Accident Tolerant Fuel Research Articles

FeCrAl fuel/clad chemical interaction in light water reactor environments

This article investigates the fuel-cladding chemical interaction (FCCI) behavior of two commercial FeCrAl alloys, APMT composition (Fe-21Cr-5Al-3Mo wt.%) and C35M (Fe-13Cr-5Al-2Mo-0.2Si-0.03Y wt.%), after neutron irradiation. “H-cup” diffusion multiples of FeCrAl alloys and ceramic UO2 fuel were irradiated at a temperature of ∼300 °C to a total estimated burnup of 26 GWd/tHM. Post-irradiation Examination results demonstrate the excellent degradation resistance of FeCrAl alloys as accident tolerant fuel (ATF) cladding materials in light water reactor conditions. The study concludes that there was no irradiation-induced defects observed in either of the two commercial FeCrAl claddings. The formation of amorphous Al/U mixed oxide was observed at the fuel-clad interface, which can serve as a tritium permeation barrier and protect against potential chemical attack from the fuel. The study attributed the formation of amorphous Al/U mixed oxide to the low temperature and limited time of neutron irradiation. APMT forms more distinct Cr and Cr-Fe intermetallic at the FeCrAl-UO2 interface than C35M due to the higher bulk Cr:Al ratio.

Searching for the viability of using thorium-based accident-tolerant fuel for VVER-1200

This study explores the feasibility of employing (U, Th)-based accident tolerant fuels (ATFs), specifically (0.8UO2, 0.2ThO2), (0.8UN, 0.2ThN), and (0.8UC, 0.2ThC). The investigation assesses the overall performance of these proposed fuel materials in comparison to the conventional UO2, focusing on deep neutronic and thermal-hydraulic (Th) analyses. Neutronic analysis utilized the MCNPX code, while COMSOL Multiphysics was employed for thermal-hydraulic analysis. The primary objective of this research is to overcome the limitations associated with traditional UO2 fuel by exploring alternative fuel materials that offer advantages in terms of abundance and potential improvements in performance and safety. Given the limited abundance of UO2, long-term sustainable nuclear energy production faces challenges. From a neutronic standpoint, the U–Th based fuels demonstrated remarkable fuel cycle lengths, except (0.8UN, 0.2ThN), which exhibited the minimum fuel cycle length and, consequently, the lowest fuel burn-up. Regarding thermal-hydraulic performance, (0.8UN, 0.2ThN) exhibited outstanding performance with significant margins against fuel melting compared to the other materials. Overall, when considering the integrated performance, the most favourable results were obtained with the use of the (0.8UC, 0.2ThC) fuel configurations. This study contributes valuable insights into the potential benefits of (U, Th)-based ATFs as a promising avenue for enhanced nuclear fuel performance.

Investigation on the Irradiation-thermal–mechanical coupling behaviors of the solid Microencapsulated fuel rod

Solid Microencapsulated Fuel (SMF) is a kind of new accident-tolerant fuel (ATF), where the TRISO coated fuel particles are dispersed directly into a type of matrix in the form of a solid rod. SMF is expected to provide improved performance because of the elimination of cladding tube and associated failure mechanisms. This study aims to investigate the irradiation-thermal–mechanical coupling behavior of SMF rod under normal operation of the 200 MW Nuclear Heating Reactor (NHR200-II) using finite element method (FEM). Two kinds of SMFs have been evaluated and compared -with the fuel particles dispersed into a silicon carbide matrix or a zirconium matrix. The finite element models, considering comprehensive coupling behavior (including irradiation-induced dimensional change, irradiation-induced creep, thermal expansion, fission gas pressure, heat generation and conduction, elasticity and plasticity, etc.), were developed based on the ABAQUS platform and accurately verified. The study also investigated the impacts of key parameters on the coupling behavior of SMF, including particle kernel diameter, non-fuel zone width, extent of particle aggregation, and operational conditions. The results revealed that SMFs exhibit excellent heat transfer capability, resistance to deformation, and a stress-relaxed state under normal operation. Specifically, SMF-Zr exhibited a higher centerline temperature but lower levels of deformation and stress compared to SMF-SiC. The effects of fuel kernel diameter and non-fuel zone width were found to be limited, whereas higher particle aggregation and harsher operational conditions indeed increase temperature, deformation, and stress, particularly during the initial irradiation stage. Moreover, a higher creep coefficient of pyrolytic carbon leads to a more relaxed stress state. This study lays a fundamental basis for the design and optimization of this kind of solid particle-dispersed fuel.

Improved and Innovative Accident-Tolerant Nuclear Fuel Materials Considered for Retrofitting Light Water Reactors—A Review

Since 2011, there has been an international effort to evaluate the behavior of newer fuel rod materials for the retrofitting of existing light water reactors (LWR). These materials include concepts for the cladding of the fuel and for the fuel itself. The materials can be broadly categorized into evolutionary or improved existing materials and revolutionary or innovative materials. The purpose of the newer materials or accident-tolerant fuels (ATF) is to make the LWRs more resistant to loss-of-coolant accidents and thus increase their operation safety. The benefits and detriments of the three main concepts for the cladding are discussed. These include (i) coatings for existing zirconium alloys; (ii) monolithic iron–chromium–aluminum alloys; and (iii) composites based on silicon carbide. The use of ATF materials may help extend the life of currently operating LWRs, while also being a link to material development for future commercial reactors.

A comprehensive DFT study on the physical properties of XCrAl (X= Fe, Co, Ni, Cu) half-Heusler alloys for applications in high-temperature technology

In the present work, structural, electronic, magnetic, elastic, mechanical, thermal and optical properties of XCrAl (X = Fe, Co, Ni, Cu) Half-Heusler (H–H) alloys were explored. First-principles study based on the non-polarized and spin-polarized density functional theory (DFT) were employed to get the geometrically relaxed unit cell structures of H–H alloys of investigation. The larger lattice constant values were found due to the spin polarization and the magnetic moments were calculated from the spin density of states (DOS). The higher values of magnetic moments make XCrAl alloys suitable for magnetic device applications. The computational electronic structure analyses at zero applied pressure confirm the zero energy band gap of XCrAl alloys which displayed metallic nature for all the compounds. From the elastic constants calculations, all the studied alloys were found mechanically and thermally stable. The values of melting temperatures were obtained in the range from 2072 to 1883 K and the calculated Debye temperatures followed the same manner of variation. Among the studied compounds, FeCrAl exhibits highest melting temperature and lattice thermal conductivity with the value of 12.94 Wm−1K−1 at 300 K which is in good agreement with the other reported results. These investigations suggest that, FeCrAl is relatively more suitable in accident-tolerant fuel (ATF) cladding material applications over the common Zr-based cladding rods. The frequency dependent optical properties of the studied H–H alloys were also explored comprehensively and the examinations indicate that, XCrAl (X = Fe, Co, Ni, Cu) alloys can also be the good choice for ultraviolet (UV) based optical device applications as they have high absorption coefficient of >200 × 104 cm−1 and maximum reflectivity of >94% in the UV region.

Design and experimental investigation of the high-entropy alloys AlCrFeNiCu and AlCrFeNbMo

To improve the service safety of accident-tolerant fuels in nuclear power plants, the high-entropy alloys (HEAs) AlCrFeNiCu and AlCrFeNbMo were designed and prepared through mechanical alloying (MA) and spark plasma sintering (SPS) processes in this work. The phase structures of AlCrFeNiCu and AlCrFeNbMo were predicted and verified. The microstructure, phase structure, and properties of alloy powders and sintered alloy were investigated under various ball-milling conditions. Results show that the phase formation rule of HEA is an empirical criterion that still needs experimental verification. The HEA powder was refined and had good fluidity after MA. Body-centered cubic (BCC) AlCrFeNbMo powder can be prepared at high milling speeds and long milling times, while the structure of AlCrFeNiCu powder was a BCC plus face-centered cubic (FCC) duplex phase structure. The spark plasma-sintered alloy had fine grains (＜1 μm) and a multiphase structure. AlCrFeNiCu was mainly composed of FCC, B2, and a trace of σ phase, while AlCrFeNbMo was composed of BCC, FCC, and σ phase. The density, hardness, compressive strength, yield strength and strain of AlCrFeNiCu and AlCrFeNbMo alloys were 6.54 g cm−3, 447.57 HV, 1727 MPa, 1642 MPa and 15% and 6.32 g cm−3,1188.82 HV, 2183 MPa, 1943 MPa, and 19%, respectively, which are higher than reported for most HEAs that were prepared by SPS. The fracture mechanism is the typical brittle fracture. Because of solution strengthening and precipitation strengthening, AlCrFeNiCu and AlCrFeNbMo have excellent mechanical properties.

CrN–NbN nanolayered coatings for enhanced accident tolerant fuels in BWR

The accident tolerant fuel (ATF) concept has emerged in the years after the 2011 Fukushima accident as part of a renewed effort in research for light water reactors. The primary focus is to further improve safety measures under and beyond design basis accident conditions, and to improve fuel cladding performance in normal operation. The application of a coating on zirconium claddings can achieve both these aims without extensive changes to the reactor design. Metallic chromium coatings have been profusely studied as solution for pressurized water reactors, but the search for an effective ATF coating able to withstand the environment inside boiling water reactors (BWRs) is still ongoing. In this work, two different versions of a novel nitride coating composition were studied. Zirconium claddings coated with 8 µm thick layers of superlattice CrN–NbN and a nanolayered CrN–NbN were tested in autoclave under BWR operating conditions for 60 days. Scanning electron microscopy, transmission electron microscopy, energy dispersive x-ray spectroscopy, electron back-scattered diffraction, x-ray diffraction, and atom probe tomography were employed to characterize as-deposited and autoclaved samples of these two materials. During exposure, both coating versions formed a stable, dense and passivating oxide scale (200–300 nm thick) on the surface, demonstrating improved oxidation protection under operating conditions. Some differences in the oxide growth mechanism were observed between the superlattice and the nanolayered CrN–NbN coatings, which allowed to glimpse at the effect of the layer thickness on the oxidation protection provided by these coatings. The nano-structured morphology of both coatings remained unaffected by the autoclave test, but a 35 nm thick Zr-Cr-N phase was found at the coating-substrate interface of the superlattice CrN–NbN coated cladding.

On the creep mechanisms and macroscopic creep rate modeling of high-uranium-density composite fuels

UN-U3Si2 composite fuel is a promising candidate of Accident Tolerant Fuels (ATFs). Its in-pile creep performance will have an important impact on the irradiation-induced thermo-mechanical coupling behavior and safety of fuel elements. Based on the existing mechanism-based creep models of UN and U3Si2, the finite element analysis was performed for the multi-scale creep behaviors of various UN-U3Si2 composite fuels, and the influences of the U3Si2 volume fraction, applied stress, temperature, fission rate, and grain size were investigated. The dominant creep mechanisms under different conditions were proposed, and the mechanism-based macroscopic creep rate model for the UN-U3Si2 composite fuels were correspondingly developed, quantitatively correlating with the macroscopic creep rate with the U3Si2 vol fraction, temperature, fission rate and equivalent stress. The creep rate predictions obtained by this model were in good agreement with the results of the finite element simulation. The research results indicate that: (1) the irradiation or thermal diffusion creep contributions from the UN and U3Si2 phases were dominated under various conditions; (2) the dislocation creep contributions from the dispersed-phase of U3Si2 were appreciable at the temperatures ranged from 500 K to 1300 K, while those of the matrix-phase of UN appeared only at an extremely high temperature of 1300 K; (3) different from the ordinary inert-matrix dispersion fuels, the macroscopic creep rates of UN-U3Si2 composite fuels were predicted to be almost independent of burnup, due to the slight difference in the irradiation swelling of UN and U3Si2; (4) the total creep contribution proportions of the U3Si2 or UN phase were almost equal to the initial volume fractions, while the creep contribution ratios within the UN or U3Si2 phase vary under different conditions. The evolutions of various creep contributions with the fission density were attributed to the fuel-swelling induced additional stresses.

Quantification of 1.75 MeV Xe5+ induced defects in zirconia doped ceria (Ce0.8Zr0.2O2)

This study investigates the structural stability of the CeO2-ZrO2 system when subjected to high doses of irradiation (a few hundred displacements per atom). The goal is to explore their potential use in safe immobilization of spent nuclear fuel and development of accident-tolerant fuels for next-generation nuclear reactors. Highly dense pellets were synthesized using a solid-state reaction and then irradiated with 1.75 MeV Xe5+ ions at ion fluences ranging from 1 × 1015 to 1 × 1017 ions/cm2. Structural and microstructural analyses were conducted using glancing angle x-ray diffraction, micro-Raman spectroscopy, and scanning electron microscopy. The results indicate that Ce0.8Zr0.2O2 has a high tolerance against irradiation-induced phase transformation or amorphization but does generate irradiation-induced point defects. Each energetic ion produced a deformed region with a voluminous swelling of ∼0.61 ± 0.09 nm and a damaged zone of ∼0.09 ± 0.02 nm, as calculated from the irradiation-caused peak broadening that is explained by a three-step damage accumulation model. The electron microscopy studies show that grain boundaries serve as a sink for defects, and an increase in grain size was observed due to defect accumulation inside the grain's volume. Overall, the study shows that polycrystalline fluorite-structured Ce0.8Zr0.2O2 is a promising nuclear material for advanced energy systems as it did not show significant structural damage such as amorphization and grain fragmentation, even on irradiation at a high dose of ∼428 dpa.

Multi-objective optimization design of TRISO-based fully ceramic microencapsulated fuel

Fully Ceramic Microencapsulated (FCM) fuel is a new type of material in nuclear energy field. It has become a promising accident tolerant fuel (ATF) candidate due to its advantages of high radiation stability, high fission product containment capacity and excellent thermo-mechanical performance. During operation, the temperature, stress, deformation and other key parameters will change significantly, which directly affect the safety of nuclear reactor. Therefore, it is necessary to carry out a multi-physics fully-coupled analysis suitable for the multi-scale, multi-component, multi-phase and nonlinear coupling processes in the reactor, so as to sort out the influences of material properties, radiation behavior, thermo-mechanical performance from numerous influencing factors. In this paper, based on the irradiation-thermo-mechanical coupling method, the performance of TRISO-based FCM fuel under irradiation conditions was analyzed, key thermo-mechanical parameters and other influencing factors of the fuel were obtained, failure behavior of the fuel element was evaluated. On this basis, the theoretical and numerical model of multi-objective optimization method were established, the optimum parameters including material properties, fuel geometry and particle arrangement were found by optimization method. The optimization results showed that the multi-objective optimization results could provide multiple Pareto optimal solutions for the FCM fuel. Specifically, when the TRISO particle spacing is 200 µm and 150 µm, the objective function difference between the optimal heat transfer solutions is only about 1.7%, meanwhile the mechanical properties vary significantly (the difference between the objective functions of the optimal solutions is about 12.4% ∼26.5%), indicating the response of stress is more sensitive than temperature. When the TRISO particles are tightly packed (particle spacing less than 100 µm), the difference of objective function between the optimal solutions is less than 5%, illustrating the optimization has limited improvement on the thermo-mechanical performance of the fuel, namely the optimization space is limited.

Effect of Cr coating on the mechanical integrity of Accident Tolerant Fuel cladding under ring compression test

The mechanical behavior of Cr-coated Zr-Nb alloy cladding was investigated with Ring Compression Test (RCT) and Finite Element Analysis (FEA). The post-RCT specimens were examined by SEM analysis with a focus on crack characterization. The experimental result shows that Cr-coating less than 20 μm does not introduce notable changes in the load-displacement behavior of the cladding structure until the point at which the Cr-coated cladding undergoes premature failure. FEA results demonstrate that the premature failure of the Cr-coated cladding cannot be explained by local stress and strain fields as they are not affected by the presence of the coating if the crack of the coating layer is not considered. Under RCT, a crack is initially formed across the coating thickness and acts as a pre-existing flaw that concentrates stresses, growing towards the underlying Zr-Nb alloy. Once the crack depth attains the critical size, it unstably grows, resulting in cladding failure which is shown as an earlier load drop compared to the uncoated cladding. FEA results using eXtended Finite Element Method (XFEM) demonstrate the stress concentration by the coating crack which results in the premature failure of the Cr-coated cladding in RCT. Increasing coating thickness decreases crack density, thereby reducing the uncovered area of the underlying cladding in case where the coating experiences extended cracking. On the other hand, the thicker coating, when cracked, can induce extended damage to the underlying cladding by introducing larger cracks. The net benefit of these two competing effects needs to be considered to optimize the thickness of Cr-coating.

Thermomechanical evaluation of ATF fuel rods regarding irradiation effects

Fuel clad is the main barrier against the release of radioactive materials. For decades zirconium-based alloys have been used as nuclear fuel clads. Fukushima accident proposed the idea of designing and developing accident tolerant fuels (ATF). This research tries to evaluate the effect of irradiation damage on ATF clads under operating condition of a pressurized water reactor. In this work, the SiC and FeCrAl clads are investigated as two candidates for replacement with conventional cladding, Zircaloy-4. The neutronics of a WWER1000 reactor is modeled based on the Monte Carlo method. The neutronic evaluation of the core with desired clads is performed over three years in different spatial points. Using calculated neutron spectrum, the SPECTER and SPECOMP codes are applied to evaluate the damage caused by the irradiation in the term of displacement per atom (DPA). The thermomechanical evaluation of the fuel rod considering cladding irradiation effects and burnup are performed by a developed thermomechanical code. Also, stress, strain, creep, fuel-clad contact, thickness of oxide layer and … are calculated and the results are compared considering irradiation effects.

Microstructure and Corrosion Behavior of the Modified Layers Grown In Situ by Plasma Nitriding Technology on the Surface of Zr Metal

Preparing protecting coatings on the surface of Zr claddings has been regarded as one of the accident tolerant fuel (ATF) strategies. In this study, a series of nitride-modified layers were in situ grown by hollow cathode plasma nitriding on the surface of Zr metal. The influence of nitriding currents and time on the phases, composition, microstructure and corrosion resistance of the modified layers was investigated by X-ray diffraction (XRD), X-ray Photoemission Spectroscopy (XPS), transmission electron microscope (TEM), scanning electron microscopy (SEM) with energy dispersive spectrometer (EDS) and potentiodynamic polarization curves. The ZrO2 layer with loose microstructure and cracks prefers to form under low nitriding current of 0.4 A, which also causes poor corrosion resistance. The high temperature caused by high nitriding currents (0.6 A and 0.8 A) promote the formation of compact nanocrystalline layers, made up of nitride and oxynitride. Below the nanocrystalline layer, it is Zr2N caused by N penetration. Besides this, a double-layer structure of the nanocrystalline layer, i.e., an equiaxed crystal zone with a grain size of ~10–50 nm on the surface and a long strip grain region beneath it was observed. The compact nitride/oxynitride layer with excellent interface bonding can improve the corrosion resistance effectively.

The elemental weakening & aggregation effect on the hydrogen adsorption on FeCrAl (110) surface

FeCrAl alloys are promising candidates for Accident Tolerant Fuel cladding in Light Water Reactors. Understanding the adsorption mechanism of hydrogen is important for understanding the hydrogen embrittlement of FeCrAl alloys. In our previous work, we find that the adsorption site can significantly influence the H adsorption energy (Eads.). However, the widely adopted Special Quasi-random Structure (SQS) method for constructing FeCrAl structure hinders comprehensive investigations on all the possible H adsorption sites. In this work, a manually created surface structure is used to systematically study all the possible H adsorption sites. It is found that when Cr/Al atoms are dispersed on the surface, the HCr/Al bonding can be weakened. The weaker HCr/Al interactions are attributed to the charge deficit of Cr/Al caused by their relatively higher electronegativity than Fe. In contrast, when Cr/Al atoms aggregate on the surface, the HCr/Al bonding can be reinforced. The reinforcement comes from the increased charge density of Cr/Al due to aggregation.

Review on performance of chromium-coated zirconium and its failure mechanisms

The conventional Zircaloy cladding may react with coolant in light water reactors and elicit severe accident at high temperature, which constraint the safe and efficient development of nuclear energy. To avoid the tragedy happened in Fukushima, the concept of Accident Tolerant Fuel (ATF) was developed around the globe to secure the steady operation for nuclear power plants. As a short-term solution, deposition of protective coatings upon the cladding tubes are proved to be effective. Among the coatings, Chromium is one of the most outstanding choices. This paper reviews the recent study results about the performance and failure mechanisms of Cr-coated Zr cladding, including mechanical, irradiation, high-temperature oxidation tests, etc. under normal and/or accident conditions. The experiments and tests collectively suggest that Cr coating possesses excellent comprehensive properties for integrity protection of claddings and provides a promising future of the commercial application in reactor cores. Finally, this review proposes ambiguities and questions needed to be analyzed and clarified in future works.

The elemental effects on the H2 dissociative adsorption on FeCrAl (110) surface

As a promising candidate for accident-tolerant fuel (ATF) cladding materials, FeCrAl alloys are susceptible to hydrogen embrittlement (HE) under high-temperature radioactive water environments. Understanding the HE mechanism of FeCrAl is crucial, and studying H2 dissociative adsorption can provide valuable insights since it is the first step in the HE processes. In this paper, we investigated the H2 dissociative adsorption on Fe (110) and FeCrAl (110) surfaces using first-principles calculations based on Density Functional Theory (DFT). Our results indicate that the decomposition of H2 molecules on FeCrAl (110) surfaces is significantly affected by surface elemental effects compared to Fe (110) surfaces. Specifically, Al atoms weaken both the H2 decomposition and the H binding strength of Al-containing sites. Cr atoms decrease the decomposition tendency of H2 in certain configurations, but Cr aggregation enhances the binding strength of H atoms on Cr-containing sites. The mechanism underlying the Al/Cr weakening effects on H2 decomposition is due to the charge deficiency through alloying. Our findings have generalized implications for studying the interactions between iron-based alloys and hydrogen in various applications such as hydrogen storage, transportation, and prevention.

Diffusion of chromium of Cr-coated Zircaloy accident tolerant fuel cladding: Model development and experimental validation

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, ZrCr2 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.

Construction of Cr coatings with different columnar structure on Zircaloy-4 alloys to optimize the high-temperature steam oxidation behavior for accident tolerant fuel claddings

To optimize the oxidation behavior of chromium-coated Zircaloy-4 alloys under 1200 °C high-temperature steam for accident tolerant fuel (ATF) claddings, Cr coatings with different columnar structure are constructed by changing bias voltage using high power impulse magnetron sputtering and corresponding microstructure, mechanical and high-temperature steam oxidation properties are studied systematically. With the bias voltage increase from 0 V to − 150 V, the microstructure of Cr coatings changes from porous columnar structure to compact columnar structure. Especial for bias voltage is − 90 V, the tiny compact grains uniform distribute on the surface micrographs and optimal compact columnar structure are observed on the cross-section micrographs. The hardness and elastic modulus of as-prepared Cr coatings increase firstly and then decrease, the hardness of Cr coatings with optimal compact columnar structure is 8.65 GPa, which is elevated ∼2.28 times than Cr coatings with porous columnar structure. 1200 °C high-temperature steam oxidation test results show an interesting phenomenon that the thickness of Cr layer reduces initial and then increase with the exposure time prolong from 1 h to 4 h. The main steam oxidation mechanism, in the early oxidation stage, is formation the compact Cr2O3 oxide layer. Then, with the exposure time extend, the reduction reaction plays the significant part for steam oxidation mechanism, which is consumption of outmost Cr2O3 oxide layer and formation of subsurface Cr layer. The construction of Cr coatings with compact columnar structure and relevant oxidation mechanism are expected to offer a new idea for improving 1200 °C steam oxidation resistance of chromium-coated Zircaloy-4 alloys.

Effect of temperature on the fretting wear behavior of Cr-coated Zircaloy cladding in high-temperature pressurized water

Cr-based coatings are promising candidates to protect the claddings on accident-tolerant fuel (ATF) rods in pressurized water reactors (PWRs). In this study, the effect of water temperature on the fretting wear behavior of an N36 Zirconium alloy cladding was investigated. In addition, the effect of applying a protective Cr coating to the cladding was also analyzed. Fretting wear tests were carried out using a unique apparatus that simulated the high-temperature and high-pressure (HTHP) water environment in a real PWR. Results show that changes in temperature alter the wear mechanism of cladding tubes. Furthermore, a Cr-coating improves the wear resistance of an N36 cladding significantly. The main wear mechanisms of Cr-coated cladding in HTHP water were observed to be abrasive and adhesive wear. The protective effect of a Cr coating on the cladding in a low-temperature water environment (25 °C, 90 °C) was less than it was in a high-temperature water environment (260 °C, 315 °C), indicating that Cr coatings can improve wear resistance in HTHP water.

Eutectic reaction and oxidation behavior of Cr-coated Zircaloy-4 accident-tolerant fuel cladding under various heating rates

The Cr-coated zirconium alloy claddings are near-term approach for accident tolerant fuel (ATF) claddings. The performance of magnetron-sputtered Cr-coated zirconium alloys is investigated under conditions simulating severe accidents. The transient oxidation experiments were performed using a thermogravimetric analyzer (TGA) with five different heating rates (2–50 K/min) up to 1380 °C, which is above the Zr-Cr eutectic temperature. The weight gain of the Cr-coated specimens was found to be significantly lower than that of the uncoated reference up to approximately 1330 °C. However, after the occurrence of the Zr-Cr eutectic reaction and failure of the coating at 1330∼1380 °C, the oxidation rate rapidly increased and the weight gain rate of the Cr-coated specimens was 1.5 to 10 times higher than that of the uncoated specimens. The difference was greater at higher heating rates. These results suggest that the Cr-coating has a limited effect in preventing oxidation and degradation of cladding in case of eutectic reaction. The oxidation was faster at higher heating rate, and it may even cause runaway oxidation of cladding due to the temperature rise caused by oxidation heat. These findings are important for understanding the capabilities and degradation behaviors of Cr-coated cladding under severe accidental scenarios.