The lead-magnesium eutectic (Pb-Mg, 97.5 wt% Pb-2.5 wt% Mg) has been identified as a promising coolant for next-generation nuclear systems. The compatibility between structural materials and candidate coolants is a critical feasibility factor for advanced nuclear systems. Corrosion studies were conducted on 316L steel in Pb-Mg at 450 degrees C and 550 degrees C for exposure durations up to 2000 h. The 316L steel exhibited significant dissolution attack, primarily due to selective leaching of nickel from the matrix. The dissolution mechanism is fundamentally identical at both temperatures, exhibiting the same linear kinetics in the early stages. Due to the increase in corrosion depth, the corrosion rate gradually decreases over time. The Ni dissolution rate constant in Pb-Mg is calculated and an Arrhenius relationship with temperature is established. This study elucidates the corrosion behavior and mechanism of 316L steel in Pb-Mg, providing valuable insights for its application in advanced nuclear energy systems.

SiC-TiC (20 vol%) composites were irradiated with single Si and sequential He+Si ions. Microstructural evolution and nano-hardness alteration caused by irradiation were investigated. Grazing incidence X-ray diffraction (GIXRD) analysis showed peak shift and broadening after both single Si and sequential He+Si irradiation. Raman spectra showed a decrease in intensity and broadening of Si-C peaks and emergence of Si-Si and C-C bonds after irradiation, indicating irradiation induced formation of homonuclear bonds within the SiC network. Transmission electron microscopy (TEM) observations demonstrated that the He-ions irradiation at room temperature caused amorphization of SiC. In contrast, TiC retained crystalline structure with formation of black dot defects and clusters. Furthermore, subsequent Si-ions irradiation resulted in bubble growth in SiC but not in TiC. Nanoindentation measurement revealed that both single Si and sequential He+Si irradiations resulted in hardening of SiC-TiC composites, with a higher hardening degree observed in the latter case. The present work could provide an insight for evaluating the practicability of ceramic composites as nuclear materials.

Chromium-coated zircaloy is one of the most promising candidate materials to enhance the accident tolerance of nuclear fuel. By introducing a large number of yttrium oxide nanoparticles into the chromium coating (i.e., ODS-Cr coating), its thermal stability and radiation tolerance are remarkably improved. However, the effect of Y addition on its oxidation behavior needs to be urgently addressed. Here, the Xe-ions irradiated ODS-Cr coatings with different Y content were exposed to 1200 degrees C steam. Microstructural examinations show that, compared to pure Cr (which formed a similar to 2 mu m thick oxide scale), Y-doped coatings exhibited smaller size needle-shaped Cr2O3 blades and a substantially thinner oxide scale of 0.2-0.6 mu m, alongside fewer voids at the scale-coating interface. Additionally, the irradiation-induced cavities displayed much higher thermal stability. The improved oxidation resistance of ODS-Cr coatings indicates that the outward Cr diffusion at high temperatures was strictly suppressed by nano-oxides segregated at grain boundaries. These findings reveal the critical role of Y in tuning both the oxidation and radiation resistance of Cr-based protective coatings, providing valuable insights for the development of accident-tolerant fuel.

SIMP steel, a reduced activation ferritic/martensitic heat–resistant steel, is a promising candidate for structural applications in future nuclear fusion reactors. Understanding the evolution of microstructure in SIMP steel is of practical importance for improving the structural integrity of components made from this material. In this work, we study the effects of cooling rate on the evolution of microstructures in the SIMP steel, including martensitic transformation and indentation deformation. At low cooling rates, the martensitic transformation exhibits characteristics of multi–step transformation, while these characteristics disappear at high cooling rates. A modified Koistinen-Marburger model was introduced to analyze the temporal evolution of martensite in the SIMP steel at stage IV under three different cooling rates (0.5, 1, and 5  K/s). The characteristic undercooling temperature at stage IV follows a power-law dependence of the cooling rate with a power index of 0.17. The dependence of indentation hardness on the characteristic block width exhibits two distinct regimes: a Hall-Petch-like relationship for block widths less than or equal to 4.1  μm, and a possible inverse Hall-Petch-like relationship for block widths larger than or equal to 4.1  μm.

In the design of the spallation target of CiADS, T91 steel was selected for the most critical component, the beam window. A Si-modified martensitic steel (SIMP) has been currently developed, which has better corrosion resistance and is planning to replace T91 as a material of beam window in the future. In order to assess the effect of irradiation damage on the microstructure evolution and corrosion behavior of the SIMP, the pre-irradiated samples with irradiation doses from 5 x 10(15) to 3 x 10(17) He/cm(2) exposed in static lead-bismuth eutectic (LBE) with saturated oxygen at 350 degrees C for 4000 h were investigated. Results show that pre-irradiation neither change the double-layer structure of the oxide layer nor significantly affect the corrosion rate. This may be due to the slow diffusion of elements at low temperatures, and the low irradiation dose not reaching the threshold for triggering accelerated corrosion.

Ceramic composites with high temperature strengths and low neutron cross-sections are promising candidates for core materials in advanced nuclear systems. In present work, SiC-20 vol% ZrC composites were irradiated with 500 keV He-ions at 25, 500 and 800 degrees C to evaluate the effect of irradiation temperature on the structural damage and bubble evolution in ceramic composites. XRD and Raman spectra analysis give that the irradiation resulted in structural damages of both SiC and ZrC. TEM observations reveal the formation of helium bubbles and defect clusters after irradiation. Moreover, the occurrence of micro-cracks in ZrC grains and amorphization of SiC are observed for the samples irradiated at room temperature. Nanoindentation test showed that there is irradiation induced hardening or softening of the composites which depends on the irradiation temperature or fluence. The correlation between microstructural evolution and mechanical properties response is discussed.

Lead-cooled fast reactors exhibit strong inherent safety performance and good economic features, while material degradation due to corrosion and irradiation is still challenging. Oxide dispersion-strengthened steels are one of the promising candidates for fuel cladding materials. The effects of both irradiation and corrosion on ODS steel need to be further studied. In this work, MX-ODS steel was irradiated by Fe ions at 500 °C up to 46 dpa. Later, the as-received specimen and the irradiated specimen were used to conduct corrosion tests in oxygen-saturated Pb at 550 °C for 1 h. In the as-received specimen, discontinuous oxides penetrated by Pb and Pb in contact with steel matrix were observed, demonstrating unsatisfactory corrosion resistance of the material. However, in the irradiated specimen after corrosion experiment, a protective oxide layer formed and prevented Pb attack. The oxidation behavior differences between the two specimens can be attributed to the defects produced by irradiation and the structural discrepancy in oxides caused by the formation process. A possible mechanism of irradiation on the corrosion is discussed. In the as-received specimen, Fe atoms loss led to voids in the oxides, and lead penetrated the oxides through these voids. In the irradiated specimen, defects left by previous irradiation helped to form a more uniform oxide layer. The adhesive outer magnetite oxide and the Fe ions generated from where grain boundary oxidation developed retarded the presence of voids and made the oxide layer protective.

The eighth experiment of the SINQ Target Irradiation Program (STIP-VIII) included a total of 941 specimens from 25 steels, 8 zircaloys, 6 tungsten based alloys (W-alloys) and 2 silicon carbide (SiC) composites. Specimens were prepared in 2016 and 2017. Irradiation was carried out in SINQ Target-13 in two periods: June 26 to December 21, 2018 and July 30 to December 23, 2020. The total proton charge received by Target-13 is 7.8 Ah, corresponding to 1.75×1023 protons. The irradiated specimens were unpacked in 2022 and 2023 and more than 98 % of the specimens were recovered. The specimens were irradiated at temperatures between 100 and 450 °C. The maximum doses are 10.4 dpa for steels, 14.8 dpa for zircaloys, 11.5 dpa for W-alloys, and 3.5 dpa for SiC composites. The He-to-dpa ratio is in the range of 22–42 appm/dpa for steels, 13–16 appm/dpa for zircaloys, 30–36 appm/dpa for W-alloys and 90–125 appm/dpa for SiC composites.

Structural materials applied in lead-cooled fast reactors will inevitably undergo plastic deformation,so it is import-ant to recognize and understand the influence of plastic deformation on the corrosion behavior of structural materials.In this work,the corrosion behavior of SIMP steels with different rolling deformations in liquid LBE at 550 ℃ was investigated.It was found that the rolled samples exhibited higher oxidation rates than the original samples during the pre-corrosion period,due to the fact that the grain boundaries and dislocations created by rolling promote the diffusion of atoms,which in turn ac-celerates the growth of the oxide layer.In the late stage of corrosion,the diffusion of Fe and O elements is inhibited due to the formation of Cr and Si-rich oxides.As a result,the rolled samples exhibit low oxidation rates relative to the original samples.

Pb83Mg17 is expected to be a potential coolant for lead cooled fast reactors because of its less toxic radioactive products and possible acceptable corrosion rate without additional oxygen control in comparison with LBE. In order to evaluate the compatibility of structural steels and Pb-Mg, two F/M steels (SIMP, T91) and two austenitic steels (316L, 15-15Ti) were tested in Pb-Mg at 350, 450 and 550 °C with exposure time ranging from 100 h to 2000 h in this work. Compared with the slight corrosion or no obvious corrosion of F/M steels, austenitic steels have suffered very serious dissolution corrosion which is significantly influenced by temperature and exposure time. The corrosion rate of austenitic steels is controlled by dissolution or diffusion mechanism at different temperatures.

The nucleation and growth of 100-textured and “toothed” columnar grains in Fe–0.5

The radial non-uniformity of an ADS fuel rod power density on the peak plane is as high as 10.46 %, which is derived from the reactor physics calculations. In order to investigate the influence of the above-mentioned radial power inhomogeneity on the fuel temperature distribution, this paper constructs a set of two-dimensional comparison examples, where one example uses a uniform heat source and another uses a non-uniform heat source distribution, with taking the fuel segment and its corresponding cladding segment in the region where the peak plane is located as the research object. The finite element software COMSOL is used to conduct the heat transfer analysis. The results of the study show that the fuel temperature under radial non-uniform power distribution is almost the same as that under uniform power. Therefore, this radial non-uniformity can be completely ignored when the research object of temperature is considered. The quantitative calculation carried out in this research can provide certain data support for the engineering research of accelerator driven subcritical system, and can also provide certain guidance for the performance analysis of such fuel elements.

A novel Cr coating with uniformly dispersed yttrium-based nano-oxides (i.e., ODS-Cr coating) was synthesized. Structural analyses show that, after heat treatments, the grain coarsening was not detected, and mechanical properties remained almost intact in ODS-Cr coatings while the grain size increased six times and mechanical properties degraded remarkably for pure Cr coatings. Irradiation-induced cavities in ODS-Cr coatings were also greatly suppressed compared with those in pure Cr coating. These results indicate that the ODS-Cr coating has a better mechanical properties and irradiation resistance than traditional pure Cr, and highlight its potential application as protective coatings for ATF solutions.

A comparison investigation on the oxidation behavior of SIMP and T91 steels exposed to oxygen saturated LBE up to 10000 h at 450 degrees C was conducted. The results show that the surface oxides formed on both steels are mainly composed of a Fe3O4 outer layer and a (Fe,Cr)3O4 inner layer. Under the same conditions, SIMP steel displays a better corrosion resistance than its counterpart, which was attributed to higher Si content in SIMP steel. The Si effect on reducing oxidation rate was discussed by first-principles calculations.

Refractory high-entropy protective coatings are of interest for nuclear fuel cladding applications due to their corrosion resistant properties and irradiation resistance at elevated temperature. Here, TiZrNbTaV metallic and (TiZrNbTaV)N films were deposited by magnetron co-sputtering. The metal elemental contents of both films were nearly equiatomic. These films were irradiated by Xe ions at room temperature and 500 degrees C, and examined by X-ray diffraction and transmission electron microscopy. The as-deposited (TiZrNbTaV)N film showed a single NaCl-type fcc phase and a pronounced columnar growth structure, which could remain intact after irradiation treatments. In contrast, the as-deposited TiZrNbTaV film exhibited an amorphous structure and formed a bcc phase structure after irradiation at 500 degrees C. The TiZrNbTaV film after irradiation at 500 degrees C composed of depth -dependent size of grains. This distribution of grain size is consistent with simulated displacement damage. The stable structure of (TiZrNbTaV)N film under high temperature irradiation indicates that these materials have potential for use as protective coatings for nuclear fuel claddings.

In this study we investigated the accelerated corrosion behavior of Fe-ion irradiated RAFM steel (SIMP) in high-temperature steam. The thickness, elementary composition, surface morphology, crystal structure and microstructure of the formed oxide films were studied. Consistent with the unirradiated sample, the oxide films of the irradiated samples consisted of an outer layer of Fe3O4, a porous inner layer of Fe-Cr spinel, and an internal oxide zone composed of (Cr, Si)2O3. The thickness of the oxide films increased with the increasing displacement damage. Compared with austenitic steels, the increase of corrosion rate of SIMP steel by displacement damage is rather moderate.

CiADS次临界反应堆采用铅铋作为冷却剂,使用绕丝定位的闭式燃料组件.螺旋绕丝能够使流动工质发生横向交混,增强不同子通道间的动量交换.本工作使用ANSYSICEM和STAR-CCM+软件划分了结构化网格和多面体网格,对CiADS燃料棒束通道横流特性进行研究,分析了结构化网格与非结构化网格对流动特性的影响.结果显示:相比于非结构化网格,结构化网格的子通道间横流量计算结果更接近LES的结果;结构化网格得到的摩擦因子系数与UCTD公式预测值最为接近,多面体网格的结果比经验公式的预测值小.同时,本研究对现有横流特性经验公式进行了改进,可更准确地预测带绕丝燃料棒束横流特性.上述研究成果提供了子通道程序优化的新思路,同时可为CiADS燃料组件热工水力设计与分析提供参考.

The study aims to understand the irradiation behavior of multilayer coatings composed of high-entropy materials. Here, we report the structural stability and elemental segregation of high-entropy TiNbZrTa/CrFeCoNi metallic and nitride multilayer coatings under 3-MeV Xe20+ ion-irradiation at room temperature and 500 °C, respectively. Transmission electron microscopy analysis shows that the microstructure of nanocrystalline CrFeCoNi high-entropy-alloy sublayers are not stable and readily transforms into amorphous state at 500 °C and/or under irradiation conditions. The elemental distribution, acquired by energy-dispersive X-ray spectroscopy under scanning transmission electron microscopy mode, shows preferential diffusion of Co and Ni into TiNbZrTa sublayers, while Fe and Cr preferentially remain within the previous CrFeCoNi sublayers. TiNbZrTaN/CrFeCoNiNx nitride multilayers exhibit a higher crystallinity and structural stability as well as resistance to diffusion at high-temperature and/or irradiation conditions than their TiNbZrTa/CrFeCoNi metallic multilayer counterparts. These findings are explained by atomic size differences, the difference in Gibbs free energy of the mixing system, and interstitial-solute-induced chemical heterogeneity. Our findings thus provide a design strategy of high entropy nitride for nuclear fuel cladding.