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Metallic micro-cell UO 2 pellet with high thermal conductivity has received attention as a promising accident-tolerant fuel. Although experimental demonstrations have been successful, studies on the potency of current metallic micro-cell UO 2 fuels for further enhancement of thermal performance are lacking. Here, we numerically investigated the thermal conductivities of micro-cell UO 2 Mo pellets in terms of the amount of Mo content, the unit cell size, and the aspect ratio of the micro-cells. The results showed good agreement with experimental measurements, and more importantly, indicated the importance of optimizing the unit cell geometries of the micro-cell pellets for greater increases in thermal conductivity. Consequently, the micro-cell UO 2 Mo pellets (5 vol% Mo) with modified geometries increased the thermal conductivity of the current UO 2 pellets by about 2.5 times, and lowered the temperature gradient within the pellets by 62.9% under a linear heat generation rate of 200 W/cm. [ABSTRACT FROM AUTHOR]

The effect of aluminum in the FeCr(Al) alloys on the corrosion behavior was studied at two temperatures (400 °C and 1200 °C). For the Fe-21Cr alloy, Cr 2 O 3 layer was formed upon exposure to both temperatures. The addition of Al altered the corrosion mechanism of the alloy. At 1200 °C, a thin α -Al 2 O 3 is formed on the Fe-21Cr-5.5Al alloy. Whereas, at low temperature, a continuous Cr 2 O 3 is formed at the top with an internal Al 2 O 3 layer. These results show that FeCrAl alloys hold promise as Fuel Cladding Materials with Al playing a crucial role under operational as well as accident-like conditions. • Effect of Al on steam corrosion of FeCr(Al) alloys was studied at 2 temperatures. • At 1200 °C, thick Cr 2 O 3 oxide layer formed on the Fe21Cr alloy • Whereas thin Al 2 O 3 layer formed on the Fe21Cr5.5Al alloy • At 400 °C, Fe21Cr5.5Al formed Cr 2 O 3 on the surface with an internal Al 2 O 3 layer. • Presence of Al improves the corrosion resistance as compared to the Fe21Cr alloy. [ABSTRACT FROM AUTHOR]

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

The influence of chromium coating on hydrogenation behavior of laser beam welds (LBW) of E110 zirconium alloy was studied at 360–900 °C and hydrogen pressure of 2 bar. The deposition of Cr coating can slow down hydrogen absorption rate of the LBW samples as activation energy for hydrogen absorption (70 kJ/mol) is higher compared to uncoated ones (61 kJ/mol). The growth of hydrides depends on the hydrogenation conditions and original microstructure of welded alloy in different zones. Nanoindentation and three-point bending tests show that hydride precipitations can cause hardening of LBW E110 alloy samples and loss of its plasticity. Some additional aspects of applying Cr coatings for protection of LBW Zr alloy in hydrogen atmosphere are presented. [ABSTRACT FROM AUTHOR]

• An annealed, 7 vol% Zr: U 3 Si 2 sample displayed an onset of oxidation in steam at T > 500°C- a higher temperature than any arc-melt fabricated silicides tested to date. • A unique microstructure is observed in a 2 vol% Zr addition to U 3 Si 2 , hypothesized here to be caused by strain induced from the incorporation of Zr into the U 3 Si 2 lattice. • Direct comparison was made between the simulated XRD patterns of the computational model system and the measured of the 14 vol%Zr sample, suggesting Zr-inclusion into uranium silicide to occur substitutionally. Research and development of high uranium density fuels supports the on-going effort to enhance reactor technology, safety, performance, and economics. Among high uranium density fuels, U 3 Si 2 is a favored fuel candidate due to its higher uranium density and improved thermal conductivity when compared to traditional UO 2. While U 3 Si 2 demonstrates an improved fuel economy, it suffers from rapid degradation when exposed to water at T > 300°C. In order for U 3 Si 2 to be implemented in water cooled reactors, the water reaction must be understood and mitigated. Reported here is the performance of Zr alloying additions to U 3 Si 2 exposed to high temperature steam oxidation environments. The investigation shows that homogenization heat treatment contributes to a delayed onset of oxidation and an improved homogeneity throughout the arc-melt fabricated alloyed compositions. Of the pre- and post-annealed samples, the post-annealed 7 vol% Zr addition exhibited the highest onset temperature of oxidation. A preliminary computational investigation to complement the experimental analysis assesses the solubility of Zr in U 3 Si 2 and its incorporation in the crystal lattice. [ABSTRACT FROM AUTHOR]

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

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

Neutronics and fuel performance analysis is done for enhanced accident tolerance fuel (ATF), with the Monte Carlo reactor physics code Serpent and INL’s fuel performance code BISON. The purpose is to evaluate the most promising ATF candidate material FeCrAl, which has excellent oxidation resistance, as fuel cladding under normal operational conditions. Due to several major disadvantages of FeCrAl coating, such as difficulty in fabrication, diametrical compression from reactor pressurization, coating spallation and inter diffusion with zirconium, a monolithic FeCrAl cladding design is suggested. To overcome the neutron penalty expected when using FeCrAl as cladding for current oxide fuel, an optimized FeCrAl cladding design from a detailed parametric study in literature is adopted, which suggests reducing the cladding thickness and slightly increasing the fuel enrichment. A neutronics analysis is done that implementing this FeCrAl cladding design in a Pressurized Water Reactor (PWR) single assembly. The results show that the PWR cycle length requirements will be matched, with a slight increase in total plutonium production. Fuel performance analysis with BISON code is carried out to investigate the effects with this FeCrAl cladding design. The results demonstrate that the application of FeCrAl cladding could improve performance. For example, the axial temperature profile is flattened. The gap closure is significantly delayed, which means the pellet cladding mechanical interaction is greatly delayed. The disadvantages for monolithic FeCrAl cladding are that: (1) fission gas release is increased; and (2) fuel temperature is increased, but the increase is less than 50 K even at high burnup. The better strength, corrosion, and embrittlement properties of FeCrAl enable the fabrication of FeCrAl cladding with thinner walls. FeCrAl cladding proves to be a good alternate for zircaloy cladding, given the advantages and insignificant disadvantages shown by fuel performance analysis. [ABSTRACT FROM AUTHOR]

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

The isothermal steam oxidation behavior of chromium-coated zirconium alloy cladding at 1200 ℃ is studied. The cladding substrate is Zr-1 Nb alloy with outer diameter of 9.5 mm and thickness of 0.57 mm. The Cr coating is prepared by magnetron sputtering method with thickness of about 11 µm. The oxidation durations are 300 s~14,400 s. After oxidation, blisters are formed on the surface due to the compressive stress in coating, and the coating surfaces are porous or exhibit a morphology with ravines and granular crystals. A four-layer structure (Cr 2 O 3 , Cr, Zr(Cr,Fe) 2 and Zr from outside to inside) is formed in the early stage of oxidation. However, Zr4+ can migrate into Cr coating through grain boundaries and reacts with Cr 2 O 3. The redox reaction between Zr and Cr 2 O 3 leads to the coating failure, resulting in the oxidation of Zr alloy substrate. Due to selective oxidation, a certain thickness of Cr coating is retained during the thickening of outer ZrO 2. The emergence of outer ZrO 2 layer is significantly delayed by Cr coating, and Cr coating is able to reduce the growth rate of outer ZrO 2 layer. Therefore, Cr coating has a protective effect in 1,200 ℃ steam environment and can improve the oxidation resistance of Zr alloy cladding. • Isothermal steam oxidation behavior of magnetron-sputtered Cr-coated Zr alloy at 1200 ℃ is studied. • The oxidation and failure mechanism of Cr coating is discussed. • The oxidation kinetics of Cr-coated Zr alloy cladding is analyzed. [ABSTRACT FROM AUTHOR]

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

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

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

[Display omitted] • 0.5Nb provides the best protection of oxide film in simulated primary coolant. • Corrosion potential does not increase with Nb. • Nb promotes the active dissolution after anodic polarization. • Cr hydroxylation is enhanced by Nb addition. XPS combined with AFM have been used to investigate the surface modifications of nuclear grade FeCrAl- x Nb (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. [ABSTRACT FROM AUTHOR]

Silicon carbide (SiC) cladding is being studied for its potential application in light water reactors for improved accident tolerance. SiC-SiC composite material undergoes irradiation-induced, temperature-dependent swelling. The circumferentially non-uniform neutron flux and power generation in the fuel rod, will produce circumferentially non-uniform swelling and thermal expansion, ultimately leading to bending of the cladding. This bending will cause interaction between the cladding and the spacer grids and can be detrimental to the structural integrity of both the cladding and spacer grids. In this study we have analyzed the bending behavior of the cladding, the stress development in the cladding and the effects of spacer grids. The sensitivity of the cladding stress to the spacer grid material, extent of circumferential variation in power have been also evaluated. The results indicate that the spacer grids do not significantly affect the cladding stresses. The stress generated in the spacer grids due to its interaction with the cladding is found to be well below the strength of the spacer grid material. The sensitivity studies showed that changing the spacer grid material have negligible impact on the stresses in both cladding and spacer grid, while the extent of circumferential variation in power can lead to significant increase in the stresses in spacer grid. [ABSTRACT FROM AUTHOR]

Abstract: The Fukushima accident in March 2011 caused by the massive earthquake and tsunami led to hydrogen explosion, core meltdown, and the subsequent release of huge radioactivity both into the atmosphere and the Pacific Ocean. In the case of volatile fission products such as 137Cs and 131I, the release fraction of the core inventory of the units 1–3 into the atmosphere is estimated to be 1.2–6.6% and 1.1–7.9%, respectively. As for gaseous fission product 133Xe, it is estimated that nearly 100% of the core inventory might have been released into the atmosphere. In addition, about 16% of the 137Cs inventory flowed into the sea when the contaminated water used for cooling the decay heat of the units 1–3 overflowed the reactors. Therefore, even though almost three years have passed since the accident, it is still having a tremendous impact not only on Japan but all over the world as well. This paper reviews the Fukushima accident from the viewpoint of radioactivity release and dispersion in the environment and its effect on public health, economy, energy policy, international relationship, and LWR fuel development. [Copyright &y& Elsevier]

• Silicon carbide (SiC) and Zirconium (Zr) are being introduced as enhanced additives to the UO 2 single crystals (UO 2 SCs) pellet respectively, which are supposed to imply an important application in future ATF. • The reactions of UO 2 SCs-SiC and UO 2 SCs-Zr fuel in a water environment was investigated to evaluate the fuel's stability under accident situations. • The results reveal the SiC additives facilitate pellet showed good oxidation resistance in water at 1200 °C, while the Zr additives pellet is completely oxidized to ZrO 2 powder and UO 2 particles. Uranium dioxide (UO 2) is currently the most widely used nuclear fuel for commercial nuclear reactors. However, the Fukushima Daiichi nuclear disaster revealed the primary safety risks of this kind of fuel under accident conditions, which thereby stimulated various international programs in developing accident tolerant fuel (ATF). Silicon carbide (SiC) and Zirconium (Zr) are being introduced as additives to enhance thermal properties and fission product retention ability of the UO 2 single crystals (SCs) pellet. The oxidation resistance in water at high temperatures is essential to the deployment of ATF. In this work, two kinds of enhanced UO 2 single crystal pellets by using SiC and Zr particles as additives (denoted as UO 2 SCs-SiC and UO 2 SCs-Zr, respectively) were fabricated through spark plasma sintering (SPS) processes. The impact of steam oxidation on UO 2 SCs-SiC and UO 2 SCs-Zr was investigated through the microstructures, phase compositions and post oxidation performances characterizations. The results revealed that the pellets with SiC as additives showed good oxidation resistance in water even at 1200 °C while that with Zr as additives showed complete oxidization which resulted in the formation ZrO 2 powder. [ABSTRACT FROM AUTHOR]

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

Highlights: [•] Validated SERPENT–PARCS for analysis of fully ceramic microencapsulated (FCM) fuel. [•] Modified PARCS to incorporate UN and SiC thermo-physical properties. [•] Core kinetics simulation of an RIA for a PWR with conventional UOX or FCM fuel. [•] The results for the core power transient are similar for UOX or FCM fuel. [•] Sensitivity of the RIA response to FCM thermal model was quantified. [Copyright &y& Elsevier]

Highlights: [•] Fully ceramic microencapsulated (FCM) fuel was studied. [•] Sensitivity and benchmark calculations were performed on lattice level for FCM fuel. [•] A non-linear reactivity model was vital to represent the reactivity during burn-up. [•] Reactivity coefficients were determined for FCM versus conventional UO X fuel. [Copyright &y& Elsevier]

Cr coatings with CrN/Cr multilayers with a layer thickness of 100, 250 and 750 nm were deposited by a multi-cathode magnetron sputtering on E110 zirconium alloy in Ar and Ar + N 2 atmosphere. The mechanical properties of the as-deposited coatings were measured by scratch testing and tribometry. A comparative study of crack and oxidation resistance of the Cr-coated E110 alloy was performed under thermal cycling (1000 °С) and high-temperature (1200–1400 °С) steam oxidation conditions. The influence of the CrN/Cr multilayers and their thickness on the mechanical properties and oxidation resistance of the coated E110 alloy is discussed. It is found that the CrN/Cr multilayers can limit interdiffusion between the coating and E110 alloy at high temperatures only for a short period. • Increase in wear resistance of Cr-coated Zr alloy • CrN/Cr multilayers use as a barrier layer for Cr Zr interdiffusion • Role of thickness layer of CrN/Cr multilayers on cracking resistance • Nitrogen-stabilization of α-Zr phase underneath Cr coating • Deceleration of Cr Zr interdiffusion due to α-Zr(N) layer [ABSTRACT FROM AUTHOR]

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

The present study addresses the lack of steam oxidation kinetic data for high purity uranium mononitride (UN), assesses the sensitivity of the dynamic response of UN to the density of the monolithic sample exposed to flowing steam, as well as observes the variation in the oxidation kinetic response of UN to the presence of a common fabrication contaminant, carbon. In addition, the mass spectrometry performed on the evolved gas during the reaction with steam informs the reactions thermochemistry for UN with flowing steam during a thermal ramp and during an isothermal exposure at 500 °C. It is concluded here that high purity, high density samples, (≥94% theoretical density), have a significantly delayed onset of reaction with steam compared to values previously reported, the kinetic reaction is retarded in a reducing steam atmosphere (steam + 100 ppm H 2), and the reaction products are predominantly ammonia (NH 3 gas) and UO 2 at temperatures less than 700 °C, above which UO 2+ x (0.2 ≥ x ≥ 0.0) begins to form and NH 3 (g) dissociates. [ABSTRACT FROM AUTHOR]

The Fukushima-Daiichi accident revealed that the zirconium fuel claddings have the significant safety risk of hydrogen detonation due to the strong oxidation and hydrogen release during the design basis accidents (DBA) and beyond design basis accidents (BDBA). Therefore, research and development of accident tolerant fuel (ATF) concepts that aim to improve nuclear fuel safety during normal operation, operational transients and possible accident scenarios have been boosted in the last decade. Deposition of protective coatings on Zircaloy cladding tubes has been considered as a near-term solution of enhanced ATF cladding. Among the candidate coating materials, there is no doubt that the research progress of Cr coating is the fastest around the world because of the advantages of such type of coating: excellent good chemical stability (including oxidation resistance and hydrothermal corrosion resistance), low thermal neutrons absorption cross-section, and excellent adherent. In this paper, the oxidation, diffusion, and mechanical properties of Cr-coated Zr alloys in normal operation conditions and accident conditions of nuclear reactors are reviewed. The factors that cause the failure of the coating are analyzed, and some questions that need to be clarified and further studied are proposed. • Focus on Cr-coated Zr alloy ATF cladding. • Oxidation and diffusion behavior reviewed. • Mechanical property of the coating discussed. • Failure mechanism analyzed. • Questions to be solved in the future proposed. [ABSTRACT FROM AUTHOR]

Uranium dioxide has been the primary fuel type used in light water reactors for more than 40 years and proven to be reliable and robust. However, the Fukushima-Daiichi nuclear accident has motivated new work evaluating fuels with characteristics promoting accident tolerance, including enhanced thermal conductivity. Recently, additives have been investigated to increase thermal conductivity, but research has been largely focused on non-fissile additions. This study investigated the use of fissile additives to not only increase the thermal conductivity but also increase the uranium loading. Uranium diboride was chosen as the additive for this study due to its promising corrosion behavior as well as its significantly higher thermal conductivity at 573 K (25 Wm−1K−1) when compared to UO 2 (7 Wm−1K−1). Uranium diboride powder was fabricated via the arc melting technique and a ball milling process prior to mixing with UO 2 in a 90/10 wt% UO 2 /UB 2 ratio. Green bodies were made using a uniaxial die and subjected to a traditional pressureless sintering technique at 2073 K in argon. Sintered samples were analyzed via laser flash analysis for thermal diffusivity and differential scanning calorimetry for specific heat capacity in order to calculate thermal conductivity. The samples displayed an increase of 36–55% in thermal conductivity between 323 and 1273 K when compared to the benchmark samples (pure UO 2) as reported in open literature. [ABSTRACT FROM AUTHOR]

FeCrAl alloys are candidate materials for manufacturing accident-tolerant fuel (ATF) cladding intended for light water reactors to increase fuel reliability and safety during design-basis and beyond-design-basis accident scenarios. The evolution of the materials' surface characteristics, microstructure, and mechanical properties when exposed to the Critical Heat Flux (CHF) in flow boiling testing is crucial for safety analysis while providing insights into their thermal-hydraulic performance in nuclear reactors. After CHF, the surface chemistry of two FeCrAl alloys, APMT and C26M, was studied to understand their evolution at the early stage of high-temperature excursions in short time periods. The results indicated a thin layer composed of oxides and hydroxides of Al, Cr, and Fe with varying proportions at different depths in the layer, as indicated by X-ray photoelectron spectroscopy (XPS) and depth profiling. The cross-sections prepared by focused ion beam (FIB) revealed the growth of an oxide layer, in the range of 90-180 nm thick, on the alloys' surfaces. The evolution of the materials' surface chemistry also led to a noticeable post CHF excursion increase in their wettability, with a slight increase in roughness. The investigation of the materials' mechanical properties indicated a modest increase in hardness by 10-15% as well as an increase in their yield strength, as evidenced by the microindentation and ring compression tests conducted before and after CHF testing. Scanning electron Microscopy (SEM) and X-ray diffraction (XRD) were used to investigate microstructural features of the materials and their changes after CHF treatment. [ABSTRACT FROM AUTHOR]

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

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