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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

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

The Los Alamos National Laboratory (LANL or the Laboratory) generates radioactive waste as a result of various activities. Operational or institutional waste is generated from a wide variety of research and development activities including nuclear weapons development, energy production, and medical research. Environmental restoration (ER), and decontamination and decommissioning (D and D) waste is generated as contaminated sites and facilities at LANL undergo cleanup or remediation. The majority of this waste is low-level radioactive waste (LLW) and is disposed of at the Technical Area 54 (TA-54), Area G disposal facility. U.S. Department of Energy (DOE) Order 435.1 (DOE, 2001) requires that radioactive waste be managed in a manner that protects public health and safety, and the environment. To comply with this order, DOE field sites must prepare and maintain site-specific radiological performance assessments for LLW disposal facilities that accept waste after September 26, 1988. Furthermore, sites are required to conduct composite analyses that account for the cumulative impacts of all waste that has been (or will be) disposed of at the facilities and other sources of radioactive material that may interact with the facilities. Revision 4 of the Area G performance assessment and composite analysis was issued in 2008 (LANL, 2008). These analyses estimate rates of radionuclide release from the waste disposed of at the facility, simulate the movement of radionuclides through the environment, and project potential radiation doses to humans for several on-site and off-site exposure scenarios. The assessments are based on existing site and disposal facility data and on assumptions about future rates and methods of waste disposal. The accuracy of the performance assessment and composite analysis depends upon the validity of the data used and assumptions made in conducting the analyses. If changes in these data and assumptions are significant, they may invalidate or call into question certain aspects of the analyses. For example, if the volumes and activities of waste disposed of during the remainder of the disposal facility's lifetime differ significantly from those projected, the doses projected by the analyses may no longer apply. DOE field sites are required to implement a performance assessment and composite analysis maintenance program. The purpose of this program is to ensure the continued applicability of the analyses through incremental improvement of the level of understanding of the disposal site and facility. Site personnel are required to conduct field and experimental work to reduce the uncertainty in the data and models used in the assessments. Furthermore, they are required to conduct periodic reviews of waste receipts, comparing them to projected waste disposal rates. The radiological inventory for Area G was updated in conjunction with Revision 4 of the performance assessment and composite analysis (Shuman, 2008). That effort used disposal records and other sources of information to estimate the quantities of radioactive waste that have been disposed of at Area G from 1959, the year the facility started receiving waste on a routine basis, through 2007. It also estimated the quantities of LLW that will require disposal from 2008 through 2044, the year in which it is assumed that disposal operations at Area G will cease. This report documents the fourth review of Area G disposal receipts since the inventory was updated and examines information for waste placed in the ground during fiscal years (FY) 2008 through 2011. The primary objective of the disposal receipt review is to ensure that the future waste inventory projections developed for the performance assessment and composite analysis are consistent with the actual types and quantities of waste being disposed of at Area G. Toward this end, the disposal data that are the subject of this review are used to update the future waste inventory projections for the disposal facility. These projections are compared to the future inventory projections that were developed for Revision 4 of the performance assessment and composite analysis. The approach used to characterize the FY 2008 through 2011 waste is generally the same as that used to characterize the inventory for the Revision 4 analyses (Shuman, 2008). This methodology is described in Section 2. The results of the disposal receipt review are presented in Section 3 and discussed in terms of their significance to the Area G analyses.

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DOE OSTI2015

Development of disposal container designs for disposal of high level waste and legacy spent fuel in the UK - 15471

Radioactive Waste Management Limited (RWM), a fully owned subsidiary of the Nuclear Decommissioning Authority (NDA), is responsible for the implementation of geological disposal in the UK. Three illustrative concepts of a geological disposal facility (GDF), corresponding to three generic geological environments - higher strength rock, lower strength sedimentary rock and evaporite rock - have been developed to demonstrate the viability of geological disposal of intermediate level waste, high level waste and spent fuel. It is envisaged that HLW which has been encapsulated in Waste Vitrification Plant (WVP) canisters, advanced gas-cooled reactor (AGR) SF which has been stripped of their spacers and packaged into slotted fuel cans, and PWR SF will be packaged into disposal containers for disposal. A set of schematic designs of these disposal containers was developed in the late 2000's. The design concept for the higher strength rock environment was based on the joint Posiva/SKB disposal container design, and the design concept for the lower strength sedimentary rock and evaporate rock illustrative environments was based on the disposal container concept from Switzerland's National Cooperative for the Disposal of Radioactive Waste (Nagra). The size, mass and dimensions of the disposal containers in the scheme designs vary significantly. The construction method and the lid closing method also vary between the disposal containers for the different waste contents. A design study has indicated that optimization of the number of HLW WVP canisters and the number of AGR slotted fuel cans in the disposal containers, standardizing construction method and closure arrangement across the containers for each geology type, providing handling features at both ends of the disposal container, and balancing of shielding around the containers, could bring significant advantage to the overall GDF programme by reducing the number of disposal containers, reducing GDF footprint, and reducing the extent of excavations, spoil generated and materials that are required for construction. A project to develop the scheme designs to conceptual level was carried in 2011 to 2013. In this project, the above changes were taken into account, technical underpinning of the designs was carried out and manufacturing method was considered. Two disposal container variants for different host rock environments have been developed: - Variant 1, for compatibility with the higher strength rock illustrative disposal system design. This is constructed from copper, with a cast iron insert and is based on the joint Posiva/SKB disposal container design. - Variant 2, for compatibility with the lower strength sedimentary rock and evaporite rock illustrative disposal system designs. This is a steel container based on the NAGRA disposal container design. Each variant has three internal configurations to accommodate 3 HLW vitrified canisters, 16 consolidated cans of AGR SF and 4 PWR spent fuel assemblies. These designs have been shown to bring significant benefit to the GDF designs, bringing a significant reduction to the operating programme and GDF programme cost, including reducing the number of disposal containers by 33% for HLW, 50% for AGR SF and reducing the size of the GDF footprint by 25%. (authors)

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DOE OSTI2015

Regulatory inspection of the provision of disposability assessment and waste packaging advice in the UK to provide confidence that higher activity waste stored on licensed sites will be suitable for safe handling, transport and disposal - 15088

Radioactive Waste Management Limited (RWM), the developer and future operator of a geological disposal facility (GDF) in the United Kingdom, operates a disposability assessment process to minimise the risk that current conditioning and packaging of higher activity wastes (HAW) will result in packages incompatible with geological disposal in the future. Specialists in nuclear and environmental safety from the Office for Nuclear Regulation (ONR) and the Environment Agency (EA) in 2013/14 completed an inspection of the RWM's process of disposability assessment. The inspection gave the regulators confidence that RWM has in place a systematic process to provide an important means of reducing risks associated with waste packaging in advance of an operational GDF and that the process of disposability assessment is providing the necessary information for licensees to assure future HAW disposal to the GDF. This work forms part of our continuing programme to review RWM's work related to geological disposal of higher activity radioactive waste (HAW). (authors)

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DOE OSTI2018

Preliminary disposal limits, plume interaction factors, and final disposal limits

In the 2008 E-Area Performance Assessment (PA), each final disposal limit was constructed as the product of a preliminary disposal limit and a plume interaction factor. The following mathematical development demonstrates that performance objectives are generally expected to be satisfied with high confidence under practical PA scenarios using this method. However, radionuclides that experience significant decay between a disposal unit and the 100-meter boundary, such as H-3 and Sr-90, can challenge performance objectives, depending on the disposed-of waste composition, facility geometry, and the significance of the plume interaction factor. Pros and cons of analyzing single disposal units or multiple disposal units as a group in the preliminary disposal limits analysis are also identified.

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DOE OSTI2015

A study on the criticality safety for the direct disposal of used nuclear fuel in Japan: Application of burnup credit to the criticality safety evaluation for the disposal canister - 14049

Japan has recently started the technical development of the direct disposal of used nuclear fuel (UNF) to prepare various disposal options. Unlike the disposal of radioactive waste from UNF reprocessing, criticality safety is important here because of the presence of a certain amount of fissile nuclides in UNF. This paper provides an outline of the research to be addressed in this field and the relevant studies in Japan. It especially includes the first result of a criticality safety evaluation of a disposal canister model adopting burnup credit. In this evaluation, it was assumed that a steel disposal canister loaded with fuel assemblies was surrounded by a clay buffer material, and the fuel assemblies and the disposal canister were intact. Conditions representative of discharged pressurized-water reactor (PWR) fuel were applied in this study - initial U-235 enrichment of 4.5 wt% and burnup of 45 GWd/t. The isotopic composition of the burnt fuel was evaluated using SWAT3.1 code. In the depletion calculation, conservative parameters were applied within the typical PWR operation range. The effective neutron multiplication factor (k{sub eff}) was then calculated using a continuous-energy Monte Carlo code MVP-2.0, adopting the JENDL-4.0 nuclear data library. Uncertainty in k{sub eff} due to depletion calculation errors was evaluated using the correction factors of isotopic composition, which were obtained via multiple comparisons between the calculated and measured isotopic concentrations of irradiated fuels. In addition, the effects of axial burnup profile and horizontal burnup gradient on k{sub eff} were evaluated using the open burnup distribution database. The k{sub eff} including the uncertainties and conservatism was found to be below 0.95 for both actinide-only and actinide and fission product considerations when the fuel assemblies and the disposal canister were assumed to be intact. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Panel Session 104: US DOE LLW On-Site Disposal: LLW Disposal Facility Federal Review Group Review Process

The US DOE Low-Level Waste Disposal Facility Federal Review Group (LFRG) as the regulatory authority for onsite low-level waste disposal facilities for the complex per DOE Order 435.1. This panel focused on the review process for technical basis documents including the Performance Assessments that are used to issue the disposal authorization statement Presentations proposed include: 1 - Low-Level Waste Disposal Overview (Jhon Carilli, US DOE EM Nevada (USA)); 2 - Factors Influencing Allowable Waste Concentrations and Inventories at Disposal Facilities (Roger Seitz, IEI (USA)); 3 - Low-Level Waste Disposal Facility Federal Review Group - LFRG (Jaffet Ferrer-Torres, US DOE (USA))

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DOE OSTI2025

Resolving Issues Regarding Disposal of Safeguarded Nuclear Materials: Concepts for Near Surface and Intermediate Depth Disposal sites

This document presents two case studies on nuclear waste disposal to explore the safeguards challenges and potential solutions present under each scenario. Both involve the disposal of nuclear materials, either for waste containing safeguarded nuclear materials or waste where safeguards on the nuclear materials have been previously terminated. The first case study is that of a low-level waste (LLW) repository that is situated near the surface. This facility will accept safeguarded nuclear materials for disposal. Examples of this type of activity are rare, but the potential for more facilities of this type is very possible as States become more comfortable with considering safeguards measures in perpetuity on deep geological repositories. The second study extends this to the case of terminated wastes at an intermediate depth disposal site. In this case, the reasons why terminated nuclear materials could be found in these disposal facilities and how obligations under the Comprehensive Safeguards Agreement (CSA) and Additional Protocol (AP) are met when the disposed waste items are intermediate level waste (ILW) or high-level waste (HLW).

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2019

Operational Safety Analysis for Disposal Period of the Future Belgian Geological Disposal Facility - 19249

For more than 40 years, ONDRAF/NIRAS (the federal agency responsible for managing radioactive waste and enriched fissile materials in Belgium) has been studying the concept of a geological disposal facility (GDF) in poorly indurated clays as an option for the long-term management of both category B waste (low-level and intermediate-level long-lived waste - LILW-LL) and category C waste (high-level waste - HLW and spent fuel). In the absence of a policy decision, the proposed Belgian reference option for disposal of such waste is a GDF in poorly indurated clays, located at a depth between 200 m and 600 m. ONDRAF/NIRAS has developed a reference disposal concept for research, development and demonstration (RD and D) purposes. At the current stage of this program, the primary objective of ONDRAF/NIRAS is to establish the safety and feasibility of constructing, operating and closing a Geological Disposal Facility (GDF) at the conceptual design level, without any presumption about the exact location of the facility. This will be evaluated in a first version of the safety and feasibility case (SFC-1 V1) for the geological disposal of category B and C radioactive wastes. Analysis of the operational safety is an important part of the SFC-1 V1. ONDRAF/NIRAS, supported by BGE Technology GmbH, has prepared an operational safety analysis of the current GDF reference design and the processes occurring during the operation and closure period. A SWIFT methodology (Structured What If Technique) is used to identify risks and the measures required to mitigate their consequences. In parallel to this SWIFT analysis, specific attention is also given to fire risk, especially in the blind-ending disposal galleries (DG). Based on understanding of fire risk analysis, recommendations suggest using exclusively electrically driven equipment to reduce fire loads and minimize the fire hazard. In addition, strict working procedures could prevent the presence of workers between the DG dead end and equipment presenting significant fire load. The use of mobile refuge chambers should also be part of the safety concept. (authors)

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DOE OSTI2017

Unreviewed Disposal Question Evaluation: Waste Disposal in Engineered Trenches 3 and 4

Revision 0 of this UDQE addressed the proposal to place Engineered Trench #3 (ET#3) in the footprint designated for Slit Trench #12 (ST#12) and operate using ST#12 disposal limits. Similarly, Revision 1 evaluates whether ET#4 can be located in and operated to Slit Trench #13 (ST#13) disposal limits. Both evaluations conclude that the proposed operations result in an acceptably small risk of exceeding a SOF of 1.0 and approve these actions from a performance assessment (PA) perspective. Because ET#3 will be placed in the location previously designated for ST#12, Solid Waste Management (SWM) requested that the Savannah River National Laboratory (SRNL) determine if the ST#12 limits could be employed as surrogate disposal limits for ET#3 operations. SRNL documented in this Unreviewed Disposal Question Evaluation (UDQE) that the use of ST#12 limits as surrogates for the new ET#3 disposal unit will provide reasonable assurance that Department of Energy (DOE) 435.1 performance objectives and measures (USDOE, 1999) will be protected. Therefore, new ET#3 inventory limits as determined by a Special Analysis (SA) are not required.

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DOE OSTI2019

Why Cleanup Acceptance Is Running into Conflicts - Off-site Disposal Versus On Site Disposal - 19068

For years the U.S. Department of Energy (DOE) identified several sites to receive shipments of various types of wastes. However, with the backlog at the Waste Isolation Pilot Plant (WIPP), the inability to license Yucca Mountain, and the agency's desire to expedite cleanup, DOE is trying to 'dispose of' or 'store' more waste belonging to the Federal government on-site. Some states and communities that were 'promised' disposal of DoE's waste in geologic repositories and licensed landfills in arid locations, are now being told that on-site disposal is the best path forward. Competition for resources is forcing more sites to focus on near term cost savings of on-site disposal, rather than develop comprehensive life-cycle estimates representing a more realistic picture of costs associated with isolating radioactive and other hazardous materials on-site in perpetuity. The paper explores the issues and possible solutions in a case study on proposed on-site disposal of DOE waste, and concludes that DOE needs community support and community acceptance to successfully implement the agency's cleanup mission. Ongoing dialogue with the community saves both time and money but more importantly, cooperation helps ensure that effective long-term cleanup occurs. (author)

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DOE OSTI2017

237 Np analytical method using 239 Np tracers and application to a contaminated nuclear disposal facility

Environmental 237Np analyses are challenged by low 237Np concentrations and lack of an available yield tracer; we report a rapid, inexpensive 237Np analytical approach employing the short lived 239Np (t1/2 = 2.3 days) as a chemical yield tracer followed by 237Np quantification using inductively coupled plasma-mass spectrometry. 239Np tracer is obtained via separation from a 243Am stock solution and standardized using gamma spectrometry immediately prior to sample processing. Rapid digestions using a commercial, 900 watt “Walmart” microwave and Parr microwave vessels result in 99.8 ± 0.1% digestion yields, while chromatographic separations enable Np/U separation factors on the order of 106 and total Np yields of 95 ± 4% (2σ). Application of this method to legacy soil samples surrounding a radioactive disposal facility (the Subsurface Disposal Area at Idaho National Laboratory) reveal the presence of low level 237Np contamination within 600 meters of this site, with maximum 237Np concentrations on the order of 103 times greater than nuclear weapons testing fallout levels.

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DOE OSTI2015

Development of a backfill material for LILW and HLW disposal galleries in the current Belgian geological disposal concept - 15374

In Belgium, it is planned to dispose of long-lived and high-level waste in an underground facility. After disposal of the waste, the galleries will be backfilled to provide stability to the galleries and to limit the amount of voids in the repository. To achieve those goals, the backfill material has to have a good flowability, a negligible bleeding, and a limited shrinkage. A limited grain size is also required to allow the injection of the backfill material. Despite the fact that the backfill supports the gallery lining, its strength must be low enough to enable the retrieval of the waste packages. The backfill material has to be chemically compatible with the Boom Clay and the waste packages. This means that it should not unduly perturb the clay or disposal packages. The thermal conductivity of the backfill material in the galleries containing high-level and thus heat-generating waste must be high enough to allow sufficient dissipation of the decay heat into the surrounding clay. Based on these objectives, material requirements were specified and the development of a backfill mixture was carried out. Initially, the mix composition was optimized in the laboratory. Thereafter, the backfill process of a gallery section was simulated. The investigations illustrate that this mixture can be transported via pipelines through the shaft and drifts and would fill completely the backfill sections in the galleries. Measurements of the porosity, the pore solution composition, the thermal material properties, and the strength illustrate the compliancies with the requirements and the feasibility of backfilling the disposal galleries. (authors)

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DOE OSTI2015

Cavern/Vault Disposal Concepts and Thermal Calculations for Direct Disposal of 37-PWR Size DPCs

This report provides two sets of calculations not presented in previous reports on the technical feasibility of spent nuclear fuel (SNF) disposal directly in dual-purpose canisters (DPCs): 1) thermal calculations for reference disposal concepts using larger 37-PWR size DPC-based waste packages, and 2) analysis and thermal calculations for underground vault-type storage and eventual disposal of DPCs. The reader is referred to the earlier reports (Hardin et al. 2011, 2012, 2013; Hardin and Voegele 2013) for contextual information on DPC direct disposal alternatives.

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DOE OSTI2019

Unreviewed Disposal Question Evaluation: Disposal of the Oak Ridge National Laboratory Contaminated Debris at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This Unreviewed Disposal Question Evaluation (UDQE) assesses whether the Oak Ridge National Laboratory (ORNL) Contaminated Debris (ORNLDEBRIS001, Revision 11 [ORNL 2019a]) 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 waste stream meets all performance objectives in U.S. Department of Energy (DOE) Manual DOE M 435.1-1, Radioactive Waste Management Manual, Chapter IV, Section P (DOE 2011). The ORNL Contaminated Debris waste stream is recommended for acceptance with conditions on the total activity disposed of tritium ( 3 H), niobium-94 ( 94 Nb), technetium-99 ( 99 Tc), cesium-137 ( 137 Cs), iridium-192 ( 192 mIr), radium-226 ( 226 Ra), uranium-234 ( 234 U), and 238 U. The total activity disposed shall not exceed the following limits: 3 H: 2E12 Bq, 94 Nb: 2E9 Bq, 99 Tc: 8E10 Bq, 137 Cs: 1E11 Bq, 192 mIr: 1E9 Bq, 226 Ra: 3E11 Bq, 234 U: 2E9 Bq, 238 U: 5E9 Bq.

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DOE OSTI2012

Waste disposal technology transfer matching requirement clusters for waste disposal facilities in China

Highlights: Black-Right-Pointing-Pointer We outline the differences of Chinese MSW characteristics from Western MSW. Black-Right-Pointing-Pointer We model the requirements of four clusters of plant owner/operators in China. Black-Right-Pointing-Pointer We examine the best technology fit for these requirements via a matrix. Black-Right-Pointing-Pointer Variance in waste input affects result more than training and costs. Black-Right-Pointing-Pointer For China technology adaptation and localisation could become push, not pull factors. - Abstract: Even though technology transfer has been part of development aid programmes for many decades, it has more often than not failed to come to fruition. One reason is the absence of simple guidelines or decision making tools that help operators or plant owners to decide on the most suitable technology to adopt. Practical suggestions for choosing the most suitable technology to combat a specific problem are hard to get and technology drawbacks are not sufficiently highlighted. Western counterparts in technology transfer or development projects often underestimate or don't sufficiently account for the high investment costs for the imported incineration plant; the differing nature of Chinese MSW; the need for trained manpower; and the need to treat flue gas, bunker leakage water, and ash, all of which contain highly toxic elements. This article sets out requirements for municipal solid waste disposal plant owner/operators in China as well as giving an attribute assessment for the prevalent waste disposal plant types in order to assist individual decision makers in their evaluation process for what plant type might be most suitable in a given situation. There is no 'best' plant for all needs and purposes, and requirement constellations rely on generalisations meaning they cannot be blindly applied, but an alignment of a type of plant to a type of owner or operator can realistically be achieved. To this end, a four-step approach is suggested and a technology matrix is set out to ease the choice of technology to transfer and avoid past errors. The four steps are (1) Identification of plant owner/operator requirement clusters; (2) Determination of different municipal solid waste (MSW) treatment plant attributes; (3) Development of a matrix matching requirement clusters to plant attributes; (4) Application of Quality Function Deployment Method to aid in technology localisation. The technology transfer matrices thus derived show significant performance differences between the various technologies available. It is hoped that the resulting research can build a bridge between technology transfer research and waste disposal research in order to enhance the exchange of more sustainable solutions in future.

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DOE OSTI2016

Geological Disposal of Nuclear Waste in Tuff: Yucca Mountain (USA) [Geological Disposal in Tuff: Yucca Mountain]

For more than three decades, the US Department of Energy has investigated the potential for permanent disposal of high-level radioactive waste and spent nuclear fuel in a deep-mined repository at Yucca Mountain, Nevada (USA). A detailed license application submitted to the US Nuclear Regulatory Commission in 2008 provides full documentation of the case for permanent disposal of nuclear waste in tuff. Furthermore, the aridity of the site and great depth to the water table provide a disposal environment and a design concept unique among deep-mined repositories currently or previously proposed worldwide.

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DOE OSTI2015

Robust disposal concept, uniform regulations, and trust in the messenger; three fundamental building blocks for consent-based HLW-disposal solutions in the USA - 15104

Pending enabling legislation, the USA's HLW-disposal program had been held in political abeyance since 2009 at the crossroads to one or both of the Yucca Mountain candidate (since 1987) HLW-repository in Nevada and/or the consent-based siting of the new HLW-disposal and centralized HLW-storage facilities proposed in 2012; a road not previously travelled. The consent-based siting approach would require the following gap between the reality (A) and the requirement (B) for public trust, acceptance and sustained support (PASS) to be bridged: A. Less than 1% of the general public in any given country fully comprehends the scientific-underpinnings and the projected health risks imposed by a HLW repository. B. More than 50% of the general public in a given geographical area would have to accept to host it from start to end. The following PASS-building blocks, successfully used at the WIPP-site in New Mexico and abroad to get from A to B during the past 30+ years, are thus highly recommended: 1. Defining the consent footprint upon host-county/state boundaries and/or radiation/health-risks and/or to identify/designate the directly affected parties (DAPs). 2. Using databases and lessons-learned by other repository programs to advantage. 3. Presenting the information in terms and/or by examples more readily understood by laypeople, because both their acceptance and opposition/rejection are influenced by: a. The type, number, problems, and status of 'similar' disposal concepts; and b. The level of i. Radiation protection provided by applicable regulations; ii. Involvement and decision-making authority vested in the DAPs; and iii. Their trust in the 'spokesperson/messenger. (authors)

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