核动力工程

The Key Technologies of Medical Isotope Test Reactor

, Available online , doi: 10.13832/j.jnpe.2025.06.0260

Abstract(5) PDF(0)

Abstract:
This article briefly introduces the application status, production principles, and production methods of isotopes, as well as the development overview of homogeneous solution-type reactors both domestically and internationally. It systematically elaborates on the system composition and design situation of isotope production experimental reactors, specifically including the reactor and its major systems, isotope extraction process systems, and supporting systems. Meanwhile, it provides a detailed explanation of the main technical issues related to reactor and isotopes extraction processes, such as reactivity stability, radiation protection design, prevention of fuel solution precipitation, corrosion resistance of structural materials, critical safety of fuel solutions, isotope extraction processes, uranium recovery technology, and fuel purification technology.

Research on Communication Method for Long-Distance Multi-Device Response Time Test of Reactor Protection System

, Available online , doi: 10.13832/j.jnpe.2025.03.0096

Abstract(18) PDF(1)

Abstract:
The response time test of the reactor protection system is crucial for ensuring the safety of the reactor. To address the challenges posed by measuring the response time of the Tricon platform reactor protection system, which involve multiple remote devices and long-distance transmission, this paper proposes a new communication method for response time testing. This method enables four-wire communication for measuring the response time of the Tricon platform reactor protection system by establishing a star topology suitable for devices distributed across different areas, particularly in situations where the rooms have high isolation or the devices are located at considerable distances. It uses four core wires to transmit both data and synchronization time signals, solving the problem of measuring the response time of multiple devices distributed across different areas. Furthermore, it has been successfully validated at the Fuqing Nuclear Power Plant, demonstrating the feasibility of the method.

Research on Machine Learning Methods for Predicting Pressure-Composition- Temperature Isotherms of Sponge Zirconium

, Available online , doi: 10.13832/j.jnpe.2025.04.0165

Abstract(31) PDF(4)

Abstract:
As one of the structural materials for nuclear fuel element fabrication, sponge zirconium reacts with environmental hydrogen, leading to material hydrogen embrittlement. This phenomenon significantly jeopardizes the operational safety and structural reliability of nuclear reactors. Pressure-Composition-Temperature (PCT) Isotherms provide critical insights for regulating the thermodynamic and kinetic behaviors of hydrogen absorption in sponge zirconium. Three predictive models for the PCT isotherms of sponge zirconium were established in this study based on experimentally measured data and data augmentation procedures: polynomial model, support vector regression (SVR) model, and artificial neural network (ANN) model. Compared with the traditional polynomial model, superior prediction accuracy was demonstrated by both the ANN and SVR models, showing MAE reductions of 73.14%% and 63.53% on the test set, respectively. Among the developed models, the ANN approach was demonstrated to exhibit optimal performance in predicting PCT curves under unknown temperature conditions, showing superior generalization capability with a test set R² value exceeding 0.98. An effective approach for precise PCT curve prediction was established through this study for metal-hydrogen systems.

Electromagnetic Characteristics of CRDM Model Based on Multiple Magnetic Circuit Coupling for Power Drive System

, Available online , doi: 10.13832/j.jnpe.2025.05.0196

Abstract(20) PDF(1)

Abstract:
To solve the problem of large deviation of electromagnetic lifting mile calculation caused by oversimplification of the equivalent magnetic circuit model of the traditional magnetic lifting control rod drive mechanism (CRDM), this study adopts an equivalent magnetic circuit model of the magnetic lifting CRDM based on multi-magnetic circuit coupling, which takes into account the coupling influence between multiple coils on the basis of the traditional magnetic circuit model and establishes an equivalent magnetic circuit model of the whole drive mechanism. Considering the edge effect of the air-gap flux, the air-gap reluctance calculation formula is improved; the quantitative relationship between the reluctance other than the air-gap reluctance and the air-gap length is analyzed, and the loop equations are established based on the Kirchhoff's law of the magnetic circuit. The research results show that the improved equivalent magnetic circuit model substantially improves the calculation accuracy of electromagnetic lifting force. Therefore, the equivalent magnetic circuit model of CRDM for magnetic lifting based on multiple magnetic circuit coupling established in this study can be used for the calculation of electromagnetic lifting force.

The research on the disposal technique of cross linked polyethylene High Integrity Container

, Available online , doi: 10.13832/j.jnpe.2025.03.0133

Abstract(22) PDF(0)

Abstract:
This paper investigates the current status of disposal technologies for cross-linked polyethylene High-Integrity Containers (HICs) abroad, analyzes the relevant requirements for the disposal of cross-linked polyethylene HICs, and formulates a specific solution based on these requirements. A near surface disposal solution has been designed, which involves storage HICs into shafts mixed with metal drums in cement cell units. This solution addresses issues such as the structural load-bearing of HICs, resin decomposition and gas (H₂)release due to radiation, and the Impact of radiation exposure between HIC drums and long-term durability. Compared to the American approach of placing HICs into shielded containers and the domestic practice of using cement solidification processes, this solution results in a smaller final disposal volume for waste drums, contributing to waste volume reduction. The proposed design ensures the long-term safety disposal of HICs.

, Available online , doi: 10.13832/j.jnpe.2025.05.0203

Abstract(13) PDF(2)

Abstract:
The lead-bismuth-cooled fast reactor (LFR) employs lead-bismuth eutectic alloy as the reactor coolant, which is characterized by a high boiling point and chemical stability. It also offers outstanding safety advantages such as favorable neutronic properties and strong natural circulation capability. As such, it has been selected by the Generation IV International Forum (GIF) as one of the priority reactor types for development. Due to the large number of fuel assemblies and the geometrically complex core structure of the LFR, directly applying traditional computational fluid dynamics (CFD) methods to construct a high-fidelity geometric model of the entire core and generate control volume meshes for simulation would result in extremely high computational resource demands. To address this challenge, the Nuclear Thermal-Hydraulics Laboratory (NuTHeL) at Xi’an Jiaotong University has developed the CorTAF series of full-core three-dimensional thermal-hydraulic analysis codes based on an open-source CFD platform. These codes are capable of subchannel-level resolution and adaptable to various reactor types. In particular, CorTAF-LBE has been specifically tailored to the structural features of LFR cores. It enables high-fidelity coupled simulations of flow and heat transfer between fuel rods and coolant under limited computational resources. This makes it possible to accurately capture key thermal-hydraulic parameters under complex core operating conditions, thereby supporting structural optimization and safety margin assessment. NuTHeL has carried out extensive code verification and model improvement efforts based on experimental data and international benchmark problems. This has enabled CorTAF-LBE to perform core safety analyses under various accident conditions and to support cross-scale coupling calculations among multiple systems. For example, by embedding a flow blockage module, the code can reveal the characteristics of non-uniform temperature distribution induced by local blockages and assess the mitigating effects of inter-wrapper flow on thermal accumulation. It also elucidates the transient evolution of thermal stratification in the upper plenum at different shutdown stages, as well as the fine-scale distribution of surface thermal stresses in key locations and their impact on structural integrity. This paper introduces the fundamental principles, code framework, and typical applications of CorTAF-LBE, summarizes the team's previous research achievements, and provides an outlook for future work.

Determination of Iron, Chromium, Niobium, and Tin in Zirconium by Laser Ablation Inductively Coupled Plasma Mass Spectrometry

, Available online , doi: 10.13832/j.jnpe.2025.04.0155

Abstract(20) PDF(0)

Abstract:
A direct analytical method for determining iron (Fe), chromium (Cr), niobium (Nb), and tin (Sn) in zirconium metal samples was established using laser ablation inductively coupled plasma mass spectrometry (LA-ICP-MS). Matrix-matched calibration standards were prepared with zirconium alloy-certified reference materials to eliminate potential matrix interference effects during measurements. The detection limits for Cr, Fe, Nb, and Sn were determined as 0.06 mg/g, 0.11 mg/g, 0.84 mg/g, and 0.88 mg/g, respectively. The developed method was validated by analyzing two real zirconium samples and comparing the results with those obtained through conventional wet digestion-ICP-MS. Statistical evaluation using Student's t-test demonstrated no significant differences between the two methods (p > 0.05), confirming the reliability of the LA-ICP-MS-based approach.

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(34) PDF(7)

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(38) 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(59) PDF(7)

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(28) 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(54) PDF(10)

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(56) PDF(6)

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(43) PDF(2)

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(64) PDF(6)

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.

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