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At least 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Brittle failure analysis and modeling of high-burnup PWR fuel cladding alloys

The aim of this research is the development of methods for predicting mechanical behavior and identification of limiting conditions to prevent brittle failure of high-burnup (HBU) pressure water reactor (PWR) fuel cladding alloys. A finite element (FE) model of the ring compression test (RCT) was created to analyze the failure behavior of zirconium-based alloys with radial hydrides during the RCT. An elastic-plastic material model describes the zirconium alloy. The stress-strain curve needed for the elastic-plastic material model was derived by inverse finite element analyses. Cohesive zone modeling is used to reproduce sudden load drops during RCT loading. Based on the failure mechanism in non-irradiated ZIRLO (R) claddings, a micro-mechanical model was developed that distinguishes between brittle failure along hydrides and ductile failure of the zirconium matrix. Two different cohesive laws representing these types of failure are present in the same cohesive interface. The key differences between these constitutive laws are the cohesive strength, the stress at which damage initiates, and the cohesive energy, which is the damage energy dissipated by the cohesive zone. Statistically generated matrix-hydride distributions were mapped onto the cohesive elements and simulations with focus on the first load drop were performed. Computational results are in good agreement with the RCT results conducted on high-burnup M5 (R) samples. It could be shown that crack initiation and propagation strongly depend on the specific configuration of hydrides and matrix material in the fracture area.

Simbruner, Kai↗
DOE OSTI2023

Formation and characterization of hydride rim structures in Zircaloy-4 nuclear fuel cladding tubes

Zirconium alloy tubes used as nuclear fuel cladding are subject to oxidation and subsequent hydrogen pickup during their long service in commercial light water reactors. The hydrogen picked up in the cladding can precipitate as a brittle hydride rim feature on the cladding outer surface. Here, to better understand the effect of hydride rims on the fracture behavior of stress-relieved zirconium alloy cladding tubes, a procedure to produce these rim-like structures has been developed and is described herein. Extensive characterization of the ‘as hydrided’ tubes is performed. The hydrogen charging apparatus consists of a tube furnace with a quartz chamber that is connected to a vacuum pump as well as a bottle of pure hydrogen gas. The charging station uses a static charge of hydrogen as opposed to a flowing gas. The chief advantage of this approach is the ability to monitor the pressure drop in the hydriding chamber and correlate this pressure drop to a known rate of hydrogen pickup in the cladding tube. It was found that the hydrogen partial pressure, metal temperature, and surface treatment all clearly played a role in whether a hydride rim was formed. Extensive characterization of the hydride rims shows they consist of needle like platelets of δ phase hydrides (ZrH1.66) oriented in the circumferential direction with a radial spacing of several microns in a sandwich-like structure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Investigating the stability of reoriented hydrides and their reprecipitation using in-situ heating experiments

The behavior of radial hydrides in Zr-2.5Nb pressure tube samples at elevated temperature has been studied. The aim of this work is to investigate the stability of stress-reoriented hydrides under heating without an external applied load. Using in-situ heating at beamline, 1-ID at Argonne National Laboratory, USA, it was shown that two populations of hydrides existed in the sample, naturally oriented and reoriented hydrides. At a relatively low temperature of 180 °C it was observed that the reoriented hydrides completely dissolved, leaving only the naturally oriented hydrides in the sample, when heated without an applied load. On subsequent cooling of the sample, only naturally oriented hydrides were observed. These observations were confirmed using in-situ scanning electron microscopy heating experiments and post-mortem transmission electron microscopy studies. The results presented here offer a potential route for ductility recovery of pressure tube materials where radial hydrides have been shown to significantly reduce fracture toughness of components.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Development of Hydrogen Permeation Barrier and Matrix Interaction Protection Coatings for TZM Enclosure of Hydride Moderators (Microreactor Program M3 Report)

It is important to develop a high-performance enclosure solution for hydride-based microreactor moderation components to effectively suppress hydride thermal decomposition and prevent hydrogen loss in harsh in-reactor environments. Currently, the DOE-NE Microreactor Program (MR program) is investigating the titanium-zirconium-molybdenum (TZM) canned yttrium hydride (YH 2-x ) solution, which provides acceptable hydrogen retainment performance up to ~700 °C. However, improvements in the enclosure are necessary for increasing long-term operation temperature beyond 700 °C and enhancing short-term survivability at even higher temperatures during power transients. One potential solution is the addition of a dedicated hydrogen permeation barrier, which can potentially improve the enclosure's performance. Furthermore, the hydride moderator components will be inserted into the microreactor matrix, which may consist of materials that are not chemically inert against TZM. Therefore, the TZM might need to be protected from the external environment through the application of another surface protection coating. To address these critical issues, two types of barrier coatings optimized for TZM-based moderator enclosure has been developed. A multilayer coating approach has been used to identify TZM-compatible barrier materials. Behavior of these coating designs against H 2 permeation and long-term high-temperature graphite matrix interactions has been examined in detail. Additionally, the developed coatings' behavior when exposed to both high-temperature thermal cycling and heavy-ion irradiation has also been investigated. Overall, the development of these barrier coatings is essential for achieving maturity for the metal hydride moderator technology. The results in this report will provide crucial insights into the behavior of the developed coatings and their potential for improving the long-term performance of TZM-based moderator enclosures.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Benchmark Gap Assessment for the Manufacturing of High-Assay Low-Enriched Uranium Fuels

This document develops basic critical conditions for spheres—moderated and unmoderated, as well as reflected and unreflected—in consideration of nuclear criticality safety of a potential fuel production facility producing high-assay low-enriched uranium (HALEU) fuel of several different types like tristructural-isotropic (TRISO), uranium metal and alloys, oxide and non-metallic forms. In addition to spherical arrangements, TRISO particle manufacturing process–specific equipment is modeled as it would be for the criticality safety analysis. The objective is to develop representative systems that can then be used for comparison with existing benchmarks. SCALE/TSUNAMI is used to assess the similarity index between these systems to assess validation gaps for possible fuel production applications of proposed advanced reactors. Several different fuel types were evaluated, including TRISO, uranium metal, uranium molybdenum, uranium zirconium, uranium dioxide, uranium nitride, uranium hydride, U-ZrH, and uranium chloride. This selection of fuel types covers a breadth of proposed reactor types, as well as intermediate steps in the production and fabrication of the fuel

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Study of hydride reorientation and ductility of SRA ZIRLO and PRXA ZIRLO using ring compression testing (RCT)

Before dry storage, spent nuclear fuel is dried at high temperatures, which induces hydride dissolution. Radial hydride particles precipitate when cooled under a stress above a threshold stress. This process can severely degrade the ductility of cladding tubes. Zirconium alloys with different microstructures, such as grain boundary density, were quantified and studied to examine their effect on hydride reorientation. The ring compression loading mode creates a range of stress states depending on the location within the ring with respect to the applied load, producing various hydride morphologies that can be used to study the effects of microstructures on hydride reorientation. Finite element modeling was performed to calculate the hoop stress, which is the dominant stress, during hydride reorientation. These results were coupled with hydride morphology to determine the threshold stress for hydride reorientation, depending on the microstructure. In conclusion, the samples were then mechanically tested using ring compression to evaluate the ductility of zirconium alloys as a function of different hydride morphologies.

Hydride embrittlement↗
DOE OSTI2026

PIE-Enabled Study of Aqueous Corrosion & Zr Hydriding in Cr-Coated Cladding: M3GV-23PN0101132

Chromium-coated zirconium alloy cladding is under investigation as an accident-tolerant fuel (ATF) concept to extend performance to higher burnups via improved oxidation resistance and reduced hydrogen pickup. However, hydrothermal corrosion behavior and hydrogen transport mechanisms governing in-service hydriding of these coatings remain poorly understood. In a collaborative study between Pacific Northwest National Laboratory (PNNL), Idaho National Laboratory (INL), and the University of Huddersfield, cold spray (CS) and physical vapor deposition (PVD) Cr-coated Optimized ZIRLO™ cladding samples were characterized after exposure to PWR-simulated water chemistry under both in-core (neutron irradiation) and out-of-core (aqueous-only) conditions at the MIT Research Reactor. Multi-scale characterization was performed independently at PNNL and INL to evaluate coating integrity, microstructural evolution, Cr/Zr interface chemistry, and hydride formation. Both laboratories observed a consistent divergence in hydriding behavior: out-of-core CS Cr-coated cladding exhibited elevated hydride concentrations exceeding those of uncoated cladding, whereas in-core CS samples showed substantially suppressed hydriding. In-core specimens also displayed irradiation-specific features, including nanoscale voids within the Cr coating and radiation-induced segregation clusters in the Zr substrate. CS coatings retained an interdiffusion-free Cr/Zr bond with no intermetallic layer, whereas PVD coatings exhibited inferior quality with visible cracks and pores. An automated image-based hydride quantification method systematically overestimated bulk hydrogen content relative to inert gas fusion measurements, underscoring the need for standardized sample preparation and reference standards. Interpretation of hydride nucleation and growth is complicated by the open inner-diameter of the specimens, which provides additional hydrogen pathways. The results motivate further foundational studies to support predictive models of hydrothermal corrosion and hydrogen transport in unirradiated and irradiated Cr-coated cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Pressure Composition Temperature Curves of Hydride Moderators

Microreactors are compact and mobile reactors which operate at low power and high temperatures. These reactors generate heat which is then converted to electricity for use in a variety of special purpose applications. Microreactors are of interest due to their light weight, compact size and the ability to meet the needs for high energy in difficult or remote areas. Metal hydrides have been studied for nuclear and aviation applications. The emergence of microreactors the interest in metal hydrides is significant due to their ability to moderate neutrons in a light weight and solid form. In case of an off normal event, more accurate data is needed for the absorption/desorption rates of the hydrides due to the high uncertainties on hydrogen redistribution in the moderator. The literature shows considerable discrepancies in the PCT curves for yttrium and zirconium. The uncertainty is much more notable at higher stoichiometries of the hydride phase at high temperature. To understand the in-pile performance, a detailed understating of the impurities and microstructure is needed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1994

A low cost igniter utilizing an SCB and titanium sub-hydride potassium perchlorate pyrotechnic

A conventional NSI (NASA Standard Initiator) normally employs a hot-wire ignition element to ignite ZPP (zirconium potassium perchlorate). With minor modifications to the interior of a header similar to an NSI device to accommodate an SCB (semiconductor bridge), a low cost initiator was obtained. In addition, the ZPP was replaced with THKP (titanium sub-hydride potassium perchlorate) to obtain increased overall gas production and reduced static-charge sensitivity. This paper reports on the all-fire and no-fire levels obtained and on a dual mix device that uses THKP as the igniter mix and a thermite as the output mix.

R. W. Bickes, Jr.↗
DOE OSTI2023

The deformation and burst behavior of Zircaloy-4 cladding tubes with hydride rim features subject to internal pressure loads

This study is motivated by the desire to conduct a separate effects test replicating specific aspects of an in-pile pulse reactor test on high exposure light water reactor (LWR) fuel rods. The principal purpose of the tests was to determine whether temperature and or hydrogen concentration thresholds exist in assessing the vulnerability of a stress relieved zirconium alloy cladding tubes to low strain ruptures. A secondary objective was to determine whether observations of the cladding stress during deformation could aid in the understanding of the crack formation and propagation in cladding tubes with hydride rim features. Here, the study utilized artificially hydrided Zircaloy-4 cladding tubes and an isothermal, internal pressure test that was controlled with an in-situ strain measurement. The test results show a clear transition in failure vulnerability between room temperature and 150 °C. Below 150 °C, cladding tubes rupture at very low uniform plastic strain—even at relatively modest cladding hydrogen levels. At and above 150 °C, cladding tubes only rupture at low uniform plastic strain if the cladding hydrogen content is greater than 500 ppm and usually experience at least some modest amount of plastic deformation prior to rupture. Post-test metallographs from non-ruptured cladding tubes show numerous primary cracks forming in the hydride rim and arresting at the metal/hydride interface. This observation combined with the decrease in bulk yielding of hydrided cladding tubes when compared to fresh non-hydrided tubes leads to the conclusion that the fracture stress of the hydride rims is related to and bounded by the yield stress of the zirconium metal. Some evidence is presented for a change in crack propagation mechanism from one of void formation on hydride plates and linkage at temperatures below 150 °C to more classical ductile shearing beneath the primary crack tip at temperatures above 150 °C. This apparent change in crack propagation mechanism may help explain the observed temperature and hydrogen concentration thresholds.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

BISON BWR Fuel Modeling Capability: Material Models (FY2020)

Until this year the BISON fuel performance code has been developed primarily for pressurized water reactor (PWR), accident tolerant fuel (ATF), and advanced reactor (e.g., tri-structural isotropic (TRISO), metallic) analyses. The main focus of this work is to develop and document required material models that would aid predicting cladding failure with a specific hydride distribution to be investigated for boiling water reactor (BWR) applications in the next fiscal year. In terms of materials, the primary differences between a BWR and PWR is the use of gadolinia doping of the UO 2 fuel, the use of Zircaloy-2 for the cladding, and the addition of a liner on the inner surface of the cladding for improved pellet-clad mechanical interaction (PCMI)performance. The liner is typically made of pure zirconium or a lower tin content zirconium alloy (e.g. Zr-0.3%Sn). This work details the development and documentation of material and behavioral models for BWR analysis such as the addition of material and behavior models for gadolinia doped UO 2 , pure zirconium, Zr-0.3%Sn, and Zircaloy-2 as well as the ease-of-use additions to allow internal meshing capabilities to include liners. Another important aspect affecting cladding performance is hydrogen diffusion and hydride precipitation in the cladding. A model was added previously for non-lined, homogeneous claddings. In this work, instructions are provided on how to apply the model to BWR liner cladding. Representative experimental values are provided for the hydrogen diffusion parameters and precipitation-dissolution kinetics for typical liner materials of interest.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Tritium adsorption and absorption on (100) and (001) surfaces of pure and tin defective zirconium

Zirconium alloys such as zircaloy-4 are used as tritium (T) getter materials in tritium-producing burnable absorber rods (TPBARs) due to their ability to capture T, thereby forming metal hydrides. Developing an understanding of T adsorption onto zircaloy prior to diffusion into the subsurface is relevant for rational tritium getter and TPBAR design, to improve material properties for nuclear applications. Herein, density functional theory calculations revealed the preferred binding sites for T adsorption on Zr(001) and Zr(100). The energy barriers of T transfer, along the surface and from the surface to the subsurface were computed. The adsorption properties of Zr(001) were found to be superior to those of Zr(100). Surface tin impurities were found to strongly repel T. The presence of subsurface and surface tin resulted in higher absorption energy barriers for both the forward and reverse processes. Based on the calculated energy barriers, a surface to surface T diffusion coefficient of 9.53 × 10 -10 m 2 s -1 is expected for pristine Zr(001). Finally, a surface to subsurface T diffusion coefficient on the order of 10 -13 m 2 s -1 is predicted in pristine Zr, decreasing to 10 -19 m 2 s -1 for the transfer with a subsurface tin impurity.

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

Designing an impedance-based technique for studying corrosion on cladding materials

In nuclear cladding materials, changes in structure and chemistry over the lifetime of claddings can have a significant impact on fuel and reactor performance. Currently, chemical evolution over the lifetime of the fuel and cladding materials can only be inferred using post-irradiation examination. As an alternative to post-irradiation measurements, electrochemical impedance spectroscopy (EIS) can provide an in-pile measurement of changes in chemistry in oxides and hydrides. The objective of this study is to develop an impedance based sensing technology for measuring spatial- and time-resolved in-pile changes in cladding chemistry in order to understand the mechanisms of hydride formation, hydride dealloying and associated cladding corrosion so that a mitigation strategy and technology can be developed for reactor safe operation. Two prototype impedance-based sensors were designed and assembled using Zr-4 as working electrode and Pt as counter and reference electrode. One sensor is tubular type and the other one is plate-type. These sensors will be tested in static autoclave to investigate the effects of various environmental conditions (including temperature, pressure, and water chemistry) on cladding material corrosion mechanisms. These in-situ EIS sensors would provide corrosion-layer parameters in high-temperature water, such as (1) total layer thickness, (2) the evolution of oxide structures as uniform layer dielectric or sublayer(s), and (3) the possible electrochemistry of the volume corrosion process and the solid/liquid interface, such as the cyclic nature of corrosion of Zr alloys in high-temperature water. This work was focused on electrochemical sensor development, targeting cladding oxides/hydrides under thermodynamic equilibria; characterization of samples using Kelvin probe force microscopy (SKPFM), Raman spectroscopy, and scanning electron microscopy (SEM) with energy-dispersive X-ray spectroscopy (EDS) to provide high-resolution co-localized characterization of zirconium oxide and metal/oxide interface.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Study of Tritium Diffusivity in Pure and Sn-Defective Zr: A First-Principles Density Functional Theory Approach

Zirconium alloys (e.g., zircaloy-4) are used as tritium ( 3 H) getter materials in tritium-producing burnable absorber rods (TPBARs) owing to their ability to capture 3 H and chemically convert them into metal hydrides. Understanding of 3 H diffusion mechanisms in zircaloy is crucial for the optimal design of material performance in nuclear technology. Here, for this work, we perform first-principles density functional theory calculations to study the 3 H diffusion mechanism in pure and impure Zr with a low concentration of tin (Sn) atoms to determine the impact of the presence of Sn on the movement of 3 H atoms through the material. First, we calculated the diffusion barriers for 3 H in pure Zr by taking different migration pathways. We then introduced a low concentration of Sn impurity and systematically explored the impurity effect on the diffusion barriers for 3 H. Using our calculated diffusion energy barriers, we further obtained the diffusion coefficients and analyzed the results by comparing them with the experimental and previously calculated values. A diffusion coefficient of the order of 10 –8 m 2 /s is predicted. We also found that the presence of a Sn impurity could reduce the diffusivity up to 4 orders of magnitude. In conclusion, our results could serve as guidelines for further experimental investigations.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Hydrogen embrittlement of Zircaloy-4 fabricated by ultrasonic additive manufacturing

Ultrasonic additive manufacturing (UAM) was successfully applied to the zirconium material system to create a planar geometry. Following fabrication, SS-J3 type tensile specimens of the UAM and wrought Zircaloy-4 with a nominal gage section of 5×1.2×0.75 mm were machined for hydriding studies and mechanical testing. The SS-J3 type tensile specimens were gas-charged with hydrogen using a custom system that precisely controls hydrogen gas flow rate, hydrogen partial pressure, and temperature. To avoid altering the UAM material, the maximum process temperature was limited to 550°C. Using different initial hydrogen gas pressures and flow rates, various hydrogen contents (70–1755 wppm) were achieved for Zircaloy-4 specimens. Tensile testing shows that, regardless of hydrogen content, all UAM specimens measured yield strengths in the range of 557±16 MPa and ultimate tensile strengths of 660±4MPa. However, total elongation clearly decreased as a function of increasing hydrogen content. Further, at the lowest hydrogen content, the total plastic elongation measured 21.5%, and this value decreased to less than 2% when the hydrogen content was increased to 1000 wppm. Cross-sectional optical microscopy images revealed that the hydride distributions are randomly oriented. For the low hydrogen content specimen, these randomly distributed hydrides are isolated from each other. A sandwiched structure, consisting of high-density and low-density layers, was also observed for the specimens with hydrogen content between 200 and 400 wppm. As the hydrogen content increases, the hydrides diffuse into the low-density hydride layers to form a network across the whole specimen. Although the tensile properties of UAM and wrought Zircaloy-4 exhibit the same behavior with increasing hydrogen content, the distinctions in grain orientations led to differences in the hydride orientations at low and intermediate hydrogen concentrations.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Predictive machine learning approaches for the microstructural behavior of multiphase zirconium alloys

Abstract Zirconium alloys are widely used in harsh environments characterized by high temperatures, corrosivity, and radiation exposure. These alloys, which have a hexagonal closed packed (h.c.p.) structure thermo-mechanically degrade, when exposed to severe operating environments due to hydride formation. These hydrides have a different crystalline structure, than the matrix, which results in a multiphase alloy. To accurately model these materials at the relevant physical scale, it is necessary to fully characterize them based on a microstructural fingerprint, which is defined here as a combination of features that include hydride geometry, parent and hydride texture and crystalline structure of these multiphase alloys. Hence, this investigation will develop a reduced order modeling approach, where this microstructural fingerprint is used to predict critical fracture stress levels that are physically consistent with microstructural deformation and fracture modes. Machine Learning (ML) methodologies based on Gaussian Process Regression, random forests, and multilayer perceptrons (MLP) were used to predict material fracture critical stress states. MLPs, or neural networks, had the highest accuracy on held-out test sets across three predetermined strain levels of interest. Hydride orientation, grain orientation or texture, and hydride volume fraction had the greatest effect on critical fracture stress levels and had partial dependencies that were highly significant, and in comparison hydride length and hydride spacing have less effects on fracture stresses. Furthermore, these models were also used accurately predicted material response to nominal applied strains as a function of the microstructural fingerprint.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Molecular dynamics studies of lattice defect effects on tritium diffusion in zirconium

Tritium diffusion in α-Zr containing point defects such as vacancies or self-interstitial atoms (SIAs) is simulated using molecular dynamics. Point defects rapidly aggregate to form extended defects, such as 3D nanoclusters and Frank loops. The geometry of extended defects is affected by the presence of tritium. At low temperature and in the absence of tritium, vacancies aggregate to form stacking fault pyramids. Addition of tritium at these temperatures promotes aggregation of vacancies to form 3D nanoclusters, within which the tritium concentration can be sufficiently high to suggest that these defects may serve as nucleation sites for hydride precipitation. Trapping of tritium in vacancy nanocluster reduces the calculated bulk diffusivity by an amount proportional to the vacancy concentration. At high temperature, vacancy clusters change shape to form planar basal dislocation loops, which bind tritium less strongly, leading to a sharp reduction in the fraction of trapped tritium and a corresponding increase in tritium diffusivity at high temperature. In contrast, SIAs increase tritium diffusion through α-Zr. Analysis of atomic trajectories shows that tritium does not interact directly with SIAs. In conclusion, diffusion enhancement is instead related to expansion of the lattice.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Westinghouse Accident Tolerant Fuel Phase 2B with Higher Enriched and Higher Burnup Add-On Project Final Technical Report Deliverable Volume 1

The Westinghouse Electric Company LLC (Westinghouse) accident tolerant fuel (ATF) program funded by Westinghouse and the US Department of Energy (DOE) utilized chromium (Cr) coated zirconium alloy cladding with doped UO 2 (ADOPT(TM)) and uranium silicide (U 3 Si 2 ) high density/high thermal conductivity fuel for its lead test rod (LTR) program with irradiation beginning in 2019. ADOPT is a Cr 2 O 3 +Al 2 O 3 doped UO 2 pellet with increased oxidation resistance and density, and increased resistance to fission gas release. Due to the issues with U 3 Si 2 fuel reaction in pressurized water reactor (PWR) environments, uranium nitride (U 15 N) has been substituted for U 3 Si 2 as a long-term fuel option. Cr coated cladding with ADOPT fuel is the near-term Westinghouse EnCore fuel product. The long-term EnCore ® fuel product is SiGA ® SiC/SiC composite cladding with high density/high thermal conductivity UN fuel. In 2020, the higher burnup and higher enriched and (HBHE) program was integrated into the ATF program. The objective of this expanded program is to extend ATF burnups to at least 75 MWd/kgU to economically facilitate 24-month cycles in PWRs. The ATF program now includes ADOPT fuel with enrichments >5% 235 U. Over the past several years, Westinghouse has tested the Cr coated zirconium (Zr) and silicon carbide (SiC) claddings in Westinghouse Churchill autoclaves and the Massachusetts Institute of Technology (MIT) reactor. Additionally, the Cr coated cladding has been tested in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and in LTR programs at the Doel 4 and Bryon 2 commercial reactors. High temperature tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried out to determine the time and temperature limits for the Cr coated zirconium claddings. These tests indicate that Cr coated Zr cladding can take temperatures up to about 1500°C for short periods of time without becoming totally oxidized. The main issue is the formation of the Cr-Zr eutectic at 1333°C resulting in the migration of this eutectic inward with the formation of ZrCr 2 which rapidly oxidizes to ZrO 2 on the outside of the tube. The Cr coated Zr cladding also delays the bursting of the tube, reduces the burst area, and reduces hydriding and embrittlement of the Zr which is a major benefit for reducing fuel fragmentation, redistribution, and dispersal (FFRD), the major licensing issue faced by HBHE. The manufacturing parameters for the SiC have been found to have a significant effect on the corrosion rate of the SiC in light water reactor (LWR) conditions and significant improvements have been made in the yield, quality, and oxidation resistance of the SiC cladding by identifying and tightening manufacturing process parameters. Current autoclave results for SiC composite claddings indicate that a corrosion rate of fewer than 2 micrometers per year can be achieved which meets corrosion requirements under normal operating conditions. High temperature steam oxidation tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried and determined that SiC cladding can withstand temperatures up to about 1800°C to 1900°C before excessive corrosion reactions begin, and that the SiC will not balloon and burst. ADOPT fuel pellets were incorporated into the Westinghouse ATF program because of their increased density and oxidation resistance. A topical report supporting the use of ADOPT pellets has been submitted to the NRC and will be approved shortly. ADOPT pellets have been a product in Europe for over 15 years. The additives work to make larger grain sizes which appear to reduce fission gas release during transients as well as increase the density of the pellet. Fuel rod and assembly design in preparation for the lead test rod (LTR) and lead test assembly (LTA) programs is underway as well as licensing efforts with the Nuclear Regulatory Commission (NRC). Finally, accident analyses coupled with economic evaluations for both operating savings as well as fuel savings have been initiated. Modular Accident Analysis Program 5 (MAAP5) calculations indicate that both Cr coated cladding and SiC composite cladding can increase the time to fuel rod loss of geometry for up to two hours longer than current Zr based cladding in a station blackout scenario. This additional time is due to the much higher oxidation resistance of the SiC and Cr coated cladding. These two hours can be used to implement additional response options instituted through the Diverse and Flexible Mitigation Capability (FLEX) program by reactor operators. Both ATF cladding options also reduce hydrogen production which dramatically reduces primary system and containment pressure and the risk of fission product release beyond containment in the unlikely event of an accident. The lower pressure in the system allows more time to feed cooling water to the core, resulting in the avoidance of fuel melting. The coping time is extended indefinitely if the modest water flow provided by FLEX continues. Experimental work on methods to rapidly fabricate SiC composite structures with high density and reduce the fabrication price of SiC fibers while maintaining a high level of performance is needed. Methods for mitigating or stopping the Cr-Zr eutectic are also needed to further improve the oxidation resistance of Cr coated Zr. Minimal (<1%) swelling of U 3 Si 2 and subsequent fission gas release has been demonstrated up to 20 MWd/kgU. Irradiation experiments with U 3 Si 2 fuel in ATR to determine these properties at 40 MWd/kgU were completed but post irradiation examinations were not done since U 3 Si 2 has been replaced by UN. UN has three issues that are being addressed. The first is increasing the oxidation resistance so that there is not excessive reaction up to ~1500°C. While UN increases the pellet density and additives to the UN postponed the temperature at which rapid oxidation occurred by ~100°C to 200°C, this is not enough, and pellet coating options are now being pursued. The second is developing a method to manufacture that does not require the multi-step process of UF 6 to UO 2 to UC to UN. Efforts are underway looking at UF 6 to UN 2 to UN and UF 6 to UF 4 to UN 2 to UN reactions. Finally, an economically acceptable process for potentially enriching the 15 N content of the nitrogen used to make UN to >95% 15N was identified though not experimentally validated. In addition to the work supported by the DOE, research and testing activities are being carried on in a world-wide effort funded by many countries such as Sweden, United Kingdom, Belgium, Netherlands, Spain, Germany, Japan, and France. This work is being facilitated through the Westinghouse led Collaboration for Advanced Research on Accident Tolerant Fuel (CARAT) program. Annual meetings were organized in 2018 and 2019 by Westinghouse as a venue for presentation of this work and to provide for the cross-fertilization of ideas among the many outstanding researchers in the ATF area. No meetings were held in 2020 and 2021 due to the COVID-19 related restrictions on traveling and gatherings. Since SiC, coated cladding, and high-density fuel options are not currently used in the nuclear industry; support from the government and industry members is needed to further the significant effort of setting new standards. The same is true for the Nuclear Regulatory Commission (NRC) which must review and approve the commercial use of these new fuels since all current regulations are oriented toward Zr/UO 2 fuel. Several meetings have been held with the NRC to generate a fast-track approach to licensing ATF using a combination of atomic modeling, in-rod sensors, and in-reactor testing. This approach was memorialized in an “Accelerated Fuel Qualification White Paper” prepared by the Accelerated Fuel Qualification Working Group lead by General Atomics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗