Search Results (269)

The core outlet temperature is an important safety parameter for pressurized water reactors (PWRs), and its accurate determination depends on highly reliable thermocouple sensors. In this study, two typical commercial K-type thermocouple alloys (de1 and de2) were used to fabricate mineral-insulated metal-sheathed (MIMS) thermocouples. A comparative severe accelerated aging test at 500 °C for 168 h was performed to investigate electromotive force (EMF) stability under high-temperature exposure conditions. After aging, both alloy types exhibited obvious positive EMF drift, and the maximum drift magnitude appeared at the calibration temperature of 400 °C. The measured irreversible EMF drift of de2 thermocouples was lower than that of de1 thermocouples. For the same de2 alloy, thermocouples fabricated with pre-oxidized thermoelement wires presented smaller irreversible drift than those made of bright wires. Reversible EMF drift and heating–cooling calibration hysteresis were more significant for de1 specimens. Stabilization treatment at 570 °C for 2 h followed by furnace cooling effectively reduced the thermal hysteresis of both alloys during temperature cycling. This study provides straightforward experimental data and practical processing references for optimizing the manufacturing route of nuclear-grade MIMS thermocouples with improved high-temperature EMF stability. Full article

Autonomous robotic systems designed for extraterrestrial in situ resource utilization (ISRU) will play a central role in enabling a sustained human presence on the Moon and Mars. These robots are expected to perform tasks such as regolith excavation, water extraction, oxygen production, and propellant generation under extremely harsh environmental conditions, including large temperature variations, abrasive dust, high radiation levels, and significant communication delays with Earth. Consequently, their onboard electrical systems must operate with high reliability, autonomy, and fault tolerance. A critical enabling technology for these systems is the isolated power conversion architecture, which distributes energy from primary power sources to multiple robotic subsystems, including mobility actuators, drilling systems, sensors, computing units, and thermal management modules. Future lunar and Martian missions are expected to rely on a combination of alternative energy sources, including solar photovoltaic arrays with energy storage, fuel cells, radioisotope power systems, and nuclear surface power reactors, which can provide continuous and high-density energy independent of sunlight availability. These diverse power sources require flexible and highly efficient isolated DC–DC power conversion architectures capable of managing wide input voltage ranges while ensuring electrical isolation, safety, and system stability across distributed robotic platforms. This literature review surveys recent developments in autonomous isolated power conversion architectures suitable for lunar and Martian resource extraction robots. The review examines advanced converter topologies such as resonant converters, phase-shifted full-bridge converters, dual-active bridge converters, and modular multiport power converters designed for high efficiency, high power density, and scalable power distribution. Emphasis is placed on converter architectures capable of interfacing with nuclear-powered systems and other high-energy-density sources while supporting distributed loads in robotic mining and processing systems. In addition, the paper reviews emerging autonomous control strategies, including adaptive digital control, intelligent power management, fault detection and self-recovery mechanisms, and distributed power architectures capable of maintaining stable operation under dynamic load conditions. The role of wide-bandgap semiconductor technologies, including silicon carbide (SiC) and gallium nitride (GaN), is also examined, highlighting their potential to enable higher switching frequencies, improved efficiency, reduced system mass, and enhanced thermal performance in vacuum environments. Finally, system-level considerations for integrating isolated power conversion within robotic ISRU platforms are discussed, including redundancy strategies, power bus architectures, electromagnetic compatibility, thermal management, and long-duration reliability requirements. By consolidating advances across power electronics, autonomous control, and space power systems, this review identifies key research gaps and outlines design directions for next-generation autonomous power conversion systems capable of supporting scalable lunar and Martian resource extraction infrastructures powered by both renewable and nuclear energy sources. Full article

A large quantity of irradiated nuclear graphite has been generated worldwide and must be managed as radioactive waste. In Germany, the final disposal of irradiated graphite (i-graphite) remains challenging due to limited knowledge of the radionuclide inventory and strict regulatory constraints. The presence of gamma-emitting radionuclides raises concerns regarding radiation exposure during handling, for example, in reactor decommissioning operations. Decontamination processes are therefore being investigated to enable the selective removal of key radionuclides, with the dual objective of facilitating waste acceptance and enabling the potential reuse of treated graphite in technical applications. This study focuses on the selective removal of volatile gamma-emitting radionuclides using a thermal treatment approach. In this work, a steam-assisted thermal treatment was applied to i-graphite samples from the Forschungsreaktor Jülich 2 (FRJ-2). The influence of key process parameters, including temperature, treatment duration, and relative humidity, was systematically investigated. The results show complete removal of ⁶⁰Co and ¹³⁷Cs, while partial removal of approximately 40% was achieved for ^152,154,155Eu and ²⁴¹Am. No release of ¹³³Ba was observed. These findings demonstrate the potential of high-temperature treatment as an effective method for reducing the radionuclide inventory of irradiated graphite, thereby contributing to improved waste management strategies and safer handling during decommissioning. Full article

Silicon carbide (SiC), a third-generation wide-bandgap semiconductor, demonstrates prominent application advantages for extreme-environment sensing scenarios including deep-space exploration, nuclear reactor monitoring, and fusion device diagnosis, which benefit from its excellent radiation resistance, high-temperature stability, and chemical inertness. This review systematically investigates the action mechanisms of different radiation environments on the electrical and mechanical properties of SiC-based sensors, with emphasis on the regulatory effects of radiation-induced defects on key sensing parameters, including piezoresistive properties, charge-collection efficiency, leakage current, and sensitivity. In addition, this paper discusses the response behavior and research progress of SiC sensors applied in mixed radiation fields. Existing research confirms that although high-fluence radiation can induce lattice defects and further result in the degradation of SiC sensor sensing performance, SiC still retains remarkable advantages in intrinsic radiation resistance. The sensing reliability of SiC in extreme environments can be further improved via device-structure optimization and material-modification strategies. This review is expected to provide a theoretical reference for the development and design of SiC sensors applied in advanced nuclear energy, aerospace, and nuclear medicine fields. Full article

Molten salts are increasingly regarded as promising fluids for high-temperature heat transfer, thermal energy storage, and advanced reaction processes, including concentrated solar power (CSP), molten salt oxidation (MSO), and next-generation nuclear reactors. Among these materials, the ternary eutectic mixture Li₂CO₃–Na₂CO₃–K₂CO₃ (32.12–33.36–34.52 wt%) has emerged as a leading candidate due to its wide operating temperature range and favourable thermodynamic properties. Despite its relevance, substantial inconsistencies and gaps remain in the available thermophysical property data, posing challenges for reliable design, modelling, and industrial deployment. This work revisits the Li₂CO₃–Na₂CO₃–K₂CO₃ eutectic through a critical assessment of the literature from its reported melting point at 670 K (397 °C) up to approximately 1200 K (927 °C). Using a methodology inspired by IUPAC-supported strategies previously applied to common liquids such as water and hydrocarbons, we examine the quantity, quality, and coherence of existing measurements. Reference correlations are proposed only where the data are sufficiently robust to justify them. The analysis highlights a pressing need for more accurate and comprehensive measurements—particularly for heat capacity, thermal conductivity, and viscosity—to enable the development of reliable standard reference correlations. Brief recommendations are given on the measurement methods that should be used in high-temperature measurements, namely for heat capacity, viscosity, and thermal conductivity. Reliable thermophysical property data for (LiNaK)₂CO₃ remain limited and inconsistent, despite its relevance for high-temperature energy applications. Density data are comparatively robust, but heat capacity, thermal conductivity, and viscosity still require high-accuracy measurements at elevated temperatures. Addressing these data deficiencies is essential for advancing the safe and efficient use of molten carbonates in high-temperature energy technologies. Full article

In this study, double-pass hot compression tests were conducted to systematically investigate the effects of hot deformation parameters on the static recrystallization (SRX) behavior of SA-508M Gr.3 steel used for nuclear reactor pressure vessels. The deformation temperatures were set to 950, 1050, and 1150 °C, with strain rates of 0.01, 0.1, and 1 s⁻¹. The first-pass strains were 0.05, 0.10, and 0.15; the inter-pass time was fixed at 60 s; and the second-pass strain was maintained at 0.05. Based on the experimental data, a kinetic model describing SRX softening behavior was established. The activation energy for SRX was determined to be 81.45 kJ·mol⁻¹, and the Avrami exponent was 0.5742. The characteristic time for 50% recrystallization (t_0.5) was quantified under different deformation conditions. In addition, the microstructural evolution of SRX after double-pass hot compression was characterized using electron backscatter diffraction (EBSD). The results show that increasing the deformation temperature and strain rate leads to opposite trends in the flow stress during double-pass deformation, with the flow stress decreasing with temperature and increasing with strain rate. Meanwhile, inter-pass static softening is enhanced, resulting in a pronounced stress drop during the second pass. An increase in the first-pass strain further intensifies the stress drop and enhances the extent of SRX. EBSD analysis reveals consistent microstructural evolution: with increasing deformation temperature, strain rate, and the first-pass strain, the misorientation distribution shifts from low-angle grain boundaries (LAGBs) to high-angle grain boundaries (HAGBs), indicating an increased degree of SRX. These findings provide a theoretical basis and experimental support for process parameter optimization and engineering applications of SA-508M Gr.3 steel. Full article

Mo–Re alloys serve as critical structural components for high-temperature nuclear reactors, and their irradiation degradation is closely related to the evolution of vacancy-type defects. In this study, heavy-ion and He-ion irradiations were performed under RT to introduce an average displacement damage of 3.5 dpa within the 1 μm-thick surface layer of Mo–Re alloys with Re content up to 47 wt.%. PALS, SPB-DBS and CDB techniques were employed to characterize the size, concentration, depth distribution and local chemical environment of irradiation-induced vacancy-type defects. The results demonstrate that the longer lifetime component of irradiated Mo–Re alloys ranged from 262 to 280 ps, corresponding to medium-sized vacancy clusters. The S parameter of all specimens increased significantly from approximately 0.42 to 0.50, with negligible differences (<0.01) among various Mo–Re alloys. No distinct characteristic peak of Re was observed near 17 × 10⁻³ m₀c at the vacancy sites, which was inconsistent with simulation predictions. Mo–Re alloys exhibit similar vacancy-type defect features to pure Mo, implying weak interactions between Re solute atoms and vacancy-type defects under RT irradiation. Full article

Durability tests were performed on concrete representative of nuclear power plant structures exposed to boric acid leakage. Boric acid is widely used in pressurized water reactor systems for neutron absorption and chemical control; therefore, accidental leakage or evaporation-induced concentration of borated water can create an important durability concern for adjacent concrete structures. Three boric acid concentrations (0, 2500 and 18,000 ppm) and two temperature levels (20 and 60 °C) were considered. The evolution of cube compressive strength and elastic modulus with exposure age was examined over 180 days. The results show a slight increase in concrete strength and stiffness during the early stage of exposure, followed by a continuous increase in the loss ratios of cube compressive strength and elastic modulus at later ages. The maximum measured strength–loss ratio was 4.684%, and the maximum measured elastic modulus–loss ratio was 1.755%. A service life prediction model based on loss ratio parameters was then developed from the test results. The fitted examples gave high coefficients of determination (R² = 0.9960 for cube compressive strength and R² = 0.9947 for elastic modulus), indicating that the proposed model can support durability assessment and aging management of boric-acid-corroded concrete structures in nuclear power plants. Full article

It has been observed in prior research that high thermal impact—resulting from a large temperature difference between hot water vapor and cold liquid water—can enhance the thermal performance of cavitation-induced low-energy nuclear reactions (LENRs) in water, with an estimated increase in the coefficient of performance (COP) of approximately 50% for every 100 °C temperature rise. The temperature of the hot water vapor is primarily determined by the boiler output, which typically represents the highest temperature source and plays a dominant role in reactor performance. In this study, a flow oscillator was designed as an thermal conditioning component for these potential LENR reactor systems using linear flow network analysis (LFNA) to generate flow resonance that elevates the hot vapor temperature, thereby increasing thermal impact and improving LENR performance. LFNA is based on the linearization of the fluid flow equations governing mass and momentum transport and utilizes a fluid-electric circuit analogy. For a fluid flow system, various components can be modeled using analogs of electrical resistance, capacitance, and inductance (R, C, and L), allowing the system behavior to be analyzed similarly to an RLC circuit. Through this analogy, flow resonance phenomena can be predicted, potentially enabling the generation of high-temperature and high-pressure responses that are beneficial to LENR processes. The analytical model was experimentally validated and subsequently applied in the LENR reactor design. The analytical result shows that an output temperature difference exceeding 350 °C can be achieved using a 0.5 m pulse tube at a 46 Hz triggering frequency with 20 kPa perturbation, which indicates a potential COP enhancement of 175% based on prior studies. The result provides a potential mechanism to significantly enhance the thermal impact conditions and promote LENR performance in water-based reactor systems. Full article

Zirconium carbide (ZrC) is a leading candidate for advanced nuclear reactor components due to its ultra-high melting point, thermomechanical stability, and low neutron absorption. However, its irradiation damage behavior and mechanism remains underexplored. In this work, dense pressureless-sintered ZrC ceramics with low-neutron-absorption MoSi₂ additives were irradiated with 500 keV He²⁺ ions at room temperature to peak damage levels of 0.30, 1.49, and 2.97 dpa. The changes in their microstructure, bonding states, and property were analyzed via TEM, GIXRD, Raman spectroscopy, nanoindentation, and TDTR. ZrC retained crystallinity regardless of high-density black-spot defects, while MoSi₂ exhibited severe amorphization and swelling. Lattice expansion and partial Zr-C bond breakage with C-C bond formation were confirmed, with maximum hardening at 1.49 dpa and significant elastic modulus reduction at 2.97 dpa. Thermal conductivity decreased modestly and showed minimal dose dependence, indicating a saturation effect. These results elucidate defect evolution in pressureless-sintered ZrC-MoSi₂ ceramics and support its application in high-irradiation nuclear environments. Full article

Reaction-bonded SiC (RBSC) ceramics exhibit limited hydrothermal corrosion resistance due to the presence of residual silicon. This study presents a strategy to enhance the corrosion resistance of RBSC through homogeneous incorporation of MoSi₂ and concurrent reduction in residual silicon content. Three material systems were fabricated via reactive melt infiltration: conventional RBSC with a SiC/C preform (SC), a SiC–MoSi₂ composite incorporating commercial Mo₂C powder via physical mixing (MC), and a SiC–MoSi₂ composite derived from a Mo₂C/C precursor synthesized by a molten salt method (MS). The Mo₂C/C composite synthesized at 1150 °C exhibited fine, uniformly distributed Mo₂C particles coated on carbon black, contrasting with the agglomerated distribution in commercial Mo₂C mixtures. During reactive sintering at 1600 °C, Mo₂C reacted with molten Si to form MoSi₂, reducing residual Si content. Sample MS achieved the lowest residual Si (8.77 ± 0.45 vol.%), followed by MC (12.43 ± 0.86 vol.%) and SC (19.17 ± 1.01 vol.%). All samples achieved near-full densification (open porosity < 0.1%), with bulk densities of 2.96 ± 0.05, 3.03 ± 0.03, and 3.07 ± 0.03 g/cm³ for SC, MC, and MS, respectively. Microstructurally, MS displayed homogeneous MoSi₂ dispersion, while MC showed partial MoSi₂ aggregation, and SC contained continuous residual Si regions. Hydrothermal corrosion tests at 345 °C and 15 MPa for 9 days demonstrated that corrosion resistance followed the order MS > MC > SC. After 9 days, weight loss was 22.3970 ± 1.2059 mg/cm² (SC), 17.6370 ± 0.8266 mg/cm² (MC), and 15.4347 ± 0.7807 mg/cm² (MS), with corrosion depths of 393.17 ± 27.46, 267.40 ± 24.44, and 224.60 ± 25.13 μm, respectively. The enhanced performance of MS arises from two synergistic factors: reduced residual Si minimizes large corrosion pores, while uniform distribution of MoSi₂ facilitates the formation of a stable, dissolution-resistant composite oxide layer composed of MoO₃ and SiO₂, in which MoO₃ restrains excessive dissolution of SiO₂ through a pinning effect. These findings demonstrate that combining residual Si reduction with homogeneous MoSi₂ incorporation via molten salt-synthesized precursors offers an effective strategy for improving hydrothermal corrosion resistance of reaction-bonded SiC-based materials for applications in high-temperature and high-pressure aqueous environments such as nuclear water reactors. Full article

The Zircaloy-4 alloy is a key structural material for nuclear reactor cores. However, its behavior under warm deformation conditions and during phase transformations requires in-depth investigation to improve technologies for producing ultrafine-grained (UFG) structures using severe plastic deformation methods. This work presents a comprehensive study of the rheological properties, phase stability, and microstructural evolution of the alloy in the temperature range from 20 to 950 °C at strain rates of 0.5 and 15 s⁻¹. The experimental part included plastometric testing, dilatometric analysis, and microstructural characterization. It was established that the optimal window for plastic deformation corresponds to warm deformation at 650 °C. Dilatometric analysis confirmed that heating to 650 °C ensures the preservation of a stable initial α-phase structure, since the formation of secondary phases and the α→β transformation are initiated at higher temperatures, namely 694 °C (onset) and 847 °C (completion). At 650 °C, the deformation resistance decreases by approximately 70% compared to cold processing, while the strain-rate sensitivity of the flow stress is minimized. EBSD analysis showed that deformation under these conditions leads to intensive grain fragmentation via mechanisms of dynamic recovery and the initial stages of continuous dynamic recrystallization. The decisive role of the kinetic factor was demonstrated: reducing the strain rate to 0.5 s⁻¹ promotes the formation of a finer and more homogeneous grain structure. In contrast, high strain-rate deformation (15 s⁻¹) results in coarser grains and increased non-relaxed intragranular residual stresses. The obtained results provide a physical basis for optimizing thermomechanical processing regimes and can be used to produce UFG structures in zirconium alloys without the risk of phase degradation. Full article

Nuclear energy is a clean and efficient energy source crucial for the future energy supply. The harsh conditions in reactors, including high temperature, high pressure, and intense neutron irradiation, cause structural materials to accumulate irradiation damage, leading to performance degradation. Austenitic stainless steel, due to its superior mechanical properties, irradiation resistance, and corrosion resistance, has been extensively utilized as a core structural material in light water reactors and emerged as a candidate material for Generation IV nuclear reactors. Therefore, understanding irradiation damage and macroscopic properties evolution in austenitic stainless steels is critical for enhancing the safety and long-term service life of reactor core materials. This review began by elucidating the application of charged particles in irradiation studies, emphasizing the prevailing substitution of neutron irradiation with proton irradiation experiments in current studies. Subsequently, the work systematically synthesized irradiation damages and their consequential impacts on macroscopic properties. Finally, it consolidated the progress and provided prospects for research on improving the resistance of austenitic stainless steel to irradiation-induced segregation, irradiation hardening, irradiation swelling, and irradiation-corrosion synergies. Full article

The article presents a comparative thermodynamic and economic analysis of nuclear power plants using high-temperature gas-cooled reactors (HTGRs) and small modular pressurized water reactors (SMRs). HTGRs, with their ability to achieve steam temperatures exceeding 650 °C, offer significantly higher electricity generation efficiency (approximately 52%) compared to SMRs and traditional PWRs, which achieve around 32%. The study underscores the importance of economic efficiency in investment decisions, noting that while SMRs are still in the conceptual phase, their future construction is uncertain, with realistic deployment expected by the late 2030s or early 2040s. The analysis highlights the superior thermodynamic performance of HTGRs due to their hierarchical dual-cycle gas–steam technology, compared to the single-cycle Clausius–Rankine process used in SMRs and PWRs. The paper contributes new insights into the comparative advantages and challenges of these nuclear technologies, particularly emphasizing the advanced safety features of HTGRs and the inherent design challenges associated with scaling down PWR technology for SMRs. Full article

Nuclear energy has emerged as a crucial technological solution for ensuring energy security and achieving carbon neutrality goals, given its ultra-high energy density and near-zero carbon emissions against the backdrop of rapid socioeconomic development, increasing energy demands, and accelerated global transition toward low-carbon energy structures. As the core component for energy conversion in nuclear reactors, fuel elements critically determine reactor efficiency and safety performance, with the fission product retention capability of silicon carbide layers in multilayer-coated fuel particles having been thoroughly validated through high-temperature gas-cooled reactor irradiation tests. The precise sphericity control of large-sized UO₂ fuel kernels represents a fundamental requirement for enhancing tristructural isotropic (TRISO) fuel particle performance and advancing Generation IV nuclear power plant development. This study presents a sphericity control strategy based on sol–gel processing that synergistically integrates physicochemical regulation of gelling media with multi-field washing flow field optimization. By implementing silicone oil-mediated interfacial tension gradient control, we effectively suppressed gel sphere destabilization while developing an innovative three-phase sequential washing technique involving kerosene washing, anhydrous ethanol interfacial transition, and ammonia solution replacement, which significantly enhanced mass transfer diffusion in stagnant liquid films and revolutionized fuel microsphere washing technology with improved efficiency and quality. Experimental results demonstrate that this integrated approach increases kernel sphericity qualification to 99.8%, reduces washing solution consumption by 79%, and achieves an average sphericity of 1.03. The research establishes a coupling mechanism between gelling media and multi-field washing processes, elucidating the synergistic effect between interfacial tension regulation and washing optimization, thereby providing both theoretical foundations and engineering application basis for the precision manufacturing of high-performance nuclear fuels. Full article