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At least 235 records · Page 13

Process Model Development and Experimental Investigation for Spent Fuel Disposal in Crystalline Rocks: FY20 Report.

The U.S. Department of Energy Office of Spent Fuel Waste Disposition (SFWD) established in fiscal year 2010 (FY10) the Spent Fuel Waste Science & Technology (SFWST) Program (formerly the Used Fuel Disposition Campaign - UFDC) program to conduct the research and development (R&D) activities related to storage, transportation and disposal of used nuclear fuel and high level nuclear waste. The Mission of the SFWST is: To identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel and wastes generated by existing and future nuclear fuel cycles. Significant progress has been made in FY20 in both experimental and modeling arenas in evaluation of used fuel disposal in crystalline rocks, especially in model demonstration using field data. The work covers a wide range of research topics identified in the R&D plan.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NNSA-IAEC Science and Technology Working Group Area V - Waste Management and Subsurface Science, (Q1 FY2021)

The objective of Topic Area V is to evaluate the feasibility and safety of an intermediate-depth borehole for disposal of IAEC nuclear waste. The IAEC nuclear waste is currently stored in above ground and/or in near-surface facilities that provide short-term radiological worker safety but do not provide long-term, full security and safeguards protection for the nuclear material. Disposal of the IAEC nuclear waste several hundred meters below the surface in a sealed borehole would provide long-term security and safeguards due to its depth underground, as well as long-term radiological safety, due to its isolation in a stable geological formation.

54 ENVIRONMENTAL SCIENCES↗

Area V: Waste Management and Subsurface Science (FY23 Q2)

The objective of Topic Area V is to evaluate the feasibility and safety of an intermediate-depth borehole for disposal of IAEC nuclear waste. The IAEC nuclear waste is currently stored in above ground and/or in near-surface facilities that provide short-term radiological worker safety but do not provide long-term, full security and safeguards protection for the nuclear material. Disposal of the IAEC nuclear waste several hundred meters below the surface in a sealed borehole would provide long-term security and safeguards due to its depth underground, as well as long-term radiological safety, due to its isolation in a stable geological formation.

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Area V: Waste Management and Subsurface Science Q3 Report

The objective of Topic Area V is to evaluate the feasibility and safety of an intermediate-depth borehole for disposal of IAEC nuclear waste. The IAEC nuclear waste is currently stored in above ground and/or in near-surface facilities that provide short-term radiological worker safety but do not provide long-term, full security and safeguards protection for the nuclear material. Disposal of the IAEC nuclear waste several hundred meters below the surface in a sealed borehole would provide long-term security and safeguards due to its depth underground, as well as long-term radiological safety, due to its isolation in a stable geological formation.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

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↗

Panel Session 57: US Nuclear Power Plant Waste Management - US LLW Disposal Issues

This panel focused on senior utility managers discussing specific disposal site issues with the disposal site operators. Representatives from each of the operating commercial disposal sites began by presenting an update of their facility, license, and waste acceptance criteria (WAC) followed by a panel discussion in response to audience questions. Panelists with presentations: Current and Emerging US NRC Regulatory Topics (Joe Weismann); WCS Update (Dan Burns); EnergySolutions Clive Facility Support for US LLW Disposal Issues (Vern C. Rogers)

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Estimation of stress and stress-induced permeability change in a geological nuclear waste repository in a thermo-hydrologically coupled simulation

Geologic disposal is a promising solution for a safe permanent isolation of accumulated high-level nuclear waste from nuclear powerplants. The behavior of host rock is highly coupled thermally, hydromechanically, and chemically. Numerical simulations of such coupled phenomena for the extremely long term (>100,000 years) and large length scale (>1 km) of geologic disposal remain to be computationally challenging. In this study, a methodology has been developed to approximate stress and stress-induced permeability change in host rock using only thermo-hydrological (TH) variables. A coupled thermo-hydromechanical (THM) simulation is carried out using TOUGH-FLAC simulator to model THM behaviors of a generic nuclear waste repository, in order to evaluate the performance of the developed methodology, which is implemented in a coupled TH simulation. Results show that stress and permeability change estimated by the developed methodology in the TH-coupled simulation match those calculated in the THM-coupled simulation over the simulated timeframe of over 10,000 years. Details about errors in stress and permeability estimates accrued by the developed methodology are discussed here. The developed methodology will help incorporate stress-induced permeability change into existing TH simulators for the long-term radionuclide transport in geologic disposal.

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Micro- and Nanoscale Surface Analysis of Late Iron Age Glass from Broborg, a Vitrified Swedish Hillfort

Abstract Archaeological glasses with prolonged exposure to biogeochemical processes in the environment can be used to understand glass alteration, which is important for the safe disposal of vitrified nuclear waste. Samples of mafic and felsic glasses with different chemistries, formed from melting amphibolitic and granitoid rocks, were obtained from Broborg, a Swedish Iron Age hillfort. Glasses were excavated from the top of the hillfort wall and from the wall interior. A detailed microscopic, spectroscopic, and diffraction study of surficial textures and chemistries were conducted on these glasses. Felsic glass chemistry was uniform, with a smooth surface showing limited chemical alteration (<150 nm), irrespective of the position in the wall. Mafic glass was heterogeneous, with pyroxene, spinel, feldspar, and quartz crystals in the glassy matrix. Mafic glass surfaces in contact with topsoil were rougher than those within the wall and had carbon-rich material consistent with microbial colonization. Limited evidence for chemical or physical alteration of mafic glass was found; the thin melt film that coated all exposed surfaces remained intact, despite exposure to hydraulically unsaturated conditions, topsoil, and associated microbiome for over 1,500 years. This supports the assumption that aluminosilicate nuclear waste glasses will have a high chemical durability in near-surface disposal facilities.

Matthews, Bethany E. (ORCID:0000000171932583)↗

GDSA Framework Development and Process Model Integration FY2024

The Disposal Research & Development (Disposal R&D) Campaign of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Spent Fuel & High-Level Waste Disposition is conducting research and development (R&D) on geologic disposal of spent nuclear fuel (SNF) and high-level nuclear waste (HLW). A high priority for Disposal R&D is disposal system modeling (Sassani et al. 2023). The Geologic Disposal Safety Assessment (GDSA) work package is charged with developing a disposal system modeling and analysis capability for evaluating generic disposal system performance for nuclear waste in geologic media.

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New Tank Mapping Method Improves Waste Removal Process

Savannah River Mission Completion is the Liquid Waste (LW) contractor at the Savannah River Site (SRS). The LW mission is tasked with treating and disposing of legacy nuclear waste. There are multiple facilities involved in this work, including the Concentration, Storage, and Transfer Facilities (CSTF), the Defense Waste Processing Facility (DWPF), the Salt Waste Processing Facility (SWPF), and the Saltstone Production Facility (SPF). The CSTF includes 43 underground waste tanks used to store and support processing of radioactive liquid waste. Waste removal activities, such as salt dissolution campaigns and sludge agitation, are conducted within the CSTF waste tanks to convert the waste into a form that allows for downstream processing at other LW facilities. While performing these waste removal campaigns, camera inspections are performed to assess the quantity and distribution of the remaining waste within the waste tank (i.e. saltcake or sludge). Understanding the quantity and distribution of the salt/sludge within the waste tanks allows for improved waste removal strategies (e.g. mixing pump operation) and refined safety controls. Typically, several camera inspections are performed during a waste removal transfer to verify the elevation of the visible salt/sludge mounds against the known elevation of the liquid surface. The camera inspection footage must then be interpreted by a trained engineer who will develop a 2-D map that depicts the waste distribution at various elevations within the waste tank. This tank mapping is then used in conjunction with conservative assumptions to evaluate the volume of saltcake or sludge that is present within the waste tank.

Mini, Melany↗

Scenario development for safety assessment in deep geologic disposal of high-level radioactive waste and spent nuclear fuel: A review

Radiation and radioactive substances result in the production of radioactive wastes which require safe management and disposal to avoid risks to human health and the environment. To ensure permanent safe disposal, the performance of a deep geological repository for radioactive waste is assessed against internationally agreed risk-based standards. Assessing postclosure safety of the future system's evolution includes screening of features, events, and processes (FEPs) relevant to the situation, their subsequent development into scenarios, and finally the development and execution of safety assessment (SA) models. Global FEP catalogs describe important natural and man-made repository system features and identify events and processes that may affect these features into the future. By combining FEPs, many of which are uncertain, different possible future system evolution scenarios are derived. Repository licensing should consider both the reference or “base” evolution as well as alternative futures that may lead to radiation release, pollution, or exposures. Scenarios are used to derive and consider both base and alternative evolutions, often through production of scenario-specific SA models and the recombination of their results into an assessment of the risk of harm. Furthermore, while the FEP-based scenario development process outlined here has evolved somewhat since its development in the 1980s, the fundamental ideas remain unchanged. A spectrum of common approaches is given here (e.g., bottom–up vs. top–down scenario development, probabilistic vs. bounding handling of uncertainty), related to how individual numerical models for possible futures are converted into a determination as to whether the system is safe (i.e., how aleatoric uncertainty and scenarios are integrated through bounding or Monte Carlo approaches).

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Geologic Disposal Safety Assessment (GDSA) Biosphere Model Development

The Spent Fuel and Waste Science and Technology Campaign of the U.S. Department of Energy Office of Nuclear Energy, Office of Spent Fuel and Waste Disposition is conducting research and development on geologic disposal of spent nuclear fuel and high-level nuclear waste. This work includes the Geologic Disposal Safety Assessment (GDSA) program which is charged with development of generic deep geologic repository concepts and system performance assessment models. One part of the GDSA framework is the development of a biosphere model capable of assessing doses to potential receptors exposed to radionuclides released from geologic disposal sites. As part of the GDSA framework, a biosphere model compatible with the PFLOTRAN massively parallel subsurface flow and reactive transport code is under development. The PFLOTRAN model provides the radionuclide source term for the biosphere model. The GDSA Biosphere model then assesses the potential movement of radionuclides through the surface biosphere and the subsequent exposure to a human receptor living in the biosphere. The biosphere model includes pathways originating from the groundwater as well as pathways originating from surface water bodies that have a water exchange with a contaminated groundwater body. The pathways for human exposure include consumption of drinking water, irrigated crops, meat animals, aquatic vegetation, and animals, etc.; external exposure from irrigated ground surfaces, surface water bodies, recreational activities, etc.; and inadvertent exposures such as ingestion of contaminated soils or shower water, etc. The GDSA Biosphere model was designed to be flexible and generic in order to accommodate a variety of different sites and climate states. This presentation will present the on the purpose, design, and development progress of the GDSA Biosphere Model.

GDSA, biosphere, repository↗

M3SF-24LL010302062-NEA-TDB Management and International Collaborations in Sorption and Thermodynamic Modeling

This progress report (Level 3 Milestone Number M3SF-24LL010302062) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Crystalline International Collaborations Work Package Number SF-24LL01030206. The activity is focused on our long-term commitment to engaging our partners in international nuclear waste repository research. This includes participation in the Nuclear Energy Agency Thermochemical Database (NEA-TDB) Project and development of methodologies for integrating US and international thermodynamic databases for use in SFWST Generic Disposal System Assessment (GDSA) efforts. A continuing focus for FY24 efforts is to support the US participation in the NEA-TDB effort. The focus of FY24 activities was the development of an agreement for a Phase 7 activity that will start in Q1 of 2025. Mavrik Zavarin is now the US representative on both the Management Board and the Executive group to the NEA-TDB. He is also the POC for the Cements State of the Art Report that is undergoing peer review in FY24. In FY24, we used our position on the NEA-TDB MB and EG to facilitate the integration of NEA-TDB thermochemical data with LLNL’s SUPCRTNE thermodynamic database that supports the SFWST GDSA activities. This effort is coordinated with the Argillite work package SUPCRTNE database development efforts (Wolery, 2024). The goal is to provide a downloadable database that will be hosted on LLNL’s thermodynamics website which incorporates NEA-TDB data into the LLNL database where appropriate. We also began engagement with the EURAD2 program that was initiated in FY24 by our European collaborators at the Karlsruhe Institute of Technology (KIT), Germany. The primary focus of the engagement is with WP20: DITUSC Thermodynamic database evaluation program. A kickoff meeting for this activity is planned for early FY25. Finally, we have been selected to co-host (with Clemson University) the International Conference on Chemistry and Migration Behaviour of Actinides and Fission Products in the Geosphere in 2025 (Migration2025). The meeting will be held September 21-26, 2025, in New Orleans, Louisiana, and will focus on international efforts to understand the risks of radionuclide releases into the environment. This central focus of this conference is on international efforts to develop safe disposal options for nuclear wastes. As such, we are developing a theme focused on US underground nuclear waste repository science.

58 GEOSCIENCES↗

New Beginnings: Optimising Waste and End-state at UK Nuclear Sites to Enable Environmental Permit surrender (the GRR) - 20193

The decommissioning of nuclear sites involves the management of large volumes of radioactive waste, other conventional waste, and the clean-up of land and groundwater contamination. In the UK, the disposal of radioactive waste from nuclear sites is regulated by the Environment Agency (England), Natural Resources Wales (Wales) and the Scottish Environment Protection Agency (Scotland). All three environment agencies seek to make sure that the generation and disposal of radioactive waste is optimised: ensuring impacts to people are 'as low as reasonably achievable' (ALARA). With a large proportion of UK nuclear sites embarking on decommissioning, the environment agencies have collaboratively developed new guidance: the 'Management of radioactive waste from the decommissioning of nuclear sites: Guidance on the Requirements for Release from radioactive substances regulation' (referred to as the GRR)[1]. The environment agencies worked in liaison with the Office for Nuclear Regulation (ONR), the UK nuclear safety regulator, and the UK government body responsible, the Nuclear Decommissioning Authority (NDA). Prior to finalising the guidance, a 12 month trial was completed at three of the NDA's nuclear sites, enabling the environment agencies to test their draft GRR guidance against 'real world scenarios' and the challenge of incorporating its requirements into decommissioning plans. Experience gained during the trial was incorporated into the current version of the GRR, published in July 2018. The GRR represents the 'new beginning' of this paper's title because for the first time, the UK environmental regulators have expressed in detail their requirements for surrender of environmental permits for nuclear sites. The GRR provides a framework for a UK nuclear site to plan, assess and justify its management of radioactive waste and deliver through decommissioning, an optimal, environmentally safe and sustainable end-state. This paper will describe the requirements and technical standards within the GRR, the establishment of defensible clean-up criteria and the production of a Waste Management Plan (WMP), supported by a Site-Wide Environmental Safety Case (SWESC). It will describe the experience gleaned by operators and regulators during the trial and it will highlight current work by regulators to implement the new approach proportionately. Finally, this paper will look forward and outline planned amendments to legislation in the UK affecting nuclear sites that should enable the full benefits of the GRR to be realised in the future. (authors)

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

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

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

Applications for special-purpose minerals at a lunar base

Maintaining a colony on the Moon will require the use of lunar resources to reduce the number of launches necessary to transport goods from the Earth. It may be possible to alter lunar materials to produce minerals or other materials that can be used for applications in life support systems at a lunar base. For example, mild hydrothermal alteration of lunar basaltic glasses can produce special-purpose minerals (e.g., zeolites, smectites, and tobermorites) that in turn may be used in life support, construction, waste renovation, and chemical processes. Zeolites, smectites, and tobermorites have a number of potential applications at a lunar base. Zeolites are hydrated aluminosilicates of alkali and alkaline earth cations that possess infinite, three-dimensional crystal structures. They are further characterized by an ability to hydrate and dehydrate reversibly and to exchange some of their constituent cations, both without major change of structure. Based on their unique absorption, cation exchange, molecular sieving, and catalytic properties, zeolites may be used as a solid support medium for the growth of plants, as an adsorption medium for separation of various gases (e.g., N2 from O2), as catalysts, as molecular sieves, and as a cation exchanger in sewage-effluent treatment, in radioactive waste disposal, and in pollution control. Smectites are crystalline, hydrated 2:1 layered aluminosilicates that also have the ability to exchange some of their constituent cations. Like zeolites, smectites may be used as an adsorption medium for waste renovation, as adsorption sites for important essential plant growth cations in solid support plant growth mediums (i.e., 'soils'), as cation exchangers, and in other important application. Tobermorites are cystalline, hydrated single-chained layered silicates that have cation-exchange and selectivity properties between those of smectites and most zeolites. Tobermorites may be used as a cement in building lunar base structures, as catalysts, as media for nuclear and hazardous waste disposal, as exchange media for waste-water treatment, and in other potential applications. Special-purpose minerals synthesized at a lunar base may also have important applications at a space station and for other planetary missions. New technologies will be required at a lunar base to develop life support systems that are self-sufficient, and the use of special-purpose minerals may help achieve this self-sufficiency.

Ming, Douglas W.↗

Simulations of Criticality Control Overpack Container Compaction at the Waste Isolation Pilot Plant

Criticality Control Overpack (CCO) containers are being considered for the disposal of defense-related nuclear waste at the Waste Isolation Pilot Plant (WIPP). At WIPP, these containers would be placed in underground disposal rooms, which will naturally close and compact the containers closer to one another over several centuries. This report details simulations to predict the final container configuration as an input to nuclear criticality assessments. Each container was discretely modeled, including the plywood and stainless steel pipe inside the 55-gallon drum, in order to capture its complex mechanical behavior. Although these high-fidelity simulations were computationally intensive, several different material models were considered in an attempt to reasonably bound the horizontal and vertical compaction percentages. When exceptionally strong materials were used for the containers, the horizontal and vertical closure respectively stabilized at 43:9 % and 93:7 %. At the other extreme, when the containers completely degraded and the clay seams between the salt layers were glued, the horizontal and vertical closure reached respective final values of 48:6 % and 100 %.

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