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DOE OSTI2025

SiC cladding R&D execution plan

To advance the development of accident-tolerant fuel cladding for light-water reactors, the Advanced Fuels Campaign actively addresses the R&D gap between the specialized technological progress in SiC-based composite materials, led by industrial partners, and the fundamental scientific understanding of material behavior under reactor operating conditions. The laboratory activities are systematically organized based on the Development Strategy document, which outlines critical barriers to the use of SiC composites in nuclear energy applications for the coming decade and beyond. To complement the strategy, this execution plan specifies a detailed year-by-year work package and establishes key milestones for assessing progress in technological advancement. This plan has been formulated in collaboration with principal investigators at national laboratories, stakeholders, and industry partners with the goal of closing technology gaps and readying SiC-based cladding technologies for full-scale commercial testing. The research conducted at the national laboratories is categorized into three primary subjects: (1) addressing challenges associated with the performance of SiC cladding under normal operating environments, (2) identifying the failure limits of SiC cladding under accident conditions, and (3) modeling fuel performance for SiC cladding systems. Comprehensive laboratory activities integrate the development of advanced modeling tools for SiC-based cladding systems with experimental validation via integrated testing of multiphysics phenomena. This document lays out the laboratory plan to support the development of SiC-based technologies, which have the potential for significant operational and safety benefits for the U.S. nuclear industry. This plan aligns with industry developer needs, relevant Nuclear Energy University Programs, and Nuclear Energy Advanced Modeling and Simulation Programs to ensure SiC cladding technologies are matured as efficiently as possible and will be maintained and updated to accommodate the dynamically evolving progress of development, capability needs, and resource availability.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Orientation-Dependent Surface Faceting in UO{sub 2}

The surface behavior of UO{sub 2} plays an important role throughout the lifetime of this widely-used commercial nuclear fuel, from manufacturing to waste storage. UO{sub 2} surface properties affect fuel swelling behavior by determining the morphology, size and distribution of fission gas bubbles. Additionally, UO{sub 2} surface characteristics influence fracture behavior, which is controlled by the thermodynamics of fracture surfaces. Therefore, detailed understanding of the fundamental surface behavior will help advance prediction of the fuel performance of UO{sub 2}. Due to significant anisotropy in surface tension, faceting features form after annealing in both UO{sub 2} and many other crystalline materials. Knowledge of the correlation between crystallographic orientation and surface faceting features will expand the fundamental knowledge of UO{sub 2} surface characteristics, leading to a better understanding of fission gas swelling and fracture mechanisms. Inherent characterization technique limited previous studies on the crystallographic orientation of UO{sub 2} surface facets to low Miller index surfaces in single crystal specimens. In this study, synchrotron Laue microdiffraction was utilized to examine the crystallographic orientation of grains on the coarse surface of a polycrystalline UO{sub 2} specimen. The results were then compared with the surface morphology, obtained from scanning electron microscopy (SEM), so that a general correlation between the surface faceting features and crystallographic orientation in UO{sub 2} can be established. Both the fundamental surface characteristics and the crystallographic orientation measurements obtained in this study help advance the experiment validation of MARMOT, a mesoscale multi-physics code that predicts the coevolution of microstructure and properties within reactor fuel during its lifetime in the reactor. The MARMOT code is an indispensable component of the MOOSE-BISON-MARMOT (MBM) code suite that has been developed by Idaho National Laboratory to enable next generation fuel performance modeling capability as a part of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program Fuels Product Line (FPL). In this study, the surface-faceting features of UO{sub 2} were found to be dominated by the triple-plane structure and its variants. With this knowledge, by measuring the directions of the <110> edges, the crystallographic orientation of a surface UO{sub 2} grain can be precisely deduced in most regions of the standard stereogram triangle (SST) of the surface normal. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Electron microscopy analysis of high burnup commercial fuel

Modern electron microscopy analysis can provide detailed insight into the microstructure and chemistry of high burnup UO{sub 2} fuel. This more complete understanding of fuel evolution and performance in reactor can provide critical insight necessary to validate fuel performance models and optimize reactor operations. Preliminary results from the investigation of a polished cross-section of high burnup UO{sub 2} fuel (seven cycles) from the H.B. Robinson pressurized water reactor will be discussed. The analysis focuses on scanning electron microscopy (SEM) of the complete cross-section, while scanning transmission electron microscopy (S/TEM) was performed on targeted features identified by SEM where specimens were fabricated by traditional focused ion beam lift-out techniques. Analysis of the complete cross-section allowed for investigation of the microstructural variations from centerline to fuel/clad interface, pellet/pellet interface, and cladding. Secondary electron imaging and backscatter electron imaging was performed to identify variations in the distribution of fission product secondary phases and bubbles/voids. The analysis of the pellet/pellet interface is a unique feature of this sample where initial analysis shows segregation of fission product phases at the pellet/pellet interface and also indicates an interface microstructure lacking of bubbles and voids. X-ray spectrum imaging was performed by both SEM and STEM techniques to identify the composition of the secondary phases present in the fuel. Multivariate statistical analysis (MVSA) of the X-ray spectrum images has been applied to improve the sensitivity of the analysis and highlight minor variations in the fuel chemistry and secondary phases comprised of fission product as well as the distribution of sub-surface Xe gas bubbles in the high burnup structure. The results presented demonstrate the level of microstructural understanding that is attainable with a combination of SEM techniques along with S/TEM techniques.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Assessing High Burnup U-19Pu-10Zr Fuel Performance against Historical and Modeled Behavior

Advancing the deployment of sodium-cooled fast reactors (SFRs) requires thorough testing of metallic fuel pins under accident conditions to establish safe operational limits of high burnup fuel. To conduct transient testing, a comprehensive understanding of steady-state fuel behavior obtained through both experimental characterization and accurate predictive capabilities is needed. This study comparatively assesses the steady-state irradiation performance of two high burnup U-19Pu-10Zr fuel pins, DP-36 and DP-40, irradiated under prototypic fast reactor conditions in preparation for planned safety testing at the Transient Reactor Test Facility. Since DP-40 was designated for use in the test and DP-36 serves as its sibling pin, non-destructive, engineering-scale post-irradiation examinations (PIE) were conducted on both pins while destructive examinations were performed exclusively on DP-36. The results were then assessed against historical performance data from similar fuel pins irradiated in the Experimental Breeder Reactor-II. Additionally, the steady-state irradiation of each pin was modeled using the BISON fuel performance code to assess the accuracy of current modeling capabilities in predicting the baseline irradiation behavior. Non-destructive examinations included neutron radiography to measure fuel column elongation, gamma scanning to verify pin integrity and fission product migration, and profilometry to assess dimensional changes. Benchmarking against existing PIE data revealed consistent patterns in axial fuel column growth and cladding diametral strain, though both pins exhibited longer low-density “fluff” structures, which can have implications for core reactivity and source term calculations. Destructive examinations on DP-36 included fission gas release analysis and sectioning for optical microscopy, which showed more complex constituent redistribution patterns than the traditionally accepted 3-ring model. The axial evolution of fractional areas and porosities of each of the redistributed zones were quantified and presented. Modeling comparisons showed agreement in fractional fission gas release but consistently overestimated axial and radial swelling and disagreed with measured axial porosity patterns. These conservative overpredictions suggested that the pins would appear closer to failure or operational limits at the start of transient tests, potentially leading to higher strain accumulation during the transient. While conservative estimates provide safety margins, they can negatively impact fuel economics. A review of the swelling models identified areas for improvement in the gaseous swelling, solid swelling, and fuel hot-pressing models when applied to ternary fuel. The results of this study highlight the critical importance of conducting pre-test characterization on both test and sibling pins to accurately capture steady-state fuel behavior, providing a precise baseline for post-test evaluations and essential inputs for transient modeling of the planned experiments. The analysis also revealed significant data gaps that require further investigation to enhance the understanding and prediction of fuel swelling and pore dynamics. Collecting comprehensive data across different irradiation conditions, burnup levels, and fuel compositions are essential for refining existing models and developing mechanistic models for both binary and ternary metallic fuels, ultimately improving the integration of modeling and experimental approaches in accident testing.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

AGR-3/4 Irradiation Test Predictions using PARFUME

PARFUME, a fuel performance modeling code used for high temperature gas reactors, was used to model the AGR-3/4 irradiation test using as-run physics and thermal hydraulics data. The AGR-3/4 test is the combined third and fourth planned irradiations of the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The AGR-3/4 test train consists of twelve separate and independently controlled and monitored capsules. Each capsule contains four compacts filled with both uranium oxycarbide (UCO) unaltered “driver” fuel particles and UCO designed-to-fail (DTF) fuel particles. The DTF fraction was specified to be 1×10-2. This report documents the calculations performed to predict failure probability of TRISO-coated fuel particles during the AGR-3/4 experiment. In addition, this report documents the calculated source term from both the driver fuel and DTF particles. The calculations include the modeling of the AGR-3/4 irradiation that occurred from December 2011 to April 2014 in the Advanced Test Reactor (ATR) over a total of ten ATR cycles including seven normal cycles, one low power cycle, one unplanned outage cycle, and one Power Axial Locator Mechanism cycle. Results show that failure probabilities are predicted to be low, resulting in zero fuel particle failures per capsule. The primary fuel particle failure mechanism occurred as a result of localized stresses induced by the calculated IPyC cracking. Assuming 1,872 driver fuel particles per compact, failure probability calculated by PARFUME leads to no predicted particle failure in the AGR-3/4 driver fuel. In addition, the release fraction of fission products Ag, Cs, and Sr were calculated to vary depending on capsule location and irradiation temperature. The maximum release fraction of Ag occurs in Capsule 7 reaching up to 56% for the driver fuel and 100% for the DTF fuel. The release fraction of the other two fission products, Cs and Sr, are much smaller and in most cases less than 1% for the driver fuel. The notable exception occurs in Capsule 7 where the release fraction for Cs and Sr reach up to 0.73% and 2.4%, respectively, for the driver fuel. For the DTF fuel in Capsule 7, the release fraction for Cs and Sr are estimated to be 100% and 5%, respectively.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2025

Pre-Transient Characterization of MLOF-1 Test Pin

This study focuses on the pre-transient characterization of U-10Zr test and sibling fuel pins for the THOR-M-LOF test series. Using neutron radiography, element contact profilometry (ECP), precise gamma scan (PGS), and gas assay, sampling, and recharge (GASR) analysis, it was confirmed that the fuel pins were intact and suitable for testing. Key fuel behaviors quantified include axial elongation, diametral strain, fluff structure geometry, axial isotope distribution, and fission gas release. Any deviations from historically expected behaviors were investigated and attributed to factors other than the irradiation behavior of the fuel pin. These pre-transient measurements establish a baseline for future post-transient analysis, which will be used to inform fuel performance models and safety criteria for sodium-cooled fast reactors (SFRs). The results will enhance understanding of transient fuel behavior and expand limited data on LOF scenarios.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Temperature Effect over Gas Bubble Evolution in Uranium-10 wt.% Molybdenum alloy Irradiated by Swift Xe Ions

Results from post-irradiation examination (PIE) of U-10Mo binary metallic fuel irradiated by 84 MeV Xe ions at two different temperatures (450°C & 275°C) have been evaluated and reported. The main goal of this study was to understand gas bubble formation (due to accumulation of Xe gas) and corresponding size evolution dependent upon net Xe deposition at two different temperatures, for the U-10Mo fuel samples. From the post ion irradiation examination analysis, it can be concluded that at 450°C the Xe gas bubbles seems to coarsen more easily compared to 275°C. The results of ion irradiation testing can provide initial data that may be beneficial for fuel system down-selections, and for the development and validation of fuel performance models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

FAST - A Unique Experiment and Simplified Approach to Accelerated Integral Fuel Testing

Idaho National Laboratory has a long and internationally recognized history of performing irradiation testing of advanced fast reactor fuels in the Experimental Breeder Reactor II (EBR-II) and the Advanced Test Reactor (ATR). Since EBR-II’s decommissioning in the early 1990’s, irradiation testing has been exclusively performed in ATR. The testing of fast reactor fuels in ATR is challenging given its thermal neutron spectrum and coolant temperatures far below those prototypic of fast reactors. Adept researchers have overcome these obstacles through the use of Cadmium shrouds which attenuate all thermalized neutrons and an experiment design that fosters fuel and cladding temperatures prototypic of fast neutron reactors. Although this design has successfully demonstrated temperatures and temperature gradients that closely mimic those seen in fast reactors, irradiation performance is highly sensitive to fabrication tolerances and the time requirements necessary for high fuel burnup, where irradiation performance data of fast fuel is lacking, are significant. This work summarizes two novel experiment designs and their associated physics analyses for testing fast reactor fuels in ATR; an accelerated irradiation experiment and a fast spectrum experiment. The accelerated irradiation experiment does not utilize a Cadmium shroud, has been shown to triple the burnup rate, yet yields temperatures and radial power profiles prototypic of fast reactor fuels. The fast spectrum experiment utilizes a booster fuel element that that hardens the incident neutron spectrum and also yields temperatures and radial power profiles prototypic of fast reactor fuels. These experiment designs efficiently reduce irradiation time requirements, are more economical, and significantly increase ATR throughput. Fast reactor experiments that capitalize on these designs expedite irradiation performance data for fast reactor fuels and provide information invaluable to fuel performance modeling and simulation validation over a wide range of parameters.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

TRISO fuel performance analysis: Uncertainty quantification toward optimization

Tri-structural isotropic (TRISO) fuel particles are a fuel form being considered for potential use in next-generation nuclear reactors (i.e., high-temperature gas-cooled reactors). Though the TRISO fuel manufacturing process has continually advanced in recent years, particle comparisons still reveal statistical variations and uncertainties in terms of geometric configurations and material properties. Given that the physical processes ongoing in TRISO fuel particles during reactor operation are highly correlated with each other, a small degree of uncertainty in one model may lead to significant uncertainty in another. This makes appropriate uncertainty quantification of TRISO fuel particles essential. However, one may wonder about the extent to which the current version of TRISO particles has been optimized, and whether any room remains for further improvements. This paper quantifies TRISO fuel performance model uncertainties that stem from geometric and material data. For this analysis, the BISON code was used, and the Advanced Gas Reactor (AGR)-2 experiment served as a reference case. A total of 10 5 calculations was performed for the uncertainty and optimization analysis, altering the geometric and material data within their uncertainty range. Lastly, the optimization potential of TRISO particles is evaluated from a fuel performance perspective.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

BISON modeling of reactivity-initiated accident experiments in a static environment

In conjunction with the restart of the TREAT reactor and the design of test vehicles, modeling and simulation efforts are being used to model the response of Accident Tolerant Fuel (ATF) concepts under reactivity-initiated accident (RIA) conditions. The purpose of this work is to model a baseline case of a 10 cm long UO{sub 2}- Zircaloy fuel rodlet in the Multi-SERTTA capsule in PWR conditions, using BISON and RELAP5 over a range of energy depositions and with varying reactor power pulse widths. The results provide an initial best estimate of fuel performance modeling during a RIA event in TREAT with a focus on the effect of varying the pulse width and energy deposition on both thermal and mechanical parameters that are important for predicting failure of the fuel rodlet. The combined BISON/RELAP5 model captures coupled thermal and mechanical effects on the fuel-to-cladding gap conductance, cladding-to-coolant heat transfer coefficient and water temperature and pressure that would not be capable in each code individually. These combined effects allow for a more accurate modeling of the thermal and mechanical response in the fuel rodlet and thermal-hydraulics of the test vehicle. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Microstructure and Fission Product Distribution Examination in the UCO kernel of TRISO Fuel Particles

The effect of irradiation and fabrication process on the UCO fuel kernel, kernel-buffer interlayer and fission products, has not been fully studied for UCO fuel kernels of TRISO coated particles. Specifically, the potential correlation(s) with fission product retention, composition and distribution within these structures for the tristructural isotropic (TRISO) fuel performance at normal and accident conditions, are not well reported in literature. Structural and chemical information in the irradiated UCO fuel kernel, e.g., the stoichiometry variations, may further contribute to fuel performance modeling under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. The microstructural and micro-chemical evolution of selected irradiated UCO fuel kernels, kernel buffer interfaces and recoil zones were characterized using Scanning Transmission Electron Microscopy (STEM) and Energy Dispersive Spectroscopy (EDS) elemental mapping. The fuel kernel of particles from Advanced Gas Reactor (AGR)-2 irradiated to 10.8% fissions per initial metal atom (FIMA), were examined with the following main findings: (1) large fission gas bubbles were mainly identified in the zirconium and molybdenum enriched but oxygen depleted phases in the center of fuel kernel, (2) significant inter-diffusion between the UCO fuel kernel and pyro-carbon buffer layer was observed as a high density of sub-micron sized carbon inclusions was identified at near the interface in the kernel side and some isolated UCO particles were located in the buffer layer, (3) pores filled with free carbon were also observed in the recoil zone specimen. Fuel kernels from two AGR-1 compacts irradiated to 11.4 and 17.4% FIMA respectively, were examined with the following main findings: (1) two primary phases consisting of a “high-Z (atomic mass)” UC(O) phase (rocksalt; a=0.496 nm) that also contained Zr, Mo, and/or Ru and a “low-Z” UO2(C) phase (fluorite structure; a = 0.547nm) in which trace amounts of I, Nd, Pr, and/or Er could be detected. This latter phase contained many precipitates of the “high-Z” phase. Conversely, this latter phase did not contain precipitates of the UO2-type material, (2) Xe bubbles were found almost exclusively in the “high-Z” phase and in the particular example shown, the Zr-containing variant. Further work in this area should address whether this potential Xe preference for just one of the high-Z phases is indeed a real effect.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

A Multiphysics Evaluation of Annular Uranium-Zirconium Metallic Fuels [Poster]

This study examines the performance of U-10Zr annular metallic fuel rodlets which were experimentally evaluated as part of the Advanced Fuels Campaign (AFC). The AFC mission is to develop novel fuel technologies and facilitate the implementation of those technologies by industry partners. A key objective is to improve steady-state and transient performance over current fuel types. The experiments of interest in this study included annular metallic U-Zr fuel rodlets within HT-9 cladding which were placed in SS-316 capsules and inserted in the Advanced Test Reactor (ATR). Certain mechanical and thermal conditions cannot be directly evaluated through experiments and fuel performance modeling is used to shed light on this evolution over time. In this study, BISON Multiphysics simulations are leveraged to investigate the state of the fuel system throughout and after the experimental conditions.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

FASTGRASS implementation in BISON and Fission gas behavior characterization in UO 2 and connection to validating MARMOT

This activity is supported by the US Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Product Line (FPL). Two major accomplishments in FY 15 are summarized in this report: (1) implementation of the FASTGRASS module in the BISON code; and (2) a Xe implantation experiment for large-grained UO 2 . Both BISON AND MARMOT codes have been developed by Idaho National Laboratory (INL) to enable next generation fuel performance modeling capability as part of the NEAMS Program FPL. To contribute to the development of the Moose-Bison-Marmot (MBM) code suite, we have implemented the FASTGRASS fission gas model as a module in the BISON code. Based on rate theory formulations, the coupled FASTGRASS module in BISON is capable of modeling LWR oxide fuel fission gas behavior and fission gas release. In addition, we conducted a Xe implantation experiment at the Argonne Tandem Linac Accelerator System (ATLAS) in order to produce the needed UO 2 samples with desired bubble morphology. With these samples, further experiments to study the fission gas diffusivity are planned to provide validation data for the Fission Gas Release Model in MARMOT codes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Uranium carbide properties for advanced fuel modeling – A review

Uranium carbides are receiving renewed interest as a preferred nuclear fuel composition for advanced reactors due to their numerous favorable properties. Like many other refractory and transition metal carbides these compounds exist in both hypo- and hyper-stoichiometric compositions which results in significant variations in physical, thermal, and mechanical properties. This manuscript surveys both historic and recent literature to compile important properties data mainly as a function of temperature and carbon content to support fuel performance modeling. Expressions and fits for such properties are also suggested when allowed by complete data sets. Finally, the manuscript also attempts to highlight gaps and discrepancies in the reported data to assist the nuclear fuels community identify key areas that may require further attention in terms of future research and development.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Status of Multiple Channel Fuel Performance Capabilities Within the SAS4A/SASSYS-1 Safety Analysis Software

SAS4A/SASSYS-1 (SAS) is a fast-running simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced liquid-metal-cooled nuclear reactors. It is a critical element of safety analysis capabilities for the U.S. Department of Energy and is utilized within industry to perform the transient safety analyses required to support the licensing of Liquid Metal-cooled Fast Reactors (LMFRs). Although SAS is exceptionally fast for most transient scenarios, fuel performance calculations, along with the associated pre-transient characterization of the fuel pin, may be required for transient scenarios where fuel pin failure is hypothesized. Both the pre-transient characterization and the transient fuel performance calculation are necessary to properly quantify margins to potential fuel failure and assess the time spent potentially exceeding such margins during events. While safety analysis calculations with fuel performance models provide a more detailed characterization of the reactor during a transient, the pre-transient characterization can be time-consuming and computationally expensive. Often, large numbers of fuel pins have been exposed to similar pre-transient irradiation conditions. Similarly, the same pre-transient fuel characterization may be applicable to numerous transient conditions. This provides an opportunity to optimize the SAS computational framework such that pre-transient fuel characterization can be shared across multiple channels (fuel pins) and across multiple simulations, thus dramatically reducing overall computational costs. This report summarizes progress toward enhancing the SAS computational framework to support shared, multiple channel fuel performance characterizations intended to significantly reduce computational costs. Preliminary testing has shown that the computational time saved by using the pre-transient sharing capability is approximately equal to the time it takes to perform the pre-transient characterization.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Density functional theory calculations of the thermodynamic and kinetic properties of point defects in β -U

Density functional theory (DFT) calculations of the thermodynamic and kinetic properties of point defects in the β phase of uranium are reported. Defect energies and entropies were calculated using 2 x 2 x 2 supercells and the Generalized Gradient Approximation (GGA) of Perdew-Burke-Ernzerhof (PBE) for the exchange-correlation potential. Due to computational cost, calculations of the vibrational properties governing entropies were performed by only displacing atoms within (roughly) the 3rd or 4th nearest neighbor shell of the defect, which implicitly assumes that atoms beyond this distance are unaffected by the defect. Migration barriers were estimated by nudged elastic band (NEB) calculations. The low symmetry of the β-U phase (the unit cell is tetragonal and contains 30 atoms) results in many point defect configurations and even more migration pathways. A connectivity map, starting from the most stable point defects, was developed in order to identify the rate-limiting step controlling the net diffusion rate in each crystallographic direction. The uranium self-diffusivity tensor was calculated by combining the defect formation energies, entropies, migration barriers and attempt frequencies. The fastest diffusion rate was determined to be a vacancy mechanism in the z crystallographic direction. The predicted uranium self-diffusivity for this mechanism agrees well with available experimental data. Finally, the diffusion mechanisms and rates identified in this study will inform fuel performance models of swelling and gas evolution.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

In-situ TEM study of the ion irradiation behavior of U 3 Si 2 and U 3 Si 5

We report that U 3 Si 2 and U 3 Si 5 are two important uranium silicide phases currently under extensive investigation as potential fuel forms or components for light water reactors (LWRs) to enhance accident tolerance. In this paper, their irradiation behaviors are studied by ion beam irradiations with various ion mass and energies, and their microstructure evolution is investigated by in-situ transmission electron microscopy (TEM). U 3 Si 2 can easily be amorphized by ion beam irradiations (by 1MeV Ar 2+ or Kr 2+ ) at room temperature with the critical amorphization dose less than 1 dpa. The critical amorphization temperatures of U 3 Si 2 irradiated by 1MeV Kr 2+ and 1MeV Ar 2+ ion are determined as 580±10K and 540±5K, respectively. In contrast, U 3 Si 5 remains crystalline up to 8 dpa at room temperature and is stable against ion irradiation-induced amorphization up to ~50 dpa by either 1MeV Kr 2+ or 150KeV Kr + at 623K. In conclusion, these results provide valuable experimental data to guide future irradiation experiments, support the relevant post irradiation examination, and serve as the experimental basis for the validation of advanced fuel performance models

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Metallic Fuel Benchmark Simulations Based on the X430 Experiments

To gauge the accuracy of current BISON metallic fuel performance models and identify models that require further development, Oak Ridge National Laboratory has developed benchmark problems based on two pins of the X430 experiment. These benchmarks recreate the experimental conditions of the X430 series of experiments with as much accuracy as possible given the available data. The benchmarks were implemented in the BISON nuclear fuel performance code and the simulation results were compared with the experimental results to quantify the simulation accuracy. Findings in this work were consistent with previous benchmark problems. The temperature profiles, burnup, and fission gas behavior were all as accurate as can reasonably be expected. The cladding radial growth was accurate enough to not affect the results of other measurements. The fuel axial growth was accurate for U-Zr fuel but not for U-Pu-Zr fuel, which appears to be caused by a lack of U-Pu-Zr data. Swelling correlations have not been developed for U-Pu-Zr fuel because there were not enough data on this fuel type, so BISON used the U-Zr correlations. Finally, recommendations were made for near-term BISON development.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗