This paper investigates the accelerating structure parameters of a compact high-voltage accelerator neutron source. A three-dimensional model of the acceleration system was established using SIMION. Following the optimization of the electrode aperture geometry to reduce the peak electric field intensity, the effects of the accelerating voltage, electrode aperture, and electrode gap on beam transmission, target beam spot size, and beam power density were systematically simulated and analyzed. The results indicate that the regulation of the acceleration voltage has a significant influence on beam focusing characteristics. Specifically, increasing the accelerating voltage enhances the radial confinement of the beam, leading to a substantial reduction in the beam spot size on the target. The size of the accelerating electrode aperture modifies the potential gradient within the accelerating field, thereby facilitating beam focusing. Furthermore, the inter-electrode distance should be controlled within 60-65 mm. In this range, the maximum electric field intensity remains below 6.2 MV/m, effectively reducing the risk of high-voltage breakdown. Simultaneously, the beam spot diameter on the target is maintained at approximately 50 mm, preventing arcing phenomena caused by beam impingement on the electrodes. Consequently, the peak power density is controlled below 5.2 W/mm2, extending the service life of the target. These findings provide a critical theoretical basis and optimization guidelines for designing electrode structures in high-voltage high-current accelerators.
The China Fusion Engineering Demonstration Reactor (CFEDR) serves as a vital bridge toward commercial fusion energy, necessitating a highly efficient Power Conversion System (PCS) to manage its multi-grade heat and pulsed operation. This study presents the thermodynamic design and optimization of a steam Rankine cycle integrated with Thermal Energy Storage (TES) for the CFEDR. Based on a continuous high-temperature thermal input of 1081.5 MW and a low-temperature input of 276.75MW, thermodynamic models were established to evaluate different cycle configurations and parameter sensitivities. Among the two complete candidate engineering schemes considered, the subcritical three-cylinder Rankine cycle scheme exhibited better overall thermodynamic performance under the adopted design conditions, achieving a net electrical power output of 487.53 MW and a cycle efficiency of 45.08%. Parametric analyses yielded optimal main steam and exhaust pressures of 17MPa and 5kPa, respectively, balancing thermal efficiency, component stress, and cooling water demands. Furthermore, the integration of low-grade waste heat into the feedwater preheating system was investigated. While replacing low-pressure regenerators with a low-grade heat preheater is thermodynamically viable, terminal temperature difference constraints restrict the maximum utilizable low-grade heat to approximately 220MW. A counter-case analysis confirmed that drastically adjusting main steam parameters to accommodate total low-grade heat absorption significantly penalizes overall net power. Consequently, this study establishes an optimized Rankine cycle baseline and highlights the necessity for complementary bottoming cycles to fully exploit residual low-grade thermal energy.
Metal hydrides possess excellent neutron moderating power and favorable thermal conductivity, making them ideal high-temperature neutron moderator materials for micro-nuclear reactors. Zirconium hydride and yttrium hydride are two typical solid moderator materials applied in nuclear energy systems. Compared with zirconium hydride, yttrium hydride exhibits superior high-temperature thermal stability, which makes it more adaptable to the application requirements of the current mainstream high-temperature microreactor designs and has thus attracted extensive industrial attention in recent years. However, the performance database to support engineering application of yttrium hydride in reactors remains incomplete at present. In this paper, the domestic and international research progress on the preparation, thermodynamic properties, basic physical properties and irradiation effects of yttrium hydride is systematically summarized. On this basis, the key research and development requirements for yttrium hydride neutron moderators are further deliberated, to provide comprehensive and reliable references for the design and engineering application of this rare earth-based functional material in microreactors.
Lead-cooled fast reactors (LFRs) represent a highly promising fourth-generation nuclear technology, characterized by low operating pressures, inherent safety, and compact core configurations. Within LFRs, hexagonal fuel assemblies rely heavily on wire-wrapped structures to ensure subchannel stability and enhance thermal-hydraulic performance. However, existing studies predominantly focus on isolated geometric variations, leaving the synergistic multi-parameter effects in heavy liquid metals insufficiently explored. Addressing this gap, this study conducts a comprehensive computational fluid dynamics (CFD) investigation on the Lead-Bismuth Eutectic (LBE) cooled 19-pin wire-wrapped assemblies based on the MYRRHA core. Following a rigorous validation against experimental benchmark data, a systematic parametric analysis is executed to evaluate the coupled effects of wire pitch, wire diameter, and the number of wires. Utilizing the field synergy angle and Performance Evaluation Criteria (PEC), the underlying fluid-structure interactions are explicitly elucidated to identify thermally optimized configurations. Furthermore, drawing upon the extensive CFD dataset, a new multivariate non-linear empirical model—designated as the UCTD-L correlation—is developed to predict the friction resistance coefficient. By incorporating coupled structural correction factors, the UCTD-L model reliably captures the drag characteristics of diverse geometries. Validated against an independent out-of-sample test dataset, the proposed correlation restricts the Mean Relative Error (MRE) strictly below 20%.
The core of a specific pool-type research reactor utilizes multiple closed plate-type fuel assemblies. Non-uniform coolant flow distribution among these assemblies can lead to localized overheating, posing a safety risk. This study investigates the flow distribution characteristics of the reactor core using a combined approach of theoretical modeling and Computational Fluid Dynamics (CFD) simulation, validated by full-scale experiments. The key findings are: (1) The results from the proposed methodology show a deviation of less than 2% from the experimental data, demonstrating high accuracy. (2) Both analytical and experimental results confirm a uniform flow distribution across different core assemblies, with flow distribution factors ranging from 0.97 to 1.02, thereby meeting the core thermal-hydraulic safety requirements. (3) The reactor retains a 1.82% design margin in bypass flow, which facilitates future target design and enhances operational economics. (4) The flow distribution factors stabilize once the experimental flow rate exceeds 70% of the nominal operating flow rate, indicating that the core has reached a hydraulically stable state.
Verification serves as the prerequisite for the engineering application of independently developed nuclear safety analysis software. Verification methods for reactor thermal-hydraulic system analysis software are studied, and the framework and workflow for verification were proposed. Based on the development and application of reactor thermal-hydraulic numerical algorithms, nine benchmark problems were identified and selected. Multiple nuclear power research institutions of China were invited to perform comparative calculations using independently developed software. Analysis of the computational results revealed the influence of factors such as boundary conditions, numerical algorithms, interphase friction, and virtual mass force on software verification results. Consequently, the boundary conditions and application requirements for the benchmark problems were standardized, the comparison and acceptance standards for the benchmark problems were unified, and the standards framework have been established for the verification work of China's independent developed reactor thermal-hydraulic system analysis software.
The disturbance wave characteristics of annular flow in narrow rectangular channels are crucial for the accurate prediction of dry-out type critical heat flux in nuclear reactor fuel assembly systems. To investigate the wave evolution mechanisms and associated characteristics of annular liquid films in a horizontal narrow rectangular channel, the adiabatic air-water flow experiments were carried out using a test section with dimensions of 30mm × 1mm×1000 mm. An electrical-based liquid film sensor matrix for narrow rectangular geometries was developed to enable high-frequency acquisition of liquid film thickness time-series data on the bottom wide wall. The mean amplitude and frequency of the disturbance waves on bottom wide wall of the horizontal narrow rectangular channel were quantitatively extracted by a threshold identification method. Experimental findings reveal that the mean disturbance wave amplitude exhibits a pronounced upward trend as the superficial liquid velocity rises. Meanwhile, the disturbance wave frequency demonstrates a positive correlation with superficial liquid velocity and increases moderately with superficial gas velocity, without exhibiting a strong linear dependence on gas velocity in the present narrow rectangular channel geometry. Based on these observations, predictive correlations for both amplitude and frequency of disturbance waves within horizontal narrow rectangular ducts were formulated and subsequently validated against both the present experimental dataset and independently published data from existing literature. The validation outcomes confirm that prediction deviations for both models remain bounded within ±25%, demonstrating satisfactory accuracy and practical applicability in engineering contexts. The present study advances the mechanistic understanding of interfacial fluctuation dynamics and disturbance wave development in annular flows confined within rectangular geometries, while simultaneously furnishing a theoretical foundation for assessing dry-out risks and guiding the safety design of compact nuclear reactor fuel assemblies.
This paper proposes a Bayesian framework for a distribution-free fault-tree analysis using a probability-boxes and confidence-boxes to estimate the reliability of a complex coherent system. The approach addresses key challenges arising from limited component reliability data, available as Boolean observations, “k-out-of-n” data, or reported probability values, alongside the uncertainty about the underlying distribution class and the statistical dependence between component and system failures. To overcome these issues, a calculus of distribution-free probability-boxes and confidence-boxes is implemented to characterise the aleatory and the epistemic uncertainties while also accounting for the uncertain event-level statistical dependence within the fault-tree structure. The framework enables the propagation of these uncertainties to obtain bounds on the Top-event probability. A novel interpretation of the resulting distribution-free output based on statistical performance is provided, addressing a gap in the existing literature. The proposed methodology is demonstrated through a real-world case study of a reactor coolant subsystem, with results validated against a reference top event probability, illustrating the feasibility and effectiveness of the approach.
To address the high risk of high-temperature and high-pressure physical testing, the long testing cycle, and the low efficiency of control logic integration in integrated safety system tests for advanced pressurized water reactors, this study proposes a real-time digital twin simulation platform for the AHPR1000 integrated safety system test facility based on closed-loop coupling between a mechanistic model and a virtual distributed control system (DCS). According to the requirements of steady-state operation, accident transient simulation, control logic verification, and operating condition reproduction, a full-scope thermal-hydraulic model covering 14 subsystems, including the reactor coolant system, pressurizer system, main steam system, feedwater system, safety injection system, and residual heat removal system, was established. A large-scale tag mapping, data interaction, and operating state management mechanism was further developed between the model side and the virtual DCS side. The platform supports real-time monitoring and control of more than 3000 tags and provides functions such as freeze/unfreeze, reset, snapshot saving, online parameter tuning, and multi-scenario switching. The steady-state validation results show that the platform reproduces key thermal-hydraulic parameters with good accuracy and can operate continuously and stably for more than 100 h. The transient validation results, taking a 2-inch cold-leg small-break loss-of-coolant accident as an example, demonstrate that the platform can reasonably reproduce key processes including primary loop depressurization, safety system actuation, tank level variation, and accident mitigation. The results indicate that the proposed method can effectively balance system-level physical fidelity, real-time performance, and closed-loop control integration requirements, and can provide technical support for safety system testing, control strategy verification, operator training, and digital twin applications in advanced nuclear power systems.
To investigate the impact of coolant flow within a 2×2 control rod guide tube (CRGT) assembly on the deformation and wear of control rods, this study employed a synergistic approach utilizing both experimental measurements and numerical simulations. The flow characteristics, along with control rod strain and deformation, were systematically analyzed from the upper nozzle to the continuous guide section outlet. Flow fields at critical internal locations were captured via Particle Image Velocimetry (PIV). To enhance predictive fidelity, the constants of the Realizable k-ϵ turbulence model were recalibrated using an Ensemble Kalman Filter (EnKF) algorithm. Subsequent fluid-structure interaction (FSI) simulations, grounded in these optimized constants, provided high-resolution data on unmeasured flow regions and mechanical responses. The experimental results indicated a highly non-uniform flow distribution within the CRGT. Among the flow holes on the core upper plate, the flow distribution ratio of the upper guide tube holes was 30% higher than that of the support column holes and 17% higher than that of the unobstructed flow holes. Pronounced transverse flow were identified near the hold-down springs and flange gap. After recalibration through data assimilation, the velocity error between the simulations and experiments at the gap between upper nozzle and guide tube was reduced from 10% to 4%. FSI analysis further revealed that the maximum principal strain was concentrated between the upper nozzle and the guide tube flange, where control rods were subjected to opposing fluid-induced excitation forces.
Passive Containment Cooling System (PCS) and the Secondary Passive Residual Heat Removal System (PRS) are two major passive systems that ensure the safety of nuclear power plants under accident conditions. Aiming at the engineering pain points of the existing integrated concrete water tank in HPR1000, including high construction difficulty, long construction period, and low utilization rate of lower tube bundles caused by uneven pool boiling heat transfer, this paper proposes a new separated PCS/PRS steel water tank design. A numerical model of the natural circulation heat transfer system for the separated water tank was established using the RELAP5 system code. The transient thermal-hydraulic characteristics of the separated water tank under accident conditions were systematically investigated. The heat transfer performance difference between the separated water tank and the original HPR1000 integrated water tank was quantitatively compared. The results show that the separated water tank significantly enhances the natural circulation driving force by utilizing the height difference between the upper and lower tanks, increasing the steady-state heat transfer power compared with the original scheme, and effectively improves the uneven heat transfer defect of pool boiling. The system can stably remove core residual heat within the 40-h design period, meeting the safety requirements of design extension conditions. The proposed separated steel water tank realizes factory prefabrication and integral hoisting, shortening the on-site construction period by about 30%, and provides a new technical approach for the optimization of passive safety systems in nuclear power plants.
Chalk River Unidentified Deposits (CRUD) on nuclear fuel cladding present persistent challenges in pressurized water reactors (PWRs), including cladding corrosion, elevated radiation fields, and CRUD-induced power shifts (CIPS). To better understand these effects, several Multiphysics models have been developed in recent years. Building on these advances, this study introduces a fully coupled two-dimensional Multiphysics framework that integrates two-phase film boiling, solute transport, chemical equilibria, and radiolytic decomposition, a process usually overlooked in existing film boiling models. The model captures vapor film formation within CRUD under high heat flux, along with local temperature, pressure, and solute accumulation. It shows improved agreement with WALT benchmark data compared with previous models. Results indicate that boron accumulation within the CRUD increases the local saturation temperature, while pH rise at the liquid-vapor film interface promotes LiBO2 precipitation under extreme thermal conditions. Parametric studies further reveal how deposit thickness, porosity, pore size, and chimney density influence cladding temperature, vapor film thickness, pH, and boron accumulation.
In the event of a steam generator tube rupture (SGTR) accident in a liquid lead-bismuth eutectic (LBE)-cooled reactor (LFR), a small rupture in the tube wall may lead to the leakage of high-pressure, low-temperature liquid water into a low-pressure, high-temperature pool of liquid LBE. Consequently, the liquid water quickly vaporizes, and a chain of steam bubbles may be formed following the continuous leakage of liquid water. Existing drag models, primarily developed for isolated bubbles, show significant limitations when applied to bubble chains, in which interactions between bubbles are observed. This study investigates the effect of the spacing between adjacent bubbles on the terminal velocity in a chain in stagnant LBE and develops a novel drag coefficient model that incorporates this key parameter. Bubble dynamics obtained using neutron radiography, including terminal velocity in a chain and equivalent bubble diameter of a single bubble chain rising in stagnant LBE, were used to validate the volume-of-fluid (VOF) method. Subsequently, extensive numerical simulations covering a wider range of parameters were conducted to establish a comprehensive drag coefficient database. The numerical results demonstrated that the drag coefficient of the bubbles in the bubble chain decreases as the ratio of the bubble spacing to the equivalent bubble diameter decreases. A new drag model was proposed by introducing this ratio into the functional form of the correlation along with the bubble Reynolds number and Weber number. The proposed drag coefficient model shows excellent agreement with the numerical databases and significantly outperforms existing models.
This study investigates the influence mechanisms of primary-side inlet temperature and flow rate on the axial temperature distribution of heat transfer tubes under transient conditions, based on a single-channel primary circuit numerical model of the once-through steam generator (OTSG). The results demonstrate that the temperature variation rate governs the coupled thermal inertia-flow inertia effects. In low-slope regions, the system exhibits quasi-steady-state characteristics with less than 2% variation in evaporation section length. Conversely, high-slope regions trigger rapid dry-out zone expansion, where void fraction exceeding critical thresholds induces flow regime transition. This initiates a positive feedback loop between void growth and heat transfer deterioration, significantly amplifying local overheating risks. Flow inertia regulates phase change reconstruction processes: rapid pump startup induces 0.3L evaporation front migration with temperature fluctuations 2.3 times greater than slow startup, whereas gradual startup limits migration to 0.1L with controllable temperature fields. The research reveals synergistic interaction mechanisms among thermal inertia, flow inertia, and latent heat absorption during transients, providing theoretical foundations for OTSG safety operation and transient control strategies. Key findings highlight that thermal inertial effects dominate at low perturbation rates, while flow inertia becomes decisive in rapid transients through momentum-driven phase redistribution.
Molybdenum-99 (Mo-99) is a crucial radioisotope widely used in nuclear medicine as the precursor of technetium-99 m (Tc-99 m), which plays a significant role in medical imaging and diagnosis. This study analyzes the potential of MoO3 as a target material for Mo-99 production using the IAEA 10 MW benchmark reactor developed by the International Atomic Energy Agency (IAEA). Monte Carlo simulations were performed using open-source Monte Carlo code, OpenMC, with ENDF/B-VIII nuclear data to estimate the activity of Mo-99 produced while also evaluating the local heating characteristics within the target. The analysis was conducted for natural Mo as well as enriched Mo-98 in the form of MoO3 target, ranging from 25% to 95%. The results indicate that increasing Mo-98 enrichment leads to a higher Mo-99 activity, while the target heating distribution in the MoO3 target shows a consistent power distribution across all enrichment levels, suggesting that higher enrichment does not significantly affect the target thermal load. Consequently, the existing cooling system does not have to undergo major modifications to facilitate high-enriched Mo targets. On the other hand, an optimal irradiation duration was found to be between 7 and 10 days, with the Mo-99 activity approaching saturation after approximately 12 days. These findings can be considered to prepare the reactor operation schedule while optimizing the Mo-99 production in the core.
Due to material oxidation, activation, and the presence of fission products and uranium, the metallic layer in a two-layer corium pool contains internal heat sources. This study investigated the influence of these additional heat sources on heat transfer characteristics using 3D CFD simulations, analyzing temperature distributions and wall heat flux patterns. The results reveal that the thickest solidified layer typically forms in the upper-middle section of the oxide layer rather than the cooler lower region. Beyond a certain height, the solidified layer rapidly diminishes, leading to significant wall erosion. While the Nu relationship established for conditions without metallic layer heat sources remains applicable when such sources are present, parameter c substantially impacts the metallic layer's heat transfer behavior. However, heat transfer correlations like the Nu relationship developed for heat-source-free conditions may not be valid when heat sources exist. Even experimental data from studies focusing solely on metallic layer with internal heat sources cannot be directly applied to analyze heat transfer in two-layer corium pools. The coefficient f was proposed to quantify the effect of metallic layer heat sources on heat transfer characteristics.
To support closure-model development for thermal-hydraulic analysis of SGTR/HXTR-related two-phase flow in lead-cooled fast reactors, a separate-effect experimental study was conducted on lead–bismuth eutectic (LBE)–nitrogen bubbly flow in a vertical circular pipe using a loop-type high-temperature liquid-metal two-phase-flow facility. The test section was a vertical pipe with an inner diameter of 50 mm and a length of 2000 mm. A single-sensor conductivity probe and four-sensor conductivity probes were employed for different measurement objectives. The single-sensor probe was used mainly to examine local void-fraction trends in the upper developed section, whereas the four-sensor probes were used to obtain local void fraction, bubble frequency, bubble velocity, equivalent bubble radius, and interfacial area concentration (IAC) at two representative axial sections and multiple radial positions.The measurements indicate that the void fraction exhibits clear spatial evolution along both the radial and axial directions. The radial void-fraction profile gradually changes from a relatively flat distribution in the near-inlet region to a center-peaked distribution at the upper measurement section. With increasing superficial gas velocity, both the local void fraction and the three-point area-weighted estimate of the cross-sectional void fraction increase. At relatively high superficial liquid velocity, the estimated cross-sectional void fraction tends to decrease because of the shortened gas residence time. Bubble velocity and equivalent bubble radius generally decrease along the flow direction, while the IAC varies in close correspondence with bubble frequency.These results provide a local-parameter database for the validation and improvement of interphase-interaction and interfacial-area closure models in LBE two-phase-flow simulations. The present study focuses on dispersed gas transport and bubbly-flow evolution in a high-density liquid-metal system, and should therefore be interpreted as a separate-effect bubbly-flow study relevant to SGTR/HXTR analysis rather than as a full reproduction of prototypical tube-rupture transients involving high-pressure steam release, condensation, flashing, and thermal shock.
In the progression of severe accidents in light water reactors, the retention of core melt in the reactor pressure vessel (RPV) lower head, known as In-Vessel Retention (IVR), constitutes a core strategy for ensuring nuclear power plant safety and preventing large-scale release of radioactive materials. Targeting the complex multi-component liquid-liquid phase separation phenomena within the melt pool and the resulting "thermal focusing effect," a coupled phase-field-flow-heat transfer multiphysics model based on free energy functionals was established. By simultaneously solving the Cahn-Hilliard equation, incompressible Navier-Stokes equations, and energy conservation equation in a two-dimensional axisymmetric domain, the dynamic evolution of a U-Zr-Fe-O quaternary system from an initially random uniform mixture to a macroscopic stratified state within the RPV lower head was simulated. Results demonstrate that the melt pool, driven by decay heat and interfacial energy, undergoes a "labyrinthine" bicontinuous structure evolution initiated by spinodal decomposition, ultimately forming a stable bilayer structure with a light metal layer above and a heavy oxide layer below under gravitational effects. The flow field evolves from initial small-scale chaotic convection (maximum velocity 0.0541 m/s) to large-scale circulation driven by thermal buoyancy (maximum velocity 0.00192 m/s), forming a stable configuration with an active upper convection zone and a stagnant lower dead zone. Due to the significant thermal conductivity contrast between the two stratified layers, wall heat flux exhibits a notable increase at the metal-oxide interface region, with peak heat flux reaching approximately 0.9×106 W/m2 during mid-evolution, attaining 60% of the upper CHF limit from ULPU experiments, posing a potential threat to IVR safety margins. Quantitative analysis of the phase separation time scale and its impact on wall thermal loading was conducted, providing numerical evidence and theoretical support for IVR safety margin assessment.
Addressing the issue of abnormally high silicon content in the In-containment Refueling Water Storage Tank (IRWST) of a Hualong One unit at a nuclear power plant during its first cycle and subsequent operation, this paper identifies two main pathways leading to the increase in silicon content through a systematic investigation of potential factors such as civil engineering materials, equipment operation, and the air environment, combined with leaching experiments and operational data coupling analysis. The first pathway is the release of initial exogenous sediment left over from the installation and commissioning stage during pump startup operation. The second is due to the "open-top" structural design of the IRWST; during power operation, under the unfiltered internal circulation mode of the EVR system, airborne silicon-containing particulates in the reactor building air continuously settle and dissolve into the water pool. The results indicate that the introduction from the air environment is the key root cause leading to the continuous deterioration of this issue. Ultimately, the plant successfully restored the water quality to within limits using a nanofiltration (SRD) device.
The air Brayton cycle is distinguished by its high efficiency, flexible start-up and shutdown capabilities, and environmental friendliness—attributes that render it a suitable energy conversion system for mobile small-scale reactors. As a core component of this cycle, the air turbine plays a critical role in system performance. Given that mobile small-scale reactors operate under dynamically varying conditions, the fatigue damage of air turbines is crucial for safeguarding the safe operation of the entire cycle system. This study establishes a thermo-fluid-structure coupling and fatigue life analysis workflow based on the ANSYS platform, and investigates the effects of variable load rate and inlet temperature change rate on the fatigue damage of an air centrifugal turbine during its load transition from 75% to 100% of the design power. The results indicate that when the load rate is 12.5%Pe·min-1, the fatigue damage from a single load-transient process reaches a maximum of 2.25×10-6. When the inlet temperature change rate is 12 K·min-1, the fatigue damage from a single temperature-change process reaches a maximum of 2.119×10-6. Both the variable load rate and inlet temperature change rate exhibit a positive correlation with the fatigue damage of centrifugal air turbines.
