et al. 2010b), improving its predictive performance. We also investigated the mechanism for agglomeration of particles and impact of agglomerates on accumulation rate. In addition, the TL was measured as a function of temperature and composition.

  9. Perspectives of Future R and D on HLW Disposal in Germany

    SciTech Connect (OSTI)

    Steininger, W.J. [Forschungszentrum Karlsruhe GmbH, Project Management Agency Forschungszentrum Karlsruhe (PTKA-WTE), Eggenstein-Leopoldshafen (Germany)

    2008-07-01

    The 5. Energy Research Program of the Federal Government 'Innovation and New Technology' is the general framework for R and D activities in radioactive waste disposal. The Ministry of Economics and Technology (BMWi), the Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU) and the Ministry of Education and Research (BMBF) apply the Research Program concerning their respective responsibilities and competences. With regard to the Government's obligation to provide repositories for HLW (spent fuel and vitrified HAW) radioactive waste basic and applied R and D is needed in order to make adequate knowledge available to implementers, decision makers and stakeholders in general. Non-site specific R and D projects are funded by BMWi on the basis of its Research Concept. In the first stage (1998 -2001) most R and D issues were focused on R and D activities related to HLW disposal in rock salt. By that time the R and D program had to be revised and some prioritization was demanded due to changes in politics. In the current version (2001 -2006) emphasize was put on non-saline rocks. The current Research Concept of BMWi is presently subjected to a sort of revision, evaluation, and discussion, inter alia, by experts from several German research institutions. This activity is of special importance against the background of streamlining and focusing the research activities to future demands, priorities and perspectives with regard to the salt concept and the option of disposing of HLW in argillaceous media. Because the status of knowledge on disposal in rock salt is well advanced, it is necessary to take stock of the current state-of-the-art. In this framework some key projects are being currently carried out. The results may contribute to future decisions to be made in Germany with respect to HLW disposal. The first project deals with the development of an advanced safety concept for a HLW waste repository in rock salt. The second project (also carried out in the frame of the 6. Framework Program of the European Commission) aims at completing and optimizing the direct disposal concept for spent fuel by a full-scale demonstration of the technology of emplacement in vertical boreholes. The third project is devoted to the development of a reference concept to dispose of HLW in deep geological repository in clay in Germany. In the following a brief overview is given on the achievements, the projects, and ideas about the consequences for HLW disposal in Germany. (author)

  10. Reevaluation of Vitrified High-Level Waste Form Criteria for Potential Cost Savings at the Defense Waste Processing Facility - 13598

    SciTech Connect (OSTI)

    Ray, J.W. [Savannah River Remediation (United States)] [Savannah River Remediation (United States); Marra, S.L.; Herman, C.C. [Savannah River National Laboratory, Savannah River Site, Aiken, SC 29808 (United States)] [Savannah River National Laboratory, Savannah River Site, Aiken, SC 29808 (United States)

    2013-07-01

    At the Savannah River Site (SRS) the Defense Waste Processing Facility (DWPF) has been immobilizing SRS's radioactive high level waste (HLW) sludge into a durable borosilicate glass since 1996. Currently the DWPF has poured over 3,500 canisters, all of which are compliant with the U. S. Department of Energy's (DOE) Waste Acceptance Product Specifications for Vitrified High-Level Waste Forms (WAPS) and therefore ready to be shipped to a federal geologic repository for permanent disposal. Due to DOE petitioning to withdraw the Yucca Mountain License Application (LA) from the Nuclear Regulatory Commission (NRC) in 2010 and thus no clear disposal path for SRS canistered waste forms, there are opportunities for cost savings with future canister production at DWPF and other DOE producer sites by reevaluating high-level waste form requirements and compliance strategies and reducing/eliminating those that will not negatively impact the quality of the canistered waste form. (authors)

  11. Reevaluation Of Vitrified High-Level Waste Form Criteria For Potential Cost Savings At The Defense Waste Processing Facility

    SciTech Connect (OSTI)

    Ray, J. W.; Marra, S. L.; Herman, C. C.

    2013-01-09

    At the Savannah River Site (SRS) the Defense Waste Processing Facility (DWPF) has been immobilizing SRS's radioactive high level waste (HLW) sludge into a durable borosilicate glass since 1996. Currently the DWPF has poured over 3,500 canisters, all of which are compliant with the U. S. Department of Energy's (DOE) Waste Acceptance Product Specifications for Vitrified High-Level Waste Forms (WAPS) and therefore ready to be shipped to a federal geologic repository for permanent disposal. Due to DOE petitioning to withdraw the Yucca Mountain License Application (LA) from the Nuclear Regulatory Commission (NRC) in 2010 and thus no clear disposal path for SRS canistered waste forms, there are opportunities for cost savings with future canister production at DWPF and other DOE producer sites by reevaluating high-level waste form requirements and compliance strategies and reducing/eliminating those that will not negatively impact the quality of the canistered waste form.

  12. End of Year 2010 SNF & HLW Inventories | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    End of Year 2010 SNF & HLW Inventories End of Year 2010 SNF & HLW Inventories Map of the United States of America that shows the location of approximately 64,000 MTHM of Spent ...

  13. Hanford Makes Progress Toward Vitrifying Waste with Facility's

    Energy Savers [EERE]

    Groundbreaking | Department of Energy Makes Progress Toward Vitrifying Waste with Facility's Groundbreaking Hanford Makes Progress Toward Vitrifying Waste with Facility's Groundbreaking March 16, 2016 - 12:30pm Addthis Workers excavate for the Effluent Management Facility site at Hanford’s Waste Treatment and Immobilization Plant. Workers excavate for the Effluent Management Facility site at Hanford's Waste Treatment and Immobilization Plant. RICHLAND, Wash. - EM's Office of River

  14. HLW-OVP-94-00n High Level Waste Management Division HLW System...

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

    4-00n High Level Waste Management Division HLW System Plan Revision 3 (U) Westinghouse Savannah River Company Aiken, South Carolina May 31, 1994 Westinghouse Savannah River Company ...

  15. HIGH ALUMINUM HLW GLASSES FOR HANFORDS WTP

    SciTech Connect (OSTI)

    KRUGER AA; JOSEPH I; BOWMAN BW; GAN H; KOT W; MATLACK KS; PEGG IL

    2009-08-19

    The world's largest radioactive waste vitrification facility is now under construction at the United State Department of Energy's (DOE's) Hanford site. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is designed to treat nearly 53 million gallons of mixed hazardous and radioactive waste now residing in 177 underground storage tanks. This multi-decade processing campaign will be one of the most complex ever undertaken because of the wide chemical and physical variability of the waste compositions generated during the cold war era that are stored at Hanford. The DOE Office of River Protection (ORP) has initiated a program to improve the long-term operating efficiency of the WTP vitrification plants with the objective of reducing the overall cost of tank waste treatment and disposal and shortening the duration of plant operations. Due to the size, complexity and duration of the WTP mission, the lifecycle operating and waste disposal costs are substantial. As a result, gains in High Level Waste (HLW) and Low Activity Waste (LAW) waste loadings, as well as increases in glass production rate, which can reduce mission duration and glass volumes for disposal, can yield substantial overall cost savings. EnergySolutions and its long-term research partner, the Vitreous State Laboratory (VSL) of the Catholic University of America, have been involved in a multi-year ORP program directed at optimizing various aspects of the HLW and LAW vitrification flow sheets. A number of Hanford HLW streams contain high concentrations of aluminum, which is challenging with respect to both waste loading and processing rate. Therefore, a key focus area of the ORP vitrification process optimization program at EnergySolutions and VSL has been development of HLW glass compositions that can accommodate high Al{sub 2}O{sub 3} concentrations while maintaining high processing rates in the Joule Heated Ceramic Melters (JHCMs) used for waste vitrification at the WTP. This paper, reviews the achievements of this program with emphasis on the recent enhancements in Al{sub 2}O{sub 3} loadings in HLW glass and its processing characteristics. Glass formulation development included crucible-scale preparation and characterization of glass samples to assess compliance with all melt processing and product quality requirements, followed by small-scale screening tests to estimate processing rates. These results were used to down-select formulations for subsequent engineering-scale melter testing. Finally, further testing was performed on the DM1200 vitrification system installed at VSL, which is a one-third scale (1.20 m{sup 2}) pilot melter for the WTP HLW melters and which is fitted with a fully prototypical off-gas treatment system. These tests employed glass formulations with high waste loadings and Al{sub 2}O{sub 3} contents of {approx}25 wt%, which represents a near-doubling of the present WTP baseline maximum Al{sub 2}O{sub 3} loading. In addition, these formulations were processed successfully at glass production rates that exceeded the present requirements for WTP HLW vitrification by up to 88%. The higher aluminum loading in the HLW glass has an added benefit in that the aluminum leaching requirements in pretreatment are reduced, thus allowing less sodium addition in pretreatment, which in turn reduces the amount of LAW glass to be produced at the WTP. The impact of the results from this ORP program in reducing the overall cost and schedule for the Hanford waste treatment mission will be discussed.

  16. Waste Treatment and Immobilation Plant HLW Waste Vitrification...

    Office of Environmental Management (EM)

    6 Technology Readiness Assessment for the Waste Treatment and Immobilization Plant (WTP) HLW Waste Vitrification Facility L. Holton D. Alexander C. Babel H. Sutter J. Young August ...

  17. Widening the envelope of UK HLW vitrification - Experimental studies with high waste loadings and new product formulations on a full scale non-active vitrification plant

    SciTech Connect (OSTI)

    Short, R.; Gribble, N. [Nexia Solutions, Sellafield, Cumbria, CA20 1PG (United Kingdom); Riley, A. [Sellafield Ltd, Sellafield, Seascale, Cumbria, CA20 1PG, UK (United Kingdom)

    2008-07-01

    The Vitrification Test Rig is a full scale waste vitrification plant that processes non-radioactive liquid HLW simulants based on the active waste streams produced by the reprocessing plants in the UK. Previous work on the rig has primarily concerned increasing the operational envelopes for the active waste vitrification plants at Sellafield to accommodate higher throughputs of Blended waste streams, higher waste oxide incorporation rates in the vitrified products, and the incorporation of legacy waste streams from early reactor commissioning and reprocessing operations at Sellafield. Recent operations have focussed on four main areas; dilute liquid feeds, very high Magnox waste stream incorporation levels, alternative base glass formulations and providing an operational envelope for 28 %w/w Magnox waste vitrification. This paper details the work performed and the major findings of that work. In summary: The VTR has been successfully used to determine operational envelopes and product quality for several HLW feed variations that will allow WVP to increase overall plant throughput via increased waste loading in canisters, increased HLW feed rates or a combination of both. The VTR has also demonstrated the ability to go to waste incorporations, feed rates and glass compositions that are currently beyond WVP specified limits, but that are feasible for future vitrification regimes. In addition, the VTR has trialled dilute feeds similar to those that are likely to be received by WVP in the future and the data obtained from these experiments will allow WVP to prepare adequately for the high throughput challenge of such feeds. Furthermore, new equipment has been trialled on the VTR in water feed mode to determine its suitability and operational limitations for WVP. Future operations will, in the short term, be concerned with increasing the throughput of WVP and are likely to focus on HLW decommissioning operations waste streams in the longer term. (authors)

  18. Waste Treatment and Immobilation Plant HLW Waste Vitrification Facility

    Office of Environmental Management (EM)

    6 Technology Readiness Assessment for the Waste Treatment and Immobilization Plant (WTP) HLW Waste Vitrification Facility L. Holton D. Alexander C. Babel H. Sutter J. Young August 2007 Prepared by the U.S. Department of Energy Office of River Protection Richland, Washington, 99352 07-DESIGN-046 Technology Readiness Assessment for the Waste Treatment and Immobilization Plant (WTP) HLW Waste Vitrification Facility L. Holton D. Alexander C. Babel H. Sutter J. Young August 2007 Prepared by the U.S.

  19. Idaho Waste Vitrification Facilities Project Vitrified Waste Interim Storage Facility

    SciTech Connect (OSTI)

    Bonnema, Bruce Edward

    2001-09-01

    This feasibility study report presents a draft design of the Vitrified Waste Interim Storage Facility (VWISF), which is one of three subprojects of the Idaho Waste Vitrification Facilities (IWVF) project. The primary goal of the IWVF project is to design and construct a treatment process system that will vitrify the sodium-bearing waste (SBW) to a final waste form. The project will consist of three subprojects that include the Waste Collection Tanks Facility, the Waste Vitrification Facility (WVF), and the VWISF. The Waste Collection Tanks Facility will provide for waste collection, feed mixing, and surge storage for SBW and newly generated liquid waste from ongoing operations at the Idaho Nuclear Technology and Engineering Center. The WVF will contain the vitrification process that will mix the waste with glass-forming chemicals or frit and turn the waste into glass. The VWISF will provide a shielded storage facility for the glass until the waste can be disposed at either the Waste Isolation Pilot Plant as mixed transuranic waste or at the future national geological repository as high-level waste glass, pending the outcome of a Waste Incidental to Reprocessing determination, which is currently in progress. A secondary goal is to provide a facility that can be easily modified later to accommodate storage of the vitrified high-level waste calcine. The objective of this study was to determine the feasibility of the VWISF, which would be constructed in compliance with applicable federal, state, and local laws. This project supports the Department of Energys Environmental Management missions of safely storing and treating radioactive wastes as well as meeting Federal Facility Compliance commitments made to the State of Idaho.

  20. FEASIBILITY AND EXPEDIENCE TO VITRIFY NPP OPERATIONAL WASTE

    SciTech Connect (OSTI)

    LIFANOV, F.A.; OJOVAN, M.I.; STEFANOVSKY, S.V.; BURCL, R.

    2003-02-27

    Operational radioactive waste is generated during routine operation of NPP. Process waste is mainly generated by treatment of water from reactor or ancillaries including spent fuel storage pools and some decontamination operations. Typical process wastes of pressurized water reactors (PWR or WWER) are borated water concentrates, whereas typical process wastes of boiling and RBMK type reactors are water concentrates with no boron content. NPP operational wastes are classified as low and intermediate level waste (LILW). NPP operational waste must be solidified in order to ensure safe conditions of storage and disposal. Currently the most promising solidification method for this waste is the vitrification technology. Vitrification of NPP operational waste is a relative new option being developed for last years. Nevertheless there is already accumulated operational experience on vitrifying low and intermediate level waste in Russian Federation at Moscow SIA ''Radon'' vitrification plant. This plant uses the most advanced type induction high frequency melters that facilitate the melting process and significantly reduce the generation of secondary waste and henceforth the overall cost. The plant was put into operation by the end of 1999. It has three operating cold crucible melters with the overall capacity up to 75 kg/h. The vitrification technology comprises a few stages, starting with evaporation of excess water from liquid radioactive waste, followed by batch preparation, glass melting, and ending with vitrified waste blocks and some relative small amounts of secondary waste. First of all since the original waste contain as main component water, this water is removed from waste through evaporation. Then the remaining salt concentrate is mixed with necessary technological additives, thus a glass-forming batch is formed. The batch is fed into melters where the glass melting occurs. From here there are two streams: one is the glass melt containing the most part of radioactivity and second is the off gas flow, which contains off gaseous and aerosol airborne. The melt glass is fed into containers, which are slowly cooled in an annealing tunnel furnace to avoid accumulation of mechanical stresses in the glass. Containers with glass are the final processing product containing the overwhelming part of waste contaminants. The second stream from melter is directed to gas purification system, which is a rather complex system taking into account the necessity to remove from off gas not only radionuclides but also the chemical contaminants. Operation of this purification system leads to generation of a small amount of secondary waste. This waste stream slightly contaminated with volatilized radionuclides is recycled in the same technological scheme. As a result only non-radioactive materials are produced. They are either discharged into environment or reused. Based on the experience gained during operation of vitrification plant one can conclude on high efficiency achieved through vitrification method. Another significant argument on vitrifying NPP operational waste is the minimal impact of vitrified radioactive waste onto environment. Solidified waste shall be disposed of into a near surface disposal facility. Waste forms disposed of in a near-surface wet repository eventually come into contact with groundwater. Engineered structures used or designed to prevent or postpone such contact and the subsequent radionuclide release are complex and often too expensive. Vitrification technologies provide waste forms with excellent resistance to corrosion and gave the basic possibility of maximal simplification of engineered barrier systems. The most simple disposal option is to locate the vitrified waste form packages directly into earthen trenches provided the host rock has the necessary sorption and confinement properties. Such an approach will significantly make simpler the disposal facilities thus contributing both to enhancing safety and economic al efficiency.

  1. HLW Canister and Can-In-Canister Drop Calculation

    SciTech Connect (OSTI)

    H. Marr

    1999-09-15

    The purpose of this calculation is to evaluate the structural response of the standard high-level waste (HLW) canister and the HLW canister containing the cans of immobilized plutonium (''can-in-canister'' throughout this document) to the drop event during the handling operation. The objective of the calculation is to provide the structure parameter information to support the canister design and the waste handling facility design. Finite element solution is performed using the commercially available ANSYS Version (V) 5.4 finite element code. Two-dimensional (2-D) axisymmetric and three-dimensional (3-D) finite element representations for the standard HLW canister and the can-in-canister are developed and analyzed using the dynamic solver.

  2. HIGH LEVEL WASTE (HLW) VITRIFICATION EXPERIENCE IN THE US: APPLICATION OF GLASS PRODUCT/PROCESS CONTROL TO OTHERHLW AND HAZARDOUS WASTES

    SciTech Connect (OSTI)

    Jantzen, C; James Marra, J

    2007-09-17

    Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. At the Savannah River Site (SRS) actual HLW tank waste has successfully been processed to stringent product and process constraints without any rework into a stable borosilicate glass waste since 1996. A unique 'feed forward' statistical process control (SPC) has been used rather than statistical quality control (SQC). In SPC, the feed composition to the melter is controlled prior to vitrification. In SQC, the glass product is sampled after it is vitrified. Individual glass property models form the basis for the 'feed forward' SPC. The property models transform constraints on the melt and glass properties into constraints on the feed composition. The property models are mechanistic and depend on glass bonding/structure, thermodynamics, quasicrystalline melt species, and/or electron transfers. The mechanistic models have been validated over composition regions well outside of the regions for which they were developed because they are mechanistic. Mechanistic models allow accurate extension to radioactive and hazardous waste melts well outside the composition boundaries for which they were developed.

  3. Long-term management of high-level radioactive waste (HLW) and...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF) Long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF) ...

  4. Small Scale Mixing Demonstration Batch Transfer and Sampling Performance of Simulated HLW - 12307

    SciTech Connect (OSTI)

    Jensen, Jesse; Townson, Paul; Vanatta, Matt

    2012-07-01

    The ability to effectively mix, sample, certify, and deliver consistent batches of High Level Waste (HLW) feed from the Hanford Double Shell Tanks (DST) to the Waste treatment Plant (WTP) has been recognized as a significant mission risk with potential to impact mission length and the quantity of HLW glass produced. At the end of 2009 DOE's Tank Operations Contractor, Washington River Protection Solutions (WRPS), awarded a contract to EnergySolutions to design, fabricate and operate a demonstration platform called the Small Scale Mixing Demonstration (SSMD) to establish pre-transfer sampling capacity, and batch transfer performance data at two different scales. This data will be used to examine the baseline capacity for a tank mixed via rotational jet mixers to transfer consistent or bounding batches, and provide scale up information to predict full scale operational performance. This information will then in turn be used to define the baseline capacity of such a system to transfer and sample batches sent to WTP. The Small Scale Mixing Demonstration (SSMD) platform consists of 43'' and 120'' diameter clear acrylic test vessels, each equipped with two scaled jet mixer pump assemblies, and all supporting vessels, controls, services, and simulant make up facilities. All tank internals have been modeled including the air lift circulators (ALCs), the steam heating coil, and the radius between the wall and floor. The test vessels are set up to simulate the transfer of HLW out of a mixed tank, and collect a pre-transfer sample in a manner similar to the proposed baseline configuration. The collected material is submitted to an NQA-1 laboratory for chemical analysis. Previous work has been done to assess tank mixing performance at both scales. This work involved a combination of unique instruments to understand the three dimensional distribution of solids using a combination of Coriolis meter measurements, in situ chord length distribution measurements, and electro-resistive tomography. This current work utilized the same instruments to monitor simulated waste transfers. This paper will discuss some of the scaling compromises when it came to the scaled sampling system design, handling of large quantities of material for sampling, and present data for the discuss of likely behavior of the full scale DST based on scaling correlations using a scale ratio exponent (SRE) from 0.25 to 0.45 and the behavior observed in the SSMD platform. This does not establish a scaling factor for DST mixing using paired jet mixers but is an attempt to envelope the likely performance ranges in terms of certification sampling bias, certification sample root-mean-square-deviation, and bath to batch relative standard deviation. (authors)

  5. DEVELOPMENT OF GLASS MATRICES FOR HLW RADIOACTIVE WASTES

    SciTech Connect (OSTI)

    Jantzen, C.

    2010-03-18

    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 borosilicate glass or phosphate glass. One of the primary reasons that glass has become the most widely used immobilization media is the relative simplicity of the vitrification process, e.g. melt waste plus glass forming frit additives and cast. A second reason that glass has become widely used for HLW is that the short range order (SRO) and medium range order (MRO) found in glass atomistically bonds the radionuclides and governs the melt properties such as viscosity, resistivity, sulphate solubility. The molecular structure of glass controls contaminant/radionuclide release by establishing the distribution of ion exchange sites, hydrolysis sites, and the access of water to those sites. The molecular structure is flexible and hence accounts for the flexibility of glass formulations to waste variability. Nuclear waste glasses melt between 1050-1150 C which minimizes the volatility of radioactive components such as Tc{sup 99}, Cs{sup 137}, and I{sup 129}. Nuclear waste glasses have good long term stability including irradiation resistance. Process control models based on the molecular structure of glass have been mechanistically derived and have been demonstrated to be accurate enough to control the world's largest HLW Joule heated ceramic melter in the US since 1996 at 95% confidence.

  6. Water Quantity | Open Energy Information

    Open Energy Info (EERE)

    Quantity Jump to: navigation, search Retrieved from "http:en.openei.orgwindex.php?titleWaterQuantity&oldid612364" Feedback Contact needs updating Image needs updating...

  7. HLW Return from France to Germany - 15 Years of Experience in Public Acceptance and Technical Aspects - 12149

    SciTech Connect (OSTI)

    Graf, Wilhelm

    2012-07-01

    Since in 1984 the national reprocessing concept was abandoned the reprocessing abroad was the only existing disposal route until 1994. With the amendment of the Atomic Energy Act in 2001 spent fuel management changed completely since from 1 June 2005 any delivery of spent fuel to reprocessing plants was prohibited and the direct disposal of spent fuel became mandatory. Until 2005 the total amount of spent fuel to be reprocessed abroad added up to 6080 t HM, 5309 t HM thereof in France. The waste generated from reprocessing - alternatively an equivalent amount of radioactive material - has to be returned to the country of origin according to the commercial contracts signed between the German utilities and COGEMA, now AREVA NC, in France and BNFL, now INS in UK. In addition the German and the French government exchanged notes with the obligation of both sides to enable and support the return of reprocessing residues or equivalents to Germany. The return of high active vitrified waste from La Hague to the interim storage facility at Gorleben was demanding from the technical view i. e. the cask design and the transport. Unfortunately the Gorleben area served as a target for nuclear opponents from the first transport in 1996 to the latest one in 2011. The protection against sabotage of the railway lines and mass protests needed highly improved security measures. In France and Germany special working forces and projects have been set up to cope with this extraordinary situation. A complex transport organization was established to involve all parties in line with the German and French requirements during transport. The last transport of vitrified residues from France has been completed successfully so far thus confirming the efficiency of the applied measures. Over 15 years there was and still is worldwide no comparable situation it is still unique. Summing up, the exceptional project handling challenge that resulted from the continuous anti-nuclear civil disobedience in Germany over the whole 15-year long project running time could be faced efficiently. It has to be concluded that despite of all problems the anti-nuclear activities have caused so far, all transports of vitrified HLW have always been completed successfully by adapting the commonly established safety, security and public acceptance measures to the special conditions and needs in Germany and coordinating the activities of all parties involved but at the expense of high costs for industry and government and a challenging operational complexity. Apart from an anticipatory project planning a good communication between all involved industrial parties and the French and the German government was the key to the effective management of such shipments and to minimize the radiological, economic, environmental, public and political impact. The future will show how efficiently the gained experience can be used for further return projects which are to be realized since no reprocessed waste has yet been returned from UK and neither the medium-level nor the low-level radioactive waste has been transferred from France to Germany. (author)

  8. Melter Throughput Enhancements for High-Iron HLW

    SciTech Connect (OSTI)

    Kruger, A. A.; Gan, Hoa; Joseph, Innocent; Pegg, Ian L.; Matlack, Keith S.; Chaudhuri, Malabika; Kot, Wing

    2012-12-26

    This report describes work performed to develop and test new glass and feed formulations in order to increase glass melting rates in high waste loading glass formulations for HLW with high concentrations of iron. Testing was designed to identify glass and melter feed formulations that optimize waste loading and waste processing rate while meeting all processing and product quality requirements. The work included preparation and characterization of crucible melts to assess melt rate using a vertical gradient furnace system and to develop new formulations with enhanced melt rate. Testing evaluated the effects of waste loading on glass properties and the maximum waste loading that can be achieved. The results from crucible-scale testing supported subsequent DuraMelter 100 (DM100) tests designed to examine the effects of enhanced glass and feed formulations on waste processing rate and product quality. The DM100 was selected as the platform for these tests due to its extensive previous use in processing rate determination for various HLW streams and glass compositions.

  9. Vitrified chiral-nematic liquid crystalline films for selective reflection and circular polarization

    SciTech Connect (OSTI)

    Katsis, D.; Chen, P.H.M.; Mastrangelo, J.C.; Chen, S.H.; Blanton, T.N.

    1999-06-01

    Nematic and left-handed chiral-nematic liquid crystals comprising methoxybiphenylbenzoate and (S)-(-)-1-phenylethylamine pendants to a cyclohexane core were synthesized and characterized. Although pristine samples were found to be polycrystalline, thermal quenching following heating to and annealing at elevated temperatures permitted the molecular orders characteristic of liquid crystalline mesomorphism to be frozen in the glassy state. Left at room temperature for 6 months, the vitrified liquid crystalline films showed no evidence of recrystallization. An orientational order parameter of 0.65 was determined with linear dichroism of a vitrified nematic film doped with Exalite 428 at a mole fraction of 0.0025. Birefringence dispersion of a blank vitrified nematic film was determined using a phase-difference method complemented by Abbe refractometry. A series of vitrified chiral-nematic films were prepared to demonstrate selective reflection and circular polarization with a spectral region tunable from blue to the infrared region by varying the chemical composition. The experimentally measured circular polarization spectra were found to agree with the Good-Karali theory in which all four system parameters were determined a priori: optical birefringence, average refractive index, selective reflection wavelength, and film thickness.

  10. RECENT PROCESS IMPROVEMENTS TO INCREASE HLW THROUGHPUT AT THE DWPF

    SciTech Connect (OSTI)

    Herman, C

    2007-02-14

    The Savannah River Site's (SRS) Defense Waste Processing Facility (DWPF), the world's largest operating high level waste (HLW) vitrification plant, began stabilizing about 35 million gallons of SRS liquid radioactive waste by-product in 1996. The DWPF has since filled over 2000 canisters with about 4000 pounds of radioactive glass in each canister. In the past few years there have been several process and equipment improvements at the DWPF to increase the rate at which the waste can be stabilized. These improvements have either directly increased waste processing rates or have desensitized the process and therefore minimized process upsets and thus downtime. These improvements, which include glass former optimization, increased waste loading of the glass, the melter heated bellows liner, and glass surge protection software, will be discussed in this paper.

  11. EMPIRICAL MODEL FOR FORMULATION OF CRYSTAL-TOLERANT HLW GLASSES

    SciTech Connect (OSTI)

    KRUGER AA; MATYAS J; HUCKLEBERRY AR; VIENNA JD; RODRIGUEZ CA

    2012-03-07

    Historically, high-level waste (HLW) glasses have been formulated with a low liquideus temperature (T{sub L}), or temperature at which the equilibrium fraction of spinel crystals in the melt is below 1 vol % (T{sub 0.01}), nominally below 1050 C. These constraints cannot prevent the accumulation of large spinel crystals in considerably cooler regions ({approx} 850 C) of the glass discharge riser during melter idling and significantly limit the waste loading, which is reflected in a high volume of waste glass, and would result in high capital, production, and disposal costs. A developed empirical model predicts crystal accumulation in the riser of the melter as a function of concentration of spinel-forming components in glass, and thereby provides guidance in formulating crystal-tolerant glasses that would allow high waste loadings by keeping the spinel crystals small and therefore suspended in the glass.

  12. Long-term management of high-level radioactive waste (HLW) and spent

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    nuclear fuel (SNF) | Department of Energy Long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF) Long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF) GC-52 provides legal advice to DOE regarding the long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF). SNF is nuclear fuel that has been used as fuel in a reactor to generate nuclear energy but that has been removed from the reactor as no

  13. Technology survey for real-time monitoring of plutonium in a vitrifier off-gas system

    SciTech Connect (OSTI)

    Berg, J.M.; Veirs, D.K.

    1996-01-01

    We surveyed several promising measurement technologies for the real-time monitoring of plutonium in a vitrifier off-gas system. The vitrifier is being developed by Westinghouse Savannah River Corp. and will be used to demonstrate vitrification of plutonium dissolved in nitric acid for fissile material disposition. The risk of developing a criticality hazard in the off-gas processing equipment can be managed by using available measurement technologies. We identified several potential technologies and methods for detecting plutonium that are sensitive enough to detect the accumulation of a mass sufficient to form a criticality hazard. We recommend gross alpha-monitoring technologies as the most promising option for Westinghouse Savannah River Corp. to consider because that option appears to require the least additional development. We also recommend further consideration for several other technologies because they offer specific advantages and because gross alpha-monitoring could prove unsuitable when tested for this specific application.

  14. Stabilization of vitrified wastes: Task 4. Topical report, October 1994--September 1995

    SciTech Connect (OSTI)

    Nowok, J.W.; Pflughoeft-Hassett, D.F.; Hassett, D.J.; Hurley, J.P.

    1995-09-01

    The goal of this task was to work with private industry to refine existing vitrification processes to produce a more stable vitrified product. The initial objectives were to (1) demonstrate a waste vitrification procedure for enhanced stabilization of waste materials and (2) develop a testing protocol to understand the long-term leaching behavior of the stabilized waste form. The testing protocol was expected to be based on a leaching procedure called the synthetic groundwater leaching procedure (SGLP). This task will contribute to the US DOE`s identified technical needs in waste characterization, low-level mixed-waste processing, disposition technology, and improved waste forms. The proposed work was to proceed over 4 years in the following steps: literature surveys to aid in the selection and characterization of test mixtures for vitrification, characterization of optimized vitrified test wastes using advanced leaching protocols, and refinement and demonstration of vitrification methods leading to commercialization. For this year, literature surveys were completed, and computer modeling was performed to determine the feasibility of removing heavy metals from a waste during vitrification, thereby reducing the hazardous nature of the vitrified material and possibly producing a commercial metal concentrate. This report describes the following four subtasks: survey of vitrification technologies; survey of cleanup sites; selection and characterization of test mixtures for vitrification and crystallization; and selection of crystallization methods based on thermochemistry modeling.

  15. Redox Control For Hanford HLW Feeds VSL-12R2530-1, REV 0

    SciTech Connect (OSTI)

    Kruger, A. A. [Department of Energy, Office of River Protection, Richland, Washington (United States); Matlack, Keith S. [The Catholic University of America, Washington, DC (United States); Pegg, Ian L. [The Catholic University of America, Washington, DC (United States); Kot, Wing K. [The Catholic University of America, Washington, DC (United States); Joseph, Innocent [The Catholic University of America, Washington, DC (United States)

    2012-12-13

    The principal objectives of this work were to investigate the effects of processing simulated Hanford HLW at the estimated maximum concentrations of nitrates and oxalates and to identify strategies to mitigate any processing issues resulting from high concentrations of nitrates and oxalates. This report provides results for a series of tests that were performed on the DM10 melter system with simulated C-106/AY-102 HLW. The tests employed simulated HLW feeds containing variable amounts of nitrates and waste organic compounds corresponding to maximum concentrations proj ected for Hanford HLW streams in order to determine their effects on glass production rate, processing characteristics, glass redox conditions, melt pool foaming, and the tendency to form secondary phases. Such melter tests provide information on key process factors such as feed processing behavior, dynamic effects during processing, processing rates, off-gas amounts and compositions, foaming control, etc., that cannot be reliably obtained from crucible melts.

  16. Classroom simulation of public involvement in H.L.W. issues featuring STS concepts

    SciTech Connect (OSTI)

    Bieniawski, Z.T.

    1996-12-01

    This paper reports on a classroom experiment, now in its third year, conducted by the author for students involvement in High-Level Waste disposal. The project involved designing a methodology for HLW management in the United States.

  17. Practical Thermodynamic Quantities for Aqueous Vanadium- and...

    Office of Scientific and Technical Information (OSTI)

    Practical Thermodynamic Quantities for Aqueous Vanadium- and Iron-Based Flow Batteries. Citation Details In-Document Search Title: Practical Thermodynamic Quantities for Aqueous...

  18. FINAL REPORT TESTS ON THE DURAMELTER 1200 HLW PILOT MELTER SYSTEM USING AZ-101 HLW SIMULANTS VSL-02R0100-2 REV 1 2/17/03

    SciTech Connect (OSTI)

    KRUGER AA; MATLACK KS; KOT WK; BARDAKCI T; GONG W; D'ANGELO NA; SCHATZ TR; PEGG IL

    2011-12-29

    This document provides the final report on data and results obtained from a series of nine tests performed on the one-third scale DuraMelter{trademark} 1200 (DM1200) HLW Pilot Melter system that has been installed at VSL with an integrated prototypical off-gas treatment system. That system has replaced the DM1000 system that was used for HLW throughput testing during Part B1 [1]. Both melters have similar melt surface areas (1.2 m{sup 2}) but the DM1200 is prototypical of the present RPP-WTP HLW melter design whereas the DM1000 was not. These tests were performed under a corresponding RPP-WTP Test Specification and associated Test Plans. The nine tests reported here were preceded by an initial series of short-duration tests conducted to support the start-up and commissioning of this system. This report is a followup to the previously issued Preliminary Data Summary Reports. The DM1200 system was deployed for testing and confirmation of basic design, operability, flow sheet, and process control assumptions as well as for support of waste form qualification and permitting. These tests include data on processing rates, off-gas treatment system performance, recycle stream compositions, as well as process operability and reliability. Consequently, this system is a key component of the overall HLW vitrification development strategy. The primary objective of the present series of tests was to determine the effects of a variety of parameters on the glass production rate in comparison to the RPP-WTP HL W design basis of 400 kg/m{sup 2}/d. Previous testing on the DMIOOO system [1] concluded that achievement of that rate with simulants of projected WTP melter feeds (AZ-101 and C-106/AY-102) was unlikely without the use of bubblers. As part of those tests, the same feed that was used during the cold-commissioning of the West Valley Demonstration Project (WVDP) HLW vitrification system was run on the DM1000 system. The DM1000 tests reproduced the rates that were obtained at the larger WVDP facility, lending confidence to the tests results [1]. Since the inclusion or exclusion of a bubbler has significant design implications, the Project commissioned further tests to address this issue. In an effort to identify factors that might increase the glass production rate for projected WTP melter feeds, a subsequent series of tests was performed on the DM100 system. Several tests variables led to glass production rate increases to values significantly above the 400 kg/m2/d requirement. However, while small-scale melter tests are useful for screening relative effects, they tend to overestimate absolute glass production rates, particularly for un-bubbled tests. Consequently, when scale-up effects were taken into account, it was not clear that any of the variables investigated would conclusively meet the 400 kg/m{sup 2}/d requirement without bubbling. The present series of tests was therefore performed on the DM1200 one-third scale HLW pilot melter system to provide the required basis for a final decision on whether bubblers would be included in the HLW melter. The present tests employed the same AZ-101 waste simulant and glass composition that was used for previous testing for consistency and comparability with the results from the earlier tests.

  19. Disposal of defense spent fuel and HLW at the Idaho Chemical Processing Plant

    SciTech Connect (OSTI)

    Ermold, L.F.; Loo, H.H.; Klingler, R.D.; Herzog, J.D.; Knecht, D.A.

    1993-06-01

    Irradiated nuclear fuel has been reprocessed at the Idaho Chemical Processing Plant (ICPP) since 1953 to recover uranium-235 and krypton-85 for the US Department of Energy (DOE). The resulting acidic high-level radioactive waste (HLW) has been solidified to a calcine since 1963 and stored in stainless steel underground bins enclosed by concrete vaults. Several different types of unprocessed irradiated DOE-owned fuels are also in storage at the ICPP. In April, 1992, DOE announced that spent fuel would no longer be reprocessed to recover enriched uranium and called for a shutdown of the reprocessing facilities at the ICPP. A new Spent Fuel and HLW Technology Development program was subsequently initiated to develop technologies for immobilizing ICPP spent fuels and HLW for disposal, in accordance with the Nuclear Waste Policy Act. The Program elements include Systems Analysis, Graphite Fuel Disposal, Other Spent Fuel Disposal, Sodium-Bearing Liquid Waste Processing, Calcine Immobilization, and Metal Recycle/Waste Minimization. This paper presents an overview of the ICPP radioactive wastes and current spent fuels, with an emphasis on the description of HLW and spent fuels requiring repository disposal.

  20. HLW Melter Control Strategy Without Visual Feedback VSL-12R2500-1 Rev 0

    SciTech Connect (OSTI)

    Kruger, A A.; Joseph, Innocent; Matlack, Keith S.; Callow, Richard A.; Abramowitz, Howard; Pegg, Ian L.; Brandys, Marek; Kot, Wing K.

    2012-11-13

    Plans for the treatment of high level waste (HL W) at the Hanford Tank Waste Treatment and Immobilization Plant (WTP) are based upon the inventory of the tank wastes, the anticipated performance of the pretreatment processes, and current understanding of the capability of the borosilicate glass waste form [I]. The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat and mass transfer and increase glass melting rates. The WTP HLW melter has a glass surface area of 3.75 m{sup 2} and depth of ~ 1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HL W waste processing rates can potentially be achieved by increasing the melter operating temperature above 1150?C and by increasing the waste loading in the glass product. Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage.

  1. Interagency cooperation in the development of a cost-effective transportation and disposal solution for vitrified radium bearing material

    SciTech Connect (OSTI)

    Smith, M.L.; Nixon, D.A.; Stone, T.J.; Tope, W.G.; Vogel, R.A.; Allen, R.B.; Schofield, W.D.

    1996-02-01

    Fernald radium bearing ore residue waste, stored within Silos 1 and 2 (K-65) and Silo 3 waste, will be vitrified for disposal at the Nevada Test Site (NTS). A comprehensive, parametric evaluation of waste form, shielding requirements, packaging, and transportation alternatives was completed to identify the safest, most cost-effective approach. The impacts of waste loading, waste form, regulatory requirements, NTS waste acceptance criteria, as-low-as-resonably-achievable principles, and material handling costs were factored into the recommended approach. Through cooperative work between the U.S. Department of Energy (DOE) and the U.S. Department of Transportation (DOT), the vitrified K-65 and Silo 3 radioactive material will be classified consistent with the regulations promulgated by DOT in the September 28, 1995 Federal Register. These new regulations adopt International Atomic Energy Agency language to promote a consistent approach for the transportation and management of radioactive material between the international community and the DOT. Use of the new regulations allows classification of the vitrified radioactive material from the Fernald silos under the designation of low specific activity-II and allows the development of a container that is optimized to maximize payload while minimizing internal void space, external surface radiation levels, and external volume. This approach minimizes the required number of containers and shipments, and the related transportation and disposal costs.

  2. A Review of Iron Phosphate Glasses and Recommendations for Vitrifying Hanford Waste

    SciTech Connect (OSTI)

    Delbert E. Ray; Chandra S. Ray

    2013-11-01

    This report contains a comprehensive review of the research conducted, world-wide, on iron phosphate glass over the past ~30 years. Special attention is devoted to those iron phosphate glass compositions which have been formulated for the purpose of vitrifying numerous types of nuclear waste, with special emphasis on the wastes stored in the underground tanks at Hanford WA. Data for the structural, chemical, and physical properties of iron phosphate waste forms are reviewed for the purpose of understanding their (a) outstanding chemical durability which meets all current DOE requirements, (b) high waste loadings which can exceed 40 wt% (up to 75 wt%) for several Hanford wastes, (c) low melting temperatures, can be as low as 900°C for certain wastes, and (d) high tolerance for “problem” waste components such as sulfates, halides, and heavy metals (chromium, actinides, noble metals, etc.). Several recommendations are given for actions that are necessary to smoothly integrate iron phosphate glass technology into the present waste treatment plans and vitrification facilities at Hanford.

  3. Methods of vitrifying waste with low melting high lithia glass compositions

    DOE Patents [OSTI]

    Jantzen, Carol M.; Pickett, John B.; Cicero-Herman, Connie A.; Marra, James C.

    2001-01-01

    The invention relates to methods of vitrifying waste and for lowering the melting point of glass forming systems by including lithia formers in the glass forming composition in significant amounts, typically from about 0.16 wt % to about 11 wt %, based on the total glass forming oxides. The lithia is typically included as a replacement for alkali oxide glass formers that would normally be present in a particular glass forming system. Replacement can occur on a mole percent or weight percent basis, and typically results in a composition wherein lithia forms about 10 wt % to about 100 wt % of the alkali oxide glass formers present in the composition. The present invention also relates to the high lithia glass compositions formed by these methods. The invention is useful for stabilization of numerous types of waste materials, including aqueous waste streams, sludge solids, mixtures of aqueous supernate and sludge solids, combinations of spent filter aids from waste water treatment and waste sludges, supernate alone, incinerator ash, incinerator offgas blowdown, or combinations thereof, geological mine tailings and sludges, asbestos, inorganic filter media, cement waste forms in need of remediation, spent or partially spent ion exchange resins or zeolites, contaminated soils, lead paint, etc. The decrease in melting point achieved by the present invention desirably prevents volatilization of hazardous or radioactive species during vitrification.

  4. Collaboration, Automation, and Information Management at Hanford High Level Radioactive Waste (HLW) Tank Farms

    SciTech Connect (OSTI)

    Aurah, Mirwaise Y.; Roberts, Mark A.

    2013-12-12

    Washington River Protection Solutions (WRPS), operator of High Level Radioactive Waste (HLW) Tank Farms at the Hanford Site, is taking an over 20-year leap in technology, replacing systems that were monitored with clipboards and obsolete computer systems, as well as solving major operations and maintenance hurdles in the area of process automation and information management. While WRPS is fully compliant with procedures and regulations, the current systems are not integrated and do not share data efficiently, hampering how information is obtained and managed.

  5. A decision-making students project for site selection of HLW repository

    SciTech Connect (OSTI)

    Bieniawski, Z.T.

    1995-12-01

    With the escalating costs, constant delays and adverse public opinion, the YMP is not only under fire but the DOE and the other parties involved have failed to convince the public that they know what they are doing and that theirs is the best way. For example, while the whole world is evaluating alternative HLW disposal sites, why is the United States characterizing just one site? A student class at Penn State decided that they may have a better concept of what should be done.

  6. A Review and Analysis of European Industrial Experience in Handling LWR Spent Fuel and Vitrified High-Level Waste

    SciTech Connect (OSTI)

    Blomeke, J.O.

    2001-07-10

    The industrial facilities that have been built or are under construction in France, the United Kingdom, Sweden, and West Germany to handle light-water reactor (LWR) spent fuel and canisters of vitrified high-level waste before ultimate disposal are described and illustrated with drawings and photographs. Published information on the operating performance of these facilities is also given. This information was assembled for consideration in planning and design of similar equipment and facilities needed for the Federal Waste Management System in the United States.

  7. MELT RATE ENHANCEMENT FOR HIGH ALUMINUM HLW (HIGH LEVEL WASTE) GLASS FORMULATION FINAL REPORT 08R1360-1

    SciTech Connect (OSTI)

    KRUGER AA; MATLACK KS; KOT W; PEGG IL; JOSEPH I; BARDAKCI T; GAN H; GONG W; CHAUDHURI M

    2010-01-04

    This report describes the development and testing of new glass formulations for high aluminum waste streams that achieve high waste loadings while maintaining high processing rates. The testing was based on the compositions of Hanford High Level Waste (HLW) with limiting concentrations of aluminum specified by the Office of River Protection (ORP). The testing identified glass formulations that optimize waste loading and waste processing rate while meeting all processing and product quality requirements. The work included preparation and characterization of crucible melts and small scale melt rate screening tests. The results were used to select compositions for subsequent testing in a DuraMelter 100 (DM100) system. These tests were used to determine processing rates for the selected formulations as well as to examine the effects of increased glass processing temperature, and the form of aluminum in the waste simulant. Finally, one of the formulations was selected for large-scale confirmatory testing on the HLW Pilot Melter (DM1200), which is a one third scale prototype of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) HLW melter and off-gas treatment system. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for Department of Energy (DOE) to increase waste loading and processing rates for high-iron HLW waste streams as well as previous tests conducted for ORP on the same high-aluminum waste composition used in the present work and other Hanford HLW compositions. The scope of this study was outlined in a Test Plan that was prepared in response to an ORP-supplied statement of work. It is currently estimated that the number of HLW canisters to be produced in the WTP is about 13,500 (equivalent to 40,500 MT glass). This estimate is based upon the inventory of the tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form. The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat transfer and glass melting rate. The WTP HLW melter has a glass surface area of 3.75 m{sup 2} and depth of {approx}1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HLW waste processing rates can potentially be achieved by increasing the melter operating temperature above 1150 C and by increasing the waste loading in the glass product. Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage. The current estimates and glass formulation efforts have been conservative in terms of achievable waste loadings. These formulations have been specified to ensure that the glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet WTP Contract terms. The WTP's overall mission will require the immobilization of tank waste compositions that are dominated by mixtures of aluminum (Al), chromium (Cr), bismuth (Bi), iron (Fe), phosphorous (P), zirconium (Zr), and sulfur (S) compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. Recently, DOE has initiated a testing program to develop and characterize HLW glasses with higher waste loadings. Results of this work have demonstrated the feasibility of increases in wasteloading from about 25 wt% to 33-50 wt% (based on oxide loading) in the glass depending on the waste stream. It is expected that these higher waste loading glasses will reduce the HLW canister production requirement by about 25% or more.

  8. The Production of Advanced Glass Ceramic HLW Forms using Cold Crucible Induction Melter

    SciTech Connect (OSTI)

    Veronica J Rutledge; Vince Maio

    2013-10-01

    Cold Crucible Induction Melters (CCIMs) will favorably change how High-Level radioactive Waste (from nuclear fuel recovery) is treated in the 21st century. Unlike the existing Joule-Heated Melters (JHMs) currently in operation for the glass-based immobilization of High-Level Waste (HLW), CCIMs offer unique material features that will increase melt temperatures, increase throughput, increase mixing, increase loading in the waste form, lower melter foot prints, eliminate melter corrosion and lower costs. These features not only enhance the technology for producing HLW forms, but also provide advantageous attributes to the waste form by allowing more durable alternatives to glass. This paper discusses advantageous features of the CCIM, with emphasis on features that overcome the historical issues with the JHMs presently utilized, as well as the benefits of glass ceramic waste forms over borosilicate glass waste forms. These advantages are then validated based on recent INL testing to demonstrate a first-of-a-kind formulation of a non-radioactive ceramic-based waste form utilizing a CCIM.

  9. Method and Apparatus for Measuring Radiation Quantities

    DOE Patents [OSTI]

    Roberts, N.O.

    1955-01-25

    This patent application describes a compact dosimeter for measuring X-ray and gamma radiation by the use of solutions which undergo a visible color change upon exposure to a predetermined quantity of radiation.

  10. DM100 AND DM1200 MELTER TESTING WITH HIGH WASTE LOADING GLASS FORMULATIONS FOR HANFORD HIGH-ALUMINUM HLW STREAMS

    SciTech Connect (OSTI)

    KRUGER AA; MATLACK KS; KOT WK; PEGG IL; JOSEPH I

    2009-12-30

    This Test Plan describes work to support the development and testing of high waste loading glass formulations that achieve high glass melting rates for Hanford high aluminum high level waste (HLW). In particular, the present testing is designed to evaluate the effect of using low activity waste (LAW) waste streams as a source of sodium in place ofchemical additives, sugar or cellulose as a reductant, boehmite as an aluminum source, and further enhancements to waste processing rate while meeting all processing and product quality requirements. The work will include preparation and characterization of crucible melts in support of subsequent DuraMelter 100 (DM 100) tests designed to examine the effects of enhanced glass formulations, glass processing temperature, incorporation of the LAW waste stream as a sodium source, type of organic reductant, and feed solids content on waste processing rate and product quality. Also included is a confirmatory test on the HLW Pilot Melter (DM1200) with a composition selected from those tested on the DM100. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for Department of Energy's (DOE's) Office of River Protection (ORP) to increase waste loading and processing rates for high-iron HLW waste streams as well as previous tests conducted for ORP on the same waste composition. This Test Plan is prepared in response to an ORP-supplied statement of work. It is currently estimated that the number of HLW canisters to be produced in the Hanford Tank Waste Treatment and Immobilization Plant (WTP) is about 12,500. This estimate is based upon the inventory ofthe tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form. The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat transfer and glass melting rate. The WTP HLW melter has a glass surface area of 3.75 m{sup 2} and depth of {approx}1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HLW waste processing rates can potentially be achieved by increasing the melter operating temperature above 1150 C and by increasing the waste loading in the glass product Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage. The current estimates and glass formulation efforts have been conservative in terms of achievable waste loadings. These formulations have been specified to ensure that the glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet WTP contract requirements. The WTP's overall mission will require the immobilization oftank waste compositions that are dominated by mixtures of aluminum (Al), chromium (Cr), bismuth (Bi), iron (Fe), phosphorous (P), zirconium (Zr), and sulfur (S) compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. Recently, DOE has initiated a testing program to develop and characterize HLW glasses with higher waste loadings. Results of this work have demonstrated the feasibility of increases in waste-loading from about 25 wt% to 33-50 wt% (based on oxide loading) in the glass depending on the waste stream. It is expected that these higher waste loading glasses will reduce the HLW canister production requirement by about 25% or more.

  11. Concept and Design of the JAEA KMS for Geological Disposal of HLW

    SciTech Connect (OSTI)

    Makino, Hitoshi; Osawa, Hideaki; Nakano, Katsushi; Naito, Morimasa; Umeki, Hiroyuki; Takase, Hiroyasu; McKinley, Ian G.

    2007-07-01

    The information explosion resulting from modern technology is identified as a critical problem for deep geological disposal of high-level radioactive waste (HLW). A paradigm shift is needed in the basic concept for information management. This recognition had led to the development of a 'next generation' Knowledge Management System (the JAEA KMS) that makes maximum use of recent developments in Information Technology (IT) and the methodology of Knowledge Engineering (KE) as applied in other technical fields. This paper provides a brief outline of the key concepts of the JAEA KMS and then overviews recent progress towards development of an operational system, including a 'wish list' of expected functions of the JAEA KMS, a perspective on applicability of existing methodologies and an introduction to the concept of an 'intelligent assistant'. (authors)

  12. INTEGRATED DM 1200 MELTER TESTING OF HLW C-106/AY-102 COMPOSITION USING BUBBLERS VSL-03R3800-1 REV 0 9/15/03

    SciTech Connect (OSTI)

    KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D'ANGELO NA; KOT WK; PEGG IL

    2011-12-29

    This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of simulated HLW C-106/AY-102 feed. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW C-106/AY-102 feed; determine the effect of bubbling rate on production rate; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and to perform pre- and post test inspections of system components.

  13. DETERMINATION OF HLW GLASS MELT RATE USING X-RAY COMPUTED TOMOGRAPHY

    SciTech Connect (OSTI)

    Choi, A.; Miller, D.; Immel, D.

    2011-10-06

    The purpose of the high-level waste (HLW) glass melt rate study is two-fold: (1) to gain a better understanding of the impact of feed chemistry on melt rate through bench-scale testing, and (2) to develop a predictive tool for melt rate in support of the on-going frit development efforts for the Defense Waste Processing Facility (DWPF). In particular, the focus is on predicting relative melt rates, not the absolute melt rates, of various HLW glass formulations solely based on feed chemistry, i.e., the chemistry of both waste and glass-forming frit for DWPF. Critical to the successful melt rate modeling is the accurate determination of the melting rates of various HLW glass formulations. The baseline procedure being used at the Savannah River National Laboratory (SRNL) is to; (1) heat a 4 inch-diameter stainless steel beaker containing a mixture of dried sludge and frit in a furnace for a preset period of time, (2) section the cooled beaker along its diameter, and (3) measure the average glass height across the sectioned face using a ruler. As illustrated in Figure 1-1, the glass height is measured for each of the 16 horizontal segments up to the red lines where relatively large-sized bubbles begin to appear. The linear melt rate (LMR) is determined as the average of all 16 glass height readings divided by the time during which the sample was kept in the furnace. This 'visual' method has proved useful in identifying melting accelerants such as alkalis and sulfate and further ranking the relative melt rates of candidate frits for a given sludge batch. However, one of the inherent technical difficulties of this method is to determine the glass height in the presence of numerous gas bubbles of varying sizes, which is prevalent especially for the higher-waste-loading glasses. That is, how the red lines are drawn in Figure 1-1 can be subjective and, therefore, may influence the resulting melt rates significantly. For example, if the red lines are drawn too low, a significant amount of glassy material interspersed among the gas bubbles will be excluded, thus underestimating the melt rate. Likewise, if they are drawn too high, many large voids will be counted as glass, thus overestimating the melt rate. As will be shown later in this report, there is also no guarantee that a given distribution of glass and gas bubbles along a particular sectioned plane will always be representative of the entire sample volume. Poor reproducibility seen in some LMR data may be related to these difficulties of the visual method. In addition, further improvement of the existing melt rate model requires that the overall impact of feed chemistry on melt rate be reflected on measured data at a greater quantitative resolution on a more consistent basis than the visual method can provide. An alternate method being pursued is X-ray computed tomography (CT). It involves X-ray scanning of glass samples, performing CT on the 2-D X-ray images to build 3-D volumetric data, and adaptive segmentation analysis of CT results to not only identify but quantify the distinct regions within each sample based on material density and morphologies. The main advantage of this new method is that it can determine the relative local density of the material remaining in the beaker after the heat treatment regardless of its morphological conditions by selectively excluding all the voids greater than a given volumetric pixel (voxel) size, thus eliminating much of the subjectivity involved in the visual method. As a result, the melt rate data obtained from CT scan will give quantitative descriptions not only on the fully-melted glass, but partially-melted and unmelted feed materials. Therefore, the CT data are presumed to be more reflective of the actual melt rate trends in continuously-fed melters than the visual data. In order to test the applicability of X-ray CT scan to the HLW glass melt rate study, several new series of HLW simulant/frit mixtures were melted in the Melt Rate Furnace (MRF) and the contents of each cooled but un-sectioned beaker were CT scanned and analyzed.

  14. New York Quantity of Production Associated with Reported Wellhead...

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) New York Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet)...

  15. South Dakota Quantity of Production Associated with Reported...

    Gasoline and Diesel Fuel Update (EIA)

    Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) South Dakota Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet)...

  16. U.S. Quantity of Production Associated with Reported Wellhead...

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) U.S. Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade...

  17. North Dakota Quantity of Production Associated with Reported...

    U.S. Energy Information Administration (EIA) Indexed Site

    Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) North Dakota Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet)...

  18. FINAL REPORT START-UP AND COMMISSIONING TESTS ON THE DURAMELTER 1200 HLW PILOT MELTER SYSTEM USING AZ-101 HLW SIMULANTS VSL-01R0100-2 REV 0 1/20/03

    SciTech Connect (OSTI)

    KRUGER AA; MATLACK KS; KOT WK; BRANDYS M; WILSON CN; SCHATZ TR; GONG W; PEGG IL

    2011-12-29

    This document provides the final report on data and results obtained from commissioning tests performed on the one-third scale DuraMelter{trademark} 1200 (DM 1200) HLW Pilot Melter system that has been installed at VSL with an integrated prototypical off-gas treatment system. That system has replaced the DM1000 system that was used for HLW throughput testing during Part BI [1]. Both melters have similar melt surface areas (1.2 m{sup 2}) but the DM1200 is prototypical of the present RPP-WTP HLW melter design whereas the DM1000 was not. These tests were performed under a corresponding RPP-WTP Test Specification and associated Test Plan. This report is a followup to the previously issued Preliminary Data Summary Report. The DM1200 system will be used for testing and confirmation of basic design, operability, flow sheet, and process control assumptions as well as for support of waste form qualification and permitting. This will include data on processing rates, off-gas treatment system performance, recycle stream compositions, as well as process operability and reliability. Consequently, this system is a key component of the overall HLW vitrification development strategy. The results presented in this report are from the initial series of short-duration tests that were conducted to support the start-up and commissioning of this system prior to conducting the main body of development tests that have been planned for this system. These tests were directed primarily at system 'debugging,' operator training, and procedure refinement. The AZ-101 waste simulant and glass composition that was used for previous testing was selected for these tests.

  19. Final Report - Sulfate Solubility in RPP-WTP HLW Glasses, VSL-06R6780-1, Rev. 0

    SciTech Connect (OSTI)

    Kruger, Albert A.; Pegg, I. L.; Feng, A.; Gan, H.; Kot, W. K.

    2013-12-03

    This report describes the results of work and testing specified by Test Specifications 24590-HLW-TSP-RT-01-006 Rev 1, Test Plans VSL-02T7800-1 Rev 1 and Test Exceptions 24590-HLW-TEF-RT-05-00007. The work and any associated testing followed established quality assurance requirements and were conducted as authorized. The descriptions provided in this report are an accurate account of both the conduct of the work and the data collected. Results required by the Test Plans are reported. Also reported are any unusual or anomalous occurrences that are different from the starting hypotheses. The test results and this report have been reviewed and verified.

  20. Final Report - Testing of Optimized Bubbler Configuration for HLW Melter VSL-13R2950-1, Rev. 0, dated 6/12/2013

    SciTech Connect (OSTI)

    Kruger, Albert A.; Pegg, I. L.; Callow, R. A.; Joseph, I.; Matlack, K. S.; Kot, W. K.

    2013-11-13

    The principal objective of this work was to determine the glass production rate increase and ancillary effects of adding more bubbler outlets to the current WTP HLW melter baseline. This was accomplished through testing on the HLW Pilot Melter (DM1200) at VSL. The DM1200 unit was selected for these tests since it was used previously with several HLW waste streams including the four tank wastes proposed for initial processing at Hanford. This melter system was also used for the development and optimization of the present baseline WTP HLW bubbler configuration for the WTP HLW melter, as well as for MACT testing for both HLW and LAW. Specific objectives of these tests were to: Conduct DM1200 melter testing with the baseline WTP bubbling configuration and as augmented with additional bubblers. Conduct DM1200 melter testing to differentiate the effects of total bubbler air flow and bubbler distribution on glass production rate and cold cap formation. Collect melter operating data including processing rate, temperatures at a variety of locations within the melter plenum space, melt pool temperature, glass melt density, and melter pressure with the baseline WTP bubbling configuration and as augmented with additional bubblers. Collect melter exhaust samples to compare particulate carryover for different bubbler configurations. Analyze all collected data to determine the effects of adding more bubblers to the WTP HLW melter to inform decisions regarding future lid re-designs. The work used a high aluminum HLW stream composition defined by ORP, for which an appropriate simulant and high waste loading glass formulation were developed and have been previously processed on the DM1200.

  1. Final Report - Crystal Settling, Redox, and High Temperature Properties of ORP HLW and LAW Glasses, VSL-09R1510-1, Rev. 0, dated 6/18/09

    SciTech Connect (OSTI)

    Kruger, Albert A.; Wang, C.; Gan, H.; Pegg, I. L.; Chaudhuri, M.; Kot, W.; Feng, Z.; Viragh, C.; McKeown, D. A.; Joseph, I.; Muller, I. S.; Cecil, R.; Zhao, W.

    2013-11-13

    The radioactive tank waste treatment programs at the U. S. Department of Energy (DOE) have featured joule heated ceramic melter technology for the vitrification of high level waste (HLW). The Hanford Tank Waste Treatment and Immobilization Plant (WTP) employs this same basic technology not only for the vitrification of HLW streams but also for the vitrification of Low Activity Waste (LAW) streams. Because of the much greater throughput rates required of the WTP as compared to the vitrification facilities at the West Valley Demonstration Project (WVDP) or the Defense Waste Processing Facility (DWPF), the WTP employs advanced joule heated melters with forced mixing of the glass pool (bubblers) to improve heat and mass transport and increase melting rates. However, for both HLW and LAW treatment, the ability to increase waste loadings offers the potential to significantly reduce the amount of glass that must be produced and disposed and, therefore, the overall project costs. This report presents the results from a study to investigate several glass property issues related to WTP HLW and LAW vitrification: crystal formation and settling in selected HLW glasses; redox behavior of vanadium and chromium in selected LAW glasses; and key high temperature thermal properties of representative HLW and LAW glasses. The work was conducted according to Test Plans that were prepared for the HLW and LAW scope, respectively. One part of this work thus addresses some of the possible detrimental effects due to considerably higher crystal content in waste glass melts and, in particular, the impact of high crystal contents on the flow property of the glass melt and the settling rate of representative crystalline phases in an environment similar to that of an idling glass melter. Characterization of vanadium redox shifts in representative WTP LAW glasses is the second focal point of this work. The third part of this work focused on key high temperature thermal properties of representative WTP HLW and LAW glasses over a wide range of temperatures, from the melter operating temperature to the glass transition.

  2. Final Report - Management of High Sulfur HLW, VSL-13R2920-1, Rev. 0, dated 10/31/2013

    SciTech Connect (OSTI)

    Kruger, Albert A.; Gan, H.; Pegg, I. L.; Feng, Z.; Gan, H,; Joseph, I.; Matlack, K. S.

    2013-11-13

    The present report describes results from a series of small-scale crucible tests to determine the extent of corrosion associated with sulfur containing HLW glasses and to develop a glass composition for a sulfur-rich HLW waste stream, which was then subjected to small-scale melter testing to determine the maximum acceptable sulfate loadings. In the present work, a new glass formulation was developed and tested for a projected Hanford HLW composition with sulfate concentrations high enough to limit waste loading. Testing was then performed on the DM10 melter system at successively higher waste loadings to determine the maximum waste loading without the formation of a separate sulfate salt phase. Small scale corrosion testing was also conducted using the glass developed in the present work, the glass developed in the initial phase of this work [26], and a high iron composition, all at maximum sulfur concentrations determined from melter testing, in order to assess the extent of Inconel 690 and MA758 corrosion at elevated sulfate contents.

  3. FINAL REPORT INTEGRATED DM1200 MELTER TESTING USING AZ 102 AND C 106/AY-102 HLW SIMULANTS: HLW SIMULANT VERIFICATION VSL-05R5800-1 REV 0 6/27/05

    SciTech Connect (OSTI)

    KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D'ANGELO NA; BRANDYS M; KOT WK; PEGG IL

    2011-12-29

    The principal objectives of the DM1200 melter tests were to determine the effects of feed rheology, feed solid content, and bubbler configuration on glass production rate and off-gas system performance while processing the HLW AZ-101 and C-106/AY-102 feed compositions; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components, as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and perform pre- and post test inspections of system components. The specific objectives (including test success criteria) of this testing, along with how each objective was met, are outlined in a table. The data provided in this Final Report address the impacts of HLW melter feed rheology on melter throughput and validation of the simulated HLW melter feeds. The primary purpose of this testing is to further validate/verify the HLW melter simulants that have been used for previous melter testing and to support their continued use in developing melter and off-gas related processing information for the Project. The primary simulant property in question is rheology. Simulants and melter feeds used in all previous melter tests were produced by direct addition of chemicals; these feed tend to be less viscous than rheological the upper-bound feeds made from actual wastes. Data provided here compare melter processing for the melter feed used in all previous DM100 and DM1200 tests (nominal melter feed) with feed adjusted by the feed vendor (NOAH Technologies) to be more viscous, thereby simulating more closely the upperbounding feed produced from actual waste. This report provides results of tests that are described in the Test Plan for this work. The Test Plan is responsive to one of several test objectives covered in the WTP Test Specification for this work; consequently, only part of the scope described in the Test Specification was addressed in this particular Test Plan. For the purpose of comparison, the tests reported here were performed with AZ-102 and C-106/AY-102 HLW simulants and glass compositions that are essentially the same as those used for recent DM1200 tests. One exception was the use of an alternate, higher-waste-loading C-106/AY-102 glass composition that was used in previous DM100 tests to further evaluate the performance of the optimized bubbler configuration.

  4. U.S. Department of Energy (DOE) initiated performance enhancements to the Hanford waste treatment and immobilization plant (WTP) high-level waste vitrification (HLW) system

    SciTech Connect (OSTI)

    Bowan, Bradley [Energy Solutions, LLC (United States); Gerdes, Kurt [United States Department of Energy (United States); Pegg, Ian [Vitreous State Laboratory, Catholic University of America, 400 Hannan Hall 620 Michigan Avenue, NE Washington, DC 20064 (United States); Holton, Langdon [Pacific Northwest National Laboratory, PO Box 999, Richland WA 99352 (United States)

    2007-07-01

    Available in abstract form only. Full text of publication follows: The U.S Department of Energy is currently constructing, at the Hanford, Washington Site, a Waste Treatment and Immobilization Plant (WTP) for the treatment and immobilization, by vitrification, of stored underground tank wastes. The WTP is comprised of four major facilities: a Pretreatment facility to separate the tank waste into high level waste (HLW) and low activity waste (LAW); a HLW vitrification facility to immobilize the HLW fraction; a LAW vitrification facility to immobilize the LAW fraction and an analytical Laboratory to support the treatment facilities. DOE has strategic objectives to optimize the performance of the WTP facilities, and waste forms, in order to reduce the overall schedule and cost for the treatment of the Hanford tank wastes. One key part of this strategy is to maximize the loading of inorganic waste components in the final glass product (waste loading). For the Hanford tank wastes, this is challenging because of the compositional diversity of the wastes generated over several decades. This paper presents the results of an initial series of HLW waste loading enhancement tests, using diverse HLW compositions that are projected for treatment at the WTP. Specifically, results of glass formulation development and melter testing with simulated Hanford HLW containing high concentrations of troublesome components such as bismuth, aluminum, aluminum-sodium, and chromium will be presented. (authors)

  5. Development Of High Waste-Loading HLW Glasses For High Bismuth Phosphate Wastes, VSL-12R2550-1, Rev 0

    SciTech Connect (OSTI)

    Kruger, A. A.; Pegg, Ian L.; Gan, Hao; Kot, Wing K.

    2012-12-13

    This report presents results from tests with new glass formulations that have been developed for several high Bi-P HLW compositions that are expected to be processed at the WTP that have not been tested previously. WTP HLW feed compositions were reviewed to select waste batches that are high in Bi-P and that are reasonably distinct from the Bi-limited waste that has been tested previously. Three such high Bi-P HLW compositions were selected for this work. The focus of the present work was to determine whether the same type of issues as seen in previous work with high-Bi HLW will be seen in HLW with different concentrations of Bi, P and Cr and also whether similar glass formulation development approaches would be successful in mitigating these issues. New glass compositions were developed for each of the three representative Bi-P HLW wastes and characterized with respect to key processing and product quality properties and, in particular, those relating to crystallization and foaming tendency.

  6. Impact Of Particle Agglomeration On Accumulation Rates In The Glass Discharge Riser Of HLW Melter

    SciTech Connect (OSTI)

    Kruger, A. A.; Rodriguez, C. A.; Matyas, J.; Owen, A. T.; Jansik, D. P.; Lang, J. B.

    2012-11-12

    The major factor limiting waste loading in continuous high-level radioactive waste (HLW) melters is an accumulation of particles in the glass discharge riser during a frequent and periodic idling of more than 20 days. An excessive accumulation can produce robust layers a few centimeters thick, which may clog the riser, preventing molten glass from being poured into canisters. Since the accumulation rate is driven by the size of particles we investigated with x-ray microtomography, scanning electron microscopy, and image analysis the impact of spinel forming components, noble metals, and alumina on the size, concentration, and spatial distribution of particles, and on the accumulation rate. Increased concentrations of Fe and Ni in the baseline glass resulted in the formation of large agglomerates that grew over the time to an average size of ~185+-155 {mu}m, and produced >3 mm thick layer after 120 h at 850 deg C. The noble metals decreased the particle size, and therefore significantly slowed down the accumulation rate. Addition of alumina resulted in the formation of a network of spinel dendrites which prevented accumulation of particles into compact layers.

  7. Final Report - Melt Rate Enhancement for High Aluminum HLW Glass Formulation, VSL-08R1360-1, Rev. 0, dated 12/19/08

    SciTech Connect (OSTI)

    Kruger, Albert A.; Pegg, I. L.; Chaudhuri, M.; Gong, W.; Gan, H.; Matlack, K. S.; Bardakci, T.; Kot, W.

    2013-11-13

    The principal objective of the work reported here was to develop and identify HLW glass compositions that maximize waste processing rates for the aluminum limted waste composition specified by ORP while maintaining high waste loadings and acceptable glass properties. This was accomplished through a combination of crucible-scale tests, confirmation tests on the DM100 melter system, and demonstration at pilot scale (DM1200). The DM100-BL unit was selected for these tests since it was used previously with the HLW waste streams evaluated in this study, was used for tests on HLW glass compositions to support subsequent tests on the HLW Pilot Melter, conduct tests to determine the effect of various glass properties (viscosity and conductivity) and oxide concentrations on glass production rates with HLW feed streams, and to assess the volatility of cesium and technetium during the vitrification of an HLW AZ-102 composition. The same melter was selected for the present tests in order to maintain comparisons between the previously collected data. These tests provide information on melter processing characteristics and off-gas data, including formation of secondary phases and partitioning. Once DM100 tests were completed, one of the compositions was selected for further testing on the DM1200; the DM1200 system has been used for processing a variety of simulated Hanford waste streams. Tests on the larger melter provide processing data at one third of the scale of the actual WTP HLW melter and, therefore, provide a more accurate and reliable assessment of production rates and potential processing issues. The work focused on maximizing waste processing rates for high aluminum HLW compositions. In view of the diversity of forms of aluminum in the Hanford tanks, tests were also conducted on the DM100 to determine the effect of changes in the form of aluminum on feed properties and production rate. In addition, the work evaluated the effect on production rate of modest increases in melter operating temperature. Glass composition development was based on one of the HLW waste compositions specified by ORP that has a high concentration of aluminum. Small-scale tests were used to provide an initial screening of various glass formulations with respect to melt rates; more definitive screening was provided by the subsequent DM100 tests. Glass properties evaluated included: viscosity, electrical conductivity, crystallinity, gross glass phase separation and the 7- day Product Consistency Test (ASTM-1285). Glass property limits were based upon the reference properties for the WTP HLW melter. However, the WTP crystallinity limit (< 1 vol% at 950oC) was relaxed slightly as a waste loading constraint for the crucible melts.

  8. Laboratory Testing of Bulk Vitrified Low-Activity Waste Forms to Support the 2005 Integrated Disposal Facility Performance Assessment

    SciTech Connect (OSTI)

    Pierce, Eric M.; McGrail, B. Peter; Bagaasen, Larry M.; Rodriguez, Elsa A.; Wellman, Dawn M.; Geiszler, Keith N.; Baum, Steven R.; Reed, Lunde R.; Crum, Jarrod V.; Schaef, Herbert T.

    2006-06-30

    The purpose of this report is to document the results from laboratory testing of the bulk vitri-fied (BV) waste form that was conducted in support of the 2005 integrated disposal facility (IDF) performance assessment (PA). Laboratory testing provides a majority of the key input data re-quired to assess the long-term performance of the BV waste package with the STORM code. Test data from three principal methods, as described by McGrail et al. (2000a; 2003a), are dis-cussed in this testing report including the single-pass flow-through test (SPFT) and product con-sistency test (PCT). Each of these test methods focuses on different aspects of the glass corrosion process. See McGrail et al. (2000a; 2003a) for additional details regarding these test methods and their use in evaluating long-term glass performance. In addition to evaluating the long-term glass performance, this report discusses the results and methods used to provided a recommended best estimate of the soluble fraction of 99Tc that can be leached from the engineer-ing-scale BV waste package. These laboratory tests are part of a continuum of testing that is aimed at improving the performance of the BV waste package.

  9. Laboratory Testing of Bulk Vitrified Low-Activity Waste Forms to Support the 2005 Integrated Disposal Facility Performance Assessment

    SciTech Connect (OSTI)

    Pierce, Eric M.; McGrail, B. Peter; Bagaasen, Larry M.; Rodriguez, Elsa A.; Wellman, Dawn M.; Geiszler, Keith N.; Baum, Steven R.; Reed, Lunde R.; Crum, Jarrod V.; Schaef, Herbert T.

    2005-03-31

    The purpose of this report is to document the results from laboratory testing of the bulk vitri-fied (BV) waste form that was conducted in support of the 2005 integrated disposal facility (IDF) performance assessment (PA). Laboratory testing provides a majority of the key input data re-quired to assess the long-term performance of the BV waste package with the STORM code. Test data from three principal methods, as described by McGrail et al. (2000a; 2003a), are dis-cussed in this testing report including the single-pass flow-through test (SPFT) and product con-sistency test (PCT). Each of these test methods focuses on different aspects of the glass corrosion process. See McGrail et al. (2000a; 2003a) for additional details regarding these test methods and their use in evaluating long-term glass performance. In addition to evaluating the long-term glass performance, this report discusses the results and methods used to provided a recommended best estimate of the soluble fraction of 99Tc that can be leached from the engineer-ing-scale BV waste package. These laboratory tests are part of a continuum of testing that is aimed at improving the performance of the BV waste package.

  10. Vitrification of HLW Produced by Uranium/Molybdenum Fuel Reprocessing in COGEMA's Cold Crucible Melter

    SciTech Connect (OSTI)

    Do Quang, R.; Petitjean, V.; Hollebecque, F.; Pinet, O.; Flament, T.; Prod'homme, A.

    2003-02-25

    The performance of the vitrification process currently used in the La Hague commercial reprocessing plants has been continuously improved during more than ten years of operation. In parallel COGEMA (industrial Operator), the French Atomic Energy Commission (CEA) and SGN (respectively COGEMA's R&D provider and Engineering) have developed the cold crucible melter vitrification technology to obtain greater operating flexibility, increased plant availability and further reduction of secondary waste generated during operations. The cold crucible is a compact water-cooled melter in which the radioactive waste and the glass additives are melted by direct high frequency induction. The cooling of the melter produces a solidified glass layer that protects the melter's inner wall from corrosion. Because the heat is transferred directly to the melt, high operating temperatures can be achieved with no impact on the melter itself. COGEMA plans to implement the cold crucible technology to vitrify high level liquid waste from reprocessed spent U-Mo-Sn-Al fuel (used in gas cooled reactor). The cold crucible was selected for the vitrification of this particularly hard-to-process waste stream because it could not be reasonably processed in the standard hot induction melters currently used at the La Hague vitrification facilities : the waste has a high molybdenum content which makes it very corrosive and also requires a special high temperature glass formulation to obtain sufficiently high waste loading factors (12 % in molybdenum). A special glass formulation has been developed by the CEA and has been qualified through lab and pilot testing to meet standard waste acceptance criteria for final disposal of the U-Mo waste. The process and the associated technologies have been also being qualified on a full-scale prototype at the CEA pilot facility in Marcoule. Engineering study has been integrated in parallel in order to take into account that the Cold Crucible should be installed remotely in one of the R7 vitrification cell. This paper will present the results obtained in the framework of these qualification programs.

  11. Category 3 threshold quantities for hazard categorization of nonreactor facilities

    SciTech Connect (OSTI)

    Mandigo, R.L.

    1996-02-13

    This document provides the information necessary to determine Hazard Category 3 threshold quantities for those isotopes of interest not listed in WHC-CM-4-46, Section 4, Table 1.''Threshold Quantities.''

  12. FINAL REPORT DM1200 TESTS WITH AZ 101 HLW SIMULANTS VSL-03R3800-4 REV 0 2/17/04

    SciTech Connect (OSTI)

    KRUGER AA; MATLACK KS; BARDAKCI T; D'ANGELO NA; GONG W; KOT WK; PEGG IL

    2011-12-29

    This report documents melter and off-gas performance results obtained on the DM 1200 HLW Pilot Melter during processing of simulated HLW AZ-101 feed. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW AZ-101 feed; determine the effect of bubbling rate and feed solids content on production rate; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and to perform pre- and post-test inspections of system components. The test objectives (including test success criteria), along with how they were met, are outlined in a table.

  13. Crystallization In High Level Waste (HLW) Glass Melters: Operational Experience From The Savannah River Site

    SciTech Connect (OSTI)

    Fox, K. M.

    2014-02-27

    processing strategy for the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The basis of this alternative approach is an empirical model predicting the crystal accumulation in the WTP glass discharge riser and melter bottom as a function of glass composition, time, and temperature. When coupled with an associated operating limit (e.g., the maximum tolerable thickness of an accumulated layer of crystals), this model could then be integrated into the process control algorithms to formulate crystal tolerant high level waste (HLW) glasses targeting higher waste loadings while still meeting process related limits and melter lifetime expectancies. This report provides a review of the scaled melter testing that was completed in support of the Defense Waste Processing Facility (DWPF) melter. Testing with scaled melters provided the data to define the DWPF operating limits to avoid bulk (volume) crystallization in the un-agitated DWPF melter and provided the data to distinguish between spinels generated by K-3 refractory corrosion versus spinels that precipitated from the HLW glass melt pool. This report includes a review of the crystallization observed with the scaled melters and the full scale DWPF melters (DWPF Melter 1 and DWPF Melter 2). Examples of actual DWPF melter attainment with Melter 2 are given. The intent is to provide an overview of lessons learned, including some example data, that can be used to advance the development and implementation of an empirical model and operating limit for crystal accumulation for WTP. Operation of the first and second (current) DWPF melters has demonstrated that the strategy of using a liquidus temperature predictive model combined with a 100 °C offset from the normal melter operating temperature of 1150 °C (i.e., the predicted liquidus temperature (TL) of the glass must be 1050 °C or less) has been successful in preventing any detrimental accumulation of spinel in the DWPF melt pool, and spinel has not been observed in any of the pour stream glass samples. Spinel was observed at the bottom of DWPF Melter 1 as a result of K-3 refractory corrosion. Issues have occurred with accumulation of spinel in the pour spout during periods of operation at higher waste loadings. Given that both DWPF melters were or have been in operation for greater than 8 years, the service life of the melters has far exceeded design expectations. It is possible that the DWPF liquidus temperature approach is conservative, in that it may be possible to successfully operate the melter with a small degree of allowable crystallization in the glass. This could be a viable approach to increasing waste loading in the glass assuming that the crystals are suspended in the melt and swept out through the riser and pour spout. Additional study is needed, and development work for WTP might be leveraged to support a different operating limit for the DWPF. Several recommendations are made regarding considerations that need to be included as part of the WTP crystal tolerant strategy based on the DWPF development work and operational data reviewed here. These include: Identify and consider the impacts of potential heat sinks in the WTP melter and glass pouring system; Consider the contributions of refractory corrosion products, which may serve to nucleate additional crystals leading to further accumulation; Consider volatilization of components from the melt (e.g., boron, alkali, halides, etc.) and determine their impacts on glass crystallization behavior; Evaluate the impacts of glass REDuction/OXidation (REDOX) conditions and the distribution of temperature within the WTP melt pool and melter pour chamber on crystal accumulation rate; Consider the impact of precipitated crystals on glass viscosity; Consider the impact of an accumulated crystalline layer on thermal convection currents and bubbler effectiveness within the melt pool; Evaluate the impact of spinel accumulation on Joule heating of the WTP melt pool; and Include noble metals in glass melt experiments because of their potential to act as nucleation sites for spinel crystallization.

  14. BENEFITS OF VIBRATION ANALYSIS FOR DEVELOPMENT OF EQUIPMENT IN HLW TANKS - 12341

    SciTech Connect (OSTI)

    Stefanko, D.; Herbert, J.

    2012-01-10

    Vibration analyses of equipment intended for use in the Savannah River Site (SRS) radioactive liquid waste storage tanks are performed during pre-deployment testing and has been demonstrated to be effective in reducing the life-cycle costs of the equipment. Benefits of using vibration analysis to identify rotating machinery problems prior to deployment in radioactive service will be presented in this paper. Problems encountered at SRS and actions to correct or lessen the severity of the problem are discussed. In short, multi-million dollar cost saving have been realized at SRS as a direct result of vibration analysis on existing equipment. Vibration analysis of equipment prior to installation can potentially reduce inservice failures, and increases reliability. High-level radioactive waste is currently stored in underground carbon steel waste tanks at the United States Department of Energy (DOE) Savannah River Site and at the Hanford Site, WA. Various types of rotating machinery (pumps and separations equipment) are used to manage and retrieve the tank contents. Installation, maintenance, and repair of these pumps and other equipment are expensive. In fact, costs to remove and replace a single pump can be as high as a half million dollars due to requirements for radioactive containment. Problems that lead to in-service maintenance and/or equipment replacement can quickly exceed the initial investment, increase radiological exposure, generate additional waste, and risk contamination of personnel and the work environment. Several different types of equipment are considered in this paper, but pumps provide an initial example for the use of vibration analysis. Long-shaft (45 foot long) and short-shaft (5-10 feet long) equipment arrangements are used for 25-350 horsepower slurry mixing and transfer pumps in the SRS HLW tanks. Each pump has a unique design, operating characteristics and associated costs, sometimes exceeding a million dollars. Vibration data are routinely collected during pre-installation tests and screened for: Critical speeds or resonance, Imbalance of rotating parts, Shaft misalignment, Fluid whirl or lubrication break down, Bearing damages, and Other component abnormalities. Examples of previous changes in operating parameters and fabrication tolerances and extension of equipment life resulting from the SRS vibration analysis program include: (1) Limiting operational speeds for some pumps to extend service life without design or part changes; (2) Modifying manufacturing methods (tightening tolerances) for impellers on slurry mixing pumps based on vibration data that indicated hydraulic imbalance; (3) Identifying rolling element mounting defects and replacing those components in pump seals before installation; and (4) Identifying the need for bearing design modification for SRS long-shaft mixing pump designs to eliminate fluid whirl and critical speeds which significantly increased the equipment service life. In addition, vibration analyses and related analyses have been used during new equipment scale-up tests to identify the need for design improvements for full-scale operation / deployment of the equipment in the full size tanks. For example, vibration analyses were recently included in the rotary micro-filtration scale-up test program at SRNL.

  15. ARM - Evaluation Product - Critical soil quantities for describing...

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

    ProductsCritical soil quantities for describing land properties ARM Data Discovery Browse Data Documentation Use the Data File Inventory tool to view data availability at the file...

  16. fiberConnector-Quantities-18Oct2006.xls

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

    attached highlights the quantities and lengths needed for the Minerva detector for the following: Fiber connectors Clear fibers WLS fibers Reviewed by: Robert Flight, PE Sr....

  17. Advances in Glass Formulations for Hanford High-Alumimum, High-Iron and Enhanced Sulphate Management in HLW Streams - 13000

    SciTech Connect (OSTI)

    Kruger, Albert A.

    2013-01-16

    The current estimates and glass formulation efforts have been conservative in terms of achievable waste loadings. These formulations have been specified to ensure that the glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet Hanford Tank Waste Treatment and Immobilization Plant (WTP) Contract terms. The WTP?s overall mission will require the immobilization of tank waste compositions that are dominated by mixtures of aluminum (Al), chromium (Cr), bismuth (Bi), iron (Fe), phosphorous (P), zirconium (Zr), and sulphur (S) compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. Recently, DOE has initiated a testing program to develop and characterize HLW glasses with higher waste loadings and higher throughput efficiencies. Results of this work have demonstrated the feasibility of increases in waste loading from about 25 wt% to 33-50 wt% (based on oxide loading) in the glass depending on the waste stream. In view of the importance of aluminum limited waste streams at Hanford (and also Savannah River), the ability to achieve high waste loadings without adversely impacting melt rates has the potential for enormous cost savings from reductions in canister count and the potential for schedule acceleration. Consequently, the potential return on the investment made in the development of these enhancements is extremely favorable. Glass composition development for one of the latest Hanford HLW projected compositions with sulphate concentrations high enough to limit waste loading have been successfully tested and show tolerance for previously unreported tolerance for sulphate. Though a significant increase in waste loading for high-iron wastes has been achieved, the magnitude of the increase is not as substantial as those achieved for high-aluminum, high-chromium, high-bismuth or sulphur. Waste processing rate increases for high-iron streams as a combined effect of higher waste loadings and higher melt rates resulting from new formulations have been achieved.

  18. Initiating the Validation of CCIM Processability for Multi-phase all Ceramic (SYNROC) HLW Form: Plan for Test BFY14CCIM-C

    SciTech Connect (OSTI)

    Vince Maio

    2014-08-01

    This plan covers test BFY14CCIM-C which will be a first–of–its-kind demonstration for the complete non-radioactive surrogate production of multi-phase ceramic (SYNROC) High Level Waste Forms (HLW) using Cold Crucible Induction Melting (CCIM) Technology. The test will occur in the Idaho National Laboratory’s (INL) CCIM Pilot Plant and is tentatively scheduled for the week of September 15, 2014. The purpose of the test is to begin collecting qualitative data for validating the ceramic HLW form processability advantages using CCIM technology- as opposed to existing ceramic–lined Joule Heated Melters (JHM) currently producing BSG HLW forms. The major objectives of BFY14CCIM-C are to complete crystalline melt initiation with a new joule-heated resistive starter ring, sustain inductive melting at temperatures between 1600 to 1700°C for two different relatively high conductive materials representative of the SYNROC ceramic formation inclusive of a HLW surrogate, complete melter tapping and pouring of molten ceramic material in to a preheated 4 inch graphite canister and a similar canister at room temperature. Other goals include assessing the performance of a new crucible specially designed to accommodate the tapping and pouring of pure crystalline forms in contrast to less recalcitrant amorphous glass, assessing the overall operational effectiveness of melt initiation using a resistive starter ring with a dedicated power source, and observing the tapped molten flow and subsequent relatively quick crystallization behavior in pans with areas identical to standard HLW disposal canisters. Surrogate waste compositions with ceramic SYNROC forming additives and their measured properties for inductive melting, testing parameters, pre-test conditions and modifications, data collection requirements, and sampling/post-demonstration analysis requirements for the produced forms are provided and defined.

  19. Groundwater Quantity Regulation in Vermont: A Path Forward |...

    Open Energy Info (EERE)

    Groundwater Quantity Regulation in Vermont: A Path Forward Jump to: navigation, search OpenEI Reference LibraryAdd to library Legal Document- Secondary Legal SourceSecondary Legal...

  20. Final Report - Effects of High Spinel and Chromium Oxide Crystal Contents on Simulated HLW Vitrification in DM100 Melter Tests, VSL-09R1520-1, Rev. 0, dated 6/22/09

    SciTech Connect (OSTI)

    Kruger, Albert A.; Matlack, K. S.; Kot, W.; Pegg, I. L.; Chaudhuri, M.; Lutze, W.

    2013-11-13

    The principal objective of the work was to evaluate the effects of spinel and chromium oxide particles on WTP HLW melter operations and potential impacts on melter life. This was accomplished through a combination of crucible-scale tests, settling and rheological tests, and tests on the DM100 melter system. Crucible testing was designed to develop and identify HLW glass compositions with high waste loadings that exhibit formation of crystalline spinel and/or chromium oxide phases up to relatively high crystal contents (i.e., > 1 vol%). Characterization of crystal settling and the effects on melt rheology was performed on the HLW glass formulations. Appropriate candidate HLW glass formulations were selected, based on characterization results, to support subsequent melter tests. In the present work, crucible melts were formulated that exhibit up to about 4.4 vol% crystallization.

  1. U.S. NWTRB March Meeting to Focus on DOE R&D Related to Salt as a Geologic Medium for Disposal of SNF and HLW

    Broader source: Energy.gov [DOE]

    The U.S. Nuclear Waste Technical Review Board will hold a public meeting in Albuquerque, NM, on Wednesday, March 19, 2014. The main topic of the meeting is the U.S. Department of Energy (DOE) research and development (R&D) activities related to salt as a geologic medium for disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW).

  2. Arizona Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Arizona Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 26 10 0 0 0 0 1,360 1990's 2,125 1,225 730 548 691 500 405 401 411 439 2000's 332 266 243 426 306 211 588 634 503 695 2010's 165 - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid

  3. Illinois Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Illinois Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 1,030 1,530 1,324 1,887 1,371 1,338 1,477 1990's 677 466 346 250 333 0 0 0 0 0 2000's 0 0 NA 0 NA NA NA NA NA NA 2010's NA - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid

  4. Indiana Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Indiana Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 135 394 367 365 217 412 416 1990's 399 232 174 192 107 249 360 526 615 855 2000's 899 1,064 1,309 1,464 3,401 3,135 2,921 3,606 4,701 4,927 2010's 6,802 - = No Data Reported; -- = Not Applicable; NA = Not Available;

  5. Maryland Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Maryland Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 31 60 39 20 44 29 34 1990's 22 29 33 28 26 22 0 118 63 18 2000's 34 32 22 48 34 46 NA NA NA NA 2010's NA - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual

  6. Missouri Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Missouri Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 4 4 4 4 4 4 1990's 7 19 27 14 8 16 25 5 0 2000's 0 0 0 0 0 0 2010's 0 - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Release Date:

  7. TESTS WITH HIGH-BISMUTH HLW GLASSES FINAL REPORT VSL-10R1780-1 REV 0 12/13/10

    SciTech Connect (OSTI)

    MATLACK KS; KRUGER AA; JOSEPH I; GAN H; KOT WK; CHAUDHURI M; MOHR RK; MCKEOWN DA; BARDAKEI T; GONG W; BUECCHELE AC; PEGG IL

    2011-01-05

    This Final Report describes the testing of glass formulations developed for Hanford High Level Waste (HLW) containing high concentrations of bismuth. In previous work on high-bismuth HLW streams specified by the Office of River Protection (ORP), fully compliant, high waste loading compositions were developed and subjected to melter testing on the DM100 vitrification system. However, during heat treatment according to the Hanford Tank Waste Treatment and Immobilization Plant (WTP) HLW canister centerline cooling (CCC) curves, crucible melts of the high-bismuth glasses were observed to foam. Clearly, such an occurrence during cooling of actual HLW canisters would be highly undesirable. Accordingly, the present work involves larger-scale testing to determine whether this effect occurs under more prototypical conditions, as well as crucible-scale tests to determine the causes and potentially remediate the observed foaming behavior. The work included preparation and characterization of crucible melts designed to determine the underlying causes of the foaming behavior as well as to assess potential mitigation strategies. Testing was also conducted on the DM1200 HLW Pilot melter with a composition previously tested on the DM100 and shown to foam during crucible-scale CCC heat treatment. The DM1200 tests evaluated foaming of glasses over a range of bismuth concentrations poured into temperature-controlled, 55-gallon drums which have a diameter that is close to that of the full-scale WTP HLW canisters. In addition, the DM1200 tests provided the first large-scale melter test data on high-bismuth WTP HLW compositions, including information on processing rates, cold cap behavior and off-gas characteristics, and data from this waste composition on the prototypical DM1200 off-gas treatment system. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for ORP on the same waste composition. The scope of this study was outlined in a Test Plan that was prepared in response to an ORP-supplied statement of work. The present glass formulation and melter testing work was aimed at one of the four waste streams previously specified by the Office of River Protection (ORP). Such testing supports the ORP basis for projection of the amount of Immobilized High Level Waste (IHLW) to be produced at Hanford and evaluation of the likely potential for future enhancements of the WTP over and above the present well-developed baseline. It should be noted that the compositions of the four ORP-specified waste streams differ significantly from those of the feed tanks (AZ-101, AZ-102, C-16/AY-102, and C-104/AY-101) that have been the focus of the extensive technology development and design work performed for the WTP baseline. In this regard, the work on the ORP-specified compositions is complementary to and necessarily of a more exploratory nature than the work in support of the current WTP baseline.

  8. New Design for an HLW Repository (for Spent Fuel and Waste from Reprocessing) in a Salt Formation in Germany - 12213

    SciTech Connect (OSTI)

    Bollingerfehr, Wilhelm; Filbert, Wolfgang; Lerch, Christian; Mueller-Hoeppe, Nina; Charlier, Frank

    2012-07-01

    In autumn 2010, after a 10-year moratorium, exploration was resumed in Gorleben, the potential site for a German HLW repository. At the same time, the Federal Government launched a two-year preliminary safety analysis to assess whether the salt dome at Gorleben is suitable to host all heat-generating radioactive waste generated by German NPPs based on the waste amounts expected at that time. The revised Atomic Energy Act of June 2011 now stipulates a gradual phase-out of nuclear energy production by 2022, which is 13 years earlier than expected in 2010. A repository design was developed which took into account an updated set of data on the amounts and types of expected heat-generating waste, the documented results of the exploration of the Gorleben salt dome, and the new 'Safety Requirements Governing the Final Disposal of Heat-Generating Radioactive Waste' of 30 September, 2010. The latter has a strong influence on the conceptual designs as it requires that retrievability of all waste containers is possible within the repository lifetime. One design considered that all waste containers will be disposed of in horizontal drifts of a geologic repository, while the other design considered that all waste containers will be disposed of in deep vertical boreholes. For both options (emplacement in drifts/emplacement in vertical boreholes), the respective design includes a selection of waste containers, the layout of drifts, respectively lined boreholes, a description of emplacement fields, and backfilling and sealing measures. The design results were described and displayed and the differences between the two main concepts were elaborated and discussed. For the first time in both repository designs the requirement was implemented to retrieve waste canisters during the operational phase. The measures to fulfill this requirement and eventually the consequences were highlighted. It was pointed out that there arises the need to keep transport- and storage casks in adequate numbers and interim storage facilities available until the repository is closed. (authors)

  9. Tennessee Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Tennessee Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 3,950 5,022 4,686 3,464 2,707 2,100 1,900 1990's 2,067 1,856 1,770 1,660 1,990 1,820 1,690 1,510 1,420 1,230 2000's 1,150 2,000 2,050 1,803 2,100 2,200 2,663 3,942 4,700 5,478 2010's 5,144 - = No Data Reported; --

  10. Texas Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Texas Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 6,227,995 6,630,246 6,367,936 6,465,964 6,414,021 6,386,544 6,276,968 1990's 6,476,032 6,066,256 5,893,069 5,769,437 5,834,671 5,592,323 4,684,140 4,716,304 4,777,945 5,719,128 2000's 5,869,901 5,159,233 5,166,315

  11. Utah Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Utah Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 63,158 74,698 52,324 21,491 48,654 49,378 58,356 1990's 57,098 62,241 86,682 93,894 154,907 153,804 168,944 174,275 190,230 194,413 2000's 218,283 215,527 250,118 202,784 250,261 267,766 319,268 NA 276,340 389,830

  12. Virginia Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Virginia Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 4,342 8,928 15,041 15,427 19,223 18,424 17,935 1990's 14,283 14,906 24,734 37,840 50,259 49,818 0 0 0 0 2000's 0 0 0 0 NA NA NA NA NA NA 2010's NA - = No Data Reported; -- = Not Applicable; NA = Not Available; W =

  13. West Virginia Quantity of Production Associated with Reported Wellhead

    U.S. Energy Information Administration (EIA) Indexed Site

    Value (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) West Virginia Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 130,078 143,730 144,883 135,431 160,000 174,942 177,192 1990's 95,271 198,605 202,775 171,024 55,756 50,439 0 0 0 0 2000's 0 0 NA 0 NA NA NA NA NA NA 2010's NA - = No Data Reported; -- = Not Applicable;

  14. Arkansas Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Arkansas Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 78,097 75,575 86,552 68,206 42,688 102,046 42,226 1990's 99,456 83,864 85,177 122,596 24,326 180,117 76,671 71,449 61,012 54,382 2000's 55,057 16,901 161,871 166,329 183,299 190,533 193,491 269,886 446,551 680,613

  15. California Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) California Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 282,639 343,079 361,739 329,366 346,720 327,399 283,509 1990's 275,738 211,841 195,515 76,381 199,649 263 37,823 219,216 264,810 382,715 2000's 323,864 328,778 309,399 293,691 276,520 274,817 278,933 268,016

  16. Colorado Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Colorado Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 139,820 143,552 126,037 163,684 164,557 191,544 216,737 1990's 242,997 271,159 314,105 388,016 441,343 511,513 559,473 637,375 696,321 705,477 2000's 735,332 800,712 819,205 989,678 1,058,383 1,106,993 1,170,819

  17. Florida Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Florida Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 25,630 18,897 13,162 3,004 1,893 1,883 1,437 1990's 1,443 2,096 3,849 2,612 4,940 3,545 0 0 0 0 2000's 0 0 NA 0 NA NA NA NA NA NA 2010's NA - = No Data Reported; -- = Not Applicable; NA = Not Available; W = Withheld

  18. Kansas Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Kansas Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 366,189 494,641 532,092 465,695 458,426 577,815 588,757 1990's 555,187 630,155 640,583 668,640 714,659 721,436 712,796 678,652 603,586 553,419 2000's 525,430 480,145 454,901 418,893 397,121 377,229 372,029 366,859

  19. New Mexico Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) New Mexico Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 884,517 925,298 880,307 676,886 790,639 752,629 833,593 1990's 949,735 1,029,824 1,274,220 1,489,052 1,510,804 1,480,327 1,553,103 1,540,157 1,483,370 1,511,671 2000's 1,685,664 1,670,644 1,614,045 1,576,639

  20. Ohio Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Ohio Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 149,096 184,651 180,458 180,287 164,960 166,690 159,730 1990's 154,619 146,189 143,381 135,939 130,855 125,085 119,251 116,246 108,542 102,505 2000's 98,551 97,272 103,158 120,081 119,847 83,523 86,315 88,095 84,858

  1. Oklahoma Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Oklahoma Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 1,730,061 1,985,869 1,936,341 1,917,493 2,004,797 2,106,632 2,185,204 1990's 2,186,153 2,119,161 1,937,224 2,005,971 1,879,257 1,765,788 1,751,487 1,452,233 1,644,531 1,577,961 2000's 1,612,890 1,477,058 1,456,375

  2. Oregon Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Oregon Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 3 2,790 4,080 4,600 3,800 4,000 2,500 1990's 2,815 2,741 2,580 4,003 3,221 1,923 1,439 1,173 1,067 1,291 2000's 1,214 1,069 837 688 467 433 NA 390 751 751 2010's 1,376 - = No Data Reported; -- = Not Applicable; NA =

  3. Pennsylvania Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Pennsylvania Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 118,372 166,342 150,234 159,889 163,318 167,089 191,774 1990's 177,609 152,500 138,675 189,443 187,113 177,139 0 0 0 0 2000's 0 0 0 0 NA NA NA NA NA NA 2010's NA - = No Data Reported; -- = Not Applicable; NA =

  4. Kentucky Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Kentucky Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 46,720 61,518 73,126 80,195 70,125 44,725 72,417 1990's 75,333 78,904 79,690 86,966 73,081 74,754 81,435 79,547 81,868 76,770 2000's 81,545 81,723 88,259 87,609 94,259 92,795 95,320 95,437 114,116 NA 2010's 135,355

  5. Louisiana Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Louisiana Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 3,149,192 3,650,412 3,179,306 2,986,468 3,243,795 3,158,903 3,066,789 1990's 3,780,551 3,355,867 3,404,963 3,454,646 3,562,360 3,709,015 3,976,305 5,398,216 5,410,523 5,265,670 2000's 3,587,815 1,529,733 1,365,925

  6. Michigan Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Michigan Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 138,910 144,537 131,855 127,287 146,996 146,145 155,988 1990's 106,193 189,497 190,637 199,746 216,268 238,203 245,740 305,950 278,076 277,364 2000's 296,556 275,036 274,476 236,987 259,681 261,112 NA NA 153,130

  7. Mississippi Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Mississippi Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 211,116 206,871 178,426 197,217 195,299 196,912 148,167 1990's 149,012 126,637 129,340 131,450 105,646 95,349 88,805 98,075 88,723 83,232 2000's 70,965 76,986 112,979 133,901 145,692 52,923 60,531 73,460 96,641

  8. Montana Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Montana Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 47,751 47,534 46,113 42,203 42,814 47,748 52,044 1990's 45,998 48,075 50,359 58,810 51,953 46,739 46,868 50,409 51,967 55,780 2000's 67,294 78,493 86,075 86,027 90,771 101,666 106,843 110,942 802,619 293,941 2010's

  9. Nebraska Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Nebraska Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 2,091 2,300 1,944 1,403 1,261 910 878 1990's 793 785 1,177 1,375 2,098 1,538 1,332 1,194 1,285 1,049 2000's 879 883 892 1,168 1,172 1,172 NA 1,555 3,082 2,908 2010's 2,231 - = No Data Reported; -- = Not Applicable;

  10. Alabama Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Alabama Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Alabama Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 59,051 56,685 42,925 34,164 35,674 45,488 41,614 1990's 37,229 35,972 51,219 75,474 70,489 54,964 493,069 583,370 560,414 544,020 2000's 521,215 376,241 370,753 348,722 304,212 285,237 274,176 259,062

  11. Alaska Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Alaska Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Alaska Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 211,569 211,579 222,637 304,841 271,120 228,284 192,760 1990's 191,798 200,557 206,259 224,786 201,891 227,797 193,278 191,017 192,982 186,727 2000's 189,896 197,735 200,871 199,616 413,667 502,887 494,323

  12. Wyoming Quantity of Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    (Million Cubic Feet) Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Wyoming Quantity of Production Associated with Reported Wellhead Value (Million Cubic Feet) Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9 1980's 395,656 447,615 416,565 352,858 407,863 471,095 623,915 1990's 690,356 711,799 765,254 63,667 14,283 12,449 27,821 719,933 1,004,020 1,079,375 2000's 1,240,038 1,359,868 1,533,724 1,561,322 1,724,725 1,729,760

  13. A Comparison of Water Vapor Quantities from Model Short-Range...

    Office of Scientific and Technical Information (OSTI)

    Water Vapor Quantities from Model Short-Range Forecasts and ARM Observations Citation Details In-Document Search Title: A Comparison of Water Vapor Quantities from Model ...

  14. ROLE OF MANGANESE REDUCTION/OXIDATION (REDOX) ON FOAMING AND MELT RATE IN HIGH LEVEL WASTE (HLW) MELTERS (U)

    SciTech Connect (OSTI)

    Jantzen, C; Michael Stone, M

    2007-03-30

    High-level nuclear waste is being immobilized at the Savannah River Site (SRS) by vitrification into borosilicate glass at the Defense Waste Processing Facility (DWPF). Control of the Reduction/Oxidation (REDOX) equilibrium in the DWPF melter is critical for processing high level liquid wastes. Foaming, cold cap roll-overs, and off-gas surges all have an impact on pouring and melt rate during processing of high-level waste (HLW) glass. All of these phenomena can impact waste throughput and attainment in Joule heated melters such as the DWPF. These phenomena are caused by gas-glass disequilibrium when components in the melter feeds convert to glass and liberate gases such as H{sub 2}O vapor (steam), CO{sub 2}, O{sub 2}, H{sub 2}, NO{sub x}, and/or N{sub 2}. During the feed-to-glass conversion in the DWPF melter, multiple types of reactions occur in the cold cap and in the melt pool that release gaseous products. The various gaseous products can cause foaming at the melt pool surface. Foaming should be avoided as much as possible because an insulative layer of foam on the melt surface retards heat transfer to the cold cap and results in low melt rates. Uncontrolled foaming can also result in a blockage of critical melter or melter off-gas components. Foaming can also increase the potential for melter pressure surges, which would then make it difficult to maintain a constant pressure differential between the DWPF melter and the pour spout. Pressure surges can cause erratic pour streams and possible pluggage of the bellows as well. For these reasons, the DWPF uses a REDOX strategy and controls the melt REDOX between 0.09 {le} Fe{sup 2+}/{summation}Fe {le} 0.33. Controlling the DWPF melter at an equilibrium of Fe{sup +2}/{summation}Fe {le} 0.33 prevents metallic and sulfide rich species from forming nodules that can accumulate on the floor of the melter. Control of foaming, due to deoxygenation of manganic species, is achieved by converting oxidized MnO{sub 2} or Mn{sub 2}O{sub 3} species to MnO during melter preprocessing. At the lower redox limit of Fe{sup +2}/{summation}Fe {approx} 0.09 about 99% of the Mn{sup +4}/Mn{sup +3} is converted to Mn{sup +2}. Therefore, the lower REDOX limits eliminates melter foaming from deoxygenation.

  15. INCONEL 690 CORROSION IN WTP (WASTE TREATMENT PLANT) HLW (HIGH LEVEL WASTE) GLASS MELTS RICH IN ALUMINUM & BISMUTH & CHROMIUM OR ALUMINUM/SODIUM

    SciTech Connect (OSTI)

    KRUGER AA; FENG Z; GAN H; PEGG IL

    2009-11-05

    Metal corrosion tests were conducted with four high waste loading non-Fe-limited HLW glass compositions. The results at 1150 C (the WTP nominal melter operating temperature) show corrosion performance for all four glasses that is comparable to that of other typical borosilicate waste glasses, including HLW glass compositions that have been developed for iron-limited WTP streams. Of the four glasses tested, the Bi-limited composition shows the greatest extent of corrosion, which may be related to its higher phosphorus content. Tests at higher suggest that a moderate elevation of the melter operating temperature (up to 1200 C) should not result in any significant increase in Inconel corrosion. However, corrosion rates did increase significantly at yet higher temperatures (1230 C). Very little difference was observed with and without the presence of an electric current density of 6 A/inch{sup 2}, which is the typical upper design limit for Inconel electrodes. The data show a roughly linear relationship between the thickness of the oxide scale on the coupon and the Cr-depletion depth, which is consistent with the chromium depletion providing the material source for scale growth. Analysis of the time dependence of the Cr depletion profiles measured at 1200 C suggests that diffusion of Cr in the Ni-based Inconel alloy controls the depletion depth of Cr inside the alloy. The diffusion coefficient derived from the experimental data agrees within one order of magnitude with the published diffusion coefficient data for Cr in Ni matrices; the difference is likely due to the contribution from faster grain boundary diffusion in the tested Inconel alloy. A simple diffusion model based on these data predicts that Inconel 690 alloy will suffer Cr depletion damage to a depth of about 1 cm over a five year service life at 1200 C in these glasses.

  16. A comparison of methods for representing sparsely sampled random quantities.

    SciTech Connect (OSTI)

    Romero, Vicente Jose; Swiler, Laura Painton; Urbina, Angel; Mullins, Joshua

    2013-09-01

    This report discusses the treatment of uncertainties stemming from relatively few samples of random quantities. The importance of this topic extends beyond experimental data uncertainty to situations involving uncertainty in model calibration, validation, and prediction. With very sparse data samples it is not practical to have a goal of accurately estimating the underlying probability density function (PDF). Rather, a pragmatic goal is that the uncertainty representation should be conservative so as to bound a specified percentile range of the actual PDF, say the range between 0.025 and .975 percentiles, with reasonable reliability. A second, opposing objective is that the representation not be overly conservative; that it minimally over-estimate the desired percentile range of the actual PDF. The presence of the two opposing objectives makes the sparse-data uncertainty representation problem interesting and difficult. In this report, five uncertainty representation techniques are characterized for their performance on twenty-one test problems (over thousands of trials for each problem) according to these two opposing objectives and other performance measures. Two of the methods, statistical Tolerance Intervals and a kernel density approach specifically developed for handling sparse data, exhibit significantly better overall performance than the others.

  17. W, F, and I : Three quantities basic to radiation physics. (Conference...

    Office of Scientific and Technical Information (OSTI)

    W, F, and I : Three quantities basic to radiation physics. Citation Details In-Document Search Title: W, F, and I : Three quantities basic to radiation physics. You are ...

  18. Memo, "Incorporation of HLW Glass Shell V2.0 into the Flowsheets," to ED Lee, CCN: 184905, October 20, 2009

    SciTech Connect (OSTI)

    Gimpel, Rodney F.; Kruger, Albert A.

    2013-12-18

    Efforts are being made to increase the efficiency and decrease the cost of vitrifying radioactive waste stored in tanks at the U.S. Department of Energy Hanford Site. The compositions of acceptable and processable high-level waste (HL W) glasses need to be optimized to minimize the waste-form volume and, hence, to reduce cost. A database of glass properties of waste glass and associated simulated waste glasses was collected and documented in PNNL 18501, Glass Property Data and Models for Estimating High-Level Waste Glass Volume and glass property models were curve-fitted to the glass compositions. A routine was developed that estimates HL W glass volumes using the following glass property models: II Nepheline, II One-Percent Crystal Temperature (T1%), II Viscosity (11) II Product Consistency Tests (PCT) for boron, sodium, and lithium, and II Liquidus Temperature (TL). The routine, commonly called the HL W Glass Shell, is presented in this document. In addition to the use of the glass property models, glass composition constraints and rules, as recommend in PNNL 18501 and in other documents (as referenced in this report) were incorporated. This new version of the HL W Glass Shell should generally estimate higher waste loading in the HL W glass than previous versions.

  19. FINAL REPORT INTEGRATED DM1200 MELTER TESTING OF BUBBLER CONFIGURATIONS USING HLW AZ-101 SIMULANTS VSL-04R4800-4 REV 0 10/5/04

    SciTech Connect (OSTI)

    KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D'ANGELO NA; LUTZE W; CALLOW RA; BRANDYS M; KOT WK; PEGG IL

    2011-12-29

    This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of AZ-101 HLW simulants. The tests reported herein are a subset of six tests from a larger series of tests described in the Test Plan for the work; results from the other tests have been reported separately. The solids contents of the melter feeds were based on the WTP baseline value for the solids content of the feeds from pretreatment which changed during these tests from 20% to 15% undissolved solids resulting in tests conducted at two feed solids contents. Based on the results of earlier tests with single outlet 'J' bubblers, initial tests were performed with a total bubbling rate of 651 pm. The first set of tests (Tests 1A-1E) addressed the effects of skewing this total air flow rate back and forth between the two installed bubblers in comparison to a fixed equal division of flow between them. The second set of tests (2A-2D) addressed the effects of bubbler depth. Subsequently, as the location, type and number of bubbling outlets were varied, the optimum bubbling rate for each was determined. A third (3A-3C) and fourth (8A-8C) set of tests evaluated the effects of alternative bubbler designs with two gas outlets per bubbler instead of one by placing four bubblers in positions simulating multiple-outlet bubblers. Data from the simulated multiple outlet bubblers were used to design bubblers with two outlets for an additional set of tests (9A-9C). Test 9 was also used to determine the effect of small sugar additions to the feed on ruthenium volatility. Another set of tests (10A-10D) evaluated the effects on production rate of spiking the feed with chloride and sulfate. Variables held constant to the extent possible included melt temperature, plenum temperature, cold cap coverage, the waste simulant composition, and the target glass composition. The feed rate was increased to the point that a constant, essentially complete, cold cap was achieved, which was used as an indicator of a maximized feed rate for each test. The first day of each test was used to build the cold cap and decrease the plenum temperature. The remainder of each test was split into two- to six-day segments, each with a different bubbling rate, bubbler orientation, or feed concentration of chloride and sulfur.

  20. Table A26. Total Quantity of Purchased Energy Sources by Census...

    U.S. Energy Information Administration (EIA) Indexed Site

    Total Quantity of Purchased Energy Sources by Census Region and" " Economic ... ","(1000","(trillion","Row" "Economic Characteristics(a)","Btu)","kWh)","(1000 ...

  1. A comparison of water vapor quantities from model short-range...

    Office of Scientific and Technical Information (OSTI)

    Technical Report: A comparison of water vapor quantities from model short-range forecasts and ARM observations Citation Details In-Document Search Title: A comparison of water ...

  2. FINAL REPORT INTEGRATED DM1200 MELTER TESTING OF REDOX EFFECTS USING HLW AZ-101 AND C-106/AY-102 SIMULANTS VSL-04R4800-1 REV 0 5/6/

    SciTech Connect (OSTI)

    KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D'ANGELO NA; LUTZE W; BIZOT PM; CALLOW RA; BRANDYS M; KOT WK; PEGG IL

    2011-12-29

    This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of AZ-101 and C-106/AY-102 HLW simulants. The tests reported herein are a subset of three tests from a larger series of tests described in the Test Plan for the work; results from the remaining tests will be reported separately. Three nine day tests, one with AZ-101 and two with C-106/AY-102 feeds were conducted with variable amounts of added sugar to address the effects of redox. The test with AZ-101 included ruthenium spikes to also address the effects of redox on ruthenium volatility. One of tests addressed the effects of increased flow-sheet nitrate levels using C-106/AY-102 feeds. With high nitrate/nitrite feeds (such as WTP LAW feeds), reductants are required to prevent melt foaming and deleterious effects on glass production rates. Sugar is the baseline WTP reductant for this purpose. WTP HLW feeds typically have relatively low nitrate/nitrite content in comparison to the organic carbon content and, therefore, have typically not required sugar additions. However, HLW feed variability, particularly with respect to nitrate levels, may necessitate the use of sugar in some instances. The tests reported here investigate the effects of variable sugar additions to the melter feed as well as elevated nitrate levels in the waste. Variables held constant to the extent possible included melt temperature, bubbling rate, plenum temperature, cold cap coverage, the waste simulant composition, and the target glass composition. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW feeds with variable amounts of added sugar and increased nitrate levels; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and perform pre- and post test inspections of system components. The specific objectives (including test success criteria) of this testing, along with how each objective was met, are outlined in a table.

  3. FINAL REPORT DETERMINATION OF THE PROCESSING RATE OF RPP WTP HLW SIMULANTS USING A DURAMELTER J 1000 VITRIFICATION SYSTEM VSL-00R2590-2 REV 0 8/21/00

    SciTech Connect (OSTI)

    KRUGER AA; MATLACK KS; KOT WK; PEREZ-CARDENAS F; PEGG IL

    2011-12-29

    This report provides data, analysis, and conclusions from a series of tests that were conducted at the Vitreous State Laboratory of The Catholic University of America (VSL) to determine the melter processing rates that are achievable with RPP-WTP HLW simulants. The principal findings were presented earlier in a summary report (VSL-00R2S90-l) but the present report provides additional details. One of the most critical pieces of information in determining the required size of the RPP-WTP HLW melter is the specific glass production rate in terms of the mass of glass that can be produced per unit area of melt surface per unit time. The specific glass production rate together with the waste loading (essentially, the ratio of waste-in to glass-out, which is determined from glass formulation activities) determines the melt area that is needed to achieve a given waste processing rate with due allowance for system availability. As a consequence of the limited amount of relevant information, there exists, for good reasons, a significant disparity between design-base specific glass production rates for the RPP-WTP LAW and HLW conceptual designs (1.0 MT/m{sup 2}/d and 0.4 MT/m{sup 2}/d, respectively); furthermore, small-scale melter tests with HLW simulants that were conducted during Part A indicated typical processing rates with bubbling of around 2.0 MT/m{sup 2}/d. This range translates into more than a factor of five variation in the resultant surface area of the HLW melter, which is clearly not without significant consequence. It is clear that an undersized melter is undesirable in that it will not be able to support the required waste processing rates. It is less obvious that there are potential disadvantages associated with an oversized melter, over and above the increased capital costs. A melt surface that is consistently underutilized will have poor cold cap coverage, which will result in increased volatilization from the melt (which is generally undesirable) and increased plenum temperatures due to increased thermal radiation from the melt surface (which mayor may not be desirable but the flexibility to choose may be lost). Increased volatilization is an issue both in terms of the increased challenge to the off-gas system as well as for the ability to effectively close the recycle loops for volatile species that must be immobilized in the glass product, most notably technetium and cesium. For these reasons, improved information is needed on the specific glass production rates of RPP-WTP HLW streams in DuraMelterJ systems over a range of operating conditions. Unlike the RPP-WTP LAW program, for which a pilot melter system to provide large-scale throughout information is already in operation, there is no comparable HLW activity; the results of the present study are therefore especially important. This information will reduce project risk by reducing the uncertainty associated with the amount of conservatism that mayor may not be associated with the baseline RPP-WTP HLW melter sizing decision. After the submission of the first Test Plan for this work, the RPP-WTP requested revisions to include tests to determine the processing rates that are achievable without bubbling, which was driven by the potential advantages of omitting bubblers from the HLW melter design in terms of reduced maintenance. A further objective of this effort became the determination of whether the basis of design processing rate could be achieved without bubbling. Ideally, processing rate tests would be conducted on a full-scale RPP-WTP melter system with actual HLW materials, but that is clearly unrealistic during Part B1. As a practical compromise the processing rate determinations were made with HL W simulants on a DuraMelter J system at as close to full scale as possible and the DM 1000 system at VSL was selected for that purpose. That system has a melt surface area of 1.2 m{sup 2}, which corresponds to about one-third scale based on the specific glass processing rate of 0.4 MT/m{sup 2}/d assumed in the RPP-WTP HLW conceptual design, but would correspon

  4. A Comparison of Water Vapor Quantities from Model Short-Range Forecasts and

    Office of Scientific and Technical Information (OSTI)

    ARM Observations (Technical Report) | SciTech Connect Water Vapor Quantities from Model Short-Range Forecasts and ARM Observations Citation Details In-Document Search Title: A Comparison of Water Vapor Quantities from Model Short-Range Forecasts and ARM Observations (in English; Croatian) Model evolution and improvement is complicated by the lack of high quality observational data. To address a major limitation of these measurements the Atmospheric Radiation Measurement (ARM) program was

  5. HLW System Integrated Project Team

    Office of Environmental Management (EM)

    l W S Hi h l W S High Level Waste System High Level Waste System Integrated Project Team ... and skilled kf Developing and deploying t h l i This document is intended for planning ...

  6. HLW-OVP-96 C

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

    6 C 0083 High Level Waste Management Division. High-Level*Waste System Plan Revision 7 (U)* . .. October 1996 Westinghouse Savannah River Company Savannah River Site . Aiken, .SC 29808 -~--~---------------------~-~-----------~---------~---~- Prepared for the U. S. Department of Energy under contract no. DE-AC09-96SR18500 Westinghouse Savannah River Company Mr. A. L. Watkins, Assistant Manager High Level Waste U. S. Department of Energy Savannah River Operations Office P. O. Box A . Aiken, SC

  7. FINAL REPORT REGULATORY OFF GAS EMISSIONS TESTING ON THE DM1200 MELTER SYSTEM USING HLW AND LAW SIMULANTS VSL-05R5830-1 REV 0 10/31/05

    SciTech Connect (OSTI)

    KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D'ANGELO NA; BRANDYS M; KOT WK; PEGG IL

    2011-12-29

    The operational requirements for the River Protection Project - Waste Treatment Plant (RPP-WTP) Low Activity Waste (LAW) and High Level Waste (HLW) melter systems, together with the feed constituents, impose a number of challenges to the off-gas treatment system. The system must be robust from the standpoints of operational reliability and minimization of maintenance. The system must effectively control and remove a wide range of solid particulate matter, acid mists and gases, and organic constituents (including those arising from products of incomplete combustion of sugar and organics in the feed) to concentration levels below those imposed by regulatory requirements. The baseline design for the RPP-WTP LAW primary off-gas system includes a submerged bed scrubber (SBS), a wet electrostatic precipitator (WESP), and a high efficiency particulate air (HEPA) filter. The secondary off-gas system includes a sulfur-impregnated activated carbon bed (AC-S), a thermal catalytic oxidizer (TCO), a single-stage selective catalytic reduction NOx treatment system (SCR), and a packed-bed caustic scrubber (PBS). The baseline design for the RPP-WTP HLW primary off-gas system includes an SBS, a WESP, a high efficiency mist eliminator (HEME), and a HEPA filter. The HLW secondary off-gas system includes a sulfur-impregnated activated carbon bed, a silver mordenite bed, a TCO, and a single-stage SCR. The one-third scale HLW DM1200 Pilot Melter installed at the Vitreous State Laboratory (VSL) was equipped with a prototypical off-gas train to meet the needs for testing and confirmation of the performance of the baseline off-gas system design. Various modifications have been made to the DM1200 system as the details of the WTP design have evolved, including the installation of a silver mordenite column and an AC-S column for testing on a slipstream of the off-gas flow; the installation of a full-flow AC-S bed for the present tests was completed prior to initiation of testing. The DM1200 system was reconfigured to enable testing of the baseline HLW or LAW off-gas trains to perform off-gas emissions testing with both LAW and HLW simulants in the present work. During 2002 and 2003, many of these off-gas components were tested individually and in an integrated manner with the DM1200 Pilot Melter. Data from these tests are being used to support engineering design confirmation and to provide data to support air permitting activities. In fiscal year 2004, the WTP Project was directed by the Office of River Protection (ORP) to comply with Environmental Protection Agency (EPA) Maximum Achievable Control Technology (MACT) requirements for organics. This requires that the combined melter and off-gas system have destruction and removal efficiency (DRE) of >99.99% for principal organic dangerous constituents (PODCs). In order to provide confidence that the melter and off-gas system are able to achieve the required DRE, testing has been directed with both LAW and HLW feeds. The tests included both 'normal' and 'challenge' WTP melter conditions in order to obtain data for the potential range of operating conditions for the WTP melters and off-gas components. The WTP Project, Washington State Department of Ecology, and ORP have agreed that naphthalene will be used for testing to represent semi-volatile organics and allyl alcohol will be used to represent volatile organics. Testing was also performed to determine emissions of halides, metals, products of incomplete combustion (PICs), dioxins, furans, coplanar PCBs, total hydrocarbons, and COX and NOX, as well as the particle size distribution (PSD) of particulate matter discharged at the end of the off-gas train. A description of the melter test requirements and analytical methods used is provided in the Test Plan for this work. Test Exceptions were subsequently issued which changed the TCO catalyst, added total organic emissions (TOE) to exhaust sampling schedule, and allowing modification of the test conditions in response to attainable plenum temperatures as well as temperature increases in the sulfur impregnated activated carbon (AC-S) column. Data are provided in this final report for all the required emission samples as well as melter and off-gas conditions during all the sampling periods. Appended to this report are previously issued VSL Letter Reports on method development for monitoring allyl alcohol in melter exhaust streams, on the results of characterization of the selected AC-S carbon media (Donnau BAT37), and on DM1200 off-line tests on the AC-S bed; also appended are reports from Air Tech on emissions sampling, and reports from Keika Ventures on validation of analytical data provided by Severn Trent Laboratories of Knoxville, Tennessee.

  8. Final Report - Glass Formulation Development and Testing for DWPF High AI2O3 HLW Sludges, VSL-10R1670-1, Rev. 0, dated 12/20/10

    SciTech Connect (OSTI)

    Kruger, Albert A.; Pegg, I. L.; Kot, W. K.; Gan, H.; Matlack, K. S.

    2013-11-13

    The principal objective of the work described in this Final Report is to develop and identify glass frit compositions for a specified DWPF high-aluminum based sludge waste stream that maximizes waste loading while maintaining high production rate for the waste composition provided by ORP/SRS. This was accomplished through a combination of crucible-scale, vertical gradient furnace, and confirmation tests on the DM100 melter system. The DM100-BL unit was selected for these tests. The DM100-BL was used for previous tests on HLW glass compositions that were used to support subsequent tests on the HLW Pilot Melter. It was also used to process compositions with waste loadings limited by aluminum, bismuth, and chromium, to investigate the volatility of cesium and technetium during the vitrification of an HLW AZ-102 composition, to process glass formulations at compositional and property extremes, and to investigate crystal settling on a composition that exhibited one percent crystals at 963{degrees}C (i.e., close to the WTP limit). The same melter was selected for the present tests in order to maintain comparisons between the previously collected data. The tests provide information on melter processing characteristics and off-gas data, including formation of secondary phases and partitioning. Specific objectives for the melter tests are as follows: Determine maximum glass production rates without bubbling for a simulated SRS Sludge Batch 19 (SB19). Demonstrate a feed rate equivalent to 1125 kg/m{sup 2}/day glass production using melt pool bubbling. Process a high waste loading glass composition with the simulated SRS SB19 waste and measure the quality of the glass product. Determine the effect of argon as a bubbling gas on waste processing and the glass product including feed processing rate, glass redox, melter emissions, etc.. Determine differences in feed processing and glass characteristics for SRS SB19 waste simulated by the co-precipitated and direct-hydroxide methods. The above tests were proposed based on previous tests for WTP in which there were few differences in the melter processing characteristics, such as processing rate and melter emissions, between precipitated and direct hydroxide simulants, even though there were differences in rheological properties. To the extent this similarity is found also for simulants for SRS HLW, the direct hydroxide methods may offer the potential for faster, simpler, and cheaper simulant production. There was no plan to match the yield stress and particle size of the direct hydroxide simulant to that of the precipitated simulant because that would have increased the preparation cost and complexity and defeated the purpose of the tests. These objectives were addressed by first developing a series of glass frits and then conducting a crucible scale study to determine the waste loading achievable for the waste composition and to select the preferred frit. Waste loadings were increased until the limits of a glass property were exceeded experimentally. Glass properties for evaluation included: viscosity, electrical conductivity, crystallinity (including liquidus temperature and nepheline formation after canister centerline cooling (CCC) heat-treatment), gross glass phase separation, and the 7- day Product Consistency Test (PCT, ASTM-1285) response. Glass property limits were based upon the constraints used for DWPF process control.

  9. SAFETY ANALYSIS REPORT FOR PACKAGING, MODEL 9977, ADDENDUM 3, JUSTIFICATION FOR SMALL GRAM QUANTITY CONTENTS

    SciTech Connect (OSTI)

    Abramczyk, G.

    2011-10-31

    This Addendum establishes a new family of content envelopes consisting of small quantities of radioactive materials. These content envelopes and specific packing configurations are shown to be subcritical. However, the dose rates of some payloads must be measured and shown to comply with applicable radiation limits. Authorization for shipment of the content envelop requires acceptance of this Addendum by the DOE-HQ certifying official as a supplement to the 9977 SARP Revision 2 and DOE-HQ�s subsequent revision of the CoC Revision 10 (which is based on SARP Addendum 2 and SARP Addendum 4) to authorize the additional content envelope. The Small Gram Quantity Content Envelopes and packing configurations will be incorporated in the next revision of the 9977 SARP.

  10. Blast furnace injection of massive quantities of coal with enriched air or pure oxygen

    SciTech Connect (OSTI)

    Ponghis, N.; Dufresne, P.; Vidal, R.; Poos, A. )

    1993-01-01

    An extensive study of the phenomena associated with the blast furnace injection of massive quantities of coal is described. Trials with conventional lances or oxy-coal injectors and hot blast at different oxygen contents - up to 40% - or with cold pure oxygen were realized at coal to oxygen ratios corresponding to a range of 150 to 440 kg. Pilot scale rigs, empty or filled with coke, as well as industrial blast furnaces were utilized.

  11. DELPHI expert panel evaluation of Hanford high level waste tank failure modes and release quantities

    SciTech Connect (OSTI)

    Dunford, G.L.; Han, F.C.

    1996-09-30

    The Failure Modes and Release Quantities of the Hanford High Level Waste Tanks due to postulated accident loads were established by a DELPHI Expert Panel consisting of both on-site and off-site experts in the field of Structure and Release. The Report presents the evaluation process, accident loads, tank structural failure conclusion reached by the panel during the two-day meeting.

  12. Level: National Data; Row: NAICS Codes; Column: Reasons that Made Quantity Unswitchable;

    U.S. Energy Information Administration (EIA) Indexed Site

    0 Reasons that Made Electricity Unswitchable, 2006; Level: National Data; Row: NAICS Codes; Column: Reasons that Made Quantity Unswitchable; Unit: Million kWh. Total Amount of Total Amount of Equipment is Not Switching Unavailable Long-Term Unavailable Combinations of NAICS Electricity Consumed Unswitchable Capable of Using Adversely Affects Alternative Environmenta Contract Storage for Another Columns F, G, Code(a) Subsector and Industry as a Fuel Electricity Fuel Use Another Fuel the Products

  13. Level: National Data; Row: NAICS Codes; Column: Reasons that Made Quantity Unswitchable;

    U.S. Energy Information Administration (EIA) Indexed Site

    1 Reasons that Made Natural Gas Unswitchable, 2006; Level: National Data; Row: NAICS Codes; Column: Reasons that Made Quantity Unswitchable; Unit: Billion cubic feet. Total Amount of Total Amount of Equipment is Not Switching Unavailable Long-Term Unavailable Combinations of NAICS Natural Gas Unswitchable Capable of Using Adversely Affects Alternative Environmenta Contract Storage for Another Columns F, G, Code(a) Subsector and Industry Consumed as a FueNatural Gas Fuel Use Another Fuel the

  14. Level: National Data; Row: NAICS Codes; Column: Reasons that Made Quantity Unswitchable;

    U.S. Energy Information Administration (EIA) Indexed Site

    2 Reasons that Made Coal Unswitchable, 2006; Level: National Data; Row: NAICS Codes; Column: Reasons that Made Quantity Unswitchable; Unit: Million short tons. Total Amount of Total Amount of Equipment is Not Switching Unavailable Long-Term Unavailable Combinations of NAICS Coal Consumed Unswitchable Capable of Using Adversely Affects Alternative Environmenta Contract Storage for Another Columns F, G, Code(a) Subsector and Industry as a Fuel Coal Fuel Use Another Fuel the Products Fuel

  15. Level: National Data; Row: NAICS Codes; Column: Reasons that Made Quantity Unswitchable;

    U.S. Energy Information Administration (EIA) Indexed Site

    3 Reasons that Made LPG Unswitchable, 2006; Level: National Data; Row: NAICS Codes; Column: Reasons that Made Quantity Unswitchable; Unit: Million barrels. Total Amount of Total Amount of Equipment is Not Switching Unavailable Long-Term Unavailable Combinations of NAICS LPG Consumed Unswitchable Capable of Using Adversely Affects Alternative Environmenta Contract Storage for Another Columns F, G, Code(a) Subsector and Industry as a Fuel LPG Fuel Use Another Fuel the Products Fuel

  16. Table A31. Quantity of Electricity Sold to Utility and Nonutility Purchasers

    U.S. Energy Information Administration (EIA) Indexed Site

    Quantity of Electricity Sold to Utility and Nonutility Purchasers by Census Region," " Census Division, and Economic Characteristics of the Establishment, 1994" " (Estimates in Million Kilowatthours)" ,,,,"RSE" " "," ","Utility ","Nonutility","Row" "Economic Characteristics(a)","Total Sold","Purchaser(b)","Purchaser(c)","Factors" ,"Total United

  17. Value of Demand Response: Quantities from Production Cost Modeling (Presentation), NREL (National Renewable Energy Laboratory)

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

    Value Study Desk Manual Value Study Desk Manual Updated September 26, 2012. PDF icon Memo from Robert Myers regarding DOE Benefit Value Desk Manual PDF icon Value Study Desk Manual More Documents & Publications Contractor Human Resources Management VALUE STUDY VALUE STUDY

    Value of Demand Response: Quantities from Production Cost Modeling Marissa Hummon PLMA Spring 2014 April 15-16, 2014 Denver, CO NREL/PR-6A20-61815 2 Background DOE-led, multiple national laboratory research project

  18. Practical thermodynamic quantities for aqueous vanadium- and iron-based flow batteries

    SciTech Connect (OSTI)

    Hudak, Nicholas S.

    2013-12-31

    A simple method for experimentally determining thermodynamic quantities for flow battery cell reactions is presented. Equilibrium cell potentials, temperature derivatives of cell potential (dE/dT), Gibbs free energies, and entropies are reported here for all-vanadium, iron–vanadium, and iron–chromium flow cells with state-of-the-art solution compositions. Proof is given that formal potentials and formal temperature coefficients can be used with modified forms of the Nernst Equation to quantify the thermodynamics of flow cell reactions as a function of state-of-charge. Such empirical quantities can be used in thermo-electrochemical models of flow batteries at the cell or system level. In most cases, the thermodynamic quantities measured here are significantly different from standard values reported and used previously in the literature. The data reported here are also useful in the selection of operating temperatures for flow battery systems. Because higher temperatures correspond to lower equilibrium cell potentials for the battery chemistries studied here, it can be beneficial to charge a cell at higher temperature and discharge at lower temperature. As a result, proof-of-concept of improved voltage efficiency with the use of such non-isothermal cycling is given for the all-vanadium redox flow battery, and the effect is shown to be more pronounced at lower current densities.

  19. FINAL REPORT MELTER TESTS WITH AZ-101 HLW SIMULANT USING A DURAMELTER 100 VITRIFICATION SYSTEM VSL-01R10N0-1 REV 1 2/25/02

    SciTech Connect (OSTI)

    KRUGER AA; MATLACK KS; KOT WK; PEGG IL

    2011-12-29

    This report provides data, analyses, and conclusions from a series of tests that were conducted at the Vitreous State Laboratory of The Catholic of America (VSL) to determine the processing rates that are achievable with AZ-101 HLW simulants and corresponding melter feeds on a DuraMelter 100 (DM100) vitrification system. One of the most critical pieces of information in determining the required size of the RPP-WTP HLW melter is the specific glass production rate in terms of the mass of glass that can be produced per unit area of melt surface per unit time. The specific glass production rate together with the waste loading (essentially, the ratio of waste-in to glass-out, which is determined from glass formulation activities) determines the melt area that is needed to achieve a given waste processing rate with due allowance for system availability. Tests conducted during Part B1 (VSL-00R2590-2) on the DM1000 vitrification system installed at the Vitreous State Laboratory of The Catholic University of America showed that, without the use of bubblers, glass production rates with AZ-101 and C-106/AY-102 simulants were significantly lower than the Project design basis rate of 0.4 MT/m{sup 2}/d. Conversely, three-fold increases over the design basis rate were demonstrated with the use of bubblers. Furthermore, an un-bubbled control test using a replica of the melter feed used in cold commissioning tests at West Valley reproduced the rates that were observed with that feed on the WVDP production melter. More recent tests conducted on the DM1200 system, which more closely represents the present RPP-WTP design, are in general agreement with these earlier results. Screening tests conducted on the DM10 system have provided good indications of the larger-scale processing rates with bubblers (for both HL W and LAW feeds) but significantly overestimated the DM1000 un-bubbled rate observed for C-106/AY-102 melter feeds. This behavior is believed to be a consequence of the role of heat transfer in rate attainment and the much greater role of wall effects in heat transfer when the melt pool is not agitated. The DM100 melter used for the present tests has a surface area of 0.108 m{sup 2}, which is approximately 5 times larger than that of the DM10 (0.021 m{sup 2}) and approximately 11 times smaller than that of the DM1000 (1.2 m{sup 2}) (the DM1000 has since been replaced by a pilot-scale prototypical HLW melter, designated the DM1200, which has the same surface area as the DM1000). Testing on smaller melters is the most economical method for obtaining data over a wide range of operating conditions (particularly at extremes) and for guiding the more expensive tests that are performed at pilot-scale. Thus, one objective of these tests was to determine whether the DM100 melters are sufficiently large to reproduce the un-bubbled melt rates observed at the DM1000 scale, or to determine the extent of any off-set. DM100-scale tests can then be used to screen feed chemistry variations that may serve to increase the un-bubbled production rates prior to confirmation at pilot scale. Finally, extensive characterization data obtained on simulated HLW melter feeds formed from various glass forming additives indicated that there may be advantages in terms of feed rheology and stability to the replacement of some of the hydroxides by carbonates. A further objective of the present tests was therefore to identify any deleterious processing effects of such a change before adopting the carbonate feed as the baseline. Data from the WVDP melter using acidified (nitrated) feeds, and without bubbling, showed productions rates that are higher than those observed with the alkaline RPP feeds at the VSL. Therefore, the effect of feed acidification on production rate also was investigated. This work was performed under Test Specification, 'TSP-W375-00-00019, Rev 0, 'HLW-DM10 and DM100 Melter Tests' dated November 13, 2000 and the corresponding Test Plan. It should be noted, however, that the RPP-WTP Project directed a series of changes to the Test Plan as the result

  20. ITEM NO. SUPPLIES/SERVICES QUANTITY UNIT UNIT PRICE AMOUNT NAME OF OFFEROR OR CONTRACTOR

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

    ITEM NO. SUPPLIES/SERVICES QUANTITY UNIT UNIT PRICE AMOUNT NAME OF OFFEROR OR CONTRACTOR 2 4 CONTINUATION SHEET REFERENCE NO. OF DOCUMENT BEING CONTINUED PAGE OF COMPUTER SCIENCES CORPORATION (A) (B) (C) (D) (E) (F) DE-AC06-04RL14383/108 c. As a result of this modification, the total obligated funding on the contract remains at $78,915,927.97. All other terms and conditions remain unchanged. FOB: Destination Period of Performance: 01/06/2004 to 09/30/2013 NSN 7540-01-152-8067 OPTIONAL FORM 336

  1. The behavior, quantity, and location of undissolved gas in Tank 241-SY-101

    SciTech Connect (OSTI)

    Brewster, M.E.; Gallagher, N.B.; Hudson, J.D.; Stewart, C.W.

    1995-10-01

    Mitigation of episodic flammable gas releases from Hanford Waste Tank 241-SY-101 was accomplished in July 1993 with the installation of a mixer pump that prevents gas retention. But is has not been possible until recently to measure the effects of mixing on the waste or how much gas remains and where it is located. Direct measurements of the void fraction and rheology of the mixed waste by the void fraction instrument (VFI) and ball rheometer along with previous data provide estimates of the location, quantity, and behavior of undissolved gas in the tank. This report documents the compilation and integration of the information that enables this understanding.

  2. ITEM NO. SUPPLIES/SERVICES QUANTITY UNIT UNIT PRICE AMOUNT NAME OF OFFEROR OR CONTRACTOR

    National Nuclear Security Administration (NNSA)

    ITEM NO. SUPPLIES/SERVICES QUANTITY UNIT UNIT PRICE AMOUNT NAME OF OFFEROR OR CONTRACTOR 2 2 CONTINUATION SHEET REFERENCE NO. OF DOCUMENT BEING CONTINUED PAGE OF BABCOCK & WILCOX TECHNICAL SERVICES Y-12, LLC (A) (B) (C) (D) (E) (F) DE-AC05-00OR22800/240 incorporated into the contract at Part III, List of Documents, Exhibits and Other Attachments, Section J Attachments, Attachment E Baseline List of Required Compliance Documents. RCN NNSA-43 and NNSA-44 are attached to this modification

  3. Practical thermodynamic quantities for aqueous vanadium- and iron-based flow batteries

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Hudak, Nicholas S.

    2013-12-31

    A simple method for experimentally determining thermodynamic quantities for flow battery cell reactions is presented. Equilibrium cell potentials, temperature derivatives of cell potential (dE/dT), Gibbs free energies, and entropies are reported here for all-vanadium, iron–vanadium, and iron–chromium flow cells with state-of-the-art solution compositions. Proof is given that formal potentials and formal temperature coefficients can be used with modified forms of the Nernst Equation to quantify the thermodynamics of flow cell reactions as a function of state-of-charge. Such empirical quantities can be used in thermo-electrochemical models of flow batteries at the cell or system level. In most cases, the thermodynamic quantitiesmore » measured here are significantly different from standard values reported and used previously in the literature. The data reported here are also useful in the selection of operating temperatures for flow battery systems. Because higher temperatures correspond to lower equilibrium cell potentials for the battery chemistries studied here, it can be beneficial to charge a cell at higher temperature and discharge at lower temperature. As a result, proof-of-concept of improved voltage efficiency with the use of such non-isothermal cycling is given for the all-vanadium redox flow battery, and the effect is shown to be more pronounced at lower current densities.« less

  4. PCP METHODOLOGY FOR DETERMINING DOSE RATES FOR SMALL GRAM QUANTITIES IN SHIPPING PACKAGINGS

    SciTech Connect (OSTI)

    Nathan, S.

    2011-08-23

    The Small Gram Quantity (SGQ) concept is based on the understanding that small amounts of hazardous materials, in this case radioactive materials, are significantly less hazardous than large amounts of the same materials. This study describes a methodology designed to estimate an SGQ for several neutron and gamma emitting isotopes that can be shipped in a package compliant with 10 CFR Part 71 external radiation level limits regulations. These regulations require packaging for the shipment of radioactive materials perform, under both normal and accident conditions, the essential functions of material containment, subcriticality, and maintain external radiation levels within regulatory limits. 10 CFR 71.33(b)(1)(2)&(3) state radioactive and fissile materials must be identified and their maximum quantity, chemical and physical forms be included in an application. Furthermore, the U.S. Federal Regulations require application contain an evaluation demonstrating the package (i.e., the packaging and its contents) satisfies the external radiation standards for all packages (10 CFR 71.31(2), 71.35(a), & 71.47). By placing the contents in a He leak-tight containment vessel, and limiting the mass to ensure subcriticality, the first two essential functions are readily met. Some isotopes emit sufficiently strong photon radiation that small amounts of material can yield a large external dose rate. Quantifying of the dose rate for a proposed content is a challenging issue for the SGQ approach. It is essential to quantify external radiation levels from several common gamma and neutron sources that can be safely placed in a specific packaging, to ensure compliance with federal regulations. The Packaging Certification Program (PCP) Methodology for Determining Dose Rate for Small Gram Quantities in Shipping Packagings described in this report provides bounding mass limits for a set of proposed SGQ isotopes. Methodology calculations were performed to estimate external radiation levels for the 9977 shipping package using the MCNP radiation transport code to develop a set of response multipliers (Green's functions) for 'dose per particle' for each neutron and photon spectral group. The source spectrum for each isotope generated using the ORIGEN-S and RASTA computer codes was folded with the response multipliers to generate the dose rate per gram of each isotope in the 9977 shipping package and its associated shielded containers. The maximum amount of a single isotope that could be shipped within the regulatory limits contained in 10 CFR 71.47 for dose rate at the surface of the package is determined. If a package contains a mixture of isotopes, the acceptability for shipment can be determined by a sum of fractions approach. Furthermore, the results of this analysis can be easily extended to additional radioisotopes by simply evaluating the neutron and/or photon spectra of those isotopes and folding the spectral data with the Green's functions provided.

  5. PACKAGING CERTIFICATION PROGRAM METHODOLOGY FOR DETERMINING DOSE RATES FOR SMALL GRAM QUANTITIES IN SHIPPING PACKAGINGS

    SciTech Connect (OSTI)

    Nathan, S.; Loftin, B.; Abramczyk, G.; Bellamy, S.

    2012-05-09

    The Small Gram Quantity (SGQ) concept is based on the understanding that small amounts of hazardous materials, in this case radioactive materials (RAM), are significantly less hazardous than large amounts of the same materials. This paper describes a methodology designed to estimate an SGQ for several neutron and gamma emitting isotopes that can be shipped in a package compliant with 10 CFR Part 71 external radiation level limits regulations. These regulations require packaging for the shipment of radioactive materials, under both normal and accident conditions, to perform the essential functions of material containment, subcriticality, and maintain external radiation levels within the specified limits. By placing the contents in a helium leak-tight containment vessel, and limiting the mass to ensure subcriticality, the first two essential functions are readily met. Some isotopes emit sufficiently strong photon radiation that small amounts of material can yield a large dose rate outside the package. Quantifying the dose rate for a proposed content is a challenging issue for the SGQ approach. It is essential to quantify external radiation levels from several common gamma and neutron sources that can be safely placed in a specific packaging, to ensure compliance with federal regulations. The Packaging Certification Program (PCP) Methodology for Determining Dose Rate for Small Gram Quantities in Shipping Packagings provides bounding shielding calculations that define mass limits compliant with 10 CFR 71.47 for a set of proposed SGQ isotopes. The approach is based on energy superposition with dose response calculated for a set of spectral groups for a baseline physical packaging configuration. The methodology includes using the MCNP radiation transport code to evaluate a family of neutron and photon spectral groups using the 9977 shipping package and its associated shielded containers as the base case. This results in a set of multipliers for 'dose per particle' for each spectral group. For a given isotope, the source spectrum is folded with the response for each group. The summed contribution from all isotopes determines the total dose from the RAM in the container.

  6. A COMPLETE HISTORY OF THE HIGH-LEVEL WASTE PLANT AT THE WEST VALLEY DEMONSTRATION PROJECT

    SciTech Connect (OSTI)

    Petkus, Lawrence L.; Paul, James; Valenti, Paul J.; Houston, Helene; May, Joseph

    2003-02-27

    The West Valley Demonstration Project (WVDP) vitrification melter was shut down in September 2002 after being used to vitrify High Level Waste (HLW) and process system residuals for six years. Processing of the HLW occurred from June 1996 through November 2001, followed by a program to flush the remaining HLW through to the melter. Glass removal and shutdown followed. The facility and process equipment is currently in a standby mode awaiting deactivation. During HLW processing operations, nearly 24 million curies of radioactive material were vitrified into 275 canisters of HLW glass. At least 99.7% of the curies in the HLW tanks at the WVDP were vitrified using the melter. Each canister of HLW holds approximately 2000 kilograms of glass with an average contact dose rate of over 2600 rem per hour. After vitrification processing ended, two more cans were filled using the Evacuated Canister Process to empty the melter at shutdown. This history briefly summarizes the initial stages of process development and earlier WVDP experience in the design and operation of the vitrification systems, followed by a more detailed discussion of equipment availability and failure rates during six years of operation. Lessons learned operating a system that continued to function beyond design expectations also are highlighted.

  7. A Ceramic membrane to Recycle Caustic

    Office of Environmental Management (EM)

    High-Level Waste (HLW) tanks must be maintained in a caustic environment to inhibit corrosion. Consequently, they contain large quantities of NaOH. Ultimately the HLW will be...

  8. A GREEN'S FUNCTION APPROACH FOR DETERMINING DOSE RATES FOR SMALL GRAM QUANTITIES IN SHIPPING PACKAGINGS

    SciTech Connect (OSTI)

    Nathan, S.

    2012-06-14

    The Small Gram Quantity (SGQ) concept is based on the understanding that small amounts of hazardous materials, in this case radioactive materials (RAM), are significantly less hazardous than large amounts of the same materials. This paper describes a methodology designed to estimate an SGQ for several neutron and gamma emitting isotopes that can be shipped in a package in compliance with 10 CFR Part 71 external radiation level limits regulations. The neutron and photon sources were calculated using both ORIGEN-S and RASTA. The response from a unit source in each neutron and photon group was calculated using MCNP5 with each unshielded and shielded container configuration. Effects of self-shielding on both neutron and photon response were evaluated by including either plutonium oxide or iron in the source region for the case with no shielded container. For the cases of actinides mixed with light elements, beryllium is the bounding light element. The added beryllium (10 to 90 percent of the actinide mass) in the cases studied represents between 9 and 47 percent concentration of the total mixture mass. For beryllium concentrations larger than 50 percent, the increase in the neutron source term and dose rate tend to increase at a much lower rate than at concentrations lower than 50%. The intimately mixed actinide-beryllium form used in these models is very conservative and thus the limits presented in this report are practical bounds on the mass that can be safely shipped. The calculated dose rate from one gram of each isotope was then used to determin the maximum amount of a single isotope that could be shipped in the Model 9977 Package (or packagings having the same or larger external dimensions as well as similar structural materials) and have the external radiation level within the regulatory dose limits at the surface of the package. The estimates of the mass limits presented would also serve as conservative limits for both the Models 9975 and 9978 packages. If a package contains a mixture of isotopes, the acceptability for shipment can be determined by a sum of fractions approach. It should be noted that the SGQ masses presented in this report represent limits that would comply with the external radiation limits under 10CFR Part 71. They do not necessarily bound lower limits that may be required to comply with other factors such as heat load of the package.

  9. Approximate models for the study of exponential changed quantities: Application on the plasma waves growth rate or damping

    SciTech Connect (OSTI)

    Xaplanteris, C. L.; Xaplanteris, L. C.; Leousis, D. P.

    2014-03-15

    Many physical phenomena that concern the research these days are basically complicated because of being multi-parametric. Thus, their study and understanding meets with big if not unsolved obstacles. Such complicated and multi-parametric is the plasmatic state as well, where the plasma and the physical quantities that appear along with it have chaotic behavior. Many of those physical quantities change exponentially and at most times they are stabilized by presenting wavy behavior. Mostly in the transitive state rather than the steady state, the exponentially changing quantities (Growth, Damping etc) depend on each other in most cases. Thus, it is difficult to distinguish the cause from the result. The present paper attempts to help this difficult study and understanding by proposing mathematical exponential models that could relate with the study and understanding of the plasmatic wavy instability behavior. Such instabilities are already detected, understood and presented in previous publications of our laboratory. In other words, our new contribution is the study of the already known plasmatic quantities by using mathematical models (modeling and simulation). These methods are both useful and applicable in the chaotic theory. In addition, our ambition is to also conduct a list of models useful for the study of chaotic problems, such as those that appear into the plasma, starting with this paper's examples.

  10. ORP Grand Challenges 2015 HLW Direct Vitrification

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

    4/59 Emissions and Performance Benchmarking of a Prototype Dimethyl Ether-Fueled Heavy-Duty Truck February 2014 Prepared by James P. Szybist Oak Ridge National Laboratory Samuel McLaughlin Volvo North America Suresh Iyer The Pennsylvania State University DOCUMENT AVAILABILITY Reports produced after January 1, 1996, are generally available free via the U.S. Department of Energy (DOE) Information Bridge. Web site http://www.osti.gov/bridge Reports produced before January 1, 1996, may be purchased

  11. ID HLW FEIS amended ROD.pdf

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

  12. Summary - WTP HLW Waste Vitrification Facility

    Office of Environmental Management (EM)

    ( adequacy (W processing H corrosion res stainless ste Testing activ contacted wi halogens), s Testing the p blending liqu requirements established. maturity of certain key...

  13. Microsoft PowerPoint - Reporting Small Quantities to NMMSS_Brian Horn_Suzanne Ani [Compatibility Mode]

    National Nuclear Security Administration (NNSA)

    Small Quantities to NMMSS Brian Horn, Suzanne Ani, Nuclear Regulatory Commission Overview  Introduction  Reportable Units  One-Party Transactions (A-M)  Reporting to NMMSS 2 Introduction  NRC regulations require each licensee who transfers or receives SNM to complete a DOE/NRC Form 741, Nuclear Material Transaction Report  DOE also requires NRC licensees to report to NMMSS all receipts, transfers, and inventories of government-owned, -leased, or -loaned material in their

  14. Definition of Small Gram Quantity Contents for Type B Radioactive Material Transportation Packages: Activity-Based Content Limitations

    SciTech Connect (OSTI)

    Sitaraman, S; Kim, S; Biswas, D; Hafner, R; Anderson, B

    2010-10-27

    Since the 1960's, the Department of Transportation Specification (DOT Spec) 6M packages have been used extensively for transportation of Type B quantities of radioactive materials between Department of Energy (DOE) facilities, laboratories, and productions sites. However, due to the advancement of packaging technology, the aging of the 6M packages, and variability in the quality of the packages, the DOT implemented a phased elimination of the 6M specification packages (and other DOT Spec packages) in favor of packages certified to meet federal performance requirements. DOT issued the final rule in the Federal Register on October 1, 2004 requiring that use of the DOT Specification 6M be discontinued as of October 1, 2008. A main driver for the change was the fact that the 6M specification packagings were not supported by a Safety Analysis Report for Packaging (SARP) that was compliant with Title 10 of the Code of Federal Regulations part 71 (10 CFR 71). Therefore, materials that would have historically been shipped in 6M packages are being identified as contents in Type B (and sometimes Type A fissile) package applications and addenda that are to be certified under the requirements of 10 CFR 71. The requirements in 10 CFR 71 include that the Safety Analysis Report for Packaging (SARP) must identify the maximum radioactivity of radioactive constituents and maximum quantities of fissile constituents (10 CFR 71.33(b)(1) and 10 CFR 71.33(b)(2)), and that the application (i.e., SARP submittal or SARP addendum) demonstrates that the external dose rate (due to the maximum radioactivity of radioactive constituents and maximum quantities of fissile constituents) on the surface of the packaging (i.e., package and contents) not exceed 200 mrem/hr (10 CFR 71.35(a), 10 CFR 71.47(a)). It has been proposed that a 'Small Gram Quantity' of radioactive material be defined, such that, when loaded in a transportation package, the dose rates at external points of an unshielded packaging not exceed the regulatory limits prescribed by 10 CFR 71 for non-exclusive shipments. The mass of each radioisotope presented in this paper is limited by the radiation dose rate on the external surface of the package, which per the regulatory limit should not exceed 200 mrem/hr. The results presented are a compendium of allowable masses of a variety of different isotopes (with varying impurity levels of beryllium in some of the actinide isotopes) that, when loaded in an unshielded packaging, do not result in an external dose rate on the surface of the package that exceeds 190 mrem/hr (190 mrem/hr was chosen to provide 5% conservatism relative to the regulatory limit). These mass limits define the term 'Small Gram Quantity' (SGQ) contents in the context of radioactive material transportation packages. The term SGQ is isotope-specific and pertains to contents in radioactive material transportation packages that do not require shielding and still satisfy the external dose rate requirements. Since these calculated mass limits are for contents without shielding, they are conservative for packaging materials that provide some limited shielding or if the contents are placed into a shielded package. The isotopes presented in this paper were chosen as the isotopes that Department of Energy (DOE) sites most likely need to ship. Other more rarely shipped isotopes, along with industrial and medical isotopes, are planned to be included in subsequent extensions of this work.

  15. THERMAL ANALYSIS OF GEOLOGIC HIGH-LEVEL RADIOACTIVE WASTE PACKAGES

    SciTech Connect (OSTI)

    Hensel, S.; Lee, S.

    2010-04-20

    The engineering design of disposal of the high level waste (HLW) packages in a geologic repository requires a thermal analysis to provide the temperature history of the packages. Calculated temperatures are used to demonstrate compliance with criteria for waste acceptance into the geologic disposal gallery system and as input to assess the transient thermal characteristics of the vitrified HLW Package. The objective of the work was to evaluate the thermal performance of the supercontainer containing the vitrified HLW in a non-backfilled and unventilated underground disposal gallery. In order to achieve the objective, transient computational models for a geologic vitrified HLW package were developed by using a computational fluid dynamics method, and calculations for the HLW disposal gallery of the current Belgian geological repository reference design were performed. An initial two-dimensional model was used to conduct some parametric sensitivity studies to better understand the geologic system's thermal response. The effect of heat decay, number of co-disposed supercontainers, domain size, humidity, thermal conductivity and thermal emissivity were studied. Later, a more accurate three-dimensional model was developed by considering the conduction-convection cooling mechanism coupled with radiation, and the effect of the number of supercontainers (3, 4 and 8) was studied in more detail, as well as a bounding case with zero heat flux at both ends. The modeling methodology and results of the sensitivity studies will be presented.

  16. Quantity | Open Energy Information

    Open Energy Info (EERE)

    Property:EstimatedTimeHigh Property:EstimatedTimeLow Property:EstimatedTimeMedian F Property:FirstWellDepth Property:FirstWellFlowRate G Property:GeneratingCapacity...

  17. Reportable Quantity-Calculator

    Broader source: Energy.gov [DOE]

    Any time a hazardous substance as defined under the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA or more commonly known as Superfund) is released to the environment...

  18. SHIELDING AND DETECTOR RESPONSE CALCULATIONS PERTAINING TO CATEGORY 1 QUANTITIES OF PLUTONIUM AND HAND-HELD PLASTIC SCINTILLATORS

    SciTech Connect (OSTI)

    Couture, A.

    2013-06-07

    Nuclear facilities sometimes use hand-held plastic scintillator detectors to detect attempts to divert special nuclear material in situations where portal monitors are impractical. MCNP calculations have been performed to determine the neutron and gamma radiation field arising from a Category I quantity of weapons-grade plutonium in various shielding configurations. The shields considered were composed of combinations of lead and high-density polyethylene such that the mass of the plutonium plus shield was 22.7 kilograms. Monte-Carlo techniques were also used to determine the detector response to each of the shielding configurations. The detector response calculations were verified using field measurements of high-, medium-, and low- energy gamma-ray sources as well as a Cf-252 neutron source.

  19. Quantities and characteristics of the contact-handled low-level mixed waste streams for the DOE complex

    SciTech Connect (OSTI)

    Huebner, T.L.; Wilson, J.M.; Ruhter, A.H.; Bonney, S.J.

    1994-08-01

    This report supports the Integrated Thermal Treatment System (ITTS) Study initiated by the Department of Energy (DOE) Office of Technology Development (EM-50), which is a system engineering assessment of a variety of mixed waste treatment process. The DOE generates and stores large quantities of mixed wastes that are contaminated with both chemically hazardous and radioactive species. The treatment of these mixed wastes requires meeting the standards established by the Environmental Protection Agency for the specific hazardous contaminants regulated under the Resource Conservation and Recovery Act while also providing adequate control of the radionuclides. The thrust of the study is to develop preconceptual designs and life-cycle cost estimates for integrated thermal treatment systems ranging from conventional incinerators, such as rotary kiln and controlled air systems, to more innovative but not yet established technologies, such as molten salt and molten metal waste destruction systems. Prior to this engineering activity, the physical and chemical characteristics of the DOE low-level mixed waste streams to be treated must be defined or estimated. This report describes efforts to estimate the DOE waste stream characteristics.

  20. PLUTONIUM/HIGH-LEVEL VITRIFIED WASTE BDBE DOSE CALCULATION

    SciTech Connect (OSTI)

    D.C. Richardson

    2003-03-19

    In accordance with the Nuclear Waste Policy Amendments Act of 1987, Yucca Mountain was designated as the site to be investigated as a potential repository for the disposal of high-level radioactive waste. The Yucca Mountain site is an undeveloped area located on the southwestern edge of the Nevada Test Site (NTS), about 100 miles northwest of Las Vegas. The site currently lacks rail service or an existing right-of-way. If the Yucca Mountain site is found suitable for the repository, rail service is desirable to the Office of Civilian Waste Management (OCRWM) Program because of the potential of rail transportation to reduce costs and to reduce the number of shipments relative to highway transportation. A Preliminary Rail Access Study evaluated 13 potential rail spur options. Alternative routes within the major options were also developed. Each of these options was then evaluated for potential land use conflicts and access to regional rail carriers. Three potential routes having few land use conflicts and having access to regional carriers were recommended for further investigation. Figure 1-1 shows these three routes. The Jean route is estimated to be about 120 miles long, the Carlin route to be about 365 miles long, and Caliente route to be about 365 miles long. The remaining ten routes continue to be monitored and should any of the present conflicts change, a re-evaluation of that route will be made. Complete details of the evaluation of the 13 routes can be found in the previous study. The DOE has not identified any preferred route and recognizes that the transportation issues need a full and open treatment under the National Environmental Policy Act. The issue of transportation will be included in public hearings to support development of the Environmental Impact Statement (EIS) proceedings for either the Monitored Retrievable Storage Facility or the Yucca Mountain Project or both.

  1. Iron Phosphate Glasses: An Alternative for Vitrifying Certain Nuclear Wastes

    SciTech Connect (OSTI)

    Delbert E. Day; Chandra S. Ray; Cheol-Woon Kim

    2004-12-28

    Vitrification of nuclear waste in a glass is currently the preferred process for waste disposal. DOE currently approves only borosilicate (BS) type glasses for such purposes. However, many nuclear wastes, presently awaiting disposal, have complex and diverse chemical compositions, and often contain components that are poorly soluble or chemically incompatible in BS glasses. Such problematic wastes can be pre-processed and/or diluted to compensate for their incompatibility with a BS glass matrix, but both of these solutions increases the wasteform volume and the overall cost for vitrification. Direct vitrification using alternative glasses that utilize the major components already present in the waste is preferable, since it avoids pre-treating or diluting the waste, and, thus, minimizes the wasteform volume and overall cost.

  2. Properties of vitrified Rocky Flats TRUW with different waste loadings

    SciTech Connect (OSTI)

    Eddy, T.L.; Sears, J.W.; Grandy, J.D.; Miley, D.V.; Erickson, A.W.; Fransworth, R.N.; Larsen, E.D.

    1994-07-01

    One of the major waste streams at the Idaho National Laboratory (INEL) is a combination of the Rocky Flats Plant 1st and 2nd stage sludges (hydrated metal oxides or H-series), which constitutes about 20 wt % of the buried waste. A similar mass fraction is in interim storage. The buried waste is commingled with about five times as much soil that has become contaminated as the containers have deteriorated. The purpose of this paper is to report on waste form property variations of the H-series waste melted with various fractions of soil, plus volatile and hazardous metals and transuranic surrogates. Optimally, the waste form will minimize the bulk leach rate, maximize the volume reduction, minimize the additives needed, and stabilize the transuranic nuclides. Topics to be discussed include the input and final compositions, the melting and crystallization processes, the test results, and conclusions.

  3. Hanford Makes Progress Toward Vitrifying Waste with Facility...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    March 16, 2016 - 12:30pm Addthis Workers excavate for the Effluent Management ... WTP to support DFLAW," explained Jason Young, federal project director for the WTP ...

  4. DOE Statement on Savannah River Site Vitrified Waste Concentrations

    Office of Energy Efficiency and Renewable Energy (EERE)

    The Office of Environmental Management has decided not to move forward at this time with its February decision to direct contractors to start planning for higher concentrations of plutonium in...

  5. Vitrified magnesia dissolution and its impact on plutonium residue processing

    SciTech Connect (OSTI)

    Keith W. Fife; Jennifer L. Alwin; Coleman A. Smith; Michael D. Mayne; David A. Rockstraw

    2000-03-01

    Aqueous chloride operations at the Los Alamos Plutonium Facility cannot directly dispose of acidic waste solutions because of compatibility problems with existing disposal lines. Consequently, all hydrochloric acid must be neutralized and filtered prior to exiting the facility. From a waste minimization standpoint, the use of spent magnesia pyrochemical crucibles as the acid neutralization agent is attractive since this process would take a stream destined for transuranic waste and use it as a reagent in routine plutonium residue processing. Since Los Alamos National Laboratory has several years of experience using magnesium hydroxide as a neutralizing agent for waste acid from plutonium processing activities, the use of spent magnesia pyrochemical crucibles appeared to be an attractive extension of this activity. In order to be competitive with magnesium hydroxide, however, size reduction of crucible shards had to be performed effectively within the constraints of glovebox operations, and acid neutralization time using crucible shards had to be comparable to neutralization times observed when using reagent-grade magnesium hydroxide. The study utilized non-plutonium-contaminated crucibles for equipment evaluation and selection and used nonradioactive acid solutions for completing the neutralization experiments. This paper discusses experience in defining appropriate size reduction equipment and presents results from using the magnesia crucibles for hydrochloric acid neutralization, a logical precursor to introduction into glovebox enclosures.

  6. Systems study of the feasibility of high-level nuclear-waste fractionation for thermal stress control in a geologic repository: main report

    SciTech Connect (OSTI)

    McKee, R.W.; Elder, H.K.; McCallum, R.F.; Silviera, D.J.; Swanson, J.L.; Wiles, L.E.

    1983-06-01

    This study assesses the benefits and costs of fractionating the cesium and strontium (Cs/Sr) components in commercial high-level waste (HLW) to a separate waste stream for the purpose of reducing geologic-repository thermal stresses in the region of the HLW. System costs are developed for a broad range of conditions comparing the Cs/Sr fractionation concept with disposal of 10-year-old vitrified HLW and vitrified HLW aged to achieve (through decay) the same heat output as the fractionated high-level waste (FHLW). All comparisons are based on a 50,000 metric ton equivalent (MTE) system. The FHLW and the Cs/Sr waste are both disposed of as vitrified waste but emplaced in separate areas of a basalt repository. The FHLW is emplaced in high-integrity packages at relatively high waste loading but low heat loading, while the Cs/Sr waste is emplaced in minimum-integrity packages at relatively high heat loading in a separate region of the repository. System cost comparisons are based on minimum cost combinations of canister diameter, waste concentration, and canister spacing in a basalt repository. The effects on both long- and near-term safety considerations are also addressed. The major conclusion is that the Cs/Sr fractionation concept offers the prospect of a substantial total system cost advantage for HLW disposal if reduced HLW package temperatures in a basalt repository are desired. However, there is no cost advantage if currently designated maximum design temperatures are acceptable. Aging the HLW for 50 to 100 years can accomplish similar results at equivalent or lower costs. 37 figures, 58 tables.

  7. An indirect sensing technique for diesel fuel quantity control. Technical progress report, October 1--December 31, 1998

    SciTech Connect (OSTI)

    MacCarley, C.A.

    1999-01-26

    Work has proceeded intensely with the objective of completing the commercial prototype system prior to the end of the contract period. At the time of this report, testing and refinement of the commercial version of the system has not been completed. During this reporting period, several major milestones were reached and many significant lessons were learned. These are described. The experimental retrofit system has achieved all performance objectives in engine dynamometer tests. The prototype commercial version of the system will begin demonstration service on the first of several Santa Maria Area Transit (SMAT) transit buses on February 1, 1999. The commercial system has been redesignated the Electronic Diesel Smoke Reduction System (EDSRS) replacing the original internal pseudonym ADSC. The focus has been narrowed to a retrofit product suitable for installation on existing mechanically-governed diesel engines. Included in this potential market are almost all diesel-powered passenger cars and light trucks manufactured prior to the introduction of the most recent clean diesel engines equipped with particulate traps and electronic controls. Also included are heavy-duty trucks, transit vehicles, school buses, and agricultural equipment. This system is intended to prevent existing diesel engines from overfueling to the point of visible particulate emissions (smoke), while allowing maximum smoke-limited torque under all operating conditions. The system employs a microcontroller and a specialized exhaust particulate emission sensor to regulate the maximum allowable fuel quantity via an adaptive throttle-limit map. This map specifies a maximum allowable throttle position as a function of engine speed, turbocharger boost pressure and engine coolant temperature. The throttle position limit is mechanized via a servo actuator inserted in the throttle cable leading to the injection pump.

  8. Supplemental Immobilization Cast Stone Technology Development and Waste Form Qualification Testing Plan

    SciTech Connect (OSTI)

    Westsik, Joseph H.; Serne, R. Jeffrey; Pierce, Eric M.; Cozzi, Alex; Chung, Chul-Woo; Swanberg, David J.

    2013-05-31

    The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being constructed to treat the 56 million gallons of radioactive waste stored in 177 underground tanks at the Hanford Site. The WTP includes a pretreatment facility to separate the wastes into high-level waste (HLW) and low-activity waste (LAW) fractions for vitrification and disposal. The LAW will be converted to glass for final disposal at the Integrated Disposal Facility (IDF). The pretreatment facility will have the capacity to separate all of the tank wastes into the HLW and LAW fractions, and the HLW Vitrification Facility will have the capacity to vitrify all of the HLW. However, a second immobilization facility will be needed for the expected volume of LAW requiring immobilization. A number of alternatives, including Cast Stone—a cementitious waste form—are being considered to provide the additional LAW immobilization capacity.

  9. ALUMINUM AND CHROMIUM LEACHING WORKSHOP WHITEPAPER

    SciTech Connect (OSTI)

    McCabe, D; Jeff Pike, J; Bill Wilmarth, B

    2007-04-25

    A workshop was held on January 23-24, 2007 to discuss the status of processes to leach constituents from High Level Waste (HLW) sludges at the Hanford and Savannah River Sites. The objective of the workshop was to examine the needs and requirements for the HLW flowsheet for each site, discuss the status of knowledge of the leaching processes, communicate the research plans, and identify opportunities for synergy to address knowledge gaps. The purpose of leaching of non-radioactive constituents from the sludge waste is to reduce the burden of material that must be vitrified in the HLW melter systems, resulting in reduced HLW glass waste volume, reduced disposal costs, shorter process schedules, and higher facility throughput rates. The leaching process is estimated to reduce the operating life cycle of SRS by seven years and decrease the number of HLW canisters to be disposed in the Repository by 1000 [Gillam et al., 2006]. Comparably at Hanford, the aluminum and chromium leaching processes are estimated to reduce the operating life cycle of the Waste Treatment Plant by 20 years and decrease the number of canisters to the Repository by 15,000-30,000 [Gilbert, 2007]. These leaching processes will save the Department of Energy (DOE) billions of dollars in clean up and disposal costs. The primary constituents targeted for removal by leaching are aluminum and chromium. It is desirable to have some aluminum in glass to improve its durability; however, too much aluminum can increase the sludge viscosity, glass viscosity, and reduce overall process throughput. Chromium leaching is necessary to prevent formation of crystalline compounds in the glass, but is only needed at Hanford because of differences in the sludge waste chemistry at the two sites. Improving glass formulations to increase tolerance of aluminum and chromium is another approach to decrease HLW glass volume. It is likely that an optimum condition can be found by both performing leaching and improving formulations. Disposal of the resulting aluminum and chromium-rich streams are different at the two sites, with vitrification into Low Activity Waste (LAW) glass at Hanford, and solidification in Saltstone at SRS. Prior to disposal, the leachate solutions must be treated to remove radionuclides, resulting in increased operating costs and extended facility processing schedules. Interim storage of leachate can also add costs and delay tank closure. Recent projections at Hanford indicate that up to 40,000 metric tons of sodium would be needed to dissolve the aluminum and maintain it in solution, which nearly doubles the amount of sodium in the entire current waste tank inventory. This underscores the dramatic impact that the aluminum leaching can have on the entire system. A comprehensive view of leaching and the downstream impacts must therefore be considered prior to implementation. Many laboratory scale tests for aluminum and chromium dissolution have been run on Hanford wastes, with samples from 46 tanks tested. Three samples from SRS tanks have been tested, out of seven tanks containing high aluminum sludge. One full-scale aluminum dissolution was successfully performed on waste at SRS in 1982, but generated a very large quantity of liquid waste ({approx}3,000,000 gallons). No large-scale tests have been done on Hanford wastes. Although the data to date give a generally positive indication that aluminum dissolution will work, many issues remain, predominantly because of variable waste compositions and changes in process conditions, downstream processing, or storage limitations. Better approaches are needed to deal with the waste volumes and limitations on disposal methods. To develop a better approach requires a more extensive understanding of the kinetics of dissolution, as well as the factors that effect rates, effectiveness, and secondary species. Models of the dissolution rate that have been developed are useful, but suffer from limitations on applicable compositional ranges, mineral phases, and particle properties that are difficult to measure. The experimental

  10. Quantity and quality of stormwater runoff recharged to the Floridan aquifer system through two drainage wells in the Orlando, Florida area

    SciTech Connect (OSTI)

    German, E.R.

    1989-01-01

    Quantity and quality of inflow to two drainage wells in the Orlando, Fla., area were determined for the period April 1982 through March 1983. The wells, located at Lake Midget and at Park Lake, are used to control the lake levels during rainy periods. The lakes receive stormwater runoff from mixed residential-commercial areas of about 64 acres (Lake Midget) and 96 acres (Park Lake) and would frequently flood adjacent areas if the wells did not drain the excess stormwater. These lakes and wells are typical of stormwater drainage systems in the area.

  11. Kristallin-I performance assessment: First results (Book) | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    Book: Kristallin-I performance assessment: First results Citation Details In-Document Search Title: Kristallin-I performance assessment: First results The Kristallin-I performance assessment indicates that the Swiss concept for disposal of vitrified HLW deep in the crystalline basement of Northern Switzerland will offer sufficient safety. This conclusion is based on a scenario analysis and an associated consequence analysis using an extensive model chain. The planned system of engineered

  12. Evaluation of Shortline Railroads & SNF/HLW Rail Shipment Inspections...

    Office of Environmental Management (EM)

    of Spent Nuclear Fuel Task: Identify Shortline Railroads Serving Nuclear Power Plants Establish Contact Information with Railroads Officials Field Review of each ...

  13. Waste Treatment and Immobilation Plant HLW Waste Vitrification Facility |

    Energy Savers [EERE]

    Energy Waste Treatment Facility Saves Taxpayers Nearly $20 Million Waste Treatment Facility Saves Taxpayers Nearly $20 Million December 11, 2012 - 1:40pm Addthis A new enclosure for processing radioactive casks has put Oak Ridge on a path to finishing cleanup work two years ahead of schedule, saving nearly $20 million. | Photo courtesy of the Office of Environmental Management. A new enclosure for processing radioactive casks has put Oak Ridge on a path to finishing cleanup work two years

  14. H:\\cindy_pratt\\hlw rod.tif

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

    DOE considered the information in the Final EIS, a related Supplement Analysis (DOEIS-0287-SA-OI) (SA), and comments received on the Federal Register Notice (70 FR 44598; August ...

  15. Optimization of Deep Borehole Systems for HLW Disposal

    SciTech Connect (OSTI)

    Driscoll, Michael; Baglietto, Emilio; Buongiorno, Jacopo; Lester, Richard; Brady, Patrick; Arnold, B. W.

    2015-09-09

    This is the final report on a project to update and improve the conceptual design of deep boreholes for high level nuclear waste disposal. The effort was concentrated on application to intact US legacy LWR fuel assemblies, but conducted in a way in which straightforward extension to other waste forms, host rock types and countries was preserved. The reference fuel design version consists of a vertical borehole drilled into granitic bedrock, with the uppermost kilometer serving as a caprock zone containing a diverse and redundant series of plugs. There follows a one to two kilometer waste canister emplacement zone having a hole diameter of approximately 40-50 cm. Individual holes are spaced 200-300 m apart to form a repository field. The choice of verticality and the use of a graphite based mud as filler between the waste canisters and the borehole wall liner was strongly influenced by the expectation that retrievability would continue to be emphasized in US and worldwide repository regulatory criteria. An advanced version was scoped out using zinc alloy cast in place to fill void space inside a disposal canister and its encapsulated fuel assembly. This excludes water and greatly improves both crush resistance and thermal conductivity. However the simpler option of using a sand fill was found adequate and is recommended for near-term use. Thermal-hydraulic modeling of the low permeability and porosity host rock and its small (≤ 1%) saline water content showed that vertical convection induced by the waste’s decay heat should not transport nuclides from the emplacement zone up to the biosphere atop the caprock. First order economic analysis indicated that borehole repositories should be cost-competitive with shallower mined repositories. It is concluded that proceeding with plans to drill a demonstration borehole to confirm expectations, and to carry out priority experiments, such as retention and replenishment of in-hole water is in order.

  16. Development of a Rotary Microfilter for SRS HLW Applications

    SciTech Connect (OSTI)

    MICHAEL, POIRIER

    2004-11-24

    The processing rate of Savannah River Site high level waste decontamination processes are limited by the flow rate of the solid-liquid separation. The baseline process, using a 0.1 micron cross flow filter, produces 0.02 gpm/ft2 of filtrate under expected operating conditions. Savannah River National Laboratory personnel identified the rotary microfilter as a technology that could significantly increase filter flux, with throughput improvements of as much as 10X for that specific operation. With funding from the Department of Energy Office of Cleanup Technologies, SRNL personnel are evaluating and developing the rotary microfilter for radioactive service at SRS. This work includes pilot-scale and actual waste testing to evaluate system reliability, the impact of radiation on system components, the filter flux for a variety of waste streams, and relative performance for alternative filter media.

  17. Method for quantitative determination and separation of trace amounts of chemical elements in the presence of large quantities of other elements having the same atomic mass

    DOE Patents [OSTI]

    Miller, C.M.; Nogar, N.S.

    1982-09-02

    Photoionization via autoionizing atomic levels combined with conventional mass spectroscopy provides a technique for quantitative analysis of trace quantities of chemical elements in the presence of much larger amounts of other elements with substantially the same atomic mass. Ytterbium samples smaller than 10 ng have been detected using an ArF* excimer laser which provides the atomic ions for a time-of-flight mass spectrometer. Elemental selectivity of greater than 5:1 with respect to lutetium impurity has been obtained. Autoionization via a single photon process permits greater photon utilization efficiency because of its greater absorption cross section than bound-free transitions, while maintaining sufficient spectroscopic structure to allow significant photoionization selectivity between different atomic species. Separation of atomic species from others of substantially the same atomic mass is also described.

  18. A high-entropy-wind r-process study based on nuclear-structure quantities from the new finite-range droplet model FRDM(2012)

    SciTech Connect (OSTI)

    Kratz, Karl-Ludwig; Farouqi, Khalil; Mller, Peter E-mail: kfarouqi@lsw.uni-heidelberg.de

    2014-09-01

    Attempts to explain the source of r-process elements in our solar system (S.S.) by particular astrophysical sites still face entwined uncertainties, stemming from the extrapolation of nuclear properties far from stability, inconsistent sources of different properties (e.g., nuclear masses and ?-decay properties), and the poor understanding of astrophysical conditions, which are hard to disentangle. In this paper we present results from the investigation of r-process in the high-entropy wind (HEW) of core-collapse supernovae (here chosen as one of the possible scenarios for this nucleosynthesis process), using new nuclear-data input calculated in a consistent approach, for masses and ?-decay properties from the new finite-range droplet model FRDM(2012). The accuracy of the new mass model is 0.56 MeV with respect to AME2003, to which it was adjusted. We compare the new HEW r-process abundance pattern to the latest S.S. r-process residuals and to our earlier calculations with the nuclear-structure quantities based on FRDM(1992). Substantial overall and specific local improvements in the calculated pattern of the r-process between A ? 110 and {sup 209}Bi, as well as remaining deficiencies, are discussed in terms of the underlying spherical and deformed shell structure far from stability.

  19. Permitting plan for the high-level waste interim storage

    SciTech Connect (OSTI)

    Deffenbaugh, M.L.

    1997-04-23

    This document addresses the environmental permitting requirements for the transportation and interim storage of solidified high-level waste (HLW) produced during Phase 1 of the Hanford Site privatization effort. Solidified HLW consists of canisters containing vitrified HLW (glass) and containers that hold cesium separated during low-level waste pretreatment. The glass canisters and cesium containers will be transported to the Canister Storage Building (CSB) in a U.S. Department of Energy (DOE)-provided transportation cask via diesel-powered tractor trailer. 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 of Tank Waste Remediation Systems (TWRS) immobilized HLW (IHLW) and other canistered high-level waste forms; and (2) interim storage and disposal of TWRS immobilized low-activity tank waste (ILAW). An environmental requirements checklist and narrative was developed to identify the permitting path forward for the HLW interim storage (HLWIS) project (See Appendix B). This permitting plan will follow the permitting logic developed in that checklist.

  20. 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.

  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. Method for selective recovery of PET-usable quantities of [.sup.18 F] fluoride and [.sup.13 N] nitrate/nitrite from a single irradiation of low-enriched [.sup.18 O] water

    DOE Patents [OSTI]

    Ferrieri, Richard A.; Schlyer, David J.; Shea, Colleen

    1995-06-13

    A process for simultaneously producing PET-usable quantities of [.sup.13 N]NH.sub.3 and [.sup.18 F]F.sup.- for radiotracer synthesis is disclosed. The process includes producing [.sup.13 N]NO.sub.2.sup.- /NO.sub.3.sup.- and [.sup.18 F]F.sup.- simultaneously by exposing a low-enriched (20%-30%) [.sup.18 O]H.sub.2 O target to proton irradiation, sequentially isolating the [.sup.13 N]NO.sub.2.sup.- /NO.sub.3.sup.- and [.sup.18 F]F.sup.- from the [.sup.18 O]H.sub.2 O target, and reducing the [.sup.13 N]NO.sub.2.sup.- /NO.sub.3.sup.- to [.sup.13 N]NH.sub.3. The [.sup.13 N]NH.sub.3 and [.sup.18 F]F.sup.- products are then conveyed to a laboratory for radiotracer applications. The process employs an anion exchange resin for isolation of the isotopes from the [.sup.18 O]H.sub.2 O, and sequential elution of [.sup.13 N]NO.sub.2.sup.- /NO.sub.3.sup.- and [ .sup.18 F]F.sup.- fractions. Also the apparatus is disclosed for simultaneously producing PET-usable quantities of [.sup.13 N]NH.sub.3 and [.sup.18 F]F.sup.- from a single irradiation of a single low-enriched [.sup.18 O]H.sub.2 O target.

  3. 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.

  4. 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.

  5. 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.

  6. Superconducting open-gradient magnetic separation for the pretreatment of radioactive or mixed waste vitrification feeds. 1997 annual progress report

    SciTech Connect (OSTI)

    Doctor, R.; Nunez, L.; Cicero-Herman, C.A.; Ritter, J.A.; Landsberger, S.

    1997-01-01

    'Vitrification has been selected as a final waste form technology in the US for long-term storage of high-level radioactive wastes (HLW). However, a foreseeable problem during vitrification in some waste feed streams lies in the presence of elements (e.g., transition metals) in the HLW that may cause instabilities in the final glass product. The formation of spinel compounds, such as Fe{sub 3}O{sub 4} and FeCrO{sub 4}, results in glass phase separation and reduces vitrifier lifetime, and durability of the final waste form. A superconducting open gradient magnetic separation (OGMS) system maybe suitable for the removal of the deleterious transition elements (e.g. Fe, Co, and Ni) and other elements (lanthanides) from vitrification feed streams due to their ferromagnetic or paramagnetic nature. The OGMS systems are designed to deflect and collect paramagnetic minerals as they interact with a magnetic field gradient. This system has the potential to reduce the volume of HLW for vitrification and ensure a stable product. In order to design efficient OGMS and High gradient magnetic separation (HGMS) processes, a fundamental understanding of the physical and chemical properties of the waste feed streams is required. Using HLW simulant and radioactive fly ash and sludge samples from the Savannah River Technology Center, Rocky Flats site, and the Hanford reservation, several techniques were used to characterize and predict the separation capability for a superconducting OGMS system.'

  7. Property:DayQuantity | Open Energy Information

    Open Energy Info (EERE)

    are not known. Acceptable units (and their conversions) are: 1 day,Day,days,Days,DAY,DAYS,d,D 24 hour,hours,Hour,Hours,hr,hrs,HOUR,HOURS,HR,HRS 1440 minute,minutes,Minute,Minutes,m...

  8. Fuel quantity modulation in pilot ignited engines

    DOE Patents [OSTI]

    May, Andrew

    2006-05-16

    An engine system includes a first fuel regulator adapted to control an amount of a first fuel supplied to the engine, a second fuel regulator adapted to control an amount of a second fuel supplied to the engine concurrently with the first fuel being supplied to the engine, and a controller coupled to at least the second fuel regulator. The controller is adapted to determine the amount of the second fuel supplied to the engine in a relationship to the amount of the first fuel supplied to the engine to operate in igniting the first fuel at a specified time in steady state engine operation and adapted to determine the amount of the second fuel supplied to the engine in a manner different from the relationship at steady state engine operation in transient engine operation.

  9. ARM - Lesson Plans: Measuring Quantities of Gas

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

    may otherwise be confusing Materials Each group of students will need the following: Food coloring or ink 1-liter measure 4 containers (approximately 2-liter capacity) Teat...

  10. Method for selective recovery of PET-usable quantities of [{sup 18}F] fluoride and [{sup 13}N] nitrate/nitrite from a single irradiation of low-enriched [{sup 18}O] water

    DOE Patents [OSTI]

    Ferrieri, R.A.; Schlyer, D.J.; Shea, C.

    1995-06-13

    A process for simultaneously producing PET-usable quantities of [{sup 13}N]NH{sub 3} and [{sup 18}F]F{sup {minus}} for radiotracer synthesis is disclosed. The process includes producing [{sup 13}N]NO{sub 2}{sup {minus}}/NO{sub 3}{sup {minus}}and [{sup 18}F]F{sup {minus}} simultaneously by exposing a low-enriched (20%-30%) [{sup 18}O]H{sub 2}O target to proton irradiation, sequentially isolating the [{sup 13}N]NO{sub 2}{sup {minus}}/NO{sub 3}{sup {minus}} and [{sup 18}F]F{sup {minus}} from the [{sup 18}O]H{sub 2}O target, and reducing the [{sup 13}N]NO{sub 2}{sup {minus}}/NO{sub 3}{sup {minus}} to [{sup 13}N]NH{sub 3}. The [{sup 13}N]NH{sub 3} and [{sup 18}F]F{sup {minus}} products are then conveyed to a laboratory for radiotracer applications. The process employs an anion exchange resin for isolation of the isotopes from the [{sup 18}O]H{sub 2}O, and sequential elution of [{sup 13}N]NO{sub 2}{sup {minus}}/NO{sub 3}{sup {minus}} and [{sup 18}F]F{sup {minus}} fractions. Also the apparatus is disclosed for simultaneously producing PET-usable quantities of [{sup 13}N]NH{sub 3} and [{sup 18}F]F{sup {minus}} from a single irradiation of a single low-enriched [{sup 18}O]H{sub 2}O target. 2 figs.

  11. 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.

  12. Investigation of Radiation and Chemical Resistance of Flexible HLW Transfer Hose

    SciTech Connect (OSTI)

    E. Skidmore; Billings, K.; Hubbard, M.

    2010-03-24

    A chemical transfer hose constructed of an EPDM (ethylene-propylene diene monomer) outer covering with a modified cross-linked polyethylene (XLPE) lining was evaluated for use in high level radioactive waste transfer applications. Laboratory analysis involved characterization of the hose liner after irradiation to doses of 50 to 300 Mrad and subsequent exposure to 25% NaOH solution at 93 C for 30 days, simulating 6 months intermittent service. The XLPE liner mechanical and structural properties were characterized at varying dose levels. Burst testing of irradiated hose assemblies was also performed. Literature review and test results suggest that radiation effects below doses of 100 kGy are minimal, with acceptable property changes to 500 kGy. Higher doses may be feasible. At a bounding dose of 2.5 MGy, the burst pressure is reduced to the working pressure (1.38 MPa) at room temperature. Radiation exposure slightly reduces liner tensile strength, with more significant decrease in liner elongation. Subsequent exposure to caustic solutions at elevated temperature slightly increases elongation, suggesting an immersion/hydrolytic effect or possible thermal annealing of radiation damage. This paper summarizes the laboratory results and recommendations for field deployment.

  13. Enhanced Sulfate Management in HLW Glass Formulations VSL12R2540-1 REV 0

    SciTech Connect (OSTI)

    Kruger, A. A.; Pegg, Ian L.; Kot, Wing; Gan, Hao; Matlack, Keith S.

    2012-11-13

    The Low Activity Waste (LAW) tanks that are scheduled to provide the Hanford Tank Waste Treatment and Immobilization Plant (WTP) with waste feeds contain significant amounts of sulfate. The sulfate content in the LAW feeds is sufficiently high that a separate molten sulfate salt phase may form on top of the glass melt during the vitrification process unless suitable glass formulations are employed and sulfate levels are controlled. Since the formation of the salt phase is undesirable from many perspectives, mitigation approaches had to be developed. Considerable progress has been made and reported by the Vitreous State Laboratory (VSL) in enhancing sulfate incorporation into LAW glass melts and developing strategies to manage and mitigate the risks associated with high-sulfate feeds.

  14. Amended Record of Decision for the Idaho High-Level Waste (HLW) and

    Office of Environmental Management (EM)

    Power - Great Northern Transmission Line: Federal Register Notice, Volume 79, No. 222 - Nov. 18, 2014 | Department of Energy No. PP-398 Minnesota Power - Great Northern Transmission Line: Federal Register Notice, Volume 79, No. 222 - Nov. 18, 2014 Amended Application for Presidential Permit OE Docket No. PP-398 Minnesota Power - Great Northern Transmission Line: Federal Register Notice, Volume 79, No. 222 - Nov. 18, 2014 Minnesota Power, Great Northern Transmission Line has submitted an

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

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Atomic Energy Act of 1954 Energy Reorganization Act of 1974 Atomic Energy Defense Act Low-Level Radioactive Waste Policy Amendments Act of 1985 (42 USC 2021b) 42 USC 2021c 42 USC ...

  16. HLW-OVP-97-0068 High Level Waste Management Division High-Level...

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

    ... kiln or secondary combustion chamber for treatment ... Influents - F-Canyon Low Heat Waste (LHW) and High Heat Waste ... IV tanks, which lack internal structures, thereby Page ...

  17. Extended Development Work to Validate a HLW Calcine Waste Form via INL's Cold Crucible Induction Melter

    SciTech Connect (OSTI)

    James A. King; Vince Maio

    2011-09-01

    To accomplish calcine treatment objectives, the Idaho Clean-up Project contractor, CWI, has chosen to immobilize the calcine in a glass-ceramic via the use of a Hot-Isostatic-Press (HIP); a treatment selection formally documented in a 2010 Record of Decision (ROD). Even though the HIP process may prove suitable for the calcine as specified in the ROD and validated in a number of past value engineering sessions, DOE is evaluating back-up treatment methods for the calcine as a result of the technical, schedule, and cost risk associated with the HIPing process. Consequently DOE HQ has requested DOE ID to make INL's bench-scale cold-crucible induction melter (CCIM) available for investigating its viability as a process alternate to calcine treatment. The waste form is the key component of immobilization of radioactive waste. Providing a solid, stable, and durable material that can be easily be stored is the rationale for immobilization of radioactive waste material in glass, ceramic, or glass-ceramics. Ceramic waste forms offer an alternative to traditional borosilicate glass waste forms. Ceramics can usually accommodate higher waste loadings than borosilicate glass, leading to smaller intermediate and long-term storage facilities. Many ceramic phases are known to possess superior chemical durability as compared to borosilicate glass. However, ceramics are generally multiphase systems containing many minor phase that make characterization and prediction of performance within a repository challenging. Additionally, the technologies employed in ceramic manufacture are typically more complex and expensive. Thus, many have proposed using glass-ceramics as compromise between in the more inexpensive, easier to characterize glass waste forms and the more durable ceramic waste forms. Glass-ceramics have several advantages over traditional borosilicate glasses as a waste form. Borosilicate glasses can inadvertently devitrify, leading to a less durable product that could crack during cooling and crystals may be prone to dissolution. By designing a glass-ceramics, the risks of deleterious effects from devitrification are removed. Furthermore, glass-ceramics have higher mechanical strength and impact strengths and possess greater chemical durability as noted above. Glass-ceramics should provide a waste form with the advantages of glass - ease of manufacture - with improved mechanical properties, thermal stability, and chemical durability. This report will cover aspects relevant for the validation of the CCIM use in the production of glass-ceramic waste forms.

  18. Report - Melter Testing of New High Bismuth HLW Formulations VSL-13R2770-1

    SciTech Connect (OSTI)

    Kruger, Albert A.; Pegg, I. L.; Kot, W. K.; Gan, H.; Matlack, K. S.

    2013-11-13

    The primary objective of the work described was to test two glasses formulated for a high bismuth waste stream on the DM100 melter system. Testing was designed to determine processing characteristics and production rates, assess the tendency for foaming, and confirm glass properties. The glass compositions tested were previously developed to maintain high waste loadings and processing rates while suppressing the foaming observed in previous tests

  19. River Protection Project (RPP) Immobilized High Level Waste (HLW) Interim Storage Plan

    SciTech Connect (OSTI)

    BRIGGS, M.G.

    2000-09-22

    This document replaces HNF-1751, Revision 1. It incorporates updates to reflect changes in programmatic direction associated with the vitrification plant contract and associated DOE-ORP guidance. In addition it includes planning associated with failed/used melter and sample handling and disposition work scope. The document also includes format modifications and section numbering update consistent with CH2M HILL Hanford Group, Inc. procedures.

  20. THERMAL ANALYSIS OF WASTE GLASS MELTER FEEDS

    SciTech Connect (OSTI)

    KRUGER AA; HRMA PR; POKORNY R; PIERCE DA

    2011-10-21

    Melter feeds for high-level nuclear waste (HLW) typically contain a large number of constituents that evolve gas on heating, Multiple gas-evolving reactions are both successive and simultaneous, and include the release of chemically bonded water, reactions of nitrates with organics, and reactions of molten salts with solid silica. Consequently, when a sample of a HLW feed is subjected to thermogravimetric analysis (TGA), the rate of change of the sample mass reveals multiple overlapping peaks. In this study, a melter feed, formulated for a simulated high-alumina HLW to be vitrified in the Waste Treatment and Immobilization Plant, currently under construction at the Hanford Site in Washington State, USA, was subjected to TGA. In addition, a modified melter feed was prepared as an all-nitrate version of the baseline feed to test the effect of sucrose addition on the gas-evolving reactions. Activation energies for major reactions were determined using the Kissinger method. The ultimate aim of TGA studies is to obtain a kinetic model of the gas-evolving reactions for use in mathematical modeling of the cold cap as an element of the overall model of the waste-glass melter. In this study, we focused on computing the kinetic parameters of individual reactions without identifying their actual chemistry, The rough provisional model presented is based on the first-order kinetics.

  1. Compliance with Waste Acceptance Criteria of WIPP and NTS for Vitrified Low-Level and TRU Waste Forms

    SciTech Connect (OSTI)

    Harbour, J.R.; Andrews, M.K.

    1998-07-01

    A joint project between the Oak Ridge National Laboratory (ORNL) and the Savannah River Technology Center (SRTC) has been established to evaluate vitrification as an option for the immobilization of waste within ORNL tank farms. This paper presents details of calculations based on current best available analyses of the Oak Ridge Tanks on the limits for waste loadings imposed by the waste acceptance criteria.

  2. Low Temperature Aluminum Dissolution Of Sludge Waste

    SciTech Connect (OSTI)

    Keefer, M.T.; Hamm, B.A.; Pike, J.A. [Washington Savannah River Company, Aiken, SC (United States)

    2008-07-01

    High Level Waste (HLW) at the Savannah River Site (SRS) is currently stored in aging underground storage tanks. This waste is a complex mixture of insoluble solids, referred to as sludge, and soluble salts. Continued long-term storage of these radioactive wastes poses an environmental risk. The sludge is currently being stabilized in the Defense Waste Processing Facility (DWPF) through a vitrification process immobilizing the waste in a borosilicate glass matrix for long-term storage in a federal repository. Without additional treatment, the existing volume of sludge would produce nearly 8000 canisters of vitrified waste. Aluminum compounds, along with other non-radioactive components, represent a significant portion of the sludge mass currently planned for vitrification processing in DWPF. Removing the aluminum from the waste stream reduces the volume of sludge requiring vitrification and improves production rates. Treating the sludge with a concentrated sodium hydroxide (caustic) solution at elevated temperatures (>90 deg. C) to remove aluminum is part of an overall sludge mass reduction effort to reduce the number of vitrified canisters, shorten the life cycle for the HLW system, and reduce the risk associated with the long term storage of radioactive wastes at SRS. A projected reduction of nearly 900 canisters will be achieved by performing aluminum dissolution on six targeted sludge batches; however, a project to develop and install equipment will not be ready for operation until 2013. The associated upgrades necessary to implement a high temperature process in existing facilities are costly and present many technical challenges. Efforts to better understand the characteristics of the sludge mass and dissolution kinetics are warranted to overcome these challenges. Opportunities to further reduce the amount of vitrified waste and increase production rates should also be pursued. Sludge staged in Tank 51 as the next sludge batch for feed to DWPF consisted primarily of radioactive wastes containing a very high aluminum concentration. Based on initial laboratory testing and previous sludge characterization, aluminum in this sludge could be dissolved at low temperature (no more than 65 deg. C) in a concentrated caustic solution. The amount of aluminum predicted to dissolve under these conditions ranged from 25% to 80%. An opportunity existed to remove a significant amount of aluminum prior to vitrification in DWPF and increase the level of understanding of the effects of caustic dissolution of aluminum at lower temperatures. This paper presents the results of a real waste laboratory demonstration and full-scale implementation of a low temperature aluminum dissolution process which should be considered as a viable means to reduce radioactive sludge mass and reduce the amount of waste to be vitrified. (authors)

  3. High Level Waste System Impacts from Small Column Ion Exchange Implementation

    SciTech Connect (OSTI)

    McCabe, D. J.; Hamm, L. L.; Aleman, S. E.; Peeler, D. K.; Herman, C. C.; Edwards, T. B.

    2005-08-18

    The objective of this task is to identify potential waste streams that could be treated with the Small Column Ion Exchange (SCIX) and perform an initial assessment of the impact of doing so on the High-Level Waste (HLW) system. Design of the SCIX system has been performed as a backup technology for decontamination of High-Level Waste (HLW) at the Savannah River Site (SRS). The SCIX consists of three modules which can be placed in risers inside underground HLW storage tanks. The pump and filter module and the ion exchange module are used to filter and decontaminate the aqueous tank wastes for disposition in Saltstone. The ion exchange module contains Crystalline Silicotitanate (CST in its engineered granular form is referred to as IONSIV{reg_sign} IE-911), and is selective for removal of cesium ions. After the IE-911 is loaded with Cs-137, it is removed and the column is refilled with a fresh batch. The grinder module is used to size-reduce the cesium-loaded IE-911 to make it compatible with the sludge vitrification system in the Defense Waste Processing Facility (DWPF). If installed at the SRS, this SCIX would need to operate within the current constraints of the larger HLW storage, retrieval, treatment, and disposal system. Although the equipment has been physically designed to comply with system requirements, there is also a need to identify which waste streams could be treated, how it could be implemented in the tank farms, and when this system could be incorporated into the HLW flowsheet and planning. This document summarizes a preliminary examination of the tentative HLW retrieval plans, facility schedules, decontamination factor targets, and vitrified waste form compatibility, with recommendations for a more detailed study later. The examination was based upon four batches of salt solution from the currently planned disposition pathway to treatment in the SCIX. Because of differences in capabilities between the SRS baseline and SCIX, these four batches were combined into three batches for a total of about 3.2 million gallons of liquid waste. The chemical and radiological composition of these batches was estimated from the SpaceMan Plus{trademark} model using the same data set and assumptions as the baseline plans.

  4. "Table A32. Total Quantity of Purchased Energy Sources by...

    U.S. Energy Information Administration (EIA) Indexed Site

    ... W "," Q "," W "," Q "," W ",0,7,14.5 2911,"Petroleum Refining"," W ",1824,0,0," W ",0," W ",0,6,15.4 30,"Rubber and Miscellaneous Plastics Products",35,3350," W ...

  5. Table A23. Quantity of Purchased Electricity, Steam, and Natural...

    U.S. Energy Information Administration (EIA) Indexed Site

    ..."W","*",0,"W",16,1,10,11.9 3331," Primary Copper","W",0,0,0,"W","W","W",1.4 3334," Primary ...350,0,0,0,"W","W","*",23.5 3331," Primary Copper","W",0,0,0,"*",0,"*",1.3 3334," Primary ...

  6. "Table A22. Total Quantity of Purchased Energy Sources by...

    U.S. Energy Information Administration (EIA) Indexed Site

    ... Fibers",0,0,0,0,0,0,0,0,0,"NF" 2824," Organic Fibers Noncellulosic",0,0,0,0,0,0,0,0,0,"... such combustible energy sources as wood" "waste, hydrogen, or waste oils and tars." " ...

  7. Quantity of Natural Gas Production Associated with Reported Wellhead Value

    U.S. Energy Information Administration (EIA) Indexed Site

    Annual Download Series History Download Series History Definitions, Sources & Notes Definitions, Sources & Notes Show Data By: Data Series Area 2005 2006 2007 2008 2009 2010 View History U.S. 15,425,867 15,981,421 1980-2006 Alabama 285,237 274,176 259,062 246,747 225,666 212,769 1983-2010 Alaska 502,887 494,323 368,344 337,359 397,077 316,546 1983-2010 Arizona 211 588 634 503 695 165 1983-2010 Arkansas 190,533 193,491 269,886 446,551 680,613 936,600 1983-2010 California 274,817 278,933

  8. Table A27. Quantity of Purchased Electricity, Steam, and Natural...

    U.S. Energy Information Administration (EIA) Indexed Site

    Type" " of Supplier, Census Region, and Economic Characteristics of the Establishment," ...,"Utility","Transmission","Other","Row" "Economic Characteristics(a)","Supplier(b)","Suppl...

  9. EM's Indefinite Delivery/Indefinite Quantity Cleanup Contracts...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    http:www.emcbc.doe.govEM%20Nationwide%20IDIQ%20(COP)Awarded%20IDIQ%20Contracts.php Questions about the IDIQ contracts can be referred to the following points of contact...

  10. Low Level Waste Disposition – Quantity and Inventory

    Broader source: Energy.gov [DOE]

    This study has been prepared by the Used Fuel Disposition (UFD) campaign of the Fuel Cycle Research and Development (FCR&D) program. The purpose of this study is to provide an estimate of the...

  11. fiberConnector-Quantities-18Oct2006.xls

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

    ID and OD fiber lengths Parameters Extra Fiber Length (in) 3 OD Frames 108 ID Scint Planes 196 Fibers per Preform 10 Bare Lengths (Ordered, mm) Scratch Col This Length Spare...

  12. fiberConnector-Quantities-18Oct2006.xls

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

    Adapters to Order 7568 8583 ODUs + Ferrules for Scint Assemblies Tracking Prototype Ferrules 3487 Assume straight 20% Tracking Prototype Adapters 1717 Assume straight...

  13. "Table A33. Total Quantity of Purchased Energy Sources by...

    U.S. Energy Information Administration (EIA) Indexed Site

    "Value of Shipments and Receipts" "(million dollars)" " Under 20",1676,121525,4039,11727,661,4834,3503,354,382,5.7 " 20-49",2025,122093,10357,4443,864,7770,8135,513,402,6.4 " 50-99...

  14. Value of Demand Response: Quantities from Production Cost Modeling (Presentation)

    SciTech Connect (OSTI)

    Hummon, M.

    2014-04-01

    Demand response (DR) resources present a potentially important source of grid flexibility particularly on future systems with high penetrations of variable wind and solar power generation. However, managed loads in grid models are limited by data availability and modeling complexity. This presentation focuses on the value of co-optimized DR resources to provide energy and ancillary services in a production cost model. There are significant variations in the availabilities of different types of DR resources, which affect both the operational savings as well as the revenue for each DR resource. The results presented include the system-wide avoided fuel and generator start-up costs as well as the composite revenue for each DR resource by energy and operating reserves. In addition, the revenue is characterized by the capacity, energy, and units of DR enabled.

  15. Table 7.6 Quantity of Purchased Energy Sources, 2010;

    U.S. Energy Information Administration (EIA) Indexed Site

    93 2,673 * * 22 W 0 0 W 325193 Ethyl Alcohol 317 7,359 0 * 251 * 1 0 7 325199 Other ... W W * * W * 0 0 * 325193 Ethyl Alcohol W 72 0 * W * 0 0 0 325199 Other Basic ...

  16. Product/Process (P/P) Models For The Defense Waste Processing Facility (DWPF): Model Ranges And Validation Ranges For Future Processing

    SciTech Connect (OSTI)

    Jantzen, C.; Edwards, T.

    2015-09-25

    Radioactive high level waste (HLW) at the Savannah River Site (SRS) has successfully been vitrified into borosilicate glass in the Defense Waste Processing Facility (DWPF) since 1996. Vitrification requires stringent product/process (P/P) constraints since the glass cannot be reworked once it is poured into ten foot tall by two foot diameter canisters. A unique “feed forward” statistical process control (SPC) was developed for this control rather than statistical quality control (SQC). In SPC, the feed composition to the DWPF melter is controlled prior to vitrification. In SQC, the glass product would be sampled after it is vitrified. Individual glass property-composition models form the basis for the “feed forward” SPC. The models transform constraints on the melt and glass properties into constraints on the feed composition going to the melter in order to guarantee, at the 95% confidence level, that the feed will be processable and that the durability of the resulting waste form will be acceptable to a geologic repository.

  17. HIGH LEVEL WASTE SLUDGE BATCH 4 VARIABILITY STUDY

    SciTech Connect (OSTI)

    Fox, K; Tommy Edwards, T; David Peeler, D; David Best, D; Irene Reamer, I; Phyllis Workman, P

    2006-10-02

    The Defense Waste Processing Facility (DWPF) is preparing for vitrification of High Level Waste (HLW) Sludge Batch 4 (SB4) in early FY2007. To support this process, the Savannah River National Laboratory (SRNL) has provided a recommendation to utilize Frit 503 for vitrifying this sludge batch, based on the composition projection provided by the Liquid Waste Organization on June 22, 2006. Frit 418 was also recommended for possible use during the transition from SB3 to SB4. A critical step in the SB4 qualification process is to demonstrate the applicability of the durability models, which are used as part of the DWPF's process control strategy, to the glass system of interest via a variability study. A variability study is an experimentally-driven assessment of the predictability and acceptability of the quality of the vitrified waste product that is anticipated from the processing of a sludge batch. At the DWPF, the durability of the vitrified waste product is not directly measured. Instead, the durability is predicted using a set of models that relate the Product Consistency Test (PCT) response of a glass to the chemical composition of that glass. In addition, a glass sample is taken during the processing of that sludge batch, the sample is transmitted to SRNL, and the durability is measured to confirm acceptance. The objective of a variability study is to demonstrate that these models are applicable to the glass composition region anticipated during the processing of the sludge batch - in this case the Frit 503 - SB4 compositional region. The success of this demonstration allows the DWPF to confidently rely on the predictions of the durability/composition models as they are used in the control of the DWPF process.

  18. HLW flowsheet material balance for DWPF rad operation with Tank 51 sludge and ITP Cycle 1 precipitate

    SciTech Connect (OSTI)

    Choi, A.S.

    1995-04-19

    This document presents the details of the Savannah River Plant Flowsheet for the Rad Operation with Tank Sludge and ITP Cycle 1 Precipitate. Topics discussed include: material balance; radiolysis chemistry of tank precipitates; algorithm for ESP washing; chemistry of hydrogen and ammonia generation in CPC; batch sizes for processing feed; and total throughput of a streams during one cycle of operation.

  19. Amended Record of Decision for the Idaho High-Level Waste (HLW) and Facilities Disposition Final Environmental Impact Statement

    Office of Environmental Management (EM)

    Additional Public Scoping Meetings, and Notice of Floodplains and Wetlands Involvement for the Northern Pass Project: Federal Register Notice VOlume 78, No. 173 - September 6, 2013 | Department of Energy Amended Notice of Intent To Modify the Scope of the EIS and Conduct Additional Public Scoping Meetings, and Notice of Floodplains and Wetlands Involvement for the Northern Pass Project: Federal Register Notice VOlume 78, No. 173 - September 6, 2013 Amended Notice of Intent To Modify the

  20. Crystal Settling, Redox, and High Temperature Properties of ORP HLW and LAW Glasses, VSL-09R1510-1

    SciTech Connect (OSTI)

    Kruger, A. A.; Gan, H.; Viragh, C.; Mckeown, D. A.; Muller, I. S.; Cecil, R.; Kot, W. K.; Joseph, I.; Wang, C.; Pegg, I. L.; Chaudhuri, M.; Zhao, W.; Feng, Z.

    2015-06-08

    This report describes the results of testing specified by the Test Plans (VSL-08T1520-1 Rev 0 and VSL-08T1510-1 Rev 0). The work was performed in compliance with the quality assurance requirements specified in the Test Plans. Results required by the Test Plans are reported. The test results and this report have been reviewed for correctness, technical adequacy, completeness, and accuracy.

  1. Evaluation of Shortline Railroads & SNF/HLW Rail Shipment Inspections Tasked for the Transportation of Spent Nuclear Fuel

    Office of Environmental Management (EM)

    | Department of Energy Evaluation of Production of Oil & Gas From Oil Shale in the Piceance Basin Evaluation of Production of Oil & Gas From Oil Shale in the Piceance Basin The purpose of this paper is to provide the public and policy makers accurate estimates of energy efficiencies, water requirements, water availability, and CO2 emissions associated with the development of the 60 percent portion of the Piceance Basin where economic potential is the greatest, and where environmental

  2. Hanford immobilized LAW product acceptance: Initial Tanks Focus Area testing data package

    SciTech Connect (OSTI)

    JD Vienna; A Jiricka; BP McGrail; BM Jorgensen; DE Smith; BR Allen; JC Marra; DK Peeler; KG Brown; IA Reamer; WL Ebert

    2000-03-08

    The Hanford Site's mission has been to produce nuclear materials for the US Department of Energy (DOE) and its predecessors. A large inventory of radioactive and mixed waste, largely generated during plutonium production, exists in 177 underground single- and double-shell tanks. These wastes are to be retrieved and separated into low-activity waste (LAW) and high-level waste (HLW) fractions. The total volume of LAW requiring immobilization will include the LAW separated from the tank waste, as well as new wastes generated by the retrieval, pretreatment, and immobilization processes. Per the Tri-Party Agreement (1994), both the LAW and HLW will be vitrified. It has been estimated that vitrification of the LAW waste will result in over 500,000 metric tons or 200,000 m{sup 3} of immobilized LAW (ILAW) glass. The ILAW glass is to be disposed of onsite in a near-surface burial facility. It must be demonstrated that the disposal system will adequately retain the radionuclides and prevent contamination of the surrounding environment. This report describes a study of the impacts of systematic glass-composition variation on the responses from accelerated laboratory corrosion tests of representative LAW glasses. A combination of two tests, the product consistency test and vapor-hydration test, is being used to give indictations of the relative rate at which a glass could be expected to corrode in the burial scenario.

  3. Generalized Test Plan for the Vitrification of Simulated High-Level -Waste Calcine in the Idaho National Laboratory‘s Bench -Scale Cold Crucible Induction Melter

    SciTech Connect (OSTI)

    Vince Maio

    2011-08-01

    This Preliminary Idaho National Laboratory (INL) Test Plan outlines the chronological steps required to initially evaluate the validity of vitrifying INL surrogate (cold) High-Level-Waste (HLW) solid particulate calcine in INL's Cold Crucible Induction Melter (CCIM). Its documentation and publication satisfies interim milestone WP-413-INL-01 of the DOE-EM (via the Office of River Protection) sponsored work package, WP 4.1.3, entitled 'Improved Vitrification' The primary goal of the proposed CCIM testing is to initiate efforts to identify an efficient and effective back-up and risk adverse technology for treating the actual HLW calcine stored at the INL. The calcine's treatment must be completed by 2035 as dictated by a State of Idaho Consent Order. A final report on this surrogate/calcine test in the CCIM will be issued in May 2012-pending next fiscal year funding In particular the plan provides; (1) distinct test objectives, (2) a description of the purpose and scope of planned university contracted pre-screening tests required to optimize the CCIM glass/surrogate calcine formulation, (3) a listing of necessary CCIM equipment modifications and corresponding work control document changes necessary to feed a solid particulate to the CCIM, (4) a description of the class of calcine that will be represented by the surrogate, and (5) a tentative tabulation of the anticipated CCIM testing conditions, testing parameters, sampling requirements and analytical tests. Key FY -11 milestones associated with this CCIM testing effort are also provided. The CCIM test run is scheduled to be conducted in February of 2012 and will involve testing with a surrogate HLW calcine representative of only 13% of the 4,000 m3 of 'hot' calcine residing in 6 INL Bin Sets. The remaining classes of calcine will have to be eventually tested in the CCIM if an operational scale CCIM is to be a feasible option for the actual INL HLW calcine. This remaining calcine's make-up is HLW containing relatively high concentrations of zirconium and aluminum, representative of the cladding material of the reprocessed fuel that generated the calcine. A separate study to define the CCIM testing needs of these other calcine classifications in currently being prepared under a separate work package (WP-0) and will be provided as a milestone report at the end of this fiscal year.

  4. Vitrification of ion exchange materials. Innovative technology summary report

    SciTech Connect (OSTI)

    Not Available

    1999-07-01

    Ion exchange is a process that safely and efficiently removes radionuclides from tank waste. Cesium and strontium account for a large portion of the radioactivity in waste streams from US Department of Energy (DOE) weapons production. Crystalline silicotitanate (CST) is an inorganic sorbent that strongly binds cesium, strontium, and several other radionuclides. Developed jointly by Sandia National Laboratory and Texas A and M University, CST was commercialized through a cooperative research and development agreement with an industrial partner. Both an engineered (mesh pellets) and powdered forms are commercially available. Cesium removal is a baseline in HLW treatment processing. CST is very effective at removing cesium from HLW streams and is being considered for adoption at several sites. However, CST is nonregenerable, and it presents a significant secondary waste problem. Treatment options include vitrification of the CST, vitrification of the CST coupled with HLW, direct disposal, and low-temperature processes such as grouting. The work presented in this report demonstrates that it is effective to immobilize CST using a baseline technology such as vitrification. Vitrification produces a durable waste form. CST vitrification was not demonstrated before 1996. In FY97, acceptable glass formulations were developed using cesium-loaded CST obtained from treating supernatants from Oak Ridge Reservation (ORR) tanks, and the CST was vitrified in a research melter at the Savannah River Technology Center (SRTC). In FY98, SRS decided to reevaluate the use of in-tank precipitation using tetraphenylborate to remove cesium from tank supernatant and to consider other options for cesium removal, including CST. Hanford and Idaho National Engineering and Environmental Laboratory also require radionuclide removal in their baseline flowsheets.

  5. 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.

  6. HIGH-LEVEL WASTE FEED CERTIFICATION IN HANFORD DOUBLE-SHELL TANKS

    SciTech Connect (OSTI)

    THIEN MG; WELLS BE; ADAMSON DJ

    2010-01-14

    The ability to effectively mix, sample, certify, and deliver consistent batches of High Level Waste (HLW) feed from the Hanford Double Shell Tanks (DST) to the Waste Treatment and Immobilization Plant (WTP) presents a significant mission risk with potential to impact mission length and the quantity of HLW glass produced. DOE's River Protection Project (RPP) mission modeling and WTP facility modeling assume that individual 3785 cubic meter (l million gallon) HLW feed tanks are homogenously mixed, representatively sampled, and consistently delivered to the WTP. It has been demonstrated that homogenous mixing ofHLW sludge in Hanford DSTs is not likely achievable with the baseline design thereby causing representative sampling and consistent feed delivery to be more difficult. Inconsistent feed to the WTP could cause additional batch-to-batch operational adjustments that reduce operating efficiency and have the potential to increase the overall mission length. The Hanford mixing and sampling demonstration program will identify DST mixing performance capability, will evaluate representative sampling techniques, and will estimate feed batch consistency. An evaluation of demonstration program results will identify potential mission improvement considerations that will help ensure successful mission completion. This paper will discuss the history, progress, and future activities that will define and mitigate the mission risk.

  7. Homogeneity testing and quantitative analysis of manganese (Mn) in vitrified Mn-doped glasses by laser-induced breakdown spectroscopy (LIBS)

    SciTech Connect (OSTI)

    Unnikrishnan, V. K.; Nayak, Rajesh; Kartha, V. B.; Santhosh, C. E-mail: unnikrishnan.vk@manipal.edu; Sonavane, M. S.; Yeotikar, R. G.; Shah, M. L.; Gupta, G. P.; Suri, B. M.

    2014-09-15

    Laser-induced breakdown spectroscopy (LIBS), an atomic emission spectroscopy method, has rapidly grown as one of the best elemental analysis techniques over the past two decades. Homogeneity testing and quantitative analysis of manganese (Mn) in manganese-doped glasses have been carried out using an optimized LIBS system employing a nanosecond ultraviolet Nd:YAG laser as the source of excitation. The glass samples have been prepared using conventional vitrification methods. The laser pulse irradiance on the surface of the glass samples placed in air at atmospheric pressure was about 1.710{sup 9} W/cm{sup 2}. The spatially integrated plasma emission was collected and imaged on to the spectrograph slit using an optical-fiber-based collection system. Homogeneity was checked by recording LIBS spectra from different sites on the sample surface and analyzing the elemental emission intensities for concentration determination. Validation of the observed LIBS results was done by comparison with scanning electron microscope- energy dispersive X-ray spectroscopy (SEM-EDX) surface elemental mapping. The analytical performance of the LIBS system has been evaluated through the correlation of the LIBS determined concentrations of Mn with its certified values. The results are found to be in very good agreement with the certified concentrations.

  8. 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.

  9. 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.

  10. Data quality objectives for TWRS privatization phase 1: confirm tank T is an appropriate feed source for high-level waste feed batch X

    SciTech Connect (OSTI)

    NGUYEN, D.M.

    1999-06-01

    The U.S. Department of Energy-Richland Operations Office (DOE-RL) has initiated Phase 1 of a two-phase privatization strategy for treatment and immobilization of high-level waste (HLW) that is currently managed by the Hanford Tank Waste Remediation System (TWRS) Project. In this strategy, DOE will purchase services from a contractor-owned and operated facility under a fixed price. The Phase 1 TWRS privatization contract requires that the Project Hanford Management Contract (PHMC) contractors, on behalf of DOE, deliver HLW feed in specified quantities and composition to the Privatization Contractor in a timely manner (DOE-RL 1996). Additional requirements are imposed by the interface control document (ICD) for HLW feed (PHMC 1997). In response to these requirements, the Tank Waste Remediation System Operation and Utilization Plan (TWRSO and UP) (Kirkbride et al. 1997) was prepared by the PHMC. The TWRSO and UP, as updated by the Readiness-To-Proceed (RTP) deliverable (Payne et al. 1998), establishes the baseline operating scenario for the delivery of HLW feed to the Privatization Contractor. The scenario specifies tanks from which HLW will be provided for each feed batch, the operational activities needed to prepare and deliver each batch, and the timing of these activities. The operating scenario was developed based on current knowledge of waste composition and chemistry, waste transfer methods, and operating constraints such as tank farm logistics and availability of tank space. A project master baseline schedule (PMBS) has been developed to implement the operating scenario. The PMBS also includes activities aimed at reducing programmatic risks. One of the activities, ''Confirm Tank TI is Acceptable for Feed,'' was identified to verify the basis used to develop the scenario Additional data on waste quantity, physical and chemical characteristics, and transfer properties will be needed to support this activity. This document describes the data quality objective (DQO) process undertaken to assure appropriate data will be collected to support the activity, ''Confirm Tank T is Acceptable for HLW Feed.'' The DQO process was implemented in accordance with the TWRS DQO process (Banning 1997) with some modifications to accommodate project or tank-specific requirements and constraints.

  11. 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.

  12. Ruthenium Behavior at Phase Separation of Borosilicate Glass-12259

    SciTech Connect (OSTI)

    Enokida, Youichi [Graduate School of Engineering, Nagoya University, Nagoya, 463-8603 (Japan); Sawada, Kayo [EcoTopia Science Institute, Nagoya University, Nagoya, 463-8603 (Japan)

    2012-07-01

    The Rokkasho reprocessing plant (RRP) located in Aomori, Japan, vitrifies high level waste (HLW) into a borosilicate glass. The HLW is generated from the reprocessing of spent fuel and contains ruthenium (Ru) and other platinum group metals (PGMs). Based on the recent consequences after a huge earthquake that occurred in Japan, a hypothetical blackout was postulated for the RRP to address additional safety analysis requirements. During a prolonged blackout, the borosilicate glass could phase separate due to cooling of the glass in the melter. The Ru present in the glass matrix could migrate into separate phases and impact the durability of the borosilicate glass. The durability of the glass is important for quality assurance and performance assessment of the vitrified HLW. A fundamental study was performed at an independent university to understand the impact of a prolonged blackout. Simulated HLW glasses were prepared for the RRP, and the Ru behavior in phase separated glasses was studied. The simulated HLW glasses contained nonradioactive elements and PGMs. The glass compositions were then altered to enhance the formation of the phase-separated glasses when subjected to thermal treatment at 700 deg. C for 24 hours. The synthesized simulated glasses contained 1.1 % Ru by weight as ruthenium dioxide (RuO{sub 2}). A portion of the RuO{sub 2} formed needle-shaped crystals in the glass specimens. After the thermal treatment, the glass specimen had separated into two phases. One of the two phases was a B{sub 2}O{sub 3} rich phase, and the other phase was a SiO{sub 2} rich phase. The majority of the chemical species in the B{sub 2}O{sub 3} rich phase was leached away with the Material Characterization Center-3 (MCC-3) protocol standardized by the Pacific Northwest National Laboratory using an aqueous low-concentrated nitric acid solution, but the leaching of the Ru fraction was very limited; less than 1% of the original Ru content. The Ru leaching was much less than those of the other elements, and the needle-shaped crystals of RuO{sub 2} were observed in the B{sub 2}O{sub 3} rich phase in the specimen after the leaching test. Another experiment was performed using another glass specimen which had been prepared with the same frits, but used reagent RuO{sub 2} of granular shape at lower content (0.0073% by weight as RuO{sub 2}). The leached fractions of elements for the latter specimen increased to almost the same fraction (more than 10% of the original Ru content) as observed for boron and sodium, when the phase separated glass was leached using the MMC-3 protocol with non-acidic de-ionized water. Based on the results of this study, it was concluded that needle-shaped RuO{sub 2} crystals are contained in the B{sub 2}O{sub 3}-rich phase after phase separation of the borosilicate glass after a hypothetical blackout. The leaching fraction for the needle-shaped RuO{sub 2} present in the phase separated glass is much lower than those for boron or sodium. Ruthenium behavior has been studied for a hypothetical loss of cooling in the liquid fed ceramic melter for high level waste by taking into account the phase separation of borosilicate glass. The needle-shaped crystal of ruthenium dioxide after bi-nodal-type phase separation of the borosilicate glass at 700 deg. C migrated into the B{sub 2}O{sub 3} rich phase, but remained without dissolution by an acidic aqueous solution. Additionally, granular ruthenium dioxide can be a morphological form of ruthenium after bimodal-type phase separation of the vitrified high level waste with borosilicate glass media. After the phase separation of the borosilicate glass, the crystal shape of the ruthenium dioxide is either needle-shaped or granular, and the leachable fraction of ruthenium is relatively much lower than those of major components (boron and sodium) in the vitrified borosilicate glass. The fraction of leached ruthenium increased to almost the same fraction as observed for boron and sodium when the phase-separated glass was leached with ultrapure water. (authors)

  13. Depleted uranium: A DOE management guide

    SciTech Connect (OSTI)

    1995-10-01

    The U.S. Department of Energy (DOE) has a management challenge and financial liability in the form of 50,000 cylinders containing 555,000 metric tons of depleted uranium hexafluoride (UF{sub 6}) that are stored at the gaseous diffusion plants. The annual storage and maintenance cost is approximately $10 million. This report summarizes several studies undertaken by the DOE Office of Technology Development (OTD) to evaluate options for long-term depleted uranium management. Based on studies conducted to date, the most likely use of the depleted uranium is for shielding of spent nuclear fuel (SNF) or vitrified high-level waste (HLW) containers. The alternative to finding a use for the depleted uranium is disposal as a radioactive waste. Estimated disposal costs, utilizing existing technologies, range between $3.8 and $11.3 billion, depending on factors such as applicability of the Resource Conservation and Recovery Act (RCRA) and the location of the disposal site. The cost of recycling the depleted uranium in a concrete based shielding in SNF/HLW containers, although substantial, is comparable to or less than the cost of disposal. Consequently, the case can be made that if DOE invests in developing depleted uranium shielded containers instead of disposal, a long-term solution to the UF{sub 6} problem is attained at comparable or lower cost than disposal as a waste. Two concepts for depleted uranium storage casks were considered in these studies. The first is based on standard fabrication concepts previously developed for depleted uranium metal. The second converts the UF{sub 6} to an oxide aggregate that is used in concrete to make dry storage casks.

  14. NEXT GENERATION MELTER(S) FOR VITRIFICATION OF HANFORD WASTE STATUS AND DIRECTION

    SciTech Connect (OSTI)

    RAMSEY WG; GRAY MF; CALMUS RB; EDGE JA; GARRETT BG

    2011-01-13

    Vitrification technology has been selected to treat high-level waste (HLW) at the Hanford Site, the West Valley Demonstration Project and the Savannah River Site (SRS), and low activity waste (LAW) at Hanford. In addition, it may potentially be applied to other defense waste streams such as sodium bearing tank waste or calcine. Joule-heated melters (already in service at SRS) will initially be used at the Hanford Site's Waste Treatment and Immobilization Plant (WTP) to vitrify tank waste fractions. The glass waste content and melt/production rates at WTP are limited by the current melter technology. Significant reductions in glass volumes and mission life are only possible with advancements in melter technology coupled with new glass formulations. The Next Generation Melter (NGM) program has been established by the U.S. Department of Energy's (DOE's), Environmental Management Office of Waste Processing (EM-31) to develop melters with greater production capacity (absolute glass throughput rate) and the ability to process melts with higher waste fractions. Advanced systems based on Joule-Heated Ceramic Melter (JHCM) and Cold Crucible Induction Melter (CCIM) technologies will be evaluated for HLW and LAW processing. Washington River Protection Solutions (WRPS), DOE's tank waste contractor, is developing and evaluating these systems in cooperation with EM-31, national and university laboratories, and corporate partners. A primary NGM program goal is to develop the systems (and associated flowsheets) to Technology Readiness Level 6 by 2016. Design and testing are being performed to optimize waste glass process envelopes with melter and balance of plant requirements. A structured decision analysis program will be utilized to assess the performance of the competing melter technologies. Criteria selected for the decision analysis program will include physical process operations, melter performance, system compatibility and other parameters.

  15. FINAL REPORT DURAMELTER 100 HLW SIMULANT VALIDATION TESTS WITH C-106/AY-102 FEEDS VSL-05R5710-1 REV 0 6/2/05

    SciTech Connect (OSTI)

    KRUGER AA; MATLACK KS; GONG W; PEGG IL

    2011-12-29

    The principal objectives of the DM100 tests were to determine the processing characteristics of several C-106/AY102 feeds derived from simulants prepared by different methods, which result in different physical characteristics of the feed. The VSL simulant used in a previous test was prepared by the direct hydroxide method, which was the method used for feed preparation in the bulk of previous VSL melter testing. The NOAH Technologies Corporation modified-rheology simulant was prepared to the same composition as the VSL simulant using a method that resulted in rheological properties closer to those of certain actual waste samples. The SIPP simulant was produced by processing a co-precipitated waste simulant through a non-radioactive pilot scale semi-integrated pretreatment facility. The general intent of these tests was to provide a basis for determining whether the variations in rheology or other feed physical characteristics arising from the different methods of simulant preparation have significant effects on the processing characteristics of the feed in the melter. Completion of the test objectives is detailed in a table.

  16. 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''.

  17. ACTUAL-WASTE TESTS OF ENHANCED CHEMICAL CLEANING FOR RETRIEVAL OF SRS HLW SLUDGE TANK HEELS AND DECOMPOSITION OF OXALIC ACID

    SciTech Connect (OSTI)

    Martino, C.; King, W.; Ketusky, E.

    2012-01-12

    Savannah River National Laboratory conducted a series of tests on the Enhanced Chemical Cleaning (ECC) process using actual Savannah River Site waste material from Tanks 5F and 12H. Testing involved sludge dissolution with 2 wt% oxalic acid, the decomposition of the oxalates by ozonolysis (with and without the aid of ultraviolet light), the evaporation of water from the product, and tracking the concentrations of key components throughout the process. During ECC actual waste testing, the process was successful in decomposing oxalate to below the target levels without causing substantial physical or chemical changes in the product sludge.

  18. 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.

  19. Decontamination and decommissioning of the Chemical Process Cell (CPC): Topical report for the period January 1985-March 1987

    SciTech Connect (OSTI)

    Meigs, R. A.

    1987-07-01

    To support interim storage of vitrified High-Level Waste (HLW) at the West Valley Demonstration Project, the shielded, remotely operated Chemical Process Cell (CPC) was decommissioned and decontaminated. All equipment was removed, packaged and stored for future size reduction and decontamination. Floor debris was sampled, characterized, and vacuumed into remotely handled containers. The cell walls, ceiling, and floor were decontaminated. Three 20 Mg (22.5 ton) concrete neutron absorber cores were cut with a high-pressure water/abrasive jet cutting system and packaged for disposal. All operations were performed remotely using two overhead bridge cranes which included two 1.8 Mg (2 ton) hoists, one 14.5 Mg (16 ton) hoist, and an electromechanical manipulator or an industrial robot mounted on a mobile platform. Initial general area dose rates in the cell ranged from 1 to 50 R/h. Target levels of less than 10 mR/h general area readings were established before decontamination and decommissioning was initiated; general area dose rates between 200 mR/h and 1200 mR/h were obtained at the completion of the decontamination work. 4 refs., 11 figs., 8 tabs.

  20. Preliminary ILAW Formulation Algorithm Description, 24590 LAW RPT-RT-04-0003, Rev. 1

    SciTech Connect (OSTI)

    Kruger, Albert A.; Kim, Dong-Sang; Vienna, John D.

    2013-12-03

    The U.S. Department of Energy (DOE), Office of River Protection (ORP), has contracted with Bechtel National, Inc. (BNI) to design, construct, and commission the Hanford Tank Waste Treatment and Immobilization Plant (WTP) at the Hanford Site (DOE 2000). This plant is designed to operate for 40 years and treat roughly 50 million gallons of mixed hazardous high-level waste (HLW) stored in 177 underground tanks at the Hanford Site. The process involves separating the hight-level and low-activity waste (LAW) fractions through filtration, leaching, Cs ion exchange, and precipitation. Each fraction will be separately vitrified into borosilicate waste glass. This report documents the initial algorithm for use by Hanford WTP in batching LAW and glass-forming chemicals (GFCs) in the LAW melter feed preparation vessel (MFPV). Algorithm inputs include the chemical analyses of the pretreated LAW in the concentrate receipt vessel (CRV), the volume of the MFPV heel, and the compositions of individual GFCs. In addition to these inputs, uncertainties in the LAW composition and processing parameters are included in the algorithm.

  1. Evaluation of Flygt Propeller Xixers for Double Shell Tank (DST) High Level Waste Auxiliary Solids Mobilization

    SciTech Connect (OSTI)

    PACQUET, E.A.

    2000-07-20

    The River Protection Project (RPP) is planning to retrieve radioactive waste from the single-shell tanks (SST) and double-shell tanks (DST) underground at the Hanford Site. This waste will then be transferred to a waste treatment plant to be immobilized (vitrified) in a stable glass form. Over the years, the waste solids in many of the tanks have settled to form a layer of sludge at the bottom. The thickness of the sludge layer varies from tank to tank, from no sludge or a few inches of sludge to about 15 ft of sludge. The purpose of this technology and engineering case study is to evaluate the Flygt{trademark} submersible propeller mixer as a potential technology for auxiliary mobilization of DST HLW solids. Considering the usage and development to date by other sites in the development of this technology, this study also has the objective of expanding the knowledge base of the Flygt{trademark} mixer concept with the broader perspective of Hanford Site tank waste retrieval. More specifically, the objectives of this study delineated from the work plan are described.

  2. Summary Of Cold Crucible Vitrification Tests Results With Savannah River Site High Level Waste Surrogates

    SciTech Connect (OSTI)

    Stefanovsky, Sergey; Marra, James; Lebedev, Vladimir

    2014-01-13

    The cold crucible inductive melting (CCIM) technology successfully applied for vitrification of low- and intermediate-level waste (LILW) at SIA Radon, Russia, was tested to be implemented for vitrification of high-level waste (HLW) stored at Savannah River Site, USA. Mixtures of Sludge Batch 2 (SB2) and 4 (SB4) waste surrogates and borosilicate frits as slurries were vitrified in bench- (236 mm inner diameter) and full-scale (418 mm inner diameter) cold crucibles. Various process conditions were tested and major process variables were determined. Melts were poured into 10L canisters and cooled to room temperature in air or in heat-insulated boxes by a regime similar to Canister Centerline Cooling (CCC) used at DWPF. The products with waste loading from ~40 to ~65 wt.% were investigated in details. The products contained 40 to 55 wt.% waste oxides were predominantly amorphous; at higher waste loadings (WL) spinel structure phases and nepheline were present. Normalized release values for Li, B, Na, and Si determined by PCT procedure remain lower than those from EA glass at waste loadings of up to 60 wt.%.

  3. ITEM NO. SUPPLIES/SERVICES QUANTITY UNIT UNIT PRICE AMOUNT NAME OF OFFEROR OR CONTRACTOR

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

    IT NNSA Demonstrates its Commitment to Small Business The National Nuclear Security Administration (NNSA) recently selected three small businesses for its new Information Technology (IT) Infrastructure and Cyber Security Support Blanket Purchase Agreement. The contract covers a wide spectrum of IT and Cyber Security support for NNSA's Office of... Information Security Information security deals with requirements for the protection and control of information and matter required to be classified

  4. ITEM NO. SUPPLIES/SERVICES QUANTITY UNIT UNIT PRICE AMOUNT NAME OF OFFEROR OR CONTRACTOR

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

    13 CONTINUATION SHEET REFERENCE NO. OF DOCUMENT BEING CONTINUED PAGE OF COMPUTER SCIENCES CORPORATION (A) (B) (C) (D) (E) (F) DE-AC06-04RL14383/144 $112,970,046.65 to $113,586,970.48. All designated funding on Financial Plan Reports #12 and 13 are for work within scope of the contract and incrementally funds a portion of the work scope for FY 2011 Occupational Medical Services. d. Section J.6 - List of Applicable Directives, is revised to remove the following directive from the contract: DOE O

  5. Report on the treatability study for inerting small quantities of radioactive explosives and explosive components

    SciTech Connect (OSTI)

    Loyola, V.M.; Reber, S.D.

    1996-02-01

    As a result of Sandia`s radiation hardening testing on a variety of its explosive components, radioactive waste streams were generated and have to be disposed of as radioactive waste. Due to the combined hazards of explosives and radioactivity, Sandia`s Radioactive and Mixed Waste Management organization did not have a mechanism for disposal of these waste streams. This report documents the study done to provide a method for the removal of the explosive hazard from those waste streams. The report includes the design of the equipment used, procedures followed, results from waste stream analog tests and the results from the actual explosive inerting tests on radioactive samples. As a result of the inerting treatment, the waste streams were rendered non-explosive and, thus, manageable through normal radioactive waste disposal channels.

  6. Method and apparatus for dispensing small quantities of mercury from evacuated and sealed glass capsules

    DOE Patents [OSTI]

    Grossman, M.W.; George, W.A.; Pai, R.Y.

    1985-08-13

    A technique is disclosed for opening an evacuated and sealed glass capsule containing a material that is to be dispensed which has a relatively high vapor pressure such as mercury. The capsule is typically disposed in a discharge tube envelope. The technique involves the use of a first light source imaged along the capsule and a second light source imaged across the capsule substantially transversely to the imaging of the first light source. Means are provided for constraining a segment of the capsule along its length with the constraining means being positioned to correspond with the imaging of the second light source. These light sources are preferably incandescent projection lamps. The constraining means is preferably a multiple looped wire support. 6 figs.

  7. Method and apparatus for dispensing small quantities of mercury from evacuated and sealed glass capsules

    DOE Patents [OSTI]

    Grossman, Mark W.; George, William A.; Pai, Robert Y.

    1985-01-01

    A technique for opening an evacuated and sealed glass capsule containing a material that is to be dispensed which has a relatively high vapor pressure such as mercury. The capsule is typically disposed in a discharge tube envelope. The technique involves the use of a first light source imaged along the capsule and a second light source imaged across the capsule substantially transversely to the imaging of the first light source. Means are provided for constraining a segment of the capsule along its length with the constraining means being positioned to correspond with the imaging of the second light source. These light sources are preferably incandescent projection lamps. The constraining means is preferably a multiple looped wire support.

  8. Table A21. Quantity of Electricity Sold to Utility and Nonutility...

    U.S. Energy Information Administration (EIA) Indexed Site

    Region and Economic Characteristics of the Establishment, 1991" " (Estimates in Million Kilowatthours)" ,,,,"RSE" " "," ","Utility ","Nonutility","Row" "Economic ...

  9. Quantity, quality, and availability of waste heat from United States thermal power generation

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Gingerich, Daniel B; Mauter, Meagan S

    2015-06-10

    Secondary application of unconverted heat produced during electric power generation has the potential to improve the life-cycle fuel efficiency of the electric power industry and the sectors it serves. This work quantifies the residual heat (also known as waste heat) generated by U.S. thermal power plants and assesses the intermittency and transport issues that must be considered when planning to utilize this heat. Combining Energy Information Administration plant-level data with literature-reported process efficiency data, we develop estimates of the unconverted heat flux from individual U.S. thermal power plants in 2012. Together these power plants discharged an estimated 18.9 billion GJthmoreof residual heat in 2012, 4% of which was discharged at temperatures greater than 90 C. We also characterize the temperature, spatial distribution, and temporal availability of this residual heat at the plant level and model the implications for the technical and economic feasibility of its end use. Increased implementation of flue gas desulfurization technologies at coal-fired facilities and the higher quality heat generated in the exhaust of natural gas fuel cycles are expected to increase the availability of residual heat generated by 10.6% in 2040.less

  10. Microsoft PowerPoint - Reporting Small Quantities to NMMSS_Brian...

    National Nuclear Security Administration (NNSA)

    Reportable Units DOE-owned Other Materials: - Enriched Lithium Kilograms Li - Deuterium 0.1 kg D 2 - Pu-242 Grams Pu, Grams Pu-242 (>20% Pu-242 by weight) - Americium-241 Grams ...

  11. Source Term Analysis for the Waste Isolation Pilot Plant (WIPP) Release Quantity

    Broader source: Energy.gov [DOE]

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

  12. Contract Demand Quantity (CDQ) Close-Out of Comments - June 17...

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

    determination BPA followed the steps laid out in section 5.3.5.2 of the Tiered Rates Methodology in establishing the customers' CDQ amounts. BPA held public workshops and conducted...

  13. Pairing correlations and thermodynamical quantities in {sup 96,97}Mo

    SciTech Connect (OSTI)

    Kargar, Z.

    2007-06-15

    The nuclear level densities of {sup 96,97}Mo are calculated in the framework of superconducting theory. The parameters of nuclear level density are so chosen that the saddle point conditions are satisfied and the best fit to the experimental data yields. Then, using these parameters the energy, the entropy and the spin cut-off factor are calculated as a function of temperature. The curves show structures, reflecting the phase transition from a correlated to an uncorrelated phase. The critical temperature for quenching of pairing correlations is found at T{sub c}{approx}0.7-0.9 MeV.

  14. Determination of the Quantity of I-135 Released from the AGR Experiment Series

    SciTech Connect (OSTI)

    Scates, Dawn Marie; Walter, John Bradley; Reber, Edward Lawrence; Sterbentz, James William; Petti, David Andrew

    2014-10-01

    A series of three Advanced Gas Reactor (AGR) experiments have been conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). From 2006 through 2014, these experiments supported the development and qualification of the new U.S. tri structural isotropic (TRISO) particle fuel for Very High Temperature Reactors (VHTR). Each AGR experiment consisted of multiple fueled capsules, each plumbed for independent temperature control using a mix of helium and neon gases. The gas leaving a capsule was routed to individual Fission Product Monitor (FPM) detectors. For intact fuel particles, the TRISO particle coatings provide a substantial barrier to fission product release. However, particles with failed coatings, whether because of a minute percentage of initially defective particles, those which fail during irradiation, or those designed to fail (DTF) particles, can release fission products to the flowing gas stream. Because reactive fission product elements like iodine and cesium quickly deposit on cooler capsule components and piping structures as the effluent gas leaves the reactor core, only the noble fission gas isotopes of Kr and Xe tend to reach FPM detectors. The FPM system utilizes High Purity Germanium (HPGe) detectors coupled with a thallium activated sodium iodide NaI(Tl) scintillator. The germanium detector provides individual isotopic information, while the NaI(Tl) scintillator is used as a gross count rate meter. During irradiation, the 135mXe concentration reaching the FPM detectors is from both direct fission and by decay of the accumulated 135I. About ~2.5 hours after irradiation (ten 15.3 minute 135mXe half lives) the directly produced 135mXe has decayed and only the longer lived 135I remains as a source. Decay systematics dictate that 135mXe will be in secular equilibrium with its 135I parent, such that its production rate very nearly equals the decay rate of the parent, and its concentration in the flowing gas stream will appear to decay with the parent half life. This equilibrium condition enables the determination of the amount of 135I released from the fuel particles by measurement of the 135mXe at the FPM following reactor shutdown. In this paper, the 135I released will be reported and compared to similar releases for noble gases as well as the unexpected finding of 131I deposition from intentional impure gas injection into capsule 11 of experiment AGR 3/4.

  15. Table A18. Quantity of Electricity Sold to Utility and Nonutility...

    U.S. Energy Information Administration (EIA) Indexed Site

    ...ucts",988,940,48,16.2 2011," Meat Packing Plants",0,0,0,"NF" 2033," Canned Fruits and ...roducts",58,58,0,20.9 2011," Meat Packing Plants",0,0,0,"NF" 2033," Canned Fruits and ...

  16. Table A30. Quantity of Electricity Sold to Utility and Nonutility...

    U.S. Energy Information Administration (EIA) Indexed Site

    W "," W ",28 2011," Meat Packing Plants",0,0,0,0 2033," Canned Fruits and ... W "," W "," W ",26.6 2011," Meat Packing Plants",0,0,0,0 2033," Canned Fruits and ...

  17. Method for the rapid synthesis of large quantities of metal oxide nanowires at low temperatures

    DOE Patents [OSTI]

    Sunkara, Mahendra Kumar; Vaddiraju, Sreeram; Mozetic, Miran; Cvelbar, Uros

    2009-09-22

    A process for the rapid synthesis of metal oxide nanoparticles at low temperatures and methods which facilitate the fabrication of long metal oxide nanowires. The method is based on treatment of metals with oxygen plasma. Using oxygen plasma at low temperatures allows for rapid growth unlike other synthesis methods where nanomaterials take a long time to grow. Density of neutral oxygen atoms in plasma is a controlling factor for the yield of nanowires. The oxygen atom density window differs for different materials. By selecting the optimal oxygen atom density for various materials the yield can be maximized for nanowire synthesis of the metal.

  18. Import and Export of Category 1 and 2 Radioactive Sources Aggregated Quantities

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2008-11-10

    To formalize relevant guidance contained in the International Atomic Energy Agency (IAEA) CODEOC 2004, Code of Conduct on the Safety and Security of Radioactive Sources, January 2004 and IAEA CODEOC IMP-EXP 2005, Guidance on the Import and Export of Radioactive Sources, March 2005 and to assign responsibilities and prescribe procedures for DOE elements and contractors in support of the Import-Export Guidance. Admin Chg 1, 7-10-2013 supersedes DOE O 462.1. Certified 12-3-14.

  19. Import and Export of Category 1 and 2 Radioactive Sources and Aggregated Quantities

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2008-11-10

    This Order has been developed to provide requirements and responsibilities pertaining to the International Atomic Energy Agency CODEOC/2004, Code of Conduct on the Safety and Security of Radioactive Sources. No cancellation. Admin Chg 1, 7-10-13.

  20. Source Term Analysis for the WIPP Release Quantity 5-28-14

    Office of Environmental Management (EM)

    he k nown b reached d rum, 2 ) m easured A m---241Pu---239 ratios i n t he d rum a nd a ir f ilters c ollected p ost---release, a nd 3 ) t he t ime s eries o f t he r elease....

  1. W, F, and I : Three quantities basic to radiation physics. (Conference...

    Office of Scientific and Technical Information (OSTI)

    ion pair of either sign produced'', or, in a simpler language, ''per electron liberated''. ... Country of Publication: United States Language: English Subject: 73 NUCLEAR PHYSICS AND ...

  2. Reduced hierarchical equations of motion in real and imaginary time: Correlated initial states and thermodynamic quantities

    SciTech Connect (OSTI)

    Tanimura, Yoshitaka

    2014-07-28

    For a system strongly coupled to a heat bath, the quantum coherence of the system and the heat bath plays an important role in the system dynamics. This is particularly true in the case of non-Markovian noise. We rigorously investigate the influence of system-bath coherence by deriving the reduced hierarchal equations of motion (HEOM), not only in real time, but also in imaginary time, which represents an inverse temperature. It is shown that the HEOM in real time obtained when we include the system-bath coherence of the initial thermal equilibrium state possess the same form as those obtained from a factorized initial state. We find that the difference in behavior of systems treated in these two manners results from the difference in initial conditions of the HEOM elements, which are defined in path integral form. We also derive HEOM along the imaginary time path to obtain the thermal equilibrium state of a system strongly coupled to a non-Markovian bath. Then, we show that the steady state hierarchy elements calculated from the real-time HEOM can be expressed in terms of the hierarchy elements calculated from the imaginary-time HEOM. Moreover, we find that the imaginary-time HEOM allow us to evaluate a number of thermodynamic variables, including the free energy, entropy, internal energy, heat capacity, and susceptibility. The expectation values of the system energy and system-bath interaction energy in the thermal equilibrium state are also evaluated.

  3. Department of Energy, Indefinite Delivery Indefinite Quantity, Multiple Award, Energy Savings Performance

    Office of Environmental Management (EM)

    Modernize our Nation's Electric Grid | Department of Energy Department of Energy Official Touts Bush Administration's Efforts to Modernize our Nation's Electric Grid Department of Energy Official Touts Bush Administration's Efforts to Modernize our Nation's Electric Grid August 28, 2007 - 11:08am Addthis Louisiana to increase energy efficiency with upgrades between the LaBarre and Metaire electric substations NEW ORLEANS, LA - The U.S. Department of Energy's (DOE) newly confirmed Assistant

  4. Source Term Analysis for the WIPP Release Quantity 5-28-14

    Office of Environmental Management (EM)

    Department of Energy Sour Gas Streams Safe for Carbon Sequestration, DOE-Sponsored Study Shows Sour Gas Streams Safe for Carbon Sequestration, DOE-Sponsored Study Shows September 23, 2010 - 1:00pm Addthis Washington, D.C. -- Gas streams containing high levels of both carbon dioxide (CO2) and hydrogen sulfide (H2S) can be safely used for carbon capture and storage (CCS), according to results from a field test completed by the Plains CO2 Reduction (PCOR) Partnership. The test by PCOR--one of

  5. Effect of seasonal changes in quantities of biowaste on full scale anaerobic digester performance

    SciTech Connect (OSTI)

    Illmer, P. Gstraunthaler, G.

    2009-01-15

    A 750,000 l digester located in Roppen/Austria was studied over a 2-year period. The concentrations and amounts of CH{sub 4}, H{sub 2}, CO{sub 2} and H{sub 2}S and several other process parameters like temperature, retention time, dry weight and input of substrate were registered continuously. On a weekly scale the pH and the concentrations of NH{sub 4}{sup +}-N and volatile fatty acids (acetic, butyric, iso-butyric, propionic, valeric and iso-valeric acid) were measured. The data show a similar pattern of seasonal gas production over 2 years of monitoring. The consumption of VFA and not the hydrogenotrophic CH{sub 4} production appeared to be the limiting factor for the investigated digestion process. Whereas the changes in pH and the concentrations of most VFA did not correspond with changes in biogas production, the ratio of acetic to propionic acid and the concentration of H{sub 2} appeared to be useful indicators for reactor performance. However, the most influential factors for the anaerobic digestion process were the amount and the quality of input material, which distinctly changed throughout the year.

  6. Table 7.7 Quantity of Purchased Electricity, Natural Gas, and...

    U.S. Energy Information Administration (EIA) Indexed Site

    ... 2,673 W W 22 20 2 W W W 325193 Ethyl Alcohol 7,359 6,597 762 251 81 170 2,825 1,611 ... W W 0 W W 0 0 0 0 325193 Ethyl Alcohol 72 72 0 W 1 W 0 0 0 325199 Other Basic ...

  7. Table 7.7 Quantity of Purchased Electricity, Natural Gas, and...

    U.S. Energy Information Administration (EIA) Indexed Site

    ...2,15,26,17775,0,17775,0.9 325193," Ethyl Alcohol ",1309,1209,101,29,10,20,6861,3255,"Q",4....","W",4,"W","W",0,"W",0.9 325193," Ethyl Alcohol ","*","*",0,"*","*",0,0,0,0,0.7 325199," ...

  8. Supplemental Immobilization of Hanford Low-Activity Waste: Cast Stone Augmented Formulation Matrix Tests

    SciTech Connect (OSTI)

    Cozzi, A.; Crawford, C.; Fox, K.; Hansen, E.; Roberts, K.

    2015-07-20

    More than 56 million gallons of radioactive and hazardous waste are stored in 177 underground storage tanks at the U.S. Department of Energy’s (DOE’s) Hanford Site in Washington State. The HLW will be vitrified in the HLW facility for ultimate disposal at an offsite federal repository. A portion (~35%) of the LAW will be vitrified in the LAW vitrification facility for disposal onsite at the Integrated Disposal Facility (IDF). The pretreatment and HLW vitrification facilities will have the capacity to treat and immobilize all of the wastes destined for those facilities. However, a second facility will be needed for the expected volume of LAW requiring immobilization. Cast Stone, a cementitious waste form, is being considered to provide the required additional LAW immobilization capacity. The Cast Stone waste form must be acceptable for disposal in the IDF. The Cast Stone waste form and immobilization process must be tested to demonstrate that the final Cast Stone waste form can comply with the waste acceptance criteria for the disposal facility and that the immobilization processes can be controlled to consistently provide an acceptable waste form product. A testing program was developed in fiscal year (FY) 2012 describing in detail the work needed to develop and qualify Cast Stone as a waste form for the solidification of Hanford LAW. A statistically designed test matrix was used to evaluate the effects of key parameters on the properties of the Cast Stone as it is initially prepared and after curing. For the processing properties, the water-to-dry-blend mix ratio was the most significant parameter in affecting the range of values observed for each property. The single shell tank (SST) Blend simulant also showed differences in measured properties compared to the other three simulants tested. A review of the testing matrix and results indicated that an additional set of tests would be beneficial to improve the understanding of the impacts noted in the Screening Matrix tests. A set of Cast Stone formulations were devised to augment the original screening test matrix and focus on the range of the test conditions. Fly ash and blast furnace slag were limited to either northwest or southeast and the salt solutions were narrowed to the Average and the SST Blend at the 7.8M Na concentration. To fill in the matrix, a mix ratio of 0.5 was added. In addition, two admixtures, Xypex Admix C-500 and Rheomac SF100 (silica fume), were added as an additional dry material binder in select compositions. As in the Screening Matrix, both fresh and cured properties were evaluated for the formulations. In this study, properties that were influenced by the W/DM ratio in the Screening Matrix; flow diameter, plastic viscosity, density, and compressive strength, showed consistent behavior with respect to W/DM. The leach index for highly soluble components, sodium and nitrate, were not influenced by changes in formulation or the admixtures. The leach index for both iodine and Tc-99 show an influence from the addition of the admixture, Xypex Admix C-500. Additional testing should be performed to further evaluate the influence of Xypex Admix C-500 on the leach index over a range of admixture concentrations, Cast Stone formulations, and curing and storage conditions.

  9. Technitium Management at the Hanford Site

    SciTech Connect (OSTI)

    Robbins, Rebecca A.

    2013-08-15

    Long Abstract. Full Text. The Hanford tank waste contains approx 26,000 Ci of technetium-99 (Tc-99), the majority of which is in the supernate fraction. Tc-99 is a long-lived radionuclide with a half-life of approx 212,000 years and, in its predominant pertechnetate (TcO{sub 4}) form, is highly soluble and very mobile in the vadose zone and ultimately the groundwater. Tc-99 is identified as the major dose contributor (in groundwater) by past Hanford site performance assessments and therefore considered a key radionuclide of concern at Hanford. The United States Department of Energy (DOE) River Protection Project's (RPP) long-term Tc-99 management strategy is to immobilize the Tc-99 in a waste form that will retain the Tc-99 for many thousands of years. To achieve this, the RPP flowsheet will immobilize the majority of the Tc-99 as a vitrified low-activity waste product that will be ultimately disposed on site in the Integrated Disposal Facility. The Tc-99 will be released gradually from the glass at very low rates such that the groundwater concentrations at any point in time would be substantially below regulatory limits.The liquid secondary waste will be immobilized in a low-temperature matrix (cast stone) and the solid secondary waste will be stabilized using grout. Although the Tc-99 that is immobilized in glass will meet the release rate for disposal in IDF, a proportion is driven into the secondary waste stream that will not be vitrified and therefore presents a disposal risk. If a portion of the Tc-99 were to be removed from the Hanford waste inventory and disposed off-site, (e.g., as HLW), it could lessen a major constraint on LAW waste form performance, i.e., the requirement to retain Tc-99 over thousands of years and have a positive impact on the IDF Performance Assessment. There are several technologies available at various stages of technical maturity that can be employed for Tc-99 removal. The choice of technology and the associated efficacy of the technology are dependent on the chemical fonn of the technetium in the waste, the removal location in the tlowsheet. and the ultimate disposition path chosen for the technetium product. This paper will discuss the current plans for the management of the technetium present in the Hanford tank waste. It will present the risks associated with processing technetium in the current treatment flowsheet and present potential mitigation opportunities, the status of available technetium removal technologies, the chemical speciation of technetium in the tank waste, and the available disposition paths and waste forms for technetium containing streams.

  10. Final Report - "Foaming and Antifoaming and Gas Entrainment in Radioactive Waste Pretreatment and Immobilization Processes"

    SciTech Connect (OSTI)

    Wasan, Darsh T.

    2007-10-09

    The Savannah River Site (SRS) and Hanford site are in the process of stabilizing millions of gallons of radioactive waste slurries remaining from production of nuclear materials for the Department of Energy (DOE). The Defense Waste Processing Facility (DWPF) at SRS is currently vitrifying the waste in borosilicate glass, while the facilities at the Hanford site are in the construction phase. Both processes utilize slurry-fed joule-heated melters to vitrify the waste slurries. The DWPF has experienced difficulty during operations. The cause of the operational problems has been attributed to foaming, gas entrainment and the rheological properties of the process slurries. The rheological properties of the waste slurries limit the total solids content that can be processed by the remote equipment during the pretreatment and meter feed processes. Highly viscous material can lead to air entrainment during agitation and difficulties with pump operations. Excessive foaming in waste evaporators can cause carryover of radionuclides and non-radioactive waste to the condensate system. Experimental and theoretical investigations of the surface phenomena, suspension rheology and bubble generation of interactions that lead to foaming and air entrainment problems in the DOE High Level and Low Activity Radioactive Waste separation and immobilization processes were pursued under this project. The first major task accomplished in the grant proposal involved development of a theoretical model of the phenomenon of foaming in a three-phase gas-liquid-solid slurry system. This work was presented in a recently completed Ph.D. thesis (9). The second major task involved the investigation of the inter-particle interaction and microstructure formation in a model slurry by the batch sedimentation method. Both experiments and modeling studies were carried out. The results were presented in a recently completed Ph.D. thesis. The third task involved the use of laser confocal microscopy to study the effectiveness of three slurry rheology modifiers. An effective modifier was identified which resulted in lowering the yield stress of the waste simulant. Therefore, the results of this research have led to the basic understanding of the foaming/antifoaming mechanism in waste slurries as well as identification of a rheology modifier, which enhances the processing throughput, and accelerates the DOE mission. The objectives of this research effort were to develop a fundamental understanding of the physico-chemical mechanisms that produced foaming and air entrainment in the DOE High Level (HLW) and Low Activity (LAW) radioactive waste separation and immobilization processes, and to develop and test advanced antifoam/defoaming/rheology modifier agents. Antifoams/rheology modifiers developed from this research ere tested using non-radioactive simulants of the radioactive wastes obtained from Hanford and the Savannah River Site (SRS).

  11. IMPACT OF NOBLE METALS AND MERCURY ON HYDROGEN GENERATION DURING HIGH LEVEL WASTE PRETREATMENT AT THE SAVANNAH RIVER SITE

    SciTech Connect (OSTI)

    Stone, M; Tommy Edwards, T; David Koopman, D

    2009-03-03

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site vitrifies radioactive High Level Waste (HLW) for repository internment. The process consists of three major steps: waste pretreatment, vitrification, and canister decontamination/sealing. HLW consists of insoluble metal hydroxides (primarily iron, aluminum, calcium, magnesium, manganese, and uranium) and soluble sodium salts (carbonate, hydroxide, nitrite, nitrate, and sulfate). The pretreatment process in the Chemical Processing Cell (CPC) consists of two process tanks, the Sludge Receipt and Adjustment Tank (SRAT) and the Slurry Mix Evaporator (SME) as well as a melter feed tank. During SRAT processing, nitric and formic acids are added to the sludge to lower pH, destroy nitrite and carbonate ions, and reduce mercury and manganese. During the SME cycle, glass formers are added, and the batch is concentrated to the final solids target prior to vitrification. During these processes, hydrogen can be produced by catalytic decomposition of excess formic acid. The waste contains silver, palladium, rhodium, ruthenium, and mercury, but silver and palladium have been shown to be insignificant factors in catalytic hydrogen generation during the DWPF process. A full factorial experimental design was developed to ensure that the existence of statistically significant two-way interactions could be determined without confounding of the main effects with the two-way interaction effects. Rh ranged from 0.0026-0.013% and Ru ranged from 0.010-0.050% in the dried sludge solids, while initial Hg ranged from 0.5-2.5 wt%, as shown in Table 1. The nominal matrix design consisted of twelve SRAT cycles. Testing included: a three factor (Rh, Ru, and Hg) study at two levels per factor (eight runs), three duplicate midpoint runs, and one additional replicate run to assess reproducibility away from the midpoint. Midpoint testing was used to identify potential quadratic effects from the three factors. A single sludge simulant was used for all tests and was spiked with the required amount of noble metals immediately prior to performing the test. Acid addition was kept effectively constant except to compensate for variations in the starting mercury concentration. SME cycles were also performed during six of the tests.

  12. Pulse Jet Mixer Overblow Testing for Assessment of Loadings During Multiple Overblows

    SciTech Connect (OSTI)

    Pfund, David M.; Bontha, Jagannadha R.; Michener, Thomas E.; Nigl, Franz; Yokuda, Satoru T.; Leigh, Richard J.; Golovich, Elizabeth C.; Baumann, Aaron W.; Kurath, Dean E.; Hoza, Mark; Combs, William H.; Fort, James A.; Bredt, Ofelia P.

    2009-07-20

    The U.S. Department of Energy (DOE) Office of River Protections Waste Treatment Plant (WTP) is being designed and built to pretreat and then vitrify a large portion of the wastes in Hanfords 177 underground waste storage tanks. The WTP consists of three primary facilities: pretreatment, low-activity waste (LAW) vitrification, and high-level waste (HLW) vitrification. The pretreatment facility will receive waste feed from the Hanford tank farms and separate it into 1) a high-volume, low-activity liquid stream stripped of most solids and radionuclides and 2) a much smaller volume of HLW slurry containing most of the solids and most of the radioactivity. Many of the vessels in the pretreatment facility will contain pulse jet mixers (PJMs) that will provide some or all of the mixing in the vessels. This technology was selected for use in so-called black cell regions of the WTP, where maintenance capability will not be available for the operating life of the WTP. PJM technology was selected for use in these regions because it has no moving mechanical parts that require maintenance. The vessels with the most concentrated slurries will also be mixed with air spargers and/or steady jets in addition to the mixing provided by the PJMs. This report contains the results of single and multiple PJM overblow tests conducted in a large, ~13 ft-diameter 15-ft-tall tank located in the high bay of the Pacific Northwest National Laboratory (PNNL) 336 Building test facility. These single and multiple PJM overblow tests were conducted using water and a clay simulant to bound the lower and upper rheological properties of the waste streams anticipated to be processed in the WTP. Hydrodynamic pressures were measured at a number of locations in the test vessel using an array of nine pressure sensors and four hydrophones. These measurements were made under normal and limiting vessel operating conditions (i.e., maximum PJM fluid emptying velocity, maximum and minimum vessel contents for PJM operation, and maximum and minimum rheological properties). Test data collected from the PJM overblow tests were provided to Bechtel National, Inc. (BNI) for assessing hydrostatic, dynamic, and acoustic pressure loadings on in-tank structures during 1) single overblows; 2) multiple overlapping overblows of two to four PJMs; 3) simultaneous overblows of pairs of PJMs.

  13. Pulse Jet Mixer Overblow Testing for Assessment of Loadings During Multiple Overblows

    SciTech Connect (OSTI)

    Pfund, David M.; Bontha, Jagannadha R.; Michener, Thomas E.; Nigl, Franz; Yokuda, Satoru T.; Leigh, Richard J.; Golovich, Elizabeth C.; Baumann, Aaron W.; Kurath, Dean E.; Hoza, Mark; Combs, William H.; Fort, James A.; Bredt, Ofelia P.

    2008-03-03

    The U.S. Department of Energy (DOE) Office of River Protections Waste Treatment Plant (WTP) is being designed and built to pretreat and then vitrify a large portion of the wastes in Hanfords 177 underground waste storage tanks. The WTP consists of three primary facilities: pretreatment, low-activity waste (LAW) vitrification, and high-level waste (HLW) vitrification. The pretreatment facility will receive waste feed from the Hanford tank farms and separate it into 1) a high-volume, low-activity liquid stream stripped of most solids and radionuclides and 2) a much smaller volume of HLW slurry containing most of the solids and most of the radioactivity. Many of the vessels in the pretreatment facility will contain pulse jet mixers (PJMs) that will provide some or all of the mixing in the vessels. This technology was selected for use in so-called black cell regions of the WTP, where maintenance capability will not be available for the operating life of the WTP. PJM technology was selected for use in these regions because it has no moving mechanical parts that require maintenance. The vessels with the most concentrated slurries will also be mixed with air spargers and/or steady jets in addition to the mixing provided by the PJMs. This report contains the results of single and multiple PJM overblow tests conducted in a large, ~13 ft-diameter 15-ft-tall tank located in the high bay of the Pacific Northwest National Laboratory (PNNL) 336 Building test facility. These single and multiple PJM overblow tests were conducted using water and a clay simulant to bound the lower and upper rheological properties of the waste streams anticipated to be processed in the WTP. Hydrodynamic pressures were measured at a number of locations in the test vessel using an array of nine pressure sensors and four hydrophones. These measurements were made under normal and limiting vessel operating conditions (i.e., maximum PJM fluid emptying velocity, maximum and minimum vessel contents for PJM operation, and maximum and minimum rheological properties). Test data collected from the PJM overblow tests were provided to Bechtel National, Inc. (BNI) for assessing hydrostatic, dynamic, and acoustic pressure loadings on in-tank structures during 1) single overblows; 2) multiple overlapping overblows of two to four PJMs; 3) simultaneous overblows of pairs of PJMs.

  14. Calculates Neutron Production in Canisters of High-level Waste

    Energy Science and Technology Software Center (OSTI)

    1993-01-15

    ALPHN calculates the (alpha,n) neutron production rate of a canister of vitrified high-level waste. The user supplies the chemical composition of the glass or glass-ceramic and the curies of the alpha-emitting actinides present. The output of the program gives the (alpha,n) neutron production of each actinide in neutrons per second and the total for the canister. The (alpha,n) neutron production rates are source terms only; that is, they are production rates within the glass andmore » do not take into account the shielding effect of the glass. For a given glass composition, the user can calculate up to eight cases simultaneously; these cases are based on the same glass composition but contain different quantities of actinides per canister.« less

  15. Secondary Waste Form Down-Selection Data Package—Fluidized Bed Steam Reforming Waste Form

    SciTech Connect (OSTI)

    Qafoku, Nikolla; Westsik, Joseph H.; Strachan, Denis M.; Valenta, Michelle M.; Pires, Richard P.

    2011-09-12

    The Hanford Site in southeast Washington State has 56 million gallons of radioactive and chemically hazardous wastes stored in 177 underground tanks (ORP 2010). The U.S. Department of Energy (DOE), Office of River Protection (ORP), through its contractors, is constructing the Hanford Tank Waste Treatment and Immobilization Plant (WTP) to convert the radioactive and hazardous wastes into stable glass waste forms for disposal. Within the WTP, the pretreatment facility will receive the retrieved waste from the tank farms and separate it into two treated process streams. These waste streams will be vitrified, and the resulting waste canisters will be sent to offsite (high-level waste [HLW]) and onsite (immobilized low-activity waste [ILAW]) repositories. As part of the pretreatment and ILAW processing, liquid secondary wastes will be generated that will be transferred to the Effluent Treatment Facility (ETF) on the Hanford Site for further treatment. These liquid secondary wastes will be converted to stable solid waste forms that will be disposed of in the Integrated Disposal Facility (IDF). To support the selection of a waste form for the liquid secondary wastes from WTP, Washington River Protection Solutions (WRPS) has initiated secondary waste form testing work at Pacific Northwest National Laboratory (PNNL). In anticipation of a down-selection process for a waste form for the Solidification Treatment Unit to be added to the ETF, PNNL is developing data packages to support that down-selection. The objective of the data packages is to identify, evaluate, and summarize the existing information on the four waste forms being considered for stabilizing and solidifying the liquid secondary wastes. At the Hanford Site, the FBSR process is being evaluated as a supplemental technology for treating and immobilizing Hanford LAW radioactive tank waste and for treating secondary wastes from the WTP pretreatment and LAW vitrification processes.

  16. Integrated Pilot Plant for a Large Cold Crucible Induction Melter

    SciTech Connect (OSTI)

    Do Quang, R.; Jensen, A.; Prod'homme, A.; Fatoux, R.; Lacombe, J.

    2002-02-26

    COGEMA has been vitrifying high-level liquid waste produced during nuclear fuel reprocessing on an industrial scale for over 20 years, with two main objectives: containment of the long lived fission products and reduction of the final volume of waste. Research performed by the French Atomic Energy Commission (CEA) in the 1950s led to the selection of borosilicate glass as the most suitable containment matrix for waste from spent nuclear fuel and to the development of the induction melter technology. This was followed by the commissioning of the Marcoule Vitrification Facility (AVM) in 1978. The process was implemented at a larger scale in the late 1980s in the R7 and T7 facilities of the La Hague reprocessing plant. COGEMA facilities have produced more than 11,000 high level glass canisters, representing more than 4,500 metric tons of glass and 4.5 billion curies. To further improve the performance of the vitrification lines in the R7 and T7 facilities, the CEA and COGEMA have been developing the Cold Crucible Melter (CCM) technology since the 1980s. This technology benefits from the 20 years of COGEMA HLW vitrification experience and ensures a virtually unlimited equipment service life and extensive flexibility in dealing with different types of waste. The high specific power directly transferred by induction to the melt allows high operating temperatures without any impact on the process equipment. In addition, the mechanical stirring of the melter significantly reduces operating constraints. COGEMA is already providing the CCM technology to international customers for nuclear and non-nuclear applications and plans to implement it in the La Hague vitrification plant for the vitrification of highly concentrated and corrosive solutions produced by uranium/molybdenum fuel reprocessing. The paper presents the CCM project that led to the building and start-up of this evolutionary and flexible pilot plant. It also describes the plant's technical characteristics and reports commissioning results.

  17. Protection of Operators and Environment - the Safety Concept of the Karlsruhe Vitrification Plant VEK

    SciTech Connect (OSTI)

    Fleisch, J.; Kuttruf, H.; Lumpp, W.; Pfeifer, W.; Roth, G.; Weisenburger, S.

    2002-02-26

    The Karlsruhe Vitrification Plant (VEK) plant is a milestone in decommissioning and complete dismantling of the former Karlsruhe Reprocessing Plant WAK, which is in an advanced stage of disassembly. The VEK is scheduled to vitrify approx. 70 m3 of the highly radioactive liquid waste (HLW) resulting from reprocessing. Site preparation, civil work and component manufacturing began in 1999. The building will be finalized by mid of 2002, hot vitrification operation is currently scheduled for 2004/2005. Provisions against damages arising from construction and operation of the VEK had to be made in accordance with the state of the art as laid down in the German Atomic Law and the Radiation Protection Regulations. For this purpose, the appropriate analysis of accidents and their external and internal impacts were investigated. During the detailed design phase, a failure effects analysis was carried out, in which single events were studied with respect to the objectives of protection and ensuring activity containment, limiting radioactive discharges to the environment and protecting of the staff. Parallel to the planning phase of the VEK plant a cold prototype test facility (PVA) covering the main process steps was constructed and operated at the Institut fuer Nukleare Entsorgung (INE) of FZK. This pilot operation served to demonstrate the process technique and its operation with a simulated waste solution, and to test the main items of equipment, but was conducted also to use the experimental data and experience to back the safety concept of the radioactive VEK plant. This paper describes the basis of the safety concept of the VEK plant and results of the failure effect analysis. The experimental simulation of the failure scenarios, their effect on the process behavior, and the controllability of these events as well as the effect of the results on the safety concept of VEK are discussed. Additionally, an overview of the actual status of civil work and manufacturing of the technical equipment is given.

  18. RADIOACTIVE DEMONSTRATIONS OF FLUIDIZED BED STEAM REFORMING WITH ACUTAL HANFORD LOW ACTIVITY WASTES VERIFYING FBSR AS A SUPPLEMENTARY TREATMENT

    SciTech Connect (OSTI)

    Jantzen, C.; Crawford, C.; Burket, P.; Bannochie, C.; Daniel, G.; Nash, C.; Cozzi, A.; Herman, C.

    2012-01-12

    The U.S. Department of Energy's Office of River Protection is responsible for the retrieval, treatment, immobilization, and disposal of Hanford's tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level waste (HLW) and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the cleanup mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA). Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. Fluidized Bed Steam Reforming (FBSR) is one of the supplementary treatments being considered. FBSR offers a moderate temperature (700-750 C) continuous method by which LAW and other secondary wastes can be processed irrespective of whether they contain organics, nitrates/nitrites, sulfates/sulfides, chlorides, fluorides, and/or radio-nuclides like I-129 and Tc-99. Radioactive testing of Savannah River LAW (Tank 50) shimmed to resemble Hanford LAW and actual Hanford LAW (SX-105 and AN-103) have produced a ceramic (mineral) waste form which is the same as the non-radioactive waste simulants tested at the engineering scale. The radioactive testing demonstrated that the FBSR process can retain the volatile radioactive components that cannot be contained at vitrification temperatures. The radioactive and nonradioactive mineral waste forms that were produced by co-processing waste with kaolin clay in an FBSR process are shown to be as durable as LAW glass.

  19. ESTIMATING HIGH LEVEL WASTE MIXING PERFORMANCE IN HANFORD DOUBLE SHELL TANKS

    SciTech Connect (OSTI)

    THIEN MG; GREER DA; TOWNSON P

    2011-01-13

    The ability to effectively mix, sample, certify, and deliver consistent batches of high level waste (HLW) feed from the Hanford double shell tanks (DSTs) to the Waste Treatment and Immobilization Plant (WTP) presents a significant mission risk with potential to impact mission length and the quantity of HLW glass produced. The Department of Energy's (DOE's) Tank Operations Contractor (TOC), Washington River Protection Solutions (WRPS) is currently demonstrating mixing, sampling, and batch transfer performance in two different sizes of small-scale DSTs. The results of these demonstrations will be used to estimate full-scale DST mixing performance and provide the key input to a programmatic decision on the need to build a dedicated feed certification facility. This paper discusses the results from initial mixing demonstration activities and presents data evaluation techniques that allow insight into the performance relationships of the two small tanks. The next steps, sampling and batch transfers, of the small scale demonstration activities are introduced. A discussion of the integration of results from the mixing, sampling, and batch transfer tests to allow estimating full-scale DST performance is presented.

  20. RCC Contract No. DE-AC06-05RL14655 TABLE B.2 SCHEDULE OF QUANTITIES...

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

    ... Impact for Hanford Site Beryllium Work Permit Implementation (Defintized by Mod 466 ... RTD for 100-N-84:5 (100-N Sanitary Sewer Pipeline) from Contract (Definitized by Mod 484 ...

  1. Growth of microscopic cones on titanium cathodes of sputter-ion pumps driven by sorption of large argon quantities

    SciTech Connect (OSTI)

    Porcelli, Tommaso; Siviero, Fabrizio; Bongiorno, Gero A.; Michelato, Paolo; Pagani, Carlo

    2015-09-15

    Microscopic cones have been observed on titanium cathodes of sputter-ion pumps (SIPs) after pump operation. The cones were studied by means of scanning electron microscopy and energy dispersive x-ray analysis. Size and morphology of these cones are clearly correlated with the nature and the relative amount of each gas species pumped by each SIP during its working life. In particular, their growth was found to be fed by sputtering mechanisms, mostly during Ar pumping, and to be driven by the electromagnetic field applied to the Penning cells of each SIP. Experimental findings suggest that the formation and extent of such conic structures on cathode surfaces might play a leading role in the onset of phenomena typically related to the functioning of SIPs, e.g., the so-called argon instability.

  2. Studies of minute quantities of natural abundance molecules using 2D heteronuclear correlation spectroscopy under 100kHz MAS

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Nishiyama, Y.; Kobayashi, T.; Malon, M.; Singappuli-Arachchige, D.; Slowing, I. I.; Pruski, M.

    2015-02-16

    Two-dimensional 1H{13C} heteronuclear correlation solid-state NMR spectra of naturally abundant solid materials are presented, acquired using the 0.75-mm magic angle spinning (MAS) probe at spinning rates up to 100 kHz. In spite of the miniscule sample volume (290 nL), high-quality HSQC-type spectra of bulk samples as well as surface-bound molecules can be obtained within hours of experimental time. The experiments are compared with those carried out at 40 kHz MAS using a 1.6-mm probe, which offered higher overall sensitivity due to a larger rotor volume. The benefits of ultrafast MAS in such experiments include superior resolution in 1H dimensionmore » without resorting to 1H–1H homonuclear RF decoupling, easy optimization, and applicability to mass-limited samples. As a result, the HMQC spectra of surface-bound species can be also acquired under 100 kHz MAS, although the dephasing of transverse magnetization has significant effect on the efficiency transfer under MAS alone.« less

  3. Urban Mining: Quality and quantity of recyclable and recoverable material mechanically and physically extractable from residual waste

    SciTech Connect (OSTI)

    Di Maria, Francesco Micale, Caterina; Sordi, Alessio; Cirulli, Giuseppe; Marionni, Moreno

    2013-12-15

    Highlights: Material recycling and recovery from residual waste by physical and mechanical process has been investigated. About 6% of recyclable can be extracted by NIR and 2-3Dimension selector. Another 2% of construction materials can be extracted by adopting modified soil washing process. Extracted material quality is quite high even some residual heavy metal have been detected by leaching test. - Abstract: The mechanically sorted dry fraction (MSDF) and Fines (<20 mm) arising from the mechanical biological treatment of residual municipal solid waste (RMSW) contains respectively about 11% w/w each of recyclable and recoverable materials. Processing a large sample of MSDF in an existing full-scale mechanical sorting facility equipped with near infrared and 2-3 dimensional selectors led to the extraction of about 6% w/w of recyclables with respect to the RMSW weight. Maximum selection efficiency was achieved for metals, about 98% w/w, whereas it was lower for Waste Electrical and Electronic Equipment (WEEE), about 2% w/w. After a simulated lab scale soil washing treatment it was possible to extract about 2% w/w of inert exploitable substances recoverable as construction materials, with respect to the amount of RMSW. The passing curve showed that inert materials were mainly sand with a particle size ranging from 0.063 to 2 mm. Leaching tests showed quite low heavy metal concentrations with the exception of the particles retained by the 0.5 mm sieve. A minimum pollutant concentration was in the leachate from the 10 and 20 mm particle size fractions.

  4. Method and an apparatus for non-invasively determining the quantity of an element in a body organ

    DOE Patents [OSTI]

    Vartsky, D.; Ellis, K.J.; Cohn, S.H.

    1980-06-27

    An apparatus and a method for determining in a body organ the amount of an element with the aid of a gaseous gamma ray source, where the element and the source are paired in predetermined pairs, and with the aid of at least one detector selected from the group consisting of Ge(Li) and NaI(Tl). Gamma rays are directed towards the organ, thereby resonantly scattering the gamma rays from nuclei of the element in the organ; the intensity of the gamma rays is detected by the detector; and the amount of the element in the organ is then substantially proportional to the detected intensity of the gamma rays.

  5. HANFORD RIVER PROTECTION PROJECT ENHANCED MISSION PLANNING THROUGH INNOVATIVE TOOLS LIFECYCLE COST MODELING AND AQUEOUS THERMODYNAMIC MODELING - 12134

    SciTech Connect (OSTI)

    PIERSON KL; MEINERT FL

    2012-01-26

    Two notable modeling efforts within the Hanford Tank Waste Operations Simulator (HTWOS) are currently underway to (1) increase the robustness of the underlying chemistry approximations through the development and implementation of an aqueous thermodynamic model, and (2) add enhanced planning capabilities to the HTWOS model through development and incorporation of the lifecycle cost model (LCM). Since even seemingly small changes in apparent waste composition or treatment parameters can result in large changes in quantities of high-level waste (HLW) and low-activity waste (LAW) glass, mission duration or lifecycle cost, a solubility model that more accurately depicts the phases and concentrations of constituents in tank waste is required. The LCM enables evaluation of the interactions of proposed changes on lifecycle mission costs, which is critical for decision makers.

  6. West Valley Demonstration Project Prepares to Relocate High-Level Waste |

    Office of Environmental Management (EM)

    Department of Energy HLW Waste Vitrification Facility Waste Treatment and Immobilation Plant HLW Waste Vitrification Facility Full Document and Summary Versions are available for download PDF icon Waste Treatment and Immobilation Plant HLW Waste Vitrification Facility PDF icon Summary - WTP HLW Waste Vitrification Facility More Documents & Publications Waste Treatment and Immobilization Plant (WTP) Analytical Laboratory (LAB), Balance of Facilities (BOF) and Low-Activity Waste

  7. Advanced waste form and Melter development for treatment of troublesome high-level wastes

    SciTech Connect (OSTI)

    Marra, James; Kim, Dong -Sang; Maio, Vincent

    2015-10-01

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these “troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.The Hanford site AZ-101 tank waste composition represents a waste group that is waste loading limited primarily due to high concentrations of Fe2O3 (also with high Al2O3 concentrations). Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of more than 20% could be achieved for this waste composition, and by extension higher loadings for wastes in the same group. An extended duration CCIM melter test was conducted on an AZ-101 waste simulant using the CCIM platform at the Idaho National Laboratory (INL). The melter was continually operated for approximately 80 hours demonstrating that the AZ-101 high waste loading glass composition could be readily processed using the CCIM technology. The resulting glass was close to the targeted composition and exhibited excellent durability in both the as poured state and after being slowly cooled according to the canister centerline cooling (CCC) profile. Glass formulation development was also completed on other Hanford tank wastes that were identified to further challenge waste loading due to the presence of appreciable quantities (>750 g) of plutonium in the waste tanks. In addition to containing appreciable Pu quantities, the C-102 waste tank and the 244-TX waste tank contain high concentrations of aluminum and iron, respectively that will further challenge vitrification processing. Glass formulation testing also demonstrated that high waste loadings could be achieved with these tank compositions using the attributes afforded by the CCIM technology.

  8. Geological aspects of the nuclear waste disposal problem

    SciTech Connect (OSTI)

    Laverov, N.P.; Omelianenko, B.L.; Velichkin, V.I.

    1994-06-01

    For the successful solution of the high-level waste (HLW) problem in Russia one must take into account such factors as the existence of the great volume of accumulated HLW, the large size and variety of geological conditions in the country, and the difficult economic conditions. The most efficient method of HLW disposal consists in the maximum use of protective capacities of the geological environment and in using inexpensive natural minerals for engineered barrier construction. In this paper, the principal trends of geological investigation directed toward the solution of HLW disposal are considered. One urgent practical aim is the selection of sites in deep wells in regions where the HLW is now held in temporary storage. The aim of long-term investigations into HLW disposal is to evaluate geological prerequisites for regional HLW repositories.

  9. IG-0549.pub

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    ... DOE preferred vitrification because it is assumed that borosilicate glass, or vitrified waste, would be the only waste form acceptable in the high-level waste repository. ...

  10. EIS-0082-S2: Notice of Intent to Prepare a Supplemental Environmental...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    from liquid high-level radioactive waste before vitrifying the high-activity fraction of the waste in the Defense Waste Processing Facility and disposing of the...

  11. Reaction of Inconel 690 and 693 in Iron Phosphate Melts: Alternative Glasses for Waste Vitrification

    SciTech Connect (OSTI)

    Day, Delbert E.

    2005-09-13

    The corrosion resistance of candidate materials used for the electrodes (Inconel 690 & 693) and the melt contact refractory (Monofrax K-3) in a Joule Heated Melter (JHM) has been investigated at the University of Missouri-Rolla (UMR) during the period from June 1, 2004 to August 31, 2005. This work was supported by the U.S. Department of Energy (DOE) Office of Biological and Environmental Research (DE-FG02-04ER63831). The unusual properties and characteristics of iron phosphate glasses, as viewed from the standpoint of alternative glasses for vitrifying nuclear and hazardous wastes which contain components that make them poorly suited for vitrification in borosilicate glass, were recently discovered at UMR. The expanding national and international interest in iron phosphate glasses for waste vitrification stems from their rapid melting and chemical homogenization which results in higher furnace output, their high waste loading that varies from 32 wt% up to 75 wt% for the Hanford LAW and HLW, respectively, and the outstanding chemical durability of the iron phosphate wasteforms which meets all present DOE requirements (PCT and VHT). The higher waste loading in iron phosphate glasses, compared to the baseline borosilicate glass, can reduce the time and cost of vitrification considerably since a much smaller mass of glass will be produced, for example, about 43% less glass when the LAW at Hanford is vitrified in an iron phosphate glass according to PNNL estimates. In view of the promising performance of iron phosphate glasses, information is needed for how to best melt these glasses on the scale needed for practical use. Melting iron phosphate glasses in a JHM is considered the preferred method at this time because its design could be nearly identical to the JHM now used to melt borosilicate glasses at the Defense Waste Processing Facility (DWPF), Westinghouse Savannah River Co. Therefore, it is important to have information for the corrosion of candidate electrode and refractory materials in iron phosphate melts in a JHM. During the period from June 1, 2004 to August 31, 2005, the corrosion resistance of coupons of Inconel 690 & 693 metals and Monofrax K-3 refractory, partially submerged in several iron phosphate melts at 950-1200?C, has been investigated to determine whether iron phosphate glasses could be melted in a JHM equipped with such electrodes and refractory in the same manner as now being used to melt borosilicate glass. These representative iron phosphate melts, which contained 30 wt% Hanford LAW and 40 wt% Idaho SBW simulants, did not corrode the Inconel 690 to any greater extent than what has been reported for Inconel 690 in the borosilicate melt in the JHM at DWPF. Inconel 693 appeared to be an even better candidate for use in iron phosphate melts since its corrosion rate (1.8 to 25.4 ?m/day) was only about one half that (5.4 to 45.4 ?m/day) of Inconel 690. The dynamic corrosion measured for the candidate refractory, Monofrax K-3, by iron phosphate melts is quite encouraging since the measured corrosion (0.011 to 0.132 mm/day at 9.2 rpm) is less than the corrosion (0.137 mm/day) that has been reported in the JHM used to melt borosilicate glass at DWPF. During the period covered by this final report, the results of the research on iron phosphate glasses have been described in seven technical papers and have been presented at one national meeting. In addition to the principal investigator, one research professor and one undergraduate research aide were supported by this project.

  12. Fluidized Bed Steam Reforming (FBSR) Mineralization for High Organic and Nitrate Waste Streams for the Global Nuclear Energy Partnership (GNEP)

    SciTech Connect (OSTI)

    Jantzen, C.M.; Williams, M.R. [Savannah River National Laboratory, Aiken, SC (United States)

    2008-07-01

    Waste streams that may be generated by the Global Nuclear Energy Partnership (GNEP) Advanced Energy Initiative may contain significant quantities of organics (0-53 wt%) and/or nitrates (0-56 wt%). Decomposition of high nitrate streams requires reducing conditions, e.g. organic additives such as sugar or coal, to reduce the NOx in the off-gas to N{sub 2} to meet the Clean Air Act (CAA) standards during processing. Thus, organics will be present during waste form stabilization regardless of which GNEP processes are chosen, e.g. organics in the feed or organics for nitrate destruction. High organic containing wastes cannot be stabilized with the existing HLW Best Developed Available Technology (BDAT) which is HLW vitrification (HLVIT) unless the organics are removed by preprocessing. Alternative waste stabilization processes such as Fluidized Bed Steam Reforming (FBSR) operate at moderate temperatures (650-750 deg. C) compared to vitrification (1150-1300 deg. C). FBSR converts organics to CAA compliant gases, creates no secondary liquid waste streams, and creates a stable mineral waste form that is as durable as glass. For application to the high Cs-137 and Sr-90 containing GNEP waste streams a single phase mineralized Cs-mica phase was made by co-reacting illite clay and GNEP simulated waste. The Cs-mica accommodates up to 30% wt% Cs{sub 2}O and all the GNEP waste species, Ba, Sr, Rb including the Cs-137 transmutation to Ba-137. For reference, the cesium mineral pollucite (CsAlSi{sub 2}O{sub 6}), currently being studied for GNEP applications, can only be fabricated at {>=}1000 deg. C. Pollucite mineralization creates secondary aqueous waste streams and NOx. Pollucite is not tolerant of high concentrations of Ba, Sr or Rb and forces the divalent species into different mineral host phases. The pollucite can accommodate up to 33% wt% Cs{sub 2}O. (authors)

  13. FLUIDIZED BED STEAM REFORMING MINERALIZATION FOR HIGH ORGANIC AND NITRATE WASTE STREAMS FOR THE GLOBAL NUCLEAR ENERGY PARTNERSHIP

    SciTech Connect (OSTI)

    Jantzen, C; Michael Williams, M

    2008-01-11

    Waste streams that may be generated by the Global Nuclear Energy Partnership (GNEP) Advanced Energy Initiative may contain significant quantities of organics (0-53 wt%) and/or nitrates (0-56 wt%). Decomposition of high nitrate streams requires reducing conditions, e.g. organic additives such as sugar or coal, to reduce the NO{sub x} in the off-gas to N{sub 2} to meet the Clean Air Act (CAA) standards during processing. Thus, organics will be present during waste form stabilization regardless of which GNEP processes are chosen, e.g. organics in the feed or organics for nitrate destruction. High organic containing wastes cannot be stabilized with the existing HLW Best Developed Available Technology (BDAT) which is HLW vitrification (HLVIT) unless the organics are removed by preprocessing. Alternative waste stabilization processes such as Fluidized Bed Steam Reforming (FBSR) operate at moderate temperatures (650-750 C) compared to vitrification (1150-1300 C). FBSR converts organics to CAA compliant gases, creates no secondary liquid waste streams, and creates a stable mineral waste form that is as durable as glass. For application to the high Cs-137 and Sr-90 containing GNEP waste streams a single phase mineralized Cs-mica phase was made by co-reacting illite clay and GNEP simulated waste. The Cs-mica accommodates up to 30% wt% Cs{sub 2}O and all the GNEP waste species, Ba, Sr, Rb including the Cs-137 transmutation to Ba-137. For reference, the cesium mineral pollucite (CsAlSi{sub 2}O{sub 6}), currently being studied for GNEP applications, can only be fabricated at {ge} 1000 C. Pollucite mineralization creates secondary aqueous waste streams and NO{sub x}. Pollucite is not tolerant of high concentrations of Ba, Sr or Rb and forces the divalent species into different mineral host phases. The pollucite can accommodate up to 33% wt% Cs{sub 2}O.

  14. Development And Initial Testing Of Off-Gas Recycle Liquid From The WTP Low Activity Waste Vitrification Process - 14333

    SciTech Connect (OSTI)

    McCabe, Daniel J.; Wilmarth, William R.; Nash, Charles A.; Taylor-Pashow, Kathryn M.; Adamson, Duane J.; Crawford, Charles L.; Morse, Megan M.

    2014-01-07

    The Waste Treatment and Immobilization Plant (WTP) process flow was designed to pre-treat feed from the Hanford tank farms, separate it into a High Level Waste (HLW) and Low Activity Waste (LAW) fraction and vitrify each fraction in separate facilities. Vitrification of the waste generates an aqueous condensate stream from the off-gas processes. This stream originates from two off-gas treatment unit operations, the Submerged Bed Scrubber (SBS) and the Wet Electrospray Precipitator (WESP). Currently, the baseline plan for disposition of the stream from the LAW melter is to recycle it to the Pretreatment facility where it gets evaporated and processed into the LAW melter again. If the Pretreatment facility is not available, the baseline disposition pathway is not viable. Additionally, some components in the stream are volatile at melter temperatures, thereby accumulating to high concentrations in the scrubbed stream. It would be highly beneficial to divert this stream to an alternate disposition path to alleviate the close-coupled operation of the LAW vitrification and Pretreatment facilities, and to improve long-term throughput and efficiency of the WTP system. In order to determine an alternate disposition path for the LAW SBS/WESP Recycle stream, a range of options are being studied. A simulant of the LAW Off-Gas Condensate was developed, based on the projected composition of this stream, and comparison with pilot-scale testing. The primary radionuclide that vaporizes and accumulates in the stream is Tc-99, but small amounts of several other radionuclides are also projected to be present in this stream. The processes being investigated for managing this stream includes evaporation and radionuclide removal via precipitation and adsorption. During evaporation, it is of interest to investigate the formation of insoluble solids to avoid scaling and plugging of equipment. Key parameters for radionuclide removal include identifying effective precipitation or ion adsorption chemicals, solid-liquid separation methods, and achievable decontamination factors. Results of the radionuclide removal testing indicate that the radionuclides, including Tc-99, can be removed with inorganic sorbents and precipitating agents. Evaporation test results indicate that the simulant can be evaporated to fairly high concentration prior to formation of appreciable solids, but corrosion has not yet been examined.

  15. PJM Controller Testing with Prototypic PJM Nozzle Configuration

    SciTech Connect (OSTI)

    Bontha, Jagannadha R.; Nigl, Franz; Weier, Dennis R.; Leigh, Richard J.; Johnson, Eric D.; Wilcox, Wayne A.; Pfund, David M.; Baumann, Aaron W.; Wang, Yeefoo

    2009-08-21

    The U.S. Department of Energy (DOE) Office of River Protections Waste Treatment Plant (WTP) is being designed and built to pre-treat and then vitrify a large portion of the wastes in Hanfords 177 underground waste storage tanks. The WTP consists of three primary facilitiespretreatment, low-activity waste (LAW) vitrification, and high-level waste (HLW) vitrification. The pretreatment facility will receive waste piped from the Hanford tank farms and separate it into a high-volume, low-activity liquid stream stripped of most solids and radionuclides and a much smaller volume of HLW slurry containing most of the solids and most of the radioactivity. Many of the vessels in the pretreatment facility will contain pulse jet mixers (PJM) that will provide some or all of the mixing in the vessels. Pulse jet mixer technology was selected for use in black cell regions of the WTP, where maintenance cannot be performed once hot testing and operations commence. The PJMs have no moving mechanical parts that require maintenance. The vessels with the most concentrated slurries will also be mixed with air spargers and/or steady jets in addition to the mixing provided by the PJMs. Pulse jet mixers are susceptible to overblows that can generate large hydrodynamic forces, forces that can damage mixing vessels or their internal parts. The probability of an overblow increases if a PJM does not fill completely. The purpose of the testing performed for this report was to determine how reliable and repeatable the primary and safety (or backup) PJM control systems are at detecting drive overblows (DOB) and charge vessel full (CVF) conditions. Testing was performed on the ABB 800xA and Triconex control systems. The controllers operated an array of four PJMs installed in an approximately 13 ft diameter 15 ft tall tank located in the high bay of the Pacific Northwest National Laboratory (PNNL) 336 Building test facility. The PJMs were fitted with 4 inch diameter discharge nozzles representative of the nozzles to be used in the WTP. This work supplemented earlier controller tests done on PJMs with 2 inch nozzles (Bontha et al. 2007). Those earlier tests enabled the selection of appropriate pressure transmitters with associated piping and resulted in an alternate overblow detection algorithm that uses data from pressure transmitters mounted in a water flush line on the PJM airlines. Much of that earlier work was only qualitative, however, due to a data logger equipment failure that occurred during the 2007 testing. The objectives of the current work focused on providing quantitative determinations of the ability of the BNI controllers to detect DOB and CVF conditions. On both control systems, a DOB or CVF is indicated when the values of particular internal functions, called confidence values, cross predetermined thresholds. There are two types of confidence values; one based on a transformation of jet pump pair (JPP) drive and suction pressures, the other based on the pressure in the flush line. In the present testing, we collected confidence levels output from the ABB and Triconex controllers. These data were analyzed in terms of the true and noise confidence peaks generated during multiple cycles of DOB and CVF events. The distributions of peak and noise amplitudes were compared to see if thresholds could be set that would enable the detection of DOB and CVF events at high probabilities, while keeping false detections to low probabilities. Supporting data were also collected on PJM operation, including data on PJM pressures and levels, to provide direct experimental evidence of when PJMs were filling, full, driving, or overblowing.

  16. 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.

  17. DETERMINATION OF THE QUANTITY OF I-135 RELEASED FROM THE AGR-1 TEST FUELS AT THE END OF ATR OPERATING CYCLE 138B

    SciTech Connect (OSTI)

    J. K. Hartwell; D. M. Scates; J. B. Walter; M. W. Drigert

    2007-05-01

    The AGR-1 experiment is a multiple fueled-capsule irradiation experiment being conducted in the Advanced Test Reactor (ATR) in support of the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The experiment began irradiation in the ATR with a cycle that reached full power on December 26, 2006 and ended with shutdown of the reactor for a brief outage on February 10, 2007 at 0900. The AGR-1 experiment will continue cyclical irradiation for about 2.5 years. In order to allow estimation of the amount of radioiodine released during the first cycle, purge gas flow to all capsules continued for about 4 days after reactor shutdown. The FPMS data acquired during part of that shutdown flow period has been analyzed to elucidate the level of 135I released during the operating cycle.

  18. Bench-scale vitrification studies with Savannah River Site mercury contaminated soil

    SciTech Connect (OSTI)

    Cicero, C.A.; Bickford, D.F.

    1995-12-31

    The Savannah River Technology Center (SRTC) has been charted by the Department of Energy (DOE)--Office of Technology Development (OTD) to investigate vitrification technology for the treatment of Low Level Mixed Wastes (LLMW). In fiscal year 1995, mercury containing LLMW streams were targeted. In order to successfully apply vitrification technology to mercury containing LLMW, the types and quantities of glass forming additives necessary for producing homogeneous glasses from the wastes have to be determined and the treatment for the mercury portion must also be determined. Selected additives should ensure that a durable and leach resistant waste form is produced, while the mercury treatment should ensure that hazardous amounts of mercury are not released into the environment. The mercury containing LLMW selected for vitrification studies at the SRTC was mercury contaminated soil from the TNX pilot-plant facility at the Savannah River Site (SRS). Samples of this soil were obtained so bench-scale vitrification studies could be performed at the SRTC to determine the optimum waste loading obtainable in the glass product without sacrificing durability and leach resistance. Vitrifying this waste stream also required offgas treatment for the capture of the vaporized mercury.

  19. Support for the in situ vitrification treatability study at the Idaho National Engineering Laboratory: FY 1988 summary

    SciTech Connect (OSTI)

    Oma, K.H.; Reimus, M.A.H.; Timmerman, C.L.

    1989-02-01

    The objective of this project is to determine if in situ vitrification (ISV) is a viable, long-term confinement technology for previously buried solid transuranic and mixed waste at the Radioactive Waste Management Complex (RWMC). The RWMC is located at the Idaho National Engineering Laboratory (INEL). In situ vitrification is a thermal treatment process that converts contaminated soils and wastes into a durable glass and crystalline form. During processing, heavy metals or other inorganic constituents are retained and immobilized in the glass structure, and organic constituents are typically destroyed or removed for capture by an off-gas treatment system. The primary FY 1988 activities included engineering-scale feasibility tests on INEL soils containing a high metals loading. Results of engineering-scale testing indicate that wastes with a high metals content can be successfully processed by ISV. The process successfully vitrified soils containing localized metal concentrations as high as 42 wt % without requiring special methods to prevent electrical shorting within the melt zone. Vitrification of this localized concentration resulted in a 15.9 wt % metals content in the entire ISV test block. This ISV metals limit is related to the quantity of metal that accumulates at the bottom of the molten glass zone. Intermediate pilot-scale testing is recommended to determine metals content scale-up parameters in order to project metals content limits for large-scale ISV operation at INEL.

  20. Microsoft Word - RPP-PLAN-47325 Rev 0.docx

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

    Regulations CMS IP Corrective Measures Study Implementation Plan DOE U.S. Department of Energy FR Federal Register HFFACO Hanford Federal Facility Agreement and Consent Order HLW...

  1. Action Item Review and Status

    Office of Environmental Management (EM)

    Board Action Items Action Item Resolution Action Item Strategic Planning Initiative Optimization Study Resolution Presentation by S. Schneider (HLW System Integrated Project...

  2. Management Overview

    Office of Environmental Management (EM)

    (FEPs) Analysis * Scenario Development * PA Model 4 ... HLW resulting from atomic energy defense activities is ... Energy Agency, Vienna, Austria. NWPA (Nuclear Waste ...

  3. EIS-0287: Notice of Preferred Sodium Bearing Waste Treatment...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Preferred Sodium Bearing Waste Treatment Technology EIS-0287: Notice of Preferred Sodium Bearing Waste Treatment Technology Idaho High-Level Waste (HLW) and Facilities Disposition...

  4. Enterprise Assessments Operational Awareness Record, Waste Treatment...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Observation of Waste Treatment and Immobilization Plant High Level Waste Facility Radioactive Liquid Waste Disposal System Hazards Analysis Activities (EA-WTP-HLW-2014-08-18(a))...

  5. IMPACT OF PARTICLE AGGLOMERATION ON ACCUMULATION RATES IN THE...

    Office of Scientific and Technical Information (OSTI)

    IMPACT OF PARTICLE AGGLOMERATION ON ACCUMULATION RATES IN THE GLASS DISCHARGE RISER OF HLW MELTER Citation Details In-Document Search Title: IMPACT OF PARTICLE AGGLOMERATION ON ...

  6. Hanford ETR Tank Waste Treatment and Immobilization Plant - Hanford...

    Office of Environmental Management (EM)

    to include the foremost experts from the chemical processing industry, the glass industry, ... process system GFC glass-forming chemical HEPA high-efficiency particulate air HLW ...

  7. Supplemnental Volume - Independent Oversight Assessment of the...

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

    ... Training HLW High Level Waste HPA Human ... Integrated Safety Management System LAW Low Activity Waste NCR Nonconformance ... appraisals, safety criteria in work activities, ...

  8. Tank Closure and Waste Management Environmental Impact Statement...

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

    ... and storage; and waste management and disposal. ... tank farms into high-level radioactive waste (HLW) and low-activity waste (LAW) ... This section primarily discusses criteria and ...

  9. Independent Oversight Review of the Hanford Site Waste Treatment...

    Office of Environmental Management (EM)

    and Emergency Management Evaluations Office of ... Criteria, Review and Approach Document DOE U.S. Department of Energy FWCL Field Welding Checklist HLW High-Level Waste HSS ...

  10. Office of Enterprise Assessments Review of the West Valley Demonostrat...

    Energy Savers [EERE]

    ... West Valley, LLC CRAD Criteria and Review Approach ... EMCBC Environmental Management Consolidated Business ... Head End Vent HLW High Level Waste I&C Instrumentation and ...

  11. Office of Enterprse Assessments Review of the West Valley Demonstratio...

    Energy Savers [EERE]

    ... LLC Ci Curies CRAD Criteria, Review, and Approach ... Guide EMIP Emergency Management Implementing Procedure ... Material HLW High Level Waste HSS Office of Health, ...

  12. Enterprise Assessments Review of the Hanford Site Waste Treatment...

    Office of Environmental Management (EM)

    ... Management CRAD Criteria, Review and ... Document Management System EIA ... High-Level Waste HMH HLW Melter Handling ITS Important to Safety LAW Low-Activity ...

  13. Microsoft Word - M-2 WTP Contract Section J - Conformed Thru...

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

    DFLAW Design Completion Criteria Incentive Definitions ... HLW High-Level Waste HUBZone Historically Underutilized Business Zone HWMA Hazardous Waste Management Act ICD ...

  14. Information Request Yucca Mountain Site

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    ... ".... requiring both engineering and geology to contribute to isolation can be used to ... suggested exclusive reliance on either geology or engineering for the isolation of HLW." ...

  15. Seismic design evaluation guidelines for buried piping for the...

    Office of Scientific and Technical Information (OSTI)

    Seismic design evaluation guidelines for buried piping for the DOE HLW Facilities Citation Details In-Document Search Title: Seismic design evaluation guidelines for buried piping ...

  16. Coupled Model for Heat and Water Transport in a High Level Waste Repository in Salt

    Broader source: Energy.gov [DOE]

    This report summarizes efforts to simulate coupled thermal-hydrological-chemical (THC) processes occurring within a generic hypothetical high-level waste (HLW) repository in bedded salt.

  17. I:'"

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

    -_ / 1 (',\ ! / I ( ~' ' / I:'" 7 ( /, -. \ ( , /.- .I. ) ( '" / ' i, I - / I HLW System Plan - Revision 0 (U) Table of Contents Page Mission Statement 4 Executive Summary 4 HLW/DWPF System Description 4 1.0 Principles/Guidelines 5 1.1 Regulatory Guidelines 5 1.2 DOE Guidance/AOP/FYP 6 1.3 Process Considerations 6 1.4 Waste Removal Sequence 7 2.0 Assumptions 2.1 Facility Startup and Operation 8 2.1.1 DWPF 8 2.1.2 HLW Processing Facilities 8 2.1.3 HLW Tank Space Gain 9 2.1.4 Canyon

  18. Waste Processing Annual Technology Development Report 2007 |...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    More Documents & Publications System Planning for Low-Activity Waste at Hanford Waste Treatment and Immobilation Plant HLW Waste Vitrification Facility Caustic Recovery Technology

  19. Technology Readiness Assessments | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    August 1, 2007 Waste Treatment and Immobilation Plant HLW Waste Vitrification Facility ... (BOF) and Low-Activity Waste Vitrification Facilities (LAW) Full Document and ...

  20. Physical Modeling of Spinel Crystals Settling at Low Reynolds...

    Office of Scientific and Technical Information (OSTI)

    during the high-level waste (HLW) vitrification process poses a potential danger to ... SLUDGES; SPINELS; SURFACE AREA; VITRIFICATION; WASTES Settling; Spinel; High-level ...

  1. EIS-0303: Record of Decision | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    waste (HLW) tanks and associated equipment such as evaporator systems, transfer pipelines, diversion boxes, and pump pits. DOE has selected the preferred alternative...

  2. Review of the Hanford Site Waste Treatment and Immobilization...

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

    ... of quality control (QC) tests performed on samples of concrete placed in the HLW facility. ... which maintain the 177 underground storage tanks; and the WTP, which is responsible ...

  3. Report on Separate Disposal of Defense High- Level Radioactive...

    Office of Environmental Management (EM)

    Radioactive Waste March 2015 This page left blank. i EXECUTIVE SUMMARY Purpose This report considers whether a separate repository for high-level radioactive waste (HLW) ...

  4. 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.

  5. EIS-0356: Retrieval, Treatment and Disposal of Tank Wastes and Closure of Single-Shell Tanks at the Hanford Site, Richland, WA

    Broader source: Energy.gov [DOE]

    This EIS analyzes DOE's proposed retrieval, treatment, and disposal of the waste being managed in the high-level waste (HLW) tank farms at the Hanford Site near Richland, Washington, and closure of the 149 single-shell tanks (SSTs) and associated facilities in the HLW tank farms.

  6. Examining the role of canister cooling conditions on the formation of nepheline from nuclear waste glasses

    SciTech Connect (OSTI)

    Christian, J. H.

    2015-09-01

    Nepheline (NaAlSiO₄) crystals can form during slow cooling of high-level waste (HLW) glass after it has been poured into a waste canister. Formation of these crystals can adversely affect the chemical durability of the glass. The tendency for nepheline crystallization to form in a HLW glass increases with increasing concentrations of Al₂O₃ and Na₂O.

  7. Text of Plan

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

    1 April 2000 An Integrated System at the Savannah River Site Converting Waste to Glass Converting Waste to Glass HLW-2000-00019 HLW-2000-00019 High Level Waste System Plan Revision 11 Page i Table of Contents TABLE OF CONTENTS ............................................................................................................... I EXECUTIVE SUMMARY............................................................................................................ 1 COMPARISON: REQUIREMENTS CASE

  8. Examining the role of canister cooling conditions on the formation of nepheline from nuclear waste glasses

    SciTech Connect (OSTI)

    Christian, J. H.

    2015-09-01

    Nepheline (NaAlSiO?) crystals can form during slow cooling of high-level waste (HLW) glass after it has been poured into a waste canister. Formation of these crystals can adversely affect the chemical durability of the glass. The tendency for nepheline crystallization to form in a HLW glass increases with increasing concentrations of Al?O? and Na?O.

  9. Pilot-Scale Testing of In Situ Vitrification of Arnold Engineering Development Center Site 10 Contaminated Soils

    SciTech Connect (OSTI)

    Timmerman, C. L.; Peterson, M. E.

    1990-02-01

    Process verification testing using in situ vitrification (ISV) was successfully performed in a pilot-scale test using soils containing fuel oils and heavy metals from Site 10 Installation Restoration Program (IRP) at the Arnold Engineering Development Center (AEDC) located in the southern portion of middle Tennessee. This effort was directed through the U.S. Department of Energy ' s Hazardous Waste Remedial Action Program (HAZWRAP) Office managed by Martin Marietta Energy Systems. In situ vitrification is a thermal treatment process that converts contaminated soils and wastes into a durable product containing glass and crystalline phases. During processing, heavy metals or other inorganic constituents are retained and immobilized in the glass structure; organic constituents are typically destroyed or removed and captured by the off-gas treatment system. The objective of this test is to verify the applicability of the ISV process for stabilization of the contaminated soil at Site 10 . The pilotscale ISV testing results, reported herein, indicate that the AEDC Site 10 Fire Training Area may be successfully processed by ISV. Site 10 is a fire training pit that is contaminated with fuel oils and heavy metals from fire training exercises. Actual site material was processed by ISV to verify its feasible application to those soils . Initial feasibility bench-scale testing and analyses of the soils determined that a lower-melting, electrically conductive fluxing additive (such as sodium carbonate) is required as an additive to the soil for ISV processing to work effecti vely. The actual Site 10 soils showed a larger degree of compositional variation than the soil used for the bench-scale test . This variation dictates that each vitrification setting should be analyzed to determine the composition as. a function of depth and location . This data will dictate the amount (if any) of fluxing add itives of sodium and calci um to bring the melt composition to the recommended quantity of 5 wt% sodium and 5 wt% calcium oxide. Each variable additive adjustment would result in a vitrification melt composition of 5 wt% calcium and sodium oxide content . The pilot -scale operation created a vitrified block weighing 15 metric t onnes (t) and measuring 1.5 m (5 ft) deep and 2.4 m (8 ft) on each side. The quantity of fluxing additives and the method of placing the fluxing additives in the surface cover soil limited the operating electrical system providing power to the ISV melt. The power limitation created enhanced lateral growth of the block and resulted in a shallower depth . This method of adding fluxes demonstrated that ISV operating efficiency would be greatly improved if the fluxes were injected or mixed with the entire designated vitrification volume. However, the volume vitrified contained a sufficient quantity of hazardous contaminants to allow for an effective verification evaluation of ISV processing of the AEDC Site 10. Analytical efforts for this project were directed towards evaluating the organic destruction and thermal transport effects of ISV processing on the Site 10 contaminated soil. No thermal transport of hydrocarbon contaminants to the surrounding soil were detected. These results continue to confirm the organic destruction and nontransport mechanisms presented in this report . Off-gas releases of the hydrocarbons indicated an 89 wt% destruction efficiency by the ISV process exclusive of off-gas treatment. The destruction and removal efficiency of the overall ISV system was 99.85 wt%. Leach testing using extraction procedure (EP) toxicity and toxic characteristics leach procedure (TCLP) showed that all metals of concern were below leach testing release limits, indicating that the ISV process produces a nonhazardous product . These favorable results indicate that ISV can be used to effectively treat and remediate the contaminated soils at the AEDC Site 10 .

  10. Characteristics of potential repository wastes. Volume 2

    SciTech Connect (OSTI)

    Not Available

    1992-07-01

    The LWR spent fuels discussed in Volume 1 of this report comprise about 99% of all domestic non-reprocessed spent fuel. In this report we discuss other types of spent fuels which, although small in relative quantity, consist of a number of diverse types, sizes, and compositions. Many of these fuels are candidates for repository disposal. Some non-LWR spent fuels are currently reprocessed or are scheduled for reprocessing in DOE facilities at the Savannah River Site, Hanford Site, and the Idaho National Engineering Laboratory. It appears likely that the reprocessing of fuels that have been reprocessed in the past will continue and that the resulting high-level wastes will become part of defense HLW. However, it is not entirely clear in some cases whether a given fuel will be reprocessed, especially in cases where pretreatment may be needed before reprocessing, or where the enrichment is not high enough to make reprocessing attractive. Some fuels may be canistered, while others may require special means of disposal. The major categories covered in this chapter include HTGR spent fuel from the Fort St. Vrain and Peach Bottom-1 reactors, research and test reactor fuels, and miscellaneous fuels, and wastes generated from the decommissioning of facilities.

  11. An economic analysis of a light and heavy water moderated reactor synergy: burning americium using recycled uranium

    SciTech Connect (OSTI)

    Wojtaszek, D.; Edwards, G.

    2013-07-01

    An economic analysis is presented for a proposed synergistic system between 2 nuclear utilities, one operating light water reactors (LWR) and another running a fleet of heavy water moderated reactors (HWR). Americium is partitioned from LWR spent nuclear fuel (SNF) to be transmuted in HWRs, with a consequent averted disposal cost to the LWR operator. In return, reprocessed uranium (RU) is supplied to the HWRs in sufficient quantities to support their operation both as power generators and americium burners. Two simplifying assumptions have been made. First, the economic value of RU is a linear function of the cost of fresh natural uranium (NU), and secondly, plutonium recycling for a third utility running a mixed oxide (MOX) fuelled reactor fleet has been already taking place, so that the extra cost of americium recycling is manageable. We conclude that, in order for this scenario to be economically attractive to the LWR operator, the averted disposal cost due to partitioning americium from LWR spent fuel must exceed 214 dollars per kg, comparable to estimates of the permanent disposal cost of the high level waste (HLW) from reprocessing spent LWR fuel. (authors)

  12. HIGH-LEVEL WASTE GLASS FORMULATION MODEL SENSITIVITY STUDY 2009 GLASS FORMULATION MODEL VERSUS 1996 GLASS FORMULATION MODEL

    SciTech Connect (OSTI)

    BELSHER JD; MEINERT FL

    2009-12-07

    This document presents the differences between two HLW glass formulation models (GFM): The 1996 GFM and 2009 GFM. A glass formulation model is a collection of glass property correlations and associated limits, as well as model validity and solubility constraints; it uses the pretreated HLW feed composition to predict the amount and composition of glass forming additives necessary to produce acceptable HLW glass. The 2009 GFM presented in this report was constructed as a nonlinear optimization calculation based on updated glass property data and solubility limits described in PNNL-18501 (2009). Key mission drivers such as the total mass of HLW glass and waste oxide loading are compared between the two glass formulation models. In addition, a sensitivity study was performed within the 2009 GFM to determine the effect of relaxing various constraints on the predicted mass of the HLW glass.

  13. Assessment of Disposal Options for DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel

    Broader source: Energy.gov [DOE]

    The Assessment of Disposal Options for DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel report assesses the technical options for the safe and permanent disposal of high-level radioactive waste (HLW) and spent nuclear fuel (SNF) managed by the Department of Energy. Specifically, it considers whether DOE-managed HLW and SNF should be disposed of with commercial SNF and HLW in one geologic repository or whether there are advantages to developing separate geologic disposal pathways for some DOE-managed HLW and SNF. The report recommends that the Department begin implementation of a phased, adaptive, and consent-based strategy with development of a separate mined repository for some DOE-managed HLW and cooler DOE-managed SNF.

  14. Written Statement of Mark Whitney Acting Assistant Secretary...

    Energy Savers [EERE]

    4,000 canisters of vitrified high-level waste and closed six of the site's underground storage tanks. At 1 our Portsmouth, Ohio, and Paducah, Kentucky, sites, we have designed,...

  15. EIS-0220-SA-01: Supplement Analysis | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    impacts of stabilizing LAP material be first processing it through the HB-Line and H-Canyon, and then vitrifying the resulting solution in DWPF. DOE finds that the impact...

  16. STATISTICAL EVALUATION OF SMALL SCALE MIXING DEMONSTRATION SAMPLING AND BATCH TRANSFER PERFORMANCE - 12093

    SciTech Connect (OSTI)

    GREER DA; THIEN MG

    2012-01-12

    The ability to effectively mix, sample, certify, and deliver consistent batches of High Level Waste (HLW) feed from the Hanford Double Shell Tanks (DST) to the Waste Treatment and Immobilization Plant (WTP) presents a significant mission risk with potential to impact mission length and the quantity of HLW glass produced. DOE's Tank Operations Contractor, Washington River Protection Solutions (WRPS) has previously presented the results of mixing performance in two different sizes of small scale DSTs to support scale up estimates of full scale DST mixing performance. Currently, sufficient sampling of DSTs is one of the largest programmatic risks that could prevent timely delivery of high level waste to the WTP. WRPS has performed small scale mixing and sampling demonstrations to study the ability to sufficiently sample the tanks. The statistical evaluation of the demonstration results which lead to the conclusion that the two scales of small DST are behaving similarly and that full scale performance is predictable will be presented. This work is essential to reduce the risk of requiring a new dedicated feed sampling facility and will guide future optimization work to ensure the waste feed delivery mission will be accomplished successfully. This paper will focus on the analytical data collected from mixing, sampling, and batch transfer testing from the small scale mixing demonstration tanks and how those data are being interpreted to begin to understand the relationship between samples taken prior to transfer and samples from the subsequent batches transferred. An overview of the types of data collected and examples of typical raw data will be provided. The paper will then discuss the processing and manipulation of the data which is necessary to begin evaluating sampling and batch transfer performance. This discussion will also include the evaluation of the analytical measurement capability with regard to the simulant material used in the demonstration tests. The paper will conclude with a discussion of the analysis results illustrating the relationship between the pre-transfer samples and the batch transfers, which support the recommendation regarding the need for a dedicated feed sampling facility.

  17. Economic Feasibility of Electrochemical Caustic Recycling at the Hanford Site

    SciTech Connect (OSTI)

    Poloski, Adam P.; Kurath, Dean E.; Holton, Langdon K.; Sevigny, Gary J.; Fountain, Matthew S.

    2009-03-01

    This report contains a review of potential cost benefits of NaSICON Ceramic membranes for the separation of sodium from Hanford tank waste. The primary application is for caustic recycle to the Waste Treatment and Immobilization Plant (WTP) pretreatment leaching operation. The report includes a description of the waste, the benefits and costs for a caustic-recycle facility, and Monte Carlo results obtained from a model of these costs and benefits. The use of existing cost information has been limited to publicly available sources. This study is intended to be an initial evaluation of the economic feasibility of a caustic recycle facility based on NaSICON technology. The current pretreatment flowsheet indicates that approximately 6,500 metric tons (MT) of Na will be added to the tank waste, primarily for removing Al from the high-level waste (HLW) sludge (Kirkbride et al. 2007). An assessment (Alexander et al. 2004) of the pretreatment flowsheet, equilibrium chemistry, and laboratory results indicates that the quantity of Na required for sludge leaching will increase by 6,000 to 12,000 MT in order to dissolve sufficient Al from the tank-waste sludge material to maintain the number of HLW canisters produced at 9,400 canisters as defined in the Office of River Protection (ORP) System Plan (Certa 2003). This additional Na will significantly increase the volume of LAW glass and extend the processing time of the Waste Treatment and Immobilization Plant (WTP). Future estimates on sodium requirements for caustic leaching are expected to significantly exceed the 12,000-MT value and approach 40,000-MT of total sodium addition for leaching (Gilbert, 2007). The cost benefit for caustic recycling is assumed to consist of four major contributions: 1) the cost savings realized by not producing additional immobilized low-activity waste (ILAW) glass, 2) caustic recycle capital investment, 3) caustic recycle operating and maintenance costs, and 4) research and technology costs needed to deploy the technology. In estimating costs for each of these components, several parameters are used as inputs. Due to uncertainty in assuming a singular value for each of these parameters, a range of possible values is assumed. A Monte Carlo simulation is then performed where the range of these parameters is exercised, and the resulting range of cost benefits is determined.

  18. Glass Science Could Boost Hanford Cleanup | Department of Energy

    Energy Savers [EERE]

    Glass Science Could Boost Hanford Cleanup Glass Science Could Boost Hanford Cleanup October 29, 2015 - 12:30pm Addthis A canister filled with nonradioactive glass sits on display. A canister filled with nonradioactive glass sits on display. A sample of vitrified glass at EM's Office of River Protection. A sample of vitrified glass at EM's Office of River Protection. A glass scientist works with molten glass. A glass scientist works with molten glass. A canister filled with nonradioactive glass

  19. 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.

  20. 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.

  1. Office of River Protection (ORP) Mission Completion Strategy

    SciTech Connect (OSTI)

    WIEGMAN, S.A.

    2002-02-24

    DOE's Office of River Protection (ORP) is readying itself to commence construction of a Waste Treatment Plant (WTP) that will start the process of turning Hanford tank waste into glass. The plant is state-of-the art and includes reasonable flexibility to improve operations as technology and operational understandings improve. During its 40 year design life the plant has the capability to treat half of the total volume of tank waste and reduce risk to the public by up to ninety percent. Looking beyond initial processing towards the project end state, however, it is apparent that ORP's baseline approach is part of the issue raised by the DOE Secretary when he said that $300 billion and 75 years is too costly and too long for DOE's environmental cleanups. ORP has reviewed its cost and schedule drivers and has started identifying areas where better technologies and risk-based strategies could substantially decrease its life cycle cost and schedule. Specific technologies under consideration will be discussed along with expected return on investment. ORP is totally committed to taking all steps necessary during cleanup to protect human health and the environment and to comply with appropriate regulations and commitments. But, ORP is also very conscious of the fact that the history of Hanford production and tank farm operations has resulted in very large tank-to-tank variabilities in the waste constituents. Not all tank wastes demand the same high level of rigor in treatment as provided by the WTP in order to protect people and the environment. Parallel treatment paths, keyed to the hazards and chemical challenges each tank presents, need to be developed. The WTP vitrification capabilities should be deployed for the higher risk wastes that require vitrification. By getting wastes in the proper paths for treatment based upon their chemical characteristics and inherent risks, ORP will be able to both accelerate the cleanup schedule and bring its life cycle and annual funding requirements into line. The WTP needs to be managed and its throughput enhanced to vitrify all of the HLW and approximately 50% of the low-level tank waste by about 2030. That represents the lion's share of the current and long-term risk presented by the tanks. For much of the low activity waste currently in the tanks, parallel treatment technologies are required that protect people and the environment but require less time and less cost than the total vitrification option presents. Any such technologies that ORP deploys must have sound, defensible bases with the prerequisite QA pedigrees. Providing parallel paths for lower risk wastes will allow ORP to avoid the 20-30 year treatment schedule that lower risk tanks would otherwise face. Potential parallel paths will be described. ORP also needs to deploy and test technologies to demonstrate that its tank farms can be successfully closed. Starting such a demonstration during the period while the plant is under construction will allow ORP to start developing critical data that it will need for permanent closures at a later date. It will take many years of testing such demonstration activities and monitoring to develop confidence in tank closure approaches. If ORP starts such an effort, practicing on smaller, more benign tanks, it will reap significant institutional benefits in the near-term and have far better information when it is ready to start to close entire tank farms in the future.

  2. Feasibility Evaluation and Retrofit Plan for Cold Crucible Induction Melter Deployment in the Defense Waste Processing Facility at Savannah River Site

    SciTech Connect (OSTI)

    Barnes, A.B. [Savannah River National Laboratory, Washington Savannah River Company, Aiken, SC (United States); Iverson, D.C.; Adkins, B.J. [Liquid Waste Operations, Washington Savannah River Company, Aiken, SC (United States); Tchemitcheff, E. [AREVA NC Inc., Richland Office, Richland, WA (United States)

    2008-07-01

    Cold crucible induction melters (CCIM) have been proposed as an alternative technology for waste glass melting at the Defense Waste Processing Facility (DWPF) at Savannah River Site (SRS) as well as for other waste vitrification facilities. Proponents of this technology cite high temperature operation, high tolerance for noble metals and aluminum, high waste loading, high throughput capacity, and low equipment cost as the advantages over existing Joule Heated Melter (JHM) technology. The CCIM uses induction heating to maintain molten glass at high temperature. A water-cooled helical induction coil is connected to an AC current supply, typically operating at frequencies from 100 kHz to 5 MHz. The oscillating magnetic field generated by the oscillating current flow through the coil induces eddy currents in conductive materials within the coil. Those oscillating eddy currents, in turn, generate heat in the material. In the CCIM, the induction coil surrounds a 'Cold Crucible' which is formed by metal tubes, typically copper or stainless steel. The tubes are constructed such that the magnetic field does not couple with the crucible. Therefore, the field generated by the induction coil couples primarily with the conductive medium (hot glass) within. The crucible tubes are water cooled to maintain their temperature between 100 deg. C to 200 deg. C so that a protective layer of molten glass and/or batch material, referred to as a 'skull', forms between them and the hot, corrosive melt. Because the protective skull is the only material directly in contact with the molten glass, the CCIM doesn't have the temperature limitations of traditional refractory lined JHM. It can be operated at melt temperatures in excess of 2000 deg. C, allowing processing of high waste loading batches and difficult-to-melt compounds. The CCIM is poured through a bottom drain, typically through a water-cooled slide valve that starts and stops the pour stream. To promote uniform temperature distribution and increase heat transfer to the slurry fed High Level Waste (HLW) sludge, the CCIM may be equipped with bubblers and/or water cooled mechanical agitators. The DWPF could benefit from use of CCIM technology, especially in light of our latest projections of waste volume to be vitrified. Increased waste loading and increased throughput could result in substantial life cycle cost reduction. In order to significantly surpass the waste throughput capability of the currently installed JHM, it may be necessary to install two 950 mm CCIMs in the DWPF Melt Cell. A cursory evaluation of system design requirements and modifications to the facility that may be required to support installation and operation of two 950 mm CCIMs was performed. Based on this evaluation, it appears technically feasible to position two CCIMs in the Melt Cell of the DWPF within the existing footprint of the current melter. Interfaces with support systems and controls including Melter Feed, Power, Melter Cooling Water, Melter Off-gas, and Canister Operations must be designed to support dual CCIM operations. This paper describes the CCIM technology and identifies technical challenges that must be addressed in order to implement CCIMs in the DWPF. (authors)

  3. FEASIBILITY EVALUATION AND RETROFIT PLAN FOR COLD CRUCIBLE INDUCTION MELTER DEPLOYMENT IN THE DEFENSE WASTE PROCESSING FACILITY AT SAVANNAH RIVER SITE 8118

    SciTech Connect (OSTI)

    Barnes, A; Dan Iverson, D; Brannen Adkins, B

    2008-02-06

    Cold crucible induction melters (CCIM) have been proposed as an alternative technology for waste glass melting at the Defense Waste Processing Facility (DWPF) at Savannah River Site (SRS) as well as for other waste vitrification facilities. Proponents of this technology cite high temperature operation, high tolerance for noble metals and aluminum, high waste loading, high throughput capacity, and low equipment cost as the advantages over existing Joule Heated Melter (JHM) technology. The CCIM uses induction heating to maintain molten glass at high temperature. A water-cooled helical induction coil is connected to an AC current supply, typically operating at frequencies from 100 KHz to 5 MHz. The oscillating magnetic field generated by the oscillating current flow through the coil induces eddy currents in conductive materials within the coil. Those oscillating eddy currents, in turn, generate heat in the material. In the CCIM, the induction coil surrounds a 'Cold Crucible' which is formed by metal tubes, typically copper or stainless steel. The tubes are constructed such that the magnetic field does not couple with the crucible. Therefore, the field generated by the induction coil couples primarily with the conductive medium (hot glass) within. The crucible tubes are water cooled to maintain their temperature between 100 C to 200 C so that a protective layer of molten glass and/or batch material, referred to as a 'skull', forms between them and the hot, corrosive melt. Because the protective skull is the only material directly in contact with the molten glass, the CCIM doesn't have the temperature limitations of traditional refractory lined JHM. It can be operated at melt temperatures in excess of 2000 C, allowing processing of high waste loading batches and difficult-to-melt compounds. The CCIM is poured through a bottom drain, typically through a water-cooled slide valve that starts and stops the pour stream. To promote uniform temperature distribution and increase heat transfer to the slurry fed High Level Waste (HLW) sludge, the CCIM may be equipped with bubblers and/or water cooled mechanical agitators. The DWPF could benefit from use of CCIM technology, especially in light of our latest projections of waste volume to be vitrified. Increased waste loading and increased throughput could result in substantial life cycle cost reduction. In order to significantly surpass the waste throughput capability of the currently installed JHM, it may be necessary to install two 950 mm CCIMs in the DWPF Melt Cell. A cursory evaluation of system design requirements and modifications to the facility that may be required to support installation and operation of two 950 mm CCIMs was performed. Based on this evaluation, it appears technically feasible to position two CCIMs in the Melt Cell of the DWPF within the existing footprint of the current melter. Interfaces with support systems and controls including Melter Feed, Power, Melter Cooling Water, Melter Off-gas, and Canister Operations must be designed to support dual CCIM operations. This paper describes the CCIM technology and identifies technical challenges that must be addressed in order to implement CCIMs in the DWPF.

  4. RADIOACTIVE DEMONSTRATION OF FINAL MINERALIZED WASTE FORMS FOR HANFORD WASTE TREATMENT PLANT SECONDARY WASTE BY FLUIDIZED BED STEAM REFORMING USING THE BENCH SCALE REFORMER PLATFORM

    SciTech Connect (OSTI)

    Crawford, C.; Burket, P.; Cozzi, A.; Daniel, W.; Jantzen, C.; Missimer, D.

    2012-02-02

    The U.S. Department of Energy's Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford's tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. In addition, the WTP LAW vitrification facility off-gas condensate known as WTP Secondary Waste (WTP-SW) will be generated and enriched in volatile components such as {sup 137}Cs, {sup 129}I, {sup 99}Tc, Cl, F, and SO{sub 4} that volatilize at the vitrification temperature of 1150 C in the absence of a continuous cold cap (that could minimize volatilization). The current waste disposal path for the WTP-SW is to process it through the Effluent Treatment Facility (ETF). Fluidized Bed Steam Reforming (FBSR) is being considered for immobilization of the ETF concentrate that would be generated by processing the WTP-SW. The focus of this current report is the WTP-SW. FBSR offers a moderate temperature (700-750 C) continuous method by which WTP-SW wastes can be processed irrespective of whether they contain organics, nitrates, sulfates/sulfides, chlorides, fluorides, volatile radionuclides or other aqueous components. The FBSR technology can process these wastes into a crystalline ceramic (mineral) waste form. The mineral waste form that is produced by co-processing waste with kaolin clay in an FBSR process has been shown to be as durable as LAW glass. Monolithing of the granular FBSR product is being investigated to prevent dispersion during transport or burial/storage, but is not necessary for performance. A Benchscale Steam Reformer (BSR) was designed and constructed at the SRNL to treat actual radioactive wastes to confirm the findings of the non-radioactive FBSR pilot scale tests and to qualify the waste form for applications at Hanford. BSR testing with WTP SW waste surrogates and associated analytical analyses and tests of granular products (GP) and monoliths began in the Fall of 2009, and then was continued from the Fall of 2010 through the Spring of 2011. Radioactive testing commenced in 2010 with a demonstration of Hanford's WTP-SW where Savannah River Site (SRS) High Level Waste (HLW) secondary waste from the Defense Waste Processing Facility (DWPF) was shimmed with a mixture of {sup 125/129}I and {sup 99}Tc to chemically resemble WTP-SW. Prior to these radioactive feed tests, non-radioactive simulants were also processed. Ninety six grams of radioactive granular product were made for testing and comparison to the non-radioactive pilot scale tests. The same mineral phases were found in the radioactive and non-radioactive testing.

  5. LOW ACTIVITY WASTE FEED SOLIDS CARACTERIZATION AND FILTERABILITY TESTS

    SciTech Connect (OSTI)

    McCabe, D.; Crawford, C.; Duignan, M.; Williams, M.; Burket, P.

    2014-04-03

    The primary treatment of the tank waste at the DOE Hanford site will be done in the Waste Treatment and Immobilization Plant (WTP) that is currently under construction. The baseline plan for the WTP Pretreatment facility is to treat the waste, splitting it into High Level Waste (HLW) feed and Low Activity Waste (LAW) feed. Both waste streams are then separately vitrified as glass and sealed in canisters. The LAW glass will be disposed onsite in the Integrated Disposal Facility (IDF). There are currently no plans to treat the waste to remove technetium in the WTP Pretreatment facility, so its disposition path is the LAW glass. Options are being explored to immobilize the LAW portion of the tank waste, i.e., the LAW feed from the WTP Pretreatment facility. Removal of {sup 99}Tc from the LAW Feed, followed by off-site disposal of the {sup 99}Tc, would eliminate a key risk contributor for the IDF Performance Assessment (PA) for supplemental waste forms, and has potential to reduce treatment and disposal costs. Washington River Protection Solutions (WRPS) is developing some conceptual flow sheets for LAW treatment and disposal that could benefit from technetium removal. One of these flowsheets will specifically examine removing {sup 99}Tc from the LAW feed stream to supplemental immobilization. The conceptual flow sheet of the {sup 99}Tc removal process includes a filter to remove insoluble solids prior to processing the stream in an ion exchange column, but the characteristics and behavior of the liquid and solid phases has not previously been investigated. This report contains results of testing of a simulant that represents the projected composition of the feed to the Supplemental LAW process. This feed composition is not identical to the aqueous tank waste fed to the Waste Treatment Plant because it has been processed through WTP Pretreatment facility and therefore contains internal changes and recycle streams that will be generated within the WTP process. Although a Supplemental LAW feed simulant has previously been prepared, this feed composition differs from that simulant because those tests examined only the fully soluble aqueous solution at room temperature, not the composition formed after evaporation, including the insoluble solids that precipitate after it cools. The conceptual flow sheet for Supplemental LAW immobilization has an option for removal of {sup 99}Tc from the feed stream, if needed. Elutable ion exchange has been selected for that process. If implemented, the stream would need filtration to remove the insoluble solids prior to processing in an ion exchange column. The characteristics, chemical speciation, physical properties, and filterability of the solids are important to judge the feasibility of the concept, and to estimate the size and cost of a facility. The insoluble solids formed during these tests were primarily natrophosphate, natroxalate, and a sodium aluminosilicate compound. At the elevated temperature and 8 M [Na+], appreciable insoluble solids (1.39 wt%) were present. Cooling to room temperature and dilution of the slurry from 8 M to 5 M [Na+] resulted in a slurry containing 0.8 wt% insoluble solids. The solids (natrophosphate, natroxalate, sodium aluminum silicate, and a hydrated sodium phosphate) were relatively stable and settled quickly. Filtration rates were in the range of those observed with iron-based simulated Hanford tank sludge simulants, e.g., 6 M [Na+] Hanford tank 241-AN-102, even though their chemical speciation is considerably different. Chemical cleaning of the crossflow filter was readily accomplished with acid. As this simulant formulation was based on an average composition of a wide range of feeds using an integrated computer model, this exact composition may never be observed. But the test conditions were selected to enable comparison to the model to enable improving its chemical prediction capability.

  6. RADIOACTIVE DEMONSTRATIONS OF FLUIDIZED BED STEAM REFORMING AS A SUPPLEMENTARY TREATMENT FOR HANFORD'S LOW ACTIVITY WASTE AND SECONDARY WASTES

    SciTech Connect (OSTI)

    Jantzen, C.; Crawford, C.; Cozzi, A.; Bannochie, C.; Burket, P.; Daniel, G.

    2011-02-24

    The U.S. Department of Energy's Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford's tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. The Supplemental Treatment chosen will immobilize that portion of the retrieved LAW that is not sent to the WTP's LAW Vitrification facility into a solidified waste form. The solidified waste will then be disposed on the Hanford site in the Integrated Disposal Facility (IDF). In addition, the WTP LAW vitrification facility off-gas condensate known as WTP Secondary Waste (WTP-SW) will be generated and enriched in volatile components such as Cs-137, I-129, Tc-99, Cl, F, and SO4 that volatilize at the vitrification temperature of 1150 C in the absence of a continuous cold cap. The current waste disposal path for the WTP-SW is to recycle it to the supplemental LAW treatment to avoid a large steady state accumulation in the pretreatment-vitrification loop. Fluidized Bed Steam Reforming (FBSR) offers a moderate temperature (700-750 C) continuous method by which LAW and/or WTP-SW wastes can be processed irrespective of whether they contain organics, nitrates, sulfates/sulfides, chlorides, fluorides, volatile radionuclides or other aqueous components. The FBSR technology can process these wastes into a crystalline ceramic (mineral) waste form. The mineral waste form that is produced by co-processing waste with kaolin clay in an FBSR process has been shown to be as durable as LAW glass. Monolithing of the granular FBSR product is being investigated to prevent dispersion during transport or burial/storage but is not necessary for performance. A Benchscale Steam Reformer (BSR) was designed and constructed at the Savannah River National Laboratory (SRNL) to treat actual radioactive wastes to confirm the findings of the non-radioactive FBSR pilot scale tests and to qualify the waste form for applications at Hanford. Radioactive testing commenced in 2010 with a demonstration of Hanford's WTP-SW where Savannah River Site (SRS) High Level Waste (HLW) secondary waste from the Defense Waste Processing Facility (DWPF) was shimmed with a mixture of I-125/129 and Tc-99 to chemically resemble WTP-SW. Ninety six grams of radioactive product were made for testing. The second campaign commenced using SRS LAW chemically trimmed to look like Hanford's LAW. Six hundred grams of radioactive product were made for extensive testing and comparison to the non-radioactive pilot scale tests. The same mineral phases were found in the radioactive and non-radioactive testing.

  7. SUMMARY PLAN FOR BENCH-SCALE REFORMER AND PRODUCT TESTING TREATABILITY STUDIES USING HANFORD TANK WASTE

    SciTech Connect (OSTI)

    DUNCAN JB

    2010-08-19

    This paper describes the sample selection, sample preparation, environmental, and regulatory considerations for shipment of Hanford radioactive waste samples for treatability studies of the FBSR process at the Savannah River National Laboratory and the Pacific Northwest National Laboratory. The U.S. Department of Energy (DOE) Hanford tank farms contain approximately 57 million gallons of wastes, most of which originated during the reprocessing of spent nuclear fuel to produce plutonium for defense purposes. DOE intends to pre-treat the tank waste to separate the waste into a high level fraction, that will be vitrified and disposed of in a national repository as high-level waste (HLW), and a low-activity waste (LAW) fraction that will be immobilized for on-site disposal at Hanford. The Hanford Waste Treatment and Immobilization Plant (WTP) is the focal point for the treatment of Hanford tank waste. However, the WTP lacks the capacity to process all of the LAW within the regulatory required timeframe. Consequently, a supplemental LAW immobilization process will be required to immobilize the remainder of the LAW. One promising supplemental technology is Fluidized Bed Steam Reforming (FBSR) to produce a sodium-alumino-silicate (NAS) waste form. The NAS waste form is primarily composed of nepheline (NaAlSiO{sub 4}), sodalite (Nas[AlSiO{sub 4}]{sub 6}Cl{sub 2}), and nosean (Na{sub 8}[AlSiO{sub 4}]{sub 6}SO{sub 4}). Semivolatile anions such as pertechnetate (TcO{sub 4}{sup -}) and volatiles such as iodine as iodide (I{sup -}) are expected to be entrapped within the mineral structures, thereby immobilizing them (Janzen 2008). Results from preliminary performance tests using surrogates, suggests that the release of semivolatile radionuclides {sup 99}Tc and volatile {sup 129}I from granular NAS waste form is limited by Nosean solubility. The predicted release of {sup 99}Tc from the NAS waste form at a 100 meters down gradient well from the Integrated Disposal Facility (IDF) was found to be comparable to immobilized low-activity waste glass waste form in the initial supplemental LAW treatment technology risk assessment (Mann 2003). To confirm this hypothesis, DOE is funding a treatability study where three actual Hanford tank waste samples (containing both {sup 99}Tc and {sup 125}I) will be processed in Savannah River National Laboratory's (SRNL) Bench-Scale Reformer (BSR) to form the mineral product, similar to the granular NAS waste form, that will then be subject to a number of waste form qualification tests. In previous tests, SRNL have demonstrated that the BSR product is chemically and physically equivalent to the FBSR product (Janzen 2005). The objective of this paper is to describe the sample selection, sample preparation, and environmental and regulatory considerations for treatability studies of the FBSR process using Hanford tank waste samples at the SNRL. The SNRL will process samples in its BSR. These samples will be decontaminated in the 222-S Laboratory to remove undissolved solids and selected radioisotopes to comply with Department of Transportation (DOT) shipping regulations and to ensure worker safety by limiting radiation exposure to As Low As Reasonably Achievable (ALARA). These decontamination levels will also meet the Nuclear Regulatory Commission's (NRC's) definition of low activity waste (LAW). After the SNRL has processed the tank samples to a granular mineral form, SRNL and Pacific Northwest National Laboratory (PNNL) will conduct waste form testing on both the granular material and monoliths prepared from the granular material. The tests being performed are outlined in Appendix A.

  8. Bases, Assumptions, and Results of the Flowsheet Calculations for the Decision Phase Salt Disposition Alternatives

    SciTech Connect (OSTI)

    Dimenna, R.A.; Jacobs, R.A.; Taylor, G.A.; Durate, O.E.; Paul, P.K.; Elder, H.H.; Pike, J.A.; Fowler, J.R.; Rutland, P.L.; Gregory, M.V.; Smith III, F.G.; Hang, T.; Subosits, S.G.; Campbell, S.G.

    2001-03-26

    The High Level Waste (HLW) Salt Disposition Systems Engineering Team was formed on March 13, 1998, and chartered to identify options, evaluate alternatives, and recommend a selected alternative(s) for processing HLW salt to a permitted wasteform. This requirement arises because the existing In-Tank Precipitation process at the Savannah River Site, as currently configured, cannot simultaneously meet the HLW production and Authorization Basis safety requirements. This engineering study was performed in four phases. This document provides the technical bases, assumptions, and results of this engineering study.

  9. app_d

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

    D-83 DOE/EIS-0287 Idaho HLW & FD EIS Document 37, Public Comment Hearing, February 8, 2000, Pocatello, ID Page 1 of 6 Idaho HLW & FD EIS Document 36, Public Comment Hearing, February 9, 2000, Jackson, WY Page 54 of 54 DOE/EIS-0287 D-84 Appendix D Document 37, Public Comment Hearing, February 8, 2000, Pocatello, ID Page 2 of 6 Document 37, Public Comment Hearing, February 8, 2000, Pocatello, ID Page 3 of 6 - New Information - D-85 DOE/EIS-0287 Idaho HLW & FD EIS Document 37, Public

  10. ch_5

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

    25 DOE/EIS-0287 Idaho HLW & FD EIS 5.3 Facility Disposition Impacts Section 5.3 presents a discussion of potential impacts associated with the disposition of exist- ing HLW management facilities at INEEL and disposition of new facilities that would be built in support of the proposed waste processing alternatives. The discussion includes (1) the potential impacts of short-term actions in dispo- sitioning new and existing HLW management facilities, (2) the potential long-term impacts from the

  11. ch_5

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

    45 DOE/EIS-0287 Idaho HLW & FD EIS 5.3.4.2 Existing Facilities Associated with High-Level Waste Management The facilities in this group are those that have historically been used at the INTEC to generate, treat, and store HLW. Because of the number of facilities involved, DOE has grouped them in functional groups for purposes of analysis (see Table 3-3). DOE analyzed the HLW tanks and bin sets for closure under all five disposition sce- narios; however, facilities that support the Tank Farm

  12. RIVER PROTECTION PROJECT MISSION ANALYSIS WASTE BLENDING STUDY

    SciTech Connect (OSTI)

    SHUFORD DH; STEGEN G

    2010-04-19

    Preliminary evaluation for blending Hanford site waste with the objective of minimizing the amount of high-level waste (HLW) glass volumes without major changes to the overall waste retrieval and processing sequences currently planned. The evaluation utilizes simplified spreadsheet models developed to allow screening type comparisons of blending options without the need to use the Hanford Tank Waste Operations Simulator (HTWOS) model. The blending scenarios evaluated are expected to increase tank farm operation costs due to increased waste transfers. Benefit would be derived from shorter operating time period for tank waste processing facilities, reduced onsite storage of immobilized HLW, and reduced offsite transportation and disposal costs for the immobilized HLW.

  13. Commercial waste treatment program annual progress report for FY 1983

    SciTech Connect (OSTI)

    McElroy, J.L.; Burkholder, H.C. (comps.)

    1984-02-01

    This annual report describes progress during FY 1983 relating to technologies under development by the Commercial Waste Treatment Program, including: development of glass waste form and vitrification equipment for high-level wastes (HLW); waste form development and process selection for transuranic (TRU) wastes; pilot-scale operation of a radioactive liquid-fed ceramic melter (LFCM) system for verifying the reliability of the reference HLW treatment proces technology; evaluation of treatment requirements for spent fuel as a waste form; second-generation waste form development for HLW; and vitrification process control and product quality assurance technologies.

  14. High Level Waste Management Division . H L W System Plan

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

    . H L W System Plan Revision 1 (U) Westinghouse Savannah River Company Aiken, South Carolina . August 4, 1993 1.0 2.0 3.0 4.0 5.0 6.0 HLW System Plan - Revision 1 (U) Table of Contents Mission Statement Purpose Executive Summary 3.1 Reference Date 3.2 Key Milestones 3.3 Operational Plan Summary 3.4 Key Issues and Assumptions 3.5 HLW System Plan Management .HLW System Description Principles and Guidelines 5.1 Safety Documentation 5.2 Regulatory Permits 5.3 Long Range Planning 5.4 Roadmaps 5.5 DOE

  15. High Level Waste Management Division High. Level Waste System Plan

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

    HLW -OVP-98-0037 High Level Waste Management Division High. Level Waste System Plan Revision 9 (U) April 1998 Westinghouse Savannah River Company Savannah River Site Aiken, SC 29808 Prepared for the U. S. Department of Energy under contract no. DE-AC09-96SR18500 HLW -OVP-98-0037 High Level Waste Management Division High Level Waste System Plan Revision 9 (U) Contributors: A. S. Choi P. Paul F. E. Wise Prepared by: ?1M.J II£) ~ N. R. Davis Approved by: HLW System Integration Manager ll\1-'-ft

  16. Energy Conservation Standards Activities - Report to Congress - August 2012

    Office of Environmental Management (EM)

    End of Year 2010 SNF & HLW Inventories End of Year 2010 SNF & HLW Inventories Map of the United States of America that shows the location of approximately 64,000 MTHM of Spent Nuclear Fuel (SNF) & 275 High-Level Radioactive Waste (HLW) Canisters. PDF icon Slide 1 More Documents & Publications Assessment of Disposal Options for DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel FY 2007 Total System Life Cycle Cost, Pub 2008 Microsoft Word - TSLCC 2007_5_05_08 rev

  17. ColorMac.cdd

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

    REVISION 13 March 2002 An Integrated System at the Savannah River Site Waste Immobilization Waste Immobilization HLW-2002-00025 HLW-2002-00025 Retention: Permanent offer to NARA when no longer needed by the Dept. Disposal Auth: DOE 1-9.a Track #: 10048 High Level Waste Division High Level Waste System Plan Revision 13 (U) Prepared by: T. B. Caldwell D. P. Chew H. H. Elder M. J. Mahoney K. B. Way W. A. Wilson F. E. Wise Approved by: M. J. Mahoney Date HLW Systems Integration Manager S. S. Cathey

  18. Quantity of 135I Released from the AGR 1, AGR 2, and AGR 3/4 Experiments and Discovery of 131I at the FPMS Traps during the AGR-3/4 Experiment

    SciTech Connect (OSTI)

    Dawn Scates

    2014-09-01

    A series of three Advanced Gas Reactor (AGR) experiments have been conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). From 2006 through 2014, these experiments supported the development and qualification of the new U.S. tristructural isotropic (TRISO) particle fuel for Very High Temperature Reactors (VHTR). Each AGR experiment consisted of multiple fueled capsules, each plumbed for independent temperature control using a mix of helium and neon gases. The gas leaving a capsule was routed to individual Fission Product Monitor (FPM) detectors. For intact fuel particles, the TRISO particle coatings provide a substantial barrier to fission product release. However, particles with failed coatings, whether because of a minute percentage of initially defective particles, those which fail during irradiation, or those designed to fail (DTF) particles, can release fission products to the flowing gas stream. Because reactive fission product elements like iodine and cesium quickly deposit on cooler capsule components and piping structures as the effluent gas leaves the reactor core, only the noble fission gas isotopes of Kr and Xe tend to reach FPM detectors. The FPM system utilizes High Purity Germanium (HPGe) detectors coupled with a thallium activated sodium iodide NaI(Tl) scintillator. The HPGe detector provides individual isotopic information, while the NaI(Tl) scintillator is used as a gross count rate meter. During irradiation, the 135mXe concentration reaching the FPM detectors is from both direct fission and by decay of the accumulated 135I. About 2.5 hours after irradiation (ten 15.3 minute 135mXe half lives) the directly produced 135mXe has decayed and only the longer lived 135I remains as a source. Decay systematics dictate that 135mXe will be in secular equilibrium with its 135I parent, such that its production rate very nearly equals the decay rate of the parent, and its concentration in the flowing gas stream will appear to decay with the parent half life. This equilibrium condition enables the determination of the amount of 135I released from the fuel particles by measurement of the 135mXe at the FPM following reactor shutdown. In this paper, the 135I released will be reported and compared to similar releases for noble gases as well as the unexpected finding of 131I deposition from intentional impure gas injection into capsule 11 of experiment AGR 3/4.

  19. The benefits of a fast reactor closed fuel cycle in the UK

    SciTech Connect (OSTI)

    Gregg, R.; Hesketh, K.

    2013-07-01

    The work has shown that starting a fast reactor closed fuel cycle in the UK, requires virtually all of Britain's existing and future PWR spent fuel to be reprocessed, in order to obtain the plutonium needed. The existing UK Pu stockpile is sufficient to initially support only a modest SFR 'closed' fleet assuming spent fuel can be reprocessed shortly after discharge (i.e. after two years cooling). For a substantial fast reactor fleet, most Pu will have to originate from reprocessing future spent PWR fuel. Therefore, the maximum fast reactor fleet size will be limited by the preceding PWR fleet size, so scenarios involving fast reactors still require significant quantities of uranium ore indirectly. However, once a fast reactor fuel cycle has been established, the very substantial quantities of uranium tails in the UK would ensure there is sufficient material for several centuries. Both the short and long term impacts on a repository have been considered in this work. Over the short term, the decay heat emanating from the HLW and spent fuel will limit the density of waste within a repository. For scenarios involving fast reactors, the only significant heat bearing actinide content will be present in the final cores, resulting in a 50% overall reduction in decay energy deposited within the repository when compared with an equivalent open fuel cycle. Over the longer term, radiological dose becomes more important. Total radiotoxicity (normalised by electricity generated) is lower for scenarios with Pu recycle after 2000 years. Scenarios involving fast reactors have the lowest radiotoxicity since the quantities of certain actinides (Np, Pu and Am) eventually stabilise. However, total radiotoxicity as a measure of radiological risk does not account for differences in radionuclide mobility once in repository. Radiological dose is dominated by a small number of fission products so is therefore not affected significantly by reactor type or recycling strategy (since the fission product will primarily be a function of nuclear energy generated). However, by reprocessing spent fuel, it is possible to immobilise the fission product in a more suitable waste form that has far more superior in-repository performance. (authors)

  20. app_d

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

    Idaho HLW & FD EIS Document 23, Department of Health & Human Services (Kenneth W. Holt), Atlanta, GA Page 10 of 21 Document 23, Department of Health & Human Services (Kenneth W. ...

  1. Technical Exchange on Improved Design and Performance of High Level Waste Melters - Final Report

    SciTech Connect (OSTI)

    SK Sundaram; ML Elliott; D Bickford

    1999-11-19

    SIA Radon is responsible for management of low- and intermediate-level radioactive waste (LILW) produced in Central Russia. In cooperation with Minatom organizations Radon carries out R and D programs on treatment of simulated high level waste (HLW) as well. Radon scientists deal with a study of materials for LILW, HLW, and Nuclear Power Plants (NPP) wastes immobilization, and development and testing of processes and technologies for waste treatment and disposal. Radon is mostly experienced in LILW vitrification. This experience can be carried over to HLW vitrification especially in field of melting systems. The melter chosen as a basic unit for the vitrification plant is a cold crucible. Later on Radon experience in LILW vitrification as well as our results on simulated HLW vitrification are briefly described.

  2. Enterprise Assessments Operational Awareness Record, Waste Treatment and Immobilization Plant – December 2014

    Broader source: Energy.gov [DOE]

    Operational Awareness Record for the Observation of Waste Treatment and Immobilization Plant High Level Waste Facility Radioactive Liquid Waste Disposal System Hazards Analysis Activities (EA-WTP-HLW-2014-08-18(a))

  3. EIS-0303: Final Environmental Impact Statement | Department of...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    EIS-0303: Final Environmental Impact Statement High-Level Waste Tank Closure DOE proposes to close the high-level waste (HLW) tanks at the Savannah River Site (SRS) in accordance ...

  4. A pilot test of partitioning for the simulated highly saline high level waste

    SciTech Connect (OSTI)

    Chen, Jing; Wang, Jianchen; Jing, Shan

    2007-07-01

    It is a problem how to treat the highly saline high level waste (HLW). A partitioning process for HLW was developed at INET. The partitioning process includes the removal of actinides by TRPO extraction, the removal of Sr by crown ether extraction, and the removal of Cs by ion exchange. A 72-hour test was carried out in a pilot facility using the simulated HLW. Nd and Zr were used to simulate Am and Pu, respectively. The decontamination factors are >3000, >500, >1000, {approx}150 and {approx}94 for U, Nd, Zr, Sr and Cs, respectively. The results meet the requirement to change the highly saline HLW into a non-{alpha} and intermediate level waste. (authors)

  5. EIS-0287: Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement, EIS-0287 (September 2002)

    Broader source: Energy.gov [DOE]

    This EIS analyzes the potential environmental consequences of alternatives for managing high-level waste (HLW) calcine, mixed transuranic waste/sodium bearing waste (SBW) and newly generated liquid...

  6. TEC Working Group Topic Groups Routing

    Broader source: Energy.gov [DOE]

    The Routing Topic Group has been established to examine topics of interest and relevance concerning routing of shipments of spent nuclear fuel (SNF) and high-level radioactive waste (HLW) to a...

  7. Microsoft Word - FY2014_HAB_WorkPlan_v6.doc

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

    Liaisons. 8302013 v.6 Page 7 of 11 Priority Topics - "A" List Policy Level Issues Potential HAB Action Committee Assignment (lead in bold) Status LAW and HLW Direct Feed...

  8. High Level Waste System Plan Revision 9

    SciTech Connect (OSTI)

    Davis, N.R.; Wells, M.N.; Choi, A.S.; Paul, P.; Wise, F.E.

    1998-04-01

    Revision 9 of the High Level Waste System Plan documents the current operating strategy of the HLW System at SRS to receive, store, treat, and dispose of high-level waste.

  9. Tank Waste Corporate Board | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    The following documents are associated with the Tank Waste Corporate Board Meeting held on November 6th, 2008. Note: (Please contact Steven Ross at steven.ross@em.doe.gov for a HLW ...

  10. Tank Waste Corporate Board Meeting 11/06/08 | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    The following documents are associated with the Tank Waste Corporate Board Meeting held on November 6th, 2008. Note: (Please contact Steven Ross at steven.ross@em.doe.gov for a HLW ...

  11. DOE/ID-Number

    Office of Environmental Management (EM)

    for Spent Nuclear Fuel and High-Level Waste Prepared for U.S. Department of Energy ... management of spent nuclear fuel (SNF) a and high-level waste (HLW) for 54 years. ...

  12. Summary.qxd

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

    (such as land use restrictions or building permits). S-11 DOEEIS-0287 Idaho HLW & FD EIS this EIS is the commitment to have all calcine treated and ready for shipment out of...

  13. Summary.qxd

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

    3 DOEEIS-0287 Idaho HLW & FD EIS cussed below. Options excluded from DOE's Preferred Alternative are, storage of calcine in the bin sets for an indefinite period under the...

  14. EIS-0287: Idaho High-Level Waste & Facilities Disposition

    Broader source: Energy.gov [DOE]

    This EIS analyzes the potential environmental consequences of alternatives for managing high-level waste (HLW) calcine, mixed transuranic waste/sodium bearing waste (SBW) and newly generated liquid...

  15. Office of Enterprise Assessments Operational Awareness Record...

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

    Operational Awareness Record Report Number: EA-WTP-HLW-2014-10-20 Site: Hanford Site Subject: Observation of the Waste Treatment and Immobilization Plant High Level Waste Facility ...

  16. SAND2010-8200

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

    United States", Annu. Rev. Energy Environ., Vol. 26, p.201-235. McKinley, I.G. (1997) "Engineering for robustness: An approach to optimising HLW disposal concepts", Waste...

  17. Hanford Tank Waste Residuals

    Office of Environmental Management (EM)

    Hanford Tank Waste Residuals DOE HLW Corporate Board November 6, 2008 Chris Kemp, DOE ORP Bill Hewitt, YAHSGS LLC Hanford Tanks & Tank Waste * Single-Shell Tanks (SSTs) - 27 million ...

  18. Biography N. Mller-Hoeppe

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

    N. Mller-Hoeppe Dr. Mueller-Hoeppe presented at the first USGerman Workshop on the topic of AN ENGINEERING APPROACH FOR A SAFETY CONCEPT FOR DISPOSING OFF HLW IN SALT. Dr....

  19. Microsoft Word - FY13_WM_Committee_Work_Plan_Final_Draft

    Office of Environmental Management (EM)

    waste, e.g., Transuranic (TRU) and High Level Waste (HLW). It is well underway but CAB should continue to monitor b. Subsurface remediation in the event that buried waste is...

  20. SR0203

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

    31, 2002 Media Contact: Julie Petersen (803) 725-2889 DOE Announces Availability Of Final EIS On SRS High-Level Waste Tank Closures Flag Ribbon Art AIKEN, SC; May 31 - The Department of Energy (DOE) Savannah River Site (SRS) announced today the availability of the Final Environmental Impact Statement (EIS) on the proposed closing of additional high-level waste (HLW) tanks at SRS. The EIS evaluated three alternatives regarding closure of the HLW tanks: (1) Stabilize Tanks; (2) Clean and Remove

  1. Preliminary waste form characteristics report Version 1.0. Revision 1

    SciTech Connect (OSTI)

    Stout, R.B.; Leider, H.R.

    1991-10-11

    This report focuses on radioactive waste form characteristics that will be used to design a waste package and an engineered barrier system (EBS) for a suitable repository as part of the Yucca Mountain Project. The term waste form refers to irradiated reactor fuel, other high-level waste (HLW) in various physical forms, and other radioactive materials (other than HLW) which are received for emplacement in a geologic repository. Any encapsulating of stabilizing matrix is also referred to as a waste form.

  2. Waste Treatment and Immobilization Plant Progress

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

    Waste Treatment and Immobilization Plant Progress Hanford Advisory Board requested action:  Based on progress discussions, the Hanford Advisory Board will develop and advocate an effective public communication strategy for use by the Waste Treatment and Immobilization Plant Assistant Manager/Federal Project Director Progress discussions on the following:  High-level waste (HLW) authorization to proceed with full production engineering:  HLW Safety Design Strategy approval and

  3. 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.

  4. Managing Category I and II ACM During Decontamination and Demolition Presentation

    Office of Environmental Management (EM)

    Management of nuclear materials and non-HLW Management of nuclear materials and non-HLW GC-52 provides legal advice to DOE regarding the consolidation and disposition of nuclear materials, including plutonium, uranium, and nuclear waste in accordance with applicable statutes, DOE Orders and international commitments. Advice encompasses issues related to mixed oxide fuel, waste incidental-to-reprocessing, transuranic waste, low-level waste, greater-than-class C waste and sealed sources.

  5. Folie 1

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

    Bernhard Droste 1 SNF/HLW Dual and Multi Purpose Casks Issues Bernhard Droste BAM Federal Institute for Materials Research and Testing Berlin, Germany bernhard.droste@bam.de BAM/Sandia Workshop Albuquerque, NM, USA, October 6-8, 2014 Bernhard Droste BAM/Sandia Workshop 2 Presentation Outline - Design, Transport, Storage of DPCs for SNF and HLW in Germany - Measurement and Demonstration Programs - Integrated DPC Safety Case Approach, IAEA - Aging Considerations - Inspections before Transport

  6. SLUDGE HEEL REMOVAL BY ALUMINUM DISSOLUTION AT SAVANNAH RIVER SITE 12390

    SciTech Connect (OSTI)

    Keefer, M.

    2012-01-12

    High Level Waste (HLW) at the Savannah River Site (SRS) is currently stored in aging underground storage tanks. This waste is a complex mixture of insoluble solids, referred to as sludge, and soluble salts. Continued long-term storage of these radioactive wastes poses an environmental risk. Operations are underway to remove and disposition the waste, clean the tanks and fill with grout for permanent closure. Heel removal is the intermediate phase of the waste retrieval and tank cleaning process at SRS, which is intended to reduce the volume of waste prior to treatment with oxalic acid. The goal of heel removal is to reduce the residual amount of radioactive sludge wastes to less than 37,900 liters (10,000 gallons) of wet solids. Reducing the quantity of residual waste solids in the tank prior to acid cleaning reduces the amount of acid required and reduces the amount of excess acid that could impact ongoing waste management processes. Mechanical heel removal campaigns in Tank 12 have relied solely on the use of mixing pumps that have not been effective at reducing the volume of remaining solids. The remaining waste in Tank 12 is known to have a high aluminum concentration. Aluminum dissolution by caustic leaching was identified as a treatment step to reduce the volume of remaining solids and prepare the tank for acid cleaning. Dissolution was performed in Tank 12 over a two month period in July and August, 2011. Sample results indicated that 16,440 kg of aluminum oxide (boehmite) had been dissolved representing 60% of the starting inventory. The evolution resulted in reducing the sludge solids volume by 22,300 liters (5900 gallons), preparing the tank for chemical cleaning with oxalic acid.

  7. Development of the high-level waste high-temperature melter feed preparation flowsheet for vitrification process testing

    SciTech Connect (OSTI)

    Seymour, R.G.

    1995-02-17

    High-level waste (HLW) feed preparation flowsheet development was initiated in fiscal year (FY) 1994 to evaluate alternative flowsheets for preparing melter feed for high-temperature melter (HTM) vitrification testing. Three flowsheets were proposed that might lead to increased processing capacity relative to the Hanford Waste Vitrification Plant (HWVP) and that were flexible enough to use with other HLW melter technologies. This document describes the decision path that led to the selection of flowsheets to be tested in the FY 1994 small-scale HTM tests. Feed preparation flowsheet development for the HLW HTM was based on the feed preparation flowsheet that was developed for the HWVP. This approach allowed the HLW program to build upon the extensive feed preparation flowsheet database developed under the HWVP Project. Primary adjustments to the HWVP flowsheet were to the acid adjustment and glass component additions. Developmental background regarding the individual features of the HLW feed preparation flowsheets is provided. Applicability of the HWVP flowsheet features to the new HLW vitrification mission is discussed. The proposed flowsheets were tested at the laboratory-scale at Pacific Northwest Laboratory. Based on the results of this testing and previously established criteria, a reductant-based flowsheet using glycolic acid and a nitric acid-based flowsheet were selected for the FY 1994 small-scale HTM testing.

  8. GeoMelt{sup R} ICV{sup TM} Treatment of Sellafield Pond Solids Waste - 13414

    SciTech Connect (OSTI)

    Witwer, Keith; Woosley, Steve; Campbell, Brett [Kurion, Inc., GeoMelt Division, 3015 Horn Rapids Road, Richland, Washington (United States)] [Kurion, Inc., GeoMelt Division, 3015 Horn Rapids Road, Richland, Washington (United States); Wong, Martin; Hill, Joanne [AMEC Inc., Birchwood Park, 601 Faraday Street, Birchwood, Warrington, WA3 6GN (United Kingdom)] [AMEC Inc., Birchwood Park, 601 Faraday Street, Birchwood, Warrington, WA3 6GN (United Kingdom)

    2013-07-01

    Kurion, Inc., in partnership with AMEC Ltd., is demonstrating its GeoMelt{sup R} In-Container Vitrification (ICV){sup TM} Technology to Sellafield Ltd. (SL). SL is evaluating the proposition of directly converting a container (skip/box/drum) of raw solid ILW into an immobilized waste form using thermal treatment, such that the resulting product is suitable for interim storage at Sellafield and subsequent disposal at a future Geological Disposal Facility. Potential SL feed streams include sludges, ion-exchange media, sand, plutonium contaminated material, concrete, uranium, fuel cladding, soils, metals, and decommissioning wastes. The solid wastes have significant proportions of metallic constituents in the form of containers, plant equipment, structural material and swarf arising from the nuclear operations at Sellafield. GeoMelt's proprietary ICV process was selected for demonstration, with the focus being high and reactive metal wastes arising from solid ILW material. A composite surrogate recipe was used to demonstrate the technology towards treating waste forms of diverse types and shapes, as well as those considered difficult to process; all the while requiring few (if any) pre-treatment activities. Key strategic objectives, along with their success criterion, were established by SL for this testing, namely: 1. Passivate and stabilize the raw waste simulant, as demonstrated by the entire quantity of material being vitrified, 2. Immobilize the radiological and chemo-toxic species, as demonstrated via indicative mass balance using elemental analyses from an array of samples, 3. Production of an inert and durable product as evidenced by transformation of reactive metals to their inert oxide forms and satisfactory leachability results using PCT testing. Two tests were performed using the GeoMelt Demonstration Unit located at AMEC's Birchwood Park Facilities in the UK. Post-melt examination of the first test indicated some of the waste simulant had not fully processed, due to insufficient processing time and melt temperature. A second test, incorporating operational experience from the first test, was performed and resulted in all of the 138 kg of feed material being treated. The waste simulant portion, at 41 kg, constituted 30 wt% of the total feed mass, with over 90% of this being made up of various reactive and non-reactive metals. The 95 liters of staged material was volume reduced to 41 liters, providing a 57% overall feed to product volume reduction in a fully passivated two-phase glass/metal product. The GeoMelt equipment operated as designed, vitrifying the entire batch of waste simulant. Post-melt analytical testing verified that 91-99+% of the radiological tracer metals were uniformly distributed within the glass/cast refractory/metal product, and the remaining fraction was captured in the offgas filtration systems. PCT testing of the glass and inner refractory liner showed leachability results that outperform the DOE regulatory limit of 2 g/m{sup 2} for the radiological species of interest (Sr, Ru, Cs, Eu, Re), and by more than an order of magnitude better for standard reference analytes (B, Na, Si). (authors)

  9. ARM - Instruments

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

    govInstrumentsDerived Quantities and Models

  10. Life extension program for the modular caustic side solvent extraction unit at Savannah River Site

    SciTech Connect (OSTI)

    Samadi-Dezfouli, Azadeh

    2012-11-14

    Caustic Side Solvent Extraction (CSSX) is currently used at the U.S. Department of Energy (DOE) Savannah River Site (SRS) for removal of cesium from the high-level salt-wastes stored in underground tanks. At SRS, the CSSX process is deployed in the Modular CSSX Unit (MCU). The CSSX technology utilizes a multi-component organic solvent and annular centrifugal contactors to extract cesium from alkaline salt waste. Coalescers and decanters process the Decontaminated Salt Solution (DSS) and Strip Effluent (SE) streams to allow recovery and reuse of the organic solvent and to limit the quantity of solvent transferred to the downstream facilities. MCU is operated in series with the Actinide Removal Process (ARP) which removes strontium and actinides from salt waste utilizing monosodium titanate. ARP and MCU were developed and implemented as interim salt processing until future processing technology, the CSSX-based Salt Waste Processing Facility (SWPF), is operational. SWPF is slated to come on-line in October 2014. The three year design life of the ARP/MCU process, however, was reached in April 2011. Nevertheless, most of the individual process components are capable of operating longer. An evaluation determined ARP/MCU can operate until 2015 before major equipment failure is expected. The three year design life of the ARP/MCU Life Extension (ARP/MCU LE) program will bridge the gap between current ARP/MCU operations and the start of SWPF operation. The ARP/MCU LE program introduces no new technologies. As a portion of this program, a Next Generation Solvent (NGS) and corresponding flowsheet are being developed to provide a major performance enhancement at MCU. This paper discusses all the modifications performed in the facility to support the ARP/MCU Life Extension. It will also discuss the next generation chemistry, including NGS and new stripping chemistry, which will increase cesium removal efficiency in MCU. Possible implementation of the NGS chemistry in MCU accomplishes two objectives. MCU serves as a demonstration facility for improved flowsheet deployment at SWPF; operating with NGS and boric acid validates improved cesium removal performance and increased throughput as well as confirms Defense Waste Processing Facility (DWPF) ability to vitrify waste streams containing boron. NGS implementation at MCU also aids the ARP/MCU LE operation, mitigating the impacts of delays and sustaining operations until other technology is able to come on-line.

  11. TREATMENT OF GASEOUS EFFLUENTS ISSUED FROM RECYCLING A REVIEW OF THE CURRENT PRACTICES AND PROSPECTIVE IMPROVEMENTS

    SciTech Connect (OSTI)

    Patricia Paviet-Hartmann; William Kerlin; Steven Bakhtiar

    2010-11-01

    The objectives of gaseous waste management for the recycling of nuclear used fuel is to reduce by best practical means (ALARA) and below regulatory limits, the quantity of activity discharged to the environment. The industrial PUREX process recovers the fissile material U(VI) and Pu(IV) to re-use them for the fabrication of new fuel elements e.g. recycling plutonium as a Mixed Oxide (MOX) fuel or recycling uranium for new enrichment for Pressurized Water Reactor (PWR). Meanwhile the separation of the waste (activation and fission product) is performed as a function of their pollution in order to store and avoid any potential danger and release towards the biosphere. Raffinate, that remains after the extraction step and which contains mostly all fission products and minor actinides is vitrified, the glass package being stored temporarily at the recycling plant site. Hulls and end pieces coming from PWR recycled fuel are compacted by means of a press leading to a volume reduced to 1/5th of initial volume. An organic waste treatment step will recycle the solvent, mainly tri-butyl phosphate (TBP) and some of its hydrolysis and radiolytic degradation products such as dibutyl phosphate (HDPB) and monobutyl phosphate (H2MBP). Although most scientific and technological development work focused on high level waste streams, a considerable effort is still under way in the area of intermediate and low level waste management. Current industrial practices for the treatment of gaseous effluents focusing essentially on Iodine-129 and Krypton-85 will be reviewed along with the development of novel technologies to extract, condition, and store these fission products. As an example, the current industrial practice is to discharge Kr-85, a radioactive gas, entirely to the atmosphere after dilution, but for the large recycling facilities envisioned in the near future, several techniques such as 1) cryogenic distillation and selective absorption in solvents, 2) adsorption on activated charcoal, 3) selective sorption on chemical modified zeolites, or 4) diffusion through membranes with selective permeability are potential technologies to retain the gas.

  12. Defense High Level Waste Disposal Container System Description

    SciTech Connect (OSTI)

    2000-10-12

    The Defense High Level Waste Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms (IPWF)) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. U.S. Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as 'co-disposal'. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by which to identify the disposal container and its contents. Different materials will be selected for the disposal container inner and outer cylinders. The two metal cylinders, in combination with the Emplacement Drift System, drip shield, and natural barrier, will support the design philosophy of defense-in-depth. The use of materials with different properties prevents a single mode failure from breaching the waste package. The inner cylinder and inner cylinder lids will be constructed of stainless steel and the outer cylinder and outer cylinder lids will be a barrier made of high-nickel alloy. The defense HLW disposal container interfaces with the emplacement drift environment and the internal waste by transferring heat from the canisters to the external environment and by protecting the canisters and their contents from damage/degradation by the external environment. The disposal container also interfaces with the canisters by limiting access of moderator and oxidizing agents to the waste. A loaded and sealed disposal container (waste package) interfaces with the Emplacement Drift System's emplacement drift waste package supports upon which the waste packages are placed. The disposal container interfaces with the Canister Transfer System, Waste Emplacement /Retrieval System, Disposal Container Handling System, and Waste Package Remediation System during loading, handling, transfer, emplacement, and retrieval for the disposal container/waste package.

  13. 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.

  14. Crystallization in high-level waste glass: A review of glass theory and noteworthy literature

    SciTech Connect (OSTI)

    Christian, J. H.

    2015-08-18

    There is a fundamental need to continue research aimed at understanding nepheline and spinel crystal formation in high-level waste (HLW) glass. Specifically, the formation of nepheline solids (K/NaAlSiO4) during slow cooling of HLW glass can reduce the chemical durability of the glass, which can cause a decrease in the overall durability of the glass waste form. The accumulation of spinel solids ((Fe, Ni, Mn, Zn)(Fe, Cr)2O4), while not detrimental to glass durability, can cause an array of processing problems inside HLW glass melters. In this review, the fundamental differences between glass and solid-crystals are explained using kinetic, thermodynamic, and viscosity arguments, and several highlights of glass-crystallization research, as it pertains to high-level waste vitrification, are described. In terms of mitigating spinel in the melter and both spinel and nepheline formation in the canister, the complexity of HLW glass and the intricate interplay between thermal, chemical, and kinetic factors further complicates this understanding. However, new experiments seeking to elucidate the contributing factors of crystal nucleation and growth in waste glass, and the compilation of data from older experiments, may go a long way towards helping to achieve higher waste loadings while developing more efficient processing strategies. Higher waste loadings and more efficient processing strategies will reduce the overall HLW Hanford Tank Waste Treatment and Immobilization Plant (WTP) vitrification facilities mission life.

  15. Radioactive waste disposal package

    DOE Patents [OSTI]

    Lampe, Robert F. (Bethel Park, PA)

    1986-01-01

    A radioactive waste disposal package comprising a canister for containing vitrified radioactive waste material and a sealed outer shell encapsulating the canister. A solid block of filler material is supported in said shell and convertible into a liquid state for flow into the space between the canister and outer shell and subsequently hardened to form a solid, impervious layer occupying such space.

  16. Method of making nanostructured glass-ceramic waste forms

    DOE Patents [OSTI]

    Gao, Huizhen; Wang, Yifeng; Rodriguez, Mark A.; Bencoe, Denise N.

    2014-07-08

    A waste form for and a method of rendering hazardous materials less dangerous is disclosed that includes fixing the hazardous material in nanopores of a nanoporous material, reacting the trapped hazardous material to render it less volatile/soluble, and vitrifying the nanoporous material containing the less volatile/soluble hazardous material.

  17. In-situ vitrification of soil

    DOE Patents [OSTI]

    Brouns, Richard A.; Buelt, James L.; Bonner, William F.

    1983-01-01

    A method of vitrifying soil at or below a soil surface location. Two or more conductive electrodes are inserted into the soil for heating of the soil mass between them to a temperature above its melting temperature. Materials in the soil, such as buried waste, can thereby be effectively immobilized.

  18. Final Vitrification Melter And Vessels Evaluation Documentation

    Broader source: Energy.gov [DOE]

    DOE has prepared final evaluations and made waste incidental to reprocessing determinations for the vitrification melter and feed vessels (the concentrator feed makeup tank and the melter feed hold tank), used by DOE’s West Valley Demonstration Project as part of the process to vitrify waste from prior commercial reprocessing of spent nuclear fuel.

  19. Method of making nanostructured glass-ceramic waste forms

    DOE Patents [OSTI]

    Gao, Huizhen; Wang, Yifeng; Rodriguez, Mark A.; Bencoe, Denise N.

    2012-12-18

    A method of rendering hazardous materials less dangerous comprising trapping the hazardous material in nanopores of a nanoporous composite material, reacting the trapped hazardous material to render it less volatile/soluble, sealing the trapped hazardous material, and vitrifying the nanoporous material containing the less volatile/soluble hazardous material.

  20. Final Report - IHLW PCT, Spinel T1%, Electrical Conductivity, and Viscosity Model Development, VSL-07R1240-4

    SciTech Connect (OSTI)

    Kruger, Albert A.; Piepel, Gregory F.; Landmesser, S. M.; Pegg, I. L.; Heredia-Langner, Alejandro; Cooley, Scott K.; Gan, H.; Kot, W. K.

    2013-11-13

    This report is the last in a series of currently scheduled reports that presents the results from the High Level Waste (HLW) glass formulation development and testing work performed at the Vitreous State Laboratory (VSL) of the Catholic University of America (CUA) and the development of IHLW property-composition models performed jointly by Pacific Northwest National Laboratory (PNNL) and VSL for the River Protection Project-Waste Treatment and Immobilization Plant (RPP-WTP). Specifically, this report presents results of glass testing at VSL and model development at PNNL for Product Consistency Test (PCT), one-percent crystal fraction temperature (T1%), electrical conductivity (EC), and viscosity of HLW glasses. The models presented in this report may be augmented and additional validation work performed during any future immobilized HLW (IHLW) model development work. Completion of the test objectives is addressed.

  1. Studies on the stripping of transuranic elements from loaded TRPO by N,N-Dimethyl-3-oxa-glutaramic acid

    SciTech Connect (OSTI)

    Chen, Jing; Wang, Jianchen; Duan, Wuhua

    2008-07-01

    The partitioning and transmutation of long-lived nuclides such as minor actinides from HLW is a method to reduce the long-term radiotoxicity of high-level waste (HLW). The TRPO partitioning process to remove actinides from HLW was developed in China. In the original TRPO process, Am and lanthanides, Pu, and Np are stripped by 5.5 M HNO{sub 3} and 0.6 M oxalic acid from the loaded solvent, respectively. In order to simplify the stripping of transuranic elements, a new compound N,N-dimethyl-3-oxa-glutaramic acid (DOGA) was synthesized. Two pilot tests were carried out in the centrifugal-contactor facility. Nd and Zr were used to simulate Am and Pu, respectively. Stripping of >99.9% Zr and >99.9% Nd was achieved using DOGA from the loaded 30% TRPO-kerosene. (authors)

  2. Spent Nuclear Fuel Transportation: An Examination of Potential Lessons Learned From Prior Shipping Campaigns

    SciTech Connect (OSTI)

    Marsha Keister; Kathryn McBride

    2006-08-01

    The Nuclear Waste Policy Act of 1982 (NWPA), as amended, assigned the Department of Energy (DOE) responsibility for developing and managing a Federal system for the disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW). The Office of Civilian Radioactive Waste Management (OCRWM) is responsible for accepting, transporting, and disposing of SNF and HLW at the Yucca Mountain repository in a manner that protects public health, safety, and the environment; enhances national and energy security; and merits public confidence. OCRWM faces a near-term challengeto develop and demonstrate a transportation system that will sustain safe and efficient shipments of SNF and HLW to a repository. To better inform and improve its current planning, OCRWM has extensively reviewed plans and other documents related to past high-visibility shipping campaigns of SNF and other radioactive materials within the United States. This report summarizes the results of this review and, where appropriate, lessons learned.

  3. Operating experience during high-level waste vitrification at the West Valley Demonstration Project

    SciTech Connect (OSTI)

    Valenti, P.J.; Elliott, D.I.

    1999-01-01

    This report provides a summary of operational experiences, component and system performance, and lessons learned associated with the operation of the Vitrification Facility (VF) at the West Valley Demonstration Project (WVDP). The VF was designed to convert stored high-level radioactive waste (HLW) into a stable waste form (borosilicate glass) suitable for disposal in a federal repository. Following successful completion on nonradioactive test, HLW processing began in July 1995. Completion of Phase 1 of HLW processing was reached on 10 June 1998 and represented the processing of 9.32 million curies of cesium-137 (Cs-137) and strontium-90 (Sr-90) to fill 211 canisters with over 436,000 kilograms of glass. With approximately 85% of the total estimated curie content removed from underground waste storage tanks during Phase 1, subsequent operations will focus on removal of tank heel wastes.

  4. 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.

  5. Advanced waste form and melter development for treatment of troublesome high-level wastes

    SciTech Connect (OSTI)

    Marra, James; Kim, Dong -Sang; Maio, Vincent

    2015-09-02

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these "troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approached to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.

  6. Effect of vitrification temperature upon the solar average absorptance properties of Pyromark Series 2500 black paint

    SciTech Connect (OSTI)

    Nelson, C.; Mahoney, A.R.

    1986-06-01

    A significant drop in production efficiency has occurred over time at the Solar One facility at Barstow, California, primarily as a result of the degradation of the Pyromark Series 2500 black paint used as the absorptive coating on the receiver panels. As part of the investigation of the problem, the solar-averaged adsorptance properties of the paint were determined as a function of vitrification temperature, since it is known that a significant amount of the panel surface area at Solar One was vitrified at temperatures below those recommended by the paint manufacturer (540/sup 0/C, 1000/sup 0/F). Painted samples initially vitrified at 230/sup 0/C (450/sup 0/F), 315/sup 0/C (600/sup 0/F), 371/sup 0/C (700/sup 0/F), and 480/sup 0/C (900/sup 0/F) exhibited significantly lower solar-averaged absorptance values (0.02 absorptance units) compared to samples vitrified at 540/sup 0/C (1000/sup 0/F). Thus, Solar One began its service life below optimal levels. After 140 h of thermal aging at 370/sup 0/C (700/sup 0/F) and 540/sup 0/C (1000/sup 0/F), all samples regardless of their initial vitrification temperatures, attained the same solar-averaged absorptance value (..cap alpha../sub s/ = 0.973). Therefore, both the long-term low-temperature vitrification and the short-term high-temperature vitrification can be used to obtain optimal or near-optimal absorptance of solar flux. Futher thermal aging of vitrified samples did not result in paint degradation, clearly indicating that high solar flux is required to produce this phenomenon. The panels at Solar One never achieved optimal absorptance because their exposure to high solar flux negated the effect of long-term low-temperature vitrification during operation. On future central receiver projects, every effort should be made to properly vitrify the Pyromark coating before its exposure to high flux conditions.

  7. WTP Pretreatment Facility Potential Design Deficiencies--Sliding Bed and Sliding Bed Erosion Assessment

    SciTech Connect (OSTI)

    Hansen, E. K.

    2015-05-06

    This assessment is based on readily available literature and discusses both Newtonian and non-Newtonian slurries with respect to sliding beds and erosion due to sliding beds. This report does not quantify the size of the sliding beds or erosion rates due to sliding beds, but only assesses if they could be present. This assessment addresses process pipelines in the Pretreatment (PT) facility and the high level waste (HLW) transfer lines leaving the PT facility to the HLW vitrification facility concentrate receipt vessel.

  8. Tank Waste Remediation System optimized processing strategy

    SciTech Connect (OSTI)

    Slaathaug, E.J.; Boldt, A.L.; Boomer, K.D.; Galbraith, J.D.; Leach, C.E.; Waldo, T.L.

    1996-03-01

    This report provides an alternative strategy evolved from the current Hanford Site Tank Waste Remediation System (TWRS) programmatic baseline for accomplishing the treatment and disposal of the Hanford Site tank wastes. This optimized processing strategy performs the major elements of the TWRS Program, but modifies the deployment of selected treatment technologies to reduce the program cost. The present program for development of waste retrieval, pretreatment, and vitrification technologies continues, but the optimized processing strategy reuses a single facility to accomplish the separations/low-activity waste (LAW) vitrification and the high-level waste (HLW) vitrification processes sequentially, thereby eliminating the need for a separate HLW vitrification facility.

  9. U. S. Department of Energy Savannah River Operations Office - DOE-SR News

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

    Releases Savannah River Site Liquid Waste System Plans Archive Savannah River Site Liquid Waste System Plans Archive SRR-LWP-2009-00001 R-18 (June 2013) Adobe Acrobat PDF SRR-LWP-2009-00001 R-17 (February 2012) Adobe Acrobat PDF SRR-LWP-2009-00001 R-16 (December 2010) Adobe Acrobat PDF SRR-LWP-2009-00001 R-15 (January 2010) Adobe Acrobat PDF LWO-PIT-2007-00062 R-14.1 (October 2007) Adobe Acrobat PDF HLW-2002-00025 R-13 (March 2002) Adobe Acrobat PDF HLW-2001-00040 R-12 (March 2001) Adobe

  10. app_b

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

    B Alternative Selection Process B-iii DOE/EIS-0287 Idaho HLW & FD EIS TABLE OF CONTENTS Section Page Appendix B Alternative Selection Process B-1 B.1 Introduction B-1 B.2 Purpose B-1 B.3 Identification of Candidate Alternatives B-2 B.3.1 Analysis of Previous INEEL and other HLW DOE Studies B-2 B.3.2 Consideration of Pu blic Comments B-4 B.3.2.1 Overall Public Concerns B-4 B.3.2.2 Public Comments Applied to Alternative Development B-4 B.3.3 Candidate Alternatives B-5 B.3.3.1 Alternatives

  11. app_d

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

    DOE/EIS-0287 Idaho HLW & FD EIS Document 1, Darryl D. Siemer, Idaho Falls, ID Page 1 of 18 Document 1, Darryl D. Siemer, Idaho Falls, ID Page 2 of 18 - New Information - DOE/EIS-0287 D-2 Appendix D Document 1, Darryl D. Siemer, Idaho Falls, ID Page 3 of 18 Document 1, Darryl D. Siemer, Idaho Falls, ID Page 4 of 18 - New Information - D-3 DOE/EIS-0287 Idaho HLW & FD EIS Document 1, Darryl D. Siemer, Idaho Falls, ID Page 5 of 18 Document 1, Darryl D. Siemer, Idaho Falls, ID Page 6 of 18 -

  12. app_d

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

    5 DOE/EIS-0287 Idaho HLW & FD EIS Document 1, Darryl D. Siemer, Idaho Falls, ID Page 9 of 18 Document 1, Darryl D. Siemer, Idaho Falls, ID Page 10 of 18 - New Information - DOE/EIS-0287 D-6 Appendix D Document 1, Darryl D. Siemer, Idaho Falls, ID Page 11 of 18 Document 1, Darryl D. Siemer, Idaho Falls, ID Page 12 of 18 - New Information - D-7 DOE/EIS-0287 Idaho HLW & FD EIS Document 1, Darryl D. Siemer, Idaho Falls, ID Page 13 of 18 Document 1, Darryl D. Siemer, Idaho Falls, ID Page 14

  13. app_d

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

    Appendix D Document 10, Jeffrey Joel, Kelly, WY Page 1 of 1 Document 11, Avril Currier, Jackson, WY Page 1 of 2 - New Information - D-19 DOE/EIS-0287 Idaho HLW & FD EIS Document 11, Avril Currier, Jackson, WY Page 2 of 2 Document 12, Cisco Oldani, Jackson, WY Page 1 of 1 - New Information - DOE/EIS-0287 D-20 Appendix D Document 13, Paul & Ann Ruttle, Jackson, WY Page 1 of 2 Document 13, Paul & Ann Ruttle, Jackson, WY Page 2 of 2 - New Information - D-21 DOE/EIS-0287 Idaho HLW &

  14. app_d

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

    7 DOE/EIS-0287 Idaho HLW & FD EIS Document 24, Snake River Alliance (Jay Hormel), Bliss, ID Page 1 of 1 Document 25, Ruthann Saphier, Ketchum, ID Page 1 of 1 - New Information - DOE/EIS-0287 D-38 Appendix D Document 26, Wayne Ross, Richland, WA Page 1 of 1 Document 27, State of Oregon (Ken Niles), Portland, OR Page 1 of 3 - New Information - D-39 DOE/EIS-0287 Idaho HLW & FD EIS Document 27, State of Oregon (Ken Niles), Portland, OR Page 2 of 3 Document 27, State of Oregon (Ken Niles),

  15. app_d

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

    41 DOE/EIS-0287 Idaho HLW & FD EIS Document 30, Joyce E. Batezel, Moose, WY Page 1 of 1 Document 31, Tri-City Industrial Development Council (Harold Heacock), Kennewick, WA Page 1 of 2 - New Information - DOE/EIS-0287 D-42 Appendix D Document 31, Tri-City Industrial Development Council (Harold Heacock), Kennewick, WA Page 2 of 2 Document 32, U.S. Department of Commerce (Susan B. Fruchter), Washington, D.C. Page 1 of 2 - New Information - D-43 DOE/EIS-0287 Idaho HLW & FD EIS Document 32,

  16. app_d

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

    09 DOE/EIS-0287 Idaho HLW & FD EIS Document 43, Bruce, J. Mincher, Idaho Falls, ID Page 1 of 1 Document 44, Anne Newcomb, Wilson, WY Page 1 of 1 - New Information - Document 45, Public Comment Hearing, February 15, 2000, Twin Falls, ID Page 2 of 13 DOE/EIS-0287 D-110 Appendix D Document 45, Public Comment Hearing, February 15, 2000, Twin Falls, ID Page 1 of 13 - New Information - D-111 DOE/EIS-0287 Idaho HLW & FD EIS - New Information - Document 45, Public Comment Hearing, February 15,

  17. app_d

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

    17 DOE/EIS-0287 Idaho HLW & FD EIS - New Information - Document 47, Kemble & Mildred Stout, Spokane, WA Page 1 of 1 Document 48, U.S. Department of Interior (Preston A. Sleeger), Portland, OR Page 1 of 1 DOE/EIS-0287 D-118 Appendix D - New Information - Document 49, Lynn Sims, Portland, OR Page 1 of 1 Document 50, Public Comment Hearing, February 17, 2000, Boise, ID Page 1 of 6 D-119 DOE/EIS-0287 Idaho HLW & FD EIS - New Information - Document 50, Public Comment Hearing, February 17,

  18. app_d

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

    3 DOE/EIS-0287 Idaho HLW & FD EIS - New Information - Document 52, Hanford Advisory Board (Merilyn Reeves), Richland, WA Page 1 of 3 Document 52, Hanford Advisory Board (Merilyn Reeves), Richland, WA Page 2 of 3 DOE/EIS-0287 D-124 Appendix D - New Information - Document 52, Hanford Advisory Board (Merilyn Reeves), Richland, WA Page 3 of 3 Document 53, Public Comment Hearing, February 24, 2000, Pasco, WA Page 1 of 6 D-125 DOE/EIS-0287 Idaho HLW & FD EIS - New Information - Document 53,

  19. app_d

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

    7 DOE/EIS-0287 Idaho HLW & FD EIS - New Information - Document 54, INEEL Citizens Advisory Board (Stan Hobson), Idaho Falls, ID Page 1 of 9 Document 53, Public Comment Hearing, February 24, 2000, Pasco, WA Page 6 fo 6 DOE/EIS-0287 D-128 Appendix D - New Information - Document 54, INEEL Citizens Advisory Board (Stan Hobson), Idaho Falls, ID Page 2 of 9 Document 54, INEEL Citizens Advisory Board (Stan Hobson), Idaho Falls, ID Page 3 of 9 D-129 DOE/EIS-0287 Idaho HLW & FD EIS - New

  20. app_d

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

    53 DOE/EIS-0287 Idaho HLW & FD EIS - New Information - Document 58, COGEMA, Inc. (Rhonnie Smith), Idaho Falls, ID Page 1 of 13 Document 57, Studsvik, Inc. (Thomas Oliver), Columbia, SC Page 31 of 31 DOE/EIS-0287 D-154 Appendix D - New Information - Document 58, COGEMA, Inc. (Rhonnie Smith), Idaho Falls, ID Page 2 of 13 Document 58, COGEMA, Inc. (Rhonnie Smith), Idaho Falls, ID Page 3 of 13 D-155 DOE/EIS-0287 Idaho HLW & FD EIS - New Information - Document 58, COGEMA, Inc. (Rhonnie

  1. app_d

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

    2 Appendix D - New Information - Document 61, Jim Willison, Aiken, SC Page 1 of 5 Document 61, Jim Willison, Aiken, SC Page 2 of 5 D-163 DOE/EIS-0287 Idaho HLW & FD EIS - New Information - Document 61, Jim Willison, Aiken, SC Page 3 of 5 Document 61, Jim Willison, Aiken, SC Page 4 of 5 DOE/EIS-0287 D-164 Appendix D - New Information - Document 61, Jim Willison, Aiken, SC Page 5 of 5 Document 62, Shoshone-Bannock Tribes (Claudeo Broncho), Fort Hall, ID Page 1 of 7 D-165 DOE/EIS-0287 Idaho HLW

  2. app_d

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

    199 DOE/EIS-0287 Idaho HLW & FD EIS - New Information - Document 80, Melissa Clark Rhodes, Jackson, WY Page 11 of 19 Document 80, Melissa Clark Rhodes, Jackson, WY Page 12 of 19 DOE/EIS-0287 D-200 Appendix D - New Information - Document 80, Melissa Clark Rhodes, Jackson, WY Page 13 of 19 Document 80, Melissa Clark Rhodes, Jackson, WY Page 14 of 19 D-201 DOE/EIS-0287 Idaho HLW & FD EIS - New Information - Document 80, Melissa Clark Rhodes, Jackson, WY Page 15 of 19 Document 80, Melissa

  3. app_d

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

    203 DOE/EIS-0287 Idaho HLW & FD EIS - New Information - Document 80, Melissa Clark Rhodes, Jackson, WY Page 19 of 19 Document 81, Dennis Donnelly, Pocatello, ID Page 1 of 2 DOE/EIS-0287 D-204 Appendix D - New Information - Document 81, Dennis Donnelly, Pocatello, ID Page 2 of 2 Document 82, U.S. Department of the Interior (Preston A. Sleeger), Portland, OR, Page 1 of 3 D-205 DOE/EIS-0287 Idaho HLW & FD EIS - New Information - Document 82, U.S. Department of the Interior (Preston A.

  4. Selection of Pretreatment Processes for Removal of Radionuclides from Hanford Tank Waste

    SciTech Connect (OSTI)

    CARREON, R.

    2002-01-01

    The U.S. Department of Energy's (DOE's), Office of River Protection (ORP) located at Hanford Washington has established a contract (1) to design, construct, and commission a new Waste Treatment and Immobilization Plant (WTP) that will treat and immobilize the Hanford tank wastes for ultimate disposal. The WTP is comprised of four major elements, pretreatment, LAW immobilization, HLW immobilization, and balance of plant facilities. This paper describes the technologies selected for pretreatment of the LAW and HLW tank wastes, how these technologies were selected, and identifies the major technology testing activities being conducted to finalize the design of the WTP.

  5. ch_2

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

    0 2.0 B B ackgr ackgr ound ound The Idaho National Engineering and Environmental Laboratory (INEEL) cur- rently manages waste associated with the processing of spent nuclear reactor fuel, including high-level waste (HLW). This waste is being managed to reduce the risk to human health and the environment. This Environmental Impact Statement (often referred to as the Idaho HLW & FD EIS or simply "this EIS") describes tech- nologies and methods the U.S. Department of Energy (DOE) is

  6. ch_2

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

    10 Background 2.2 High-Level Waste Overview 2.2.1 HIGH-LEVEL WASTE DESCRIPTION According to Section 2(12) of the Nuclear Waste Policy Act (42 USC 10101), high-level radioac- tive waste means: In July 1999, DOE issued Order 435.1 Radioactive Waste Management. This Order and its associated Manual and Guidance set forth the authorities, responsibilities, and requirements for the management of DOE's inventory of HLW, transuranic waste, and low-level waste. Specific to HLW, DOE uses the Nuclear Waste

  7. ch_3

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

    13 DOE/EIS-0287 Idaho HLW & FD EIS except the pillar and panel tanks) would be full of mixed transuranic waste in approximately 2017. Other facilities depending on the capacity of the Tank Farm for operation eventually would be shut down due to their inability to discharge liquid waste. Under this alternative, DOE would not meet its commitment to cease use of the Tank Farm by 2012 or to make its mixed HLW road ready by 2035. Facilities required for the No Action Alternative include the bin

  8. Tank waste remediation system optimized processing strategy with an altered treatment scheme

    SciTech Connect (OSTI)

    Slaathaug, E.J.

    1996-03-01

    This report provides an alternative strategy evolved from the current Hanford Site Tank Waste Remediation System (TWRS) programmatic baseline for accomplishing the treatment and disposal of the Hanford Site tank wastes. This optimized processing strategy with an altered treatment scheme performs the major elements of the TWRS Program, but modifies the deployment of selected treatment technologies to reduce the program cost. The present program for development of waste retrieval, pretreatment, and vitrification technologies continues, but the optimized processing strategy reuses a single facility to accomplish the separations/low-activity waste (LAW) vitrification and the high-level waste (HLW) vitrification processes sequentially, thereby eliminating the need for a separate HLW vitrification facility.

  9. Full page fax print

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

    5 (U) April 26, 1995 Westinghouse Savannah River Company Savannah River S~e Aiken, SC 29808 HLW-OVP-95-0031 @ Westinghouse Savannah River Company. April 27, 1995 A. B. Scoa, Jr. Va Preoident and General...,. Hgh Level Was .. ManagamenI DiviIion P. O. Box 61' Allen, SC 29602 HLW-OVP-95-0031 Keywords: Retention: High Level W88te SyslemPIan Production Plan 25 years, offer to NARA at end of retention. Disposal Auth: DOEF 1324.5 95-0002 Track .: 7605 File Record: 773-52A, Records Mr. Steven D.

  10. High Level Waste Management Division High-Level Waste System Plan

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

    6 (U) December 20, 1995 Westinghouse Savannah River Company Savannah River Site Aiken, SC 29808 HLW-OVP-95-0102 Westinghouse Savannah River Company 2 0 GtC 1995 Mr. A. L. Watkins, Assistant Manager High Level Waste U. S. Department of Energy Savannah River Operations Office P. O. Box A Aiken, SC 29802 Dear Mr. Watkins: A. B. Scott, Jr. Vice President and General Manager High Level Waste Management Division P. O. Box 616 Aiken, SC 29802 HLW-OVP-95-0102 HIGH LEVEL WASTE SYSTEM PLAN. REVISION 6 (U)

  11. TESTING OF ENHANCED CHEMICAL CLEANING OF SRS ACTUAL WASTE TANK 5F AND TANK 12H SLUDGES

    SciTech Connect (OSTI)

    Martino, C.; King, W.

    2011-08-22

    Forty three of the High Level Waste (HLW) tanks at the Savannah River Site (SRS) have internal structures that hinder removal of the last approximately five thousand gallons of waste sludge solely by mechanical means. Chemical cleaning can be utilized to dissolve the sludge heel with oxalic acid (OA) and pump the material to a separate waste tank in preparation for final disposition. This dissolved sludge material is pH adjusted downstream of the dissolution process, precipitating the sludge components along with sodium oxalate solids. The large quantities of sodium oxalate and other metal oxalates formed impact downstream processes by requiring additional washing during sludge batch preparation and increase the amount of material that must be processed in the tank farm evaporator systems and the Saltstone Processing Facility. Enhanced Chemical Cleaning (ECC) was identified as a potential method for greatly reducing the impact of oxalate additions to the SRS Tank Farms without adding additional components to the waste that would extend processing or increase waste form volumes. In support of Savannah River Site (SRS) tank closure efforts, the Savannah River National Laboratory (SRNL) conducted Real Waste Testing (RWT) to evaluate an alternative to the baseline 8 wt. % OA chemical cleaning technology for tank sludge heel removal. The baseline OA technology results in the addition of significant volumes of oxalate salts to the SRS tank farm and there is insufficient space to accommodate the neutralized streams resulting from the treatment of the multiple remaining waste tanks requiring closure. ECC is a promising alternative to bulk OA cleaning, which utilizes a more dilute OA (nominally 2 wt. % at a pH of around 2) and an oxalate destruction technology. The technology is being adapted by AREVA from their decontamination technology for Nuclear Power Plant secondary side scale removal. This report contains results from the SRNL small scale testing of the ECC process using SRS sludge tank sample material. A Task Technical and Quality Assurance Plan (TTQAP) details the experimental plan as outlined by the Technical Task Request (TTR). The TTR identifies that the data produced by this testing and results included in this report will support the technical baseline with portions having a safety class functional classification. The primary goals for SRNL RWT are as follows: (1) to confirm ECC performance with real tank sludge samples, (2) to determine the impact of ECC on fate of actinides and the other sludge metals, and (3) to determine changes, if any, in solids flow and settling behavior.

  12. Final Report: RPP-WTP Semi-Integrated Pilot Plant

    SciTech Connect (OSTI)

    Duignan, M. R.; Adamson, D. J.; Calloway, T. B.; Fowley, M. D.; Qureshi, Z. H.; Steimke, J. L.; Williams, M. R.; Zamecnik, J. R.

    2005-06-01

    In August 2004 the last of the SIPP task testing ended--a task that formally began with the issuance of the RPP-WTP Test Specification in June 2003. The planning for the task was a major effort in itself and culminated with the input of all stakeholders, DOE, Bechtel National, Inc., Washington Group International, in October 2003 at Hanford, WA (Appendix A). This report documents the activities carried out as a result of that planning. Campaign IV, the fourth and final step towards the Semi-Integrated Pilot Plant (SIPP) task, conducted by the Savannah River National Laboratory (SRNL) at the Savannah River Site, was to take the several recycle streams produced in Campaign III, the third step of the task, and combine them with other simulated recycle and chosen waste streams. (Campaign III was fed recycles from Campaign II, as Campaign II was fed by Campaign I.) The combined stream was processed in a fashion that mimicked the pretreatment operations of the DOE River Protection Project--Waste Treatment and Immobilization Plant (RPP-WTP) with the exception of the Ion Exchange Process. The SIPP task is considered semi-integrated because it only deals with the pretreatment operations of the RPP-WTP. That is, the pilot plant starts by receiving waste from the tank farm and ends when waste is processed to the point of being sent for vitrification. The resulting pretreated LAW and HLW simulants produced by the SIPP were shipped to VSL (Vitreous State Laboratory) and successfully vitrified in pilot WTP melters. Within the SIPP task these steps are referred to as Campaigns and there were four Campaigns in all. Campaign I, which is completely different than other campaigns, subjected a simulant of Hanford Tank 241-AY-102/C-106 (AY102) waste to cross-flow ultrafiltration only and in that process several important recycle streams were produced as a result of washing the simulant and cleaning the cross-flow filter. These streams were fed to subsequent campaigns and that work was the subject of the issued Campaign I interim report (Duignan et al., 2004a or Appendix I-1). The streams created in Campaign I were used for Campaign II, and during Campaign II more of the same recycle streams were produced, with the addition of recycle streams created during the pilot-scale ion exchange unit operation (Duignan et al., 2004b or Appendix I-2). Campaign III used the recycles from Campaign II and was the first campaign to use all the recycle streams (Duignan et al., 2004c or Appendix I-3). The operation of each of the subsequent campaigns, i.e., II, III, and IV, while different from Campaign I, are very similar to each other, and can be best understood as the process of operating a series of Pretreatment Unit Operations in a somewhat prototypic manner. That is, while Campaign I studied the operation of a single, albeit important, Pretreatment Unit Operation, i.e., Ultrafiltration, subsequent campaigns were to study the four major unit operations that make-up the RPP-WTP Pretreatment Facility. They are: Waste Feed Evaporation Process (FEP), Ultrafiltration Process (UFP), Cesium Ion Exchange Process (CIX), and the Treated LAW Evaporation Process (TLP). Each of the campaigns operated basically as a separate subtask, but as with Campaign I, the recycle streams produced in one campaign were fed into the subsequent campaign. Therefore, all four campaigns were chemically connected through these recycle streams, which carry over effects of the preceding campaign. The results of Campaign IV operations are the subject of this fourth and final report. Separate reports were issued after each of the previous campaigns, but they were treated as interim because of being limited to the results obtained from a single campaign (or past campaigns) and further limited to only highlights of that single campaign. This final report not only discusses the Campaign IV results but compares those with the previous campaigns. Also included is a more comprehensive discussion of the overall task activities, as well as abridged versions of the full databases of the accumulated results and the equipment used during the year-long SIPP task.

  13. NITRATE DESTRUCTION LITERATURE SURVEY AND EVALUATION CRITERIA

    SciTech Connect (OSTI)

    Steimke, J.

    2011-02-01

    This report satisfies the initial phase of Task WP-2.3.4 Alternative Sodium Recovery Technology, Subtask 1; Develop Near-Tank Nitrate/Nitrite Destruction Technology. Some of the more common anions in carbon steel waste tanks at SRS and Hanford Site are nitrate which is corrosive, and nitrite and hydroxide which are corrosion inhibitors. At present it is necessary to periodically add large quantities of 50 wt% caustic to waste tanks. There are three primary reasons for this addition. First, when the contents of salt tanks are dissolved, sodium hydroxide preferentially dissolves and is removed. During the dissolution process the concentration of free hydroxide in the tank liquid can decrease from 9 M to less than 0.2 M. As a result, roughly half way through the dissolution process large quantities of sodium hydroxide must be added to the tank to comply with requirements for corrosion control. Second, hydroxide is continuously consumed by reaction with carbon dioxide which occurs naturally in purge air used to prevent buildup of hydrogen gas inside the tanks. The hydrogen is generated by radiolysis of water. Third, increasing the concentration of hydroxide increases solubility of some aluminum compounds, which is desirable in processing waste. A process that converts nitrate and nitrite to hydroxide would reduce certain costs. (1) Less caustic would be purchased. (2) Some of the aluminum solid compounds in the waste tanks would become more soluble so less mass of solids would be sent to High Level Vitrification and therefore it would be not be necessary to make as much expensive high level vitrified product. (3) Less mass of sodium would be fed to Saltstone at SRS or Low Level Vitrification at Hanford Site so it would not be necessary to make as much low level product. (4) At SRS less nitrite and nitrate would be sent to Defense Waste Processing Facility (DWPF) so less formic acid would be consumed there and less hydrogen gas would be generated. This task involves literature survey of technologies to perform the nitrate to hydroxide conversion, selection of the most promising technologies, preparation of a flowsheet and design of a system. The most promising technologies are electrochemical reduction of nitrates and chemical reduction with hydrogen or ammonia. The primary reviewed technologies are listed and they aredescribed in more detail later in the report: (1) Electrochemical destruction; (2) Chemical reduction with agents such as ammonia, hydrazine or hydrogen; (3) Hydrothermal reduction process; and (4) Calcination. Only three of the technologies on the list have been demonstrated to generate usable amounts of caustic; electrochemical reduction and chemical reduction with ammonia, hydrazine or hydrogen and hydrothermal reduction. Chemical reduction with an organic reactant such as formic acid generates carbon dioxide which reacts with caustic and is thus counterproductive. Treatment of nitrate with aluminum or other active metals generates a solid product. High temperature calcination has the potential to generate sodium oxide which may be hydrated to sodium hydroxide, but this is unproven. The following criteria were developed to evaluate the most suitable option. The numbers in brackets after the criteria are relative weighting factors to account for importance: (1) Personnel exposure to radiation for installation, routine operation and maintenance; (2) Non-radioactive safety issues; (3) Whether the technology generates caustic and how many moles of caustic are generated per mole of nitrate plus nitrite decomposed; (4) Whether the technology can handle nitrate and nitrite at the concentrations encountered in waste; (5) Maturity of technology; (6) Estimated annual cost of operation (labor, depreciation, materials, utilities); (7) Capital cost; (8) Selectivity to nitrogen as decomposition product (other products are flammable and/or toxic); (9) Impact of introduced species; (10) Selectivity for destruction of nitrate vs. nitrite; and (11) Cost of deactivation and demolition. Each technology was given a score from one

  14. Bases, Assumptions, and Results of the Flowsheet Calculations for the Decision Phase Salt Disposition Alternatives

    SciTech Connect (OSTI)

    Elder, H.H.

    2001-07-11

    The HLW salt waste (salt cake and supernate) now stored at the SRS must be treated to remove insoluble sludge solids and reduce the soluble concentration of radioactive cesium radioactive strontium and transuranic contaminants (principally Pu and Np). These treatments will enable the salt solution to be processed for disposal as saltstone, a solid low-level waste.

  15. Preliminary waste acceptance criteria for the ICPP spent fuel and waste management technology development program

    SciTech Connect (OSTI)

    Taylor, L.L.; Shikashio, R.

    1993-09-01

    The purpose of this document is to identify requirements to be met by the Producer/Shipper of Spent Nuclear Fuel/High-LeveL Waste SNF/HLW in order for DOE to be able to accept the packaged materials. This includes defining both standard and nonstandard waste forms.

  16. GLASS FORMULATION FOR THE HANFORD TANK WASTE TREATMENT AND IMMOBILIZATION PLANT (WTP)

    SciTech Connect (OSTI)

    KRUGER AA; VIENNA JD; KIM DS; JAIN V

    2009-05-27

    A computational method for formulating Hanford HLW glasses was developed that is based on empirical glass composition-property models, accounts for all associated uncertainties, and can be solved in Excel{sup R} in minutes. Calculations for all waste form processing and compliance requirements included. Limited experimental validation performed.

  17. Evaluation of Options for Permanent Geologic Disposal of Spent NuclearFuel and High-Level Radioactive Waste

    Broader source: Energy.gov [DOE]

    [In Support of a Comprehensive National Nuclear Fuel Cycle Strategy, Volumes I and II (Appendices)] This study provides a technical basis for informing policy decisions regarding strategies for the management and permanent disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW) in the United States requiring geologic isolation.

  18. US Department of Energy Storage of Spent Fuel and High Level Waste

    SciTech Connect (OSTI)

    Sandra M Birk

    2010-10-01

    ABSTRACT This paper provides an overview of the Department of Energy's (DOE) spent nuclear fuel (SNF) and high level waste (HLW) storage management. Like commercial reactor fuel, DOE's SNF and HLW were destined for the Yucca Mountain repository. In March 2010, the DOE filed a motion with the Nuclear Regulatory Commission (NRC) to withdraw the license application for the repository at Yucca Mountain. A new repository is now decades away. The default for the commercial and DOE research reactor fuel and HLW is on-site storage for the foreseeable future. Though the motion to withdraw the license application and delay opening of a repository signals extended storage, DOE's immediate plans for management of its SNF and HLW remain the same as before Yucca Mountain was designated as the repository, though it has expanded its research and development efforts to ensure safe extended storage. This paper outlines some of the proposed research that DOE is conducting and will use to enhance its storage systems and facilities.

  19. Contract No. DE-AC27-OIRV14136 Modification No. A185 SF-30 Continuatio...

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

    1111243 5,987,799.34 HLW 1250 1111244 198,741,241.83 PT 1250 1111245 324,194,585.43 PD 1250 1110462 300,000.00 Budgetary Control Points Thru A180 6,202,247,873.17 Page 2 of...

  20. Generic Deep Geologic Disposal Safety Case

    Broader source: Energy.gov [DOE]

    The Generic Deep Geologic Disposal Safety Case presents generic information that is of use in understanding potential deep geologic disposal options (e.g., salt, shale, granite, deep borehole) in the U.S. for used nuclear fuel (UNF) from reactors and high-level radioactive waste (HLW).

  1. Crystallization in high-level waste glass: A review of glass theory and noteworthy literature

    SciTech Connect (OSTI)

    Christian, J. H.

    2015-08-01

    There is a fundamental need to continue research aimed at understanding nepheline and spinel crystal formation in high-level waste (HLW) glass. Specifically, the formation of nepheline solids (K/NaAlSiO₄) during slow cooling of HLW glass can reduce the chemical durability of the glass, which can cause a decrease in the overall durability of the glass waste form. The accumulation of spinel solids ((Fe, Ni, Mn, Zn)(Fe,Cr)₂O₄), while not detrimental to glass durability, can cause an array of processing problems inside of HLW glass melters. In this review, the fundamental differences between glass and solid-crystals are explained using kinetic, thermodynamic, and viscosity arguments, and several highlights of glass-crystallization research, as it pertains to high-level waste vitrification, are described. In terms of mitigating spinel in the melter and both spinel and nepheline formation in the canister, the complexity of HLW glass and the intricate interplay between thermal, chemical, and kinetic factors further complicates this understanding. However, new experiments seeking to elucidate the contributing factors of crystal nucleation and growth in waste glass, and the compilation of data from older experiments, may go a long way towards helping to achieve higher waste loadings while developing more efficient processing strategies.

  2. Performance Enhancements to the Hanford Waste Treatment and Immobilization Plant Low-Activity Waste Vitrification System

    SciTech Connect (OSTI)

    Hamel, W. F. [Office of River Protection, U.S. Department of Energy, 2400 Stevens Drive, Richland, WA 99354 (United States); Gerdes, K. [U.S. Department of Energy, 19901 Germantown Road, Germantown, MD 20874 (United States); Holton, L. K. [Pacific Northwest National Laboratory, PO Box 999, Richland, WA 99352 (United States); Pegg, I.L. [Vitreous State Laboratory, The Catholic University of America, 620 Michigan Avenue NE, Washington, DC 20064 (United States); Bowan, B.W. [Duratek, Inc., 10100 Old Columbia Road, Columbia, Maryland 21046 (United States)

    2006-07-01

    The U.S Department of Energy Office of River Protection (DOE-ORP) is constructing a Waste Treatment and Immobilization Plant (WTP) for the treatment and vitrification of underground tank wastes stored at the Hanford Site in Washington State. The WTP comprises four major facilities: a pretreatment facility to separate the tank waste into high level waste (HLW) and low-activity waste (LAW) process streams, a HLW vitrification facility to immobilize the HLW fraction; a LAW vitrification facility to immobilize the LAW fraction, and an analytical laboratory to support the operations of all four treatment facilities. DOE has established strategic objectives to optimize the performance of the WTP facilities and the LAW and HLW waste forms to reduce the overall schedule and cost for treatment and vitrification of the Hanford tank wastes. This strategy has been implemented by establishing performance expectations in the WTP contract for the facilities and waste forms. In addition, DOE, as owner-operator of the WTP facilities, continues to evaluate 1) the design, to determine the potential for performance above the requirements specified in the WTP contract; and 2) improvements in production of the LAW and HLW waste forms. This paper reports recent progress directed at improving production of the LAW waste form. DOE's initial assessment, which is based on the work reported in this paper, is that the treatment rate of the WTP LAW vitrification facility can be increased by a factor of 2 to 4 with a combination of revised glass formulations, modest increases in melter glass operating temperatures, and a second-generation LAW melter with a larger surface area. Implementing these improvements in the LAW waste immobilization capability can benefit the LAW treatment mission by reducing the cost of waste treatment. (authors)

  3. US DOE Initiated Performance Enhancements to the Hanford Waste Treatment and Immobilization Plant (WTP) Low-activity Waste Vitrification (LAW) System

    SciTech Connect (OSTI)

    Hamel, William F.; Gerdes, Kurt D.; Holton, Langdon K.; Pegg, Ian L.; Bowen, Brad W.

    2006-03-03

    The U.S Department of Energy Office of River Protection (DOE-ORP) is constructing a Waste Treatment and Immobilization Plant (WTP) for the treatment and vitrification of underground tank wastes stored at the Hanford Site in Washington State. The WTP comprises four major facilities: a pretreatment facility to separate the tank waste into high level waste (HLW) and low-activity waste (LAW) process streams, a HLW vitrification facility to immobilize the HLW fraction; a LAW vitrification facility to immobilize the LAW fraction, and an analytical laboratory to support the operations of all four treatment facilities. DOE has established strategic objectives to optimize the performance of the WTP facilities and the LAW and HLW waste forms to reduce the overall schedule and cost for treatment and vitrification of the Hanford tank wastes. This strategy has been implemented by establishing performance expectations in the WTP contract for the facilities and waste forms. In addition, DOE, as owner-operator of the WTP facilities, continues to evaluate 1) the design, to determine the potential for performance above the requirements specified in the WTP contract; and 2) improvements in production of the LAW and HLW waste forms. This paper reports recent progress directed at improving production of the LAW waste form. DOEs initial assessment, which is based on the work reported in this paper, is that the capacity of the WTP LAW vitrification facility can be increased by a factor of 2 to 4 with a combination of revised glass formulations, modest increases in melter glass operating temperatures, and a second-generation LAW melter with a larger surface area. Implementing these improvements in the LAW waste immobilization capability can benefit the LAW treatment mission by reducing both processing time and cost.

  4. Vitrification of high level nuclear waste inside ambient temperature disposal containers using inductive heating: The SMILE system

    SciTech Connect (OSTI)

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

    1996-03-01

    A new approach, termed SMILE (Small Module Inductively Loaded Energy), for the vitrification of high level nuclear wastes (HLW) is described. Present vitrification systems liquefy the HLW solids and associated frit material in large high temperature melters. The molten mix is then poured into small ({approximately}1 m{sup 3}) disposal canisters, where it solidifies and cools. SMILE eliminates the separate, large high temperature melter. Instead, the BLW solids and frit melt inside the final disposal containers, using inductive heating. The contents then solidify and cool in place. The SMILE modules and the inductive heating process are designed so that the outer stainless can of the module remains at near ambient temperature during the process cycle. Module dimensions are similar to those of present disposal containers. The can is thermally insulated from the high temperature inner container by a thin layer of refractory alumina firebricks. The inner container is a graphite crucible lined with a dense alumina refractory that holds the HLW and fiit materials. After the SMILE module is loaded with a slurry of HLW and frit solids, an external multi-turn coil is energized with 30-cycle AC current. The enclosing external coil is the primary of a power transformer, with the graphite crucible acting as a single turn ``secondary.`` The induced current in the ``secondary`` heats the graphite, which in turn heats the HLW and frit materials. The first stage of the heating process is carried out at an intermediate temperature to drive off remnant liquid water and water of hydration, which takes about 1 day. The small fill/vent tube to the module is then sealed off and the interior temperature raised to the vitrification range, i.e., {approximately}1200C. Liquefaction is complete after approximately 1 day. The inductive heating then ceases and the module slowly loses heat to the environment, allowing the molten material to solidify and cool down to ambient temperature.

  5. PROCESSING ALTERNATIVES FOR DESTRUCTION OF TETRAPHENYLBORATE

    SciTech Connect (OSTI)

    Lambert, D; Thomas Peters, T; Samuel Fink, S

    2007-02-27

    Two processes were chosen in the 1980's at the Savannah River Site (SRS) to decontaminate the soluble High Level Waste (HLW). The In Tank Precipitation (ITP) process (1,2) was developed at SRS for the removal of radioactive cesium and actinides from the soluble HLW. Sodium tetraphenylborate was added to the waste to precipitate cesium and monosodium titanate (MST) was added to adsorb actinides, primarily uranium and plutonium. Two products of this process were a low activity waste stream and a concentrated organic stream containing cesium tetraphenylborate and actinides adsorbed on monosodium titanate (MST). A copper catalyzed acid hydrolysis process was built to process (3, 4) the Tank 48H cesium tetraphenylborate waste in the SRS's Defense Waste Processing Facility (DWPF). Operation of the DWPF would have resulted in the production of benzene for incineration in SRS's Consolidated Incineration Facility. This process was abandoned together with the ITP process in 1998 due to high benzene in ITP caused by decomposition of excess sodium tetraphenylborate. Processing in ITP resulted in the production of approximately 1.0 million liters of HLW. SRS has chosen a solvent extraction process combined with adsorption of the actinides to decontaminate the soluble HLW stream (5). However, the waste in Tank 48H is incompatible with existing waste processing facilities. As a result, a processing facility is needed to disposition the HLW in Tank 48H. This paper will describe the process for searching for processing options by SRS task teams for the disposition of the waste in Tank 48H. In addition, attempts to develop a caustic hydrolysis process for in tank destruction of tetraphenylborate will be presented. Lastly, the development of both a caustic and acidic copper catalyzed peroxide oxidation process will be discussed.

  6. Preconceptual Design Description for Caustic Recycle Facility

    SciTech Connect (OSTI)

    Sevigny, Gary J.; Poloski, Adam P.; Fountain, Matthew S.; Kurath, Dean E.

    2008-04-12

    The U.S. Department of Energy plans to vitrify both high-level and low-activity waste at the Hanford Site in southeastern Washington State. One aspect of the planning includes a need for a caustic recycle process to separate sodium hydroxide for recycle. Sodium is already a major limitation to the waste-oxide loading in the low-activity waste glass to be vitrified at the Waste Treatment Plant, and additional sodium hydroxide will be added to remove aluminum and to control precipitation in the process equipment. Aluminum is being removed from the high level sludge to reduce the number of high level waste canisters produced. A sodium recycle process would reduce the volume of low-activity waste glass produced and minimize the need to purchase new sodium hydroxide, so there is a renewed interest in investigating sodium recycle. This document describes an electrochemical facility for recycling sodium for the WTP.

  7. 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.

  8. Development of a glass polymer composite sewer pipe from waste glass. Final report

    SciTech Connect (OSTI)

    Rayfiel, R.; Kukacka, L.E.

    1980-02-01

    A range of polymer-aggregate composites for applications in industry which appear to be economically attractive and contribute to energy conservation were developed at BNL. Waste glass is the aggregate in one such material, which is called glass-polymer-composite (GPC). This report assays the economics and durability of GPC in piping for storm drains and sewers. The properties of the pipe are compared statistically with the requirements of industrial specifications. These establish the raw materials requirements. The capital and operating costs for producing pipe are then estimated. Using published sales values for competing materials, the return on investment is calculated for two cases. The ultimate energy requirement of the raw materials in GPC is compared with the corresponding requirement for vitrified clay pipe. The strengths of GPC, reinforced concrete, vitrified clay and asbestos cement pipe are compared after extended exposure to various media. The status of process and product development is reviewed and recommendations are made for future work.

  9. Submerged combustion melting processes for producing glass and similar materials, and systems for carrying out such processes

    SciTech Connect (OSTI)

    Charbonneau, Mark William

    2015-08-04

    Processes of controlling submerged combustion melters, and systems for carrying out the methods. One process includes feeding vitrifiable material into a melter vessel, the melter vessel including a fluid-cooled refractory panel in its floor, ceiling, and/or sidewall, and heating the vitrifiable material with a burner directing combustion products into the melting zone under a level of the molten material in the zone. Burners impart turbulence to the molten material in the melting zone. The fluid-cooled refractory panel is cooled, forming a modified panel having a frozen or highly viscous material layer on a surface of the panel facing the molten material, and a sensor senses temperature of the modified panel using a protected thermocouple positioned in the modified panel shielded from direct contact with turbulent molten material. Processes include controlling the melter using the temperature of the modified panel. Other processes and systems are presented.

  10. Hanford Dangerous Waste Permit

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

    Integrated Disposal Facility Operating Unit #11 Aerial view of IDF looking south. Note semi-truck trailer for scale. There are risks to groundwater in the future from secondary waste, according to modeling. Secondary waste would have to be significantly mitigated before it could be disposed at IDF. Where did the waste come from? No waste is stored here yet. IDF will receive vitrified waste when the Waste Treatment Plant starts operating. It may also receive secondary waste resulting from

  11. SETTLING OF SPINEL IN A HIGH-LEVEL WASTE GLASS MELTER

    SciTech Connect (OSTI)

    Pavel Hrma; Pert Schill; Lubomir Nemec

    2002-01-07

    High-level nuclear waste is being vitrified, i.e., converted to a durable glass that can be stored in a safe repository for hundreds of thousands of years. Waste vitrification is accomplished in reactors called melters to which the waste is charged together with glass-forming additives. The mixture is electrically heated to a temperature as high as 1150 decrees C to create a melt that becomes glass on cooling.

  12. Audit Report: IG-0894 | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    4 Audit Report: IG-0894 September 30, 2013 Department of Energy Quality Assurance: Design Control for the Waste Treatment and Immobilization Plant at the Hanford Site The Department is constructing the $12.2 billion Waste Treatment and Immobilization Plant (WTP) to vitrify approximately 56 million gallons of radioactive and chemically hazardous waste stored at the Hanford Site. To ensure the vitrification process is safe for workers, the public and the environment, the Department required the

  13. Identification and summary characterization of materials potentially requiring vitrification: Background information

    SciTech Connect (OSTI)

    Croff, A.G.

    1996-05-13

    This document contains background information for the Workshop in general and the presentation entitled `Identification and Summary Characterization of Materials Potentially Requiring Vitrification` that was given during the first morning of the workshop. summary characteristics of 9 categories of US materials having some potential to be vitrified are given. This is followed by a 1-2 page elaborations for each of these 9 categories. References to more detailed information are included.

  14. Summary - Demonstration Bulk Vitrification System (DBVS) for Low-Actvity Waste at Hanford

    Office of Environmental Management (EM)

    DBVS ETR Report Date: September 2006 ETR-3 United States Department of Energy Office of Environmental Management (DOE-EM) External Technical Review of the Demonstration Bulk Vitrification System (DBVS) for Low Activity Waste (LAW) at Hanford Why DOE-EM Did This Review The Department of Energy (DOE) is charged with the safe retrieval, treatment and disposal of 53 million gallons of Hanford radioactive waste. The Waste Treatment Plant (WTP) is being designed to treat and vitrify the High Level

  15. Manufacture of ceramic tiles from fly ash

    DOE Patents [OSTI]

    Hnat, James G.; Mathur, Akshay; Simpson, James C.

    1999-01-01

    The present invention relates to a process for forming glass-ceramic tiles. Fly ash containing organic material, metal contaminants, and glass forming materials is oxidized under conditions effective to combust the organic material and partially oxidize the metallic contaminants and the glass forming materials. The oxidized glass forming materials are vitrified to form a glass melt. This glass melt is then formed into tiles containing metallic contaminants.

  16. Manufacture of ceramic tiles from fly ash

    DOE Patents [OSTI]

    Hnat, J.G.; Mathur, A.; Simpson, J.C.

    1999-08-10

    The present invention relates to a process for forming glass-ceramic tiles. Fly ash containing organic material, metal contaminants, and glass forming materials is oxidized under conditions effective to combust the organic material and partially oxidize the metallic contaminants and the glass forming materials. The oxidized glass forming materials are vitrified to form a glass melt. This glass melt is then formed into tiles containing metallic contaminants. 6 figs.

  17. ADVANCED VITRIFICATION SYSTEM (RIC AVS) RESEARCH AND DEVELOPMENT PROJECT

    SciTech Connect (OSTI)

    J.R. Powell; M. Reich

    2003-06-30

    The objective of this AVS testing program is to use bench-scale test equipment to produce a vitrified product at maximum waste loading from the specified AZ-101 waste simulant and conduct a TTT analysis using laboratory scale melts to show compliance with the DOE Waste Acceptance Product Specifications for Vitrified High-Level Waste Forms (WAPS). The vitrified product complies with the following WAPS. A borosilicate glass with a waste loading of 60.9-wt% was produced from a slurry feed of AZ101 simulant. Glass durability testing, glass characterization testing, and testing methodology were performed in accordance with the Department of Energy approved Test Plan. The glass has two crystalline phases and good uniformity of composition. The Product Consistency Test on the 6 location-specific samples are at least 1 to 2 orders of magnitude below the mean PCT results for the EA glass. Standard deviations were less than 10% of measured values. The glass transition temperature averaged 658 {+-} 9 C. A TTT diagram was produced. There was measured cesium loss of about 2%, and compliance with the Universal Treatment Standards.

  18. Used fuel disposition campaign international activities implementation plan.

    SciTech Connect (OSTI)

    Nutt, W. M. (Nuclear Engineering Division)

    2011-06-29

    The management of used nuclear fuel and nuclear waste is required for any country using nuclear energy. This includes the storage, transportation, and disposal of low and intermediate level waste (LILW), used nuclear fuel (UNF), and high level waste (HLW). The Used Fuel Disposition Campaign (UFDC), within the U.S. Department of Energy (DOE), Office of Nuclear Energy (NE), Office of Fuel Cycle Technology (FCT), is responsible for conducting research and development pertaining to the management of these materials in the U.S. Cooperation and collaboration with other countries would be beneficial to both the U.S. and other countries through information exchange and a broader participation of experts in the field. U.S. participation in international UNF and HLW exchanges leads to safe management of nuclear materials, increased security through global oversight, and protection of the environment worldwide. Such interactions offer the opportunity to develop consensus on policy, scientific, and technical approaches. Dialogue to address common technical issues helps develop an internationally recognized foundation of sound science, benefiting the U.S. and participating countries. The UNF and HLW management programs in nuclear countries are at different levels of maturity. All countries utilizing nuclear power must store UNF, mostly in wet storage, and HLW for those countries that reprocess UNF. Several countries either utilize or plan to utilize dry storage systems for UNF, perhaps for long periods of time (several decades). Geologic disposal programs are at various different states, ranging from essentially 'no progress' to selected sites and pending license applications to regulators. The table below summarizes the status of UNF and HLW management programs in several countriesa. Thus, the opportunity exists to collaborate at different levels ranging from providing expertise to those countries 'behind' the U.S. to obtaining access to information and expertise from those countries with more mature programs. The U.S. fuel cycle is a once through fuel cycle involving the direct disposal of UNF, as spent nuclear fuel, in a geologic repository (previously identified at Yucca Mountain, Nevada), following at most a few decades of storage (wet and dry). The geology at Yucca Mountain, unsaturated tuff, is unique among all countries investigating the disposal of UNF and HLW. The decision by the U.S. Department of Energy to no longer pursue the disposal of UNF at Yucca Mountain and possibly utilize very long term storage (approaching 100 years or more) while evaluating future fuel cycle alternatives for managing UNF, presents a different UNF and HLW management R&D portfolio that has been pursued in the U.S. In addition, the research and development activities managed by OCRWM have been transferred to DOE-NE. This requires a reconsideration of how the UFDC will engage in cooperative and collaborative activities with other countries. This report presents the UFDC implementation plan for international activities. The DOE Office of Civilian Radioactive Waste Management (OCRWM) has cooperated and collaborated with other countries in many different 'arenas' including the Nuclear Energy Agency (NEA) within the Organization for Economic Co-operation and Development (OECD), the International Atomic Energy Agency (IAEA), and through bilateral agreements with other countries. These international activities benefited OCRWM through the acquisition and exchange of information, database development, and peer reviews by experts from other countries. DOE-NE cooperates and collaborates with other countries in similar 'arenas' with similar objectives and realizing similar benefits. However the DOE-NE focus has not typically been in the area of UNF and HLW management. This report will first summarize these recent cooperative and collaborative activities. The manner that the UFDC will cooperate and collaborate in the future is expected to change as R&D is conducted regarding long-term storage and the potential disposal of UNF and HLW in different geolo

  19. INEEL Summary on Calcination

    SciTech Connect (OSTI)

    Gombert, Dirk

    2003-12-01

    Irradiated nuclear fuel reprocessing to recover 235U and 80Kr began at the INEEL in 1953. The resulting acidic high-level liquid radioactive waste (HLW) was stored in stainless steel tanks in underground concrete vaults. A fluidized-bed calcination process was developed during the 1950s to form a granular calcine solid from the acidic HLW with a seven-fold volume reduction. An engineering-scale demonstration, the Waste Calcining Facility (WCF) was constructed and operated in 1963. After the successful demonstration of the process, the WCF was continued as a production facility through 1981, Calcining 15,000 m3 of HLW to 2,160 m3 of calcine.1 The New Waste Calcining Facility (NWCF) was designed and constructed based on the operating experience of the WCF, and began operation in 1982. With a rated capacity of 3,000 gallons/day, the NWCF continued waste processing operations through May of 2000, resulting in an additional 2,226 m3 of calcine (total current inventory of 4,386 m3).2 During waste processing at the NWCF, sodium-bearing waste (SBW) from decontamination activities was blended with HLW to minimize alkali (sodium and potassium) concentrations in the calciner feed solution. This was necessary due to the propensity of sodium and potassium nitrates to melt in the calciner, causing the bed to agglomerate and interfere with fluidization. However, near the end of HLW processing, work was initiated to modify the calcination process to treat SBW directly, blending it with chemical additives such as aluminum nitrate rather than lower alkali content HLW liquids. The result of this development effort was to increase the operating temperature of the calciner from 500°C to 600°C. The 600°C SBW flowsheet was successfully demonstrated at the NWCF during two separate trials during 1999 and 2000.3, 4 The conclusion from these demonstrations was that operating the existing NWCF at 600°C is a viable method for solidifying SBW, and this concept is currently being evaluated as one option for preparing the SBW for disposal. To restart the NWCF for SBW processing, applicable environmental waste processing and/or air permits will be required. It is anticipated that the NWCF will be regulated as an incinerator; thus, compliance with maximum achievable control technology (MACT) emission limits will be required. To meet these stringent standards, an upgrade to the off-gas treatment train will be required. Specifically, emission of CO, total hydrocarbons (THC), and mercury must be mitigated. In addition, NOx abatement is necessary to eliminate interferences with the instrumentation required for air monitoring to demonstrate compliance.

  20. FINAL REPORT SUMMARY OF DM 1200 OPERATION AT VSL VSL-06R6710-2 REV 0 9/7/06

    SciTech Connect (OSTI)

    KRUGER AA; MATLACK KS; DIENER G; BARDAKCI T; PEGG IL

    2011-12-29

    The principal objective of this report was to summarize the testing experience on the DuraMelter 1200 (DMI200), which is the High Level Waste (HLW) Pilot Melter located at the Vitreous State Laboratory (VSL). Further objectives were to provide descriptions of the history of all modifications and maintenance, methods of operation, problems and unit failures, and melter emissions and performance while processing a variety of simulated HL W and low activity waste (LAW) feeds for the Hanford Waste Treatment and Immobilization Plant (WTP) and employing a variety of operating methods. All of these objectives were met. The River Protection Project - Hanford Waste Treatment and Immobilization Plant (RPP-WTP) Project has undertaken a 'tiered' approach to vitrification development testing involving computer-based glass formulation, glass property-composition models, crucible melts, and continuous melter tests of increasing, more realistic scales. Melter systems ranging from 0.02 to 1.2 m{sup 2} installed at the Vitreous State Laboratory (VSL) have been used for this purpose, which, in combination with the 3.3 m{sup 2} low activity waste (LAW) Pilot Melter at Duratek, Inc., span more than two orders of magnitude in melt surface area. In this way, less-costly small-scale tests can be used to define the most appropriate tests to be conducted at the larger scales in order to extract maximum benefit from the large-scale tests. For high level waste (HLW) vitrification development, a key component in this approach is the one-third scale DuraMelter 1200 (DM 1200), which is the HLW Pilot Melter that has been installed at VSL with an integrated prototypical off-gas treatment system. That system replaced the DM1000 system that was used for HLW throughput testing during Part B1. Both melters have similar melt surface areas (1.2 m{sup 2}) but the DM1200 is prototypical of the present RPP-WTP HLW melter design whereas the DM1000 was not. In particular, the DM1200 provides for testing on a vitrification system with the specific train of unit operations that has been selected for both HLW and LAW RPP-WTP off-gas treatment.

  1. Production of Synroc ceramics from titanate gel microspheres

    SciTech Connect (OSTI)

    Sizgek, E.; Bartlett, J.R.; Woolfrey, J.L.; Vance, E.R.

    1994-12-31

    Synroc is a multi-component titanate ceramic, designed to immobilise High Level Waste (HLW) from nuclear fuel reprocessing plants. Synroc precursor powders have been previously produced by various methods, such as oxide and alkoxide-hydrolysis routes. However, various technological aspects of HLW processing make the use of free-flowing, dust-free, highly sinterable precursor powders desirable. Such powders have been produced by spray-drying colloidal precursors, yielding microspherical particles with controlled porosity. These particles were readily impregnated with 20 wt% simulated high-level nuclear waste solutions, calcined at 1023 K and subsequently hot-pressed to produce dense Synroc monoliths. This paper discusses the preparation and fabrication of Synroc monoliths from the microspheres and their physical properties. The resulting microstructures and leaching characteristics of the Synroc monoliths are also presented.

  2. Characteristics of potential repository wastes. Volume 1

    SciTech Connect (OSTI)

    Not Available

    1992-07-01

    This document, and its associated appendices and microcomputer (PC) data bases, constitutes the reference OCRWM data base of physical and radiological characteristics data of radioactive wastes. This Characteristics Data Base (CDB) system includes data on spent nuclear fuel and high-level waste (HLW), which clearly require geologic disposal, and other wastes which may require long-term isolation, such as sealed radioisotope sources. The data base system was developed for OCRWM by the CDB Project at Oak Ridge National Laboratory. Various principal or official sources of these data provided primary information to the CDB Project which then used the ORIGEN2 computer code to calculate radiological properties. The data have been qualified by an OCRWM-sponsored peer review as suitable for quality-affecting work meeting the requirements of OCRWM`s Quality Assurance Program. The wastes characterized in this report include: light-water reactor (LWR) spent fuel and immobilized HLW.

  3. RADIOISOTOPE INVENTORY FOR TSPA-SR

    SciTech Connect (OSTI)

    C. Leigh; R. Rechard

    2001-01-30

    The total system performance assessment for site recommendation (TSPA-SR), on Yucca Mountain, as a site (if suitable) for disposal of radioactive waste, consists of several models. The Waste Form Degradation Model (i.e, source term) of the TSPA-SR, in turn, consists of several components. The Inventory Component, discussed here, defines the inventory of 26 radioisotopes for three representative waste categories: (1) commercial spent nuclear fuel (CSNF), (2) US Department of Energy (DOE) spent nuclear fuel (DSNF), and (3) high-level waste (HLW). These three categories are contained and disposed of in two types of waste packages (WPs)--CSNF WPs and co-disposal WPs, with the latter containing both DSNF and HLW. Three topics are summarized in this paper: first, the transport of radioisotopes evaluated in the past; second, the development of the inventory for the two WP types; and third, the selection of the most important radioisotopes to track in TSPA-SR.

  4. Mixing of process heels, process solutions, and recycle streams: Results of the small-scale radioactive tests

    SciTech Connect (OSTI)

    GJ Lumetta; JP Bramson; OT Farmer III; LR Greenwood; FV Hoopes; MA Mann; MJ Steele; RT Steele; RG Swoboda; MW Urie

    2000-05-17

    Various recycle streams will be combined with the low-activity waste (LAW) or the high-level waste (HLW) feed solutions during the processing of the Hanford tank wastes by BNFL, Inc. In addition, the LAW and HLW feed solutions will also be mixed with heels present in the processing equipment. This report describes the results of a test conducted by Battelle to assess the effects of mixing specific process streams. Observations were made regarding adverse reactions (mainly precipitation) and effects on the Tc oxidation state (as indicated by K{sub d} measurements with SuperLig{reg_sign} 639). The work was conducted according to test plan BNFL-TP-29953-023, Rev. 0, Small Scale Mixing of Process Heels, Solutions, and Recycle Streams. The test went according to plan, with only minor deviations from the test plan. The deviations from the test plan are discussed in the experimental section.

  5. Statement of work for conceptual design of solidified high-level waste interim storage system project (phase I)

    SciTech Connect (OSTI)

    Calmus, R.B., Westinghouse Hanford

    1996-12-17

    The U.S. Department of Energy (DOE) has embarked upon a course to acquire Hanford Site tank waste treatment and immobilization services using privatized facilities. This plan contains a two phased approach. Phase I is a ``proof-of-principle/commercial demonstration- scale`` effort and Phase II is a full-scale production effort. In accordance with the planned approach, interim storage (IS) and disposal of various products from privatized facilities are to be DOE furnished. The path forward adopted for Phase I solidification HLW IS entails use of Vaults 2 and 3 in the Spent Nuclear Fuel Canister Storage Building, to be located in the Hanford Site 200 East Area. This Statement of Work describes the work scope to be performed by the Architect-Engineer to prepare a conceptual design for the solidified HLW IS System.

  6. USE OF AN EQUILIBRIUM MODEL TO FORECAST DISSOLUTION EFFECTIVENESS, SAFETY IMPACTS, AND DOWNSTREAM PROCESSABILITY FROM OXALIC ACID AIDED SLUDGE REMOVAL IN SAVANNAH RIVER SITE HIGH LEVEL WASTE TANKS 1-15

    SciTech Connect (OSTI)

    KETUSKY, EDWARD

    2005-10-31

    This thesis details a graduate research effort written to fulfill the Magister of Technologiae in Chemical Engineering requirements at the University of South Africa. The research evaluates the ability of equilibrium based software to forecast dissolution, evaluate safety impacts, and determine downstream processability changes associated with using oxalic acid solutions to dissolve sludge heels in Savannah River Site High Level Waste (HLW) Tanks 1-15. First, a dissolution model is constructed and validated. Coupled with a model, a material balance determines the fate of hypothetical worst-case sludge in the treatment and neutralization tanks during each chemical adjustment. Although sludge is dissolved, after neutralization more is created within HLW. An energy balance determines overpressurization and overheating to be unlikely. Corrosion induced hydrogen may overwhelm the purge ventilation. Limiting the heel volume treated/acid added and processing the solids through vitrification is preferred and should not significantly increase the number of glass canisters.

  7. 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.

  8. 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.

  9. Engineered materials characterization report for the Yucca Mountain Site Characterization Project. Volume 2, Design data

    SciTech Connect (OSTI)

    Konynenburg, R.A.; McCright, R.D.; Roy, A.K.; Jones, D.A.

    1995-08-01

    This is Volume 2 of the Engineered Materials Characterization Report which presents the design data for candidate materials needed in fabricating different components for both large and medium multi-purpose canister (MPC) disposal containers, waste packages for containing uncanistered spent fuel (UCF), and defense high-level waste (HLW) glass disposal containers. The UCF waste package consists of a disposal container with a basket therein. It is assumed that the waste packages will incorporate all-metallic multibarrier disposal containers to accommodate medium and large MPCs, ULCF, and HLW glass canisters. Unless otherwise specified, the disposal container designs incorporate an outer corrosion-allowance metal barrier over an inner corrosion-resistant metal barrier. The corrosion-allowance barrier, which will be thicker than the inner corrosion-resistant barrier, is designed to undergo corrosion-induced degradation at a very low rate, thus providing the inner barrier protection from the near-field environment for a prolonged service period.

  10. Glass Property Data and Models for Estimating High-Level Waste Glass Volume

    SciTech Connect (OSTI)

    Vienna, John D.; Fluegel, Alexander; Kim, Dong-Sang; Hrma, Pavel R.

    2009-10-05

    This report describes recent efforts to develop glass property models that can be used to help estimate the volume of high-level waste (HLW) glass that will result from vitrification of Hanford tank waste. The compositions of acceptable and processable HLW glasses need to be optimized to minimize the waste-form volume and, hence, to save cost. A database of properties and associated compositions for simulated waste glasses was collected for developing property-composition models. This database, although not comprehensive, represents a large fraction of data on waste-glass compositions and properties that were available at the time of this report. Glass property-composition models were fit to subsets of the database for several key glass properties. These models apply to a significantly broader composition space than those previously publised. These models should be considered for interim use in calculating properties of Hanford waste glasses.

  11. NEUTRALIZATIONS OF HIGH ALUMINUM LOW URANIUM USED NUCLEAR FUEL SOLUTIONS CONTAINING GADOLINIUM AS A NEUTRON POISON

    SciTech Connect (OSTI)

    Taylor-Pashow, K.

    2011-06-08

    H-Canyon will begin dissolving High Aluminum - Low Uranium (High Al/Low U) Used Nuclear Fuel (UNF) following approval by DOE which is anticipated in CY2011. High Al/Low U is an aluminum/enriched uranium UNF with small quantities of uranium relative to aluminum. The maximum enrichment level expected is 93% {sup 235}U. The High Al/Low U UNF will be dissolved in H-Canyon in a nitric acid/mercury/gadolinium solution. The resulting solution will be neutralized and transferred to Tank 39H in the Tank Farm. To confirm that the solution generated could be poisoned with Gd, neutralized, and discarded to the Savannah River Site (SRS) high level waste (HLW) system without undue nuclear safety concerns the caustic precipitation of simulant solutions was examined. Experiments were performed with three simulant solutions representative of the H-Canyon estimated concentrations in the final solutions after dissolution. The maximum U, Gd, and Al concentration were selected for testing from the range of solution compositions provided. Simulants were prepared in three different nitric acid concentrations, ranging from 0.5 to 1.5 M. The simulant solutions were neutralized to four different endpoints: (1) just before a solid phase was formed (pH 3.5-4), (2) the point where a solid phase was obtained, (3) 0.8 M free hydroxide, and (4) 1.2 M free hydroxide, using 50 wt % sodium hydroxide (NaOH). The settling behavior of the neutralized solutions was found to be slower compared to previous studies, with settling continuing over a one week period. Due to the high concentration of Al in these solutions, precipitation of solids was observed immediately upon addition of NaOH. Precipitation continued as additional NaOH was added, reaching a point where the mixture becomes almost completely solid due to the large amount of precipitate. As additional NaOH was added, some of the precipitate began to redissolve, and the solutions neutralized to the final two endpoints mixed easily and had expected densities of typical neutralized waste. Based on particle size and scanning electron microscopy analyses, the neutralized solids were found to be homogeneous and less than 20 microns in size. The majority of solids were less than 4 microns in size. Compared to previous studies, a larger percentage of the Gd was found to precipitate in the partially neutralized solutions (at pH 3.5-4). In addition the Gd:U mass ratio was found to be at least 1.0 in all of the solids obtained after partial or full neutralization. The hydrogen to U (H:U) molar ratios for two accident scenarios were also determined. The first was for transient neutralization and agitator failure. Experimentally this scenario was determined by measuring the H:U ratio of the settled solids. The minimum H:U molar ratio for solids from fully neutralized solutions was 388:1. The second accident scenario is for the solids drying out in an unagitiated pump box. Experimentally, this scenario was determined by measuring the H:U molar ratio in centrifuged solids. The minimum H:U atom ratios for centrifuged precipitated solids was 250:1. It was determined previously that a 30:1 H:Pu atom ratio was sufficient for a 1:1 Gd:Pu mass ratio. Assuming a 1:1 equivalence with {sup 239}Pu, the results of these experiments show Gd is a viable poison for neutralizing U/Gd solutions with the tested compositions.

  12. CHEMICAL COMPOSITION AND PCT DATA FOR THE INITIAL SET OF HANFORD ENHANCED WASTE LOADING GLASSES

    SciTech Connect (OSTI)

    Fox, K.; Edwards, T.

    2014-06-02

    In this report, the Savannah River National Laboratory provides chemical analyses and Product Consistency Test results for 20 simulated high level waste glasses fabricated by the Pacific Northwest National Laboratory. The results of these analyses will be used as part of efforts to revise or extend the validation ranges of the current Hanford Waste Treatment and Immobilization Plant glass property models to cover a broader span of waste compositions. The measured chemical composition data are reported and compared with the targeted values for each component for each glass. Two components of the study glasses, fluorine and silver, were not measured since each of these species would have required the use of an additional preparation method and their measured values were likely to be near or below analytical detection limits. Some of the glasses were difficult to prepare for chemical analysis. A sodium peroxide fusion dissolution method was successful in completely dissolving the glasses. Components present in the glasses in minor concentrations can be difficult to measure using this dissolution method due to dilution requirements. The use of a lithium metaborate preparation method for the minor components (planned for use since it is typically successful in digesting Defense Waste Processing Facility HLW glasses) resulted in an unacceptable amount of undissolved solids remaining in the sample solutions. An acid dissolution method was used instead, which provided more thorough dissolution of the glasses, although a small amount of undissolved material remained for some of the study glasses. The undissolved material was analyzed to determine those components of the glasses that did not fully dissolve. These components (e.g., calcium and chromium) were present in sufficient quantities to be reported from the measurements resulting from the sodium peroxide fusion preparation method, which did not leave undissolved material. Overall, the analyses resulted in sums of oxides that ranged from about 98 to 101.5 wt % for the study glasses, indicating excellent recovery of all the components in the chemical composition analyses. Comparisons of the targeted and measured chemical compositions indicated that, in general, the measured values for the glasses met the targeted concentrations. Exceptions were Cr{sub 2}O{sub 3}, MgO, and P{sub 2}O{sub 5}. The measured values for Cr{sub 2}O{sub 3} were somewhat low when compared to the targeted values for all of the study glasses targeting Cr{sub 2}O{sub 3} concentrations above 0.5 wt %. Many of the measured MgO and P{sub 2}O{sub 5} values were below the targeted values for those glasses that contained these components. Two of the study glasses exhibited differences from the targeted compositions that may indicate a batching error. Glasses EWG-HAI-Centroid-2 and EWG-OL-1672 had measured values for Al{sub 2}O{sub 3} and SiO{sub 2} that were lower than the targeted values, and measured values for B{sub 2}O{sub 3} that were higher than the targeted values. Glass EWG-HAI-Centroid-2 also had a measured value for Fe{sub 2}O{sub 3} that was lower than the targeted value. A review of the PCT data, including standards and blanks, revealed no issues with the performance of the tests. The PCT results were normalized to both the targeted and measured compositions of the study glasses. Comparisons of the normalized PCT results for both the quenched and Canister Centerline Cooled versions of the study glasses are made with the Environmental Assessment benchmark glass for reference.

  13. Cleanup of Nuclear Licensed Facility 57

    SciTech Connect (OSTI)

    Jeanjacques, Michel; Bremond, Marie Pierre; Marchand, Carole; Poyau, Cecile; Viallefont, Cecile; Gautier, Laurent; Masure, Frederic

    2008-01-15

    This summary describes the operations to clean up the equipment of the Nuclear Licensed Facility 57 (NLF 57). Due to the diversity of the research and development work carried out on the reprocessing of spent fuel in it, this installation is emblematic of many of the technical and organizational issues liable to be encountered in the final closure of nuclear facilities. The French atomic energy commission's center at Fontenay aux Roses (CEA-FAR) was created in 1946 to house pile ZOE. Laboratories for fuel cycle research were installed in existing buildings at the site. Work was later concentrated on spent fuel reprocessing, in a pilot workshop referred to as the 'Usine Pu'. In the early sixties, after the dismantling of these first generation facilities, a radiochemistry laboratory dedicated to research and development work on reprocessing was constructed, designated Building 18. During the same decade, more buildings were added: Building 54, storehouses and offices, Building 91, a hall and laboratories for chemical engineering research on natural and depleted uranium. Together, these three building constitute NLF 57. Building 18 architecture featured four similar modules. Each module had three levels: a sub-level consisting of technical galleries and rooms for the liquid effluent tanks, a ground floor and roof space in which the ventilation was installed. Offices, change rooms, four laboratories and a hall were situated on the ground floor. The shielded lines were installed in the laboratories and the halls. Construction of the building took place between 1959 and 1962, and its commissioning began in 1961. The research and development programs performed in NLF 57 related to studies of the reprocessing of spent fuel, including dry methods and the Purex process, techniques for the treatment of waste (vitrification, alpha waste decontamination, etc.) as well as studies and production of transuranic elements for industry and research. In addition to this work, the necessary methods of analysis for monitoring it were also developed. The research and development program finally ended on 30 June 1995. The NLF 57 cleanup program was intended to reduce the nuclear and conventional hazards and minimize the quantities of HLW and MLW during the subsequent dismantling work. To facilitate the organization of the cleanup work, it was divided into categories by type: - treatment and removal of nuclear material, - removal of radioactive sources, - treatment and removal of aqueous liquid waste, - treatment and removal of organic effluents, - treatment and removal of solid waste, - pumping out of the PETRUS tank, - flushing and decontamination of the tanks, - cleanup of Buildings 18 and 91/54. To estimate the cost of the operations and to monitor the progress of the work, an indicator system was put in place based on work units representative of the operation. The values of the work units were periodically updated on the basis of experience feedback. The cleanup progress is now 92% complete (06/12/31): - treatment and removal of nuclear material: 100%, - removal of radioactive sources: 100%, - treatment and removal of aqueous liquid waste: 64%, - treatment and removal of organic effluents: 87%, - treatment and removal of solid waste: 99%, - pumping out of the PETRUS tank: 69%, - flushing and decontamination of tank: 75%, - section cleaning of Buildings 18 and 91/: 90%. The DRSN/SAFAR is the delegated Project Owner for cleanup and dismantling operations. It is also the prime contractor for the cleanup and dismantling operations. SAFAR itself is responsible for operations relating to the CEA activity and those with technical risks (Removal of nuclear materials, Removal of radioactive sources, Pumping out plutonium and transuranic contaminated solvent and Flushing and decontamination of tanks and pipes). All other operations are sub-contracted to specialist companies. The NLF57 cleanup program as executed is capable of attaining activity levels compatible with a future dismantling operation using known and mastered techniques and producing a

  14. April 2013 Most Viewed Documents for Geosciences | OSTI, US Dept of Energy,

    Office of Scientific and Technical Information (OSTI)

    Office of Scientific and Technical Information April 2013 Most Viewed Documents for Geosciences Sloshing analysis of viscous liquid storage tanks Uras, R.Z. (1995) 83 Seismic design evaluation guidelines for buried piping for the DOE HLW Facilities Lin, Chi-Wen [Consultant, Martinez, CA (United States)]; Antaki, G. [Westinghouse Savannah River Co., Aiken, SC (United States)]; Bandyopadhyay, K. [Brookhaven National Lab., Upton, NY (United States)]; Bush, S.H. [Review & Synthesis

  15. Generic Repository Concepts and Thermal Analysis for Advanced Fuel Cycles -

    Office of Scientific and Technical Information (OSTI)

    12477 (Conference) | SciTech Connect Generic Repository Concepts and Thermal Analysis for Advanced Fuel Cycles - 12477 Citation Details In-Document Search Title: Generic Repository Concepts and Thermal Analysis for Advanced Fuel Cycles - 12477 A geologic disposal concept for spent nuclear fuel (SNF) or high-level waste (HLW) consists of three components: waste inventory, geologic setting, and concept of operations. A set of reference geologic disposal concepts has been developed by the U.S.

  16. Hanford Dangerous Waste Permit

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

    Waste Treatment and Immobilization Plant (vit plant) Operating Unit #10 Aerial view of construction, July 2011 Where will the waste go? LAW canisters will go to shallow disposal at Hanford's Integrated Disposal Facility. HLW canisters will go to a For scale, here's the parking lot! Safe disposition of our nation's most dangerous waste relies on the vit plant's safe completion and ability to process waste for 20+ years. * Permitted for storage and treatment of Hanford's tank waste in unique

  17. app_d

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

    72 Appendix D - New Information - Document 68, Environmental Defense Institute (Chuck Broscious), Troy, ID Page 1 of 32 Document 68, Environmental Defense Institute (Chuck Broscious), Troy, ID Page 2 of 32 D-173 DOE/EIS-0287 Idaho HLW & FD EIS - New Information - Document 68, Environmental Defense Institute (Chuck Broscious), Troy, ID Page 3 of 32 Document 68, Environmental Defense Institute (Chuck Broscious), Troy, ID Page 4 of 32 DOE/EIS-0287 D-174 Appendix D - New Information - Document

  18. app_d

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

    0 Appendix D - New Information - Document 68, Environmental Defense Institute (Chuck Broscious), Troy, ID Page 17 of 32 Document 68, Environmental Defense Institute (Chuck Broscious), Troy, ID Page 18 of 32 D-181 DOE/EIS-0287 Idaho HLW & FD EIS - New Information - Document 68, Environmental Defense Institute (Chuck Broscious), Troy, ID Page 19 of 32 Document 68, Environmental Defense Institute (Chuck Broscious), Troy, ID Page 20 of 32 DOE/EIS-0287 D-182 Appendix D - New Information -

  19. ch_3

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

    47 DOE/EIS-0287 Idaho HLW & FD EIS has been provided to the public, committed DOE to restoring the existing contaminated groundwater plume outside the INTEC security fence to meet the current drinking water stan- dard of 4 millirem per year. A performance assessment would be developed for each facility or group of facilities under consideration for disposition, to determine which of the three disposition alternatives would be implemented. The performance assessment results would be used to

  20. FY 2007 Total System Life Cycle Cost, Pub 2008 | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    FY 2007 Total System Life Cycle Cost, Pub 2008 FY 2007 Total System Life Cycle Cost, Pub 2008 The Analysis of the Total System Life Cycle Cost (TSLCC) of the Civilian Radioactive Waste Management Program presents the Office of Civilian Radioactive Waste Management's (OCRWM) May 2007 total system cost estimate for the disposal of the Nation's spent nuclear fuel (SNF) and high-level radioactive waste (HLW). The TSLCC analysis provides a basis for assessing the adequacy of the Nuclear Waste Fund

  1. Waste Treatment and Immobilation Plant Pretreatment Facility | Department

    Energy Savers [EERE]

    of Energy Pretreatment Facility Waste Treatment and Immobilation Plant Pretreatment Facility Full Document and Summary Versions are available for download PDF icon Waste Treatment and Immobilation Plant Pretreatment Facility PDF icon Summary - WTP Pretreatment Facility More Documents & Publications Waste Treatment and Immobilation Plant HLW Waste Vitrification Facility Compilation of TRA Summaries Hanford ETR Tank Waste Treatment and Immobilization Plant - Hanford Tank Waste Treatment

  2. High-level waste management technology program plan

    SciTech Connect (OSTI)

    Harmon, H.D.

    1995-01-01

    The purpose of this plan is to document the integrated technology program plan for the Savannah River Site (SRS) High-Level Waste (HLW) Management System. The mission of the SRS HLW System is to receive and store SRS high-level wastes in a see and environmentally sound, and to convert these wastes into forms suitable for final disposal. These final disposal forms are borosilicate glass to be sent to the Federal Repository, Saltstone grout to be disposed of on site, and treated waste water to be released to the environment via a permitted outfall. Thus, the technology development activities described herein are those activities required to enable successful accomplishment of this mission. The technology program is based on specific needs of the SRS HLW System and organized following the systems engineering level 3 functions. Technology needs for each level 3 function are listed as reference, enhancements, and alternatives. Finally, FY-95 funding, deliverables, and schedules are s in Chapter IV with details on the specific tasks that are funded in FY-95 provided in Appendix A. The information in this report represents the vision of activities as defined at the beginning of the fiscal year. Depending on emergent issues, funding changes, and other factors, programs and milestones may be adjusted during the fiscal year. The FY-95 SRS HLW technology program strongly emphasizes startup support for the Defense Waste Processing Facility and In-Tank Precipitation. Closure of technical issues associated with these operations has been given highest priority. Consequently, efforts on longer term enhancements and alternatives are receiving minimal funding. However, High-Level Waste Management is committed to participation in the national Radioactive Waste Tank Remediation Technology Focus Area. 4 refs., 5 figs., 9 tabs.

  3. HANFORD MEDIUM & LOW CURIE WASTE PRETREATMENT PROJECT PHASE 1 LAB REPORT

    SciTech Connect (OSTI)

    HAMILTON, D.W.

    2006-01-30

    A fractional crystallization (FC) process is being developed to supplement tank waste pretreatment capabilities provided by the Waste Treatment and Immobilization Plant (WTP). FC can process many tank wastes, separating wastes into a low-activity fraction (LAW) and high-activity fraction (HLW). The low-activity fraction can be immobilized in a glass waste form by processing in the bulk vitrification (BV) system.

  4. DOE/ID-Number FCRD-USED-2011-000184

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

    1 Performance Assessment Modeling of a Generic SNF/HLW Repository in Salt with Coupled Thermal-Hydrologic Effects - 15423 S. David Sevougian*, Geoff A. Freeze*, W. Payton Gardner*, Glenn E. Hammond*, Paul E. Mariner*, and Robert J. MacKinnon* *Advanced Nuclear Energy Programs Group, sdsevou@sandia.gov Sandia National Laboratories P.O. Box 5800, M.S. 0747 Albuquerque, NM 87185 ABSTRACT This paper describes advances in performance assessment modeling of deep geologic repositories facilitated by a

  5. ICPP tank farm closure study. Volume 1

    SciTech Connect (OSTI)

    Spaulding, B.C.; Gavalya, R.A.; Dahlmeir, M.M.

    1998-02-01

    The disposition of INEEL radioactive wastes is now under a Settlement Agreement between the DOE and the State of Idaho. The Settlement Agreement requires that existing liquid sodium bearing waste (SBW), and other liquid waste inventories be treated by December 31, 2012. This agreement also requires that all HLW, including calcined waste, be disposed or made road ready to ship from the INEEL by 2035. Sodium bearing waste (SBW) is produced from decontamination operations and HLW from reprocessing of SNF. SBW and HLW are radioactive and hazardous mixed waste; the radioactive constituents are regulated by DOE and the hazardous constituents are regulated by the Resource Conservation and Recovery Act (RCRA). Calcined waste, a dry granular material, is produced in the New Waste Calcining Facility (NWCF). Two primary waste tank storage locations exist at the ICPP: Tank Farm Facility (TFF) and the Calcined Solids Storage Facility (CSSF). The TFF has the following underground storage tanks: four 18,400-gallon tanks (WM 100-102, WL 101); four 30,000-gallon tanks (WM 103-106); and eleven 300,000+ gallon tanks. This includes nine 300,000-gallon tanks (WM 182-190) and two 318,000 gallon tanks (WM 180-181). This study analyzes the closure and subsequent use of the eleven 300,000+ gallon tanks. The 18,400 and 30,000-gallon tanks were not included in the work scope and will be closed as a separate activity. This study was conducted to support the HLW Environmental Impact Statement (EIS) waste separations options and addresses closure of the 300,000-gallon liquid waste storage tanks and subsequent tank void uses. A figure provides a diagram estimating how the TFF could be used as part of the separations options. Other possible TFF uses are also discussed in this study.

  6. Office of Enterprise Assessments Operational Awareness Record for the Observation of the Waste Treatment and Immobilization Plant High Level Waste Faciity Concentrate Receipt/Melter Feed/Glass Formers Reagent Hazard Analysis and Review of the Radioactive Liquid Disposal Hazards Analysis Event Tables - March 2015

    Energy Savers [EERE]

    Operational Awareness Record Report Number: EA-WTP-HLW-2014-10-20 Site: Hanford Site Subject: Observation of the Waste Treatment and Immobilization Plant High Level Waste Facility Concentrate Receipt/Melter Feed/ Glass Formers Reagent Hazards Analysis Activities and Review of the Radioactive Liquid Disposal Hazards Analysis Event Tables Dates of Activity: 10/20/14 - 11/06/14 Report Preparer: James O. Low Activity Description/Purpose: Bechtel National, Incorporated (BNI) is implementing a Safety

  7. Summary.qxd

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

    3 DOE/EIS-0287 Idaho HLW & FD EIS 6.4.4 HEALTH AND SAFETY Airborne contamination is the principal transport pathway through which radioactive materials from the INEEL affect workers and the public. The SNF and INEL EIS evaluated radiation releases and subsequent offsite doses associated with INEEL operations. Doses have always been small and within applicable radiation pro- tection standards. In 1996, for example, the col- lective radiological dose to the population within 50 miles of the

  8. Tank Waste Committee - Transcribed Flipcharts

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

    - Transcribed Flipcharts January 8, 2014 Framework 1. Is there anything in the Framework document the Board should respond to? o Any concepts the Board does not support? 2. Anything missing from IM discussion draft? o Agree/disagree with IM recommendation? Page 1 Framework * Missing: schedule & cost * Agree & support o Early LAW o Support TRU waste retrieval  Provided the State of New Mexico concurs  Determine not HLW (process knowledge)  As long as meets all applicable

  9. Plutonium immobilization plant using glass in existing facilities at the Savannah River Site

    SciTech Connect (OSTI)

    DiSabatino, A., LLNL

    1998-06-01

    The Plutonium Immobilization Plant (PIP) accepts plutonium (Pu) from pit conversion and from non-pit sources and, through a glass immobilization process, converts the plutonium into an immobilized form that can be disposed of in a high level waste (HLW) repository. The objective is to make an immobilized form, suitable for geologic disposal, in which the plutonium is as inherently unattractive and inaccessible as the plutonium in spent fuel from commercial reactors.

  10. Hanford Tank Farm interim storage phase probabilistic risk assessment outline

    SciTech Connect (OSTI)

    Not Available

    1994-05-19

    This report is the second in a series examining the risks for the high level waste (HLW) storage facilities at the Hanford Site. The first phase of the HTF PSA effort addressed risks from Tank 101-SY, only. Tank 101-SY was selected as the initial focus of the PSA because of its propensity to periodically release (burp) a mixture of flammable and toxic gases. This report expands the evaluation of Tank 101-SY to all 177 storage tanks. The 177 tanks are arranged into 18 farms and contain the HLW accumulated over 50 years of weapons material production work. A centerpiece of the remediation activity is the effort toward developing a permanent method for disposing of the HLW tank`s highly radioactive contents. One approach to risk based prioritization is to perform a PSA for the whole HLW tank farm complex to identify the highest risk tanks so that remediation planners and managers will have a more rational basis for allocating limited funds to the more critical areas. Section 3 presents the qualitative identification of generic initiators that could threaten to produce releases from one or more tanks. In section 4 a detailed accident sequence model is developed for each initiating event group. Section 5 defines the release categories to which the scenarios are assigned in the accident sequence model and presents analyses of the airborne and liquid source terms resulting from different release scenarios. The conditional consequences measured by worker or public exposure to radionuclides or hazardous chemicals and economic costs of cleanup and repair are analyzed in section 6. The results from all the previous sections are integrated to produce unconditional risk curves in frequency of exceedance format.

  11. A New Path Forward for WTP AL Boldt and RI Smith

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

    Dick Smith and Al Boldt - thoughts to share with the Tank Waste Committee Not a committee work product A New Path Forward for WTP AL Boldt and RI Smith February 3, 2014 Introduction The "Framework" document, issued by the Department of Energy (DOE) in September of 2013, is purported to show the path forward for completion and operation of the Hanford Waste Treatment Plant (WTP) for treatment of Hanford tank wastes. Construction on two principal facilities (HLW and Pretreatment) was

  12. Microsoft PowerPoint - River Protection Project HAB.ppt [Read-Only]

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

    River Protection Project Waste Treatment & Disposition January 2009 1 2 Insights From External Review Analyses * No need to make LAW ST decision now: make in 2015 - 2017 once Na/Al uncertainties reduced, M-2/M-3 Pretreatment issues resolved, and improved overall integrated RPP system understanding in place (e.g., System Plan 4) * WTP LAW Facility alone cannot complete LAW immobilization mission in same timeframe as HLW mission even with 3 rd melter or higher capacity melters. Third melter

  13. High Level Waste ManagemenfDivision ..

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

    High Level Waste ManagemenfDivision .. . . HLWSystem Plan Revision 2(U) Westinghouse Savannah River Company . Aiken; South Carolina Jam,lary 14,1994 HIGH LEVEL WASTE SYSTEM PLAN REVISION 2 _--JANUARY 14, 1994 APPROVAL SHEET Deputy General Manager High Level Waste Management Westinghouse Savannah River Company fO ..... R. E. Erickson Director,- Vitrification Projects Division U. S. Department of Energy, Headquarters Date I Date Date " " HLW System Plan - Revision 2 (U) Table of Contents

  14. Solvent Extraction of Tc and Cs from Alkaline Nitrate Wastes

    SciTech Connect (OSTI)

    Bonnesen, P.V.; Conner, C.; Delmau, L.H.; Haverlock, T.J.; Leonard, R.A.; Lumetta, G.J.; Moyer, B.A.; Sachleben, R.A.

    1999-07-11

    This paper summarizes progress at three collaborating US national laboratories on the extraction of the fission products {sup 99}Tc and {sup 137}Cs from alkaline high-level wastes (HLW). Efficient, economical processes for Tc and Cs extraction (SRTALK and alkaline-side CSEX, respectively) have been developed, and testing has progressed through batch tests on actual wastes and continuous countercurrent centrifugal-contactor tests on simulants.

  15. Microsoft Word - 2015 12.7 WTP Communications Approach White Paper DRAFT.docx

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

    Waste Treatment and Immobilization Plant (WTP) Communications Approach Draft White Paper - last revised 12/4/15 Issue Managers: Suyama, Mattson, Niles, Hudson, Leckband Summary The Hanford Advisory Board, following discussions conducted by the Board's Tank Waste and Public Involvement and Communication committees with the U.S. Department of Energy (DOE) Office of River Protection (ORP), prepared an assessment and recommendations for a communications approach regarding the High Level Waste (HLW)

  16. Microsoft Word - 2016 2.8 WTP Communications Approach White Paper DRAFT.docx

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

    Waste Treatment and Immobilization Plant (WTP) Communications Approach Draft White Paper - last revised 2/8/16 Issue Managers: Suyama, Mattson, Niles, Hudson, Leckband Summary The Hanford Advisory Board, following discussions conducted by the Board's Tank Waste and Public Involvement and Communication committees with the U.S. Department of Energy, Office of River Protection (DOE), has prepared this assessment and recommendations for a communications approach regarding the High Level Waste (HLW)

  17. Decomposition of tetraphenylborate precipitates used to isolate Cs-137 from Savannah River Site high-level waste

    SciTech Connect (OSTI)

    Ferrara, D.M.; Bibler, N.E.; Ha, B.C.

    1993-03-01

    This paper presents results of the radioactive demonstration of the Precipitate Hydrolysis Process (PHP) that will be performed in the Defense Waste Processing Facility (DWPF) at the Savannah River Site. The PHP destroys the tetraphenylborate precipitate that is used at SRS to isolate Cs-137 from caustic High-Level Waste (HLW) supernates. This process is necessary to decrease the amount of organic compounds going to the melter in the DWPF. Actual radioactive precipitate containing Cs-137 was used for this demonstration.

  18. Radiation and transmutation effects relevant to solid nuclear waste forms

    SciTech Connect (OSTI)

    Vance, E.R.; Roy, R.; Pillay, K.K.S.

    1981-03-15

    Radiation effects in insulating solids are discussed in a general way as an introduction to the quite sparse published work on radiation effects in candidate nuclear waste forms other than glasses. Likely effects of transmutation in crystals and the chemical mitigation strategy are discussed. It seems probable that radiation effects in solidified HLW will not be serious if the actinides can be wholly incorporated in such radiation-resistant phases as monazite or uraninite.

  19. PowerPoint Presentation

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

    West Valley, New York - Processed over 600,000 gallons of legacy commercial high- level waste (HLW) - Completed production of 278 glass canisters in 2002 - Defense Waste Processing Facility (DWPF), Savannah River Site (SRS) - Processing 32 million gallons of DOE defense-related nuclear waste - Produced over 3,780 glass canisters since 1996 4 Operations and Commissioning Staff are Highly Experienced  Other DOE nuclear facilities - SRS and Hanford Tank Farms - Integrated Waste Treatment Unit

  20. Assessment of Available Particle Size Data to Support an Analysis of the Waste Feed Delivery System Transfer System

    SciTech Connect (OSTI)

    JEWETT, J.R.

    2000-08-10

    Available data pertaining to size distribution of the particulates in Hanford underground tank waste have been reviewed. Although considerable differences exist between measurement methods, it may be stated with 95% confidence that the median particle size does not exceed 275 {micro}m in at least 95% of the ten tanks selected as sources of HLW feed for Phase 1 vitrification in the RPP. This particle size is recommended as a design basis for the WFD transfer system.