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All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Integrating Immersive Visualization in Molten-Salt Reactor Waste Management for Experimental Design and Planning

Molten-salt reactors (MSRs) represent a promising solution for next-generation nuclear energy, offering advantages in safety, fuel efficiency, and waste minimization. However, their liquid-fueled design presents unique challenges for spent fuel management, making post-shutdown waste characterization essential for developing effective strategies. Despite this need, there is a notable absence of visualization platforms specifically tailored to the unique characteristics and analytical requirements of MSR waste management. Existing tools in the nuclear industry are primarily designed for reactor operations or generic data exploration and lack both integration with MSR-specific multiphysics frameworks and the ability to simultaneously visualize time-dependent thermal fields, chemical composition evolution, and radiation distribution patterns. To address these limitations, this paper presents an immersive virtual reality (VR) visualization platform that processes and displays high-fidelity multiphysics simulation output from the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework in real-time, using Unity. The platform visualizes MSR waste characteristics such as nuclide decay, salt cooling, and corrosion by using Exodus II output data and running on a VR headset. It includes a user-friendly interface with features such as visibility toggling, cross-sectional slicing, and time-series animation for exploring simulation data. These capabilities support experimental design, stakeholder engagement, and public communication by making complex reactor behavior more accessible and understandable. By enhancing spatial reasoning and reducing cognitive load, this immersive environment fosters more effective communication and decision-making in MSR waste management.

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DOE OSTI2020

Panel Session 3: Prospects for Shared Solutions for Solving the WM Challenges for Small and Emerging Nuclear Countries

This panel focused on sharing WM solutions for small and emerging nuclear countries. As countries with substantial inventories of spent nuclear fuel pursue their national repository programs with varying degrees of progress, countries with only a few reactors and those at the beginning stages of their nuclear programs face a far different and substantial challenge toward developing a national solution. This session brought together a panel of experts from government and industry to explore shared solutions and next steps building off the recent progress made by international organizations and the recent efforts and lessons learned. Panelists with presentations: The Rationale for a Multinational Repository (Sean Tyson); Perspectives on the Future of the Multi-National Repository Concept (Alan Brownstein); Why Would any Country Accept Foreign Spent Nuclear Fuel? (Charles McCombie); Dual Track Strategy for Small Countries (Leon Kegel)

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DOE OSTI2020

Industry overview of TRANSCEND - 20271

The United Kingdom has a legacy of nuclear facilities to decommission and radioactive waste to manage and dispose of. The TRANSCEND programme is an independent research programme with significant participation from industrial organizations that are involved in that decommissioning programme and radioactive waste management. The participation has led to a programme that is both academically robust enough to attract significant funding from the EPSRC, the universities themselves and relevant enough to be supported financially by the industrial partners who see the needs and drivers. The programme offers benefits in terms of the development of people and in enabling and promoting innovation in both hardware and software tools. Industrial engagement is fundamental to being able to capitalise on that innovation. (authors)

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DOE OSTI2020

Preliminary Evaluation of Removing SNF from Nuclear Power Plant Sites - Pilgrim Site Visit - 20048

The U.S. Department of Energy Office of Integrated Waste Management (DOE-IWM)a conducted an evaluation of removing spent nuclear fuel (SNF) from the Pilgrim site on November 4-8, 2019. The Pilgrim site is located on the western shore of Cape Cod Bay in the Town of Plymouth, Massachusetts, about 61 km southeast of Boston and 71 km east of Providence, Rhode Island. Participants in the site visit included the U.S. Department of Energy, Oak Ridge National Laboratory, Pacific Northwest National Laboratory, the state of Massachusetts, the Federal Railroad Administration, the U.S. Coast Guard, the U.S. Army Corps of Engineers, the Mashpee Wampanoag Tribe, the Consolidated Group of Tribes and Organizations, the Council of State Governments-Eastern Regional Conference, the Pilgrim Nuclear Decommissioning Citizens Advisory Panel, and the Massachusetts Coastal Railroad. The Pilgrim site was found to have two transportation mode options for the removal of SNF, offsite rail access and onsite barge access. Two offsite heavy haul truck to rail transload locations were evaluated, one in Middleborough, Massachusetts, about 35-47 km from the Pilgrim site, and a second location in Middleborough about 36-43 km from the Pilgrim site. A passenger railroad terminates in Plymouth, Massachusetts, about 11 km from the Pilgrim site; however, this location is not suitable for freight rail involving large SNF transportation casks. There is also an onsite barge facility located on the Pilgrim site that was used during construction and would require refurbishment to be used. (authors)

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DOE OSTI2020

Waste Shipments to and from the Nevada National Security Site, Radioactive Waste Management Complex (Q3 Transportation Report, FY20) Transportation Report

This report satisfies the U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office (NNSA/NFO) commitment to prepare a quarterly summary of waste shipments to the Nevada National Security Site (NNSS) Radioactive Waste Management Complex (RWMC) in Area 5.

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DOE OSTI2021

Nuclear Fuels and Reprocessing Technologies: A U.S. Perspective

Reprocessing and/or waste management issues are of concern to the “back end” of the nuclear fuel cycle. Of course, there are a great many “nuclear fuel cycle” scenarios to consider; if not in practice, then at least in theory. The simplest conceptually is the “once through” fuel cycle in which the spent fuel is discarded. The more complex fuel cycle scenarios involve reprocessing spent nuclear fuels and a family of nuclear reactor technologies to accommodate burning and breeding for various military and commercial needs. Therefore, the selection of a specific “fuel cycle” is what ultimately imposes the engineering requirements of the reprocessing and waste management technologies. No one part is independent of the other parts in a fuel cycle flowsheet; all parts are fully integrated. This paper presents a summary of radiochemical processes, nuclear reactor technologies, nuclear fuel types, and the reprocessing technologies that serve the different nuclear fuel types. Comprehending how this series of topics are related to each other is a prerequisite to understanding the requirements of any reprocessing strategy. The summary materials presented here are selective, as opposed to comprehensive. More detailed information on any one subject can be found in the reference materials.

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DOE OSTI1989

Nuclear Safety [Vol. 30, No. 3, July-September 1989]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 325 Safety of Framatome Advanced Nuclear Steam Supply Systems Designs by J. A. Charles and D. Lange, 333 Book Review of Nuclear Accidents: Intervention Levels for the Protection of the Public by H. B. Piper; ACCIDENT ANALYSIS: 335 Living PRA Computer Systems by S. C. Dinsmore and H.-P. Balfanz, 343 Summary of ICAP Assessments of RELAP5/MOD2 by W. E. Driskell and R. G. Hanson; CONTROL AND INSTRUMENTATION: 352 Thermal Performance Monitoring System at Maanshan Nuclear Power Plant by H.-J. Chao, Y.-P. Lin, G.-H. Jou, L.-Y. Liao, and Y.-B. Chen; DESIGN FEATURES: 358 Warning Systems for Nuclear Power Plant Emergencies by J. H. Sorensen and D. S. Mileti; WASTE AND SPENT FUEL MANAGEMENT: 371 Activities Related to Waste Management Compiled by E G. Silver; OPERATING EXPERIENCES: 382 Steam Generator Tube Performance: Experience with Water-Cooled Nuclear Power Reactors During 1985 by O. S. Tatone and R. L. Tapping, 400 Systems Interaction Analyses: Concepts and Techniques (Part II) by M. D. Muhlheim and G. A. Murphy, 413 Reactor Shutdown Experience Compiled by J. W. Cletcher, 416 Operating U.S. Power Reactors Compiled by E G. Silver; RECENT DEVELOPMENTS: 440 General Administrative Activities Compiled by E G. Silver, 460 Reports, Standards, and Safety Guides by D. S. Queener, 466 Status of Power-Reactor Licensing Activities Compiled by E G. Silver, 470 Proposed Rule Changes as of Mar. 31, 1989; ANNOUNCEMENTS: 334 Proceedings Published, 351 CEC Seminar on Methods and Codes for Assessing the Off-Site Consequences of Nuclear Accidents, 357 Short Course on Multiphase Flow and Heat Transfer: Bases and Applications in A: The Nuclear Power Industry B: The Process Industries, 357 International Conference on Probabilistic Safety Assessment and Management, 478 International Topical Meeting on the Safety, Status, and Future of Non-Commercial Reactors and Irradiation Facilities, 475 The Authors.

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DOE OSTI2020

The Evaluation of Chemical Decontamination for Contaminated Carbon Steel Scrap by Multiphase Treatment Process - 20249

There is a significant growth in the field of decommissioning of nuclear facilities due the fact that for a lot of them their end of life is nearing. In Taiwan, the government plans to steadily reduce the dependence on nuclear power and moves gradually towards a nuclear-free homeland. Therefore, the development of decommissioning technology becomes the significant issue for the next coming years. The decommissioning of a nuclear power plant is a long process and practical work including radiological surveys, decontamination, dismantling, and the subsequent waste management. Chemical decontamination is an important task in the decommissioning of nuclear power plants and is usually performed by circulating the chemical solution in selected systems or full systems in order to reduce dose levels. However, the decontamination of segmented metal parts is also considered to reduce the contamination to such levels that they may be disposed of at a lower classification or be unconditionally released, and it helps to achieve the goal of waste minimization and provide a more economical cost for disposal management. Chemical decontamination technology is used to remove the contaminated oxides deposited on the surface of metals by dissolving them with chemical mechanisms, such as chelation, acidic or alkaline dissolution, redox reaction and so on. Carbon steel is the major material used in systems and pipelines in the nuclear power plant, thus it is also commonly selected as the target to be decontaminated in the decommissioning process. In this study, the effectiveness of chemical decontamination is evaluated by using multiphase treatment process including the reduction and the oxidation steps. In the step of redox reaction, the oxidation state of the substance is very important. Especially metals of some oxidation states show solubility in water, which has a decisive influence on the chemical decontamination. The experiment of decontamination of metal scraps is carried out by the immersion method within a chemical solution. The relative operation parameters such as temperature, pH, and concentration of the chemical solution are recorded to demonstrate the effect of the decontamination process. (authors)

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DOE OSTI2020

Grain boundary facilitated dissolution of nanocrystalline NpO 2 (s) from legacy waste processing

Dissolution of actinide dioxides, including neptunium dioxide (NpO 2 (s)), is paramount for the prediction of the environmental fate of nuclear materials. Quantifying dissolution rates, as well as understanding qualitative dissolution mechanisms, informs performance assessment for geologic disposal of spent nuclear fuel and management of legacy radioactive waste. The aim of this research was to measure the dissolution rate of nanocrystalline NpO 2 (s), produced through legacy nuclear waste processing, under oxidizing conditions, as well as to characterize surface alteration to the material. The solid phase was characterized using electron microscopy techniques (SEM/STEM) and X-ray photoelectron spectroscopy (XPS), indicating preferential dissolution of Np-hydroxide contained in the grain boundaries of NpO 2 (s) and fragmentation of grains from the matrix. The oxidative dissolution was monitored over 40 weeks, yielding a two-step kinetic dissolution model involving hydration of NpO 2 (s) and subsequent oxidation and dissolution of the hydroxide phase. Here, the proposed dissolution models for nanocrystalline NpO 2 (s) suggest that microstructural features such as grain boundaries are key factors affecting dissolution, including release of colloidal particles, and ultimately, environmental fate and transport of nuclear materials.

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DOE OSTI2020

Enhanced Sustainability in Treatment of Contaminated Metals for Clearance and Recycling - 20290

Cyclife Sweden AB has, for more than 30 years, provided metal treatment services to the domestic and international nuclear industry with the aim of creating value through metal clearance and recycling. These operations are to a large extent based on the concept that the materials, after pre-treatment, including decontamination, will be melted for clearance. During the melting process, some nuclides separate from the metal while the remaining radioactivity becomes fully homogenized in the metal bath and the resulting ingots, along with the samples taken before casting At a time when activities to fight Climate Challenge are in focus, any savings in energy consumption, in addition to the positive environmental effect of recycling metals back to industry, is more than welcome. If the clearance of certain metallic items can be achieved safely, without using melting, it is the environmentally preferred option provided it carries a reasonable cost. It is even better if an object can go through a decontamination and clearance process prior to re-use inside or outside the nuclear sector. Cyclife Sweden is currently investing in a direct clearance process, to be operated in parallel with the melting process. This new process is expected to be fully operational in the late summer of 2020 and includes pre-treatment, decontamination and heavily shielded clearance measurement installations. The treatment will also allow clearance measurements of larger objects without the need for segmentation, along with allowing objects composed of different materials, like electric motors can be accepted for treatment. This paper gives an overview of the circular economy aspects in a nuclear back-end context, the Cyclife approach to further implement a circular economy approach in the processes, Material Acceptance Criteria for treatment, the facility set-up for the new process and how the melting route and the direct route can be combined in an efficient industrial scale process. To put the new capabilities into perspective, the two different treatment and clearance approaches, i.e. direct clearance and melting, are to be compared. (authors)

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DOE OSTI2020

Treatment of Problematic Reactive Metal Wastes Using GeoMelt ICV - WM2020 Conference Paper

Decommissioning of sodium-cooled reactors and fast reactor technologies has generated a number of reactive metal waste configurations that are problematic to treat and typically lack cost effective treatment methods and disposition options. As a result, Veolia Nuclear Solutions, under contract with Idaho National Laboratory (owned by the U.S. Department of Energy and managed and operated by Battelle Energy Alliance, LLC) demonstrated its GeoMelt® In-Container Vitrification (ICV)™ technology to safely convert sodium metal to a non-reactive vitrified oxide form. The demonstration project, supported by glass formulation and crucible testing, consisted of a series of ICV™ melts that processed elemental sodium into stable non-reactive glass. INL is currently implementing GeoMelt® technology as a means to safely and reliably convert radioactive reactive metal residues that contaminate sodium cooled reactor components into waste forms that comply with existing disposition pathways. Reactive metal wastes require treatment in order to remove the Resource Conservation and Recovery Act (RCRA) reactivity and ignitability characteristics to comply with land disposal restrictions. GeoMelt®, which is an alternative to other potential treatment approaches, provides a robust approach that chemically converts the reactive metals to an inert oxide while also immobilizing radionuclides in a vitrified waste form with durability equal to or better than vitrified nuclear fuel reprocessing wastes (very robust and inert waste forms). Most other treatment approaches generate hydrogen gas which is problematic. In 2016, Veolia Nuclear Solutions first demonstrated the effectiveness of the GeoMelt® ICV™ process in deactivating reactive sodium metal. Crucible, bench-scale, and engineering-scale demonstrations were conducted on several surrogate waste configurations with various ratios of sodium metal and glass-formers. Each ratio and configuration demonstrated complete deactivation of the surrogate sodium metal. Follow-on work in 2017 demonstrated the deactivation of reactive sodium by GeoMelt® ICV™ at a higher waste loading relative to previously demonstrated work performed in 2016; the higher waste loading optimized glass chemistry while enhancing the economical full-scale treatment of reactive metals. Additionally, follow-on demonstration testing in 2018 and 2019 focused on more complex shapes and other reactive-metals (mocked up Experimental Breeder Reactor II [EBR-II] subassembly, sodium filled heat exchanger, and a can containing sodium potassium alloy) which were all performed at engineering-scale. Since 2018, a 10-metric ton full-scale GeoMelt unit (GeoMelt® Richland) was designed, installed, and commissioned at Perma-Fix Northwest in Richland, Washington for the treatment of reactive metal wastes. As of September 2019, over 900 55-gallon drums containing a total of around 3,500 lb of sodium with low levels of radioactivity have been treated at GeoMelt ® Richland, with resulting glass monoliths disposed at the Nevada National Security Site (NNSS). A full-scale radiological demonstration melt on an actual EBR-II subassembly has also been performed using the full-scale melter in 2019. The GeoMelt® technology is a proven radioactive waste treatment technology capable of immobilizing radioactive wastes, including bulk rubble such as drums and other steel vessels usually without pretreatment. Utilizing the GeoMelt® technology to treat reactive metals eliminates pretreatment steps resulting from having to separate the reactive metal from steel containers or jackets as GeoMelt® can easily operate at temperatures sufficient to melt the steel and expose the reactive metal for treatment. Eliminating handling steps of reactive metals is a significant safety advantage since reactive metals are pyrophoric. The results generated as a part of the 2018-2019 demonstration program are presented in the paper.

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DOE OSTI2021

On-line Waste Library Supporting Information

The On-Line Waste Library is a website that contains information regarding United States Department of Energy-managed high-level waste, spent nuclear fuel, and other wastes that are likely candidates for deep geologic disposal, with links to supporting documents for the data. This report provides supporting information for the data for which an already published source was not available.

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DOE OSTI2021

Demonstration of the On-the-Fly Shielding Analysis Method: Spent Fuel and Waste Disposition

This report documents work performed supporting the US Department of Energy (DOE) Office of Nuclear Energy (NE) Spent Fuel and Waste Disposition (SFWD) Integrated Waste Management activities under work breakdown structure element 1.08.02.04.01, “Data and Tools Development, Validation, and Maintenance.” In particular, this report fulfills milestone M3SF-21OR020401016, “Implement on-the-fly dose analysis methodology in UNF-ST&DARDS” within work package SF-21OR02040101, “Commercial SNF Characterization - ORNL.” The Used Nuclear Fuel - Storage, Transportation & Disposal Analysis Resource and Data System (UNFST& DARDS) enables automated dose rate calculations for spent nuclear fuel (SNF) transportation packages and storage casks using a Monte Carlo radiation transport code. The explicit method uses a detailed model of the SNF system and its contents. Therefore, a dose rate calculation is required for each as-loaded transportation package or storage cask because the SNF assemblies within a canister typically have unique irradiation characteristics. An alternate method, referred to as the “on-the-fly” shielding analysis method, has been proposed that requires only a set of Monte Carlo dose rate calculations for each transportation packaging/storage cask design. The results of the Monte Carlo dose rate calculations are independent of the SNF assembly irradiation and decay characteristics. The dose rate values may then be combined with the radiation source strength of the SNF assemblies associated with a particular transportation packaging/storage cask design to determine actual dose rates. This report presents on-the-fly dose rate calculations for a representative SNF storage cask and verification of the on-the-fly dose rate calculation results by comparison with reference dose rate calculations using the explicit Monte Carlo dose rate calculation. The on-the-fly shielding analysis method was implemented in UNF-ST&DARDS. A Python program was developed to process the MAVRIC dose rate results obtained by source particle type, energy group, and fuel geometry region. A Python processor created binary files, which were saved as a special UNF-ST&DARDS library for on-the- fly shielding analyses. UNF-ST&DARDS uses the precalculated on-the-fly binary libraries generated by the Python data processor and directly executes the Python code for on-the-fly dose analysis. This Python code unzips the pre-generated binary files mentioned above, reads the data, and combines them with user-specified sources for dose and uncertainty calculations. The Python programs were verified using Excel calculations and by comparison with the values obtained with the MAVRIC post-processing utilities applied to the 3dmap files. This method can currently be used to determine dose rates for as-loaded HI-STORM FW storage casks. The UNF-ST&DARDS analysis wizard for on-the-fly shielding analysis is described in this report.

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DOE OSTI2020

Systematic Analysis of Drivers and Barriers of Cross-Sectorial Learning Between Nuclear and Oil and Gas Decommissioning Projects - 20139

Project and megaproject management in the energy sector have historically focused on the planning and delivery of the construction of infrastructure. Therefore, policies are predominantly oriented to support the construction of infrastructure rather than its decommissioning. However, globally, several facilities have reached (or will soon reach) their end-of-life and need to be decommissioned. These facilities span across the energy sector, including nuclear power plants and oil and gas rigs, whose decommissioning present unprecedented techno-socio-economic challenges. Moreover, globally, their decommissioning cost is estimated to reach hundreds of billions of US dollars and this budget keeps increasing, with limited knowledge-transfer across industries to mitigate the spiraling increase of this figure. Cross-sectorial knowledge-transfer is a way to tackle this matter and improve the planning and delivery of decommissioning and waste management projects. The aim of this research is to build a road-map based on cross-sectorial industrial experience, designed to promote the sharing of good practices between projects both within and across industries dealing with major decommissioning and waste management challenges. To reach this aim, the tacit knowledge of senior practitioners on several large scale nuclear and oil and gas decommissioning projects is made explicit by the means of semi-structured interviews, systematically analyzed through directed content analysis. The results of this analysis include a list of good practices that could and should be transferred to nuclear decommissioning projects, including the need to promote the adoption of incentives explicitly linked to the project deliverables, and to develop long-term plans to support smooth transition from operations to decommissioning. Suggestions about how the knowledge-transfer can be promoted within and across organizations are also discussed. In this way, this research provides practical guidance for project managers and policymakers on how to support the planning and delivery of projects with improved performance. (authors)

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DOE OSTI2021

Annual Summary Report for the Remote-Handled Low-Level Waste Disposal Facility—FY 2020

This Fiscal Year (FY) 2020 annual summary report documents the continued adequacy of the performance assessment (PA), the composite analysis (CA) and associated operating disposal-authorization statement (ODAS) technical basis documents for the Remote Handled (RH) Low-Level Waste (LLW) Disposal Facility at Idaho National Laboratory. Annual review of the adequacy of the PA and CA for the RHLLW Disposal Facility ensures that conclusions of the analyses remain valid in accordance with requirements of Department of Energy (DOE) Order 435.1 Change 1, “Radioactive Waste Management.” In FY 2020, no significant operational changes or other activities occurred that would cause deviation from the assumptions in the PA and CA pertaining to disposal geometry, verification of waste characteristics, tracking disposal inventories against total limits, facility closure design, or institutional controls. Fifteen total waste-canister shipments were received at the RHLLW Disposal Facility, and 15 total waste canisters were emplaced.

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DOE OSTI2022

NEVADA NATIONAL SECURITY SITE 2021 WASTE MANAGEMENT MONITORING REPORT AREA 3 AND AREA 5 RADIOACTIVE WASTE MANAGEMENT SITES

Environmental monitoring data are collected at and around the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) within the Nevada National Security Site (NNSS). This report summarizes the 2021 environmental data to provide an overall evaluation of RWMS performance and to support environmental compliance and performance assessment (PA) activities.

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DOE OSTI2020

Panel Session 1: WMS 2020 Plenary Session

The plenary panel focused on government and industry world leaders speaking on the pressing issues facing radioactive waste management in 2020 and beyond. Panelists included an OECD/ NEA representative for the international community, a Huntington Ingalls Industries representative for the commercial sector, and the US DOE Senior Advisor to the Under Secretary of Science representing DOE Environmental Management. Panelists with presentations: WM2020 Speech (Michael Lempke); Optimization: A Better Path toward Managing Nuclear Decommissioning and Legacies (William Magwood); Update On the DOE Environmental Management Program (William White)

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