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DOE OSTI2024

Model Development for Thermal-Hydrology Simulations of a Full-Scale Heater Experiment in Opalinus Clay

Disposal of commercial spent nuclear fuel in a geologic repository is studied. In situ heater experiments in underground research laboratories provide a realistic representation of subsurface behavior under disposal conditions. Here, this study describes process model development and modeling analysis for a full-scale heater experiment in opalinus clay host rock. The results of thermal-hydrology simulation, solving coupled nonisothermal multiphase flow, and comparison with experimental data are presented. The modeling results closely match the experimental data.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

NETL’s Analog Studies to Geologic Storage of CO 2 – Overview

This document provides a high-level overview and key takeaways of the three analog studies to geologic storage of CO 2 : underground natural gas storage, deep well waste disposal using U.S. EPA UIC Class I wells, and CO 2 EOR. The three analog studies are available on NETL’s website under the Collection Name: NETL’s Analog Studies to Geologic Storage of CO 2 .

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2020

A Dose Assessment Model for Radioactive Waste Exposed by Gully Erosion at West Valley - 20513

The Western New York Nuclear Service Center (WNYNSC), located approximately 48 km south of Buffalo, New York, is the site of a former nuclear fuel reprocessing and radioactive waste disposal facility. Spent nuclear fuel was processed there from 1966 to 1972, leaving behind radioactive and chemical wastes in two disposal areas and a waste tank farm. Site operations also resulted in releases of radioactivity to site soils, groundwater, and to surface waters draining the site. The New York State Energy Research and Development Authority (NYSERDA) and the U.S. Department of Energy (DOE) are collaborating in a process of decision making for decommissioning those facilities remaining at the WNYNSC following the completion of Phase 1 decommissioning. Neptune and Company, Inc. (Neptune) was contracted to develop a probabilistic performance assessment (PPA) computer model to assist the agencies in this process. The PPA Model includes a contaminant transport component focusing on the movement of contaminants within and among environmental media including groundwater and surface water transport, contaminant translocation by plants and animals, diffusion, and erosion. The model also includes evaluation of potential exposure and health effects for a Resident Farmer exposure scenario, where the Resident Farmer represents a critical group, described as that group of individuals reasonably expected to receive the greatest exposure to residual radioactivity for any applicable set of circumstances. The West Valley Site is located in the glaciated Allegheny Plateau region of western New York State. The waste reprocessing and disposal areas were constructed on a relatively fat area of plateau dissected by drainages of Buttermilk Creek, including Erdman Brook, Franks Creek, and Quarry Creek. An important aspect of the contaminant transport component of the PPA model is consideration of erosive processes such as slumping of the stream slopes and the advance of gullies from these streams. These erosion processes remove material from the plateau, growing the size of the creek valleys and making them wider and deeper. Of particular interest for the impact of erosion is radioactive waste contained in the Nuclear Regulatory Commission (NRC)-Licensed Disposal Area (NDA), the New York State-Licensed Disposal Area (SDA), and residual radiological inventory in the underground storage tanks at the Waste Tank Farm (WTF). The PPA Model is organized around geographically-defined facilities which were constructed upon the plateau, including the NDA, SDA, and WTF. Ongoing stream erosion processes will potentially transfer radioactive waste and residual inventory from these facilities to the ground surface on adjacent hillslope areas where erosion has breached the facility. Hence, it is important to evaluate the consequences of potential exposures to a Resident Farmer on the hillslopes below a breached facility. Two interrelated aspects of the dose assessment model related to hillslope exposure are discussed: 1) representation of the physical processes related to transport of radionuclides from facilities onto the hillslopes, and from the hillslopes into adjoining creeks where contaminated material migrates downstream with surface water and sediment, and; 2) adaptation of the activities associated with the Resident Farmer scenario to assess potential exposures to contamination in the hillslope areas. This discussion will cover the conceptual basis of the hillslope exposure and transport models, and also implementation in the PPA computer model. (authors)

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DOE OSTI2020

ANDRA's Underground Research Laboratory in Bure: Major Role in the Cigeo Development - 20005

The Industrial Center for Geological Disposal, also called Cigeo, is the deep geological disposal facility project developed by Andra since 1991. It is intended for the final disposal of High-Level Waste (HLW) and Intermediate Level Waste-Long Lived (ILW-LL) generated in France by the nuclear industry. Cigeo is located in the east of France (Meuse/Haute-Marne site). The disposal will be implemented in a 140 to 160 m thick clay layer at about 500 m depth. The license application file will be submitted in 2020. If the license is granted, the construction of the pilot phase of Cigeo (ramp, shafts, drifts, initial disposal vaults) could start in 2025 at the earliest. Since 2000, the development of the safety case of Cigeo for post closure has been supported by a three-stage construction, design, and scientific and technological experiment program performed in the French Underground Research Laboratory (URL) in Bure. These stages of the URL activities supported the iterative interactions between the knowledge acquired by scientific and technological R and D program, the design process and the safety assessment. The stages focused on assessing the suitability of the disposal concept, providing the basis for safety options and construction design, and preparing for licensing using large scale demonstrations. Recently, Andra launched the fourth development stage of the URL to implement a new set of technological experiments aiming at consolidating the design options of disposal cells and galleries for the pilot phase of Cigeo and at assessing monitoring technologies. Recently, Andra launched the fourth development stage of the URL to implement a new set of technological experiments aiming at consolidating the design of the pilot phase of Cigeo and at assessing design options and monitoring technologies. Removal of a segment ring, construction of an X drift crossing, improved construction techniques for HLW vaults, and construction of an ILW-LL prototype disposal vault are example activities during the fourth phase. In the future, the URL in Bure will remain a unique location to carry out research on promising technical solutions, to reduce Cigeo's construction and operation risks and strengthen the long-term safety assessment. (authors)

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DOE OSTI1978

An Assessment of the Nevada Test Site for Low-Level Waste Management

This report presents a summary of description information, facilities, and operating practices at the Nevada Test Site. It is concluded that NTS is not only a feasible location for surface and near-surface storage of low-level wastes, but also a potential site for underground storage of higher level wastes.

052002* -- Nuclear Fuels-- Waste Disposal & Storag↗

DOE OSTI2020

toughio v1.2.2

TOUGH is a widely used general purpose numerical simulator designed for fluid and heat flows of multiphase, multicomponent fluid mixtures in porous and fractured media, which has been applied to many real-world problems such as underground geological storage, geothermal reservoir engineering or nuclear waste disposal, to name a few. `toughio` is a lightweight, object-oriented and vectorized Python library that aims to provide user-friendly routines to facilitate pre- and post-processing of a TOUGH simulation. Although its primary target is the latest version TOUGH3, `toughio` can also be used to read/write TOUGH2 input/output files. Currently, to the best of our knowledge, only PyTOUGH offers an exhaustive number of features to carry out a complete TOUGH simulation using a scripting language. `toughio` and PyTOUGH share the same objectives, yet with different approaches and philosophies. On the one hand, a PyTOUGH mesh is represented as a MULGRAPH geometry where elements can be unstructured horizontally but only layered vertically. It also checks that the model parameters are valid to ensure that a TOUGH simulation runs smoothly. On the other hand, although it provides basic meshing features, `toughio` mostly relies on common third-party softwares (e.g. Abaqus, FLAC3D, Gmsh, LaGriT) to generate the mesh by importing and converting it to a TOUGH mesh, which also conveniently facilitates the coupling of TOUGH with any other simulator that also supports the same mesh formats. Besides, `toughio` is designed to be just a TOUGH input and output file reader/writer. It assumes that the user knows what he is doing and does not verify the validity of the model parameters (`toughio` does verify the functional arguments).

Luu, Keurfon↗

DOE OSTI2019

Fundamental Study on Sorption Behavior of Europium to Mica Flakes - 19116

It is well known that biotite exists with flake form in actual underground environment, dominating sorption of cationic radionuclides in a geological disposal system consisting of crystalline rock. The behavior of sorption of radionuclides will differ depending on the difference in shape, while most of the previous studies have used powder biotite. This study focused on the influence of flake structure of phyllosilicate minerals such as biotite on the diffusion and sorption processes of radionuclides, examining the sorption behaviors of Eu on biotite, phlogopite and muscovite in flake form. These flakes (5 mm x 6 mm) were prepared for the sorption experiments and the initial concentrations of the Eu ions were set to 1 mM. The sorption experiments were conducted by mixing 0.4 g of biotite flakes and 4 mL at pH 5 for 7 days. As a result, the sorption amount of Eu increased little by little with internal diffusion in all minerals, respectively. The diffusion process of Eu into the layer structure could be also observed by secondary ion-microprobe mass spectrometry (SIMS) analysis. The apparent diffusion coefficient of Eu into each flake was estimated by fitting a 2-D diffusion model to the sorption data. The estimated diffusion values were clearly smaller than that of matrix diffusion in granite. Hence it was suggested that the so-called Kd-model, which is a local equilibrium model, is not always applicable to the estimation of the retardation effect (i.e., migration rate) of radionuclides in rock matrix. In other words, the migration rate might depend also on the diffusion process of radionuclide in the inside of phyllosilicate minerals in addition of the macro diffusion process in rock matrix. Besides, sorption amount and diffusion coefficient of biotite, phlogopite and muscovite were largely different. Not only the tetrahedral structure adjacent to the potassium but also the octahedral structure might affect the restraint of the interlayer ion (K ions) and the sorption behavior. (authors)

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DOE OSTI2019

CO 2 Leakage During EOR Operations – Analog Studies to Geologic Storage of CO 2

This study focuses on CO 2 EOR operations and CO 2 geologic storage in saline-bearing formations; both individually and in relation to each other. This study is the third of three that evaluate analog industries of CO 2 storage (first focuses on underground natural gas storage and second on deep well waste disposal using U.S. EPA UIC Class I wells). Several pieces of information can be found throughout the study including overview of CO 2 EOR history and side-by-side comparisons of major synergistic features between CO 2 EOR and CO 2 storage in saline-bearing formations. Best practices and lessons learned from CO 2 EOR operations can provide perspective from which future CO 2 storage R&D pursuits and field projects can benefit. A summary document providing a high-level overview and key takeaways of this study is available on NETL’s website under the Collection Name: NETL’s Analog Studies to Geologic Storage of CO 2 .

58 GEOSCIENCES↗

DOE OSTI2022

Review of recent developments in iodine wasteform production

Radioiodine capture and immobilization is not only important to consider during the operation of reactors (i.e., I-131), during nuclear accidents (i.e., I-131 and I-129) or nuclear fuel reprocessing (i.e., I-131 and I-129), but also during disposal of nuclear wastes (i.e., I-129). Most disposal plans for I-129-containing waste forms (including spent nuclear fuel) propose to store them in underground repositories. Here, iodine can be highly mobile and, given its radiotoxicity, needs to be carefully managed to minimize long-term environmental impacts arising from disposal. Typically, any process that has been used to capture iodine from reprocessing or in a reactor is not suitable for direct disposal, rather conversion into a wasteform for disposal is required. The objectives of these materials are to use either chemical immobilization or physical encapsulation to reduce the leaching of iodine by groundwaters. Some of the more recent ideas have been to design capture materials that better align with disposal concepts, making the industrial processing requirements easier. Research on iodine capture materials and wasteforms has been extensive. This review will act as both an update on the state of the research since the last time it was comprehensively summarized, and an evaluation of the industrial techniques required to create the proposed iodine wasteforms in terms of resulting material chemistry and applicability.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

NASA NTRS2004

Permanent Sequestration of Emitted Gases in the Form of Clathrate Hydrates

Underground sequestration has been proposed as a novel method of permanent disposal of harmful gases emitted into the atmosphere as a result of human activity. The method was conceived primarily for disposal of carbon dioxide (CO2, greenhouse gas causing global warming), but could also be applied to CO, H2S, NOx, and chorofluorocarbons (CFCs, which are super greenhouse gases). The method is based on the fact that clathrate hydrates (e.g., CO2 6H2O) form naturally from the substances in question (e.g., CO2) and liquid water in the pores of sub-permafrost rocks at stabilizing pressures and temperatures. The proposed method would be volumetrically efficient: In the case of CO2, each volume of hydrate can contain as much as 184 volumes of gas. Temperature and pressure conditions that favor the formation of stable clathrate hydrates exist in depleted oil reservoirs that lie under permafrost. For example, CO2-6H2O forms naturally at a temperature of 0 C and pressure of 1.22 MPa. Using this measurement, it has been calculated that the minimum thickness of continuous permafrost needed to stabilize CO2 clathrate hydrate is only about 100 m, and the base of the permafrost is known to be considerably deeper at certain locations (e.g., about 600 m at Prudhoe Bay in Alaska). In this disposal method, the permafrost layers over the reservoirs would act as impermeable lids that would prevent dissociation of the clathrates and diffusion of the evolved gases up through pores.

Duxbury, N.↗

DOE OSTI2020

Retrieval of Transuranic Drums with Deflagration to Detonation Potential from a Vault at the Oak Ridge National Laboratory - 20207

During the final phase of a long-term project, a total of 80 TRU waste drums at ORNL have been retrieved from a below-grade storage vault in which they have remained untouched for about 25 years. The drums, which house material generated from past ORNL operations, are now stored and ready for eventual transfer to the Transuranic Waste Processing Center, located on Highway 95 in Oak Ridge. The Transuranic Waste Processing Center will process and ship the drums to WIPP in Carlsbad, New Mexico, for disposal. Workers lifted the drums one-by-one from the underground storage vault with a 110-ton crane over a one-week period in October 2017. Although the physical work was performed in one week, years of detailed planning preceded the fieldwork. A key obstacle needed to be overcome before the project could safely proceed. This obstacle involved the need to ensure controls were developed for the Deflagration to Detonation Transition (DDT) potential. DDT refers to a phenomenon in ignitable mixtures of a flammable gas and air (or oxygen) when a sudden transition takes place from a deflagration type of combustion to a detonation type of explosion. Basically, these drums, under certain circumstances, have a potential to detonate and overpressurize. Extensive analysis of the drums verified that detonation, while unlikely, was possible. Of the total 103 drums that were retrieved or moved in preparation for retrieval, 18 had DDT potential. Extensive security requirements were also necessary, due to the nature of the material, and this added another layer of complexity to the fieldwork. Prior to retrieving the drums from the underground storage vault, space needed to be made for them within the various above-grade facilities. Inventory and security requirements dictated what could be stored where, and the first step of the field work involved a number of drum movements in preparation for storage of the newly retrieved drums. This effort involved moving 21 drums already in the storage area to facilitate the addition of the newly retrieved ones. The retrieval project required a total of 107 critical lifts using the 110-ton crane, under 3 critical lift plans. The TRU waste storage facility is a Category 2 Nuclear Facility and is managed under an extensive Documented Safety Analysis. The TRU retrieval project was conducted in accordance with DOE Order 425.1D and included three Implementation Verification Reviews, two Management Assessments, one Readiness Assessment, and seven mock-ups in preparation for the field work. Despite being stored for 25 years, the drums were in good shape due to the integrity of the facilities where the drums were stored. Of the 80 drums that were retrieved, 12 were placed in special containers called 'overpacks' based on their content. The drums were removed safely and without incident, thanks to a lot of preparation and the efforts of a highly skilled team. Lessons learned through the DDT control development and the drum retrieval and overpacking process could be beneficial to others having to manage similar drums of TRU waste. (authors)

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DOE OSTI2019

Development of a Trace Detection Method for Mercury Speciation in Hanford Tank Waste - 19014

The U.S. Department of Energy's (DOE) Hanford Site in southeast Washington State stores 56 million gallons of chemical and radioactive waste stored in underground tanks. Mercury species in liquid and solid fractions of Hanford tank waste pose a risk to the implementation and execution of DoE's River Protection Project mission due to mercury concentration limits imposed under waste acceptance criteria, toxicity characteristic leaching procedure restrictions for secondary waste disposal, and air emission permits. Although there is currently no processing of tank waste at Hanford, waste processing is imminent, and representative mercury inventory tracking will be important to anticipate and mitigate stack emissions and secondary liquid waste disposal issues. A methodology to accurately determine total mercury and mercury speciation in Hanford tank waste, while maintaining low mercury detection limits, currently does not exist at Hanford. Moreover, the mercury species in the waste solids are unknown and may be waste type-specific. The development of an analytical method to separate and speciate mercury in tank waste was undertaken by the Pacific Northwest National Laboratory (PNNL) while leveraging an existing non-radiological, ultra-trace mercury detection technique frequently performed at PNNL's Marine Sciences Laboratory in Sequim, Washington. The ultimate goal of this work is to develop an analytical method to obtain detailed mercury speciation information and update flowsheet modeling assumptions related to mercury partitioning at Hanford. The first phase of testing was completed in the development of an analytical methodology for the low-level determination of total mercury, elemental mercury, methylmercury, and dimethylmercury in Hanford tank waste. This paper presents Phase I testing which focused on determining whether mercury species can be preconcentrated from aqueous Hanford tank waste solution onto a solid substrate, while simultaneously eliminating or very significantly reducing the parent radionuclide activity. This was demonstrated by developing protocols for the separation and preconcentration of total mercury and methylmercury onto solid substrates, then measuring the substrates for radiological content at PNNL's Radiochemical Processing Laboratory on the Hanford Site. The mercury content was successfully separated onto stable substrates and the radionuclide activity reduction was significant enough that the preconcentrated samples could be subjected to quantification using far less restrictive handling and radiological controls than would be required for the parent sample (i.e., contamination area fume hood vs. hot cell). This mercury separation method required no sample dilution, which enables maximum mercury detectability, while still allowing mercury speciation. (authors)

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DOE OSTI2025

Integrity Monitoring, Prediction and Assessment, Corrosion Control, and Repair of the Hanford Storage Tanks – 25213

Shukla, Pavan K. [Savannah River National Laborato↗

DOE OSTI2024

Feasibility Study for Cathodic Protection of Waste Tanks at Hanford

The Hanford Nuclear Reservation site contains approximately 211 million liters of radioactive and chemically hazardous waste arising from nuclear weapons production, beginning with World War II, and continuing through the Cold War. The waste is stored in 177 carbon-steel underground storage tanks, of which 149 are single-shell tanks (SSTs) and the remaining are double-shell tanks (DSTs). The mission of an ongoing River Protection Project is to retrieve the waste from the underground storage tanks and then treat and immobilize (i.e., vitrify) it for disposal. Waste from the older SSTs is being progressively retrieved into the newer DSTs for storage pending treatment, immobilization, and disposal. Waste chemistry controls are in place to mitigate general corrosion, pitting corrosion, and stress corrosion cracking (SCC) and minimize corrosion risks to the carbon steel tank liners of DST at the Hanford Site. The chemistry control program focuses on preserving liner integrity while balancing the need for caustic/nitrite additions that have large downstream impacts on vitrification. Some of the DSTs store salt cake or sludge waste layers contain out-of-specification interstitial liquids that contact the primary tank bottom or sidewall. Remediation of such layers through chemical additions is not practical. At the present time, the most significant leak integrity concern at Hanford is for corrosion of the tank bottoms of DSTs where the solids layer is out of specification because of hydroxide ion depletion and consequently, pH values that are less than 12. Cathodic protection (CP) has been proposed as a possible way to protect the liners of such tanks against corrosion. The objective of this project was to investigate the feasibility of applying CP to Hanford DSTs to mitigate internal corrosion of the carbon steel tank primary liners. The scope of the project included (1) a literature review on the application of CP to storage tank systems, in general, and to waste tanks at the Savannah River Site (SRS) and Hanford, (2) a review of the finite element analysis (FEA) model recently developed to evaluate potentials for a DST, and (3) the development of a boundary element analysis (BEA) model for CP for a DST. Tank AN-107, one of the Hanford DSTs, was selected for this work because (1) the bottommost waste layer in the tank is out-of-specification, and (2) electrochemical and resistivity data, based on recent waste sampling and testing, are available for this tank. The overarching conclusion of this study is that, based on the available information, it is feasible to apply CP to mitigate internal corrosion of the primary liners of the Hanford DSTs. However, significant additional information and work are required before a CP system can be designed. Recommendations for next steps in the design of the CP system will be presented.

WIERSMA, BRUCE↗

DOE OSTI2019

Mechanical Waste Gathering System (MWGS), Alternate Retrieval of Hanford Tank Waste - 19188

Hanford has 149 underground single shell tanks; of which several have known or suspected leak sites. Sluicing with large volumes of solution in compromised leaking tanks is not desirable due to the risk of environmental release. To address the complex retrieval of hardpan nuclear waste in Hanford's single shell tanks, a Mechanical Waste Gathering System is being developed and tested, which breaks up the waste and moves it to the surfaces without the introduction of liquid to the tank bottom. To improve the removal of hardpan tank waste, WRPS has leveraged industry knowledge and previous experience to identify a solution. This technology will provide another option for the technology toolbox at Hanford. (authors)

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DOE OSTI2020

Grimsel Test Site - A Successful International Underground Research Laboratory for Many Decades - 20429

For more than 35 years, Nagra and its partners from around the world have been conducting underground research projects at the Grimsel Test Site (GTS, www.grimsel.com) to contribute to the development and confirmation of safe geological disposal concepts and for the characterization of suitable host rock formations. Over the years, the results of this internationally recognized research program have been, and continue to be, incorporated directly into exploration programs, modelling, safety, and engineering feasibility studies on options for deep geological repositories. Each project of the GTS program involves field-testing, laboratory studies, design and modelling tasks, and integrates all scientific and technical aspects. Each project phase is planned with a duration of three to five years to facilitate practical and administrative aspects and allow flexibility for updating the overall project plans with the latest findings. Scientific and engineering interaction among the different projects is ensured via an international steering committee meeting. Hosting an IAEA level C radiation- controlled zone, which allows use of radionuclides, including actinides such as thorium, uranium, neptunium, plutonium and americium, in in-situ experiments is one of the reasons why GTS also developed as a center of excellence for work with radioactive tracers under realistic in-situ boundary conditions. Last year, a new five-year program (2019 to 2023) started which includes projects with a planning horizon of decades. The new five-year program includes a new phase of in-situ experiments using radionuclides such as migration experiments in the Colloid Formation and Migration project (CFM), the Long-Term Diffusion experiment (LTD) and the newly established C-14 and I-129 Migration in cement project (CIM). The 'High Temperature effects on Bentonite' (HotBENT) project is starting in the current phase and is studying the effects of elevated temperatures (>175 deg. C) on bentonite materials. As a generic underground research laboratory (URL) it is expected that the GTS will provide in the coming years a platform for international collaboration, knowledge development and knowledge transfer for the next generation of scientists and engineers in the area of radioactive waste disposal and geosciences. A key role regarding knowledge transfer and training is provided by the well-established Grimsel Training Center (GTC), which (beside many URL related issues) also covers many general aspects of radioactive waste management. In this paper we provide an overview of the current program at the GTS, focusing on the experiments that study the migration of radionuclides through engineered barrier materials and the geosphere. (authors)

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DOE OSTI2020

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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DOE OSTI2020

Engineered Barrier System Research Activities at LBNL (FY2020 Progress Report)

The design of a radioactive waste repository typically involves a multi-barrier system, including the natural barrier system (NBS), i.e., the host rock and its surrounding subsurface environment, and an engineered barrier system (EBS). The EBS is to be constructed from the man-made, engineered materials placed within a repository. The repository includes the waste form, waste canisters, buffer materials, backfill, and seals. The most common buffer material for EBS is compacted bentonite, which features low permeability and high retardation of radionuclides. Extensive studies concerning the behavior of bentonite backfill in crystalline and argillite/shale geologic environments for nuclear waste disposal have been conducted by means of laboratory experiments, numerical modeling, and large-scale in situ tests in Underground Research Laboratories (URLs) in Switzerland, France, Belgium, and Japan. This report includes the results of LBNL’s research activities conducted according to the objectives and scope of the work packages “SF-20LB01030802 Engineered Barrier System R&D - LBNL” and “SF-20LB01030806 Engineered Barrier System International Collaborations - LBNL” of the Spent Fuel and Waste Science and Technology (SFWST) (formerly called Used Fuel Disposal) Campaign of the Department of Energy’s (DOE) Office of Nuclear Energy. LBNL research studies included laboratory scale tests and modeling of the evolution of the EBS bentonite and associated coupled processes, and impacts of high temperature on parameters and processes relevant to the performance of crystalline and argillite repositories, including the evaluation of the technical basis for applying the maximum allowable temperature. The results of these studies are being addressing the technical elements necessary to evaluate the EBS design concepts. Emphasis is on the study of thermal, mechanical, and chemical processes that influence the performance of EBS, and the development of modeling capability for reliable assessment of these processes, and ultimately supporting the development of the GDSA model with detailed coupled THMC process models.

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