Challenges in Explicit Modeling of TRISO in MCNP [Slides]
Current capabilities in MCNP allow for full modeling of TRISO particles to handle uncertainties.
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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.
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Core Physics Characteristics of Extended Enrichment and High Burnup Boiling Water Reactor Fuel
This paper presents the highlights of boiling water reactor (BWR) core physics studies performed at Oak Ridge National Laboratory as part of a series of studies conducted to compare low-enriched uranium (LEU) with LEU+ fuel. The studies analyzed isotopic fuel content, lattice parameters (Phase 1), and core physics (Phase 2) to identify challenges in operation, storage, and transportation for BWRs and pressurized water reactors (PWRs). Because of a lack of publicly available lattice and core designs for modern BWR fuel assemblies and reactor cores, several optimized lattice designs were generated, and different core loading strategies were investigated. Twelve optimized lattice designs with 235 U enrichments ranging from 1.6% to 9% and gadolinia loadings ranging from 3 to 8 wt% were used to model axial enrichment and geometry variations in fuel assemblies for core designs. Each core shares a common set of approximations in design and analysis to allow for consistent comparisons between LEU and LEU+ fuel. The objective is to highlight anticipated changes in core behavior with respect to the reference LEU core. The results of this study show that the differences in LEU and LEU+ core reactor physics characteristics are less significant than the differences in lattice physics characteristics reported in the Phase 1 studies.
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The Deimos Experiment: Advanced Reactor Testbed
Advanced reactor initiatives are growing significantly through programs nationwide. This research area includes small modular reactors, microreactors, and space reactors. Many of the reactors being designed are untested concepts. They include unique moderators, varying fuel types, high temperatures, and compact configurations. The shift in fuel type, from highly enriched uranium (HEU) to high assay, low enriched uranium (HALEU), is particularly important as it has driven many of the other changes. For example, lower enrichment requires advanced moderators, which in turn require different reflectors to make the systems compact. The change in materials including the transition from HEU to HALEU affects the temperature feedback of the systems. Additionally, these advanced reactor concepts generally have a thermal neutron spectrum in contrast to earlier fast spectrum advanced reactor. With the extensive changes from previous reactor designs, validation experiments are needed. The National Criticality Experiments Research Center (NCERC) is uniquely equipped to perform such experiments. The Deimos experiment, designed for execution at NCERC, will serve as a testbed for advanced reactor concepts. It will use HALEU fuel in a graphite matrix, provide the ability to use advanced moderators, and allow measurements of temperature reactivity coefficients (TRCs).
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A Comprehensive Assessment of the Viability of Small Modular Reactors in Africa: A Nuclear Security and Nonproliferation Perspective
In response to escalating global energy demands driven by industrialization and the pressing need for decarbonization, this paper explores the potential of Small Modular Reactors (SMRs) as a sustainable energy solution in Africa. Focusing on nuclear security and non-proliferation concerns, the study assesses Africa's energy landscape, emphasizing the need for diverse and reliable power sources. While highlighting the scalability and cost-effectiveness of SMRs, the analysis acknowledges potential challenges associated with their introduction, particularly concerning nuclear security and non-proliferation. Given the recommendation to use High Assay Low Enriched Uranium (HALEU) in some SMRs, it is important to critically consider the security implications of transporting nuclear materials, the proximity of the public to the plant, and the time taken to respond to planned assaults. The possibility of using HALEU makes the nuclear material more prone to adversaries such as sabotage, theft, and terrorism. Utilizing PESTLE analysis, this research seeks to outline the detailed political, economic, social, technological, environmental, and legal readiness of Africa to embrace the first-of-a-kind technology (SMR) while fulfilling its mandate to the Non-Proliferation Treaty (NPT). Examining regulatory frameworks, international cooperation, and safety protocols, the study underscores the importance of regional collaboration to prevent the misuse of nuclear technology for military and malicious intent. Drawing insights from successful case studies, the paper concludes by synthesizing key findings and proposing recommendations for policymakers and stakeholders. These recommendations encompass regulatory enhancement, capacity building, technology transfer, and diplomatic efforts to strengthen nuclear security, non-proliferation, and safeguards in Africa.
Prah, Christina↗
Isotopic Characterization of HALEU from EBR-II Driver Fuel Processing
Provide the isotopic characteristics of the High Assay Low Enriched Uranium (HALEU) product recovered during the used nuclear fuel treatment operations occurring at Idaho National Laboratory. Revision one of this document (TEV-3537 Rev 1) provides isotopic characterization for material that has been recast using the drip cast crucible gravity casting technique used to transform traditional EBR-II uranium ingots into regulus.
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History and status of spent fuel treatment at the INL Fuel Conditioning Facility
Experimental Breeder Reactor II (EBR-II) had an illustrious operational history from 1964 to 1994 as one of the first sodium-cooled fast reactors to use sodium-bonded metallic driver fuels and blankets. EBR-II was a testbed for fuel development and reactor design principles. Following shutdown, it was decided to use electrometallurgical methods to treat the spent fuel and blanket materials for final disposition. Here, this paper provides a brief history of U.S. liquid metal cooled reactor development with emphasis on EBR-II, and a technical description of the electrometallurgical methods.
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Strategies for thorium fuel cycle transition in the SD-TMSR
Liquid-fueled Molten Salt Reactor (MSR) systems represent advances in safety, economics, and sustainability. The MSR has been designed to operate with a Th/ 233 U fuel cycle with 233 U used as startup fissile material. Since 233 U does not exist in nature, we must examine other available fissile materials to start up these reactor concepts. This work investigates the fuel cycle and neutronics performance of the Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR) with different fissile material loadings at startup: High Assay Low Enriched Uranium (HALEU) (19.79%), Pu mixed with HALEU (19.79%), reactor-grade Pu (a mixture of Pu isotopes chemically extracted from Pressurized Water Reactor (PWR) spent nuclear fuel (SNF) with 33 GW d/tHM burnup), transuranic elements (TRU) from Light Water Reactor (LWR) SNF, and 233 U. The MSR burnup routine provided by SERPENT-2 is used to simulate the online reprocessing and refueling in the SD-TMSR. The effective multiplication factor, fuel salt composition evolution, and net production of 233 U are studied in the present work. Additionally, the neutron spectrum shift during the reactor operation is calculated. The results show that the continuous flow of reactorgrade Pu helps transition to the thorium fuel cycle within a relatively short time (≈ 4.5 years) compared to 26 years for 233 U startup fuel. Finally, using TRU as the initial fuel materials offers the possibility of operating the SD-TMSR for an extended period of time (≈ 40 years) without any external feed of 233 U.
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A High-Assay Low-Enriched Uranium Fuel Transportation Concept
The uranium 235U enrichment commonly used in fuel production for U.S. light water nuclear reactors typically does not exceed 5 wt%. In contrast, many of the currently investigated advanced reactor concepts demand fuel with higher enrichments. This includes high-assay low-enriched uranium (HALEU), characterized by a 235U enrichment of 5 to 20 wt%. The necessity of HALEU transportation in the fuel production cycle leads to new challenges caused by various technical and regulatory hurdles. Current U.S. Nuclear Regulatory Commission–approved transportation package designs for UF6 with enrichments above 5 wt% provide relatively small payloads [=116 kg (250 lb)]. Furthermore, in accordance with 10 CFR 71.55, package design activities for fissile material enriched above 5 wt% need to consider water infiltration in the containment as part of the criticality safety evaluations. This study presents a transportation package concept for HALEU advanced nuclear reactor fuel with a significantly higher payload of up to 376 kg (830 lb) of fissile material per package and up to 1881 kg (4149 lb) of HALEU per legal weight truck. The anticipated chemical form of the transported material is UO2 downblended from available highly enriched uranium. The concept utilizes a combination of existing transportation packaging, 18 inner canisters, and a novel basket design that includes a borated aluminum flux trap. Criticality and shielding evaluations; fundamental structural, confinement, and thermal assessments; and studies on package operations are presented. The results of this study build significant confidence in the technical feasibility of a high-capacity HALEU transportation package concept while demonstrating the concept’s potential to meet U.S. regulatory requirements.
Eidelpes, Elmar↗
Assessment of Existing Transportation Packages for Use with HALEU
Commercial light water reactor operators and fuel vendors in the United States are pursuing changes to fuel that include increased 235 U enrichment. Economic studies generally anticipate maximum near-term fuel assembly designs with up to 8 wt.% 235 U. Many next-generation nuclear reactor designs require high-assay low-enriched uranium (HALEU) (19.75 wt.% > 235 U > 5 wt.%) fuel. One necessary element for the commercial-scale use of HALEU is the ability to safely transport large quantities of enriched fuel material in multiple forms. However, there is uncertainty as to whether subcriticality requirements can be satisfied with existing package designs and whether existing critical benchmark experiment data are sufficient to support criticality safety code validation for HALEU transportation applications. This study assesses the potential to use currently licensed transportation packages for the transportation of increased enrichment unirradiated U fuel forms. The assessment uses selected package designs that represent the five categories of fuel form-boiling water reactor pins and assemblies, pressurized water reactor pins and assemblies, UF 6 , U-metal and tristructural isotropic (TRISO) particles, and UO 2 pellets or powder-and focuses on demonstrating subcriticality and identifying benchmark critical experiments appropriate for use in criticality computer code validation. Key quantities of interest that relate to subcriticality are limiting conditions (e.g., optimum moderation), package or package array k eff , package capacity, and package transportation array size. The SCALE TSUNAMI-3D and TSUNAMI Indices and Parameters codes are used for sensitivity and uncertainty calculations and for identification of candidate critical benchmark experiments for code validation. The primary metric for identifying candidate benchmarks is the similarity coefficient, c k . For each fuel form category, a representative package is evaluated. Results provided for each package evaluation include enrichment and packaging limits (e.g., maximum transportation array size as a function of enrichment) and benchmark critical experiment similarity coefficients. Results indicate that there are viable means for increasing enrichments into the HALEU range across the spectrum of fuel forms with differing increase amounts available for different packages. Sources of subcriticality margin to offset increased enrichment reactivity include reduced transportation array size, reduced fissile mass, burnable absorber credit, and safety analysis margin harvesting. For all packages except the DN-30, numerous critical benchmark experiment candidates for validation were identified.
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Potential Fuel Cycle Cost Reductions of Once Through HALEU Reactors
This report looks to identify potential cost reduction opportunities for SFR and HTGR reactors using HALEU fuel. This analysis focuses on how learning rates and experience from other industries could translate to future HALEU fueled reactor fuel cycles. Various fuel loading and residence scenarios are evaluated to estimate areas of potential cost savings. Cost savings from location optimization is explored for both fresh and spent nuclear fuel. Additional analysis was completed to estimate cost savings through improved labor productivity.
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Advanced Fuel Cycle – Cost Basis Report: Module C3 High-Assay Low-Enriched Uranium (HALEU) Enrichment and Deconversion/ Metallization
This is a cost module that provides estimates of the cost of HALEU enrichment.
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Demonstration of In Situ Down Blending During ZIRCEX Process
The Material Recovery Pilot Plant has successfully demonstrated in situ down-blending of HEU fuel to support INL’s commitment to HALEU production technologies. Depleted uranium pellets, along with an HEU fuel piece were placed in the hydrochlorinator and subsequently declad and oxidized. UOx product was recovered, analyzed, and demonstrated to have reduced enrichment within the targeted range of 20 – 30 % U-235. This first attempt at in situ down-blending will be followed by continued process refinement over the remaining fuel operations at MRPP.
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AFC Metal Fuel Workshop Presentations
Compilation of PowerPoint slide decks from presentations given at the 3rd Annual AFC Metal Fuels Workshop. The purpose of this workshop is to bring the national metallic fuel stakeholders together to discuss status and needs of metallic fuel and review the DOE AFC Metal Fuel R&D. Desired Outcomes - Advance Metallic Fuel Technology: 1. Stakeholder Community Building 2. National Program R&D Accountability, Planning Support, and Increased Stakeholder Advocation.
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Initial Evaluation of Microreactor Disposition Options
The United States Department of Energy is supporting the U.S advanced reactor industry through funding, legislation and regulatory development to actively pursue several microreactor design concepts. The Idaho National Laboratory (INL) is strategically positioned to support demonstration of microreactor technology in the next three to five years. This report provides and initial evaluation of the disposition options for microreactor spent nuclear fuel (SNF) generated as part of the microreactor technology demonstration program. Currently available information constitutes the basis for the options identified and discussions thereof. In the absence of detailed microreactor design information, assumptions were made to facilitate the identification of disposition options Particulars pertaining to any component of an identified disposition pathway are naturally highly microreactor design specific and in general such details are not provided. The diverse nature of potential microreactor SNF is reflected in the diverse nature of DOE owned SNF stored at INL. Therefore, it is anticipated that DOE currently stores and manages fuels that can serve as analogs for most microreactor fuel concepts. As such disposition options for microreactor SNF are expected to be much the same as that for these existing fuels. Two generic microreactor concepts have been selected for the purposes of this options assessment. The selected reactor concepts are a tristructural isotropic (TRISO) fueled high temperature gas reactor concept and a sodium/potassium bonded heat pipe reactor with uranium oxide fuel. Both concepts are assumed to be using high-assay low enriched uranium (HALEU) as the initial fuel composition. Interim storage, treatment and neutralization, material recovery, packaging and extended dry storage options are identified these reactors. The disposition options include existing INL facilities and capabilities and new facilities and capabilities developed as part of the microreactor program or as part of DOE's overall strategy for the eventual transfer of all SNF at INL to a permanent repository.
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Comparison of Nuclear Thermal Propulsion Reactor Cores Using Different Fuel Forms, Neutron Spectra, and Uranium Enrichments
A comparison of 300 MWt HEU and HALEU NTP cores using different fuel forms is presented to demonstrate trends and potentially inform fuel development. Compared to HEU, HALEU cores require additional in-core moderation. Beryllium was chosen as the reference in-core moderator material; metal hydrides may also be considered. Typical NTP fuel forms such as carbide and carbonitride solid solutions, carbide-graphite composites, CERMETs and CERCERs were considered together with less typical fuel forms such as UC and U metal. Be-moderated HALEU cores are heavier than HEU cores (1.9-2.7 and 0.9-2.0 metric tons, respectively). Furthermore, HALEU loading is very sensitive to the amount of in-core moderation which translates into tight fabrication tolerances. HALEU cores requires more reactivity control mechanisms than HEU cores to compensate for temperature feed-back. Using enrichments only slightly above HALEU (e.g., 25%) may provide a compromise between the administrative preference for using HALEU and the engineering preference for simplifying the reactor systems by, for example, not using in-core moderators.
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Devices and systems for material transportation
Various embodiments relate to devices for transporting high-assay low-enriched uranium (HALEU). A device may include at least one section, wherein each section of the at least one section includes a number of storage tubes. Each storage tube, which is configured to receive and hold a container, extends from adjacent a first end of the section toward a second, opposite end of the section. Each section further includes a number of flux traps, wherein each storage tube of the number of storage tubes is at least partially surrounded by a flux trap of the number of flux traps Associated systems are also disclosed.
Eidelpes, Elmar F.↗
Estimated HALEU Requirements for Advanced Reactors to Support a Net-Zero Emissions Economy by 2050
The primary purpose of this report is to provide a documented estimate of High-Assay Low-Enriched Uranium (HALEU) needs that may be used in Department of Energy discussions with stakeholders as part of preparations to establish the HALEU Availability Program authorized by the Energy Act of 2020. This report employed a two-step process to estimate HALEU needs for commercial advanced reactors.
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Study on Application of Distributed Network of Sensors with List Mode for NMAC Literature Review
Nuclear material accounting and control (NMAC) for nuclear security detects, deters, and resolves questions related to unauthorized removal (i.e. theft) or misuse of nuclear material. NMAC also serves as a key insider threat mitigation measure and aids in recovery of nuclear material that is missing. Effective nuclear security depends on NMAC for timely and accurate information about nuclear material types, quantities, and locations. Bulk nuclear material processing facilities, however, present unique challenges for effective NMAC due to the presence of large quantities of material in-process and the accumulation of residual material holdup within process equipment. These holdup accumulations can obscure accurate physical inventory taking and complicate efforts to resolve NMAC irregularities at the facility level. Bulk material monitoring systems often rely on material balance calculations and indirect measurement techniques, which may mask protracted theft of smaller amounts of nuclear material. These monitoring limitations have generated increased interest in continuous monitoring technologies, including distributed non-destructive assay (NDA) sensor networks capable of providing real-time or near-real-time measurement of material movement and accumulation within bulk processing environments. Recent advancements in distributed networks of NDA radiation detectors and sensing technologies provide an opportunity to address these limitations. Although such distributed sensor networks have been implemented in select facilities for IAEA Safeguards applications, their potential for supporting NMAC functions specifically tailored to nuclear security objectives remains largely unexplored. Furthermore, emerging list-mode data acquisition technologies have reached high technology readiness levels, enabling time-correlated detection of nuclear events across multiple temporal scales. These capabilities provide enhanced opportunities for accurate holdup measurement, continuous process monitoring, and improved detection of material theft or misuse over time. The increasing global expansion of civil nuclear power and development of related bulk material processing facilities, including those supporting high-assay low-enriched uranium (HALEU) and other advanced reactor fuel fabrication, further increases the need for advanced measurement and monitoring strategies for NMAC.
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