核动力工程

Numerical study on the structure and operational characteristic of plate-type passive autocatalytic recombiners

, Available online , doi: 10.13832/j.jnpe.2025.04.0150

Abstract(0) PDF(0)

Abstract:
Passive autocatalytic recombiners (PARs) utilizes the hydrogen catalytic oxidation reaction mechanism to mitigate the risk of hydrogen deflagration during severe accidents in large pressurized water reactor containment. This study developed a mass transfer-reaction coupled numerical model for H₂/O₂ mixed gas on a platinum (Pt)-based catalyst, systematically analyzing the effects of operational parameters (concentration, temperature, flow rate) and structural parameters (plate spacing, height, thickness) on the comprehensive performance of the plate-type PARs. The results indicate that increasing hydrogen concentration and gas temperature can directly enhance the catalytic reaction rate. When the flow rate increases from 0.2 m/s to 1.0 m/s and the plate spacing enlarges from 8 mm to 15 mm, the increased hydrogen mass flow rate accelerates the catalytic reaction rate, but the hydrogen elimination rate decreases by 16.23% and 9.59%, respectively. The high-activity region at the leading edge of the catalytic plate (﹤30 mm)exhibits a reaction rate an order of magnitude faster than that of the middle and rear sections (﹥75 mm). The variation in the catalytic plate thickness has a minimal impact on hydrogen concentration, and the average temperature difference of the catalytic plate is less than 20 K.

Research on Underwater Local Dry Ring Laser Lap Welding of 316L Stainless Steel Patch

, Available online , doi: 10.13832/j.jnpe.2025.12.0198

Abstract(0) PDF(0)

Abstract:
Abstract：This study focuses on the lap repair of cladding in spent fuel pool in nuclear power plants, employing an independently designed underwater laser welding torch. We conducted overlap fillet experiments on 3mm-thick 316L austenitic stainless steel plates at a water depth of 0.5m using an adjustable ring-mode laser system. As laser power increases, the throat size of the weld first decreases and then increases before decreasing again, the penetration depth of the base plate shows an increasing trend, the leg size of the weld gradually increases, the wetting angle gradually decreases, and the spreadability of the weld is improved. Under the experimental conditions, the weld metal exhibits a Ferrite-Austenite(FA) solidification mode, and the weld structure consists of γ-austenite and a significant amount of residual δ-ferrite. By adjusting the proportion of the ring laser, it is found that the central power has a significant effect on the growth of columnar crystals at the bottom of the melt pool, while the ring power affects the number of equiaxed crystals in the upper part of the melt pool. Increasing the proportion of ring power refines the equiaxed crystals at the center of the weld and makes the weld structure more uniform, thereby increasing the microhardness at the center of the weld. This study provides important process parameters and relationships between microstructure and properties for underwater laser welding repair of the bottom plate of spent fuel pool in nuclear power stations.

Research on the Numerical Simulation of Neutron Noise based on the Unstructured-Grid Variational Nodal Method

, Available online , doi: 10.13832/j.jnpe.2025.02.0064

Abstract(15) PDF(2)

Abstract:
Neutron noise is the stochastic fluctuation of the neutron field under the influence of various weak perturbations, attracting attention as an indirect monitoring tool for nuclear reactors. This paper establishes a frequency-domain numerical simulation method for neutron noise based on the unstructured-grid variational nodal method using the first-order perturbation theory and Fourier transform techniques. This work extends the functionality of the general-purpose neutron transport code VITAS. The accuracy of the frequency-domain method is verified by comparing its results with those from time-domain methods in the MOX case and the C4V benchmark. Numerical results indicate that the frequency-domain method can achieve a precision comparable to the existing time-domain method under weak perturbations. The relative error in the amplitude of neutron noise is less than 2%, and the relative error in the phase of neutron noise is less than 1%.

Study and Application of Construction Significance Determination Process for Nuclear Power Plant

, Available online , doi: 10.13832/j.jnpe.2025.04.0144

Abstract(12) PDF(0)

Abstract:
In order to evaluate the effectiveness of licensee oversight and quality assurance efforts associated with construction activities, the method of Construction Significance Determination Process (CSDP) was developed based on risk-informed approach and technology. The CSDP evaluation model for the VVER-1200 reactor and application process were established, which have been applied during the construction supervision for the VVER-1200 nuclear power plants. The application was carried out for two typical cases. The results showed that quantitative evaluation for abnormal matters can be achieved, and CSDP can help regulators focus on the most critical areas for safety, so as to optimize regulatory resources and improve regulatory efficiency.

Development and Process of CorTAF: A Three-Dimensional Cross-Scale Multi-Physics Coupling Analysis Code for Nuclear Reactor Core

, Available online , doi: 10.13832/j.jnpe.2025.05.0206

Abstract(37) PDF(5)

Abstract:
　　
　　The reactor core is a critical component of nuclear power systems with a complex geometric structure, and it experiences strong coupling effects between various physical fields. High-precision thermal-hydraulic and multi-physics coupling analysis of the core is essential for ensuring the safety design and safety analysis of advanced nuclear power systems. The Nuclear Reactor Thermal-Hydraulic Laboratory (NuTHeL) at Xi'an Jiaotong University has developed a full-core neutronic-thermal-fluid-deposition multi-physics coupling analysis model, and has independently developed the CorTAF code for three-dimensional analysis at the full-core channel level based on the open-source CFD platform. Based on CorTAF code, a cross-scale coupling strategy between the core model of CorTAF and the detailed three-dimensional thermal-hydraulic model of the pressure vessel was proposed, which allows CFD-based multi-physics calculations and predictions for the entire pressure vessel. Validation and verification work has also been conducted based on international benchmark problems. In recent years, the research team has continually developed and refined the mathematical and physical models of the code. Currently, the CorTAF code supports cross-scale coupling calculations for multiple reactor types (PWR, LFR, SFR), physical fields (Neutronics, Thermal hydraulic, Deposition), and system structures (Core, Lower plenum, Upper plenum). This paper reviews the development process of the CorTAF series codes, presents their main functions and applications in PWR calculation, summarizes the current computational results, and discusses the future direction of the program's development.
　　The reactor core is a critical component of nuclear power systems with a complex geometric structure, and it experiences strong coupling effects between various physical fields. High-precision thermal-hydraulic and multi-physics coupling analysis of the core is essential for ensuring the safety design and safety analysis of advanced nuclear power systems. The Nuclear Reactor Thermal-Hydraulic Laboratory (NuTHeL) at Xi'an Jiaotong University has developed a full-core neutronic-thermal-fluid-deposition multi-physics coupling analysis model, and has independently developed the CorTAF code for three-dimensional analysis at the full-core channel level based on the open-source CFD platform. Based on CorTAF code, a cross-scale coupling strategy between the core model of CorTAF and the detailed three-dimensional thermal-hydraulic model of the pressure vessel was proposed, which allows CFD-based multi-physics calculations and predictions for the entire pressure vessel. Validation and verification work has also been conducted based on international benchmark problems. In recent years, the research team has continually developed and refined the mathematical and physical models of the code. Currently, the CorTAF code supports cross-scale coupling calculations for multiple reactor types (PWR, LFR, SFR), physical fields (Neutronics, Thermal hydraulic, Deposition), and system structures (Core, Lower plenum, Upper plenum). This paper reviews the development process of the CorTAF series codes, presents their main functions and applications in PWR calculation, summarizes the current computational results, and discusses the future direction of the program's development.

Multiscale Numerical Study on the Tensile Deformation Behavior of Small Size Specimen

, Available online

Abstract(34) PDF(3)

Abstract:
Based on the uniaxial tensile testing of small size specimen made from domestic A508-III steel, a macroscopic mechanical constitutive model and a ductile damage model were established, and the crystal plasticity parameters at the microscale were calibrated. A multiscale numerical model for the uniaxial tensile behavior of small size specimen was established by combining macroscopic finite element method with microscopic crystal plasticity. The tensile mechanical response behavior of small size specimen was explored at the macroscopic scale, and the plastic deformation mechanism was explained at the microscopic scale. The results show that the tensile testing of small size specimen exhibits a certain discreteness and the tensile fracture surface shows obvious ductile fracture characteristics. At the initial stage of tension, plastic deformation is primarily achieved through the uniform dislocation motion. As strain increases, the strain localization and stress concentration at the grain scale become evident at the microscale, especially after necking, the non-uniform plastic deformation becomes more significant, and dislocations begin to show obvious localization phenomena. The geometrically necessary dislocation (GND) density increases rapidly from 16 μm⁻² at a strain of 8% to 65 μm⁻² at a strain of 10%. Throughout the plastic deformation process, the evolution ofstatistically stored dislocation (SSD) density plays a dominant role. Due to the uneven stress distribution and dislocation pile-up, an "orange peel effect" is observed on the specimen surface. Grain boundary have a significantly impact on the dislocation evolution, and the dislocation accumulation at grain boundary follows the rule that the larger the grain misorientation, the higher the dislocation density.

Numerical Simulation Study on the Influence of Cr-Coated Zirconium Alloy Cladding on Activated Corrosion Products in PWR

, Available online

Abstract(23) PDF(0)

Abstract:
Cr-coated zirconium alloy claddings, as accident-tolerant fuel cladding schemes, have garnered significant attention in the nuclear materials field due to their excellent anti-oxidation performance, low thermal neutron absorption cross-section, and superior thermomechanical properties. This study focuses on the CPR1000 nuclear power unit, replacing all fuel claddings with Cr-coated zirconium alloys to systematically assess the impact of Cr coatings on activated corrosion products in the primary circuit. Through numerical simulation methods, this research thoroughly analyzes the deposition characteristics of radioactive nuclides in steam generators, main pipelines, and reactor cores. The study results reveal that the corrosion release of Cr coatings has a greater influence on the activated corrosion products inside the reactor compared to outside. Importantly, the application of Cr coatings does not alter the dominant position of ⁶⁰Co, this indicates that their impact on existing reactor operation modes is manageable. Such findings provide crucial theoretical and data-based support for the practical application of Cr-coated zirconium alloy claddings in nuclear power plants.

Current status and prospects of rare nuclides irradiation production technology through ultra-high flux reactors

, Available online

Abstract(49) PDF(3)

Abstract:
²³⁸Pu, ²⁵²Cf and other rare nuclides have important applications in nuclear energy, aerospace and other fields. The irradiation production of rare nuclides faces challenges such as complex conversion chain, large fission loss, and extremely low yield, and typically requires irradiating targets under ultra-high neutron flux. Ultra-high flux reactor is the most important facility for large-scale preparation of rare nuclides. Currently, our country doesn’t have the capacity for large-scale preparation of rare nuclides and relies entirely on imports. Realizing the independent and stable supply of rare nuclides is of great significance for the development of strategic key areas in China. The key technologies for producing rare nuclides through irradiation in ultra-high flux reactors include preparation techniques for super-heavy target materials, optimization design techniques for irradiation targets, irradiation techniques in ultra-high flux reactors, separation and purification techniques, etc. This paper analyzes the current status and key technologies for producing rare nuclides through irradiation in ultra-high flux reactors, and looks forward to the development strategy of rare nuclides preparation in China.

Experimental and CFD Simulation of the RPV Head Plenum

Chen Yongchao, Wei Xingfang, Liu Yanwu, Fang Jian, Ran Xiaobing

, Available online , doi: 10.13832/j.jnpe.2024.070004

Abstract(42) HTML (26) PDF(5)

Abstract:
This paper aims to investigate the flow characteristics of the RPV Head Plenum region in the HPR1000 Pressurized Water Reactor (PWR) to provide support for addressing the erosion of the Thermal Sleeve in operational CPR1000 PWR plants and optimizing the RPV Head Plenum region structures in HPR1000 PWR plants. This paper uses Computational Fluid Dynamics (CFD) methods to conduct numerical simulations of the RPV Head Plenum region and also carries out hydraulic model experiments of the RPV Head Plenum to obtain the flow distribution and hydraulic characteristics of key areas within the RPV Head Plenum region. The theoretical analysis and experimental results show that: The CFD results of the key areas in the RPV Head Plenum and the measured lateral velocity deviations from the experiments are within 10%. Under normal operating conditions, the overall flow velocity within the RPV Head Plenum is relatively low, but higher velocities are observed near the RPV Head nozzles and nearby surface. The coolant water in RPV Head Plenum flows downward through the holes at the top of CRGT and then flows into the Upper Plenum, which is consistent with the expectation of the HPR1000 “Cold RPV head”. Additionally, a vortex with higher flow velocity is present near the Thermal Sleeve Bell Mouth in the central region than in the peripheral area of the RPV Head Plenum, resulting in more severe fluid impact and wear on the thermal sleeves.

Large-eddy Simulation of Temperature Fluctuation Characteristics of Lead Bismuth Eutectic Alloy in Triple Jet

Guo Chao, Xu Jiangming, Liu Songtao, Miao Yiran

, Available online , doi: 10.13832/j.jnpe.2024.060038

Abstract(38) HTML (30) PDF(0)

Abstract:
In order to study influence of velocity on LBE temperature fluctuations, temperature fluctuation characteristics of lead bismuth eutectic alloy in triple jet was simulated. Firstly, three turbulent models were used to analyze temperature fluctuations characteristics of sodium in triple jet. The temperature and velocity fields at different times were evaluated and temperature fluctuation characteristics were predicted successfully by large eddy simulation (LES) and LES is suitable for analyzing temperature fluctuations characteristics of liquid metal. Secondly, LES was employed to simulate temperature fluctuation characteristics of lead bismuth eutectic alloy(LBE) in triple jet at different velocities and velocity ratios. Finally, by analyzing the influence of different velocity ratios and different velocities, it is found that the amplitude and frequency of temperature fluctuation increases as the velocity increases due to that increase of velocity enhanced turbulence. The temperature fluctuation characteristics of LBE alloy in triple jet were analyzed and it could be anticipated to support the subsequent investigation of temperature fluctuations of LBE cooled fast reactor.

Analysis of Heat Transfer Characteristics of Steam Generator with Axial Economizer Based on RELAP5

Huang Zhongyuan, Wang Xiaoding, Li Zhenzhong, Liu Haidong, Chen Deqi

, Available online , doi: 10.13832/j.jnpe.2024.050032

Abstract(33) HTML (25) PDF(2)

Abstract:
The vertical natural-circulation steam generator, as a crucial equipment in pressurized water reactor nuclear power plants, enhancing its heat transfer performance is vital to the economy of the entire power plant. This study selects the steam generator of AP1000 as the research object and employs the RELAP5 system analysis program to calculate and analyze both the conventional steam generator and the steam generator with axial economizer. The heat transfer mechanism of the steam generator with axial economizer is studied, and focus on the effects of different heights of the divider plate and the recirculated water distribution ratio on the heat transfer characteristics. The results show that the steam generator with axial economizer can significantly increase the heat transfer temperature difference between the primary and secondary sides, thereby effectively enhancing the overall heat transfer efficiency. In addition, the study finds that increasing the height of the divider plate can improve the heat transfer capacity to a certain extent, and there is an optimal height at which the heat transfer power reaches its peak. Moreover, reducing the recirculated water distribution ratio helps to increase the heat transfer temperature difference, further improving the heat transfer performance. This research provides reference for the engineering analysis and design of the steam generator with axial economizer.

Numerical Simulation Research on Natural Circulation Flow of the Reactor Coolant System

Zhang Mingqian, Lin Run, Li Zhenguang

, Available online , doi: 10.13832/j.jnpe.2024.050038

Abstract(49) HTML (28) PDF(5)

Abstract:
Computational Fluid Dynamics(CFD) program is employed to enable the high-fidelity modeling of the reactor coolant system (RCS) for a typical three-loop pressurized water reactor, and the completed model of the RCS is build including reactor vessel and internals, core, steam generator, primary pump and linking pipe. The three-dimensional, global and localized flow features have been investigated under natural circulation flow condition with lower core thermal power, and the temperature at different locations are compared with the measured values from the operating nuclear power plant in order to verify the accurate description of the developed CFD model. The results show that the natural circulation flow rate is about 4.5% of the full power flow rate while the temperature of the core outlet is stable, and the residual core heat could be effectively removed. The phenomenon of the thermal stratification in the reactor pressure vessel head dome shows that the measured temperature value of the detector position in nuclear power plant could not provide the highest value. The coolant from different loops could be more fully mixed due to the local convection flow. There is a swirling flow at the outlet of primary pump, and the tangential velocity near the pipe wall is large while the local convection occurs at the central area. This analysis practice provides an effective evaluation for the system-level three-dimensional thermal hydraulics phenomena of the reactor coolant system.

Experimental Study of Geyser Boiling in High-temperature Sodium Heat Pipe at Inclined Angle Condition

Yang Siyuan, Ma Yugao, Wen Qinglong, Wen Shuang, Ding Shuhua, He Linfeng, Yuan Bo

, Available online , doi: 10.13832/j.jnpe.2024.060020

Abstract(57) HTML (24) PDF(3)

Abstract:
To study the geyser boiling phenomenon in the start-up process of high-temperature alkali metal heat pipe and provide reference operating conditions for the safe operation of the heat pipe reactor, sodium metal was used as the working medium to investigate the influence factors and mechanism of geyser boiling during the start-up process of the heat pipe. The results show that the heating power and inclination angle of the heat pipe have important effects on geyser boiling. Under the condition of 90° inclination angle, the heating power increases from 600 W to 750 W, the geyser boiling period varies from 736 s to 29 s, and the temperature amplitude ranges from 35℃ to 18℃. Geyser boiling is easy to occur under medium heating power conditions but will not happen when the inclination of the heat pipe is 0°. With the increase of inclination angle, the heating power of the start and stop of the geyser boiling decreases. When the inclination angle is 45°, 60°, 90° respectively, the geyser boiling starts at 250, 200, 150 W and stops at 600, 450, 350 W. The geyser boiling period varies at the same heating power from different inclination angles, but the temperature amplitude changes little. The geyser boiling intensity decreases and the power range of geyser boiling advances with the shortening of the condensing section. The results of this study laid a foundation for further research on the geyser boiling mechanism of alkali metal heat pipes, and provide important data and theoretical support for the design optimization of alkali metal heat pipes and the safe operation of heat pipe reactors.

Online First