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DOE OSTI2012

Application of Analytical Heat Transfer Models of Multi-layered Natural and Engineered Barriers in Potential High-Level Nuclear Waste Repositories - 12435

A combination of transient heat transfer analytical solutions for a finite line source, a series of point sources, and a series of parallel infinite line sources were combined with a quasi-steady-state multi-layered cylindrical solution to simulate the temperature response of a deep geologic radioactive waste repository with multi-layered natural and engineered barriers. This evaluation was performed to provide information to scientists and decision makers to compare candidate geologic media for a repository (crystalline rock [granite], clay, salt, and deep borehole), and to provide input for the future evaluation of the trade-off between pre-emplacement surface storage time, waste package size, and repository footprint. This approach was selected in favor of the finite element solution typically used to analyze the temperature response because it allowed rapid comparison of a large number of alternative disposal options and design configurations. More than 100 combinations of waste form, geologic environment, repository design configuration, and surface storage times were analyzed and compared. The analytical solution approach used to analyze the repository temperature response allowed rapid comparison of a large number of alternative disposal options and design configurations. More than 100 combinations of waste form, geologic environment, repository design configuration, and surface storage times were analyzed and compared. This approach allowed investigation of the sensitivity of the results to combinations of parameters that show that there is much flexibility to be gained in terms of spent fuel management options by varying a few key parameters. This initial analysis used representative design concepts and thermal constraints based on international design concepts, and it also included waste forms representing future fuel cycles with high burnup fuels. Unlike repository designs with large open tunnels and pre-closure ventilation, all of the disposal concepts analyzed in this study used enclosed emplacement modes, where the waste packages were in direct contact with encapsulating engineered or natural materials. The deep borehole repository concept limits the size of the SNF waste packages and may require rod consolidation to fit within the drill casing diameter. A single assembly waste package, assuming rod consolidation, was evaluated in the current analysis. Similar size restrictions apply for the HLW canisters. At this time no thermal constraints have been defined for the deep borehole repository concept. Representative EBS materials and properties were evaluated. However, changes in EBS design concepts and materials can also have significant effects on the maximum waste package surface temperature. Increased thermal conductivity of the buffer layer can be achieved by using an engineered buffer consisting of a mixture of graphite, sand, and bentonite [14]. One of the advantages of the analytical model is that it highlights the sensitivity of the results to the parameters that define the repository layout, including spacing between axial and lateral neighboring waste packages and drifts. It is clear that repository layout adjustments can be made to reduce the calculated peak temperatures. The results also show that significant reductions in required surface storage times can be achieved if higher thermal constraints can be justified Additional studies are planned to evaluate the trade-offs between surface storage times, repository layout parameters, and variations in EBS design concepts. Model validation and uncertainties will also be addressed. It is expected that shorter surface storage times and more optimized repository design configurations may be achieved. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Processes in Salt Repositories for Radioactive Waste Disposal

This document summarizes the key processes (thermal, hydrological, mechanical, and chemical; THMC) impacting the features of a deep geological repository for radioactive waste in salt. Some processes are natural and on-going whether the repository is there or not, and other processes are driven by the perturbation associated with the repository. The features considered here include both engineered and natural components of the repository system. The engineered barrier system (EBS) in a salt repository is quite different from those implemented for a repository in clay or crystalline rocks, because it is comprised mostly of granular salt and salt-compatible cements, rather than bentonite. When compared to other rocks (i.e., silicates), salt has unique properties that make it an excellent potential host rock. Openings and fractures in salt creep closed readily. Salt has high thermal conductivity, which can reduce peak temperatures. Additionally, far away from the excavations the porosity of salt is unconnected, which leads to essentially zero advective or diffusive transport. The small amount of hypersaline brine occurring in salt minimizes microbial activity, reduces colloid-assisted transport, and eliminates in-package criticality (i.e., chloride is a neutron poison). At the end of the report, we present a brief outline for a potential salt repository, including considerations avoided in previous repository disposal concepts. We propose considering higher-temperature processes in future disposal concepts, rather than trying to minimize the thermal perturbation of the repository. Since hot salt is drier, a dry repository would limit corrosion, gas generation, and solute transport. Openings and fractures creep shut faster in hot salt. Therefore, higher temperatures could be seen as beneficial, rather than something to minimize, through increased spacing between waste packages (increasing repository costs).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

The Challenges of Repository Operation in an Existing Mine - 19339

A generally accepted principle related to long term management of radioactive waste/ spent fuel is the isolation in a geological repository. In Germany, the disposal operation of low-level waste (LLW) and intermediate-level waste (ILW) started in the Asse II mine in 1967 and in the former German Democratic Republic in the Morsleben repository in 1971. After the German reunification, the responsibility for the existing sites changed several times until the Bundesgesellschaft fuer Endlagerung (BGE) became new license holder for both sites in 2017. The BGE is responsible to carry out the operative tasks of searching for a high-level waste (HLW) repository, the planning and construction of the Konrad repository for the emplacement of low- and intermediate-level waste (LILW), the operation of the Asse II mine including the retrieval of the emplaced waste, and the preparation for the closure of the Morsleben repository. Nevertheless, the challenges of repository operation in an existing mine have not changed in all these years, as the regulatory framework is mainly written for Nuclear Power Plants (NPPs) and is difficult to adapt for a repository mine. In particular, conditions in a mine are substantially different from 'clean' above-ground nuclear installations. Dust due to mining activities, relatively high temperatures, and a corrosive atmosphere in a salt mine might be present. Therefore, the usability of the repository for several decades during the different operational phases, i.e. planning, construction, emplacement, and closure, can only be ensured by continuous maintenance. Furthermore, the atomic law is applicable for above-ground facilities and the mining law for mines of every description. The difficulty of geological repositories is to ensure that both laws are going to be applied above and below ground with no regulatory framework tailored especially for such sites. A special case is the Asse II mine where the mining law was applied from the very beginning, before the atomic law needed to be established later on and now requires the retrieval of the radioactive waste before decommissioning. In this presentation, similarities and differences in the legal framework of the repositories in the responsibility of BGE and some operational experiences throughout the years are described. This applies to the radiation protection regime, the excavation ratio, the host rock stability and the backfilling concepts of the mining areas. Further challenges arise from the demonstration of long term safety according to the state-of-the-art of science and technology for either existing repositories/ mines or those under construction or planned. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Evaluation of Bentonite Engineered Barrier Performance under Repository Conditions

Nuclear energy has been an important source of power throughout the world for several decades, but with that power comes substantial amounts of waste. The USA alone is projected to produce 77,100 MTHM (34,700 MCi) of commercial spent fuel by the year 2020. There has been a growing sense of urgency to determine a method of disposal for all types of nuclear waste. Underground storage has been proposed as the safest means of long term storage and disposal by both the Academy of Sciences and the Blue Ribbon Commission with a focus on deep geologic disposal. Because deep geological repositories (DGR) pose as the safest method of disposal due to the isolation of the waste and the low-permeability of the surrounding material, it is crucial to understand the science behind these repositories. Many scientists have concern with DGRs because of all the unanswered scientific questions. With a global inventory of 300,000 MTHM, it is imperative to understand how DGRs will retain the nuclear waste over a time span of up to a projected 1 million years. Because of the amount of nuclear waste increases each year, there is an inherent need to study the viability of a repository in order to begin implementation of that repository. Based on the half-lives of each radionuclide in nuclear waste, after approximately 100 years of storage, the heat production from the decay of radionuclides becomes dominated by the long-lived actinides. Because the decay of the nuclear waste results in an elevated temperature within the canister, That mixed with water intrusion can corrode the canister and allow radionuclides to leach into the surrounding environment. In an effort to retain these radionuclides within the repositories, most all deep geologic repository designs incorporate an engineered multi-barrier approach. Once the radionuclides begin to leach out of the canisters, the next barrier to retain them is a layer of compacted clay such as a smectite. A mechanistic understanding of mobility at elevated temperatures is crucial in order to evaluate the performance of the engineered clay barrier system within a deep geological repository. Because the clay barrier is compacted so tightly, the only method of mobility for radionuclides is through molecular diffusion. Assessing the factors that affect the rates at which various radionuclides will diffuse through these clays is important in the performance assessment of the repository system as a whole. Neptunium-237 has been identified as a major actinide of concern for long term storage, dominating the radioactivity emitted from the waste between 10,000 to 100,000 years of storage. Under oxidizing conditions, this radionuclide exists mainly as Np(V). In this oxidation state, the neptunium exists chemically as the NpO{sub 2}{sup +} ion which exhibits enhanced mobility characteristics such as high solubility and low affinity for sorption to surfaces. Because of this, {sup 237}Np(V) has been identified as the most mobile radionuclide in nuclear waste and thus a major radionuclide of concern, motivating research into the mobility of {sup 237}Np(V) through the compacted clay barrier. Bentonite is one of the most common backfill materials used in repositories. This clayey rock's main constituent is the mineral Na-montmorillonite [(Ca{sub 0.06}Na{sub 0.14}K{sub 0.05})(Al{sub 1.57}Fe{sup 3+} {sub 0.17}Mg{sub 0.25})(Al{sub 0.04}Si{sub 3.96}O{sub 10} - (OH){sub 2})] which gives this material the advantageous properties required for a repository such as a high swelling capacity. Because of this high affinity for absorbing water, montmorillonite is the ideal candidate for backfill material in a deep geological repository. Adsorption of radionuclides onto the surface of the compacted clay is a major mechanism for retardation of mobility. Because diffusion and adsorption are integrated into this process, it is important to characterize Np adsorption onto the surface of compacted montmorillonite. Previous studies have quantified the adsorption of Np onto montmorillonite at variable ionic strengths and temperatures. Furthermore, several studies have focused on the diffusion of Np through compacted clay, but no other studies have determined diffusion coefficients at elevated temperatures of up to 80 deg. C. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Integrating repositories with fuel cycles: The airport authority model

The organization of the fuel cycle is a legacy of World War II and the cold war. Fuel cycle facilities were developed and deployed without consideration of the waste management implications. This led to the fuel cycle model of a geological repository site with a single owner, a single function (disposal), and no other facilities on site. Recent studies indicate large economic, safety, repository performance, nonproliferation, and institutional incentives to collocate and integrate all back-end facilities. Site functions could include geological disposal of spent nuclear fuel (SNF) with the option for future retrievability, disposal of other wastes, reprocessing with fuel fabrication, radioisotope production, other facilities that generate significant radioactive wastes, SNF inspection (navy and commercial), and related services such as SNF safeguards equipment testing and training. This implies a site with multiple facilities with different owners sharing some facilities and using common facilities - the repository and SNF receiving. This requires a different repository site institutional structure. We propose development of repository site authorities modeled after airport authorities. Airport authorities manage airports with government-owned runways, collocated or shared public and private airline terminals, commercial and federal military facilities, aircraft maintenance bases, and related operations - all enabled and benefiting the high-value runway asset and access to it via taxi ways. With a repository site authority the high value asset is the repository. The SNF and HLW receiving and storage facilities (equivalent to the airport terminal) serve the repository, any future reprocessing plants, and others with needs for access to SNF and other wastes. Non-public special-built roadways and on-site rail lines (equivalent to taxi ways) connect facilities. Airport authorities are typically chartered by state governments and managed by commissions with members appointed by the state governor, county governments, and city governments. This structure (1) enables state and local governments to work together to maximize job and tax benefits to local communities and the state, (2) provides a mechanism to address local concerns such as airport noise, and (3) creates an institutional structure with large incentives to maximize the value of the common asset, the runway. A repository site authority would have a similar structure and be the local interface to any national waste management authority. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Performance Assessment of a Generic Nuclear Waste Repository in Shale - 18392

A performance assessment (PA) is an important component of a comprehensive safety analysis for a nuclear waste repository. In a PA, probabilistic simulations of the total repository system are performed, and results are evaluated against performance metrics. Uncertainty and sensitivity analyses help prioritize additional research and model development. The United States Department of Energy (DOE) has been developing a state of the art PA simulation software tool-kit, Geologic Disposal Safety Assessment (GDSA) Framework, that couples increasingly higher fidelity models of subsystem processes into total system PA simulations. Over the past several years, PAs of generic repositories in three geologic media (salt, shale, and crystalline rock) have demonstrated ongoing developments in capability. The current PA of a nuclear waste repository in a generic shale formation showcases GDSA Framework, including capabilities in domain discretization (Cubit), multi-physics simulations (PFLOTRAN), uncertainty and sensitivity analysis (Dakota), and visualization (Paraview). The generic shale reference case considers the disposal of 22,000 metric tons heavy metal of commercial spent nuclear fuel (SNF) in a generic shale formation. PA simulations account for the thermal load and radionuclide inventory of the waste form, components of the engineered barrier system including waste package and buffer, and components of the natural barrier system including the host rock shale and underlying and overlying stratigraphic units. Two repository layouts are considered, one for emplacement of waste packages containing 12 pressurized water reactor (PWR) assemblies, and one for 4-PWR waste package emplacement. Model domains are half-symmetry and contain between 7 and 22 million grid cells. Grid refinement captures the detail of individual waste packages, emplacement drifts, access drifts, and shafts. Simulations are run in a high-performance computing (HPC) environment on 512 or 2048 processes. The governing equations describing coupled heat and fluid flow and reactive transport are solved with PFLOTRAN, an open-source, massively parallel multiphase flow and reactive transport code. Additional simulated processes include waste package degradation; waste form dissolution; radioactive decay and ingrowth in aqueous, solid, and adsorbed phases; and a simple biosphere based on a pumping well. Simulations are run to 106 y; dose and radionuclide concentrations are observed within aquifers at a point approximately 5 km downgradient of the repository. Dakota is used to sample likely ranges of input parameters including waste form and waste package degradation rates and properties of engineered and natural materials to quantify uncertainty in predicted concentrations and sensitivity to input parameters. Given the assumptions of the reference case, results of 12-PWR and 4-PWR simulations are very similar. Dose at the pumping well 5-km downgradient of the repository does not exceed 10-9 Sv/y. I-129 is the largest contributor to dose. The dose due to Cl-36, the next largest contributor, is orders of magnitude smaller. At the pumping well, the concentration of I-129 5-km downgradient does not exceed 10-14 mol/L. Rank correlation coefficients indicate that the maximum concentration of I-129 within the aquifer is sensitive to shale (repository host rock) porosity and aquifer permeability. These results affirm that HPC-capable codes can be used to simulate important multi-physics couplings directly in a total system performance assessment of a geologic repository, and can be used in prioritization of future research and model development. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Review of the Safety Analyses of the Konrad Repository - 20530

The Bundesgesellschaft fuer Endlagerung (BGE) is the license holder of all geological repository sites in Germany since 2017. Besides the construction of the Konrad repository for the emplacement of low- and intermediate-level waste (LILW), the operation of the Asse II mine including the retrieval of the emplaced waste, and the preparation for the closure of the Morsleben repository, BGE is responsible to carry out the operative tasks of selecting a site for a high-level waste (HLW) repository. The Konrad repository is Germany's first repository for radioactive waste with negligible heat generation, which was fully licensed under atomic law within the scope of a plan approval procedure from the original application in 1982 until the license was granted in 2002 and finally confirmed by the Federal Administrative Court in 2007. The emplacement of LILW was licensed up to a volume of 303,000 m{sup 3} with a total β/γ-activity of 5 x 10{sup 18} Bq and a total α-activity of 1.5 x 10{sup 17} Bq. Currently, the former iron ore mine Konrad is being converted to a repository and is expected to come into operation in 2027. But what if there is no safety case in place nor a regulatory framework tailored for a change of documents in order to ensure the safety according to the state-of-the-art of science and technology during the evolution of a long-term project such as a geological repository? In this presentation, a review procedure of the safety analyses, that have been the basis of the plan approval decision for the Konrad repository at that time, will be described. Furthermore, the first results of this graduated review procedure for updating the safety analyses will be presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Status of DOE-Managed Waste Repository Safety Assessment Information Needs

The U.S. DOE evaluated the need for disposing of only DOE-Managed high-level-waste (HLW, which includes DOE-managed spent nuclear fuel [SNF] in this report) in a geologic repository by using the Nuclear Waste Policy Act 1982 (NWPA, as amended) and stated (DOE, 2015) “A geologic repository for permanent disposal of Defense HLW could be sited, licensed, constructed, and operated more quickly than a Common NWPA Repository and would provide valuable experience to reduce the cost of a future repository and the time needed to develop it.” Based in part on that report, the President of the United States issued a memorandum stating that such a repository was required (see website with content at https://www.whitehouse.gov/the-press-office/2015/03/24/presidential-memorandum-disposal-defensehigh-level-radioactive-waste-se). This furthered the recommendation from DOE (2014) to “…begin implementation of a phased, adaptive, and consent-based strategy with development of a separate mined repository for some DOE-managed HLW and cooler DOE-managed SNF, potentially including some portion of the inventory of naval SNF.” The work in this report summarizes the status of activities within the Used Fuel Disposition Campaign (UFDC) to begin assessing the information needs for evaluating the safety of such a repository, considering both the waste forms being disposed and the repository concepts being considered. This report is milestone M3FT-15SN0824011 performed under the UFDC work package FT-15SN082401 to provide a status of the progress made towards this objective.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

RANGERS: Modeling Report on Integrity and Performance Assessment of Engineered Barrier Systems in a Salt Repository for HLW/SNF

The Engineered Barrier System (EBS) plays an important role in ensuring the long-term safety and containment of high-level waste (HLW) and spent nuclear fuel (SNF) in deep geological repositories in salt formation. As part of a multi-barrier system, the EBS works alongside the natural barrier, which is the salt formation itself and the technical barrier comprising the disposal casks. The primary function of the EBS is to maintain containment during a defined period until the backfill used in the repository made of crushed salt, develops its sealing capacity through compaction. Over the time, the backfill eventually compacts to a state of low porosity and permeability, acting as a long-term seal. However, until this process is complete, the EBS must retain its structural and functional integrity. Regulatory guidelines in Germany currently require the EBS to remain effective for up to next ice age, that is expected in 50,000 years. The significant hydro-geological and topographic changes expected during an ice age could make it impossible to accurately predict the hydro-chemical conditions within the repository system at that time. In response to these challenges, BGE TECHNOLOGY GmbH (BGE TEC) and Sandia National Laboratories (SNL) have jointly developed a comprehensive methodology for the design and safety assessment of engineered barrier systems within the scope of the RANGERS project. This methodology is tailored for repositories in salt formations. The developed methodology provides a structured approach for designing and assessing the performance of the EBS in salt-based repositories. It begins with defining a sealing concept based on the geological characteristics of the selected site and the overall repository design. The entire repository system, comprising the geological site, repository infrastructure, and EBS, is then subjected to a Features, Events, and Processes (FEP) analysis, focusing solely on those FEPs that affect the EBS. The derived FEPs help identify the loads and stresses acting on the EBS, which serve as the foundation for conducting an integrity assessment. This analysis helps predict the EBS’s evolution and performance over the regulatory time frame, feeding into integrated performance assessment simulations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Engineering for Operation of a Future Belgian Deep Geological Repository for ILW and HLW - 12379

In Belgium, an advanced conceptual design is being elaborated for deep geologic disposal of high level waste (HLW) and for low and intermediate level waste (LILW) not amenable for surface disposal. The concept is based on a shielded steel and concrete container for disposal of HLW, i.e., the Super-container. LILW will be disposed of in separately designed concrete caissons. The reference host rock is the Boom Clay, a poorly indurated clay formation in northeastern Belgium. Investigations into the potential host rock are conducted at the HADES underground research laboratory in Mol, Belgium. In 2009 the Belgian Agency for Management of Radioactive Waste and Enriched Fissile Materials (ONDRAF/NIRAS) initiated a four year research project aimed at confirming the fundamental feasibility of building and operating a repository. The goal of the program is to demonstrate at a detailed conceptual level that the proposed geologic disposal system can be safely constructed, operated, and progressively closed. Part of the broader research efforts being conducted includes evaluations optimization of the waste transportation shaft, subsurface transportation system, ventilation system, and evaluation of backfilling and sealing concepts for the repository design. The potential for implementation of a waste retrieval strategy encompassing the first 100 years after emplacement is also considered. In the framework of a four year research program aimed at confirming the fundamental feasibility of building and operating a repository in poorly indurated clay design studies have been underway to optimize the waste transportation shaft, subsurface transportation system, and ventilation system. Additionally backfilling and sealing concepts proposed for the potential repository have been reviewed in conjunction with impacts related to the potential future inclusion of a retrievability requirement in governing regulations. The main engineering challenges in the Belgian repository concept are size limitations on the underground facilities imposed by the mechanical behavior of the candidate host rock type (i.e., poorly indurated clay) and the resulting ground support requirements. Underground excavations in the Boom Clay require a significant level of ground support to ensure the openings remain stable. A concrete lining system has been developed to address this engineering requirement. As a result strict size limits are imposed on both the diameter of the tunnels and the dimensions of the shaft stations resulting in unique design challenges requiring maximal optimization of the available space. Ongoing studies indicate that a significant (20%) reduction in shaft diameter can be achieved by diagonally orienting the hoist guide rails with respect to the cage, optimizing the counter weight dimensions, and reconfiguring the auxiliary hoisting system as a single rope system. Reliable subsurface transportation of waste packages can be achieved through a hybrid rail/wheel system powered by a battery operated electric locomotive. Key components of the system, including the battery-powered locomotive and a turntable used for transitioning waste shipments from the access gallery into disposal galleries without the need for constructing turnouts, have been successfully demonstrated at the Gorleben exploratory facility and the Konrad repository in Germany, respectively. By optimizing the available space in the disposal galleries and limiting the introduction of hazardous gases by using electric powered systems combined with the relatively small number of workers envisioned in the Belgian repository concept adequate ventilation can be achieved to ensure safe operational conditions. The proposed sealing and backfill systems in the Belgian repository concept should provide adequate safeguards as currently planned. Should a future retrievability requirement be imposed on the design it appears likely that a partial backfilling strategy could be employed. The key component in ensuring retrievability in the design would be the selection of a backfill that combines the safety functions required in the current design concept with a material strength amenable to removal by hydro-jet excavation techniques. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Logistics Modeling of Emplacement Rate and Duration of Operations for Generic Geologic Repository Concepts - 16529

This study (1) identifies 16 potential geologic repository concepts for disposal of commercial spent nuclear fuel (SNF) and (2) evaluates the achievable repository waste emplacement rate and time needed to complete disposal for these concepts. Total repository capacity is assumed to be 140,000 MTU of spent fuel. The results provide important input to rough-order-of-magnitude (ROM) cost estimates for SNF disposal alternatives. The disposal concepts cover three categories of host geologic media: crystalline, salt, and argillaceous rock. Four waste package sizes are considered: 4PWR/9BWR; 12PWR/21BWR; 21PWR/44BWR, and dual purpose canisters (DPCs) packaged for disposal. Each concept is associated with package thermal power limits for emplacement and repository closure (the same for some concepts). To estimate the repository emplacement rate and duration of operations, logistical simulations were performed using assumptions on interim SNF storage, disposition of SNF stored in DPCs, repository opening date, and throughput rates for storage and disposal facilities. The simulations demonstrate that all concepts and scenarios, with few exceptions, are similar with regard to achievable emplacement rate and duration of operations. Throughput of 3,000 MTU per year can be maintained during the first 44 years of repository operations. Most of the inventory (99%) can be emplaced within 54 years of repository operations (the remaining 1% could require additional aging or smaller packages). Estimating ROM disposal cost was not an objective of this analysis, however, an example of disposal cost is provided to demonstrate key differences between concepts. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Safety Concept and Safety Assessment for a HLW Repository in Bedded Salt in Germany - 18253

The German site selection act for a repository for vitrified HLW and SNF demands the consideration of various host rock formations for deep geological disposal. One of the eligible and presently investigated host rock formations is bedded rock salt. Within the framework of the national R and D project KOSINA the scientific bases for the design and safety assessment of a repository for HLW and SNF in bedded salt formations were provided. The research findings presented in this paper include the formulation of a safety concept and safety demonstration concept as well as exemplary radiological consequences analysis for a generic repository in bedded rock salt formations. The safety concept describes in a verbal, argument-based way how the natural conditions, processes and technical measures at large yield a status of safety. Fundamental principle of the safety concept is the concentration and containment of the radioactive and other pollutants in the waste in the containment providing rock zone (CRZ), which ensures isolation from the biosphere. Based on the principle of safe containment inside the CRZ guiding principles, design objectives and strategic measures are derived. The safety demonstration concept describes the means, such as analyses and arguments, which are used in the safety case to demonstrate the safety of a repository system in bedded salt on the basis of the safety concept and its guiding principles and design objectives. The decisive elements are: the demonstration of integrity of the geological barrier, the demonstration of integrity of the geotechnical barriers, the scenario analysis, and the evaluation of the identified scenarios. These elements are complemented in the safety demonstration by concepts on how to consider criticality, non-radiological protection goals and operational safety. The radiological consequences analysis primarily consists of the calculation of the radiological safety indicator for different test cases and scenarios for the various repository design and waste emplacement options. Within the KOSINA project the disposal concepts of drift disposal, horizontal and vertical borehole disposals were investigated in detail by means of computational flow and transport modelling of the repository mine (near field) and assessment of the radiological consequences (biosphere model). The applied software package RepoTREND is introduced; the modelling concept and employed radiological indicators are explained. Examples of the modelling results for the reference cases of the drift disposal repository concept for bedded salt will be given, demonstrating the capabilities of the model and the applicability to close-to-reality repository systems for deep geological disposal. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Panel Session 62: Progress on Deep Repository Programs Around the World

This panel session focused on the progress of deep repository programs, worldwide. While efforts on a deep geological repository for SNF/UNF and HLW in the USA were terminated in 2010, several other repository programs continue to make considerable progress and deserve to be recognized. Panelists were invited from national, multi-national, and international repository programs and organizations to report upon the status and future as well as the perceived reasons underlying the success and the failure of components of their respective repository program. This panel provided a forum for interested parties to present, exchange, and take advantage of, as appropriate, repository progress, challenges, concepts/designs and other lessons learned around the world, including site characterization, site selection, research and development, licensing, construction, operation, and public acceptance and participation. Presentations proposed include: 1 - Cigeo (French Deep Geological Repository Project) update 2018 (Daniel Delort, Andra (France)); 2 - Progress on Deep Geological Repository Programs Around the World (Bruce McKirdy, Radioactive Waste Management (United Kingdom)); 3 - Geological disposal facility in Belgium (Didier Leonard, ONDRAF/NIRAS (Belgium))

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Proceedings of the 10th US/German Workshop on Salt Repository Research Design and Operation. Rev. 1

The 10th US/German Workshop on Salt Repository Research, Design, and Operation was hosted by RESPEC and South Dakota School of Mines & Technology, both located in the Black Hills of South Dakota. Over 60 registered participants representing Germany, United States, the Netherlands, and the United Kingdom availed themselves to excellent facilities on the School of Mines campus. As the 10th annual workshop, this occasion is a milestone of the modern era of collaboration between the US and Germany, which has extended to other countries with potential for radioactive waste disposal in salt. Thrust areas covered in the annual workshops are selected by participants and typically include elements of continuity from year to year. This year's themes included siting, modeling challenges, seal systems and materials, operational safety, and special topics. Two major breakout sessions addressed test sample conditioning and natural closure of salt openings. When the new-generation US/German workshops were conceived, one goal was to identify challenging issues related to salt repository sciences and then conduct open discussions in special breakout sessions. These timely sessions achieve the workshop paradigm and provide in-depth dialogue on important salt repository considerations. In many respects, general themes of these salt repository workshops reflect advances in the scientific basis for nuclear-waste disposal in salt formations and develop naturally as a consequence of unremitting attention largely as a result of the workshop commitment. Technical capabilities in the laboratory and field continue to improve in concert with accumulating experience. Because we are working together, mechanical deformation at the micro-scale can be interpreted at a large scale, which is fundamental to predictive modeling of salt repository evolution. This document records the Proceedings of the 2019 gathering of salt repository nations and has been compiled by RESPEC, demonstrating an adopted protocol in which the host organization creates that year's Proceedings. To assist with organization and compilation, individual chapters are summarized by volunteer subject-matter experts from the participatory audience; these contributors are recognized in the Acknowledgements. All formal presentations are included in this document, thus providing a resource for referencing and many excellent photographic images. Appendices include the agenda, list of participants, abstracts, and presentations. A primary purpose for recording workshop activities is to create and sustain an accessible record of salt repository research. These archives also illustrate transparent development of research areas each year.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

A Prototype Performance Assessment Model for Generic Deep Borehole Repository for High-Level Nuclear Waste - 12132

A deep borehole repository is one of the four geologic disposal system options currently under study by the U.S. DOE to support the development of a long-term strategy for geologic disposal of commercial used nuclear fuel (UNF) and high-level radioactive waste (HLW). The immediate goal of the generic deep borehole repository study is to develop the necessary modeling tools to evaluate and improve the understanding of the repository system response and processes relevant to long-term disposal of UNF and HLW in a deep borehole. A prototype performance assessment model for a generic deep borehole repository has been developed using the approach for a mined geological repository. The preliminary results from the simplified deep borehole generic repository performance assessment indicate that soluble, non-sorbing (or weakly sorbing) fission product radionuclides, such as I-129, Se-79 and Cl-36, are the likely major dose contributors, and that the annual radiation doses to hypothetical future humans associated with those releases may be extremely small. While much work needs to be done to validate the model assumptions and parameters, these preliminary results highlight the importance of a robust seal design in assuring long-term isolation, and suggest that deep boreholes may be a viable alternative to mined repositories for disposal of both HLW and UNF. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Salt Disposal Investigations to Study Thermally Hot Radioactive Waste In A Deep Geologic Repository in Bedded Rock Salt - 12488

A research program is proposed to investigate the behavior of salt when subjected to thermal loads like those that would be present in a high-level waste repository. This research would build upon results of decades of previous salt repository program efforts in the US and Germany and the successful licensing and operation of a repository in salt for disposal of defense transuranic waste. The proposal includes a combination of laboratory-scale investigations, numerical simulations conducted to develop validated models that could be used for future repository design and safety case development, and a thermal field test in an underground salt formation with a configuration that replicates a small portion of a conceptual repository design. Laboratory tests are proposed to measure salt and brine properties across and beyond the range of possible repository conditions. Coupled numerical models will seek to describe phenomenology (thermal, mechanical, and hydrological) observed in the laboratory tests. Finally, the field test will investigate many phenomena that have been variously cited as potential issues for disposal of thermally hot waste in salt, including buoyancy effects and migration of pre-existing trapped brine up the thermal gradient (including vapor phase migration). These studies are proposed to be coordinated and managed by the Carlsbad Field Office of DOE, which is also responsible for the operation of the Waste Isolation Pilot Plant (WIPP) within the Office of Environmental Management. The field test portion of the proposed research would be conducted in experimental areas of the WIPP underground, far from disposal operations. It is believed that such tests may be accomplished using the existing infrastructure of the WIPP repository at a lower cost than if such research were conducted at a commercial salt mine at another location. The phased field test is proposed to be performed over almost a decade, including instrumentation development, several years of measurements during heating and then subsequent cooling periods, and the eventual forensic mining back of the test bed to determine the multi-year behavior of the simulated waste/rock environment. Funding possibilities are described, and prospects for near term start-up are discussed. Mining of the access drifts required to create the test area in the WIPP underground began in November 2011. Because this mining uses existing WIPP infrastructure and labor, it is estimated to take about two years to complete the access drifts. WIPP disposal operations and facility maintenance activities will take priority over the SDI field test area mining. Funding of the SDI proposal was still being considered by DOE's Offices of Environmental Management and Nuclear Energy at the time this paper was written, so no specific estimates of the progress in 2012 have been included. (authors)

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DOE OSTI2015

The role of Aspo Hard Rock laboratory for the Swedish nuclear fuel and waste management's research, developing and demonstration programme for a final repository of spent nuclear fuel - 15378

The Aspo Hard Rock Laboratory (HRL), located in the Simpevarp area in the municipality of Oskarshamn in Sweden, Figure 1, constitutes an important part of the Swedish Nuclear Fuel and Waste Management (SKB) work with design and construction of a deep geological repository for final disposal of spent nuclear fuel. One of the fundamental reasons behind SKB's decision to construct an underground laboratory was to create an opportunity for research, development and demonstration in a realistic and undisturbed rock environment down to repository depth. Most of the research has concerned with processes of importance for the long-term safety of a future final repository and the capability to model the processes taking place at repository depth. Demonstration addresses the performance of the engineered barriers and practical means of constructing a repository and emplacing the canisters containing spent fuel. SKB will in the coming years more focus on using the Aspo HRL for demonstrating technology for and function of important parts of the repository system before constructing and operate the spent nuclear fuel repository in Forsmark. Aspo HRL includes both a research village with office space for 100 persons and different laboratories, e.g. Chemistry and Material Science Laboratory and Bentonite Laboratory, and an underground research laboratory to a depth of 460 m. The Aspo HRL is also open for other organisations to be used for their own projects and experiments not only related to nuclear waste management research. (authors)

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DOE OSTI2018

Elaborating a First Version of a KIF For The Manche Surface Repository in France - 18060

The concept of a Key Information File developed in the framework of the RK and M international initiative has been tested against the case of the Manche surface repository. Work is still in progress. The aim is to issue a final version before end of year 2018, with the file for the decennial safety assessment update. The KIF will be part of the regulatory required Summary Memory File. Testing the KIF brought several lessons, on the elaboration process of such document, on the adaptation of the KIF concept to a surface repository, on the necessary evolution of the document over time. The Manche repository is the first repository operated by Andra, and the one at its most advanced stage. Therefore, it is de facto a pilot for similar issues encountered on other Andra's repositories. The difficulties encountered, the errors made and analyzed are the basis for improved provisions in other cases. It has been the case for the repository design, when the second surface repository was implemented in the Aube district. It is now the case for the cover design, for monitoring and for memory provisions. (authors)

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