Search NASA⌕ Search

SEARCH · Search NASA

Results for “Waste Repository”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Panel Session 89: Progress on Deep Repository Around the World

This panel focused on the progress of deep repository programs, worldwide. While efforts on a deep geological repository for SNF/UNF and HLW in the USA were terminated in 2010, several other repository programs continue to make considerable progress 10 and deserve to be recognized. Panelists from national, multi-national, and international repository programs and organizations reported on the status and future plans as well as the perceived reasons underlying the success and the failure of components of their respective repository program. This session provided a forum for interested parties to present, exchange, and take advantage of, as appropriate, repository progress, challenges, concepts/designs and other lessons learned around the world, including site characterization, site selection, research and development, licensing, construction, operation, and public acceptance and participation. Panelists with presentations: Switzerland's Geological Repository Program for Radioactive Waste (Tim Vietor); Update on NWMO's Progress in Implementing Canada's Plan (Sarah Hirschorn); Cigeo Project Progress and Next Milestones (Corinne Bauer)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Optimising National WM Strategies through early consideration of Nuclear Safeguards - 18607

International Atomic Energy Agency (IAEA) safeguards stand alongside safety and security as one of the three pillars for prudent management of all nuclear material, including nuclear waste. Although safety considerations, including public and environmental radiation protection, are the traditional drivers of waste management strategies, there is a requirement for nuclear safeguards in shaping and quantifying long-term solutions for special classes of radioactive waste. This is especially important when considering the disposal of nuclear material in a repository that could be released from regulatory oversight for radiation protection after a period of institutional control, as some safeguards oversight would still be required. Failure to consider nuclear safeguards at the early stages of planning may alternatively lead to excessive volumes for disposal or the need to recondition already treated wastes. At the very least, safeguards measures should be considered before countries seek to dispose of legacy nuclear materials, decommissioning wastes and other materials containing uranium, thorium or plutonium in an operating repository. Nuclear safeguards are the set of tools and techniques used by the IAEA to verify that countries comply with their international obligations not to divert nuclear materials to nuclear explosive devices. Under the Nuclear Non-Proliferation Treaty (NPT), nuclear safeguards apply to all uranium, thorium and plutonium held in non-weapons NPT states, with limited exceptions. In nuclear-weapon states, much of their material is likewise subject to IAEA safeguards through conditions in agreements with respective suppliers of the uranium. Under IAEA safeguards, such nuclear material must remain accounted for and accessible for periodic verification-analogous to the safety concepts of reversibility and retrievability. Nevertheless, safeguards requirements can be terminated - provided that the State and the [IAEA] agree that such nuclear material is practicably irrecoverable - through dilution or immobilisation in solid form (such as cement, glass or ceramic). This means candidate material can be disposed of as non- nuclear material, with no further safeguards requirements, provided this criterion is met. Unlike radiological wastes where repository design strongly influences pathways for radionuclide migration and overall safety performance, repository design is not a factor in considering the termination of safeguards on nuclear material. This paper looks at similarities and differences in the safety, security and safeguards regulatory requirements for the management of radioactive waste and nuclear material in Australia. In particular, it outlines recent experience gained from participation in Australia's waste acceptance criteria working group for a national low-level waste repository and intermediate-level waste store, as well as from Australia's overall waste management strategy. These items of work have coincided with the IAEA update of its internal policy on termination of safeguards on nuclear material. Australia's experience shows how inclusion of safeguards criteria in the planning process is leading to optimisations in choice of facility concepts, project sequencing and equitable distribution of economic risk. By considering the three pillars-safety, security and safeguards-early on in the planning process, Australia's waste management strategy will provide a benchmark for other states in establishing and updating their own waste acceptance criteria for the disposal of nuclear material. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

FIU Project 2: Environmental Remediation Science & Technology [Slides]

FIU’s research under this project involves conducting basic and applied science to fill knowledge gaps and validate potential remediation technologies for contaminated soil and groundwater and the assessment of the fate and transport of contaminants in the environment. The aim of FIU’s research is to reduce the potential for contaminant mobility or toxicity in the surface and subsurface through the development and application of state-of-the-art scientific and environmental remediation technologies at the Hanford Site, Savannah River Site (SRS), and the Waste Isolation Pilot Plant (WIPP), which is the Nation’s only mined geologic repository for permanent disposal of transuranic waste. FIU collaborates with scientists from Pacific Northwest National Laboratory (PNNL), Savannah River National Laboratory (SRNL), Savannah River Ecology Laboratory (SREL), Los Alamos National Laboratory (LANL) and the DOE Carlsbad Field Office (CBFO) in order to plan and execute research that is synergistic with the work being conducted at the sites, and that supports the resolution of critical science and engineering needs which leads to a better understanding of the long-term behavior of subsurface contaminants. The knowledge gained through this research will be used to transform experimental and modeling innovations into practical applications deployed at the sites to support EM’s primary goal of expediting the closure of major contaminated soil and groundwater sites and waste units. Collaborative relationships between FIU and the national laboratories have provided large benefits over the years to FIU, the national laboratories, the DOE complex, and the DOE EM mission. By working closely with the national laboratories, FIU’s research is not only closely aligned with the cleanup mission priorities at the DOE sites, but complements and supports ongoing work at the national laboratories for screening of new remedial technologies. This coordination and leveraging of research efforts results in time- and cost-savings, and will accelerate progress of the DOE EM environmental restoration mission.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

What's the big deal with TA-55's trash?

There's a beast of hazardous waste, and we're learning how to tame it. Plutonium and production activities are ongoing at TA-55, which means the beast of hazardous waste is always looming around the corner. The Laboratory must dispose of this delicate type of waste properly and swiftly. Learning from previous snags and pitfalls in an incredibly complex process, TA-55 leadership and staff have worked together to greatly improve and streamline waste processes. The result: As of the last week of fiscal year 2020, 42 shipments containing 1,275 containers of transuranic (TRU) waste were sent to the Waste Isolation Pilot Plant (WIPP) in southern New Mexico, the nation’s only repository for defense-generated TRU waste. In FY 2019, we shipped less than half that amount.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Systematic Calculation of Pu Pourbaix Diagrams: Modelling Support for the LANL ACRSP Team

The report enclosed entitled “Systematic Calculation of Pu Pourbaix Diagrams: Modelling Support of the LANL ACRSP Team” was prepared by Amphos21 under contract to the Los Alamos National Laboratory Actinide Chemistry and Repository Science Program (ACRSP) team to support an enhanced understanding of the actinide and brine chemistry in the Waste Isolation Pilot Plant (WIPP) transuranic repository. In this report, the ThermoChimie chemistry model and database, with some slight modifications, were used to calculate Pourbaix diagrams to investigate the effects of ionic strength, organic complexation, and metal competition on the predominance diagrams for aqueous and solid plutonium species under the anoxic and reducing conditions expected in the repository. These diagrams were connected to iron chemistry phase diagrams since this iron chemistry is expected to define and control redox properties in the WIPP repository concept.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Simulations of Pipe Overpack Container Compaction at the Waste Isolation Pilot Plant

The Waste Isolation Pilot Plant (WIPP) is an operating geologic repository in southeastern New Mexico for transuranic (TRU) waste from nuclear defense activities. Past nuclear criticality concerns have generally been low at the WIPP due to the low initial concentration of fissile material and the natural tendency of fissile solute to disperse during fluid transport in porous media (Rechard et al. 2000). On the other hand, the list of acceptable WIPP waste types has expanded over the years to include Criticality Control Overpack (CCO) containers and Pipe Overpack (POP) containers. Containers bound for WIPP are bundled together in hexagon shaped 7-packs (six containers surround one container in the center). Two 7-packs are often combined into a TRUPACT-II package for a total of 14 containers. Most TRUPACT-II packages are restricted to a maximum fissile mass equivalent to plutonium (FMEP) between 0.1 and 0.38 kg, but a CCO TRUPACT-II package and a POP TRUPACT-II package are respectively permitted to have 5.32 kg and 2.80 kg FMEP (see Section 3 of US DOE (2013)). Consequently, CCO container criticality after emplacement at the WIPP was evaluated in Saylor and Scaglione (2018), and Oak Ridge National Laboratories is currently at work on POP container criticality analyses.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

WIPP PA Sensitivity Studies in Support of CRA-2014 Completeness Determination and Compliance Decision - 18209

The Waste Isolation Pilot Plant (WIPP) has been developed by the U.S. Department of Energy (DOE) for the geologic (deep underground) disposal of transuranic (TRU) waste. Containment of TRU waste at the WIPP is regulated by the U.S. Environmental Protection Agency (EPA). The DOE demonstrates compliance with the containment requirements by means of performance assessment (PA) calculations. WIPP PA calculations estimate the probability and consequence of potential radionuclide releases from the repository to the accessible environment for a regulatory period of 10,000 years after facility closure. Since TRU waste was first received at the WIPP in 1999, compliance recertification applications (CRAs) have been required by the EPA on five-year cycle. The latest CRA (CRA-2014) was submitted by the DOE to the EPA in 2014. As part of the application, a formal set of PA calculations (CRA-2014 PA) was performed and results were documented in detail. The CRA-2014 PA incorporated several parameter updates based on new data, as well as the implementation of a different model for panel closures, barriers planned for the opening of each waste panel. The results presented as part of the CRA-2014 showed via the required complementary cumulative distribution function (CCDF) release curves that the estimated releases were within compliance limits. As part of their process for reviewing the completeness of the CRA-2014 [1], the EPA formally requested that the DOE perform sensitivity studies based on EPA-requested changes to model parameters. The DOE performed and documented the results of three sensitivity studies described below: CRA14-SEN2 [2], CRA14-SEN3 [3], and CRA14-SEN4 [4]. In each case, the new analysis results were compared to CRA- 2014 PA (CRA14) results to determine the impact of the EPA-requested changes. CRA14-SEN2: the application of EPA-requested modified parameters in the operations and experimental (non-waste) areas of the repository to simulate an accelerated (instantaneous) creep closure, the inclusion of capillary pressure effects on relative permeability, and an increase in initial/residual brine saturation and residual gas saturation. The modifications to the repository model resulted in increased pressures and decreased brine saturations in waste areas and increased pressures and brine saturations in the operations and experimental areas. However, the overall effects on total high-probability mean releases from the repository were entirely insignificant. It was concluded that the modeling assumptions associated with the operations and experimental areas of the repository have an insignificant effect on the prediction of total releases from the repository and/or adequacy of the current (CRA14) model to demonstrate compliance with the regulatory limits. CRA14-SEN3: the application of EPA-requested modified parameters in the panel closure areas and operations and experimental (non-waste) areas of the repository to simulate an accelerated (instantaneous) creep closure, the inclusion of capillary pressure effects on relative permeability, and an increase in initial/residual brine saturation and residual gas saturation. The modifications to the repository model predominantly result in increased pressures and decreased brine saturations in waste areas, increased pressures and brine saturations in the operations and experimental areas, and decreased brine and gas flows across all panel closures. The pressure increases in repository waste regions yielded slightly decreased brine saturations (on average) in those areas. However, the overall effects on total high-probability mean releases from the repository are entirely insignificant. It was concluded that the modeling assumptions made for this sensitivity study which were associated with the panel closures and north end areas of the repository have only a small effect on the prediction of total releases from the repository and/or adequacy of the current (CRA14) model to demonstrate compliance with the regulatory limits. CRA14-SEN4: the application of several changes to the current WIPP PA model (CRA14), including: parameter changes, use of an updated code, and a revised computational grid for the BRAGFLO code. The modifications to the repository model resulted in increased releases in all primary release mechanisms. The impacts of each EPA-requested change to CRA14 were analyzed with respect to each release mechanism in the CRA14-SEN4 study. Overall, total high-probability predicted mean releases from the repository were increased by about 15%. Total low-probability predicted mean releases were increased by about 107%. It was concluded that the EPA-requested changes to the CRA14 result in increases to the predicted total releases from the repository. However, releases calculated in the CRA14-SEN4 analysis remained below regulatory limits, demonstrating continued compliance of the WIPP. In January 2017, the EPA determined that CRA-2014 was complete and then began a technical evaluation of CRA-2014 to determine whether the WIPP continued to comply with the radiation protection standards for disposal. During their evaluation, the EPA utilized the sensitivity study results to inform their ultimate determination in July 2017 that the WIPP continues to comply with regulatory standards. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Advances in 3D Geologic Modeling of Alluvial Basins with a Focus on Facies and Property Modeling

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. There are hundreds of alluvial basins and sub-basins scattered across the arid western United States (Figure 1-1). Precipitation and infiltration rates are relatively low with high evapotranspiration, resulting in vertical separation between repository and water table and thus longer transport paths 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. A thick geologic host medium, which serves as the natural barrier system (NBS) in the conceptual model framework of a geologic disposal system, 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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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 OSTI2017

International Collaboration Activities on Engineered Barrier Systems

The Spent Fuel Waste Science and Technology (SFWST) campaign from the DOE Fuel Cycle and Technology (FCT) program has been engaging in international collaborations between repository R&D programs for nuclear waste disposal to leverage on the extensive research investigations and laboratory/field data of engineered barrier system (EBS) components (e.g., near-field) and characterization of transport phenomena in the host rock (e.g., far-field) processes from state-of-the-art underground research laboratories (URL) experiments. Thermal heating from radionuclide decay in the waste canisters will generate increases in temperature that will drive chemical and transport processes in the near- and far-field domains of the repository. URL sites provide the ideal setting to conduct heater test experiments to simulate the thermal effects of heat-generating nuclear waste in disposal galleries and surrounding host rock.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

DECOVALEX-2023: Task E Final Report

This is the Task E final report for DECOVALEX-2023. Task E is focused on understanding thermal, two-phase hydrological, and mechanical (TH2M) processes, especially related to predicting brine migration in the excavation damaged zone around a heated excavation in salt. Salt is attractive as a disposal medium for radioactive waste because it is self-healing and is essentially impermeable and essentially non-porous in the far field (away from excavations). Investigation of the short-term (days to years) near-field (centimeters to tens of meters) behavior of salt is important for radioactive waste disposal because this early period strongly controls the amount of brine in a salt repository. Brine leads to corrosion of waste forms and waste packages, and possible dissolution of radionuclides with brine transport being a potential transport vector to the accessible environment. The main test case used in Task E is the ongoing Brine Availability Test in Salt (BATS) heater test located underground at the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico, USA. The Task was divided into a series of Steps. Step 0 was an introduction to processes in salt, that included matching historical unheated brine inflow data from boreholes at WIPP and matching temperature observations during BATS heater test 1a. Step 1 included validation of models against a thermo-poroelastic analytical solution relevant to heated boreholes in salt, and two-phase flow around an excavation in salt. Step 2 required all the individual components covered in steps 0 and 1 to come together to match observed brine inflow behavior during the BATS 1a heater test. There were a range of approaches from the teams, from mechanistic to prescriptive. Given the uncertainties in the problem, some teams used one- or two-dimensional models of the processes, while other teams included more geometrical complexity in three-dimensional models. The key learning points from Task E have been: • Heat conduction through salt typically requires non-linear thermal conductivity (as a function of temperature), but most models do a good job matching observations, given appropriate adjustments to the applied power and some thermocouple locations. • Thermal pressurization requires coupled thermal-hydrological-mechanical (THM) responses that consider the thermal expansion of the fluid and solid phases. • Initialization of two-phase flow models around a borehole or excavation in salt are more realistically represented as “wetting up”, rather than “drying down” (i.e., the initial state after excavation is mostly dry, rather than mostly wet). • The BATS 1a heater test includes a significant release of brine after the end of heating, which requires a large increase in permeability to recreate. Task E has been a great learning experience for all the teams involved, and feedback from the modeling teams has led to changes in the design of follow-on BATS experiments, which are now ongoing underground at WIPP. There was a balance throughout the task between freedom to model phenomena how each team saw fit, and prescriptiveness in problem design to bring the modeling teams closer together to allow attribution of smaller differences between models to different modeling choices. The modeling approaches seem to go through two phases: an early phase of discovery or testing, and a later phase of refinement and improvement. In future modeling efforts, different field data could be used (e.g., BATS 2) and more time should be included in the processes for teams to make multiple model refinement or even significant changes to their conceptual model or setup, based on lessons learned from the modeling exercise.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Final Report: Vitrification of Inorganic Ion-Exchange Media, VSL-16R3710-1

One of the primary roles of waste pretreatment at the Hanford Tank Waste Treatment and Immobilization Plant (WTP) is to separate the majority of the radioactive components from the majority of the nonradioactive components in retrieved tank wastes, producing a high level waste (HLW) stream and a low activity waste (LAW) stream. This separation process is a key element in the overall strategy to reduce the volume of HLW that requires vitrification and subsequent disposal in a national deep geological repository for high level nuclear waste. After removal of the radioactive constituents, the LAW stream, which has a much larger volume but smaller fraction of radioactivity than the HLW stream, will be immobilized and disposed of in near surface facilities at the Hanford site.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Potential for Transport of Cesium as Bio-colloids in a High Ionic Strength System

The Waste Isolation Pilot Plant (WIPP) is a deep geologic repository for long-term disposal of transuranic, radioactive waste that is a byproduct from the nation's nuclear defense program. Most likely release scenario → human intrusion (potentially through drilling) (US DOE 1995; 1996). Brine may proceed through the Rustler formation (the most transmissive layer) (Perkins et al., 1999). There is a need to investigate halophilic microbes including their mobility in the Rustler formation and the potential for enhancing transport of cesium as a bio-colloid. Objectives: To quantify the change in mobility of Cs by interaction with microbes (Chromohalobacter) that may be present near Waste Isolation Pilot Plant. What is Chromohalobacter?: a halophilic microbe, isolated from near the WIPP site, able to thrive in the high salt concentrations relevant to the repository. Does the microbe growth phase effect uptake of Cs? Experiments conducted with actively growing and stationary phase Chromohalobacter. Uptake may differ if Cs{sup +} is mistaken for K{sup +} in an actively growing (Log) phase versus stationary phase microbes (not actively reproducing). Materials: 1'' Teflon column packed with 1 gram of dolomite [355-500 μm, CaMg(CO{sub 3}){sub 2}] connected to a syringe pump and fraction collector via Teflon tubing. Chromohalobacter (100 million cells/mL) and dolomite collected near the WIPP. Synthetic brine - 15% NaCl (w/v ∼2.78 M) + 3 mM NaHCO{sub 3}. Methodology: Negative Control 5,000 ppb Cs initially injected into mini-columns with brine. Injection solution was then switched to only brine. Cs Experiments with Stationary Phase Chromo: Set 1: Viable stationary Chromo injected into minicolumns after reaction with 200 ppb Cs. Injection solution was then switched to only brine (without Cs or microbes). Set 2: Stressed stationary Chromo initially injected into mini-columns after reaction with 200 ppb Cs. Then, switched to brine only. Cs Experiments with Log phase Chromo: Set 1: Chromo injected into mini-columns after reaction with 5,000 ppb Cs. Injection then switched to only brine (without Cs or microbes). Set 2: Chromo grown with 5,000 ppb Cs and spiked with additional Cs before injection into columns. Preliminary Conclusions: Cs was not taken up or mobilized by Chromohalobacter in the actively growing or stationary phases or under stressed conditions. Competition may have occurred between Cs and K{sup +} at lower ionic strength, significantly greater concentrations of K{sup +} were present. Results suggest that Chromohalobacter can selectively uptake K{sup +} in the presence of Cs{sup +}. Future Research: K{sup +} and Cs{sup +} uptake will be monitored at high ionic strength in K{sup +} depleted media to consider uptake of Cs in the absence of K.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Designing a repository in domal salt: The influence of design variants in different modelling environments

To understand the long-term environmental impact of disposing radioactive waste of in a deep geological repository and to optimise its design, performance assessments are used. In this study, four teams (COVRA, GRS, Quintessa, and DOE) modified the previously developed generic repository of DECOVALEX task F2 to identify commonalities and differences between the teams for specific changes in repository design. The teams tested six design modifications: (1) Replacing concrete abutments with run-of-mine salt; (2) Replacing the salt seal with a concrete abutment and using run-of-mine salt instead for the two concrete abutments in each drift seal; (3) Halving the size of the infrastructure area; (4) Using run-of-mine salt instead of gravel for backfilling the infrastructure area; (5) Disposal of spent nuclear fuel without the POLLUX-10 containers (6); Lower initial saturation of the spent nuclear fuel and vitrified waste disposal drifts. Despite different modelling strategies used, models agreed that a smaller infrastructure area has a limited effect on radionuclide transport. Responses to the absence of the two concrete abutments in each seal, the use of single large concrete abutments (200 m each), or the use of run-of-mine salt in the infrastructure area differ between teams due to differing modelling assumptions. Based on these results, the estimated efficacy of containment depends strongly on the model assumptions of each team. More specifically, it appears to depend on the compaction model used and therefore on the backfill material used in different areas of the repository. However, the drift seal appears to be a critical design element in all models, effectively limiting radionuclide transport by hydrologically disconnecting sections of the repository. Additional beneficial design choices include the use of dry salt in disposal drifts to limit radionuclide transport and reducing the infrastructure area costs and minimizing host rock damage.

DECOVALEX↗
DOE OSTI2020

Reduction of the Waste Occupied Area by Nuclide Separation and Horizontal Emplacement of the Waste Packages - 20422

Two of the key objectives for realizing the advanced nuclear fuel cycle are to reduce the volume of high-level radioactive waste and to reduce the foot-print of the geological repository. To achieve these objectives, research and development on nuclide separation technologies is a prerequisite. In this context, it is necessary to consider the vitrification conditions and the design of the geological repository where separation technologies would be implemented to achieve effective reductions in the volume of vitrified waste and the foot-print of the geological repository. In this study, the effects of the vitrification conditions on the number of vitrified waste units and the disposal area were evaluated, based on the assumption that separation of Cs and Sr from the spent nuclear fuel had been performed. In addition, with reference to the design of the geological repository, the impact on the foot-print of the repository for different emplacement methods of the waste packages was examined. As a result, it was clarified that the waste-occupied area could be reduced by horizontal emplacement of the vitrified waste with low heat generation. In addition, it was demonstrated that a low waste-loaded vitrified waste contributed to a reduction in the foot-print of the geological repository in the case of horizontal emplacement, although the number of vitrified waste units increased. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Panel Session 26: Decommissioning, Issues and Possibilities/ Opportunities

The panel focused on the latest Issues and Opportunities in D and D around the world. The format of this panel was aiming at an open exchange of experience and ideas based on actual decommissioning and gave the panelists and the audience an opportunity to share lessons learned as well as networking during and after the session. For the session, the panelists with experience from different fields in decommissioning such as dismantling, waste management, regulatory aspects and repository management participated. Interesting topics to discuss with the audience were: 1) Logistics (from cradle to grave); 2) Organizational aspects of decommissioning; 3) Regulatory aspects and issues; 4) Definition of the back-end with regard to packages and Waste Acceptance Criteria (WAC); and 5) Identification and management of difficult-to-handle D and D waste (e.g. damaged fuel, legacy/forgotten waste inside Nuclear Power Plants, etc. Presentation proposed: Decommissioning - Issues, Possibilities and Opportunities (David Peattie, Nuclear Decommissioning Authority (United Kingdom))

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Response of bentonite microbial communities to stresses relevant to geodisposal of radioactive waste

Microbes have been isolated previously from bentonite materials that may be used as barriers for the disposal of radioactive waste. Actively respiring microbes in such barrier materials, within a repository environment, have the potential to adversely affect waste container corrosion rates. Additionally, they could potentially alter the properties of the bentonite barrier itself. This is of significance, since the integrity of the waste container and properties of the bentonite barrier are required to fulfil defined safety functions. Furthermore to help identify the critical factors that affect microbial activity in bentonite materials, this study examines the impact of a range of parameters that could affect microbial metabolism in a geodisposal environment. Several bentonites from different sources (bentonite mined from locations in Spain and the USA, along with commercially-sourced bentonite) were subjected to increased pressure (74 MPa, 30 s), heat (90 °C, 24 h), and irradiation (1000 Gy, 24.17 Gy min –1 ), before incubation in growth media selective for sulfate-reducing bacteria (SRB) or iron-reducing bacteria (IRB).

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2017

Legacy Waste Containers for Low and Intermediate Level Radioactive Waste in Germany - Actual Situation and Challenges - 17070

Low and intermediate level radioactive waste has been produced in Germany for several decades. While a part of the waste had been deposited in the Morleben repository and the Asse mine which have been operated for just a limited period of time, the major amount is conditioned and packaged in temporarily stored waste containers. Those containers are designated for final disposal in the KONRAD repository being currently under construction and expected to start operation in 2022. As a consequence, an assessment of the waste container documentation and safety is strongly needed to receive an approval for the Konrad repository. This requalification is challenging because those old containers have to meet the same safety requirements as new containers. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗