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DOE OSTI2023

Temperature-dependent microbial reactions by indigenous microbes in bentonite under Fe(III)- and sulfate-reducing conditions

Bentonite is a promising buffer material for constructing spent nuclear fuel (SNF) repositories. However, indigenous microbes in bentonite can be introduced to the repository and subsequent sealing of the repository develops anoxic conditions over time which may stimulate fermentation and anaerobic respiration, possibly affecting bentonite structure and SNF repository stability. Moreover, the microbial activity in the bentonite can be impacted by the heat generated from radionuclides decay. Therefore, to investigate the temperature effect on microbial activities in bentonite, we created microcosms with WRK bentonil (a commercial bentonite) using lactate as the electron donor, and sulfate and/or ferrihydrite (Fe(III)) as electron acceptors with incubation at 18 °C and 50 °C. Indigenous WRK microbes reduced sulfate and Fe(III) at both temperatures but with different rates and extents. Lactate was metabolized to acetate at both temperatures, but only to propionate at 18 °C during early-stage microbial fermentation. More Fe(III)-reduction at 18 °C but more sulfate-reduction at 50 °C was observed. Thermophilic and/or metabolically flexible microbes were involved in both fermentation and Fe(III)/ sulfate reduction. In conclusion, our findings illustrate the necessity of considering the influence of temperature on microbial activities when employing bentonite as an engineered buffer material in construction of SNF repository barriers.

59 BASIC BIOLOGICAL SCIENCES↗

DOE OSTI2024

Dataset for "Evaluating Deep Learning Approaches for Predictions in Unmonitored Basins with Continental-scale Stream Temperature Models" Willard et al. (2024)

This data release provides all data and code used in the paper " "Evaluating Deep Learning Approaches for Predictions in Unmonitored Basins with Continental-scale Stream Temperature Models" Willard et al. (2024)" to model stream temperature, evaluate, and assess results. The associated manuscript explores current open questions in prediction in ungauged and unmonitored basins concerning top-down versus bottom-up approaches, tradeoffs between data available and input requirements, and the appropriate representation of catchment attributes as inputs to deep learning models. Modeling was done primarily with long short-term memory (LSTM) models, and stream site coverage spans 1362 locations across the conterminous United States. The data is organized into these items items:Code repository and data for the paper " "Evaluating Deep Learning Approaches for Predictions in Unmonitored Basins with Continental-scale Stream Temperature Models" Willard et al. (2024)".Code: stream_temp_ml_regionalization.zip contains the code repositoryData to run the code: - data_dir.zip -- contains all files that should be moved to the "DATA_DIR" variable defined in the "set_env_vars.sh" script in the code repository- metadata_dir.zip -- contains all files that should be moved to the "METADATA_DIR" variable defined in the "set_env_vars.sh" script in the code repository- error_analysis_attribute_and_groundwater_dir.zip - workflows for the extended error analysis by stream attribute and groundwater influenceData produced by the code and used in the paper:- outputs_dir.zip - contains model output and results (outputs_dir/results), model weights (outputs_dir/models), and all other outputs used for the paper including feature importances.To cite this code, please use the following BibTeX or MLA entries:bibtex:@misc{willard2024streamdata, author = {Jared Willard and Fabio Ciulla and Helen Weierbach and Vipin Kumar and Charuleka Varadharajan}, title = {Dataset for "Evaluating Deep Learning Approaches for Predictions in Unmonitored Basins with Continental-scale Stream Temperature Models"}, year = {2024}, doi = {10.15485/2448016}, publisher = {ESS-DIVE Repository}, url = {https://doi.org/10.15485/2448016}}MLA: Willard, Jared, et al. Dataset for "Evaluating Deep Learning Approaches for Predictions in Unmonitored Basins with Continental-scale Stream Temperature Models". 2024. ESS-DIVE Repository, doi:10.15485/2448016.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2011

Disposal systems evaluations and tool development : Engineered Barrier System (EBS) evaluation.

Key components of the nuclear fuel cycle are short-term storage and long-term disposal of nuclear waste. The latter encompasses the immobilization of used nuclear fuel (UNF) and radioactive waste streams generated by various phases of the nuclear fuel cycle, and the safe and permanent disposition of these waste forms in geological repository environments. The engineered barrier system (EBS) plays a very important role in the long-term isolation of nuclear waste in geological repository environments. EBS concepts and their interactions with the natural barrier are inherently important to the long-term performance assessment of the safety case where nuclear waste disposition needs to be evaluated for time periods of up to one million years. Making the safety case needed in the decision-making process for the recommendation and the eventual embracement of a disposal system concept requires a multi-faceted integration of knowledge and evidence-gathering to demonstrate the required confidence level in a deep geological disposal site and to evaluate long-term repository performance. The focus of this report is the following: (1) Evaluation of EBS in long-term disposal systems in deep geologic environments with emphasis on the multi-barrier concept; (2) Evaluation of key parameters in the characterization of EBS performance; (3) Identification of key knowledge gaps and uncertainties; and (4) Evaluation of tools and modeling approaches for EBS processes and performance. The above topics will be evaluated through the analysis of the following: (1) Overview of EBS concepts for various NW disposal systems; (2) Natural and man-made analogs, room chemistry, hydrochemistry of deep subsurface environments, and EBS material stability in near-field environments; (3) Reactive Transport and Coupled Thermal-Hydrological-Mechanical-Chemical (THMC) processes in EBS; and (4) Thermal analysis toolkit, metallic barrier degradation mode survey, and development of a Disposal Systems Evaluation Framework (DSEF). This report will focus on the multi-barrier concept of EBS and variants of this type which in essence is the most adopted concept by various repository programs. Empasis is given mainly to the evaluation of EBS materials and processes through the analysis of published studies in the scientific literature of past and existing repository research programs. Tool evaluations are also emphasized, particularly on THCM processes and chemical equilibria. Although being an increasingly important aspect of NW disposition, short-term or interim storage of NW will be briefly discussed but not to the extent of the EBS issues relevant to disposal systems in deep geologic environments. Interim storage will be discussed in the report Evaluation of Storage Concepts FY10 Final Report (Weiner et al. 2010).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2011

Granite disposal of U.S. high-level radioactive waste.

This report evaluates the feasibility of disposing U.S. high-level radioactive waste in granite several hundred meters below the surface of the earth. The U.S. has many granite formations with positive attributes for permanent disposal. Similar crystalline formations have been extensively studied by international programs, two of which, in Sweden and Finland, are the host rocks of submitted or imminent repository license applications. This report is enabled by the advanced work of the international community to establish functional and operational requirements for disposal of a range of waste forms in granite media. In this report we develop scoping performance analyses, based on the applicable features, events, and processes (FEPs) identified by international investigators, to support generic conclusions regarding post-closure safety. Unlike the safety analyses for disposal in salt, shale/clay, or deep boreholes, the safety analysis for a mined granite repository depends largely on waste package preservation. In crystalline rock, waste packages are preserved by the high mechanical stability of the excavations, the diffusive barrier of the buffer, and favorable chemical conditions. The buffer is preserved by low groundwater fluxes, favorable chemical conditions, backfill, and the rigid confines of the host rock. An added advantage of a mined granite repository is that waste packages would be fairly easy to retrieve, should retrievability be an important objective. The results of the safety analyses performed in this study are consistent with the results of comprehensive safety assessments performed for sites in Sweden, Finland, and Canada. They indicate that a granite repository would satisfy established safety criteria and suggest that a small number of FEPs would largely control the release and transport of radionuclides. In the event the U.S. decides to pursue a potential repository in granite, a detailed evaluation of these FEPs would be needed to inform site selection and safety assessment.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2012

Advanced Nuclear Fuel Cycle Effects on the Treatment of Uncertainty in the Long-Term Assessment of Geologic Disposal Systems - EBS Input

The Used Fuel Disposition (UFD) Campaign within the Department of Energy's Office of Nuclear Energy (DOE-NE) Fuel Cycle Technology (FCT) program has been tasked with investigating the disposal of the nation's spent nuclear fuel (SNF) and high-level nuclear waste (HLW) for a range of potential waste forms and geologic environments. The planning, construction, and operation of a nuclear disposal facility is a long-term process that involves engineered barriers that are tailored to both the geologic environment and the waste forms being emplaced. The UFD Campaign is considering a range of fuel cycles that in turn produce a range of waste forms. The UFD Campaign is also considering a range of geologic media. These ranges could be thought of as adding uncertainty to what the disposal facility design will ultimately be; however, it may be preferable to thinking about the ranges as adding flexibility to design of a disposal facility. For example, as the overall DOE-NE program and industrial actions result in the fuel cycles that will produce waste to be disposed, and the characteristics of those wastes become clear, the disposal program retains flexibility in both the choice of geologic environment and the specific repository design. Of course, other factors also play a major role, including local and State-level acceptance of the specific site that provides the geologic environment. In contrast, the Yucca Mountain Project (YMP) repository license application (LA) is based on waste forms from an open fuel cycle (PWR and BWR assemblies from an open fuel cycle). These waste forms were about 90% of the total waste, and they were the determining waste form in developing the engineered barrier system (EBS) design for the Yucca Mountain Repository design. About 10% of the repository capacity was reserved for waste from a full recycle fuel cycle in which some actinides were extracted for weapons use, and the remaining fission products and some minor actinides were encapsulated in borosilicate glass. Because the heat load of the glass was much less than the PWR and BWR assemblies, the glass waste form was able to be co-disposed with the open cycle waste, by interspersing glass waste packages among the spent fuel assembly waste packages. In addition, the Yucca Mountain repository was designed to include some research reactor spent fuel and naval reactor spent fuel, within the envelope that was set using the commercial reactor assemblies as the design basis waste form. This milestone report supports Sandia National Laboratory milestone M2FT-12SN0814052, and is intended to be a chapter in that milestone report. The independent technical review of this LLNL milestone was performed at LLNL and is documented in the electronic Information Management (IM) system at LLNL. The objective of this work is to investigate what aspects of quantifying, characterizing, and representing the uncertainty associated with the engineered barrier are affected by implementing different advanced nuclear fuel cycles (e.g., partitioning and transmutation scenarios) together with corresponding designs and thermal constraints.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

EBS Report - LANL Experimental update of buffer/backfill at elevated P,T

The Used Fuel Disposition campaign is presently investigating various generic repository options for disposal of used fuel. Of interest are the disposal of high heat load canisters (up to 300 °C), which may allow for a reduced repository footprint. The focus of this experimental work is to increase our understanding on the stability of bentonite barriers under different geochemical and mineralogical conditions other than what has been previously investigated. Experiments were performed at 150 – 160 bars at temperatures up to 300 °C for five to six weeks. Unprocessed Wyoming bentonite was saturated with a K-Ca-Na-Cl-rich water at a 9:1 water:rock ratio. Experiments contained various steel plates or Cu-foil and were buffered to low Eh conditions using magnetite and metallic iron. Other experiments for international EBS research included mixed buffers and a heated bentonite with 20 wt. % free moisture for LBNL. The mixed buffer experiments (bentonite-graphite, bentonite-quartz sand) were conducted to investigate potential mineralogic/petrologic changes to mixed buffer systems proposed by the German nuclear repository program. Reaction products associated with the international EBS research were characterized by LANL with thermal-mechanical properties to be characterized by LBNL. The pH, K + , and Ca 2+ concentrations dropped, while SiO 2(aq) , Na + , and SO 4 2- concentrations increased throughout the experiments. The alkali and alkaline earth metals aqueous concentrations appear to be buffered via the montmorillonite and clinoptilolite exchange reactions. Illite or illite/smectite mixed-layer formation is significantly retarded in the closed system due to Na+ and SiO2(aq) accumulation and limited K + supply. Precursor clinoptilolite underwent extensive dissolution during the six week, 300 °C experiments subsequently producing a high-silicon analcime in addition to authigenic silica phases. Analcime and feldspar formation partially sequesters aqueous Al 3+ , thereby potentially inhibiting illitization. Associated with the zeolite alteration is a ~ 17 % volume decrease (assuming quartz formation) that translates into ~ 2% volume loss in the bulk bentonite. Pyrite decomposition is first observed at ~210 °C, generating available H 2 S (aq,g) that reacts with metal plates or evolves as a gas. The copper rapidly degrades in the presence of H 2 S (aq,g) , resulting in the formation of a chalcocite crust on the copper. A layer of Fe-saponite (or rarely, chlorite) forms at the steel bentonite interface, potentially acting as a passivating agent to retard further metal corrosion. Zeolite mineralogy alteration, metal-bentonite interface reactions, and illitization retardation under these experimental repository conditions are important considerations for the long-term stability evaluation of a used fuel repository.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

International Approaches for Nuclear Waste Disposal in Geological Formations: Report on Fifth Worldwide Review

The goal of the Fifth Worldwide Review is to document evolution in the state-of-the-art of approaches for nuclear waste disposal in geological formations since the Fourth Worldwide Review that was released in 2006. The last ten years since the previous Worldwide Review has seen major developments in a number of nations throughout the world pursuing geological disposal programs, both in preparing and reviewing safety cases for the operational and long-term safety of proposed and operating repositories. The countries that are approaching implementation of geological disposal will increasingly focus on the feasibility of safely constructing and operating their repositories in short- and long terms on the basis existing regulations. The WWR-5 will also address a number of specific technical issues in safety case development along with the interplay among stakeholder concerns, technical feasibility, engineering design issues, and operational and post-closure safety. Preparation and publication of the Fifth Worldwide Review on nuclear waste disposal facilitates assessing the lessons learned and developing future cooperation between the countries. The Report provides scientific and technical experiences on preparing for and developing scientific and technical bases for nuclear waste disposal in deep geologic repositories in terms of requirements, societal expectations and the adequacy of cases for long-term repository safety. The Chapters include potential issues that may arise as repository programs mature, and identify techniques that demonstrate the safety cases and aid in promoting and gaining societal confidence. The report will also be used to exchange experience with other fields of industry and technology, in which concepts similar to the design and safety cases are applied, as well to facilitate the public perception and understanding of the safety of the disposal approaches relative to risks that may increase over long times frames in the absence of a successful implementation of final dispositioning.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

International Collaboration Activities on Disposal in Argillite R&D: DECOVALEX19 Task C – Geochemical and Reactive Transport Modeling

International collaborations on nuclear waste disposal R&D are an integral part of the Spent Fuel Waste Science and Technology (SFWST) campaign within the DOE Fuel Cycle and Technology (FCT) program. These partnerships with international repository R&D programs provide key opportunities to participate in experiments developing laboratory/field data (underground research laboratories (URL)) of engineered barrier system (EBS) interactions (e.g., near-field) and characterization of transport phenomena in the host rock (e.g., far-field). The results of these experiments are used in the evaluation of coupled processes and their representation via state-of-the-art simulation approaches to evaluate repository performance. During the thermal heating period, increases in temperature from radionuclide decay in the spent fuel (SF) waste canisters will increase temperature in the surrounding EBS driving chemical and transport processes in the near- and far-field domains of the repository. URL heater-tests for extended periods of times (e.g., years) provide key information and data on thermal effects affecting engineered barriers in response to temperature and water saturation levels. Groundwater interactions with cementitious barriers are also important to in-drift chemistry and EBS performance during post-closure. Descriptions of the various URL experiments for various disposal design concepts according to the host country repository program and relevance to the US program is given elsewhere (Birkholzer et al.,2017;Jové Col& et al., 2016). The DECOVALEX-2019 Task C involves collaboration with the GREET (Groundwater REcovery Experiment in Tunnel) at the Mizunami URL, Japan ,which targets the development of monitoring methodologies of groundwater in granitic rock with applications to THMC simulations (Iwatsuki et al., 2005;Iwatsuki et al.,2015,2017). Some of the goals of GREET is to conduct a facility-scale geochemical characterization study of short- and long-term effects of tunnel excavation activities, impacts on groundwater flow and transport, and influences on groundwater chemistry (Iwatsuki et al.,2015). The data obtained from these URL activities is then used in the development and evaluation of THC models to support post-closure safety and performance assessments of the repository environment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Investigations of Dual-Purpose Canister Direct Disposal Feasibility (FY14)

Results reported here continue to support the FY13 conclusion that direct disposal of DPCs is technically feasible, at least for some DPCs, and for some disposal concepts (geologic host media). Much of the work performed has reached a point where site-specific information would be needed for further resolution. Several activities in FY14 have focused on clay/shale media because of potential complications resulting from low thermal conductivity, limited temperature tolerance, and the need to construct hundreds of kilometers of emplacement drifts that will remain stable for at least 50 years. Technologies for rapid excavation and liner installation have significantly advanced in the past 20 years. Tunnel boring machines are the clear choice for large-scale excavation. The first TBM excavations, including some constructed in clay or shale media, are now approaching 50 years of service. Open-type TBMs are a good choice but the repository host formation would need to have sufficient compressive strength for the excavation face to be self-supporting. One way to improve the strength-stress relationship is to reduce the repository depth in soft formations (e.g., 300 m depth). The fastest construction appears to be possible using TBMs with a single-pass liner made of pre-fabricated concrete segments. Major projects have been constructed with prefabricated segmented liner systems, and with cast-in-place concrete liners. Cost comparisons show that differences in project management and financing may be larger cost factors than the choice of liner systems. Costs for large-scale excavation and construction in clay/shale media vary widely but can probably be limited to $10,000 per linear meter, which is similar to previous estimates for repository construction. Concepts for disposal of DPC-based waste packages in clay/shale media are associated with thermal management challenges because of the relatively low thermal conductivity and limited temperature tolerance. Peak temperature limits of 100°C or lower for clay-rich materials have been selected by some international programs, but a limit above 100°C could help to shorten the duration of surface decay storage and repository ventilation. The effects of locally higher peak temperatures on repository performance need to be evaluated (in addition to the effects at lower temperatures). This report describes a modeling approach that couples the TOUGH2 and FLAC3D codes to represent thermally driven THM processes, as a demonstration of the types of models needed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Cost estimation inputs for spent nuclear fuel geologic disposal concepts (Revision 1)

A set of 16 geologic disposal concepts is described in sufficient detail for rough-order-of-magnitude repository cost estimates, for disposal of spent nuclear reactor fuel in generic crystalline, argillaceous, and salt host geologic media. The description includes total length, diameter, and volume for all underground shafts, ramps, drifts and large-diameter borings. Basic types of ground support are specified. Total repository capacity is assumed to be approximately 140,000 MT of spent fuel, but concepts are described in terms of modular panels each containing 10,000 MT. Waste packaging is described, and the materials and outer dimensions for disposal overpacks are given. The manner of emplacement is specified, with any additional fixturing, lining, buffer materials, and backfill needed. Thermal limits for waste package emplacement or repository closure are given, as appropriate for enclosed and open emplacement modes, respectively. The 16 disposal concepts are based on disposal concept studies performed for the Used Fuel Disposition R&D program between 2011 and 2014, and the accompanying engineering analyses. Revision 1 results from review and checking of repository spacings, drift length estimates, repository plan area estimates, closure power limits, and estimates of fuel age at closure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Preliminary Analysis of Postclosure DPC Criticality Consequences

One of the objectives of the United States (U.S.) Department of Energy's (DOE) Office of Nuclear Energy's Spent Fuel and Waste Science and Technology Campaign is to better understand the technical basis, risks, and uncertainty associated with the safe and secure disposition of spent nuclear fuel (SNF) and high-level radioactive waste. Commercial nuclear power generation in the U.S. has resulted in thousands of metric tons of SNF, the disposal of which is the responsibility of the DOE (Nuclear Waste Policy Act 1982). Any repository licensed to dispose the SNF must meet requirements regarding the longterm performance of that repository. For an evaluation of the long-term performance of the repository, one of the events that may need to be considered is the SNF achieving a critical configuration. Of particular interest is the potential behavior of SNF in dual-purpose canisters (DPCs), which are currently being used to store and transport SNF but were not designed for permanent geologic disposal. A two-phase study has been initiated to begin examining the potential consequences, with respect to longterm repository performance, of criticality events that might occur during the postclosure period in a hypothetical repository containing DPCs. Phase I, a scoping phase, consisted of developing an approach intended to be a starting point for the development of the modeling tools and techniques that may eventually be required either to exclude criticality from or to include criticality in a performance assessment (PA) as appropriate; Phase I is documented in Price et al. (2019). The Phase I approach guided the analyses and simulations done in Phase II to further the development of these modeling tools and techniques as well as the overall knowledge base. The purpose of this report is to document the results of the analyses conducted during Phase II. The remainder of Section 1 presents the background, objective, and scope of this report, as well as the relevant key assumptions used in the Phase II analyses and simulations. Subsequent sections discuss the analyses that were conducted (Section 2), the results of those analyses (Section 3), and the summary and conclusions (Section 4). This report fulfills the Spent Fuel and Waste Science and Technology Campaign deliverable M2SF-20SN010305061.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Postclosure Criticality Consequence Analysis - Scoping Phase

Commercial generation of energy via nuclear power plants in the United States (U.S.) has generated thousands of metric tons of spent nuclear fuel (SNF), the disposal of which is the responsibility of the U.S. Department of Energy (DOE) (Nuclear Waste Policy Act of 1982). Any repository licensed to dispose of the SNF must meet requirements regarding the long-term performance of the repository. In evaluating the long-term performance of the repository, one of the events that may need to be considered is the SNF achieving a critical configuration. Of particular interest is the potential behavior of SNF in dual-purpose canisters (DPCs), which are currently being used to store the SNF but were not designed for permanent disposal. As part of a multiyear plan that is currently being developed for the DOE, a two-phase study has been initiated to examine the potential consequences, with respect to long-term repository performance, of criticality events that might occur during the postclosure period in a hypothetical repository containing DPCs. Phase I, a scoping phase, consists of generating an approach intended to be a starting point for the development of the modeling tools and techniques that may eventually be required either to exclude criticality from or include criticality in a performance assessment (PA) as appropriate. The Phase I approach will be used to guide the analyses and simulations done in Phase II to further the development of these modeling tools and techniques as well as the overall knowledge base. The purpose of this report is to document the approach created during Phase I. The study discussed herein focuses on the consequences of criticality in a DPC; it does not address the probability of occurrence of a criticality event. This approach examines two types of criticality events for SNF disposed of in a single type of DPC: a steady-state criticality and a transient criticality. The steady-state critical event is characterized by a relatively low constant power output over 10,000 years, while the transient critical event is characterized by a power spike that lasts on the order of seconds. Possible effects of the criticality are an increase in the radionuclide inventory; an increase in temperature; and a change in the chemistry inside the waste package, along with a change in radionuclide solubilities, fuel degradation rates, and steel corrosion rates. Additionally, for transient criticality the possibility of mechanical damage to the engineered and natural barriers also exists.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Development of the IES Plug-and-Play Framework

This report discusses the status of the flexible plug-and-play framework development currently ongoing that aims to integrate Modelica/Dymola with the Risk Analysis and Virtual ENvironment (RAVEN) software in terms of both Functional Mock-Up Interface (FMI)/Functional Mock-Up Unit (FMU) construction and repository structures that aim to ease the sharing and simulation of complex dynamic models. This report aims to provide an overview of all the performed activities resolving around the deployment of methods, software infrastructures, guidelines and workflow for the construction and usage of models, encapsulated using the FMI/FMU protocols and standards. In particular, the report is organized in three main macro-subjects, which are connected to each other: - FMI/FMU adaptors for modelica models - HYBRID repository new structure and open-source deployment - RAVEN FMI/FMU exporting capabilities and Artificial Intelligence (AI)-based analysis acceleration. The first part of the report discusses the FMI/FMU adaptors that have been created within the HYBRID repository to allow users to quickly export models, such as FMUs. Several examples are shown that highlight the step-by-step process of converting an existing Modelica model into an FMU for use within the Dymola platform. Simulation results demonstrate that, while minor differences may occur, the overall control, trends, and solution integrity are maintained between standard Modelica simulation and FMU simulation results. However, it is worth noting that, for small systems, the FMU results have a slower simulation time than the Modelica only simulation. Using this process, a company can provide models that contain proprietary information to entities without disclosing any of the information about the model that could be considered business sensitive. Such an ability would allow institutions to bypass the necessity of “whitewashing” data. In the second part of the report, the new structure of the HYBRID repository is discussed with a major focus on the series of updates that has been completed. These updates include the addition of Modelica system-level regression tests and software quality assurance documentation that ensure that modifications to the Modelica models do not alter system-level model results. The third and final part of the report aims to report the work that has been performed for the deployment of methods and workflows for the construction of RAVEN AI-based models compliant with the FMI/FMU standard. Such development represents the key for the deployment of the concept of “Flexible ecosystem” since it allows for the replacement of high-fidelity modelica models (or any other FMI/FMU compliant model) with RAVEN generated AI surrogate models. Overall, extensive work has been completed on developing FMUs and FMIs from existing models, understanding the requirements and limitations of FMUs, and open-sourcing the HYBRID repository with an integrated regression system.

42 ENGINEERING↗

DOE OSTI2021

Development of the IES Plug-and-Play Framework

14 SOLAR ENERGY↗

DOE OSTI2022

Post-closure Nuclear Criticality Safety Evaluations for Disposition of Criticality Control Overpacks at the Waste Isolation Pilot Plant

The Waste Isolation Pilot Plant (WIPP) is a geological repository in southern New Mexico that provides for disposal of transuranic (TRU) wastes from atomic energy defense activities. The Sandia National Laboratories (Sandia) Report, Consideration of Nuclear Criticality When Disposing of Transuranic Waste at the Waste Isolation Pilot Plant, addresses nuclear criticality safety based on the projected inventory characteristics for the initial compliance certification application of WIPP in 1996. As the inventory, waste forms, and disposal package designs change, revised or new analyses are necessary to demonstrate acceptability for these configurations within the WIPP safety basis and compliance with 10,000-year post-closure standards of the US Environmental Protection Agency (EPA). Saylor and Scaglione evaluated criticality control overpacks (CCOs) in 2017 based on conservative assumptions for post-closure repository structural conditions with resulting effects on containers and container spacing, The Saylor and Scaglione evaluation of CCOs addressed a single waste configuration that represents the Surplus Plutonium Disposition Program’s dilute and dispose waste form and composition. This initial CCO study demonstrated that 50 grams of boron carbide (B 4 C) per CCO is sufficient to ensure post-closure criticality safety based on a well-mixed waste composition, and Oak Ridge National Laboratory (ORNL) subsequently determined that this amount of B 4 C does not require constraints on moisture or plastic present as moderator. The Saylor and Scaglione analysis conservatively assumes repository room closure that eliminates all space between fissile gram equivalent (FGE) 239 Pu masses. The close-packed array was selected based on limited availability of repository salt creep modeling results at that time. In 2019, Brickner provided additional evaluations for pipe overpack containers (POCs), building on the conservative basis provided by Saylor and Scaglione. Brickner’s 2019 analysis made use of new geomechanical data for post-closure spacing that rely on advances in repository modeling as documented in the work by Reedlunn and Bean. This current CCO evaluation for generic waste materials expands on earlier work performed at ORNL and includes evaluation of CCOs across a much broader range of possible waste compositions and geometries. This evaluation is intended to provide input for the required feature, event and process (FEP) screening to determine if post-closure criticality must be included as an event in the 10,000-year regulatory evaluation. As such, the approach to modeling post-closure criticality presented in this report has been coordinated with the Sandia team responsible for FEP screening. The resulting analysis supports disposition of fissile materials in the CCO containing up to 380 FGE 239 Pu and expands conditions acceptable for disposal of fissile material in CCOs. This evaluation builds on the methodology of Saylor and Scaglione and Brickner, using the most recently available geomechanical data for CCO spacing under salt creep compaction scenarios provided by Reedlunn and Bean. The broad range of fissile material configurations analyzed in this report are intended to account for configurations that may occur during the post-closure disposal time period, and it also includes waste configurations that are not physically possible to support analysis of conditions that influence neutron fluence.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

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 OSTI2022

Incorporating Heterogeneity into 3D Geologic Models for the Alluvial Basin Reference Case

The unsaturated zone alluvium reference case is one of several geologic systems under consideration by the U.S. Department of Energy Office of Nuclear Energy for hosting repositories for spent nuclear fuel and associated waste (Sevougian et al., 2019). As noted by Mariner et al. (2018), the generic alluvial basin offers positive attributes that merit its consideration as a reference case by the Spent Fuel and Waste Science and Technology (SFWST) campaign. Hundreds of alluvial basins and sub-basins are scattered across the arid western United States (Figure 1-1). Precipitation and infiltration rates are relatively low with high evapotranspiration, resulting in deep regional water tables. Deep regional water tables allow for significant vertical separation between proposed repository depths and groundwater and thus longer transport paths from a hypothetical repository to an aquifer. Accumulations of alluvial sediments within these basins are typically on the order of hundreds of meters, and locally may exceed 1,000 m in thickness, as is the case for the Deming sub-basin in southern New Mexico. In alluvial basins, alluvium creates a thick geologic host medium that serves as the natural barrier system (NBS) in the conceptual model framework of a geologic disposal system and isolates the waste packages from receptors in the biosphere. Further, alluvial basin fill is typically comprised of stacked playa and lacustrine deposits along the basin axis (Perry et al., 2018). Characterized by low permeability, these layers protect the biosphere above the repository and the groundwater resources below the repository by limiting flow and transport of radionuclides.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Improved Creep Testing Approach for Bentonite EBS

In the United States, the nuclear reactor contributes 18.7 percent of the nation’s total electricity generation and produces a significant amount of spent nuclear fuel (SNF) and high-level waste. Because SNF is being stored for longer periods than initially envisioned, the U.S. Department of Energy Office of Nuclear Energy, Office of Spent Fuel and Waste Science and Technology is assessing the technical performance of the SNF storage systems after extended durations. The concept of a sustainable geologic repository for SNF disposal relies heavily on the safety provided by multiple barriers, such as repository host rock, overlying rock formation, and human-made engineered barrier systems (EBS). One of the major components of EBS is bentonite-based buffer materials that isolate the nuclear-waste canisters from the host rock and fill the void left in the horizontally drilled boreholes. The resaturated bentonite has a swelling characteristic, which is useful for self-sealing the microcracks in the host rock and supporting the canister’s heavy weight. However, factors such as nonuniformity in the saturation, swelling pressure of bentonite, and the high temperature at the canister-bentonite interface may lead to instability of the waste canister inside the borehole. Therefore, evaluating the long-term deformation of bentonite-based material is necessary. RESPEC Company, LLC recently completed a preliminary study for the U.S. Department of Energy (DE SC0022804) that is attempting to understand the time-dependent deformation in the consolidated sand/bentonite (SB) specimens through triaxial creep experiments and improve the understanding of EBS performance under anticipated repository conditions. The project included developing a standard testing procedure for preparing the consolidated core specimen from a mixture of sand and bentonite in a 50:50 ratio by weight, fabricating a new pressure vessel in-house, modifying existing creep equipment to perform triaxial creep experiments on consolidated SB specimens, and performing six long-term triaxial creep experiments under low-deviatoric stress at two different temperature conditions, which is representative of hypothetical conditions that might be encountered in a geologic repository within a reasonable rate of success in completing creep experiments. The consolidated SB specimen was prepared using the novel specimen preparation procedure and had physical characteristics (e.g., moisture content and bulk density) reasonably comparable with the properties of bentonite-based buffer, which is proposed to be used in the EBS in the Swedish KBS-3H design and the Swiss design of SNF disposal in the horizontal drifts. The newly fabricated pressure vessel could withstand the confining pressure of up to 15 megapascals (MPa) while maintaining the test temperature of up to 90 degrees Celsius (°C) for long-term creep experiments. The preliminary results from the triaxial creep experiments suggested that the consolidated SB specimen may experience time-dependent deformation at a low-deviatoric stress state and room temperature condition. Based on the deformation recorded in the SB specimen from axial and radial linear variable differential transformers (LVDTs) and physical characteristic changes, it is apparent that the time-dependent deformation is a combination of consolidation and creep, which is influenced by the level of deviatoric stresses and temperature. The systematic creep testing of bentonite-based specimens helped develop a standard testing procedure applicable for analyzing the similar characteristics of shale or clay-like materials in different geologic conditions outside repository science, including in civil engineering and infrastructure projects, underground and open pit mining, and deep drilling.

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