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At least 649 records · Page 36

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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

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)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

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.↗

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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

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↗

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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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↗

Geochemical Modeling Using OLI Systems Mixed Solvent Electrolyte Model in Support of the Waste Isolation Pilot Plant

Aqueous electrolyte thermodynamic models were developed for synthetic Salado (GWB) and Castile (ERDA-6) brines to inform the experimental design and to facilitate laboratory data interpretation in geochemical studies supporting disposal in the Waste Isolation Pilot Plant (WIPP). The thermodynamic models, which include the relevant vapor-liquid-solid equilibria for the oxidation-reduction conditions of interest, enable computation of system parameters such as activity-based pH, E h [vs. standard hydrogen electrode (SHE)], ionic strength, total dissolved solids, and solution density; the chemical speciation of all phases; saturation tendencies for thermodynamically feasible solids; and changes in brine chemistry because of the addition of H 2 (g), argon, MgO, Fe(0), and other waste form components. The impacts of oxidation-reduction potential (ORP) and pH on aqueous speciation and dominant solid phases were also assessed with the aid of real-solution Pourbaix diagrams for key elements of interest. In addition to the OLI modeling results, limited measurement data for pH and ORP standards are provided, analyzed, and discussed.

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A Cable Car Transfer System for the Cigeo Access Ramps - An Original Approach - 17035

Andra is currently implementing the detailed engineering phase of the step-wise design of Cigeo (the French HLW and IL-LLW Deep Geological Repository). Various technical issues are at stake, including the definition and development of a key component: a funicular ('cable-car') system dedicated to the transfer (down the access ramp and leading to the underground infrastructures) of the shielding casks housing the waste disposal containers (packages). The present paper starts with a brief description of the Cigeo architecture (i.e. the surface utilities, the service and nuclear ramps, the various shafts and finally the underground facilities). A focus is then made on the 'nuclear ramp' dedicated to the transfer of the waste containers lodged inside the casks (those latter weighing some 100 to 130 metric tons each) and its main geometrical characteristics (slope, length, diameter, structure). Then a discussion follows to detail the rationale having led to the choice of a somehow 'sophisticated' funicular transfer system, at the expense of what could be called 'more conventional' and more straightforward transport solutions (like a transfer system by trucks or by rack and pinion trains). A systematic comparison is finally made between the various solutions, explaining and rating the pros and cons of each concept, via a multi-criteria analysis. This technical choice (and the related justification approach) was submitted to an Experts review at the end of the Preliminary Design phase of Cigeo and comforted by their strong support before the design freeze of this solution was integrated in Cigeo. The following part of the paper is then focused on a description of the main technical components and the mechanical performance of the cable-car system, its current state of design and the set of qualification tests (of key safety components) to be implemented before its effective construction. This article concludes with a timeline of the qualification tests and developments to come. The positive outcomes of this future technical test campaign will also pave the way for the future instruction of the Cigeo licensing file, scheduled as of mid-2018. (authors)

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Fracture Toughness Evaluation for Spent Nuclear Fuel Clad Systems Using Spiral Notch Torsion Fracture Toughness Test

Many radioactive materials within the nuclear fuel cycle present a significant hazard. Such materials include spent nuclear fuel, high-level waste, legacy waste and other nuclear materials. These materials are often held in long-term storage as an interim stage within their lifecycle, including reuse or disposal. Safety and security of spent nuclear fuel (SNF) interim storage installations and follow-on SNF transportations to its final repository sites are very important, due to a great concentration of fission products, actinides and activation products. Fracture mechanics approach in applying to SNF system reliability investigation, especially for the high burn up (HBU) SNF, during SNF long term dry storage or SNF transportation is necessary due to the inherited flaws and inhomogeneity structures existed in a SNF system from the reactor operation. For instance, such as hydride and oxide formation, surface flaw induced by SNF assembly contact interactions, or internal flaws induced by pellet-clad mechanical interaction (PCMI). However, none of the existing fracture toughness data deal with fuel cladding specific geometry or SNF material conditions, such as cladding structure with the segment fuel pellets configuration and the PCMI mechanism. Consequently, the application of existing fracture toughness data to SNF system reliability investigation can be error prone; thus, the development of an industry-wide consensus fracture mechanics approach is needed. The objective of this research is to develop an in-situ fracture mechanics testing protocol, including the associated analytical procedure, that is suitable for evaluating the SNF fracture toughness. Fracture toughness data was obtained under quasi-static fracture loading using Oak Ridge National Laboratory (ORNL) developed Spiral Notch Torsion Test (SNTT) technology carried out on a biaxial tension/torsion tester. These data will be used to support SNF reliability investigation during long-tern SNF dry storage stage or in the follow-on SNF system transport to the final repository sites. SNTT has been a recent breakthrough in measuring fracture toughness for different materials, including metals, ceramics, concrete and polymers composites. Due to its high geometry constraint and unique loading condition, SNTT can be used to measure the fracture toughness with smaller specimens without concern of size effects. The application of SNTT to brittle materials has been proved to be successful. The micro-cracks induced by original notches in brittle materials could ensure crack growth in SNTT samples. Therefore, no fatigue pre cracks are needed for brittle materials specimens. The application of SNTT to the ductile material to generate valid toughness data will require a test sample with sufficient pre-crack length to increase the sample localized constraint at crack front. Fatigue pre-crack growth techniques with the associated compliance function estimated was developed for estimating the crack penetration depth to monitor the fatigue crack growth evolution. In order to extend SNTT approach to a thin shell cladding structure material, a new testing protocol and the associated analytical procedure for estimating the SNF fracture toughness was developed. The detailed SNTT approach and its estimated fracture toughness for the Zr-4 cladding structure with segment alumina inserts are presented in this report. The SNTT test results indicate that SNTT method is a reliable test approach with good repeatability in applying to Zr-4 cladding material. For the medium and the short crack length, the estimate J Q upon fracture for the baseline Zr-4 cladding is at 285.7 lb/in with 2-sigma uncertainty of 18.59 lb/in., and the associated K Q is at 61.4 Ksi√in. For a long crack length, the crack initialization orientation is apparently deviated from the principle tensile stress contour – 45° spiral crack front (i.e., Mode I + Mode III failure mechanism); and the estimated J Q is at 108 lb/in., the associated K Q is at 37.7 Ksi√in. The estimate Mode I J Q (along the 45° pitch principle tensile stress contour) is at 200 lb/in.; which indicate a significant reduction in the fracture toughness in a tubing structure under mixed mode loading condition, Mode I + Mode III, compared to that of Mode I (tensile stress dominated) alone.

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Sol–gel synthesis of iodosodalite precursors and subsequent consolidation with a glass binder made from oxides and sol–gel routes

Radioiodine accumulates in aqueous solutions and off-gas streams during nuclear fuel reprocessing due to its solubility in aqueous solutions and volatility, respectively. Additionally, radioiodine is highly mobile in geological environments and iodine-129 has been found to be one of the largest long-term contributors to chronic dose to humans in scenarios of radionuclide migration from geological repositories. Most of the radioiodine can be captured during fuel reprocessing in off-gas streams using solid sorbents and scrubbing solutions. Once iodine is captured, it must be stored in a durable form for eventual disposal. Iodosodalite has been investigated as a waste form for radioiodine, however these synthesis processes typically result in mixed products and iodine volatilization during consolidation. Therefore, this paper proposes a novel approach to synthesizing iodosodalite utilizing a sol-gel method. This method was chosen to lower processing temperatures and improve product yield. Preliminary experiments conducted to determine the viability of this synthetic method are presented. Additionally, consolidation of sol-gel derived iodosodalite with a glass-binder was explored using three different methods: 1) incorporating the glass-binder during gel preparation using alkoxide precursors; 2) separately preparing the glass binder using a sol-gel method; and 3) separately preparing the glass-binder using a melt-quench technique. Glass-bonded iodosodalite was successfully synthesized using these novel sol-gel approaches.

36 MATERIALS SCIENCE↗

Steel Corrosion Impacts on Radionuclide Immobilization

This progress report (Level 4 Milestone Number M4SF-21LL010302052) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Crystalline Activity Number SF-21LL01030205. The research is focused on actinide and radionuclide sequestration in steel corrosion products. Fuel matrix degradation models suggest that the near field is likely to be reducing at the time of canister breaching, steel corrosion, and radionuclide release, but more oxidizing conditions may prevail in the far field. The incorporation of radionuclides into corrosion phases may limit the rate of radionuclide release by sequestering a portion of the radionuclide source term. For these reasons there is a need to evaluate the incorporation of Pu and other radionuclides into various Fe-oxide phases, and to understand the behavior of coprecipitated phases during mineral recrystallization processes and during re-oxidation events. Radionuclide coprecipitation with Fe minerals may impact long-term repository performance and is an ongoing research focus at Lawrence Livermore National Laboratory.

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La Hague STE2 Sludge Retrieval and Conditioning Strategy - 19229

About 9000 m{sup 2} of Long-Lived Intermediate Level Waste LL-ILW STE2 sludges have been generated by the reprocessing operations of UNGG and later UOx fuel in La Hague (UP2-400) and subsequent liquid effluent treatment. They were placed into a set of adjoining tanks, called the STE2 silos. Today, the STE2 Sludges have to be retrieved and conditioned for final disposal in the deep geological repository under development in France (Cigeo). To this end, since the early 2000's, different retrieval and conditioning scenarios have been evaluated. This paper tracks the history of the technical and non-technical key factors having led to the choice of the baseline scenario developed until the end of 2016, describes the technical achievements and makes a state-of the art of the current scenario consisting in retrieving and temporarily storing STE2 sludges and, meanwhile, studying, researching and testing final conditioning processes and technologies in the aim of achieving a technical and economic optimum. (authors)

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Decovalex-2019 (Executive Summary)

The DECOVALEX Project is an on-going international research collaboration, established in 1992, to advance the understanding and modeling of coupled Thermal (T), Hydrological (H), Mechanical (M) and Chemical (C) processes in geological in geological systems. DECOVALEX was initially motivated by the recognition that prediction of these coupled effects is an essential part of the performance and safety assessment of geologic disposal systems for radioactive waste and spent nuclear fuel. Later it was realized that these processes also play a critical role in other subsurface engineering activities, such as subsurface CO 2 storage, enhanced geothermal systems, and unconventional oil and gas production through hydraulic fracturing. Research teams from many countries (e.g., Canada, China, Czech Republic, Finland, France, Germany, Japan, Republic of Korea, Spain, Sweden, Switzerland, Taiwan, United Kingdom, and the United States) various institutions have participated in the DECOVALEX Project over the years, providing a wide range of perspectives and solutions to these complex problems. These institutions represent radioactive waste management organizations, national research institutes, regulatory agencies, universities, as well as industry and consulting groups. The overall aim of DECOVALEX-2019 was to increase the understanding of various thermo-hydro-mechanical-chemical processes of importance for radionuclide release and transport from a repository to the biosphere and how they can be described and modelled using mathematical models. The scientific and technical objectives are: to increase the basic understanding of T-H-M-C coupled processes in fractured rocks (crystalline, sedimentary, argillaceous) and buffer materials; to investigate the predictive capabilities of different codes to field experiments and to perform verification of codes; to exchange experimental data, and improve the understanding of the constitutive behavior of crystalline and argillaceous rock masses and buffer materials; and to perform THMC calculations in a performance/safety assessment context.

58 GEOSCIENCES↗

Brine Availability Test in Salt (BATS) (FY25 Final Closeout)

This fiscal year 2025 (FY25) closeout report summarizes the final experiments conducted and data collected in phase 2 of the Brine Availability Test in Salt (BATS 2). The design and interpretation of the test was previously funded by the US Department of Energy’s Office of Nuclear Energy (DOE-NE) Spent Fuel and Waste Disposition Program, under the Disposal Research and Development (R&D) program of the Office of Spent Fuel & Waste Science and Technology (SFWST) but they are not funding it anymore (since May 2025). The experiment was located underground at the Waste Isolation Pilot Plant (WIPP), southeastern New Mexico, which is a DOE Office of Environmental Management (DOE-EM) site managed by the Carlsbad Field Office (CBFO). DOE-EM funds the WIPP Test Coordination Office (TCO), which provided critical implementation support for the continued execution of BATS. A high-level summary of processes expected in a salt repository are given in Kuhlman & Mills (2025).

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