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DOE OSTI2021

Can we safely go to 200 °C? An integrated approach to assessing impacts to the engineered barrier system in a high-temperature repository

Abstract. This presentation gives on overview of the complex thermo-hydro-mechanical-chemical (THMC) processes occurring during the disposal of heat-producing high-level radioactive waste in geologic repositories. A specific focus is on the role of compacted bentonite, which is commonly used as an engineered backfill material for emplacement tunnels because of its low permeability, high swelling pressure, and radionuclide retention capacity. Laboratory and field tests integrated with THMC modeling have provided an effective way to deepen our understanding of temperature-related perturbations in the engineered barrier system; however, most of this work has been conducted for maximum temperatures around 100 ∘C. In contrast, some international disposal programs have recently started investigations to understand whether local temperatures in the bentonite of up to 200 ∘C could be tolerated with no significant changes to safety relevant properties. Raising the maximum temperature is attractive for economical and safety reasons but faces the challenge of exposing the bentonite to significant temperature increases. Strong thermal gradients may induce complex moisture transport processes while geochemical processes, such as cementation and perhaps also illitization effects may occur, all of which could strongly affect the bentonite and near-field rock properties. Here, we present initial investigations of repository behavior exposed to strongly elevated temperatures. We will start discussing our current knowledge base for temperature effects in repositories exposed to a maximum temperature of 100 ∘C, based on data and related modeling analysis from a large heater experiment conducted for over 18 years in the Grimsel Test Site in Switzerland. We then show results from coupled THMC simulations of a nuclear waste repository in a clay formation exposed to a maximum temperature of 200 ∘C. We also explore preliminary data from a bench-scale laboratory mock-up experiment, which was designed to represent the strong THMC gradients occurring in a “hot” repository, and we finally touch on a full-scale field heater test to be conducted soon in the Grimsel Test Site underground research laboratory in Switzerland (referred to as HotBENT).

Birkholzer, Jens T. (ORCID:0000000279891912)↗

DOE OSTI2025

Development of an ERT‐Based Framework for Bentonite Buffers Monitoring From Laboratory Tests: 2. Quantitative Moisture Dynamics Estimation Model

Abstract The long‐term containment of high‐level radioactive waste in geological disposal repositories relies on Engineered Barrier Systems (EBS), with bentonite clay emerging as a candidate material due to its unique properties. Understanding moisture dynamics within bentonite buffers is crucial for EBS performance, as it directly influences the material's swelling capacity, thermal and hydraulic conductivity, mechanical properties, and long‐term evolution under complex thermal‐hydrological‐mechanical (THM) processes. This study develops an advanced Electrical Resistivity Tomography (ERT)‐based framework to quantitatively monitor moisture dynamics under THM conditions. Our framework extends the Waxman‐Smits model to incorporate the coupled effects of temperature, water content, fluid chemistry, and mechanical changes on bentonite's electrical properties. Utilizing HotBENT‐Lab data from our companion paper, which includes electrical conductivity, CT density, and thermocouple measurements, this study offers a novel methodological framework bridging different scales of the model. Our results show that the extended model can estimate water content from ERT data, capturing spatial and temporal variations in moisture distribution within bentonite columns. However, the model tends to overestimate water content compared to CT density‐derived measurements. We address this discrepancy by incorporating a simplified swelling effect model, which improves agreement between ERT and CT density‐based water content estimates. We also discuss model limitations, including simplified treatment of swelling and micropore effects, and propose a conceptual framework for transitioning from laboratory to field applications, addressing challenges such as parameter scalability, field validation methods, and integration of diverse data sources. This ERT‐based framework can potentially advance real‐world moisture monitoring of bentonite‐based EBS in nuclear waste repositories. Plain Language Summary Safely containing high‐level radioactive waste depends on barriers made from materials like bentonite clay, which is effective because it swells and seals in the waste. To ensure these barriers work well over time, it's important to understand how moisture moves through the clay. Our study developed a new method using ERT to monitor moisture levels in bentonite under conditions that mimic those in actual storage sites, including changes in temperature, water content, and mechanical stress. This study improved an existing model to better account for how these factors affect the clay, allowing us to create more accurate moisture maps. Initially, the proposed model overestimated the amount of water in the clay, but its accuracy was improved by factoring in how the clay swells when wet. This study also identified some limitations of the model and suggested ways to adapt it for use in real‐world waste storage sites. This new approach could lead to better monitoring and safety checks for nuclear waste storage systems, helping to ensure long‐term containment. Key Points This work develops an ERT‐based framework extending the Waxman‐Smits model to monitor bentonite moisture dynamics during coupled THM processes The extended model accurately estimates water content from Electrical Resistivity Tomography data, incorporating swelling effects to improve precision This work proposes a conceptual framework for transitioning from laboratory to field applications, advancing EBS monitoring in nuclear waste repositories

Chen, Hang↗

DOE OSTI2022

DPC Direct Disposal Postclosure Thermal Modeling

Performance of geologic radioactive waste repositories depends on near-field and far-field processes, including km-scale flow and transport in engineered and natural barriers, that may require simulations of up to 1 M years of regulatory period. For a relatively short time span (less than 1000 years), the thermohydro-mechanical-chemical (THMC) coupled processes caused by heat from the waste package will influence near-field multiphase flow, chemical/reactive transport, and mechanical behaviors in the repository system. This study integrates the heat-driven perturbations in thermo-hydro-mechanical characteristics into thermo-hydro-chemical simulations using PFLOTRAN to reduce dimensionality and improve computational efficiency by implementing functions of stress-dependent permeability and saturation-temperature-dependent thermal conductivity. These process couplings are developed for spent nuclear fuel in dual-purpose canisters in two different hypothetical repositories: a shale repository and a salt repository.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

A Rapid Sequential Separation Method for Determination of Actinides and TENORM in Fracking Wastes - 20223

With recent advances in unconventional drilling technology in the US and around the world, the risks of environmental contamination and exposure due to technologically enhanced naturally occurring radioactive materials (TENORM) from fracking wastes have also surfaced. Although the Permian Basin (southeastern New Mexico, western Texas) has long been a modest producer of oil, the advent of fracking technology a decade ago turned it into hot property for producers. A 2018 assessment of undiscovered, technically recoverable continuous oil and gas resources by the U.S. Geological Survey estimates an average 46.3 billion barrels of oil and 281 trillion cubic feet of gas. This is likely to increase the use of unconventional drilling and TENORM generation. TENORM concentrations in oil and gas exploration and production waste can be several hundred to several thousand times higher than background TENORM concentrations. Both the thorium and uranium decay chains contribute to airborne radionuclides arising from the radon gas escaping the ground and subsequently decaying as airborne particulates. Additionally, radon decay products {sup 210}Pb and {sup 210}Po can build up in scale on the internal surfaces of oil and gas handling pipes and in sludge in refineries, becoming potential inhalation and ingestion hazards for workers. Furthermore, southeastern NM is also home to world's only licensed and operating transuranic nuclear waste repository, the DoE's Waste Isolation Pilot Plant (WIPP). Plutonium isotopes ({sup 239+240}Pu) and {sup 241}Am, are expected to account for more than 99% of the total radioactivity scheduled for disposal in the WIPP repository. Thus, an improved understanding of the environmental fate and transport of TENORM, liberated by unconventional drilling, is essential to assess how best to protect individuals and the environment. In this context, accurate measurement of TENORM and actinides in environmental and biological samples is essential. In this presentation, a new sequential method for the separation and pre-concentration of actinides (Pu, Am, Np) and TENORM (Po, U and Th) derived from oil and gas exploration is proposed. The TEVA method involves a rapid co-precipitation step to remove matrix interferences from the samples, followed by plutonium oxidation state adjustment to Pu (IV) and an incubation period of ∼ 1 hour at 50-60 deg. C to allow the resultant Po (II) to oxidize into Po(IV). The polonium, neptunium, thorium and plutonium are then separated on a TEVA column, while americium and uranium are separated on a TRU column. After separation, the alpha counting source was prepared by micro-precipitation with copper sulfide (CuS) for polonium and neodymium fluoride (NdF{sub 3}) for actinides. The efficiency and reliability of the procedures were tested by analyzing filter, drinking water and frack sand samples. The method is simple, robust and can be performed quickly with excellent removal of interferences, high chemical recovery and very good alpha peak resolution. (authors)

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2020

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↗

DOE OSTI2025

Past Approaches for Spent Nuclear Fuel, High-Level, and Transuranic Waste Disposal in the United States—Part 1: Safety Criteria and Treatment of Uncertainty

The United States (US), with its 50-year experience in developing deep geologic disposal for transuranic waste, spent nuclear fuel (SNF), and high-level radioactive waste (HLW), has much to share with other countries. Yet, other countries are better able to translate the US experience and corresponding policy decisions into solutions sensible for their country when they understand the compliance requirements in US laws and regulations. This paper presents past approaches in the generic and site-specific standards of the US Environmental Protection Agency (EPA) and implementing regulations of the US Nuclear Regulatory Commission (NRC) using the framework provided by (1) key questions of the Blue Ribbon Commission on America’s Nuclear Future, and (2) international consensus standards. Both the 1985 EPA generic standards, as updated in 1993 and applied at the operating Waste Isolation Pilot Plant in southern New Mexico for transuranic waste from atomic energy defense activities, and the EPA 2008 site-specific standards and implementing regulations, as applied at the proposed Yucca Mountain repository in southern Nevada for SNF and HLW, adopt the strategy of using quantitative, probabilistic analysis to assess performance and compliance.

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DOE OSTI2021

Modeling and Analysis of the Transport and Disposal of Beryllium Moderator Blocks and Greater than Class C (GTCC) Waste

Radioactive waste in the United states is categorized based on its source, radioactivity, and security risk. The categorization method employed by the Nuclear Regulatory Committee (NRC) for low-level radioactive waste is not currently applied to any waste generated by the Department of Energy (DOE) or disposed of in DOE facilities. As a result, there is a large volume of DOE-generated waste with characteristics similar to the NRC’s “Greater than Class C” (GTCC) waste category which presently do not have a path for long-term disposal. Much of this waste cannot undergo the standard commercial disposition process due to it being categorized as Transuranic (TRU) waste under DOE guidelines, due to elevated concentrations of key fission products. Possible solutions to this dilemma are explored in the Environmental Impact Statement for the “Disposal of Greater-Than-Class-C Low Level Radioactive Wave” (EIS-0375). This paper analyzes several EIS possible paths for disposal for this “orphaned” waste. One example that highlights this categorization issue is the spent beryllium cladding that has been extracted from the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) throughout its operational lifetime. These beryllium blocks surround the reactor’s main chambers and components and serve as neutron moderators. This paper will use this particular waste form to analyze the economic and technological viability of transporting and storing this material under the storage and transportation criteria of the Waste isolation Pilot Plant (WIPP), which was deemed the most feasible disposition path according to the EIS. The example of this waste form used in this paper that highlights this categorization issue is the spent beryllium cladding extracted from Idaho National Laboratory’s (INL) Advanced Test Reactor (ATR). These beryllium blocks surround the reactor’s components and serve as neutron moderators, and thus have accumulated high concentration of high-activity transuranic isotopes. This paper analyzed the viability of transporting and storing this material under the storage and transportation criteria of the five primary disposition paths covered in EIS-0375. An array of calculations was carried out to assess the viability of the various disposition paths for the beryllium shipments as well as other waste shipments that fall within the GTCC category. Modeling with MCNP 6.2 was conducted to determine whether the beryllium blocks could be safely stored and transported within a 72-B cask; the standard shipping container for remote-handled, transuranic waste, while also meeting regulatory limits at various disposition paths. A cost analysis of the transportation and long-term disposal paths mentioned in the EIS was also carried out using available data and information from similar waste shipments. Lastly, a geochemical analysis of the various geological repository discussed in the EIS was also carried out using the Geochemists Workbench Release 14. Long-term disposal of the beryllium blocks at the Waste Isolation Pilot Plant (WIPP) proved to be the most cost-effective long-term disposition option of the ones considered in the EIS. A dose rate calculation at both contact and remote-handling distances indicate that the analyzed beryllium shipments should not exceed the exposure limits at any of the considered locations. Greater-than-class-C waste has been left in a regulatory state of limbo for years, resulting in backlogs of inventory across several research sites in the United States. Due to its high activity and presence of transuranic isotopes, it is imperative to ensure that it remains inaccessible and sequestered both in its short-term interim as well as a long-term geologic time scale. The information and assessments done in the paper could potentially serve as a reference for any future shipments of this waste form at the WIPP facility as well as other disposition paths that may be considered in the future.

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DOE OSTI2020

Geology and Design of Major Spent Fuel Repositories

This report summarizes the geology and design concepts of four major spent fuel repositories from around the world. The crystalline rock with bentonite backfill is currently implemented in Finland with another repository being constructed in Sweden. The salt formation design focuses on the Waste Isolation Pilot Plant (WIPP) repository in the United States. The volcanic tuff open air design is based on the Yucca Mountain location. The volcanic tuff open air design at Yucca Mountain is only a concept as this repository has not been approved for operations. The last spent fuel repository design, clay/shale formation with backfill, focuses on the Cigeo Project in France, which is also pending.

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DOE OSTI2020

Cigeo Design Options, Risks and Opportunities Management: A Continuous Project Management Goal - 20015

Andra is currently completing the detailed engineering phase of the stepwise design development of Cigeo, the French HLW and IL-LLW Deep Geological Repository. If licensed in the coming years, Cigeo's progressive construction, industrial operation and closure will last over more than a century. Such a duration, associated with the requirements in terms of reversibility, lead to the choice of an incremental design that allows, for next operator generations, to integrate improvements derived from scientific and technical progress as well as from experience feedback from its operation. This incremental design approach, associated with a permanent risks and opportunities management process, will continue after licensing of Cigeo as a tool for the governance of the reversibility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

Low Activity Waste Tuning Feed Material Testing VSL-18R4350-1 (Final Report)

High level waste (HLW) from the Hanford tank farms will be retrieved and transferred to the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The waste will be pretreated to separate the soluble chemical salts (mostly sodium nitrate/nitrite and sodium hydroxide) from the bulk of the radioactive materials. This decontaminated salt solution is the low activity waste (LAW) stream which constitutes most of the total waste volume. The remaining HLW contains most of the radioactive materials but a fraction of the volume. Each of these waste streams is directed to an independent processing facility where the waste is mixed with glass forming chemicals (mostly silica and borax or boric acid) and fed into the melters for stabilization by conversion into glass. The molten glass is poured into stainless steel containers to produce packages for disposal: local shallow burial for the LAW containers and a future geologic repository for the HLW containers. The LAW facility melters produce significant quantities of NOx-laden off-gas that require abatement in accordance with 10 CFR 830 and air emission requirements. The NOx emissions also pose a safety risk. The LAW facility commissioning sequence requires that the melters be operated and process control loops tuned prior to introducing waste or waste feed materials that produce NOx. Therefore, a temporary or permanent system architecture is needed that provides a feed supply to the melters for the purpose of melter testing and off-gas tuning that does not result in the production of significantly hazardous off-gas products.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Time-Temperature-Transformation (TTT) Diagram for a Sludge Batch 9 Glass Composition Based on Coupled-Operation with the Salt Waste Processing Facility

The amorphous structure of a glass waste form has the potential to rearrange into crystalline phases at temperatures between the liquidus temperature and the glass transition temperature (Tg). Certain phases that can form will be detrimental to the durability of the glass and it is important to know the conditions that promote devitrification. The canister-centerline-cooling (CCC) profile is used to replicate the area within the center of the Defense Waste Processing Facility (DWPF) canister with the slowest cooling during the initial cool down after pouring, which has the greatest potential for crystallization. Other time-temperature conditions that cause significant changes in either phase structure or phase composition are identified by a time-temperature-transformation (TTT) study. The phase stability of a waste form must be determined as a part of the Waste Acceptance Product Specifications (WAPS) if it is to eventually be stored in a geologic repository. This requires the creation of a TTT diagram and analysis of the Tg, as defined by the Department of Energy (DOE). The previous TTT study for a DWPF glass waste form was completed in 2010 prior to coupled operation with the Salt Waste Processing Facility (SWPF). SWPF transfers two high activity waste streams to DWPF for vitrification: a cesium-containing strip effluent and a stream containing monosodium titanate/sludge solids. These SWPF streams were first transferred to DWPF for vitrification during Sludge Batch 9 (SB9) in 2021. The impact of these SWPF streams on crystallization behavior was not determined in previous studies and the need for data to satisfy WAPS Specification 1.4 for SB9 was identified.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

DPC Direct Disposal R&D Independent Technical Review: Plan

Developing and evaluating approaches for direct geologic disposal of commercial spent nuclear fuel (SNF) in dual-purpose canisters (DPCs) is a cross-cutting multi-disciplinary activity that is directly tied to the implementation of DPCs by the nuclear industry. The ultimate goal of the DPC direct disposal R&D program is to facilitate and maximize safe, cost-effective, licensed direct disposal. Independent Technical Review (ITR) is needed to maximize the impact of the R&D program on future implementation. The review will involve a team of experts representing the nuclear industry, repository sciences, and licensing. The team will be charged to review a set of representative technical reports and other information, and answer a set of questions that focus on R&D steering.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

DECOVALEX-2023: Task B Final Report

In all repository concepts for the geological disposal of radioactive waste, an engineered barrier system (EBS) is used to encapsulate the waste canister, or, to act as borehole or gallery seals. These systems are often based on bentonite clays due to their low permeability and high swelling capacity enabling the closure of engineering voids. However, in all repository concepts gases will be generated through the corrosion of metallic materials (under anoxic conditions), the radioactive decay of waste and the radiolysis of water. Thus, understanding the processes and mechanisms controlling the advective movement of gas (as a discrete phase) in clay-based materials is a key aspect when assessing the impact of gas flow in a repository safety case.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Machine learning surrogates for surface complexation model of uranium sorption to oxides

Abstract The safety assessments of the geological storage of spent nuclear fuel require understanding the underground radionuclide mobility in case of a leakage from multi-barrier canisters. Uranium, the most common radionuclide in non-reprocessed spent nuclear fuels, is immobile in reduced form (U(IV) and highly mobile in an oxidized state (U(VI)). The latter form is considered one of the most dangerous environmental threats in the safety assessments of spent nuclear fuel repositories. The sorption of uranium to mineral surfaces surrounding the repository limits their mobility. We quantify uranium sorption using surface complexation models (SCMs). Unfortunately, numerical SCM solvers often encounter convergence problems due to the complex nature of convoluted equations and correlations between model parameters. This study explored two machine learning surrogates for the 2-pK Triple Layer Model of uranium retention by oxide surfaces if released as U(IV) in the oxidizing conditions: random forest regressor and deep neural networks. Our surrogate models, particularly DNN, accurately reproduce SCM model predictions at a fraction of the computational cost without any convergence issues. The safety assessment of spent fuel repositories, specifically the migration of leaked radioactive waste, will benefit from having ultrafast AI/ML surrogates for the computationally expensive sorption models that can be easily incorporated into larger-scale contaminant migration models. One such model is presented here.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

DECOVALEX-2019 (Task E Final Report)

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. This document is the final report of Task E which was proposed and coordinated by Andra, the National Radioactive Waste Management Agency in France, presenting the technical definitions of the problems studied, approaches applied, achievements made and outstanding issues for future research. The purpose of Task E of the DECOVALEX-2019 project is to investigate upscaling THM modelling from small-scale experiments (some cubic meters) to full-scale experiments (some ten cubic meters) and finally to the scale of the waste repository (cubic kilometers). To achieve this aim, the data of two in-situ heating experiments performed by Andra (the French National Radioactive Waste Management Agency) in the Meuse/Haute-Marne Underground Research Laboratory (MHM URL) have formed the basis for the understanding of the THM behavior of the COx at different scales. The first experiment provided the reference values of the THM parameters by means of a calibration exercise and they were used for a blind prediction and an interpretative analysis of the second one.

58 GEOSCIENCES↗

DOE OSTI2024

Value of abstraction in performance assessment – When is a higher level of detail necessary?

In this study, different approaches in performance assessment (PA) of the long-term safety of a repository for radioactive waste were examined. This investigation was carried out as part of the DECOVALEX-2023 project, an international collaborative effort for research and model comparison. One specific task of the DECOVALEX-2023 project was the Salt Performance Assessment Modelling task (Salt PA), which aimed at comparing various models and methods employed in the performance assessment of deep geological repositories in salt. In the context of the Salt PA task, three distinct teams from SNL (United States), Quintessa Ltd (United Kingdom), and GRS (Germany) examined the consequences of employing different levels of abstractions when modelling the repository's geometry and implementing various features and processes, using the example of a simple hypothetical repository structure in domal salt. Each team applied their own tools: PFLOTRAN (SNL), QPAC (Quintessa) and LOPOS (GRS). These differ essentially regarding numerical concept and degree of detail in the representation of the underlying physical processes. The discussion focused on when simplifications can be appropriately applied and what consequences result from them. Furthermore, it was explored when and if a higher level of fidelity in geometry or physical processes is required.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

Evaluation of Desiccation Behavior in Basalt Microfiber–Reinforced Bentonite Clay for Geological Repositories of Nuclear Spent Fuel Using Digital Image Correlation

ABSTRACT Secure storage of nuclear spent fuel (NSF) is of great concern for protecting public health and safety. The preferred long-term solution is underground containment in geological repositories, where one or more engineered barrier materials (EBM) encapsulate the NSF and separate it from the natural rock. Bentonite clay is commonly used as an EBM due to its many advantageous properties including low hydraulic conductivity, which ensures limitation of water infiltration to the system and the subsequent risk of corrosion in NSF canisters. However, bentonite clay subjected to heating from nuclear decay may form desiccation cracking. This study conducted disk-shaped free shrinkage tests and ring-shaped restrained shrinkage tests of bentonite clay samples reinforced with basalt microfibers. Digital image correlation was used as a noncontact full-field displacement measurement to track the time-evolving shrinkage and desiccation cracking phenomena and make quantified comparisons between plain bentonite and bentonite with varying contents of basalt microfibers (i.e., 0.0, 0.5, 1.0, and 1.5 % wt.). Results indicate that plain bentonite and basalt microfiber-reinforced samples showed similar free shrinkage behavior, while desiccation cracking behavior was significantly altered by adding basalt microfibers. Microfiber reinforcement effectively reduced major cracks through a “crack-bridging” effect while causing minor cracks to initiate earlier and at higher moisture contents than plain bentonite. Results infer that reinforcing plain bentonite with inorganic microfibers can potentially control desiccation cracking, leading to safer and improved nuclear waste management.

Materials Science↗

DOE OSTI2023

Modeling Geologic Waste Repository Systems Below Residual Saturation

The heat generated by high-level radioactive waste can pose numerical and physical challenges to subsurface flow and transport simulators if the liquid water content in a region near the waste package approaches residual saturation due to evaporation. Here, residual saturation is the fraction of the pore space occupied by liquid water when the hydraulic connectivity through a porous medium is lost, preventing the flow of liquid water. While conventional capillary pressure models represent residual saturation using asymptotically large values of capillary pressure, here, residual saturation is effectively modeled as a tortuosity effect alone. Treating the residual fluid as primarily dead-end pores and adsorbed films, relative permeability is independent of capillary pressure below residual saturation. To test this approach, PFLOTRAN is then used to simulate thermal-hydrological conditions resulting from direct disposal of a dual-purpose canister in unsaturated alluvium using both conventional asymptotic and revised, smooth models. Importantly, while the two models have comparable results over 100 000 years, the number of flow steps required is reduced by approximately 94%.

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