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In order to meet the growing energy demand, more environmentally friendly and efficient GEN-IV reactors have emerged. Additionally, nuclear structural materials need larger more safety margins for accident scenarios as a result of the Fukushima accident. In order to extend the failure time and lessen the effect of accidents, this design method for accident-tolerant fuel materials calls for cladding materials to maintain good corrosion resistance and mechanical properties during a beyond design basis accident (BDBA). Accidents involving nuclear reactors would undoubtedly result in higher temperatures, which would make it much harder for materials to withstand corrosion. Oxide dispersion strengthened (ODS) FeCrAl alloys have shown promise as candidate materials because of their extraordinarily slow reaction rates under high-temperature steam. However, the addition of the Al element renders the alloy's high-temperature mechanical properties insufficient. In particular, studies on the alloy's creep properties are extremely rare, despite the fact that the creep properties are crucial in the real service environment. Therefore, this paper focuses on the creep properties of ODS FeCrAl alloy, summarizes and analyzes the research results of this material, and provides a reference for future research and applications. [ABSTRACT FROM AUTHOR]

From a safety point of view, the fuel-cladding of the current design of the KLT-40S reactor still carries a potential risk in the event of a loss-of-coolant accident (LOCA) allowing the formation of hydrogen gas. The concept of accident tolerant fuels (ATF) offers a variety of new safer fuel-cladding materials, one of which is U3Si2-FeCrAl, a potential fuel-cladding combination according to various research sources. In this research, a study of neutronic parameters (1) cycle length, (2) reactivity feedback coefficient, and (3) reactor proliferation resistance was performed with ATF material U3Si2-FeCrAl as fuel-cladding in the KLT-40S reactor core. Modeling and simulation of the ATF-fueled KLT-40S reactor core were performed using KENO-VI and TRITON modules from SCALE code. The results showed that replacement of the fuel-cladding material with the ATF material in the KLT-40S reactor resulted in a shorter cycle length, and the enrichment required to reproduce the original cycle length was above the safeguard limit. The fuel temperature, moderator temperature, and void reactivity coefficient were negative, although not as negative as the original ones. The spent fuel produced at the end of the cycle had good proliferation resistance, although not as good as the original one. [ABSTRACT FROM AUTHOR]

Summary: Among various accident tolerant fuel (ATF) concepts, U3Si2–FeCrAl fuel systems are considered as ones of the most potential fuel for the next generation light water reactor. In this work, a multi‐objective optimization framework based on multi‐objective genetic algorithm (GA) and Gaussian process (GP) was established for the optimal design of U3Si2–FeCrAl ATF fuel element. The coupling of GA and surrogate GP model could take advantage of the high fidelity of multi‐physics modeling and the global searching capability of GA, therefore allowing to perform a relatively fast and accurate multi‐objective optimization design for the U3Si2–FeCrAl ATF fuel element. Maximum fuel temperature, maximum cladding Von–Mises stress, and maximum plenum pressure are considered as the three objective functions for the optimization problem in this study. According to the Pareto approximation surface obtained in the optimization process, a lowest maximum fuel temperature of 871 K, a lowest maximum cladding stress of 87.9 MPa, and a lowest plenum pressure of 3.1 MPa for the U3Si2–FeCrAl ATF fuel element could be achieved based on different design parameters. The multi‐objective optimization framework developed in this work could perform as an efficient tool for the multi‐objective optimization design of the ATF fuel element. [ABSTRACT FROM AUTHOR]

In the development for advanced nuclear fuels for LWRs, the focus has traditionally centered on a limited array of materials like doped-UO 2 , UN, and U 3 Si 2 , leaving the vast majority of potential uranium compounds unexplored. To expand the search and expedite the discovery of advanced LWR fuels, we developed a machine learning (ML) classification model dedicated to identifying uranium compounds with excellent thermal conductivity. However, the model's predictive accuracy and the feasibility of such an ML-assisted approach to advanced nuclear fuel discovery have yet to be empirically validated. Therefore, this study examines the thermal properties of two uranium compounds, UFe 3 B 2 and USiNi, that are predicted to outperform UO 2 at high temperatures. Our findings confirm that the model precisely predicted the thermal conductivities of both compounds, thereby adding credibility to its reliability. However, the experimental process unveiled several challenges, such as the fabrication of single-phase USiNi and microcrack-free UFe 3 B 2 samples, which our ML model, focused primarily on thermal conductivity, could not anticipate. These challenges underscore the need for a more comprehensive ML framework that considers the intricacies of sample fabrication. This advancement is crucial for the more efficient development of practical, accident-tolerant, advanced LWR fuels. • Validated ML model's predictions for uranium compounds' thermal conductivity. • First report of UFe 3 B 2 and USiNi thermophysical and mechanical properties. • Identified unforeseen experimental challenges beyond ML prediction. • Emphasized need for comprehensive ML framework for advanced nuclear fuel development. [ABSTRACT FROM AUTHOR]

In this paper, two kinds of composites constructed by UO 3 nanoflakes (which is UO 3 ·yNH 3 ·xH 2 O, precisely) and different graphene oxide (GO) nanosheets (homemade GO, GO-h, and commercial GO, GO-c) were prepared via solution chemical reaction. The two kinds of GO share similar morphology and possess the same species of functional groups. Nevertheless, GO-h possesses higher oxidization degree and thus could adsorb more uranyl ions, and the corresponding composite shows a much more regular shape: UO 3 nanoflakes and GO nanosheets cross each other, which would help overcome the disadvantage of graphene aggregation. As a result, homemade reduced graphene oxide (RGO-h) sheets inside the final UO 2 pellets (named as UO 2 /GO-h) could construct a well-bulit thermally conductive network to enhance its thermal conductivity. However, similar RGO-c network cannot be observed in the UO 2 /GO-c pellest originated from GO-c. Therefore, only the thermal conductivity of UO 2 /GO-h pellet acquires a dramatic increase, 37.30% relative to UO 2 pellet at 1200 °C, showing that UO 2 /GO-h possesses great potential for the development of novel UO 2 -based ATF fuels. [ABSTRACT FROM AUTHOR]

• This paper investigates the core characteristics of a reactor loaded with ATF. • Cr-coated and LAS doped fuel pellet ATF were considered. • The overall impact on core performance is not significant exculding cycle length. Following the Fukushima Daiichi accident, efforts have been made in a variety of disciplines to respond to nuclear accidents. From the perspective of nuclear fuel, various approaches have been attempted, and development has been made as an Accident Tolerant Fuel (ATF). Advanced fuel technology, often known as advanced fuel concepts, is ATF. The ATF is a type of fuel designed to be more resistant to damage in the event of a loss-of-coolant accident (LOCA) or other type of incident. Some ATF designs make use of materials that can better withstand high temperatures and pressures, which can help prevent fuel melting and release of radioactive materials. Currently, we are in the stage of developing a method to apply Cr coating on zircaloy cladding or a new cladding made of SiC. Additionally, some minerals like Cr, Mn, or La 2 O 3 -Al 2 O 3 -SiO 2 (LAS) are doped into pellet materials to alter their characteristics in order to create ATF fuel. In this paper, the characteristics of a core loaded with Cr-coated clad, and LAS doped fuel pellet ATF was investigated. The multiplication factor, power distribution, peaking factor, moderator temperature coefficient, and cycle length of the ATF-loaded core were calculated and compared to the results of the ATF-free core. The neutron absorption increases when the ATF is loaded due to the Cr coating and LAS doped fuel pellet, resulting in a difference when compared to the reference case. However, this study confirmed that the difference attributable to ATF was not significant except for the cycle length. To compensate for the reduced cycle length due to ATF, the cycle length and peaking factor were evaluated for the core with increased fuel enrichment and compared with the results of the conventional core. [ABSTRACT FROM AUTHOR]

Summary: This article is mainly concerned with the impacts of various accident tolerant fuel (ATF) claddings on neutronics performances of a soluble‐boron‐free (SBF) small modular reactor (SMR) core. There are two ATF cladding concepts which are evaluated here: (a) coating Zircaloy‐4 cladding with a thin layer of Cr or Cr alloys; (b) high‐strength and oxidation‐resistant claddings: stainless steel and FeCrAl. Comparisons between Zircaloy‐4 and ATF claddings are done in terms of the cycle length, discharge burnup, pin peaking factor (PPF), burnup reactivity, and spectral change. Moreover, the ATF claddings are also compared in view of the rim effect, neutron absorption by the cladding and He production in the cladding. In addition, a linear reactivity model is used to estimate the required U‐235 enrichment so that the cycle length with ATF claddings should be equivalent to that with the reference Zircaloy‐4 case. Furthermore, impacts of a selected ATF cladding are then analyzed in a centrally shielded burnable absorber (CSBA)‐loaded FA in terms of PPF, burnup reactivity, and spectral change. Based on the CSBA‐loaded FA analysis, a minor modification of burnable absorber loading strategy in the SBF autonomous transportable on‐demand reactor module core is proposed to adopt the selected ATF cladding without compromising the core performance. The lattice calculations are done using the Monte Carlo Serpent 2 code with the ENDF/B‐VII.1 nuclear library, while the 3‐D multi‐physics core calculations are performed using a Monte Carlo‐diffusion hybrid procedure. [ABSTRACT FROM AUTHOR]

Recently, uranium carbonitride solid solution has been recognized as a promising nuclear reactor fuel due to its excellent physicochemical properties. However, due to the complexity of U-N-C ternary system, the research on the structure and properties of uranium carbonitride is still insufficient. In this work, a series of promising accident-tolerant ternary compounds in the U-N-C system were predicted through a combination of first-principles calculations with an evolutionary algorithm. Fifteen dynamically and mechanically stable novel U-N-C ternary compounds with seven chemical compositions, including U-N-C solid solutions and new intermetallic structures, were suggested to be promising for experimental synthesis. For U-N-C solid solutions with rock-salt structures, it was revealed that the lattice expansion and bond length elongation caused by the substitution of carbon atoms for nitrogen atoms can lead to a decrease in elastic moduli and melting point and an increase in toughness. Moreover, two newly identified structures P 6 3 / mcm -U 3 N 2 C and P 2/ m -U 4 NC were considered to be very promising accident-tolerant fuels due to their high mechanical strength, excellent melting point, low thermal expansion coefficient, and high thermal conductivity. The larger atomic number density of P 6 3 / mcm -U 3 N 2 C enables a higher elastic modulus and a satisfactory melting point than those of rock-salt structures. While for P 2/ m -U 4 NC, its unique shorter and stronger U-N/C and U-U bonding leads to its highest mechanical strength and appropriate melting point. Furthermore, the linear thermal expansion coefficients of P 2/ m -U 4 NC and P 6 3 / mcm -U 3 N 2 C are comparable to those of UN and UO 2 , respectively, and their thermal conductivities increase with temperature, much higher than that of UO 2. [Display omitted] [ABSTRACT FROM AUTHOR]

Demonstration of hermetic SiC fiber–reinforced SiC matrix composite cladding under normal operating environments has been identified as one of the most critical feasibility issues for accident-tolerant fuel cladding in light-water reactors. This study provides critical experimental data needed for understanding the effects of irradiation on hermeticity. SiC composite and monolithic tubes were neutron-irradiated to 2 displacements per atom with and without a nominal radial heat flux of 0.6 MW/m2 to produce a simulated in-pile stress state for the normal operation of a light-water reactor. The through-thickness temperature gradient under irradiation results in a gradient in swelling, which causes a significant stress buildup. Such irradiation-induced stress was modeled using a commercial finite element analysis code. The radial heat flux–irradiation synergism was experimentally investigated by constructing a special irradiation capsule and evaluating the helium hermeticity of the specimens. The simulated stress state exhibited a near equi-biaxial tensile axial and hoop stress of ∼150 MPa at the inner surface of the SiC composite tube. This stress level is potentially beyond the matrix cracking stress. Degradation of hermeticity of the SiC composite tubes was observed after irradiation, indicating irradiation-induced cracking, whereas the irradiated monolithic SiC tubes remained hermetic. The results indicate that loss of hermeticity caused by radiation-induced microcracking is a potential issue for SiC composite cladding, depending on the magnitude of the temperature gradients. Coating the outer surface of the cladding was identified as a mitigation strategy that might overcome the cracking issue. [ABSTRACT FROM AUTHOR]

• The impact of several ATF materials on 5 × 5 rod bundle is investigated. • MCT and MFCT are reduced by FeCrAl, superior to HNLS/ML-A and SA3/PyC150-A. • MFCT is significantly reduced by UO 2 + BeO/SiC fuels. • Both CSM and FPSM are significantly improved with utilization of ATF fuels and claddings. Accident tolerant fuel (ATF) materials need to be studied by subchannel analysis under a variety of transient conditions before being applied to Light-Water-Reactors (LWRs). The present study assesses the thermal behaviors of various ATF claddings and fuels under four different transient conditions which will give rise to the occurrence of Departure from Nucleate Boiling (DNB) phenomenon. A 5 × 5 rod bundle model is built to conduct the subchannel analysis, and maximum cladding temperature (MCT) and maximum fuel centerline temperature (MFCT) of different ATF cladding and fuel groups are obtained. The simulation results show that both MCT and MFCT are reduced with the utilization of FeCrAl cladding, superior to HNLS/ML-A and SA3/PyC150-A claddings that only reduce MCT but increase MFCT. Significant MFCT reduction but slight MCT increase are achieved in most transient cases with the application of UO 2 + BeO/SiC fuel pellets. Due to the different safety limit of various ATF materials, safety margin analysis is performed as well. The results show FeCrAl possibly is the best alternative cladding material among claddings studied, because both cladding safety margin (CSM) and fuel pellet safety margin (FPSM) are improved. Compared to single UO 2 fuel, a larger fuel pellet safety margin is still provided by UO 2 + BeO/SiC fuel pellets. [ABSTRACT FROM AUTHOR]

• The thermal hydraulic behaviors of several ATF materials under RIA transient are investigated. • PCT is reduced by AMPT cladding but raised by SA3/PyC50-A cladding. • PFCT can be significantly reduced by ATF fuels but PCT increases. • Lower PFCT and PCT can possibly be obtained for ATF systems with 10% increase of coolant channel lateral length. This paper describes use of sub-channel analysis code COBRA-EN to assess the thermal hydraulic performance and influence of different coolant channel geometries on several accident tolerant fuel (ATF) materials. The results show that FeCrAl-based alloy (AMPT) cladding could reduce peak cladding temperature (PCT) and peak fuel centerline temperature (PFCT), but irradiated SiC/SiC composite (SA3/PyC50-A) cladding raised them. Significant PFCT reduction could be achieved by using ATF fuels, but higher PCT with heat transfer deterioration. In order to analyze the potential impact of coolant channel geometry size, four geometry models were subjected to sensitivity analysis. The results show that previous coolant channel geometry size was the best design for UO 2. Heat transfer deterioration on cladding surface of ATF fuel/cladding systems could be sufficiently improved with 10% increase of coolant channel geometry size. So, 10% increase of coolant channel geometry size for ATF fuel/cladding systems possible is a better design. [ABSTRACT FROM AUTHOR]

Highlights • Reactions of many fission product (FP) species with SiC are energetically favorable. • Reactions are diffusion limited, thus not expected for normal SiC-CMC applications. • SiC quality and microstructure impacts interactions with FP species. • Complexity in the LWR fuel system must be considered to predict SiC-CMC performance. Abstract Silicon carbide (SiC) fiber reinforced SiC ceramic matrix composites (CMC) represent an alternative cladding material for accident tolerant fuel light water reactor (LWR) applications, however, data and technology gaps exist that need to be understood before SiC-CMCs can be considered. One technology gap is the understanding of fuel-clad chemical interactions (FCCI) in SiC-CMC clad LWR systems. A review of SiC interactions with relevant elements to the LWR fuel system is discussed. Insights are drawn from research focusing on the use of SiC for high temperature electronics applications and high temperature structural materials applications as well as insights from tristructural-isotropic (TRISO) coated particle fuel development, where SiC is a critical component of the TRISO design. The interactions discussed in this review are framed relative to the expected conditions for SiC-CMC cladding application. The general observation is that limited risk of FCCI exists under normal operating conditions for SiC-CMC cladding systems in LWR applications. [ABSTRACT FROM AUTHOR]

Highlights • Transmutation of minor actinides in potential accident tolerant fuel-cladding system U3Si2-FeCrAl(SiC) is suggested. • Critical loading concentrations of minor actinides are determined by constraining the cycle length. • Higher transmutation efficiencies are found in candidate accident tolerant fuels comparing with traditional fuel. • One U3Si2-SiC assembly can transmute 237Np produced from six normal assemblies while other properties are kept. Abstract An advanced transmutation method is suggested that the long-lived Minor Actinides (MAs) in the spent fuel can be efficiently transmuted in the candidate Accident Tolerant Fuel (ATF). The transmutation of MAs is investigated through the Monte Carlo simulations in two potential fuel-claddings of ATF, U 3 Si 2 -FeCrAl and U 3 Si 2 -SiC. The critical loadings of MAs are determined through the Linear Reactivity Model (LRM) in order to keep the same reactivity as the current UO 2 -zircaloy system at the End of Cycle (EOC). In all cases, excellent transmutation efficiencies are found for the most important three MAs, 237Np, 241Am, and 243Am, of which the total transmutation rates are around 60%, 90%, and 60%, respectively. If only the longest-lived isotope 237Np is considered, one U 3 Si 2 -SiC assembly can transmute 237Np from six normal assemblies. The loading of MAs has little influences on the neutronic properties, such as the power distributions inside an assembly and inside a fuel rod. The transmutation of MAs in the ATF assembly is shown to be more efficient and safe comparing with the normal assembly, while other important properties are kept, such as the cycle length and the power distribution. [ABSTRACT FROM AUTHOR]

As a part of the effort to develop accident-tolerant fuel (ATF) that would delay potential deleterious consequences of loss-of-coolant-accidents (LOCA), TiN and TiAlN coating application on nuclear fuel claddings were shown as a promising path to enhance corrosion resistance of nuclear fuel cladding. Recently, ytterbium doping was shown to contribute to improving TiAlN coating corrosion performance. However, overall cladding coating performance evaluation requires investigation of neutronic penalties and mechanical properties in addition to corrosion resistance. The current study is composed of two parts. First, the neutronic impact of adding monolithic TiN, monolithic TiAlN and 8-layer TiAlN/TiN coatings (with and without Yb dopants) to the exterior of conventional zirconium-alloy cladding was investigated. Second, an experimental investigation was performed through the deposition of select coating design architecture using cathodic arc physical vapor deposition process onto ZIRLO®1 sheets followed by mechanical testing to examine the adhesion and hardness of the coatings. Characterizations were performed using optical microscopy (OM), scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), electron probe microanalysis (EPMA), Raman spectroscopy and X-ray diffraction (XRD). It is concluded that to minimize the neutronic penalties, a limitation on the concentration of Yb-dopant is necessary. Mechanical testing showed that hardness decreased with increasing Yb content. The adhesion of the TiAlN coatings was not affected by doping but was affected by the cathode composition. [ABSTRACT FROM AUTHOR]

Abstract Silicon carbide fiber-reinforced silicon carbide matrix (SiC-SiC) composites are being considered as components in light water reactor cores to improve accident tolerance, including channel boxes and fuel cladding. In the nuclear reactor environment, core components like a channel box will be exposed to neutron and other radiation damage and temperature gradients. To ensure reliable and safe operation of a SiC-SiC channel box, it is important to assess its deformation behavior under in-reactor conditions including the expected neutron flux and temperature distributions. In particular, this work has evaluated the effect of non-uniform dimensional changes caused by spatially varying neutron flux and temperatures on the deformation behavior of the channel box over the course of one year. These analyses have been performed using the fuel performance modeling code BISON and the commercial finite element analysis code Abaqus, based on fast flux and temperature boundary conditions that have been calculated using the neutronics and thermal-hydraulics codes Serpent and CTF, respectively. The dependence of dimensions and thermophysical properties on fast flux and temperature has been incorporated into the material models. These initial results indicate significant bowing of the channel box with a lateral displacement greater than 6.5 mm. The channel box bowing behavior is time dependent and driven by the temperature dependence of the SiC irradiation-induced swelling and the neutron flux/fluence gradients. The bowing behavior gradually recovers during the course of the operating cycle as the swelling of the SiC-SiC material saturates. However, the bending relaxation due to temperature gradients does not fully recover and residual bending remains after the swelling saturates in the entire channel box. [ABSTRACT FROM AUTHOR]

Corrosion of zirconium alloy fuel cladding in nuclear reactors is one of the main limiting factors to component lifespan. Protective coatings may be applied to reduce the corrosion rate during regular operation and can have the added benefit of mitigating follow-on risks associated with loss-of-coolant accidents. This study considers the use of electrochemical testing to rapidly evaluate the corrosion protection of chromium coatings on zirconium alloys. By correlating these electrochemical testing results with the microstructural characteristics and mechanical properties of the coating, it is possible to qualitatively predict the performance of these coatings during long-term autoclave testing and eventual in-reactor use. A series of chromium coatings were deposited using cathodic arc physical vapor deposition (CA-PVD) with varying substrate bias. Electrochemical testing showed that the coatings provided improvements in corrosion protection over the uncoated baseline zirconium alloys, with further improvements observed for substrates to which voltage bias was applied. This is attributed to the increased adatom mobility during the deposition process, allowing for the development of a dense, uniform microstructure of the chromium coating. These results were compared to autoclave exposure data over a 56-day period to show the effectiveness of electrochemical testing towards predicting significant differences in corrosion resistance under regular operating conditions. This approach provided limited qualitative insight on long-term performance in its present state, but shows promise for further development. [ABSTRACT FROM AUTHOR]

The U.S. Department of Energy’s Accident Tolerant Fuel program is focused on extending the time for fuel failure during postulated severe accidents compared to the standard UO 2 -Zr alloy fuel system. This paper investigates the feasibility of two different chromium-coated cladding concepts, one of which is zirconium-alloy based and the other is composite-SiC based. Both claddings had 50 µm coatings, deducted from the base layer thicknesses. The claddings were studied, using the multi-physics fuel performance tool BISON, under steady-state PWR operating conditions as well as under two transients: a power ramp and a loss-of-coolant accident (LOCA). The chromium-coated claddings showed comparable thermo-mechanical performance to the reference Zircaloy-4 cladding. As chromium is reported to provide an order of magnitude improvement in oxidation resistance, it is expected to be a better alternative in accident scenarios. Simulation results for both concepts show that further experimental investigation as well as modeling of beyond design-basis accidents is warranted. [ABSTRACT FROM AUTHOR]

A mechanistic model that calculates Cr diffusion for Cr-coated Zircaloy cladding based on Finite Difference Method (FDM) has been developed and validated. The developed model has been incorporated into TRANOX 2.0. The Cr-coated cladding was divided into multilayers (Zr, ZrCr 2 and Cr), and chromium diffusion coefficients of each phase were obtained by experiments. The prediction of the model was validated with Cr distribution measured by EPMA through extensive high temperature annealing experiments conducted in the temperature range of 1000–1300 °C. The developed model can predict diffusional loss coating thickness of Cr-coated Accident Tolerant Fuel. [Display omitted] • A mechanistic model that calculates Cr diffusion for Cr-coated Zircaloy ATF cladding • Accurate prediction for phase thickness and Cr distribution under high temperature • Analyzing the diffusional loss coating thickness under high temperature • ZrCr 2 precipitation hardening the Zr matrix [ABSTRACT FROM AUTHOR]

• Magnetron sputtered Cr coatings on Zircaloy-4 provide increased resistance to cladding rupture during simulated LOCA. • Defects in Cr coatings can inhibit resistance to cladding rupture. • Crack-free Cr coatings on Zry-4 show similar rupture temperatures as predicted for uncoated Zry-4 with ∼78% thicker cladding walls. Cr-coated zirconium alloys represent a modern approach to enhance cladding safety during accident scenarios. Two high-power impulse magnetron sputtered Cr-coated Zry-4 systems were subjected to simulated loss-of-coolant accident conditions to investigate cladding performance. The first Cr-coating (4.8 µm thick) was deposited onto Zry-4 cladding and exhibited through-thickness cracking while the second Cr-coating (6.8 µm thick) was deposited with improved deposition parameters onto polished Zry-4 and exhibited no cracking. During burst testing, the coating with a higher density of defects failed to consistently reduce oxidation and exhibited similar burst behavior as Zry-4. In contrast, the second Cr-coating reduced ZrO 2 formation through formation of Cr 2 O 3 and displayed enhanced burst temperatures by ∼80 °C compared to Zry-4. Utilizing an empirical relation for burst behavior of zirconium alloys, the 6.8 µm Cr/Zry-4 system displayed enhanced burst temperatures equivalent to an effective 0.464 mm increase in Zry-4 wall thickness, highlighting the value of continuous Cr coatings for accident scenarios. [ABSTRACT FROM AUTHOR]

Iron-chromium-aluminum (FeCrAl) alloys are candidates to be used as nuclear fuel cladding for increased accident tolerance. An analysis of the response of FeCrAl under normal operating and loss of coolant conditions has been performed using fuel performance modeling. In particular, recent information on FeCrAl material properties and phenomena from separate effects tests has been implemented in the BISON fuel performance code and analyses of integral fuel rod behavior with FeCrAl cladding have been performed. BISON simulations included both light water reactor normal operation and loss-of-coolant accidental transients. In order to model fuel rod behavior during accidents, a cladding failure criterion is desirable. For FeCrAl alloys, a failure criterion is developed using recent burst experiments under loss of coolant like conditions. The added material models are utilized to perform comparative studies with Zircaloy-4 under normal operating conditions and oxidizing and non-oxidizing out-of-pile loss of coolant conditions. The results indicate that for all conditions studied, FeCrAl behaves similarly to Zircaloy-4 with the exception of improved oxidation performance. Further experiments are required to confirm these observations. [ABSTRACT FROM AUTHOR]

Five CrNb coatings containing different Nb contents were deposited on zircalloy substrate by magnetron sputtering technology. The effect of Nb content on microstructure, mechanical property, high-temperature corrosion and oxidation resistance of the coatings were investigated systematically. With increasing Nb content, the surface wetting characteristics of the coating changed from hydrophilic to hydrophobic, the wear performance and the nanohardness were enhanced. The Nb addition has little effect on the corrosion resistance of the coatings. The optimal mechanical and high temperature oxidation properties were obtained for the coating with 21at% Nb content. • With increasing Nb content, the surface wetting characteristics of the coating changed from hydrophilic to hydrophobic. • The wear performance and the nanohardness were enhanced with increasing of Nb content. • The formation of Nb 2 O 5 mitigate the out-diffusion of Cr and the inner-diffusion of O. • The optimal mechanical and high temperature oxidation properties were obtained for the coating with 21at% Nb content. [ABSTRACT FROM AUTHOR]

In severe accident conditions with loss of active cooling in the core, zirconium alloys, used as fuel cladding materials for current light water reactors (LWR), undergo a rapid oxidation by high temperature steam with consequent hydrogen generation. Novel fuel technologies, named accident tolerant fuels (ATF), seek to improve the endurance of severe accident conditions in LWRs by eliminating or at least mitigating such detrimental steam-cladding interaction. Most ATF concepts are expected to work within the design framework of current and future light water reactors, and for that reason they must match or exceed the performance of conventional fuel in normal conditions. This study analyzed the neutronic performance of ATF when employed in both pressurized and boiling water reactors. Two concepts were evaluated: (1) coating the exterior of zirconium-alloy cladding with thin ceramics to limit the zirconium available for reaction with high-temperature steam; (2) replacing zirconium alloys with alternative materials possessing slower oxidation kinetics and reduced hydrogen production. Findings show that ceramic coatings should remain 10–30 μm thick to limit the neutronic penalty. Alternative cladding materials, with the exception of SiC, enhance neutron loss compared to zirconium-alloys. An extensive parametric analysis concluded that reference performance metrics can be met by employing 300-μm or less thick cladding or increasing fuel enrichment by up to 1.74% depending on material and geometry. [ABSTRACT FROM AUTHOR]

Among the challenges posed by accident-tolerant fuels (ATFs) is the demonstration of their safe behaviour under normal and postulated accident conditions. In order to accurately predict the response of novel materials in such scenarios, the development of modelling capabilities is paramount. This paper provides an integral characterization of FeCrAl cladding under steady-state conditions up to high burnup, highlighting the major differences with respect to Zircaloy-4 and the key properties responsible for their distinct behaviour. To that end, the fuel performance code FRAPCON-4.0 has been extended to FeCrAl cladding simulation. Such extension has been carried out according to a critical review of the thermo-mechanical data and correlations reported in the open literature for several FeCrAl alloys. The main outcome of this study is that there are no major performance differences between FeCrAl and Zircaloy-4 at end-of-life, except for their stress states, which differ due to different creep deformation rates. Therefore, modelling the thermo-mechanical response of FeCrAl cladding would benefit from an accurate description of its creep behaviour, not only in the case of foreseeable conditions during reactor operation but also for the potential implications that it may have under transient scenarios. • The steady-state thermo-mechanical performance of FeCrAl cladding has been assessed. • A critical review of thermo-mechanical data on FeCrAl alloys has been carried out. • The fuel performance code FRAPCON-4.0 has been extended to FeCrAl simulation. • Creep is the main property affecting the integral behaviour of FeCrAl cladding. [ABSTRACT FROM AUTHOR]

Silicon carbide (SiC) is being investigated for accident tolerant fuel cladding applications due to its high temperature strength, exceptional stability under irradiation, and reduced oxidation compared to Zircaloy under accident conditions. An engineered cladding design combining monolithic SiC and SiC–SiC composite layers could offer a tough, hermetic structure to provide improved performance and safety, with a failure rate comparable to current Zircaloy cladding. Modeling and design efforts require a thorough understanding of the properties and structure of SiC-based cladding. Furthermore, both fabrication and characterization of long, thin-walled SiC–SiC tubes to meet application requirements are challenging. In this work, mechanical and thermal properties of unirradiated, as-fabricated SiC-based cladding structures were measured, and permeability and dimensional control were assessed. In order to account for the tubular geometry of the cladding designs, development and modification of several characterization methods were required. [ABSTRACT FROM AUTHOR]

Silicon carbide (SiC) fiber, SiC matrix composites (SiC/SiC) are being considered as a cladding material for light water reactors in order to improve safety performance. Engineered, multi-layer cladding designs consisting of both monolithic SiC (mSiC) and SiC/SiC have been examined as promising concepts to meet both strength and impermeability requirements. A new model has been developed to calculate stresses and failure probabilities for multi-layer cladding consisting of SiC-based materials in reactor operating conditions. The results show that stresses in SiC-based cladding are dominated by temperature-dependent irradiation-induced swelling, with the largest stresses occurring during the cold shutdown conditions. Failure probabilities are driven by the resulting tensile stresses at the cladding inner wall, while the outer wall is subject to compressive stresses. This indicates that the inner SiC/SiC, outer mSiC concept has the lowest failure probability, as the pseudo-plastic deformation of the composite reduces tensile loading and the compressed monolith provides a reliable, impermeable barrier to fission product release. [ABSTRACT FROM AUTHOR]

A study analyzed the neutronics of alternate cladding materials in a pressurized water reactor (PWR) environment. Austenitic type 310 (310SS) and 304 stainless steels, ferritic Fe-20Cr-5Al (FeCrAl) and APMT™ alloys, and silicon carbide (SiC)-based materials were considered and compared with Zircaloy-4. SCALE 6.1 was used to analyze the associated neutronics penalty/advantage, changes in reactivity coefficients, and spectral variations once a transition in the cladding was made. In the cases examined, materials containing higher absorbing isotopes invoked a reduction in reactivity due to an increase in neutron absorption in the cladding. Higher absorbing materials produced a harder neutron spectrum in the fuel pellet, leading to a slight increase in plutonium production. A parametric study determined the geometric conditions required to match cycle length requirements for each alternate cladding material in a PWR. A method for estimating the end of cycle reactivity was implemented to compare each model to that of standard Zircaloy-4 cladding. By using a thinner cladding of 350 μm and keeping a constant outer diameter, austenitic stainless steels require an increase of no more than 0.5 wt% enriched 235 U to match fuel cycle requirements, while the required increase for FeCrAl was about 0.1%. When modeling SiC (with slightly lower thermal absorption properties than that of Zircaloy), a standard cladding thickness could be implemented with marginally less enriched uranium (∼0.1%). Moderator temperature and void coefficients were calculated throughout the depletion cycle. Nearly identical reactivity responses were found when coolant temperature and void properties were perturbed for each cladding material. By splitting the pellet into 10 equal areal sections, relative fission power as a function of radius was found to be similar for each cladding material. FeCrAl and 310SS cladding have a slightly higher fission power near the pellet’s periphery due to the harder neutron spectrum in the system, causing more 239 Pu breeding. An economic assessment calculated the change in fuel pellet production costs for use of each cladding. Implementing FeCrAl alloys would increase fuel pellet production costs about 15% because of increased 235 U enrichment and the additional UO 2 pellet volume enabled by using thinner cladding. [ABSTRACT FROM AUTHOR]

The corrosion behaviors of Mo 0.5 NbTiVCr 0.25 and Mo 0.5 NbTiV 0.5 Zr 0.25 multi principal element alloys were investigated in borated and lithiated water at 360 ℃. The two multi principal element alloys showed superior corrosion resistance than the M5 zirconium alloy due to the thinner inner oxide layer of the former, which may be related to composition of the inner oxides. The inner oxides primarily consisted of Ti and Nb elements, while the outer oxides primarily consisted of Ti, Nb and V elements. The corrosion behavior was dominated by the oxidation of Ti and Nb, and the corrosion process was explored. • Mo 0.5 NbTiVCr 0.25 and Mo 0.5 NbTiV 0.5 Zr 0.25 HEAs are designed to improve the accident tolerance of nuclear fuel cladding. • The two HEAs show double-layer oxide scale in borated and lithiated high temperature water. • The two HEAs show superior corrosion resistance than the M5 zirconium alloy due to the more protective oxide scale. • The corrosion behavior was dominated by the oxidation of Ti and Nb, and the corrosion process was explored. [ABSTRACT FROM AUTHOR]

• Coated ZrN sphere-UO 2 as surrogates for UN-UO 2 accident tolerant fuel composites. • Spheres coated with Mo or W nanopowders by developed processes and W by CVD. • Use of spark plasma sintering to fabricate the composites at 1773 K and 80 MPa. • Both coating methods provide dense layers of Mo and W. • Mo and W layers can protect the nitride fuel against interactions with UO 2. Uranium nitride (UN) spheres embedded in uranium dioxide (UO 2) matrix is considered an innovative accident tolerant fuel (ATF). However, the interaction between UN and UO 2 restricts the applicability of such composite in light water reactors. A possibility to limit this interaction is to separate the two materials with a diffusion barrier that has a high melting point, high thermal conductivity, and reasonably low neutron cross-section. Recent density functional theory calculations and experimental results on interface interactions in UN-X-UO 2 systems (X = V, Nb, Ta, Cr, Mo, W) concluded that Mo and W are promising coating candidates. In this work, we develop and study different methods of coating ZrN spheres, used as a surrogate material for UN spheres: first, using Mo or W nanopowders (wet and binder); and second, using chemical vapour deposition (CVD) of W. ZrN-UO 2 composites containing 15 wt% of coated ZrN spheres were consolidated by spark plasma sintering (1773 K, 80 MPa) and characterised by SEM/FIB-EDS and EBSD. The results show dense Mo and W layers without interaction with UO 2. Wet and binder Mo methods provided coating layers of about 20 µm and 65 µm, respectively, while the binder and CVD of W methods layers of about 12 µm and 3 µm, respectively. [ABSTRACT FROM AUTHOR]

The Cr coatings (4.5–9.0 µm-thick) with porous/columnar and dense/columnar microstructure were deposited on Zr alloy. Oxidation was performed at 900, 1050, 1200 and 1400 °C in steam to define the role of coating microstructure and thickness on oxidation behavior of Cr-coated Zr alloy. The activation energy for chromium oxidation decreases from 249 to 124 kJ/mol when the coating microstructure changes from dense to porous. Both coatings types are consumed by Cr-Zr interdiffusion resulting in the similar thickness of residual Cr. The Cr-Zr interdiffusion can prevent forming the Cr oxide film and cause direct oxidation of the alloy at 1400 °C. • Effect of coating thickness on oxidation resistance of Cr-coated Zr alloy. • Higher oxidation resistance of Zr alloy with dense Сr coatings. • Сonsuming Cr coatings by oxidation in steam and Cr-Zr interdiffusion. • Stabilization of α-Zr (O) phase slows down Cr-Zr interdiffusion. [ABSTRACT FROM AUTHOR]

The microstructure and nanomechanical properties of pure Cr coatings deposited on SiC with various physical vapor deposition (PVD) methods have been investigated for the mitigation of hydrothermal corrosion of SiC-SiC f (SiC fiber-reinforced SiC matrix) composite fuel cladding in light water reactors (LWR). Cr coatings (4–7 μm thick) were deposited on unbiased SiC substrates using six variants of magnetron sputtering processes: (i) standard (S-DCMS), (ii) pulsed (P-DCMS), (iii) ion-assisted (I-DCMS), and (iv) pulsed ion-assisted direct current magnetron sputtering (PI-DCMS), (v) high-power impulse magnetron sputtering (HiPIMS), and (vi) bipolar HiPIMS (B-HiPIMS). Microstructural characterization and nanoindentation testing were used to evaluate effects of deposition technique on coating microstructure and nanomechanical properties. The application of positive reverse pulses to the sputter target induced coarsening of the columnar grain structure, while ion bombardment of substrates during deposition promoted densification of the coating microstructure. Both effects are exemplified in B-HiPIMS deposition, resulting in a high-density microstructure with compressive residual stress. Material pile-up around nano-indents and plastic work during nanoindentation were analyzed to understand deformation behavior of the coatings. The results suggest the B-HiPIMS process to be among the promising methods for the deposition of Cr coatings on SiC-SiC f for LWR cladding application. [Display omitted] • Cr coatings 4–7 μm thick were deposited on SiC using magnetron sputtering methods. • Simultaneous pulsing and ion bombardment resulted in densification of DCMS coating. • Bipolar HiPIMS produced dense coating with larger grains, compressive stress. • Pile-up formation and extensive deformation of grains reveals plasticity of bipolar HiPIMS coating. [ABSTRACT FROM AUTHOR]

The oxidation and interdiffusion behavior of Cr- and Cr/Mo-coated Zr 1Nb zirconium alloy were investigated. A single-layer Cr (8 um) and multilayer Cr (8 μm)/Mo (3 μm) coatings were deposited by magnetron sputtering. The coated Zr 1Nb alloy samples were oxidized in air at 1100 °C for 15–60 min. Both coating types had a protective scale during high-temperature (HT) oxidation. The use of Mo sublayer resulted in preventing Cr Zr interdiffusion under high temperature. Scanning electron microscopy, optical microscopy, X-ray diffraction before and after HT oxidation and in situ X-ray diffraction up to 1250 °C were used to identify interdiffusion behavior of the system of Cr/Mo/Zr. Some aspects of applying of Cr/Mo coatings for Zr nuclear fuel claddings are discussed. • Mo sublayer limits Cr Zr interdiffusion • Higher thickness of residual Cr in Cr/Mo coatings compared to single-layer Cr • Mo diffusion at interfaces results Cr 3 Mo and Mo 2 Zr formation • Mo Zr interdiffusion layer increases with oxidation time [ABSTRACT FROM AUTHOR]

The challenges and opportunities to alloyed and composite fuel architectures designed and intended to mitigate oxidation of the fuel during a cladding breach of a water-cooled reactor are discussed in this manuscript focused on the oxidation performance of uranium diboride and uranium monocarbide. Several high uranium density fuels are under consideration for deployment as accident tolerant and/or advanced technology nuclear reactor fuels, including UN, U 3 Si 2 , UB 2 , and UC. Presented here is the literature for UB 2 and UC degradation modes, thermodynamics, and oxidation performance of the pure compounds and reported alloyed and composite architectures. Furthermore, this review covers the materials and techniques for the incorporation of additives, dopants, or composite fuel architectures to improve the oxidation behavior for high uranium density fuels for use in LWRs. [ABSTRACT FROM AUTHOR]

• NITE SiC/SiC claddings were neutron-irradiated using an experimental fission reactor. • SiC/SiC claddings were irradiated in pressurized flowing water in the reactor. • Silicon dissolution was measured under various reactor operating conditions. • Post-irradiation cladding morphology was observed to evaluate various technologies. Claddings of silicon carbide (SiC)-based composites for light water reactors are expected to be a key component of accident-tolerant fuel technology for enhancing nuclear safety. Muroran Institute of Technology has developed technologies for producing SiC/SiC claddings by a nano-infiltration and transient eutectic phase (NITE) method in the MEXT-funded project named "SCARLET." As part of this program, neutron irradiation experiments were performed in a pressurized cooling water flow at the Halden Reactor in Norway. During the irradiation experiments, silicon dissolution from the SiC/SiC specimens was measured, and the relationship among the dissolution, properties of SiC/SiC claddings and reactor operating conditions was studied. The morphology of the SiC/SiC claddings was not significantly modified by the irradiation experiments; however, the joint parts were corroded and suffered damage. Observation of irradiated specimens and a rough estimate of the SiC dissolution rate under each irradiation condition indicated that the NITE SiC/SiC composite has potential as a structural material for pressurized water reactors (PWRs) that are operated under hydrogen water chemistry conditions. It is also shown that chemical-vapor-deposited (CVD)-SiC and other coating techniques are necessary for the realization of SiC reactor core. Joining and coating technologies were also tested, but appropriate methods could not be identified experimentally; thus, joining and coating techniques should be further developed. [ABSTRACT FROM AUTHOR]

• Preliminary neutronics analysis for alternative fuel cladding materials. • Effectiveness of utilizing different ATF materials on the thermal neutron flux. • Pu-239 inventory for various cladding materials during fuel cycle. This paper presents a preliminary neutronics analysis for Accident Tolerant Fuel (ATF) cladding materials for standard APR-1400 reactor A0 fuel assembly. AMPT, FeCrAl, 304 and 310 stainless steels, and SiC were considered and compared with the original Zircaloy-4 cladding material. A parametric evaluation was done for fuel and cladding materials to confirm the geometry requirements to achieve the end-of-cycle fuel reactivity, and the results were compared with the standard APR-1400 reference fuel-cladding system. Serpent Monte Carlo reactor physics code version 2.31 was utilized to evaluate the associated neutronics penalty when different cladding materials were used such as, fuel reactivity, thermal neutron spectrum, plutonium and isotopic inventory evolution, and fuel assembly linear pin power distribution. In the examined cases, fuel enrichments were fixed, and cladding thickness and pellet diameter were changed for zircaloy, cases 1 – 4. As a result, thermal neutron absorption cross section varies between materials due to higher absorbing cross section leads to hardening of the neutron spectrum in the fuel-cladding system. That leads to a slight increase in the production of actinides in the inventory such as plutonium. [ABSTRACT FROM AUTHOR]

To reduce fission gas release under both accident conditions and normal operation, Korea Atomic Energy Research Institute has developed a metallic micro-cell pellet following the 2011 Fukushima accident. One concern, however, about the metallic micro-cell pellet is that due to the presence of metallic wall that reduces both temperature and gas release, its capability of higher gas retention could result in increased gaseous swelling, leading to earlier pellet-to-cladding gap closure than the conventional UO 2 pellet. To analyze and compare the gap closure in two fuel rods in which two kinds of pellets, UO 2 and metallic micro-cell, are loaded into each of the two identical Zr claddings, we developed a model to estimate gaseous swelling in the two pellets. Then considering mainly different gaseous swelling and thermal expansion by different temperature profiles in the two pellets, we compared the pellet-to-cladding gaps as a function of burnup for the case where they would be irradiated with the same constant linear power of 30 kW/m up to 60 MWd/kgU. Due to both reduced gaseous swelling and lower thermal expansion, gap closure in the fuel rod loaded with the metallic micro-cell pellet is expected to be delayed by 6–16 MWd/kgU compared with the one loaded with the UO 2 pellet, implying pellet-cladding mechanical interaction occurrence would also be postponed by the same degree. We note that the result of this approach is preliminary until confirmed by measured data from metallic micro-cell pellets irradiated up to high burnup. • Higher gas-retention capability in the metallic micro-cell pellet could lead to earlier pellet-to-cladding gap closure than the UO 2 pellet. • Despite the higher gas-retention capability, gaseous swelling in the metallic micro-cell pellet would be reduced by its lower temperature. • Pellet-to-cladding gap closure in the metallic micro-cell pellet could be delayed by 6–16 MWd/kgU compared with the UO 2 pellet. [ABSTRACT FROM AUTHOR]

Iron-Chromium-Aluminum (FeCrAl) alloys are candidate materials for light water reactor accident tolerant fuel (ATF) rod cladding because of the unparalleled resistance of these alloys to attack by superheated steam in the case of a loss of coolant accident. Since FeCrAl alloys were never used before in a reactor environment, it was important to characterize their corrosion resistance under simulated reactor normal operation conditions. Tests were conducted for APMT and C26M alloys in three ~300°C water systems (two containing hydrogen and one containing oxygen). Results show that after 12 months of immersion the FeCrAl alloys lose a small amount of mass. The resistance to corrosion is mostly given by the development on the surface of a protective oxide mostly rich in Cr and Al. The mass loss was lower in APMT than in C26M because of the higher Cr content in the former. After 12 months residence in the ~300°C water, the FeCrAl specimens maintained their unequaled resistance to attack by 1200°C steam. [ABSTRACT FROM AUTHOR]

• The potential benefits of ATF Cr-coated-Zr were investigated in a BWR by MELCOR. • 240 calculations were performed for the Zr and ATF respectively to assess uncertainties. • The effect of ATF on hydrogen generation is substantial. • A sensitivity analysis was performed to rank the significance of material uncertainties. • The coating failure temperature is dominant for ATF material degradation modeling. The deposition of protective coatings on nuclear fuel cladding has been considered as a near-term Accident Tolerant Fuel (ATF) concept that will reduce the high-temperature oxidation rate and enhance accident tolerance of the cladding while providing additional benefits during normal operation and transients. In this study, an uncertainty analysis was employed to investigate the potential benefits of ATF Cr-coated-Zr cladding and canister for an unmitigated Short-Term Station Blackout (STSBO) sequence in a generic BWR plant using the MELCOR systems code. The MELCOR parameters that reflect the current state-of-knowledge of the relevant fuel assembly performance during core degradation were selected and characterized according to their ranges and distributions. An extensive set of simulations (240 MELCOR calculations) were performed for the Zr and Cr-coated-Zr cladding and canister materials, respectively, to determine their effect on core degradation with the associated uncertainties. The comparison between the Zr and Cr-coated-Zr calculations confirms that the use of ATF Cr-coated-Zr as cladding and canister component material in BWR might be an effective way to mitigate the accident progression and reduce the total hydrogen generation during the accident. The core degradation process was only delayed by less than a half hour, providing some additional time for compensatory actions to mitigate with the accident progression. In contrast, the effect of coated materials on total hydrogen generation was more substantial; i.e., hydrogen generation was almost reduced by half. In addition, a sensitivity analysis based on the Pearson and Spearman correlation coefficients was conducted to rank the significance of the considered parameter uncertainties. The Cr-coating failure temperature was identified as the dominant factor in the MELCOR simulations of core degradation and associated hydrogen generation. Understanding these effects will inform and guide researchers to focus on a more productive area of research and development for accident-tolerant fuel concepts and enhancement of core safety margins. [ABSTRACT FROM AUTHOR]

• The irradiation defects were significantly restricted by columnar grain boundaries. • Higher density of grain boundary contributed to reduce dislocations and voids. • CrN coating showed the best irradiation resistance among samples. Heavy ion (6 MeV Au+) irradiation studies were carried out on three candidate coatings (Cr, CrN, TiAlCrN) and Zircaloy-4 for accident tolerant fuel cladding to investigate the irradiation resistance and microstructure evolution at 400 °C. Transmission Electron Microscopy (TEM) results showed that the irradiation defects were significantly restricted by grain boundaries in the columnar grains compared to equiaxed grains and the high-density boundaries could obviously reduce the formation of voids. Fast Fourier Transformation (FFT) results of high-resolution TEM images showed that Cr and CrN coatings maintained better lattice integrity than those of TiAlCrN coating and Zircaloy-4. ATF coatings showed less irradiation defects than that of Zircaloy-4 substrate. Cr coating exhibited better irradiation resistance than that of TiAlCrN coating in terms of irradiation dislocation loops. CrN coating showed the best irradiation resistance among all samples, due to its compact crystal structure and higher density of grain boundary, which contributed to reduce dislocations and voids. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Zirconium-based alloys have served the nuclear industry for several decades due to their acceptable properties for nuclear cores of light water reactors (LWRs). However, severe accidents in LWRs have directed research and development of accident tolerant fuel (ATF) concepts that aim to improve nuclear fuel safety during normal operation, operational transients and possible accident scenarios. This review introduces the latest results in the development of protective coatings for ATF claddings based on Zr alloys, involving their behavior under normal and accident conditions in LWRs. Great attention has been paid to the protection and oxidation mechanisms of coated claddings, as well as to the mutual interdiffusion between coatings and zirconium alloys. An overview of recent developments in barrier coatings is introduced, and possible barrier layers and structure designs for suppressing mutual diffusion are proposed. [ABSTRACT FROM AUTHOR]

Silicon carbide (SiC) is widely regarded as a potential accident tolerant fuel cladding material for the next generation nuclear power system. Oxidation is the leading cause of failure of SiC in a high temperature environment. Adsorption of oxygen on SiC surface is the first stage of SiC oxidation. Effects of irradiation defects, such as vacancy and antisite atom, on the adsorption of oxygen on 3C-SiC low index surfaces are investigated by first principles calculation method. The results show that the oxygen molecule tends to dissociate spontaneously into two O atoms when adsorbed on the low index surfaces. The adsorption stability of oxygen molecule on the Si-terminated surface is higher than that on the C-terminated surface. As a whole, irradiation defects decrease the stability of oxygen adsorption on C-(100) and Si-(100) surfaces but increase the stability of oxygen adsorption on (110), (111), and ( 1 ¯ 1 ¯ 1 ¯ ) surfaces. Except for (110) surface, new covalent bonds as the precursors of oxides are formed on the SiC surfaces due to the irradiation defects, which changed the oxide species of other surfaces during the initial oxidation. Bader charge analysis and the local density of state results show that charge transfer plays a crucial role in the adsorption process. [ABSTRACT FROM AUTHOR]

• The station blackout accident of a dual-cooled annular fuel VVER-1000 is investigated. • The portable pumps is deployed to mitigate the accident. • The effect of accident tolerant fuel cladding is considered. • The RELAP5 model is used to model the annular fuel to perform the analysis. The station blackout accident is clearly a very serious problem that jeopardizes the safety barrier of nuclear reactor and ascends the core integrity steeply. The dual-cooled annular fuel has been introduced to boost the generated power density and safety margins. The fuel is cooled both internally and externally by allowing the liquid to flow on inner and outer heated surfaces. The fuel peak temperature can be reduced substantially by decreasing the fuel rod cladding heat flux leading to an increment on the DNBR. In this paper, a dual cooled annular fuel a VVER-1000 is investigated in this accident scenarios. The deployment of portable pumps for injecting coolant to dual cooled annular fuel core to mitigate this accident is also considered. The effect of accident tolerant fuel rod cladding is also investigated. The RELAP5 model is used to model the annular fuel to perform the analysis. The results are compared to traditional solid fuel geometry. [ABSTRACT FROM AUTHOR]

• The ATF concept Cr-coated-Zr was assessed with a BWR MELCOR plant model. • The Cr-coated-Zr was applied to both cladding and canister. • The effect of Cr-Zr eutectic reaction was considered. • The use of Cr-coated-Zircaloy mitigates the core degradation process. • The gross in-vessel hydrogen generation decreased by almost a factor of three. The deposition of protective coatings on nuclear fuel cladding has been considered as a near-term Accident Tolerant Fuel (ATF) concept that will reduce the high-temperature oxidation rate and enhance accident tolerance of the cladding while providing additional benefits during normal and transient. In this study, the performance of the proposed ATF concept of Cr-coated-Zircaloy is assessed using a generic Boiling Water Reactor MELCOR plant model considering a Short-term Station Blackout (STSBO) scenario. Simulation results indicate that the use of Cr-coated-Zircaloy as cladding and canister material mitigates the core degradation process as compared to the traditional Zircaloy cladding and canister design. The onset of fuel degradation and collapse is delayed by over thirty minutes, and the extent of fuel degradation is reduced. Specifically, the gross in-vessel hydrogen generation decreased by almost a factor of three. Although the eutectic reaction between Cr-coating and Zircaloy could cause an early failure of the coating, the improvement in the delay of fuel degradation is still notable. Additionally, a thicker coating is found helpful to obtain additional coping time and to decrease hydrogen generation. In addition to the eutectic formation that may compromise Cr-coated Zr, a different failure mode is identified for the Cr-coated-Zr when compared to Zircaloy; i.e., a complete melt of base material leads to component collapse before the coating is oxidized and consumed. These findings can help the industry focus on productive areas of research and development for accident-tolerant fuel concepts and enhancement of core safety margins. [ABSTRACT FROM AUTHOR]

• Oxidation resistance of UN/U 2 N 3 -UO 2 with initial 10 wt% microspheres is better than UO 2. • Oxidation onset temperatures are equivalents when using 30 wt% or 50 wt% microspheres. • UO 2 acts as a protective barrier for the nitride phases. • TGA triplicate measurements confirm good repeatability and reliability of the results. • Oxidation mechanism for UN/U 2 N 3 -UO 2 composites is proposed. Composite fuels such as UN-UO 2 are being considered to address the lower oxidation resistance of the UN fuel from a safety perspective for use in light water reactors, whilst improving the in-reactor behaviour of the more ubiquitous UO 2 fuel. An innovative UN-UO 2 accident tolerant fuel has recently been fabricated and studied: UN microspheres embedded in UO 2 matrix. In the present study, detailed oxidative thermogravimetric investigations (TGA/DSC) of high-density UN/U 2 N 3 -UO 2 composite fuels (91-97 %TD), as well as post oxidised microstructures obtained by SEM, are reported and analysed. Triplicate TGA measurements of each specimen were carried out at 5 K/min up to 973 K in a synthetic air atmosphere to assess their oxidation kinetics. The mass variation due to the oxidation reactions (%), the oxidation onset temperatures (OOTs), and the maximum reaction temperatures (MRTs) are also presented and discussed. The results show that all composites have similar post oxidised microstructures with mostly intergranular cracking and spalling. The oxidation resistance of the pellet with initially 10 wt% of UN microspheres is surprisingly better than the UO 2 reference. Moreover, there is no significant difference in the OOT (~557 K) and MRT (~615 K) when 30 wt% or 50 wt% of embedded UN microspheres are used. Therefore, the findings in this article demonstrate that the UO 2 matrix acts as a barrier to improve the oxidation resistance of the nitride phases at the beginning of life conditions. [ABSTRACT FROM AUTHOR]

SiC-based materials are currently being considered for accident tolerant fuel (ATF) cladding, however, there are concerns regarding hydrothermal corrosion rates and hermeticity. These issues may be addressed with the utilization of a Cr corrosion barrier coating (CBC) but the interaction of Cr with SiC under loss of coolant accident (LOCA) conditions needs to be investigated. In the present work, Cr coatings were deposited onto chemically vapor deposited (CVD) SiC with either high power impulse magnetron sputtering (HiPIMS), cathodic arc (CA), or a combination of both. Annealing and steam thermogravimetric analysis (TGA) was used to study reaction product formation with and without oxidation. After 1200 °C annealing, formation of a Cr x C y carbide layer at the ambient interface and a Cr x Si y C z silicide layer at the SiC interface was observed. After 1200 °C TGA exposure, similar carbide and silicide layers were observed but with a Cr 2 O 3 outer reaction layer and a Si and O rich interfacial layer forming between the carbide and silicide layers. Initial mass gain of the coatings during the 1200 °C TGA exposure was parabolic and were similar to reported rates of chromia formation. Mass gain behavior and observed transitions in chromia microstructure indicate that metallic Cr may be consumed within the first ~1 h at 1200 °C and that subsequent chromia formation occurs at the expense of the underlying Cr carbide layer. Overall, the results show no detrimental impact of Cr coatings on the steam oxidation resistance of SiC up to 4 h at 1200 °C. [ABSTRACT FROM AUTHOR]

• The phase changes of U 3 Si 2 during steam oxidation are studied in detail. • U 3 Si 2 dissociates into UO 2 and Si, and Si is partially oxidized to amorphous SiO 2. • The discrepancy between theoretical and experimental results is explained and oxidation mechanism is proposed. • U 3 Si 2 undergoes enormous structural changes and volume expansion upon oxidation. • A new fuel design is needed to enable use of U 3 Si 2 in light water reactors. U 3 Si 2 is being considered as a potential fuel candidate for water cooled power reactors. Herein, we examine the oxidation behavior and phase separation of U 3 Si 2 under flowing steam to understand the fuel degradation mechanism in the event of a steam ingress through cracks in cladding. The weight gain, microstructure evolution, and phase change after isothermal annealing were investigated for spherical U 3 Si 2 particles fabricated by centrifugal atomization. Equilibrium oxide products and their compositional changes with increasing degree of oxidation were predicted using thermodynamic calculations and compared with experimental results. It was seen that the oxidation reaction of U 3 Si 2 in steam begins with dissociation into UO 2 and Si, followed by Si being partially oxidized to amorphous SiO 2. From these results, an oxidation mechanism of U 3 Si 2 in steam is proposed. The theoretical volume expansion of U 3 Si 2 during the steam oxidation was estimated and its impact on fuel performance is also discussed. [ABSTRACT FROM AUTHOR]

Accident-tolerant fuel (ATF) cladding materials aim to improve fuel reliability and safety during accident scenarios in water-cooled reactors. These ATF cladding should also perform comparable or better than the current zirconium alloy cladding under reactor normal operating conditions. Surface-modified Zircaloy-4 was produced by depositing a protective coating of chromium by two different coating techniques, Physical Vapor Deposition (PVD) and Cold Spray (CS). The coating thickness, structure, composition, and mechanical properties of the coated Zircaloy were assessed with advanced materials characterization techniques. Results showed a strengthening effect of the Cr-coating, along with ∼83% higher hardness compared to that of the Zircaloy-4 substrate. Surface topography and surface chemistry of the Cr-coated Zircaloy-4 were assessed with contact profilometry, atomic force microscopy, contact angle goniometry, and X-ray photoelectron spectroscopy. Deposited Cr-coating significantly improved surface wettability compared to the substrate Zircaloy-4. The formation of surface oxides on the coatings by prolonged exposure to ambient air showed a further increase in the hydrophobic behavior. Surface parameters and surface chemistry after flow boiling critical heat flux tests, conducted at atmospheric pressure, evidenced a slight increase in surface roughness with notable changes in wettability associated with the formation of surface oxides. [ABSTRACT FROM AUTHOR]

As a candidate accident tolerant fuel cladding material, the high temperature oxidation behaviour of an Oxygen Dispersion Strengthened (ODS) Fe–12Cr–6Al alloy has been examined in isothermal air and steam environments. While the oxidation behaviour of FeCrAl has been extensively studied, very little has been reported in the literature on FeCrAl-ODS, which is the focus of this manuscript. By conducting Thermogravimetric Analysis (TGA), oxidation experiments were performed on specimens of Fe–12Cr–6Al-ODS at temperatures up to 1400 °C in air and up to 1215 °C in high purity steam. Oxidation kinetics were evaluated by calculating reaction rate constants and activation energies. The formation of an alumina scale was correlated to kinetics results from TGA by coupling Scanning Electron Microscopy (SEM) with Energy Dispersive X-ray Spectroscopy (EDS) line scans to investigate the chemical composition of the primary oxide layer, local oxidation effects, as well as the bulk alloy material. Electron Probe Micro-Analysis (EPMA) point scans were applied to the heavily oxidized specimen to achieve an accurate composition profiling of areas of interest. Finally, thermodynamic calculations were performed to interpret the formation of the complex oxide layer. These experiments support research efforts for the development of accident tolerant fuel technologies at Nippon Nuclear Fuel Development Co., Ltd. [ABSTRACT FROM AUTHOR]

• A detailed plant model is built in TRACE for a representative U.S. operating PWR. • Two Accident Tolerant Fuel claddings are evaluated during three PWR SBO scenarios. • Natural circulation in hotleg and steam generator impacts ATF clad relative performance. • TRACE only estimated marginal increase of the coping times with both ATF claddings. Following the Fukushima Daiichi nuclear accident in 2011, researches on Accident-tolerant fuels (ATFs) are currently of high interest in not only the nuclear industry but also governmental and international organizations. In this work, a quantitative evaluation of the performance of monolithic FeCrAl cladding and Cr-coated Zircaloy cladding has been performed for Pressurized Water Reactor (PWR) Station Blackout (SBO) accidents. A generic PWR model has been built in system thermal-hydraulics code TRACE based on the Surry Nuclear Power Station with counter-current natural circulation modelling capability for hotleg and steam generator U-tube components during the accidents. The base model results are then compared to MELCOR and RELAP simulations to verify the system component implementation in TRACE. Two PWR SBO scenarios were investigated, including: short-term SBO and long-term SBO with early reactor coolant pump (RCP) seal failure. These scenarios are defined to be very similar to the accidents studied in the State-of-the-Art Reactor Consequence Analysis (SOARCA) project. TRACE code is modified to reflect the oxidation kinetics of FeCrAl and Cr-coating. Larson-Miller creep rupture model is also implemented in TARCE using its built-in control systems to simulate the creep rupture of hotlegs. The comparison between the TRACE models with and without the counter-current flow modeling resulted in significant difference when comparing ATF cladding to Zircaloy for short term SBO, while it marginal impacted the performance during long term SBO with RCP seal failure. For short term SBO, both ATF cladding underwent hot leg creep rupture ~20 min after Zircaloy cladding. While Zircaloy and Cr-coated cladding had generated significant amount of hydrogen gas (>10 kg) before the creep rupture event, FeCrAl cladding had only generated <0.5 kg of hydrogen gas. For long term SBO with RCP seal failure, significant hydrogen generation and fuel melting was predicted before hot leg creep rupture for the ATF cladding while providing only 10–20 min additional coping time compared to Zircaloy. [ABSTRACT FROM AUTHOR]

In continuation from the earlier Part I, this paper presents the status on the investigations of Cr-coated zirconium alloys under both normal operating and hypothetical accident conditions. This paper is aimed to provide sufficient knowledge to understand the service performance and research focus of accident-tolerant fuel (ATF) coatings. [ABSTRACT FROM AUTHOR]