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DOE OSTI2024

Salt International Collaborations FY24 Update

This report summarizes the international collaborations conducted by Sandia funded by the US Department of Energy Office (DOE) of Nuclear Energy (DOE-NE) Spent Fuel and Waste Science & Technology (SFWST) as part of the Sandia National Laboratories Salt R&D and Salt International work packages. This report satisfies the level-three milestone M3SF-24SN010303063. Several stand-alone sections make up this summary report, each completed by the participants. The sections discuss granular salt reconsolidation (KOMPASS/MEASURES), engineered barriers (RANGERS), numerical model comparison (DECOVALEX), an NEA Salt Club working group on the development of scenarios as part of the performance assessment development process, and progress on seal percolation experiments with German colleagues at Gesellschaft für Anlagen- und Reaktorsicherheit (GRS). Finally, we summarize events related to the US/German Workshop on Repository Research, Design and Operations.

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DOE OSTI2015

Impact of lump characteristics on self-shielding corrections of active neutron interrogation on fissile material in waste - 15405

Non-destructive active neutron assay techniques such as active neutron coincidence counting, differential die away analysis or the californium shuffler method rely on the interrogation of fissile nuclear materials by an external neutron source. Whether it is a spontaneous fission or (α,n) neutron source, like {sup 252}Cf or AmLi; or a fusion neutron generator, all of these techniques rely on induced secondary fission events to provide prompt or delayed neutron signals that serve as the basis for the fissile material quantification. The accuracy of these fissile mass measurements is therefore dependent on the neutron interrogation efficiencies determined for calibration reference standards and whether the measured items are comparable to the calibration condition. In such favorable conditions, it is possible to establish a reliable empirical non-linear relationship between apparent (uncorrected) and real or true fissile mass. However, in waste measurements, this is unlikely to be the case; thus a calibration approach that takes account of the neutron self-shielding effect is required. By incorporating the Self-Shielding Factor (SSF) of the calibration reference standards, a linear mass calibration factor [s{sup -1}. {sup 239}Pu{sub eq} g{sup -1}] can be found; this calibration is accurate for waste in which the fissile material is 'infinitely dilute' i.e. the waste exhibits no self-shielding; this is assumed to be representative of dispersed material. The self-shielding correction and its systematic uncertainty are then left to be treated as part of the Total Measurement Uncertainty (TMU) evaluation associated with the assay result. The derivation of the TMU contribution associated with the self-shielding effect in nuclear waste is a complex problem for which a reliable evaluation has to be performed. This is usually based on assuming 'worst case scenarios' for the distribution of material that can arise in a particular waste stream. Such an approach is conservative, but often very penalizing, especially if one considers larger lumps (those containing grams of fissile material); for such lumps, the self-shielding correction becomes very significant, potentially altering the mass of {sup 239}Pu{sub eq} by over an order of magnitude. A worst case 'nuclear safety value' may not be the most appropriate choice for building a consignment best estimate or repository sentencing value for an item. In the present work the authors describe a numerical approach to study the variability of neutron self-shielding factors associated with distributions of a wide range of lump characteristics in terms of size, shape, chemical composition, density and isotopic composition. The described novel calculation tool is capable of accurately predicting the real impact of random distribution of lumps in a waste material, and allowed to show that given reasonable, defensible, assumptions on the composition of the lumps in waste stream, the systematic error component of the DDA TMU can be significantly reduced. (authors)

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DOE OSTI2018

Criticality Consequence Analysis Roadmap for Spent Nuclear Fuel Canister in a Repository

This report discusses an initial criticality consequence analysis approach of commercial spent fuel canisters in a repository that will be developed to support a performance assessment (PA) of a repository with criticality events. A repository PA with criticality events may be needed to support direct disposal of currently loaded dual-purpose canisters. The ultimate goal is to develop a criticality consequence analysis model that can interface with any repository PA model. The criticality consequence analysis approach describes multi-physics coupling between neutronics, thermal hydraulics, and mechanics analysis codes. This report also discusses the sensitivity and uncertainty analysis approach for quantifying conservatism in the criticality consequence outcomes, as well as a defensible multi-physics validation approach to support any future licensing effort. This criticality consequence analysis approach will be refined in the future to support specific regulatory frameworks for a specific repository PA if needed.

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DOE OSTI2026

Report priority gaps in high temperature thermodynamic data (Interim Progress Report)

This interim progress report (Level 4 Milestone Number M4SF-26LL010203023) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Argillite Host Rock Properties & Processes SF-26LL01020302. Our focus is to assess gaps in data availability and understanding for radionuclide thermodynamics within the context of a “hot repository” concept and expand SUPCRT-NE database development to address higher temperatures needed for a DPC DGR disposal concept. The database is intended to inform the argillite GDSA baseline model. The leading European thermochemical database (Thermochimie) is only applicable to temperatures below 80°C. Thus, a US effort to integrate and expand upon other international thermodynamics database efforts is needed, particular if a “hot repository” concept moves forward.

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DOE OSTI2015

Real-Time Electrochemical Measurements Of Carbon Steel In Ground Water With Sulfate Reducing Bacteria Enrichment - 15604

A new laboratory system for studying the microbiological corrosion of decommissioning waste was designed and developed during this project. The material in this part of the laboratory studies was carbon steel (AISI/SAE 1005/UNS G10050). In order to estimate the corrosive effects of microbial activity both biotic and abiotic (containing the biocides methyl-isothiazolinone and glutaraldehyde) mesocosms were prepared. To enhance the effects of sulfate reducing bacteria an enrichment of these bacteria isolated from the native drill hole water from the repository site was inoculated in to the biotic mesocosms. Microbiological corrosion was studied using new electrochemical technologies, such as multi-electrode arrays sensors, and molecular biology methods in the laboratory. (authors)

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DOE OSTI2015

Safety assessment for the radon-type surface disposal facility at Saakadze, Georgia - 15406

The Saakadze disposal facility is located about 40 km from Tbilisi and was planned and constructed during Soviet times as Radon-type surface facility to accommodate up to 600 m{sup 3} of low and intermediate level waste generated on the territory of Georgia. The disposal facility was operated from 1963 until its closure in 1995. The vast majority of waste in the Saakadze facility has been disposed of in two near surface concrete vaults with typical Radon-type dimensions, which cover layers already show significant structural damages. In order to assess the present safety of the site and provide the competent Georgian authorities with the necessary basis to take decisions in regard to the future of the repository, the European Commission launched and financially supported a project, which was assigned to a consortium of TUV-Nord ENSYS GmbH and DBE Technology GmbH. The safety assessment of Saakadze facility, which was an essential part of this project and which is the subject of this paper was carried out by DBE Technology GmbH. The Government of Georgia has been gradually increasing efforts to improve the regulatory processes and the legislation on nuclear and radiation safety, and, in particular, on safe management of radioactive waste and sealed radioactive sources. Currently, however, there are still areas where specific Georgian laws and regulations do not exist. Accordingly, in certain aspects IAEA recommendations and best practices were used as guidelines for the methodology of the safety assessment. In spite of the large uncertainties to be considered, especially in regard to the inventory, preliminary results of the safety assessment indicate that scenarios related to the groundwater pathway do not lead to radiological impacts that are above or near to limiting values. However, human intrusion scenarios might lead to dose rates for the critical group that are exceeding these values. This is mainly due to the fact that today there is no proper engineered cover on top of the concrete vaults. Based on the comparison of existing options and considering the results of the long-term safety assessment, it is recommended to backfill the remaining voids inside the concrete vaults and to construct an engineered cover to reduce the probability of human intrusion scenarios and to minimize their potential impact. (authors)

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DOE OSTI2015

Parameter Uncertainty for Repository Thermal Analysis

This report is one follow-on to a study of reference geologic disposal design concepts (Hardin et al. 2011a). Based on an analysis of maximum temperatures, that study concluded that certain disposal concepts would require extended decay storage prior to emplacement, or the use of small waste packages, or both. The study used nominal values for thermal properties of host geologic media and engineered materials, demonstrating the need for uncertainty analysis to support the conclusions. This report is a first step that identifies the input parameters of the maximum temperature calculation, surveys published data on measured values, uses an analytical approach to determine which parameters are most important, and performs an example sensitivity analysis. Using results from this first step, temperature calculations planned for FY12 can focus on only the important parameters, and can use the uncertainty ranges reported here. The survey of published information on thermal properties of geologic media and engineered materials, is intended to be sufficient for use in generic calculations to evaluate the feasibility of reference disposal concepts. A full compendium of literature data is beyond the scope of this report. The term “uncertainty” is used here to represent both measurement uncertainty and spatial variability, or variability across host geologic units. For the most important parameters (e.g., buffer thermal conductivity) the extent of literature data surveyed samples these different forms of uncertainty and variability. Finally, this report is intended to be one chapter or section of a larger FY12 deliverable summarizing all the work on design concepts and thermal load management for geologic disposal (M3FT-12SN0804032, due 15Aug2012).

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DOE OSTI2018

ENRESA's Spent Nuclear Fuel and Special Waste Database - 18115

Empresa Nacional de Residuos Radiactivos S.A. (ENRESA, by its initials in Spanish) Spent Nuclear Fuel (SNF) and Special Waste (SW) database development was initiated in a common effort between ENRESA and UNESA (the Spanish Association of the Electricity Industry) as a consequence of new regulatory requirements in the frame of radioactive waste management plans. Although these requirements were only mandatory for operators of nuclear power plants, ENRESA deemed necessary to have a centralized database which collected the data from the plant operators. The most relevant features of this database are: - The information collected covers not only the general characteristics of the SNF and SW but also the operation details needed for their characterization. - An application, named Gestion de Elementos Combustibles y Residuos Especiales (GECYRE, by its initials in Spanish) has been implemented for easing the queries and the generation of predefined reports. - An agile data transfer scheme, based on XML files, is used for an effective communication between ENRESA and the operators. The database is currently in production phase since 2014. During this phase it has demonstrated its effectiveness as an information repository to address the SNF and SW interim storage projects in Spain foreseen in the near future. (authors)

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DOE OSTI2019

Overview of Canada's Plan for the Long-Term Management of Used Nuclear Fuel - 19537

In June 2007, the Government of Canada approved the Nuclear Waste Management Organisation's (NWMO) recommended Adaptive Phased Management (APM) approach for the long-term management of Canada's used nuclear fuel. As per this approach, used nuclear fuel will be isolated and contained in a centralized Deep Geological Repository (DGR) located in a stable crystalline or sedimentary rock formation at a nominal depth of 500 metres below ground surface. The DGR concept contains multiple barriers to safely contain and isolate the used nuclear fuel over the long-term. The used nuclear fuel bundles will be encapsulated in a copper-coated steel Used Fuel Container (UFC) and pre-packaged in a bentonite clay buffer box above ground. The UFC and bentonite clay buffer box assemblies will then be emplaced in excavated tunnel openings and sealed with clay-based backfill material and room seals, which completes the primary Engineered Barrier System (EBS). Since 2011, the NWMO has executed a design-development program to validate the design and manufacturing technologies for the EBS. Currently, the NWMO are executing a proof test program for the EBS for the purpose of demonstrating serial production. As the prototypes are produced, they will be subject to full-scale in-room emplacement trials within a custom-built above-ground mock emplacement room. The proof test program will be executed using APM reference designs, materials, processes, and procedures. An overview of this program is discussed. (authors)

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DOE OSTI2018

Preliminary Preclosure Accident Consequence Analysis for the APM Conceptual Design - 18090

The Nuclear Waste Management Organization (NWMO) is implementing Adaptive Phased Management (APM), Canada's plan for the long-term management of used nuclear fuel. The APM approach encompasses centralized containment and isolation of the used fuel in a Deep Geological Repository (DGR) in a suitable rock formation, such as crystalline rock or sedimentary rock, in an informed and willing host. This paper describes a preliminary analysis of the potential dose consequences to the public under accident conditions associated with the operation of a DGR and related surface processing facilities. These results are for a generic site and considers exposure to a person standing at a potential fenceline location under conservative atmospheric conditions. Radiological dose consequences to the public were calculated for accidents identified in the preliminary hazard identification. The presence or absence of ventilation system HEPA filters was also considered in combination with specific accident scenarios. A Gaussian dispersion model considering the most stable atmospheric condition (Pasquill stability category F) was used to estimate the exposure of a member of the public in the direct plume path at the location of maximum airborne radionuclide concentrations. The public dose was calculated via the inhalation, air immersion and ground exposure pathways. For all accidents considered, the calculated public dose was low, up to about 0.8 mSv for a fall of an UFTP with container breach. The total dose to the public was also calculated at various distances from the Used Fuel Packaging Plant, Ventilation Shaft, and Main Shaft. For all accidents considered, the calculated dose to the public was low, below 1 mSv even at the minimum 100 m distance evaluated. Sensitivity cases were also carried out to determine the effect of stack release height, the effluent exit velocity, and the release orientation on the calculated minimum site boundary distance. (authors)

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DOE OSTI2021

Salt International Collaborations FY2021 Update

This report summarizes the international collaboration work conducted by Sandia and funded by the US Department of Energy Office (DOE) of Nuclear Energy Spent Fuel and Waste Science & Technology (SFWST) as part of the Sandia National Laboratories Salt R&D and Salt International work packages. This report satisfies the level-three milestone M3SF-20SN010303062. Several stand-alone sections make up this summary report, each completed by the participants. The sections discuss international collaborations on geomechanical benchmarking exercises (WEIMOS), granular salt reconsolidation (KOMPASS), engineered barriers (RANGERS), and model comparison (DECOVALEX). Lastly, the report summarizes a newly developed working group on the development of scenarios as part of the performance assessment development process, and the activities related to the Nuclear Energy Agency (NEA) Salt club and the US/German Workshop on Repository Research, Design and Operations.

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DOE OSTI2019

Dismantling of the Spent Resins Hot Compaction Unit at Tihange NPP, Belgium - 19139

Design, construction and commissioning of the Spent Ion Exchange Resins Hot Compaction Unit at Tihange Nuclear Power Plant took place from 2005 till 2011. In this conditioning process, batches of radioactive resin beads were first dried to less than 4% moisture by a conical vacuum dryer. Resins were then poured in compactable 200-liter drums and super-compacted. Resulting pellets were emplaced in standard 400-liter drums, the packages qualified for final disposal in Belgium, and eventually grouted. The Belgian Waste Management Agency (ONDRAF/NIRAS) was consulted before initiating the project and issued a statement of no objection. The technology turned out to be used for the first time in the particular case of bead shaped resins (unlike powdered resins) and with a grouting requirement. Finally, commissioning tests ended in a safe and reproducible waste form, compliant with prevailing Waste Acceptance Criteria. Volume Reduction Factors of about 2 were recorded after drying and compaction. New tests were carried out to support the Qualification File of the process, which was submitted to ONDRAF/NIRAS. After performing many trials using inactive resins, the Operator dried two batches of radioactive resins and super-compacted four drums: this was a planned qualification step, aiming at the validation of process parameters. In the meantime, ONDRAF/NIRAS made parallel progress in the design of the future Near Surface Repository, where conditioned resins will be disposed of, and found some incompatibility between encapsulated dried resin beads and water intrusion hazards inside the repository. Based on the risk that dried resin beads will swell if they absorb water, causing mechanical stresses likely to compromise the integrity of the disposal site, ONDRAF/NIRAS informed Tihange NPP that the Hot Compaction Unit would not be qualified as spent resins conditioning process. Further to this irrevocable decision, the Operator undertook three actions: - The first one was the qualification and purchase of waste containers to provide additional temporary storage capacity on site for its operational spent resins; - The second one is the selection of a new spent resins conditioning process and the performance of the related qualification tests to demonstrate it meets the updated Waste Acceptance Criteria; - The last one is the dismantling of the Hot Compaction Unit to clear the space and enable the installation of new equipment. This latter task was carried out in 2018, involving decontamination and free release measurements of all the materials, a very limited fraction of it having to be further processed as radioactive waste. Nevertheless, Tihange NPP wishes to keep the super-compactor and its hydraulic unit for a possible resale or reuse for the treatment of mixed waste. (authors)

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DOE OSTI2015

Validation Study for Crediting Chlorine in Criticality Analyses for US Spent Nuclear Fuel Disposition

Spent nuclear fuel (SNF) management practices in the United States rely on dry storage systems that include both canister- and cask-based systems. The United States Department of Energy Used Fuel Disposition Campaign is examining the feasibility of direct disposal of dual-purpose (storage and transportation) canisters (DPCs) in a geological repository. One of the major technical challenges for direct disposal is the ability to demonstrate the subcriticality of the DPCs loaded with SNF for the repository performance period (e.g., 10,000 years or more) as the DPCs may undergo degradation over time. Specifically, groundwater ingress into the DPC (i.e., flooding) could allow the system to achieve criticality in scenarios where the neutron absorber plates in the DPC basket have degraded. However, as was shown by Banerjee et al., some aqueous species in the groundwater provide noticeable reactivity reduction for these systems. For certain amounts of particular aqueous species (e.g., chlorine, lithium) in the groundwater, subcriticality can be demonstrated even for DPCs with complete degradation of the neutron absorber plates or a degraded fuel basket configuration. It has been demonstrated that chlorine is the leading impurity, as indicated by significant neutron absorption in the water that is available in reasonable quantities for the deep geological repository media under consideration. This paper presents the results of an investigation of the available integral experiments worldwide that could be used to validate DPC disposal criticality evaluations, including credit for chlorine. Due to the small number of applicable critical configurations, validation through traditional trending analysis was not possible. The bias in the eigenvalue of the application systems due only to the chlorine was calculated using TSURFER analysis and found to be on the order of 100 percent mille (1 pcm = 10 -5 k eff ). This study investigated the design of a series of critical configurations with varying amounts of chlorine to address validation gaps. Such integral experiments would support the crediting of the chlorine neutron-absorption properties in groundwater and the demonstration of subcriticality for DPCs in deep geologic repositories with sufficient chlorine availability.

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DOE OSTI2015

Validation Study for Crediting Chlorine in Criticality Analyses for Spent Nuclear Fuel Disposition

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DOE OSTI2017

Westinghouse Quiver - A simple and safe system for the complete handling and storing of failed BWR and PWR fuel rods and fragments - 17049

Background: Extracted leaking fuel rods from fuel assemblies have been stored in the spent fuel pools at the Nuclear Power Plants for many years. Sending those fuel rods to an intermediate storage has been very difficult due to the uncertainty of the technical requirements. The Westinghouse Quiver is a container that can be used from the first moment when the leaking fuel rod is extracted from the Fuel Assembly. The leaking fuel rod can be placed in the Westinghouse Quiver. When the Quiver is filled, it can be sealed and dried and then treated as a regular fuel assembly during intermediate storage and final repository. Purpose of System: The purpose of the system is to store leaking fuel rods, broken rods and fuel rod fragments leak tight in the fuel pool and after vacuum drying and permanent sealing at an intermediate storage facility. The Quiver is designed to be fully compatible with a regular fuel assembly. That means all handling of the Quiver in a Nuclear Power Plant, during transportation or storage is also identical to that of a regular fuel assembly. Fulfilling these criteria together with functional testing, the Quiver is considered acceptable for long time storage of leaking fuel rods, both in dry or wet condition. The Quiver has been qualified in Switzerland for use in the spent fuel pool, intermediate storage, and for a final repository. The system will be implemented as well in Sweden during 2016-2019. (authors)

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DOE OSTI2022

GDSA Framework Development and Process Model Integration FY2022

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Spent Fuel & Waste Disposition (SFWD) is conducting research and development (R&D) on geologic disposal of spent nuclear fuel (SNF) and high-level nuclear waste (HLW). A high priority for SFWST disposal R&D is disposal system modeling (Sassani et al. 2021). The SFWST Geologic Disposal Safety Assessment (GDSA) work package is charged with developing a disposal system modeling and analysis capability for evaluating generic disposal system performance for nuclear waste in geologic media. This report describes fiscal year (FY) 2022 advances of the Geologic Disposal Safety Assessment (GDSA) performance assessment (PA) development groups of the SFWST Campaign. The common mission of these groups is to develop a geologic disposal system modeling capability for nuclear waste that can be used to assess probabilistically the performance of generic disposal options and generic sites. The modeling capability under development is called GDSA Framework (pa.sandia.gov). GDSA Framework is a coordinated set of codes and databases designed for probabilistically simulating the release and transport of disposed radionuclides from a repository to the biosphere for post-closure performance assessment. Primary components of GDSA Framework include PFLOTRAN to simulate the major features, events, and processes (FEPs) over time, Dakota to propagate uncertainty and analyze sensitivities, meshing codes to define the domain, and various other software for rendering properties, processing data, and visualizing results.

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DOE OSTI2018

Calcined Waste Storage at the Idaho Nuclear Technology and Engineering Center

This comprehensive report provides definitive information regarding the volume, mass, and composition (chemical and radioactivity) of the calcined waste stored at the Idaho Nuclear Technology and Engineering Center on the Idaho National Laboratory Site. The calcine composition data are required for current regulatory compliance and safe operation of the INTEC’s six Calcined Solids Storage Facilities (CSSFs), as well as future retrieval, treatment, and shipping the calcine to an off-Site facility (such as a geologic repository). This report also contains a description of the calcine storage bins that compose each of the CSSFs, emphasizing the features that may be of concern to future calcine retrieval. The CSSFs were designed by different architectural engineering firms and built at different times. Each CSSF has a unique design, reflecting varying design criteria and lessons learned from historical CSSF operation. The varying CSSF designs will affect future calcine retrieval processes and equipment. This revision (5) of the report presents refinements of the databases used to calculate the volume, mass, chemical content, and radioactivity of calcined waste produced and stored within the CSSFs. This revision corrects minor errors and inconsistencies noted in an extensive validation and verification effort of the prior version of the Microsoft Excel workbooks upon which the calcine composition described in this report is based. Historical calcine samples are insufficient in number and scope of analysis to fully characterize the entire inventory of calcine in the CSSFs. Sample data exist for all of the liquid wastes that were calcined. This report provides calcine composition data based on liquid waste sample analyses, the volume of liquid waste calcined, calciner operating data, and CSSF operating data using several large Microsoft Excel (Microsoft 2010) databases and spreadsheets. The Excel spreadsheets are collectively called the Historical Processing Model (HPM). The calcine composition determined by this method compares favorably with historical calcine sample data.

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DOE OSTI2013

Report on International Collaboration Involving the FE Heater and HG-A Tests at Mont Terri

Nuclear waste programs outside of the US have focused on different host rock types for geological disposal of high-level radioactive waste. Several countries, including France, Switzerland, Belgium, and Japan are exploring the possibility of waste disposal in shale and other clay-rich rock that fall within the general classification of argillaceous rock. This rock type is also of interest for the US program because the US has extensive sedimentary basins containing large deposits of argillaceous rock. LBNL, as part of the DOE-NE Used Fuel Disposition Campaign, is collaborating on some of the underground research laboratory (URL) activities at the Mont Terri URL near Saint-Ursanne, Switzerland. The Mont Terri project, which began in 1995, has developed a URL at a depth of about 300 m in a stiff clay formation called the Opalinus Clay. Our current collaboration efforts include two test modeling activities for the FE heater test and the HG-A leak-off test. This report documents results concerning our current modeling of these field tests. The overall objectives of these activities include an improved understanding of and advanced relevant modeling capabilities for EDZ evolution in clay repositories and the associated coupled processes, and to develop a technical basis for the maximum allowable temperature for a clay repository.

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