Benchmarking SiCf/SiC, FeCrAl, and Cr-coated Zr alloy claddings: evaluating thermomechanical properties and high-temperature oxidation resistance of next-generation nuclear claddings

Theory-guided bottom-up design of the FeCrAl alloys as accident tolerant fuel cladding materials

Zirconium-based alloy claddings used for current light water reactors (LWRs) possess a variety of desirable features in steady-state normal operation, however, constraints regarding fast degradation, rapid exothermic reaction with high-temperature steam associated with hydrogen generation in accident scenarios motivate the requisite to develop enhanced accident tolerant fuel (ATF) claddings. One reasonable solution to improve the accident tolerance of the zirconium alloy cladding in accidental conditions while preserving its excellent behavior under normal operating conditions is external surface modification via such as coatings deposition. In addition, protective coatings applied to the zirconium alloy claddings offer the potential benefits of drastically reduced corrosion and degradation during normal operation, which are expected for application within the design framework of both current and future generation LWRs. The Mn+1AXn (MAX) phase materials comprise an extended family of layered, hexagonal ternary carbides and nitrides. They combine many attractive properties of both ceramics and metals stemming from their unique layered crystal structures and bonding characteristics; certain Al-MAX phases also possess excellent high-temperature oxidation resistance and chemical compatibility with select coolants such as hot water and molten lead. The objectives of this work are to synthesize and to evaluate three Al-containing MAX phase carbides (Ti2AlC, Zr2AlC and Cr2AlC) as potential protective coatings on Zircaloy-4 substrates with an emphasis on their high-temperature oxidation performance in steam. Oxidation of one commercial bulk Al-MAX phase Ti2AlC (Maxthal 211®), as reference/benchmark material, in steam in the temperature range of 1400°C - 1600°C was investigated to validate its high-temperature oxidation resistance and ascertain its potential as protective coatings. Oxidation of bulk Ti2AlC MAX phase ceramic at 1400°C and 1500°C formed a continuous coarse α-Al2O3 scale with randomly distributed Al2TiO5 isolated areas on the surface. The oxide scale thickening rate of Ti2AlC is more than three orders of magnitude lower than that of Zircaloy-4 at 1400°C and the maximum tolerance temperature of Ti2AlC in steam is approximately 1555°C. Therefore, these findings hold great promise of Al-containing MAX phase carbides, especially Ti2AlC, as oxidation resistant coating on zirconium-based alloy claddings. An innovative two-step approach has been established, i.e. first magnetron sputtering of nanoscale elemental multilayer stacks and subsequently thermal annealing in argon, for potential growth of phase-pure MAX phase coatings. The crystallization behavior and phase evolution of the as-deposited multilayers during annealing were systematically investigated using a diverse range of characterization and analytical techniques. The mechanical properties of designated coatings were evaluated by means of scratch tests and nanoindentation. Thermal annealing of the nanoscale elemental multilayer stacks (transition metal layer/carbon layer/aluminum layer) in argon revealed that onset crystallization temperatures of the Ti2AlC and Cr2AlC MAX phase from competing binary carbides and intermetallic phases locate at around 660°C and 480°C, respectively. Phase-pure Ti2AlC and Cr2AlC coatings were successfully fabricated, but the formation of a mixed ternary Zr(Al)C carbide rather than the Zr2AlC MAX phase was confirmed. Both MAX phase coatings have a basal-plane preferred orientation with the c-axis perpendicular to the sample surface and the multilayer stacks. The Zr/C/Al coatings crystallized to a cubic, solid solution Zr(Al)C phase with a B1 NaCl crystal structure. The phase evolution during annealing appears associated with the thermodynamic stability of corresponding MAX phases and their counterpart binary carbides. Coatings of altered designs with respect to introducing diffusion barrier or overlayer were deposited on Zircaloy-4 substrates. The coating thicknesses are 5~6 μm. Oxidation performance and degradation of these coatings during exposure in steam at elevated temperatures were investigated by thermogravimetric analysis and examining the microstructural evolution of the coating-substrate system. Steam oxidation tests found no protective effect of the Ti2AlC and Zr(Al)C based coatings with significant spallation and cracking from around 1000°C. Growth of an Al2O3-rich layer with TiO2 or ZrO2 layer beneath for the Ti2AlC and Zr(Al)C based coatings, respectively, was observed rather than a dense alumina layer. The failure of the Ti2AlC and Zr(Al)C based coatings from 1000°C can be attributed to the low thickness of the coatings, high interdiffusion rate between coating and substrate and potential phase transformation of the oxide products. The Cr2AlC-based coatings possess superior oxidation resistance up to at least 1200°C and autonomous self-healing capability with a thin and dense α-Al2O3 layer growth. Another design with a thin Cr overlayer above the Cr2AlC coating was further developed to eliminate potential fast hydrothermal dissolution of Al during normal operation. Moreover, first neutron radiography investigations of hydrogen permeability through the Ti2AlC and Cr2AlC MAX phase coatings on Zircaloy-4 substrates were reported. Hydrogen permeation experiments through non-oxidized and pre-oxidized Ti2AlC and Cr2AlC MAX phase coatings on Zircaloy-4 evidenced that both coatings are robust hydrogen diffusion barriers and impede hydrogen permeation into the matrix efficiently. The unique microstructural features of the coatings, namely free of columnar growth and highly basal-plane textured grains owing to the two-step approach, improve their efficiency in limiting hydrogen permeation as a barrier.

After the Fukushima Daiichi Accident, the safety features such as accident tolerant fuel (ATF) and diverse and flexible coping strategies (FLEX) for existing nuclear fleets are being investigated by the US Department of Energy under the Light Water Reactor Sustainability Program. This research is being conducted to quantify the risk-benefit of these safety features. Dynamic probabilistic risk assessment (DPRA)-based response-surface approach has been presented to quantify the FLEX and ATF benefits by estimating the risk associated with each option. ATFs with multilayered silicon carbide (SiC), iron-chromium-aluminum, and chromium-coated zirconium cladding were considered in this study. While these ATF candidates perform better than the current zirconium cladding (Zr), they may introduce additional failure modes in some operating conditions. The fuel failure analysis modules (FAMs) were developed to investigate ATF performance. The dynamic risk assessments were performed using RAVEN, a DPRA tool, coupled with RELAP5 and FAMs. A cumulative distribution function-based index provided a mean of comparing the benefits of safety enhancements. For medium break loss of coolant accidents, FLEX operational timing window for each fuel type was estimated. Among these ATF candidates, SiC-type ATF was the most beneficial candidate for an increased safety margin than Zr-based fuel and was found to complement FLEX strategies in terms of risk and coping time.

Based on lessons learned from the Fukushima Daiichi nuclear power plant accident, pursuit of accident tolerant fuel (ATF) has been discussed by many institutions in the world. Toshiba identified a silicon carbide (SiC) ceramic as the most promising material for accident tolerant fuel. Since SiC has less active characteristics in the presence of high temperature water steam (H2O) and is expected to be tolerant of severe accident conditions. Moreover, SiC has a smaller neutron absorption cross-section which is advantageous feature in terms of neutron economy. Zirconium alloys (Zry) are one of the main structural materials in LWR core. In high temperature H2O environment under severe accident conditions, Zry rapidly reacts with H2O and oxidation reaction accompanied by release of hydrogen gas occurs. Since SiC may inhibit the progress of oxidation reaction compared to Zry metal alloys, hydrogen and heat generation is expected to decrease in the case of core uncovered accident conditions. In order to confirm the advantage of SiC over Zry as core materials, transient analysis and safety analysis are carried out. For transient analysis, analyses of temperature behavior of cladding at plant transient condition are carried out with best-estimate transient analysis code. This analysis confirmed the effect of physical properties differences between SiC and Zry on cladding temperature behavior. Moreover to indicate the effectiveness of SiC under the core uncovered condition with oxidation reaction, safety analysis with latest “MAAP” code is carried out and the whole plant behavior during severe accident sequence is simulated. This analysis showed the effectiveness of SiC to mitigate the oxidation reaction. As the result of these analyses, the advantage of SiC over Zry can be perceived. And also, future challenges of SiC application as ATF can be clarified through these analyses.

Neutronic evaluation of coating and cladding materials for accident tolerant fuels

Development, implementation, and validation of material and behavior models for accident tolerant fuel (ATF) concepts in the Bison fuel performance code began in 2014 in response to the events that occurred at the Fukushima Daichii nuclear power plant in March 2011. Early on the focus was on U3Si2 fuel and FeCrAl cladding as part of a high impact problem through the Nuclear Energy and Advanced Modeling Simulation (NEAMS) program. Then, developments for Cr2O3-doped UO2 fuel, and SiC-SiC and Cr-coated zirconium-based claddings began based upon industry interests. In late fiscal year 2018 the Consortium for Advanced Simulation of Light Water Reactors (CASL) took over further ATF work in Bison in support of the Nuclear Regulatory Commission (NRC) engagement. Discussions with the NRC identified their list of priority fuel and cladding concepts, which included Cr2O3-doped UO2 and U3Si2 fuels, and Cr-coated zirconium-based and FeCrAl claddings. In particular, the NRC suggested that reports similar in form to NUREG/CR-7024 [1] that was developed for traditional LWR materials UO2 and zirconium-based claddings (i.e., Zircaloy-4, M5®, ZIRLO) be created for the priority ATF concepts. The approach to ATF capability development in Bison since the beginning has been two-fold: (1) empirical correlations and (2) multiscale model development. Both approaches have uncertainty inherent to them. Uncertainty in empirical correlations is bounded by the experimental data upon which with the correlation was developed. Models developed through a multiscale approach have uncertainty associated with the lower length scale calculations and input parameters that must be propagated to the engineering scale model in Bison. In this report, the recommended models, their range of applicability (e.g., temperature, burnup), and associated uncertainty for the NRC priority fuel concepts Cr2O3-doped UO2 and U3Si2 are presented in a manner similar to the aforementioned NUREG. In addition, the Cr2O3-doped UO2 models are validated to the Halden Reactor tess IFA-677.1 rods 1 and 5 and IFA-716.1 rod 1. For U3Si2 the models are validated to the recent post-irradiation examination (PIE) data ATF-13 and ATF-15 rods that were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory. Finally, an uncertainty quantification (UQ) and sensitivity analysis (SA) is completed on a single rod for each fuel concept that takes into account the defined uncertainty in select models. The results of the UQ and SA analyses indicate that given the large uncertainty in some of the input ATF models that the uncertainty on the Bison predictions of fuel performance metrics of interest bound the available experimental data. Correlation coefficients are also reported that identified the uncertain inputs with the strongest correlation (positive or negative) with the outputs of interest. Fuel performance metrics investigated include fuel elongation, rod internal pressure, fuel centerline temperature, and fission gas release.

A review of irradiation-induced hardening in FeCrAl alloy systems for accident-tolerant fuel cladding

Unveiling the latest progresses in chromium-coated Zircaloy cladding ATF materials: Fabrication techniques, performance metrics, and beyond

The possible deployment of Accident Tolerant Fuels (ATF) for currently-operating Light Water Reactors (LWR) has prompted interest in the use of Studsvik’s CMS5 code system to support the analysis of such advanced ATF core designs. Various ATF concepts have been proposed; for example, uranium silicide (U3Si2) fuel, together with iron-based (FeCrAl) cladding. The purpose of this work is to showcase the application of the CMS5 code system, which includes the CASMO5 advanced lattice physics code and the SIMULATE5 three-dimensional nodal simulator, to the analysis of a U3Si2/FeCrAl ATF concept. Given that the CMS5 code system was designed from inception to enable the analysis of advanced core designs, only minor changes to the CASMO5 lattice physics code and SIMULATE5 core simulator are necessary. The current CASMO5 586 energy-group nuclear data library provides all the necessary data to support the generation of homogenized data for downstream use by SIMULATE5 for ATF. The SIMULATE5 nodal code, which features a simplified fuel pin model, requires updating various thermophysical properties corresponding to the U3Si2/SiC ATF fuel and the gaseous conductance models. An equilibrium core for the Integral Inherently Safe (I2S) LWR design developed by the Georgia Institute of Technology was selected. The results of the CMS5 simulation were compared with those in the literature and were found to be in good agreement, giving us confidence that the CMS5 package can be used in the modeling of LWR systems with ATF technology.

Evaluation of accident tolerant cladding materials in a severe accident of the BNPP

Development of Accident Tolerant Fuel Cladding (ATFC) materials as the improvement for nuclear safety has increased intensively over the past decade as the lesson-learned from Fukushima Daiichi nuclear accident. Interaction of the zirconium alloy as the cladding material of the reactor with water vapor at high-temperature because of abnormal condition generated the hydrogen gas as the product of oxidation. Silicon carbide ceramic (SiC) is one of ATFC candidates to overcome the issue due to its high-temperature corrosion resistance. The neutronic investigation of SiC fuel cladding for LWR (Light Water Reactor) was done to analyse the feasibility of the material for the reactor from reactor physics view point. SRAC (Standard Reactor Analysis Code) as a deterministic neutronic code with cell and core calculation was used. In this study, the criticality of the reactor system with SiC fuel cladding was calculated and compared with Zircaloy-4 as the referenced standard cladding material for LWR especially PWR (Pressurized Water Reactor). Furthermore, neutron energy spectrum behavior at BOL (Beginning of Life) and EOL (End of Life) and excess reactivity behavior of the LWR core as a function of fuel burnup were calculated. The result showed that the reactivity of SiC cladding reactor core was slightly lower than Zircaloy-4 cladding reactor core. Furthermore, SiC cladding material demonstrated the negative temperature coefficient of reactivity of the reactor core examined, which is an important feature of the safety characteristic of LWR. It is shown that the use of SiC fuel cladding is feasible to be used the LWR reactor from reactor physics view point.

Research and development on accident tolerant fuel (ATF) have been widely pursued by the nuclear industry around the world. Westinghouse is developing new accident tolerant fuel design, EnCore® Fuel. Chromium (Cr) coated zirconium alloy and silicon-carbide (SiC) fuel cladding are the excellent candidates for application in fuel cladding of light water reactors (LWRs)[1, 2]. The ATF fuel is being developed to deliver beyond-design-basis (BDB) and design-basisaltering (DB) safety margins, withstanding far more severe conditions than the current fuel. The products in this portfolio have enhanced temperature performance enabling survival during BDB and DB accidents. The products also minimize interactions of fuel and fuel rod materials with water. The Cr-coated zirconium alloy cladding exhibits sustained high temperature performance and inhibits the detrimental effects of the zirconium-steam reaction. Westinghouse’s silicon-carbide (SiC) fuel cladding offers safety benefits in severe accident conditions, particularly compared to the significant hydrogen-and-heat producing reactions that occur above 1200° C for zirconium fuel cladding. These advancements will increase safety margins and enable the transition to high burnup facilities as well. To support the accident tolerant fuel (ATF) development, the Westinghouse Ultra High Temperature (UHT) test facility was built at the Westinghouse Churchill site in early 2016. The UHT test facility has been upgraded recently. The current UHT reactor was designed to promote better sealing of the reaction chamber and a smaller uncertainty on cladding temperature measurement. Simulating large break LOCA steam conditions, the UHT oxidation tests have been performed with SiC Optimized ZIRLO® and Cr-coated zirconium alloy cladding. In the EnCore Fuel program, one of the key tasks is to test the safety features and margin gain of the EnCore Fuel. This paper describes the designs, safety features and baseline commissioning tests for the upgraded Ultra High Temperature (UHT) test facility.

After the Fukushima Daiichi nuclear accident in 2011, the performance of nuclear fuel during accidents became a matter of great concern. To address this, a new type of fuel technology called accident‐tolerant fuel (ATF) has been developed with the goal of enhancing the ability of light water reactors (LWRs) to withstand severe accident conditions. Iron‐based alloys have been suggested as potential candidates for fuel cladding due to their favourable thermomechanical properties, lower reactivity with steam, and lower hydrogen generation. This study evaluates the neutronic performance of C26M (a 2nd generation nuclear grade FeCrAl alloy), APMT™, 310SS, and 304SS cladding materials by comparing them with Zircaloy‐4 cladding in a 3D PWR core at the beginning of the cycle (BOC) using OpenMC code. The results revealed that the neutronic penalty varied for different alternative cladding materials where C26M exhibited the lowest neutronic penalty value of ‐12551 pcm, while 310SS demonstrated the highest with a value of ‐17855 pcm. Additionally, important parameters in the reactor core such as neutron spectrum, reactivity coefficients, boron worth, control rod bank worth, power distribution, and radial thermal neutron flux distribution are evaluated and discussed. The analysis results showed that C26M provided a significantly higher level of neutronic performance compared to APMT™, 304SS, and 310SS. Although this study primarily focused on the neutronic performance of PWRs at BOC, future research should encompass fuel depletion analysis to delve deeper into the potential of alternative cladding materials.

Neutron cross section sensitivity and uncertainty analysis of candidate accident tolerant fuel concepts

High-temperature oxidation behavior and corrosion resistance of in-situ TiC and Mo reinforced AlCoCrFeNi-based high entropy alloy coatings by laser cladding