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Note: This page contains sample records for the topic "waste disposition packaging" from the National Library of EnergyBeta (NLEBeta).
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1

EM Makes Significant Progress on Dispositioning Transuranic Waste...  

Office of Environmental Management (EM)

EM Makes Significant Progress on Dispositioning Transuranic Waste at Idaho Site EM Makes Significant Progress on Dispositioning Transuranic Waste at Idaho Site December 24, 2013 -...

2

Waste disposal package  

DOE Patents [OSTI]

This is a claim for a waste disposal package including an inner or primary canister for containing hazardous and/or radioactive wastes. The primary canister is encapsulated by an outer or secondary barrier formed of a porous ceramic material to control ingress of water to the canister and the release rate of wastes upon breach on the canister. 4 figs.

Smith, M.J.

1985-06-19T23:59:59.000Z

3

Assistant Manager for Waste Disposition  

Broader source: Energy.gov [DOE]

The incumbent of this position is responsible for providing overall leadership and direction for oversight of assigned contractor and Federal programs and activities associated with the disposition...

4

Radioactive waste disposal package  

DOE Patents [OSTI]

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.

Lampe, Robert F. (Bethel Park, PA)

1986-01-01T23:59:59.000Z

5

DRAFT EM SSAB Chairs Meeting Waste Disposition Strategies...  

Office of Environmental Management (EM)

EM HQ Updates Waste Disposition Overview Christine Gelles Associate Deputy Assistant Secretary for Waste Management Office of Environmental Management EM SSAB Chairs Meeting 5...

6

On Going TRU Waste Disposition  

SciTech Connect (OSTI)

The ongoing effort to contain dangerous, radioactive TRU waste. Under the Recovery Act, the Savannah River Site is able to safely test and transport these items to WIPP in Carlsbad, New Mexico.

Cody, Tom

2010-01-01T23:59:59.000Z

7

On Going TRU Waste Disposition  

ScienceCinema (OSTI)

The ongoing effort to contain dangerous, radioactive TRU waste. Under the Recovery Act, the Savannah River Site is able to safely test and transport these items to WIPP in Carlsbad, New Mexico.

Cody, Tom

2012-06-14T23:59:59.000Z

8

Naval Waste Package Design Report  

SciTech Connect (OSTI)

A design methodology for the waste packages and ancillary components, viz., the emplacement pallets and drip shields, has been developed to provide designs that satisfy the safety and operational requirements of the Yucca Mountain Project. This methodology is described in the ''Waste Package Design Methodology Report'' Mecham 2004 [DIRS 166168]. To demonstrate the practicability of this design methodology, four waste package design configurations have been selected to illustrate the application of the methodology. These four design configurations are the 21-pressurized water reactor (PWR) Absorber Plate waste package, the 44-boiling water reactor (BWR) waste package, the 5-defense high-level waste (DHLW)/United States (U.S.) Department of Energy (DOE) spent nuclear fuel (SNF) Co-disposal Short waste package, and the Naval Canistered SNF Long waste package. Also included in this demonstration is the emplacement pallet and continuous drip shield. The purpose of this report is to document how that design methodology has been applied to the waste package design configurations intended to accommodate naval canistered SNF. This demonstrates that the design methodology can be applied successfully to this waste package design configuration and support the License Application for construction of the repository.

M.M. Lewis

2004-03-15T23:59:59.000Z

9

ACCELERATION OF LOS ALAMOS NATIONAL LABORATORY TRANSURANIC WASTE DISPOSITION  

SciTech Connect (OSTI)

One of Los Alamos National Laboratory's (LANL's) most significant risks is the site's inventory of transuranic waste retrievably stored above and below-ground in Technical Area (TA) 54 Area G, particularly the dispersible high-activity waste stored above-ground in deteriorating facilities. The high activity waste represents approximately 50% (by activity) of the total 292,000 PE-Ci inventory remaining to be disposed. The transuramic waste inventory includes contact-handled and remote-handled waste packaged in drums, boxes, and oversized containers which are retrievably stored both above and below-ground. Although currently managed as transuranic waste, some of the inventory is low-level waste that can be disposed onsite or at approved offsite facilities. Dispositioning the transuranic waste inventory requires retrieval of the containers from above and below-ground storage, examination and repackaging or remediation as necessary, characterization, certification and loading for shipment to the Waste Isolation Pilot Plant in Carlsbad New Mexico, all in accordance with well-defined requirements and controls. Although operations are established to process and characterize the lower-activity contact-handled transuranic waste containers, LAN L does not currently have the capability to repack high activity contact-handled transuranic waste containers (> 56 PE-Ci) or to process oversized containers with activity levels over 0.52 PE-Ci. Operational issues and compliance requirements have resulted in less than optimal processing capabilities for lower activity contact-handled transuranic waste containers, limiting preparation and reducing dependability of shipments to the Waste Isolation Pilot Plant. Since becoming the Los Alamos National Laboratory contract in June 2006, Los Alamos National Security (LANS) L.L.C. has developed a comprehensive, integrated plan to effectively and efficiently disposition the transuranic waste inventory, working in concert with the Department of Energy Los Alamos Site Office, Carlsbad Field Office and the Department of Energy Headquaeters. Rather than simply processing containers as retrieved, the plan places priority on efficient curie disposition, a direct correlation to reducing risk. Key elements of the approch include balancing inventory and operational risks, tailoring methods to meet requirements, optimizing existing facilities, equipment and staff, and incorporating best practices from other Department of Energy sites. With sufficient funding this will enable LANL to ship the above-ground high activity contact-handled transuranic waste offsite by the end of Fiscal Year (FY) 2007 and to disposition the remaining above- and below-ground contact-handled and remote-handled transuranic waste inventory by December 2010. Nearly 70% of the contact-handled transuranic waste containers, including the high activity waste, require processing and repackaging before characterization and certification for shipment to the Waste Isolation Pilot Plant. LANL is employing a balanced risk approach that accomplishes significant long-term risk reduction by accepting short-term increased facility operations risk under well-developed and justified interim controls. Reviews of facility conditions and additional analyses show that the Waste Characterization, Reduction and Repackaging Facility and the Radioassay and Nondestructive Testing Facility are the most appropriate facilities to safetly remediate, repackage, and ship lower activity and the remaining high activity drums. Updated safety documentation supporting limited Hazard Category 2 operations in these facilities has been developed. Once approved, limited-term operations to process the high activity drums can begin in early 2007, building upon the experience base established performing Hazard Category 3 operations processing lower activity waste in these facilities. LANL is also implementing a series of actions to improve and sustain operations for processing contact-handled transuranic waste inventory. Building 412 Decontamination and Volume Facility and Dom

O'LEARY, GERALD A. [Los Alamos National Laboratory

2007-01-04T23:59:59.000Z

10

Tritium waste package  

DOE Patents [OSTI]

A containment and waste package system for processing and shipping tritium xide waste received from a process gas includes an outer drum and an inner drum containing a disposable molecular sieve bed (DMSB) seated within outer drum. The DMSB includes an inlet diffuser assembly, an outlet diffuser assembly, and a hydrogen catalytic recombiner. The DMSB absorbs tritium oxide from the process gas and converts it to a solid form so that the tritium is contained during shipment to a disposal site. The DMSB is filled with type 4A molecular sieve pellets capable of adsorbing up to 1000 curies of tritium. The recombiner contains a sufficient amount of catalyst to cause any hydrogen add oxygen present in the process gas to recombine to form water vapor, which is then adsorbed onto the DMSB.

Rossmassler, Rich (Cranbury, NJ); Ciebiera, Lloyd (Titusville, NJ); Tulipano, Francis J. (Teaneck, NJ); Vinson, Sylvester (Ewing, NJ); Walters, R. Thomas (Lawrenceville, NJ)

1995-01-01T23:59:59.000Z

11

Tritium waste package  

DOE Patents [OSTI]

A containment and waste package system for processing and shipping tritium oxide waste received from a process gas includes an outer drum and an inner drum containing a disposable molecular sieve bed (DMSB) seated within the outer drum. The DMSB includes an inlet diffuser assembly, an outlet diffuser assembly, and a hydrogen catalytic recombiner. The DMSB absorbs tritium oxide from the process gas and converts it to a solid form so that the tritium is contained during shipment to a disposal site. The DMSB is filled with type 4A molecular sieve pellets capable of adsorbing up to 1000 curies of tritium. The recombiner contains a sufficient amount of catalyst to cause any hydrogen and oxygen present in the process gas to recombine to form water vapor, which is then adsorbed onto the DMSB. 1 fig.

Rossmassler, R.; Ciebiera, L.; Tulipano, F.J.; Vinson, S.; Walters, R.T.

1995-11-07T23:59:59.000Z

12

EIS-0287: Idaho High-Level Waste and Facilities Disposition Final...  

Office of Environmental Management (EM)

: Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement, EIS-0287 (September 2002) EIS-0287: Idaho High-Level Waste and Facilities Disposition...

13

Waste Package Materials Performance Peer Review | Department...  

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

Waste Package Materials Performance Peer Review Waste Package Materials Performance Peer Review A consensus peer review of the current technical basis and the planned experimental...

14

FUEL CYCLE POTENTIAL WASTE FOR DISPOSITION  

SciTech Connect (OSTI)

The United States (U.S.) currently utilizes a once-through fuel cycle where used nuclear fuel (UNF) is stored on-site in either wet pools or in dry storage systems with ultimate disposal in a deep mined geologic repository envisioned. Within the Department of Energy's (DOE) Office of Nuclear Energy (DOE-NE), the Fuel Cycle Research and Development Program (FCR&D) develops options to the current commercial fuel cycle management strategy to enable the safe, secure, economic, and sustainable expansion of nuclear energy while minimizing proliferation risks by conducting research and development of advanced fuel cycles, including modified open and closed cycles. The safe management and disposition of used nuclear fuel and/or nuclear waste is a fundamental aspect of any nuclear fuel cycle. Yet, the routine disposal of used nuclear fuel and radioactive waste remains problematic. Advanced fuel cycles will generate different quantities and forms of waste than the current LWR fleet. This study analyzes the quantities and characteristics of potential waste forms including differing waste matrices, as a function of a variety of potential fuel cycle alternatives including: (1) Commercial UNF generated by uranium fuel light water reactors (LWR). Four once through fuel cycles analyzed in this study differ by varying the assumed expansion/contraction of nuclear power in the U.S; (2) Four alternative LWR used fuel recycling processes analyzed differ in the reprocessing method (aqueous vs. electro-chemical), complexity (Pu only or full transuranic (TRU) recovery) and waste forms generated; (3) Used Mixed Oxide (MOX) fuel derived from the recovered Pu utilizing a single reactor pass; and (4) Potential waste forms generated by the reprocessing of fuels derived from recovered TRU utilizing multiple reactor passes.

Jones, R.; Carter, J.

2010-10-13T23:59:59.000Z

15

FUEL CYCLE POTENTIAL WASTE FOR DISPOSITION  

SciTech Connect (OSTI)

The United States (U.S.) currently utilizes a once-through fuel cycle where used nuclear fuel (UNF) is stored on-site in either wet pools or in dry storage systems with ultimate disposal in a deep mined geologic repository envisioned. Within the Department of Energy's (DOE) Office of Nuclear Energy (DOE-NE), the Fuel Cycle Research and Development Program (FCR&D) develops options to the current commercial fuel cycle management strategy to enable the safe, secure, economic, and sustainable expansion of nuclear energy while minimizing proliferation risks by conducting research and development of advanced fuel cycles, including modified open and closed cycles. The safe management and disposition of used nuclear fuel and/or nuclear waste is a fundamental aspect of any nuclear fuel cycle. Yet, the routine disposal of used nuclear fuel and radioactive waste remains problematic. Advanced fuel cycles will generate different quantities and forms of waste than the current LWR fleet. This study analyzes the quantities and characteristics of potential waste forms including differing waste matrices, as a function of a variety of potential fuel cycle alternatives including: (1) Commercial UNF generated by uranium fuel light water reactors (LWR). Four once through fuel cycles analyzed in this study differ by varying the assumed expansion/contraction of nuclear power in the U.S. (2) Four alternative LWR used fuel recycling processes analyzed differ in the reprocessing method (aqueous vs. electro-chemical), complexity (Pu only or full transuranic (TRU) recovery) and waste forms generated. (3) Used Mixed Oxide (MOX) fuel derived from the recovered Pu utilizing a single reactor pass. (4) Potential waste forms generated by the reprocessing of fuels derived from recovered TRU utilizing multiple reactor passes.

Carter, J.

2011-01-03T23:59:59.000Z

16

YUCCA MOUNTAIN WASTE PACKAGE CLOSURE SYSTEM  

SciTech Connect (OSTI)

The method selected for dealing with spent nuclear fuel in the US is to seal the fuel in waste packages and then to place them in an underground repository at the Yucca Mountain Site in Nevada. This article describes the Waste Package Closure System (WPCS) currently being designed for sealing the waste packages.

G. Housley; C. Shelton-davis; K. Skinner

2005-08-26T23:59:59.000Z

17

Safety Analysis Report for packaging (onsite) steel waste package  

SciTech Connect (OSTI)

The steel waste package is used primarily for the shipment of remote-handled radioactive waste from the 324 Building to the 200 Area for interim storage. The steel waste package is authorized for shipment of transuranic isotopes. The maximum allowable radioactive material that is authorized is 500,000 Ci. This exceeds the highway route controlled quantity (3,000 A{sub 2}s) and is a type B packaging.

BOEHNKE, W.M.

2000-07-13T23:59:59.000Z

18

Safety evaluation for packaging (onsite) concrete-lined waste packaging  

SciTech Connect (OSTI)

The Pacific Northwest National Laboratory developed a package to ship Type A, non-transuranic, fissile excepted quantities of liquid or solid radioactive material and radioactive mixed waste to the Central Waste Complex for storage on the Hanford Site.

Romano, T.

1997-09-25T23:59:59.000Z

19

MANAGING HANFORD'S LEGACY NO-PATH-FORWARD WASTES TO DISPOSITION  

SciTech Connect (OSTI)

The U.S. Department of Energy (DOE) Richland Operations Office (RL) has adopted the 2015 Vision for Cleanup of the Hanford Site. This vision will protect the Columbia River, reduce the Site footprint, and reduce Site mortgage costs. The CH2M HILL Plateau Remediation Company's (CHPRC) Waste and Fuels Management Project (W&FMP) and their partners support this mission by providing centralized waste management services for the Hanford Site waste generating organizations. At the time of the CHPRC contract award (August 2008) slightly more than 9,000 m{sup 3} of waste was defined as 'no-path-forward waste.' The majority of these wastes are suspect transuranic mixed (TRUM) wastes which are currently stored in the low-level Burial Grounds (LLBG), or stored above ground in the Central Waste Complex (CWC). A portion of the waste will be generated during ongoing and future site cleanup activities. The DOE-RL and CHPRC have collaborated to identify and deliver safe, cost-effective disposition paths for 90% ({approx}8,000 m{sup 3}) of these problematic wastes. These paths include accelerated disposition through expanded use of offsite treatment capabilities. Disposal paths were selected that minimize the need to develop new technologies, minimize the need for new, on-site capabilities, and accelerate shipments of transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico.

WEST LD

2011-01-13T23:59:59.000Z

20

A little here, a little there, a fairly big problem everywhere: Small quantity site transuranic waste disposition alternatives  

SciTech Connect (OSTI)

Small quantities of transuranic (TRU) waste represent a significant challenge to the waste disposition and facility closure plans of several sites in the Department of Energy (DOE) complex. This paper presents the results of a series of evaluations, using a systems engineering approach, to identify the preferred alternative for dispositioning TRU waste from small quantity sites (SQSs). The TRU waste disposition alternatives evaluation used semi-quantitative data provided by the SQSs, potential receiving sites, and the Waste Isolation Pilot Plant (WIPP) to select and recommend candidate sites for waste receipt, interim storage, processing, and preparation for final disposition of contact-handled (CH) and remote-handled (RH) TRU waste. The evaluations of only four of these SQSs resulted in potential savings to the taxpayer of $33 million to $81 million, depending on whether mobile systems could be used to characterize, package, and certify the waste or whether each site would be required to perform this work. Small quantity shipping sites included in the evaluation included the Battelle Columbus Laboratory (BCL), University of Missouri Research Reactor (MURR), Energy Technology Engineering Center (ETEC), and Mound Laboratory. Candidate receiving sites included the Idaho National Engineering and Environmental Laboratory (INEEL), the Savannah River Site (SRS), Los Alamos National Laboratory (LANL), Oak Ridge (OR), and Hanford. At least 14 additional DOE sites having TRU waste may be able to save significant money if cost savings are similar to the four evaluated thus far.

D. Luke; D. Parker; J. Moss; T. Monk (INEEL); L. Fritz (DOE-ID); B. Daugherty (SRS); K. Hladek (WM Federal Services Hanford); S. Kosiewicx (LANL)

2000-02-27T23:59:59.000Z

Note: This page contains sample records for the topic "waste disposition packaging" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

A Little Here, A Little There, A Fairly Big Problem Everywhere: Small Quantity Site Transuranic Waste Disposition Alternatives  

SciTech Connect (OSTI)

Small quantities of transuranic (TRU) waste represent a significant challenge to the waste disposition and facility closure plans of several sites in the Department of Energy (DOE) complex. This paper presents the results of a series of evaluations, using a systems engineering approach, to identify the preferred alternative for dispositioning TRU waste from small quantity sites (SQSs). The TRU waste disposition alternatives evaluation used semi-quantitative data provided by the SQSs, potential receiving sites, and the Waste Isolation Pilot Plant (WIPP) to select and recommend candidate sites for waste receipt, interim storage, processing, and preparation for final disposition of contact-handled (CH) and remote-handled (RH) TRU waste. The evaluations of only four of these SQSs resulted in potential savings to the taxpayer of $33 million to $81 million, depending on whether mobile systems could be used to characterize, package, and certify the waste or whether each site would be required to perform this work. Small quantity shipping sites included in the evaluation included the Battelle Columbus Laboratory (BCL), University of Missouri Research Reactor (MURR), Energy Technology Engineering Center (ETEC), and Mound. Candidate receiving sites included the Idaho National Engineering and Environmental Laboratory (INEEL), the Savannah River Site (SRS), Los Alamos National Laboratory (LANL), Oak Ridge (OR), and Hanford. At least 14 additional DOE sites having TRU waste may be able to save significant money if cost savings are similar to the four evaluated thus far.

Luke, Dale Elden; Parker, Douglas Wayne; Moss, J.; Monk, Thomas Hugh; Fritz, Lori Lee; Daugherty, B.; Hladek, K.; Kosiewicx, S.

2000-03-01T23:59:59.000Z

22

CH Packaging Operations for High Wattage Waste  

SciTech Connect (OSTI)

This document provides instructions for assembling the following CH packaging payload: Drum payload assembly Standard Waste Box (SWB) assembly Ten-Drum Overpack (TDOP)

Washington TRU Solutions LLC

2006-01-06T23:59:59.000Z

23

Waste Disposition Update by Doug Tonkay  

Office of Environmental Management (EM)

Douglas Tonkay Office of Disposal Operations October 20, 2011 o Continue to manage waste inventories in a safe and compliant manner. o Address high risk waste in a cost-...

24

Waste Disposition Update by Christine Gelles  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergy Cooperation | Department of EnergyDepartmentEnergy WRPSWaste Disposition Update

25

Waste Disposition Update by Doug Tonkay  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergy Cooperation | Department of EnergyDepartmentEnergy WRPSWaste Disposition

26

Accelerating the disposition of transuranic waste from LANL - 9495  

SciTech Connect (OSTI)

Los Alamos National Laboratory (LANL) was established during World War II with a single mission -- to design and build an atomic bomb. In the 65 years since, nuclear weapons physics, design and engineering have been the Laboratory's primary and sustaining mission. Experimental and process operations -- and associated cleanout and upgrade activities -- have generated a significant inventory of transuranic (TRU) waste that is stored at LANL's Technical Area 54, Material Disposal Area G (MDA G). When the Waste Isolation Pilot Plant (WIPP) opened its doors in 1999, LANL's TRU inventory totaled about 10,200 m{sup 3}, with a plutonium 239-equivalent curie (PE Ci) content of approximately 250,000 curies. By December 2008, a total of about 2,300 m3 (61,000 PE Ci) had been shipped to WIPP from LANL. This has resulted in a net reduction of about 1,000 m{sup 3} of TRU inventory over that time frame. This paper presents progress in dispositioning legacy and newly-generated transuranic waste (TRU) from ongoing missions at the LANL. The plans for, and lessons learned, in dispositioning several hundred high-activity TRU waste drums are reviewed. This waste population was one of the highest risks at LANL. Technical challenges in disposition of the high-activity drums are presented. These provide a preview of challenges to be addressed in dispositioning the remaining 6,800 m{sup 3} of TRU stored above ground and 2,400 m{sup 3} of TRU waste that is 'retrievably' stored below-grade. LANL is using subcontractors for much of this work and has formed a strong partnership with WIPP and its contractor to address this cleanup challenge.

Shepard, Mark D [Los Alamos National Laboratory; Stiger, Susan G [Los Alamos National Laboratory; Blankenhorn, James A [Los Alamos National Laboratory; Rael, George J [Los Alamos National Laboratory; Moody, David C [U.S DOE

2009-01-01T23:59:59.000Z

27

CERAMIC WASTE FORM DATA PACKAGE  

SciTech Connect (OSTI)

The purpose of this data package is to provide information about simulated crystalline waste forms that can be used to select an appropriate composition for a Cold Crucible Induction Melter (CCIM) proof of principle demonstration. Melt processing, viscosity, electrical conductivity, and thermal analysis information was collected to assess the ability of two potential candidate ceramic compositions to be processed in the Idaho National Laboratory (INL) CCIM and to guide processing parameters for the CCIM operation. Given uncertainties in the CCIM capabilities to reach certain temperatures throughout the system, one waste form designated 'Fe-MP' was designed towards enabling processing and another, designated 'CAF-5%TM-MP' was designed towards optimized microstructure. Melt processing studies confirmed both compositions could be poured from a crucible at 1600{degrees}C although the CAF-5%TM-MP composition froze before pouring was complete due to rapid crystallization (upon cooling). X-ray diffraction measurements confirmed the crystalline nature and phase assemblages of the compositions. The kinetics of melting and crystallization appeared to vary significantly between the compositions. Impedance spectroscopy results indicated the electrical conductivity is acceptable with respect to processing in the CCIM. The success of processing either ceramic composition will depend on the thermal profiles throughout the CCIM. In particular, the working temperature of the pour spout relative to the bulk melter which can approach 1700{degrees}C. The Fe-MP composition is recommended to demonstrate proof of principle for crystalline simulated waste forms considering the current configuration of INL's CCIM. If proposed modifications to the CCIM can maintain a nominal temperature of 1600{degrees}C throughout the melter, drain, and pour spout, then the CAF-5%TM-MP composition should be considered for a proof of principle demonstration.

Amoroso, J.; Marra, J.

2014-06-13T23:59:59.000Z

28

Engineered waste-package-system design specification  

SciTech Connect (OSTI)

This report documents the waste package performance requirements and geologic and waste form data bases used in developing the conceptual designs for waste packages for salt, tuff, and basalt geologies. The data base reflects the latest geotechnical information on the geologic media of interest. The parameters or characteristics specified primarily cover spent fuel, defense high-level waste, and commercial high-level waste forms. The specification documents the direction taken during the conceptual design activity. A separate design specification will be developed prior to the start of the preliminary design activity.

Not Available

1983-05-01T23:59:59.000Z

29

EM Waste and Materials Disposition & Transportation  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergy CooperationRequirementsDraft Environmental ImpactEM Active Sitesof Energy WasteOn

30

Waste Package Component Design Methodology Report  

SciTech Connect (OSTI)

This Executive Summary provides an overview of the methodology being used by the Yucca Mountain Project (YMP) to design waste packages and ancillary components. This summary information is intended for readers with general interest, but also provides technical readers a general framework surrounding a variety of technical details provided in the main body of the report. The purpose of this report is to document and ensure appropriate design methods are used in the design of waste packages and ancillary components (the drip shields and emplacement pallets). The methodology includes identification of necessary design inputs, justification of design assumptions, and use of appropriate analysis methods, and computational tools. This design work is subject to ''Quality Assurance Requirements and Description''. The document is primarily intended for internal use and technical guidance for a variety of design activities. It is recognized that a wide audience including project management, the U.S. Department of Energy (DOE), the U.S. Nuclear Regulatory Commission, and others are interested to various levels of detail in the design methods and therefore covers a wide range of topics at varying levels of detail. Due to the preliminary nature of the design, readers can expect to encounter varied levels of detail in the body of the report. It is expected that technical information used as input to design documents will be verified and taken from the latest versions of reference sources given herein. This revision of the methodology report has evolved with changes in the waste package, drip shield, and emplacement pallet designs over many years and may be further revised as the design is finalized. Different components and analyses are at different stages of development. Some parts of the report are detailed, while other less detailed parts are likely to undergo further refinement. The design methodology is intended to provide designs that satisfy the safety and operational requirements of the YMP. Four waste package configurations have been selected to illustrate the application of the methodology during the licensing process. These four configurations are the 21-pressurized water reactor absorber plate waste package (21-PWRAP), the 44-boiling water reactor waste package (44-BWR), the 5 defense high-level radioactive waste (HLW) DOE spent nuclear fuel (SNF) codisposal short waste package (5-DHLWDOE SNF Short), and the naval canistered SNF long waste package (Naval SNF Long). Design work for the other six waste packages will be completed at a later date using the same design methodology. These include the 24-boiling water reactor waste package (24-BWR), the 21-pressurized water reactor control rod waste package (21-PWRCR), the 12-pressurized water reactor waste package (12-PWR), the 5 defense HLW DOE SNF codisposal long waste package (5-DHLWDOE SNF Long), the 2 defense HLW DOE SNF codisposal waste package (2-MC012-DHLW), and the naval canistered SNF short waste package (Naval SNF Short). This report is only part of the complete design description. Other reports related to the design include the design reports, the waste package system description documents, manufacturing specifications, and numerous documents for the many detailed calculations. The relationships between this report and other design documents are shown in Figure 1.

D.C. Mecham

2004-07-12T23:59:59.000Z

31

Generation!and!Disposition!of!Municipal!Solid!Waste! (MSW)!in!the!United!States!A!National!Survey!  

E-Print Network [OSTI]

! 1! ! Generation!and!Disposition!of!Municipal!Solid!Waste! (MSW Waste (MSW) Generation and Disposition in the U.S., in collaboration with Ms. Nora Goldstein of Bio in 2012 and in 2013 EEC and BioCycle agreed that the 2013 Survey of Waste Generation and Disposition

32

44-BWR WASTE PACKAGE LOADING CURVE EVALUATION  

SciTech Connect (OSTI)

The objective of this calculation is to evaluate the required minimum burnup as a function of initial boiling water reactor (BWR) assembly enrichment that would permit loading of spent nuclear fuel into the 44 BWR waste package configuration as provided in Attachment IV. This calculation is an application of the methodology presented in ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003). The scope of this calculation covers a range of enrichments from 0 through 5.0 weight percent (wt%) U-235, and a burnup range of 0 through 40 GWd/MTU. This activity supports the validation of the use of burnup credit for commercial spent nuclear fuel applications. The intended use of these results will be in establishing BWR waste package configuration loading specifications. Limitations of this evaluation are as follows: (1) The results are based on burnup credit for actinides and selected fission products as proposed in YMP (2003, Table 3-1) and referred to as the ''Principal Isotopes''. Any change to the isotope listing will have a direct impact on the results of this report. (2) The results of 100 percent of the current BWR projected waste stream being able to be disposed of in the 44-BWR waste package with Ni-Gd Alloy absorber plates is contingent upon the referenced waste stream being sufficiently similar to the waste stream received for disposal. (3) The results are based on 1.5 wt% Gd in the Ni-Gd Alloy material and having no tuff inside the waste package. If the Gd loading is reduced or a process to introduce tuff inside the waste package is defined, then this report would need to be reevaluated based on the alternative materials.

J.M. Scaglione

2004-08-25T23:59:59.000Z

33

A Robust Power Remote Manipulator for Use in Waste Sorting, Processing, and Packaging - 12158  

SciTech Connect (OSTI)

Disposition of radioactive waste is one of the Department of Energy's (DOE's) highest priorities. A critical component of the waste disposition strategy is shipment of Transuranic (TRU) waste from DOE's Oak Ridge Reservation to the Waste Isolation Plant Project (WIPP) in Carlsbad, New Mexico. This is the mission of the DOE TRU Waste Processing Center (TWPC). The remote-handled TRU waste at the Oak Ridge Reservation is currently in a mixed waste form that must be repackaged in to meet WIPP Waste Acceptance Criteria (WAC). Because this remote-handled legacy waste is very diverse, sorting, size reducing, and packaging will require equipment flexibility and strength that is not possible with standard master-slave manipulators. To perform the wide range of tasks necessary with such diverse, highly contaminated material, TWPC worked with S.A. Technology (SAT) to modify SAT's Power Remote Manipulator (PRM) technology to provide the processing center with an added degree of dexterity and high load handling capability inside its shielded cells. TWPC and SAT incorporated innovative technologies into the PRM design to better suit the operations required at TWPC, and to increase the overall capability of the PRM system. Improving on an already proven PRM system will ensure that TWPC gains the capabilities necessary to efficiently complete its TRU waste disposition mission. The collaborative effort between TWPC and S.A. Technology has yielded an extremely capable and robust solution to perform the wide range of tasks necessary to repackage TRU waste containers at TWPC. Incorporating innovative technologies into a proven manipulator system, these PRMs are expected to be an important addition to the capabilities available to shielded cell operators. The PRMs provide operators with the ability to reach anywhere in the cell, lift heavy objects, perform size reduction associated with the disposition of noncompliant waste. Factory acceptance testing of the TWPC Powered Remote Manipulators has completed at SAT's Colorado facility, and on-site training at TWPC is scheduled to start in early 2012. (authors)

Cole, Matt; Martin, Scott [S.A. Technology, Loveland, Colorado 80537, Transuranic Waste Processing Center, Lenoir City, TN 37771 (United States)

2012-07-01T23:59:59.000Z

34

WASTE PACKAGE REMEDIATION SYSTEM DESCRIPTION DOCUMENT  

SciTech Connect (OSTI)

The Waste Package Remediation System remediates waste packages (WPs) and disposal containers (DCs) in one of two ways: preparation of rejected DC closure welds for repair or opening of the DC/WP. DCs are brought to the Waste Package Remediation System for preparation of rejected closure welds if testing of the closure weld by the Disposal Container Handling System indicates an unacceptable, but repairable, welding flaw. DC preparation of rejected closure welds will require removal of the weld in such a way that the Disposal Container Handling System may resume and complete the closure welding process. DCs/WPs are brought to the Waste Package Remediation System for opening if the Disposal Container Handling System testing of the DC closure weld indicates an unrepairable welding flaw, or if a WP is recovered from the subsurface repository because suspected damage to the WP or failure of the WP has occurred. DC/WP opening will require cutting of the DC/WP such that a temporary seal may be installed and the waste inside the DC/WP removed by another system. The system operates in a Waste Package Remediation System hot cell located in the Waste Handling Building that has direct access to the Disposal Container Handling System. One DC/WP at a time can be handled in the hot cell. The DC/WP arrives on a transfer cart, is positioned within the cell for system operations, and exits the cell without being removed from the cart. The system includes a wide variety of remotely operated components including a manipulator with hoist and/or jib crane, viewing systems, machine tools for opening WPs, and equipment used to perform pressure and gas composition sampling. Remotely operated equipment is designed to facilitate DC/WP decontamination and hot cell equipment maintenance, and interchangeable components are provided where appropriate. The Waste Package Remediation System interfaces with the Disposal Container Handling System for the receipt and transport of WPs and DCs. The Waste Handling Building System houses the system, and provides the facility, safety, and auxiliary systems required to support operations. The system receives power from the Waste Handling Building Electrical System. The system also interfaces with the various DC systems.

N.D. Sudan

2000-06-22T23:59:59.000Z

35

EM Makes Significant Progress on Dispositioning Transuranic Waste at Idaho Site  

Broader source: Energy.gov [DOE]

IDAHO FALLS, Idaho – EM and contractor CH2M-WG, IDAHO, LLC (CWI) made significant progress in 2013 dispositioning transuranic (TRU) waste and helping ship it out of Idaho.

36

Industrial Waste Landfill IV upgrade package  

SciTech Connect (OSTI)

The Y-12 Plant, K-25 Site, and ORNL are managed by DOE`s Operating Contractor (OC), Martin Marietta Energy Systems, Inc. (Energy Systems) for DOE. Operation associated with the facilities by the Operating Contractor and subcontractors, DOE contractors and the DOE Federal Building result in the generation of industrial solid wastes as well as construction/demolition wastes. Due to the waste streams mentioned, the Y-12 Industrial Waste Landfill IV (IWLF-IV) was developed for the disposal of solid industrial waste in accordance to Rule 1200-1-7, Regulations Governing Solid Waste Processing and Disposal in Tennessee. This revised operating document is a part of a request for modification to the existing Y-12 IWLF-IV to comply with revised regulation (Rule Chapters 1200-1-7-.01 through 1200-1-7-.08) in order to provide future disposal space for the ORR, Subcontractors, and the DOE Federal Building. This revised operating manual also reflects approved modifications that have been made over the years since the original landfill permit approval. The drawings referred to in this manual are included in Drawings section of the package. IWLF-IV is a Tennessee Department of Environmental and Conservation/Division of Solid Waste Management (TDEC/DSWM) Class 11 disposal unit.

Not Available

1994-03-29T23:59:59.000Z

37

Generation!and!Disposition!of!Municipal!Solid!Waste! (MSW)!in!the!United!States!A!National!Survey!  

E-Print Network [OSTI]

! 1! ! Generation!and!Disposition!of!Municipal!Solid!Waste! (MSW on Municipal Solid Waste (MSW) Generation and Disposition in the U.S., in collaboration with Ms. Nora Goldstein was not carried out in 2012 and in 2013 EEC and BioCycle agreed that the 2013 Survey of Waste Generation

Columbia University

38

DHLW Glass Waste Package Criticality Analysis (SCPB:N/A)  

SciTech Connect (OSTI)

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to determine the viability of the Defense High-Level Waste (DHLW) Glass waste package concept with respect to criticality regulatory requirements in compliance with the goals of the Waste Package Implementation Plan (Ref. 5.1) for conceptual design. These design calculations are performed in sufficient detail to provide a comprehensive comparison base with other design alternatives. The objective of this evaluation is to show to what extent the concept meets the regulatory requirements or indicate additional measures that are required for the intact waste package.

J.W. Davis

1996-03-29T23:59:59.000Z

39

Complications Associated with Long-Term Disposition of Newly-Generated Transuranic Waste: A National Laboratory Perspective  

SciTech Connect (OSTI)

The Idaho National Laboratory (INL) is a multipurpose national laboratory delivering specialized science and engineering solutions for the U.S. Department of Energy (DOE). Sponsorship of INL was formally transferred to the DOE Office of Nuclear Energy, Science and Technology (NE) by Secretary Spencer Abraham in July 2002. The move to NE, and designation as the DOE lead nuclear energy laboratory for reactor technology, supports the nation’s expanding nuclear energy initiatives, placing INL at the center of work to develop advanced Generation IV nuclear energy systems; nuclear energy/hydrogen coproduction technology; advanced nuclear energy fuel cycle technologies; and providing national security answers to national infrastructure needs. As a result of the Laboratory’s NE mission, INL generates both contact-handled and remote-handled transuranic (TRU) waste from ongoing operations. Generation rates are relatively small and fluctuate based on specific programs and project activities being conducted; however, the Laboratory will continue to generate TRU waste well into the future in association with the NE mission. Currently, plans and capabilities are being established to transfer INL’s contact-handled TRU waste to the Advanced Mixed Waste Treatment Plant (AMWTP) for certification and disposal to the Waste Isolation Pilot Plant (WIPP). Remote-handled TRU waste is currently placed in storage at the Materials and Fuels Complex (MFC). In an effort to minimize future liabilities associated with the INL NE mission, INL is evaluating and assessing options for the management and disposition of all its TRU waste on a real-time basis at time of generation. This paper summarizes near-term activities to minimize future re handling of INL’s TRU waste, as well as, potential complications associated with the long-term disposition of newly-generated TRU waste. Potential complications impacting the disposition of INL newly-generated TRU waste include, but are not limited to: 1) required remote-handled TRU packaging configuration(s) vs. current facility capabilities, 2) long-term NE mission activities, 3) WIPP certification requirements, and 4) budget considerations.

B.J. Orchard; L.A. Harvego; T.L. Carlson; R.P. Grant

2009-03-01T23:59:59.000Z

40

CH Packaging Operations for High Wattage Waste at LANL  

SciTech Connect (OSTI)

This procedure provides instructions for assembling the following CH packaging payload: Drum payload assembly Standard Waste Box (SWB) assembly Ten-Drum Overpack (TDOP).

Washington TRU Solutions LLC

2005-04-04T23:59:59.000Z

Note: This page contains sample records for the topic "waste disposition packaging" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

CH Packaging Operations for High Wattage Waste at LANL  

SciTech Connect (OSTI)

This procedure provides instructions for assembling the following CH packaging payload: Drum payload assembly Standard Waste Box (SWB) assembly Ten-Drum Overpack (TDOP).

Washington TRU Solutions LLC

2005-04-13T23:59:59.000Z

42

Commercial Spent Nuclear Fuel Waste Package Misload Analysis  

SciTech Connect (OSTI)

The purpose of this calculation is to estimate the probability of misloading a commercial spent nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or burnup) outside the waste package design. The waste package designs are based on the expected commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1 and Table 1). For this calculation, a misloaded waste package is defined as a waste package that has a fuel assembly(s) loaded into it with an enrichment and/or burnup outside the waste package design. An example of this type of misload is a fuel assembly designated for the 21-PWR Control Rod waste package being incorrectly loaded into a 21-PWR Absorber Plate waste package. This constitutes a misloaded 21-PWR Absorber Plate waste package, because the reactivity (i.e., enrichment and/or burnup) of a 21-PWR Control Rod waste package fuel assembly is outside the design of a 21-PWR Absorber Plate waste package. These types of misloads (i.e., fuel assembly with enrichment and/or burnup outside waste package design) are the only types that are evaluated in this calculation. This calculation utilizes information from ''Frequency of SNF Misload for Uncanistered Fuel Waste Package'' (CRWMS M&O 1998) as the starting point. The scope of this calculation is limited to the information available. The information is based on the whole population of fuel assemblies and the whole population of waste packages, because there is no information about the arrival of the waste stream at this time. The scope of this calculation deviates from that specified in ''Technical Work Plan for: Risk and Criticality Department'' (BSC 2002a, Section 2.1.30) in that only waste package misload is evaluated. The remaining issues identified (i.e., flooding and geometry reconfiguration) will be addressed elsewhere. The intended use of the calculation is to provide information and inputs to the Preclosure Safety Analysis Department. Before using the results of this calculation, the reader is cautioned to verify that the assumptions made in this calculation regarding the waste stream, the loading process, and the staging of the spent nuclear fuel assemblies are applicable.

A. Alsaed

2005-07-28T23:59:59.000Z

43

Commercial Spent Nuclear Fuel Waste Package Misload Analysis  

SciTech Connect (OSTI)

The purpose of this calculation is to estimate the probability of misloading a commercial spent nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or burnup) outside the waste package design. The waste package designs are based on the expected commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1 and Table 1). For this calculation, a misloaded waste package is defined as a waste package that has a fuel assembly(s) loaded into it with an enrichment and/or burnup outside the waste package design. An example of this type of misload is a fuel assembly designated for the 21-PWR Control Rod waste package being incorrectly loaded into a 21-PWR Absorber Plate waste package. This constitutes a misloaded 21-PWR Absorber Plate waste package, because the reactivity (i.e., enrichment and/or burnup) of a 21-PWR Control Rod waste package fuel assembly is outside the design of a 21-PWR Absorber Plate waste package. These types of misloads (i.e., fuel assembly with enrichment and/or burnup outside waste package design) are the only types that are evaluated in this calculation. This calculation utilizes information from ''Frequency of SNF Misload for Uncanistered Fuel Waste Package'' (CRWMS M&O 1998) as the starting point. The scope of this calculation is limited to the information available. The information is based on the whole population of fuel assemblies and the whole population of waste packages, because there is no information about the arrival of the waste stream at this time. The scope of this calculation deviates from that specified in ''Technical Work Plan for: Risk and Criticality Department'' (BSC 2002a, Section 2.1.30) in that only waste package misload is evaluated. The remaining issues identified (i.e., flooding and geometry reconfiguration) will be addressed elsewhere. The intended use of the calculation is to provide information and inputs to the Preclosure Safety Analysis Department. Before using the results of this calculation, the reader is cautioned to verify that the assumptions made in this calculation regarding the waste stream, the loading process, and the staging of the spent nuclear fuel assemblies are applicable.

J.K. Knudson

2003-10-02T23:59:59.000Z

44

Depleted uranium as a backfill for nuclear fuel waste package  

DOE Patents [OSTI]

A method for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package.

Forsberg, Charles W. (Oak Ridge, TN)

1998-01-01T23:59:59.000Z

45

Depleted uranium as a backfill for nuclear fuel waste package  

DOE Patents [OSTI]

A method is described for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package. 6 figs.

Forsberg, C.W.

1998-11-03T23:59:59.000Z

46

Superfund Policy Statements and Guidance Regarding Disposition of Radioactive Waste in Non-NRC Licensed Disposal Facilities - 13407  

SciTech Connect (OSTI)

This talk will discuss EPA congressional testimony and follow-up letters, as well as letters to other stakeholders on EPA's perspectives on the disposition of radioactive waste outside of the NRC licensed disposal facility system. This will also look at Superfund's historical practices, and emerging trends in the NRC and agreement states on waste disposition. (author)

Walker, Stuart [U.S. Environmental Protection Agency (United States)] [U.S. Environmental Protection Agency (United States)

2013-07-01T23:59:59.000Z

47

Lessons Learned from Three Mile Island Packaging, Transportation and Disposition that Apply to Fukushima Daiichi Recovery  

SciTech Connect (OSTI)

Following the massive earthquake and resulting tsunami damage in March of 2011 at the Fukushima Daiichi nuclear power plant in Japan, interest was amplified for what was done for recovery at the Three Mile Island Unit 2 (TMI-2) in the United States following its meltdown in 1979. Many parallels could be drawn between to two accidents. This paper presents the results of research done into the TMI-2 recovery effort and its applicability to the Fukushima Daiichi cleanup. This research focused on three topics: packaging, transportation, and disposition. This research work was performed as a collaboration between Japan’s Central Research Institute of Electric Power Industry (CRIEPI) and the Idaho National Laboratory (INL). Hundreds of TMI-2 related documents were searched and pertinent information was gleaned from these documents. Other important information was also obtained by interviewing employees who were involved first hand in various aspects of the TMI-2 cleanup effort. This paper is organized into three main sections: (1) Transport from Three Mile Island to Central Facilities Area at INL, (2) Transport from INL Central Receiving Facility to INL Test Area North (TAN) and wet storage at TAN, and (3) Transport from TAN to INL Idaho Nuclear Technology and Engineering Center (INTEC) and Dry Storage at INTEC. Within each of these sections, lessons learned from performing recovery activities are presented and their applicability to the Fukushima Daiichi nuclear power plant cleanup are outlined.

Layne Pincock; Wendell Hintze; Dr. Koji Shirai

2012-07-01T23:59:59.000Z

48

Savannah River Site Achieves Transuranic Waste Disposition Goal...  

Office of Environmental Management (EM)

liquid waste contractor, Savannah River Remediation (SRR): Closed two more underground tanks containing radioactive waste, helping reduce a significant environmental risk to South...

49

Packaging and transportation manual. Chapter on the packaging and transportation of hazardous and radioactive waste  

SciTech Connect (OSTI)

The purpose of this chapter is to outline the requirements that Los Alamos National Laboratory employees and contractors must follow when they package and ship hazardous and radioactive waste. This chapter is applied to on-site, intra-Laboratory, and off-site transportation of hazardous and radioactive waste. The chapter contains sections on definitions, responsibilities, written procedures, authorized packaging, quality assurance, documentation for waste shipments, loading and tiedown of waste shipments, on-site routing, packaging and transportation assessment and oversight program, nonconformance reporting, training of personnel, emergency response information, and incident and occurrence reporting. Appendices provide additional detail, references, and guidance on packaging for hazardous and radioactive waste, and guidance for the on-site transport of these wastes.

NONE

1998-03-01T23:59:59.000Z

50

Remote Handling Equipment for a High-Level Waste Waste Package Closure System  

SciTech Connect (OSTI)

High-level waste will be placed in sealed waste packages inside a shielded closure cell. The Idaho National Laboratory (INL) has designed a system for closing the waste packages including all cell interior equipment and support systems. This paper discusses the material handling aspects of the equipment used and operations that will take place as part of the waste package closure operations. Prior to construction, the cell and support system will be assembled in a full-scale mockup at INL.

Kevin M. Croft; Scott M. Allen; Mark W. Borland

2006-04-01T23:59:59.000Z

51

General Corrosion and Localized Corrosion of Waste Package Outer Barrier  

SciTech Connect (OSTI)

The waste package design for the License Application is a double-wall waste package underneath a protective drip shield (BSC 2004 [DIRS 168489]; BSC 2004 [DIRS 169480]). The purpose and scope of this model report is to document models for general and localized corrosion of the waste package outer barrier (WPOB) to be used in evaluating waste package performance. The WPOB is constructed of Alloy 22 (UNS N06022), a highly corrosion-resistant nickel-based alloy. The inner vessel of the waste package is constructed of Stainless Steel Type 316 (UNS S31600). Before it fails, the Alloy 22 WPOB protects the Stainless Steel Type 316 inner vessel from exposure to the external environment and any significant degradation. The Stainless Steel Type 316 inner vessel provides structural stability to the thinner Alloy 22 WPOB. Although the waste package inner vessel would also provide some performance for waste containment and potentially decrease the rate of radionuclide transport after WPOB breach before it fails, the potential performance of the inner vessel is far less than that of the more corrosion-resistant Alloy 22 WPOB. For this reason, the corrosion performance of the waste package inner vessel is conservatively ignored in this report and the total system performance assessment for the license application (TSPA-LA). Treatment of seismic and igneous events and their consequences on waste package outer barrier performance are not specifically discussed in this report, although the general and localized corrosion models developed in this report are suitable for use in these scenarios. The localized corrosion processes considered in this report are pitting corrosion and crevice corrosion. Stress corrosion cracking is discussed in ''Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material'' (BSC 2004 [DIRS 169985]).

K.G. Mon

2004-10-01T23:59:59.000Z

52

Contact-Handled and Remote-Handled Transuranic Waste Packaging  

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

Provides specific instructions for packaging and/or repackaging contact-handled transuranic (CH-TRU) and remote-handled transuranic (RH-TRU) waste in a manner consistent with DOE O 435.1, Radioactive Waste Management, DOE M 435.1-1 Chg 1, Radioactive Waste Management Manual, CH-TRU and RH-TRU waste transportation requirements, and Waste Isolation Pilot Plant (WIPP) programmatic requirements. Does not cancel other directives.

2011-08-09T23:59:59.000Z

53

Potential dispositioning flowsheets for ICPP SNF and wastes  

SciTech Connect (OSTI)

The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Laboratory (INEL), has reprocessed irradiated nuclear fuels for the US Department of Energy (DOE) since 1953. This activity resulted mainly in the recovery of uranium and the management of the resulting wastes. The acidic radioactive high-level liquid waste was routinely stored in stainless steel tanks and then calcined to form a dry granular solid. The calcine is stored in stainless steel bins that are housed in underground concrete vaults. In April 1992, the DOE discontinued the practice of reprocessing irradiated nuclear fuels. This decision has left a legacy of 1.8 million gallons of radioactive liquid wastes (1.5 million gallons of radioactive sodium-bearing liquid wastes and 0.3 million gallons of high-level liquid waste), 3800 cubic meters of calcine waste, and 289 metric tons of heavy metal within unprocessed spent nuclear fuel (SNF) left in inventory at the ICPP. The nation`s radioactive waste policy has been established by the Nuclear Waste Policy Act (NWPA), which requires the final disposal of SNF and radioactive waste in accordance with US Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) standards. In accordance with these regulations and other legal agreements between the State of Idaho and the DOE, the DOE must, among other requirements, (1) complete a final Environmental Impact Statement by April 30, 1995, (2) evaluate and test sodium-bearing waste pre-treatment technologies, (3) select the sodium-bearing and calcine waste pre-treatment technology, if necessary, by June 1, 1995, and (4) select a technology for converting calcined waste into an appropriate disposal form by June 1, 1995.

Olson, A.L. [ed.; Anderson, P.A.; Bendixsen, C.L. [and others

1995-11-01T23:59:59.000Z

54

DRAFT EM SSAB Chairs Meeting Waste Disposition Strategies...  

Office of Environmental Management (EM)

EM SSAB Chairs Meeting Christine M. Gelles Deputy Assistant Secretary for Waste Management Office of Environmental Management 5 November 2013 Educational Session 1 - Discussion on...

55

Providing Innovative Waste Management Disposition for the DOE...  

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

Trains for Waste Shipments to Yucca Mountain." Under this policy DOE will use dedicated train service - train service dedicated to one Commodity - for its rail transport of spent...

56

DESIGN ANALYSIS FOR THE NAVAL SNF WASTE PACKAGE  

SciTech Connect (OSTI)

The purpose of this analysis is to demonstrate the design of the naval spent nuclear fuel (SNF) waste package (WP) using the Waste Package Department's (WPD) design methodologies and processes described in the ''Waste Package Design Methodology Report'' (CRWMS M&O [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000b). The calculations that support the design of the naval SNF WP will be discussed; however, only a sub-set of such analyses will be presented and shall be limited to those identified in the ''Waste Package Design Sensitivity Report'' (CRWMS M&O 2000c). The objective of this analysis is to describe the naval SNF WP design method and to show that the design of the naval SNF WP complies with the ''Naval Spent Nuclear Fuel Disposal Container System Description Document'' (CRWMS M&O 1999a) and Interface Control Document (ICD) criteria for Site Recommendation. Additional criteria for the design of the naval SNF WP have been outlined in Section 6.2 of the ''Waste Package Design Sensitivity Report'' (CRWMS M&O 2000c). The scope of this analysis is restricted to the design of the naval long WP containing one naval long SNF canister. This WP is representative of the WPs that will contain both naval short SNF and naval long SNF canisters. The following items are included in the scope of this analysis: (1) Providing a general description of the applicable design criteria; (2) Describing the design methodology to be used; (3) Presenting the design of the naval SNF waste package; and (4) Showing compliance with all applicable design criteria. The intended use of this analysis is to support Site Recommendation reports and assist in the development of WPD drawings. Activities described in this analysis were conducted in accordance with the technical product development plan (TPDP) ''Design Analysis for the Naval SNF Waste Package (CRWMS M&O 2000a).

T.L. Mitchell

2000-05-31T23:59:59.000Z

57

Idaho High-Level Waste & Facilities Disposition, Final Environmental Impact Statement  

SciTech Connect (OSTI)

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 waste at the Idaho National Engineering and Environmental Laboratory (INEEL) in liquid and solid forms. This EIS also analyzes alternatives for the final disposition of HLW management facilities at the INEEL after their missions are completed. After considering comments on the Draft EIS (DOE/EIS-0287D), as well as information on available treatment technologies, DOE and the State of Idaho have identified separate preferred alternatives for waste treatment. DOE's preferred alternative for waste treatment is performance based with the focus on placing the wastes in forms suitable for disposal. Technologies available to meet the performance objectives may be chosen from the action alternatives analyzed in this EIS. The State of Idaho's Preferred Alternative for treating mixed transuranic waste/SBW and calcine is vitrification, with or without calcine separations. Under both the DOE and State of Idaho preferred alternatives, newly generated liquid waste would be segregated after 2005, stored or treated directly and disposed of as low-level, mixed low-level, or transuranic waste depending on its characteristics. The objective of each preferred alternative is to enable compliance with the legal requirement to have INEEL HLW road ready by a target date of 2035. Both DOE and the State of Idaho have identified the same preferred alternative for facilities disposition, which is to use performance-based closure methods for existing facilities and to design new facilities consistent with clean closure methods.

N /A

2002-10-11T23:59:59.000Z

58

DATA QUALITY OBJECTIVES SUMMARY REPORT FOR WASTE DISPOSITION OF FY2004 ISRM INJECTION & MONITORING WELLS  

SciTech Connect (OSTI)

The purpose of this data quality objective (DQO) summary report is to develop a sampling plan for waste disposition of soil cuttings and other drilling-related wastes that will result from the drilling of 21 injection wells and one groundwater monitoring well west of the 184-D Powerhouse Ash Pit in the 100-D Area of the Hanford Site. The 21 In Situ Redox Manipulation (ISRM) wells will inject treatment solutions to assist in intercepting and preventing the discharge of a hexavalent chromium plume to the Columbia River. The monitoring well will help establish groundwater chemistry downgradient of the ISRM zone. The proposed well locations are shown.

THOMAS, G.

2004-03-03T23:59:59.000Z

59

Transportable Vitrification System RCRA Closure Practical Waste Disposition Saves Time And Money  

SciTech Connect (OSTI)

The Transportable Vitrification System (TVS) was a large-scale vitrification system for the treatment of mixed wastes. The wastes contained both hazardous and radioactive materials in the form of sludge, soil, and ash. The TVS was developed to be moved to various United States Department of Energy (DOE) facilities to vitrify mixed waste as needed. The TVS consists of four primary modules: (1) Waste and Additive Materials Processing Module; (2) Melter Module; (3) Emissions Control Module; and (4) Control and Services Module. The TVS was demonstrated at the East Tennessee Technology Park (ETTP) during September and October of 1997. During this period, approximately 16,000 pounds of actual mixed waste was processed, producing over 17,000 pounds of glass. After the demonstration was complete it was determined that it was more expensive to use the TVS unit to treat and dispose of mixed waste than to direct bury this waste in Utah permitted facility. Thus, DOE had to perform a Resource Conservation and Recovery Act (RCRA) closure of the facility and find a reuse for as much of the equipment as possible. This paper will focus on the following items associated with this successful RCRA closure project: TVS site closure design and implementation; characterization activities focused on waste disposition; pollution prevention through reuse; waste minimization efforts to reduce mixed waste to be disposed; and lessons learned that would be integrated in future projects of this magnitude.

Brill, Angie; Boles, Roger; Byars, Woody

2003-02-26T23:59:59.000Z

60

DRAFT EM SSAB Chair's Meeting Waste Disposition Strategies Update  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelinesProvedDecember 2005Department ofDOE AccidentWaste Isolation Pilotat4 VolumeDOJ

Note: This page contains sample records for the topic "waste disposition packaging" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

Waste and Materials Disposition Information | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergyENERGYWomenthe House Committee onsupports high impact projectsMatt8 Image:Waste and Materials

62

EM Waste and Materials Disposition & Transportation | Department of Energy  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergy CooperationRequirementsDraft Environmental ImpactEM Active Sitesof Energy WasteOnEM

63

Disposition of nuclear waste using subcritical accelerator-driven systems  

SciTech Connect (OSTI)

Spent fuel from nuclear power plants contains large quantities of Pu, other actinides, and fission products (FP). This creates challenges for permanent disposal because of the long half-lives of some isotopes and the potential for diversion of the fissile material. Two issues of concern for the US repository concept are: (1) long-term radiological risk peaking tens-of-thousands of years in the future; and (2) short-term thermal loading (decay heat) that limits capacity. An accelerator-driven neutron source can destroy actinides through fission, and can convert long-lived fission products to shorter-lived or stable isotopes. Studies over the past decade have established that accelerator transmutation of waste (ATW) can have a major beneficial impact on the nuclear waste problem. Specifically, the ATW concept the authors are evaluating: (1) destroys over 99.9% of the actinides; (2) destroys over 99.9% of the Tc and I; (3) separates Sr-90 and Cs-137; (4) separates uranium from the spent fuel; (5) produces electric power.

Venneri, F.; Li, N.; Williamson, M.; Houts, M.; Lawrence, G.

1998-12-31T23:59:59.000Z

64

Disposition of Nuclear Waste Using Subcritical Accelerator-Driven Systems  

SciTech Connect (OSTI)

ATW destroys virtually all the plutonium and higher actinides without reprocessing the spent fuel in a way that could lead to weapons material diversion. An ATW facility consists of three major elements: (1) a high-power proton linear accelerator; (2) a pyrochemical spent fuel treatment i waste cleanup system; (3) a liquid lead-bismuth cooled burner that produces and utilizes an intense source-driven neutron flux for transmutation in a heterogeneous (solid fuel) core. The concept is the result of many years of development at LANL as well as other major international research centers. Once demonstrated and developed, ATW could be an essential part of a global non-proliferation strategy for countries that could build up large quantities of plutonium from their commercial reactor waste. ATW technology, initially proposed in the US, has received wide and rapidly increasing attention abroad, especially in Europe and the Far East with major programs now being planned, organized and tided. Substantial convergence presently exists on the technology choices among the programs, opening the possibility of a strong and effective international collaboration on the phased development of the ATW technology.

Doolen, G.D.; Venneri, F.; Li, N.; Williamson, M.A.; Houts, M.; Lawrence, G.

1998-06-27T23:59:59.000Z

65

Quality assurance for radioactive waste packages -- A general approach  

SciTech Connect (OSTI)

Radioactive waste packages must fulfill the requirements resulting from regulations concerning handling, treatment, conditioning, transportation, storage and disposal so that the goal of radioactive waste management can be achieved. Usually in different parts of waste management different quality systems are used, and different quality assurance measures are performed. In the paper, these problems ar elucidated and it is explained by means of the quality assurance performed for the disposal of radioactive waste in Germany how the fulfillment of the requirements of the repository can be ensured.

Martens, B.R. [Bundesamt fuer Strahlenschutz, Saltzgitter (Germany)

1993-12-31T23:59:59.000Z

66

Measurement of radionuclides in waste packages  

DOE Patents [OSTI]

A method is described for non-destructively assaying the radionuclide content of solid waste in a sealed container by analysis of the waste's gamma-ray spectrum and neutron emissions. Some radionuclides are measured by characteristic photopeaks in the gamma-ray spectrum; transuranic nuclides are measured by neutron emission rate; other radionuclides are measured by correlation with those already measured.

Brodzinski, R.L.; Perkins, R.W.; Rieck, H.G.; Wogman, N.A.

1984-09-12T23:59:59.000Z

67

Measurement of radionuclides in waste packages  

DOE Patents [OSTI]

A method is described for non-destructively assaying the radionuclide content of solid waste in a sealed container by analysis of the waste's gamma-ray spectrum and neutron emissions. Some radionuclides are measured by characteristic photopeaks in the gamma-ray spectrum; transuranic nuclides are measured by neutron emission rate; other radionuclides are measured by correlation with those already measured.

Brodzinski, Ronald L. (Richland, WA); Perkins, Richard W. (Richland, WA); Rieck, Henry G. (Richland, WA); Wogman, Ned A. (Richland, WA)

1986-01-01T23:59:59.000Z

68

WASTE PACKAGE OPERATIONS FY99 CLOSURE METHODS REPORT  

SciTech Connect (OSTI)

The waste package (WP) closure weld development task is part of a larger engineering development program to develop waste package designs. The purpose of the larger waste package engineering development program is to develop nuclear waste package fabrication and closure methods that the Nuclear Regulatory Commission will find acceptable and will license for disposal of spent nuclear fuel (SNF), non-fuel components, and vitrified high-level waste within a Monitored Geologic Repository (MGR). Within the WP closure development program are several major development tasks, which, in turn, are divided into subtasks. The major tasks include: WP fabrication development, WP closure weld development, nondestructive examination (NDE) development, and remote in-service inspection development. The purpose of this report is to present the objectives, technical information, and work scope relating to the WP closure weld development.and NDE tasks and subtasks and to report results of the closure weld and NDE development programs for fiscal year 1999 (FY-99). The objective of the FY-99 WP closure weld development task was to develop requirements for closure weld surface and volumetric NDE performance demonstrations, investigate alternative NDE inspection techniques, and develop specifications for welding, NDE, and handling system integration. In addition, objectives included fabricating several flat plate mock-ups that could be used for NDE development, stress relief peening, corrosion testing, and residual stress testing.

M. C. Knapp

1999-09-23T23:59:59.000Z

69

Strategy for experimental validation of waste package performance assessment  

SciTech Connect (OSTI)

A strategy for the experimental validation of waste package performance assessment has been developed as part of a program supported by the Repository Technology Program. The strategy was developed by reviewing the results of laboratory analog experiments, in-situ tests, repository simulation tests, and material interaction tests. As a result of the review, a listing of dependent and independent variables that influence the ingress of water into the near-field environment, the reaction between water and the waste form, and the transport of radionuclides from the near-field environment was developed. The variables necessary to incorporate into an experimental validation strategy were chosen by identifying those which had the greatest effect of each of the three major events, i.e., groundwater ingress, waste package reactions, and radionuclide transport. The methodology to perform validation experiments was examined by utilizing an existing laboratory analog approach developed for unsaturated testing of glass waste forms. 185 refs., 9 figs., 2 tabs.

Bates, J.K.; Abrajano, T.A. Jr.; Wronkiewicz, D.J.; Gerding, T.J.; Seils, C.A.

1990-07-01T23:59:59.000Z

70

WAPDEG Analysis of Waste Package and Drip shield Degradation  

SciTech Connect (OSTI)

As directed by ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]), an analysis of the degradation of the engineered barrier system (EBS) drip shields and waste packages at the Yucca Mountain repository is developed. The purpose of this activity is to provide the TSPA with inputs and methodologies used to evaluate waste package and drip shield degradation as a function of exposure time under exposure conditions anticipated in the repository. This analysis provides information useful to satisfy ''Yucca Mountain Review Plan, Final Report'' (NRC 2003 [DIRS 163274]) requirements. Several features, events, and processes (FEPs) are also discussed (Section 6.2, Table 15). The previous revision of this report was prepared as a model report in accordance with AP-SIII.10Q, Models. Due to changes in the role of this report since the site recommendation, it no longer contains model development. This revision is prepared as a scientific analysis in accordance with AP-SIII.9Q, ''Scientific Analyses'' and uses models previously validated in (1) ''Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material'' (BSC 2004 [DIRS 169985]); (2) ''General Corrosion and Localized Corrosion of Waste Package Outer Barrier'' (BSC 2004 [DIRS 169984]); and (3) ''General Corrosion and Localized Corrosion of Drip Shield'' (BSC 2004 [DIRS 169845]). The integrated waste package degradation (IWPD) analysis presented in this report treats several implementation-related issues, such as defining the number and size of patches per waste package that undergo stress corrosion cracking; recasting the weld flaw analysis in a form as implemented in the Closure Weld Defects (CWD) software; and, general corrosion rate manipulations (e.g., change of scale in Section 6.3.4). The weld flaw portion of this report takes input from an engineering calculation (BSC 2004 [DIRS 170024]) and uses standard mathematical methods to enable easier implementation. The IWPD analysis also provides guidance on implementation of early failures (importance sampling and multinomial distribution usage). These manipulations are evident from standard scientific practices, approaches, or methods and do not require changes to the previously validated models. The IWPD analysis itself (Section 6.4), not the resultant curves from executing the IWPD analysis presented in Section 6.5 (which are for illustrative purposes), is used directly in total system performance assessment (TSPA). The IWPD analysis simulates general corrosion and stress corrosion cracking of the waste package outer barrier and general corrosion of the drip shield. The effects of igneous and seismic events and localized corrosion on drip shield and waste package performance are not evaluated in this report. The outputs of this report are inputs and methodologies used by TSPA to evaluate waste package and drip shield degradation as a function of exposure time under exposure conditions anticipated in the repository. The analyses presented in this report are for the current repository design (BSC 2004 [DIRS 168489]).

K. Mon

2004-09-29T23:59:59.000Z

71

PROJECT STRATEGY FOR THE REMEDIATION AND DISPOSITION OF LEGACY TRANSURANIC WASTE AT THE SAVANNAH RIVER SITE, South Carolina, USA  

SciTech Connect (OSTI)

This paper discusses the Savannah River Site Accelerated Transuranic (TRU) Waste Project that was initiated in April of 2009 to accelerate the disposition of remaining legacy transuranic waste at the site. An overview of the project execution strategy that was implemented is discussed along with the lessons learned, challenges and improvements to date associated with waste characterization, facility modifications, startup planning, and remediation activities. The legacy waste was generated from approximately 1970 through 1990 and originated both on site as well as at multiple US Department of Energy sites. Approximately two thirds of the waste was previously dispositioned from 2006 to 2008, with the remaining one third being the more hazardous waste due to its activity (curie content) and the plutonium isotope Pu-238 quantities in the waste. The project strategy is a phased approach beginning with the lower activity waste in existing facilities while upgrades are made to support remediation of the higher activity waste. Five waste remediation process lines will be used to support the full remediation efforts which involve receipt of the legacy waste container, removal of prohibited items, venting of containers, and resizing of contents to fit into current approved waste shipping containers. Modifications have been minimized to the extent possible to meet the accelerated goals and involve limited upgrades to address life safety requirements, radiological containment needs, and handling equipment for the larger waste containers. Upgrades are also in progress for implementation of the TRUPACT III for the shipment of Standard Large Boxes to the Waste Isolation Pilot Plant, the US TRU waste repository. The use of this larger shipping container is necessary for approximately 20% of the waste by volume due to limited size reduction capability. To date, approximately 25% of the waste has been dispositioned, and several improvements have been made to the overall processing plan as well as facility processing rates. These lessons learned, challenges, and improvements will be discussed to aid other sites in their efforts to conduct similar activities.

Rodriguez, M.

2010-12-17T23:59:59.000Z

72

Generic Degraded Congiguration Probability Analysis for DOE Codisposal Waste Package  

SciTech Connect (OSTI)

In accordance with the technical work plan, ''Technical Work Plan For: Department of Energy Spent Nuclear Fuel Work Packages'' (CRWMS M&O 2000c), this Analysis/Model Report (AMR) is developed for the purpose of screening out degraded configurations for U.S. Department of Energy (DOE) spent nuclear fuel (SNF) types. It performs the degraded configuration parameter and probability evaluations of the overall methodology specified in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000, Section 3) to qualifying configurations. Degradation analyses are performed to assess realizable parameter ranges and physical regimes for configurations. Probability calculations are then performed for configurations characterized by k{sub eff} in excess of the Critical Limit (CL). The scope of this document is to develop a generic set of screening criteria or models to screen out degraded configurations having potential for exceeding a criticality limit. The developed screening criteria include arguments based on physical/chemical processes and probability calculations and apply to DOE SNF types when codisposed with the high-level waste (HLW) glass inside a waste package. The degradation takes place inside the waste package and is long after repository licensing has expired. The emphasis of this AMR is on degraded configuration screening and the probability analysis is one of the approaches used for screening. The intended use of the model is to apply the developed screening criteria to each DOE SNF type following the completion of the degraded mode criticality analysis internal to the waste package.

S.F.A. Deng; M. Saglam; L.J. Gratton

2001-05-23T23:59:59.000Z

73

Secondary Waste Form Down Selection Data Package – Ceramicrete  

SciTech Connect (OSTI)

As part of high-level waste pretreatment and immobilized low activity waste processing, liquid secondary wastes will be generated that will be transferred to the Effluent Treatment Facility on the Hanford Site for further treatment. These liquid secondary wastes will be converted to stable solid waste forms that will be disposed in the Integrated Disposal Facility. Currently, four waste forms are being considered for stabilization and solidification of the liquid secondary wastes. These waste forms are Cast Stone, Ceramicrete, DuraLith, and Fluidized Bed Steam Reformer. The preferred alternative will be down selected from these four waste forms. Pacific Northwest National Laboratory is developing data packages to support the down selection process. The objective of the data packages is to identify, evaluate, and summarize the existing information on the four waste forms being considered for stabilization and solidification of the liquid secondary wastes. The information included will be based on information available in the open literature and from data obtained from testing currently underway. This data package is for the Ceramicrete waste form. Ceramicrete is a relatively new engineering material developed at Argonne National Laboratory to treat radioactive and hazardous waste streams (e.g., Wagh 2004; Wagh et al. 1999a, 2003; Singh et al. 2000). This cement-like waste form can be used to treat solids, liquids, and sludges by chemical immobilization, microencapsulation, and/or macroencapsulation. The Ceramicrete technology is based on chemical reaction between phosphate anions and metal cations to form a strong, dense, durable, low porosity matrix that immobilizes hazardous and radioactive contaminants as insoluble phosphates and microencapsulates insoluble radioactive components and other constituents that do not form phosphates. Ceramicrete is a type of phosphate-bonded ceramic, which are also known as chemically bonded phosphate ceramics. The Ceramicrete binder is formed through an acid-base reaction between calcined magnesium oxide (MgO; a base) and potassium hydrogen phosphate (KH{sub 2}PO{sub 4}; an acid) in aqueous solution. The reaction product sets at room temperature to form a highly crystalline material. During the reaction, the hazardous and radioactive contaminants also react with KH{sub 2}PO{sub 4} to form highly insoluble phosphates. In this data package, physical property and waste acceptance data for Ceramicrete waste forms fabricated with wastes having compositions that were similar to those expected for secondary waste effluents, as well as secondary waste effluent simulants from the Hanford Tank Waste Treatment and Immobilization Plant were reviewed. With the exception of one secondary waste form formulation (25FA+25 W+1B.A. fabricated with the mixed simulant did not meet the compressive strength requirement), all the Ceramicrete waste forms that were reviewed met or exceeded Integrated Disposal Facility waste acceptance criteria.

Cantrell, Kirk J.; Westsik, Joseph H.

2011-08-31T23:59:59.000Z

74

Engineering evaluation of alternatives for the disposition of Niagara Falls Storage Site, its residues and wastes  

SciTech Connect (OSTI)

The final disposition scenarios selected by DOE for assessment in this document are consistent with those stated in the Notice of Intent to prepare an Environmental Impact Statement (EIS) for the Niagara Falls Storage Site (NFSS) (DOE, 1983d) and the modifications to the alternatives resulting from the public scoping process. The scenarios are: take no action beyond interim remedial measures other than maintenance and surveillance of the NFSS; retain and manage the NFSS as a long-term waste management facility for the wastes and residues on the site; decontaminate, certify, and release the NFSS for other use, with long-term management of the wastes and residues at other DOE sites; and partially decontaminate the NFSS by removal and transport off site of only the more radioactive residues, and upgrade containment of the remaining wastes and residues on site. The objective of this document is to present to DOE the conceptual engineering, occupational radiation exposure, construction schedule, maintenance and surveillance requirements, and cost information relevant to design and implementation of each of the four scenarios. The specific alternatives within each scenario used as the basis for discussion in this document were evaluated on the bases of engineering considerations, technical feasibility, and regulatory requirements. Selected alternatives determined to be acceptable for each of the four final disposition scenarios for the NFSS were approved by DOE to be assessed and costed in this document. These alternatives are also the subject of the EIS for the NFSS currently being prepared by Argonne National Laboratory (ANL). 40 figures, 38 tables.

Not Available

1984-01-01T23:59:59.000Z

75

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

SciTech Connect (OSTI)

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.

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

2011-09-12T23:59:59.000Z

76

21-PWR WASTE PACKAGE WITH ABSORBER PLATES LOADING CURVE EVALUATION  

SciTech Connect (OSTI)

The objective of this calculation is to evaluate the required minimum burnup as a function of initial pressurized water reactor (PWR) assembly enrichment that would permit loading of spent nuclear fuel into the 21 PWR waste package with absorber plates design as provided in Attachment IV. This calculation is an example of the application of the methodology presented in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003). The scope of this calculation covers a range of enrichments from 0 through 5.0 weight percent U-235, and a burnup range of 0 through 45 GWd/MTU. Higher burnups were not necessary because 45 GWd/MTU was high enough for the loading curve determination. This activity supports the validation of the use of burnup credit for commercial spent nuclear fuel applications. The intended use of these results will be in establishing PWR waste package configuration loading specifications. Limitations of this evaluation are as follows: (1) The results are based on burnup credit for actinides and selected fission products as proposed in YMP (2003, Table 3-1) and referred to as the ''Principal Isotopes''. Any change to the isotope listing will have a direct impact on the results of this report. (2) The results are based on 1.5 wt% Gd in the Ni-Gd Alloy material and having no tuff inside the waste package. If the Gd loading is reduced or a process to introduce tuff inside the waste package is defined, then this report would need to be reevaluated based on the alternative materials. This calculation is subject to the ''Quality Assurance Requirements and Description'' (QARD) (DOE 2004) because it concerns engineered barriers that are included in the ''Q-List'' (BSC 2004k, Appendix A) as items important to safety and waste isolation.

J.M. Scaglione

2004-12-17T23:59:59.000Z

77

Independent Assessment of the Savannah River Site High-Level Waste Salt Disposition Alternatives Evaluation  

SciTech Connect (OSTI)

This report presents the results of the Independent Project Evaluation (IPE) Team assessment of the Westinghouse Savannah River Company High-Level Waste Salt Disposition Systems Engineering (SE) Team's deliberations, evaluations, and selections. The Westinghouse Savannah River Company concluded in early 1998 that production goals and safety requirements for processing SRS HLW salt to remove Cs-137 could not be met in the existing In-Tank Precipitation Facility as currently configured for precipitation of cesium tetraphenylborate. The SE Team was chartered to evaluate and recommend an alternative(s) for processing the existing HLW salt to remove Cs-137. To replace the In-Tank Precipitation process, the Savannah River Site HLW Salt Disposition SE Team downselected (October 1998) 140 candidate separation technologies to two alternatives: Small-Tank Tetraphenylborate (TPB) Precipitation (primary alternative) and Crystalline Silicotitanate (CST) Nonelutable Ion Exchange (backup alternative). The IPE Team, commissioned by the Department of Energy, concurs that both alternatives are technically feasible and should meet all salt disposition requirements. But the IPE Team judges that the SE Team's qualitative criteria and judgments used in their downselection to a primary and a backup alternative do not clearly discriminate between the two alternatives. To properly choose between Small-Tank TPB and CST Ion Exchange for the primary alternative, the IPE Team suggests the following path forward: Complete all essential R and D activities for both alternatives and formulate an appropriate set of quantitative decision criteria that will be rigorously applied at the end of the R and D activities. Concurrent conceptual design activities should be limited to common elements of the alternatives.

J. T. Case (DOE-ID); M. L. Renfro (INEEL)

1998-12-01T23:59:59.000Z

78

Supplement Analysis for the Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement  

SciTech Connect (OSTI)

In October 2002, DOE issued the Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement (Final EIS) (DOE 2002) that provided an analysis of the potential environmental consequences of alternatives/options for the management and disposition of Sodium Bearing Waste (SBW), High-Level Waste (HL W) calcine, and HLW facilities at the Idaho Nuclear Technology and Engineering Center (INTEC) located at the Idaho National Engineering and Environmental Laboratory (INEEL), now known as the Idaho National Laboratory (INL) and referred to hereafter as the Idaho Site. Subsequent to the issuance of the Final EIS, DOE included the requirement for treatment of SBW in the Request for Proposals for Environmental Management activities on the Idaho Site. The new Idaho Cleanup Project (ICP) Contractor identified Steam Reforming as their proposed method to treat SBW; a method analyzed in the Final EIS as an option to treat SBW. The proposed Steam Reforming process for SBW is the same as in the Final EIS for retrieval, treatment process, waste form and transportation for disposal. In addition, DOE has updated the characterization data for both the HLW Calcine (BBWI 2005a) and SBW (BBWI 2004 and BBWI 2005b) and identified two areas where new calculation methods are being used to determine health and safety impacts. Because of those changes, DOE has prepared this supplement analysis to determine whether there are ''substantial changes in the proposed action that are relevant to environmental concerns'' or ''significant new circumstances or information'' within the meaning of the Council of Environmental Quality and DOE National Environmental Policy Act (NEPA) Regulations (40 CFR 1502.9 (c) and 10 CFR 1021.314) that would require preparation of a Supplemental EIS. Specifically, this analysis is intended to determine if: (1) the Steam Reforming Option identified in the Final EIS adequately bounds impacts from the Steam Reforming Process proposed by the new ICP Contractor using the new characterization data, (2) the new characterization data is significantly different than the data presented in the Final EIS, (3) the new calculation methods present a significant change to the impacts described in the Final EIS, and (4) would the updated characterization data cause significant changes in the environmental impacts for the action alternatives/options presented in the Final EIS. There are no other aspects of the Final EIS that require additional review because DOE has not identified any additional new significant circumstances or information that would warrant such a review.

N /A

2005-06-30T23:59:59.000Z

79

ANALYSIS OF DAMAGE TO WASTE PACKAGES CAUSED BY SEISMIC EVENTS DURING POST-CLOSURE  

SciTech Connect (OSTI)

This paper presents methodology and results of an analysis of damage due to seismic ground motion for waste packages emplaced in a nuclear waste repository at Yucca Mountain, Nevada. A series of three-dimensional rigid body kinematic simulations of waste packages, pallets, and drip shields subjected to seismic ground motions was performed. The simulations included strings of several waste packages and were used to characterize the number, location, and velocity of impacts that occur during seismic ground motion. Impacts were categorized as either waste package-to-waste package (WP-WP) or waste package-to-pallet (WP-P). In addition, a series of simulations was performed for WP-WP and WP-P impacts using a detailed representation of a single waste package. The detailed simulations were used to determine the amount of damage from individual impacts, and to form a damage catalog, indexed according to the type, angle, location and force/velocity of the impact. Finally, the results from the two analyses were combined to estimate the total damage to a waste package that may occur during an episode of seismic ground motion. This study addressed two waste package types, four levels of peak ground velocity (PGV), and 17 ground motions at each PGV. Selected aspects of waste package degradation, such as effective wall thickness and condition of the internals, were also considered. As expected, increasing the PGV level of the vibratory ground motion increases the damage to the waste packages. Results show that most of the damage is caused by WP-P impacts. TAD-bearing waste packages with intact internals are highly resistant to damage, even at a PGV of 4.07 m/s, which is the highest level analyzed.

Alves, S W; Blair, S C; Carlson, S R; Gerhard, M; Buscheck, T A

2008-05-27T23:59:59.000Z

80

Gamma and Neutron Radiolysis in the 21-PWR Waste Package  

SciTech Connect (OSTI)

The objective of this calculation is to compute gamma and neutron dose rates in order to determine the maximum radiolytic production of nitric acid and other chemical species inside the 21-PWR (pressurized-water reactor) waste package (WP). The scope of this calculation is limited to the time period between 5,000 and 100,000 years after emplacement. The information provided by the sketches attached to this calculation is that of the potential design for the type of WP considered in this calculation. The results of this calculation will be used to evaluate nitric acid corrosion of fuel cladding from radiolysis in the 21-PWR WP. This calculation was performed in accordance with the Technical Work Plan for: Waste Package Design Description for LA (Civilian Radioactive Waste Management System (CRWMS) Management and Operating Contractor (M&O) 2000a). AP-3.124, Calculations, is used to perform the calculation and develop the document. This calculation is associated with the total system performance assessment (TSPA) of which the spent fuel cladding integrity is to be evaluated.

J.S. Tang

2001-05-03T23:59:59.000Z

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81

End of FY10 report - used fuel disposition technical bases and lessons learned : legal and regulatory framework for high-level waste disposition in the United States.  

SciTech Connect (OSTI)

This report examines the current policy, legal, and regulatory framework pertaining to used nuclear fuel and high level waste management in the United States. The goal is to identify potential changes that if made could add flexibility and possibly improve the chances of successfully implementing technical aspects of a nuclear waste policy. Experience suggests that the regulatory framework should be established prior to initiating future repository development. Concerning specifics of the regulatory framework, reasonable expectation as the standard of proof was successfully implemented and could be retained in the future; yet, the current classification system for radioactive waste, including hazardous constituents, warrants reexamination. Whether or not consideration of multiple sites are considered simultaneously in the future, inclusion of mechanisms such as deliberate use of performance assessment to manage site characterization would be wise. Because of experience gained here and abroad, diversity of geologic media is not particularly necessary as a criterion in site selection guidelines for multiple sites. Stepwise development of the repository program that includes flexibility also warrants serious consideration. Furthermore, integration of the waste management system from storage, transportation, and disposition, should be examined and would be facilitated by integration of the legal and regulatory framework. Finally, in order to enhance acceptability of future repository development, the national policy should be cognizant of those policy and technical attributes that enhance initial acceptance, and those policy and technical attributes that maintain and broaden credibility.

Weiner, Ruth F.; Blink, James A. (Lawrence Livermore National Laboratory, Livermore, CA); Rechard, Robert Paul; Perry, Frank (Los Alamos National Laboratory, Los Alamos, NM); Jenkins-Smith, Hank C. (University of Oklahoma, Norman, OK); Carter, Joe (Savannah River Nuclear Solutions, Aiken, SC); Nutt, Mark (Argonne National Laboratory, Argonne, IL); Cotton, Tom (Complex Systems Group, Washington DC)

2010-09-01T23:59:59.000Z

82

Cleanup Verification Package for the 118-C-1, 105-C Solid Waste Burial Ground  

SciTech Connect (OSTI)

This cleanup verification package documents completion of remedial action for the 118-C-1, 105-C Solid Waste Burial Ground. This waste site was the primary burial ground for general wastes from the operation of the 105-C Reactor and received process tubes, aluminum fuel spacers, control rods, reactor hardware, spent nuclear fuel and soft wastes.

M. J. Appel and J. M. Capron

2007-07-25T23:59:59.000Z

83

Nuclear waste management technical support in the development of nuclear waste form criteria for the NRC. Task 1. Waste package overview  

SciTech Connect (OSTI)

In this report the current state of waste package development for high level waste, transuranic waste, and spent fuel in the US and abroad has been assessed. Specifically, reviewed are recent and on-going research on various waste forms, container materials and backfills and tentatively identified those which are likely to perform most satisfactorily in the repository environment. Radiation effects on the waste package components have been reviewed and the magnitude of these effects has been identified. Areas requiring further research have been identified. The important variables affecting radionuclide release from the waste package have been described and an evaluation of regulatory criteria for high level waste and spent fuel is presented. Finally, for spent fuel, high level, and TRU waste, components which could be used to construct a waste package having potential to meet NRC performance requirements have been described and identified.

Dayal, R.; Lee, B.S.; Wilke, R.J.; Swyler, K.J.; Soo, P.; Ahn, T.M.; McIntyre, N.S.; Veakis, E.

1982-02-01T23:59:59.000Z

84

Cleanup Verification Package for the 118-B-1, 105-B Solid Waste Burial Ground  

SciTech Connect (OSTI)

This cleanup verification package documents completion of remedial action, sampling activities, and compliance criteria for the 118-B-1, 105-B Solid Waste Burial Ground. This waste site was the primary burial ground for general wastes from the operation of the 105-B Reactor and P-10 Tritium Separation Project and also received waste from the 105-N Reactor. The burial ground received reactor hardware, process piping and tubing, fuel spacers, glassware, electrical components, tritium process wastes, soft wastes and other miscellaneous debris.

J. M. Capron

2008-01-21T23:59:59.000Z

85

Potential role of ABC-assisted repositories in U.S. plutonium and high-level waste disposition  

SciTech Connect (OSTI)

This paper characterizes the issues involving deep geologic disposal of LWR spent fuel rods, then presents results of an investigation to quantify the potential role of Accelerator-Based Conversion (ABC) in an integrated national nuclear materials and high level waste disposition strategy. The investigation used the deep geological repository envisioned for Yucca Mt., Nevada as a baseline and considered complementary roles for integrated ABC transmutation systems. The results indicate that although a U.S. geologic waste repository will continue to be required, waste partitioning and accelerator transmutation of plutonium, the minor actinides, and selected long-lived fission products can result in the following substantial benefits: plutonium burndown to near zero levels, a dramatic reduction of the long term hazard associated with geologic repositories, an ability to place several-fold more high level nuclear waste in a single repository, electricity sales to compensate for capital and operating costs.

Berwald, David; Favale, Anthony; Myers, Timothy; McDaniel, Jerry [Grumman Aerospace Corporation, Bethpage New York 11714 (United States); Bechtel Corporation, 50 Beal St., San Francisco, California 94105 (United States)

1995-09-15T23:59:59.000Z

86

Data Packages for the Hanford Immobilized Low Activity Tank Waste Performance Assessment 2001 Version [SEC 1 THRU 5  

SciTech Connect (OSTI)

Data package supporting the 2001 Immobilized Low-Activity Waste Performance Analysis. Geology, hydrology, geochemistry, facility, waste form, and dosimetry data based on recent investigation are provided. Verification and benchmarking packages for selected software codes are provided.

MANN, F.M.

2000-03-02T23:59:59.000Z

87

THE SUCCESSFUL UTILIZATION OF COMMERCIAL TREATMENT CAPABILITIES TO DISPOSITION HANFORD NO-PATH-FORWARD SUSPECT TRANSURANIC WASTES  

SciTech Connect (OSTI)

The U.S. Department of Energy (DOE) Richland Operations Office (RL) has adopted the 2015 Vision for Cleanup of the Hanford Site. The CH2M HILL Plateau Remediation Company's (CHPRC) Waste and Fuels Management Project (W&FMP) and their partners support this mission by providing centralized waste management services for the Hanford Site waste generating organizations. At the time of the CHPRC contract award (August 2008) slightly more than 9,000 cubic meters (m{sup 3}) of legacy waste was defined as ''no-path-forward waste.'' A significant portion of this waste (7,650 m{sup 3}) comprised wastes with up to 50 grams of special nuclear materials (SNM) in oversized packages recovered during retrieval operations and large glove boxes removed from Hanford's Plutonium Finishing Plant (PFP). Through a collaborative effort between the DOE, CHPRC, and Perma-Fix Environmental Services, Inc. (PESI), pathways for these problematic wastes were developed and are currently being implemented.

BLACKFORD LT; CATLOW RL; WEST LD; COLLINS MS; ROMINE LD; MOAK DJ

2012-01-30T23:59:59.000Z

88

Waste Form Release Data Package for the 2001 Immobilized Low-Activity Waste Performance Assessment  

SciTech Connect (OSTI)

This data package documents the experimentally derived input data on the representative waste glasses LAWABP1 and HLP-31 that will be used for simulations of the immobilized lowactivity waste disposal system with the Subsurface Transport Over Reactive Multiphases (STORM) code. The STORM code will be used to provide the near-field radionuclide release source term for a performance assessment to be issued in March of 2001. Documented in this data package are data related to 1) kinetic rate law parameters for glass dissolution, 2) alkali-H ion exchange rate, 3) chemical reaction network of secondary phases that form in accelerated weathering tests, and 4) thermodynamic equilibrium constants assigned to these secondary phases. The kinetic rate law and Na+-H+ ion exchange rate were determined from single-pass flow-through experiments. Pressurized unsaturated flow and vapor hydration experiments were used for accelerated weathering or aging of the glasses. The majority of the thermodynamic data were extracted from the thermodynamic database package shipped with the geochemical code EQ3/6. However, several secondary reaction products identified from laboratory tests with prototypical LAW glasses were not included in this database, nor are the thermodynamic data available in the open literature. One of these phases, herschelite, was determined to have a potentially significant impact on the release calculations and so a solubility product was estimated using a polymer structure model developed for zeolites. Although this data package is relatively complete, final selection of ILAW glass compositions has not been done by the waste treatment plant contractor. Consequently, revisions to this data package to address new ILAW glass formulations are to be regularly expected.

McGrail, B. Peter; Icenhower, Jonathan P.; Martin, Paul F.; Schaef, Herbert T.; O'Hara, Matthew J.; Rodriguez, Eugenio; Steele, Jackie L.

2001-02-01T23:59:59.000Z

89

Value Engineering Study for Closing Waste Packages Containing TAD Canisters  

SciTech Connect (OSTI)

The Office of Civilian Radioactive Waste Management announced their intention to have the commercial utilities package spent nuclear fuel in shielded, transportable, ageable, and disposable containers prior to shipment to the Yucca Mountain repository. This will change the conditions used as a basis for the design of the waste package closure system. The environment is now expected to be a low radiation, low contamination area. A value engineering study was completed to evaluate possible modifications to the existing closure system using the revised requirements. Four alternatives were identified and evaluated against a set of weighted criteria. The alternatives are (1) a radiation-hardened, remote automated system (the current baseline design); (2) a nonradiation-hardened, remote automated system (with personnel intervention if necessary); (3) a nonradiation-hardened, semi-automated system with personnel access for routine manual operations; and (4) a nonradiation-hardened, fully manual system with full-time personnel access. Based on the study, the recommended design is Alternative 2, a nonradiation-hardened, remote automated system. It is less expensive and less complex than the current baseline system, because nonradiation-hardened equipment can be used and some contamination control equipment is no longer needed. In addition, the inclusion of remote automation ensures throughput requirements are met, provides a more reliable process, and provides greater protection for employees from industrial accidents and radiation exposure than the semi-automated or manual systems. Other items addressed during the value engineering study as requested by OCRWM include a comparison to industry canister closure systems and corresponding lessons learned; consideration of closing a transportable, ageable, and disposable canister; and an estimate of the time required to perform a demonstration of the recommended closure system.

Colleen Shelton-Davis

2005-11-01T23:59:59.000Z

90

Determination of Radioisotope Content by Measurement of Waste Package Dose Rates - 13394  

SciTech Connect (OSTI)

The objective of this communication is to report the observed correlation between the calculated air kerma rates produced by radioactive waste drums containing untreated ion-exchange resin and activated charcoal slurries with the measured radiation field of each package. Air kerma rates at different distances from the drum surface were calculated with the activity concentrations previously determined by gamma spectrometry of waste samples and the estimated mass, volume and geometry of solid and liquid phases of each waste package. The water content of each waste drum varies widely between different packages. Results will allow determining the total activity of wastes and are intended to complete the previous steps taken to characterize the radioisotope content of wastes packages. (authors)

Souza, Daiane Cristini B.; Gimenes Tessaro, Ana Paula; Vicente, Roberto [Nuclear and Energy Research Institute Brazil, Radioactive Waste Management Department IPEN/GRR, Sao Paulo. SP. (Brazil)] [Nuclear and Energy Research Institute Brazil, Radioactive Waste Management Department IPEN/GRR, Sao Paulo. SP. (Brazil)

2013-07-01T23:59:59.000Z

91

Analysis of waste package environment as related to initiation and propagation of localized corrosion  

SciTech Connect (OSTI)

Models are presented for prediction of waste package lifetime for the proposed high-level nuclear waste repository located at Yucca Mountain, Nevada. The waste package environment is expected to range from saturated, highly-alkaline brine to bicarbonate waters of low ionic strength. Localized corrosion will be predicted by comparing predicted corrosion potentials of the container with critical potentials for pitting and crevice corrosion. The critical potentials will be estimated by a combined program of experiments and mechanistic modeling.

Walton, J.C.; Sridhar, N.; Cragnolino, G.; Nair, P. [Southwest Research Inst., San Antonio, TX (United States). Center for Nuclear Waste Regulatory Analyses

1993-12-31T23:59:59.000Z

92

Ceramic package fabrication for YMP nuclear waste disposal  

SciTech Connect (OSTI)

The purpose of this work is to develop alternate materials/design concepts to metal barriers for the Nevada Nuclear Waste Storage Investigations Project. There is some potential that site conditions may prove to be too aggressive for successful employment of the metal alloys under current consideration or that performance assessment models will predict metal container degradation rates that are inconsistent with the goal of substantially complete containment included in the NRC regulations. In the event that the anticipated lifetimes of metal containers are considered inadequate, alternate materials (i.e. ceramics or ceramic/metal composites) will be chosen due to superior corrosion resistance. This document was prepared using information taken from the open literature, conversations and correspondence with vendors, news releases and data presented at conferences to determine what form such a package might take. This discussion presents some ceramic material selection criteria, alternatives for the materials which might be used and alternatives for potential fabrication routes. This includes {open_quotes}stand alone{close_quotes} ceramic components and ceramic coatings/linings for metallic structures. A list of companies providing verbal or written information concerning the production of ceramic or ceramic lined waste containers appears at the end of this discussion.

Wilfinger, K.

1994-08-01T23:59:59.000Z

93

Aging and Phase Stability of Waste Package Outer Barrier  

SciTech Connect (OSTI)

This report was prepared in accordance with ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). This report provides information on the phase stability of Alloy 22, the current waste package outer barrier material. The goal of this model is to determine whether the single-phase solid solution is stable under repository conditions and, if not, how fast other phases may precipitate. The aging and phase stability model, which is based on fundamental thermodynamic and kinetic concepts and principles, will be used to provide predictive insight into the long-term metallurgical stability of Alloy 22 under relevant repository conditions. The results of this model are used by ''General Corrosion and Localized Corrosion of Waste Package Outer Barrier'' as reference-only information. These phase stability studies are currently divided into three general areas: Tetrahedrally close-packed (TCP) phase and carbide precipitation in the base metal; TCP and carbide precipitation in welded samples; and Long-range ordering reactions. TCP-phase and carbide precipitates that form in Alloy 22 are generally rich in chromium (Cr) and/or molybdenum (Mo) (Raghavan et al. 1984 [DIRS 154707]). Because these elements are responsible for the high corrosion resistance of Alloy 22, precipitation of TCP phases and carbides, especially at grain boundaries, can lead to an increased susceptibility to localized corrosion in the alloy. These phases are brittle and also tend to embrittle the alloy (Summers et al. 1999 [DIRS 146915]). They are known to form in Alloy 22 at temperatures greater than approximately 600 C. Whether these phases also form at the lower temperatures expected in the repository during the 10,000-year regulatory period must be determined. The kinetics of this precipitation will be determined for both the base metal and the weld heat-affected zone (HAZ). The TCP phases (P, {mu}, and {sigma}) are present in the weld metal in the as-welded condition. It may be possible to eliminate these phases through a solution anneal heat treatment, but that may not be possible for the closure weld because the spent nuclear fuel cladding cannot be heated to more than 350 C. The effects of any stress mitigation techniques (such as laser peening or solution heat treating) that may be used to reduce the tensile stresses on the closure welds must also be determined. Cold-work will cause an increase in dislocation density, and such an increase in dislocation density may cause an increase in diffusion rates that control precipitation kinetics (Porter et al. 1992 [DIRS 161265]; Tawancy et al. 1983 [DIRS 104991]). Long-range order (LRO) occurs in nickel (Ni)-Cr-Mo alloys (such as Alloy 22) at temperatures less than approximately 600 C. This ordering has been linked to an increased susceptibility of Ni-Cr-Mo alloys to stress corrosion cracking and hydrogen embrittlement (Tawancy et al. 1983 [DIRS 104991]). These analyses provide information on the rate at which LRO may occur in Alloy 22 under repository conditions. Determination of the kinetics of transformations through experimental techniques requires that the transformations being investigated be accelerated due to the fact that the expected service life is at least 10,000 years. Phase transformations are typically accelerated through an increase in temperature. The rate of transformation is determined at the higher temperature and is extrapolated to the lower temperatures of interest.

F. Wong

2004-09-28T23:59:59.000Z

94

FABRICATION AND DEPLOYMENT OF THE 9979 TYPE AF RADIOACTIVE WASTE PACKAGING FOR THE DEPARTMENT OF ENERGY  

SciTech Connect (OSTI)

This paper summarizes the development, testing, and certification of the 9979 Type A Fissile Packaging that replaces the UN1A2 Specification Shipping Package eliminated from Department of Transportation (DOT) 49 CFR 173. The DOT Specification Package was used for many decades by the U.S. nuclear industry as a fissile waste container until its removal as an authorized container by DOT. This paper will discuss stream lining procurement of high volume radioactive material packaging manufacturing, such as the 9979, to minimize packaging production costs without sacrificing Quality Assurance. The authorized content envelope (combustible and non-combustible) as well as planned content envelope expansion will be discussed.

Blanton, P.; Eberl, K.

2013-10-10T23:59:59.000Z

95

AXIAL SOURCE PROFILE EFFECT ON WASTE PACKAGE TRANSPORTER SHIELDING  

SciTech Connect (OSTI)

The purpose of this scoping calculation is to support preliminary design of the Waste Package (WP) transporter radiation shield configuration. Spent Nuclear Fuel (SNF) is highly radioactive and site personnel must be protected during the period that the WPs are emplaced. Personnel protection is accomplished via a heavily shielded WP transporter that moves the waste from the surface to the emplacement drift. All previous WP transporter shielding calculations have assumed a Design Basis Fuel (DBF) in which the fuel burnup is uniform (e.g. Ref. 7.3, Ref. 7.4, and Ref. 7.12). In reality, SNF burnup varies significantly from one end of the fuel assembly to the other. Since source strengths are dependent upon fuel burnup, a model which varies the fuel burnup along the assembly axis will produce a more accurate depiction of the radiation field surrounding the WP transporter. The objective of this calculation is to determine the need for using the actual axial profile, as opposed to the uniform burnup assumption, in the WP transporter shield design. The scope of the calculation is as follows: (1) Determine the impact of axial source term variation on WP transporter contact dose rates. (2) Determine appropriate shielding modifications to account for expected dose rate peaking effects. Consistent with the previous subsurface shielding analyses, this calculation considers the bounding 21 Pressurized Water Reactor (PWR) WP only. The calculation will need to be revised and extended to Boiling Water Reactor (BWR) SNF upon selection of the WP design for the License Application (LA) and availability of the source terms from the WP Operations Group.

A. Nielsen

1999-06-30T23:59:59.000Z

96

Nondestructive assay and nondestructive examination of remote-handled transuranic waste at the ORNL waste handling and packaging plant  

SciTech Connect (OSTI)

The purpose of this investigation is to examine the use of an electron linear accelerator (LINAC) in the performance of nondestructive assay (NDA) and nondestructive examination (NDE) measurements of remote-handled transuranic wastes. The system will be used to perform waste characterization and certification activities at the Oak Ridge National Laboratory's proposed Waste Handling and Packaging Plant. The NDA and NDE technologies which were developed for contact-handled wastes are inadequate to perform such measurements on high gamma and neutron dose-rate wastes. A single LINAC will provide the interrogating fluxes required for both NDA and NDE measurements of the wastes. 11 refs., 6 figs.

Schultz, F.J.; Caldwell, J.T. (Oak Ridge National Lab., TN (USA); Pajarito Scientific Corp. (USA))

1989-01-01T23:59:59.000Z

97

Operations to be Performed in the Waste Package Dry Remediation Cell  

SciTech Connect (OSTI)

Describes planned and proposed operations for remediating damaged and/or out-of-compliance waste packages, casks, DPCs, overpacks, and containers at the Yucca Mountain Dry Transfer Facility.

Norman E. Cole; Randy K. Elwood

2003-10-01T23:59:59.000Z

98

Yucca Mountain Waste Package Closure System Robotic Welding and Inspection System  

SciTech Connect (OSTI)

The Waste Package Closure System (WPCS), for the closure of radioactive waste in canisters for permanent storage of spent nuclear fuel (SNF) and high-level waste in the Yucca Mountain Repository was designed, fabricated, and successfully demonstrated at the Idaho National Laboratory (INL). This article focuses on the robotic hardware and tools necessary to remotely weld and inspect the closure lid welds. The system was operated remotely and designed for use in a radiation field, due to the SNF contained in the waste packages being closed.

C. I. Nichol; D. P. Pace; E. D. Larsen; T. R. McJunkin; D. E. Clark; M. L. Clark; K. L. Skinner; A. D. Watkins; H. B. Smartt

2011-10-01T23:59:59.000Z

99

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

SciTech Connect (OSTI)

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

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

1994-09-01T23:59:59.000Z

100

DISPOSITION PATHS FOR ROCKY FLATS GLOVEBOXES: EVALUATING OPTIONS  

SciTech Connect (OSTI)

The Kaiser-Hill Company, LLC has the responsibility for closure activities at the Rocky Flats Environmental Technology Site (RFETS). One of the challenges faced for closure is the disposition of radiologically contaminated gloveboxes. Evaluation of the disposition options for gloveboxes included a detailed analysis of available treatment capabilities, disposal facilities, and lifecycle costs. The Kaiser-Hill Company, LLC followed several processes in determining how the gloveboxes would be managed for disposition. Currently, multiple disposition paths have been chosen to accommodate the needs of the varying styles and conditions of the gloveboxes, meet the needs of the decommissioning team, and to best manage lifecycle costs. Several challenges associated with developing a disposition path that addresses both the radiological and RCRA concerns as well as offering the most cost-effective solution were encountered. These challenges included meeting the radiological waste acceptance criteria of available disposal facilities, making a RCRA determination, evaluating treatment options and costs, addressing void requirements associated with disposal, and identifying packaging and transportation options. The varying disposal facility requirements affected disposition choices. Facility conditions that impacted decisions included radiological and chemical waste acceptance criteria, physical requirements, and measurement for payment options. The facility requirements also impacted onsite activities including management strategies, decontamination activities, and life-cycle cost.

Lobdell, D.; Geimer, R.; Larsen, P.; Loveland, K.

2003-02-27T23:59:59.000Z

Note: This page contains sample records for the topic "waste disposition packaging" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


101

Conceptual waste package interim product specifications and data requirements for disposal of borosilicate glass defense high-level waste forms in salt geologic repositories  

SciTech Connect (OSTI)

The conceptual waste package interim product specifications and data requirements presented are applicable specifically to the normal borosilicate glass product of the Defense Waste Processing Facility (DWPF). They provide preliminary numerical values for the defense high-level waste form parameters and properties identified in the waste form performance specification for geologic isolation in salt repositories. Subject areas treated include containment and isolation, operational period safety, criticality control, waste form/production canister identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available.

Not Available

1983-06-01T23:59:59.000Z

102

A Fruit of Yucca Mountain: The Remote Waste Package Closure System  

SciTech Connect (OSTI)

Was the death of the Yucca Mountain repository the fate of a technical lemon or a political lemon? Without caution, this debate could lure us away from capitalizing on the fruits of the project. In March 2009, Idaho National Laboratory (INL) successfully demonstrated the Waste Package Closure System, a full-scale prototype system for closing waste packages that were to be entombed in the now abandoned Yucca Mountain repository. This article describes the system, which INL designed and built, to weld the closure lids on the waste packages, nondestructively examine the welds using four different techniques, repair the welds if necessary, mitigate crack initiating stresses in the surfaces of the welds, evacuate and backfill the packages with an inert gas, and perform all of these tasks remotely. As a nation, we now have a proven method for securely sealing nuclear waste packages for long term storage—regardless of whether or not the future destination for these packages will be an underground repository. Additionally, many of the system’s features and concepts may benefit other remote nuclear applications.

Kevin Skinner; Greg Housley; Colleen Shelton-Davis

2011-11-01T23:59:59.000Z

103

Session 35 - Panel: Remaining US Disposition Issues for Orphan or Small Volume Low Level and Low Level Mixed Waste Streams  

SciTech Connect (OSTI)

Faced with closure schedules as a driving force, significant progress has been made during the last 2 years on the disposition of DOE mixed waste streams thought previously to be problematic. Generators, the Department of Energy and commercial vendors have combined to develop unique disposition paths for former orphan streams. Recent successes and remaining issues will be discussed. The session will also provide an opportunity for Federal agencies to share lessons learned on low- level and mixed low-level waste challenges and identify opportunities for future collaboration. This panel discussion was organized by PAC member Dick Blauvelt, Navarro Research and Engineering Inc who served as co-chair along with Dave Eaton from INL. In addition, George Antonucci, Duratek Barnwell and Rich Conley, AFSC were invited members of the audience, prepared to contribute the Barnwell and DOD perspective to the issues as needed. Mr. Small provide information regarding the five year 20K M3 window of opportunity at the Nevada Test Site for DOE contractors to dispose of mixed waste that cannot be received at the Energy Solutions (Envirocare) site in Utah because of activity levels. He provided a summary of the waste acceptance criteria and the process sites must follow to be certified to ship. When the volume limit or time limit is met, the site will undergo a RCRA closure. Ms. Gelles summarized the status of the orphan issues, commercial options and the impact of the EM reorganization on her program. She also announced that there would be a follow-on meeting in 2006 to the very successful St. Louis meeting of last year. It will probably take place in Chicago in July. Details to be announced. Mr. McKenney discussed progress made at the Hanford Reservation regarding disposal of their mixed waste inventory. The news is good for the Hanford site but not good for the rest of the DOE complex since shipment for out of state of both low level and low level mixed waste will continue to be prohibited until the completion of a new NEPA study. This is anticipated to take several years. Bill Franz from Portsmouth and Dave Eaton representing the INL provided the audience with information regarding some of the problematic mixed waste streams at their respective sites. Portsmouth has some unique radiological issues with isotopes such as Tc-99 while the INL is trying to deal with mixed waste in the 10-100 nCi/g range. Kaylin Loveland spoke of the new,Energy Solutions organization and provided information on mixed waste treatment capabilities at the Clive site. Mike Lauer described the licensing activities at the WCS site in Texas where they are trying to eventually have disposal capabilities for Class A, B and C mixed waste from both DOE and the commercial sector. The audience included about 75 WM'06 attendees who asked some excellent questions and provided an active and informative exchange of information on the topic. (authors)

Blauvelt, Richard [Navarro Engineering Research Inc. (United States); Small, Ken [Doe Nevada (United States); Gelles, Christine [DOE EM HQ (United States); McKenney, Dale [Fluor Hanford (United States); Franz, Bill [LATA Portsmouth (United States); Loveland, Kaylin [Energy Solutions Inc. (United States); Lauer, Mike [Waste Control Specialists (United States)

2006-07-01T23:59:59.000Z

104

Packaging waste recycling in Europe: Is the industry paying for it?  

SciTech Connect (OSTI)

Highlights: • We study the recycling schemes of France, Germany, Portugal, Romania and the UK. • The costs and benefits of recycling are compared for France, Portugal and Romania. • The balance of costs and benefits depend on the perspective (strictly financial/economic). • Financial supports to local authorities ought to promote cost-efficiency. - Abstract: This paper describes and examines the schemes established in five EU countries for the recycling of packaging waste. The changes in packaging waste management were mainly implemented since the Directive 94/62/EC on packaging and packaging waste entered into force. The analysis of the five systems allowed the authors to identify very different approaches to cope with the same problem: meet the recovery and recycling targets imposed by EU law. Packaging waste is a responsibility of the industry. However, local governments are generally in charge of waste management, particularly in countries with Green Dot schemes or similar extended producer responsibility systems. This leads to the need of establishing a system of financial transfers between the industry and the local governments (particularly regarding the extra costs involved with selective collection and sorting). Using the same methodological approach, the authors also compare the costs and benefits of recycling from the perspective of local public authorities for France, Portugal and Romania. Since the purpose of the current paper is to take note of who is paying for the incremental costs of recycling and whether the industry (i.e. the consumer) is paying for the net financial costs of packaging waste management, environmental impacts are not included in the analysis. The work carried out in this paper highlights some aspects that are prone to be improved and raises several questions that will require further research. In the three countries analyzed more closely in this paper the industry is not paying the net financial cost of packaging waste management. In fact, if the savings attained by diverting packaging waste from other treatment (e.g. landfilling) and the public subsidies to the investment on the “recycling system” are not considered, it seems that the industry should increase the financial support to local authorities (by 125% in France, 50% in Portugal and 170% in Romania). However, in France and Portugal the industry is paying local authorities more than just the incremental costs of recycling (full costs of selective collection and sorting minus the avoided costs). To provide a more definitive judgment on the fairness of the systems it will be necessary to assess the cost efficiency of waste management operators (and judge whether operators are claiming costs or eliciting “prices”)

Ferreira da Cruz, Nuno, E-mail: nunocruz@ist.utl.pt; Ferreira, Sandra; Cabral, Marta; Simões, Pedro; Marques, Rui Cunha

2014-02-15T23:59:59.000Z

105

Vertical Drop of the Naval SNF Long Waste Package On Unyielding Surface  

SciTech Connect (OSTI)

The purpose of this calculation is to determine the structural response of a Naval SNF (Spent Nuclear Fuel) Long Waste Package (WP) subjected to 2 m-vertical drop on unyielding surface (US). The scope of this document is limited to reporting the calculation results in terms of maximum stress intensities. This calculation is associated with the waste package design; calculation is performed by the Waste Package Design group. AP-3.12Q, Revision 0, ICN 0, Calculations, is used to perform the calculation and develop the document. The finite element calculation is performed by using the commercially available ANSYS Version (V) 5.4 finite element code. The result of this calculation is provided in terms of maximum stress intensities.

S. Mastilovic

2006-07-21T23:59:59.000Z

106

Survey of waste package designs for disposal of high-level waste/spent fuel in selected foreign countries  

SciTech Connect (OSTI)

This report presents the results of a survey of the waste package strategies for seven western countries with active nuclear power programs that are pursuing disposal of spent nuclear fuel or high-level wastes in deep geologic rock formations. Information, current as of January 1989, is given on the leading waste package concepts for Belgium, Canada, France, Federal Republic of Germany, Sweden, Switzerland, and the United Kingdom. All but two of the countries surveyed (France and the UK) have developed design concepts for their repositories, but none of the countries has developed its final waste repository or package concept. Waste package concepts are under study in all the countries surveyed, except the UK. Most of the countries have not yet developed a reference concept and are considering several concepts. Most of the information presented in this report is for the current reference or leading concepts. All canisters for the wastes are cylindrical, and are made of metal (stainless steel, mild steel, titanium, or copper). The canister concepts have relatively thin walls, except those for spent fuel in Sweden and Germany. Diagrams are presented for the reference or leading concepts for canisters for the countries surveyed. The expected lifetimes of the conceptual canisters in their respective disposal environment are typically 500 to 1,000 years, with Sweden's copper canister expected to last as long as one million years. Overpack containers that would contain the canisters are being considered in some of the countries. All of the countries surveyed, except one (Germany) are currently planning to utilize a buffer material (typically bentonite) surrounding the disposal package in the repository. Most of the countries surveyed plan to limit the maximum temperature in the buffer material to about 100{degree}C. 52 refs., 9 figs.

Schneider, K.J.; Lakey, L.T.; Silviera, D.J.

1989-09-01T23:59:59.000Z

107

Scale-up considerations relevant to experimental studies of nuclear waste-package behavior  

SciTech Connect (OSTI)

Results from a study that investigated whether testing large-scale nuclear waste-package assemblages was technically warranted are reported. It was recognized that the majority of the investigations for predicting waste-package performance to date have relied primarily on laboratory-scale experimentation. However, methods for the successful extrapolation of the results from such experiments, both geometrically and over time, to actual repository conditions have not been well defined. Because a well-developed scaling technology exists in the chemical-engineering discipline, it was presupposed that much of this technology could be applicable to the prediction of waste-package performance. A review of existing literature documented numerous examples where a consideration of scaling technology was important. It was concluded that much of the existing scale-up technology is applicable to the prediction of waste-package performance for both size and time extrapolations and that conducting scale-up studies may be technically merited. However, the applicability for investigating the complex chemical interactions needs further development. It was recognized that the complexity of the system, and the long time periods involved, renders a completely theoretical approach to performance prediction almost hopeless. However, a theoretical and experimental study was defined for investigating heat and fluid flow. It was concluded that conducting scale-up modeling and experimentation for waste-package performance predictions is possible using existing technology. A sequential series of scaling studies, both theoretical and experimental, will be required to formulate size and time extrapolations of waste-package performance.

Coles, D.G.; Peters, R.D.

1986-04-01T23:59:59.000Z

108

DOE/EIS-0287 Idaho High-Level Waste & Facilities Disposition...  

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

-5 . No offsite transportation would occur. At INEEL - The estimated number of latent cancer fatalities in the population within 50 miles of INTEC related to waste processing...

109

Calculation of the Naval Long and Short Waste Package Three-Dimensional Thermal Interface Temperatures  

SciTech Connect (OSTI)

The purpose of this calculation is to evaluate the thermal performance of the Naval Long and Naval Short spent nuclear fuel (SNF) waste packages (WP) in the repository emplacement drift. The scope of this calculation is limited to the determination of the temperature profiles upon the surfaces of the Naval Long and Short SNF waste package for up to 10,000 years of emplacement. The temperatures on the top of the outside surface of the naval canister are the thermal interfaces for the Naval Nuclear Propulsion Program (NNPP). The results of this calculation are intended to support Licensing Application design activities.

H. Marr

2006-10-25T23:59:59.000Z

110

SECOND WASTE PACKAGE PROBABILISTIC CRITICALITY ANALYSIS: GENERATION AND EVALUATION OF INTERNAL CRITICIALITY CONFIGURATIONS  

SciTech Connect (OSTI)

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development (WPD) department to provide an evaluation of the criticality potential within a waste package having sonic or all of its contents degraded by corrosion and removal of neutron absorbers. This analysis is also intended to provide an estimate of the consequences of any internal criticality, particularly in terms of any increase in radionuclide inventory. These consequence estimates will be used as part of the WPD input to the Total System Performance Assessment. The ultimate objective of this analysis is to augment the information gained from the Initial Waste Package Probabilistic Criticality Analyses (Ref. 5.8 and 5.9, hereafter referred to as IPA) to a degree which will support preliminary waste package design recommendations intended to reduce the risk of waste package criticality and the risk to total repository system performance posed by the consequences of any criticality. The IPA evaluated the criticality potential under the assumption that the waste package basket retained its structural integrity, so that the assemblies retained their initial separation, even when the neutron absorbers had been leached from the basket. This analysis is based on the more realistic condition that removal of the neutron absorbers is a consequence of the corrosion of the steel in which they are contained, which has the additional consequence of reducing the structural support between assemblies. The result is a set of more reactive configurations having a smaller spacing between assemblies, or no inter-assembly spacing at all. Another difference from the IPA is the minimal attention to probabilistic evaluation given in this study. Although the IPA covered a time horizon to 100,000 years, the lack of consideration of basket degradation modes made it primarily applicable to the first 10,000 years. In contrast, this study, by focusing on the degraded modes of the basket, is primarily applicable to the post 10,000 year time horizon. The principal probability related objective of this analysis is to estimate the earliest time at which a criticality can possibly occur, as a function of several environmental and material performance parameters, the specific values of which are subject to great uncertainty. It should be noted that the longer the possibility of criticality is delayed, the less will be the probability of any criticality at all, because of the increased probability that the waste package bottom will be penetrated by corrosion so that the waste package can no longer hold enough water to provide the necessary moderation to support the criticality.

P. Gottlieb, J.R. Massari, J.K. McCoy

1996-03-27T23:59:59.000Z

111

Potential vertical movement of large heat-generating waste packages in salt.  

SciTech Connect (OSTI)

With renewed interest in disposal of heat-generating waste in bedded or domal salt formations, scoping analyses were conducted to estimate rates of waste package vertical movement. Vertical movement is found to result from thermal expansion, from upward creep or heave of the near-field salt, and from downward buoyant forces on the waste package. A two-pronged analysis approach was used, with thermal-mechanical creep modeling, and coupled thermal-viscous flow modeling. The thermal-mechanical approach used well-studied salt constitutive models, while the thermal-viscous approach represented the salt as a highly viscous fluid. The Sierra suite of coupled simulation codes was used for both approaches. The waste package in all simulations was a right-circular cylinder with the density of steel, in horizontal orientation. A time-decaying heat generation function was used to represent commercial spent fuel with typical burnup and 50-year age. Results from the thermal-mechanical base case showed approximately 27 cm initial uplift of the package, followed by gradual relaxation closely following the calculated temperature history. A similar displacement history was obtained with the package density set equal to that of salt. The slight difference in these runs is attributable to buoyant displacement (sinking) and is on the order of 1 mm in 2,000 years. Without heat generation the displacement stabilizes at a fraction of millimeter after a few hundred years. Results from thermal-viscous model were similar, except that the rate of sinking was constant after cooldown, at approximately 0.15 mm per 1,000 yr. In summary, all calculations showed vertical movement on the order of 1 mm or less in 2,000 yr, including calculations using well-established constitutive models for temperature-dependent salt deformation. Based on this finding, displacement of waste packages in a salt repository is not a significant repository performance issue.

Clayton, Daniel James; Martinez, Mario J.; Hardin, Ernest L.

2013-05-01T23:59:59.000Z

112

Review of waste package verification tests. Semiannual report, October 1983-March 1984. Volume 4  

SciTech Connect (OSTI)

The current study is part of an ongoing task to specify tests that may be used to verify that engineered waste package/repository systems comply with NRC radionuclide containment and controlled release performance objectives. Work covered in this report includes crushed tuff packing material for use in a high-level waste tuff repository. Ranges of repository conditions relevant to its testing and other factors important for its performance are discussed. 23 refs., 5 figs., 3 tabs.

Jain, H.; Veakis, E.; Soo, P.

1985-06-01T23:59:59.000Z

113

Annotated bibliography for the design of waste packages for geologic disposal of spent fuel and high-level waste  

SciTech Connect (OSTI)

This bibliography identifies documents that are pertinent to the design of waste packages for geologic disposal of nuclear waste. The bibliography is divided into fourteen subject categories so that anyone wishing to review the subject of leaching, for example, can turn to the leaching section and review the abstracts of reports which are concerned primarily with leaching. Abstracts are also cross referenced according to secondary subject matter so that one can get a complete list of abstracts for any of the fourteen subject categories. All documents which by their title alone appear to deal with the design of waste packages for the geologic disposal of spent fuel or high-level waste were obtained and reviewed. Only those documents which truly appear to be of interest to a waste package designer were abstracted. The documents not abstracted are listed in a separate section. There was no beginning date for consideration of a document for review. About 1100 documents were reviewed and about 450 documents were abstracted.

Wurm, K.J.; Miller, N.E.

1982-11-01T23:59:59.000Z

114

Evaluation and compilation of DOE waste package test data: Biannual report, August 1986-January 1987  

SciTech Connect (OSTI)

This report summarizes results of the National Bureau of Standards (NBS) evaluations of Department of Energy (DOE) activities on waste packages designed for containment of radioactive high-level nuclear waste (HLW). The waste package is a proposed engineered barrier that is part of a permanent repository for HLW. Metal alloys are the principal barriers within the engineered system. Technical discussions are given for the corrosion of metals proposed for the canister, particularly carbon and stainless steels, and copper. In the section on tuff, the current level of understanding of several canister materials is questioned. Within the Basalt Waste Isolation Project (BWIP) section, discussions are given on problems concerning groundwater, materials for use in the metallic overpack, and diffusion through the packing. For the proposed salt site, questions are raised on the work on both ASTM A216 Steel and Ti-Code 12. NBS work related to the vitrification of HLW borosilicate glass at the West Valley Demonstration Project (WVDP) and the Defense Waste Processing Facility (DWPF) is covered. NBS reviews of selected DOE technical reports and a summary of current waste-package activities of the Materials Characterization Center (MCC) is presented. Using a database management system, a computerized database for storage and retrieval of reviews and evaluations of HLW data has been developed and is described. 17 refs., 2 figs., 2 tabs.

Interrante, C.; Escalante, E.; Fraker, A.; Harrison, S.; Shull, R.; Linzer, M.; Ricker, R.; Ruspi, J.

1987-10-01T23:59:59.000Z

115

Public Comment Period for Portsmouth Site D&D and Waste Disposition...  

Office of Environmental Management (EM)

2014 8:00AM EST to March 11, 2015 5:00PM EDT Public Comment Period for the Process Buildings and Complex Facilities Decontamination and Decommissioning and Site-Wide Waste...

116

Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment  

SciTech Connect (OSTI)

This data package documents the experimentally derived input data on the representative waste glasses; LAWA44, LAWB45, and LAWC22. This data will be used for Subsurface Transport Over Reactive Multi-phases (STORM) simulations of the Integrated Disposal Facility (IDF) for immobilized low-activity waste (ILAW). The STORM code will be used to provide the near-field radionuclide release source term for a performance assessment to be issued in July 2005. Documented in this data package are data related to 1) kinetic rate law parameters for glass dissolution, 2) alkali (Na+)-hydrogen (H+) ion exchange rate, 3) chemical reaction network of secondary phases that form in accelerated weathering tests, and 4) thermodynamic equilibrium constants assigned to these secondary phases. The kinetic rate law and Na+-H+ ion exchange rate were determined from single-pass flow-through experiments. Pressurized unsaturated flow (PUF) and product consistency (PCT) tests where used for accelerated weathering or aging of the glasses in order to determine a chemical reaction network of secondary phases that form. The majority of the thermodynamic data used in this data package were extracted from the thermody-namic database package shipped with the geochemical code EQ3/6, version 8.0. Because of the expected importance of 129I release from secondary waste streams being sent to IDF from various thermal treatment processes, parameter estimates for diffusional release and solubility-controlled release from cementitious waste forms were estimated from the available literature.

Pierce, Eric M.; McGrail, B. Peter; Rodriguez, Elsa A.; Schaef, Herbert T.; Saripalli, Prasad; Serne, R. Jeffrey; Krupka, Kenneth M.; Martin, P. F.; Baum, Steven R.; Geiszler, Keith N.; Reed, Lunde R.; Shaw, Wendy J.

2004-09-01T23:59:59.000Z

117

Waste package performance assessment: Deterministic system model, program scope and specification  

SciTech Connect (OSTI)

Integrated assessments of the performance of nuclear waste package designs must be made in order to qualify waste package designs with respect to containment time and release-rate requirements. PANDORA is a computer-based model of the waste package and of the processes affecting it over the long terms, specific to conditions at the proposed Yucca Mountain, Nevada, site. The processes PANDORA models include: changes in inventories due to radioactive decay, gamma radiation dose rate in and near the package, heat transfer, mechanical behavior, groundwater contact, corrosion, waste form alteration, and radionuclide release. The model tracks the development and coupling of these processes over time. The process models are simplified ones that focus on major effects and on coupling. This report documents our conceptual model development and provides a specification for the computer program. The current model is the first in a series. Succeeding models will use guidance from results of preceding models in the PANDORA series and will incorporate results of recently completed experiments and calculations on processes affecting performance. 22 refs., 21 figs., 9 tabs.

O`Connell, W.J.; Drach, R.S.

1986-10-02T23:59:59.000Z

118

Effect of aged waste package and basalt on radioelement release  

SciTech Connect (OSTI)

Results of experiments are described that combine backfill, radioactive waste, and repository host rock in a single flowing groundwater stream in a manner analogous to a hydraulic breach of a waste repository. The experimental design is used to identify the chemical interactions that would occur if repository components were breached by flowing water. The results indicate that of three parameters studied, the alteration of the repository components as might occur upon aging had the most substantial influence on the migration of radioactive elements dissolved from the solid radioactive waste. The other two parameters, the metal alloy used in the apparatus and an ionizing radiation field imposed on the experimental apparatus, had little or no measurable effect on radioactive element transport by flowing water. Inasmuch as the alteration of the repository materials represent aging in an actual repository, it is concluded that changes with age may detrimentally affect the ability of a repository to isolate plutonium and neptunium, and possibly other radioactive elements in nuclear waste. 37 references, 2 figures, 2 tables.

Seitz, M.G.; Vandegrift, G.F.; Bowers, D.L.; Gerding, T.J.

1983-01-01T23:59:59.000Z

119

Waste Generator Instructions: Key to Successful Implementation of the US DOE's 435.1 for Transuranic Waste Packaging Instructions (LA-UR-12-24155) - 13218  

SciTech Connect (OSTI)

In times of continuing fiscal constraints, a management and operation tool that is straightforward to implement, works as advertised, and virtually ensures compliant waste packaging should be carefully considered and employed wherever practicable. In the near future, the Department of Energy (DOE) will issue the first major update to DOE Order 435.1, Radioactive Waste Management. This update will contain a requirement for sites that do not have a Waste Isolation Pilot Plant (WIPP) waste certification program to use two newly developed technical standards: Contact-Handled Defense Transuranic Waste Packaging Instructions and Remote-Handled Defense Transuranic Waste Packaging Instructions. The technical standards are being developed from the DOE O 435.1 Notice, Contact-Handled and Remote-Handled Transuranic Waste Packaging, approved August 2011. The packaging instructions will provide detailed information and instruction for packaging almost every conceivable type of transuranic (TRU) waste for disposal at WIPP. While providing specificity, the packaging instructions leave to each site's own discretion the actual mechanics of how those Instructions will be functionally implemented at the floor level. While the Technical Standards are designed to provide precise information for compliant packaging, the density of the information in the packaging instructions necessitates a type of Rosetta Stone that translates the requirements into concise, clear, easy to use and operationally practical recipes that are waste stream and facility specific for use by both first line management and hands-on operations personnel. The Waste Generator Instructions provide the operator with step-by-step instructions that will integrate the sites' various operational requirements (e.g., health and safety limits, radiological limits or dose limits) and result in a WIPP certifiable waste and package that can be transported to and emplaced at WIPP. These little known but widely productive Waste Generator Instructions (WGIs) have been used occasionally in the past at large sites for treatment and packaging of TRU waste. The WGIs have resulted in highly efficient waste treatment, packaging and certification for disposal of TRU waste at WIPP. For example, a single WGI at LANL, combined with an increase in gram loading, resulted in a mind boggling 6,400% increase in waste loading for {sup 238}Pu heat source waste. In fact, the WGI combined with a new Contact Handled (CH) TRU Waste Content (TRUCON) Code provided a massive increase in shippable wattage per Transuranic Package Transporter-II (TRUPACT-II) over the previously used and more restrictive TRUCON Code that have been used previously for the heat source waste. In fact, the use of the WGI process at LANL's TA-55 facility reduced non-compliant drums for WIPP certification and disposal from a 13% failure rate down to a 0.5% failure rate and is expected to further reduce the failure rate to zero drums per year. The inherent value of the WGI is that it can be implemented in a site's current procedure issuance process and it provides documented proof of what actions were taken for each waste stream packaged. The WGI protocol provides a key floor-level operational component to achieve goal alignment between actual site operations, the WIPP TRU waste packaging instructions, and DOE O 435.1. (authors)

French, David M. [LANL EES-12, Carlsbad, NM, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)] [LANL EES-12, Carlsbad, NM, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Hayes, Timothy A. [LANL EES-12, Carlsbad, NM, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)] [LANL EES-12, Carlsbad, NM, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Pope, Howard L. [Aspen Resources Ltd., Inc., P.O. Box 3038, Boulder, CO 80307 (United States)] [Aspen Resources Ltd., Inc., P.O. Box 3038, Boulder, CO 80307 (United States); Enriquez, Alejandro E. [LANL NCO-4, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)] [LANL NCO-4, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Carson, Peter H. [LANL NPI-7, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)] [LANL NPI-7, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)

2013-07-01T23:59:59.000Z

120

DRAFT EM SSAB Chairs Meeting Waste Disposition Strategies Update  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelinesProvedDecember 2005Department ofDOE AccidentWaste Isolation Pilotat4 VolumeDOJEM SSAB

Note: This page contains sample records for the topic "waste disposition packaging" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

DRAFT EM SSAB Chairs Meeting Waste Disposition Strategies Update  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelinesProvedDecember 2005Department ofDOE AccidentWaste Isolation Pilotat4 VolumeDOJEM SSAB EM

122

DOE Chooses Idaho Treatment Group, LLC to Disposition Waste at the Advanced  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Career Scientists'Montana. DOCUMENTSof Energy DOE Challenge Home Recommended QualityDOEMixed Waste

123

Low-level Waste Safely Dispositioned Under Runoff Cover at SRS | Department  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power.pdf11-161-LNG | DepartmentEnergy InvitationLegaltoLizLow- LEVEL WASTE DISPOSALof

124

Review of DOE Waste Package Program. Semiannual report, October 1984-March 1985. Volume 8  

SciTech Connect (OSTI)

A large number of technical reports on waste package component performance were reviewed over the last year in support of the NRC`s review of the Department of Energy`s (DOE`s) Environmental Assessment reports. The intent was to assess in some detail the quantity and quality of the DOE data and their relevance to the high-level waste repository site selection process. A representative selection of the reviews is presented for the salt, basalt, and tuff repository projects. Areas for future research have been outlined. 141 refs.

Davis, M.S. (ed.)

1985-12-01T23:59:59.000Z

125

Evaluation and compilation of DOE waste package test data: Biannual report, August 1987--January 1988  

SciTech Connect (OSTI)

This report summarizes results of the National Bureau of Standards (NBS) evaluations on waste packages designed for containment of radioactive high-level nuclear waste (HLW). The waste package is a proposed engineered barrier that is part of a permanent repository for HLW. Metal alloys are the principal barriers within the engineered system. Since enactment of the Budget Reconciliation Act for Fiscal Year 1988, the Yucca Mountain, Nevada, site (in which tuff is the geologic medium) is the only site that will be characterized for use as high-level nuclear waste repository. During the reporting period of August 1987 to January 1988, five reviews were completed for tuff, and these were grouped into the categories: ferrous alloys, copper, groundwater chemistry, and glass. Two issues are identified for the Yucca Mountain site: the approach used to calculate corrosion rates for ferrous alloys, and crevice corrosion was observed in a copper-nickel alloy. Plutonium can form pseudo-colloids that may facilitate transport. NBS work related to the vitrification of HLW borosilicate glass at the West Valley Demonstration Project (WVDP) and the Defense Waste Processing Facility (DWPF) and activities of the DOE Materials Characterization Center (MCC) for the 6-month reporting period are also included. 27 refs., 3 figs.

Interrante, C.; Escalante, E.; Fraker, A.; Ondik, H.; Plante, E.; Ricker, R.; Ruspi, J.

1988-08-01T23:59:59.000Z

126

Effects of tuff waste package components on release from 76-68 simulated waste glass: Final report  

SciTech Connect (OSTI)

An experimental matrix has been conducted that will allow evaluation of the effects of waste package constituents on the waste form release behavior in a tuff repository environment. Tuff rock and groundwater were used along with 304L, 316, and 1020M ferrous metals to evaluate release from uranium-doped MCC 76-68 simulated waste glass. One of the major findings was that in the absence of 1020M mild steel, tuff rock powder dominates the system. However, when 1020M mild steel is present, it appears to dominate the system. The rock-dominated system results in suppressed glass-water reaction and leaching while the 1020M-dominated system results in enhanced leaching - but the metal effectively scavenges uranium from solution. The 300-series stainless steels play no significant role in affecting glass leaching characteristics. 6 refs., 28 figs., 5 tabs.

McVay, G.L.; Robinson, G.R.

1984-04-01T23:59:59.000Z

127

Microsoft PowerPoint - EM SSAB Chairs Webinar - Marcinowski Waste...  

Office of Environmental Management (EM)

Chair's Meeting Waste Disposition Strategies Update www.energy.govEM 1 Waste Disposition Strategies Update Frank Marcinowski Deputy Assistant Secretary for Waste Management Office...

128

Mission Need Statement: Calcine Disposition Project Major Systems Acquisition Project  

SciTech Connect (OSTI)

This document identifies the need to establish the Calcine Disposition Project to determine and implement the final disposition of calcine including characterization, retrieval, treatment (if necessary), packaging, loading, onsite interim storage pending shipment to a repository or interim storage facility, and disposition of related facilities.

J. T. Beck

2007-04-26T23:59:59.000Z

129

Contaminant Release Data Package for Residual Waste in Single-Shell Hanford Tanks  

SciTech Connect (OSTI)

The Hanford Federal Facility Agreement and Consent Order requires that a Resource Conservation and Recovery Act (RCRA) Facility Investigation report be submitted to the Washington State Department of Ecology. The RCRA Facility Investigation report will provide a detailed description of the state of knowledge needed for tank farm performance assessments. This data package provides detailed technical information about contaminant release from closed single-shell tanks necessary to support the RCRA Facility Investigation report. It was prepared by Pacific Northwest National Laboratory (PNNL) for CH2M HILL Hanford Group, Inc., which is tasked by the U.S. Department of Energy (DOE) with tank closure. This data package is a compilation of contaminant release rate data for residual waste in the four Hanford single-shell tanks (SSTs) that have been tested (C-103, C-106, C-202, and C-203). The report describes the geochemical properties of the primary contaminants of interest from the perspective of long-term risk to groundwater (uranium, technetium-99, iodine-129, chromium, transuranics, and nitrate), the occurrence of these contaminants in the residual waste, release mechanisms from the solid waste to water infiltrating the tanks in the future, and the laboratory tests conducted to measure release rates.

Deutsch, William J.; Cantrell, Kirk J.; Krupka, Kenneth M.

2007-12-01T23:59:59.000Z

130

TECHNICAL PEER REVIEW REPORT - YUCCA MOUNTAIN: WASTE PACKAGE CLOSURE CONTROL SYSTEM  

SciTech Connect (OSTI)

The objective of the Waste Package Closure System (WPCS) project is to assist in the disposal of spent nuclear fuel (SNF) and associated high-level wastes (HLW) at the Yucca Mountain site in Nevada. Materials will be transferred from the casks into a waste package (WP), sealed, and placed into the underground facility. The SNF/HLW transfer and closure operations will be performed in an aboveground facility. The objective of the Control System is to bring together major components of the entire WPCS ensuring that unit operations correctly receive, and respond to, commands and requests for data. Integrated control systems will be provided to ensure that all operations can be performed remotely. Maintenance on equipment may be done using hands-on or remote methods, depending on complexity, exposure, and ease of access. Operating parameters and nondestructive examination results will be collected and stored as permanent electronic records. Minor weld repairs must be performed within the closure cell if the welds do not meet the inspection acceptance requirements. Any WP with extensive weld defects that require lids to be removed will be moved to the remediation facility for repair.

NA

2005-10-25T23:59:59.000Z

131

Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material  

SciTech Connect (OSTI)

Stress corrosion cracking is one of the most common corrosion-related causes for premature breach of metal structural components. Stress corrosion cracking is the initiation and propagation of cracks in structural components due to three factors that must be present simultaneously: metallurgical susceptibility, critical environment, and static (or sustained) tensile stresses. This report was prepared according to ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). The purpose of this report is to provide an evaluation of the potential for stress corrosion cracking of the engineered barrier system components (i.e., the drip shield, waste package outer barrier, and waste package stainless steel inner structural cylinder) under exposure conditions consistent with the repository during the regulatory period of 10,000 years after permanent closure. For the drip shield and waste package outer barrier, the critical environment is conservatively taken as any aqueous environment contacting the metal surfaces. Appendix B of this report describes the development of the SCC-relevant seismic crack density model (SCDM). The consequence of a stress corrosion cracking breach of the drip shield, the waste package outer barrier, or the stainless steel inner structural cylinder material is the initiation and propagation of tight, sometimes branching, cracks that might be induced by the combination of an aggressive environment and various tensile stresses that can develop in the drip shields or the waste packages. The Stainless Steel Type 316 inner structural cylinder of the waste package is excluded from the stress corrosion cracking evaluation because the Total System Performance Assessment for License Application (TSPA-LA) does not take credit for the inner cylinder. This document provides a detailed description of the process-level models that can be applied to assess the performance of Alloy 22 (used for the waste package outer barrier) and Titanium Grade 7 (used for the drip shield) that are subjected to the effects of stress corrosion cracking. The use of laser peening or other residual stress mitigation techniques is considered as a means of mitigating stress corrosion cracking in the waste package final closure lid weld.

G. Gordon

2004-10-13T23:59:59.000Z

132

Waste Package Neutron Absorber, Thermal Shunt, and Fill Gas Selection Report  

SciTech Connect (OSTI)

Materials for neutron absorber, thermal shunt, and fill gas for use in the waste package were selected using a qualitative approach. For each component, selection criteria were identified; candidate materials were selected; and candidates were evaluated against these criteria. The neutron absorber materials evaluated were essentially boron-containing stainless steels. Two candidates were evaluated for the thermal shunt material. The fill gas candidates were common gases such as helium, argon, nitrogen, carbon dioxide, and dry air. Based on the performance of each candidate against the criteria, the following selections were made: Neutron absorber--Neutronit A978; Thermal shunt--Aluminum 6061 or 6063; and Fill gas--Helium.

V. Pasupathi

2000-01-28T23:59:59.000Z

133

Roadmapping - A Tool for Resolving Science and Technology Issues Related to Processing, Packaging, and Shipping Nuclear Materials and Waste  

SciTech Connect (OSTI)

Roadmapping is an effective methodology to identify and link technology development and deployment efforts to a program's or project's needs and requirements. Roadmapping focuses on needed technical support to the baselines (and to alternatives to the baselines) where the probability of success is low (high uncertainty) and the consequences of failure are relatively high (high programmatic risk, higher cost, longer schedule, or higher ES&H risk). The roadmap identifies where emphasis is needed, i.e., areas where investments are large, the return on investment is high, or the timing is crucial. The development of a roadmap typically involves problem definition (current state versus the desired state) and major steps (functions) needed to reach the desired state. For Nuclear Materials (NM), the functions could include processing, packaging, storage, shipping, and/or final disposition of the material. Each function is examined to determine what technical development would be needed to make the function perform as desired. This requires a good understanding of the current state of technology and technology development and validation activities to ensure the viability of each step. In NM disposition projects, timing is crucial! Technology must be deployed within the project window to be of value. Roadmaps set the stage to keep the technology development and deployment focused on project milestones and ensure that the technologies are sufficiently mature when needed to mitigate project risk and meet project commitments. A recent roadmapping activity involved a 'cross-program' effort, which included NM programs, to address an area of significant concern to the Department of Energy (DOE) related to gas generation issues, particularly hydrogen. The roadmap that was developed defined major gas generation issues within the DOE complex and research that has been and is being conducted to address gas generation concerns. The roadmap also provided the basis for sharing ''lessons learned'' from R&D efforts across DOE programs to increase efficiency and effectiveness in addressing gas generation issues. The gas generation roadmap identified pathways that have significant risk, indicating where more emphasis should be placed on contingency planning. Roadmapping further identified many opportunities for sharing of information and collaboration. Roadmapping will continue to be useful in keeping focused on the efforts necessary to mitigate the risk in the disposition pathways and to respond to the specific needs of the sites. Other areas within NM programs, including transportation and disposition of orphan and other nuclear materials, are prime candidates for additional roadmapping to assure achievement of timely and cost effective solutions for the processing, packaging, shipping, and/or final disposition of nuclear materials.

Luke, Dale Elden; Dixon, Brent Wayne; Murphy, James Anthony

2002-06-01T23:59:59.000Z

134

Environment on the Surfaces of the Drip Shield and Waste Package Outer Barrier  

SciTech Connect (OSTI)

This report provides supporting analysis of the conditions at which an aqueous solution can exist on the drip shield or waste package surfaces, including theoretical underpinning for the evolution of concentrated brines that could form by deliquescence or evaporation, and evaluation of the effects of acid-gas generation on brine composition. This analysis does not directly feed the total system performance assessment for the license application (TSPA-LA), but supports modeling and abstraction of the in-drift chemical environment (BSC 2004 [DIRS 169863]; BSC 2004 [DIRS 169860]). It also provides analyses that may support screening of features, events, and processes, and input for response to regulatory inquiries. This report emphasizes conditions of low relative humidity (RH) that, depending on temperature and chemical conditions, may be dry or may be associated with an aqueous phase containing concentrated electrolytes. Concentrated solutions at low RH may evolve by evaporative concentration of water that seeps into emplacement drifts, or by deliquescence of dust on the waste package or drip shield surfaces. The minimum RH for occurrence of aqueous conditions is calculated for various chemical systems based on current understanding of site geochemistry and equilibrium thermodynamics. The analysis makes use of known characteristics of Yucca Mountain waters and dust from existing tunnels, laboratory data, and relevant information from the technical literature and handbooks.

T. Wolery

2005-02-22T23:59:59.000Z

135

Radioactive waste isolation in salt: peer review of Westinghouse Electric Corporation's report on reference conceptual designs for a repository waste package  

SciTech Connect (OSTI)

This report documents the findings of the peer panel constituted by Argonne National Laboratory to review Region A of Westinghouse Electric Corporation's report entitled Waste Package Reference Conceptual Designs for a Repository in Salt. The panel determined that the reviewed report does not provide reasonable assurance that US Nuclear Regulatory Commission (NRC) requirements for waste packages will be met by the proposed design. It also found that it is premature to call the design a ''reference design,'' or even a ''reference conceptual design.'' This review report provides guidance for the preparation of a more acceptable design document.

Rote, D.M.; Hull, A.B.; Was, G.S.; Macdonald, D.D.; Wilde, B.E.; Russell, J.E.; Kruger, J.; Harrison, W.; Hambley, D.F.

1985-10-01T23:59:59.000Z

136

Evaluation of Calcine Disposition Path Forward  

SciTech Connect (OSTI)

This document describes an evaluation of the baseline and two alternative disposition paths for the final disposition of the calcine wastes stored at the Idaho Nuclear Technology and Engineering Center at the Idaho National Engineering and Environmental Laboratory. The pathways are evaluated against a prescribed set of criteria and a recommendation is made for the path forward.

Birrer, S.A.; Heiser, M.B.

2003-02-26T23:59:59.000Z

137

Evaluation of Calcine Disposition - Path Forward  

SciTech Connect (OSTI)

This document describes an evaluation of the baseline and two alternative disposition paths for the final disposition of the calcine wastes stored at the Idaho Nuclear Technology and Engineering Center at the Idaho National Engineering and Environmental Laboratory. The pathways are evaluated against a prescribed set of criteria and a recommendation is made for the path forward.

Steve Birrer

2003-02-01T23:59:59.000Z

138

Records Disposition  

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

To assign responsibilities and authorities and to prescribe policies, procedures, standards, and guidelines for the orderly disposition of records of the Department of Energy (DOE) and its management and operating contractors. Cancels DOE O 1324.2 dated 5-28-80. Chg 1 dated 4-9-92. Canceled by DOE O 1324.2B dated 1-12-95.

1988-09-13T23:59:59.000Z

139

Secondary Waste Form Down-Selection Data Package—DuraLith  

SciTech Connect (OSTI)

This data package developed for the DuraLith wasteform includes information available in the open literature and from data obtained from testing currently underway. DuraLith is an alkali-activated geopolymer waste form developed by the Vitreous State Laboratory at The Catholic University of America (VSL-CUA) for encapsulating liquid radioactive waste. A DuraLith waste form developed for treating Hanford secondary waste liquids is prepared by alkali-activation of a mixture of ground blast furnace slag and metakaolinite with sand used as a filler material. Based on optimization tests, solid waste loading of {approx}7.5% and {approx}14.7 % has been achieved using the Hanford secondary waste S1 and S4 simulants, respectively. The Na loading in both cases is equivalent to {approx}6 M. Some of the critical parameters for the DuraLith process include, hydrogen generation and heat evolution during activator solution preparation using the waste simulant, heat evolution during and after mixing the activator solution with the dry ingredients, and a working window of {approx}20 minutes to complete the pouring of the DuraLith mixture into molds. Results of the most recent testing indicated that the working window can be extended to {approx}30 minutes if 75 wt% of the binder components, namely, blast furnace slag and metakaolin are replaced by Class F fly ash. A preliminary DuraLith process flow sheet developed by VSL-CUA for processing Hanford secondary waste indicated that 10 to 22 waste monoliths (each 48 ft3 in volume) can be produced per day. There are no current pilot-scale or full-scale DuraLith plants under construction or in operation; therefore, the cost of DuraLith production is unknown. The results of the non-regulatory leach tests, EPA Draft 1313 and 1316, Waste Simulant S1-optimized DuraLith specimens indicated that the concentrations of RCRA metals (Ag, Cd, Cr, Hg, and Pb) in the leachates were well below the Universal Treatment Standard limits in 40 CFR 268.48. The data from the EPA draft 1315 leach test showed that LI values for COCs, namely 99Tc and I, ranged from 8.2 to 11.4 and 4.3 to 7.5, respectively. These values indicate that 99Tc meets the WAC LI requirement of 9.0 whereas, the LI values for I does not meet the WAC requirement of 11.0. Results of Toxicity Characteristic Leaching Procedure (TCLP)(EPA Method 1311) conducted on Waste Simulant S1-optimized DuraLith specimens, indicated that the concentrations of RCRA metals (Ag, As, Cd, Cr, Hg, and Pb) in the leachates were well below the Universal Treatment Standard limits in 40 CFR 268.48. The data from the ANSI/ANS 16.1 leach test showed that LI values for COC, namely Re (as a Tc surrogate), ranged from 8.06 to 10.81. The LI value for another COC, namely I, was not measured in this test. The results of the compressive strength testing of Waste Simulant S1-optimized DuraLith specimens indicated that the monoliths were physically robust with compressive strengths ranging from 115.5 MPa (16757 psi) to 156.2 MPA (22667 psi).

Mattigod, Shas V.; Westsik, Joseph H.

2011-09-15T23:59:59.000Z

140

Evaluation and compilation of DOE waste package test data: Biannual report, February 1987--July 1987  

SciTech Connect (OSTI)

The waste package is a proposed engineering barrier that is part of a permanent repository for HLW. Metal alloys are the principal barriers within the engineered system. Technical discussions are given for the corrosion of metals proposed for the canister, particularly carbon steels, stainless steels, and copper. The current level of understanding of several canister materials is questioned for the candidate repository in tuff. Three issues are addressed, the possibility of the stress-induced failure of Zircaloy, the possible corrosion of copper and copper alloys, and the lack of site-specific characterization data. Discussions are given on problems concerning localized corrosion and environmentally assisted cracking of AISI 1020 steel at elevated temperatures (150{degree}C). For the proposed salt site, the importance of the duration of corrosion tests and some of the conditions that may preclude prompt initiation of needed long-term testing are two issues that are discussed. 31 refs., 5 figs.

Interrante, C.; Escalante, E.; Fraker, A.; Hall, D.; Harrison, S.; Liggett, W.; Linzer, M.; Ricker, R.; Ruspi, J.; Shull, R.

1988-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "waste disposition packaging" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

assay package insert: Topics by E-print Network  

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

disposal. An important task for waste management organizations is to translate general waste acceptance requirements into detailed waste package specifications. Waste package...

142

Records Disposition  

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

To assign responsibilities and authorities and to prescribe policies, procedures, standards, and guidelines for the orderly disposition of records of the Department of Energy (DOE) and its operating and onsite service contractors. Cancels DOE O 1324.1 dated 7-10-78. Chg 1 dated 7-2-81. Chg 2 dated 11-9-82. Canceled by DOE O 1324.2A dated 9-13-88.

1980-05-28T23:59:59.000Z

143

THERMAL EVALUATION OF THE USE OF BWR MOX SNF IN THE WASTE PACKAGE DESIGN (SCPB: N/A)  

SciTech Connect (OSTI)

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) as specified in the Waste Package Implementation Plan (pp. 4-8,4-11,4-24, 5-1, and 5-13; Ref. 5.10) and Waste Package Plan (pp. 3-15,3-17, and 3-24; Ref. 5.9). The design data request addressed herein is: (1) Characterize the conceptual 40 BWR and 24 BWR Multi-Purpose Canister (MPC) Waste Package (WP) design to show that the design is feasible for use in the MGDS environment when loaded with BWR MOX SNF. (2) Characterize the conceptual 44 BWR and 24 BWR Uncanistered Fuel (UCF) Waste Package (WP) design to show that the design is feasible for use in the MGDS environment when loaded with BWR MOX SNF. The purpose of this analysis is to respond to a concern that the long-term disposal thermal issues for the WP Design, if used with SNF designed for a MOX fuel cycle, do not preclude WP compatibility with the MGDS. The objective of this analysis is to provide thermal parameter information for the conceptual WP design with disposal container which is loaded with BWR MOX SNF under nominal MGDS repository conditions. The results are intended to show that the design has a reasonable chance to meet the MGDS design requirements for normal MGDS operation, and to provide the required guidance to determining the major design issues for future design efforts, and to show that the BWR MOX SNF loaded WP performance is similar to an WP loaded with commercial BWR SNF.

H. Wang

1997-01-23T23:59:59.000Z

144

Isotope production potential at Sandia National Laboratories: Product, waste, packaging, and transportation  

SciTech Connect (OSTI)

The U.S. Congress directed the U.S. Department of Energy to establish a domestic source of molybdenum-99, an essential isotope used in nuclear medicine and radiopharmacology. An Environmental Impact Statement for production of {sup 99}Mo at one of four candidate sites is being prepared. As one of the candidate sites, Sandia National Laboratories is developing the Isotope Production Project. Using federally approved processes and procedures now owned by the U.S. Department of Energy, and existing facilities that would be modified to meet the production requirements, the Sandia National Laboratories` Isotope Project would manufacture up to 30 percent of the U.S. market, with the capacity to meet 100 percent of the domestic need if necessary. This paper provides a brief overview of the facility, equipment, and processes required to produce isotopes. Packaging and transportation issues affecting both product and waste are addressed, and the storage and disposal of the four low-level radioactive waste types generated by the production program are considered. Recommendations for future development are provided.

Trennel, A.J.

1995-12-31T23:59:59.000Z

145

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

SciTech Connect (OSTI)

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

Tyacke, M.

1993-08-01T23:59:59.000Z

146

Hydrogen Concentration in the Inner-Most Container within a Pencil Tank Overpack Packaged in a Standard Waste Box Package  

SciTech Connect (OSTI)

A set of steady state diffusion flow equations, for the hydrogen diffusion from one bag to the next bag (or one plastic waste container to another), within a set of nested waste bags (or nested waste containers), are developed and presented. The input data is then presented and justified. Inputting the data for each volume and solving these equations yields the steady state hydrogen concentration in each volume. The input data (permeability of the bag surface and closure, dimensions and hydrogen generation rate) and equations are analyzed to obtain the hydrogen concentrations in the innermost container for a set of containers which are analyzed for the TRUCON code for the general waste containers and the TRUCON code for the Pencil Tank Overpacks (PTO) in a Standard Waste Box (SWB).

Marusich, Robert M.

2012-01-25T23:59:59.000Z

147

Pyrolysis behavior of different type of materials contained in the rejects of packaging waste sorting plants  

SciTech Connect (OSTI)

Highlights: Black-Right-Pointing-Pointer Study of the influence of materials in the pyrolysis of real plastic waste samples. Black-Right-Pointing-Pointer Inorganic compounds remain unaltered. Black-Right-Pointing-Pointer Cellulosic components give rise to an increase in char formation. Black-Right-Pointing-Pointer Cellulosic components promote the production of aqueous phase. Black-Right-Pointing-Pointer Cellulosic components increase CO and CO{sub 2} contents in the gases. - Abstract: In this paper rejected streams coming from a waste packaging material recovery facility have been characterized and separated into families of products of similar nature in order to determine the influence of different types of ingredients in the products obtained in the pyrolysis process. The pyrolysis experiments have been carried out in a non-stirred batch 3.5 dm{sup 3} reactor, swept with 1 L min{sup -1} N{sub 2}, at 500 Degree-Sign C for 30 min. Pyrolysis liquids are composed of an organic phase and an aqueous phase. The aqueous phase is greater as higher is the cellulosic material content in the sample. The organic phase contains valuable chemicals as styrene, ethylbenzene and toluene, and has high heating value (HHV) (33-40 MJ kg{sup -1}). Therefore they could be used as alternative fuels for heat and power generation and as a source of valuable chemicals. Pyrolysis gases are mainly composed of hydrocarbons but contain high amounts of CO and CO{sub 2}; their HHV is in the range of 18-46 MJ kg{sup -1}. The amount of CO-CO{sub 2} increases, and consequently HHV decreases as higher is the cellulosic content of the waste. Pyrolysis solids are mainly composed of inorganics and char formed in the process. The cellulosic materials lower the quality of the pyrolysis liquids and gases, and increase the production of char.

Adrados, A., E-mail: aitziber.adrados@ehu.es [Chemical and Environmental Engineering Department, School of Engineering of Bilbao, Alameda. Urquijo s/n, 48013 Bilbao (Spain); De Marco, I.; Lopez-Urionabarrenechea, A.; Caballero, B.M.; Laresgoiti, M.F. [Chemical and Environmental Engineering Department, School of Engineering of Bilbao, Alameda. Urquijo s/n, 48013 Bilbao (Spain)

2013-01-15T23:59:59.000Z

148

Evaluation and compilation of DOE waste package test data; Volume 8: Biannual report, August 1989--January 1990  

SciTech Connect (OSTI)

This report summarizes evaluations by the National Institute of Standards and Technology (NIST) of some of the Department of Energy (DOE) activities on waste packages designed for containment of radioactive high-level nuclear waste (HLW) for the six-month period, August 1989--January 1990. This includes reviews of related materials research and plans, information on the Yucca Mountain, Nevada disposal site activities, and other information regarding supporting research and special assistance. Short discussions are given relating to the publications reviewed and complete reviews and evaluations are included. Reports of other work are included in the Appendices.

Interrante, C.G. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of High-Level Waste Management; Fraker, A.C.; Escalante, E. [National Inst. of Standards and Technology (MSEL), Gaithersburg, MD (United States). Metallurgy Div.

1993-06-01T23:59:59.000Z

149

FINITE-ELEMENT ANALYSIS OF ROCK FALL ON UNCANISTERED FUEL WASTE PACKAGE DESIGNS (SCPB: N/A)  

SciTech Connect (OSTI)

The objective of this analysis is to explore the Uncanistered Fuel (UCF) Tube Design waste package (WP) resistance to rock falls. This analysis will also be used to determine the size of rock that can strike the WP without causing failure in the containment barriers from a height based on the starter tunnel dimensions. The purpose of this analysis is to document the models and methods used in the calculations.

Z. Ceylan

1996-10-18T23:59:59.000Z

150

Hydrogen Concentration in the Inner-Most Container within a Pencil Tank Overpack Packaged in a Standard Waste Box Package  

SciTech Connect (OSTI)

The purpose of this report is to evaluate hydrogen generation within Pencil Tank Overpacks (PTO) in a Standard Waste Box (SWB), to establish plutonium (Pu) limits for PTOs based on hydrogen concentration in the inner-most container and to establish required configurations or validate existing or proposed configurations for PTOs. The methodology and requirements are provided in this report.

Marusich, Robert M.

2013-08-15T23:59:59.000Z

151

Selection of candidate container materials for the conceptual waste package design for a potential high level nuclear waste repository at Yucca Mountain  

SciTech Connect (OSTI)

Preliminary selection criteria have been developed, peer-reviewed, and applied to a field of 41 candidate materials to choose three alloys for further consideration during the advanced conceptual design phase of waste package development for a potential high level nuclear waste repository at Yucca Mountain, Nevada. These three alloys are titanium grade 12, Alloy C-4, and Alloy 825. These selections are specific to the particular conceptual design outlined in the Site Characterization Plan. Other design concepts that may be considered in the advanced conceptual design phase may favor other materials choices.

Van Konynenburg, R.A.; Halsey, W.G.; McCright, R.D.; Clarke, W.L. Jr. [Lawrence Livermore National Lab., CA (United States)] [Lawrence Livermore National Lab., CA (United States); Gdowski, G.E. [KMI, Inc., Albuquerque, NM (United States)] [KMI, Inc., Albuquerque, NM (United States)

1993-02-01T23:59:59.000Z

152

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

SciTech Connect (OSTI)

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

Kapoor, A. [DOE; Gordon, S. [NSTec; Goldston, W. [Energy Solutions

2013-07-08T23:59:59.000Z

153

Salt Waste Processing Initiatives  

Office of Environmental Management (EM)

1 Patricia Suggs Salt Processing Team Lead Assistant Manager for Waste Disposition Project Office of Environmental Management Savannah River Site Salt Waste Processing Initiatives...

154

Report to Congress on the potential use of lead in the waste packages for a geologic repository at Yucca Mountain, Nevada  

SciTech Connect (OSTI)

In the Report of the Senate Committee on Appropriations accompanying the Energy and Water Appropriation Act for 1989, the Committee directed the Department of Energy (DOE) to evaluate the use of lead in the waste packages to be used in geologic repositories for spent nuclear fuel and high-level waste. The evaluation that was performed in response to this directive is presented in this report. This evaluation was based largely on a review of the technical literature on the behavior of lead, reports of work conducted in other countries, and work performed for the waste-management program being conducted by the DOE. The initial evaluation was limited to the potential use of lead in the packages to be used in the repository. Also, the focus of this report is post closure performance and not on retrievability and handling aspects of the waste package. 100 refs., 8 figs., 15 tabs.

NONE

1989-12-01T23:59:59.000Z

155

Determining the Appropriate Package and Transportation Methodology for the Detroit Edison, Fermi II Msrs and Associated Components  

SciTech Connect (OSTI)

During the spring of 2005, Detroit Edison, Enrico Fermi II Nuclear Power Station (Fermi) decided to disposition two MSRs and associated components scheduled for replacement in the spring of 2006 during the MSR Replacement Outage. Of concern to Fermi was the proper packaging and transportation methodology when dis-positioning a component measuring approximately 110' in length and 13' in diameter and weighing over 300 tons. Upon removal from the Turbine Deck the retired MSRs and associated components were turned over to the Rad Waste Group for packaging and final disposition. Fermi requested quotations from vendors to package, transport, and disposition the MSRs and associated components. However, multiple Vendors informed Fermi that the size and weight of the MSRs were questionable in passing permitting requirements and would require segmentation and volume reduction on site or at a waste processor. Fermi contracted with MHF Logistical Solutions (MHF-LS) based on their ability to receive clearances for shipping the MSRs in one piece via two heavy haul rail conveyances acting as a bolstered load with professionally engineered blocking and bracing configured to support the retired MSRs. (authors)

Weber, B. [Detroit Edison Company/Enrico Fermi II Nuclear Power Station, Newport, MI (United States); Dempsey, S. [MHF Logistical Solutions, Oak Ridge, TN (United States)

2007-07-01T23:59:59.000Z

156

Potential for the localized corrosion of alloy 22 Waste Packages in Multiple-Salt Deliquescent Brines in the Yucca Mountain Repository  

SciTech Connect (OSTI)

It has been postulated that the deliquescence of multiple-salt systems in dust deposits and the consequent localized corrosion in high-temperature brines could lead to premature failure of the Alloy 22 waste packages in the Yucca Mountain repository. EPRI has developed a decision tree approach to determine if the various stages leading to waste package failure are possible and whether the safety of the repository system could be compromised as a result. Through a series of arguments, EPRI has shown that it is highly unlikely that the multiple-salt deliquescent brines will form in the first place and, even if they did, that they would not be thermodynamically stable, that the postulated brines are not corrosive and would not lead to the initiation of localized corrosion of Alloy 22, that even if localized corrosion did initiate that the propagation would stifle and cease long before penetration of the waste package outer barrier, and that even if premature waste package failures did occur from this cause that the safety of the overall system would not be compromised. EPRI concludes, therefore, that the postulated localized corrosion of the waste packages due to high-temperature deliquescent brines is neither a technical nor a safety issue of concern for the Yucca Mountain repository. (authors)

King, F. [Integrity Corrosion Consulting, Ltd., Calgary, AB (Canada); Arthur, R.; Apted, M. [Monitor Scientific LLC, Denver, CO (United States); Kessler, J.H. [Electric Power Research Institute, Charlotte, NC (United States)

2007-07-01T23:59:59.000Z

157

CH Packaging Operations Manual  

SciTech Connect (OSTI)

This procedure provides instructions for assembling the CH Packaging Drum payload assembly, Standard Waste Box (SWB) assembly, Abnormal Operations and ICV and OCV Preshipment Leakage Rate Tests on the packaging seals, using a nondestructive Helium (He) Leak Test.

Washington TRU Solutions LLC

2005-06-13T23:59:59.000Z

158

Plutonium Disposition Program | National Nuclear Security Administrati...  

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

Disposition Plutonium Disposition Program Plutonium Disposition Program The U.S.-Russia Plutonium Management and Disposition Agreement (PMDA), which entered into force on...

159

Data Package for Secondary Waste Form Down-Selection—Cast Stone  

SciTech Connect (OSTI)

Available literature on Cast Stone and Saltstone was reviewed with an emphasis on determining how Cast Stone and related grout waste forms performed in relationship to various criteria that will be used to decide whether a specific type of waste form meets acceptance criteria for disposal in the Integrated Disposal Facility (IDF) at Hanford. After the critical review of the Cast Stone/Saltstone literature, we conclude that Cast Stone is a good candidate waste form for further consideration. Cast stone meets the target IDF acceptance criteria for compressive strength, no free liquids, TCLP leachate are below the UTS permissible concentrations and leach rates for Na and Tc-99 are suiteably low. The cost of starting ingredients and equipment necessary to generate Cast Stone waste forms with secondary waste streams are low and the Cast Stone dry blend formulation can be tailored to accommodate variations in liquid waste stream compositions. The database for Cast Stone short-term performance is quite extensive compared to the other three candidate waste solidification processes. The solidification of liquid wastes in Cast Stone is a mature process in comparison to the other three candidates. Successful production of Cast Stone or Saltstone has been demonstrated from lab-scale monoliths with volumes of cm3 through m3 sized blocks to 210-liter sized drums all the way to the large pours into vaults at Savannah River. To date over 9 million gallons of low activity liquid waste has been solidified and disposed in concrete vaults at Savannah River.

Serne, R. Jeffrey; Westsik, Joseph H.

2011-09-05T23:59:59.000Z

160

CH Packaging Operations Manual  

SciTech Connect (OSTI)

This procedure provides instructions forassembling the following CH packaging payload: Drum payload assembly Standard Waste Box (SWB) assembly Ten-Drum Overpack (TDOP)

Washington TRU Solutions LLC

2007-11-29T23:59:59.000Z

Note: This page contains sample records for the topic "waste disposition packaging" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

CH Packaging Operations Manual  

SciTech Connect (OSTI)

This procedure provides instructions forassembling the following CH packaging payload: Drum payload assembly Standard Waste Box (SWB) assembly Ten-Drum Overpack (TDOP)

Washington TRU Solutions LLC

2007-08-22T23:59:59.000Z

162

CH Packaging Operations Manual  

SciTech Connect (OSTI)

Introduction - This procedure provides instructions for assembling the following CH packaging payload: -Drum payload assembly -Standard Waste Box (SWB) assembly -Ten-Drum Overpack (TDOP).

Washington TRU Solutions LLC

2003-06-26T23:59:59.000Z

163

CH Packaging Operations Manual  

SciTech Connect (OSTI)

Introduction - This procedure provides instructions forassembling the following CH packaging payload: Drum payload assembly Standard Waste Box (SWB) assembly Ten-Drum Overpack (TDOP)

Washington TRU Solutions LLC

2007-05-15T23:59:59.000Z

164

CH Packaging Operations Manual  

SciTech Connect (OSTI)

Introduction - This procedure provides instructions forassembling the following CH packaging payload: Drum payload assembly Standard Waste Box (SWB) assembly Ten-Drum Overpack (TDOP)

Washington TRU Solutions LLC

2006-12-18T23:59:59.000Z

165

Geology Data Package for the Single-Shell Tank Waste Management Areas at the Hanford Site  

SciTech Connect (OSTI)

This data package discusses the geology of the single-shell tank (SST) farms and the geologic history of the area. The focus of this report is to provide the most recent geologic information available for the SST farms. This report builds upon previous reports on the tank farm geology and Integrated Disposal Facility geology with information available after those reports were published.

Reidel, Steve P.; Chamness, Mickie A.

2007-01-01T23:59:59.000Z

166

Vendor Assessment for the Waste Package Closure System (Yucca Mtn. Project)  

SciTech Connect (OSTI)

The Idaho National Engineering and Environmental Laboratory (INEEL) has been tasked with developing, designing, constructing, and operating a full-scale prototype of the work package closure system. As a precursor to developing the conceptual design, all commercially available equipment was assessed to identify any existing technology gaps. This report presents the results of that assessment for all major equipment.

Colleen Shelton-Davis

2003-09-01T23:59:59.000Z

167

Vendor Assessment for the Waste Package Closure System (Yucca Mountain Project)  

SciTech Connect (OSTI)

The Idaho National Engineering and Environmental Laboratory (INEEL) has been tasked with developing, designing, constructing, and operating a full-scale prototype of the work package closure system. As a precursor to developing the conceptual design, all commercially available equipment was assessed to identify any existing technology gaps. This report presents the results of that assessment for all major equipment.

Shelton-Davis, C.V.

2003-09-26T23:59:59.000Z

168

Improving D&D Planning and Waste Management with Cutting and Packaging Simulation  

SciTech Connect (OSTI)

The increased amount of decontamination and decommissioning (D&D) being performed throughout the world not only strains nuclear cleanup budgets, but places severe demands on the capacities of nuclear waste disposal sites. Although budgets and waste disposal sites have been able to accommodate the demand thus far, the increasing number of large facilities being decommissioned will cause major impacts to the waste disposal process. It is thus imperative that new and innovative technologies are applied within the D&D industry to reduce costs and waste disposal requirements for the decommissioning of our inventory of large and aging nuclear facilities. One of the most significant problems reactor owner’s deal with is the accurate determination of the types and volumes of wastes that will be generated during decommissioning of their facilities. Waste disposal costs, restrictions, and transportation issues can account for as much as 30% of the total costs to decommission a facility and thus it is very important to have accurate waste volume estimates. The use of simulation technologies to estimate and reduce decommissioning waste volumes provides a new way to manage risks associated with this work. Simulation improves the process by allowing facility owners to obtain accurate estimates of the types and amounts of waste prior to starting the actual D&D work. This reduces risk by permitting earlier and better negotiations with the disposal sites, and more time to resolve transportation issues. While simulation is a tool to be used by the D&D contractors, its real value is in reducing risks and costs to the reactor owners.

Richard H. Meservey; Jean-Louis Bouchet

2005-08-01T23:59:59.000Z

169

Tank vapor sampling and analysis data package for tank 241-C-106 waste retrieval sluicing system process test phase III  

SciTech Connect (OSTI)

This data package presents sampling data and analytical results from the March 28, 1999, vapor sampling of Hanford Site single-shell tank 241-C-106 during active sluicing. Samples were obtained from the 296-C-006 ventilation system stack and ambient air at several locations. Characterization Project Operations (CPO) was responsible for the collection of all SUMMATM canister samples. The Special Analytical Support (SAS) vapor team was responsible for the collection of all triple sorbent trap (TST), sorbent tube train (STT), polyurethane foam (PUF), and particulate filter samples collected at the 296-C-006 stack. The SAS vapor team used the non-electrical vapor sampling (NEVS) system to collect samples of the air, gases, and vapors from the 296-C-006 stack. The SAS vapor team collected and analyzed these samples for Lockheed Martin Hanford Corporation (LMHC) and Tank Waste Remediation System (TWRS) in accordance with the sampling and analytical requirements specified in the Waste Retrieval Sluicing System Vapor Sampling and Analysis Plan (SAP) for Evaluation of Organic Emissions, Process Test Phase III, HNF-4212, Rev. 0-A, (LMHC, 1999). All samples were stored in a secured Radioactive Materials Area (RMA) until the samples were radiologically released and received by SAS for analysis. The Waste Sampling and Characterization Facility (WSCF) performed the radiological analyses. The samples were received on April 5, 1999.

LOCKREM, L.L.

1999-08-13T23:59:59.000Z

170

ESTIMATION OF RADIOLYTIC GAS GENERATION RATE FOR CYLINDRICAL RADIOACTIVE WASTE PACKAGES - APPLICATION TO SPENT ION EXCHANGE RESIN CONTAINERS  

SciTech Connect (OSTI)

Radioactive waste packages containing water and/or organic substances have the potential to radiolytically generate hydrogen and other combustible gases. Typically, the radiolytic gas generation rate is estimated from the energy deposition rate and the radiolytic gas yield. Estimation of the energy deposition rate must take into account the contributions from all radionuclides. While the contributions from non-gamma emitting radionuclides are relatively easy to estimate, an average geometry factor must be computed to determine the contribution from gamma emitters. Hitherto, no satisfactory method existed for estimating the geometry factors for a cylindrical package. In the present study, a formulation was developed taking into account the effect of photon buildup. A prototype code, called PC-CAGE, was developed to numerically solve the integrals involved. Based on the selected dimensions for a cylinder, the specified waste material, the photon energy of interest and a value for either the absorption or attenuation coefficient, the code outputs values for point and average geometry factors. These can then be used to estimate the internal dose rate to the material in the cylinder and hence to calculate the radiolytic gas generation rate. Besides the ability to estimate the rates of radiolytic gas generation, PC-CAGE can also estimate the dose received by the container material. This is based on values for the point geometry factors at the surface of the cylinder. PC-CAGE was used to calculate geometry factors for a number of cylindrical geometries. Estimates for the absorbed dose rate in container material were also obtained. The results for Ontario Power Generation's 3 m3 resin containers indicate that about 80% of the source gamma energy is deposited internally. In general, the fraction of gamma energy deposited internally depends on the dimensions of the cylinder, the material within it and the photon energy; the fraction deposited increases with increasing dimensions of the cylinder and decreases with increasing photon energy.

Husain, A.; Lewis, Brent J.

2003-02-27T23:59:59.000Z

171

Preliminary Study of Radioactive Waste Package Made of High-Strength and Ultra Low-Permeability Concrete for Geological Disposal of TRU Wastes  

SciTech Connect (OSTI)

We have been developing a radioactive waste package made of high-strength and ultra low-permeability concrete (HSULPC) for geological disposal of TRU wastes, which is expected to be much more impervious to water than conventional concrete. In this study, basic data for the HSULPC regarding its the impervious character and the thermodynamics during cement hydration were obtained through water permeability measurements using cold isostatic pressing (CIP) and adiabatic concrete hydration experiments, respectively. Then, a prediction tool to find concrete package construction conditions to avoid thermal cracking was developed, which could deal with coupled calculations of cement hydration, heat transfer, stress, and cracking. The developed tool was applied to HSULPC hydration on a small-scale cylindrical model to examine whether there was any effect on cracking which depended on the ratio of concrete cylinder thickness to its inner diameter. The results were compared to experiments. For concrete with a compressive strength of 200MPa, the water permeability coefficient was 4 x 10{sup 19} m/s. Dependences of activation energy and frequency factor on degree of cement hydration had a sharp peaking due to the nucleation rate-determining step, and a gradual increase region due to the diffusion rate-determining step. From analyses of the small-scale cylindrical model, dependences of the maximum principal stress on the radius were obtained. When the ratio of the concrete thickness to the heater diameter was around 1, the risk of cracking was predicted to be minimized. These numerical predictions from the developed tool were verified by experiments.

Matsuo, T.; Kawasaki, T.; Sakamoto, H.; Asano, E.; Takei, A.; Shibuya, K.; Katagiri, M.

2003-02-27T23:59:59.000Z

172

SELECTION OF SURPLUS PLUTONIUM MATERIALS FOR DISPOSITION TO WIPP  

SciTech Connect (OSTI)

The U.S. Department of Energy (DOE) is preparing a Surplus Plutonium Disposition (SPD) Supplemental Environmental Impact Statement (SEIS). Included in the evaluation are up to 6 metric tons (MT) of plutonium in the form of impure oxides and metals for which a disposition plan has not been decided, among options that include preparation as feed for the Mixed Oxide Fuel Fabrication Facility; disposing to high-level waste through the Savannah River Site (SRS) HB Line and H Canyon; can-in-canister disposal using the SRS Defense Waste Processing Facility; and preparation for disposal at the Waste Isolation Pilot Plant (WIPP). DOE and SRS have identified at least 0.5 MT of plutonium that, because of high levels of chemical and isotopic impurities, is impractical for disposition by methods other than the WIPP pathway. Characteristics of these items and the disposition strategy are discussed.

Allender, J.; Mcclard, J.; Christopher, J.

2012-06-08T23:59:59.000Z

173

Characterizing Surplus US Plutonium for Disposition - 13199  

SciTech Connect (OSTI)

The United States (US) has identified 61.5 metric tons (MT) of plutonium that is permanently excess to use in nuclear weapons programs, including 47.2 MT of weapons-grade plutonium. Surplus inventories will be stored safely by the Department of Energy (DOE) and then transferred to facilities that will prepare the plutonium for permanent disposition. The Savannah River National Laboratory (SRNL) operates a Feed Characterization program for the Office of Fissile Materials Disposition (OFMD) of the National Nuclear Security Administration (NNSA) and the DOE Office of Environmental Management (DOE-EM). SRNL manages a broad program of item tracking through process history, laboratory analysis, and non-destructive assay. A combination of analytical techniques allows SRNL to predict the isotopic and chemical properties that qualify materials for disposition through the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF). The research also defines properties that are important for other disposition paths, including disposal to the Waste Isolation Pilot Plant (WIPP) as transuranic waste (TRUW) or to high-level waste (HLW) systems. (authors)

Allender, Jeffrey S. [Savannah River National Laboratory, Aiken SC 29808 (United States)] [Savannah River National Laboratory, Aiken SC 29808 (United States); Moore, Edwin N. [Moore Nuclear Energy, LLC, Savannah River Site, Aiken SC 29808 (United States)] [Moore Nuclear Energy, LLC, Savannah River Site, Aiken SC 29808 (United States)

2013-07-01T23:59:59.000Z

174

Characterizing surplus US plutonium for disposition  

SciTech Connect (OSTI)

The United States (US) has identified 61.5 metric tons (MT) of plutonium that is permanently excess to use in nuclear weapons programs, including 47.2 MT of weapons-grade plutonium. Surplus inventories will be stored safely by the Department of Energy (DOE) and then transferred to facilities that will prepare the plutonium for permanent disposition. The Savannah River National Laboratory (SRNL) operates a Feed Characterization program for the Office of Fissile Materials Disposition (OFMD) of the National Nuclear Security Administration (NNSA) and the DOE Office of Environmental Management (DOE-EM). SRNL manages a broad program of item tracking through process history, laboratory analysis, and non-destructive assay. A combination of analytical techniques allows SRNL to predict the isotopic and chemical properties that qualify materials for disposition through the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF). The research also defines properties that are important for other disposition paths, including disposal to the Waste Isolation Pilot Plant (WIPP) as transuranic waste (TRUW) or to high-level waste (HLW) systems.

Allender, Jeffrey S.; Moore, Edwin N.

2013-02-26T23:59:59.000Z

175

Degradation mode survey candidate titanium-base alloys for Yucca Mountain project waste package materials. Revision 1  

SciTech Connect (OSTI)

The Yucca Mountain Site Characterization Project (YMP) is evaluating materials from which to fabricate high-level nuclear waste containers (hereafter called waste packages) for the potential repository at Yucca Mountain, Nevada. Because of their very good corrosion resistance in aqueous environments titanium alloys are considered for container materials. Consideration of titanium alloys is understandable since about one-third (in 1978) of all titanium produced is used in applications where corrosion resistance is of primary importance. Consequently, there is a considerable amount of data which demonstrates that titanium alloys, in general, but particularly the commercial purity and dilute {alpha} grades, are highly corrosion resistant. This report will discuss the corrosion characteristics of Ti Gr 2, 7, 12, and 16. The more highly alloyed titanium alloys which were developed by adding a small Pd content to higher strength Ti alloys in order to give them better corrosion resistance will not be considered in this report. These alloys are all two phase ({alpha} and {beta}) alloys. The palladium addition while making these alloys more corrosion resistant does not give them the corrosion resistance of the single phase {alpha} and near-{alpha} (Ti Gr 12) alloys.

Gdowski, G.E.

1997-12-01T23:59:59.000Z

176

Radioactive Waste Radioactive Waste  

E-Print Network [OSTI]

#12;Radioactive Waste at UF Bldg 831 392-8400 #12;Radioactive Waste · Program is designed to;Radioactive Waste · Program requires · Generator support · Proper segregation · Packaging · labeling #12;Radioactive Waste · What is radioactive waste? · Anything that · Contains · or is contaminated

Slatton, Clint

177

Remaining Sites Verification Package for the 120-F-1 Glass Dump Waste Site, Waste Site Reclassification Form 2008-028  

SciTech Connect (OSTI)

The 120-F-1 waste site consisted of two dumping areas located 660 m southeast of the 105-F Reactor containing laboratory equipment and bottles, demolition debris, light bulbs and tubes, small batteries, small drums, and pesticide contaminated soil. It is probable that 108-F was the source of the debris but the material may have come from other locations within the 100-F Area. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

J. M. Capron

2008-06-27T23:59:59.000Z

178

Recharge Data Package for Hanford Single-Shell Tank Waste Management Areas  

SciTech Connect (OSTI)

Pacific Northwest National Laboratory (PNNL) assists CH2M HILL Hanford Group, Inc., in its preparation of the Resource Conservation and Recovery Act (RCRA) Facility Investigation report. One of the PNNL tasks is to use existing information to estimate recharge rates for past and current conditions as well as future scenarios involving cleanup and closure of tank farms. The existing information includes recharge-relevant data collected during activities associated with a host of projects, including those of RCRA, the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA), the CH2M HILL Tank Farm Vadose Zone Project, and the PNNL Remediation and Closure Science Project. As new information is published, the report contents can be updated. The objective of this data package was to use published data to provide recharge estimates for the scenarios being considered in the RCRA Facility Investigation. Recharge rates were estimated for areas that remain natural and undisturbed, areas where the vegetation has been disturbed, areas where both the vegetation and the soil have been disturbed, and areas that are engineered (e.g., surface barrier). The recharge estimates supplement the estimates provided by PNNL researchers in 2006 for the Hanford Site using additional field measurements and model analysis using weather data through 2006.

Fayer, Michael J.; Keller, Jason M.

2007-09-24T23:59:59.000Z

179

EQ6 Calculation for Chemical Degradation of Shippingport LWBR (TH/U Oxide) Spent Nuclear Fuel Waste Packages  

SciTech Connect (OSTI)

The Monitored Geologic Repository (MGR) Waste Package Department of the Civilian Radioactive Waste Management System Management & Operating contractor (CRWMS M&O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the Shippingport Light Water Breeder Reactor (LWBR) (Ref. 1). The Shippingport LWBR SNF has been considered for disposal at the potential Yucca Mountain site. Because of the high content of fissile material in the SNF, the waste package (WP) design requires special consideration of the amount and placement of neutron absorbers and the possible loss of absorbers and SNF materials over geologic time. For some WPs, the outer shell corrosion-resistant material (CRM) and the corrosion-allowance inner shell may breach (Refs. 2 and 3), allowing the influx of water. Water in the WP will moderate neutrons, increasing the likelihood of a criticality event within the WP; and the water may, in time, gradually leach the fissile components and neutron absorbers from the WP, further affecting the neutronics of the system. This study presents calculations of the long-term geochemical behavior of WPs containing a Shippingport LWBR SNF seed assembly, and high-level waste (HLW) glass canisters arranged according to the codisposal concept (Ref. 4). The specific study objectives were to determine: (1) The extent to which criticality control material, suggested for this WP design, will remain in the WP after corrosion/dissolution of the initial WP configuration (such that it can be effective in preventing criticality); (2) The extent to which fissile uranium and fertile thorium will be carried out of the degraded WP by infiltrating water (such that internal criticality is no longer possible, but the possibility of external criticality may be enhanced); and (3) The nominal chemical composition for the criticality evaluations of the WP design, and to suggest the range of parametric variations for additional evaluations. The scope of this calculation, the chemical compositions (and subsequent criticality evaluations), of the simulations are limited to time periods up to 3.17 x 10{sup 5} years. This longer time frame is closer to the one million year time horizon recently recommended by the National Academy of Sciences to the Environmental Protection Agency for performance assessment related to a nuclear repository (Ref. 5). However, it is important to note that after 100,000 years, most of the materials of interest (fissile and absorber materials) will have either been removed from the WP, reached a steady state, or been transmuted. The calculation included elements with high neutron-absorption cross sections, notably gadolinium (Gd), as well as the fissile materials. The results of this analysis will be used to ensure that the type and amount of criticality control material used in the WP design will prevent criticality.

S. Arthur

2000-09-14T23:59:59.000Z

180

Safety analysis report for packaging (Onsite) for the Hanford Ecorok packaging  

SciTech Connect (OSTI)

This safety analysis report for packaging approves the Hanford Ecorok packaging for shipping contaminated water purification filters from K Basins to the Central Waste Complex.

Mercado, M.S.

1996-02-23T23:59:59.000Z

Note: This page contains sample records for the topic "waste disposition packaging" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


181

Savannah River Site Eastern Transportation Hub: A Concept For a DOE Eastern Packaging, Staging and Maintenance Center - 13143  

SciTech Connect (OSTI)

The Department of Energy (DOE) is working to de-inventory sites and consolidate hazardous materials for processing and disposal. The DOE administers a wide range of certified shipping packages for the transport of hazardous materials to include Special Nuclear Material (SNM), radioactive materials, sealed sources and radioactive wastes. A critical element to successful and safe transportation of these materials is the availability of certified shipping packages. There are over seven thousand certified packagings (i.e., Type B/Type AF) utilized within the DOE for current missions. The synergistic effects of consolidated maintenance, refurbishment, testing, certification, and costing of these services would allow for efficient management of the packagings inventory and to support anticipated future in-commerce shipping needs. The Savannah River Site (SRS) receives and ships radioactive materials (including SNM) and waste on a regular basis for critical missions such as consolidated storage, stabilization, purification, or disposition using H-Canyon and HB-Line. The Savannah River National Laboratory (SRNL) has the technical capability and equipment for all aspects of packaging management. SRS has the only active material processing facility in the DOE complex and is one of the sites of choice for nuclear material consolidation. SRS is a logical location to perform maintenance and periodic testing of the DOE fleet of certified packagings. This initiative envisions a DOE Eastern Packaging Staging and Maintenance Center (PSMC) at the SRS and a western hub at the Nevada National Security Site (NNSS), an active DOE Regional Disposal Site. The PSMC's would be the first place DOE would go to meet their radioactive packaging needs and the primary locations projects would go to disposition excess packaging for beneficial reuse. These two hubs would provide the centralized management of a packaging fleet rather than the current approach to design, procure, maintain and dispose of packagings on a project-by-project basis. This initiative provides significant savings in packaging costs and acceleration of project schedules. In addition to certified packaging, the PSMC would be well suited for select designs of 7A Type A packaging and Industrial Packaging. (authors)

England, Jeffery L. [Savannah River National Laboratory, Aiken, South Carolina (United States)] [Savannah River National Laboratory, Aiken, South Carolina (United States); Adams, Karen; Maxted, Maxcine; Ruff Jr, Clarence [U.S. Department of Energy, Savannah River Site, Aiken, SC (United States)] [U.S. Department of Energy, Savannah River Site, Aiken, SC (United States); Albenesius, Andrew; Bowers, Mark D.; Fountain, Geoffrey; Hughes, Michael [Savannah River Nuclear Solutions, Aiken, SC (United States)] [Savannah River Nuclear Solutions, Aiken, SC (United States); Gordon, Sydney [National Security Technologies, LLC, Las Vegas, NV (United States)] [National Security Technologies, LLC, Las Vegas, NV (United States); O'Connor, Stephen [U.S. Department of Energy, HQ DOE, EM-33, Germantown MD (United States)] [U.S. Department of Energy, HQ DOE, EM-33, Germantown MD (United States)

2013-07-01T23:59:59.000Z

182

EQ6 Calculations for Chemical Degradation Of N Reactor (U-Metal) Spent Nuclear Fuel Waste Packages  

SciTech Connect (OSTI)

The Monitored Geologic Repository (MGR) Waste Package Department of the Civilian Radioactive Waste Management System Management & Operating Contractor (CRWMS M&O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the N Reactor, a graphite moderated reactor at the Department of Energy's (DOE) Hanford Site (ref. 1). The N Reactor core was fueled with slightly enriched (0.947 wt% and 0.947 to 1.25 wt% {sup 235}U in Mark IV and Mark IA fuels, respectively) U-metal clad in Zircaloy-2 (Ref. 1, Sec. 3). Both types of N Reactor SNF have been considered for disposal at the proposed Yucca Mountain site. For some WPs, the outer shell and inner shell may breach (Ref. 3) allowing the influx of water. Water in the WP will moderate neutrons, increasing the likelihood of a criticality event within the WP; and the water may, in time, gradually leach the fissile components from the WP, further affecting the neutronics of the system. This study presents calculations of the long-term geochemical behavior of WPs containing two multi-canister overpacks (MCO) with either six baskets of Mark IA or five baskets of Mark IV intact N Reactor SNF rods (Ref. 1, Sec. 4) and two high-level waste (HLW) glass pour canisters (GPCs) arranged according to the codisposal concept (Ref. 4). The specific study objectives were to determine: (1) The extent to which fissile uranium will remain in the WP after corrosion/dissolution of the initial WP configuration (2) The extent to which fissile uranium will be carried out of the degraded WP by infiltrating water (such that internal criticality is no longer possible, but the possibility of external criticality may be enhanced); and (3) The nominal chemical composition for the criticality evaluations of the WP design, and to suggest the range of parametric variations for additional evaluations. The scope of this calculation, the chemical compositions (and subsequent criticality evaluations) of the simulations, is limited to time periods up to 6.35 x 10{sup 5} years. This longer time frame is closer to the one million year time horizon recently recommended by the National Academy of Sciences to the Environmental Protection Agency for performance assessment related to a nuclear repository (Ref. 5). However, it is important to note that after 100,000 years, most of the materials of interest (fissile materials) will have either been removed from the WP, reached a steady state, or been transmuted.

P. Bernot

2001-02-27T23:59:59.000Z

183

Waste Receiving and Packaging, Module 2A, Supplemental Design Requirements Document  

SciTech Connect (OSTI)

The Supplemental Design Requirements Document (SDRD) is used to communicate plant design information from Westinghouse Hanford Company (WHC) to the US Department of Energy (DOE) and the cognizant Architect Engineer (A/E). Information in the SDRD serves two purposes: to convey design requirements that are too detailed for inclusion in a Functional Design Criteria (FDC) report; and to serve as a means of change control for design commitments in the Conceptual Design Report. The mission of WRAP 2A on the Hanford site is the treatment of contact handled low level mixed waste (MW) for final disposal. The overall systems engineering steps used to reach construction and operation of WRAP 2A are depicted in Figure 1. The WRAP 2A SDRD focuses on the requirements to address the functional analysis provided in Figure 1. This information is provided in sections 2 through 5 of this SDRD. The mission analysis and functional analysis are to be provided in a separate supporting document. The organization of sections 2 through 5 corresponds to the requirements identified in the WRAP 2A functional analysis.

Lamberd, D.L.; Boothe, G.F.; Hinkle, A.L.; Horgos, R.M.; LeClair, M.D.; Nash, C.R.; Ocampo, V.P.; Pauly, T.R.; Stroup, J.L.; Weingardt, K.M.

1994-04-26T23:59:59.000Z

184

Waste Treatment And Immobilization Plant U. S. Department Of Energy Office Of River Protection Submerged Bed Scrubber Condensate Disposition Project - Abstract # 13460  

SciTech Connect (OSTI)

The Hanford Waste Treatment and Immobilization Plant (WTP) will generate an off-gas treatment system secondary liquid waste stream [submerged bed scrubber (SBS) condensate], which is currently planned for recycle back to the WTP Low Activity Waste (LAW) melter. This SBS condensate waste stream is high in Tc-99, which is not efficiently captured in the vitrified glass matrix. A pre-conceptual engineering study was prepared in fiscal year 2012 to evaluate alternate flow paths for melter off-gas secondary liquid waste generated by the WTP LAW facility. This study evaluated alternatives for direct off-site disposal of this SBS without pre-treatment, which mitigates potential issues associated with recycling.

Yanochko, Ronald M [Washington River Protection Solutions, Richland, WA (United States); Corcoran, Connie [AEM Consulting, LLC, Richland, WA (United States)

2012-11-15T23:59:59.000Z

185

IN-PACKAGE CHEMISTRY ABSTRACTION  

SciTech Connect (OSTI)

This report was developed in accordance with the requirements in ''Technical Work Plan for Postclosure Waste Form Modeling'' (BSC 2005 [DIRS 173246]). The purpose of the in-package chemistry model is to predict the bulk chemistry inside of a breached waste package and to provide simplified expressions of that chemistry as a function of time after breach to Total Systems Performance Assessment for the License Application (TSPA-LA). The scope of this report is to describe the development and validation of the in-package chemistry model. The in-package model is a combination of two models, a batch reactor model, which uses the EQ3/6 geochemistry-modeling tool, and a surface complexation model, which is applied to the results of the batch reactor model. The batch reactor model considers chemical interactions of water with the waste package materials, and the waste form for commercial spent nuclear fuel (CSNF) waste packages and codisposed (CDSP) waste packages containing high-level waste glass (HLWG) and DOE spent fuel. The surface complexation model includes the impact of fluid-surface interactions (i.e., surface complexation) on the resulting fluid composition. The model examines two types of water influx: (1) the condensation of water vapor diffusing into the waste package, and (2) seepage water entering the waste package as a liquid from the drift. (1) Vapor-Influx Case: The condensation of vapor onto the waste package internals is simulated as pure H{sub 2}O and enters at a rate determined by the water vapor pressure for representative temperature and relative humidity conditions. (2) Liquid-Influx Case: The water entering a waste package from the drift is simulated as typical groundwater and enters at a rate determined by the amount of seepage available to flow through openings in a breached waste package.

E. Thomas

2005-07-14T23:59:59.000Z

186

Used Fuel Disposition Campaign Preliminary Quality Assurance...  

Energy Savers [EERE]

Used Fuel Disposition Campaign Preliminary Quality Assurance Implementation Plan Used Fuel Disposition Campaign Preliminary Quality Assurance Implementation Plan The primary...

187

Packaging, Transportation and Recycling of NPP Condenser Modules - 12262  

SciTech Connect (OSTI)

Perma-Fix was awarded contract from Energy Northwest for the packaging, transportation and disposition of the condenser modules, water boxes and miscellaneous metal, combustibles and water generated during the 2011 condenser replacement outage at the Columbia Generating Station. The work scope was to package the water boxes and condenser modules as they were removed from the facility and transfer them to the Perma-Fix Northwest facility for processing, recycle of metals and disposition. The condenser components were oversized and overweight (the condenser modules weighed ?102,058 kg [225,000 lb]) which required special equipment for loading and transport. Additional debris waste was packaged in inter-modals and IP-1 boxes for transport. A waste management plan was developed to minimize the generation of virtually any waste requiring landfill disposal. The Perma-Fix Northwest facility was modified to accommodate the ?15 m [50-ft] long condenser modules and equipment was designed and manufactured to complete the disassembly, decontamination and release survey. The condenser modules are currently undergoing processing for free release to a local metal recycler. Over three millions pounds of metal will be recycled and over 95% of the waste generated during this outage will not require land disposal. There were several elements of this project that needed to be addressed during the preparation for this outage and the subsequent packaging, transportation and processing. - Staffing the project to support 24/7 generation of large components and other wastes. - The design and manufacture of the soft-sided shipping containers for the condenser modules that measured ?15 m X 4 m X 3 m [50 ft X 13 ft X 10 ft] and weighed ?102,058 kg [225,000 lbs] - Developing a methodology for loading the modules into the shipping containers. - Obtaining a transport vehicle for the modules. - Designing and modifying the processing facility. - Movement of the modules at the processing facility. If any of these issues were not adequately resolved prior to the start of the outage, costly delays would result and the re-start of the power plant could be impacted. The main focus of this project was to find successful methods for keeping this material out of the landfills and preserving the natural resources. In addition, this operation provided a significant cost savings to the public utility by minimizing landfill disposal. The onsite portion of the project has been completed without impact to the overall outage schedule. By the date of presentation, the majority of the waste from the condenser replacement project will have been processed and recycled. The goals for this project included helping Energy Northwest maintain the outage schedule, package and characterize waste compliantly, perform transportation activities in compliance with 49CFR (Ref-1), and minimize the waste disposal volume. During this condenser replacement project, over three millions pounds of waste was generated, packaged, characterized and transported without injury or incident. It is anticipated that 95% of the waste generated during this project will not require landfill disposal. All of the waste is scheduled to be processed, decontaminated and recycled by June of 2012. (authors)

Polley, G.M. [Perma-Fix Environmental Services, 575 Oak Ridge Turnpike, Oak Ridge, TN 37830 (United States)

2012-07-01T23:59:59.000Z

188

Waste Treatment and Immobilization Plant U. S. Department of Energy Office of River Protection Submerged Bed Scrubber Condensate Disposition Project - 13460  

SciTech Connect (OSTI)

The Hanford Waste Treatment and Immobilization Plant (WTP) will generate an off-gas treatment system secondary liquid waste stream [submerged bed scrubber (SBS) condensate], which is currently planned for recycle back to the WTP Low Activity Waste (LAW) melter. This SBS condensate waste stream is high in Tc-99, which is not efficiently captured in the vitrified glass matrix [1]. A pre-conceptual engineering study was prepared in fiscal year 2012 to evaluate alternate flow paths for melter off-gas secondary liquid waste generated by the WTP LAW facility [2]. This study evaluated alternatives for direct off-site disposal of this SBS without pre-treatment, which mitigates potential issues associated with recycling. This study [2] concluded that SBS direct disposal is a viable option to the WTP baseline. The results show: - Off-site transportation and disposal of the SBS condensate is achievable and cost effective. - Reduction of approximately 4,325 vitrified WTP Low Activity Waste canisters could be realized. - Positive WTP operational impacts; minimal WTP construction impacts are realized. - Reduction of mass flow from the LAW Facility to the Pretreatment Facility by 66%. - Improved Double Shell Tank (DST) space management is a benefit. (authors)

Yanochko, Ronald M. [Washington River Protection Solutions, P.O. Box 850, Richland, Washington 99352 (United States)] [Washington River Protection Solutions, P.O. Box 850, Richland, Washington 99352 (United States); Corcoran, Connie [AEM Consulting, LLC, 1201 Jadwin Avenue, Richland, Washington 99352 (United States)] [AEM Consulting, LLC, 1201 Jadwin Avenue, Richland, Washington 99352 (United States)

2013-07-01T23:59:59.000Z

189

Recovering Americium and Curium from Mark-42 Target Materials- New Processing Approaches to Enhance Separations and Integrate Waste Stream Disposition - 12228  

SciTech Connect (OSTI)

Oak Ridge National Laboratory (ORNL) is investigating flowsheets to enhance processing efficiencies and to address waste streams associated with recovery of americium (Am) and curium (Cm) from Mark-42 (Mk-42) target materials stored at ORNL. The objective of this work was to identify the most effective flowsheet with which to process the 104 Mk-42 oxide capsules holding a total of 80 g of plutonium (Pu), 190 g of Cm, 480 g of Am, and 5 kg of lanthanide (Ln) oxides for the recovery and purification of the Am/Cm for future use as feedstock for heavy actinide production. Studies were also conducted to solidify the process flowsheet waste streams for disposal. ORNL is investigating flowsheets to enhance processing efficiencies and address waste streams associated with recovery of Am and Cm from Mk-42 target materials stored at ORNL. A series of small-scale runs are being performed to demonstrate an improved process to recover Am/Cm and to optimize the separations of Ln fission products from the Am/Cm constituents. The first of these runs has been completed using one of the Am/Cm/Ln oxide capsules stored at ORNL. The demonstration run showed promising results with a Ln DF of 40 for the Am/Cm product and an Am/Cm DF of 75 for the Ln product. In addition, the total losses of Am, Cm, and Ln to the waste solvents and raffinates were very low at <0.2%, 0.02%, and 0.04%, respectively. However, the Ln-actinide separation was less than desired. For future Reverse TALSPEAK demonstration runs, several parameters will be adjusted (flow rates, the ratio of scrub to strip stages, etc.) to improve the removal of Ln from the actinides. The next step will also include scale-up of the processing flowsheet to use more concentrated solutions (15 g/L Ln versus 5 g/L) and larger volumes and to recycle the HDEHP solvent. This should improve the overall processing efficiency and further reduce losses to waste streams. Studies have been performed with simulated wastes to develop solidification processes for disposal of the secondary waste streams generated by this flowsheet. Formulations were successfully developed for all the waste simulants. Additional tests with actual waste will be the next step in this effort. Future plans are to process all of the remaining 103 capsules in storage at ORNL. A nine-capsule test is now under way to provide additional information to scale-up the process to a target 20-capsule batch size for future processing runs. (authors)

Patton, Brad D.; Benker, Dennis; Collins, Emory D.; Mattus, Catherine H.; Robinson, Sharon M.; Wham, Robert M. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

2012-07-01T23:59:59.000Z

190

Disposition of Radioisotope Thermoelectric Generators Currently Located at the Oak Ridge National Laboratory - 12232  

SciTech Connect (OSTI)

Under the American Recovery and Reinvestment Act (ARRA), the U.S. Department of Energy (DOE) awarded SEC Federal Services Corporation (SEC) a 34-building demolition and disposal (D and D) project at the Oak Ridge National Laboratory (ORNL) that included the disposition of six Strontium (Sr-90) powered Radioisotope Thermoelectric Generators (RTGs) stored outside of ORNL Building 3517. Disposition of the RTGs is very complex both in terms of complying with disposal facility waste acceptance criteria (WAC) and U.S. Department of Transportation (DOT) requirements for packaging and transportation in commerce. Two of the RTGs contain elemental mercury which requires them to be Land Disposal Restrictions (LDR) compliant prior to disposal. In addition, all of the RTGs exceed the Class C waste concentration limits under Nuclear Regulatory Commission (NRC) Waste Classification Guidelines. In order to meet the LDR requirements and Nevada National Security Site (NNSS) WAC, a site specific treatability variance for mercury was submitted to the U.S. Environmental Protection Agency (EPA) to allow macro-encapsulation to be an acceptable treatment standard for elemental mercury. By identifying and confirming the design configuration of the mercury containing RTGs, the SEC team proved that the current configuration met the macro-encapsulation standard of 40 Code of Federal Regulations (CFR) 268.45. The SEC Team also worked with NNSS to demonstrate that all radioisotope considerations are compliant with the NNSS low-level waste (LLW) disposal facility performance assessment and WAC. Lastly, the SEC team determined that the GE2000 Type B cask met the necessary size, weight, and thermal loading requirements for five of the six RTGs. The sixth RTG (BUP-500) required a one-time DOT shipment exemption request due to the RTG's large size. The DOT exemption justification for the BUP-500 relies on the inherent robust construction and material make-up of the BUP- 500 RTG. DOE-ORO, SEC, and the entire SEC RTG team are nearing the conclusion of the Sr-90 RTG disposition challenge - a legacy now 50 years in the making. Over 600,000 Ci of Sr-90 waste await disposal and its removal from ORNL will mark an historical moment in the clean-up of the cold-war legacy in the ORNL central industrial area. Elimination (i.e., removal) of the RTGs will reduce security risks at ORNL and disposal will permanently eliminate security risks. The RTGs will eventually decay to benign levels within a reasonable timeframe relative to radiological risks posed by long-lived isotopes. The safety authorization basis at ORNL Building 3517 will be reduced enabling greater operational flexibility in future clean-out and D and D campaigns. Upon disposition the Department of Energy will realize reduced direct and indirect surveillance and maintenance costs that can be reapplied to accelerated and enhanced clean-up of the Oak Ridge Reservation. At present, waste profiles for the RTGs are developed and under review by NNSS RWAP staff and approval authorities. Disposition schedule is driven by the availability of compliant shipping casks necessary to safely transport the RTGs from ORNL to NNSS. The first disposal of the RCA RTG is expected in April 2012 and the remaining RTGs disposed in 2012 and 2013. (authors)

Glenn, J. [U.S. Department of Energy, Oak Ridge Operations Office, 200 Administrative Road, Oak Ridge, TN 37830 (United States); Patterson, J.; DeRoos, K. [SEC Federal Services Corporation (SEC), 2800 Solway Road, Knoxville, TN 37931 (United States); Patterson, J.E.; Mitchell, K.G. [Strata-G, LLC, 2027 Castaic Lane, Knoxville, TN 37932 (United States)

2012-07-01T23:59:59.000Z

191

Background studies in support of a feasibility assessment on the use of copper-base materials for nuclear waste packages in a repository in tuff  

SciTech Connect (OSTI)

This report combines six work units performed in FY`85--86 by the Copper Development Association and the International Copper Research Association under contract with the University of California. The work includes literature surveys and state-of-the-art summaries on several considerations influencing the feasibility of the use of copper-base materials for fabricating high-level nuclear waste packages for the proposed repository in tuff rock at Yucca Mountain, Nevada. The general conclusion from this work was that copper-base materials are viable candidates for inclusion in the materials selection process for this application. 55 refs., 48 figs., 22 tabs.

Van Konynenburg, R.A. [Lawrence Livermore National Lab., CA (USA); Kundig, K.J.A.; Lyman, W.S.; Prager, M.; Meyers, J.R.; Servi, I.S. [CDA/INCRA Joint Advisory Group, Greenwich, CT (USA)

1990-06-01T23:59:59.000Z

192

Multi-Year Work Plan to De-Inventory TRU Waste Stored at LANL - 12121  

SciTech Connect (OSTI)

The Los Alamos National Laboratory (LANL) continues to accelerate disposition of transuranic (TRU) waste stored at its Technical Area 54 (TA-54) Area G waste management facility. The current focus is on complete removal of all non-cemented above-grade Legacy and newly-generated TRU waste that was in storage on October 1, 2011, by no later than June 30, 2014. This inventory of above-grade TRU is defined as 3,706 m{sup 3} of material. Legacy TRU waste containers were placed into storage up to 40 years ago, and most of the older containers must be remediated to address compliance issues before the waste can be characterized, certified as meeting the Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria (WAC), and shipped for disposition. More than half of the remaining TRU waste volume stored above grade is contained within oversize boxes that contain waste items that must be repackaged or size reduced. Facilities and major types of equipment needed to remediate and characterize the TRU waste inventory are largely in place, but two additional oversize box processing lines are being implemented in 2012. Multiple work shifts are planned for most remediation lines in 2013. An integrated risk-based project management schedule for all disposition activities has been developed that is based on a 'Solution Package' approach. Inventories of containers that have issues in common were compiled into about 15 waste categories and about 70 'Solution Packages' that identify all of the activities needed to disposition the inventory of TRU waste in storage. Scheduled activities include all precursor activities to begin remediation, remediation processing, characterization and certification to the WIPP WAC, and shipping of containers to WIPP. Processing of the 3,706 m{sup 3} is projected to result in about 4,500 55-gallon (208 L) drums and 1,000 standard waste boxes that will be shipped to WIPP. About 385 shipments from LANL to WIPP are projected before June 30, 2014, to ship these containers, at a rate of 5 to 6 shipments a week. (authors)

Johns-Hughes, K.W.; Clemmons, J.S.; Hargis, K.M.; Christensen, D.V.; Shepard, M.D. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, New Mexico 87545 (United States); Bishop, M.L. [U.S. Department of Energy, National Nuclear Security Administration, Los Alamos Site Office, 3747 W. Jemez Road, Los Alamos, NM 87545 (United States)

2012-07-01T23:59:59.000Z

193

324 Building special-case waste assessment in support of the 324 Building closure (TPA milestone M-89-05)  

SciTech Connect (OSTI)

Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement Milestone M-89-05 requires US Department of Energy, Richland Operations Office to complete a 324 Building Special Case Waste Assessment in Support of the 324 Building Closure. This document has been prepared with the intent of meeting this regulatory commitment. Alternatives for the Special Case Wastes located in the 324 Building were defined and analyzed. Based on the criteria of safety, environmental, complexity of interfaces, risk, cost, schedule, and long-term operability and maintainability, the best alternative was chosen. Waste packaging and transportation options are also included in the recommendations. The waste disposition recommendations for the B-Cell dispersibles/tank heels and High-Level Vault packaged residuals are to direct them to the Plutonium Uranium Extraction Facility (PUREX) Number 2 storage tunnel.

Hobart, R.L.

1998-05-12T23:59:59.000Z

194

324 Facility special-case waste assessment in support of 324 closure (TPA milestone M-89-05)  

SciTech Connect (OSTI)

Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement Milestone M-89-05, requires US Department of Energy, Richland Operations Office to complete a 324 Facility Special-Case Waste Assessment in Support of 324 Closure. This document, HNF-1270, has been prepared with the intent of meeting this regulatory commitment. Alternatives for the special-case wastes located in the 324 Building were defined and analyzed. Based on the criteria of safety, environmental, complexity of interfaces, risk, cost, schedule, and long-term operability and maintainability, the best alternative was chosen. Waste packaging and transportation options are also included in the recommendations. The waste disposition recommendations for the B-Cell dispersibles/tank heels and High-Level Vault packaged residuals are to direct them to the Plutonium Uranium Extraction Facility (PUREX) Number 2 storage tunnel.

Hobart, R.L.

1998-06-25T23:59:59.000Z

195

A user's guide to the GoldSim/BLT-MS integrated software package:a low-level radioactive waste disposal performance assessment model.  

SciTech Connect (OSTI)

Sandia National Laboratories (Sandia), a U.S. Department of Energy National Laboratory, has over 30 years experience in the assessment of radioactive waste disposal and at the time of this publication is providing assistance internationally in a number of areas relevant to the safety assessment of radioactive waste disposal systems. In countries with small radioactive waste programs, international technology transfer program efforts are often hampered by small budgets, schedule constraints, and a lack of experienced personnel. In an effort to surmount these difficulties, Sandia has developed a system that utilizes a combination of commercially available software codes and existing legacy codes for probabilistic safety assessment modeling that facilitates the technology transfer and maximizes limited available funding. Numerous codes developed and endorsed by the United States Nuclear Regulatory Commission (NRC) and codes developed and maintained by United States Department of Energy are generally available to foreign countries after addressing import/export control and copyright requirements. From a programmatic view, it is easier to utilize existing codes than to develop new codes. From an economic perspective, it is not possible for most countries with small radioactive waste disposal programs to maintain complex software, which meets the rigors of both domestic regulatory requirements and international peer review. Therefore, revitalization of deterministic legacy codes, as well as an adaptation of contemporary deterministic codes, provides a credible and solid computational platform for constructing probabilistic safety assessment models. This document is a reference users guide for the GoldSim/BLT-MS integrated modeling software package developed as part of a cooperative technology transfer project between Sandia National Laboratories and the Institute of Nuclear Energy Research (INER) in Taiwan for the preliminary assessment of several candidate low-level waste repository sites. Breach, Leach, and Transport-Multiple Species (BLT-MS) is a U.S. NRC sponsored code which simulates release and transport of contaminants from a subsurface low-level waste disposal facility. GoldSim is commercially available probabilistic software package that has radionuclide transport capabilities. The following report guides a user through the steps necessary to use the integrated model and presents a successful application of the paradigm of renewing legacy codes for contemporary application.

Knowlton, Robert G.; Arnold, Bill Walter; Mattie, Patrick D.

2007-03-01T23:59:59.000Z

196

CHALLENGES WITH RETRIEVING TRANSURANIC WASTE FROM THE HANFORD BURIAL GROUNDS  

SciTech Connect (OSTI)

The U.S. DOE's Hanford Reservation produced plutonium and other nuclear materials for the nation's defense starting in World War II. The defense mission generated wastes that were either retrievably stored (i.e. retrievably stored waste) and/or disposed of in burial grounds. Challenges have emerged from retrieving suspect TRU waste including adequacy of records, radiological concerns, container integrity, industrial hygiene and safety issues, the lack of processing/treatment facilities, and the integration of regulatory requirements. All retrievably stored waste is managed as mixed waste and assumed to be TRU waste, unless documented otherwise. Mixed waste is defined as radioactive waste that contains hazardous constituents. The Atomic Energy Act governs waste with radionuclides, and the Resource Conservation and Recovery Act (RCRA) governs waste with hazardous constituents. Waste may also be governed by the Toxic Substances Control Act (TSCA), and a portion may be managed under the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA). In 1970, TRU waste was required to be placed in 20-year retrievable storage and segregated from other Waste. Prior to that date, segregation did not occur. Because of the changing definition of TRU over the years, and the limitations of early assay equipment, all retrievably stored waste in the burial grounds is managed as suspect TRU. Experience has shown that some of this waste will be characterized as low-level (non-TRU) waste after assay. The majority of the retrieved waste is not amenable to sampling due to waste type and/or radiological issues. Key to waste retrieval and disposition are characterization, historical investigation and research, knowledge of past handling and packaging, as well as a broad understanding and application of the regulations.

SWAN, R.J.; LAKES, M.E.

2007-08-06T23:59:59.000Z

197

Remaining Sites Verification Package for the 128-B-2, 100-B Burn Pit #2 Waste Site, Waste Site Reclassification Form 2005-038  

SciTech Connect (OSTI)

The 128-B-2 waste site was a burn pit historically used for the disposal of combustible and noncombustible wastes, including paint and solvents, office waste, concrete debris, and metallic debris. This site has been remediated by removing approximately 5,627 bank cubic meters of debris, ash, and contaminated soil to the Environmental Restoration Disposal Facility. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

R. A. Carlson

2005-12-21T23:59:59.000Z

198

Remaining Sites Verification Package for the 600-233 Waste Site, Vertical Pipe Near 100-B Electrical Laydown Area, Waste Site Reclassification Form 2005-041  

SciTech Connect (OSTI)

The 600-233 waste site consisted of three small-diameter pipelines within the 600-232 waste site, including previously unknown diesel fuel supply lines discovered during site remediation. The 600-233 waste site has been remediated to achieve the remedial action objectives specified in the Remaining Sites ROD. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

R. A. Carlson

2005-12-08T23:59:59.000Z

199

Weapons-grade plutonium dispositioning. Volume 2: Comparison of plutonium disposition options  

SciTech Connect (OSTI)

The Secretary of Energy requested the National Academy of Sciences (NAS) Committee on International Security and Arms Control to evaluate disposition options for weapons-grade plutonium. The Idaho National Engineering Laboratory (INEL) offered to assist the NAS in this evaluation by investigating the technical aspects of the disposition options and their capability for achieving plutonium annihilation levels greater than 90%. This report was prepared for the NAS to document the gathered information and results from the requested option evaluations. Evaluations were performed for 12 plutonium disposition options involving five reactor and one accelerator-based systems. Each option was evaluated in four technical areas: (1) fuel status, (2) reactor or accelerator-based system status, (3) waste-processing status, and (4) waste disposal status. Based on these evaluations, each concept was rated on its operational capability and time to deployment. A third rating category of option costs could not be performed because of the unavailability of adequate information from the concept sponsors. The four options achieving the highest rating, in alphabetical order, are the Advanced Light Water Reactor with plutonium-based ternary fuel, the Advanced Liquid Metal Reactor with plutonium-based fuel, the Advanced Liquid Metal Reactor with uranium-plutonium-based fuel, and the Modular High Temperature Gas-Cooled Reactor with plutonium-based fuel. Of these four options, the Advanced Light Water Reactor and the Modular High Temperature Gas-Cooled Reactor do not propose reprocessing of their irradiated fuel. Time constraints and lack of detailed information did not allow for any further ratings among these four options. The INEL recommends these four options be investigated further to determine the optimum reactor design for plutonium disposition.

Brownson, D.A.; Hanson, D.J.; Blackman, H.S. [and others

1993-06-01T23:59:59.000Z

200

RH Packaging Operations Manual  

SciTech Connect (OSTI)

This procedure provides operating instructions for the RH-TRU 72-B Road Cask, Waste Shipping Package. In this document, ''Packaging'' refers to the assembly of components necessary to ensure compliance with the packaging requirements (not loaded with a payload). ''Package'' refers to a Type B packaging that, with its radioactive contents, is designed to retain the integrity of its containment and shielding when subject to the normal conditions of transport and hypothetical accident test conditions set forth in 10 CFR Part 71. Loading of the RH 72-B cask can be done two ways, on the RH cask trailer in the vertical position or by removing the cask from the trailer and loading it in a facility designed for remote-handling (RH). Before loading the 72-B cask, loading procedures and changes to the loading procedures for the 72-B cask must be sent to CBFO at sitedocuments@wipp.ws for approval.

Washington TRU Solutions LLC

2003-09-17T23:59:59.000Z

Note: This page contains sample records for the topic "waste disposition packaging" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

Final evaluation report for Westinghouse Hanford Company, WRAP-1,208 liter waste drum, docket 94-35-7A, type A packaging  

SciTech Connect (OSTI)

This report documents the U.S. Department of Transportation Specification 7A Type A (DOT-7A) compliance test results of the Westinghouse Hanford Company, Waste Receiving and Processing Facility, Module 1 (WRAP-1) Drum. The WRAP-1 Drum was tested for DOE-HQ in August 1994, by Los Alamos National Laboratory, under docket number 94-35-7A. Additionally, comparison and evaluation of the approved, as-tested packaging configuration was performed by WHC in September 1995. The WRAP-1 Drum was evaluated against the performance of the DOT-17C, 208 1 (55-gal) steel drums tested and evaluated under dockets 89-13-7A/90-18-7A and 94-37-7A.

Kelly, D.L., Westinghouse Hanford

1996-06-12T23:59:59.000Z

202

The Use of the Hanford Onsite Packaging and Transportation Safety Program to Meet Cleanup Milestones Under the Hanford Site Cleanup 2015 Vision and the American Recovery and Reinvestment Act of 2009 - 12403  

SciTech Connect (OSTI)

The Hanford Site presents unique challenges in meeting the U.S. Department of Energy Richland Operations Office (DOE-RL) 2015 Cleanup Vision. CH2M Hill Plateau Remediation Company (CHPRC), its subcontractors, and DOE-RL were challenged to retrieve, transport and remediate a wide range of waste materials. Through a collaborative effort by all Hanford Onsite Central Plateau Cleanup Team Members, disposition pathways for diverse and seemingly impossible to ship wastes were developed under a DOE Order 460.1C-compliant Hanford Onsite Transportation Safety Program. The team determined an effective method for transporting oversized compliant waste payloads to processing and disposition facilities. The use of the onsite TSD packaging authorizations proved to be vital to safely transporting these materials for processing and eventual final disposition. The American Recovery and Reinvestment Act of 2009 (ARRA) provided additional resources to expedite planning and execution of these important cleanup milestones. Through the innovative and creative use of the TSD, the Hanford Onsite Central Plateau Cleanup Team Members have developed and are executing an integrated project plan that enables the safe and compliant transport of a wide variety of difficult-to-transport waste items, accelerating previous cleanup schedules to meet cleanup milestones. (authors)

Lavender, John C. [CH2M HILL Plateau Remediation Company, Richland, WA 99354 (United States); Edwards, W. Scott [Areva Federal Services, Richland, WA 99354 (United States); Macbeth, Paul J.; Self, Richard J. [U.S. Department of Energy Richland Operations Office, Richland, WA 99352 (United States); West, Lori D. [Materials and Energy Corporation, Richland, WA 99354 (United States)

2012-07-01T23:59:59.000Z

203

Remaining Sites Verification Package for the 100-B-1 Surface Chemical and Solid Waste Dumping Area, Waste Site Reclassification Form 2006-003  

SciTech Connect (OSTI)

The 100-B-1 waste site was a dumping site that was divided into two areas. One area was used as a laydown area for construction materials, and the other area was used as a chemical dumping area. The 100-B-1 Surface Chemical and Solid Waste Dumping Area site meets the remedial action objectives specified in the Remaining Sites ROD. The results demonstrate that residual contaminant concentrations support future unrestricted land uses that can be represented by a rural-residential scenario. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

R. A. Carlson

2006-04-24T23:59:59.000Z

204

Data Package for Past and Current Groundwater Flow and Contamination beneath Single-Shell Tank Waste Management Areas  

SciTech Connect (OSTI)

This appendix summarizes historic and recent groundwater data collected from the uppermost aquifer beneath the 200 East and 200 West Areas. Although the area of interest is the Hanford Site Central Plateau, most of the information discussed in this appendix is at the scale of individual single-shell tank waste management areas. This is because the geologic, and thus the hydraulic, properties and the geochemical properties (i.e., groundwater composition) are different in different parts of the Central Plateau.

Horton, Duane G.

2007-03-16T23:59:59.000Z

205

Performance Based Trending of Waste Disposition  

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

standard of performance, and recommends a simple system used to provide a monthly "feedback" mechanism for data quality, which can affect multiple facets of operations. What can...

206

Hight-Level Waste & Facilities Disposition  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-SeriesFlickr FlickrGuidedCH2MLLC High-Rate,Highlights Highlights Below

207

Performance Based Trending of Waste Disposition  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankCombustion | Department ofT ib l L d F SSalesOE0000652Grow YourPerformance Audit of the Department

208

Savannah River Site Waste Disposition Project  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power.pdf11-161-LNG |September2-SCORECARD-01-24-13 Page 1 of 1Sandra L.155-WSavannah27-SWTerrel

209

LANL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement  

SciTech Connect (OSTI)

The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. LANL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO{sub 2} powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within both Category 1 and 2 areas. Technical Area (TA) 55/Plutonium Facility 4 will be used to store the bulk PuO{sub 2} powder, fabricate MOX fuel pellets, assemble rods, and store fuel bundles. Bundles will be assembled at a separate facility, several of which have been identified as suitable for that activity. The Chemistry and Metallurgy Research Building (at TA-3) will be used for analytical chemistry support. Waste operations will be conducted in TA-50 and TA-54. Only very minor modifications will be needed to accommodate the LA program. These modifications consist mostly of minor equipment upgrades. A commercial reactor operator has not been identified for the LA irradiation. Postirradiation examination (PIE) of the irradiated fuel will take place at either Oak Ridge National Laboratory or ANL-W. The only modifications required at either PIE site would be to accommodate full-length irradiated fuel rods. Results from this program are critical to the overall plutonium distribution schedule.

Fisher, S.E.; Holdaway, R.; Ludwig, S.B. [and others

1998-08-01T23:59:59.000Z

210

Update of the Used Fuel Disposition Campaign Implementation Plan  

SciTech Connect (OSTI)

This Campaign Implementation Plan provides summary level detail describing how the Used Fuel Disposition Campaign (UFDC) supports achievement of the overarching mission and objectives of the Department of Energy Office of Nuclear Energy Fuel Cycle Technologies Program The implementation plan begins with the assumption of target dates that are set out in the January 2013 DOE Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste (http://energy.gov/downloads/strategy-management-and-disposal-used-nuclear-fuel-and-high-level-radioactive-waste). These target dates and goals are summarized in section III. This implementation plan will be maintained as a living document and will be updated as needed in response to progress in the Used Fuel Disposition Campaign and the Fuel Cycle Technologies Program.

Jens Birkholzer; Robert MacKinnon; Kevin McMahon; Sylvia Saltzstein; Ken Sorenson; Peter Swift

2014-09-01T23:59:59.000Z

211

Optimal segmentation and packaging process  

DOE Patents [OSTI]

A process for improving packaging efficiency uses three dimensional, computer simulated models with various optimization algorithms to determine the optimal segmentation process and packaging configurations based on constraints including container limitations. The present invention is applied to a process for decontaminating, decommissioning (D&D), and remediating a nuclear facility involving the segmentation and packaging of contaminated items in waste containers in order to minimize the number of cuts, maximize packaging density, and reduce worker radiation exposure. A three-dimensional, computer simulated, facility model of the contaminated items are created. The contaminated items are differentiated. The optimal location, orientation and sequence of the segmentation and packaging of the contaminated items is determined using the simulated model, the algorithms, and various constraints including container limitations. The cut locations and orientations are transposed to the simulated model. The contaminated items are actually segmented and packaged. The segmentation and packaging may be simulated beforehand. In addition, the contaminated items may be cataloged and recorded.

Kostelnik, Kevin M. (Idaho Falls, ID); Meservey, Richard H. (Idaho Falls, ID); Landon, Mark D. (Idaho Falls, ID)

1999-01-01T23:59:59.000Z

212

Tank Closure and Waste Management Environmental Impact Statement...  

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

which includes disposition of the SSTs, ancillary equipment, and soils. The SST (149 tanks) and DST (28 tanks) systems contain both hazardous and radioactive waste (mixed...

213

Roadmapping the Resolution of Gas Generation Issues in Packages Containing Radioactive Waste/Materials - A Status Report  

SciTech Connect (OSTI)

Gas generation issues, particularly hydrogen, have been an area of concern for the transport and storage of radioactive materials and waste in the Department of Energy (DOE) Complex. Potentially combustible gases can be generated through a variety of reactions, including chemical reactions and radiolytic decomposition of hydrogen- containing material. Since transportation regulations prohibit shipment of explosives and radioactive materials together, it was decided that hydrogen generation was a problem that warranted the execution of a high-level roadmapping effort. This paper discusses the major gas generation issues within the DOE Complex and the research that has been and is being conducted by the transuranic (TRU) waste, nuclear materials, and spent nuclear fuels (SNF) programs within DOE’s Environmental Management (EM) organizations to address gas generation concerns. This paper presents a "program level" roadmap that links technology development to program needs and identifies the probability of success in an effort to understand the programmatic risk associated with the issue of gas generation. This paper also presents the status of the roadmap and follow-up activities.

Luke, Dale Elden; Hamp, S.

2002-02-01T23:59:59.000Z

214

Roadmapping the Resolution of Gas Generation Issues in Packages Containing Radioactive Waste/Materials - A Status Report  

SciTech Connect (OSTI)

Gas generation issues, particularly hydrogen, have been an area of concern for the transport and storage of radioactive materials and waste in the Department of Energy (DOE) Complex. Potentially combustible gases can be generated through a variety of reactions, including chemical reactions and radiolytic decomposition of hydrogen-containing material. Since transportation regulations prohibit shipment of explosives and radioactive materials together, it was decided that hydrogen generation was a problem that warranted the execution of a high-level roadmapping effort. This paper discusses the major gas generation issues within the DOE Complex and the research that has been and is being conducted by the transuranic (TRU) waste, nuclear materials, and spent nuclear fuels (SNF) programs within DOE's Environmental Management (EM) organizations to address gas generation concerns. This paper presents a ''program level'' roadmap that links technology development to program needs and identifies the probability of success in an effort to understand the programmatic risk associated with the issue of gas generation. This paper also presents the status of the roadmap and follow-up activities.

Luke, D.E. (INEEL); Hamp, S. (DOE-Albuquerque Operations Office)

2002-01-04T23:59:59.000Z

215

National Transportation Stakeholders Forum (NTSF) | Department...  

Energy Savers [EERE]

Colorado Spring 2010 NTSF Meeting, Illinois Resources NTSF Charter Waste Management Nuclear Materials & Waste Tank Waste and Waste Processing Waste Disposition Packaging and...

216

Remaining Sites Verification Package for the 1607-B1 Septic System, Waste Site Reclassification Form 2007-015  

SciTech Connect (OSTI)

The 1607-B1 Septic System includes a septic tank, drain field, and associated connecting pipelines and influent sanitary sewer lines. This septic system serviced the former 1701-B Badgehouse, 1720-B Patrol Building/Change Room, and the 1709-B Fire Headquarters. The 1607-B1 waste site received unknown amounts of nonhazardous, nonradioactive sanitary sewage from these facilities during its operational history from 1944 to approximately 1970. In accordance with this evaluation, the confirmatory sampling results support a reclassification of this site to No Action. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

L. M. Dittmer

2007-08-30T23:59:59.000Z

217

Remaining Sites Verification Package for the 126-F-2, 183-F Clearwells, Waste Site Reclassification Form 2006-017  

SciTech Connect (OSTI)

The 126-F-2 site is the clearwell facility formerly used as part of the reactor cooling water treatment at the 183-F facility. During demolition operations in the 1970s, potentially contaminated debris was disposed in the eastern clearwell structure. The site has been remediated by removing all debris in the clearwell structure to the Environmental Restoration Disposal Facility. The results of radiological surveys and visual inspection of the remediated clearwell structure show neither residual contamination nor the potential for contaminant migration beyond the clearwell boundaries. The results of verification sampling at the remediation waste staging area demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

R. A. Carlson

2006-05-04T23:59:59.000Z

218

DEVELOPMENT OF GLASS AND CRYSTALLINE CERAMIC FORMS FOR DISPOSITION OF EXCESS PLUTONIUM  

SciTech Connect (OSTI)

In the aftermath of the Cold War, the United States Department of Energy (DOE) has identified up to 50 metric tons of excess plutonium that needs to be dispositioned. The bulk of the material is slated to be blended with uranium and fabricated into a Mixed Oxide (MOX) fuel for subsequent burning in commercial nuclear reactors. Excess plutonium-containing impurity materials making it unsuitable for fabrication into MOX fuel will need to be dispositioned via other means. Glass and crystalline ceramics have been developed and studied as candidate forms to immobilize these impure plutonium feeds. A titanate-based ceramic was identified as an excellent actinide material host. This composition was based on Synroc compositions previously developed for nuclear waste immobilization. These titanate ceramics were found to be able to accommodate extremely high quantities of fissile material and exhibit excellent aqueous durability. A lanthanide borosilicate (LaBS) glass was developed to accommodate high concentrations of plutonium and to be very tolerant of impurities yet still maintain good aqueous durability. Recent testing of alkali borosilicate compositions showed promise of using these compositions to disposition lower concentrations of plutonium using existing high level waste vitrification processes. The developed waste forms all appear to be suitable for Pu disposition. Depending on the actual types and concentrations of the Pu residue streams slated for disposition, each waste form offers unique advantages.

Marra, James; Cozzi, A; Crawford, C.; Herman, C.; Marra, John; Peeler, D.

2009-09-10T23:59:59.000Z

219

Depleted uranium disposition study -- Supplement, Revision 1  

SciTech Connect (OSTI)

The Department of Energy Office of Weapons and Materials Planning has requested a supplemental study to update the recent Depleted Uranium Disposition report. This supplemental study addresses new disposition alternatives and changes in status.

Becker, G.W.

1993-11-01T23:59:59.000Z

220

RH Packaging Program Guidance  

SciTech Connect (OSTI)

The purpose of this program guidance document is to provide the technical requirements for use, operation, inspection, and maintenance of the RH-TRU 72-B Waste Shipping Package and directly related components. This document complies with the requirements as specified in the RH-TRU 72-B Safety Analysis Report for Packaging (SARP), and Nuclear Regulatory Commission (NRC) Certificate of Compliance (C of C) 9212. If there is a conflict between this document and the SARP and/or C of C, the C of C shall govern. The C of C states: "...each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application." It further states: "...each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP tasks the Waste Isolation Pilot Plant (WIPP) Management and Operating (M&O) Contractor with assuring the packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with 10 Code of Federal Regulations (CFR) §71.8, "Deliberate Misconduct." Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) shall be notified immediately. CBFO will evaluate the issue and notify the NRC if required. In accordance with 10 CFR Part 71, "Packaging and Transportation of Radioactive Material," certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21, "Reporting of Defects and Noncompliance," regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations are conducted. This document details the instructions to be followed to operate, maintain, and test the RH-TRU 72-B packaging. This Program Guidance standardizes instructions for all users. Users shall follow these instructions or equivalent approved instructions. Following these instructions assures that operations meet the requirements of the SARP.

Washington TRU Solutions LLC

2006-11-07T23:59:59.000Z

Note: This page contains sample records for the topic "waste disposition packaging" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


221

RH Packaging Program Guidance  

SciTech Connect (OSTI)

The purpose of this program guidance document is to provide the technical requirements for use, operation, inspection, and maintenance of the RH-TRU 72-B Waste Shipping Package (also known as the "RH-TRU 72-B cask") and directly related components. This document complies with the requirements as specified in the RH-TRU 72-B Safety Analysis Report for Packaging (SARP), and Nuclear Regulatory Commission (NRC) Certificate of Compliance (C of C) 9212. If there is a conflict between this document and the SARP and/or C of C, the C of C shall govern. The C of C states: "...each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application." It further states: "...each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP tasks the Waste Isolation Pilot Plant (WIPP) Management and Operating (M&O) Contractor with assuring the packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) §71.8, "Deliberate Misconduct." Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, "Packaging and Transportation of Radioactive Material," certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21, "Reporting of Defects and Noncompliance," regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations are conducted. This document details the instructions to be followed to operate, maintain, and test the RH-TRU 72-B packaging. This Program Guidance standardizes instructions for all users. Users shall follow these instructions or equivalent approved instructions. Following these instructions assures that operations meet the requirements of the SARP.

Washington TRU Solutions LLC

2008-01-12T23:59:59.000Z

222

SNM measurement uncertainites: potential impacts for materials disposition  

SciTech Connect (OSTI)

A discussion of nuclear material measurement uncertainties and impacts to the Materials Disposition (MD) Program is presented. Many of the options under consideration by the disposition program present new measurement challenges include significant material processing throughputs, a variety of material forms, unique waste streams, and difficult-to-measure matrices. There are also some questions regarding the ability to achieve International Atomic Energy Agency (IAEA) verification requirements and to achieve measurement uncertainties that are small enough to meet the IAEA loss detection goals. We present a detailed formalism for determining the measurement error for nondestructive assay systems applied to the MD Program, which is an essential component for planning the safeguards and security of these systems.

Fearey, B.L.; Burr, T.L.; Pickrell, M.M.

1996-09-01T23:59:59.000Z

223

The ultimate disposition of depleted uranium  

SciTech Connect (OSTI)

Significant amounts of the depleted uranium (DU) created by past uranium enrichment activities have been sold, disposed of commercially, or utilized by defense programs. In recent years, however, the demand for DU has become quite small compared to quantities available, and within the US Department of Energy (DOE) there is concern for any risks and/or cost liabilities that might be associated with the ever-growing inventory of this material. As a result, Martin Marietta Energy Systems, Inc. (Energy Systems), was asked to review options and to develop a comprehensive plan for inventory management and the ultimate disposition of DU accumulated at the gaseous diffusion plants (GDPs). An Energy Systems task team, under the chairmanship of T. R. Lemons, was formed in late 1989 to provide advice and guidance for this task. This report reviews options and recommends actions and objectives in the management of working inventories of partially depleted feed (PDF) materials and for the ultimate disposition of fully depleted uranium (FDU). Actions that should be considered are as follows. (1) Inspect UF{sub 6} cylinders on a semiannual basis. (2) Upgrade cylinder maintenance and storage yards. (3) Convert FDU to U{sub 3}O{sub 8} for long-term storage or disposal. This will include provisions for partial recovery of costs to offset those associated with DU inventory management and the ultimate disposal of FDU. Another recommendation is to drop the term tails'' in favor of depleted uranium'' or DU'' because the tails'' label implies that it is waste.'' 13 refs.

Not Available

1990-12-01T23:59:59.000Z

224

WASTE TO WATTS Waste is a Resource!  

E-Print Network [OSTI]

to Climate protection in light of the· Waste Framework Directive. The "energy package", e.g. the RenewablesWASTE TO WATTS Waste is a Resource! energy forum Case Studies from Estonia, Switzerland, Germany Bossart,· ABB Waste-to-Energy Plants Edmund Fleck,· ESWET Marcel van Berlo,· Afval Energie Bedrijf From

Columbia University

225

CH Packaging Program Guidance  

SciTech Connect (OSTI)

The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: "each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application." They further state: "each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP charges the Waste Isolation Pilot Plant (WIPP) management and operating (M&O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) §71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.

Washington TRU Solutions LLC

2005-02-28T23:59:59.000Z

226

Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors  

SciTech Connect (OSTI)

The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrain all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim storage, packaging, transportation, waste forms, waste treatment, decontamination and decommissioning issues; and low-level waste (LLW) and high-level waste (HLW) disposal.

Shropshire, D.E.; Herring, J.S.

2004-10-03T23:59:59.000Z

227

Fissile Materials Disposition | National Nuclear Security Administrati...  

National Nuclear Security Administration (NNSA)

uranium have become surplus to the defense needs of both the United States and Russia. The Office of Fissile Materials Disposition (FMD) plays an important role in...

228

Weapons Dismantlement and Disposition NNSS Capabilities  

SciTech Connect (OSTI)

The U.S. Department of Energy (DOE) has tasked the WDD working group to disposition the large inventory of legacy classified weapon components scattered across the complex.

Pat Arnold

2011-12-01T23:59:59.000Z

229

Summary - Major Risk Factors Integrated Facility Disposition...  

Office of Environmental Management (EM)

& ORNL, Oak Ridge, TN EM Project: Integrated Facility Disposition Project (IFDP) ETR Report Date: August 2008 ETR-15 United States Department of Energy Office of Environmental...

230

Personal Property Disposition - Community Reuse Organizations...  

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

Owen (signed) Director, Office of Worker and Community Transition Department of Energy Washington, DC 20505 January 22, 2003 Disposition of Excess Personal Property...

231

Romanian experience on packaging testing  

SciTech Connect (OSTI)

With more than twenty years ago, the Institute for Nuclear Research Pitesti (INR), through its Reliability and Testing Laboratory, was licensed by the Romanian Nuclear Regulatory Body- CNCAN and to carry out qualification tests [1] for packages intended to be used for the transport and storage of radioactive materials. Radioactive materials, generated by Romanian nuclear facilities [2] are packaged in accordance with national [3] and the IAEA's Regulations [1,6] for a safe transport to the disposal center. Subjecting these packages to the normal and simulating test conditions accomplish the evaluation and certification in order to prove the package technical performances. The paper describes the qualification tests for type A and B packages used for transport and storage of radioactive materials, during a period of 20 years of experience. Testing is used to substantiate assumption in analytical models and to demonstrate package structural response. The Romanian test facilities [1,3,6] are used to simulate the required qualification tests and have been developed at INR Pitesti, the main supplier of type A packages used for transport and storage of low radioactive wastes in Romania. The testing programme will continue to be a strong option to support future package development, to perform a broad range of verification and certification tests on radioactive material packages or component sections, such as packages used for transport of radioactive sources to be used for industrial or medical purposes [2,8]. The paper describes and contain illustrations showing some of the various tests packages which have been performed during certain periods and how they relate to normal conditions and minor mishaps during transport. Quality assurance and quality controls measures taken in order to meet technical specification provided by the design there are also presented and commented. (authors)

Vieru, G. [IAEA Technical Expert, Head, Reliability and Testing Lab., Institute for Nuclear Research (Romania)

2007-07-01T23:59:59.000Z

232

TRACKING SURPLUS PLUTONIUM FROM WEAPONS TO DISPOSITION  

SciTech Connect (OSTI)

Supporting nuclear nonproliferation and global security principles, beginning in 1994 the United States has withdrawn more than 50 metric tons (MT) of government-controlled plutonium from potential use in nuclear weapons. The Department of Energy (DOE), including the National Nuclear Security Administration, established protocols for the tracking of this "excess" and "surplus" plutonium, and for reconciling the current storage and utilization of the plutonium to show that its management is consistent with the withdrawal policies. Programs are underway to ensure the safe and secure disposition of the materials that formed a major part of the weapons stockpile during the Cold War, and growing quantities have been disposed as waste, after which they are not included in traditional nuclear material control and accountability (NMC&A) data systems. A combination of resources is used to perform the reconciliations that form the basis for annual reporting to DOE, to U.S. Department of State, and to international partners including the International Atomic Energy Agency.

Allender, J.; Beams, J.; Sanders, K.; Myers, L.

2013-07-16T23:59:59.000Z

233

Naval Petroleum Reserve No. 3 Disposition Decision Analysis and...  

Energy Savers [EERE]

Naval Petroleum Reserve No. 3 Disposition Decision Analysis and Timeline Naval Petroleum Reserve No. 3 Disposition Decision Analysis and Timeline This Report to Congress provides a...

234

Surplus Plutonium Disposition Final Environmental Impact Statement  

SciTech Connect (OSTI)

In December 1996, the U.S. Department of Energy (DOE) published the ''Storage and Disposition of Weapons-Usable Fissile Materials Final Programmatic Environmental Impact Statement (Storage and Disposition PEIS)'' (DOE 1996a). That PEIS analyzes the potential environmental consequences of alternative strategies for the long-term storage of weapons-usable plutonium and highly enriched uranium (HEU) and the disposition of weapons-usable plutonium that has been or may be declared surplus to national security needs. The Record of Decision (ROD) for the ''Storage and Disposition PEIS'', issued on January 14, 1997 (DOE 1997a), outlines DOE's decision to pursue an approach to plutonium disposition that would make surplus weapons-usable plutonium inaccessible and unattractive for weapons use. DOE's disposition strategy, consistent with the Preferred Alternative analyzed in the ''Storage and Disposition PEIS'', allows for both the immobilization of some (and potentially all) of the surplus plutonium and use of some of the surplus plutonium as mixed oxide (MOX) fuel in existing domestic, commercial reactors. The disposition of surplus plutonium would also involve disposal of both the immobilized plutonium and the MOX fuel (as spent nuclear fuel) in a potential geologic repository.

N /A

1999-11-19T23:59:59.000Z

235

ESTIMATING IMPURITIES IN SURPLUS PLUTONIUM FOR DISPOSITION  

SciTech Connect (OSTI)

The United States holds at least 61.5 metric tons (MT) of plutonium that is permanently excess to use in nuclear weapons programs, including 47.2 MT of weapons-grade plutonium. Surplus inventories will be stored safely by the Department of Energy (DOE) and then transferred to facilities that will prepare the plutonium for permanent disposition. The Savannah River National Laboratory (SRNL) operates a Feed Characterization program for the Office of Fissile Materials Disposition of the National Nuclear Security Administration and the DOE Office of Environmental Management. Many of the items that require disposition are only partially characterized, and SRNL uses a variety of techniques to predict the isotopic and chemical properties that are important for processing through the Mixed Oxide Fuel Fabrication Facility and alternative disposition paths. Recent advances in laboratory tools, including Prompt Gamma Analysis and Peroxide Fusion treatment, provide data on the existing inventories that will enable disposition without additional, costly sampling and destructive analysis.

Allender, J.; Moore, E.

2013-07-17T23:59:59.000Z

236

Waste Management in Dsseldorf Combination of separate collection,  

E-Print Network [OSTI]

Waste Management in Düsseldorf Combination of separate collection, recycling and waste-to-energy Biowaste Garden waste Light packaging Paper Glass Wood from bulky waste Bulky waste Rest / mixed waste Bio- Garden- Paper Glass Light Metals Wood Bulky Rest waste waste Card- Pack. waste board Saved CO2

Columbia University

237

Optimal segmentation and packaging process  

DOE Patents [OSTI]

A process for improving packaging efficiency uses three dimensional, computer simulated models with various optimization algorithms to determine the optimal segmentation process and packaging configurations based on constraints including container limitations. The present invention is applied to a process for decontaminating, decommissioning (D and D), and remediating a nuclear facility involving the segmentation and packaging of contaminated items in waste containers in order to minimize the number of cuts, maximize packaging density, and reduce worker radiation exposure. A three-dimensional, computer simulated, facility model of the contaminated items are created. The contaminated items are differentiated. The optimal location, orientation and sequence of the segmentation and packaging of the contaminated items is determined using the simulated model, the algorithms, and various constraints including container limitations. The cut locations and orientations are transposed to the simulated model. The contaminated items are actually segmented and packaged. The segmentation and packaging may be simulated beforehand. In addition, the contaminated items may be cataloged and recorded. 3 figs.

Kostelnik, K.M.; Meservey, R.H.; Landon, M.D.

1999-08-10T23:59:59.000Z

238

EIS-0283: Surplus Plutonium Disposition Environmental Impact Statement  

Broader source: Energy.gov [DOE]

This EIS analyzes the potential environmental impacts associated with alternatives for the disposition of surplus plutonium.

239

Ceremony Theming Packages Package One  

E-Print Network [OSTI]

will incur a surcharge of $50.00 per additional half hour Weddings at The University of Western Australia #12: _____________________________________________________________________________________________________ Weddings at The University of Western Australia #12;Ceremony Theming Packages Payment Details CREDIT CARD I, ______________________________________, authorise The University Club of Western Australia P/L, to debit the amount

Tobar, Michael

240

Proliferation resistance criteria for fissile material disposition  

SciTech Connect (OSTI)

The 1994 National Academy of Sciences study {open_quotes}Management and Disposition of Excess Weapons Plutonium{close_quotes} defined options for reducing the national and international proliferation risks of materials declared excess to the nuclear weapons program. This report proposes criteria for assessing the proliferation resistance of these options. The criteria are general, encompassing all stages of the disposition process from storage through intermediate processing to final disposition including the facilities, processing technologies and materials, the level of safeguards for these materials, and the national/subnational threat to the materials.

Close, D.A.; Fearey, B.L.; Markin, J.T.; Rutherford, D.A. [Los Alamos National Lab., NM (United States); Duggan, R.A.; Jaeger, C.D.; Mangan, D.L.; Moya, R.W.; Moore, L.R. [Sandia National Labs., Albuquerque, NM (United States); Strait, R.S. [Lawrence Livermore National Lab., CA (United States)

1995-04-01T23:59:59.000Z

Note: This page contains sample records for the topic "waste disposition packaging" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

RH Packaging Program Guidance  

SciTech Connect (OSTI)

The purpose of this program guidance document is to provide technical requirements for use, operation, inspection, and maintenance of the RH-TRU 72-B Waste Shipping Package and directly related components. This document complies with the requirements as specified in the RH-TRU 72-B Safety Analysis Report for Packaging (SARP), and Nuclear Regulatory Commission (NRC) Certificate of Compliance (C of C) 9212. If there is a conflict between this document and the SARP and/or C of C, the SARP and/or C of C shall govern. The C of C states: ''...each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, ''Operating Procedures,'' of the application.'' It further states: ''...each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, ''Acceptance Tests and Maintenance Program of the Application.'' Chapter 9.0 of the SARP tasks the Waste Isolation Pilot Plant (WIPP) Management and Operating (M&O) contractor with assuring the packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC approved, users need to be familiar with 10 CFR {section} 71.11, ''Deliberate Misconduct.'' Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. CBFO will evaluate the issue and notify the NRC if required. This document details the instructions to be followed to operate, maintain, and test the RH-TRU 72-B packaging. This Program Guidance standardizes instructions for all users. Users shall follow these instructions. Following these instructions assures that operations are safe and meet the requirements of the SARP. This document is available on the Internet at: ttp://www.ws/library/t2omi/t2omi.htm. Users are responsible for ensuring they are using the current revision and change notices. Sites may prepare their own document using the word-for-word steps in th is document, in sequence, including Notes and cautions. Site specific information may be included as necessary. The document, and revisions, must then be submitted to CBFO at sitedocuments@wipp.ws for approval. A copy of the approval letter from CBFO shall be available for audit purposes. Users may develop site-specific procedures addressing preoperational activities, quality assurance (QA), hoisting and rigging, and radiation health physics to be used with the instructions contained in this document. Users may recommend changes to this document by submitting their recommendations (in writing) to the WIPP M&O Contractor RH Packaging Maintenance Engineer for evaluation. If approved, the change(s) will be incorporated into this document for use by ALL users. Before first use and every 12 months after, user sites will be audited to this document to ensure compliance. They will also be audited within one year from the effective date of revisions to this document.

Washington TRU Solutions, LLC

2003-08-25T23:59:59.000Z

242

Rev August 2006 Radiation Safety Manual Section 14 Radioactive Waste  

E-Print Network [OSTI]

Rev August 2006 Radiation Safety Manual Section 14 ­ Radioactive Waste Page 14-1 Section 14 Radioactive Waste Contents A. Proper Collection, Disposal, and Packaging and Putrescible Animal Waste.........................14-8 a. Non-Radioactive Animal Waste

Wilcock, William

243

A HOLISTIC APPROACH FOR DISPOSITION OF LONG-LIVED RADIOACTIVE MATERIALS  

SciTech Connect (OSTI)

During the past 45 years, one of the most challenging scientific, engineering, socio-economic, and political tasks and obligations of our time has been to site and develop technical, politically acceptable, solutions to the safe disposition of long-lived radioactive materials (LLRMs). However, at the end of the year 2002, the Waste Isolation Pilot Plant (WIPP) site in the United States of America (USA) hosts the world's only operating LLRM-disposal system, which (1) is based on the LLRM-disposal principles recommended by the National Academy of Sciences (NAS) in 1957, i.e., deep geological disposal in a ''stable'' salt vault/repository, (2) complies with the nation's ''Environmental Radiation Protection Standards for the Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes'', and (3) may receive 175,584 cubic meters (m3) of transuranic radioactive waste (TRUW)a. Pending the scheduled opening of repositories for once-used nuclear fuel (OUNF) in the USA, Sweden, and Finland in the years 2010, 2015, and 2017, respectively, LLRM-disposal solutions remain the missing link in all national LLRM-disposition programs. Furthermore, for a variety of reasons, many nations with nuclear programs have chosen a ''spectator'' stance in terms of enhancing the global nuclear safety culture and the nuclear renaissance, and have either ''slow-tracked'' or deferred their LLRM-disposal programs to allow time for an informed national consensus to evolve based on LLRM-disposition experiences and solutions gained elsewhere. In the meantime, LLRMs will continue to amass in different types and levels of safeguarded storage facilities around the world. In an attempt to contribute to the enhancement of the global nuclear safety culture and the nuclear renaissance, the authors developed the sample holistic approach for synergistic disposition of LLRMs comprising LLRM-disposition components considered either ''proven'' or ''promising'' by the authors. The fundamental principles of the holistic approach are: (1) Risk minimization; (2) Minimization of the LLRM volume requiring deep geological disposal; and (3) LLRM-disposition flexibility. An integral element of these principles is to allow time for LLRM-disposition solutions to evolve/mature technically, financially, and politically. Furthermore, contingent upon the desired outcome(s), available financial, scientific, and technical resources, and political will, these components may be implemented separately or in combinations by one or a group of nations.

Eriksson, Leif G.; Dials, George E.; Parker, Frank L.

2003-02-27T23:59:59.000Z

244

Remaining Sites Verification Package for the 331 Life Sciences Laboratory Drain Field Septic System, Waste Site Reclassification Form 2008-020  

SciTech Connect (OSTI)

The 331 Life Sciences Laboratory Drain Field (LSLDF) septic system waste site consists of a diversion chamber, two septic tanks, a distribution box, and a drain field. This septic system was designed to receive sanitary waste water, from animal studies conducted in the 331-A and 331-B Buildings, for discharge into the soil column. However, field observations and testing suggest the 331 LSLDF septic system did not receive any discharges. In accordance with this evaluation, the confirmatory sampling results support a reclassification of the 331 LSLDF waste site to No Action. This site does not have a deep zone or other condition that would warrant an institutional control in accordance with the 300-FF-2 ROD under the industrial land use scenario.

J. M. Capron

2008-10-16T23:59:59.000Z

245

Geochemical Processes Data Package for the Vadose Zone in the Single-Shell Tank Waste Management Areas at the Hanford Site  

SciTech Connect (OSTI)

This data package discusses the geochemistry of vadose zone sediments beneath the single-shell tank farms at the U.S. Department of Energy’s (DOE’s) Hanford Site. The purpose of the report is to provide a review of the most recent and relevant geochemical process information available for the vadose zone beneath the single-shell tank farms and the Integrated Disposal Facility. Two companion reports to this one were recently published which discuss the geology of the farms (Reidel and Chamness 2007) and groundwater flow and contamination beneath the farms (Horton 2007).

Cantrell, Kirk J.; Zachara, John M.; Dresel, P. Evan; Krupka, Kenneth M.; Serne, R. Jeffrey

2007-09-28T23:59:59.000Z

246

Assessment of the impacts of spent fuel disassembly alternatives on the Nuclear Waste Isolation System. [Preparing and packaging spent fuel assemblies for geologic disposal  

SciTech Connect (OSTI)

The objective of this report was to evaluate four possible alternative methods of preparing and packaging spent fuel assemblies for geologic disposal against the Reference Process of unmodified spent fuel. The four alternative processes were: (1) End fitting removal, (2) Fission gas venting and resealing, (3) Fuel bundle disassembly and close packing of fuel pins, and (4) Fuel shearing and immobilization. Systems analysis was used to develop a basis of comparison of the alternatives. Conceptual processes and facility layouts were devised for each of the alternatives, based on technology deemed feasible for the purpose. Assessments were made of 15 principal attributes from the technical, operational, safety/risk, and economic considerations related to each of the alternatives, including both the surface packaging and underground repository operations. Specific attributes of the alternative processes were evaluated by assigning a number for each that expressed its merit relative to the corresponding attribute of the Reference Process. Each alternative process was then ranked by summing the numbers for attributes in each of the four assessment areas and collectively. Fuel bundle disassembly and close packing of fuel pins was ranked the preferred method of disposal of spent fuel. 63 references, 46 figures, 46 tables.

Not Available

1984-07-01T23:59:59.000Z

247

LLNL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement  

SciTech Connect (OSTI)

The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of Fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO{sub 2} and UO{sub 2}), typically containing 95% or more UO{sub 2}. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. LLNL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO{sub 2} powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within a Category 1 area. Building 332 will be used to receive and store the bulk PuO{sub 2} powder, fabricate MOX fuel pellets, and assemble fuel rods. Building 334 will be used to assemble, store, and ship fuel bundles. Only minor modifications would be required of Building 332. Uncontaminated glove boxes would need to be removed, petition walls would need to be removed, and minor modifications to the ventilation system would be required.

O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

1998-08-01T23:59:59.000Z

248

Hanford Tank Waste - Near Source Treatment of Low Activity Waste  

SciTech Connect (OSTI)

Abstract only. Treatment and disposition of Hanford Site waste as currently planned consists of 100+ waste retrievals, waste delivery through up to 8+ miles of dedicated, in-ground piping, centralized mixing and blending operations- all leading to pre-treatment combination and separation processes followed by vitrification at the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The sequential nature of Tank Farm and WTP operations requires nominally 15-20 years of continuous operations before all waste can be retrieved from many Single Shell Tanks (SSTs). Also, the infrastructure necessary to mobilize and deliver the waste requires significant investment beyond that required for the WTP. Treating waste as closely as possible to individual tanks or groups- as allowed by the waste characteristics- is being investigated to determine the potential to 1) defer, reduce, and/or eliminate infrastructure requirements, and 2) significantly mitigate project risk by reducing the potential and impact of single point failures. The inventory of Hanford waste slated for processing and disposition as LAW is currently managed as high-level waste (HLW), i.e., the separation of fission products and other radionuclides has not commenced. A significant inventory of this waste (over 20M gallons) is in the form of precipitated saltcake maintained in single shell tanks, many of which are identified as potential leaking tanks. Retrieval and transport (as a liquid) must be staged within the waste feed delivery capability established by site infrastructure and WTP. Near Source treatment, if employed, would provide for the separation and stabilization processing necessary for waste located in remote farms (wherein most of the leaking tanks reside) significantly earlier than currently projected. Near Source treatment is intended to address the currently accepted site risk and also provides means to mitigate future issues likely to be faced over the coming decades. This paper describes the potential near source treatment and waste disposition options as well as the impact these options could have on reducing infrastructure requirements, project cost and mission schedule.

Ramsey, William Gene

2013-08-15T23:59:59.000Z

249

NRC comprehensive records disposition schedule. Revision 3  

SciTech Connect (OSTI)

Title 44 US Code, ``Public Printing and Documents,`` regulations issued by the General Service Administration (GSA) in 41 CFR Chapter 101, Subchapter B, ``Management and Use of Information and Records,`` and regulations issued by the National Archives and Records Administration (NARA) in 36 CFR Chapter 12, Subchapter B, ``Records Management,`` require each agency to prepare and issue a comprehensive records disposition schedule that contains the NARA approved records disposition schedules for records unique to the agency and contains the NARA`s General Records Schedules for records common to several or all agencies. The approved records disposition schedules specify the appropriate duration of retention and the final disposition for records created or maintained by the NRC. NUREG-0910, Rev. 3, contains ``NRC`s Comprehensive Records Disposition Schedule,`` and the original authorized approved citation numbers issued by NARA. Rev. 3 incorporates NARA approved changes and additions to the NRC schedules that have been implemented since the last revision dated March, 1992, reflects recent organizational changes implemented at the NRC, and includes the latest version of NARA`s General Records Schedule (dated August 1995).

NONE

1998-02-01T23:59:59.000Z

250

Remaining Sites Verification Package for the 100-F-26:13, 108-F Drain Pipelines, Waste Site Reclassification Form 2005-011  

SciTech Connect (OSTI)

The 100-F-26:13 waste site is the network of process sewer pipelines that received effluent from the 108-F Biological Laboratory and discharged it to the 188-F Ash Disposal Area (126-F-1 waste site). The pipelines included one 0.15-m (6-in.)-, two 0.2-m (8-in.)-, and one 0.31-m (12-in.)-diameter vitrified clay pipe segments encased in concrete. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

L. M. Dittmer

2008-03-03T23:59:59.000Z

251

Remaining Sites Verification Package for the 116-F-8, 1904-F Outfall Structure and the 100-F-42, 1904-F Spillway, Waste Site Reclassification Form 2006-038  

SciTech Connect (OSTI)

The 116-F-8 waste site is the former 1904-F Outfall Structure used to discharge reactor cooling water effluent fro mthe 107-F Retention Basin to the Columbia River. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

L. M. Dittmer

2006-09-25T23:59:59.000Z

252

U.S. and Russia Sign Plutonium Disposition Agreement | National...  

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

Our Jobs Working at NNSA Blog Home About Us Our History NNSA Timeline U.S. and Russia Sign Plutonium Disposition Agreement U.S. and Russia Sign Plutonium Disposition...

253

UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD  

E-Print Network [OSTI]

the Department of Energy's (DOE) work related to the packaging and transport of such waste. Consistent with its and waste package performance. In light of continuing technical challenges, the Board believes that the DOEcon144vf UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD 2300 Clarendon Boulevard, Suite 1300

254

Los Alamos National Laboratory Waste Management Program  

SciTech Connect (OSTI)

Los Alamos National Laboratory's (LANL) waste management program is responsible for disposition of waste generated by many of the LANL programs and operations. LANL generates liquid and solid waste that can include radioactive, hazardous, and other constituents. Where practical, LANL hazardous and mixed wastes are disposed through commercial vendors; low-level radioactive waste (LLW) and radioactive asbestos-contaminated waste are disposed on site at LANL's Area G disposal cells, transuranic (TRU) waste is disposed at the Waste Isolation Pilot Plant (WIPP), and high-activity mixed wastes are disposed at the Nevada Test Site (NTS) after treatment by commercial vendors. An on-site radioactive liquid waste treatment facility (RLWTF) removes the radioactive constituents from liquid wastes and treated water is released through an NPDES permitted outfall. LANL has a very successful waste minimization program. Routine hazardous waste generation has been reduced over 90% since 1993. LANL has a DOE Order 450.1-compliant environmental management system (EMS) that is ISO 14001 certified; waste minimization is integral to setting annual EMS improvement objectives. Looking forward, under the new LANL management and operating contractor, Los Alamos National Security (LANS) LLC, a Zero Liquid Discharge initiative is being planned that should eliminate flow to the RLWTF NPDES-permitted outfall. The new contractor is also taking action to reduce the number of permitted waste storage areas, to charge generating programs directly for the cost to disposition waste, and to simplify/streamline the waste system. (authors)

Lopez-Escobedo, G.M.; Hargis, K.M.; Douglass, C.R. [Los Alamos National Laboratory, NM (United States)

2007-07-01T23:59:59.000Z

255

Remaining Sites Verification Package for the 100-F-33, 146-F Aquatic Biology Fish Ponds, Waste Site Reclassification Form 2006-021  

SciTech Connect (OSTI)

The 100-F-33, 146-F Aquatice Biology Fish Ponds waste site was an area with six small rectangular ponds and one large circular pond used to conduct tests on fish using various mixtures of river and reactor effluent water. The current site conditions achieve the remedial action objectives specified in the Remaining Sites ROD. The results of verification and applicable confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

L. M. Dittmer

2006-08-25T23:59:59.000Z

256

Remaining Sites Verification Package for the 126-B-3, 184-B Coal Pit Dumping Area, Waste Site Reclassification Form 2005-028  

SciTech Connect (OSTI)

The 126-B-3 waste site is the former coal storage pit for the 184-B Powerhouse. During demolition operations in the 1970s, the site was used for disposal of demolition debris from 100-B/C Area facilities. The site has been remediated by removing debris and contaminated soils. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

L. M. Dittmer

2006-08-07T23:59:59.000Z

257

Remaining Sites Verification Package for the 116-F-8, 1904-F Outfall Structure and the 100-F-42, 1904-F Spillway, Waste Site Reclassification Form 2006-045  

SciTech Connect (OSTI)

The 100-F-42 waste site is the portion of the former emergency overflow spillway for the 1904-F Outfall Structure formerly existing above the ordinary high water mark of the Columbia River. The spillway consisted of a concrete flume designed to discharge effluent from the 107-F Retention Basin in the event that flows could not be completely discharged via the river outfall pipelines. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

L. M. Dittmer

2006-09-26T23:59:59.000Z

258

Remaining Sites Verification Package for 132-H-1, 116-H Reactor Stack Burial Site, Waste Site Reclassification Form 2006-053  

SciTech Connect (OSTI)

The 132-H-1 waste site includes the 116-H exhaust stack burial trench and the buried stack foundation (which contains an embedded vertical 15-cm (6-in) condensate drain line). The 116-H reactor exhaust stack and foundation were decommissioned and demolished using explosives in 1983, with the rubble buried in situ beneath clean fill at least 1 m (3.3 ft) thick. Residual concentrations support future land uses that can be represented by a rural-residential scenario and pose no threat to groundwater or the Columbia River based on RESRAD modeling.

L. M. Dittmer

2007-06-26T23:59:59.000Z

259

Remaining Sites Verification Package for the 1607-F1 Sanitary Sewer System (124-F-1) and the 100-F-26:8 (1607-F1) Sanitary Sewer Pipelines Waste Sites, Waste Site Reclassification Form 2005-004  

SciTech Connect (OSTI)

The 100-F-26:8 waste site consisted of the underground pipelines that conveyed sanitary waste water from the 1701-F Gatehouse, 1709-F Fire Station, and the 1720-F Administrative Office to the 1607-F1 septic tank. The site has been remediated and presently exists as an open excavation. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

L. M. Dittmer

2008-03-14T23:59:59.000Z

260

Cost Estimation Package  

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

This chapter focuses on the components (or elements) of the cost estimation package and their documentation.

1997-03-28T23:59:59.000Z

Note: This page contains sample records for the topic "waste disposition packaging" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


261

Remaining Sites Verification Package for the 100-F-46, 119-F Stack Sampling French Drain, Waste Site Reclassification Form 2008-021  

SciTech Connect (OSTI)

The 100-F-46 french drain consisted of a 1.5 to 3 m long, vertically buried, gravel-filled pipe that was approximately 1 m in diameter. Also included in this waste site was a 5 cm cast-iron pipeline that drained condensate from the 119-F Stack Sampling Building into the 100-F-46 french drain. In accordance with this evaluation, the confirmatory sampling results support a reclassification of this site to No Action. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

J. M. Capron

2008-08-08T23:59:59.000Z

262

Remaining Sites Verification Package for the 100-C-9:1 Main Process Sewer Collection Line, Waste Site Reclassification Form 2004-012  

SciTech Connect (OSTI)

The 100-C-9:1 main process sewer pipeline, also known as the twin box culvert, was a dual reinforced process sewer that collected process effluent from the 183-C and 190-C water treatment facilities, discharging at the 132-C-2 Outfall. For remedial action purposes, the 100-C-9:1 waste site was subdivided into northern and southern sections. The 100-C-9:1 subsite has been remediated to achieve the remedial action objectives specified in the Remaining Sites ROD. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

L. M. Dittmer

2007-06-11T23:59:59.000Z

263

Remaining Sites Verification Package for the 100-B-18, 184-B Powerhouse Debris Pile, Waste Site Reclassification Form 2007-020  

SciTech Connect (OSTI)

The 100-B-18 Powerhouse Debris Pile contained miscellaneous demolition waste from the decommissioning activities of the 184-B Powerhouse. The debris covered an area roughly 15 m by 30 m and included materials such as concrete blocks, mixed aggregate/concrete slabs, stone rubble, asphalt rubble, traces of tar/coal, broken fluorescent lights, brick chimney remnants, and rubber hoses. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

L. M. Dittmer

2007-11-30T23:59:59.000Z

264

Study of plutonium disposition using the GE Advanced Boiling Water Reactor (ABWR)  

SciTech Connect (OSTI)

The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the U.S. to disposition 50 to 100 metric tons of excess of plutonium in parallel with a similar program in Russia. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing long-term diversion resistance to this material. The NAS study {open_quotes}Management and Disposition of Excess Weapons Plutonium{close_quotes} identified light water reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a U.S. disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a 1350 MWe GE Advanced Boiling Water Reactor (ABWR) is utilized to convert the plutonium to spent fuel. The ABWR represents the integration of over 30 years of experience gained worldwide in the design, construction and operation of BWRs. It incorporates advanced features to enhance reliability and safety, minimize waste and reduce worker exposure. For example, the core is never uncovered nor is any operator action required for 72 hours after any design basis accident. Phase 1 of this study was documented in a GE report dated May 13, 1993. DOE`s Phase 1 evaluations cited the ABWR as a proven technical approach for the disposition of plutonium. This Phase 2 study addresses specific areas which the DOE authorized as appropriate for more in-depth evaluations. A separate report addresses the findings relative to the use of existing BWRs to achieve the same goal.

NONE

1994-04-30T23:59:59.000Z

265

Radioactive material package seal tests  

SciTech Connect (OSTI)

General design or test performance requirements for radioactive materials (RAM) packages are specified in Title 10 of the US Code of Federal Regulations Part 71 (US Nuclear Regulatory Commission, 1983). The requirements for Type B packages provide a broad range of environments under which the system must contain the RAM without posing a threat to health or property. Seals that provide the containment system interface between the packaging body and the closure must function in both high- and low-temperature environments under dynamic and static conditions. A seal technology program, jointly funded by the US Department of Energy Office of Environmental Restoration and Waste Management (EM) and the Office of Civilian Radioactive Waste Management (OCRWM), was initiated at Sandia National Laboratories. Experiments were performed in this program to characterize the behavior of several static seal materials at low temperatures. Helium leak tests on face seals were used to compare the materials. Materials tested include butyl, neoprene, ethylene propylene, fluorosilicone, silicone, Eypel, Kalrez, Teflon, fluorocarbon, and Teflon/silicone composites. Because most elastomer O-ring applications are for hydraulic systems, manufacturer low-temperature ratings are based on methods that simulate this use. The seal materials tested in this program with a fixture similar to a RAM cask closure, with the exception of silicone S613-60, are not leak tight (1.0 {times} 10{sup {minus}7} std cm{sup 3}/s) at manufacturer low-temperature ratings. 8 refs., 3 figs., 1 tab.

Madsen, M.M.; Humphreys, D.L.; Edwards, K.R.

1990-01-01T23:59:59.000Z

266

ANL-W MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement  

SciTech Connect (OSTI)

The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement (EIS). This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO{sub 2} and UO{sub 2}), typically containing 95% or more UO{sub 2}. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. The paper describes the following: Site map and the LA facility; process descriptions; resource needs; employment requirements; wastes, emissions, and exposures; accident analysis; transportation; qualitative decontamination and decommissioning; post-irradiation examination; LA fuel bundle fabrication; LA EIS data report assumptions; and LA EIS data report supplement.

O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

1997-08-01T23:59:59.000Z

267

Disposition and transportation of surplus radioactive low specific activity nitric acid. Volume 1, Environmental Assessment  

SciTech Connect (OSTI)

DOE is deactivating the PUREX plant at Hanford; this will involve the disposition of about 692,000 liters (183,000 gallons) of surplus nitric acid contaminated with low levels of U and other radionuclides. The nitric acid, designated as low specific activity, is stored in 4 storage tanks at PUREX. Five principal alternatives were evaluated: transfer for reuse (sale to BNF plc), no action, continued storage in Hanford upgraded or new facility, consolidation of DOE surplus acid, and processing the LSA nitric acid as waste. The transfer to BNF plc is the preferred alternative. From the analysis, it is concluded that the proposed disposition and transportation of the acid does not constitute a major federal action significantly affecting the quality of the human environment within the meaning of NEPA; therefore an environmental impact statement is not required.

NONE

1995-05-01T23:59:59.000Z

268

ICDF Complex Operations Waste Management Plan  

SciTech Connect (OSTI)

This Waste Management Plan functions as a management and planning tool for managing waste streams generated as a result of operations at the Idaho CERCLA Disposal Facility (ICDF) Complex. The waste management activities described in this plan support the selected remedy presented in the Waste Area Group 3, Operable Unit 3-13 Final Record of Decision for the operation of the Idaho CERCLA Disposal Facility Complex. This plan identifies the types of waste that are anticipated during operations at the Idaho CERCLA Disposal Facility Complex. In addition, this plan presents management strategies and disposition for these anticipated waste streams.

W.M. Heileson

2006-12-01T23:59:59.000Z

269

Reactor-Based Plutonium Disposition: Opportunities, Options, and Issues  

SciTech Connect (OSTI)

The end of the Cold War has created a legacy of surplus fissile materials (plutonium and highly enriched uranium) in the United States (U.S.) and the former Soviet Union. These materials pose a danger to national and international security. During the past few years, the U.S. and Russia have engaged in an ongoing dialog concerning the safe storage and disposition of surplus fissile material stockpiles. In January 1997, the Department of Energy (DOE) announced the U. S. would pursue a dual track approach to rendering approximately 50 metric tons of plutonium inaccessible for use in nuclear weapons. One track involves immobilizing the plutonium by combining it with high-level radioactive waste in glass or ceramic ''logs''. The other method, referred to as reactor-based disposition, converts plutonium into mixed oxide (MOX) fuel for nuclear reactors. The U.S. and Russia are moving ahead rapidly to develop and demonstrate the technology required to implement the MOX option in their respective countries. U.S. MOX fuel research and development activities were started in the 1950s, with irradiation of MOX fuel rods in commercial light water reactors (LWR) from the 1960s--1980s. In all, a few thousand MOX fuel rods were successfully irradiated. Though much of this work was performed with weapons-grade or ''near'' weapons-grade plutonium--and favorable fuel performance was observed--the applicability of this data for licensing and use of weapons-grade MOX fuel manufactured with modern fuel fabrication processes is somewhat limited. The U.S. and Russia are currently engaged in an intensive research, development, and demonstration program to support implementation of the MOX option in our two countries. This paper focuses on work performed in the U.S. and provides a brief summary of joint U.S./Russian work currently underway.

Greene, S.R.

1999-07-17T23:59:59.000Z

270

Relationship of Surplus Plutonium Disposition Alternatives and...  

National Nuclear Security Administration (NNSA)

Valley Authority WIPP Waste Isolation Pilot Plant WSB Waste Solidification Building Note: Appendices C and D provide details about the analyses of human health effects...

271

Borehole Data Package for RCRA Well 299-W22-47 at Single-Shell Tank Waste Management Area S-SX, Hanford Site, Washington  

SciTech Connect (OSTI)

One new Resource Conservation and Recovery Act (RCRA) groundwater assessment well was installed at single-shell tank Waste Management Area (WMA) S-SX in fiscal year (FY) 2005 to fulfill commitments for well installations proposed in Hanford Federal Facility Agreement and Consent Order, Milestone M-24-57 (2004). The need for the new well, well 299-W22-47, was identified during a data quality objectives process for establishing a RCRA/ Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA)/Atomic Energy Act (AEA) integrated 200 West and 200 East Area Groundwater Monitoring Network. This document provides a compilation of all available geologic data, spectral gamma ray logs, hydrogeologic data and well information obtained during drilling, well construction, well development, pump installation, aquifer testing, and sample collection/analysis activities. Appendix A contains the Well Summary Sheets, the Well Construction Summary Report, the geologist's Borehole Log, well development and pump installation records, and well survey results. Appendix B contains analytical results from groundwater samples collected during drilling. Appendix C contains complete spectral gamma ray logs and borehole deviation surveys.

Horton, Duane G.; Chamness, Mickie A.

2006-04-17T23:59:59.000Z

272

Plutonium stabilization and disposition focus area, FY 1999 and FY 2000 multi-year program plan  

SciTech Connect (OSTI)

Consistent with the Environmental Management`s (EM`s) plan titled, ``Accelerating Cleanup: Paths to Closure``, and ongoing efforts within the Executive Branch and Congress, this Multi-Year Program Plan (MYPP) for the Plutonium Focus Area was written to ensure that technical gap projects are effectively managed and measured. The Plutonium Focus Area (PFA) defines and manages technology development programs that contribute to the effective stabilization of nuclear materials and their subsequent safe storage and final disposition. The scope of PFA activities includes the complete spectrum of plutonium materials, special isotopes, and other fissile materials. The PFA enables solutions to site-specific and complex-wide technology issues associated with plutonium remediation, stabilization, and preparation for disposition. The report describes the current technical activities, namely: Plutonium stabilization (9 studies); Highly enriched uranium stabilization (2 studies); Russian collaboration program (2 studies); Packaging and storage technologies (6 studies); and PFA management work package/product line (3 studies). Budget information for FY 1999 and FY 2000 is provided.

NONE

1998-03-01T23:59:59.000Z

273

Draft Environmental Assessment on the Remote-handled Waste Disposition  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: Vegetation Proposed Newcatalyst phasesDataTranslocationDiurnal CycleDonald1Research:Project available for

274

Hanford Tank Waste Retrieval, Treatment and Disposition Framework |  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Career Scientists'Montana.ProgramJulietip sheetK-4In 2013 many autoThis roadFour Months Ahead

275

WASTE DISPOSITION PROJECT MAKES GREAT STRIDES AT THE IDAHO SITE |  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergyENERGYWomen Owned SmallOf The 2012Nuclear GuideReportVictor KaneContractfromPurchase

276

Low Level Waste Disposition - Quantity and Inventory | Department of  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:YearRound-UpHeatMulti-Dimensionalthe10IO1OP001Long-Term Storage ofEnergy

277

Performance Based Trending of Waste Disposition Project Data Quality |  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:Year in3.pdfEnergyDepartment ofOil's ImpactAppendix Al StatementChallenges |Reactive3

278

EM Waste and Materials Disposition & Transportation | Department of Energy  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power SystemsResources DOE ZeroThreeEnergyDepartment0: DOE512:Shines with Five DOE

279

Disposal Systems Evaluations and Tool Development - Engineered Barrier System Evaluation (Work Package LL1015080425)  

SciTech Connect (OSTI)

The Disposal Systems Evaluation Framework (DSEF) will use a logical process for developing one or more disposal system concepts (also referred to as repository system in this report) for any given waste form and geologic setting combination. In the Features, Events, and Processes (FEPs) group of work packages, there are seven categories of waste forms and eight categories of geologic setting being studied. The DSEF will also establish a Used Fuel Disposition Campaign (UFDC) knowledge management system to organize high-level information, data, and assumptions, thereby facilitating consistency in high-level system simulation and economic analyses. This system likely will be housed with the INL-based documentation system. Attention is given to lessons oearned from the systems used at the Waste Isolation Pilot Plant (WIPP) and the Yucca Mountain Project (YMP). Where reference material from other programs (e.g., international) is used or cited, the knowledge-management system imports the reference material directly or refer to it in bibliography form. Alternative data sets (e.g., from other programs) will also be utilized to evaluate their influence on DSEF analyses for given waste form and disposal-system combinations. The knowledge-management system can also be used to maintain the results of DSEF realizations, enabling the comparison and ranking of various waste-form/disposal-system-environment/disposal-system-design options. Finally, the UFDC knowledge-management system will be able to provide a compendium of 'templates' that can be utilized, in a labor-efficient fashion, to build parallel DSEF analyses (e.g., 'one offs'). The DSEF will not be a stand-alone, push-the-button and wait for the results, item of software. it will use osftware (probably EXCEL, initially), to guide the team members through a logical process of evaluating combinations of waste-form, disposal-syste-environment, and disposal-system design. In later stages, it will utilize software developed in the field of knowledge engineering and knowledge-management systems (Umeki et al. 2009). At certain points in the logical process, the DSEF software will point the evaluate to other software tools to do analyses needed to move the process forward. In the development of the DSEF, they will be mindful to make it no more complex than necessary to evaluate the system being considered. The DSEF will organize and document the work such that multiple realizations for different combinations can be compared and contrasted.

Blink, J A; Buscheck, T A; Halsey, W G; Wolery, T

2010-03-19T23:59:59.000Z

280

TRU waste certification compliance requirements for contact-handled wastes retrieved from storage for shipment to the WIPP  

SciTech Connect (OSTI)

Compliance requirements are presented for certifying that unclassified, contact-handled (CH) transuranic (TRU) solid wastes retrieved from storage at DOE sites meet the Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria (WAC). All applicable DOE Orders must continue to be met. The compliance requirements for certified waste retrieved from certified storage are addressed in another document. The compliance requirements are divided into four sections, primarily determined by the general feature that the requirements address. These sections are General Requirements, Waste Container Requirements, Waste Form Requirements, and Waste Package Requirements. The waste package is the combination of waste container and waste.

Not Available

1982-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "waste disposition packaging" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


281

Waste Management Information System (WMIS) User Guide  

SciTech Connect (OSTI)

This document provides the user of the Waste Management Information System (WMIS) instructions on how to use the WMIS software. WMIS allows users to initiate, track, and close waste packages. The modular design supports integration and utilization of data throuh the various stages of waste management. The phases of the waste management work process include generation, designation, packaging, container management, procurement, storage, treatment, transportation, and disposal.

R. E. Broz

2008-12-22T23:59:59.000Z

282

The ultimate disposition of depleted uranium  

SciTech Connect (OSTI)

Depleted uranium (DU) is produced as a by-product of the uranium enrichment process. Over 340,000 MTU of DU in the form of UF{sub 6} have been accumulated at the US government gaseous diffusion plants and the stockpile continues to grow. An overview of issues and objectives associated with the inventory management and the ultimate disposition of this material is presented.

Lemons, T.R. [Uranium Enrichment Organization, Oak Ridge, TN (United States)

1991-12-31T23:59:59.000Z

283

Remaining Sites Verification Package for the 600-111, P-11 Critical Mass Laboratory Crib, and UPR-600-16, Fire and Contamination Spread Waste Sites, Waste Site Reclassification Form 2004-065  

SciTech Connect (OSTI)

The 600-111, P-11 Critical Mass Laboratory Crib waste site, also referred to as the P-11 Facility, included the 120 Experimental Building, the 123 Control Building, and the P-11 Crib. The facility was constructed in 1949 and was used as a laboratory for plutonium criticality studies. In accordance with this evaluation, the confirmatory and verification sampling results support a reclassification of this site to Interim Closed Out. The results of confirmatory and verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

J. M. Capron

2008-10-28T23:59:59.000Z

284

Remaining Sites Verification Package for the 1607-F1 Sanitary Sewer System (124-F-1) and the 100-F-26:8 (1607-F1) Sanitary Sewer Pipelines Waste Sites, Waste Site Reclassification Form 2004-130  

SciTech Connect (OSTI)

The 1607-F1 Sanitary Sewer System (124-F-1), consisted of a septic tank, drain field, and associated pipelines that received sanitary waste water from the 1701-F Gatehouse, 1709-F Fire Station, and the 1720-F Administrative Office via the 100-F-26:8 pipelines. The septic tank required remedial action based on confirmatory sampling. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

L. M. Dittmer

2008-03-14T23:59:59.000Z

285

Global Nuclear Energy Partnership Waste Treatment Baseline  

SciTech Connect (OSTI)

The Global Nuclear Energy Partnership program (GNEP) is designed to demonstrate a proliferation-resistant and sustainable integrated nuclear fuel cycle that can be commercialized and used internationally. Alternative stabilization concepts for byproducts and waste streams generated by fuel recycling processes were evaluated and a baseline of waste forms was recommended for the safe disposition of waste streams. Waste forms are recommended based on the demonstrated or expected commercial practicability and technical maturity of the processes needed to make the waste forms, and performance of the waste form materials when disposed. Significant issues remain in developing technologies to process some of the wastes into the recommended waste forms, and a detailed analysis of technology readiness and availability may lead to the choice of a different waste form than what is recommended herein. Evolving regulations could also affect the selection of waste forms.

Dirk Gombert; William Ebert; James Marra; Robert Jubin; John Vienna

2008-05-01T23:59:59.000Z

286

Request For Records Disposition Authority-Nuclear Weapons | Department...  

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

Authority-Nuclear Weapons Request For Records Disposition Authority-Nuclear Weapons This document identifies the nuclear weapon records generated by the Department of Energy's...

287

PROGRESS IN REDUCING THE NUCLEAR THREAT: UNITED STATES PLUTONIUM CONSOLIDATION AND DISPOSITION  

SciTech Connect (OSTI)

Following the end of the Cold War, the United States identified 61.5 metric tons (MT) of plutonium and larger quantities of enriched uranium that are permanently excess to use in nuclear weapons programs. The Department of Energy (DOE) also began shutting down, stabilizing, and removing inventories from production facilities that were no longer needed to support weapons programs and non-weapons activities. The storage of 'Category I' nuclear materials at Rocky Flats, Sandia National Laboratories, and several smaller sites has been terminated to reduce costs and safeguards risks. De-inventory continues at the Hanford site and the Lawrence Livermore National Laboratory. Consolidation of inventories works in concert with the permanent disposition of excess inventories, including several tonnes of plutonium that have already been disposed to waste repositories and the preparation for transfers to the planned Mixed Oxide (MOX) Fuel Fabrication Facility (for the bulk of the excess plutonium) and alternative disposition methods for material that cannot be used readily in the MOX fuel cycle. This report describes status of plutonium consolidation and disposition activities and their impacts on continuing operations, particularly at the Savannah River Site.

Allender, J.; Koenig, R.; Davies, S.

2009-06-01T23:59:59.000Z

288

Development of an alternate pathway for materials destined for disposition to WIPP  

SciTech Connect (OSTI)

The Los Alamos National Laboratory currently has an inventory of process residues that may be viable candidates for disposition to the Waste Isolation Pilot Project (WIPP) located at Carlsbad, New Mexico. A recent 'Attractiveness Level D' exemption allows for the discard of specified intractable materials regardless of the percent plutonium. However, the limits with respect to drum loadings must be met. Cementation is a key component of the aqueous nitrate flowsheet and serves as a 'bleed-off' stream for impurities separated from the plutonium during processing operations. The main 'feed' to the cementation operations are the 'bottoms' from the evaporation process. In the majority of cases, the cemented bottoms contain less than the allowed amount per drum for WIPP acceptance. This project would expand the route to WIPP for items that have no defined disposition path, are difficult to process, have been through multiple passes, have no current recovery operations available to recover the plutonium and that are amenable to cementation. This initial work will provide the foundation for a full scale disposition pathway of the candidate materials. Once the pathway has been expanded and a cementation matrix developed, routine discard activities will be initiated.

Ayers, Georgette Y [Los Alamos National Laboratory; Mckerley, Bill [Los Alamos National Laboratory; Veazey, Gerald W [Los Alamos National Laboratory; Ricketts, Thomas E [Los Alamos National Laboratory

2010-01-01T23:59:59.000Z

289

Cleanup Verification Package for the 118-F-6 Burial Ground  

SciTech Connect (OSTI)

This cleanup verification package documents completion of remedial action for the 118-F-6 Burial Ground located in the 100-FR-2 Operable Unit of the 100-F Area on the Hanford Site. The trenches received waste from the 100-F Experimental Animal Farm, including animal manure, animal carcasses, laboratory waste, plastic, cardboard, metal, and concrete debris as well as a railroad tank car.

H. M. Sulloway

2008-10-02T23:59:59.000Z

290

Idaho Nuclear Technology and Engineering Center (INTEC) Sodium Bearing Waste - Waste Incidental to Reprocessing Determination  

SciTech Connect (OSTI)

U.S. Department of Energy Manual 435.1-1, Radioactive Waste Management, Section I.1.C, requires that all radioactive waste subject to Department of Energy Order 435.1 be managed as high-level radioactive waste, transuranic waste, or low-level radioactive waste. Determining the radiological classification of the sodium-bearing waste currently in the Idaho Nuclear Technology and Engineering Center Tank Farm Facility inventory is important to its proper treatment and disposition. This report presents the technical basis for making the determination that the sodium-bearing waste is waste incidental to spent fuel reprocessing and should be managed as mixed transuranic waste. This report focuses on the radiological characteristics of the sodiumbearing waste. The report does not address characterization of the nonradiological, hazardous constituents of the waste in accordance with Resource Conservation and Recovery Act requirements.

Jacobson, Victor Levon

2002-08-01T23:59:59.000Z

291

Remaining Sites Verification Package for the 600-111, P-11 Critical Mass Laboratory Crib, and UPR-600-16, Fire and Contamination Spread Waste Sites, Waste Site Reclassification Form 2008-045  

SciTech Connect (OSTI)

The UPR-600-16, Fire and Contamination Spread waste site is an unplanned release that occurred on December 4, 1951, when plutonium contamination was spread by a fire that ignited inside the 120 Experimental Building. The 120 Experimental Building was a laboratory building that was constructed in 1949 and used for plutonium criticality studies as part of the P-11 Project. In November 1951, a criticality occurred in the 120 Experimental Building that resulted in extensive plutonium contamination inside the building. The confirmatory evaluation supports a reclassification of this site to Interim Closed Out. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of the extensive radiological survey of the surface soil and the confirmatory and verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

J. M. Capron

2008-10-28T23:59:59.000Z

292

Waste acceptance criteria for the Waste Isolation Pilot Plant. Revision 4  

SciTech Connect (OSTI)

This Revision 4 of the Waste Acceptance Criteria (WAC), WIPP-DOE-069, identifies and consolidates existing criteria and requirements which regulate the safe handling and preparation of Transuranic (TRU) waste packages for transportation to and emplacement in the Waste Isolation Pilot Plant (WIPP). This consolidation does not invalidate any existing certification of TRU waste to the WIPP Operations and Safety Criteria (Revision 3 of WIPP-DOE--069) and/or Transportation: Waste Package Requirements (TRUPACT-II Safety Analysis Report for Packaging [SARP]). Those documents being consolidated, including Revision 3 of the WAC, currently support the Test Phase.

Not Available

1991-12-01T23:59:59.000Z

293

PROCEDURE FOR OPENING PACKAGES CONTAINING RADIONUCLIDES Laboratory personnel should open and inspect packages immediately upon receipt.  

E-Print Network [OSTI]

as soon as possible to the vendor. Return of radioactive material to the vendor must be coordinated of absorbing material. 5. The inner packaging which includes the liner, shield, and absorbent materials may be contaminated; they are to be discarded in the radioactive waste container unless shown to be uncontaminated

Slatton, Clint

294

Cleanup Verification Package for the 618-8 Burial Ground  

SciTech Connect (OSTI)

This cleanup verification package documents completion of remedial action for the 618-8 Burial Ground, also referred to as the Solid Waste Burial Ground No. 8, 318-8, and the Early Solid Waste Burial Ground. During its period of operation, the 618-8 site is speculated to have been used to bury uranium-contaminated waste derived from fuel manufacturing, and construction debris from the remodeling of the 313 Building.

M. J. Appel

2006-08-10T23:59:59.000Z

295

EXPERIENCE FROM TWO SMALL QUANTITY RH-TRU WASTE SITES IN NAVIGATING THROUGH AN EVOLVING REGULATORY LANDSCAPE  

SciTech Connect (OSTI)

Two small quantity transuranic (TRU) waste generator sites have gained considerable experience in navigating through a changing regulatory landscape in their efforts to remove the TRU waste from their sites and proceed with site remediation. The Battelle Columbus Laboratories Decommissioning Project (BCLDP) has the objectives of decontaminating nuclear research buildings and associated grounds and remediating to a level of residual contamination allowing future use without radiological restrictions. As directed by Congress, BCLDP must complete decontamination and decommissioning activities by the end of Fiscal Year (FY) 2006. This schedule requires the containerization of all TRU waste in 2002. BCLDP will generate a total of approximately 27 cubic meters (m3) of remote-handled (RH) TRU waste. Similarly, the Energy Technology Engineering Center (ETEC) is scheduled to close in 2006 pursuant to an agreement between the U.S. Department of Energy (DOE) and Boeing Canoga Park, the management and operating contractor for ETEC. ETEC had 11.0 m3 of RH-TRU and contact-handled (CH) TRU waste in storage, with the requirement to remove this waste in 2002 in order to meet their site closure schedule. The individual milestones for BCLDP and ETEC necessitated the establishment of site-specific programs to direct packaging and characterization of RH-TRU waste before the regulatory framework for the WIPP disposal of RH-TRU waste is finalized. The lack of large infrastructure for characterization activities, as well as the expedited schedules needed to meet regulatory milestones, provided both challenges and opportunities that are unique to small quantity sites. Both sites have developed unique programs for waste characterization based on the same premise, which directs comprehensive waste data collection efforts such that additional characterization will not be required following the finalization of the WIPP RH-TRU waste program requirements. This paper details the BCLDP program evolution in terms of strategy, innovative solutions to waste characterization, and development of alternative transportation options. Preliminary indications from various regulatory and oversight agencies and professional organizations are that the BCLDP RH-TRU waste characterization program is the ''model WIPP certification program'' and will satisfy anticipated regulatory expectations. This paper also summarizes how BCLDP lessons learned and their development of new resources were applied to the RH-TRU waste characterization and disposition program at ETEC.

Biedscheid, Jennifer; Devarakonda, Murthy; Eide, Jim; Kneff, Dennis

2003-02-27T23:59:59.000Z

296

Waste characterization for the F/H Effluent Treatment Facility in support of waste certification  

SciTech Connect (OSTI)

The Waste Acceptance Criteria (WAC) procedures define the rules concerning packages of solid Low Level Waste (LLW) that are sent to the E-area vaults (EAV). The WACs tabulate the quantities of 22 radionuclides that require manifesting in waste packages destined for each type of vault. These quantities are called the Package Administrative Criteria (PAC). If a waste package exceeds the PAC for any radionuclide in a given vault, then specific permission is needed to send to that vault. To avoid reporting insignificant quantities of the 22 listed radionuclides, the WAC defines the Minimum Reportable Quantity (MRQ) of each radionuclide as 1/1000th of the PAC. If a waste package contains less than the MRQ of a particular radionuclide, then the package`s manifest will list that radionuclide as zero. At least one radionuclide has to be reported, even if all are below the MRQ. The WAC requires that the waste no be ``hazardous`` as defined by SCDHEC/EPA regulations and also lists several miscellaneous physical/chemical requirements for the packages. This report evaluates the solid wastes generated within the F/H Effluent Treatment Facility (ETF) for potential impacts on waste certification.

Brown, D.F.

1994-10-17T23:59:59.000Z

297

Waste segregation procedures and benefits  

SciTech Connect (OSTI)

Segregation is a critical first step in handling hazardous and radioactive materials to minimize the generation of regulated wastes. In addition, segregation can significantly reduce the complexity and the total cost of managing waste. Procedures at Sandia National Laboratories, Albuquerque require that wastes be segregated, first, by waste type (acids, solvents, low level radioactive, mixed, classified, etc.). Higher level segregation requirements, currently under development, are aimed at enhancing the possibilities for recovery, recycle and reapplication; reducing waste volumes; reducing waste disposal costs, and facilitating packaging storage, shipping and disposal. 2 tabs.

Fish, J.D.; Massey, C.D.; Ward, S.J.

1990-01-01T23:59:59.000Z

298

Mixed waste: Proceedings  

SciTech Connect (OSTI)

This volume contains the peer-reviewed and edited versions of papers submitted for presentation a the Second International Mixed Waste Symposium. Following the tradition of the First International Mixed Waste Symposium, these proceedings were prepared in advance of the meeting for distribution to participants. The symposium was organized by the Mixed Waste Committee of the American Society of Mechanical Engineers. The topics discussed at the symposium include: stabilization technologies, alternative treatment technologies, regulatory issues, vitrification technologies, characterization of wastes, thermal technologies, laboratory and analytical issues, waste storage and disposal, organic treatment technologies, waste minimization, packaging and transportation, treatment of mercury contaminated wastes and bioprocessing, and environmental restoration. Individual abstracts are catalogued separately for the data base.

Moghissi, A.A.; Blauvelt, R.K.; Benda, G.A.; Rothermich, N.E. [eds.] [Temple Univ., Philadelphia, PA (United States). Dept. of Environmental Safety and Health

1993-12-31T23:59:59.000Z

299

TRU waste certification compliance requirements for acceptance of contact-handled wastes retrieved from storage to be shipped to the WIPP. Revision 1  

SciTech Connect (OSTI)

Compliance requirements are presented for certifying that unclassified, contact-handled (CH) transuranic (TRU) solid defense wastes retrieved from storage at DOE sites meet the Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria (WAC). All applicable DOE orders must continue to be met. The compliance requirements for certified waste retrieved from certified storage are addressed in another document. The compliance requirements are divided into four sections, primarily determined by the general feature that the requirements address. These sections are General Requirements, Waste Container Requirements, Waste Form Requirements, and Waste Package Requirements. The waste package is the combination of waste container and waste. 2 refs., 1 fig.

Not Available

1985-09-01T23:59:59.000Z

300

Overview of Fords Thermoelectric Programs: Waste Heat Recovery...  

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

Overview of progress in TE waste heat recovery from sedan gasoline-engine exhaust, TE HVAC system in hybrid sedan, and establishing targets for cost, power density, packaging,...

Note: This page contains sample records for the topic "waste disposition packaging" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


301

Formulation and Analysis of Compliant Grouted Waste Forms for SHINE Waste Streams  

SciTech Connect (OSTI)

Optional grouted waste forms were formulated for waste streams generated during the production of 99Mo to be compliant with low-level radioactive waste regulations. The amounts and dose rates of the various waste form materials that would be generated annually were estimated and used to determine the effects of various waste processing options, such as the of number irradiation cycles between uranium recovery operations, different combinations of waste streams, and removal of Pu, Cs, and Sr from waste streams for separate disposition (which is not evaluated in this report). These calculations indicate that Class C-compliant grouted waste forms can be produced for all waste streams. More frequent uranium recovery results in the generation of more chemical waste, but this is balanced by the fact that waste forms for those waste streams can accommodate higher waste loadings, such that similar amounts of grouted waste forms are required regardless of the recovery schedule. Similar amounts of grouted waste form are likewise needed for the individual and combined waste streams. Removing Pu, Cs, and Sr from waste streams lowers the waste form dose significantly at times beyond about 1 year after irradiation, which may benefit handling and transport. Although these calculations should be revised after experimentally optimizing the grout formulations and waste loadings, they provide initial guidance for process development.

Ebert, William; Pereira, Candido; Heltemes, Thad A.; Youker, Amanda; Makarashvili, Vakhtang; Vandegrift, George F.

2014-01-01T23:59:59.000Z

302

Hanford Site annual dangerous waste report, calendar year 1995  

SciTech Connect (OSTI)

This report is a compilation of data on the disposition of hazardous wastes generated on the Hanford Reservation. This information is on EPA requirement every two years. Wastes include: tank simulant waste; alkaline batteries; lead-based paints; organic solvents; light bulbs containing lead and/or mercury; monitoring well drilling wastes; soils contaminated with trace metals, halogenated organics, or other pollutants; Ni-Cd batteries; pesticides; waste oils and greases; wastes from the cleanup of fuel/gasoline spills; filters; metals; and other.

NONE

1995-12-31T23:59:59.000Z

303

Safety Evaluation for Packaging (onsite) T Plant Canyon Items  

SciTech Connect (OSTI)

This safety evaluation for packaging (SEP) evaluates and documents the ability to safely ship mostly unique inventories of miscellaneous T Plant canyon waste items (T-P Items) encountered during the canyon deck clean off campaign. In addition, this SEP addresses contaminated items and material that may be shipped in a strong tight package (STP). The shipments meet the criteria for onsite shipments as specified by Fluor Hanford in HNF-PRO-154, Responsibilities and Procedures for all Hazardous Material Shipments.

OBRIEN, J.H.

2000-07-14T23:59:59.000Z

304

Proliferation resistance criteria for fissile material disposition issues  

SciTech Connect (OSTI)

The 1994 National Acdaemy of Sciences study ``Management and Disposition of Excess Weapons Plutonium`` defined options for reducing the national and international proliferation risks of materials declared excess to the nuclear weapons program. This paper proposes criteria for assessing the proliferation resistance of these options as well defining the ``Standards`` from the report. The criteria are general, encompassing all stages of the disposition process from storage through intermediate processing to final disposition including the facilities, processing technologies and materials, the level of safeguards for these materials, and the national/subnational threat to the materials.

Rutherford, D.A.; Fearey, B.L.; Markin, J.T.; Close, D.A. [Los Alamos National Lab., NM (United States); Tolk, K.M.; Mangan, D.L. [Sandia National Labs., Albuquerque, NM (United States); Moore, L. [Lawrence Livermore National Lab., CA (United States)

1995-09-01T23:59:59.000Z

305

WASTE DISPOSAL SECTION CORNELL UNIVERSITY  

E-Print Network [OSTI]

radioactive products as regular trash. All packages must be free of contamination, radiation symbols2/07 WASTE DISPOSAL SECTION CORNELL UNIVERSITY PROCEDURE for DISPOSAL of RADIOACTIVE MATERIALS This procedure has been developed to ensure the safety of those individuals who handle radioactive waste

Pawlowski, Wojtek

306

Iron phosphate compositions for containment of hazardous metal waste  

DOE Patents [OSTI]

An improved iron phosphate waste form for the vitrification, containment and long-term disposition of hazardous metal waste such as radioactive nuclear waste is provided. The waste form comprises a rigid iron phosphate matrix resulting from the cooling of a melt formed by heating a batch mixture comprising the metal waste and a matrix-forming component. The waste form comprises from about 30 to about 70 weight percent P{sub 2}O{sub 5} and from about 25 to about 50 weight percent iron oxide and has metals present in the metal waste chemically dissolved therein. The concentration of iron oxide in the waste form along with a high proportion of the iron in the waste form being present as Fe{sup 3+} provide a waste form exhibiting improved chemical resistance to corrosive attack. A method for preparing the improved iron phosphate waste forms is also provided. 21 figs.

Day, D.E.

1998-05-12T23:59:59.000Z

307

Iron phosphate compositions for containment of hazardous metal waste  

DOE Patents [OSTI]

An improved iron phosphate waste form for the vitrification, containment and long-term disposition of hazardous metal waste such as radioactive nuclear waste is provided. The waste form comprises a rigid iron phosphate matrix resulting from the cooling of a melt formed by heating a batch mixture comprising the metal waste and a matrix-forming component. The waste form comprises from about 30 to about 70 weight percent P.sub.2 O.sub.5 and from about 25 to about 50 weight percent iron oxide and has metals present in the metal waste chemically dissolved therein. The concentration of iron oxide in the waste form along with a high proportion of the iron in the waste form being present as Fe.sup.3+ provide a waste form exhibiting improved chemical resistance to corrosive attack. A method for preparing the improved iron phosphate waste forms is also provided.

Day, Delbert E. (Rolla, MO)

1998-01-01T23:59:59.000Z

308

Cleanup Verification Package for the 118-F-1 Burial Ground  

SciTech Connect (OSTI)

This cleanup verification package documents completion of remedial action for the 118-F-1 Burial Ground on the Hanford Site. This burial ground is a combination of two locations formerly called Minor Construction Burial Ground No. 2 and Solid Waste Burial Ground No. 2. This waste site received radioactive equipment and other miscellaneous waste from 105-F Reactor operations, including dummy elements and irradiated process tubing; gun barrel tips, steel sleeves, and metal chips removed from the reactor; filter boxes containing reactor graphite chips; and miscellaneous construction solid waste.

E. J. Farris and H. M. Sulloway

2008-01-10T23:59:59.000Z

309

Biohazardous Waste Disposal GuidelinesDescriptionStorage& LabelingTreatmentDisposal  

E-Print Network [OSTI]

Waste Sharps Waste Solid Lab Waste Liquid Waste Any of these devices if contaminated with biohazardousBiohazardous Waste Disposal GuidelinesDescriptionStorage& packaging LabelingTreatmentDisposal Mixed container. Container must be leakproof, ridgid, puncture resistant, clearly marked for biohazardous waste

Wikswo, John

310

EA-1962: Analysis for Below Grade Suspect Transuranic (TRU) Waste at Technical Area (TA)-54  

Broader source: Energy.gov [DOE]

DOE is preparing an EA to evaluate the legacy suspect transuranic (TRU) waste at Area G for the purposes of reclassification of waste type and determination of a final disposal path. Per DOE Order 435.1, Change 1, Radioactive Waste Management, and its associated guide, legacy waste at Los Alamos National Laboratory that contained TRU waste was stored and managed as TRU waste. The waste was given an interim classification for the purposes of applying the most restrictive standard until the waste could be adequately characterized and a final determination on the disposition classification was made.

311

Americium/Curium Disposition Life Cycle Planning Study  

SciTech Connect (OSTI)

At the request of the Department of Energy Savannah River Office (DOE- SR), Westinghouse Savannah River Company (WSRC) evaluated concepts to complete disposition of Americium and Curium (Am/Cm) bearing materials currently located at the Savannah River Site (SRS).

Jackson, W.N. [Westinghouse Savannah River Company, AIKEN, SC (United States); Krupa, J.; Stutts, P.; Nester, S.; Raimesch, R.

1998-04-30T23:59:59.000Z

312

EIS-0327: Disposition of Scrap Metals Programmatic EIS  

Broader source: Energy.gov [DOE]

This EIS will evaluate the environmental impacts of policy alternatives for the disposition of scrap metals (primarily carbon steel and stainless steel) that may have residual surface radioactivity. DOE is cancelling this EIS.

313

Final Demolition and Disposition of 209-E Critical Mass Laboratory - 12267  

SciTech Connect (OSTI)

The 209-E Critical Mass Laboratory was constructed in 1960 to provide a heavy shielded reactor room where quantities of plutonium or uranium in solution could be brought to near-critical configurations under carefully controlled and monitored conditions. In the late 1980's, the responsible contractor, Pacific Northwest National Laboratory (PNNL), was directed by the Department of Energy (DOE) to prepare the facility for unoccupied status. The facility was demolished under a Removal Action Work Plan pursuant to the Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA). The funding for this project was provided by the American Recovery and Reinvestment Act (ARRA). The primary rooms of concern with regards to contamination in 209-E facility, which is over 9,000 square feet, are the criticality assembly room (CAR), the mix room, and the change room. The CAR contained two reactor hoods (HO-140 and HO-170), which each had a high efficiency particulate air (HEPA) filter system. The CAR contained 13 tanks ranging from 38 L (10 gal) to 401 L (106 gal). Tanks TK-109 and TK-110 are below grade, and were removed as part of this demolition and disposition remedy. Nonradiological and radiological hazardous substances were removed, decontaminated, or fixed in place, prior to demolition. Except for the removal of below grade tanks TK-109 and TK-110, the facility was demolished to slab-on-grade. PNNL performed stabilization and deactivation activities that included removal of bulk fissile material and chemicals, flushing tanks, stabilizing contamination within gloveboxes and hoods, and packaging and removing waste. The removal of the contaminated plutonium equipment and materials from the 209E facility presented a number of challenges similar in nature to those associated with the inventory reduction and cleanup activities at the Plutonium Finishing Plant. Although there were no bulk fissile materials or chemicals within the facility, there were residual radiological materials (isotopes of plutonium and americium) in the tanks and hoods. The complexity of the remedy was present because of the various configurations of the tanks and hoods, combined with the residual contaminants. Because of the weight and dimensional configuration, size reduction of the slab tanks, as well as removal and disposal of the different material used for moderation and absorption, were two examples of challenges that were resolved to complete the remedy. One of the key methods developed and implemented at the facility was the design and construction of a shroud to allow the cutting of the Pu contaminated tanks. The shroud design, development and implementation at the 209E Project was an example of enhanced work planning and task hazards analysis with worker involvement. This paper will present the lessons learned from the 209E facility inventory reduction activities including the shroud and other methodologies used. The initial Lessons Learned discussion for this project was scheduled for late January 2012. This facility is the first open-air demolition of a highly contaminated plutonium-contaminated facility accomplished by CH2M Hill under the Plateau Remediation Contract. The demolition was completed without spread of contamination to the workers and the surrounding area. As with any project of this complexity, there are significant accomplishments, as well as experience that can be applied to future demolition of plutonium-contaminated facilities on the Hanford Site. These experiences will be documented at a later date. (authors)

Woolery, Wade [US Department of Energy, Richland WA (United States); Dodd, Edwin III [CH2M Hill Plateau Remediation Company, Richland WA (United States)

2012-07-01T23:59:59.000Z

314

Plan and schedule for disposition and regulatory compliance for miscellaneous streams. Revision 1  

SciTech Connect (OSTI)

On December 23, 1991, the U.S. Department of Energy, Richland Operations Office (RL) and the Washington State Department of Ecology (Ecology) agreed to adhere to the provisions of Department of Ecology Consent Order No. DE 91NM-177 (Consent Order). The Consent Order lists regulatory milestones for liquid effluent streams at the Hanford Site to comply with the permitting requirements of Washington Administrative Code (WAC) 173-216 (State Waste Discharge Permit Program) or WAC 173-218 (Washington Underground Injection Control Program) where applicable. Hanford Site liquid effluent streams discharging to the soil column have been categorized in the Consent Order as follows: Phase I Streams Phase II Streams Miscellaneous Streams. Phase I and Phase II Streams are addressed in two RL reports: {open_quotes}Plan and Schedule to Discontinue Disposal of Contaminated Liquids into the Soil Column at the Hanford Site{close_quotes} (DOE-RL 1987), and {open_quotes}Annual Status of the Report of the Plan and Schedule to Discontinue Disposal of Contaminated Liquids into the Soil Column at the Hanford Site{close_quotes}. Miscellaneous Streams are those liquid effluent streams discharged to the ground that are not categorized as Phase I or Phase II Streams. Miscellaneous Streams discharging to the soil column at the Hanford Site are subject to the requirements of several milestones identified in the Consent Order. This document provides a plan and schedule for the disposition of Miscellaneous Streams. The disposition process for the Miscellaneous Streams is facilitated using a decision tree format. The decision tree and corresponding analysis for determining appropriate disposition of these streams is presented in this document.

NONE

1994-12-01T23:59:59.000Z

315

Nuclear Material Packaging Manual  

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

The manual provides detailed packaging requirements for protecting workers from exposure to nuclear materials stored outside of an approved engineered contamination barrier. No cancellation. Certified 11-18-10.

2008-03-07T23:59:59.000Z

316

Packaging and Transportation Safety  

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

Establishes safety requirements for the proper packaging and transportation of offsite shipments and onsite transfers of hazardous materials andor modal transport. Cancels DOE 1540.2 and DOE 5480.3

1995-09-27T23:59:59.000Z

317

Packaging and Transportation Safety  

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

Establishes safety requirements for the proper packaging and transportation of Department of Energy (DOE) offsite shipments and onsite transfers of hazardous materials and for modal transport. Canceled by DOE 460.1A

1995-09-27T23:59:59.000Z

318

Packaging and Transportation Safety  

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

Establishes safety requirements for the proper packaging and transportation of Department of Energy (DOE) offsite shipments and onsite transfers of hazardous materials and for modal transport. Cancels DOE O 460.1.

1996-10-02T23:59:59.000Z

319

CH Packaging Program Guidance  

SciTech Connect (OSTI)

The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: ''each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.'' They further state: ''each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.'' Chapter 9.0 of the SARP charges the WIPP management and operating (M&O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with 10 CFR 71.11. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. CBFO will evaluate the issue and notify the NRC if required. This document provides the instructions to be followed to operate, maintain, and test the TRUPACT-II and HalfPACT packaging. The intent of these instructions is to standardize operations. All users will follow these instructions or equivalent instructions that assure operations are safe and meet the requirements of the SARPs.

Washington TRU Solutions LLC

2003-04-30T23:59:59.000Z

320

CH Packaging Program Guidance  

SciTech Connect (OSTI)

The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT Shipping Package, and directly related components. This document complies with the minimum requirements as specified in TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event there is a conflict between this document and the SARP or C of C, the SARP and/or C of C shall govern. C of Cs state: ''each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.'' They further state: ''each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.'' Chapter 9.0 of the SAR P charges the WIPP Management and Operation (M&O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with 10 CFR 71.11. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. CBFO will evaluate the issue and notify the NRC if required. This document details the instructions to be followed to operate, maintain, and test the TRUPACT-II and HalfPACT packaging. The intent of these instructions is to standardize these operations. All users will follow these instructions or equivalent instructions that assure operations are safe and meet the requirements of the SARPs.

Washington TRU Solutions LLC

2002-03-04T23:59:59.000Z

Note: This page contains sample records for the topic "waste disposition packaging" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


321

Transuranic Waste Processing Center (TWPC) Legacy Tank RH-TRU Sludge Processing and Compliance Strategy - 13255  

SciTech Connect (OSTI)

The U.S. Department of Energy (DOE) needs to safely and efficiently treat its 'legacy' transuranic (TRU) waste and mixed low-level waste (LLW) from past research and defense activities at the Oak Ridge National Laboratory (ORNL) so that the waste is prepared for safe and secure disposal. The TWPC operates an Environmental Management (EM) waste processing facility on the Oak Ridge Reservation (ORR). The TWPC is classified as a Hazard Category 2, non-reactor nuclear facility. This facility receives, treats, and packages low-level waste and TRU waste stored at various facilities on the ORR for eventual off-site disposal at various DOE sites and commercial facilities. The Remote Handled TRU Waste Sludge held in the Melton Valley Storage Tanks (MVSTs) was produced as a result of the collection, treatment, and storage of liquid radioactive waste originating from the ORNL radiochemical processing and radioisotope production programs. The MVSTs contain most of the associated waste from the Gunite and Associated Tanks (GAAT) in the ORNL's Tank Farms in Bethel Valley and the sludge (SL) and associated waste from the Old Hydro-fracture Facility tanks and other Federal Facility Agreement (FFA) tanks. The SL Processing Facility Build-outs (SL-PFB) Project is integral to the EM cleanup mission at ORNL and is being accelerated by DOE to meet updated regulatory commitments in the Site Treatment Plan. To meet these commitments a Baseline (BL) Change Proposal (BCP) is being submitted to provide continued spending authority as the project re-initiation extends across fiscal year 2012 (FY2012) into fiscal year 2013. Future waste from the ORNL Building 3019 U-233 Disposition project, in the form of U-233 dissolved in nitric acid and water, down-blended with depleted uranyl nitrate solution is also expected to be transferred to the 7856 MVST Annex Facility (formally the Capacity Increase Project (CIP) Tanks) for co-processing with the SL. The SL-PFB project will construct and install the necessary integrated systems to process the accumulated MVST Facilities SL inventory at the TWPC thus enabling safe and effective disposal of the waste. This BCP does not include work to support current MVST Facility Surveillance and Maintenance programs or the ORNL Building 3019 U-233 Disposition project, since they are not currently part of the TWPC prime contract. The purpose of the environmental compliance strategy is to identify the environmental permits and other required regulatory documents necessary for the construction and operation of the SL- PFB at the TWPC, Oak Ridge, TN. The permits and other regulatory documents identified are necessary to comply with the environmental laws and regulations of DOE Orders, and other requirements documented in the SL-PFB, Safety Design Strategy (SDS), SL-A-AD-002, R0 draft, and the Systems, Function and Requirements Document (SFRD), SL-X-AD-002, R1 draft. This compliance strategy is considered a 'living strategy' and it is anticipated that it will be revised as design progresses and more detail is known. The design basis on which this environmental permitting and compliance strategy is based is the Wastren Advantage, Inc., (WAI), TWPC, SL-PFB (WAI-BL-B.01.06) baseline. (authors)

Rogers, Ben C.; Heacker, Fred K.; Shannon, Christopher [Wastren Advantage, Inc., Transuranic Waste Processing Center, 100 WIPP Road, Lenoir City, Tennessee 37771 (United States)] [Wastren Advantage, Inc., Transuranic Waste Processing Center, 100 WIPP Road, Lenoir City, Tennessee 37771 (United States); and others

2013-07-01T23:59:59.000Z

322

Waste characterization at Los Alamos National Laboratory  

SciTech Connect (OSTI)

Most industries generate limited types of solid wastes of a result of their manufacturing processes. The Los Alamos National Laboratory (LANL), a research and development facility, generates a large variety of solid wastes, some exotic. Over 50,000 distinct waste streams are currently generated in the 43 square mile area defining LANL. These wastes include refuse, medical, infectious, hazardous, radioactive, and mixed wastes. LANL is subject to federal and State oversight on matters concerning management of solid wastes. In order to assure regulatory agencies such as the New Mexico Environment Department (NMED) and the US Environmental Protection Agency (EPA) that the Laboratory is properly managing and disposing all solid wastes. LANL has undertaken an extensive waste characterization program to identify sources and ultimate disposition of all solid wastes. Given the number of solid waste streams expected, LANL has taken a two-pronged approach to characterizing wastes: (a) physical identification of all sources of solid wastes including interviews with waste generators; and (b) characterization of wastes from the point of generation. The former approach consists of canvassing all structures within the LANL complex, interviewing waste generators, and identifying sources of waste generation. Data gathered by these interviews are compiled in a database in order to identify the types and rates of waste generation and correct mismanagement of wastes identified during the interviews. The latter approach consists of characterizing all solid wastes which are controlled administratively or subject to stricter controls than municipal solid wastes (i.e., infectious, hazardous, radioactive, and mixed wastes). This characterization forms the basis by which LANL will manage solid waste in accordance to NMED/EPA regulations and US Department of Energy Orders. 8 refs., 3 figs.

Corpion, J.C.; Grieggs, A.R.

1991-01-01T23:59:59.000Z

323

An approach for sampling solid heterogeneous waste at the Hanford Site waste receiving and processing and solid waste projects  

SciTech Connect (OSTI)

This paper addresses the problem of obtaining meaningful data from samples of solid heterogeneous waste while maintaining sample rates as low as practical. The Waste Receiving and Processing Facility, Module 1, at the Hanford Site in south-central Washington State will process mostly heterogeneous solid wastes. The presence of hazardous materials is documented for some packages and unknown for others. Waste characterization is needed to segregate the waste, meet waste acceptance and shipping requirements, and meet facility permitting requirements. Sampling and analysis are expensive, and no amount of sampling will produce absolute certainty of waste contents. A sampling strategy is proposed that provides acceptable confidence with achievable sampling rates.

Sexton, R.A.

1993-03-01T23:59:59.000Z

324

Used fuel disposition campaign international activities implementation plan.  

SciTech Connect (OSTI)

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

Nutt, W. M. (Nuclear Engineering Division)

2011-06-29T23:59:59.000Z

325

NEVADA TEST SITE WASTE ACCEPTANCE CRITERIA  

SciTech Connect (OSTI)

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

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

2005-07-01T23:59:59.000Z

326

Initial Package Design Concepts Integrated Product Team (IPT) Summary Report  

SciTech Connect (OSTI)

Initially, the question of transporting TRU waste to WIPP was raised as part of the EM Integration activities. The issue was re-examined as part of the system-wide view to re-engineer the TRU waste program. Consequently, the National Transportation Program and the National TRU Waste Program, in a cooperative effort, made a commitment to EM-20 to examine the feasibility of using rail to transport TRU waste material to WIPP. In December of 1999 Mr. Philip Altomare assembled a team of subject matter experts (SME) to define initial concepts for a Type B package capable of shipping TRU waste by rail (see Attachment 1 for a list of team members). This same team of experts also provided input to a preliminary study to determine if shipping TRU waste by rail could offer cost savings or other significant advantages over the current mode of operation using TRUPACT-II packages loaded on truck. As part of the analysis, the team also identified barriers to implementing rail shipments to WIPP and outlined a path forward. This report documents the findings of the study and its initial set of recommendations. As the study progressed, it was expanded to include new packages for truck as well as rail in recognition of the benefits of shipping large boxes and contaminated equipment.

Moss, J.; Luke, Dale Elden

2000-03-01T23:59:59.000Z

327

High Efficiency Integrated Package  

SciTech Connect (OSTI)

Solid-state lighting based on LEDs has emerged as a superior alternative to inefficient conventional lighting, particularly incandescent. LED lighting can lead to 80 percent energy savings; can last 50,000 hours – 2-50 times longer than most bulbs; and contains no toxic lead or mercury. However, to enable mass adoption, particularly at the consumer level, the cost of LED luminaires must be reduced by an order of magnitude while achieving superior efficiency, light quality and lifetime. To become viable, energy-efficient replacement solutions must deliver system efficacies of ? 100 lumens per watt (LPW) with excellent color rendering (CRI > 85) at a cost that enables payback cycles of two years or less for commercial applications. This development will enable significant site energy savings as it targets commercial and retail lighting applications that are most sensitive to the lifetime operating costs with their extended operating hours per day. If costs are reduced substantially, dramatic energy savings can be realized by replacing incandescent lighting in the residential market as well. In light of these challenges, Cree proposed to develop a multi-chip integrated LED package with an output of > 1000 lumens of warm white light operating at an efficacy of at least 128 LPW with a CRI > 85. This product will serve as the light engine for replacement lamps and luminaires. At the end of the proposed program, this integrated package was to be used in a proof-of-concept lamp prototype to demonstrate the component’s viability in a common form factor. During this project Cree SBTC developed an efficient, compact warm-white LED package with an integrated remote color down-converter. Via a combination of intensive optical, electrical, and thermal optimization, a package design was obtained that met nearly all project goals. This package emitted 1295 lm under instant-on, room-temperature testing conditions, with an efficacy of 128.4 lm/W at a color temperature of ~2873K and 83 CRI. As such, the package’s performance exceeds DOE’s warm-white phosphor LED efficacy target for 2013. At the end of the program, we assembled an A19 sized demonstration bulb housing the integrated package which met Energy Star intensity variation requirements. With further development to reduce overall component cost, we anticipate that an integrated remote converter package such as developed during this program will find application in compact, high-efficacy LED-based lamps, particularly those requiring omnidirectional emission.

Ibbetson, James

2013-09-15T23:59:59.000Z

328

Used fuel disposition research and development roadmap - FY10 status.  

SciTech Connect (OSTI)

Since 1987 the U.S. has focused research and development activities relevant to the disposal of commercial used nuclear fuel and U.S. Department of Energy (DOE) owned spent nuclear fuel and high level waste on the proposed repository at Yucca Mountain, Nevada. At the same time, the U.S. successfully deployed a deep geologic disposal facility for defense-related transuranic waste in bedded salt at the Waste Isolation Pilot Plant. In 2009 the DOE established the Used Fuel Disposition Campaign (UFDC) within the Office of Nuclear Energy. The Mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel and wastes generated by existing and future nuclear fuel cycles. The U.S. national laboratories have participated on these programs and has conducted research and development related to these issues to a limited extent. However, a comprehensive research and development (R&D) program investigating a variety of geologic media has not been a part of the U.S. waste management program since the mid 1980s. Such a comprehensive R&D program is being developed in the UFDC with a goal of meeting the UFDC Grand Challenge to provide a sound technical basis for absolute confidence in the safety and security of long-term storage, transportation, and disposal of used nuclear fuel and wastes from the nuclear energy enterprise. The DOE has decided to no longer pursue the development of a repository at Yucca Mountain, Nevada. Since a repository site will ultimately have to be selected, sited, characterized, designed, and licensed, other disposal options must now be considered. In addition to the unsaturated volcanic tuff evaluated at Yucca Mountain, several different geologic media are under investigation internationally and preliminary assessments indicate that disposal of used nuclear fuel and high level waste in these media is feasible. Considerable progress has been made in the U.S. and other nations in understanding disposal concepts in different geologic media, but gaps in knowledge still exist. A principal aspect of concern to the UFDC as it considers the broad issues of siting a repository in different geologic media are the marked differences in the regulatory bases for assessing suitability and safety of a repository between the U.S. and other nations. Because the probability based - risked informed nature of the current U.S. regulations for high-level radioactive waste and spent nuclear fuel is sufficiently different from other regulations, information gained in previous studies, while useful, likely need to be supplemented to enable more convincing communication with the public, better defense of the numerical models, and stronger safety cases. Thus, it was recognized when the UFDC was established that there were readily identified disposal-related R&D opportunities to address knowledge gaps. An effort to document these research opportunities was a key component of Fiscal Year (FY) 2010 engineered system, natural system, and system-level modeling activities for a range of disposal environments. A principal contribution to identifying these gaps was a workshop held to gather perspectives from experts both within and external to the UFDC regarding R&D opportunities. In the planning for FY2010 it was expected that these activities would culminate with a UFDC research and development roadmap that would identify the knowledge gaps, discuss the R&D needed to fill these gaps, and prioritize the proposed R&D over both the near- and long-term. A number of knowledge gaps and needed R&D were identified and are discussed in this report. However, these preliminary R&D topics have not been evaluated in detail nor have they been prioritized to support future planning efforts. This will be completed in FY11 and the final UFDC Research and Development Roadmap will be completed. This report discusses proposed R&D topics in three areas related to repository siting, design, and performance: natural systems

Nutt, W. M. (Nuclear Engineering Division)

2010-10-01T23:59:59.000Z

329

Release Data Package for the 2004 Composite Analysis  

SciTech Connect (OSTI)

This data package contains all the key parameter data necessary for implementation of the Release Module to conduct the 2004 Composite Analysis. A composite analysis is required by DOE Order 435.1 to ensure public safety through the management of active and planned low-level radioactive waste disposal facilities associated with Hanford.

Riley, Robert G.; Lopresti, Charles A.

2004-08-02T23:59:59.000Z

330

SLIGHTLY IRRADIATED FUEL (SIF) INTERIM DISPOSITION PROJECT  

SciTech Connect (OSTI)

CH2M HILL Plateau Remediation Company (CH2M HILL PRC) is proud to submit the Slightly Irradiated Fuel (SIF) Interim Disposition Project for consideration by the Project Management Institute as Project of the Year for 2010. The SIF Project was a set of six interrelated sub-projects that delivered unique stand-alone outcomes, which, when integrated, provided a comprehensive and compliant system for storing high risk special nuclear materials. The scope of the six sub-projects included the design, construction, testing, and turnover of the facilities and equipment, which would provide safe, secure, and compliant Special Nuclear Material (SNM) storage capabilities for the SIF material. The project encompassed a broad range of activities, including the following: Five buildings/structures removed, relocated, or built; Two buildings renovated; Structural barriers, fencing, and heavy gates installed; New roadways and parking lots built; Multiple detection and assessment systems installed; New and expanded communication systems developed; Multimedia recording devices added; and A new control room to monitor all materials and systems built. Project challenges were numerous and included the following: An aggressive 17-month schedule to support the high-profile Plutonium Finishing Plant (PFP) decommissioning; Company/contractor changeovers that affected each and every project team member; Project requirements that continually evolved during design and construction due to the performance- and outcome-based nature ofthe security objectives; and Restrictions imposed on all communications due to the sensitive nature of the projects In spite of the significant challenges, the project was delivered on schedule and $2 million under budget, which became a special source of pride that bonded the team. For years, the SIF had been stored at the central Hanford PFP. Because of the weapons-grade piutonium produced and stored there, the PFP had some of the tightest security on the Hanford nuclear reservation. Workers had to pass through metal detectors when they arrived at the plant and materials leaving the plant had to be scanned for security reasons. Whereas other high-security nuclear materials were shipped from the PFP to Savannah River, S.C. as part ofa Department of Energy (DOE) program to consolidate weapons-grade plutonium, it was determined that the SIF should remain onsite pending disposition to a national repository. Nevertheless, the SIF still requires a high level of security that the PFP complex has always provided. With the 60-year PFP mission of producing and storing plutonium concluded, the environmental cleanup plans for Hanford call for the demolition of the 63-building PFP complex. Consequently, if the SIF remained at PFP it not only would have interfered with the environmental cleanup plans, but would have required $100 million in facility upgrades to meet increased national security requirements imposed after the 9/11 terrorist attacks. A new smaller and more cost-effective area was needed to store this material, which led to the SIF Project. Once the SIF project was successfully completed and the SIF was safely removed from PFP, the existing Protected Area at PFP could be removed, and demolition could proceed more quickly without being encumbered by restrictive security requirements that an active Protected Area requires. The lightened PFP security level brought by safely removing and storing the SIF would also yield lowered costs for deactivation and demolition, as well as reduce overall life-cycle costs.

NORTON SH

2010-02-23T23:59:59.000Z

331

Evaluation of Used Fuel Disposition in Clay-Bearing Rock  

SciTech Connect (OSTI)

Radioactive waste disposal in shale/argillite rock formations has been widely considered given its desirable isolation properties (low permeability), geochemically reduced conditions, anomalous groundwater pressures, and widespread geologic occurrence. Clay/shale rock formations are characterized by their high content of clay minerals such as smectites and illites where diffusive transport and chemisorption phenomena predominate. These, in addition to low permeability, are key attributes of shale to impede radionuclide mobility. Shale host-media has been comprehensively studied in international nuclear waste repository programs as part of underground research laboratories (URLs) programs in Switzerland, France, Belgium, and Japan. These investigations, in some cases a decade or more long, have produced a large but fundamental body of information spanning from site characterization data (geological, hydrogeological, geochemical, geomechanical) to controlled experiments on the engineered barrier system (EBS) (barrier clay and seals materials). Evaluation of nuclear waste disposal in shale formations in the USA was conducted in the late 70’s and mid 80’s. Most of these studies evaluated the potential for shale to host a nuclear waste repository but not at the programmatic level of URLs in international repository programs. This report covers various R&D work and capabilities relevant to disposal of heat-generating nuclear waste in shale/argillite media. Integration and cross-fertilization of these capabilities will be utilized in the development and implementation of the shale/argillite reference case planned for FY15. Disposal R&D activities under the UFDC in the past few years have produced state-of-the-art modeling capabilities for coupled Thermal-Hydrological-Mechanical-Chemical (THMC), used fuel degradation (source term), and thermodynamic modeling and database development to evaluate generic disposal concepts. The THMC models have been developed for shale repository leveraging in large part on the information garnered in URLs and laboratory data to test and demonstrate model prediction capability and to accurately represent behavior of the EBS and the natural (barrier) system (NS). In addition, experimental work to improve our understanding of clay barrier interactions and TM couplings at high temperatures are key to evaluate thermal effects as a result of relatively high heat loads from waste and the extent of sacrificial zones in the EBS. To assess the latter, experiments and modeling approaches have provided important information on the stability and fate of barrier materials under high heat loads. This information is central to the assessment of thermal limits and the implementation of the reference case when constraining EBS properties and the repository layout (e.g., waste package and drift spacing). This report is comprised of various parts, each one describing various R&D activities applicable to shale/argillite media. For example, progress made on modeling and experimental approaches to analyze physical and chemical interactions affecting clay in the EBS, NS, and used nuclear fuel (source term) in support of R&D objectives. It also describes the development of a reference case for shale/argillite media. The accomplishments of these activities are summarized as follows: ? Development of a reference case for shale/argillite; ? Investigation of Reactive Transport and Coupled THM Processes in EBS: FY14; ? Update on Experimental Activities on Buffer/Backfill Interactions at elevated Pressure and Temperature; ? Thermodynamic Database Development: Evaluation Strategy, Modeling Tools, First-Principles Modeling of Clay, and Sorption Database Assessment; ? ANL Mixed Potential Model For Used Fuel Degradation: Application to Argillite and Crystalline Rock Environments.

Carlos F. Jové Colón; Philippe F. Weck; David H. Sassani; Liange Zheng; Jonny Rutqvist; Carl I. Steefel; Kunhwi Kim; Seiji Nakagawa; James Houseworth; Jens Birkholzer; Florie A. Caporuscio; Michael Cheshire; Michael S. Rearick; Mary K. McCarney; Mavrik Zavarin; Ana Benedicto; Annie B. Kersting; Mark Sutton; James Jerden; Kurt E. Frey; Jacqueline M. Copple; William Ebert

2014-08-29T23:59:59.000Z

332

DOE SEEKS CONTRACTOR TO DISPOSITION WASTE AT THE ADVANCED MIXED WASTE  

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

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333

DOE Chooses Contractor to Disposition Waste at the Advanced Mixed Waste  

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

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334

Operable Unit 3-13, Group 3, Other Surface Soils Remediation Sets 4-6 (Phase II) Waste Management Plan  

SciTech Connect (OSTI)

This Waste Management Plan describes waste management and waste minimization activities for Group 3, Other Surface Soils Remediation Sets 4-6 (Phase II) at the Idaho Nuclear Technology and Engineering Center located within the Idaho National Laboratory. The waste management activities described in this plan support the selected response action presented in the Final Record of Decision for Idaho Nuclear Technology and Engineering Center, Operable Unit 3-13. This plan identifies the waste streams that will be generated during implementation of the remedial action and presents plans for waste minimization, waste management strategies, and waste disposition.

G. L. Schwendiman

2006-07-01T23:59:59.000Z

335

Hanford site transuranic waste certification plan  

SciTech Connect (OSTI)

As a generator of transuranic (TRU) and TRU mixed waste destined for disposal at the Waste Isolation Pilot Plant (WIPP), the Hanford Site must ensure that its TRU waste meets the requirements of U.S. Department of Energy (DOE) Order 5820.2A, ''Radioactive Waste Management, and the Waste Acceptance Criteria for the Waste Isolation Pilot Plant' (DOE 1996d) (WIPP WAC). The WIPP WAC establishes the specific physical, chemical, radiological, and packaging criteria for acceptance of defense TRU waste shipments at WIPP. The WIPP WAC also requires that participating DOE TRU waste generator/treatment/storage sites produce site-specific documents, including a certification plan, that describe their management of TRU waste and TRU waste shipments before transferring waste to WIPP. The Hanford Site must also ensure that its TRU waste destined for disposal at WIPP meets requirements for transport in the Transuranic Package Transporter41 (TRUPACT-11). The U.S. Nuclear Regulatory Commission (NRC) establishes the TRUPACT-I1 requirements in the ''Safety Analysis Report for the TRUPACT-II Shipping Package'' (NRC 1997) (TRUPACT-I1 SARP).

GREAGER, T.M.

1999-05-12T23:59:59.000Z

336

TRU (transuranic) waste certification compliance requirements for acceptance of newly generated contact-handled wastes to be shipped to the Waste Isolation Pilot Plant: Revision 2  

SciTech Connect (OSTI)

Compliance requirements are presented for certifying that unclassified, newly generated (NG), contact-handled (CH) transuranic (TRU) solid wastes from defense programs meet the Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria (WAC). Where appropriate, transportation and interim storage requirements are incorporated; however, interim storage sites may have additional requirements consistent with these requirements. All applicable Department of Energy (DOE) orders must continue to be met. The compliance requirements for stored or buried waste are not addressed in this document. The compliance requirements are divided into four sections, primarily determined by the general feature that the requirements address. These sections are General Requirements, Waste Container Requirements, Waste Form Requirements, and Waste Package Requirements. The waste package is the combination of waste container and waste. 10 refs., 1 fig.

Not Available

1989-01-01T23:59:59.000Z

337

TRU (transuranic) waste certification compliance requirements for acceptance of contact-handled wastes retrieved from storage to be shipped to the Waste Isolation Pilot Plant: Revision 2  

SciTech Connect (OSTI)

Compliance requirements are presented for certifying that unclassified, contact-handled (CH) transuranic (TRU) solid defense wastes retrieved from storage at DOE sites meet the Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria (WAC). All applicable Department of Energy (DOE) orders must continue to be met. The compliance requirements for acceptance of newly generated CH waste to be shipped to the WIPP are addressed in another document. The compliance requirements are divided into four sections, primarily determined by the general feature that the requirements address. These sections are General Requirements, Waste Container Requirements, Waste Form Requirements, and Waste Package Requirements. The waste package is the combination of waste container and waste. 10 refs., 1 fig.

Not Available

1989-01-01T23:59:59.000Z

338

Packaging and Transportation Safety  

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

The order establishes safety requirements for the proper packaging and transportation of DOE, including NNSA, offsite shipments and onsite transfers of radioactive and other hazardous materials and for modal transportation. Cancels DOE O 460.1B, 5-14-10

2010-05-14T23:59:59.000Z

339

Packaging and Transportation Safety  

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

To establish safety requirements for the proper packaging and transportation of Department of Energy (DOE)/National Nuclear Security Administration (NNSA) offsite shipments and onsite transfers of hazardous materials and for modal transport. Cancels DOE O 460.1A. Canceled by DOE O 460.1C.

2003-04-04T23:59:59.000Z

340

CH Packaging Maintenance Manual  

SciTech Connect (OSTI)

This procedure provides instructions for performing inner containment vessel (ICV) and outer containment vessel (OCV) maintenance and periodic leakage rate testing on the following packaging seals and corresponding seal surfaces using a nondestructive helium (He) leak test. In addition, this procedure provides instructions for performing ICV and OCV structural pressure tests.

Washington TRU Solutions

2002-01-02T23:59:59.000Z

Note: This page contains sample records for the topic "waste disposition packaging" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

Web Technology (elective package)  

E-Print Network [OSTI]

Web Technology (elective package) Offered by: Department of Mathematics and Computer Science? Computer Science-based approaches and enabling technologies for the web. Course descriptions Human and efficient. Web Technology The web has become the major source of information retrieval and is playing

Franssen, Michael

342

Management of Legacy Spent Nuclear Fuel Wastes at the Chalk River Laboratories: The Challenges and Innovative Solutions Implemented - 13301  

SciTech Connect (OSTI)

AECL's Fuel Packaging and Storage (FPS) Project was initiated in 2004 to retrieve, transfer, and stabilize an identified inventory of degraded research reactor fuel that had been emplaced within in-ground 'Tile Hole' structures in Chalk River Laboratories' Waste Management Area in the 1950's and 60's. Ongoing monitoring of the legacy fuel storage conditions had identified that moisture present in the storage structures had contributed to corrosion of both the fuel and the storage containers. This prompted the initiation of the FPS Project which has as its objective to design, construct, and commission equipment and systems that would allow for the ongoing safe storage of this fuel until a final long-term management, or disposition, pathway was available. The FPS Project provides systems and technologies to retrieve and transfer the fuel from the Waste Management Area to a new facility that will repackage, dry, safely store and monitor the fuel for a period of 50 years. All equipment and the new storage facility are designed and constructed to meet the requirements for Class 1 Nuclear Facilities in Canada. (authors)

Schruder, Kristan [Atomic Energy of Canada Limited - Chalk River Laboratories, Chalk River, Ontario (Canada)] [Atomic Energy of Canada Limited - Chalk River Laboratories, Chalk River, Ontario (Canada); Goodwin, Derek [Rolls-Royce Civil Nuclear Canada Limited, 678 Neal Dr., Peterborough, Ontario (Canada)] [Rolls-Royce Civil Nuclear Canada Limited, 678 Neal Dr., Peterborough, Ontario (Canada)

2013-07-01T23:59:59.000Z

343

Fissile material disposition program final immobilization form assessment and recommendation  

SciTech Connect (OSTI)

Lawrence Livermore National Laboratory (LLNL), in its role as the lead laboratory for the development of plutonium immobilization technologies for the Department of Energy`s Office of Fissile Materials Disposition (MD), has been requested by MD to recommend an immobilization technology for the disposition of surplus weapons- usable plutonium. The recommendation and supporting documentation was requested to be provided by September 1, 1997. This report addresses the choice between glass and ceramic technologies for immobilizing plutonium using the can-in-canister approach. Its purpose is to provide a comparative evaluation of the two candidate technologies and to recommend a form based on technical considerations.

Cochran, S.G.; Dunlop, W.H.; Edmunds, T.A.; MacLean, L.M.; Gould, T.H. [Westinghouse Savannah River Co., Aiken, SC (United States)

1997-10-03T23:59:59.000Z

344

Surrogate formulations for thermal treatment of low-level mixed waste, Part II: Selected mixed waste treatment project waste streams  

SciTech Connect (OSTI)

This report summarizes the formulation of surrogate waste packages, representing the major bulk constituent compositions for 12 waste stream classifications selected by the US DOE Mixed Waste Treatment Program. These waste groupings include: neutral aqueous wastes; aqueous halogenated organic liquids; ash; high organic content sludges; adsorbed aqueous and organic liquids; cement sludges, ashes, and solids; chloride; sulfate, and nitrate salts; organic matrix solids; heterogeneous debris; bulk combustibles; lab packs; and lead shapes. Insofar as possible, formulation of surrogate waste packages are referenced to authentic wastes in inventory within the DOE; however, the surrogate waste packages are intended to represent generic treatability group compositions. The intent is to specify a nonradiological synthetic mixture, with a minimal number of readily available components, that can be used to represent the significant challenges anticipated for treatment of the specified waste class. Performance testing and evaluation with use of a consistent series of surrogate wastes will provide a means for the initial assessment (and intercomparability) of candidate treatment technology applicability and performance. Originally the surrogate wastes were intended for use with emerging thermal treatment systems, but use may be extended to select nonthermal systems as well.

Bostick, W.D.; Hoffmann, D.P.; Chiang, J.M.; Hermes, W.H.; Gibson, L.V. Jr.; Richmond, A.A. [Martin Marietta Energy Systems, Inc., Oak Ridge, TN (United States)] [Martin Marietta Energy Systems, Inc., Oak Ridge, TN (United States); Mayberry, J. [Science Applications International Corp., Idaho Falls, ID (United States)] [Science Applications International Corp., Idaho Falls, ID (United States); Frazier, G. [Univ. of Tennessee, Knoxville, TN (United States)] [Univ. of Tennessee, Knoxville, TN (United States)

1994-01-01T23:59:59.000Z

345

Nevada Test Site Waste Acceptance Criteria  

SciTech Connect (OSTI)

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

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

2005-10-01T23:59:59.000Z

346

Waste treatment facility passes federal inspection, completes final  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear SecurityTensile Strain Switched FerromagnetismWaste and Materials Disposition Information Waste and

347

Cleanup Verification Package for the 100-F-20, Pacific Northwest Laboratory Parallel Pits  

SciTech Connect (OSTI)

This cleanup verification package documents completion of remedial action for the 100-F-20, Pacific Northwest Laboratory Parallel Pits waste site. This waste site consisted of two earthen trenches thought to have received both radioactive and nonradioactive material related to the 100-F Experimental Animal Farm.

M. J. Appel

2007-01-22T23:59:59.000Z

348

Plutonium stabilization and packaging system  

SciTech Connect (OSTI)

This document describes the functional design of the Plutonium Stabilization and Packaging System (Pu SPS). The objective of this system is to stabilize and package plutonium metals and oxides of greater than 50% wt, as well as other selected isotopes, in accordance with the requirements of the DOE standard for safe storage of these materials for 50 years. This system will support completion of stabilization and packaging campaigns of the inventory at a number of affected sites before the year 2002. The package will be standard for all sites and will provide a minimum of two uncontaminated, organics free confinement barriers for the packaged material.

NONE

1996-05-01T23:59:59.000Z

349

Waste Management Improvement Initiatives at Atomic Energy of Canada Limited - 13091  

SciTech Connect (OSTI)

Atomic Energy of Canada Limited's (AECL) Chalk River Laboratories (CRL) has been in operation for over 60 years. Radioactive, mixed, hazardous and non-hazardous wastes have been and continue to be generated at CRL as a result of research and development, radioisotope production, reactor operation and facility decommissioning activities. AECL has implemented several improvement initiatives at CRL to simplify the interface between waste generators and waste receivers: - Introduction of trained Waste Officers representing their facilities or activities at CRL; - Establishment of a Waste Management Customer Support Service as a Single-Point of Contact to provide guidance to waste generators for all waste management processes; and - Implementation of a streamlined approach for waste identification with emphasis on early identification of waste types and potential disposition paths. As a result of implementing these improvement initiatives, improvements in waste management and waste transfer efficiencies have been realized at CRL. These included: 1) waste generators contacting the Customer Support Service for information or guidance instead of various waste receivers; 2) more clear and consistent guidance provided to waste generators for waste management through the Customer Support Service; 3) more consistent and correct waste information provided to waste receivers through Waste Officers, resulting in reduced time and resources required for waste management (i.e., overall cost); 4) improved waste minimization and segregation approaches, as identified by in-house Waste Officers; and 5) enhanced communication between waste generators and waste management groups. (authors)

Chan, Nicholas; Adams, Lynne; Wong, Pierre [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)] [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

2013-07-01T23:59:59.000Z

350

Progress of the High Level Waste Program at the Defense Waste Processing Facility - 13178  

SciTech Connect (OSTI)

The Defense Waste Processing Facility at the Savannah River Site treats and immobilizes High Level Waste into a durable borosilicate glass for safe, permanent storage. The High Level Waste program significantly reduces environmental risks associated with the storage of radioactive waste from legacy efforts to separate fissionable nuclear material from irradiated targets and fuels. In an effort to support the disposition of radioactive waste and accelerate tank closure at the Savannah River Site, the Defense Waste Processing Facility recently implemented facility and flowsheet modifications to improve production by 25%. These improvements, while low in cost, translated to record facility production in fiscal years 2011 and 2012. In addition, significant progress has been accomplished on longer term projects aimed at simplifying and expanding the flexibility of the existing flowsheet in order to accommodate future processing needs and goals. (authors)

Bricker, Jonathan M.; Fellinger, Terri L.; Staub, Aaron V.; Ray, Jeff W.; Iaukea, John F. [Savannah River Remediation, Aiken, South Carolina, 29808 (United States)] [Savannah River Remediation, Aiken, South Carolina, 29808 (United States)

2013-07-01T23:59:59.000Z

351

Waste Form Development for the Solidification of PDCF/MOX Liquid Waste Streams  

SciTech Connect (OSTI)

At the Savannah River Site, part of the Department of Energy's nuclear materials complex located in South Carolina, cementation has been selected as the solidification method for high-alpha and low-activity waste streams generated in the planned plutonium disposition facilities. A Waste Solidification Building (WSB) that will be used to treat and solidify three radioactive liquid waste streams generated by the Pit Disassembly and Conversion Facility) and the Mixed Oxide Fuel Fabrication Facility is in the preliminary design stage. The WSB is expected to treat a transuranic (TRU) waste stream composed primarily of americium and two low-level waste (LLW) streams. The acidic wastes will be concentrated in the WSB evaporator and neutralized in a cement head tank prior to solidification. A series of TRU mixes were prepared to produce waste forms exhibiting a range of processing and cured properties. The LLW mixes were prepared using the premix from the preferred TRU waste form. All of the waste forms tested passed the Toxicity Characteristic Leaching Procedure. After processing in the WSB, current plans are to dispose of the solidified TRU waste at the Waste Isolation Pilot Plant in New Mexico and the solidified LLW waste at an approved low-level waste disposal facility.

COZZI, ALEX

2004-02-18T23:59:59.000Z

352

Hanford Site Transuranic (TRU) Waste Certification Plan  

SciTech Connect (OSTI)

As a generator of transuranic (TRU) and TRU mixed waste destined for disposal at the Waste Isolation Pilot Plant (WIPP), the Hanford Site must ensure that its TRU waste meets the requirements of US. Department of Energy (DOE) 0 435.1, ''Radioactive Waste Management,'' and the Contact-Handled (CH) Transuranic Waste Acceptance Criteria for the Waste Isolation Pilot Plant (WIPP-WAC). WIPP-WAC requirements are derived from the WIPP Technical Safety Requirements, WIPP Safety Analysis Report, TRUPACT-II SARP, WIPP Land Withdrawal Act, WIPP Hazardous Waste Facility Permit, and Title 40 Code of Federal Regulations (CFR) 191/194 Compliance Certification Decision. The WIPP-WAC establishes the specific physical, chemical, radiological, and packaging criteria for acceptance of defense TRU waste shipments at WIPP. The WPP-WAC also requires that participating DOE TRU waste generator/treatment/storage sites produce site-specific documents, including a certification plan, that describe their program for managing TRU waste and TRU waste shipments before transferring waste to WIPP. Waste characterization activities provide much of the data upon which certification decisions are based. Waste characterization requirements for TRU waste and TRU mixed waste that contains constituents regulated under the Resource Conservation and Recovery Act (RCRA) are established in the WIPP Hazardous Waste Facility Permit Waste Analysis Plan (WAP). The Hanford Site Quality Assurance Project Plan (QAPjP) (HNF-2599) implements the applicable requirements in the WAP and includes the qualitative and quantitative criteria for making hazardous waste determinations. The Hanford Site must also ensure that its TRU waste destined for disposal at WPP meets requirements for transport in the Transuranic Package Transporter-11 (TRUPACT-11). The US. Nuclear Regulatory Commission (NRC) establishes the TRUPACT-11 requirements in the Safety Analysis Report for the TRUPACT-II Shipping Package (TRUPACT-11 SARP). In addition, a TRU waste is eligible for disposal at WIPP only if it has been generated in whole or in part by one or more of the activities listed in Section 10101(3) of the Nuclear Waste Policy Act. DOE sites must determine that each waste stream to be disposed of at WIPP is ''defense'' TRU waste. (See also the definition of ''defense'' TRU waste.). Only CH TRU wastes meeting the requirements of the QAPjP, WIPP-WAP, WPP-WAC, and other requirements documents described above will be accepted for transportation and disposal at WIPP.

GREAGER, T.M.

2000-12-06T23:59:59.000Z

353

Hanford Site Transuranic (TRU) Waste Certification Plan  

SciTech Connect (OSTI)

As a generator of transuranic (TRU) and TRU mixed waste destined for disposal at the Waste Isolation Pilot Plant (WIPP), the Hanford Site must ensure that its TRU waste meets the requirements of US. Department of Energy (DOE) 0 435.1, ''Radioactive Waste Management,'' and the Contact-Handled (CH) Transuranic Waste Acceptance Criteria for the Waste Isolation Pilot Plant (WIPP-WAC). WIPP-WAC requirements are derived from the WIPP Technical Safety Requirements, WIPP Safety Analysis Report, TRUPACT-II SARP, WIPP Land Withdrawal Act, WIPP Hazardous Waste Facility Permit, and Title 40 Code of Federal Regulations (CFR) 191/194 Compliance Certification Decision. The WIPP-WAC establishes the specific physical, chemical, radiological, and packaging criteria for acceptance of defense TRU waste shipments at WIPP. The WPP-WAC also requires that participating DOE TRU waste generator/treatment/storage sites produce site-specific documents, including a certification plan, that describe their program for managing TRU waste and TRU waste shipments before transferring waste to WIPP. Waste characterization activities provide much of the data upon which certification decisions are based. Waste characterization requirements for TRU waste and TRU mixed waste that contains constituents regulated under the Resource Conservation and Recovery Act (RCRA) are established in the WIPP Hazardous Waste Facility Permit Waste Analysis Plan (WAP). The Hanford Site Quality Assurance Project Plan (QAPjP) (HNF-2599) implements the applicable requirements in the WAP and includes the qualitative and quantitative criteria for making hazardous waste determinations. The Hanford Site must also ensure that its TRU waste destined for disposal at WPP meets requirements for transport in the Transuranic Package Transporter-11 (TRUPACT-11). The US. Nuclear Regulatory Commission (NRC) establishes the TRUPACT-11 requirements in the Safety Analysis Report for the TRUPACT-II Shipping Package (TRUPACT-11 SARP). In addition, a TRU waste is eligible for disposal at WIPP only if it has been generated in whole or in part by one or more of the activities listed in Section 10101(3) of the Nuclear Waste Policy Act. DOE sites must determine that each waste stream to be disposed of at WIPP is ''defense'' TRU waste. (See also the definition of ''defense'' TRU waste.). Only CH TRU wastes meeting the requirements of the QAPjP, WIPP-WAP, WPP-WAC, and other requirements documents described above will be accepted for transportation and disposal at WIPP.

GREAGER, T.M.

2000-12-01T23:59:59.000Z

354

Release Data Package for Hanford Site Assessments  

SciTech Connect (OSTI)

Beginning in fiscal year (FY) 2003, the U.S. Department of Energy (DOE) Richland Operations Office initiated activities, including the development of data packages, to support a Hanford assessment. This report describes the data compiled in FY 2003 through 2005 to support the Release Module of the System Assessment Capability (SAC) for the updated composite analysis. This work was completed as part of the Characterization of Systems Project, part of the Remediation and Closure Science Project, the Hanford Assessments Project, and the Characterization of Systems Project managed by Pacific Northwest National Laboratory. Related characterization activities and data packages for the vadose zone and groundwater are being developed under the remediation Decision Support Task of the Groundwater Remediation Project managed by Fluor Hanford, Inc. The Release Module applies release models to waste inventory data from the Inventory Module and accounts for site remediation activities as a function of time. The resulting releases to the vadose zone, expressed as time profiles of annual rates, become source terms for the Vadose Zone Module. Radioactive decay is accounted for in all inputs and outputs of the Release Module. The Release Module is implemented as the VADER (Vadose zone Environmental Release) computer code. Key components of the Release Module are numerical models (i.e., liquid, soil-debris, cement, saltcake, and reactor block) that simulate contaminant release from the different waste source types found at the Hanford Site. The Release Module also handles remediation transfers to onsite and offsite repositories.

Riley, Robert G.; Lopresti, Charles A.; Engel, David W.

2006-07-01T23:59:59.000Z

355

Transmittal of the Calculation Package that Supports the Analysis of Performance of the Environmental Management Waste Management Facility Oak Ridge, Tennessee (Based 5-Cell Design Issued 8/14/09)  

SciTech Connect (OSTI)

This document presents the results of an assessment of the performance of a build-out of the Environmental Management Waste Management Facility (EMWMF). The EMWMF configuration that was assessed includes the as-constructed Cells 1 through 4, with a groundwater underdrain that was installed beneath Cell 3 during the winter of 2003-2004, and Cell 5, whose proposed design is an Addendum to Remedial Design Report for the Disposal of Oak Ridge Reservation Comprehensive Environmental Response, Compensation, and Liability Act of 1980 Waste, Oak Ridge, Tennessee, DOE/OR/01-1873&D2/A5/R1. The total capacity of the EMWMF with 5 cells is about 1.7 million cubic yards. This assessment was conducted to determine the conditions under which the approved Waste Acceptance Criteria (WAC) for the EMWMF found in the Attainment Plan for Risk/Toxicity-Based Waste Acceptance Criteria at the Oak Ridge Reservation, Oak Ridge, Tennessee [U.S. Department of Energy (DOE) 2001a], as revised for constituents added up to October 2008, would remain protective of public health and safety for a five-cell disposal facility. For consistency, the methods of analyses and the exposure scenario used to predict the performance of a five-cell disposal facility were identical to those used in the Remedial Investigation and Feasibility Study (RI/FS) and its addendum (DOE 1998a, DOE 1998b) to develop the approved WAC. To take advantage of new information and design changes departing from the conceptual design, the modeling domain and model calibration were upaded from those used in the RI/FS and its addendum. It should be noted that this analysis is not intended to justify or propose a change in the approved WAC.

Williams M.J.

2009-09-14T23:59:59.000Z

356

Integrated Waste Management Strategy and Radioactive Waste Forms for the 21st Century  

SciTech Connect (OSTI)

The U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP) was announced in 2006. As currently envisioned, GNEP will be the basis for growth of nuclear energy worldwide, using a closed proliferation-resistant fuel cycle. The Integrated Waste Management Strategy (IWMS) is designed to ensure that all wastes generated by fuel fabrication and recycling will have a routine disposition path making the most of feedback to fuel and recycling operations to eliminate or minimize byproducts and wastes. If waste must be generated, processes will be designed with waste treatment in mind to reduce use of reagents that complicate stabilization and minimize volume. The IWMS will address three distinct levels of technology investigation and systems analyses and will provide a cogent path from (1) research and development (R&D) and engineering scale demonstration, (Level I); to (2) full scale domestic deployment (Level II); and finally to (3) establishing an integrated global nuclear energy infrastructure (Level III). The near-term focus of GNEP is on achieving a basis for large-scale commercial deployment (Level II), including the R&D and engineering scale activities in Level I that are necessary to support such an accomplishment. Throughout these levels is the need for innovative thinking to simplify, including regulations, separations and waste forms to minimize the burden of safe disposition of wastes on the fuel cycle.

Dirk Gombert; Jay Roach

2007-03-01T23:59:59.000Z

357

Waste Information Management System with 2012-13 Waste Streams - 13095  

SciTech Connect (OSTI)

The Waste Information Management System (WIMS) 2012-13 was updated to support the Department of Energy (DOE) accelerated cleanup program. The schedule compression required close coordination and a comprehensive review and prioritization of the barriers that impeded treatment and disposition of the waste streams at each site. Many issues related to waste treatment and disposal were potential critical path issues under the accelerated schedule. In order to facilitate accelerated cleanup initiatives, waste managers at DOE field sites and at DOE Headquarters in Washington, D.C., needed timely waste forecast and transportation information regarding the volumes and types of radioactive waste that would be generated by DOE sites over the next 40 years. Each local DOE site historically collected, organized, and displayed waste forecast information in separate and unique systems. In order for interested parties to understand and view the complete DOE complex-wide picture, the radioactive waste and shipment information of each DOE site needed to be entered into a common application. The WIMS application was therefore created to serve as a common application to improve stakeholder comprehension and improve DOE radioactive waste treatment and disposal planning and scheduling. WIMS allows identification of total forecasted waste volumes, material classes, disposition sites, choke points, technological or regulatory barriers to treatment and disposal, along with forecasted waste transportation information by rail, truck and inter-modal shipments. The Applied Research Center (ARC) at Florida International University (FIU) in Miami, Florida, developed and deployed the web-based forecast and transportation system and is responsible for updating the radioactive waste forecast and transportation data on a regular basis to ensure the long-term viability and value of this system. (authors)

Upadhyay, H.; Quintero, W.; Lagos, L.; Shoffner, P.; Roelant, D. [Applied Research Center, Florida International University, 10555 West Flagler Street, Suite 2100, Miami, FL 33174 (United States)] [Applied Research Center, Florida International University, 10555 West Flagler Street, Suite 2100, Miami, FL 33174 (United States)

2013-07-01T23:59:59.000Z

358

Dangerous Waste Characteristics of Contact-Handled Transuranic Mixed Wastes from the Hanford Tanks  

SciTech Connect (OSTI)

This report summarizes existing analytical data from samples taken from the Hanford tanks designated as potentially containing transuranic mixed process wastes. Process knowledge of the wastes transferred to these tanks has been reviewed to determine whether the dangerous waste characteristics now assigned to all Hanford underground storage tanks are applicable to these particular wastes. Supplemental technologies are being examined to accelerate the Hanford tank waste cleanup mission and accomplish waste treatment safely and efficiently. To date, 11 Hanford waste tanks have been designated as potentially containing contact-handled (CH) transuranic mixed (TRUM) wastes. The CH-TRUM wastes are found in single-shell tanks B-201 through B-204, T-201 through T-204, T-104, T-110, and T-111. Methods and equipment to solidify and package the CH-TRUM wastes are part of the supplemental technologies being evaluated. The resulting packages and wastes must be acceptable for disposal at the Waste Isolation Pilot Plant (WIPP). The dangerous waste characteristics being considered include ignitability, corrosivity, reactivity, and toxicity arising from the presence of 2,4,5-trichlorophenol at levels above the dangerous waste threshold. The analytical data reviewed include concentrations of sulfur, sulfate, cyanide, 2,4,5-trichlorophenol, total organic carbon, and oxalate; the composition of the tank headspace, pH, and mercury. Differential scanning calorimetry results were used to determine the energetics of the wastes as a function of temperature.

Tingey, Joel M.; Bryan, Garry H.; Deschane, Jaquetta R.

2004-08-31T23:59:59.000Z

359

Assessment of Quality Assurance Measures for Radioactive Material Transport Packages not Requiring Competent Authority Design Approval - 13282  

SciTech Connect (OSTI)

The majority of transports of radioactive materials are carried out in packages which don't need a package design approval by a competent authority. Low-active radioactive materials are transported in such packages e.g. in the medical and pharmaceutical industry and in the nuclear industry as well. Decommissioning of NPP's leads to a strong demand for packages to transport low and middle active radioactive waste. According to IAEA regulations the 'non-competent authority approved package types' are the Excepted Packages and the Industrial Packages of Type IP-1, IP-2 and IP-3 and packages of Type A. For these types of packages an assessment by the competent authority is required for the quality assurance measures for the design, manufacture, testing, documentation, use, maintenance and inspection (IAEA SSR 6, Chap. 306). In general a compliance audit of the manufacturer of the packaging is required during this assessment procedure. Their regulatory level in the IAEA regulations is not comparable with the 'regulatory density' for packages requiring competent authority package design approval. Practices in different countries lead to different approaches within the assessment of the quality assurance measures in the management system as well as in the quality assurance program of a special package design. To use the package or packaging in a safe manner and in compliance with the regulations a management system for each phase of the life of the package or packaging is necessary. The relevant IAEA-SSR6 chap. 801 requires documentary verification by the consignor concerning package compliance with the requirements. (authors)

Komann, Steffen; Groeke, Carsten; Droste, Bernhard [BAM Federal Institute for Materials Research and Testing, Unter den Eichen 44-46, 12203 Berlin (Germany)] [BAM Federal Institute for Materials Research and Testing, Unter den Eichen 44-46, 12203 Berlin (Germany)

2013-07-01T23:59:59.000Z

360

1 INSTRODUCTION In the concept of geological radioactive waste disposal,  

E-Print Network [OSTI]

1 INSTRODUCTION In the concept of geological radioactive waste disposal, argillite is being of the radioactive waste disposal, the host rock will be subjected to various thermo-hydro-mechanical loadings, thermal solicitation comes from the heat emitting from the radioactive waste packages. On one hand

Boyer, Edmond

Note: This page contains sample records for the topic "waste disposition packaging" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

Development of a path forward for special-case wastes at the Oak Ridge Reservation  

SciTech Connect (OSTI)

This report addresses the management of the inventory of existing and potential surplus equipment and materials at the Oak Ridge Reservation (ORR) that are candidates for various waste or surplus material categories, including special case waste (SCW). This inventory is called candidate equipment and materials (CEM). This report presents a logical method for disposition of this and future CEM, summarizes the inventory, and suggests preliminary dispositions for the CEM. Also, recommendations are offered for an improved CEM management strategy and actions in this and future years to implement that strategy.

Osborne-Lee, I.W. [Oak Ridge National Lab., TN (United States); Lotts, A.L. [Delta-21 Resources, Inc., Oak Ridge, TN (United States); Robinette, R.J. [H and R Technical Associates, Inc., Oak Ridge, TN (United States)

1997-10-01T23:59:59.000Z

362

Remaining Sites Verification Package for the 100-F-26:10, 1607-F3 Sanitary Sewer Pipelines (182-F, 183-F, and 151-F Sanitary Sewer Lines), Waste Site Reclassification Form 2007-028  

SciTech Connect (OSTI)

The 100-F-26:10 waste site includes sanitary sewer lines that serviced the former 182-F, 183-F, and 151-F Buildings. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

L. M. Dittmer

2007-12-03T23:59:59.000Z

363

Radioactive and nonradioactive waste intended for disposal at the Waste Isolation Pilot Plant  

SciTech Connect (OSTI)

Transuranic (TRU) waste generated by the handling of plutonium in research on or production of US nuclear weapons will be disposed of in the Waste Isolation Pilot Plant (WIPP). This paper describes the physical and radiological properties of the TRU waste that will be deposited in the WIPP. This geologic repository will accommodate up to 175,564 m{sup 3} of TRU waste, corresponding to 168,485 m{sup 3} of contact-handled (CH-) TRU waste and 7,079 m{sup 3} of remote-handled (RH-) TRU waste. Approximately 35% of the TRU waste is currently packaged and stored (i.e., legacy) waste, with the remainder of the waste to be packaged or generated and packaged in activities before the year 2033, the closure time for the repository. These wastes were produced at 27 US Department of Energy (DOE) sites in the course of generating defense nuclear materials. The radionuclide and nonradionuclide inventories for the TRU wastes described in this paper were used in the 1996 WIPP Compliance Certification Application (CCA) performance assessment calculations by Sandia National Laboratories/New Mexico (SNL/NM).

SANCHEZ,LAWRENCE C.; DREZ,P.E.; RATH,JONATHAN S.; TRELLUE,H.R.

2000-05-19T23:59:59.000Z

364

Optimizing High Level Waste Disposal  

SciTech Connect (OSTI)

If society is ever to reap the potential benefits of nuclear energy, technologists must close the fuel-cycle completely. A closed cycle equates to a continued supply of fuel and safe reactors, but also reliable and comprehensive closure of waste issues. High level waste (HLW) disposal in borosilicate glass (BSG) is based on 1970s era evaluations. This host matrix is very adaptable to sequestering a wide variety of radionuclides found in raffinates from spent fuel reprocessing. However, it is now known that the current system is far from optimal for disposal of the diverse HLW streams, and proven alternatives are available to reduce costs by billions of dollars. The basis for HLW disposal should be reassessed to consider extensive waste form and process technology research and development efforts, which have been conducted by the United States Department of Energy (USDOE), international agencies and the private sector. Matching the waste form to the waste chemistry and using currently available technology could increase the waste content in waste forms to 50% or more and double processing rates. Optimization of the HLW disposal system would accelerate HLW disposition and increase repository capacity. This does not necessarily require developing new waste forms, the emphasis should be on qualifying existing matrices to demonstrate protection equal to or better than the baseline glass performance. Also, this proposed effort does not necessarily require developing new technology concepts. The emphasis is on demonstrating existing technology that is clearly better (reliability, productivity, cost) than current technology, and justifying its use in future facilities or retrofitted facilities. Higher waste processing and disposal efficiency can be realized by performing the engineering analyses and trade-studies necessary to select the most efficient methods for processing the full spectrum of wastes across the nuclear complex. This paper will describe technologies being evaluated at Idaho National Laboratory and the facilities we’ve designed to evaluate options and support optimization.

Dirk Gombert

2005-09-01T23:59:59.000Z

365

LEVERAGING AGING MATERIALS DATA TO SUPPORT EXTENSION OF TRANSPORTATION SHIPPING PACKAGES SERVICE LIFE  

SciTech Connect (OSTI)

Nuclear material inventories are increasingly being transferred to interim storage locations where they may reside for extended periods of time. Use of a shipping package to store nuclear materials after the transfer has become more common for a variety of reasons. Shipping packages are robust and have a qualified pedigree for performance in normal operation and accident conditions but are only certified over an approved transportation window. The continued use of shipping packages to contain nuclear material during interim storage will result in reduced overall costs and reduced exposure to workers. However, the shipping package materials of construction must maintain integrity as specified by the safety basis of the storage facility throughout the storage period, which is typically well beyond the certified transportation window. In many ways, the certification processes required for interim storage of nuclear materials in shipping packages is similar to life extension programs required for dry cask storage systems for commercial nuclear fuels. The storage of spent nuclear fuel in dry cask storage systems is federally-regulated, and over 1500 individual dry casks have been in successful service up to 20 years in the US. The uncertainty in final disposition will likely require extended storage of this fuel well beyond initial license periods and perhaps multiple re-licenses may be needed. Thus, both the shipping packages and the dry cask storage systems require materials integrity assessments and assurance of continued satisfactory materials performance over times not considered in the original evaluation processes. Test programs for the shipping packages have been established to obtain aging data on materials of construction to demonstrate continued system integrity. The collective data may be coupled with similar data for the dry cask storage systems and used to support extending the service life of shipping packages in both transportation and storage.

Dunn, K. [Savannah River National Laboratory; Bellamy, S. [Savannah River National Laboratory; Daugherty, W. [Savannah River National Laboratory; Sindelar, R. [Savannah River National Laboratory; Skidmore, E. [Savannah River National Laboratory

2013-08-18T23:59:59.000Z

366

Waste Treatment Technology Process Development Plan For Hanford Waste Treatment Plant Low Activity Waste Recycle  

SciTech Connect (OSTI)

The purpose of this Process Development Plan is to summarize the objectives and plans for the technology development activities for an alternative path for disposition of the recycle stream that will be generated in the Hanford Waste Treatment Plant Low Activity Waste (LAW) vitrification facility (LAW Recycle). This plan covers the first phase of the development activities. The baseline plan for disposition of this stream is to recycle it to the WTP Pretreatment Facility, where it will be concentrated by evaporation and returned to the LAW vitrification facility. Because this stream contains components that are volatile at melter temperatures and are also problematic for the glass waste form, they accumulate in the Recycle stream, exacerbating their impact on the number of LAW glass containers. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to acceptable concentrations in the LAW glass, and reducing the halides in the Recycle is a key component of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, this stream does not have a proven disposition path, and resolving this gap becomes vitally important. This task seeks to examine the impact of potential future disposition of this stream in the Hanford tank farms, and to develop a process that will remove radionuclides from this stream and allow its diversion to another disposition path, greatly decreasing the LAW vitrification mission duration and quantity of glass waste. The origin of this LAW Recycle stream will be from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW melter off-gas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover or precipitates of scrubbed components (e.g. carbonates). The soluble components are mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet, and will not be available until the WTP begins operation, causing uncertainty in its composition, particularly the radionuclide content. This plan will provide an estimate of the likely composition and the basis for it, assess likely treatment technologies, identify potential disposition paths, establish target treatment limits, and recommend the testing needed to show feasibility. Two primary disposition options are proposed for investigation, one is concentration for storage in the tank farms, and the other is treatment prior to disposition in the Effluent Treatment Facility. One of the radionuclides that is volatile and expected to be in high concentration in this LAW Recycle stream is Technetium-99 ({sup 99}Tc), a long-lived radionuclide with a half-life of 210,000 years. Technetium will not be removed from the aqueous waste in the Hanford Waste Treatment and Immobilization Plant (WTP), and will primarily end up immobilized in the LAW glass, which will be disposed in the Integrated Disposal Facility (IDF). Because {sup 99}Tc has a very long half-life and is highly mobile, it is the largest dose contributor to the Performance Assessment (PA) of the IDF. Other radionuclides that are also expected to be in appreciable concentration in the LAW Recycle are {sup 129}I, {sup 90}Sr, {sup 137}Cs, and {sup 241}Am. The concentrations of these radionuclides in this stream will be much lower than in the LAW, but they will still be higher than limits for some of the other disposition pathways currently available. Although the baseline process will recycle this stream to the Pretreatment Facility, if the LAW facility begins operation first, this stream will not have a disposition path internal to WTP. One potential solution is to return the stream to the tank farms where it can be evaporated in the 242-A evaporator, or perhaps deploy an auxiliary evaporator to concentrate it prior to return to the tank farms. In either case, testing is needed to evaluat

McCabe, Daniel J.; Wilmarth, William R.; Nash, Charles A.

2013-08-29T23:59:59.000Z

367

An expert system framework for nondestructive waste assay  

SciTech Connect (OSTI)

Management and disposition of transuranic (RU) waste forms necessitates determining entrained RU and associated radioactive material quantities as per National RU Waste Characterization Program requirements. Technical justification and demonstration of a given NDA method used to determine RU mass and uncertainty in accordance with program quality assurance is difficult for many waste forms. Difficulties are typically founded in waste NDA methods that employ standards compensation and/or employment of simplifying assumptions on waste form configurations. Capability to determine and justify RU mass and mass uncertainty can be enhanced through integration of waste container data/information using expert system and empirical data-driven techniques with conventional data acquisition and analysis. Presented is a preliminary expert system framework that integrates the waste form data base, alogrithmic techniques, statistical analyses, expert domain knowledge bases, and empirical artificial intelligence modules into a cohesive system. The framework design and bases in addition to module development activities are discussed.

Becker, G.K.

1996-10-01T23:59:59.000Z

368

Volume reduction of hot cell plastic wastes  

SciTech Connect (OSTI)

The disposal of radioactively-contaminated solid wastes has become a national crisis. In such circumstances, it is imperative that this waste be reduced to minimum volume and be packaged to prevent pollution of the environment. The majority of the solid waste generated at the hot cell under consideration is plastic lab ware. Cutting this waste into small pieces with a hot wire technique reduced the volume 66%. Melting the waste, although more time consuming, reduced the volume 90%. The hot wire technique can also be used to cut up damaged master slave manipulator boots, greatly reducing their disposal volume.

Dykes, F W; Henscheid, J P; Lewis, L C; Lundholm, C W; Nicklas, J H

1989-09-19T23:59:59.000Z

369

Grant Application Package CFDA Number  

E-Print Network [OSTI]

Grant Application Package CFDA Number: Opportunity Title: Offering Agency: Agency Contact: Opportunity Open Date: Opportunity Close Date: CFDA Description: Opportunity Number: Competition ID

Talley, Lynne D.

370

Seal welded cast iron nuclear waste container  

DOE Patents [OSTI]

This invention identifies methods and articles designed to circumvent metallurgical problems associated with hermetically closing an all cast iron nuclear waste package by welding. It involves welding nickel-carbon alloy inserts which are bonded to the mating plug and main body components of the package. The welding inserts might be bonded in place during casting of the package components. When the waste package closure weld is made, the most severe thermal effects of the process are restricted to the nickel-carbon insert material which is far better able to accommodate them than is cast iron. Use of nickel-carbon weld inserts should eliminate any need for pre-weld and post-weld heat treatments which are a problem to apply to nuclear waste packages. Although the waste package closure weld approach described results in a dissimilar metal combination, the relative surface area of nickel-to-iron, their electrochemical relationship, and the presence of graphite in both materials will act to prevent any galvanic corrosion problem.

Filippi, Arthur M. (Pittsburgh, PA); Sprecace, Richard P. (Murrysville, PA)

1987-01-01T23:59:59.000Z

371

The Packaging Handbook -- A guide to package design  

SciTech Connect (OSTI)

The Packaging Handbook is a compilation of 14 technical chapters and five appendices that address the life cycle of a packaging which is intended to transport radioactive material by any transport mode in normal commerce. Although many topics are discussed in depth, this document focuses on the design aspects of a packaging. The Handbook, which is being prepared under the direction of the US Department of Energy, is intended to provide a wealth of technical guidance that will give designers a better understanding of the regulatory approval process, preferences of regulators in specific aspects of packaging design, and the types of analyses that should be seriously considered when developing the packaging design. Even though the Handbook is concerned with all packagings, most of the emphasis is placed on large packagings that are capable of transporting large radioactive sources that are also fissile (e.g., spent fuel). These are the types of packagings that must address the widest range of technical topics in order to meet domestic and international regulations. Most of the chapters in the Handbook have been drafted and submitted to the Oak Ridge National Laboratory for editing; the majority of these have been edited. This report summarizes the contents.

Shappert, L.B.

1995-12-31T23:59:59.000Z

372

Preliminary siting characterization Salt Disposition Facility - Site B  

SciTech Connect (OSTI)

A siting and reconnaissance geotechnical program has been completed in S-Area at the Savannah River Site in South Carolina. This program investigated the subsurface conditions for the area known as ``Salt Disposition Facility (SDF), Site B'' located northeast of H-Area and within the S-Area. Data acquired from the Site B investigation includes both field exploration and laboratory test data.

Wyatt, D.

2000-01-04T23:59:59.000Z

373

Controlling Beryllium Contaminated Material And Equipment For The Building 9201-5 Legacy Material Disposition Project  

SciTech Connect (OSTI)

This position paper addresses the management of beryllium contamination on legacy waste. The goal of the beryllium management program is to protect human health and the environment by preventing the release of beryllium through controlling surface contamination. Studies have shown by controlling beryllium surface contamination, potential airborne contamination is reduced or eliminated. Although there are areas in Building 9201-5 that are contaminated with radioactive materials and mercury, only beryllium contamination is addressed in this management plan. The overall goal of this initiative is the compliant packaging and disposal of beryllium waste from the 9201-5 Legacy Material Removal (LMR) Project to ensure that beryllium surface contamination and any potential airborne release of beryllium is controlled to levels as low as practicable in accordance with 10 CFR 850.25.

Reynolds, T. D.; Easterling, S. D.

2010-10-01T23:59:59.000Z

374

Degradation mode survey galvanic corrosion of candidate metallic materials for high-level radioactive waste disposal containers  

SciTech Connect (OSTI)

This report presents the results of a literature review on galvanic corrosion between candidate metals and alloys currently being considered for the waste package containers.

Roy, A.K.; Jones, D.A.; McCright, R.D.

1996-08-01T23:59:59.000Z

375

Processing and Disposition of Special Actinide Target Materials - 13138  

SciTech Connect (OSTI)

The Department of Energy (DOE) manages an inventory of materials that contains a range of long-lived radioactive isotopes that were produced from the 1960's through the 1980's by irradiating targets in high-flux reactors at the Savannah River Site (SRS) to produce special heavy isotopes for DOE programmatic use, scientific research, and industrial and medical applications. Among the products were californium-252, heavy curium (including Cm-246 through Cm-248), and plutonium-242 and -244. Many of the isotopes are still in demand today, and they can be recovered from the remaining targets previously irradiated at SRS or produced from the recovered isotopes. Should the existing target materials be discarded, the plutonium (Pu) and curium (Cm) isotopes cannot be replaced readily with existing production sources. Some of these targets are stored at SRS, while other target material is stored at Oak Ridge National Laboratory (ORNL) at several stages of processing. The materials cannot be stored in their present form indefinitely. Their long-term management involves processing items for beneficial use and/or for disposition, using storage and process facilities at SRS and ORNL. Evaluations are under way for disposition options for these materials, and demonstrations of improved flow sheets to process the materials are being conducted at ORNL and the Savannah River National Laboratory (SRNL). The disposition options and a management evaluation process have been developed. Processing demonstrations and evaluations for these unique materials are under way. (authors)

Robinson, Sharon M.; Patton, Brad D. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States)] [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Allender, Jeffrey S. [Savannah River National Laboratory (United States)] [Savannah River National Laboratory (United States)

2013-07-01T23:59:59.000Z

376

Technical Evaluations of Proposed Remote-Handled Transuranic Waste Characterization Requirements at WIPP  

SciTech Connect (OSTI)

Characterization, packaging, transport, handling and disposal of remotely handled transuranic (RH TRU) waste at WIPP will be different than similar operations with contact handled transuranic (CH TRU) waste. This paper presents results of technical evaluations associated with the planned disposal of remotely handled transuranic waste at the Waste Isolation Pilot Plant (WIPP).

Anastas, G.; Channell, J. K.

2002-02-26T23:59:59.000Z

377

Praxis I/O package  

SciTech Connect (OSTI)

The Praxis language specification, like Algol and Ada, does not specify any I/O statements. The intent was to provide a standard I/O package as a companion to the compiler. This would allow the user to substitute, or supplement, the I/O package, as needed, for specialized applications. Like Algol, however, Praxis provided only limited (text) I/O for several years. Ada, in contrast, provided a comprehensive standard I/O package from its inception. Digital Equipment Corporation's (DEC's) implementation of Ada, on their VAX family of computers, further supplemented this package with other packages which exploit the I/O facilities available under the VMS operating system. The Praxis I/O package described in this document has been modeled after DEC's implementation of Ada and provides a similar set of I/O facilities. Currently, the I/O package is supported only under VAX/VMS. The design of the package, however, is essentially independent of any operating system (with the exception of the module COMMAND IO). The VAX/VMS version of the I/O package fully exploits the vast I/O facilities which are provided under VAX/VMS and makes them directly available to the Praxis programmer. The design, prototype implementation, and draft documentation of the Praxis I/O Package was done by Tim Sherman as part of a University project in computer science. Subsequent work by both Tim and Fred Holloway lead to a more complete implementation, testing and development of example programs, and inclusion of the package into the Praxis compilers as their principal interface to RMS and VMS.

Holloway, F.W.; Sherman, T.A.

1988-04-07T23:59:59.000Z

378

MANAGING THE RETRIEVAL RISK OF BURIED TRANSURANIC (TRU) WASTE WITH UNIQUE CHARACTERISTICS  

SciTech Connect (OSTI)

United States-Department of Energy (DOE) sites that store transuranic (TRU) waste are almost certain to encounter waste packages with characteristics that are so unique as to warrant special precautions for retrieval. At the Hanford Site, a subgroup of stored TRU waste (12 drums) had special considerations due to the radioactive source content of plutonium oxide (PuO{sub 2}), and the potential for high heat generation, pressurization, criticality, and high radiation. These characteristics bear on the approach to safely retrieve, overpack, vent, store, and transport the waste package. Because of the potential risk to personnel, contingency planning for unexpected conditions played an effective role in work planning and in preparing workers for the field inspection activity. As a result, the integrity inspections successfully confirmed waste package configuration and waste confinement without experiencing any perturbations due to unanticipated packaging conditions. This paper discusses the engineering and field approach to managing the risk of retrieving TRU waste with unique characteristics.

WOJTASEK, R.D.; GADD, R.R.; GREENWELL, R.D.

2006-01-19T23:59:59.000Z

379

MANAGEING THE RETRIEVAL RISK OF BURIED TRANSURANIC (TRU) WASTE WITH UNIQUE CHARACTERISTICS  

SciTech Connect (OSTI)

United States-Department of Energy (DOE) sites that store transuranic (TRU) waste are almost certain to encounter waste packages with characteristics that are so unique as to warrant special precautions for retrieval. At the Hanford Site, a subgroup of stored TRU waste (12 drums) had special considerations due to the radioactive source content of plutonium oxide (PuO{sub 2}), and the potential for high heat generation, pressurization, criticality, and high radiation. These characteristics bear on the approach to safely retrieve, overpack, vent, store, and transport the waste package. Because of the potential risk to personnel, contingency planning for unexpected conditions played an effective roll in work planning and in preparing workers for the field inspection activity. As a result, the integrity inspections successfully confirmed waste package configuration and waste confinement without experiencing any perturbations due to unanticipated packaging conditions. This paper discusses the engineering and field approach to managing the risk of retrieving TRU waste with unique characteristics.

WOJTASEK, R.D.; GREENWELL, R.D.

2005-11-17T23:59:59.000Z

380

Building an R package Division of Biostatistics  

E-Print Network [OSTI]

Building an R package Yen-Yi Ho Division of Biostatistics School of Public Health, University of Minnesota Yen-Yi Ho Building an R package #12;Steps Prepare your functions, example data sets Build package in man subdirectory) Write a package vignette Build and install the R package (R CMD build) Check the R

Carlin, Bradley P.

Note: This page contains sample records for the topic "waste disposition packaging" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

ENVIROCARE OF UTAH: EXPANDING WASTE ACCEPTANCE CRITERIA TO PROVIDE LOW-LEVEL AND MIXED WASTE DISPOSAL OPTIONS  

SciTech Connect (OSTI)

Envirocare of Utah operates a low-level radioactive waste disposal facility 80 miles west of Salt Lake City in Clive, Utah. Accepted waste types includes NORM, 11e2 byproduct material, Class A low-level waste, and mixed waste. Since 1988, Envirocare has offered disposal options for environmental restoration waste for both government and commercial remediation projects. Annual waste receipts exceed 12 million cubic feet. The waste acceptance criteria (WAC) for the Envirocare facility have significantly expanded to accommodate the changing needs of restoration projects and waste generators since its inception, including acceptable physical waste forms, radiological acceptance criteria, RCRA requirements and treatment capabilities, PCB acceptance, and liquids acceptance. Additionally, there are many packaging, transportation, and waste management options for waste streams acceptable at Envirocare. Many subcontracting vehicles are also available to waste generators for both government and commercial activities.

Rogers, B.; Loveland, K.

2003-02-27T23:59:59.000Z

382

Upgrades to meet LANL SF, 121-2011, hazardous waste facility permit requirements  

SciTech Connect (OSTI)

Members of San IIdefonso have requested information from LANL regarding implementation of the revision to LANL's Hazardous Waste Facility Permit (the RCRA Permit). On January 26, 2011, LANL staff from the Waste Disposition Project and the Environmental Protection Division will provide a status update to Pueblo members at the offices of the San IIdefonso Department of Environmental and Cultural Preservation. The Waste Disposition Project presentation will focus on upgrades and improvements to LANL waste management facilities at TA-50 and TA-54. The New Mexico Environment Department issued LANL's revised Hazardous Waste Facility permit on November 30, 2010 with a 30-day implementation period. The Waste Disposition Project manages and operates four of LANL's permitted facilities; the Waste Characterization, Reduction and Repackaging Facility (WCRRF) at TA-SO, and Area G, Area L and the Radioassay and Nondestructive Testing facility (RANT) at TA-54. By implementing a combination of permanent corrective action activities and shorter-term compensatory measures, WDP was able to achieve functional compliance on December 30, 2010 with new Permit requirements at each of our facilities. One component of WOP's mission at LANL is centralized management and disposition of the Laboratory's hazardous and mixed waste. To support this mission objective, WOP has undertaken a project to upgrade our facilities and equipment to achieve fully compliant and efficient waste management operations. Upgrades to processes, equipment and facilities are being designed to provide defense-in-depth beyond the minimum, regulatory requirements where worker safety and protection of the public and the environment are concerned. Upgrades and improvements to enduring waste management facilities and operations are being designed so as not to conflict with future closure activities at Material Disposal Area G and Material Disposal Area L.

French, Sean B [Los Alamos National Laboratory; Johns - Hughes, Kathryn W [Los Alamos National Laboratory

2011-01-21T23:59:59.000Z

383

Electro-Microfluidic Packaging  

SciTech Connect (OSTI)

Electro-microfluidics is experiencing explosive growth in new product developments. There are many commercial applications for electro-microfluidic devices such as chemical sensors, biological sensors, and drop ejectors for both printing and chemical analysis. The number of silicon surface micromachined electro-microfluidic products is likely to increase. Manufacturing efficiency and integration of microfluidics with electronics will become important. Surface micromachined microfluidic devices are manufactured with the same tools as IC's (integrated circuits) and their fabrication can be incorporated into the IC fabrication process. In order to realize applications for devices must be developed. An Electro-Microfluidic Dual In-line Package (EMDIP{trademark}) was developed surface micromachined electro-microfluidic devices, a practical method for getting fluid into these to be a standard solution that allows for both the electrical and the fluidic connections needed to operate a great variety of electro-microfluidic devices. The EMDIP{trademark} includes a fan-out manifold that, on one side, mates directly with the 200 micron diameter Bosch etched holes found on the device, and, on the other side, mates to lager 1 mm diameter holes. To minimize cost the EMDIP{trademark} can be injection molded in a great variety of thermoplastics which also serve to optimize fluid compatibility. The EMDIP{trademark} plugs directly into a fluidic printed wiring board using a standard dual in-line package pattern for the electrical connections and having a grid of multiple 1 mm diameter fluidic connections to mate to the underside of the EMDIP{trademark}.

BENAVIDES, GILBERT L.; GALAMBOS, PAUL C.

2002-06-01T23:59:59.000Z

384

SRS MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement  

SciTech Connect (OSTI)

The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. Six initial site combinations were proposed: (1) Argonne National Laboratory-West (ANL-W) with support from Idaho National Engineering and Environmental Laboratory (INEEL), (2) Hanford, (3) Los Alamos National Laboratory (LANL) with support from Pantex, (4) Lawrence Livermore National Laboratory (LLNL), (5) Oak Ridge Reservation (ORR), and (6) Savannah River Site(SRS). After further analysis by the sites and DOE-MD, five site combinations were established as possible candidates for producing MOX LAs: (1) ANL-W with support from INEEL, (2) Hanford, (3) LANL, (4) LLNL, and (5) SRS. SRS has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. An alternate approach would allow fabrication of fuel pellets and assembly of fuel rods in an S and S Category 2 or 3 facility with storage of bulk PuO{sub 2} and assembly, storage, and shipping of fuel bundles in an S and S Category 1 facility. The total Category 1 approach, which is the recommended option, would be done in the 221-H Canyon Building. A facility that was never in service will be removed from one area, and a hardened wall will be constructed in another area to accommodate execution of the LA fuel fabrication. The non-Category 1 approach would require removal of process equipment in the FB-Line metal production and packaging glove boxes, which requires work in a contamination area. The Immobilization Hot Demonstration Program equipment in the Savannah River Technology Center would need to be removed to accommodate pellet fabrication. This work would also be in a contaminated area.

O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

1998-08-01T23:59:59.000Z

385

Acceptable knowledge document for INEEL stored transuranic waste -- Rocky Flats Plant waste. Revision 2  

SciTech Connect (OSTI)

This document and supporting documentation provide a consistent, defensible, and auditable record of acceptable knowledge for waste generated at the Rocky Flats Plant which is currently in the accessible storage inventory at the Idaho National Engineering and Environmental Laboratory. The inventory consists of transuranic (TRU) waste generated from 1972 through 1989. Regulations authorize waste generators and treatment, storage, and disposal facilities to use acceptable knowledge in appropriate circumstances to make hazardous waste determinations. Acceptable knowledge includes information relating to plant history, process operations, and waste management, in addition to waste-specific data generated prior to the effective date of the RCRA regulations. This document is organized to provide the reader a comprehensive presentation of the TRU waste inventory ranging from descriptions of the historical plant operations that generated and managed the waste to specific information about the composition of each waste group. Section 2 lists the requirements that dictate and direct TRU waste characterization and authorize the use of the acceptable knowledge approach. In addition to defining the TRU waste inventory, Section 3 summarizes the historical operations, waste management, characterization, and certification activities associated with the inventory. Sections 5.0 through 26.0 describe the waste groups in the inventory including waste generation, waste packaging, and waste characterization. This document includes an expanded discussion for each waste group of potential radionuclide contaminants, in addition to other physical properties and interferences that could potentially impact radioassay systems.

NONE

1998-01-23T23:59:59.000Z

386

Nevada National Security Site Waste Acceptance Criteria  

SciTech Connect (OSTI)

This document establishes the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) Nevada National Security Site Waste Acceptance Criteria (NNSSWAC). The NNSSWAC provides the requirements, terms, and conditions under which the Nevada National Security Site (NNSS) will accept low-level radioactive waste and mixed low-level waste for disposal. The NNSSWAC includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the NNSS Area 3 and Area 5 Radioactive Waste Management Complex for disposal. The NNSA/NSO and support contractors are available to assist you in understanding or interpreting this document. For assistance, please call the NNSA/NSO Waste Management Project at (702) 295-7063 or fax to (702) 295-1153.

NSTec Environmental Management

2011-01-01T23:59:59.000Z

387

Breaking the Code on Challenging Waste - 13267  

SciTech Connect (OSTI)

Mixed low-level wastes (MLLW) with no available path to treatment or disposal have been longstanding challenges for DOE facilities. Today, mixed wastes with no path to treatment or disposal frequently present themselves in the form of combinations of problematic matrixes, problematic EPA Hazardous Waste Codes, and security classification requirements. In order to successfully treat and disposition these challenging wastes, waste management personnel must be more inquisitive and challenge the status quo more than ever before. All aspects of the waste from how it was generated to how the waste is currently being managed must be revisited. Each fact, the basis of each decision, and each regulatory determination must be investigated and validated. Since many of the difficult waste streams were generated several years ago, it can be quite challenging to locate knowledgeable generators from the time of generation. Significant investigation is often required to obtain the needed information to evaluate legacy waste streams. Special attention must be paid to the little things that may not seem central to the issues being investigated. Solutions are sometimes found in these details. (authors)

Witzeman, John; Estes, Charles [URS - CH2M Oak Ridge LLC (United States)] [URS - CH2M Oak Ridge LLC (United States); White, Aaron [U.S. Department of Energy (United States)] [U.S. Department of Energy (United States)

2013-07-01T23:59:59.000Z

388

EIS-0229: Storage and Disposition of Weapons-Usable Fissile Materials  

Broader source: Energy.gov [DOE]

The EIS will evaluate the reasonable alternatives and potential environmental impacts for the proposed siting, construction, and operation of three types of facilities for plutonium disposition.

389

Site selection for the Salt Disposition Facility at the Savannah River Site  

SciTech Connect (OSTI)

The purpose of this report is to identify, assess, and rank potential sites for the proposed Salt Disposition Facility (SDF) at the Savannah River Site.

Bowers, J.A.

2000-01-03T23:59:59.000Z

390

Microsoft Word - CX-MountainAvenueDispositionFY12_WEB.doc  

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

1, 2012 REPLY TO ATTN OF: KEPR-4 SUBJECT: Environmental Clearance Memorandum Joan Kendall Realty Specialist - TERR-3 Proposed Action: Disposition of Mountain Avenue Substation and...

391

Safety evaluation for packaging CPC metal boxes  

SciTech Connect (OSTI)

This Safety Evaluation for Packaging (SEP) provides authorization for the use of Container Products Corporation (CPC) metal boxes, as described in this document, for the interarea shipment of radioactive contaminated equipment and debris for storage in the Central Waste Complex (CWC) or T Plant located in the 200 West Area. Authorization is granted until November 30, 1995. The CPC boxes included in this SEP were originally procured as US Department of Transportation (DOT) Specification 7A Type A boxes. A review of the documentation provided by the manufacturer revealed the documentation did not adequately demonstrate compliance to the 4 ft drop test requirement of 49 CFR 173.465(c). Preparation of a SEP is necessary to document the equivalent safety of the onsite shipment in lieu of meeting DOT packaging requirements until adequate documentation is received. The equivalent safety of the shipment is based on the fact that the radioactive contents consist of contaminated equipment and debris which are not dispersible. Each piece is wrapped in two layers of no less than 4 mil plastic prior to being placed in the box which has an additional 10 mil liner. Pointed objects and sharp edges are padded to prevent puncture of the plastic liner and wrapping.

Romano, T.

1995-05-15T23:59:59.000Z

392

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

SciTech Connect (OSTI)

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

Sanchez, Mike [VP Global Sales, PacTec, Inc. (United States)] [VP Global Sales, PacTec, Inc. (United States); Reeves, Wendall [National Sales Manager, PacTec, Inc. (United States)] [National Sales Manager, PacTec, Inc. (United States); Smart, Bill [Nuclear Sales Director, PacTec, Inc. (United States)] [Nuclear Sales Director, PacTec, Inc. (United States)

2013-07-01T23:59:59.000Z

393

Challenges in Modeling the Degradation of Ceramic Waste Forms  

SciTech Connect (OSTI)

We identify the state of the art, gaps in current understanding, and key research needs in the area of modeling the long-term degradation of ceramic waste forms for nuclear waste disposition. The directed purpose of this report is to define a roadmap for Waste IPSC needs to extend capabilities of waste degradation to ceramic waste forms, which overlaps with the needs of the subconsinuum scale of FMM interests. The key knowledge gaps are in the areas of (i) methodology for developing reliable interatomic potentials to model the complex atomic-level interactions in waste forms; (ii) characterization of water interactions at ceramic surfaces and interfaces; and (iii) extension of atomic-level insights to the long time and distance scales relevant to the problem of actinide and fission product immobilization.

Devanathan, Ramaswami; Gao, Fei; Sun, Xin

2011-09-01T23:59:59.000Z

394

Waste Generated from LMR-AMTEC Reactor Concept  

SciTech Connect (OSTI)

The candidate Liquid Metal Reactor-Alkali Metal Thermal -to- Electric Converter (LMR-AMTEC) is considered to be the first reactor that would use pure liquid potassium as a secondary coolant, in which potassium vapor aids in the conversion of thermal energy to electric energy. As with all energy production, the thermal generation of electricity produces wastes. These wastes must be managed in ways which safeguard human health and minimize their impact on the environment. Nuclear power is the only energy industry, which takes full responsibility for all its wastes. Based on the candidate design of the LMR-AMTEC components and the coolant types, different wastes will be generated from LMR. These wastes must be classified and characterized according to the U.S. Code of Federal Regulation, CFR. This paper defines the waste generation and waste characterization from LMR-AMTEC and reviews the applicable U.S. regulations that govern waste transportation, treatment, storage and final disposition. The wastes generated from LMR-AMTEC are characterized as: (1) mixed waste which is generated from liquid sodium contaminated by fission products and activated corrosion products; (2) hazardous waste which is generated from liquid potassium contaminated by corrosion products; (3) spent nuclear fuel; and (4) low-level radioactive waste which is generated from the packing materials (e.g. activated carbon in cold trap and purification units). The regulations and management of these wastes are summarized in this paper.

Hasan, Ahmed; Mohamed, Yasser, T.; Mohammaden, Tarek, F.

2003-02-25T23:59:59.000Z

395

Waste management project fiscal year 1998 multi-year work plan WBS 1.2  

SciTech Connect (OSTI)

The MYWP technical baseline describes the work to be accomplished by the Project and the technical standards which govern that work. The Waste Management Project manages and integrates (non-TWRS) waste management activities at the site. Activities include management of Hanford wastes as well as waste transferred to Hanford from other DOE, Department of Defense, or other facilities. This work includes handling, treatment, storage, and disposition of radioactive, nonradioactive, hazardous, and mixed solid and liquid wastes. Major Waste Management Projects are the Solid Waste Project (SW), Liquid Effluents Project (LEP), and Analytical Services. Existing facilities (e.g., grout vaults and canyons) shall be evaluated for reuse for these purposes to the maximum extent possible. The paper tabulates the major facilities that interface with this Project, identifying the major facilities that generate waste, materials, or infrastructure for this Project and the major facilities that will receive waste and materials from this Project.

Slaybaugh, R.R.

1997-08-29T23:59:59.000Z

396

TRU waste certification compliance requirements for acceptance of newly generated contact-handled wastes to be shipped to the WIPP. Revision 1  

SciTech Connect (OSTI)

Compliance requirements are presented for certifying that unclassified, newly generated, contact-handled (CH) transuranic (TRU) solid wastes from defense programs meet the Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria (WAC). Where appropriate, transportation and interim storage requirements are incorporated, however, interim storage sites may have additional requirements consistent with these requirements. All applicable DOE orders must continue to be met. The compliance requirements for stored or buried waste are not addressed in this document. The compliance requirements are divided into four sections, primarily determined by the general feature that the requirements address. These sections are General Requirements, Waste Container Requirements, Waste Form Requirements, and Waste Package Requirements. The waste package is the combination of waste container and waste. 2 refs., 1 fig.

Not Available

1985-09-01T23:59:59.000Z

397

Packaging of solid state devices  

DOE Patents [OSTI]

A package for one or more solid state devices in a single module that allows for operation at high voltage, high current, or both high voltage and high current. Low thermal resistance between the solid state devices and an exterior of the package and matched coefficient of thermal expansion between the solid state devices and the materials used in packaging enables high power operation. The solid state devices are soldered between two layers of ceramic with metal traces that interconnect the devices and external contacts. This approach provides a simple method for assembling and encapsulating high power solid state devices.

Glidden, Steven C.; Sanders, Howard D.

2006-01-03T23:59:59.000Z

398

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

SciTech Connect (OSTI)

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

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

2013-07-01T23:59:59.000Z

399

Site Selection for the Salt Disposition Facility at the Savannah River Site  

SciTech Connect (OSTI)

A site selection study was conducted to identify a suitable location for the construction and operation of a new Salt Disposition Facility (SDF) at the Savannah River Site (SRS). The facility to be sited is a single processing facility and support buildings that could house either of three technology alternatives being developed by the High Level Waste Systems Engineering Team: Small Tank Tetraphenylborate Precipitation, Crystalline Silicotitanate Non-Elutable Ion Exchange or Caustic Side Solvent Extraction. A fourth alternative, Direct Disposal in grout, is not part of the site selection study because a location has been identified that is unique to this technology (i.e., Z-Area). Facility site selection at SRS is a formal, documented process that seeks to optimize siting of new facilities with respect to facility-specific engineering requirements, sensitive environmental resources, and applicable regulatory requirements. In this manner, the prime objectives of cost minimization, environmental protection, and regulatory compliance are achieved. The results from this geotechnical characterization indicated that continued consideration be given to Site B for the proposed SDF. Suitable topography, the lack of surface hydrology and floodplain issues, no significant groundwater contamination, the presence of minor soft zones along the northeast portion of footprint, and no apparent geological structure in the Gordon Aquitard support this recommendation.

Gladden, J.B.; Rueter, K.J.; Morin, J.P.

2000-11-15T23:59:59.000Z

400

DOE plutonium disposition study: Analysis of existing ABB-CE Light Water Reactors for the disposition of weapons-grade plutonium. Final report  

SciTech Connect (OSTI)

Core reactivity and basic fuel management calculations were conducted on the selected reactors (with emphasis on the System 80 units as being the most desirable choice). Methods used were identical to those reported in the Evolutionary Reactor Report. From these calculations, the basic mission capability was assessed. The selected reactors were studied for modification, such as the addition of control rod nozzles to increase rod worth, and internals and control system modifications that might also be needed. Other system modifications studied included the use of enriched boric acid as soluble poison, and examination of the fuel pool capacities. The basic geometry and mechanical characteristics, materials and fabrication techniques of the fuel assemblies for the selected existing reactors are the same as for System 80+. There will be some differences in plutonium loading, according to the ability of the reactors to load MOX fuel. These differences are not expected to affect licensability or EPA requirements. Therefore, the fuel technology and fuel qualification sections provided in the Evolutionary Reactor Report apply to the existing reactors. An additional factor, in that the existing reactor availability presupposes the use of that reactor for the irradiation of Lead Test Assemblies, is discussed. The reactor operating and facility licenses for the operating plants were reviewed. Licensing strategies for each selected reactor were identified. The spent fuel pool for the selected reactors (Palo Verde) was reviewed for capacity and upgrade requirements. Reactor waste streams were identified and assessed in comparison to uranium fuel operations. Cost assessments and schedules for converting to plutonium disposition were estimated for some of the major modification items. Economic factors (incremental costs associated with using weapons plutonium) were listed and where possible under the scope of work, estimates were made.

Not Available

1994-06-01T23:59:59.000Z

Note: This page contains sample records for the topic "waste disposition packaging" from the National Library of EnergyBeta (NLEBeta).
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