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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Improving PCM Waste Management Through New Approaches to Screening for LLW Treatment and Disposal - 20145

A major step forward in the management of Plutonium Contaminated Material (PCM) Waste has been taken at the Sellafield site with the introduction of new capabilities to enable wastes previously conservatively designated as PCM Intermediate Level Waste (ILW) to be re-classified for treatment as Low Level Waste (LLW). The Sellafield site, located in the North West of England, within the United Kingdom (UK), has operated plutonium processing plants and waste facilities for almost 70 years. During this time management of the raw PCM waste produced has presented a variety of challenges. The waste is currently processed by high force waste compaction in the Waste Treatment Complex (WTC). The WTC converts 200 litre drums of raw PCM waste into a compacted, grout encapsulated monolith form within stainless steel containment, ready for emplacement within the UK's Geological Disposal Facility (GDF) when it becomes available. To help progress some of the wastes waiting for treatment, recent improvements have been made with the introduction of new waste assay instrumentation which enables the identification of a 'candidate' waste potentially suitable for disposal via the LLW treatment route. Opening a new LLW route for previously designated PCM has been a major advance in PCM waste management policy. It has required a range of technical assessment arrangements to be developed, tested and commissioned and involved the engagement of a wide range of stakeholders. In particular, very detailed technical analysis methods have been developed to ensure that there is a robust and conservative approach to the final sentencing of waste. Following completion of commissioning tests, the first 86 m{sup 3} of previously designated PCM waste was successfully treated through the LLW route and consigned to LLW disposal in June 2019. The direct benefit of sentencing waste to the correct waste route (LLW) is the significant cost savings across the complete range of waste management activities. This work is also strategically important for demonstrating to a range of stakeholders including workforce and regulators, that waste previously designated as PCM can be re-classified in a systematic and justifiable way without increasing risk to personnel or the wider environment. By ensuring that waste is being managed appropriately it will be possible to carry out more remediation work safely, in less time, and at a lower cost. (authors)

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DOE OSTI2020

Panel Session 46: Final Closure of Unlined LLW Disposal Facilities Worldwide (R3.5)

This panel discussed the on-going and emerging issues with the design and construction of final cover system of unlined disposal facilities (e.g., material disposal areas (MDAs), subsurface disposal areas (SDAs), burial grounds, and similar historical waste trenches, pits and vaults). The discussion focused on key technical and regulatory issues as well as lessons learned from years of engineering and construction in US and non-US locations. Panelists with presentations: Closure of Unlined LLW Disposal Sites (Craig Benson); Closure of LLW Facilities (Paul Pointon); Final Closure of Unlined LLW Disposal Facilities - Lessons Learned (Kevin Foye)

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DOE OSTI2020

Panel Session 57: US Nuclear Power Plant Waste Management - US LLW Disposal Issues

This panel focused on senior utility managers discussing specific disposal site issues with the disposal site operators. Representatives from each of the operating commercial disposal sites began by presenting an update of their facility, license, and waste acceptance criteria (WAC) followed by a panel discussion in response to audience questions. Panelists with presentations: Current and Emerging US NRC Regulatory Topics (Joe Weismann); WCS Update (Dan Burns); EnergySolutions Clive Facility Support for US LLW Disposal Issues (Vern C. Rogers)

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DOE OSTI2020

Origins and Contents of the Waste Specifications: the Example of the French LLW-SL Disposal Facility - 20016

The waste specifications published by Andra for its surface disposal facilities for Low and Intermediate, Short-Lived waste (CSA) are primarily defined to ensure the safe disposal of waste, both in the short and long term, spanning from disposal operations to hundreds of years after closure, when the man-made safety barriers have long disappeared into dust. The paper presents the methodology used at Andra to derive the CSA waste specifications, also known as Waste Acceptance Criteria, their contents, and the process in place to verify compliance of the waste with the specifications. (authors)

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DOE OSTI2020

Panel Session 99: Commercial LLW Management: Global Perspective

This panel focused on opportunities and challenges in providing commercial nuclear waste management services in a global context. The panelists commented on their approach to establishing new businesses, overcoming communication barriers acquiring / maintaining skills and capabilities and delivering commercial nuclear waste management projects in a range of regulatory and socio-economic environments. The panel discussion was followed by a Q and A session with audience participation. Panelist with presentations: BGE Technology (Thilo von Berlepsch)

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DOE OSTI2020

Key Technical Issues for Greater-Than-Class-C (GTCC) Waste Disposal - 20164

Greater-than-Class C (GTCC) waste is low-level waste (LLW) that exceeds the Class C concentrations tabulated in Title 10, Code of Federal Regulations (CFR), Part 61. Disposal of GTCC waste in the near surface (i.e. upper 30 m of the earth's surface) is prohibited in the United States. Though GTCC waste disposal is generally prohibited, the Commission can approve disposal according to 10 CFR 61.55(a)(2)(iv) on a case-by-case basis. The waste classification tables, developed in the early 1980's, were based on model projections of dose to an inadvertent intruder in agricultural, construction, and discovery exposure scenarios. The assumptions and parameters used were documented in a series of public reports. Though different disposal facility designs and site conditions were considered, the waste classification tables were based on a disposal facility design located in a specific environment. The tables were based on shallow (i.e., top few meters) trench facility designs and did not consider deeper facilities. To determine the suitability of GTCC waste disposal in the near-surface, site-specific analyses must account for differences between GTCC waste and Class A, B, and C LLW (hereafter, traditional LLW). GTCC waste can have concentrations of radionuclides that are much higher than traditional LLW. Because of these higher concentrations, processes that are typically not significant for traditional LLW may be significant with respect to disposal of GTCC waste. These processes include, but are not limited to, heat generation, criticality, and radiolysis. The form of the waste as well as the barriers to release of the waste (e.g. waste package) could be substantially different than they are for traditional LLW. These barriers need to be considered when assessing the impacts of accidents during receipt and placement and in evaluating long-term performance. Some GTCC radioactivity is either embedded in stainless steel or contained in stainless steel barriers. Stainless steel can have very low corrosion rates under a variety of environmental conditions. Finally, and possibly most importantly, GTCC waste would likely need to be disposed deeper than traditional LLW to reduce the probability of disturbance. The waste classification table values of traditional LLW are based on an inadvertent intruder excavating into the waste and bringing some of the material to the land surface. If waste is deeper than approximately 5 m, the excavation scenario becomes very unlikely. Therefore, other intruder scenarios, such as drilling exposure scenarios, need to be evaluated. This paper summarizes key technical issues for the disposal of GTCC waste. Our previous study shows that certain GTCC waste may be suitable for near surface disposal whereas others may not. This study may support ongoing technical analyses assessing potential disposal of GTCC in a near surface disposal facility. (authors)

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DOE OSTI2020

Modeling Interfaces to Support Low-Level Waste Disposal System Performance Assessments - 20366

In low-level waste (LLW) disposal sites, interfaces between cementitious materials, used as waste forms and/or engineered barriers, and the surrounding soil or backfill material are often encountered. Reactions across these interfaces may lead to chemical and structural alteration of the cementitious and backfill materials that may ultimately affect long-term performance. For example, the ingress of carbon dioxide from soil gas into a waste tank concrete shell at the Hanford Site or Savannah River Site (SRS) or a low-level waste disposal vault at SRS leads to carbonation of cement hydration products (e.g. portlandite, calcium-silicate hydrate or C-S-H, and ettringite). The result of carbonation is a decrease in pH in the cement paste portion of the tank wall or vault concrete structure that could ultimately lead to de-passivation of the embedded structural steel and, eventually, to cracking. Cracking can lead to increased ingress of water into the structure, corresponding increased release of constituents of concern, and to subsequent increased transport of these constituents to the surrounding environment. The mobility of trace constituents (e.g., radionuclides of concern) in waste forms may increase in response to changes in pH, pore structure, and mineralogical gradients. Depletion of portlandite results in subsequent decalcification of C-S-H that leads to changes in the cement strength and may lead to structural failure. At the boundary between waste forms and concrete vaults, the migration of sulfate ions from the salt waste form (e.g., the saltstone waste form used for LLW disposal at the Savannah River Site in Saltstone Disposal Units, or SDUs) into the barrier or vault concrete has been predicted to result in expansive mineral phase formation (or 'sulfate attack'). Expansive mineral formation could result in cracking and potential loss of structural integrity in the vault concrete. In the performance assessment (PA) for the SRS SDUs, prediction of both the carbonation and sulfate attack fronts are critical to assessing the long-term performance of these LLW disposal vaults. A general purpose geochemical reactive transport model has been developed in LeachXS/ORCHESTRA to evaluate typical interfaces for LLW disposal environments, including SDUs at SRS. The model may be used to predict the long-term performance of interfaces between cementitious materials and saltstone or backfill materials with respect to primary phases by considering mass transport, cement chemistry, geochemical speciation, and multi-ionic diffusion over interfaces between different materials. The present study compares the performance and interactions between vault concrete and saltstone waste form for a representative SDU scenario. The materials were characterized using pH-dependence leaching test (EPA 1313) and semi-dynamic transport tests (EPA 1315) with test results used to parameterize the interface model. The development of a credible inter model provides an important basis for supporting future PAs where carbonation and sulfate attack are important aging and degradation mechanisms. (authors)

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DOE OSTI2020

Removal of High Specific Activity Fission Products from Uranyl Sulfate Waste Solutions

The Savannah River National Laboratory (SRNL) is currently providing support to SHINE Medical Technologies (SHINE) which plans to deploy a low energy, accelerator-based neutron source to fission low enriched U in a uranyl sulfate target solution for 99 Mo production. The 99 Mo is initially separated from the fission products and target solution by an extraction column. Subsequent washing of the column will generate waste solutions containing residual U and fission product activity. A small number of high specific activity fission products (e.g., 90 Sr, 137 Cs, and 144 Ce) in these streams will likely control the classification of the low level waste (LLW). If a sufficient amount of the high specific activity isotopes are separated from the SHINE waste streams and concentrated into a waste form, it would be possible to treat a majority of the wash solutions from the column operations as a lower class of LLW (Class A versus Class B or C or Class B versus Class C). The high specific activity fission product elements could then be dispositioned as a much smaller volume of waste rather than requiring the disposal of the entire waste stream at the higher waste classification. The Savannah River Site (SRS) has experience with using monosodium titanate (MST) and crystalline silicotitanate (CST) to remove Cs and Sr from high salt content waste solutions generated during the reprocessing of nuclear fuels and targets. Both of these materials have worked very well for their intended purposes at the SRS where the fission product elements are removed from highly alkaline waste. On the other hand, SHINE waste streams from the extraction column contain H 2 SO 4 which makes the solution acidic. Additionally, the SRS waste streams do not contain other fission product elements such as transition metals or lanthanides because they precipitate upon neutralization of the SRS waste and are not present in the supernate which is dispositioned as LLW following treatment. As such, there are inherent differences between SHINE and SRS waste treatment strategies. Savannah River National Laboratory was tasked with performing scoping studies to see if MST and CST would remove Sr, Cs, and Ce from an acidic mixed metal simulant solution. Batch contact experiments were performed using MST and two CST type materials. The MST material is a 15 wt % powder in 0.15 M NaOH slurry. The MST showed low adsorption for elements of interest from acidic solution. Furthermore, the powder size makes MST non-ideal for column operations. A CST IE-911 ion exchange material had high Cs adsorption, moderate Sr, and marginal Ce adsorption. Based on adsorption of all species, the ion exchange capacity was found to be 0.032 meq/mL. A bench-top column experiment to measure elemental breakthrough curves was performed using CST IE-911 where chromatographic separations of the mixed simulant were expected to occur. While most elements behaved as expected, the lanthanide series, containing Ce, broke though the column earlier than expected. The second CST material, CST R9120, displayed high adsorption for all elements in the acidic mixed simulant solutions in a batch contact study, and had a calculated loading capacity of 0.091 meq/mL. Future studies to develop a waste treatment flowsheet should focus on CST R9120 to treat SHINE waste solutions.

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DOE OSTI2020

Flanged Tritium Waste Containers: Background and Current Path to Disposal

The MLLW Flanged Tritium Waste Container (FTWC) Project has been operational since 2007 when, during an audit of the FTWC waste stream, it was determined that four FTWC containers contained lead in amounts greater than LDR treatment standards. From June to August of 2007, four FTWC containers were shipped to TA-54 for disposal as LLW from WETF. The contents of the FTWCs were characterized as high activity LLW and were thought to be in compliance with LANL's Waste Acceptance Criteria (WAC). During the packaging of these FTWC's, 36 intact squib valves (aka actuators), containing detonated squibs were included with the waste stream After being transported to TA-54 Area G, it was determined that the explosive used in some of the squibs was Normal Lead Styphnate (NLS). A recharacterization of the entire contents in the FTWC's was performed with the conclusion that the squib valves exhibited the RCRA toxic characteristic for Lead (D008), thus resulting in the contents of the FTWCs being reclassified as a Mixed LLW.

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DOE OSTI2021

EPC-ES WIPP WAC Appendix A Equivalence Support Measurements 2019-2020

Characterization of low-level waste (LLW) at Los Alamos National Laboratory (LANL) is primarily performed by the Environmental Protection and Compliance - Environmental Stewardship (EPC-ES) Characterization Services team and/or the Nuclear Process Infrastructure (NPI--9) Nuclear Material Support Services team (hereafter referred to as the gamma spectroscopy teams). The gamma spectroscopy teams use portable high-purity germanium (HPGe) detector systems for the nondestructive assay (NDA) characterization of LLW. Controlled-approved procedures and processes for the use of such systems to assay LLW items are currently in place. Additionally, a number of performance studies have been conducted by the gamma spectroscopy teams to support the efficacy and quality of assay results generated by the established NDA process.

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DOE OSTI2021

UNREVIEWED DISPOSAL QUESTION EVALUATION: Disposal of the Idaho National Laboratory Classified Mockup Low-Level Waste at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This Unreviewed Disposal Question Evaluation (UDQE) assesses whether the Idaho National Laboratory (INL) Classified Mockup Low-Level Waste (LLW) (INEL208599QR0, Revision 0 [INL 2020]) is suitable for shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site (NNSS). Disposal of the INL Classified Mockup LLW meets all performance objectives of U.S. Department of Energy (DOE) Manual DOE M 435.1-1, Radioactive Waste Management Manual, Chapter IV, Section P (DOE 1999). The INL Classified Mockup LLW stream is recommended for acceptance without conditions.

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DOE OSTI2021

UNREVIEWED DISPOSAL QUESTION EVALUATION: Disposal of the Idaho National Laboratory Classified Mockup Low-Level Waste, Revision 1, at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This Unreviewed Disposal Question Evaluation (UDQE) assesses whether the Idaho National Laboratory (INL) Classified Mockup Low-Level Waste (LLW) (INEL208599QR0, Revision 1 [INL 2021]) is suitable for shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site (NNSS). Disposal of the INL Classified Mockup LLW meets all performance objectives of U.S. Department of Energy (DOE) Manual DOE M 435.1-1, Radioactive Waste Management Manual, Chapter IV, Section P (DOE 1999). The INL Classified Mockup LLW stream is recommended for acceptance without conditions.

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DOE OSTI2021

UNREVIEWED DISPOSAL QUESTION EVALUATION: Disposal of the Transuranic Waste Processing Center Low-Level Waste at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This Unreviewed Disposal Question Evaluation (UDQE) assesses whether the Transuranic Waste Processing Center (TWPC) Low-Level Waste (LLW), FWORCHLLW0102, Revision 12 (TWPC 2021), is suitable for shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site (NNSS). Disposal of the TWPC LLW meets all performance objectives of DOE Manual DOE M 435.1-1, Radioactive Waste Management Manual, Chapter IV, Section P (DOE 1999). The TWPC LLW waste stream is recommended for acceptance without conditions.

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DOE OSTI2020

Panel Session 36: Hot Topics in US Commercial Low-Level Waste Management

This panel focused on emerging issues in commercial LLW management in the US from the perspective of active members of the Low-Level Radioactive Waste Forum, Inc. State, compact, federal and industry officials shared their views on a variety of timely and significant topics related to low-level radioactive waste management, disposal and related issues. Session no. 36 Co-Chairs, Patricia Holahan and Daniel Shrum, introduced the panelists and provided a summary of the purpose of the session. This Panel Session focused on several current and emerging issues impacting the commercial Low-Level Radioactive Waste (LLRW) management in the US from the perspective of active members of the Low-Level Waste Forum, Inc., States, Compact, as well as Federal and nuclear industries. The panelists provided different perspectives on number of key issues such as: LLW disposal facility status and updates (e.g.; in terms of disposal capacity, plans, and waste acceptance conditions). The panelists also addressed waste disposal rates and potential impacts of ongoing regulatory developments by the NRC and the concerned States. Other timely topics pertaining to the revision of 10CFR Part 61, the Greater-than-Class C (GTCC) and Transuranic (TRU) waste, and Very Low-Level Waste (VLLW) disposal issues were also addressed at this Session. Panelists with presentations: Hot Topics in US Commercial LLW Management (Joe Weismann); Texas LLRW Compact (David Carlson); Texas Low-Level Radioactive Waste Disposal Compact Commission (John Salsman); Barnwell LLRW Disposal Facility (Benjamin Smith, Bill Carver)

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DOE OSTI2020

Panel Session 70: Nuclear Power Plant Waste Management, LLRW Processing Issues

This panel focused on senior utility managers discussing issues with commercial LLW processors. LLW processors had an opportunity to update utilities/NPPs on new or different services or technologies being offered to assist in their radioactive waste management goals, changes to their facility license, revisions to their acceptance guidelines, and emerging compliance issues. Panelists with presentations: UniTech - 'Expanding to Reduce Costs' LLW Processor Update (Gregg Johnstone); Veolia Nuclear Solutions Federal Services - Alaron Nuclear Services (Dave Garber); EnergySolutions Tennessee Processing Facilities (Fred Schulz); Complex Waste Management Issues Services and Treatment Facilities (George M. Taylor)

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DOE OSTI2020

A Preliminary Radiological Risk Assessment Model for Disposition of Remote-Handled Transuranic Wastes at Los Alamos National Laboratory Area G - 20116

The U.S. Department of Energy (DOE) operates a low-level radioactive waste (LLW) disposal site at Material Disposal Area G, in Los Alamos, New Mexico, USA. Area G has been the primary LLW disposal site for Los Alamos National Laboratory (LANL) since the 1960's. In addition to LLW, Area G is host to a variety of other wastes, the disposition of which must be determined before closure of the site. A probabilistic Radiological Risk Assessment (RRA) for Area G is used in order to support decision making regarding some wastes that are not addressed in the extant Area G Performance Assessment (PA) and Composite Analysis (CA). Between 1979 and 1987, 33 special shafts were augered into the Bandelier Tuff at Area G. This volcanic tuff is present across Pajarito Plateau on the eastern slopes of the Jemez Mountains, and varies widely in its consistency, from weakly indurated non-welded layers to welded layers that uphold the mesa cliffs of the Plateau. These mesas are home to LANL, Area G, and the townsites of Los Alamos and White Rock, with residences about 1400 m from Area G. The 33 Shafts were lined with steel casing, and contain remote-handled (RH) transuranic wastes (TRU) resulting from experiments and analysis performed in special glove boxes at the Chemistry and Metallurgy Research (CMR) facility at LANL. Some of these wastes originated as used nuclear fuel. The purpose of the Area G RRA is to evaluate the potential future risk to humans and the environment from the RH TRU in the 33 Shafts in the context of the risk associated with the surrounding wastes at Area G. The analysis is responsive to expectations outlined in DOE Order 458.1, Radiation Protection of the Public and the Environment, and is informed by the Manual and Guidance accompanying DOE O 435.1, Radioactive Waste Management. Because the waste meets the definition of TRU, the regulatory context necessarily takes into consideration the regulation governing the disposal of TRU from the U.S. Environmental Protection Agency (EPA): 40 CFR 191, Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes. Given the broader regulatory context for the RRA, the analysis is subject to different assumptions from those made in the existing DOE O 435.1 PA and CA, such as allowing for future occupation of the site. The analysis begins with a comprehensive evaluation of features, events, processes, and exposure scenarios (FEPS) for Area G and the wastes it contains. These FEPSs are screened to eliminate from further consideration those of extremely low probability and/or consequence, and a conceptual site model (CSM) is subsequently developed. The scope and structure of the Area G RRA Model is informed by this CSM, and the Area G RRA Model is developed using the GoldSim systems analysis modeling platform. This paper presents the initial version of a defensible, transparent, and reasonably realistic model, which is based on the state of knowledge of the wastes, the site, and the FEPSs that govern contaminant transport from wastes into the environment and subsequent exposures to humans and other biota. Probabilistic model input distributions represent uncertainties inherent in the real and modeled systems. The results of the Area G RRA Model inform decisions regarding the disposition of the RH TRU in the 33 Shafts. (authors)

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