Search Results (335)

As the key technology of space exploration, space power has been a major area of international research focus. A lot of research work has been carried out around the world for the space nuclear reactor using the heat pipe, liquid metal and gas cooling methods. With the development of molten salt reactor in the Generation IV reactor system, molten salt dissolving fissile material and acting as a coolant at the same time has become a new cooling scheme, which provides new ideas for the design of space nuclear reactors. In this study, a novel reactor, the liquid-solid dual-fuel space nuclear reactor (LSSNR) was preliminarily proposed, combining the molten salt fuel and cross-shaped spiral solid fuel to achieve the design goals of 30-year lifetime and an active core weight of less than 200 kg. Monte Carlo neutron transport code OpenMC based on ENDF/B-VII.1 library was employed for neutronics design in the aspect of fuel type, cladding material, reflector material and the spectral shift absorber. Then, the thickness of the control drum absorber was optimized to meet the requirement of the sufficient shutdown margin, lower solid fuel enrichment, and 30-effective-full power-years (EFPY) operation lifetime. Finally, UC solid fuel with U-235 enrichment of 80.98 wt.% and B₄C thickness of 0.75 cm were adopted in LSSNR, and BeO was adopted as the reflector and the matrix material of the control drum. A spectral shift absorber Gd₂O₃ was used to avoid the subcritical LSSNR returning to criticality in a launch accident. The k_eff with the control drum in the innermost position is 0.954949, and the k_eff reaches 1.00592 after 30 EFPY of operation. The total mass of the active core is 158.11 kg. In addition, the thermal-hydraulic feasibility of LSSNR using cross-shaped spiral fuel was analyzed based on a 4/61 reactor core model. The structure of cross-shaped spiral fuel achieves enhanced heat transfer by generating turbulence, which leads to a uniform temperature distribution of the coolant flow field and reduces local temperature peaks. Based on the LSSNR scheme, some neutronic characteristics were analyzed. Results demonstrate that the LSSNR has strongly negative reactivity coefficients due to the thermal expansion of liquid fuel, and the fission gas-induced pressure meets safety requirements. One hundred years after the end of core life, the total radioactivity of reactor core is reduced by 99% and is 7.1305 Ci. Full article

Silicon carbide (SiC) and SiC fiber-reinforced SiC matrix composites (SiC/SiC) are receiving renewed attention for use in next-generation fusion reactors due to their ability to withstand extreme conditions, including high temperatures, neutron irradiation, and plasma interactions. General Atomics Electromagnetic Systems (GA-EMS) has demonstrated significant progress in scaling up the fabrication of SiC/SiC, achieving high mechanical uniformity and meeting dimensional requirements in components up to 12 feet in length. Key developments are discussed including scale-up of the chemical vapor infiltration (CVI) process from lab-scale to full sized parts, high-dose (100 dpa) irradiation testing, nuclear-grade ceramic joining technologies, and production-focused quality control with the collective aim to establish SiC/SiC as a reliable solution for structural and functional components in fusion systems. Beyond manufacturing, the paper addresses supply chain barriers, particularly the limited availability and high cost of nuclear-grade SiC fiber. GA-EMS is developing a novel SiC fiber production method based on a thermochemical cure step that is anticipated to reduce costs compared to traditional approaches. Additionally, advancements in engineered SiC materials, such as SiC foams and tungsten-graded SiC composites, are discussed as promising solutions for specific fusion reactor components. Full article

The 2011 Fukushima accident highlighted the vulnerability of traditional Zr alloy fuel cladding under loss-of-coolant accident (LOCA) conditions, prompting the development of accident-tolerant fuel (ATF) systems. A promising near-term solution involves depositing protective coatings on existing Zr alloy cladding. Among various deposition techniques, cold spray technology has emerged as one of the leading methods due to its solid-state, low-temperature process, which minimises thermal degradation and allows for the deposition of a wide range of high-performance materials. This review provides a comprehensive examination of recent advances in cold-sprayed coatings for ATF cladding, beginning with an overview of the fundamentals of cold spray technology and its specific advantages for nuclear applications. The core of the review critically analyses three primary coating systems: Cr, FeCrAl alloys, and MAX phase composites, with a particular focus on Cr coatings, as they have been more extensively studied compared to the other two material systems. Key coating properties, including microstructure of the coating-substrate interface, mechanical properties, thermal conductivity, oxidation resistance, irradiation tolerance, and performance under normal operation and simulated LOCA conditions, are discussed in detail, with particular emphasis on the potential of cold-sprayed Cr coatings to enhance Zr alloy cladding. Cr coatings demonstrate significant improvements in oxidation resistance and irradiation stability, but also face challenges such as high-temperature interfacial reactions. To address these issues, promising solutions, such as diffusion-barrier bilayer systems, are being explored. Additionally, the review discusses FeCrAl and MAX phase composite coatings, highlighting their promising long-term performance under extreme conditions. The review concludes with recommendations for further research to optimise cold spray processes and ensure the robustness of coatings in operational reactor environments. Full article

Polylactide (PLA) has become widely adopted across biomedical, packaging, and manufacturing sectors due to its biodegradability and renewable sourcing. However, the rapid growth in PLA consumption has created urgent challenges related to waste management and the cleaning of processing equipment. This study investigates glycolysis as a promising chemical depolymerization pathway for PLA recycling and in situ reactor cleaning. A systematic analysis of four glycolysis agents (GA) (ethylene glycol, diethylene glycol, propylene glycol, and glycerol) was performed across molar PLA:GA ratios from 1:0.125 to 1:4 at 220 °C, targeting the efficient conversion of high-molecular-weight PLA (Mn ≈ 165 kDa) into low-molecular-weight oligomers. Gel permeation chromatography (GPC) demonstrated that propylene glycol exhibited the highest depolymerization efficiency, yielding oligomers with Mn as low as 200 g·mol⁻¹ even at minimal glycolysis agent ratios, while glycerol produced hydroxyl-rich oligomers optimal for subsequent lactide synthesis. Hydroxyl value (HV) measurements showed excellent agreement with theoretical values (<5% deviation), allowing us to make an assumption about an approximate, close to near-quantitative con-version. Glycolysis products with Mw below 400 g·mol⁻¹ displayed excellent water solubility, making them particularly attractive for reactor cleaning applications. Using glycerol-derived (GL) oligomers (PLA:GL = 1:0.25), purified L-lactide with a melting point of 98.1 °C and high purity (>99%) was obtained through thermocatalytic depolymerization and five recrystallization cycles, as confirmed by ¹H nuclear magnetic resonance (¹H NMR) and differential scanning calorimetry (DSC) analyses. The recovered lactide’s high purity renders it suitable for ring-opening polymerization, enabling closed-loop PLA recycling schemes. Overall, glycolysis emerges as a highly promising chemical recycling route complementary to hydrolysis and pyrolysis: propylene glycol maximizes depolymerization efficiency for cleaning applications, while glycerol optimizes oligomer functionality for lactide recovery and advanced material synthesis. Our results provide practical guidelines for selecting glycolysis agents and conditions for cleaning and recycling applications. Full article

Growing electricity demand and decarbonisation requirements pose significant challenges for coal-dependent transition economies. This study examines whether nuclear deployment can support low-carbon economic transformation using Poland’s national nuclear programme as a case study. We conduct a structured document analysis that integrates a systematic search and screening of peer-reviewed literature with an analysis of national policy and planning materials and a synthesis of publicly available project documentation for the Lubiatowo-Kopalino nuclear power plant, the Pątnów project, and the planned small modular reactor (SMR) deployments. Impacts on employment, infrastructure, technical education, technology transfer, and local supply chain participation are assessed and mapped to the sustainable development goals and the EU climate policy criteria. The analysis indicates that, if accompanied by early workforce development and supplier prequalification, nuclear investments can stimulate industrial upgrading, strengthen energy security, and deliver regional co-benefits beyond electricity generation. At the same time, scheduling slippage, governance uncertainty, and gaps in domestic capabilities in nuclear-specific components can limit these benefits. The article concludes with recommendations for national and local authorities on stakeholder engagement, local content strategy, and risk management that can be transferred to Central European economies with similar starting conditions. Full article

Materials capable of withstanding extreme environments open promising opportunities for nuclear fusion reactors. In this study, equiatomic CrFeTaVW and CrFeTiVW high-entropy alloys are investigated as interlayer materials between W and CuCrZr. Monte Carlo and Molecular Dynamics simulations predicted a bcc-type structure for both systems. Additionally, the Monte Carlo simulation predicts lower potential energy and a more stable structure for both systems than Molecular Dynamics. For CrFeTaVW, the chemical segregation values are lower in MC than in the MD simulation, whereas for CrFeTiVW, the opposite trend is observed, with MC indicating stronger segregation values. After simulation, the high-entropy alloys were prepared by planetary ball milling, consolidated by spark plasma sintering, and analyzed using X-ray diffraction, scanning electron microscopy, and thermal diffusivity. The experimental results for the milled powders confirmed the formation of a bcc structure in both alloys. The consolidated material revealed a bcc-type structure and an Fe₂Ta Laves phase for the CrFeTaVW HEA, while the CrFeTiVW HEA exhibits two different bcc-type structures. The values of CrFeTaVW and CrFeTiVW thermal diffusivity are between 3.5 and 7 mm²/s, which is consistent with the expected values for high-entropy alloys. Overall, the findings indicate that these HEAs have promising properties that can be used in extreme environments. Full article

Molten Salt Reactors (MSRs) offer significant advantages over conventional reactors but introduce unique modeling challenges due to their circulating liquid fuel and strong coupling among nuclear, chemical, and fluid transport processes. These challenges are amplified in depletion calculations, where MSR specific phenomena such as online refueling, off-gas removal, material redistribution, and other flow driven processes must be accurately represented. This work presents a novel multi-point depletion model that efficiently and accurately predicts isotopic evolution in MSRs by explicitly accounting for these characteristics. The mathematical formulation is derived from first principles and is computationally implemented in the open-source depletion code ONIX using neutronics solutions from open-source transport code OpenMC. The new model represents the entire primary loop by dividing it into interconnected depletion zones and tracks nuclide transport, irradiation, and removal mechanisms through a system of coupled ordinary differential equations. This approach enables parallel computation and improves performance over traditional sequential depletion methods. Validation of the developed model against Molten Salt Reactor Experiment data shows good agreement for salt-seeking isotopes and those without noble gas precursors, while discrepancies for other nuclides suggest underestimation of the corresponding removal rates. The depletion model was further applied to a reference Molten Salt Fast Reactor design to assess a new reprocessing scheme intended to expedite the achievement of equilibrium operation. Full article

Inconel 617 (IN617) is a promising structural material for advanced nuclear systems such as heat pipe-cooled reactors, but its fundamental defect evolution under neutron irradiation remains poorly understood. This study employs classical molecular dynamics simulations to investigate the atomic-scale irradiation damage mechanisms in a representative Ni–Cr–Co ternary model of IN617 under successive displacement cascades. The results reveal a near-linear accumulation of Frenkel pairs with dose, with the count increasing by a factor of approximately 24 from the first to the 75th cascade. A critical finding is the stark asymmetry in defect kinetics: interstitials rapidly coalesce into large clusters (with 88.4% of interstitials found in clusters of ≥ 2 atoms after 75 cascades), while vacancies remain predominantly isolated (constituting 68.8% of all vacancy defects). This disparity directly governs microstructural evolution, as interstitial cluster growth drives dislocation loop nucleation, leading to a linear rise in dislocation density to a saturated value of approximately 4.5 × 10⁻⁴ Å⁻². The saturated dislocation structure subsequently undergoes continuous reorganization through reactions between partial dislocations. These insights demonstrate that irradiation hardening in IN617 under simulated conditions is governed primarily by interstitial-type defect clustering, providing a crucial mechanistic basis for assessing its performance in radiation environments. Full article

Accurate prediction of the effective thermal conductivity (ETC) of nuclear fuels is essential for optimizing fuel performance and ensuring reactor safety. However, the experimental determination of ETC is often limited by cost and complexity, while high-fidelity simulations are computationally intensive. This study presents a novel hybrid framework that integrates experimental data, validated mesoscale finite element simulations, and machine-learning (ML) models to efficiently and accurately estimate ETC for advanced uranium-based nuclear fuels. The framework was demonstrated on three fuel systems: UO₂-BeO composites, UO₂-Mo composites, and U-10Zr metallic alloys. Mesoscale simulations incorporating microstructural features and interfacial thermal resistance were validated against experimental data, producing synthetic datasets for training and testing ML algorithms. Among the three regression methods evaluated, namely Bayesian Ridge, Random Forest, and Multi-Polynomial Regression, the latter showed the highest accuracy, with prediction errors below 10% across all fuel types. The selected multi-polynomial model was subsequently used to predict ETC over extended temperature and composition ranges, offering high computational efficiency and analytical convenience. The results closely matched those from the validated simulations, confirming the robustness of the model. This integrated approach not only reduces reliance on costly experiments and long simulation times but also provides an analytical form suitable for embedding in engineering-scale fuel performance codes. The framework represents a scalable and generalizable tool for thermal property prediction in nuclear materials. Full article

The paper presents a comparative experimental study of heat-transfer behavior in three alloys considered candidate materials for nuclear reactors: the austenitic stainless steel 316L, Zircaloy-4 (currently used in CANDU reactors), and an ODS alloy with a ferritic matrix. The investigation was conducted across five temperature intervals, each sample being subjected to a thermal shock through short-term overheating to the upper limit of its respective interval. The variation of thermal diffusivity in the three alloys was determined as a function of both measurement temperature and applied thermal shock, and trends in heat-transfer behavior were compared across the five temperature ranges. The experimental results show that up to 400 °C, Zircaloy-4 exhibits the highest thermal diffusivity, followed by the ODS alloy, with the lowest values measured for 316L steel. At approximately 450 °C, the ratio between 316L and the ODS alloy reverses. Beyond this point, increasing the temperature up to 900 °C is accompanied by a continuous rise in thermal diffusivity for both 316L stainless steel and Zircaloy-4. In contrast, for the ODS steel, increasing temperature leads to a continuous decrease in thermal diffusivity, reaching a minimum near the Curie point. The novelty of the study lies in the comparative assessment of the influence of temperature on the heat-transfer process in three alloys relevant to nuclear energy, covering the operating temperature ranges of CANDU and ALFRED reactors, as well as potential accidental overheating up to 900 °C. A particular feature of the work is the prior application of a short-duration overheating step produced using solar energy. The results are relevant not only for nuclear reactors but also for other high-temperature applications in corrosive environments. Full article

This study investigates the neutronic performance of burnable poisons (BPs) in an epithermal spectrum supercritical carbon dioxide (S-CO₂)-cooled reactor. Twelve candidate BP materials are systematically evaluated, including rare-earth oxides (e.g., HfO₂, Er₂O₃, Eu₂O₃, etc.) and boron-based compounds (B₄C and PACS). The deterministic neutron transport code KYLIN-I with the ENDF/B VI 45-group cross-section library is employed for analysis. According to the calculation results, Eu₂O₃ effectively suppresses the initial k_inf of the epithermal-spectrum fuel assembly to ~1.2 with a relatively low weight fraction (~2.6%) while maintaining a total temperature coefficient (TTC) lower than −1.4 pcm/K throughout the entire burnup period. HfO₂ and Er₂O₃, at approximately 15% weight fraction, achieve TTC values better than −2 pcm/K. Furthermore, both Eu₂O₃ and HfO₂ contribute to maintaining a low, stable power peaking factor (PPF) below 1.24 throughout the burnup process. This study provides a theoretical foundation and technical support for designing an efficient and safe S–CO₂-cooled nuclear reactor. It highlights the importance of selecting BP materials that are well-adapted to the neutron spectrum and optimizing the fuel assembly configuration accordingly. Full article

The research and development of new accident-tolerant fuel cladding materials has emerged as a critical focus in international academic and engineering fields following the Fukushima nuclear accident. Due to the outstanding resistances in corrosion and radiation as well as high-temperature creep properties, oxide dispersion-strengthened (ODS) FeCrAl alloys have been studied extensively during the past decade. Current review articles in this field have primarily focused on the effects of chemical composition on the anti-corrosion performance and species of nano-oxide. However, several key issues have not been given adequate attention, including processing methods and parameters, high-temperature stability mechanisms, post-deformation microstructural evolution and high-temperature mechanical properties. This paper reviews the progress of basic research on ODS FeCrAl alloys, including preparation methods, the effects of preparation parameters, the thermal stability and irradiation stability of oxides, the microstructural deformation, and the mechanical properties at elevated temperatures. The aspects mentioned above not only provide valuable references for understanding the effects of preparation parameters on the microstructure and properties of ODS FeCrAl alloys but also offer a comprehensive framework for the subsequent optimization of ODS FeCrAl alloys for nuclear reactor applications. Full article

We present the concept of an ultracold neutron (UCN) source with a superfluid He-4 (SF ⁴He) converter located in the thermal column of the WWR-K research reactor at the Institute of Nuclear Physics (INP) in Almaty, Kazakhstan. The conceptual design is based on the proposal of accumulating UCNs in the source and effectively transporting them to experimental setups. We propose to improve the UCN density in the source by separating the heat and UCN transport from the production volume and decreasing the temperature of the SF ⁴He converter to below about 1 K. To obtain operation temperatures below 1 K, we plan to use a He-3 pumping cryogenic system and minimize the thermal load on the UCN accumulation trap walls. Additional gain in the total number of accumulated UCNs can be achieved through the use of a material with a high critical velocity for the walls of the accumulation trap. The implementation of such a design critically depends on the availability of materials with specific UCN and cryogenic properties. This paper describes the conceptual design of the source, discusses its implementation methods and material requirements, and plans for material testing studies. Full article

There has been increasing interest in deploying Small Modular Reactors (SMRs) due to their simplicity, enhanced safety features, and economic advantages, which may facilitate the transition from coal to nuclear energy. However, the revival of nuclear power today depends on reactors meeting long-term operational security requirements, which can lead to optimized costs for nuclear energy and greater public acceptance. Therefore, it is crucial to demonstrate best practices in the operation, reliability, and stability of the systems and materials used in construction to ensure that nuclear power plants can operate safely over extended periods. Corrosion remains a critical factor affecting the safe operation of these plants. While corrosion issues have been extensively studied in traditional nuclear reactors that use water as a coolant, advanced reactors employing non-water coolants—such as liquid metals or molten salts—present new corrosion challenges. This work aims to present the various sources of corrosion in different SMR cooling systems, along with the results of corrosion processes. It also discusses the challenges related to materials used in multiple SMR designs and highlights advancements in the development of new materials suitable for use in SMRs. Full article

Growing environmental awareness has renewed interest in thorium as a nuclear fuel, underscoring the need for further studies to evaluate how reactors perform when conventional fuels are replaced with thorium-based alternatives. In this study, thermal hydraulics and solid mechanics computations were simulated using COMSOL multiphysics to investigate the safe operating conditions of a heavy water reactor with thorium-based fuel. The thermo-mechanical analysis of the fuel rod under transient heating conditions provides critical insights into strain, displacement, stress, and coolant flow behavior at elevated volumetric heat sources. After 3 s of heating, the strain distribution in the fuel exhibits a high-strain core surrounded by a low-strain rim, with peak volumetric strain increasing nearly linearly from 0.006 to 0.014 as heat generation rises. Displacement profiles confirm that radial deformation is concentrated at the outer surface, while axial elongation remains uniform and scales systematically with power. The resulting von Mises stress fields show maxima at the outer surface, increasing from ~0.06 to 0.15 GPa at the centerline with higher heat input but remaining within structural safety margins. Cladding simulations demonstrate nearly uniform axial expansion, with displacements increasing from ~0.012 mm to 0.03 mm across the investigated power range, and average strain remains negligible (≈10⁻⁴), while mean stresses increase moderately yet stay well below the yield strength of zirconium alloys, confirming safe elastic behavior. Hydrodynamic analysis shows that coolant velocity decreases smoothly along the axial direction but maintains stability, with only minor reductions under increased heat sources. Overall, the coupled thermo-mechanical and fluid-dynamic results confirm that both the fuel and cladding remain structurally stable under the studied conditions. By using COMSOL’s multiphysics capabilities, and unlike most legacy codes optimized for uranium-based fuel, this work is designed to easily incorporate non-traditional fuels such as thorium-based systems, including user-defined material properties, temperature-dependent thermal polynomial formulas, and mechanical response. Full article