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International Collaboration Activities in Geologic Disposal Research: FY2021 Progress

This report describes the FY21 status of international collaboration regarding geologic disposal research in the Spent Fuel and Waste Disposition (SFWD) Campaign. Since 2012, in an effort coordinated by Lawrence Berkeley National Laboratory, SFWD has advanced active collaboration with several international geologic disposal programs across the world. Such collaboration has allowed the SFWD Campaign to benefit from a deep knowledge base in regards to alternative repository environments developed over decades and has provided a framework for active peer-to-peer research participation in international groups which conduct, analyze, and model performance-relevant processes. Via international collaboration, the SFWD Campaign also benefits from substantial international investments in research facilities (such as underground research laboratory testing and modeling) and achieves cost savings via joint funding of expensive field experiments. To date, SFWD’s International Disposal R&D Program has established formal collaboration agreements with multiple international initiatives and various international partners, and national lab scientists associated with SFWD have conducted a large number of specific collaborative R&D activities that align well with its R&D priorities.

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2020 LANL Contributions to the BATS Test in WIPP

The DOE Office of Nuclear Energy (DOE-NE) repository research and development program seeks to provide a sound technical basis for multiple viable disposal options, increase confidence in the robustness of generic disposal concepts, and develop science and engineering tools needed to support disposal concept implementation. Sandia, Los Alamos, and Lawrence Berkeley National labs are conducting research into salt, which includes the Brine Availability Test in Salt (BATS) field test at the Waste Isolation Pilot Plant (WIPP), a DOE Office of Environmental Management facility. BATS is leveraging the existing infrastructure associated with the WIPP to advance science informing generic disposal concepts. BATS is an ongoing heated borehole experiment being conducted in the WIPP underground. The goal of this experiment is to reduce the uncertainty associated with spent nuclear fuel disposition in geologic salt formations. The BATS experiments is designed to accomplish this by increasing our understanding of brine migration in salt, assessing damaged zones from mining and drilling, simulating a post-closure environment, confirming salt properties, and providing data for model validation. Phased testing for generic salt repository R&D was proposed in 2015. Subsequently, borehole thermal testing was developed as a first step in this process. This document summarizes the work conducted on the BATS experiment during fiscal year 2020 with an emphasis on the contributions by Los Alamos National Lab.

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SNF Interim Storage Canister Corrosion and Surface Environment Investigations (FY22 Status Update)

High-level purpose of this work: This report summarizes work carried out by Sandia National Laboratories (SNL) in the fiscal year 2022 (FY22) to evaluate the potential occurrence of stress corrosion cracking (SCC) on spent nuclear fuel (SNF) dry storage canisters. The U.S. currently lacks a repository for permanent disposal of SNF; thus, dry storage systems will be in use for much longer time periods than originally intended. Gap analyses by the US Department of Energy (DOE), the Nuclear Regulatory Commission (NRC), the Nuclear Waste Technical Review Board (NWTRB), and the Electric Power Research Institute (EPRI) have all determined that an improved understanding of the occurrence and risk of canister SCC is critical to demonstrating the safety of long-term dry storage. Should canister penetration by SCC occur, the containment boundary represented by the canister would be breached. A loss of the inert environment (helium) within the canister could occur and intrusion of air and moisture could react with and damage the fuel within the canister. For this reason, the DOE is funding an effort to evaluate the potential occurrence and consequences of dry storage canister SCC and to develop prevention, mitigation, and repair technologies for this degradation mechanism.

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Lessons Learned From Public Outreach: Effective Incorporation of the Stakeholder Voice - 18447

The licensing process for any facility associated with nuclear power or aspects of the nuclear fuel cycle can be expected to encounter intense stakeholder interest, and often significant challenges. Over the last decade, staff of the Center for Nuclear Waste Regulatory Analyses (CNWRA{sup R}) at Southwest Research Institute{sup R} (SwRI{sup R}) provided comprehensive support to the U.S. Nuclear Regulatory Commission (NRC) for public outreach efforts across a wide variety of nuclear activities. Many of these projects included public involvement and consultation activities for compliance with the National Environmental Policy Act (NEPA), including scoping, public meetings, responding to public comments on draft documents, and holding small meetings with affected local, state, and tribal governments. In addition, we have supported NRC public outreach related to nuclear facility licensing activities (including the proposed repository at Yucca Mountain), rulemaking, and regulatory guidance development. This experience has offered us a broad view of the challenges faced by industry, regulators, and stakeholders. In the process, we have observed some best practices that lead to effective stakeholder communication. Communication about complex technical issues is challenging, particularly for societally controversial topics, such as civilian use of nuclear power and management of radioactive waste. Most commonly, we have observed a lack of mutual trust and respect among industry, regulators, and stakeholders that creates an atmosphere ripe for suspicion and misunderstanding. Stakeholders have, in public comments, criticized the lack of appropriate venues to raise concerns in which they feel they can get meaningful responses from the nuclear industry and regulators. Conversely, industry faces difficulty in reaching compromise with stakeholders, who sometimes misinterpret or misrepresent factual information. Within this contentious environment, regulators are constrained by their statutory framework, even as they are challenged to be more responsive, transparent, and fair to both stakeholders and industry. In spite of these obstacles, there are examples where feedback from stakeholders was heard and utilized. Because of successful collection of appropriate feedback, regulators became more adept at understanding the full range of perspectives from both industry and stakeholders, and subsequently did a better job at effectively incorporating feedback and communicating their decisions and findings in a clear and understandable way. Ultimately, the success of these examples was based on a measure of mutual respect, which was reached through effective communication and implementation of some best practices. These best practices include (i) outreach planning that is abundantly mindful of the various stakeholder points of view, site history, regional characteristics, and probable future and related actions, combined with advance identification of key stakeholders; (ii) development and distribution of information that facilitates public participation; and (iii) use of avenues by which stakeholders can provide meaningful and effective comments. Illustrative examples include (i) how concerns raised by stakeholders were heard by an applicant and resulted in a revised project proposal; (ii) use of a variety of interaction formats that incorporated input across a broad spectrum of stakeholders; (iii) utilizing stakeholder feedback received during scoping; and (iv) how a facility underestimated the impact of 'outsider' stakeholder voices that were in contrast to the known local community. By moving past the viewpoint that public outreach is a 'necessary evil' and engaging only to check a procedural box, each participant can benefit from mutually satisfactory and improved outcomes. (author)

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Ni and Cr addition to alloy waste forms to reduce radionuclide environmental releases

Reference alloy waste forms (RAW) were fabricated and underwent hybrid corrosion/immersion testing to parameterize the ANL analytical oxidative-dissolution model to enable the calculation of fractional release rates and to determine the effectiveness of Ni and Cr trim additions in reducing release rates of radionuclide surrogates. Figure 1 shows the prototypical multiphase microstructure of the alloys with each phase type contributing about equally to the exposed surface area. The waste forms tested at SRNL were variations of the RAW-6 formulation that uses HT9 as the main alloy component, and are meant to enable evaluation of the impact of Ni and Cr trim additions on the release rates of actinides and Tc-99. The test solutions were deaerated alkaline and acidic brines, ranging in pH 3 to pH 10, representing potential repositories with those conditions. The testing approach consisted of 4 major steps; 1) bare surface corrosion measurements at pH values of 3, 5, 8, and 10, 2) hybrid potentiostatic hold/exposure measurements at pH 3, 3) measurement of radionuclide concentrations and relations to anodic current from potentiostatic holds, and 4) identification of corroding phases using SEM/EDS of electrodes.

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Developments in the Regulatory Framework of Interim Storage of Spent Nuclear Fuel (SNF) in Germany - 18137

According to article 6 of the German Atomic Energy Act, storage of nuclear fuel requires a license granted by the competent authority in this field, which is the Federal Office for Safety of Nuclear Waste Management (BfE). In 1995, a license for storage of spent nuclear fuel (SNF) and vitrified high level waste (HAW) was granted for the interim storage facility in Gorleben. Dry Interim storage at the NPP sites in its current form started in 2002 in the interim storage facility next to the nuclear power plant NPP Lingen. Since this time twelve NPPs erected their own storage facilities additionally to the three central storage facilities (Ahaus, Gorleben and Rubenow). In all of these facilities the concept of dry storage in metallic dual purpose casks (DPCs) is implemented. The actual storage licenses for SNF and the HAW (which is stored in Gorleben and Rubenow) are limited to 40 years after closing the transport and storage cask. The first license to expire will be the storage license for the interim storage facility in Gorleben in the year 2034. In Germany the decision about a repository site will be taken in the year 2031. To enable the utilities to hand over the SNF and HAW to the repository regime while the current licenses are still valid would require a very fast process licensing and construction of the required facilities at this location. As a delay cannot be excluded, necessary preparations have to be taken on the side of the regulator. In the German Atomic Energy Act precautions have already been taken and consulting the German parliament (Bundestag) is mandatory if inevitable events would require an extension of the currently licensed interim storage period. The new structure with the Federal Office BfE, and two companies for final disposal 'Bundesgesellschaft fuer Endlagerung' (BGE mbH) and interim storage 'Bundesgesellschaft fuer Zwischenlagerung' (BGZmbH) was established in 2017. The Federal Office for Safety of Nuclear Waste Management (BfE) is responsible for regulation in the field of nuclear disposal since 30.07.2016. So the package design approval for DPCs and the storage licenses for SNF will be granted by this office. Furthermore, it is the regulator for the repository and will finally grant a license for the German deep geological repository. Site selection and construction and operation of the repository will be the responsibility of BGE. On January 1, 2018, the BGZ will take over all remaining interim storage facilities operating under a license due to article six of the Atomic Energy Act. (authors)

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Geomechanical in situ testing of fault reactivation in argillite repositories

Abstract. Pressurization of natural faults as a result of repository-induced effects can lead to their reactivation and permeability generation in case such features are present near disposal tunnels. Potential driving forces for such pressurization are the temperature increase caused by heat-producing high-level radioactive waste and the generation of hydrogen and other gases due to corrosion of engineered materials. We are primarily concerned about pressurization and fault activation in host rocks and faults that have very low natural permeability for pore pressure increases to dissipate, such as the argillite rocks currently investigated in Switzerland, France and other countries. This presentation discusses a series of in situ experiments of fault activation by fluid injection conducted in the argillite rock (Opalinus clay) at the Mont Terri underground research laboratory in Switzerland. A multi-model monitoring test bed was installed at Mont Terri that includes distributed fiber optics for strain, temperature, and acoustics; local fault pore pressure and three-dimensional displacement sensors; active seismic imaging; and passive seismic monitoring. The fault experiments (and their subsequent analysis via hydromechanical modeling) provide a new fundamental understanding of the coupling between pore pressure, fault deformation and permeability generation as a function of time and allow exploring how seismic and aseismic events may impact the integrity of faulted argillite host rock. Our experimental observations demonstrate that significant flow (and transport) can occur along the initially impermeable argillite fault when rupture is activated. However, the fault permeability decreases to almost its pre-activation value when fluid injection ceases and fluid pressure drops. Dilatant slip on the fault plane alone does not explain the observed increase in fault permeability; pressure-induced fault opening also plays a role, favored by the softness of the shale along with the fact that the structure of the fault zone prevents fluids from diffusing into the adjacent damage zone. Rupture initiation and permeability generation is initially aseismic, associated only with an increase in noise level and emerging tremors. Micro-earthquakes are initiated later in the experiments and typically occur away from the fluid-pressurized area. Hydromechanical models show that stress transferred from the initial aseismic deformation can build up to stress criticality and later induce seismic rupture. After presenting the experimental results, we will close the presentation with an outlook to future experimental campaigns using the fault test bed at Mont Terri. We are currently planning a controlled thermal stimulation of the fault, where instead of fluid injection as a trigger mechanism we will heat up the nearby rock volume and measure potential effects on fault stability. Lessons learned from our past and future experiments will help inform the safety assessment of geologic disposal in argillite host rock.

Birkholzer, Jens T.↗

Thermal Performance Analysis of a Geologic Borehole Repository

The Brazilian Nuclear Research Institute (IPEN) proposed a design for the disposal of Disused Sealed Radioactive Sources (DSRS) based on the IAEA Borehole Disposal of Sealed Radioactive Sources (BOSS) design that would allow the entirety of Brazil’s inventory of DSRS to be disposed in a single borehole. The proposed IPEN design allows for 170 waste packages (WPs) containing DSRS (such as Co-60 and Cs-137) to be stacked on top of each other inside the borehole. The primary objective of this work was to evaluate the thermal performance of a conservative approach to the IPEN proposal with the equivalent of two WPs and two different inside configurations using Co-60 as the radioactive heat source. The current WP configuration (heterogeneous) for the IPEN proposal has 60% of the WP volume being occupied by a nuclear radioactive heat source and the remaining 40% as vacant space. The second configuration (homogeneous) considered for this project was a homogeneous case where 100% of the WP volume was occupied by a nuclear radioactive heat source. The computational models for the thermal analyses of the WP configurations with the Co-60 heat source considered three different cooling mechanisms (conduction, radiation, and convection) and the effect of mesh size on the results from the thermal analysis. The results of the analyses yielded maximum temperatures inside the WPs for both of the WP configurations and various mesh sizes. The heterogeneous WP considered the cooling mechanisms of conduction, convection, and radiation. The temperature results from the heterogeneous WP analysis suggest that the model is cooled predominantly by conduction with effect of radiation and natural convection on cooling being negligible. From the thermal analysis comparing the two WP configurations, the results suggest that either WP configuration could be used for the design. The mesh sensitivity results verify the meshes used, and results obtained from the thermal analyses were close to being independent of mesh size. The results from the computational case and analytically-calculated case for the homogeneous WP in benchmarking were almost identical, which indicates that the computational approach used here was successfully verified by the analytical solution.

07 ISOTOPE AND RADIATION SOURCES↗

Overview of System Integration Analysis Activities for Integrated Waste Management

Spent nuclear fuel (SNF) generated by the current fleet of commercial nuclear reactors is being stored at the reactor sites in spent fuel pools (SFPs) and in dry independent spent fuel storage installations (ISFSIs). The US Department of Energy Office of Nuclear Energy (DOE-NE) is developing an Integrated Waste Management Program (IWMP) comprising a suite of options and supporting analyses to enable future informed choices. The IWMP is organized into the following five major areas: 1) Consent-Based Siting, 2) IWM Facilities and Equipment Concepts and Development, 3) Transportation Capability Analysis and Support, 4) Information Technology Solutions and Support, and 5) System Integration Analysis and Support. This paper discusses the activities ongoing in the IWMP System Integration Analysis and Support area. Two main areas of research in system integration are data and tools development, as well as system analysis assessments. One of the tools being developed in the system integration area is the Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS) tool. It is being developed as a foundational resource for DOE-NE to manage SNF data, along with several compatible analysis tools for time-dependent characterization of SNF and related systems. UNF-ST&DARDS has the unparalleled ability to track SNF through the entire back end of the fuel cycle—from the time the fuel is discharged from a reactor through its disposal in a geological repository. UNF ST&DARDS interfaces with the SCALE code system for nuclear analysis and COBRA-SFS for thermal analysis. Another main tool being developed is the Next Generation System Analysis Model (NGSAM). NGSAM is an agent-based simulation software tool expressly designed to be capable of modeling the waste management system. NGSAM has been developed to enable informed decision-making by providing the capability of analyzing various potential system options for the management of SNF and HLW. Using NGSAM, system architecture analyses are being conducted to support the future deployment of a comprehensive nuclear waste management system that considers all major back-end aspects of the nuclear fuel cycle (i.e., transportation, storage, and disposal). System analysis assessments may investigate the implications of various strategies such as different acceptance rates, acceptance queues, facility capacities and options, standardized canisters, and different assumed system operation start dates. Recently, some system analysis effort has begun to look at how the waste management system might operate for advanced reactor fuel cycles.

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A comparative analysis of numerical approaches for the description of gas flow in clay-based repository systems: From a laboratory to a large-scale gas injection test

There is nowadays a consensus among many countries that geological disposal is a favourable solution for the long-term management. Although different host formations and different barrier systems are under consideration around the world, clay-based materials form an important component for waste isolation in most national programmes. Hence, a good comprehension of the effect of gas flow on the hydro-mechanical behaviour of clay-based soils is essential, both at laboratory and field scale. Task B under the international cooperative project DECOVALEX-2023 has recently shown that, after some enhancement, models can be employed to reproduce laboratory scale tests, even with different sample geometries 37 . However, further work is required to understand whether they can be applied to simulate a large-scale experiment. Up-scaling of models for the advective transport of gas through clay-based low permeable material presents a number of problems related to the difficulty in obtaining consistent hydrogeological parameters and constitutive relationships at both laboratory and field scale. Based on a unique dataset from a large-scale gas injection test (Lasgit) performed at the Äspö Hard Rock Laboratory (Sweden), Task B within DECOVALEX-2023 has explored the refinement of these numerical strategies applied to the simulation of gas flow. Work performed within the task reveals that codes do not need to be substantially modified from the laboratory models to reproduce full-scale tests: indeed, model parameters calibrated and validated at laboratory scale have been applied to predict field scale gas flow at Lasgit, including peak gas pressure and injected cumulative gas volume. By means of (1) the introduction of interfaces between blocks to reflect the experimental configuration and the (2) adjustment of some parameters (e.g., higher permeability), the updated models are able to represent most of the key features observed in the experimental data, even at a large scale.

Tamayo-Mas, E↗

Materials for Consideration in Standardized Canister Design Activities.

This document identifies materials and material mitigation processes that might be used in new designs for standardized canisters for storage, transportation, and disposal of spent nuclear fuel. It also addresses potential corrosion issues with existing dual-purpose canisters (DPCs) that could be addressed in new canister designs. The major potential corrosion risk during storage is stress corrosion cracking of the weld regions on the 304 SS/316 SS canister shell due to deliquescence of chloride salts on the surface. Two approaches are proposed to alleviate this potential risk. First, the existing canister materials (304 and 316 SS) could be used, but the welds mitigated to relieve residual stresses and/or sensitization. Alternatively, more corrosion-resistant steels such as super-austenitic or duplex stainless steels, could be used. Experimental testing is needed to verify that these alternatives would successfully reduce the risk of stress corrosion cracking during fuel storage. For disposal in a geologic repository, the canister will be enclosed in a corrosion-resistant or corrosion-allowance overpack that will provide barrier capability and mechanical strength. The canister shell will no longer have a barrier function and its containment integrity can be ignored. The basket and neutron absorbers within the canister have the important role of limiting the possibility of post-closure criticality. The time period for corrosion is much longer in the post-closure period, and one major unanswered question is whether the basket materials will corrode slowly enough to maintain structural integrity for at least 10,000 years. Whereas there is extensive literature on stainless steels, this evaluation recommends testing of 304 and 316 SS, and more corrosion-resistant steels such as super-austenitic, duplex, and super-duplex stainless steels, at repository-relevant physical and chemical conditions. Both general and localized corrosion testing methods would be used to establish corrosion rates and component lifetimes. Finally, it is unlikely that the aluminum-based neutron absorber materials that are commonly used in existing DPCs would survive for 10,000 years in disposal environments, because the aluminum will act as a sacrificial anode for the steel. We recommend additional testing of borated and Gd-bearing stainless steels, to establish general and localized corrosion resistance in repository-relevant environmental conditions.

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Alteration Behaviors of Granite Minerals Forming Calcium Silicate Hydrate under Ca-Rich Highly Alkaline Conditions - 16160

Cementitious materials used for geological disposal systems leach large amounts of alkaline ions (Ca{sup 2+}, Na{sup +}, K{sup +}) into groundwater, and alter the pH of groundwater by up to 13. The altered groundwater contacts the host rock and forms Calcium Silicate-Hydrate (CSH) through the interaction of Ca ions and the silicate species, which in turn clogs the flow paths of the host rock. This clogging effect in the flow paths is expected to retard the migration of radionuclides. A previous study reported the heterogeneous formation of CSH on the surface of granite under Carich and highly alkaline conditions. To clarify the formation process of CSH, this study examined the CSH formation with the dissolution of some minerals contained in the granite under Ca-rich and highly alkaline conditions. In the experiments, the granite block and the crushed minerals contained in the granite were immersed in a 6.4-8.5 mM Ca(OH){sub 2} solution adjusted pH to 12.5 by adding NaOH solution. As a result, biotite and anorthite supplied more Si to the formation of CSH than other minerals. Furthermore, most of the formed CSH precipitated on the solid surface although a little CSH remained in liquid phase as supernatant. These results suggest that under Ca-rich highly alkaline conditions the CSH formed as a secondary mineral will retard the migration of radionuclides released from the repository because of the clogging effect of the flow path. (authors)

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Panel Session 109: US HLW Repository: Status and Next Steps

This panel focused on the technical, institutional and broader political issues associated with advancing a US HLW Repository program including programmatic, regulatory, legislative and funding challenges. Panelist with presentations: WM 2020 HLW Repository Status and Next Steps (Leo Blundo)

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Preliminary Study on Recycling and Reuse of Metals from Decommissioning in Construction of Disposal Vaults in Slovakia - 17217

The decommissioning of nuclear installations represents a complex process resulting in the generation of large amounts of various waste materials. Majority of them are non-contaminated, however some of them may be hazardous as well (e.g. chemicals, asbestos, oil products, etc.). During decommissioning of NPP, significant quantities of materials with various levels of radionuclides concentration arise as well. All of these materials should be managed appropriately keeping in mind safety principles, legislative requirements, lessons learned and other relevant factors. Therefore, selection of an appropriate strategy for management of the materials arising from decommissioning is crucial and strongly influences the overall effectiveness of decommissioning process. In line with international incentives for optimization of radioactive material management, concepts of recycling and reuse of slightly radioactive materials are widely discussed and applications of these concepts are analyzed. Therefore, the paper is devoted to the preliminary study of potential for recycling of re-melted metals from decommissioning in form of reinforcement bars or mesh used in the construction of ferro-concrete disposal vaults within the near surface repository. The paper presents the preliminary results and discusses the technical and non-technical aspects related to the recycling scenario. Moreover, comparison of various scenarios related to the scrap metal clearance and recycling within nuclear industry is included. Key findings and lessons learned from study may be useful for the international community dealing with waste management optimization and in particular with recycling of materials within nuclear industry. (authors)

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Developing generic design alternatives for dry storage of used nuclear fuel at an interim storage facility - 15480

The Department of Energy (DOE) is beginning to lay the groundwork for implementing interim storage of used nuclear fuel (UNF) as recommended by the Blue Ribbon Commission on America's Nuclear Future in their report to the Secretary of Energy published on January 26, 2012. These plans include activities to 1) establish one or more Interim Storage Facilities (ISF) using consent-based siting, and 2) prepare for large-scale transport of UNF. The Administration released its Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste (Strategy) on January 11, 2013. The Strategy includes a phased, adaptive, and consent-based approach to siting and implementing a comprehensive management and disposal system. The Strategy calls for an operational Pilot ISF by 2021, a Larger ISF by 2025, and a geological repository by 2048. The DOE Nuclear Fuels Storage and Transportation Planning Project (NFST) prepared a report in April 2013 titled, -A Project Concept for Nuclear Fuels Storage and Transportation.- This report developed the functions and requirements necessary for development of a Pilot and Larger ISF to meet the framework outlined in the Strategy as follows: - The UNF inventory is limited for the Pilot and Larger ISF as described in the Strategy. - It is assumed that bare fuel will be received in reusable transportation casks in significant quantities for the larger ISF and will not be solely limited to the receipt of fuel in dry storage canister (DSC) as is the Pilot ISF. This assumption was made because the method for accepting UNF from the utilities once the larger ISF begins operating has not yet been decided. - UNF storage alternative concepts identified in previous reports must be more fully developed to support alternative analysis. The purpose of this topic is to report on the development of the generic design alternatives that have been evaluated for the receipt and storage of the UNF at the initial Pilot ISF and Larger ISF. The generic design alternatives have been developed over the past few months by collaboration between the DOE and industry experts experienced in UNF facilities. The design alternatives will enable volunteer host communities to investigate generic design alternatives that could be fully developed for a site-specific location based on their unique design requirements. The design alternatives provide flexibility for additional capabilities to be developed and executed in a modular fashion such that expansion from a Pilot ISF to a Larger ISF could be achieved in an orderly manner. (authors)

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Analysis of DWPF Sludge Batch 6 (Macrobatch 7) Pour Stream Glass Samples

The Defense Waste Processing Facility (DWPF) began processing Sludge Batch 6 (SB6), also referred to as Macrobatch 7 (MB7), in June 2010. SB6 is a blend of the heel of Tank 40 from Sludge Batch 5 (SB5), H-Canyon Np transfers and SB6 that was transferred to Tank 40 from Tank 51. SB6 was processed using Frit 418. During processing of each sludge batch, the DWPF is required to take at least one glass sample to meet the objectives of the Glass Product Control Program and to complete the necessary Production Records so that the final glass product may be disposed of at a Federal Repository. Four pour stream glass samples and two Melter Feed Tank (MFT) slurry samples were collected while processing SB6. The samples were transferred to the Savannah River National Laboratory (SRNL) where they were analyzed.

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Thermo-Hydro-Mechanical Modeling of Brine Migration in a Heated Borehole Test in Bedded Salt

Abstract This research paper focuses on the thermo-hydro-mechanical (THM) modeling of brine migration in a heated borehole test conducted as part of the ongoing Brine Availability Test in Salt (BATS) at the Waste Isolation Pilot Plant (WIPP) in New Mexico. It is a component of the international collaboration project DECOVALEX-2023 (DEvelopment of COupled models and their VALidation against EXperiments), which aims to understand the THM processes governing brine flow in heated rock salt repositories through collaborative analysis by multiple research teams. Using the TOUGH–FLAC simulator, THM simulations were performed and compared with data from the BATS phase 1a. This experiment involved two identical horizontal-borehole arrays, one heated and one serving as a control, both equipped with sensor arrays. Analysis of measurements revealed water flow rate surges during heater power transitions, with the highest jump observed during cooling. Acoustic emission activity exhibited distinct patterns in response to heater power changes, suggesting that damage to rock salt is particularly pronounced during the cooling phase. The THM simulations successfully captured these phenomena, highlighting the significance of thermal effects, brine migration, and mechanical behavior in predicting brine availability in heated and damaged rock salt. Our modeling also revealed the critical interplay between heating and cooling-induced damage and its influence on flow properties, particularly affecting brine inflow estimation. Notably, we found that cooling-induced brine inflow spikes result from increased permeability due to tensile dilatancy. These findings have important implications for the development of robust containment strategies and enhance our understanding of the complex processes involved in repository performance.

58 GEOSCIENCES↗