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DOE OSTI2024

Perform Design Support with MCNP for New Measurements

This report incorporates our work carried out during our 5-month internship at LANL under an internship agreement with EAMEA (École des Applications Militaires de l’Énergie Atomique). After outlining the context in which we worked, we present our work as aid to modeling and predicting the neutronic behavior of nuclear systems, with a view to carrying out criticality experiments qualifying the MCNP code as part of innovative projects. Fourth generation reactors will enable to tackle a lot of issues such as environmental crisis, affordable energy access, and nuclear waste management. They seem to be one of the keys for a sustainable future. Most of the projects that emerge nowadays include the use of HALEU (high assay low enriched uranium) or MOX recycled fuels. Our projects are part of this dynamic and addresses concrete scientific research issues in the nuclear field. Studies of HALEU package are essential to anticipate the need, therefore the Optimus L (OPTImal Modular Universal Shipping cask technology) designed by NAC (Nuclear Assurance Corporation) international but filled with 20 % enriched uranium dioxide (UO 2 ) will be studied to support safe transportation. However, it seems there is no benchmark with a high correlation with the combination of this fuel and this package to validate MCNP simulations. As such, the study will focus on the development of new criticality safety benchmarks for this case. On the other hand, there is a great need for critical benchmarks in the intermediate energy range with MOX fuel. An IER (Integral Experiment Request) has then be requested to answer it through a partnership between French institution IRSN and U.S. Department of Energy's Nuclear Criticality Safety Program (NCSP). This experience planned for 2025 requires to gather calculated data through a MCNP model to be realized safely. Finally, a presentation of our one-week experience at the DAF as part of our discovery of criticality experiments will be introduced in Appendix 1: Week at the DAF (Device Assembly Facility)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2024

Design of Hypothetical Processes for the Production of 131 I and 99 Mo from Activation Targets

For over six decades, medical isotope production has been a high-priority focus of many research reactors across the globe. The majority of these isotopes were produced using highly-enriched uranium (HEU) or low enriched uranium (LEU) – delivering millions of doses of diagnostic and therapeutic isotopes. As a consequence of this production, however, six decades of isotope production has resulted in massive quantities of spent uranium material worldwide with no known disposition pathway creating growing proliferation concerns. Supported by the National Nuclear Security Administration’s (NNSA) Material Management and Minimization (M3) program, there has been increased focus in the production of high-priority isotopes without special nuclear materials or without uranium altogether. Isotope production via activation can potentially fulfill regional isotope demands – particularly in under-developed regions without access to isotope supply chains. The benefits of this approach would be a reduction in uranium proliferation risks, less special nuclear material wastes, and reduced risk of supply disruption in the likely event that major isotope producers will again go off-line as has happened in recent years due to a number of factors.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2024

Synthesizing Lithium Sorbents Using Acid Mine Drainage Solids

In this study, we prepared HMO from coal mine drainage (CMD) solids and examined the practicability of recovering Li from oil and gas produced waters and geothermal brines. We aim to innovate high-value added products from the existing extract REEs and CMs processing chain of legacy byproducts associated with coal production. CMD solids, which are precipitates collected from exhausted limestone gravels of passive CMD treatment beds, are enriched in rare earth elements (REEs) and other critical minerals (CMs), and therefore, considered as a viable alternative source for REEs/CMs to supply the Nation’s demands. A patented extraction process developed by NETL (TREE) has been demonstrated as a cost-effective and efficient approach to recover REEs/CMs from CMD solids. Here, we selected residuals of CMD solids from the TREE process that are enriched in manganese (Mn) (>20% w/w) and successfully prepared LMO using a solid-state method modified from Yoshizuka et al. (2002).

Cheng, Chin-Min↗

DOE OSTI2024

The University of Missouri Research Reactor (MURR) LEU Fuel Element Flow Test Preliminary Design

As part of the U.S. National Nuclear Security Administration’s (NNSA) mission to eliminate or minimize the civilian use of weapons-grade highly enriched uranium (HEU, ≥ 20 wt% U-235) fuels, the NNSA Material Management and Minimization (M3) Office of Reactor Conversion and Uranium Supply is collaborating with six U.S. high performance research reactors (USHPRR), including one critical facility, to convert from the use of HEU to low-enriched uranium (LEU, < 20 wt% U-235) fuel. Primary conversion objectives for the USHPRR are to develop LEU fuel element designs that will ensure safe reactor operations and maintain the existing experimental facilities performance. The work is being conducted through many interrelated activities by stakeholders across organizations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

High-Burnup BWR LOCA Burst Analysis Framework Development and Demonstration

Nuclear power currently contributes approximately 20% of total electricity generation in the United States and more than 10% globally. Given the increasing reliance on nuclear energy to achieve our nation’s goal of reaching net-zero carbon emissions by 2050, there is significant pressure on the existing nuclear industry to extend plant operational licenses and improve efficiency. This is crucial as the existing nuclear fleet serves as a vital bridge until new light water and advanced reactors can be developed and deployed, bolstering the supply of carbon-free energy to meet domestic demands. Operational costs primarily consist of plant operation and maintenance and fuel costs, influenced by materials and reactor core designs. These factors, coupled with heavily subsidized renewable energy markets, create a challenging economic environment for the existing light water reactor fleet, as well as for new build projects. To address these economic challenges, the nuclear industry has developed a strategic blueprint aimed at enhancing nuclear power’s economic sustainability. Past initiatives, such as efforts to eliminate fuel failures by 2010 and reduce operating costs by 30% before 2020, have laid the groundwork. Optimizing core design parameters, including burnup limits and enrichment levels, can lengthen cycles, reduce outages, reduce batch reload batch fractions and spent fuel storage requirements, and lower maintenance and operating expenses, thereby enhancing economic viability. In the United States, boiling water reactors (BWRs) comprise approximately one-third of the fleet, although much of the research and development focus has traditionally been on pressurized water reactors (PWRs). Advances in modeling and simulation, particularly through the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, are crucial to the long-term viability of BWRs, just as they are for PWRs. A key research area of the high burnup/increased enriched fuel initiative is focused on addressing loss-of-coolant-accident (LOCA)-related issues. NEAMS has dedicated significant effort to enhancing tools to better support BWRs, with a current focus on showcasing the BWR framework for high-burnup LOCA analysis. This high-fidelity work will demonstrate a best estimate pin-by-pin high-burnup BWR LOCA analysis to assess full-core cladding rupture behavior. This modeling capability will help with better understanding and realistic evaluation of fuel fragmentation, relocation, and dispersal (FFRD) phenomena at BWRs, which then could be used to prevent FFRD at BWRs without penalizing operational parameters. In addition, the results of this work will help identify strategies to identify additional margins or to potentially limit cladding rupture through core design optimizations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Rare Earth Elements (REE) and Critical Minerals (CM) in Middle Pennsylvanian-Age Coals and Associated Sediments in the Central Appalachian Basin, Eastern U.S.A.

The Central Appalachian Basin (CAB) of Kentucky, Tennessee, West Virginia and Virginia has a long history of coal mining and oil and gas extraction that has empowered the regional and national economies, the development of infrastructure, and a highly trained energy resources work force. As our societal demands for advanced technologies have rapidly increased in recent years, coal-related materials are viewed as an important new unconventional domestic source of critical minerals (CM) that are required for telecommunications, aerospace and transportation industries, electronics, the transition to low-carbon emissions energy production, and many consumer products. Coal-related materials encompass coal, associated sediments, coal mining waste materials, produced waters, and ash residues from coal-fired power plants. An important objective of the Evolve Central Appalachia Project (Evolve CAPP), sponsored by the U.S. Department of Energy (DOE) National Energy Technology Laboratory (NETL) is to assess the quantity and distribution of CM resources in the CAB region. The rare earth elements (REE) are considered highest priority, although other important CM such as niobium, gallium, and zirconium are known to occur in the Middle Pennsylvanian-age coals and sediments. Working with coal industry partners who provided access to drill cores, coal-related sediments and waste materials, over 600 samples have been collected for laboratory analysis, and over 730 materials have been scanned using portable x-ray fluorescence (pXRF) equipment. The application of pXRF provides the means for real-time semiquantitative analysis of CM content at very close spacings (typically 2-3 inch intervals) along drill core and in-situ channel samples that span the roof, coal seam, and floor rock. The comparison of pXRF geochemical data with laboratory results, geologic data, and downhole spectral gamma logs can provide high resolution input to lithologic and depositional models for future CM resource evaluations. The preliminary findings show that pXRF is capable of accurately measuring low concentrations of many of the CM with a high level of confidence (Ba, Cr, Ga, K, Nb, Rb, Sr, Th, Y), whereas for others (La, Ce, Co, Mn, Nd, Ni, Sc, Ti, V) the detection limits are very high or spectral interferences increase the uncertainty. Notably, the mean abundances of Y (34 ppm), La (86 ppm), Ga (40 ppm), and V (146 ppm) in coal underclays in the CAB region are up to 7X enriched compared with the overlying coal. These values also exceed the reported concentrations in published reference materials for upper continental crustal rocks (Rudnick and Gao, 2003), North American Shale Composite (Gromet et al., 1984), and North American coal (Finkelman, 1993). The pXRF data are in part verified by laboratory results that indicate the mean Y abundance (36 ppm) is highly correlated (R2 = 0.807) with total REE (ΣREE). The correlation is even higher (R2 = 0.957) with heavy REE (ΣHREE). Applying these correlations to the pXRF data for the coal underclays, the mean estimated values for ΣREE+Y and ΣHREE+Y are 270 ppm and 58 ppm, respectively. Although these average values are not considered high, the range of Y measured by pXRF in the coal underclays extended as high as 114 ppm, which would suggest ΣREE+Y equal to 847 ppm. The mean abundances of Zr (192 ppm) and Th (21 ppm) in coal underclays are also enriched compared with the overlying coal and these results likely reflect the presence of resistant detrital heavy minerals such as monazite, xenotime, and zircon in the underclay matrix. Several of the profiled coal seams and associated wall rocks contained thin volcanic ash layers up to 4-5 inches in thickness. The extent to which these ash fall layers provided a source for CM under the paleoenvironmental conditions that resulted in coal deposits in the CAB remains to be fully studied. Continuing investigations in the Evolve CAPP study area will include laboratory determinations of mineralogic and clay compositions, and evaluations of CM geochemical mobility in the coal and coal underclays.

Lassetter, Billy↗

DOE OSTI2025

Noninvasive Thermal Mass Flow Meter for Safeguards

This project addressed the long-standing nuclear safeguards problem of detecting in a timely manner the unauthorized production of significant quantities of highly enriched uranium in gas centrifuge enrichment plants (GCEPs).

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗

DOE OSTI2022

Improvements to the 115g/115m Cd Nuclear Data

Repeated measurements of the isotope 115m Cd by Los Alamos National Laboratory (LANL), Pacific Northwest National Laboratory (PNNL), and Atomic Weapons Establishment (AWE) have revealed a consistent discrepancy between beta and gamma counting. This suggests that a significant disparity exists between the true value of the isotope’s gamma-ray branching ratios and the values reported by the National Nuclear Data Center (NNDC). Enriched 114 Cd was irradiated using thermal neutrons to produce a high-purity 115m/g Cd source for counting and analysis using traditional singles gamma-ray spectroscopy, liquid scintillation counting, gas-proportional counting, and analysis using the new Gamma-Alpha-Beta-Gamma (GABγ) and Gamma-Alpha-Beta Radio-Isotope EvaLuator (GABRIEL) coincidence detection systems. Using the D-T fusion generated neutrons, reduces the wait time required to let the short-lived 115Cd ground state decay away by leveraging the 1:1 production ratio of the ground and metastable state of 115 Cd from the 115 In (n,p) reaction relative to the 10:1 production ratio of the ground and metastable isotopes of 115 Cd by thermal neutron capture on 114 Cd. However, the much higher flux in from the Washington State University TRIGA reactor is a more rapid method of production of 115m/gCd but requires access to enriched 114 Cd. High purity 115m/g Cd samples were analyzed on regular intervals over a period of 140-days. The activity of 115m Cd measured by beta and gamma show a clear discrepancy using the current best known gamma-ray decay branching ratios. Based on these observations and measurements, a set of new branching ratio recommendations have been produced using conventional counting techniques and the advanced GAB? and GABRIEL instruments.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2022

Occurrences of REE and CM in volcanic ash beds associated with coals in SW Wyoming & NW Colorado

Rare earth elements (REEs) and critical minerals (CMs) are vital components in the nation’s economy, national security, and the further strengthening of our renewable energy capabilities and infrastructure. Recent work has revealed more insight into previously unknown modes of occurrence for REEs, particularly in coal seams and adjacent noncoal sediments, but many data gaps persist. Temporal relationships between coal beds and adjacent tonstiens, ash beds, or other partings are widely recognized but not fully understood. Variations in REE concentration and their direct or indirect association with coal beds are not predictable. This study presents findings from examining lateral and potentially stratigraphic variation of REE and CM occurrences in coals of late Cretaceous and Paleocene age. In this study we collected and examined samples of differing ages both core and coal mine exposures for qualitative x-ray fluorescence (XRF), and multi-element geochemistry. Preliminary XRF data indicates variable levels of light REEs in a core from the Yampa coalfield of Moffat County in northwest Colorado. An absence of similar enrichments in slightly older coals in the Adaville Formation of the Fold and Thrust belt in southwest Wyoming is described from XRF and ICP-MS evaluation. Previous work has shown an association with the Yampa Bed and REE enrichment in coals of the Upper Cretaceous Williams Fork Formation. While not common and widespread, this association suggests that leaching or reworking of the ash bed could lead to elevated REEs in the adjacent coal. We investigate differences between the two data occurringas geologic compositions and post-depositional histories. In-depth petrography and precise geochemistry should reveal data that at the very least will help improve some aspects of the geologic models of coal basins and the effect of an influx of non-coal sediments.

REE, CM, ash, volcanic ash, coal↗

DOE OSTI2025

Technology Case Study: Techno-Economic and Life Cycle Analysis for Microalgae Conversion Pathways to Fuels and Products

This technology case study report details the cost and sustainability prospects for an emerging feedstock - microalgae - converted to fuels and products via a fractionation and upgrading approach termed combined algae processing (CAP). Detailed techno-economic analysis (TEA) and life cycle analysis (LCA) are conducted for the conversion of farmed algae biomass, with two primary scenarios considering the conversion of either high-compositional-quality biomass enriched in lipids (high-lipid [HL]) or lower-quality biomass enriched in protein (high-protein [HP]). Each scenario employs a different biorefinery configuration tailored towards extracting the maximum value from the given biomass composition. The HL scenario produces fuels and non-isocyanate polyurethane (NIPU) as the primary products, while the HP scenario products fuels and a residual solid coproduct which can be used as a co-feed for producing thermoplastics. The results for the HL scenario were particularly promising, with a minimum fuel selling price (MFSP) of $\$$3.68 per gasoline gallon equivalent (GGE) and fuel GHG emissions translating to 54%-76% reduction compared to petroleum fuels depending on the coproduct handling method used. In contrast, the HP scenario faced more challenges in producing biofuels economically, projecting an MFSP of $\$$7.92/GGE despite significant revenues from the residual algae solids. LCA results for the HP case reflected a 24% reduction potential in biorefinery-level GHG emissions. However, these GHG reductions were primarily associated with the thermoplastic coproduct, which accounted for 93% of all biorefinery outputs by mass. Using a process-level allocation method, carbon intensity results were less promising, indicating a net increase in fuel GHG emissions versus petroleum fuels and highlighting the reliance of this scenario on the thermoplastic coproduct.

09 BIOMASS FUELS↗

DOE OSTI2025

Integral Experiment Request 523 CED-2 Report

This report documents the final design phase of the Critical Experiment Design (CED-2) conducted as part of integral experiment request (IER) 523. The purpose of IER 523 is to determine critical configurations of 35 weight percent (wt%) enriched uranium dioxide beryllium oxide (UO 2 -BeO) material driven by an annular ring of Seven Percent Critical Experiment (7uPCX) fuel rods at Sandia National Laboratories (Sandia). The experiments will provide benchmark data on water moderated, intermediately enriched UO2 systems as well as Be nuclear data. The experiment will also provide partial validation for the beryllium oxide (BeO) thermal neutron scattering law (TSL) in the thermal energy range. Experiment design concepts, neutronic analysis results, and proposed paths for continuing the CED process are presented. This report builds on the feasibility and justification of experimental need report (CED-0) and preliminary experiment design report (CED-1) completed in December 2021 and September 2023, respectively [1, 2].

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2025

Revising the Feasibility Study for Solar PV Deployment Conducted for Pueblo of Zia in 2013

The project team comprising ICAST (International Center for Appropriate and Sustainable Technology), a national 501 c3 nonprofit, and Pueblo of Zia worked to determine the feasibility of constructing two different types of solar projects. The first is a 5 MW solar farm on tribal land, adjacent to the Tri-State Generation and Transmission substation, the utility that has exclusive rights to serve the Pueblo of Zia. The second is behind-the-meter solar installations on or near Pueblo of Zia buildings, such as the Zia Enrichment Library, the T'siya Enrichment Center, the Health and Wellness Center, the Tribal Administration Building, and Zia Pueblo Senior Center.

14 SOLAR ENERGY↗

DOE OSTI2024

Thermo-Mechanical Analysis of Irradiated MURR LEU Fuel Plates

The University of Missouri Research Reactor (MURR®) is a multi-disciplinary research and education facility providing a broad range of analytical, materials science, and irradiation services to the research community and the commercial sector. MURR is one of five U.S. high performance research reactors (USHPRR), plus one critical facility, actively collaborating with the National Nuclear Security Administration (NNSA) Material Management and Minimization (M 3 ) Reactor Conversion Program to convert from the use of highly enriched uranium (HEU, ≥ 20 wt% U 235) to low-enriched uranium (LEU, < 20 wt% U-235) fuel. All USHPRR, including MURR, completed designs with a new type of very high-density LEU fuel based on an alloy of uranium and 10-weight percent molybdenum (U-10Mo) for conversion to LEU fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Preliminary Thermo-Mechanical Analysis of Irradiated MURR LEU Fuel Element

The University of Missouri Research Reactor (MURR) is a multi-disciplinary research and education facility providing a broad range of analytical, materials science, and irradiation services to the research community and the commercial sector. MURR is one of five U.S. high performance research reactors (USHPRR), plus one critical facility, that is actively collaborating with the National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Office of Reactor Conversion and Uranium Supply to convert from the use of highly enriched uranium (HEU, ≥ 20 wt% U-235) to low-enriched uranium (LEU, < 20 wt% U-235) fuel. A new type of very high-density LEU fuel based on an alloy of uranium and 10-weight percent molybdenum (U-10Mo) is expected to allow the conversion of some USHPRR, including MURR, to LEU fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Irradiation Thermo-Mechanical Modeling and Analysis of University of Missouri Research Reactor HEU Fuel Plates

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Impacts of Irradiation Structural Behavior on Thermal Hydraulics Safety Analysis to Support MURR LEU Conversion

The University of Missouri Research Reactor (MURR) located in Columbia, Missouri is one of six U.S. High Performance Research Reactors (USHPRR), including one critical facility, that is actively collaborating with U. S. Department of Energy (DOE) National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Office of Reactor Conversion and Uranium Supply to convert from the use of highly enriched uranium (HEU; ≥20 wt% U-235) to low-enriched uranium (LEU; <20 wt% U-235) fuel. A new type of very high-density LEU fuel based on an alloy of uranium and 10 wt% molybdenum (U-10Mo) is expected to allow the conversion to LEU of MURR, as well as four other USHPRR. MURR has been working with the Reactor Conversion Pillar at Argonne to perform fuel element design and fuel cycle performance analyses, steady-state thermal hydraulics safety analysis, and accident safety analyses in preparation for the conversion of MURR and to support a preliminary safety analysis report for conversion to LEU fuels. Subsequent analyses have also been performed, including transition cycles where all-fresh LEU fuel elements are introduced upon conversion and progressing through reactor operations the core is brought to equilibrium. Thermal hydraulics safety analyses performed as part of the above have employed an assumption on channel gap reduction due to burnup-related phenomena including fuel swelling, irradiation creep, and oxide layer buildup. Recently, a series of structural analyses have been performed on the MURR LEU fuel plates and an element due to significant differences between the plate and element designs of the MURR HEU and LEU fuels. In addition, NUREG-1537 indicates that structural phenomena are to be evaluated. Two separate types of structural analyses were performed for the MURR LEU fuel element: fluid-structure interaction (FSI) and irradiation thermo-mechanical. The FSI analysis evaluated the effects of hydraulic forces on the MURR LEU fuel element to quantify the flow-induced plate deflection, and a minimal impact to the channel gap thickness was predicted under prototypic and bounding conditions. The irradiation thermo-mechanical analysis evaluated the effects of fuel swelling, irradiation creep, and thermal expansion for the MURR LEU plates and the element for prototypic thermal and irradiation conditions based on a high-fidelity approach multiphysics approach. Overall, this thermo-mechanical analysis predicts smaller gap thickness changes in previously limiting regions. Larger changes are predicted in the middle of channels, and for end channels where power density is not typically a maximum. An additional thermo-mechanical analysis was performed for the outermost HEU fuel plate, which showed a similar magnitude of deflection as the outermost LEU plate. Due to substantial differences between the channel gap reductions assumed for the previous safety analyses and those predicted by the irradiation thermo-mechanical analysis, a need to evaluate their impact on the thermal hydraulics safety analyses arose. This report presents the results from the steady-state safety analyses for normal operation as well as the accident analyses for the two most limiting accident scenarios.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Development of Decay Heat Sensitivity Analysis Capability in SCALE/ORIGEN

In this work, a decay heat sensitivity analysis capability was developed and implemented in the ORIGEN code of the SCALE nuclear modeling and simulation suite. This capability introduces improved numerical integration schemes, which overcome the challenges associated with accurately modeling the behavior of adjoint nuclide amounts during coarse time steps for both nuclide amount and decay heat sensitivity calculations. This capability significantly improves the accuracy of calculations without compromising computational efficiency compared to the existing method. Extensive verification was conducted for various benchmark problems, including a 238 Pu decay and an irradiation problem involving 135 Xe, evaluated with both coarse and fine time grids. The results show excellent agreement with reference direct perturbation solutions, reaffirming the computational accuracy of the newly proposed numerical integration methods. Furthermore, sensitivity analyses were performed for fission product inventories ( 147 Sm, 150 Sm, 155 Gd) in pressurized water reactor UO 2 and MOX fuel assemblies. These analyses demonstrated that the ORIGEN sensitivity analysis capability can capture detailed sensitivity coefficients and underlying physics in real applications. Additionally, a decay heat sensitivity analysis for high-assay low-enriched uranium fuel, including various initial 235 U enrichment and burnup points, highlights the extended capabilities of SCALE/ORIGEN in comprehensively assessing the factors influencing total decay heat. These advancements in ORIGEN offer valuable insights for reactor analysis, fuel design, and safety assessments, especially in the context of advanced nuclear fuel development and design changes.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2024

Identify and Assess Technical Challenges in Safeguards Measurements of Spent Advanced Reactor Fuels

Advanced reactor (AR) designs use various nuclear fuel types that can be significantly different than conventional light-water reactor (LWR) fuels, including differences in sizes, compositions, and chemical forms (e.g., oxide, carbide, metal). Nearly all the proposed AR fuels use high-assay low-enriched uranium (HALEU), which will have higher enrichments (5–20 wt% 235 U) than LWR fuels (currently limited to <5 wt% 235 U). In advance of the wide use of these new fuel types around the world, international safeguards organizations such as the International Atomic Energy Agency (IAEA]) are working with some of the AR vendors to formulate safeguards approaches for these AR fuel cycles. As part of the overall safeguards approach, it is important to identify the potential technical challenges in performing safeguards verification measurements of these AR fuels (both fresh and spent fuels) in advance of the widespread adoption of these new fuel types, because new safeguards technologies can take several years to develop, test, and approve for use. This report documents work performed in fiscal year 2024 based on modeling and simulation to assess the performance of the existing safeguards measurement technologies for irradiated or spent AR fuel elements or items. This work is a continuation of the work performed in fiscal year 2023 that focused on fresh AR fuels. Spent AR fuels have a distinct difference from their LWR counterparts: unlike the spent LWR fuels typically stored in a water-filled pool, some spent AR fuels—such as tristructural-isotropic (TRISO)-based fuels—will most likely be stored in air-filled hot cells. Because most safeguards measurements on spent fuel performed to date have been conducted under water, the air-filled hot cell environment could present unique challenges to safeguards measurements. Fork detector (FDET) and Cerenkov viewing device (CVD) systems have been the two primary instruments used by the IAEA for several decades to measure spent LWR fuel assemblies stored in pools for safeguards verification purposes. Because the lower refractive index of air causes Cerenkov light to be of lower intensity in air than in water, existing CVDs are likely unable to perform safeguards verification measurements for spent fuel stored in an air-filled hot cell, as is the case for the TRISO-based spent fuel elements (e.g., pebbles, graphite fuel blocks). Unlike FDET measurements, CVD measurements do not require fuel be moved, so they are a simpler and faster to take than FDET measurements. The inability to perform CVD measurements on the TRISO-based AR fuel types presents a major technical challenge in the effort to use existing technology to perform safeguards measurements on spent AR fuels. This study was mainly conducted through the modeling and simulation of an FDET or an FDET-like system on five spent AR fuel types, including one metallic fuel type and four TRISO-based fuel types in both pebble and graphite block forms in their respective storage configurations and environments. Because the various AR fuel types have significantly different dimensions, FDET systems must be adapted to accommodate them. Partial defect tests were also simulated in this study to assess the FDET’s ability to detect potential fuel diversions. The FDET measures the fuel’s total passive neutron and gamma emissions. The simulated FDET results from spent AR fuel items are compared against results from a typical spent pressurized water reactor (PWR) assembly. High-purity germanium (HPGe) gamma detector measurements were also simulated for the spent AR fuel types and the PWR assembly because the signature photopeaks have been used in LWR safeguards verifications, although HPGe is usually not used to detect diversions because of the fuel’s self-attenuation effects on those photopeaks. The results indicate that these detectors have significant challenges in performing safeguards measurements of the spent AR fuel items, including incompatibilities between AR fuel items and existing FDETs, lower neutron count rates, lower sensitivities to fuel diversions in certain AR fuel items, and significantly higher interference from a neighboring fuel item when the measurement is performed in air. These results suggest that an alternative technology or significant and timely technology development is needed to perform adequate safeguards measurements of some of these AR fuel items.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗