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At least 973 records · Page 54
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Influences of temperature on permeability changes of flow-paths altered by highly alkaline Ca-rich groundwater - 15192

Fractures (or cracks) in host rock around the geological repository are considered as main flow-paths for the migration of radionuclides. Such fractures would be altered by the alkaline components (Ca, Na, K) leaching from cementitious materials for the construction of the repository. While silicate minerals in fractures are dissolved by highly alkaline groundwater, secondary minerals would deposit on the surface of fractures because of the reaction between the alkaline components and silicate species. The authors' previous study showed the generation of CSH-gel and the clogging of micro flow-path under the condition of Ca-rich and high pH by flow experiments. The present study examined the permeability changes with temperature in micro flow-path flowing highly alkaline Ca-rich solution, considering temperature at the depth of geological disposal system. For the simulation of micro flow-path such as fractures, 'micro flow-cell', which nipped a thin Teflon sheet with a slit between granite chip and Teflon basement, were used for the experiments. Highly alkaline Ca-rich solution (pH>12 and 8.5 mM Ca(OH){sub 2}) were injected into micro flow-cell with 2.0 ml/h of flow rate under the given temperature conditions (278 K, 296 K or 313 K). As a result, the permeability became lower with the increase in temperature. In addition, the deposition rates of CSH-gel became larger with the increase in temperature, and the activation energy was estimated to be 32.0 kJ/mol. These results suggest that flow-paths around the geological repository would undergo the clogging with CSH-gel in a shorter time-period under a condition of higher temperature. (authors)

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

MY EXES and EDM for Automated Remediation Data Quality Assessments - 19053

MY Electronic data exchange and Evaluation System (EXES) is a new web-based technology providing external admin access and full control to the Web-based EXES tool developed by GD for EPA for automated environmental remediation data quality assessments. EXES replaces most manual assessment processes, reducing remediation data review time and costs and ensuring that data are of known and documented quality. This tool is available from EPA to federal agencies and remediation firms and has successfully been used to support over 190 site investigations, 105 Remedial investigations and over 20 compliance monitoring efforts. The new MY EXES functionality has been developed for project managers and decision makers to setup projects, track receipt of environmental remediation monitoring data, assess data quality, and manage data. It is designed to be flexible and configurable to meet any government agency's data review and data management needs. It is accessible 24/7 through a secure portal. MY EXES efficiently evaluates analytical data regardless of the underlying method and data reporting requirements. It is a centralized project and data assessment tool that is scalable, flexible and configurable; improving data quality while reducing cost through the following: - Provides users with the ability to create a project and to specify the analytical methods corresponding to the data that will be submitted for that project. A project can be modified and managed anytime during the project's life cycle. - Allows users to manage a library of analytical methods that can be made available for association to their projects. - Users can specify the automated tests and assessment that needs to be performed on the analytical data deliverables based on specific requirements in the project's Quality Assurance Project Plan (QAPP). - Users specify: - The format of the incoming data deliverables from the laboratory; - The type and level of review services to be performed; - The assessment deliverables to be generated - output spreadsheets and reports; - Project owner sets up system access and permissions for members of the project team. - A faster (600%) and less expensive (85%) process than the traditional data usability/validation. The EXES Data Manager application (EDM) provides a centralized data repository to view and retrieve generated reports, spreadsheets, and other data deliverables as defined by the client, eliminating the need for data transfer between project team members. (authors)

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

Deep Borehole Disposal Safety Analysis.

This report presents a preliminary safety analysis for the deep borehole disposal (DBD) concept, using a safety case framework. A safety case is an integrated collection of qualitative and quantitative arguments, evidence, and analyses that substantiate the safety, and the level of confidence in the safety, of a geologic repository. This safety case framework for DBD follows the outline of the elements of a safety case, and identifies the types of information that will be required to satisfy these elements. At this very preliminary phase of development, the DBD safety case focuses on the generic feasibility of the DBD concept. It is based on potential system designs, waste forms, engineering, and geologic conditions; however, no specific site or regulatory framework exists. It will progress to a site-specific safety case as the DBD concept advances into a site-specific phase, progressing through consent-based site selection and site investigation and characterization.

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

Oak Ridge environmental information system (OREIS): formalizing the shapes of things on the Oak Ridge Reservation - 15393

As representations of object and area boundaries on Earth's surface, shapes are important components of geographic information system platforms. The US Department of Energy Office of Environmental Management in Oak Ridge, Tennessee recognizes the importance of shapes as boundary representations and has mandated the Oak Ridge Environmental Information System as the official repository of all geographic information system shapes related to CERCLA cleanup on the Oak Ridge Reservation. (authors)

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

System impacts of accelerated transfer of used nuclear fuel to dry storage - 15220

In the wake of the September 11, 2001 terror attacks, and again after the March 11, 2011, accident at the Fukushima Dai-Ichi station from the Great East Japan Earthquake/tsunami, concern was expressed regarding security of current spent fuel storage practices in the United States. An approach recurrently proposed to reduce the potential for fire and the release of radionuclides in an accident or sabotage scenario is to reduce the loading of the reactor fuel pools, which are frequently maintained near their full capacity at many reactor sites in the U.S. We assess the impact on the waste management system of different scenarios for accelerating the schedule of used nuclear fuel (UNF) transfer from reactor pools to dry storage casks. We consider scenarios where the spent fuel pools are drawn down to 50% or 25% of full capacity over 5- or 10-year schedules. The scenarios include status quo, where UNF continues to remain in dry storage at the reactor sites, as well as scenarios with UNF transfer to interim storage and eventually to a mined geologic repository. The primary purpose is to evaluate how the different scenarios for accelerating the transfer of UNF to at-reactor dry storage would affect the performance of the entire UNF management system. Areas that were considered include an estimate of the additional system costs that could be incurred and where the existing infrastructure at the reactor sites would come under stress from the transfer demands. This study considered only the fuel-handling logistics and rough-order-of-magnitude cost at reactor sites, interim storage facility, and operations to package UNF into disposal canisters. The effects of minimum fuel age that could be transferred and the accelerated transfer starting date are also examined, but found to have relatively little effect. Impacts on safety and worker dose are not evaluated quantitatively, although effects can be inferred from increases in handling operations, etc. from the scenarios considered. (authors)

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DOE OSTI2016

Test Results and Comparison of Triaxial Strength Testing of Waste Isolation Pilot Plant Clean Salt

This memorandum documents laboratory thermomechanical triaxial strength testing of Waste Isolation Pilot Plant (WIPP) clean salt. The limited study completed independent, adjunct laboratory tests in the United States to assist in validating similar testing results being provided by the German facilities. The testing protocol consisted of completing confined triaxial, constant strain rate strength tests of intact WIPP clean salt at temperatures of 25°C and 100°C and at multiple confining pressures. The stratigraphy at WIPP also includes salt that has been labeled “argillaceous.” The much larger test matrix conducted in Germany included both the so-called clean and argillaceous salts. When combined, the total database of laboratory results will be used to develop input parameters for models, assess adequacy of existing models, and predict material behavior. These laboratory studies are also consistent with the goals of the international salt repository research program. The goal of this study was to complete a subset of a test matrix on clean salt from the WIPP undertaken by German research groups. The work was performed at RESPEC in Rapid City, South Dakota. A rigorous Quality Assurance protocol was applied, such that corroboration provides the potential of qualifying all of the test data gathered by German research groups.

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DOE OSTI2016

M4FT-16LL080302052-Update to Thermodynamic Database Development and Sorption Database Integration

This progress report (Level 4 Milestone Number M4FT-16LL080302052) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Argillite Disposal R&D Work Package Number FT-16LL08030205. The focus of this research is the thermodynamic modeling of Engineered Barrier System (EBS) materials and properties and development of thermodynamic databases and models to evaluate the stability of EBS materials and their interactions with fluids at various physico-chemical conditions relevant to subsurface repository environments. The development and implementation of equilibrium thermodynamic models are intended to describe chemical and physical processes such as solubility, sorption, and diffusion.

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

Effect of Iron on Radiolytic Hydrogen Peroxide Generation

This report gives the details on the effect of iron chemistry on H 2 O 2 decomposition under radiolytic condition at the surface of used nuclear fuel under repository conditions. Additionally, suggestions are offered on what further data or measurements would be required for model verification and applicability.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Fuel/Basket Degradation Modeling Summary

This report documents the activities in a preliminary phase of development for three models: 1) waste package breach model, 2) fuel/basket degradation model, and 3) dual-purpose canister (DPC) crush model. The waste package breach model describes the coupling of mass flow, heat transport, and canister shell deformation in response to a heat-generating (criticality) event. The fuel/basket degradation model describes potential weakening and disaggregation of the DPC structure from corrosion, possibly accelerated by seismic ground motion. Progressively degraded three-dimensional (3D) configurations of the fuel, basket, and shell are generated for future analysis of reactivity (with as-loaded DPC fuel characteristics). Another important application for the fuel/basket degradation model is validation of the two stylized degradation cases currently being used by other investigators for analysis of the as-loaded DPC inventory under disposal conditions. The DPC crush model investigates stability of a typical DPC after breach of the disposal overpack allows fluids from the repository near-field environment to penetrate and externally pressurize the canister shell. Preliminary results show that large deformation of the DPC could occur for external pressure on the order of 10 to 15 MPa, or the shell could be stable (not collapse) with pressure of 20 MPa or greater if the basket plates are fully welded at the connections.

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

Investigation of Benefits from U/TRU Recycle - Quantification and Comparison to U/Pu Recycle

The recently completed comprehensive evaluation and screening of nuclear fuel cycle options identified a number of potentially promising fuel cycles for R&D that offer what could be considered by decision-makers as having the potential for significant improvement compared to the current U.S. fuel cycle. The fuel cycles that consistently performed the best were recycle fuel cycles that used self-sustaining fast reactors operating with either U/Pu or U/TRU recycle fuel and also included options where the fast reactors provided fissile materials to support operation of thermal reactors. However, based on the evaluation criteria and metrics used in the study, there was no difference in benefit between recycle of U/Pu and U/TRU (where TRU is plutonium and the minor actinides) while there were differences in the challenges for developing and deploying such fuel cycles, with U/TRU recycle being more challenging. This observation prompted the question as to the desirability of pursuing R&D on U/TRU recycle given that there may not be an increase in benefit. As a result, activities have been pursued to further investigate the performance differences between U/Pu and U/TRU recycle based on considering issues beyond those used in the evaluation and screening study to identify, if possible, areas where there are significant benefits of U/TRU recycle compared to U/Pu recycle. These new considerations focused on several areas, but especially on the impact on disposal of the HLW, which in the case of U/Pu recycle contains all of the minor actinides along with fission products, while in the case of U/TRU recycle only contains the losses of minor actinides from the reprocessing and recycle fuel fabrication operations. This difference in content has several implications. One impact is on the time dependent decay heat which can affect handling and the use of space in a geologic repository. Another impact concerns the HLW form and volume, since presence of minor actinides may adversely affect the ability to reduce HLW volume. The short-term radioactivity and long-term radiotoxicity of the HLW is also affected, which may be of more or less importance depending on the specific geologic disposal environment. To study these potential effects, a range of waste forms and disposal environments were used in the analysis, documenting to what extent the recycle of minor actinides in addition to plutonium may offer further benefit. Another area of investigation concerned the recycle fuel, for the fast reactor and for the thermal reactors they may support. Information to date indicates that U/Pu fuel may be simpler to fabricate and has a much more extensive database than U/TRU fuel, one of the reasons for the increased challenge for developing and deploying a U/TRU fuel cycle, and also indicates that heterogeneous recycle of the minor actinides may be even more difficult as compared to homogeneous recycle. This information was reviewed and updated to reflect the most recent studies for the purpose of informing on all aspects of the differences between U/Pu and U/TRU recycle. The results of all of these investigations will be presented to provide information on the findings concerning the value of U/TRU recycle.

Fast reactor recycle↗
DOE OSTI2019

A Practical Approach of Mining Operational Risks in Cigeo - A field-test - 19349

Andra is currently completing the detailed engineering phase of the stepwise design development of Cigeo (the French HLW and IL-LLW Deep Geological Repository, aka DGR). Various technical issues are at stake, including the risk management requirements linked to the safety of 'Mining Operations' (i.e. occupational safety issues related to underground excavation/dismantling works at some 500 m of depth), at different stages of the DGR life cycle (i.e. from its construction start-up to its final closure and future monitoring/or conversely a decision of waste retrieval). These occupational safety requirements have a significant impact on Cigeo's design, its progressive construction and later closure. On a shorter term, one must demonstrate that the design bases are technically sound and that the solutions selected (to build, operate, close the DGR infrastructures or conversely retrieve the waste) are both feasible and safe. In that sense (and following an identification of operational risks during construction), Andra focused on two sensitive cases: - The construction of large diameter drift crossings, in particular those related to the connection between a transfer drift (main) and a IL-LLW disposal vault (i.e. during construction operations), - The partial dismantling of a drift concrete lining (made of concrete segment rings, erected and placed by a TBM) at time of construction of a horizontal seal (i.e. during closure operations). The discussion starts with a description of the geotechnical conditions prevailing at the (French) Bure site, in the Callovo-Oxfordian (COx) claystone formation. Then the paper presents the Cigeo underground architecture (at large) and the specific structural choices made (at the end of the detailed engineering phase) for drifts, disposal vaults and seals. The discussion on construction - deconstruction challenges for each of the critical issues cited above provides an opportunity to explain the operational sequences (and how the methodology selected enables to reduce the risks). The following part of the paper focuses the technological demonstration (an in-situ test to be planned in detail) of a partial dismantling of the concrete segments rings forming the support lining of two tunnels previously built in the Meuse/Haute-Marne URL (Underground Research Laboratory) infrastructures, combined with the construction of a specific drift crossing. In this dismantling and construction test, the technical demonstration objective is three-fold: - The capacity to lay down compressible arch segments and support (by bolts and mesh) the rock after the lining removal, - The capacity to excavate and line the drift crossing, - The structural / geomechanical behavior (evolution with time) of the rock and lining compared with that of 'neighboring' zones. This paper provides a timeline of the tests. Should the results obtained deliver some positive perspectives of technical feasibility (and convincing operational safety) Andra will incorporate this dismantling/construction solution in the 'design freeze' of the DGR underground infrastructures. The positive outcomes of this technical test campaign will also provide 'elements of proof' for the instruction of the Cigeo licensing (a process likely to take place between 2020 and 2024) by the French Nuclear Safety Authority (ASN). (authors)

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

Full-Scale Retrieval and Transfer System Mockup for Pneumatic Conveyance of High-Level Calcine Waste - 19563

The Idaho National Laboratory reprocessed spent nuclear fuel from 1953 to 1994. The reprocessing used solvent extraction to separate uranium, resulting in a high-level radioactive liquid waste. A calcination process was used to convert the liquid waste to a high-level radioactive granular solid, referred to as calcine. The calcine is stored in stainless-steel bins contained within six concrete vaults called the Calcined Solids Storage Facilities (CSSFs). The Calcine Disposition Project is tasked with demonstrating the ability to retrieve the calcine from the bins of CSSF 1 and transferring it to the bins of CSSF 6. CSSF 1 was constructed to be a final calcine repository, with no forethought of retrieval. It represents the greatest technical hurdle to calcine extraction of all the CSSFs. The challenges inherent with accessing the calcine, safely extracting it, and securely transporting it are copious and complex. To overcome those challenges in a timely and objective manner, a full-scale mockup will be constructed. The mockup includes a full-size model of a concentric group of bins from CSSF 1 and a storage vessel representing CSSF 6. Mockup equipment will include an air compressor, filtration, blower/vacuum, and retrieval devices. In addition to defining process parameters, the mockup testing scope will demonstrate the amount of residual calcine that can be expected to be in the bins following extraction. This will be key to preparing for closure of CSSF 1 following removal of the calcine. Once testing is complete, the mockup will constitute a model of the calcine extraction and transport system and will be prepared to fulfill a second advantageous objective: the operators that will carry out the calcine transfer from CSSF 1 to CSSF 6 will be trained using the mockup. The control system will be identical, the major pieces of equipment will be representative, and the procedures will be analogous; hence, high-fidelity training can take place at the mockup. Once radiological operations commence at CSSF 1, the mock-up's commission will still not be completed. If any problems arise during calcine transport, the mockup will be ready to serve as a simulator to more effectively reproduce dilemmas and deduce solutions. The mock-up's mission spans the duration of the project. Its usefulness begins with helping to identify process criteria. Operations personnel will benefit greatly from its use as a training tool. And finally, the immediate availability of a simulator to reproduce problems encountered during radiological extraction and transfer will be invaluable. The knowledge gained during this process will also be of continued benefit as future work is done to remove various types of calcine from the remaining CSSFs. The methodology of safe and secure pneumatic transfer of radioactive material may be of benefit to others throughout the U.S. Department of Energy complex. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Development of an Integrated System for the Long-term Preservation of Records, Knowledge and Memory relating to Radioactive Waste Disposal Facilities - 17099

The Nuclear Energy Agency (NEA) has co-ordinated an international initiative to develop a common approach to the preservation of records, knowledge and long-term memory (RK and M), focused on radioactive waste disposals. The work programme has run in two phases, starting in 2011, with the current phase of the project due to be completed in 2018. Preparations for final reporting are now underway. The detailed implementation of an approach to RK and M preservation for any disposal facility will always remain a national decision. However, the NEA project has produced a 'menu' of tools and techniques that can be accessed and adapted to suit national needs. The common source of these menu components will help to ensure that the memory of the repository is kept alive, that messages to future populations can be clearly understood, and that evidence derived from the disposal environment can be properly interpreted. This paper describes the vision for implementation of the RK and M initiative within national programmes. Practical testing of some components of the menu of tools and techniques is currently underway in France, Sweden and USA, with encouraging results. Other disposal facilities, both under development or in planning, are invited to 'road test' the products and provide feedback. Refinement and maintenance of the tools remains important, and the NEA is committed to ensuring the continued accessibility and availability of the project outputs into the future. (authors)

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

Solutions for management of defective fuel assemblies - 15318

Defective fuel management is a major challenge for nuclear operators when the time comes for finding a long-term solution for their used (or spent) fuel assemblies. This question of defective fuel management can be raised in several cases: - For optimizing the remaining operating costs of shutdown reactors, the spent fuel pools must be emptied as soon as possible; a solution needs therefore to be found for evacuating all the spent fuel assemblies, defective fuel assemblies and in particular defective fuel rods included; - Some utilities are today facing a saturation of their reactor spent fuel pool; evacuating the defective fuel assemblies and extracted defective fuel rods allows optimizing the space in the pool. Whereas the solutions for managing intact used fuel assemblies are well-known, a specific technical solution must be implemented for managing the defective spent fuel assemblies as well. The purpose of this paper is to describe the existing solutions for the management of all types of defective fuel assemblies. Interim Dry Storage Solution Interim dry storage of defective spent fuel assemblies is being implemented for decades in the USA. Specific operations are performed for preparing the defective fuel assemblies for interim dry storage and they are then stored with intact used fuel assemblies in the dry storage cask. The advantage of such a solution is its short-term cost effectiveness. The paper presents the existing dry interim storage solutions for defective used fuel rods, the way they are implemented, the associated operational experience. Long-term Dry Storage Solution A long-term dry storage solution for defective fuel assemblies is being developed, compliant with direct disposal requirements. The solution is based on the encapsulation of each defective fuel rod into a fuel rod capsule and an associated specific dry storage licensing approach: this is the best available technology regarding long term storage safety requirements. The paper presents this unique capsule technology, the transport and storage casks that can be used for transportation of defective fuel rods to the storage facility, and the associated transport licensing approach currently used in Europe. Treatment Solution. The most definitive way to get rid of defective fuel assemblies is their treatment (or reprocessing) in a treatment plant. There is then no need to assess the behaviour of these specific used fuel assemblies as for an interim storage facility or a final repository. The advantage of such a solution is its sustainable approach. The paper presents the existing transport and treatment solutions, the way they are implemented, the associated operational experience. (authors)

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

MANTA. An Integral Reactor Physics Experiment to Infer the Neutron Capture Cross Sections of Actinides and Fission Products in Fast and Epithermal Spectra

Neutron cross-sections characterize the way neutrons interact with matter. They are essential to most nuclear engineering projects and, even though theoretical progress has been made as far as the predictability of neutron cross-section models, measurements are still indispensable to meet tight design requirements for reduced uncertainties. Within the field of fission reactor technology, one can identify the following specializations that rely on the availability of accurate neutron cross-sections: (1) fission reactor design, (2) nuclear fuel cycles, (3) nuclear safety, (4) nuclear safeguards, (5) reactor monitoring and neutron fluence determination and (6) waste disposal and transmutation. In particular, the assessment of advanced fuel cycles requires an extensive knowledge of transuranics cross sections. Plutonium isotopes, but also americium, curium and up to californium isotope data are required with a small uncertainty in order to optimize significant features of the fuel cycle that have an impact on feasibility studies (e.g. neutron doses at fuel fabrication, decay heat in a repository, etc.). Different techniques are available to determine neutron cross sections experimentally, with the common denominator that a source of neutrons is necessary. It can either come from an accelerator that produces neutrons as a result of interactions between charged particles and a target, or it can come from a nuclear reactor. When the measurements are performed with an accelerator, they are referred to as differential since the analysis of the data provides the cross-sections for different discrete energies, i.e. σ(E i ), and for the diffusion cross sections for different discrete angles. Another approach is to irradiate a very pure sample in a test reactor such as the Advanced Test Reactor (ATR) at INL and, after a given time, determine the amount of the different transmutation products. The precise characterization of the nuclide densities before and after neutron irradiation allows to infer energy-integrated neutron cross sections, i.e. ∫₀ ∞ σ(E)φ(E)dE, where φ(E) is the neutron flux “seen” by the sample. This approach, which is usually defined and led by reactor physicists, is referred to as integral and is the object of this report. These two sources of information, i.e. differential and integral, are complementary and are used by the nuclear physicists in charge of producing the evaluated nuclear data files used by the nuclear community (ENDF, JEFF…). The generation of accurate nuclear data files requires an iterative process involving reactor physicists and nuclear data evaluators. This experimental program has been funded by the ATR National Scientific User Facility (ATR-NSUF) and by the DOE Office of Science in the framework of the Recovery Act. It has been given the name MANTRA for Measurement of Actinides Neutron TRAnsmutation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Evaluation of Engineered Barrier Systems in the Disposition of Spent Nuclear Fuel

This document is a summary of the R&D activities associated with the Engineered Barrier Systems Work Package. Multiple facets of Engineered Barrier Systems (EBS) research were examined in the course of FY18 activities. This report is focused on delivering an update on the status and progress of modelling tools and experimental methods, both of which are essential to understanding and predicting long-term repository performance as part of the safety case. Specifically, the work described herein aims to improve understanding of EBS component evolution and interactions. Utlimately, the EBS Work Package is working towards producing process models for distinct processes that can either be incorporated into performance assessment (PA), or provide critical information for implementing better constraints on barrier performance The main objective of this work is that the models being developed and refined will either be implemented directly into the Generic Disposal System Analysis platform (GDSA), or can otherwise be indirectly linked to the performance assessment by providing improved bounding conditions. In either the case, the expectation is that validated modelling tools will be developed that provide critical input to the safety case. This report covers a range of topics — modelling topics include: thermal-hydrologic-mechanical-chemical coupling (THMC) in buffer materials, comparisons of modelling approaches to optimize computational efficiency, thermal analysis for EBS/repository design, benchmarking of thermal analysis tools, and a preliminary study of buffer re-saturation processes. Experimental work reported, includes: chemical evolution and sorption behavior of clay-based buffer materials and high-pressure, high temperature studies of EBS material interactions. The work leverages international collaborations to ensure that the DOE program is active and abreast of the latest advances in nuclear waste disposal. This includes participation in the HotBENT Field Test, aimed at understanding near-field effects on EBS materials at temperatures above 100 °C, and the analysis of data and characterization of samples from the FEBEX Field Test. Both the FEBEX and HotBENT Field Tests utilize/utilized the Grimsel Test Site in Switzerland, which is situated in a granite host rock. These tests offer the opportunity to understand near field evolution of bentonite buffer at in situ conditions for either a relatively long timescale (18 years for FEBEX) or temperature above 100 °C (HotBENT). Overall, this report provides in depth descriptions of tools and capabilities to investigate nearfield performance of EBS materials (esp. bentonite buffer), as well as tools for drift-scale thermal and thermal-hydrologic analysis critical to EBS and repository design. For a more detailed description of work contained herein, please see Section 10 ("Conclusions") of this document.

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

Panel Session 3: Prospects for Shared Solutions for Solving the WM Challenges for Small and Emerging Nuclear Countries

This panel focused on sharing WM solutions for small and emerging nuclear countries. As countries with substantial inventories of spent nuclear fuel pursue their national repository programs with varying degrees of progress, countries with only a few reactors and those at the beginning stages of their nuclear programs face a far different and substantial challenge toward developing a national solution. This session brought together a panel of experts from government and industry to explore shared solutions and next steps building off the recent progress made by international organizations and the recent efforts and lessons learned. Panelists with presentations: The Rationale for a Multinational Repository (Sean Tyson); Perspectives on the Future of the Multi-National Repository Concept (Alan Brownstein); Why Would any Country Accept Foreign Spent Nuclear Fuel? (Charles McCombie); Dual Track Strategy for Small Countries (Leon Kegel)

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