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An alternative material for fuel cladding was required to prevent oxidation caused by interacting with steam, leading to improvement in core integrity. This study analyzes reactor physics parameters of various cladding material candidates for Pressurized Water Reactor (PWR) such as SS-304 austenitic stainless steel, FeCrAl alloy, APMT alloy, and silicon carbide (SiC) ceramic, as candidate Accident Tolerant Fuel (ATF). The neutronic parameters such as infinite multiplication factor (k-inf), and neutron spectrum, while temperature reactivity coefficient related to fuel temperature (DTC) and moderator temperature (MTC) is also considered, followed by a void coefficient of reactivity (VCR) of each candidate material were then compared with ZIRLO as a standard cladding material of AP1000. k-inf calculated by SRAC2006 is also compared to MCNP for various fuel assembly types. At the beginning of cycle (BOC), the 2.35% UO2 using SiC gives a higher kinf than ZIRLO at 937 pcm, while 4.45% UO2 with 88 IFBA & 9 PYREX at 796 pcm. FeCrAl, APMT, and SS-304 cladding gave a smaller k-inf compared to ZIRLO in the range of 11000-14000 pcm at 2.35% UO2 fuel assembly. The values of DTC, MTC, and VCR were still negative throughout the reactor operation which indicates that the inherent safety feature of alternative cladding was possible for this type of fuel assembly, especially for iron-based cladding material followed by an increase in fuel enrichment. [ABSTRACT FROM AUTHOR]

Since the Fukushima incident, accident tolerant fuel (ATF) has received widespread attention. Ceramic-based ATF, represented by UO2, has become the mainstream of nuclear fuel, due to its advantages such as high melting point, excellent radiation resistance, and chemical stability. However, ceramic-based ATF still faces challenges such as complex preparation processes and high costs. Progress has been made in shortening sintering time and reducing sintering temperature by adopting new sintering techniques such as spark plasma sintering and flash sintering to prepare ceramic-based ATF. This article is aimed to analyze and summarize the progress of sintering technology for ATFs based on UO2, UN, U3Si2 and UC. In addition, the characteristics and development prospects of different sintering techniques will be discussed and the future development of ATF will be looked forward. [ABSTRACT FROM AUTHOR]

This paper describes results from the mechanical evaluation of unirradiated SiC fiber–reinforced SiC matrix composite tubes under a controlled high-temperature steam environment. The experiments were aimed at identifying key material degradation behavior under environments relevant to loss-of-coolant accidents of light water reactors. Mechanical tests of the SiC composite tubes at 1000 °C under steam and inert environments were conducted using a unique test capability. The material tested was a duplex tube with a thick monolithic SiC layer on the outer surface. The tubes were subjected to preloading at ∼100 MPa in tension before exposure to high-temperature steam with up to 75 % of the preload at a constant displacement. In the presence of matrix cracks, the steam exposure caused embrittlement of the SiC composite tubes and failure at a stress level below the pretest stress. The material degradation was explained by a fiber oxidation model, which can be applied to various SiC cladding concepts. The embrittlement could be a limiting factor for using SiC cladding subjected to loss-of-coolant accident conditions. [ABSTRACT FROM AUTHOR]

Accident Tolerant Fuel (ATF) is a next-generation fuel technology developed to enhance the fundamental safety of light water reactors. Compared to existing standard fuel technologies, ATF technology can provide the same or better performance during normal operation, transient conditions, and accident scenarios of nuclear power plants, enhancing the ability of fuel elements to withstand severe accidents while reducing the cost of generating electricity over the licensed life of the plants. This paper introduces the definition, characteristics and thermal-hydraulic research objectives of ATF technology, systematically reviews the thermal-hydraulic research results of different ATF technologies in various aspects from the perspectives of near-term concepts (UO2-based composite fuels, coated Zr-based claddings, FeCrAl claddings) and long-term concepts (high-uranium-density UN fuels, FCM fuels, extruded metal fuels, SiC/SiC claddings), respectively. Mainly includes: improving fuel thermal conductivity, lowering fuel operating temperatures and stored energy, reducing fuel-cladding interactions, improving cladding-coolant reactions, reducing the production of flammable gases (e.g., hydrogen) and enhancing the retention of radioactive fission products. Finally, this paper points out the current shortcomings and challenges in the field of ATF thermal-hydraulic research and outlooks the potential future directions.

Two strategies have been extensively employed to develop advanced accident tolerant fuels (ATF): improving thermal conductivity and producing large grain sized pellets. However, there are few reports on the simultaneous utilization of both strategies. In this work, we fabricated Mo–Cr alloy reinforced UO 2 (UMC) composite pellet with both high thermal conductivity and large grain size by a simple in-situ alloying method for the first time. The average grain size of UO 2 increased from 6 μm to 113 μm. Thermal conductivity of the UMC pellet (with 2 vol% Mo) at 1200 °C increased by 36.46% compared with pure UO 2. The increase rate of thermal conductivity per 1 vol% dopant reached about 18%. The Mo–Cr alloy formed continuous micro-cell layer around UO 2 particles. Such continuous micro-cell Mo–Cr alloy layer and increase of UO 2 grain size both helped improve the thermal conductivity of UMC composite pellets. Thermal expansion coefficient of the UMC under the operating temperature decreased by 11.90% compared with pure UO 2. The oxidation resistance of the UMC pellet under high temperature steam was also improved. This work provides a new strategy towards fabricating ATF with both high thermal conductivity and large grain size. [ABSTRACT FROM AUTHOR]

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Accident tolerant fuel can improve the safety of nuclear reactor core and strengthen the ability of nuclear reactor to withstand serious accidents. Coated particle dispersed fuel has become a hot research topic in accident tolerant fuel in recent years, due to its high thermal conductivity and good radiation resistance stability. Coated particle dispersed fuel consists of threedimensional isotropic coated particles as the fuel phase and silicon carbide as the matrix phase, adopting the fuel form of coated particle dispersed in the silicon carbide matrix. However, due to the low selfdiffusion rate of SiC, it is difficult to achieve high densification by solid phase sintering, which is only controlled by diffusion. The existing sintering process mainly involves adding sintering aids or generating liquid glass phase to assist in the sintering of SiC powder. Spark plasma sintering (SPS) process can activate powder particles and promote sintering, and has been widely used in nuclear ceramic sintering in recent years. This article used SPS to prepare coated particle dispersed fuel pellets, and studied the influence of sintering process parameters on the density of the coated particle dispersed fuel pellets, such as the amount of sintering additive addition, maximum sintering temperature, maximum sintering pressure, and holding time. It is found that the best process parameters for SPS sintering of coated particle dispersed fuel pellet is 7% of sintering additive addition, holding 15 minutes at 1 850 ℃ and 45 MPa. Under such process parameter, the relative density of pellet can reach 98.5% of theoretical density. In addition, the sintering mechanism of coated particle dispersed fuel was studied and it is found that the sintering of coated particle dispersed fuel can be divided into five stages. Phase detection, metallographic detection and microstructure characterization analysis of the coated particle dispersed fuel pellets were carried out. The XRD examination results indicate that after sintering, the coated particle dispersed fuel pellet is pure β phase, which indicate that SPS rapid sintering can better control the phase transition. The metallographic examination results indicate that there is no obvious contact between TRISO particles in the coated particle dispersed fuel pellet obtained under the optimal process conditions. The SEM examination results show that the overall integrity of TRISO particles is good, indicating that the structure of TRISO particles dose not break during the sintering process. In the thermal shock test, there are no penetrating micro-cracks on the surface of the fuel pellet after a thermal shock, which maintain good integrity.

The APMT FeCrAl alloy is a cladding candidate for the accident-tolerant fuel (ATF) systems in light water reactors (LWRs) due to its ability to form a protective Al2O3 film on the surface at temperatures greater than 1000 °C. In order to understand the alumina evolution at the early stages of high-temperature exposure in short periods, an APMT tube specimen was heated up rapidly to melting temperature in a flowing 100 % water and steam environment. Using transmission electron microscopy (TEM), the presence of a ∼ 90 nm thick oxide layer was confirmed. The surface chemistry of APMT was characterized layer by layer using X-ray photoelectron spectroscopy (XPS) depth profiling. The results indicated a thin layer composed of oxides and hydroxides of Al, Cr, and Fe with varying proportions at different depths in the oxide layer. The microstructure of the oxide layer was characterized using X-ray diffraction and showed the presence of α-Al2O3 and γ-Fe2O3/Fe3O4 after ramp heating. The results depict the ability of the APMT FeCrAl alloy to form a protective oxide layer in less than a second on exposure to steam under high ramp rates.

A solid-phase metallurgy combined with spark plasma sintering technology was used to prepare U3Si2 pellets. The thermal conductivity and oxidation behavior of the pellets were studied. The pellets were highly dense (> 98% theoretical density) and had an ultra-high phase purity. The thermal conductivity of the U3Si2 pellets increased linearly with temperature from 300 to 973 K. By further annealing the pellets at 300 °C in air for two hours, the oxidation onset temperature was increased from 490 to 520 °C. A mechanism of pore passivation was proposed to account for the enhanced oxidation resistance. [ABSTRACT FROM AUTHOR]

Accident tolerant fuel (ATF) are currently being developed as part of an international effort to modify or replace zirconium-based cladding materials in light water reactors. The main objective of ATF cladding materials is to improve fuel reliability and safety during accident scenarios. Thus, these materials should perform comparable to or better than the current zirconium alloy. One of the ATF concepts explores modifying standard Zircaloy cladding by coating their outer surface. In this study, surface-modified Zr-4 is produced by depositing a protective coating of chromium through two different coating techniques, Physical Vapor Deposition (PVD) and Cold Spray (CS). The coated specimens were tested at ∼300 °C in out-of-pile autoclaves simulating Boiling Water Reactor (BWR) conditions. Results show that after one month of immersion, Cr-coated Zr-4 evidenced a small mass gain in hydrogenated environments. However, both types of coated specimens showed significant coating degradation and mass loss in oxygenated environments.

To address concerns about the degradation of unprotected internal surfaces of nuclear fuel claddings, inner-side coatings have been proposed as a complementary approach to the accident tolerant fuel protective coating concept. In addition to the increased coping time during severe events, these coatings are expected to provide enhanced protection in normal operation conditions. This study analyzes the neutronic performance of chromium (Cr) inner-side coatings in VVER-type fuel assemblies. Different aspects, such as reactivity and cycle length penalties, enrichment requirements, neutron flux, and the associated isotopic concentration changes are discussed, considering both the coating thickness and position. The results show that for the same thickness, reactivity penalties due to the use of inner-side Cr coatings will be (~30% on average) less compared to external coatings. The fuel assembly operating cycle showed reductions by ~ 5.5 effective full-power days when a 10-µm-thick internal Cr coating is introduced, while a 10-µm two-sided coated assembly possessed an ~13.6-day shorter operating cycle compared to an uncoated fuel assembly of the same specifications. The neutron flux showed slight shifts and hardening in the thermal energy region. The analysis of nuclide inventories showed relative increases in these inventories, which were proportional to the thickness. For the fissile plutonium isotope 239Pu, this relative increase reached a peak of 0.25% and 0.42% (for the 10-µm and 20-µm internal Cr coatings) at a fuel burnup of 18 MWd/kg heavy metal (HM). While for 241Pu, the observed highest relative increases for the 10-µm- and 20-µm-thick internal Cr coatings were 0.74% and 1.03%, respectively. The 135Xe isotopic concentration showed a relative increase that reached 0.2% and 0.4% for the 10-µm and 20-µm internal Cr coatings at a burnup of 34 MWd/kgHM, while the 149Sm concentration increased by 0.2% and 0.5% for the 10-µm and 20-µm internal Cr coatings, respectively. While these observed isotopic concentration changes were generally small for the studied inner-side coatings, the results showed that the changes remain subject to further increases as the amount of coating material gets higher. Therefore, it is important for the coating thickness to be optimized, taking into account the impact of such nuclide inventory changes. Possible fuel-clad gap reductions and the associated effects on heat transfer, as well as gap tolerance to fission products and fuel relocations, will require further studies, especially so that additional enrichments may be applied. [ABSTRACT FROM AUTHOR]

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Accident tolerant fuel (ATF) is expected to delay or prevent core damage by providing additional coping time under accidents involving loss of core cooling. The effect of extended coping time may vary depending on the plant response to accidents. Age-related component degradation that deteriorates plant performance over time could have an impact on the actual advantages of ATF. The potential safety benefits of two near-term ATF candidates, including Cr-coated Zr cladding and FeCrAl cladding, are assessed for a 2-in. loss-of-coolant accident with failed high-pressure safety injection using the dynamic event tree (DET) approach considering possible stress corrosion cracking of steam generator (SG) tubing under aging. The DET approach allows likelihood quantification of accident sequences leading to core damage, including stochastic variation of system response and human actions during accident mitigation. The safety benefits of the selected ATF claddings in terms of additional coping time and the core damage frequency reduction rate under specified accident situations were quantitatively estimated. The results show that the deployment of the two selected ATF claddings is expected to lead to longer coping times and lower core damage frequency due to the wider safety margin to peak cladding temperature they provide. The safety advantages would be greater as SG tube degradation proceeds. Thus, the two ATF candidates would lead to less severe consequences in terms of likelihood of core damage and susceptibility to the SG tube degradation than UO2-Zr fuel. [ABSTRACT FROM AUTHOR]

Uranium mononitride (UN) is a candidate fuel material for light water reactors with higher uranium (U) loading and thermal conductivity than uranium dioxide (UO2). However, the sintering of UN pellets is challenging as the UN powder particles oxidize rapidly at high temperatures unless the oxygen concentration is extremely low. Oxidation during sintering either reduces the relative density of the sintered UN pellet or disintegrates the sintered UN pellet to powder. To address this problem, the present work developed a rapid sintering method for producing highly densified UN surrogate pellets with minimal oxidation. Cerium nitride (CeN) is used as a surrogate for UN to reduce radiation hazards. With the custom-developed fast-heating system, the sintering process was completed within 150 s. The sintering atmosphere was flowing nitrogen (N2). The sintered CeN pellet density was 95% of the theoretical density (TD) or higher. The microstructure was uniform with a 10–25 µm grain size as demonstrated by scanning electron microscopy (SEM) and contained trivial levels of oxides as demonstrated by X-ray diffraction (XRD). The resultant pellets indicate that the rapid sintering method is a promising method to make UN fuel pellets with equivalent or higher density to pellets made by conventional sintering methods, while also being more efficient in time and costs.

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]

Solid Microencapsulated Fuel (SMF) is a type of solid fuel rod design that disperses TRISO coated fuel particles directly into a kind of matrix. SMF is expected to provide improved performance because of the elimination of cladding tube and associated failure mechanisms. This study focused on the neutronics and some of the fuel cycle characteristics of SMF by using OpenMC. Two kinds of SMFs have been designed and evaluated - fuel particles dispersed into a silicon carbide matrix and fuel particles dispersed into a zirconium matrix. A 7x7 fuel assembly with increased rod diameter transformed from the standard NHR200-II 9x9 array was also introduced to increase the heavy metal inventory. A preliminary study of two kinds of burnable poisons (Erbia & Gadolinia) in two forms (BISO and QUADRISO particles) was also included. This study found that SMF requires about 12% enriched UN TRISO particles to match the cycle length of standard fuel when loaded in NHR200-II, which is about 7% for SMF with increased rod diameter. Feedback coefficients are less negative through the life of SMF than the reference. And it is estimated that the average center temperature of fuel kernel at fuel rod centerline is about 60 K below that of reference in this paper.

The Accident Tolerant Fuel (ATF) is a new concept of fuel, which can not only withstand the consequences of the accident for a longer time, but also maintain or improve the performance under operating conditions. ISAA is a self-developed severe accident analysis code, which uses modular structures to simulate the development processes of severe accidents in nuclear plants. The basic version of ISAA is developed based on UO2–Zr fuel. To study the potential safety gain of ATF cladding, an improved version of ISAA, referred to as ISAA-ATF, is introduced to analyze the station blackout accident of PWR using ATF cladding. The results show that ATF cladding enable the core to maintain a longer time compared to zirconium alloy cladding, thereby enhancing the accident mitigation capability. Meanwhile, the generation of hydrogen is significantly reduced and delayed, which proves that ATF can improve the safety characteristics of the nuclear reactor.

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.

Cr-coated zirconium (Zr) alloy is one of the potential candidate materials for accident tolerant fuel (ATF) cladding. However, little attention has been paid in the simultaneous microstructure evolution of Cr coating and Zr alloy matrix under irradiation, which is necessary for comprehensive evaluation of the alloy irradiation performance. In the present work, in-situ ion irradiation experiment was conducted to capture the major features during simultaneous microstructure evolution of Cr coating and Zr alloy at various temperatures. It is found that at 300 ℃ the size and number density of dislocation loops increase with irradiation dose, but at higher temperature the tendency seems different, which is related to the formation of dislocation networks. Besides, the extent of precipitate amorphization in Zr alloy matrix decreases with temperature due to the recovery of irradiation defects. The present work may provide useful insights into the irradiation property of the Cr-coated Zr alloy for ATF applications. [ABSTRACT FROM AUTHOR]

The hydrogenation behavior of Cr-coated resistance upset welds (RUW) of E110 zirconium alloy was investigated at 360, 450 and 900 °C and a hydrogen pressure of 2 bar. The deposition of Cr coating, via magnetron sputtering, can decrease the hydrogen absorption rate of an RUW Zr alloy. The activation energy for the hydrogen absorption of Cr-coated specimens (84 kJ/mol) is higher in comparison with uncoated ones (71 kJ/mol), which indicates the deceleration of the hydriding of welded Zr alloys in the case of Cr coating deposition. A Cr coating can limit the formation of radially oriented hydrides and the hardening of RUW specimens at 360 and 450 °C. No significant difference in the hydrogen absorption rate was found at 900 °C. The application of Cr coating deposition to protect resistance-upset-welded Zr alloys in a hydrogen atmosphere is discussed. [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]

As an accident tolerant fuel (ATF) concept, oxide dispersion strengthened Zircaloy-4 (ODS Zry-4) cladding has been developed to enhance the mechanical properties of cladding using laser processing technology. In this study, a simulation technique was established to investigate the mechanical properties and effects of Y2O3 particles for the ODS Zry-4. A 3D representative volume element (RVE) model was developed considering the parameters of the size, shape, distribution and volume fraction (VF) of the Y2O3 particles. From the 3D RVE model, the Young's modulus, coefficient of thermal expansion (CTE) and creep strain rate of the ODS Zry-4 were effectively calculated. It was observed that the VF of Y2O3 particles had a significant effect on the aforementioned mechanical properties. In addition, the predicted properties of ODS Zry-4 were applied to a simulation model to investigate cladding deformation under a transient condition. The ODS Zry-4 cladding showed better performance, such as a delay in large deformation compared to Zry-4 cladding, which was also found experimentally. Accordingly, it is expected that the simulation approach developed here can be efficiently employed to predict more properties and to provide useful information with which to improve ODS Zry-4.

• Deformation and cracking of Cr coating were studied by in-situ tensile tests. • 450 °C annealed coating fractured in a brittle manner, similar to the as-deposited. • 800 °C annealed coating was crack-free under tension due to recrystallization. • Improvement in interface deformation compatibility helps to hinder cracking. Uncovering the deformation and cracking behavior of Cr coating were essential for the integrity of accident-tolerant fuel (ATF) claddings during both operating and accident conditions. In the present work, the crack resistance of Cr coated Zr-4 alloys was improved by an annealing process. By using novel in-situ tensile tests, the 450 °C annealed Cr coating with enlarged columnar grains fractured in a brittle manner under tension, similar to the as-deposited. But the crack density of annealed sample was reduced. For the 800 °C annealed Cr coating with large equiaxed grains, despite the formation of thin brittle ZrCr 2 interlayer, Cr coating was crack-free under even severe deformation, which benefited from recrystallization and grain growth at high temperature. Besides, the improvement in deformation compatibility between ductile Cr coating and Zr-4 substrate also played an important role in hindering initiation and propagation of surface cracks in Cr coating. [ABSTRACT FROM AUTHOR]

• The effect of cladding corrosion on heat transfer is considered. • The heat transfer model and the improved CHF model are established in COBRA-EN. • A detailed validation of the model was performed based on two benchmark experiments. • The thermal behavior of ATF under corrosion is investigated. Accident-tolerant fuel (ATF) enhances the accident tolerance of the fuels by improving its thermal properties and antioxidant radiation performance, thereby enabling the reactor to withstand severe accidents for a long time. This article applies and improves heat transfer and CHF models in the COBRA-EN by considering the impact of cladding surface corrosion on critical heat transfer between coolant and fuel rods. We conduct detailed and critical validation of the model constructed in this paper based on two benchmark experiments. We apply this model to study the thermal–hydraulic behaviors of ATFs under accident conditions. We obtain parameters such as the maximum fuel centerline temperature (MFCT), the maximum cladding surface temperature (MCT), the minimal departure from nucleate boiling ratio (MDNBR), the critical heat flux (CHF), and the average void fraction (AVF) for different ATFs. The results indicate that under most transient operating conditions, cladding corrosion delays the soaring time of the MFCT and MCT, and generally enhances CHF, with an average enhancement amplitude of over 0.15 MW/m2. At the same time, due to the effect of cladding corrosion, the MDNBR of the reactor has also been improved, which mitigates the impact of the accident to some extent. [ABSTRACT FROM AUTHOR]

• A Cr-coated Zr-alloy cladding capability has been incorporated into the ENIGMA code. • Thermo-mechanical behaviour of near-term accident tolerant fuel concept simulated. • Focus on behaviour during a design-basis loss-of-coolant accident in an LWR. • New models for cladding oxidation, hydriding & creep in LWR irradiation conditions. • Coating thinning via high-temperature diffusion of Cr into base alloy also modelled. Zirconium-alloy cladding with a chromium coating is the most advanced of the near-term concepts amongst Accident Tolerant Fuel (ATF) materials for Light Water Reactor (LWR) applications. The ENIGMA fuel performance code has been updated to model the thermo-mechanical behaviour of such cladding in both normal and off-normal operating conditions. The focus was on accurately simulating the behaviour in Loss Of Coolant Accident (LOCA) conditions to evaluate the increase in coping time during design-basis accidents. New low-temperature and high-temperature models were incorporated for cladding oxidation and hydriding and cladding creep which take into account the impact of the chromium coating on the overall cladding behaviour. Furthermore, the consumption of the chromium coating due to high-temperature diffusion of chromium into the cladding base alloy's β-Zr phase is simulated. The new models have been validated using measurements on chromium-coated cladding from irradiated rods, high-temperature annealing experiments and semi-integral LOCA tests. The validation showed good agreement between ENIGMA's predictions and the experimental data; thereby demonstrating the applicability of the new models for simulating the performance of LWR fuel rods with chromium-coated cladding in both normal operation and accident conditions. [ABSTRACT FROM AUTHOR]

• The capacity of the refractory metal interlayer to inhibit the diffusion of Cr and Zr elements was investigated. • Revealed the evolution process between layers in high temperature steam environment. • The mixed phase zone of refractory metals effectively prevents the diffusion of Cr and O elements into the Zr-4 alloys. Under the ATF design concept, Cr-based coating is the most promising zirconium alloy cladding protective coating in the case of PWR accidents. However, the primary factor affecting the oxidation resistance of Cr coatings is the interdiffusion of Cr and Zr. In this study, Cr coatings with three refractory metal interlayers of Nb, Mo, and Ta were deposited on zirconium alloy surfaces using closed-field unbalanced magnetron sputtering. The study investigated the effects of different interlayers on the microstructure, oxidation properties, and interface evolution process of Cr coatings under steam conditions at 1200 ℃, and compared them with Cr coatings without intermediate layers. The results indicate that the refractory metal interlayer influences the orientation and growth rate of Cr coating grains. The formation of a Laves phase mixed layer in the interlayer during oxidation effectively hinders the diffusion of O and Cr to the zirconium alloy matrix, enhancing its oxidation resistance. Furthermore, the diffusion rate and vacancy concentration of Mo and Ta elements to the Zr layer exceed that of diffusion to the Cr layer. During the cooling process, a precipitate phase forms in the Zr-4 matrix, while the Nb coating forms an infinite solid solution in the β-Zr. [ABSTRACT FROM AUTHOR]

A refined coating degradation model that comprehensively considers the phenomena of coating oxidation, diffusion, reduction, and volatilization is used to evaluate the performance of accident tolerant Cr-coated zirconium alloy cladding under accident conditions. The typical accident types of Station Blackout (SBO) and Large Break Loss of Coolant Accident (LBLOCA) are analyzed, with the consideration of three modes of coating failure: Failure in coating oxidation stage, Failure in oxide layer reduction stage, Failure caused by eutectic reaction. For the SBO accidents with relatively long time, the initial coating thickness of 9∼10 μm is recommended, which could provide good protection for zirconium alloy throughout the entire process and trigger the reduction of Cr 2 O 3 before reaching Cr–Zr eutectic temperature to inhibit the occurrence of eutectic reaction. For the LBLOCA accidents where the cladding is rapidly heated, eutectic reaction is the dominant factor leading to coating failure, so how to avoid eutectic reaction failure remains to be further studied. [ABSTRACT FROM AUTHOR]

The Cr/Zr inter-diffusion has been becoming a critical issue for Cr-coated Zr alloys as accident tolerant fuel (ATF) cladding. Here, ZrO 2 barrier layers against Cr/Zr inter-diffusion were successfully fabricated between the Cr layer and Zry4 substrate. Thermal decomposition of ZrO 2 is confirmed under high temperatures, and ZrO 2 is transformed into oxygen-stabilized α -Zr (O), which is effective for mitigating the Cr/Zr inter-diffusion. In-situ transmission electron microscopy (TEM) results indicate that the ZrO 2 thermal decomposition process becomes more obvious under 500 °C. In addition, Cr/ α -Zr (O) inter-diffusion leads to the thin Zr(Cr, Fe) 2 interface layer due to the high stability of the α -Zr (O) phase. • ZrO 2 barrier layers against Cr/Zr inter-diffusion were successfully fabricated for Cr-coated Zry4. • Thermal decomposition of ZrO 2 was confirmed under high temperature. • ZrO 2 was transformed into oxygen-stabilized α -Zr (O). • ZrO 2 thermal decomposition process became more obvious under 500 °C. • Cr/ α -Zr (O) inter-diffusion led to the thin Zr(Cr, Fe) 2 interface layer. [ABSTRACT FROM AUTHOR]

• Cr-Nb coating prepared on Zr alloys. • Nb coating can restrain the inward diffusion of cr. • Complete miscibility between Nb and Zr lead to the fast consumption of the Nb coating. • Early-stage microstructure evolution of the sample was analyzed. Although the Cr-coated Zr alloys are considered as the most promising accident tolerant fuel cladding materials in recent years, the Cr-Zr interdiffusion problem is a major obstacle for the widely used of the Cr coating. In this study, Cr-Nb coated Zr alloy samples were prepared by magnetron sputtering. The Nb layer was designed as a barrier layer to retard the Cr-Zr interdiffusion. By using a fast-in and fast-out equipment, short time annealing was conducted in argon environment to study the coating evolution in high temperature environment. Results show that the Nb coating can retard the inward diffusion of Cr because of the low diffusion coefficient of Cr in the Nb layer. However, the complete miscibility between Nb and Zr lead to the fast consumption of the Nb coating. After 10 min annealing at 1200 °C, the 2.5 μm Nb coating disappeared, and a Cr-Zr-Nb mixed phase formed between the residual Cr coating and Zr substrate. The early-stage microstructure evolution of the sample was analyzed, and the barrier effect of the Nb coating was summarized. [ABSTRACT FROM AUTHOR]

The demand for cleaner and more efficient new generation reactors has become increasingly urgent to solve the world’s energy supply and environmental issues such as carbon emissions. The Fukushima nuclear power plant disaster in 2011 prompted researchers to pay more attention to the safety performance of cladding tube materials in nuclear power plants under non-working conditions. Earlier, zirconium alloy, which was widely used in cladding tube materials, would cause serious accidents due to the production of explosive products after failure under the condition of beyond design-basis accident (BDBA). To avoid this problem, researchers proposed the design concept of accident tolerant fuel (ATF). ATF requires the new cladding material to retain a particular strength under the condition of BDBA and does not produce explosive products, thereby avoiding catastrophic accidents. Oxide dispersion strengthened (ODS)−FeCrAl alloy has good high-temperature strength due to its dispersion strengthening. After treatment, the presence of Al forms a thin and dense Al2O3 protective film on the surface of the alloy. This layer of Al2O3 protects the alloy, ensuring that it does not fail due to corrosion even when exposed to 1400 °C steam. This combination of high-temperature strength and corrosion resistance makes ODS−FeCrAl alloy a promising candidate for advanced reactor cladding materials like ATF. Although the introduction of aluminum improves the corrosion resistance of the alloy, it also changes the type of dispersed particles in the ODS alloy. The size of dispersed particles containing Al is usually larger than before, and their number density decrease. The state of dispersed particles in the alloy is closely related to the mechanical properties of the alloy. In this paper, the current research progress is summarized using relevant domestic and foreign documents considering the influence and control method of the microstructure of ODS−FeCrAl alloy due to the introduction of the Al element with the goal of serving as a reference for the forture development of ODS−FeCrAl alloy.

Uranium mononitride (UN) is a candidate fuel material for light water reactors with higher uranium (U) loading and thermal conductivity than uranium dioxide (UO2). However, the sintering of UN pellets is challenging as the UN powder particles oxidize rapidly at high temperatures unless the oxygen concentration is extremely low. Oxidation during sintering either reduces the relative density of the sintered UN pellet or disintegrates the sintered UN pellet to powder. To address this problem, the present work developed a rapid sintering method for producing highly densified UN surrogate pellets with minimal oxidation. Cerium nitride (CeN) is used as a surrogate for UN to reduce radiation hazards. With the custom-developed fast-heating system, the sintering process was completed within 150 s. The sintering atmosphere was flowing nitrogen (N2). The sintered CeN pellet density was 95% of the theoretical density (TD) or higher. The microstructure was uniform with a 10–25 µm grain size as demonstrated by scanning electron microscopy (SEM) and contained trivial levels of oxides as demonstrated by X-ray diffraction (XRD). The resultant pellets indicate that the rapid sintering method is a promising method to make UN fuel pellets with equivalent or higher density to pellets made by conventional sintering methods, while also being more efficient in time and costs. [ABSTRACT FROM AUTHOR]

Accident Tolerant Fuel (ATF) is a next-generation fuel technology developed to enhance the fundamental safety of light water reactors. Compared to existing standard fuel technologies, ATF technology can provide the same or better performance during normal operation, transient conditions, and accident scenarios of nuclear power plants, enhancing the ability of fuel elements to withstand severe accidents while reducing the cost of generating electricity over the licensed life of the plants. This paper introduces the definition, characteristics and thermal-hydraulic research objectives of ATF technology, systematically reviews the thermal-hydraulic research results of different ATF technologies in various aspects from the perspectives of near-term concepts (UO 2 -based composite fuels, coated Zr-based claddings, FeCrAl claddings) and long-term concepts (high-uranium-density UN fuels, FCM fuels, extruded metal fuels, SiC/SiC claddings), respectively. Mainly includes: improving fuel thermal conductivity, lowering fuel operating temperatures and stored energy, reducing fuel-cladding interactions, improving cladding-coolant reactions, reducing the production of flammable gases (e.g., hydrogen) and enhancing the retention of radioactive fission products. Finally, this paper points out the current shortcomings and challenges in the field of ATF thermal-hydraulic research and outlooks the potential future directions., Competing Interests: The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper., (© 2024 The Authors. Published by Elsevier Ltd.)

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]

• The thermal conductivity of a 12 μm-thick Cr coating deposited on a Zircaloy-4 substrate was measured in situ. • The effect of annealing-induced microstructural changes on the thermal conductivity of coating was investigated. • Coating conductivities before and after annealing are close to the bulk property. • The effect of grain boundaries is predicted to manifest when the critical dimension is on the order of 0.5 μm. The deployment of Cr coatings as accident tolerant fuel cladding inside nuclear reactors necessitates a thorough understanding of their performance, including mechanical, chemical, irradiation, and thermal properties. While significant research has delved into the former, the thermal transport properties of Cr coatings, which dictates fuel temperature and impacts fuel safety, have been relatively underexplored. This work presents an investigation into the thermal conductivity of 12 µm-thick Cr coatings fabricated by multi-arc ion plating employing high-resolution frequency and spatial domain thermoreflectance techniques. Vacuum annealing was utilized to control the morphology of coatings to explore the impact of grain size. The results indicate that both as-deposited and annealed coatings, despite possessing distinct grain morphologies, exhibit thermal conductivities similar to their bulk counterparts. The influence of boundaries on thermal transport is anticipated to become prominent as the critical dimension diminishes below 0.5 μm. Future investigation is required to address the potential impact of adverse reactions, such as oxygen ingress and irradiation, on the thermal properties of Cr coatings to ensure safe and efficient operation in nuclear reactor environments. [ABSTRACT FROM AUTHOR]

• Coating structure: 4 μm Zr coating + 9 μm Cr coating. • A Cr/Zr mixed oxide layer formed between the two coatings. • Weight gain rate was reduced by the outer Zr coating. • No cracking or exfoliation was observed on the coating. Cr-coated Zr alloys are widely regarded as promising accident-tolerant fuel (ATF) cladding materials. However, the rapid consumption of the Cr coating through both oxidation and diffusion demands a thick coating, which is detrimental to neutron economic. In this study, a thin Zr coating was fabricated on the top of the Cr coating to act as a protective layer, aiming to reduce the oxidation consumption rate of Cr coatings. In-situ weight gain measurements were performed to determine the oxidation kinetics. The microstructural evolution of the coating/substrate interface and the oxide/coating interface were analyzed. The results show that the weight gain rate of the Zr-Cr coated Zr alloy sample is lower than that of the bare Cr-coated Zr alloy samples. An equiaxed grain structure composed of Cr/Zr mixed oxides was observed in the coating, which helps to inhibit elements diffusion and reduce the oxidation rate of Cr coatings. After steam oxidation at 1000 °C for 4 h, the sample structure consisted of the Zr substrate, a Cr 2 Zr layer, a residual Cr coating, a Cr 2 O 3 layer, a Cr/Zr mixed oxides layer, and a ZrO 2 layer. [ABSTRACT FROM AUTHOR]

Uranium dioxide (UO 2) is the standard fuel used in light water reactors (LWRs). However, it has a low thermal conductivity that ultimately limits its performance both during normal operation and in accident conditions. Adding a material with high thermal conductivity is a potential approach to enhance the thermal conductivity of UO 2. Forming an interconnected structure of high-conductivity material can significantly enhance the overall thermal conductivity of the composite. Molybdenum (Mo) has been used as an additive material in UO 2 composites previously. A new method for the fabrication of interconnected UO 2− Mo composites using pre-sintered UO 2 granules to improve the continuity of Mo channels was investigated in this study. UO 2 –10 wt% Mo composites were produced using UO 2 granules and 1073 K and 1473 K pre-sintered UO 2 granules, followed by spark plasma sintering (SPS) of the mixtures at 1473 K for 5 min. The composites were characterised using scanning electron microscopy and X-ray diffractometry and their thermal conductivities were measured by the laser flash method and compared with a reference UO 2 pellet. At a maximum measurement temperature of 1073 K, a 52 % increase in thermal conductivity was observed in the composites containing UO 2 without pre-sintering, and UO 2 pre-sintered at 1073 K. The increase was 31 % for composites manufactured from UO 2 pre-sintered at 1473 K. These results suggest that higher temperature pre-sintering may be detrimental to forming interconnected Mo structures. [ABSTRACT FROM AUTHOR]

CyMe4-BTPhen and CyMe4-BTBP are the principal ligand systems used in Europe for the separation of actinides from lanthanides as a part of the SANEX process for nuclear recycling and reprocessing. We present a new approach to the synthesis of the CyMe4 fragment beginning from readily available hydroxypivalic acid. It features a cobalt-catalysed homocoupling of a neopentyl bromide to provide the key bisester precursor, thereby avoiding the requirement for technically challenging low-temperature LDA-mediated aldol chemistries.

Solid-state diffusion bonding between pure Cr and Zircaloy-4 (Zry4) was studied with and without 316 stainless steel (SS) as an interlayer at temperatures ranging from 1203 to 1573 K, followed by the detailed microstructural analysis and mechanical property evaluation at the bonding interfaces. For the samples directly bonded between Cr and Zry4, two regions were confirmed at the diffusion interface. One was a continuous block of Zr2Cr intermetallic compound formed by the inter-diffusion of Cr and Zr, and the other was a mixture of Zr2Cr and Zrα phases as the consequence of Zr-Cr eutectic reaction. For the samples with SS as an interlayer, three regions were mainly observed at the SS-Zry4 interface, consisting of (FeCr)α-, Zr(Fe, Cr)2-, and Zr2(Fe, Ni)-phase dominated layers, with (FeCr)α-phase observed at Cr-SS interface. The microstructure of the phases at diffusion interfaces were closely related to the bonding temperature. Nano-indentation results confirmed a rapid increase in hardness at Cr-Zry4 and SS-Zry4 interfaces for Cr-Zry4 and Cr-SS-Zry4 bonded specimens, respectively, due to formation of the ZrCr2/Zr(Fe, Cr)2 phase. The presence of the hard brittle intermetallic compounds at the interface infers the possible brittle fracture risk of the Cr-Zry4 interface during long-term operation under normal condition and high temperature under accident conditions.

XPS combined with AFM have been used to investigate the surface modifications of nuclear grade FeCrAl-xNb (x= 0, 0.5, 1.0, 1.5 wt.%) alloys, induced by anodic polarization in a simulated primary coolant. Electrochemical tests were performed to study the role of the minor element Nb on the corrosion potential and composition. The optimal content of Nb, obtained after the comparative analysis of the UHV-sputtered surface, native oxide-covered surface and anodic polarized surface, is 0.5 wt.%. This alloy presents a more compact Al-enriched oxide film consistent with a higher corrosion potential at −89 mV. The protectiveness of oxide film against the active dissolution can be improved by Nb addition, but does not increases linearly with Nb content. High resolution XPS analysis details the changes of Nb oxidized states, and provides the stratification of the oxide film containing the (hydr) oxides of Fe, Al and Cr.

Reflood plays an important role in preventing the serious accident process. In this paper, the reflood oxidation performances of Cr-coated Zr-Sn-Nb alloy cladding tubes at 1000 °C, 1100 °C and 1200 °C were studied by high-temperature steam oxidation followed by in-situ water quenching. The oxidation kinetics, macroscopic morphology, phase transformation and microstructural evolution were investigated. Cr coating can provide excellent anti-reflood oxidation protection for Zr-Sn-Nb alloy, resulting in an oxidation rate constant of only 1/4–1/7 of that of uncoated Zr-Sn-Nb alloys. However, the cladding tube degraded into fragments due to excessive thermal stress during the in-situ water quenching after reflood oxidation at 1200 °C for 4000 s. After reflood oxidation at 1000 °C and 1100 °C, the most prominent feature of the surface is a porous flocculation structure main due to the formation of volatile products, while after 1200 °C, it is a fine mackerel scale-like structure accompanied by protruding whiskers main owing to Cr3+ rapidly migrates outward to the outer surface via short circuit paths. After oxidation, the cross-section of Cr-coated Zr-Sn-Nb alloys exhibits a multi-layer structure. The thickness of each layer changes approximately linearly with oxidation time. However, due to the redox reaction, compared to that after oxidation for 2000s, the Cr 2 O 3 layer becomes thinner and the residual Cr coating becomes thicker after oxidation at 1200 °C for 4000 s. [ABSTRACT FROM AUTHOR]

The increasing demand for nuclear reactors as a source of clean energy forced nuclear researchers to search for ways to enhance the safety operation of nuclear reactors. This paper discusses the viability of using different thorium compounds ((233U, Th)O 2 , (235U, Th)O 2 , (RGPu, Th)O 2) as an accident tolerant fuel/advanced technology fuel (ATF) for Pebble Bed Modular Reactor 400 (PBMR-400). A comprehensive analysis of thorium-based fuel performance was carried out and compared with UO 2 to develop a novel nuclear fuel with the capability to tolerate severe accident conditions. The neutronic characteristics such as the infinity multiplication factor, initial heavy metal concentration, RGPu concentration and fission products concentration were analyzed. In addition to, the main safety parameters have been calculated to distinguish the effect of the proposed fuels on them. The thermal spectrum and thermal output power have been visualized to give a good insight view inside the core about the distribution of the heat. In general, the neutronic analysis demonstrated good behavior of the examined thorium-based fuel in the PBMR-400 and especially in the case of (RGPu, Th)O 2. PBMR-400 showed excellent performance in incinerating a large amount of reactor-grade plutonium when it was used in a kernel fuel bed. [ABSTRACT FROM AUTHOR]

Accident Tolerant Fuels, also known as Advanced Technology Fuels (ATFs), have the potential to improve the safety and economic performance of commercial nuclear reactors as well as to provide additional operational flexibility for responding to variations in system energy demand. The design, fabrication and in-pile behaviour of currently operating and innovative nuclear fuels to enable a reliable and safe operation of nuclear power plants is a priority for the nuclear industry. The Spanish nuclear professionals (public and private companies, academia, and research centres) have achieved significant advances in this area. Both irradiation programs in commercial reactors, along with experimental and modelling activities, were focused into comprehending and modelling different ATF concepts to understand and predict the behaviour under normal and off-normal conditions. The International Atomic Energy Agency (IAEA) sub-program "Nuclear Power Reactor Fuel Engineering" is strongly involved in the development and validation of computer codes for ATF fuel design and the valuable Spanish contribution to this sub-program is described in this article. The future Spanish R&D in this field is expected to be focused on further improvements on ATF product development and performance knowledge. [ABSTRACT FROM AUTHOR]

• Material model for FeCrAl implemented in MELCOR-2.2 with increased detail. • Recovered PWR LOCA scenario analyzed for high burnup ATF and reference zircaloy case. • For scenario analyzed, additional coping time from FeCrAl is important. • Increased burnup increases fission product inventory. • Lower fission product release with FeCrAl compared to zircaloy maintained at high burnup. It has been established that incremental improvements in beyond design basis accident performance can be achieved through accident tolerant fuel (ATF). However, they have the potential to recover margin with respect to conventional fuel and therefore enhance plant economics through uprate or cycle length increase. To realize this potential, it is necessary to quantify the reduction in source term due to use of ATF, and correspondingly how this is affected by increasing the cycle length and/or burnup. This requires development of a risk informed analysis methodology for ATF under high burnup conditions, which is being developed within LWRS. To this end, a MELCOR simulation of FeCrAl ATF in a recovered Large Break LOCA (LBLOCA) scenario has been developed, using a model based on the Zion Pressurized Water Reactor (PWR). The new user defined material capability, along with the inclusion of detailed neutronics- and depletion-derived parameters such as core power distribution, decay heat behavior, and fission product inventories, allows a more detailed simulation of FeCrAl clad material properties and behavior than has previously been possible using MELCOR. These detailed FeCrAl results were compared with a zircaloy clad model to investigate the differences between the two clad types and quantify the benefits of FeCrAl ATF with an 18-month cycle. Next, the fuel cycle with ATF was extended to 24 months, to determine whether any additional safety margin provided by FeCrAl ATF could be leveraged to implement high burnup FeCrAl-clad fuel while retaining the same operating and safety limits as current zircaloy-clad fuel. For the particular scenario analyzed, the delay to fuel failure from using ATF was of the same order as the LOCA recovery time, and hence significant in reducing fission product release, with the 24-month FeCrAl core performing better than the 18-month zircaloy core. It is noted that for other transients, the reduction in release due to using FeCrAl may be less significant. The material model developed here can be used in such further studies in support of determining the overall source term. [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]

This article discusses the methods of improving the corrosion resistance of fuel rod claddings from E110 zirconium alloy performed at TRINITI State Research Center using coatings (Al, Al2O3, Cr) applied by pulsed laser deposition as well as magnetron sputtering and galvanic deposition of chromium. Some results are presented regarding studies of adhesion, microstructure, and corrosion resistance of chromium coatings deposited by various methods. [ABSTRACT FROM AUTHOR]

Silicon carbide composites are expected for light water reactors. The objective is to understand the steam oxidation behavior and the high-temperature water corrosion behavior of the liquid phase sintering silicon carbide and to develop the liquid phase sintering silicon carbide composites, which are stable at the high-temperature water conditions in normal operation and the high-temperature steam conditions in a severe accident. The steam oxidation experiments were carried out at 1200 and 1400 °C. The high-temperature water corrosion experiments were carried out at 320 and 360 °C. The formation of the silicate, which is expected to have excellent resistance to the steam, was confirmed following the steam exposure at 1400 °C. High-temperature water corrosion resistance was improved by the formation of Yttrium Aluminum Garnet at the grain boundary. The particle-dispersion silicon carbide composite tubes with the modified condition were developed, and the thermal shock experiments from 1200 °C to ambient temperature were carried out. The composite tubes showed excellent oxidation and thermal shock resistance. The particle-dispersion liquid phase sintering silicon carbide composites with the modified condition are promising materials for light water reactors. [ABSTRACT FROM AUTHOR]