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DOE OSTI2021

AGR-5/6/7 Irradiation Test Final As-run Report

This document presents the as-run analysis of the Advanced Gas Reactor (AGR) 5/6/7 irradiation experiment. AGR-5/6/7 is the last of a series of experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of the development and qualification of tri-structural isotropic (TRISO) low enriched fuel for use in high-temperature gas-cooled reactors (HTGR). The test train contained five separate capsules that were independently controlled and monitored. Each capsule contained multiple 24.91-mm-long and 12.25-mm-dimeter compacts filled with low-enriched uranium carbide/oxide tri-structural isotropic fuel particles. The objectives of the AGR 5/6/7 experiment were to: 1. Irradiate reference-design fuel particles to support fuel qualification. 2. Establish operating margins for the fuel, beyond normal operating conditions. 3. Provide irradiated-fuel performance data and irradiated-fuel samples for post irradiation examination and safety testing. The primary objective of the AGR-5/6 test (Capsules 1, 2, 4, and 5) was to verify the successful performance of the reference-design fuel under normal operating conditions. The AGR-7 test (Capsule 3) was designed to explore fuel performance at higher temperatures. Its primary objective is to demonstrate the capability of the fuel to withstand conditions beyond normal operating conditions, in support of plant design and licensing. AGR 5/6/7 will also provide irradiated-fuel performance data based on the fission gas release from particles during irradiation. To achieve the test objectives, the AGR-5/6/7 experiment was irradiated in the northeast (NE) flux trap of the ATR with a planned duration of 500 effective full-power days (EFPDs). The NE flux trap was selected because its larger diameter provided greater flexibility for test-train design compared to the Large B positions used for the AGR-1 and AGR-2 irradiations, significantly enhancing test capabilities for the combined irradiation campaigns. Due to delays in the ATR schedule, the AGR-5/6/7 irradiation was significantly shorter than the originally planned 13-cycle schedule. Irradiation began on February 16, 2018 and ended on July 22, 2020, spanning nine ATR cycles (162B?168A) over two and a half years. Thus, the AGR-5/6/7 fuel compacts were irradiated for a total of approximately 360.9 EFPDs. Final burnup values, on a per-compact basis, ranged from 5.66% to 15.26% fissions per initial heavy metal atom, while fast fluence values ranged from 1.62 to 5.55 × 1025 n/m2 (E >0.18 MeV). Time-averaged volume-averaged fuel temperatures on a capsule basis at the end of irradiation ranged from 756°C in Capsule 5 to 1313°C in Capsule 3 excluding days with significantly lower temperature during the two short powered axial locator mechanism cycles, 163A and 167A. By the end of irradiation, 48 out of 54 installed thermocouples (TC) had failed (the bottom three capsules lost all TCs). During the first five cycles (162B ? 165A), fission-gas isotope release-rate-to-birthrate (R/B) ratios were stable in the 10-8?10-6 range, and no in-pile particle failures were observed based on the gross gamma counts. During this time, the high exposed kernel fraction and high fuel particle temperatures in Capsule 1 led to the maximum R/B value of around 2 ? 10-6 for Kr-85m. The fission gas release in all capsules started to increase from the second half of Cycle 166A, when a large number of in-pile particle failures occurred in Capsule 1 and a gas line problem in this capsule caused fission gas leakage at various degrees into the other four capsules. This gas line problem also prevented fission gas release measurement for Capsule 1 during the last three cycles due to gas flow isolation. By the end of irradiation, it is estimated that approximately 15 particle failures in Capsule 3, which was considered possible at the start of the irradiation, since the experiment was designed to operate beyond the high-temperature gas-cooled reactor (HTGR) normal operating temperature range. A few hundred in-pile particle failures were estimated for Capsule 1 by the end of Cycle 166A, but the total number of failures is unknown due to lack of fission gas release data in the later cycles. Additionally, four potential in-pile failures were identified for Capsule 2 during the last cycle, 168A. In contrast, no in-pile failures were identified in the top two capsules (4 and 5) based on absence of the typical spikes in gross gamma counts and low failure estimates using the AGR-3/4 R/B per exposed kernel model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

AGR-5/6/7 Irradiation Test Final As-run Report

This document presents the as-run analysis of the Advanced Gas Reactor (AGR)-5/6/7 irradiation experiment. AGR-5/6/7 is the last of a series of experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory in support of the development and qualification of tri-structural isotropic low-enriched fuel for use in high-temperature gas-cooled reactors. The test train contained five separate capsules that were independently controlled and monitored. Each capsule contained multiple 24.91-mm-long and 12.25-mm-dimeter compacts filled with low-enriched uranium carbide/oxide tri-structural isotropic fuel particles. The objectives of the AGR-5/6/7 experiment were to: • Irradiate reference-design fuel particles to support fuel qualification. • Establish operating margins for the fuel, beyond normal operating conditions. • Provide irradiated-fuel performance data and irradiated-fuel samples for post-irradiation examination and safety testing. The primary objective of the AGR-5/6 test (Capsules 1, 2, 4, and 5) was to verify the successful performance of the reference-design fuel under normal operating conditions. The AGR-7 test (Capsule 3) was designed to explore fuel performance at higher temperatures. Its primary objective was to demonstrate the capability of the fuel to withstand conditions beyond normal operating conditions, in support of plant design and licensing. AGR-5/6/7 will also provide irradiated-fuel performance data based on the fission gas release from particles during irradiation. To achieve the test objectives, the AGR-5/6/7 experiment was irradiated in the northeast flux trap of the ATR with a planned duration of 500 effective full-power days. The northeast flux trap was selected because its larger diameter provided greater flexibility for test-train design compared to the Large B positions used for the AGR-1 and AGR-2 irradiations, significantly enhancing test capabilities for the combined irradiation campaigns. Due to delays in the ATR schedule, the AGR-5/6/7 irradiation was significantly shorter than the originally planned 13-cycle schedule. Irradiation began on February 16, 2018 and ended on July 22, 2020, spanning nine ATR cycles (162B–168A) over two and a half years. Thus, the AGR-5/6/7 fuel compacts were irradiated for a total of approximately 360.9 effective full-power days. Final burnup values, on a per-compact basis, ranged from 5.66 to 15.26% fissions per initial heavy metal atom, while fast fluence values ranged from 1.62 to 5.55 × 1025 n/m2 (E >0.18 MeV). Time-averaged volume-averaged fuel temperatures on a capsule basis at the end of irradiation ranged from 756°C in Capsule 5 to 1313°C in Capsule 3 excluding days with significantly lower temperature during the two short powered axial locator mechanism cycles, 163A and 167A. By the end of irradiation, 48 out of 54 installed thermocouples had failed (the bottom three capsules lost all thermocouples). During the first five cycles (162B – 165A), the fission-gas isotope release-rate-to-birthrate (R/B) ratios were stable in the 10-8–10-6 range, and no in-pile particle failures were observed based on the gross gamma counts. During this time, the high exposed kernel fraction and high fuel particle temperatures in Capsule 1 led to a maximum R/B value of around 2 ? 10-6 for Kr-85m. The fission gas release in all capsules started to increase from the second half of Cycle 166A, when a large number of in-pile particle failures occurred in Capsule 1 and a gas line problem in this capsule caused fission gas leakage at various degrees into the other four capsules. This gas line problem also prevented a fission gas release measurement for Capsule 1 during the last three cycles due to gas flow isolation. By the end of irradiation, it is estimated that approximately 15 particles failed in Capsule 3, which was considered possible because the experiment was designed to operate beyond the high-temperature gas-cooled reactor normal operating temperature range. A few hundred in-pile particle failures were estimated for Capsule 1 by the end of Cycle 166A, but the total number of failures is unknown due to the lack of fission gas release data in the later cycles. Additionally, four potential in-pile failures were identified for Capsule 2 during the last cycle, Cycle 168A. In contrast, no in-pile failures were identified in the top two capsules (4 and 5) based on the absence of the typical spikes in gross gamma counts and low failure estimates using the AGR-3/4 R/B per exposed kernel model.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Processing of SOFC Anodes for Enhanced Intermediate Temperature Catalytic Activity at High Fuel Utilization (Final Report)

The overall objective was to infiltrate anodes with nanoparticle catalysts so that anode catalytic performance can be improved as anode operation temperature is reduced from 800°C to intermediate temperatures of 700°C and 600°C and as anode fuel utilization is increased. The addition of nickel nanoparticle catalysts into the anode improves performance by providing additional triple phase boundaries (TPBs), increasing the density of electrochemical reaction sites. To achieve this goal, both liquid phase and vapor phase methods for depositing nickel nanoparticles in the anode active layer, the region of the anode near the electrolyte where the electrochemical reaction occurs. It is imperative that nanoparticles are stable against coarsening during long-term operation, so the mechanisms and kinetics of nanoparticle stability during cell operation will be investigated. Finally, priority is given to processes that are scalable and easily transferable to industry, leveraging existing Ni-YSZ cermet anode technology. Utilizing nickel nanoparticles at different temperatures has different considerations that will be addressed. At the upper end of the intermediate temperature range (800°C), it is important that performance improvement is maintained over long periods by mitigating the instability of nanoparticles. At the middle of the temperature range (700°C), it is important that performance improvement is maintained at high fuel utilizations. At the lower end of the temperature range (600°C), where the performance of cells is poor due to increased resistances, it is important to improve anode performance by significantly increasing the electrochemical reaction site density, while nanoparticle durability is less of an issue. Finally, the objective was to explore mixed ionic and electronic (MIEC) nanoparticle catalysts like GDC to see if 2-phase boundaries can be created to increase the electrochemical reactions at the anode.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2025

Ceramic Composite Inert Matrix Fuel Forms in High-Temperature Gas-Cooled Microreactors

Here, this work optimizes micro-prismatic high-temperature gas reactor (HTGR) designs to reduce the energy-normalized mass of spent nuclear fuel (SNF) and high-level waste (HLW) produced. The optimization was performed for the current graphite moderator and an inert matrix fuel (IMF) concept employing different composite moderators in a prismatic design architecture. The fuel matrix is magnesium oxide (MgO) with entrained tristructural-isotropic (TRISO) fuel. The moderator materials, including beryllium oxide (MgO-BeO) and beryllium (MgO-Be) at 40 vol % loading and yttrium hydride (MgO-YH x=1.9 ) and zirconium hydride (MgO-ZrH x=1.9 ) at 15 vol % loading, were entrained within the MgO host matrix. A generic graphite micro-prismatic HTGR is used as the baseline point design where the external dimensions are held constant. The composite moderator designs use 19.9% enriched uranium nitride TRISO fuel and hexagonal assemblies. For each IMF concept, an optimization study was performed to maximize the discharge burnup of the fuel by varying the TRISO packing fraction and the lattice pitch of the assemblies. The mass of SNF and HLW, other waste metrics, fuel cost, environmental impact metrics, and the activity of the SNF and HLW at 100 years and 100 000 years were calculated for the optimized IMF and graphite reference designs. The IMF results were subsequently compared to those of the graphite reference and the values for a light water reactor (LWR) and a small modular LWR. For the SNF and HLW, all the IMF concepts and the graphite reference produced less waste compared to the traditional LWR designs. However, the IMF concepts outperformed the graphite reference regarding the mass of SNF and HLW. For the other waste metrics, the IMF concepts showed reductions in fuel cost with improved environmental metrics relative to the graphite reference. Overall, the IMF concepts significantly reduced the SNF and HLW produced per unit of energy generated compared to traditional LWR designs.

TRISO↗
DOE OSTI2022

Machine learning molecular dynamics simulations toward exploration of high-temperature properties of nuclear fuel materials: case study of thorium dioxide

Predicting materials properties of nuclear fuel compounds is a challenging task in materials science. Their thermodynamical behaviors around and above the operational temperature are essential for the design of nuclear reactors. However, they are not easy to measure, because the target temperature range is too high to perform various standard experiments safely and accurately. Moreover, theoretical methods such as first-principles calculations also suffer from the computational limitations in calculating thermodynamical properties due to their high calculation-costs and complicated electronic structures stemming from f-orbital occupations of valence electrons in actinide elements. Here, we demonstrate, for the first time, machine-learning molecular-dynamics to theoretically explore high-temperature thermodynamical properties of a nuclear fuel material, thorium dioxide. The target compound satisfies first-principles calculation accuracy because f-electron occupation coincidentally diminishes and the scheme meets sampling sufficiency because it works at the computational cost of classical molecular-dynamics levels. We prepare a set of training data using first-principles molecular dynamics with small number of atoms, which cannot directly evaluate thermodynamical properties but captures essential atomistic dynamics at the high temperature range. Then, we construct a machine-learning molecular-dynamics potential and carry out large-scale molecular-dynamics calculations. Consequently, we successfully access two kinds of thermodynamic phase transitions, namely the melting and the anomalous λ transition induced by large diffusions of oxygen atoms. Furthermore, we quantitatively reproduce various experimental data in the best agreement manner by selecting a density functional scheme known as SCAN. Our results suggest that the present scale-up simulation-scheme using machine-learning techniques opens up a new pathway on theoretical studies of not only nuclear fuel compounds, but also a variety of similar materials that contain both heavy and light elements, like thorium dioxide.

74 ATOMIC AND MOLECULAR PHYSICS↗
DOE OSTI2025

Oxygen Reduction on Platinum-Nickel and Platinum-Cobalt Alloy Based Catalysts for High Temperature Proton Exchange Membrane Fuel Cells

Catalyst layers, with commercial PtNi/TKK (TECNiE52; platinum 46.5 wt%), PtCo/TKK (TEC36E52; platinum 45.8 wt%) catalysts, and an in-house catalyst PtNi-NC (platinum 38.4 wt%), synthesised using acoustic mixer and a tube furnace, are prepared using three different ionomers: 20 wt% Nafion®, 60 wt% polytetrafluoroethylene (PTFE), and an in-house 5 wt% ionomer (‘Ionomer X’). These inks are bar-coated onto carbon paper gas diffusion layers (GDLs), which are cut into 3 cm × 5 cm pieces. The coated layers are tested for oxygen reduction in a commercial gas diffusion electrode (GDE) test cell, FlexCell® (Gaskatel GmbH, Germany), using 85 wt% phosphoric acid at both room temperature and at 155°C. This study evaluates the polynorbornene (PNB)-based in-house ionomer performance as binder and in-house catalyst oxygen reduction reaction (ORR) activity in comparison to commercial products. The catalyst layers are characterised using X-ray diffraction (XRD) analysis, X-ray photoelectron spectroscopy (XPS), and high-angle annular dark-field scanning transmission electron microscopy (HAADF-STEM). Among the catalyst layers prepared with PTFE ionomer and tested at 155°C, the in-house catalyst demonstrates the highest ORR activity. Ionomer X proves to be a good candidate to be used as a per- and polyfluoroalkyl-free binder for high-temperature (HT) proton exchange membrane fuel cell (PEMFC) applications.

85 wt% phosphoric acid↗
DOE OSTI2024

SCALE Analyses of Scenarios in the High-Temperature Gas-Cooled Reactor Fuel Cycle

This report demonstrates the SCALE code system’s capabilities in modeling and simulating scenarios in the High-Temperature Gas-cooled Reactor (HTGR) nuclear fuel cycle as part of a United States Nuclear Regulatory Commission (NRC) project that aims to demonstrate the capabilities of the SCALE and MELCOR codes for non-light water reactor (non-LWR) modeling and simulation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Effect of Nanoscale Ce 0.8 Gd 0.2 O 2–δ Infiltrant and Steam Content on Ni–(Y 2 O 3 ) 0.08 (ZrO 2 ) 0.92 Fuel Electrode Degradation during High-Temperature Electrolysis

Studies of Ni–yttria-stabilized zirconia (YSZ) fuel electrode degradation mechanisms in solid oxide electrolysis cells (SOECs) are complicated by the different possible Ni–YSZ microstructures and compositions, and the variations in the H 2 /H 2 O ratio encountered in an electrolysis stack. Here, in this study, we describe a life testing scheme aimed at providing survey results on degradation as a function of the H 2 O–H 2 composition, with life tests carried out at five different steam contents from 90% to 10%. A Ni–YSZ-supported symmetric cell geometry is employed both with and without infiltrated nanoscale gadolinia-doped ceria (GDC). Impedance spectroscopy is utilized to observe changes in electrochemical characteristics during the life test, and a transmission-line-based equivalent circuit is used to model the data. Post-test electrode microstructures were observed. The results suggest that the GDC infiltrant reduces the electrode polarization resistance and provides more stable electrode polarization over a range of conditions.

08 HYDROGEN↗
DOE OSTI2026

AGR-5/6/7 Ceramography Report

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established in 2002 to conduct research and development on tri-structural isotropic (TRISO)-coated particle fuel for High Temperature Gas-cooled Reactors (HTGRs). All AGR irradiations have been completed: AGR-1 (Collin 2015), AGR-2 (Collin 2014), AGR-3/4 (Collin 2016) and AGR-5/6/7 (Pham 2021). The objectives of each program are presented in Figure 1.1. The purpose of the AGR-5/6/7 program was to provide a baseline fuel qualification data set to support licensing, deployment, and operation of HTGRs in the United States. To achieve these goals, the program includes fuel fabrication, irradiations of TRISO fuels and high-temperature materials, safety testing and post-irradiation examination (PIE), fuel performance modeling, and fission product (FP) transport (Sharp 2020). The AGR-5/6/7 work continues as a final irradiation program of the Advanced Reactor Technologies (ART) program, that began in February 2018 and ended in July 2020. The AGR-5/6/7 fuel compacts were irradiated for a total of approximately 360.9 effective full power days (EFPDs) (Stempien 2023), resulting in final burn-up values, on a per-compact basis, ranging from 5.66% to 15.26% fissions per initial heavy metal atom (FIMA), and fast fluence values ranging from 1.62×1025 n/m2 to 5.55×1025 n/m2 (E >0.18 MeV) (Stempien 2022)

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Microstructural and Oxidation Effects of Nb Additions to U3Si2

U3Si2 is a long term, accident-tolerant nuclear fuel candidate for light-water reactors because of its superior thermal conductivity and increased uranium density when compared to traditional uranium dioxide (UO2). While reducing internal thermal stresses and increasing efficiency, U3Si2 exhibits energetic oxidation during certain off-normal and accident scenarios, which include coolant or steam exposure. To mitigate this, Nb is investigated as an alloy constituent to enhance corrosion resistance and increase mechanical strength. The work presented investigates the response of Nb-alloyed U3Si2 to steam atmospheres. A thermogravimetric analysis is conducted in flowing steam to T > 1000 °C to assess oxidation resistance. The phase characterization of as-melted, thermally annealed and post-oxidation compositions with up to 12 vol% Nb by powder X-ray diffraction, scanning electron microscopy, and energy dispersive spectroscopy is reported.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Influence of gasoline fuel formulation on lean autoignition in a mixed-mode-combustion (deflagration/autoignition) engine

We report stoichiometric spark-ignition engines suffer efficiency penalties due to throttling losses at low loads, a low specific-heat ratio of the stoichiometric working fluid, and limits on compression ratio due to end-gas autoignition leading to undesirable knocking. Mixed-Mode Combustion (MMC) mitigates these shortcomings by using a lean working fluid where a spark-initiated pilot-stabilized deflagrative flame front is followed by controlled end-gas autoignition. This MMC study investigates the effects of initial conditions (intake air temperature, intake pressure, equivalence ratio, and intake oxygen fraction) on autoignition tendency of four gasoline-range fuels with varying properties and composition. The use of fuels with varying octane sensitivity (S) allowed exploring the importance of low-temperature heat release in triggering autoignition. Fuels with high S were less reactive for conditions that promote low-temperature chemistry (operation at high intake air pressure or without N2 dilution). Conversely, an Alkylate fuel with low S showed a greater autoignition resistance at operating conditions that were unfavorable for low-temperature chemistry. Next, the effect of residual gas composition on autoignition tendency of fuels was examined with a chemical-kinetics model. Among the various molecules in the residual gas, nitric oxide (NO) enhanced the low-temperature chemistry and increased the autoignition tendency most significantly. The fuels’ autoignition response to increasing NO amount corroborates the experimental observations. Next, the sequential autoignition of the end-gas was assessed to be less impacted by thermal stratification because of lean mixtures showing relatively less low-temperature chemistry, when compared to stoichiometric mixtures. Next, the effect of changing equivalence ratio on the autoignition was found to be similar for all fuels, regardless of their S. With changing intake air temperature, the response of fuels’ autoignition tendency depended on the dilution level used. At high dilution (i.e. low intake [O2]), fuels’ reactivity increased with increasing intake air temperature. In contrast, for operation without dilution, the autoignition tendency of the low-S Alkylate fuel decreased with increasing intake air temperature, while that of high-S High Cycloalkane fuel still increased with increasing intake air temperature. In conclusion, conventional octane metrics (RON and MON) have utility in assessing the autoignition tendency under lean MMC operation. Moreover, the fuel requirements for MMC align with that of stoichiometric operation: i.e., high RON and high S fuels are desirable for stable non-knocking operation.

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

High Temperature Polymer Electrolyte Membrane Fuel Cell Electrode Composition: Durable Fuel Cell Power under Anhydrous Conditions

Innovators/Researchers at Los Alamos National Laboratory have developed an electrode composition that improves fuel cell performance under anhydrous conditions above 100°C. It has demonstrated great durability in humidified conditions under 100°C as well. It holds promise for use in fuel cells for heavy duty vehicles while retaining utility for portable and stationary power applications. In coming years this flexibility may be a breakthrough improvement over conventional compositions.

25 ENERGY STORAGE↗
DOE OSTI2022

Zero carbon solid-state rechargeable redox fuel for long duration and seasonal storage

Here, this work presents a unique thermochemical process for charging magnesium-manganese-oxide-based solid-state rechargeable redox fuel. High-temperature heating of the processing furnace can be driven by either renewable electricity or concentrated solar power. The simple recyclable fuel charging concept is based on a tubular falling bed reactor with countercurrent oxygen-depleted gas flow for complete heat recuperation. The main focus of this work is achieving solid flowability at high temperatures (1,450°C) and extracting chemically charged solid at ambient temperature with minimal energy loss. The operation strategies described in this work have enabled consistent magnesium-manganese-oxide particle flow up to 1,450°C. The measured extent of the thermal reduction reaction after cooling the particles is more than 90% of the fully reduced state at equilibrium. The thermal-to-chemical efficiency and overall system efficiency are 96% and 35% respectively, which are the highest reported for thermochemical fuels to date.

25 ENERGY STORAGE↗
DOE OSTI2024

BISON Simulated and Experimental Fission Product Release Comparisons from Reradiated AGR-3/4 Compacts During High Temperature Heating Tests

The fuel performance modeling code BISON was used to predict the release of fission products iodine-131 (131I), xenon-133 (133Xe), and krypton-85 (85Kr) from four re-irradiated AGR-3/4 fuel compacts containing tristructural isotropic (TRISO) coated particles during high-temperature isothermal heating tests. The AGR-3/4 fuel compacts were irradiated in the Advanced Test Reactor (ATR) as part of the third and fourth series of planned experiments to support the Advanced Gas Reactor (AGR) Program. They were subsequently stored and re-irradiated in the Neutron Radiography (NRAD) reactor for approximately five days and then stored for another five to eight days before being subjected to isothermal heating tests in the Fuel Accident Condition Simulation (FACS furnace) for 200 to 300 hours at temperatures between 1000°C and 1600°C to evaluate fission product release at elevated temperatures. New nuclide-specific fission product source term models for the three nuclides of interest were developed using the reactor multiphysics code Griffin and implemented into BISON to support this work. The new source term models were incorporated into coupled compact- and particle-scale BISON simulations, which predict spatially- and temporally-resolved radionuclide generation, radioactive decay, transport, and release throughout the entire irradiation history, including the initial ATR irradiation, NRAD re-irradiations, FACS heating tests, and intermediate periods spent in storage. The experimentally measured fission product release from the heating tests were compared to modeling release predictions calculated by BISON to evaluate how well the code compares to experimental results. Overall, the experimental measured and BISON predicted comparative release results varied but generally agreed to within 5 particle equivalents. Comparative release results identified general observations to take into consideration to help refine future models and reduce uncertainties associated with both the measurement results and predictive results. This includes developing new uranium oxycarbide (UCO) specific kernel diffusivities for the three isotopes examined to more accurately reflect the material properties of the fuel form. Deriving new diffusivities will aid in producing a more informed BISON model

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Feasibility and performance of the staged Z-pinch: A one-dimensional study with FLASH and MACH2

Z-pinch platforms constitute a promising pathway to fusion energy research. Here, we present a one-dimensional numerical study of the staged Z-pinch (SZP) concept using the FLASH and MACH2 codes. We discuss the verification of the codes using two analytical benchmarks that include Z-pinch-relevant physics, building confidence on the codes' ability to model such experiments. Then, FLASH is used to simulate two different SZP configurations: a xenon gas-puff liner (SZP1*) and a silver solid liner (SZP2). The SZP2 results are compared against previously published MACH2 results, and a new code-to-code comparison on SZP1* is presented. Using an ideal equation of state and analytical transport coefficients, FLASH yields a fuel convergence ratio (CR) of approximately 39 and a mass-averaged fuel ion temperature slightly below 1 keV for the SZP2 scheme, significantly lower than the full-physics MACH2 prediction. For the new SZP1* configuration, full-physics FLASH simulations furnish large and inherently unstable CRs (>300) but achieve fuel ion temperatures of many kilo-electron volts. While MACH2 also predicts high temperatures, the fuel stagnates at a smaller CR. The integrated code-to-code comparison reveals how magnetic insulation, heat conduction, and radiation transport affect platform performance and the feasibility of the SZP concept.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗