核动力工程

Status and Prospect of Reactor Decommissioning Technology

Zhang Hangzhou, Cao Junjie, Zhang Yongling, Sun Zhijun, Hu Dongmei, Lin Li, Wu Yao, Du Defu, Wang Shuai, Chen Xisan

2024, 45(3): 1-13. doi: 10.13832/j.jnpe.2024.03.0001

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Abstract:
Reactor decommissioning is one of the nuclear facility decommissioning work that needs to be focused on. In this paper, the overall situation of global reactor decommissioning is analyzed. The concept of decommissioning technology in a broad sense is put forward, and a reactor decommissioning technology system consisting of decommissioning management and overall technology, decommissioning special technology and common support technology is constructed. The policies, regulations and standards, technical route, project management and other decommissioning management and overall technology are discussed. The special decommissioning technologies such as safe shutdown, characteristic investigation, decontamination, cutting and dismantling, waste management and final decommissioning management are analyzed. The common supporting technologies such as digitization and intellectualization, radiation protection and monitoring are discussed. After comprehensive analysis and demonstration, the development direction of specific decommissioning technology and the development trend of reactor decommissioning technology in the future are prospected.

Study on Multi-physics Coupling Calculation of Xi’an Pulsed Reactor Based on MOOSE Platform

Hu Tianliang, Jiang Duoyu, Zhang Xinyi, Wang Zhaohao, Wang Lipeng, Cao Lu, Li Da, Chen Lixin

2024, 45(3): 14-19. doi: 10.13832/j.jnpe.2024.03.0014

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Abstract:
Nuclear reactors involve a series of multi-physics coupling processes with complex interactions. With the rapid development of high-performance computing technology, the analysis based on multi-physics coupling has been paid more and more attention. Based on the multi-physics coupling platform MOOSE, the steady-state and transient multi-physics coupling simulation of Xi'an pulsed reactor is studied. The nonlinear problems in multi-physics coupling are solved by Picard iteration and Jacobian Free Newton-Krylov (JFNK) method, and the multi-physical coupling calculation of three-dimensional neutron spatio-temporal dynamics, three-dimensional solid heat conduction and one-dimensional fluid flow and heat transfer is realized. The reactor behavior under 2 MW steady-state operation and 3.45$ \$$ (1$\$$ represents an effective delayed neutron fraction) pulse operation of Xi’an pulsed reactor is calculated, and the three-dimensional power and temperature distribution of the core are obtained. The calculated results are in good agreement with the experimental results, which proves the correctness of multi-physics coupling. The multi-physics coupling method developed in this paper has the advantages of good geometric adaptability and flexible coupling, and has the potential to be applied to other micro reactors.

Analysis of Photon Heating Behavior in Fast Reactor Based on NGAMMA

Zhang Teng, Ma Xubo, Hu Kui, Jia Guanqun, Zhao Chen, Wang Lianjie

2024, 45(3): 20-27. doi: 10.13832/j.jnpe.2024.03.0020

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Abstract:
In order to improve the accuracy of photon heat release calculation for fast reactors, this paper studies the theory of high-precision photon cross-section production and the method of photon heat release calculation, and a library of photon cross-section related to the problem based on the self-developed photon cross-section processing code NGAMMA is generated. The cross-section library mainly includes neutron-photon ratio kinetic energy release (KERMA) factor, photo-atomic reaction cross-section, prompt photon production cross-section, delayed photon production cross-section and other data. The library is verified using the fast reactor benchmark problem ZPPR-9, and the computational results show that: (1) the computational accuracy of using the newly developed 94-group photon cross-section library to obtain the problem-related photon cross-section library by merging groups is significantly improved over the previous computational accuracy of directly generating the 21-group photon cross-section by utilizing NJOY. The relative deviation of the calculated results of photon heat release from the Monte Carlo results in the blanket region is reduced from −7.88% to less than 3%, and the relative deviation in the reflector region is reduced from 14.76% to 5.05%; (2) Consideration of delayed photon significantly affects the photon heat release, and compared with the consideration of prompt photon only, the consideration of delayed photon leads to an improvement of photon heat release in the inner and outer core regions by 33.11%; (3) The approximate calculation of the heat release of delayed photons by using the scaling factor method is in good agreement with the exact calculation results, and the relative deviation in each region is within ±2%

Development and Preliminary Verification of MCAT Platform for Fuel Element Multi-physics Coupling Analysis

Qi Feipeng, Liu Zhenhai, Yin Chunyu, Luo Jian, Liu Yong, Qian Libo, Zhou Yi, Wang Haoyu, Chen Ping, Li Quan

2024, 45(3): 28-36. doi: 10.13832/j.jnpe.2024.03.0028

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Abstract:
In order to further improve the prediction accuracy of fuel performance and expand the application scope of fuel analysis tools, and based on the commercial finite element analysis code COMSOL, the system safety analysis code ARSAC and the Monte Carlo burnup calculation code RMC, a multi-physical field coupling analysis platform MCAT for typical rod fuel elements is established, which realizes the bidirectional coupling of fuel module, thermal hydraulic module and neutron physics module. The concept of modular designe is adopted for the coupling platform, the intermediate data interface is used to manage the coupling parameters and define the "boundary" of each module, and the updating and feedback of the coupling parameters are realized with the input/output text files of each module, thus avoiding the source code level modification of the existing codes. Asymmetric Picard iterative algorithm is used to realize bidirectional coupling between modules, and the physical and thermal modules are regarded as black box codes. In the process of solving fuel thermodynamics, RMC and ARSAC are called in turn to perform calculations and exchange data, and iteration is repeated until convergence. In this paper, MCAT is preliminarily verified from the aspects of module, interface and comprehensive prediction results. The results show that MCAT can accurately predict the distribution of parameters such as power, temperature, structural deformation and coolant state in fuel elements, which proves the correctness of MCAT platform in module development, coupling process construction and coding implementation.

Research on the Neutron-photon Transport and Heat Calculation Method Based on MOSASAUR Code

Hu Kui, Ma Xubo, Wang Lianjie, Zhang Bin, Zhao Chen, Zhang Teng, Chen Yixue

2024, 45(3): 37-44. doi: 10.13832/j.jnpe.2024.03.0037

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Abstract:
To accurately calculate the heat released by all fissile and non-fissile materials in the fast reactor core, with a meticulous consideration of energy deposition by neutrons, photons, and electrons within the core, so as to enhance the precision of heat generation calculations, this paper, based on the deterministic two-step method, explores and implements a neutron-photon coupled transport calculation method. By solving the fission-source neutron transport equation and the fixed-source photon transport equation, the neutron and photon flux are obtained. Prompt neutron and prompt photon heat generation rates are calculated using the KERMA factor method. The delayed photon production matrix is computed using the scaling factor method. A self-coupling method in the MOSASAUR code is employed to achieve neutron-photon transport and heat generation calculations within the fast reactor core. The power distribution of the lead-bismuth fast reactor RBEC-M benchmark is compared with the results from Monte Carlo code. The relative deviations of total power are within ±4% for fuel assemblies, within ±10% for non-fuel assemblies, and within ±10% for all assemblies' photon power. Therefore, the neutron-photon transport and heat calculation method studied in this article has a high level of accuracy for the fast reactor cores.

Research and Application of Single-Point Calibration Technology Based on Dynamic-xenon Condition

Bai Jiahe, Zheng Dongjia, Wan Chenghui, Li Zaipeng, Fang He, Pan Zefei, Wu Hongchun

2024, 45(3): 45-50. doi: 10.13832/j.jnpe.2024.03.0045

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Abstract:
During the power-elevation after refueling overhaul, the nuclear power plant will carry out calibration test of ex-core power range detector (referred to as ex-core detector) at the specified power level to indicate the accurate core power level and axial power deviation. This paper proposes an ex-core detector calibration method based on dynamic-xenon condition. Based on the single core flux measurement test under dynamic xenon condition, and combined with the dynamic xenon theoretical library provided by the core physical analysis code SPARK, the core power is reconstructed, and then the calibration of the ex-core detector is completed by single-point calibration. The long waiting time for xenon equilibrium isn’t required in the whole process, which has high economic benefits. The method is verified by the measured data of Unit 5 in Tianwan Nuclear Power Station during power elevation in the second overhaul, and the errors of power level and axial power deviation are all within ±1.0% and ±0.5%, respectively. The verification results show that the single-point calibration method proposed in this paper can complete the calibration of ex-core detector within 2 h after the specified power level is reached, and the calibration coefficient has high calculation accuracy.

Experimental Study on Gas-Liquid Counter Current Flow Limitation in Inclined Pipes

Ma Youfu, Han Linfeng, Wen Huiming, Lyu Junfu, Wang Shuo

2024, 45(3): 51-59. doi: 10.13832/j.jnpe.2024.03.0051

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Abstract:
The smooth countercurrent of gas and liquid phases in the hot leg of pressurized water reactor (PWR) is very important to prevent the core melting accident, and the hot leg is composed of horizontal pipe and inclined pipe. In order to determine the constraint mechanism of gas-liquid countercurrent in the hot leg, experiments were carried out on the characteristics of countercurrent flow limitation (CCFL) in inclined and horizontal pipes with ambient air/water as two-phase working medium, and the effects of pipe layout inclination angle (0°~30°) and pipe diameter (40~100 mm) on CCFL in the pipes were studied. The main conclusions are as follows: under the CCFL conditions, the flow pattern in the horizontal pipe shows a typical stratified flow; in the inclined pipe, as the inclination angle and diameter of the pipe increase, stratified flow gradually transits to mist flow. At the same pipe diameter, the CCFL curve characterized by superficial velocities increases with the increase of pipe inclination, which indicates that the gas-liquid countercurrent flow in the hot leg is mainly controlled by the horizontal section. At the same pipe inclination angle, the CCFL curves for both inclined and horizontal pipes increase with the increase of pipe diameter. The conventional Wallis parameters do not reflect the effect of pipe inclination on the CCFL, and also fail to accurately characterize the effect of pipe diameter on the CCFL of horizontal pipes; however, they can correlate the effect of pipe diameter on the CCFL of inclined pipes satisfactorily. Finally, an experimental correlation, which can simultaneously correlate the effects of pipe inclination and diameter, was proposed for the CCFL of inclined pipes. The research results provide basic data and experimental correlation for the safety analysis of PWR nuclear power plant.

Study on Characteristics of Xi’an Pulsed Reactor in Reflood Stage

Wang Ziming, Fu Junsen, Xiao Yao, Tian Xiaoyan, Su Chunlei, Li Da, Gu Hanyang

2024, 45(3): 60-67. doi: 10.13832/j.jnpe.2024.03.0060

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Abstract:
Reflood process is an important part of LOCA, and it is crucial to identify the transient thermal characteristics of core during reflood process for the safety analysis of break accident. Based on the subchannel analysis code CTF, the transient analysis and calculation are carried out for the reflood process of Xi'an pulsed reactor in typical large and small break loss of coolant accidents. In the analysis, the core is divided into eight channels based on power. The spatiotemporal distribution characteristics of the quenching front and core temperature are obtained, and the temperature changes of the cladding and fuel rod center at different heights of fuel rods at different positions are analyzed. The results show that the temperature of the outer fuel rods is lower and completely cooled earlier than that of the hottest rod at the center. Under both large and small break conditions, the coolant can completely submerge the core, allowing it to cool completely. Under large break conditions, due to higher decay power, it takes longer for the core to be completely cooled. The risk of cladding failure is relatively low under large break conditions, while the risk of cladding failure in small break accidents is higher compared to large break conditions. Fuel pellet melting will not occur under both large and small break conditions.

Numerical Study on Heat Transfer Characteristics of a Scaled Model for Horizontal Dry Storage System of Spent Fuel

Wang Zhengquan, Yang Ting, Wen Qinglong, Xu Shijia, Chen Kang, Cheng Cheng, Tang Qionghui

2024, 45(3): 68-75. doi: 10.13832/j.jnpe.2024.03.0068

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Abstract:
Spent fuel dry storage system, with the advantages of safety, economy, and flexibility, is one of the research hotspots in the field of nuclear industry. In this study, the horizontal storage module (HSM) of the spent fuel storage system is taken as the research object, and a three-dimensional natural convection heat transfer model of the 1/2 scale model of the concrete module and the storage container is established by using the physical modeling method combining the conjugate heat transfer technology of computational fluid dynamics (CFD) with porous media, and the flow and heat transfer process in the module is numerically simulated in the Fluent solver. The results show that: the shrinkage structure design at the bottom of the concrete module leads to a sharp increase of the airflow velocity at the inlet section, and the maximum flow velocity is 1.98 m/s. The thermal shielding plate in the concrete module can effectively reduce the temperature of the concrete. The high-temperature area is distributed in the upper part of the back wall of the internal chamber, and the maximum temperature is 108.05℃. The temperature of the fuel assembly shows a symmetric distribution in the axial and horizontal radial direction, and the maximum temperature is 321.97℃. The natural convection heat exchange in concrete module accounts for 82.3% of the total heating power, and its structural design has good heat removal ability. This study will provide important references for scaling experiments and prototype design of spent fuel dry storage systems.

Numerical Study on Laminar Mixed Convective Heat Transfer of Molten Salt along Helical Cruciform Single-Rod

Jiang Dianqiang, Zhang Dalin, Chen Kailong, Tian Wenxi, Qiu Suizheng, Su Guanghui

2024, 45(3): 76-84. doi: 10.13832/j.jnpe.2024.03.0076

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Abstract:
The concept for FuSTAR, a small fluoride salt-cooled high-temperature reactor, was proposed by Xi'an Jiaotong University, and the helical cruciform fuel assembly is adopted for FuSTAR. In order to study the mixed convection heat transfer characteristics of molten salt in the helical cruciform fuel assembly, a helical cruciform single-rod channel model is established. The computational fluid dynamics (CFD) method is used to verify the numerical calculation model with experimental data. The difference between the numerical calculation value and the experimental measurement value of 94% wall temperature data point is within ±5℃, and the relative error between the numerical calculation value and the experimental measurement value of 94% average heat transfer coefficient data point is −15%~15%. The results of mixed convection heat transfer of molten salt along helical cruciform single-rod show that the influence of natural convection on the whole mixed convection heat transfer is related to inlet temperature and heat flux. The effect of natural convection on overall convective heat transfer can be more accurately evaluated by Φ/Gz (Φ is the combination variable of dimensionless number of natural convection, and Gz is Graetz number). In addition, the correlations of laminar mixed convection heat transfer of molten salt along helical cruciform single-rod at 30≤Re≤500, 6≤Pr≤26, 600≤Gr≤42000 are fitted(Re is Reynolds number, Pr is Prandtl number, and Gr is Grashof number).

Reduced Order Modeling of Once-through Steam Generator Based on Dynamic Mode Decomposition

Xu Yifan, Peng Minjun, Xia Genglei

2024, 45(3): 85-94. doi: 10.13832/j.jnpe.2024.03.0085

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Abstract:
The operation characteristics of once-through steam generator (OTSG) have an important influence on the safety of the reactor. The large-scale and refined simulation model provides high fidelity simulation results for the thermal hydraulic characteristics and safety analysis of OTSG, but it also challenges the computing resources. Dynamic Mode Decomposition with Control (DMDc) is a data-driven model order reduction (MOR) method, which can establish a low-dimensional and accurate input-output model for the system with control inputs on the basis of dynamic mode decomposition (DMD) to replace the high-fidelity model for fast calculation. Considering that the thermal parameters of OTSG, such as steam pressure, are affected by the reactor control system in actual operation, the full-order model established by RELAP5 is used to obtain the high-fidelity simulation results of the main thermal parameters of OTSG under the conditions of rapid load reduction and rapid load increase, and the reduced-order model (ROM) of OTSG is established based on DMDc. The results show that DMDc can extract the dynamic characteristics of OTSG under variable load conditions, and the maximum relative error between the calculation results of reduced-order model and the full-order model is less than 2%. In addition, the effects of DMDc and DMD methods on OTSG reduced-order modeling are compared, which proves the superiority of DMDc method.

Simulation of Blockage Accident of LBE-Cooled Fast Reactor Fuel Assembly Based on Mesh Deformation Method

Liu Zhenglong, Qiu Hanrui, Wang Mingjun, Sun Hao, Tian Wenxi, Su Guanghui

2024, 45(3): 95-103. doi: 10.13832/j.jnpe.2024.03.0095

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Abstract:
In LBE-cooled fast reactors, the fuel rod claddings or structural materials in the reactor will fall off due to corrosion of LBE. When corroded by LBE, which will block the coolant channel, cause local heat transfer deterioration and eventually lead to cladding failure. Therefore, it is necessary to analyze the flow and heat transfer characteristics in the assembly under the condition of blockage. The structure of wire-wrapped assembly is complex, and the unstructured grid division method has a large grid quantity, which requires high computing resources. In order to reduce the grid quantity, the mesh deformation method based on radial basis function is used to deform the grid of bare rod assembly, and the hexahedral grid with wire-wrapped assembly is obtained and numerical calculation is carried out. The calculated results of the hexahedral grid are in good agreement with the experimental data, and the grid quantity is much less than that of the unstructured grid, which can realize the rapid calculation of the blockage accident of wire-wrapped assemblies. Furthermore, calculations for two different types of blockages were performed in a typical wire-wrapped fuel assembly with 61 pins. The results show that for column-type blockage, the flow field recovers faster but with a higher local temperature rise; for plate-type blockage, the flow field recovers more slowly but the local temperature rise is small.

Experimental Study on Axial Evolution Characteristics of Single-phase Turbulent Mixing in Rod Bundle Sub-channels with Grids

Liu Shasha, Ma Zaiyong, Zhang Rui, Sun Wan, Zhang Luteng, Zhu Longxiang, Pan Liangming

2024, 45(3): 104-109. doi: 10.13832/j.jnpe.2024.03.0104

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Abstract:
The turbulent mixing between rod bundle sub-channels is a key part that affects the accurate calculation of thermal parameters in the reactor core, which is of great significance to improve the accuracy of reactor safety analysis. For the turbulent mixing between rod bundle sub-channels with grids, the existing research often uses the thermal diffusion coefficient to study its average effect, and lacks detailed analysis of its axial evolution characteristics. In this paper, based on the tracer analysis method, the single-phase turbulent mixing characteristics of the two sub-channels with and without the grid were studied experimentally, and the experimental results showed that the grid had a significant enhancement effect on single-phase turbulent mixing. Compared with the non-grid condition, the enhancement effect of turbulent mixing at the grid was the strongest due to the strong disturbance of the grid and cross-flow effect, the near downstream of the grid was the weakest due to the reverse cross-flow effect, the far downstream of the grid was slightly stronger than the upstream of the grid, and the enhancement effect could last for a long distance.

Study on Heat Transfer Characteristics of Irradiation Testing Fuel Assembly Discharged from Reactor and Dewatering Process

Yue Xiao, Bu Renyue, Mei Xing, Li Kejun, Huang Min, Tian Zhilong, Tian Xiaorui

2024, 45(3): 110-115. doi: 10.13832/j.jnpe.2024.03.0110

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Abstract:
In order to establish a method to analyze the heat transfer characteristics of the fuel assembly after irradiation in the test loop of the research reactor, this study takes a PWR fuel assembly irradiated in the test loop of the High Flux Engineering Test Reactor (HFETR) as the object, simulates its discharge process and dewatering process after irradiation test by computational fluid dynamics (CFD), establishes a three-dimensional flow field and temperature field calculation model, and calculates their respective heat transfer characteristics. The results show that the maximum fluid temperature during discharging process is higher than the assembly outlet temperature. In the case of low room temperature, the heat dissipation of the dehydrated assembly is better, and the dehydration process can be carried out in advance. This method can calculate the maximum allowable time of the discharging process from the reactor and the cooling time required to achieve the dewatering condition. The method in this research can be applied to heat transfer characteristics analysis and prediction for the cooling process of fuel assemblies after irradiation in the test loop of research reactor.

Large Eddy Simulation Study on Flow and Heat Transfer in Narrow Channel with a Semi-ellipsoid Blockage

Liu Junrui, Xiong Jinbiao

2024, 45(3): 116-123. doi: 10.13832/j.jnpe.2024.03.0116

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Abstract:
In order to study the influence of semi-ellipsoid blockage on the flow and heat transfer characteristics of coolant in narrow channels, large eddy simulation was carried out for a semi-ellipsoid blockage with a height of 1/2 of the narrow side width. The time-average velocity field, turbulent kinetic energy and temperature distribution downstream of the blockage are analyzed, and the characteristics of flow and heat transfer near the blockage are studied. The downstream flow field of semi-ellipsoid blockage has obvious three-dimensional characteristics, and there are characteristic regions such as recirculation region brought by the flow around the blockage and the boundary layer separation, shearing layer, main flow region and recovery region. Combined with the time-averaged temperature field and the local Nusselt number (Nu), the influence mechanism of the blockage on the heat transfer in narrow channels is studied. It is found that the local minimum Nu and local high temperature appear in the recirculation region near the side wall of the semi-ellipsoid bottom and the recovery region of the top wall due to the accumulation of hot fluid and the lack of mainstream low temperature fluid. In general, the spanwise influence range of the semi-ellipsoid blockage on the velocity field and temperature field in the narrow channel is 5 times of the semi-diameter range.

Simulation of Thermodynamic Characteristics of Supercritical Carbon Dioxide Brayton Cycle System Based on Modelica

Zhang Liqin, Huang Yanping, Zeng Xiaokang, Gong Houjun

2024, 45(3): 124-131. doi: 10.13832/j.jnpe.2024.03.0124

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Abstract:
Modelica is an open-source object-oriented language for modeling large and complex systems, and developed by the Swedish non-profit organization Modelica Association. In this paper, Modelica language is used to simulate the thermodynamic characteristics of supercritical carbon dioxide Brayton cycle system. Based on the mechanism relationship of key equipment such as compressor, turbine, regenerator and cooler, a supercritical carbon dioxide model library based on Modelica language was developed. A simulation model of single-stage regenerative cycle system was built based on drag and drop modeling and visual interface. The steady-state solution was carried out based on the solver of Modelica platform Mworks. Compared with the calculation results of SCTRAN/CO2, the reliability of Modelica model and the feasibility of Modelica in the simulation of thermodynamic characteristics of supercritical carbon dioxide Brayton cycle system were verified, and the transient characteristics of single-stage regenerative cycle were analyzed.

Experiment Study on Critical Heat Flux of 19-Pin Helical Cruciform Fuel Assembly

Fu Junsen, Xiao Yao, Chen Shuo, Zhang Wei, Cong Tenglong, Gu Hanyang

2024, 45(3): 132-138. doi: 10.13832/j.jnpe.2024.03.0132

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Abstract:
The fuel assembly is an important component of the reactor, and the critical heat flux (CHF) is one of the most critical parameters that determine the performance of the fuel assembly. With reference to the parameters of helical cruciform fuel elements in the Nuclear Engineering Thermal-Hydraulic Laboratory of Shanghai Jiao Tong University, a 19-pin helical cruciform fuel assembly was designed and the CHF experiment was carried out. A measurement method for the CHF of uniformly heated full-length helical fuel rod bundle was developed. The CHF database of helical fuel assemblies was obtained, and the experimental results were analyzed. The results show that the critical power decreases linearly with the increase of inlet temperature, pressure and dryness, and increases with the increase of mass flow rate. A CHF prediction model for helical fuel assembly is established by introducing a circumferential non-uniform factor. The statistical distribution of experimental value (M)/predicted value (P) data presents normal distribution and is uniformly distributed around 1, which proves the reliability and accuracy of the relationship. The proposed experimental technique and model development method are universal, and can be applied to the study of CHF characteristics of similar helical fuel assemblies.

Study on Weight-gain Model of FeCrAl Alloy by Steam Oxidation at Medium and High Temperature

Liu Zhen, Zhang Xiaohong, Qiao Yingjie, He Kun, Du Peinan, Zhang Ruiqian, Du Shiyu

2024, 45(3): 139-145. doi: 10.13832/j.jnpe.2024.03.0139

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Abstract:
In order to predict the steam oxidation behavior of FeCrAl alloy at different temperatures and provide the model for the evolution simulation of the performance of FeCrAl cladding under loss of coolant accident (LOCA), a two-stage differential oxidation weight-gain model was proposed based on the reaction and diffusion control mechanisms, and a parameter calibration method was also presented. Combined with the experimental data from FeCrAl steam oxidation tests at high temperature (900-1200℃) and medium temperature (400℃), the model can uniformly describe the weight-gain behavior of FeCrAl alloy in the temperature range of 400-1200℃, and the error with experimental data is controlled within 20%. At the same time, it is observed that the critical weight-gain of the reaction-diffusion mechanism is basically unchanged at 400-900℃, but increases significantly at higher temperature, because the oxidation layer grows too fast to form the dense oxidation protective layer. In addition, considering the influence of initial oxide layer from water corrosion and the change of steam pressure during LOCA, a modified scheme of the oxidation-weight gain model is given. This study is expected to provide oxidation model and parameters for the failure behavior simulation of the FeCrAl alloy cladding under LOCA accidents.

Effect of Fe+Cr and Si Contents on Corrosion Resistance of Zircaloy-4

Yue Huifang, Pang Hua, Gao Bo, Gao Shixin, Luo Qianqian, Zhao Yanli, Jiang Yourong

2024, 45(3): 146-153. doi: 10.13832/j.jnpe.2024.03.0146

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In order to optimize the corrosion resistance of domestic Zr-4 alloy, the effects of alloying element Fe+Cr and impurity element Si on the corrosion resistance of domestic Zr-4 alloy were studied under the accelerated corrosion condition of high temperature and high pressure steam at 420℃ and 10.3 MPa. The results show that within the range of Fe+Cr content specified by ASTM (0.28 wt%-0.37 wt%, with wt% representing mass percentage), the higher the Fe+Cr content, the larger the number and size of the precipitated second phases, which is beneficial to the improvement of the corrosion resistance of the material. When the content of Fe+Cr increases from 0.28 wt% to 0.37 wt%, the corrosion weight gain of Zr-4 alloy decreases by about 30% after 126 d of corrosion in steam at 420℃. When the Si content of Zr-4 alloy is as low as 10 mg/kg during quenching at 1100℃, the coarse parallel-plate structure is formed. When the Si content is increased to 100 mg/kg, the precipitated fine Zr₃Si provides nucleation sites for α phase, which leads to the appearance of basketweave structure and fine parallel-plate structure in the microstructure. The basketweave structure can promote the distribution of the second phase in the structure to be more uniform and diffuse, so the Zr-4 alloy with high Si content shows better corrosion resistance.

Performance Design and Experimental Study of Large-load Isolators

Li Xingzhao, Ma Weijie, Liu Haowen, Du Xinxin, Han Chao, Sun Yue, Peng Fanglan

2024, 45(3): 154-160. doi: 10.13832/j.jnpe.2024.03.0154

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A large-load isolator is designed for heavy equipment of nuclear power system by combining disc spring with magnetorheological liquid phase. The magnetorheological fluid damper is used as the main damping component, and the composite component of small stiffness and large stiffness disc spring is used as the main bearing structure. Through the analysis and design of the disc spring composite component, the structural parameters of the composite component are determined, and the sample isolator is formed accordingly. The spring rate ratio and damping ratio of the isolator samples were tested. The results show that the rated load of the isolator reaches 11 tons and the natural frequency is about 6.2 Hz. The dynamic stiffness and damping ratio performance of the isolator are relatively stable in the range of main vibration isolation frequencies. However, the dynamic stiffness and damping ratio are greatly affected by the amplitude. When the vibration amplitude increases from 0.05 mm to 0.2 mm, the dynamic stiffness decreases from about 100 kN/mm to about 45 kN/mm, and the damping ratio increases from about 0.07 to about 0.19. The research in this paper can provide technical reference and support for subsequent engineering applications.

Study on Impact Buckling Simulation Analysis Method and Buckling Behavior of Fuel Assembly Spacer Grid

Liu Sheng, Li Pengzhou, Yang Yiren

2024, 45(3): 161-169. doi: 10.13832/j.jnpe.2024.03.0161

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The fuel assembly spacer grid is a key part of the nuclear reactor and its dynamic buckling behaviors are important to the reactor structural safety. In this paper, the repeated impact simulation analysis method is established for the fuel assembly spacer grid, the differences between the simulation results of single impact and repeated impact and the test results are compared, and the influence of the simulation analysis method on the dynamic buckling behavior of grid impact is discussed. It is found that the simulation analysis method of repeated impact can simulate the cumulative deformation process of repeated impact in the test, and its simulation results are in better agreement with the test results. The impact force-initial velocity curves of the repeated impact simulation and the test form a yield platform near the buckling point, and the rebound coefficient and impact dynamic stiffness in the yield platform remain stable. After the yield platform, the impact force rapidly decreases, while the rebound velocity and rebound coefficient undergo drastic changes. However, the impact force and rebound velocity of the single simple impact simulation remain stable and slowly increase after the buckling point. Before buckling, the time history curve of the impact acceleration is approximately symmetrical; as the initial velocity increases, the rebound stage of the acceleration curve tails and makes its symmetry destroyed. The buckling deformation of the grids in the repeated impact simulation and the test is a first-order buckling failure dominated by transverse shear deformation at the bottom, while the buckling shape cannot be accurately predicted by the single simple impact simulation. The repeated impact simulation analysis method proposed in this paper can establish a more accurate analysis model and reveal the dynamic mechanical behavior in the buckling test of spacer grid more accurately.

Research on Small Modular Reator Concurrent Fault Diagnosis Method Based on Knowledge Matrix Reasoning

Peng Qiao, Ma Jie, Liu Minghui, Sheng Dongjie

2024, 45(3): 170-173. doi: 10.13832/j.jnpe.2024.03.0170

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Aiming at the problem of concurrent fault diagnosis of small modular reactor (SMR), a concurrent fault diagnosis method based on knowledge matrix for knowledge expression and reasoning is proposed. This method gives an efficient expert knowledge expression, which transforms the complex logical reasoning of traditional expert system into simple matrix operation, improves the efficiency of fault diagnosis, and can meet the requirements of online concurrent fault diagnosis of SMR. The effectiveness of this method is proved by the diagnosis examples of main coolant pipeline damage and control rod failure in SMR.

Finite Element Analysis of a Special Multi-bar Detection Manipulator

Cheng Baoliang, He Gaoqing, Chen Siyuan, Chen Junrong

2024, 45(3): 174-178. doi: 10.13832/j.jnpe.2024.03.0174

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Abstract:
In order to satisfy the need of ultrasonic detection of the welding seam between the steam generator (SG) outlet pipe and the main cooling pump in nuclear power plant, a set of special multi-bar detection manipulators with multi-degree of freedom is developed. A finite element analysis is conducted on the special multi-bar detection manipulator using Solidwork’s Simulation module. The results show that the special multi-bar detection manipulator made of aluminum bars only meets the strength requirements during service and could not meet the stiffness requirements, which greatly affects the positioning accuracy during transportation and scanning. Subsequently, an aluminum tube is used to optimize the structure of the special multi-bar detection manipulator. The results show that the improved equipment can meet the strength requirements during service, and the stiffness has been significantly improved. After further lightweighting, the positioning accuracy during transportation and scanning is significantly improved, which is conducive to the operation of later detection.

Analysis of Mitigation Effect of Passive Pulse Cooling System on SBO/TLFW Accident in PWR

Wu Zhenhua, Tang Qi, Li Wei, Xu Junjun, Duan Qianni, Wu Junmei

2024, 45(3): 179-185. doi: 10.13832/j.jnpe.2024.03.0179

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Abstract:
The current accident regulations of the second-generation reactor nuclear power plant are not enough to handle the superposition accident (SBO/TLFW) of the Station-Black-Out (SBO) and the Total Loss of Feed-Water (TLFW) accidents. Passive pulse cooling is a new approach to delay the accident process by fully utilizing the existing equipment and system in the secondary circuit of the second-generation reactor nuclear power plant. In order to analyze the mitigating effect of the passive pulse cooling system on SBO/TLFW, based on the best-estimate system code RELAP5, models of the primary system, the secondary circuit and the passive pulse cooling system of the CPR1000 unit were established, and accident scenarios were analyzed for the SBO/TLFW accident. The accident processes with and without the passive pulse cooling system were compared. The calculation results show that if the passive pulse cooling system is started within eight minutes following reactor scram, the core uncovering can be delayed by 12 hours, merely relying on the water stored in the deaerator, which can significantly delay the SBO/TLFW accident process in PWR.

Design and Verification of OTSG Steam Pressure Linear Active Disturbance Rejection Cascade Control for Floating Nuclear Power Plant

Zhang Tao, Shi Bo, Wu Zhijiang, Guo Wei

2024, 45(3): 186-192. doi: 10.13832/j.jnpe.2024.03.0186

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Abstract:
The once-through steam generator (OTSG) of floating nuclear power plant has a small steam-water volume and heat storage capacity, and the steam pressure fluctuates greatly when there are external disturbances and load changes, so it is difficult for the PID controller based on the local model to achieve good control in the whole operating range. Therefore, based on the system identification method, the transfer function models of OTSG under different operating conditions are obtained. The trapezoidal membership function is used for weighting, and a fuzzy model under all operating conditions suitable for OTSG steam pressure control is established. The linear active disturbance rejection controller (LADRC) is applied to the outer loop of OTSG to form a cascade control system. Combined with frequency domain analysis method, the LADRC engineering parameter adjustment law of OTSG steady-state operation and variable-condition operation is given, and the parameters are adjusted on this basis, and the performance is compared with PID cascade control system. The simulation results show that compared with PID cascade control, LADRC cascade control has faster response speed, smaller control error, stronger anti-interference and robustness.

Study on Risk-informed SSC Safety Classification of High Temperature Gas-cooled Reactor

Ni Man, Zhao Jun, Qian Hongtao, Zhang Jiajia, Gong Yu, Xiao Jun

2024, 45(3): 193-198. doi: 10.13832/j.jnpe.2024.03.0193

Abstract(119) HTML (23) PDF(69)

Abstract:
Risk-informed safety classification is based on probabilistic safety analysis (PSA), which optimizes the traditional deterministic safety classification, thus improving the management requirements of safety classification of nuclear power plants and further enhancing the safety and economy of nuclear power plants. Based on the concept of risk-informed, this paper puts forward the safety classification process of high temperature gas-cooled reactor, and takes the accident discharge system of steam generator (SG) as an example to carry out the risk-informed safety classification study. The results show that the safety SG electromagnetic discharge valve can be classified as safety-related but with low-risk, which means that there is still room for optimization of the safety classification of the system, and can provide reference for the safety classification of structures, systems and components (SSC) in nuclear power plants in the future.

Research Analysis and Implementation of Reactor Doubling Period Algorithm in Source Range Measurement

Zhu Chaoyang, Chang Jiahao, Wang Zhentao, Xing Guilai, Li Litao

2024, 45(3): 199-205. doi: 10.13832/j.jnpe.2024.03.0199

Abstract(127) HTML (20) PDF(70)

Abstract:
The reactor period doubling time is an important monitoring parameter in reactor control, as it directly reflects the rate of change of reactor power level. It is used to determine whether the reactor is in a safe and controllable state. Neutron measurement in reactor source range uses pulse counting rate to characterize neutron fluence rate and power level, and the calculation of corresponding reactor doubling period is complicated. A digital real-time algorithm based on multi-point fitting with minimum error is proposed using digitalization techniques. This algorithm is implemented in field programmable gate array (FPGA) hardware, and functional and performance tests are conducted. The test results show that the FPGA using this algorithm can accurately output the calculated reactor period doubling time within the reactor protection threshold of ±30s, as the triggering signals for the safety protection system. This reactor period doubling time algorithm is suitable for real-time monitoring of power changes in the source range neutron measurement system, ensuring safety and controllability during startup phase.

Study on Fluid-Thermal-Solid Coupling of Damper of Sodium to Air Heat Exchanger in Sodium-Cooled Fast Reactor

Gao Bo, Zhang Yujing, Sun Baoping, Liu Sheng

2024, 45(3): 206-212. doi: 10.13832/j.jnpe.2024.03.0206

Abstract(125) HTML (26) PDF(64)

Abstract:
Aiming at the problems of blade sticking and damper body strength failure of sodium to air heat exchanger in nuclear power sodium-cooled fast reactor under high temperature working environment, a double-blade damper for sodium to air heat exchanger of sodium-cooled fast reactor was designed. The characteristics of flow field, temperature distribution of damper body, deformation and stress of damper body with different opening degrees were studied using fluid-thermal-solid coupling method. The results show that when the opening of damper is below 45 degrees, the fluid will have obvious velocity gradient and pressure gradient before and after it passes through damper. The greater the local temperature difference, the greater the thermal stress value, and the maximum stress is 206.94 MPa, which appears at the edge of the frame side sealing plate, and the maximum stress value meets the material strength requirements. The deformation and stress of damper body are mainly thermal deformation and thermal stress caused by heating. The maximum deformation of damper is 3.3368 mm, and the deformation of the frame is greater than that of the blade in all directions. The newly designed damper has no phenomenon of blade sticking.

Loop System Debugging for Production of ¹²⁵I by Continuous Cycle Reactor Irradiation

Li Bo, Li Shibin, Zhang Jinsong, Luo Ning, Xue Fu, Hu Yingjiang, Zeng Junjie, Chen Yunming

2024, 45(3): 213-218. doi: 10.13832/j.jnpe.2024.03.0213

Abstract(65) HTML (18) PDF(59)

Abstract:
The production of medical isotope ¹²⁵I by continuous cycle irradiation can not only realize the cycle irradiation of the target material, but also the ¹²⁵I produced will be captured by the iodine adsorption device. The ¹²⁵I preparation process device system had been installed on site according to the site layout of Minjiang test reactor (MJTR). In this study, the process design concept, functional composition and research and development progress of the system are discussed in detail. Meanwhile, the air tightness, functionality and stability of the whole system device were debugged and verified by using the pressure maintaining of the loop and the simulation of the cycle irradiation process. The results show that the whole system device has good air tightness, and the functionality and stability of the equipment meet the technical design requirements. The research results lay a foundation for the follow-up thermal test research.

Multifactorial Coupling Enhancement Testing and Research on Improvement of Control Mode of Nuclear-grade Electric Shut-off Valve

Wu Xiaofei, Huang Maoli, Zhang Lin, Nie Changhua, Xu Changzhe, Xu Yao, Zhuo Wenbin, Li Pengzhou

2024, 45(3): 219-223. doi: 10.13832/j.jnpe.2024.03.0219

Abstract(99) HTML (28) PDF(62)

Abstract:
As the key active equipment in the nuclear reactor system, the reliability of valves directly determines the safety of nuclear reactors and personnel. In this paper, aming at the common valves (electric shut-off valve) in nuclear power plant, and based on the analysis of previous fault data and structural characteristics, it is found that the main fault of the electric shut-off valve is that the position indicator does not match the travel position, resulting in internal leakage of the valve. On this basis, the coupling enhancement test scheme was designed to simulate the vibration strengthening effect under comprehensive conditions. The failure of electric shut-off valve was stimulated in a relatively short period of time in the test. Finally, it was proposed to change the drive control mode to solve the problem of internal leakage of electric shut-off valve according to the cause of fault. The test results show that after improving the control mode, the torque control mechanism is used to control the closing, and the valve is closed normally without leakage.

Research on Dismantling Technology for Heavy Concrete Structure of Hot Cell Decommissioning

Wang Shuai, Chen Xisan, Zhang Hangzhou, Teng Lei

2024, 45(3): 224-228. doi: 10.13832/j.jnpe.2024.03.0224

Abstract(64) HTML (16) PDF(56)

Abstract:
As a new type of non-destructive static cutting technology, diamond wire saw system has great potential in cutting and dismantling the decommissioned heavy concrete structures of nuclear facilities because of its environmental protection, high efficiency and safety. In this paper, the demolition status of heavy concrete structure in a nuclear facility is briefly introduced, and then the influence of different cooling and cutting methods on cutting and dismantling the heavy concrete structure of hot cell is studied. The results show that the lowest wear coefficient is 1.4×10⁻⁵ and the maximum cutting efficiency is 0.66 m²/h for heavy concrete with a cutting ratio of 4.2 t/m³ by water-cooled diamond wire saw and tension cutting.

Steam Leakage Detection in Nuclear Power Systems based on Gaussian Filtering and Frame Difference Method

Liu Jie, Yuan Kai, Zhou Suting, Zhang Lin, Zeng Jiusheng, Nie Changhua, Huang Yanping

2024, 45(3): 229-233. doi: 10.13832/j.jnpe.2024.03.0229

Abstract(681) HTML (29) PDF(73)

Abstract:
The nuclear power system contains high-temperature and high-pressure steam inside. Once steam leakage occurs, it can lead to a loss of system function, and even cause casualties in severe cases. Therefore, it is urgent to carry out steam leakage detection in nuclear power system. At present, steam leakage is mainly detected through pressure gauges. If there is a small leakage when the system pressure is high, but the pressure reduction is not enough to attract the attention of operators, it will lead to the failure to detect steam leakage in a timely manner. Therefore, this detection method has certain drawbacks. This article introduces computer vision technology to monitor the nuclear power system and uses frame difference method to detect steam leakage in the high-temperature and high-pressure circuit system. The results show that this method can effectively detect the steam leakage in the early stage and give an early warning for the steam leakage detection video with the frame rate less than 26. This method provides a new idea for steam leakage detection of nuclear power system and can be effectively applied to engineering field, and also provides reference for leakage detection of high-temperature and high-pressure media in other fields.

Fault Detection for Reactor Coolant Pump Based on Moving Window Kernel Principal Component Analysis

Zhang Xiuchun, Xia Hong, Liu Yongkang, Zhu Shaomin, Liu Jie, Zhang Jiyu

2024, 45(3): 234-240. doi: 10.13832/j.jnpe.2024.03.0234

Abstract(116) HTML (26) PDF(62)

Abstract:
Due to the influence of component performance decline and operation condition change, nuclear power plants (NPPs) show obvious time variability during operation, which leads to the failure of fault detection model. In order to improve the performance and in-service adaptability of traditional fault detection methods in time-varying industrial processes, this paper proposes a long-term fault detection strategy for NPPs based on kernel principal component analysis (KPCA) and moving window. In this method, the KPCA fault detection model is automatically updated by moving window technology, which solves the time-varying problem of signals in the detection process. The moving window KPCA method is applied to the long-term monitoring of the reactor coolant pump in a nuclear power plant. The results show that the moving window KPCA method has good performance in fault detection rate and false alarm rate under normal and abnormal conditions.

Research on Full-Cover Fixed Ultrasonic Inspection Technology for AP1000 Control Rod

Yan Guohua, Chen Shu, Li Wei, Xiao Aiwu, Yu Yingming

2024, 45(3): 241-245. doi: 10.13832/j.jnpe.2024.03.0241

Abstract(105) HTML (28) PDF(58)

Abstract:
The control rod assembly is the component inserted into the fuel assembly to control the reactivity of the reactor core, so its status needs to be evaluated timely to ensure the safety of the plant operation. According to the structure and size of AP1000 control rod, the full-cover fixed ultrasonic inspection technology is studied in this paper, and the special inspection device is developed. Through the test of the mock-up flaw rod, it is known that the axial measurement accuracy of the flaw is ≤1 mm and the wall thickness measurement accuracy is <0.01 mm. The device has been successfully applied in the field overhaul of AP1000 unit, and several wear and swelling defects have been found. The application results show that the full-cover fixed ultrasonic inspection technology can visually display the defects of the control rod cladding, and give information such as the position, axial length, circumferential span and maximum wear depth of the defects.

Research on Flow Induced Vibration Characteristics of Plate Fuel Assembly in High-Flow Lead-bismuth Environment

Sun Yu, Liu Jian, Wang Haoyu, Qian Sheng, Qi Huanhuan

2024, 45(3): 246-251. doi: 10.13832/j.jnpe.2024.03.0246

Abstract(124) HTML (43) PDF(77)

Abstract:
Liquid lead-bismuth alloy has the characteristics of good thermal conductivity and high heat capacity, making it an ideal coolant for the new generation of advanced reactor. In this article, a full-scale computational fluid dynamics (CFD) model for plate fuel assembly in high-flow lead-bismuth environment was established, the transient fluid dynamics analysis based on large eddy simulation (LES) turbulence model was carried out and the fluid excitation force on the fuel plate was obtained. The dynamic analysis model of fuel plate was established, the structural dynamics calculation based on time domain was carried out according to the transient fluid excitation data, and the displacement response of fuel plate was obtained. The calculation results show that the fluid excitation force on the fuel plate in the middle position is much greater than that on the two sides because of the vortex shedding formed by the hoisting structure. The displacement response of the fuel plate concentrates on its own first-order frequency, and the first-order frequency of single fuel plate is much greater than the main frequency of turbulent excitation, so there is no risk of resonance of the fuel plate under fluid excitation. Considering the influence of inlet turbulence intensity, the conservatism of the flow induced vibration analysis method based on rectangular channel power density spectrum may be insufficient. This research can provide a reference for the development of new generation high-performance fuel assemblies.

Research on Reactor Information Extraction Method Based on ROERE Model

Li Cong, Li Sijia, Xu Haoran, Yan Xiong

2024, 45(3): 252-257. doi: 10.13832/j.jnpe.2024.03.0252

Abstract(74) HTML (19) PDF(58)

Abstract:
The texts of reactor design field contain a wealth of valuable information that needs to be mined, yet the unstructured form of storage poses great challenges for information extraction. Traditional information extraction methods based on artificial rules are difficult to produce efficiency in the processing of complex data, and artificial intelligence technology is needed to overcome these problems. This paper focuses on the text data of main reactor equipment, analyzes its data characteristics, and addresses the issue of single entity overlap encountered in information extraction. By incorporating the CasRel model with added relationship information and a relation-oriented module, the improved ROERE model is developed. Experimental validation across different models demonstrates that integrating relationship information and relation-oriented modules is an effective strategy, enabling more accurate and comprehensive identification and prediction of triples, thereby enhancing the accuracy and recall of information extraction for main reactor equipment.

Study on Heat Transfer Characteristics of Sub-cooled Boiling in Passive Residual Heat Removal System

Wang Yu, Lu Qing, Chen Zhihui, Hao Chengming, Zhao Jing, Yan Siwei

2024, 45(3): 258-262. doi: 10.13832/j.jnpe.2024.03.0258

Abstract(80) HTML (21) PDF(56)

Abstract:
The sub-cooled boiling in the passive residual heat removal system (PRS) of miniaturized nuclear power plant can cause unstable flow, which will affect the safety of the reactor. Through static heat transfer analysis, the temperature limit of the outer wall of the heat transfer tube to avoid net steam generation in the flowing state is obtained, and the natural circulating flow of cooling water in PRS system is improved by increasing the outlet position elevation of the cooling water outlet pipe and the inner diameter of the cooling water pipe. The research results show that after the implementation of the improvement scheme, the natural circulation capacity of cooling water and the heat removal capacity of the system are improved, which can effectively reduce the boiling intensity of cooling water, avoid excessive steam generation and make the PRS system run stably.

Analysis and Treatment of Quality Bit Anomaly of Reactor Coolant Flow Rate

Liu Danhui, Xu Tao, Zhu Jialiang, He Zhengxi, Qin Yue, Li Zhuoyue, Shi Yadong

2024, 45(3): 263-267. doi: 10.13832/j.jnpe.2024.03.0263

Abstract(130) HTML (47) PDF(70)

Abstract:
Aiming at the phenomenon of over-range triggered quality bit anomaly of all three loop flowmeters during single pump operation in the commissioning of Hainan 1&2 units, the measurement method of loop flow rate and the measured data of flow signal in cold shutdown state are analyzed and studied based on the measurement principle of reactor coolant flow signal, shutdown protection logic and quality bit setting principle. This paper points out that the physical range of flow signal currently designed cannot envelope the relative flow rate value under various conditions. According to the analysis, it is necessary to adjust the measurement range of the flow meter so that the corresponding process flow range after converting the output 4~20 mA current is adjusted from 0~120%FP (FP is full power) to 0~129%FP (the loop flow signal shows X%FP, indicating that the current flow is X% of the relative flow rate during full power operation), and the current output of the flowmeters during the normal full power operation should be calibrated to 13.615 mA.

Study on First Cycle Loading of PWR Nuclear Power Plant Based on Multi-reactor Management

Liao Hongkuan, Hu Yuying, Yu Yingrui, Wang Dan, Duan Yongqiang, Li Tianya, He Caiyun

2024, 45(3): 268-271. doi: 10.13832/j.jnpe.2024.03.0268

Abstract(107) HTML (21) PDF(60)

Abstract:
The core loading pattern is one of the most crucial design components for the effectiveness and safety of a power plant. As new fuel assemblies are used for whole core in the traditional first cycle loading pattern, many discharged assemblies from the first cycle will unavoidably be abandoned, resulting in insufficient fuel and core economy. Therefore, it is imperative to conduct research on approach for the first cycle loading design with improved fuel economy and more rational fuel assembly utilization. In this paper, a design method for the first cycle loading of PWR nuclear power plant based on multi-reactor management is proposed. By sharing the fuel assemblies of multiple units, the first cycle of multiple units is designed as multi-reactor, so as to improve the enrichment of new fuel assemblies used in the first cycle and meet the cycle length requirements. Reducing the number of new fuel assemblies on the basis of ensuring the cycle length of each unit can significantly improve the fuel utilization rate and the burnup of the first cycle unloading assembly, realizing a significant increase in the unit's economy. A representative HPR1000 reactor is used for verification. The results demonstrate that 41 new fuel assemblies can be reduced in two units with the same cycle length under the multi-reactor mode, and all the core parameters are within the design limits. This study can provide reference for the first cycle loading design of subsequent units to enhance the economy and competitiveness of nuclear energy.

Study on Neutron Sensitivity Calculation Model of Ex-core Neutron Detector

Liu Yaolong, Chen Zhi, Huang Youjun, Lin Chao, Gao Zhiyu, Luo Tingfang

2024, 45(3): 272-278. doi: 10.13832/j.jnpe.2024.03.0272

Abstract(147) HTML (39) PDF(69)

Abstract:
Based on the Monte Carlo method, this thesis proposes a neutron sensitivity calculation model for the design and study of the performance of ex-core neutron detectors. Firstly, based on the physical principle of the detector, the influencing factors of neutron sensitivity are obtained. On this basis, the neutron field characteristics at the monitor holes are analyzed, and the neutron sensitivity calculation model is proposed. The calculation model is discussed to obtain the uncertainty analysis of the model. Finally, the model is verified, and the error between the calculated neutron sensitivity and the measured data is within the acceptable range, which verifies the feasibility of the model.

2024 Vol. 45, No. 3