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DOE OSTI2018

Identification of Waste Streams and Chemicals of Concern for CNL's Near Surface Disposal Facility - 18057

Canadian Nuclear Laboratories (CNL) has a 70-year history in nuclear Research and Development (R and D) and the original support to the Canada Deuterium Uranium (Candu) reactor fleet. The present laboratory and operations include multiple facilities including reactors, laboratories, and buildings at various locations in Canada. The onsite facilities produced and still produce numerous and complex waste streams which have historically been placed in waste storage areas at their sites. CNL has completed the design and is in the process of licensing a Near Surface Disposal Facility (NSDF) for CNL low level waste (LLW). The current plan, pending regulatory approvals would complete construction and commissioning in 2021. CNL has identified the waste streams that are proposed to be disposed of to determine which may or may not require treatment prior to disposal and the Chemicals of Concern (COCs) that require identification and control to protect the facility, workers, public and the environment. CNL gathered the information into a Waste Stream spreadsheet that identifies the waste. Each waste stream was then evaluated as to whether or not it should meet the NSDF Waste Acceptance Criteria (WAC), and if not, which treatment technologies would be appropriate to make it ready for disposal. Understanding the hazards and associated COCs of the waste is required to safely dispose the waste. COCs present at Chalk River Laboratories (CRL) need to be identified and subsequently evaluated to determine if they pose any type of risk due to their hazardous properties or quantities. The COC data base was modified into a tool to assist waste generators. CNL has developed waste stream and COC data bases to support safe disposal of radioactive waste. Identification of waste streams and COCs for disposal is important for any disposal facility. The lessons learned by CNL as to what information is important and how to obtain the information would be useful to many facilities. (authors)

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DOE OSTI2018

Small Waste in a Large Package - Meeting Requirements for Packaging, Transport and Disposal - 18092

A small amount of Class C Mixed Low Level Waste (MLLW) required packaging and disposition to an off-site facility. The AREVA TN RAM cask, designed and licensed for shipping Type B waste packages, was chosen as the most appropriate shipping solution. Due to packaging, shipping and disposal requirements, the standard TN RAM cask could not be utilized without making some modifications. Unique solutions and approaches had to be developed by the AREVA design team to complete the project within the anticipated budget and schedule. The general work scope was to characterize, remove from a pool, profile, package, certify and transport four irradiated components to an approved off-site facility. The material makeup of the components (three stainless steel plates and a lead counterweight) resulted in a MLLW classification which required special packaging and disposal methods. The AREVA design team worked with the disposal facility to identify all requirements associated with disposal of Class B/C MLLW. After defining the packaging and disposal requirements, it was determined that unique spacers and void fillers would be required. The design solution consisted of a liner insert as well as axial and radial spacers filled with inert material. Remaining voids in the liner insert were filled with recycled glass beads to fulfill the requirement for maximum 10% void space in the waste package. The axial spacers also served to center the waste load in the shipping cask to comply with the AREVA TN RAM Cask NRC license. Concentration averaging and waste consolidation were used to reach the Class C MLLW determination for shipping and disposal. This disposal method for Class C waste met the requirements of macro-encapsulation. Therefore, no additional treatment action was required when disposing of Class C MLLW. AREVA's solution provides a model for meeting macro-encapsulation requirements while also meeting void space requirements. This solution provides a model for how to adapt large shipping casks and liners to smaller waste packages that would not normally be a good fit for traditional casks. (authors)

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DOE OSTI2018

DOE/NNSA Efforts to Facilitate the Disposal of Risk Significant Sealed Sources - 18590

Commercial disposal of high-activity Class B and C disused sources has been particularly challenging in the U.S. For many years, a lack of disposal facility access, a shortage of certified of Type B transportation containers, and outdated regulatory guidance had effectively relegated a large number of risk-significant sources to indefinite storage. Progress has been made in all of these efforts. The DOE/NNSA Office of Radiological Security (ORS) and the Conference of Radiation Control Program Directors (CRCPD) Source Collection and Threat Reduction (SCATR) piloted the commercial disposal of a high-activity Cs- 137 source. To facilitate implementation of the NRC's revised waste classification guidance on sealed sources. Disposal of the 560 Ci blood irradiator at the U.S. Ecology facility in Richmond has helped to establish the availability of disposal access for similar sources under the 'alternative approach' provisions of the NRC's 2015 Concentration Averaging and Encapsulation Branch Technical Position (CA BTP). However, while these developments are significant for making disposal access available, the available information suggests that commercial sealed source users continue to keep disused sources in storage. This may be due to remaining uncertainty regarding disposal access, or due to the high cost of commercial disposal. (authors)

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DOE OSTI2020

Special Analyses for the Hanford Integrated Disposal Facility Performance Assessment - 20102

In 2014, the Department of Energy (DOE) Office of River Protection and its contractors began to develop a performance assessment for the near-surface disposal of low-level and mixed low-level waste at the Hanford Site's Integrated Disposal Facility (IDF). The IDF is a doubly-lined landfill that was constructed between 2004 and 2006 to be the disposal facility for the vitrified low-activity waste that will be produced at the Waste Treatment and Immobilization Plant (WTP). IDF is also expected to receive solid secondary waste produced at the WTP and other solid wastes from site activities. The IDF has been in a preoperational state awaiting authorization from DOE and a RCRA permit modification from the State of Washington Department of Ecology to receive waste. Both the Disposal Authorization Statement and permit modification require a performance assessment demonstrating that the system of engineered and natural features will limit releases of radionuclides and hazardous chemicals from the IDF and be protective of human health and the environment. The simulated duration is 10,000 years. Based on the analyses presented in the 2017 Integrated Disposal Facility Performance Assessment, DOE issued a conditional Operating Disposal Authorization Statement for the IDF in June 2018. The long-term performance of the IDF to be protective of human health and the environment was evaluated under the requirements of DOE Order 435.1, Radioactive Waste Management. Computer simulations were performed to evaluate whether or not the IDF would comply with DOE requirements. In the time that has passed since the performance assessment was approved by DOE, new information has been discovered that had not been considered in the performance assessment. Since this new information has not been evaluated, the potential impact of the changes have not been taken into consideration in DoE's disposal authorization. DOE and its contractors follow a change control process to screen and, when necessary, evaluate new information that could potentially impact the conclusions of the completed performance assessment. This paper will describe the change control process and provide two examples of evaluations performed following the change control process. The first example evaluates a new waste form for liquid secondary waste that was not evaluated in the performance assessment. In the performance assessment, liquid secondary waste was assumed to be solidified with grout. A new recommendation to dispose of the liquid secondary waste after drying it to a powder was evaluated. The second example evaluates inventory implications from changes to the flow sheet that estimates the feed composition to the low-activity waste vitrification facility. The changes result in higher strontium concentrations in the vitrified waste stream. (authors)

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DOE OSTI2020

Borehole Disposal of Radioactive Waste in Israel - 20388

Israel is assessing borehole disposal of radioactive waste. With limited geological options for disposal, intermediate-depth borehole disposal is being considered in the arid Yamin Plain region of the northeastern Negev desert at depths of several hundred meters below ground surface in the vadose zone. Unlike deep borehole disposal of several kilometers, which relies on emplacement below the depth of recirculating groundwater, the safety case for intermediate-depth borehole disposal relies more on the aridity of the vadose zone, the robustness of the waste package, and the other engineered barriers (e.g., seals, backfill materials) in the disposal borehole. As a tradeoff, the shallower depth may allow the use of a larger diameter borehole than would be possible for a deeper borehole. To study the suitability of the Yamin Plain region for borehole disposal, a small-diameter characterization borehole is being planned to retrieve core samples and to better understand the vadose zone geomechanical and hydrogeochemical properties and percolation flux. The information from the characterization borehole will inform the safety case, and together with performance assessment analyses will help to identify key areas of uncertainty and guide future research and development activities aimed at demonstrating the feasibility of the intermediate-depth borehole disposal concept in Israel. (authors)

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DOE OSTI2014

Unreviewed Disposal Question Evaluation: Impact of New Information since 2008 PA on Current Low-Level Solid Waste Operations

Solid low-level waste disposal operations are controlled in part by an E-Area Low-Level Waste Facility (ELLWF) Performance Assessment (PA) that was completed by the Savannah River National Laboratory (SRNL) in 2008 (WSRC 2008). Since this baseline analysis, new information pertinent to disposal operations has been identified as a natural outcome of ongoing PA maintenance activities and continuous improvement in model simulation techniques (Flach 2013). An Unreviewed Disposal Question (UDQ) Screening (Attachment 1) has been initiated regarding the continued ability of the ELLWF to meet Department of Energy (DOE) Order 435.1 performance objectives in light of new PA items and data identified since completion of the original UDQ Evaluation (UDQE). The present UDQE assesses the ability of Solid Waste (SW) to meet performance objectives by estimating the influence of new information items on a recent sum-of-fractions (SOF) snapshot for each currently active E-Area low-level waste disposal unit. A final SOF, as impacted by this new information, is projected based on the assumptions that the current disposal limits, Waste Information Tracking System (WITS) administrative controls, and waste stream composition remain unchanged through disposal unit operational closure (Year 2025). Revision 1 of this UDQE addresses the following new PA items and data identified since completion of the original UDQE report in 2013: New K d values for iodine, radium and uranium; Elimination of cellulose degradation product (CDP) factors; Updated radionuclide data; Changes in transport behavior of mobile radionuclides; Potential delay in interim closure beyond 2025; and Component-in-grout (CIG) plume interaction correction. Consideration of new information relative to the 2008 PA baseline generally indicates greater confidence that PA performance objectives will be met than indicated by current SOF metrics. For SLIT9, the previous prohibition of non-crushable containers in revision 0 of this UDQE has rendered the projected final SOF for SLIT9 less than the WITS Admin Limit. With respect to future disposal unit operations in the East Slit Trench Group, consideration of new information for Slit Trench#14 (SLIT14) reduced the current SOF for the limiting All-Pathways 200-1000 year period (AP2) by an order of magnitude and by one quarter for the Beta-Gamma 12-100 year period (BG2) pathway. On the balance, updates to K{sub d} values and dose factors and elimination of CDP factors (generally favorable) more than compensated for the detrimental impact of a more rigorous treatment of plume dispersion. These observations suggest that future operations in the East Slit Trench Group can be conducted with higher confidence using current inventory limits, and that limits could be increased if desired for future low-level waste disposal units. The same general conclusion applies to future ST’s in the West Slit Trench Group based on the Impacted Final SOFs for existing ST’s in that area.

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DOE OSTI2016

Deep Borehole Disposal Concept: Development of Universal Canister Concept of Operations

This report documents key elements of the conceptual design for deep borehole disposal of radioactive waste to support the development of a universal canister concept of operations. A universal canister is a canister that is designed to be able to store, transport, and dispose of radioactive waste without the canister having to be reopened to treat or repackage the waste. This report focuses on the conceptual design for disposal of radioactive waste contained in a universal canister in a deep borehole. The general deep borehole disposal concept consists of drilling a borehole into crystalline basement rock to a depth of about 5 km, emplacing WPs in the lower 2 km of the borehole, and sealing and plugging the upper 3 km. Research and development programs for deep borehole disposal have been ongoing for several years in the United States and the United Kingdom; these studies have shown that deep borehole disposal of radioactive waste could be safe, cost effective, and technically feasible. The design concepts described in this report are workable solutions based on expert judgment, and are intended to guide follow-on design activities. Both preclosure and postclosure safety were considered in the development of the reference design concept. The requirements and assumptions that form the basis for the deep borehole disposal concept include WP performance requirements, radiological protection requirements, surface handling and transport requirements, and emplacement requirements. The key features of the reference disposal concept include borehole drilling and construction concepts, WP designs, and waste handling and emplacement concepts. These features are supported by engineering analyses.

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DOE OSTI2018

Review of Criticality Evaluations for Direct Disposal of DPCs and Recommendations

There are currently 2,462 dual-purpose canisters (DPCs) containing spent nuclear fuel (SNF) across the United States. Repackaging DPCs into specialized disposal canisters can be financially and operationally costly with undue risks. Technical feasibility of direct disposal of DPCs has been evaluated by the Department of Energy (DOE) and industry over the past 15 years. A concerted effort most recently conducted by DOE Office of Nuclear Energy (NE) Spent Fuel and Waste Science and Technology (SFWST) research and development (R&D) programs is evaluating the technical feasibility of direct disposal of DPCs in various geologies. This report focuses on reviewing the work completed by SFWST for the criticality considerations of DPC geologic disposal. Disposal of DPCs is not only viable, but assured from a technical and assumed regulatory perspective (similar to 10 CFR 63). The analysis approach should be multi-faceted to ensure effective implementation of a licensing basis. Recommendations are provided in this report that could enhance the bases for direct disposal of DPCs by exploiting all technically attainable and regulatorily defensible options. The review objectives, including addressing several questions regarding the value of accumulating asloaded fuel and DPC design data, suitability of DPC designs for disposal, and reasonable modifications for loading of DPCs that could facilitate eventual disposal, are also addressed in this report.

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DOE OSTI2018

Advances in Geologic Disposal Safety Assessment and an Unsaturated Alluvium Reference Case

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Depat ment of Energy (DOE) Office of Nuclear Energy (NE), Office of Fuel Cycle Technology (OFCT) is conducting research and development (R&D) on geologic disposal of spent nuclear fuel (SNF) and high level nuclear waste (HLW). Two high priorities for SFWST disposal R&D are design concept development and disposal system modeling (DOE 2011, Table 6). These priorities are directly addressed in the SFWST Geologic Disposal Safety Assessment (GDSA) work package, which is charged with developing a disposal system modeling and analysis capability for evaluating disposal system performance for nuclear waste in geologic media. This report describes specific GDSA activities in fiscal year 2018 (FY 2018) toward the development of GDSA Framework, an enhanced disposal system modeling and analysis capability for geologic disposal of nuclear waste. GDSA Framework employs the PFLOTRAN thermal-hydrologic-chemical multiphysics code (Hammond et al. 2011a; Lichtner and Hammond 2012) and the Dakota uncertainty sampling and propagation code (Adams et al. 2012; Adams et al. 2013). Each code is designed for massivelyparallel processing in a high-performance computing (HPC) environment. Multi-physics representations in PFLOTRAN are used to simulate various coupled processes including heat flow, fluid flow, waste dissolution, radionuclide release, radionuclide decay and ingrowth, precipitation and dissolution of secondary phases, and radionuclide transport through engineered barriers and natural geologic barriers to the biosphere. Dakota is used to generate sets of representative realizations and to analyze parameter sensitivity.

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DOE OSTI2020

Performance Assessment for the Environmental Restoration Disposal Facility (FY 2019 Annual Status Report)

DOE O 435.1 and DOE M 435.1-1 require that a determination of continued adequacy of the performance assessment (PA), composite analysis, and disposal authorization statement (DAS) be annually, and it must consider the results of data collection and analysis from research, field studies, and monitoring as well as update any Radioactive Waste Management Basis documents. Beginning in 1996, the Environmental Restoration Disposal Facility (ERDF) has accepted low-level radioactive, hazardous, and mixed wastes that were generated during cleanup activities at the Hanford Site. ERDF is composed of a series of cells (currently 10) or disposal areas and can accommodate future design expansions as needed. During this reporting period (fiscal year 2019, October 1, 2018, to September 30, 2019). approximately 139,721 U.S. tons of waste was disposed at ERDF. From inception through September 30, 2019, ERDF has received approximately 18.5 million U.S. tons of waste, which equates to consumption of approximately 88% of the disposal volume. As a condition of the DAS, disposal operations within ERDF must be in accordance with the waste acceptance criteria (WAC) that provide specific radionuclide disposal limits, waste from restrictions, and descriptions of acceptable waste packages in compliance with DOE M 435-1.1 requirements. The ERDF WAC stipulates that waste destined for disposal at ERDF be controlled based on source, physical form, and containment concentration and activity levels

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DOE OSTI2019

Preliminary Engineering and Cost Analysis for DPC Disposal Solutions. Final Draft

There are currently (as of January, 2019) more than 2,700 dual-purpose canisters (DPCs) loaded with spent nuclear fuel (SNF) across the United States. DPCs continue to be loaded at a rate of more than 200 per year by mid-century there are likely to be more than 8,160 DPCs in service. Options for disposing of SNF loaded in DPCs include repackaging into specialized disposal canisters, directly disposing of the loaded DPCs (with or without modification), or some combination of the two. The main technical challenges for direct disposal of loaded DPCs are thermal management, handling and emplacement operations for the large, heavy packages, and postclosure criticality control. This report focuses on postclosure criticality control which is the most challenging. The challenge lies in determining how to modify DPCs so as to minimize the probability that a criticality event might occur in a repository, or if the DPCs are not modified, to understand the nature and consequences of postclosure criticality events. There are several approaches that could facilitate direct disposal of loaded DPCs with acceptable repository performance. This report describes these approaches and presents comparative analysis of the rough-order-of-magnitude (ROM) costs. Repackaging SNF in DPCs into specialized disposal canisters could be financially and operationally costly with additional radiological, operational safety, and management risks. A disposition approach that would not involve repackaging or modifications to DPCs (future or already loaded) is development of a new licensing strategy that addresses the risk (probability and consequence) from criticality events. A different approach would modify existing loaded DPCs (some or all of them), and change the loading or design of future DPCs, to decrease the probability of a criticality event in a repository below levels of concern. This report investigates the cost to modify existing loaded DPCs, and the cost to modify the loading or design of future DPCs to facilitate direct disposal. It establishes the ROM cost for repackaging SNF that has been loaded into DPCs, into specialized canisters for disposal. It also identifies technical and regulatory challenges associated with the potential design modifications and loading considerations. It is left to future analyses to compare radiological, operational safety, and management risks associated with the available approaches.

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DOE OSTI2014

US DOE-EM On-Site Disposal Cell Working Group - Fostering Communication On Performance Assessment Challenges

On-site disposal cells are in use and being considered at several U.S. Department of Energy (USDOE) sites as the final disposition for large amounts of waste associated with cleanup of contaminated areas and facilities. These facilities are typically developed with regulatory oversight from States and/or the US Environmental Protection Agency (USEPA) in addition to USDOE. The facilities are developed to meet design standards for disposal of hazardous waste as well as the USDOE performance based standards for disposal of radioactive waste. The involvement of multiple and different regulators for facilities across separate sites has resulted in some differences in expectations for performance assessments and risk assessments (PA/RA) that are developed for the disposal facilities. The USDOE-EM Office of Site Restoration formed a working group to foster improved communication and sharing of information for personnel associated with these Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) disposal cells and work towards more consistent assumptions, as appropriate, for technical and policy considerations related to performance and risk assessments in support of a Record of Decision and Disposal Authorization Statement. The working group holds teleconferences, as needed, focusing on specific topics of interest. The topics addressed to date include an assessment of the assumptions used for performance assessments and risk assessments (PA/RAs) for on-site disposal cells, requirements and assumptions related to assessment of inadvertent intrusion, DOE Manual 435.1-1 requirements, and approaches for consideration of the long-term performance of liners and covers in the context of PAs. The working group has improved communication among the staff and oversight personnel responsible for onsite disposal cells and has provided a forum to identify and resolve common concerns.

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DOE OSTI2015

Field Test to Evaluate Deep Borehole Disposal.

The U.S. Department of Energy (DOE) has embarked on the Deep Borehole Field Test (DBFT), which will investigate whether conditions suitable for disposal of radioactive waste can be found at a depth of up to 5 km in the earth’s crust. As planned, the DBFT will demonstrate drilling and construction of two boreholes, one for initial scientific characterization, and the other at a larger diameter such as could be appropriate for waste disposal (the DBFT will not involve radioactive waste). A wide range of geoscience activities is planned for the Characterization Borehole, and an engineering demonstration of test package emplacement and retrieval is planned for the larger Field Test Borehole. Characterization activities will focus on measurements and samples that are important for evaluating the long-term isolation capability of the Deep Borehole Disposal (DBD) concept. Engineering demonstration activities will focus on providing data to evaluate the concept’s operational safety and practicality. Procurement of a scientifically acceptable DBFT site and a site management contractor is now underway. The concept of deep borehole disposal (DBD) for radioactive wastes is not new. It was considered by the National Academy of Science (NAS 1957) for liquid waste, studied in the 1980’s in the U.S. (Woodward–Clyde 1983), and has been evaluated by European waste disposal R&D programs in the past few decades (for example, Grundfelt and Crawford 2014; Grundfelt 2010). Deep injection of wastewater including hazardous wastes is ongoing in the U.S. and regulated by the Environmental Protection Agency (EPA 2001). The DBFT is being conducted with a view to use the DBD concept for future disposal of smaller-quantity, DOE-managed wastes from nuclear weapons production (i.e., Cs/Sr capsules and granular solid wastes). However, the concept may also have broader applicability for nations that have a need to dispose of limited amounts of spent fuel from nuclear power reactors. For such nations the cost for disposing of volumetrically limited waste streams could be lower than mined geologic repositories.

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DOE OSTI2019

The corrosion of UK vitrified waste during disposal

The UK's vitrified high-level-waste (HLW) is destined to be disposed of in a deep geological disposal facility (GDF). Contact with groundwater will be the process by which radionuclides are released during 100,000+ years of disposal. Understanding the mechanisms and kinetics of glass dissolution allow models and predictions to be made on the long-term fate of the disposed HLW. Current standard laboratory scale experimental methodologies may be unrepresentative of the geological disposal environment. Aim: Provide insight into the dissolution behaviour of alkali-alumino-borosilicate UK vitrified waste in geological disposal environments. A 'cyclic vapor-static' dissolution methodology is proposed, capturing expected temporal behaviour. This will elucidate dissolution kinetics and mechanisms, optimize the base glass composition, waste loading content and generate significant, pragmatic disposal performance assessment data. Variants of three glasses to be investigated: waste-loaded MW25 and ZnCa MW28, and the ZnCa MW base. Preliminary Results: Suggests drying process decreases the chemical durability of the glass. Addition of ZnCa to UK HLW improves chemical durability. Increased waste-loading improves chemical durability. Banded structure in altered glass possibly attributed to dissolution/reprecipitation of secondary phases). Conclusion - Further Work: Cyclic vapor-static dissolution methodology may elucidate dissolution behaviour expected within a UK GDF. Potential to provide beneficial, pragmatic data for the performance assessment to implement geological disposal. Fundamental mechanisms to be understood.

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DOE OSTI2015

Applying understanding of likelihood and consequences to demonstrate post-closure criticality safety of a geological disposal facility - 14377

Radioactive Waste Management (RWM) is responsible for implementing geological disposal of the UK's higher activity wastes. These wastes include significant quantities of fissile nuclides and, therefore, demonstration of criticality safety under disposal conditions forms an important component of RWM's disposal system safety case. At the time of disposal, controls (e.g. fissile material limits and/or water barriers) will ensure that waste packages are sub-critical. These controls, plus the containment provided by the engineered barrier system of a geological disposal facility (GDF) will also ensure that criticality is prevented for a long time after disposal. However, over post-closure timescales conditions in the GDF will evolve, and therefore underpinning knowledge is required to demonstrate: under what conditions criticality could occur; what the likelihood of such conditions developing is; and what the consequences would be if a post-closure criticality did occur. These requirements are set out in the UK environment agencies' Guidance on Requirements for Authorisation of a GDF. This paper discusses the disposal of Pressurised Water Reactor fuel, as an example of a waste type that may require disposal. It briefly selects relevant research to combine and summarise results from three recent RWM research projects. The first 'Likelihood of Criticality' considered under what conditions post-closure criticality could occur, whilst a second 'Modelling of Consequences of Hypothetical Criticality' focused on understanding the local physical consequences (temperature rise, power and rock response) of hypothetical 'what-if' post-closure criticality. The paper then explains how this understanding is applied to make safety arguments as to why post-closure criticality is not a significant concern. This is achieved via a third, stream of work that has considered the effects of a potential criticality on pathways that might give rise to a post-closure risk in the future, such as transport of radionuclides in groundwater. Collectively RWMs post-closure criticality safety work allows us to conclude in our generic environmental safety case that a post closure criticality is a low likelihood event, and even if one of these low likelihood events was to occur its consequences would be insignificant to GDF performance. This paper also serves to give context to two companion UK post-closure criticality safety papers which communicate further information on our modelling development and more detailed results underpinning our demonstration of post-closure criticality safety. (authors)

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DOE OSTI2015

Understanding the consequences of hypothetical post-closure transient criticality in a geological disposal facility - 14439

Radioactive Waste Management (RWM) is responsible for implementing geological disposal of the UK's higher-activity radioactive wastes. These wastes include significant quantities of fissile nuclides and, therefore, demonstration of criticality safety for the wastes under disposal conditions forms an important component of RWM's disposal system safety case. At the time of disposal all waste packages will be safely sub-critical. Containment provided by the engineered barrier system of a geological disposal facility (GDF) will ensure that criticality is not possible for some time after disposal. However, as waste packages degrade, fissile and other materials may be mobilised, which could affect the potential for criticality. Although the likelihood of a critical configuration developing is demonstrably low, it cannot be assumed to be zero over long post-closure timescales. The UK environment agencies' Guidance on Requirements for Authorisation for a GDF includes a specific requirement to consider the consequences of hypothetical 'what-if' post-closure criticality. RWM has undertaken a programme of research to model and understand the consequences of hypothetical post-closure criticality events. Initially focused on the scenario of fissile material relocation and accumulation following intermediate level waste (ILW) disposal, two bespoke computer models were developed. The quasi-steady-state (QSS) model considered critical systems maintained by accumulation of fissile materials where the temperature feedback remains negative throughout the transient. The rapid transient model (RTM) considered systems with positive temperature feedback. The QSS model development included a successful comparison with the Oklo natural reactors. With a requirement to consider post-closure criticality for a wider range of potential waste types under different illustrative disposal concepts, the initial transient criticality models had some limitations. For the QSS model this was the assumption of fixed geometry, while the complex functional inputs required in the RTM were not readily available for different materials. For the QSS model this paper describes the model development to allow changing volume; this has enabled its application to 'what-if' scenarios of flooding within a spent fuel waste package and the slumping of a stack of ILW packages. For rapid transient criticality this paper explains how the development of a simpler 'bounding approach' model to complement the RTM has enabled transient criticality analysis for a wider range of host materials. The role of this research in demonstrating post-closure criticality safety of a GDF in the UK is presented in a companion paper which also draws on the results of a parallel research project to demonstrate that the likelihood of post-closure criticality is low. (authors)

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DOE OSTI2015

Multi-pack Disposal Concepts for Spent Fuel (Rev. 0)

At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media (Hardin et al., 2012). Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design (CRWMS M&O, 1999). Thermal analysis showed that, if “enclosed” concepts are constrained by peak package/buffer temperature, waste package capacity is limited to 4 PWR assemblies (or 9-BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems (EnergySolution, 2015). This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).

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DOE OSTI2016

Multi-Pack Disposal Concepts for Spent Fuel (Revision 1)

At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media. Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design. Thermal analysis showed that if “enclosed” concepts are constrained by peak package/buffer temperature, that waste package capacity is limited to 4 PWR assemblies (or 9 BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems. This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).

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