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DOE OSTI2023

Joint Geophysical and Numerical Insights of the Coupled Thermal‐Hydro‐Mechanical Processes During Heating in Salt

Abstract Salt offers an optimal medium for the permanent isolation of heat‐producing radioactive waste due to its impermeability, high thermal conductivity, and ability to close fractures through creep. A thorough understanding of the thermal‐hydrological‐mechanical (THM) processes, encompassing brine migration, is fundamental for secure radioactive waste disposal within salt formations. At the Waste Isolation Pilot Plant (WIPP), we conducted joint in situ geophysical monitoring experiments during active heating to investigate brine migration near excavations. This experiment incorporated electrical resistivity tomography (ERT) alongside high‐resolution fiber‐optic‐based distributed temperature sensing within a controlled heating experiment. Additionally, discrete element model (DEM) based numerical simulations were conducted to simulate THM processes during heating, providing a more mechanistic understanding of the coupled processes leading to the observed changes in geophysical measurements. During heating, resistivity shifts near the heater were reasonably explained by temperature effects. However, in more distant, cooler regions, the resistivity decrease exceeded predictions based solely on temperature. DEM simulations highlighted brine migration, propelled by pore pressure gradients, as the likely primary factor contributing to the additional resistivity decline beyond temperature effects. The comparison between the predicted ERT responses and observations was much improved when considering the effects of brine migration based on the DEM simulations. These geophysical and simulation findings shed light on brine migration in response to salt heating, enhancing our understanding of the coupled THM processes in salt for safe radioactive waste disposal.

58 GEOSCIENCES↗
DOE OSTI2025

Past Approaches for Spent Nuclear Fuel, Transuranic, and High-Level Waste Disposal in the United States—Part 2: Siting Process, Staged Development, and Public Preferences

This report presents pertinent aspects of the ~50-year United States experience in siting a mined geologic disposal repository for spent nuclear fuel (SNF), transuranic (TRU) waste, and high-level radioactive waste (HLW) as related to site selection and the staged process for site investigations as specified in the Nuclear Waste Policy Act of 1982 and generic and site-specific regulations of the US Department of Energy (DOE), US Environmental Protection Agency (EPA), and US Nuclear Regulatory Commission (NRC). The roles of the Environmental Impact Statement and guidance in international consensus standards by the International Atomic Energy Agency are also mentioned. The focus is on siting and developing the Waste Isolation Pilot Plant, an operating repository for TRU waste from atomic energy defense activities, and the proposed Yucca Mountain repository for commercial SNF and HLW. In the social dimension, the role of institutional stakeholders is described. Past national surveys related to waste management options for storage and disposal provide insight on public preferences of other stakeholders. The descriptions are intended to help other countries more fully understand the stages adopted for siting and developing repositories in the United States.

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DOE OSTI2020

Technical Issues that Need to Be Addressed in Preparing a Large Program to Transport SNF and HLW - 20093

The U.S. Nuclear Waste Technical Review Board (Board) has been interested for many years in the Department of Energy (DOE) activities related to transporting spent nuclear fuel (SNF) and high-level radioactive waste (HLW). The Board has conducted several public meetings and issued associated reports on this topic. In September 2019, the Board issued a report entitled Preparing for Nuclear Waste Transportation-Technical Issues that Need to Be Addressed in Preparing for a Nationwide Effort to Transport Spent Nuclear Fuel and High-Level Radioactive Waste [1] that focuses on the issues DOE will need to address to plan and implement an integrated transportation program. In its report, the Board describes 30 broad technical issues that DOE needs to address and offers three sets of findings and recommendations to DOE. (authors)

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DOE OSTI2020

Selection of Vadose Zone Flow and Transport Properties with Gravel Fraction Corrections for the Hanford Site Composite Analysis and Cumulative Impact Evaluation

This environmental calculation file (ECF) is a compilation of hydrologic and radiological/chemical properties to be used for the updated Hanford Site Composite Analysis (CA) and Cumulative Impact Evaluation (CIE) vadose zone (VZ) modeling. The CA will provide an all-pathways dose projection to a hypothetical future member of the public from all planned low-level radioactive waste disposal facilities and potential contributions from all other projected end-state sources of radioactive material left at the Hanford Site following site closure. Its primary purpose is to support the decision-making process of the U.S. Department of Energy (DOE) under DOE O 435.1 Chg 1, Radioactive Waste Management, related to managing low-level waste disposal facilities at the Hanford Site. The CIE evaluates the effects of cleanup decisions regarding groundwater quality in the Hanford Site Central Plateau (DOE/RL-2018-69, Cumulative Impact Evaluation Technical Approach Document). Due to the complexity and large number of waste sites in source operable units (OUs), the computational tools used for the CIE must be capable of representing a range of site conditions and source terms in the VZ while also efficiently computing the impact that cleanup decisions have on the underlying aquifer. Waste-site proximity between and within source OUs has resulted in contaminants commingling in the vadose and saturated zones in complex ways. Plume commingling requires cleanup decisions to be evaluated considering the surrounding waste sites and existing groundwater contamination, therefore demonstrating the need to evaluate cumulative impacts from the VZ to groundwater.

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DOE OSTI2021

Proton Nuclear Magnetic Resonance (H NMR) of Glycolate in Real Waste: Developing and Testing Analytical Methods for the Savannah River Site Liquid Waste System

In preparation for implementing the NG acid flowsheet for the SRS LWS, analytical methods for determining glycolate at low concentration are desired to support system management and safety. Previous work documented that the IC method performed well for samples with low to moderate ionic strength. That work also included a scoping effort to determine if an alternate analytical strategy using H NMR would complement and extend the capabilities of the IC method. The H NMR scoping results indicated that the method had the potential to expand glycolate analysis in LWS samples to higher ionic strength tanks/solutions. The scope of this work is to develop innovative proton NMR techniques, including demonstrating ion exchange decontamination protocols needed for application of the technique to real waste samples. Part of radioactive waste processing at SRS uses formic acid to reduce oxidized (Hg 2+ ) to more volatile elemental Hg for steam striping, collecting, and disposal. Under acidic conditions found in the Chemical Processing Cell (CPC) at the Defense Waste Processing Facility (DWPF), formic acid has a much higher hydrogen generation rate than an alternative reductant, glycolic acid. Thus, an NG acid flowsheet has been developed utilizing glycolic acid with the benefit of easing the need for headspace monitoring requirements for hydrogen and ammonia at DWPF. Low concentrations of glycolate are conservatively assumed to be in the recycle stream, which will collect in the Recycle Collection Tank (RCT). The DWPF recycle stream collected in the RCT has a distinct pathway to the LWS waste tanks that feed the 2H and 3H Evaporator. This route involves transfer of the DWPF recycle to Tank 22 in the Concentration, Storage, and Transfer Facilities (CSTF) followed by transfer to the LWS tank farm/evaporators.

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DOE OSTI2026

Developing a Process for Collaboration-Based Siting of a Federal Consolidated Interim Storage Facility in the United States: Learning from Communities Living with Legacy Waste

In June of 2023, the U.S. Department of Energy (DOE), Office of Nuclear Energy (NE) announced the selection of its Collaboration-Based Siting (CBS) Consortia, a group of 12 awardees, including the Consortium for Risk Evaluation with Stakeholder Participation (CRESP) (led by Vanderbilt University) to assist DOE-NE with the development of its process for siting a federal consolidated interim storage facility (FCISF) for spent nuclear fuel (SNF) storage. At this time, DOE NE is not soliciting interested host communities, rather the CBS Consortia are tasked with in-depth engagement, mutual learning, and capacity building to provide DOE NE with feedback on the CBS process. CRESP’s objective is to engage communities in two regions (the Pacific Northwest and the Southeast) with sites currently storing defense- and research-related SNF to foster learning concerning the best and worst practices in community participation in risk-informed decision making. This includes learning from existing structures for public input in radioactive waste management decision making [e.g. citizen advisory boards (CABs)] and other local parties about how to build and sustain trust among the parties. CRESP has been working to engage stakeholders and Tribes surrounding two DOE sites historically differing in receptiveness to engaging with DOE and trusting in DOE to accomplish its missions. CRESP’s approach to date has consisted of (1) engaging voluntarily members and former members of the DOE Office of Environmental Management’s CABs to develop a mutual understanding of their perspectives, values, and experiences related to risk and participatory decision making; (2) engaging members of those communities that would likely be part of any radioactive waste management discussions and who may provide valuable feedback for the development of the CBS process; and in the longer term, (3) engaging communities to foster knowledge sharing on understanding and definitions of risk and how risk is factored into community decision making. Our emphasis is to identify opportunities for improving risk communication frameworks, strategies, and decision making. The selection of a future FCISF site is likely to result in the creation of a structure similar to a CAB composed of members of the local community. We anticipate that these insights can help DOE NE learn from existing participatory risk communication structures in place at sites storing defenseor research-related SNF to understand best practices for fostering enduring and participatory relationships with future CABs. Within this paper, we (1) describe CRESP’s overall approach to assisting DOE NE with maturing the CBS process; (2) present a summary of preliminary phase 1 project results, including the development of a body of knowledge (describing available resources to support community engagement, risk communica tion, and participatory decision making) and community ecosystem information (leveraging demographic, social, economic, and environmental attribute mapping and advanced sentiment analysis of social media data); and (3) based on these results, provide observations for future research opportunities to support the development of CBS processes for radioactive waste management facilities, generally.

collaboration-based siting↗
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

Implementation of the Glycolate Destruction Process with Sodium Permanganate in the Defense Waste Processing Facility (DWPF) - 20141

The Savannah River Site (SRS) seeks to replace formic acid with glycolic acid as a chemical reductant in the Defense Waste Processing Facility (DWPF), where borosilicate glass is mixed with high-level radioactive waste, melted at high temperatures, then poured into stainless steel canisters for safe storage. Prior research and development have demonstrated the feasibility and advantages of the new nitric-glycolic acid flowsheet over the current nitric-formic acid flowsheet to chemically adjust the radioactive sludge slurry waste in the Sludge Receipt and Adjustment Tank (SRAT) prior to vitrification. Use of glycolic acid reduces hydrogen and ammonia generation during the reduction process, thus reducing flammability hazards and purge requirements in the DWPF processing vessels. In addition, glycolic acid will support the new Salt Waste Processing Facility (SWPF) by allowing for receipt of higher volumes of waste due to reduction of flammable gases, increased boil-up rates in the vessels, and easing the processing of future sludge batches containing high levels of mercury. However, during DWPF operations, trace amounts of glycolate are anticipated to be entrained in the recycle stream and sent to the Concentration, Storage, and Transfer Facilities (CSTF) via the Recycle Collection Tank (RCT). The RCT receives condensate from the Slurry Mix Evaporator Condensate Tank (SMECT) and Off-gas Condensate Tank (OGCT) that may contain unreacted glycolate. A literature review found the potential for hydrogen generation due to thermolysis of glycolate in caustic CSTF conditions. Savannah River National Laboratory (SRNL) conducted several tests on simulant and radioactive waste from various tanks at SRS and confirmed that glycolate added to waste at concentrations of approximately 1000 mg/L and higher generates hydrogen through thermolysis. There are potential operating scenarios in the future where glycolate concentration could exceed 1000 mg/L in the 2H Evaporator Drop Tank, the highest accumulation point for DWPF organics received via recycle. These findings presented a need to develop another process to mitigate the impact of glycolic acid to the CSTF. Several DWPF compatible options were evaluated for their ability to destroy glycolate in the DWPF recycle stream under various processing conditions, thereby mitigating its introduction to the CSTF. Downselection testing identified sodium permanganate as the most promising and simplest option to oxidize glycolate. It has been demonstrated to be effective in various RCT simulant compositions, operating temperatures, oxidant addition rates, solution pH, and initial glycolate concentrations. A feasibility study identified existing tanks within the facility that could be used for receiving and storing the commercially available sodium permanganate, as well as feeding it to the RCT for processing. The use of existing equipment as well as the ability to purchase the pre-mixed chemical reduces implementation and operational complexity. The downstream impacts of the sodium permanganate reaction were also evaluated and demonstrated to have minimal impact on other chemical species present in the RCT and CSTF, as well as on the materials of construction of the RCT and CSTF vessels and associated piping and instrumentation. Future work is in progress to validate the glycolate destruction process with testing in actual radioactive RCT waste before implementation in DWPF. (authors)

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DOE OSTI2020

Composite Analysis Solid Waste Release Data Reduction of Activity Flux from Waste Sites to the Vadose Zone for Baseline Assessment

This environmental calculation file (ECF) describes the data reduction methodology for the solid waste release datasets generated by the Composite Analysis Solid Waste Release (CASWR) model (CP-62766, Model Package Report: Composite Analysis Solid Waste Release Model [CASWR Model]) and documents the resulting reduced datasets for the updated Composite Analysis (CA) baseline assessment. This ECF is in support of the updated Hanford Site CA. The purpose of the updated Hanford Site CA is to provide an estimate of the cumulative radiological impacts from active and planned low-level radioactive waste disposals and other potentially interacting radioactive waste disposal sources that will remain following Hanford Site closure. The vadose zone modeling activities associated with the CA includes solid waste forms as contaminant sources. To support the modeling of solid waste transport from the vadose zone to the groundwater at the Hanford Site, the CASWR model was designed to generate deterministic radionuclide release rates for the Hanford Site Central Plateau solid waste disposal sites. The complete CASWR model-generated dataset for the CA Baseline Assessment consists of 2,378,831 sets of time and release rates (1,073,869 sets for the 200 East Area and 1,304,962 sets for the 200 West Area), representing a course of the simulation from year 1953 to year 12,069, 16 radiological contaminants of potential concern (COPCs), and 237 solid waste sources as documented in ECFHANFORD- 19-0112, Solid Waste Release Calculations for the Composite Analysis Baseline Assessment. For several of the STOMP vadose zone models, the number of release rate data pairs required for release at all waste sites would exceed the limitations imposed by the STOMP modelling software package used to simulate contaminant transport through the vadose zone. As a result, a data reduction of the CASWR dataset is required prior to its use as a source input for the CA vadose zone modeling activities.

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DOE OSTI2024

RANGERS: Methodology Report on Design and Performance Assessment of Engineered Barrier Systems in a Salt Repository for HLW/SNF

Salt formations are one of the potential host rocks for the final disposal of high-level radioactive waste (HLW) in deep geological repositories, both in Germany and the United States. The safe isolation of radioactive waste in these repositories relies on a multi-barrier system, combining engineered and natural barriers. The natural barrier is provided by the salt rock itself, known for its self-sealing properties and long-term stability. The engineered barrier, on the other hand, comprises sealing components strategically placed within the repository to enhance its containment capabilities. In both Germany and the United States, long-term safety assessments require demonstrating the integrity of the natural barrier for a period of up to 1 million years. Concurrently, the engineered barrier system (EBS) must maintain its structural and functional integrity until the long-term sealing, such as the granular salt backfill material, has re-consolidated to its final low porosity and permeability. Based on extensive expertise and experience with engineered barriers in salt formations, BGE TECHNOLOGY GmbH and Sandia National Laboratories have partnered to develop a robust methodology for the integrity and performance assessment of EBS in HLW repositories through the RANGERS project. This collaborative effort aims to establish a unified approach to geotechnical engineering, repository design, integrity and performance evaluation of EBS in salt repositories.

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DOE OSTI2020

On Temperature Limits for the Outer Surface of a Container in a Disposal Facility - 20313

The temperature in a disposal facility for high-level radioactive waste (HLRW) is an issue for safety. When the recommencement of the search for and selection of a site for a disposal facility for HLRW in Germany was stipulated by the Site Selection Act (StandAG 2017) in 2017, a precautionary temperature limit of 100 deg. C on the outer surface of the containers with high-level radioactive waste in the disposal facility section was set. This precautionary temperature limit shall be applied in preliminary safety analyses provided that the 'maximum physically possible temperatures' in the respective host rocks have not yet been determined due to pending research. Therefore, this issue is addressed and discussed in this paper, contributing to 'pending research' by a review of the literature. This paper briefly discusses a few examples of thermohydraulic, mechanical, chemical and biological processes in a disposal facility, because temperature limits are derived based on safety impacts regarding THMCB-processes. The temperature-dependent processes have been extracted from databases for features, events and processes (FEP-databases). Furthermore, it is discussed if the feasibility to retrieve and recover HLRW is hampered at high temperatures. A design temperature concerning single components of a disposal facility for the preservation of their features can be derived when a safety concept is established. However, the interactions of all relevant processes in a disposal concept must be considered to determine a specific temperature limit for the outer surface of the containers. Therefore, applicable temperature limits may vary for particular safety and disposal concepts in the following host rocks: rock salt, clay stone and crystalline rock. Technical solutions for retrieval and design options for recovery seem to be viable up to temperatures of 200 deg. C with different, sometimes severe, downsides according to expert judgement. Temperature limits regarding the outer surface of the containers can be derived specifically for each safety concept and design of the disposal facility in a host rock. General temperature limits without reference to specific safety concepts or the particular design of the disposal facility may narrow down the possibilities for optimisation of the disposal facility and could adversely affect the site selection process in finding the best suitable site. (authors)

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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 OSTI2026

Performance Assessment of the Inactive Trenches in the 200 West Low-Level Burial Grounds at the Hanford Site

The purpose of this document is to provide a performance assessment (PA) of the 200 West Area Low‑Level Burial Grounds (LLBGs), under the requirements of DOE O 435.1, Radioactive Waste Management1. The objective is to ensure that the radioactive waste is managed in a manner that is protective of human health and the environment. Figure ES-1 shows the location of the LLBGs in the 200 West and 200 East Areas of the Hanford Site.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2022

Hanford Site Composite Analysis: Radionuclide Selection for Groundwater Pathway Evaluation

The updated Hanford Site Composite Analysis (CA) (DOE/RL-2019-52, Composite Analysis for Low-Level Waste Disposal in the Hanford Site Central Plateau (FY 2020)) provides an all-pathways dose projection to a hypothetical future member of the public from all planned low-level radioactive waste disposal facilities and potential contributions from all other projected end-state sources of radioactive material left at the Hanford Site following site closure. Its primary purpose is to support the decision-making process of the U.S. Department of Energy (DOE) under DOE O 435.1-1, Radioactive Waste Management, related to managing low-level waste disposal facilities at the Hanford Site. A key aspect of conducting a CA is selecting the radionuclides to be analyzed. This document describes the selection process for radionuclides to be included in the quantitative analysis.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

Results of Re-evaluation of FEPs Related to Higher Fissile Content in HLW Glass at SRS

One of the objectives of the United States Department of Energy Office of Nuclear Energy’s Office of Spent Fuel and High-Level Waste Disposition is to better understand the technical bases, risks, and uncertainties associated with the safe and secure disposition of spent nuclear fuel and high-level radioactive waste. Domestic defense and research activities have generated a few thousand metric tons of spent nuclear fuel and high-level radioactive waste, much of which has been or will be processed and vitrified into high-level waste glass. The Nuclear Waste Policy Act 1982 makes the Department of Energy responsible for disposal of these materials.

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DOE OSTI2020

Review of the Safety Analyses of the Konrad Repository - 20530

The Bundesgesellschaft fuer Endlagerung (BGE) is the license holder of all geological repository sites in Germany since 2017. Besides the construction of the Konrad repository for the emplacement of low- and intermediate-level waste (LILW), the operation of the Asse II mine including the retrieval of the emplaced waste, and the preparation for the closure of the Morsleben repository, BGE is responsible to carry out the operative tasks of selecting a site for a high-level waste (HLW) repository. The Konrad repository is Germany's first repository for radioactive waste with negligible heat generation, which was fully licensed under atomic law within the scope of a plan approval procedure from the original application in 1982 until the license was granted in 2002 and finally confirmed by the Federal Administrative Court in 2007. The emplacement of LILW was licensed up to a volume of 303,000 m{sup 3} with a total β/γ-activity of 5 x 10{sup 18} Bq and a total α-activity of 1.5 x 10{sup 17} Bq. Currently, the former iron ore mine Konrad is being converted to a repository and is expected to come into operation in 2027. But what if there is no safety case in place nor a regulatory framework tailored for a change of documents in order to ensure the safety according to the state-of-the-art of science and technology during the evolution of a long-term project such as a geological repository? In this presentation, a review procedure of the safety analyses, that have been the basis of the plan approval decision for the Konrad repository at that time, will be described. Furthermore, the first results of this graduated review procedure for updating the safety analyses will be presented. (authors)

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DOE OSTI2021

Sulfidation of magnetite with incorporated uranium

Uranium (U) is a radionuclide of key environmental interest due its abundance by mass within radioactive waste and presence in contaminated land scenarios. Ubiquitously present iron (oxyhydr)oxide mineral phases, such as (nano)magnetite, have been identified as candidates for immobilisation of U via incorporation into the mineral structure. Studies of how biogeochemicalprocesses, such as sulfidation from the presence of sulfate-reducing bacteria, may affect iron (oxyhydr)oxides and impact radionuclide mobility are important in order to underpin geological disposal of radioactive waste and manage radioactively contaminated land. Here, this study utilised a highly controlled abiotic method for sulfidation of U(V) incorporated into nanomagnetite to determine the fate and speciation of U. Upon sulfidation, transient release of U into solution occurred (~8.6 % total U) for up to 3 days, despite the highly reducing conditions. As the system evolved, lepidocrocite was observed to form over a period of days to weeks. After 10 months, XAS and geochemical data showed all U was partitioned to the solid phase, as both nanoparticulate uraninite (U(IV)O2) and a percentage of retained U(V). Further EXAFS analysis showed incorporation of the residual U(V) fraction into an iron (oxyhydr)oxide mineral phase, likely nanomagnetite or lepidocrocite. Overall, these results provide new insights into the stability of U(V) incorporated iron (oxyhydr)oxides during sulfidation, confirming the longer term retention of U in the solid phase under complex, environmentally relevant conditions.

Townsend, Luke↗
DOE OSTI2022

Next Generation System Analysis Model Recently Added Features and Future Plans - Abstract

The Nuclear Waste Policy Act of 1982, as amended (NWPA 1982), established the federal government’s responsibility to accept spent nuclear fuel (SNF) and high-level radioactive waste (HLW) from waste owners and generators for ultimate disposition. SNF generated by the current fleet of commercial nuclear reactors is being stored at the reactor sites in spent fuel pools (SFPs) and in dry independent spent fuel storage installations (ISFSIs). The US Department of Energy Office of Nuclear Energy (DOE-NE) is developing an Integrated Waste Management Program (IWMP) comprising a suite of options and supporting analyses to enable future informed choices. The IWMP is applying integrated waste management system architecture analysis, system engineering, and decision analysis principles to inform potential future decisions regarding potential nuclear waste management system architectures. Architecture analyses of the IWM system are being conducted to support the future deployment of a comprehensive system for managing nuclear waste that considers all major aspects of the back end of the nuclear fuel cycle (i.e., transportation, storage, and disposal). The Next Generation System Analysis Model (NGSAM) is an agent-based simulation software tool designed for the express purpose of modeling the IWM system. NGSAM imports data from the Oak Ridge National Laboratory (ORNL) Unified Database (e.g., historic assembly information, thermal profiles for assembly heat, at-reactor dry storage loadings) to ensure that the simulation initializes with a realistic representation of the state of commercial SNF in the United States. Recent major enhancements that have been implemented into NGSAM since NGSAM was last presented at the WM2019 conference include: • Tracking of railroad escort and buffer car acquisition. • Addition of heavy haul and barge routes for some sites, as well as support for user-defined inter-modal routes. • Updates to the logic that checks the thermal maps prior to package transport. • Addition of an allocation method that predicts when reactor sites will pack assemblies from their pools for dry storage and allocates packages to those reactor sites in the preceding periods, favoring direct transport packages and reducing the number of packages that reactor sites pack for dry storage at their ISFSIs. • Addition of reactor site family operational limits, which are used to limit the number of loads from the pool and from dry storage at a given reactor site per year. • Support has been added for multiple canister loading maps and packages having multiple compatible transportation overpacks. • Updates in the handling of non-commercial fuel, including a new database containing data to support the updates. • Support for repackaging at reactor sites. • Implementing additional output reports or modifying existing reports. • User edits can now be created and edited via the NGSAM website. • Ability to load packages for dry storage at ISF pools. • Same-type package blending at DOE sites. • Support for multi-mode transloading at reactor sites. These new features have improved NGSAM capabilities and/or improve the user experience with the model and will be discussed in more detail. The initial NGSAM requirements for advanced reactor fuels, reprocessing, treatment, and conditioning are preliminary and are described at a high level in this paper: analysts will provide more specific requirements to the NGSAM team in the future. Additionally, there are many data needs associated with modeling advanced reactors in NGSAM, but many of the data or plans are still in progress and/or yet to be fully defined. However, this document describes an initial exploration of the data relevant to this program. Advanced reactor data will likely require revision as concepts evolve and new considerations are made. This is a technical paper that does not take into account contractual limitations or obligations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). For example, under the provisions of the Standard Contract, spent nuclear fuel in multi-assembly canisters is not an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract governs the obligations of the parties, and this paper in no manner supersedes, overrides, or amends the Standard Contract. This paper reflects technical work which could support future decision making by DOE. No inferences should be drawn from this paper regarding future actions by DOE, which are limited both by the terms of the Standard Contract and Congressional appropriations for the Department to fulfill its obligations under the Nuclear Waste Policy Act including licensing and construction of a spent nuclear fuel repository.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗