核动力工程

Solving Multi-Dimensional Neutron Diffusion Equation Using Deep Machine Learning Technology Based on PINN Model

Liu Dong, Luo Qi, Tang Lei, An Ping, Yang Fan

2022, 43(2): 1-8. doi: 10.13832/j.jnpe.2022.02.0001

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Abstract:
This paper elaborates the physics-informed neural network model (PINN), constructs a deep neural network as a trial function, substitutes it into the neutron diffusion equation to form a residual, and takes it as the weighted loss function of machine learning, and then approaches the numerical solution of the neutron diffusion equation by deep machine learning technique; According to the characteristics of diffusion equation, this paper puts forward innovative key technologies such as accelerated convergence method of eigenvalue equation, efficient parallel search technology of effective multiplication coefficient (k_eff), learning sample grid point uneven distribution strategy, and analyzes the sensitivity of key parameters such as neural network depth, neuron number, boundary condition loss function weight and so on. The verification calculation results show that the method has good accuracy, and the proposed key technology has remarkable results, and explores a new technical approach for the numerical solution of the neutron diffusion equation.

A Two-dimensional Coupled Neutron Transport Method for MOC and S_N via Boundary Fluence Rate Coupling

Zhang Sifan, Yuan Yuan, Liu Zhouyu, Zhou Xinyu, Cao Liangzhi

2022, 43(2): 9-16. doi: 10.13832/j.jnpe.2022.02.0009

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Abstract:
When the method of characteristics (MOC) is used to calculate the out-of-pile detector or some special heavy water moderated light water cooled experimental reactor, the dense characteristics will lead to a large waste of computing resources because the external structural material or moderator area of its active region is too large. To solve this problem, a new coupled transport method based on MOC and discrete ordinate (S_N) nodal method is proposed and implemented in the numerical reactor physics calculation program NECP-X. In this method, the calculation region is divided into MOC domain (including complex structure regions such as active region) and S_N domain (including simple structure regions such as moderator and reflector), and then the grids of the two regions are subject to hybrid sweeping and coupled through the angular fluence rate of the region interface; At the same time, some feasible methods are proposed to reduce the error caused by the fluence rate of the coupling boundary angle. Finally, the calculation effect of the coupling method is verified by the test of two-dimensional C5G7 benchmark task and whole core problem. The numerical results show that the coupling method achieves superior efficiency and accuracy.

Research on Transient Multi-Group Pin-Power Reconstruction Based on Source Expansion Method

Bai Jiahe, Wan Chenghui, Li Yunzhao, Wu Hongchun

2022, 43(2): 17-21. doi: 10.13832/j.jnpe.2022.02.0017

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Abstract:
This paper studies the calculation method of transient multi-group pin-power reconstruction of PWR. Starting from the transient fixed-source equation, the transient fixed-source term is expressed in the form of biquadratic Legendre polynomial, and the pin-power reconstruction is realized through the iterative solution of the whole reactor. The method is applied to the core physics code SPARK, and the TWIGL transient benchmark task is used to verify the accuracy of transient multi-group pin-power reconstruction. Numerical results show that the PWR transient multi-group pin-power reconstruction method proposed in this paper has high computational accuracy.

Experimental Research of Bundle and Spacer Grid Arrangement on Fuel Assembly Mixing Characteristics

Cheng Cheng, Ye Tingpu, Lu Donghua, Su Qianhua, Long Biao, Chen Zhenhui

2022, 43(2): 22-27. doi: 10.13832/j.jnpe.2022.02.0022

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Abstract:
Taking the 5×5 bundle fuel assembly assembled with the spacer grid used in CPR1000 nuclear power unit as the object, experiments on the mixing characteristics of several groups of full-length bundle fuel assembly were carried out. The effects of geometric parameters such as cold-hot rod arrangement and spacer grid arrangement on the mixing characteristics of fuel assembly are analyzed emphatically. The experimental results show that the thermal diffusion coefficient of the fuel assembly is closer to the true value when the cold-hot rods are symmetrically arranged in the center; The inter-span mixing spacer grid has a small enhancement effect on the overall thermal diffusion coefficient of fuel assembly, but its contribution to the pressure drop of bundle is very low.

Research on ¹⁰B Abundance Calculation Method of PWR Based on Bamboo-C

Liu Yu, Wan Chenghui, Huang Xing, Wu Hongchun, Zhang Shengbin, Cai Guangming

2022, 43(2): 28-31. doi: 10.13832/j.jnpe.2022.02.0028

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Abstract:
During the power operation of PWR, due to burnup effect and boronation effect, the abundance of ¹⁰B in the primary circuit boric acid solution will change continuously with time, and the nuclear power plant can not provide the real-time measured value of ¹⁰B abundance. As a result, the calculated value of the critical boron concentration in the nuclear design program is lower than the measured value in the middle of the cycle. Therefore, in order to reduce the deviation between the calculated value and the measured value of the critical boron concentration in the primary circuit of the core, this paper gives a calculation method of ¹⁰B abundance based on the Bamboo-C software independently developed by the Nuclear Engineering Computational Physics (NECP) Laboratory of Xi'an Jiaotong University. This method can consider the influence of burnup effect and boronation effect on the critical boron concentration at the same time. Using the ¹⁰B abundance calculation method mentioned in this paper, the power operation history of the M310 unit of Fuqing NPP was tracked and simulated, and the calculated values of the primary circuit critical boron concentration and B-10 abundance were compared and verified with the measured values. The verification results show that the ¹⁰B abundance calculation method mentioned in this paper can improve the numerical calculation results of core critical boron concentration, and has the conditions for industrial application.

MSR Supercritical Carbon Dioxide Brayton Cycle System and Thermodynamic Analysis

Lu Heng, Zhao Heng, Dai Ye, Chen Xingwei, Jia Guobin, Zou Yang

2022, 43(2): 32-39. doi: 10.13832/j.jnpe.2022.02.0032

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Abstract:
Molten salt reactor (MSR) can realize on-line packing and post-processing, and the outlet temperature is higher, so it shall be equipped with an innovative cycle mode that matches its outlet temperature, and can achieve higher cycle efficiency. In this paper, a supercritical carbon dioxide (SCO₂) Brayton cycle system is designed based on the small modular molten salt reactor (smTMSR-400) designed by Shanghai Institute of Applied Physics, Chinese Academy of Sciences. The effects of split ratio, compressor/turbine efficiency, outlet temperature of main compressor and heat exchange temperature difference/resistance of low temperature heat exchanger on SCO₂ Brayton cycle system are analyzed by using the control variable method. The analysis results show that: ①there is an optimal split ratio to make the temperature difference between the two sides of the low temperature heat exchanger equal; ②compared with the compressor efficiency, the equal-amplitude turbine efficiency improvement can make the system cycle efficiency and exergy efficiency higher; ③ the increase in the outlet pressure of the main compressor has a positive impact on the system, but the cycle efficiency/exergy efficiency and its slope gradually decrease; ④the heat exchange temperature difference and flow resistance of the heat exchanger bring quantifiable burden to the system cycle: for every 10 K increase in the heat exchange temperature difference, the cycle efficiency decreases by 1.85% and exergy efficiency decreases by 2.70%; When the flow resistance increases by 1 MPa, the cycle efficiency decreases by 6.58% and exergy efficiency decreases by 10.22%. At last ,this paper designs 5 physical reference schemes based on the analysis results and system exergy changes.

Research on the Key Influencing Factors of the Backflow Phenomenon on the Primary Side of the Inverted U-tube Steam Generator under Natural Circulation

Wang Tianshi, Wang Yuxuan, Zhao Pengcheng, Wang Xijie, Ling Yufan, Wang Yuqing, Zhu Enping

2022, 43(2): 40-46. doi: 10.13832/j.jnpe.2022.02.0040

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Abstract:
Backflow exists in the inverted U-tube steam generator (UTSG) under the condition of natural circulation, which affects the heat carrying capacity and natural circulation capacity of the primary circuit coolant system. Referring to the design parameters of UTSG in PWR thermal experimental device (PWR PACTEL) in Finland, this paper uses computational fluid dynamics (CFD) software Fluent to simulate the backflow in UTSG under the condition of uniform flow decline, and studies the influence of primary side operation parameters, UTSG design parameters and secondary side operation parameters on the backflow. The results show that increasing the primary side temperature, primary side operating pressure and thermal conductivity of the inverted U-tube will increase the critical mass flow rate of the UTSG, making the UTSG more prone to backflow; Increasing the water supply and temperature on the secondary side of UTSG and the roughness of the inner wall of inverted U-tube will decrease the critical mass flow rate of UTSG and restrain the occurrence of backflow. Changing the wall thickness of inverted U-tube has little effect on the backflow; Compared with changing the temperature of the secondary circuit, changing the temperature of the primary circuit has a more significant impact on the backflow. The results of this study can provide some reference for the parameter optimization of UTSG.

Evaluation of Single-phase and Two-phase Mixing Models for Rod Bundle Channel

Ye Tingpu, Lyu Lulu, Zhang Ge, He Hui, Cheng Cheng

2022, 43(2): 47-52. doi: 10.13832/j.jnpe.2022.02.0047

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Abstract:
In this study, the subchannel program is used to evaluate the single-phase and two-phase mixing model of the rod bundle channel based on the existing experimental data. The single-phase mixing mainly considers the cross flow and turbulent mixing. The cross flow is determined by the conservation equation and plays a leading role in the flow distribution. The turbulent mixing depends on the mixing coefficient. It is found that the prediction results of Sadatomi model are in good agreement with the experimental results in the study of turbulent mixing. Two-phase mixing is jointly caused by cross flow, turbulent mixing and void drift. Through the comparative analysis of the prediction results of the existing models and the experimental data, it is recommended to use the Hotta model for the void drift in the two-phase mixing, the Sadatomi model for the turbulent mixing coefficient, and the Beus model for the two-phase multiplier. This is a combined model with conservative prediction results, which is conducive to the conservative evaluation of reactor safety.

Influence of Inlet and Outlet Resistance on the Characteristics of Backflow in Inverted U-tube Bundle

Ma Songyang, Li Mingrui, Chen Wenzhen, Hao Jianli, Xiao Hongguang, Ye Lei

2022, 43(2): 53-58. doi: 10.13832/j.jnpe.2022.02.0053

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Abstract:
The inverted U-tube bundle backflow problem of the steam generator (SG), which increases the flow resistance of SG under natural circulation conditions and reduces the natural circulation capacity of the system, has a negative impact on reactor safety. In view of the above problems, taking the marine SG as the research object, the flow distribution and backflow characteristics of the tube plate before and after the improvement scheme are studied by using the methods of theoretical analysis and computational fluid dynamics (CFD) numerical simulation and by rounding the tube plate opening in the parallel inverted U-tube bundle. The results show that by rounding the tube plate opening of the short tube, the flow resistance of SG can be effectively reduced and the flow distribution of the short tube can be increased under the natural cycle condition, so as to reduce the critical flow of backflow and delay the occurrence of backflow. The research conclusion provides a feasible solution for the backflow problem in the SG inverted U-tube bundle, and can provide support for the research on solving the backflow problem.

Stability Analysis of Double-Loop Natural Circulation under Asymmetric Conditions

Zhu Enping, Wang Ting, Liu Zijing, Zhao Pengcheng, Wang Tianshi

2022, 43(2): 59-64. doi: 10.13832/j.jnpe.2022.02.0059

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In order to study the stability change of double loop natural circulation under asymmetric conditions, taking the single loop natural circulation heat carrying system as the starting point, the displacement term in the dimensionless governing equation set of single loop natural circulation is expanded by Fourier expansion, and the Jacobi matrix characterizing the single loop natural circulation heat carrying system is obtained, and verification is conducted. On this basis, the Jacobi matrix analysis model of double loop is constructed. Based on the constructed model, the stability analysis of double loop natural circulation under different load difference and resistance difference is carried out, and the influence of loop geometric characteristics on the stability boundary is carried out. The results show that for a double loop system, there are two critical Reynolds numbers. When the load difference between the left and right loops is greater than these two critical Reynolds numbers, the system will become unstable. Increasing the aspect ratio, reducing the pipe diameter, increasing the length of heating section and reducing the length of cooling section can increase the range of loop stability area, and the stability boundary of reactor loop is more sensitive to the aspect ratio and the length of heating section; Besides, within the allowable range of natural circulation loop formation, increasing the loop pressure drop can improve the system stability.

Study on Flow Pattern Evolution in Outlet Heat Removal Pipe of Open Natural Circulation System

Sun Yuxiang, Xu Jianjun, Zhou Huihui, Deng Zhiyong, Yuan Zhaofei, Cui Yinghuan, Huang Yanping

2022, 43(2): 65-69. doi: 10.13832/j.jnpe.2022.02.0065

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As the ultimate heat sink discharge loop of the new passive residual heat removal system, the safe and stable operation of the open natural circulation system is very important for removing the residual heat of the reactor core under accident conditions. In this paper, the flow pattern evolution in the outlet heat removal pipe of the open natural circulation system is observed by visual experiment. It was found that with the increase of heating power, the open natural circulation is gradually established. There are five typical flow patterns in the heat removal pipe at the outlet of the system: single-phase flow, intermittent bubble flow, dispersed bubble flow, slug flow and intermittent jet flow. The relationship between the flow pattern in the outlet heat removal pipe and the stable operation of the system is analyzed, and the source of the violent oscillation of the open natural circulation system is found, which provides reference for improving the flow stability of the open natural circulation system.

CFD Analysis on Characteristics of High Temperature Heat Pipe

Yu Qingyuan, Zhao Pengcheng, Ma Yugao

2022, 43(2): 70-76. doi: 10.13832/j.jnpe.2022.02.0070

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Abstract:
The analysis and prediction of the operating characteristics of high temperature heat pipe are of great significance to the design and application of heat pipe. In order to analyze the heat transfer characteristics of two-phase flow in high temperature heat pipe, first, the computational fluid dynamics (CFD) analysis model of sodium heat pipe is established, and the calculated values of the model are compared with the steady-state experimental data of sodium heat pipe. The absolute error between the simulation results and the temperature of the experimental measuring point is less than 40℃, and the relative error is within 5%; Second, the flow field characteristics of heat pipe under different heat transfer power and inclination angle are analyzed and studied by using the model and method in this paper. The results show that under the condition of uniform heating, the velocity in the steam chamber changes nearly linearly in the evaporation section, while in the condensation section, the decrease of gas velocity leads to the rise of pressure. At the same time, the flow pressure drop and velocity of steam decrease with the increase of heating power; Under the horizontal and inclined operating conditions of heat pipe, the liquid phase pressure drop is dominant in the two-phase flow pressure drop in heat pipe; in the gas-liquid shear effect, the gas flow velocity is the dominant effect. The model presented in this paper can provide heat pipe design and analysis methods for high temperature heat pipe applications such as heat pipe reactor.

Experimental Research on Pressure Drop Characteristics of Flow Damper in Advanced Accumulator

He Yanqiu, Yuan Zhaofei, Zhang Yan, Tan Shushi, Zan Yuanfeng, Qiao Min, Hu Qiang

2022, 43(2): 77-82. doi: 10.13832/j.jnpe.2022.02.0077

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Abstract:
Two different types of pressure drop characteristics of flow damper were obtained through the test of pressure drop characteristics of flow damper in advanced accumulator. The influence law of different geometric parameters on the pressure drop coefficient is studied, and the relationship of pressure drop coefficient is fitted. The results show that in the range of experimental parameters, the vortex pressure drop coefficient increases gradually with the increase of Reynolds number, and the mixing pressure drop coefficient decreases rapidly and then increases slowly with the increase of the flow ratio of large tube to small tube. The small tube width and flow damper diameter have a certain influence on the vortex pressure drop coefficient. The inclination of large and small tubes, the flow damper diameter and the large tube width have an effect on the mixing pressure drop coefficient, while the small tube width has no obvious effect on the mixing pressure drop coefficient. The deviation between the predicted value of the vortex pressure drop coefficient relationship and the experimental value is within ±10%, and the deviation between the predicted value of the mixing pressure drop coefficient relationship and the experimental value is large.

Study on Coupling Characteristics of Multiple Thermal Parameters during the Fuel Assembly Steady-State Irradiation in the Test Loop

Si Junping, Sun Sheng, Tong Mingyan, Lu Mengkang, Lei Jin, Jin Shuai, Wang Wanjin

2022, 43(2): 83-88. doi: 10.13832/j.jnpe.2022.02.0083

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Abstract:
The steady-state irradiation in the test loop is a key process to study the anti-irradiation performance of the fuel assembly. In view of the importance of the thermal parameters of the irradiation test loop to the test run, aiming at the steady-state irradiation in the test loop, combined with the requirements of fuel assemblies for the thermal parameters of the steady-state irradiation test, this paper analyzes the coupling characteristics of the key thermal parameters of the loop system under the changes of primary water flow and inlet temperature, the existence of heat exchange in the device and the operation mode of different heat exchangers. The research shows that with the decrease of the primary water inlet temperature and flow rate of the heat exchanger, the maximum heat exchange power of the main heat exchanger decreases significantly. Under the condition of serious deviation from the design condition, there is a risk that the heat exchange capacity of the main heat exchanger can not meet the irradiation operation of the fuel assembly. At the same time, the internal heat exchange of the irradiation device is extremely disadvantageous to the heat exchange of the main heat exchanger. Under the condition that the operating temperature of the loop is high and the internal heat exchange is strong, the design margin of the secondary water flow of the main heat exchanger shall not be lower than the heat exchange ratio in the device, and the margin shall be larger. The advantages of parallel operation of two main heat exchangers are mainly reflected in the large primary water flow, and there is a small primary water flow, which makes the heat exchange capacity of single heat exchanger running independently consistent with that of two heat exchangers running in parallel. When the flow ratio is lower than this, the heat exchange capacity of two heat exchangers running in parallel is weaker than that of single heat exchanger.

Research on Analysis for Performance and Optimization of Prismatic Dispersed Microencapsulated Fuel in Gas-Cooled Reactor

Zhao Bo, Li Quan, Li Yuanming, Huang Yongzhong, Ma Qiang, Su Min, Liu Zhenhai, Qi Feipeng, Ma Chao, Chen Hao

2022, 43(2): 89-95. doi: 10.13832/j.jnpe.2022.02.0089

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Prismatic dispersed microencapsulated fuel is a particle-reinforced composite fuel which is formed by TRISO fuel particles dispersed in metal or ceramic matrix. It has good structural stability, fission product inclusion capacity and irradiation stability, and it is one of the promising fuels in high temperature gas-cooled reactor. A prismatic dispersed microencapsulated fuel with TRISO particles dispersed in SiC matric was proposed in this paper. Based on the finite element analysis software COMSOL, the 3D thermal-fluid-solid coupling analysis model of the fuel element is established, and the performance analysis and optimization design of the fuel element are preliminarily realized. The results show that the maximum temperature of the prismatic dispersed microencapsulated fuel element is located on the outside of the fuel element, the peak stress is on the wall of the coolant channel, and the thermal stress of the fuel element is the lowest when the edge-distance ratio is 0.76 to 0.84 and the hole-distance ratio is 0.68 to 0.75. The performance analysis method and research conclusions of the prismatic dispersed microencapsulated fuel established in this paper can provide guidance and reference for the subsequent design of this type of gas-cooled reactor fuel element.

Research on Handling Method of Accelerometer Tripping of Fresh Fuel Assembly Transport Vessel

Jin Yuan, LI Weicai, Chen Jianxin, Zhou Yuemin

2022, 43(2): 96-101. doi: 10.13832/j.jnpe.2022.02.0096

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Accelerometers mounted on fresh fuel assembly transport vessel are used to monitor the abnormal shock of fuel assemblies during transport. The tripping of the accelerometer indicates that there may be a load that may damage the fuel assembly during transport. In recent years, several tripping events of fresh fuel assembly transport vessel accelerometers happened in domestic nuclear power plants. After the tripping of accelerometer of the transport vessel, it is necessary to evaluate the mechanical integrity and availability of the incident fuel assembly and judge whether it can be used in the reactor. Based on the in-depth analysis of the principle of the accelerometer and the tripping process of the accelerometer, this paper puts forward a general handling method to deal with the tripping of the accelerometer in the transport vessel of the new fuel assembly. The general handling method is used to deal with the tripping event of the new fuel assembly transport vessel accelerometer in a nuclear power plant in recent years, and the result has been adopted by the owner. The general handling method of accelerometer tripping event proposed in this paper can provide an important reference and basis for the subsequent handling of accelerometer tripping event in domestic nuclear power plants.

Study on PUREX 1B Process Simulation Based on MPMS Calculation Model

Tang Jia, Yang Yu, Lin Mingzhang, Lu Yunyun, Xiong Wei, Guo Zifang, Wu Zhihao

2022, 43(2): 102-107. doi: 10.13832/j.jnpe.2022.02.0102

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In order to simulate the U/Pu separation (1B) process of uranium plutonium redox extraction (PUREX), a mathematical model of 1B process with NH₃OH⁺-N₂H₄(HAN-HYD) as reducing extractant was established based on the framework of MPMS calculation model, using multi-stage mixer-settler as extraction equipment, and based on the chemical reaction system and empirical extraction system. The validity of the model is verified by comparing with the literature data. The mathematical model is applied to explore the effect of a low flow rate and high acid cleaning solution in 1B process under specific parameters. The results show that the introduction of the low flow rate and high acid cleaning solution will reduce the recovery of U/Pu. In order to further evaluate the effect of low flow rate and high acid cleaning solution on the separation efficiency and recovery efficiency of 1B process under different conditions, different U/Pu separation effects were obtained by changing the process parameters of low flow rate and high acid cleaning solution in the model. The calculation results show that the optimal U/Pu separation efficiency can be obtained by 1B process without introducing low flow rate and high acid cleaning solution. The mathematical model will provide useful help for the process evaluation and prediction of 1B process based on multi-stage mixer-settler.

Analysis on Wearing of Bearing for Main Pump in Nuclear Power Plant

Zhang Haiying, Huang Zhong, Jiao Shaoyang, Dai Hongling, Xu Xiaogang, Sheng Feng

2022, 43(2): 108-111. doi: 10.13832/j.jnpe.2022.02.0108

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The heat release rate of the main pump in nuclear power plant continued to rise, it was found that the bearing surface of main pump was abnormal worn through inspection. In order to analysis the causes and the level of the wearing of the bearing surface, the physical properties of the material of the bearing were tested, and the wear degree was analyzed qualitatively by using Archard formula. The results show that the larger the thermal expansion coefficient of the bearing is, the thinner the thickness of the liquid film is, which is easy to cause the damage of the liquid film and is not conducive to the cooling of the bearing. The uneven distribution of the material of bearing is easy to the wear particles produced, and the gas in the liquid film is increased, both of which make that the cavitation phenomenon between the bears more serious and the liquid film is local damaged and the dry friction is happened.

Three-dimensional Site Response Analysis Based on Equivalent-Linear Behavior of Foundation Soil

Qu Yunguang, Xu Zhengyu

2022, 43(2): 112-116. doi: 10.13832/j.jnpe.2022.02.0112

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With the increasing complexity of nuclear power plant's siting conditions, soil-structure interaction (SSI) has become an important issue to be considered in the seismic analysis. At present, the classical free-field site response analysis uses the analysis of one-dimensional layered foundation soil, such as SHAKE91, EERA and SASSI, so it is difficult to consider the heterogeneous layer factors of the soil layer. Therefore, with the increasing regulatory requirements for nuclear power safety, the refined analysis of earthquake resistance has become a trend. In this paper, the UMAT material subprogram written by the finite element program ABAQUS is used to realize the equivalent linearity of the foundation soil material and carry out the three-dimensional free-field site response analysis of the homogeneous layered soil. Comparing the calculation and analysis results with SHAKE91 calculation results, it shows that it is in good agreement under the condition of homogeneous layered soil. Therefore, this study provides a good engineering applicability for the SSI analysis of complex heterogeneous foundation conditions.

Design and Sealing Performance Verification of Graphite Seal Assembly of CRDM

Luo Qingsong, Xu Huaijing, Tang Baoqiang, Han Jiaxin, He Canyun, Guo Yong, Liu Xin, Ma Zhigang

2022, 43(2): 117-121. doi: 10.13832/j.jnpe.2022.02.0117

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The pressure housing of CRDM needs to be disassembled many times during the maintenance of control rod drive mechanism (CRDM) and core refueling. In order to solve the leakage of Ω sealing weld of the existing pressure housing and the problem that it can not be disassembled many times, the scheme of pressing graphite ring with nut is adopted in this paper, the compression resilience of graphite ring is used to prevent coolant leakage, and a kind of graphite ring sealing assembly is designed to realize rapid disassembly. The sealing performance of graphite ring sealing assembly is verified by carrying out sealing performance tests such as compression rebound test of sealing ring, stress relaxation test and leakage rate of sealing assembly. The results show that the graphite ring sealing ring assembly designed in this paper meets the design requirements and can achieve sealing performance under high temperature and high pressure environment.

Study on Numerical Simulation of Drop Impact of Spent Fuel Transfer Equipment

Yuan Liang, Yang Jie

2022, 43(2): 122-125. doi: 10.13832/j.jnpe.2022.02.0122

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Analysis of drop of spent fuel transfer equipment in nuclear power plant is the most stringent condition in overall structural safety analysis. In order to solve the problem of dynamic impact analysis and evaluation of equipment drop, the finite element analysis and simulation software LS-DYNA is used to numerically simulate the spent fuel transfer equipment, and model the drop of typical spent fuel transfer equipment. Combined with the actual plant site conditions, the drop contact ground adopts Holmquist-Johnson-Cook (HJC) model. Through simulation calculation, the acceleration curve of equipment and the deformation variables of key position are obtained. The results show that the deformation of the storage sleeve is greatly affected by the drop angle when combined with the actual ground conditions in the plant. The vertical drop of the equipment shall be avoided in the overshoot of storage and transportation. The analysis and evaluation method in this paper can provide technical support and theoretical basis for the autonomous design of spent fuel transfer equipment.

Fragility Analysis of Nuclear Power Plant Containment under Near-site Vibration

Rong Hua, Jin Song, Gong Jinxin

2022, 43(2): 126-132. doi: 10.13832/j.jnpe.2022.02.0126

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Containment structure is one of the most important structures in nuclear power plant, and its seismic fragility is the focus of probabilistic seismic safety assessment of nuclear power plant structure. Combined with nonlinear finite element analysis technology and incremental dynamic analysis method, the fragility of nuclear power plant containment under near-site vibration is analyzed. In addition, in order to overcome the limitations of the traditional global damage index of containment structure based on top displacement, an energy based global damage index is proposed and its effectiveness is verified. Finally, a fragility curve construction method considering the statistical uncertainty of seismic fragility parameters is proposed. The research results show that the global damage index of the containment structure proposed in this paper can reflect the overall deformation characteristics of the containment structure well, and its variability is smaller than that of the global damage index based on top displacement. The overall impact of statistical uncertainty on the corresponding fragility curve of containment structure under different damage performance levels can be ignored, but it has a certain impact on the lower tail of the fragility curve.

Research on Test and Theoretical Analysis Methods on Stability of LBB Circumferential Through-Wall Crack in Austenitic Stainless Steel Pipe under Dynamic Load

He Feng, Yao Di, Wang Xinjun, Li Yilei, Bai Xiaoming, Xiong Furui

2022, 43(2): 133-137. doi: 10.13832/j.jnpe.2022.02.0133

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Judging whether the circumferential through-wall crack of pipe is stable is one of the criteria to judge whether the pipe meets the Leak-Before-Break (LBB) design criteria. In order to ensure the safety and reliability of LBB technology, the stability of circumferential through-wall crack of pipe under dynamic load is studied and analyzed by experiment. The horizontal impact machine was used to carry out the impact test on the pipe with circumferential through-wall crack without operating pressure at high temperature with the loading speed of 1.22 m/s, 2 m/s, 3 m/s, 4 m/s in order to obtain the test limit load values at each strain rate, and then compared with the engineering theoretical analysis and calculation results. The results show that the failure mode of circumferential through-wall cracks in austenitic stainless steel pipe under dynamic load is plastic instability. Through the verification of the test, when LBB analysis is carried out for austenitic stainless steel pipe under dynamic load in the project, the theoretical analysis method of limit load in the standard review plan 3.6.3 leak-before-break evaluation procedures (SRP 3.6.3) of the NRC has high engineering safety. If the mechanical properties of materials under quasi-static state are adopted at the same time, the engineering safety is higher.

A Review of Research on Aerosol Hygroscopic Growth in Severe Nuclear Reactor Accidents

Wang Jinghong, Peng Wei, Yu Suyuan

2022, 43(2): 138-151. doi: 10.13832/j.jnpe.2022.02.0138

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The hygroscopic growth of soluble aerosols (hereinafter referred to as aerosols) is one of the key factors affecting the dynamic behavior of radioactive products in serious nuclear reactor accidents. In this paper, the theoretical model of hygroscopic growth, experimental measurement scheme and research progress of hygroscopic growth in nuclear safety field are summarized. Based on Köhler theory, aerosol hygroscopic growth theoretical model in the field of nuclear safety describes the relationship between aerosol thermophysical parameters and environmental parameters in the process of hygroscopic growth. On this basis, a variety of improved models are more suitable for the analysis of practical problems, and have been widely used in nuclear safety aerosol calculation programs such as NAUA-HYGROS and MELCOR. Experimental measurement is also an important means to study the hygroscopic growth characteristics of aerosols. Compared with the gravimetric method, which can measure the overall hygroscopic ability of aerosols, but the results are relatively rough, the power balance method and the HTDMA method have higher accuracy, and can measure the hygroscopic ability of single particles and multimodal particle groups in real time, and have potential application prospects in the experimental study of the aerosol hygroscopic growth characteristics of nuclear accidents. At the end of this paper, the existing applied researches on hygroscopic growth in the field of serious nuclear accidents are summarized, including the theoretical model and numerical application of typical aerosol hygroscopic growth in nuclear accidents, and the experimental study of hygroscopic growth. Numerical studies show that incorporating aerosol hygroscopic growth characteristics into the aerosol calculation program of nuclear accident can achieve more accurate prediction and analysis of aerosol behavior after nuclear accident. The hygroscopic growth characteristics of CsOH, CSI and other typical nuclear accident aerosols were obtained by relevant experiments.

The Application of Modelica Simulation Technology in Micro Gas-Cooled Reactor

Liang Yangyang, Zhang Huimin, Wang Li, Li Yunlong, Yuan Yidan, Wang Jun, Du Shuhong

2022, 43(2): 152-159. doi: 10.13832/j.jnpe.2022.02.0152

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Abstract:
Compared with traditional large-scale nuclear power plants, the functions of micro reactor systems are closely coupled and restrict each other. The traditional professional decoupling design mode is difficult to deal with it, and a full range of system simulation is needed. A system simulation model of an gas-cooled micro reactor is established by using Modelica language, and the accident analysis and calculation is carried out by taking the expected transient (ATWS) accident without emergency shutdown as an example, and compared with the results of professional core safety analysis. The results show that the change trend of reactor power is consistent, and the shutdown can be realized only by the negative reactivity caused by the increase of core temperature after ATWS accident. The research method proposed in this paper not only lays a solid foundation for the whole system modeling and simulation of gas-cooled micro reactor, but also provides a good reference for the system modeling and simulation of other types of reactors.

Discussion on DEC of High-level Radioactive Waste Liquid Storage System from the Perspective of Multiple-failure

Lyu Dan, Yang Xinjing, Wang Shijun, Yang Zhiyi, Yang Hao, Xu Chunyan, Liu Xinhua

2022, 43(2): 160-166. doi: 10.13832/j.jnpe.2022.02.0160

Abstract(194) HTML (44) PDF(22)

Abstract:
The analysis of design extension condition (DEC) is an important part of the analysis of beyond-design-basis accidents of nuclear power facilities. Currently there is no such practice in the field of post-processing facilities. The high-level radioactive waste liquid storage system of the post-processing facility is taken as the research and demonstration object. Based on engineering judgment and deterministic methods, DEC is identified from the perspective of multiple-failure. The results show that among the 29 working conditions of high-level radioactive waste liquid storage system, 14 conditions will not cause excessive release of radioactive substances to the environment, which belongs to DEC; 8 conditions may cause excessive release of radioactive material to the external environment. If the concrete pouring layer has the function of containing the waste liquid and the equipment room has the function of venting or anti-explosion, then these 8 conditions will not cause excessive release to the external environment and can also be included in DEC. The remaining 7 conditions will cause excessive release of radioactive material to the external environment. By improving the equipment reliability of exhaust filtration and chimney in the filter room of orange area, the magnitude of radioactive aerosol release to the environment shall not exceed the release level of hypothetical accident of siting.

Analysis on Deterministic Behavior Design of Safety Digital Instrumentation and Control System

Wu Qiaofeng, Liu Hongchun, Sun Shiyan, Li Yu, Wang Lin, Zhang Junqi, Wu Kunren

2022, 43(2): 167-170. doi: 10.13832/j.jnpe.2022.02.0167

Abstract(218) HTML (120) PDF(25)

Abstract:
The behavior logic of safety digital instrumentation and control (I&C) system is carried by software, but the software reliability evaluation is relatively difficult. Therefore, in order to ensure reproducibility and timeliness of safety digital I&C system, and ensure the reliability and safety of system, behavior deterministic design is required. According to the standard requirements and combined with engineering experience, this paper puts forward the deterministic design requirements of safety digital I&C system, and from the two aspects of its certainty, puts forward that the reproducibility of the system can be guaranteed through model-based formal modeling, and the timeliness of system response can be guaranteed through the allocation of response time of each link of the system. It provides a reference for the behavior deterministic design of safety digital I&C system.

Fault Diagnosis Method of Nuclear Gate Valve Based on Characteristic Analysis of Operation Process Variables

Liu Zhilong, Li Tongxi, Nie Changhua, Zhan Li, Tang Zhangchun, Liu Jie

2022, 43(2): 171-174. doi: 10.13832/j.jnpe.2022.02.0171

Abstract(280) HTML (86) PDF(40)

Abstract:
Aiming at the sticking fault of nuclear gate valve, a fault diagnosis method of gate valve based on characteristic analysis of operation process variables is proposed. The operating process of gate valve opening and closing often contains fault characteristics and changing rules. Therefore, this method first uses Shannon entropy to measure the vibration signal power spectrum of gate valve opening and closing process, calculates the mean value of power spectrum entropy as the target process variable, analyzes the characteristic changes of target process variables under the condition of gate valve health and fault degree, and then divides the fault area and non-fault area for gate valve fault diagnosis. Finally, based on the nuclear gate valve experiment, this method is experimentally verified. The results show that this method can effectively diagnose the fault of nuclear gate valve, and has a certain fault prediction ability. Therefore, the use of this method can reduce the probability of nuclear facility accidents caused by the sticking fault of the gate valve, and at the same time, this method can be applied to the fault diagnosis of the gate valve in other fields.

Experimental Study on Wind Load Performance of ACP100 Passive Containment Air Cooling System

Wang Hongliang, Yu Mingrui, Li Yunyi, Liu Changliang, Han Xu, Li Lujun

2022, 43(2): 175-180. doi: 10.13832/j.jnpe.2022.02.0175

Abstract(432) HTML (122) PDF(53)

Abstract:
Environmental wind field has important influences on the operation of passive containment air cooling system (PAS) of ACP100, a modular pressurized water reactor nuclear power plant. A small-scale model of ACP100 is built in the wind tunnel platform to explore the influence of different environmental wind direction angles and wind speed conditions on the operation of PAS. At the same time, the ACP100 test model (mainly composed of wind tunnel platform model and ACP100 small-scale model) and the original model are numerically simulated by Ansys Fluent. The results show that all environmental wind direction angles and wind speed conditions are conducive to PAS heat transfer; with the increase of the environmental wind speed, the PAS inlet and outlet pressures have a quadratic decreasing relationship with the environmental wind speed. When the environmental wind speed reaches 12.5 m/s, the pressure difference between the inlet and outlet of PAS is positively proportional to the square of the environmental wind speed; The numerical simulation results of ACP100 original model show that the greater the ACP100 wind speed, the better the PAS heat exchange effect. When the ACP100 wind speed reaches 20 m/s, the increase of PAS heat exchange power corresponding to each environmental wind direction angle is 23.0% ~56.5% compared with that under windless conditions, with an average increase of 35.6%. The research results provide a powerful reference for the design and optimization of ACP100.

Research on Thermal Safety of Intensive Spent Fuel Dry Storage Facility for Heavy Water Reactor

Xu Zhen, Ren Bing, Liu Zhan, Wang Zhe, Ye Qing, Guo Wei

2022, 43(2): 181-188. doi: 10.13832/j.jnpe.2022.02.0181

Abstract(269) HTML (58) PDF(42)

Abstract:
In order to solve the problem that Qinshan No.3 Nuclear Power Co., Ltd. (hereinafter referred to as TQNPC) plans to extend the service life of the nuclear reactor, resulting in the increase of spent fuel and insufficient capacity of existing spent fuel dry storage module, an intensive spent fuel dry storage facility (M1 spent fuel storage module) is developed based on the original No.1~6 (QM-400) spent fuel storage modules. Compared with QM-400 spent fuel storage module, M1 spent fuel storage module has larger storage capacity and higher energy density. In order to demonstrate the thermal safety of M1 spent fuel storage module, RELAP5/MOD3 program is used to establish the thermal analysis model of M1 spent fuel storage module according to the conservative initial assumptions. The temperature of each part of the module under normal operation and accident analysis under extreme weather conditions is calculated. The 3D computational fluid dynamics (CFD) program is used to verify the result of RELAP5 program. The thermal safety of M1 spent fuel storage module is verified according to the calculation results from RELAP5 and CFD programs.

Design and Verification of Algorithm of Reactor Neutron Doubling Time Based on SCADE

Yu Heng, Wang Yinli, He Zhengxi, Huang Youjun, Jiang Tianzhi, Lin Chao, Yang Zhenlei, Zhang Mi

2022, 43(2): 189-193. doi: 10.13832/j.jnpe.2022.02.0189

Abstract(274) HTML (76) PDF(35)

Abstract:
To realize the close monitoring of the nuclear fission rate of reactor during the period from fuel loading to power ascension, the correct and stable measurement of reactor neutron doubling time shall be implemented. Based on the statistical analysis of neutron fluence rate measurement, a doubling time algorithm suitable for PWR nuclear instrument system is designed and implemented by SCADE software. The in-pile test is carried out on Unit 2 of Fangjiashan nuclear power project. Tests verify the stability, timeliness and effectiveness of the algorithm. Therefore, the reactor neutron doubling time algorithm designed in this study can be applied to the processing of the neutron fluence rate measurement signal of the PWR nuclear instrumentation system.

Study on Nuclide Diffusion in Closed Environment of Marine Reactor with Large Break Loss of Coolant Accident

Zhao Fang, Zou Shuliang, Xu Shoulong, Xu Tao

2022, 43(2): 194-198. doi: 10.13832/j.jnpe.2022.02.0194

Abstract(341) HTML (67) PDF(55)

Abstract:
Based on the research method of severe accident analysis program MELCOR coupled computational fluid dynamics software CFD-FLUENT, MELCOR is used to analyze the loss of coolant accident of marine reactor. The results are used as the initial conditions of CFD-FLUENT simulation experiment to study the diffusion of radionuclides in the reactor compartment with large break loss of coolant accident of marine reactor. The results show that when the leakage time is 45 min, the radionuclides diffuse to the inlet and outlet of the coolant, but at 14 min, the radionuclides begin to spread to the containment, at 51 min, the radionuclides begin to spread from the break of the containment to the containment, and at 87 min, the radionuclides begin to spread to the adjacent compartment. The calculation results of this study can provide theoretical support and data support for emergency decision-making of nuclear accidents.

Researh on Vector Control Technology of Synchronous Reluctance Motor Control Rod Drive Mechanism

Peng Renyong, Wang Jinxin, Qing Xianguo, Liu Yiyi, Zhang Jianjian, Liu Yanan

2022, 43(2): 199-203. doi: 10.13832/j.jnpe.2022.02.0199

Abstract(413) HTML (97) PDF(52)

Abstract:
Aiming at the problem that the synchronous reluctance motor control rod drive mechanism (CRDM) is highly electromagnetically coupled and it is difficult to effectively adjust the output torque linearly, this paper studies the vector control technology of the synchronous reluctance motor CRDM. The mathematical model of the synchronous reluctance motor of the CRDM is mapped into the synchronous rotating coordinate system, and the output electromagnetic torque and the flux linkage are decoupled. And the excitation current and the torque current are independently adjusted to realize the linear control of the output torque. The vector control model of the synchronous reluctance motor CRDM was built using MATLAB/SIMULINK to verify the control scheme. The simulation results show that the control scheme of this research has better response speed, steady-state accuracy and stability.

Numerical Study on Hydrogen Flow Distribution Characteristics in Small-Scale Space

Liu Hanchen, Wu Xinzhuang, Xiang Wenjuan, Liu Jie, Wu Huiping

2022, 43(2): 204-211. doi: 10.13832/j.jnpe.2022.02.0204

Abstract(309) HTML (164) PDF(39)

Abstract:
Different from the large space of nuclear power plant containment, in small-scale space such as containment compartment and advanced small reactor, the flow of mixed gas of hydrogen and steam is limited by the wall, and the gas flow cannot fully develop, which may lead to the accumulation of hydrogen in some locations and lead to hydrogen risk. In this paper, the distribution characteristics of hydrogen flow in small-scale space are studied by means of numerical simulation and theoretical analysis. It is found that under typical working conditions, a hydrogen concentration reserve area with relatively uniform hydrogen concentration distribution is formed in the upper part of the small-scale space, and the hydrogen concentration transition zone and high air concentration zone are formed in the middle and lower areas, respectively. With the increase of the momentum of the source term gas, the ability of the source term gas to enter the upper space increases, resulting in the increase of hydrogen concentration in the upper area of the space. This study can provide support for the follow-up hydrogen risk research and analysis of advanced small reactors.

Numerical Study on the Flow Characteristics of the Parallel Main Pumps of the Vertical Canned Motor

Zhou Xingzhu, Song Yu, Yin Junlian, Wang Dezhong, Xia Shuan, Feng Lei

2022, 43(2): 212-218. doi: 10.13832/j.jnpe.2022.02.0212

Abstract(273) HTML (65) PDF(73)

Abstract:
Taking the reactor main coolant circulation pump (referred to as the main pump) of the vertical canned motor with a scale factor of 1:4 as the research object, two kinds of geometric models with two counter-rotating parallel main pumps (model 1) and two co-rotating parallel main pumps (model 2) are established. Using computational fluid dynamics (CFD) method to calculate the steady-state operation of the internal flow field of the two models of parallel main pumps, from the external characteristics of the main pumps, the inlet flow characteristics, the inflow quality, and the pressure distribution in the pipe, the comparative analysis of model 1 and model 2 is conducted. The results show that the performance of the main pumps A and B in model 1 are basically the same; the relative deviation of the flow rate of the main pumps A and B in model 2 is basically within 0.8%, the maximum value reaches 1.69%, the relative deviation of head is stable within 1%; the maximum relative deviation of the efficiency and shaft power reaches 6% and 8% respectively; compared with model 1, model 2 has better flow stability, higher inflow quality, and lower pressure distribution in the pipe, which is conducive to the long-term operation of the equipment.

Fatigue Reliability Test and Evaluation of Main Pump Spindle of Nuclear Power Plant Reactor

Zhang Jianxin, Gu Jipin, Chen Shuming, Pu Enshan, Guo Xiaoxian, Li Hailiang, Fang Jinghui

2022, 43(2): 219-225. doi: 10.13832/j.jnpe.2022.02.0219

Abstract(323) HTML (62) PDF(52)

Abstract:
In order to obtain the fatigue reliability data of the main spindle material of the nuclear power plant reactor under a given confidence level and different reliability levels, a simulation spindle with the same inner and outer diameter dimensions and processing technology as the product spindle was manufactured for sampling. The fatigue lives of six kinds of specimens including room temperature smooth, room temperature weld, room temperature notch, high temperature smooth, high temperature weld and high temperature notch were tested. The lower confidence limit equation of the fatigue reliability life with different reliability levels and the reliability fatigue limit of the six kinds of specimens were determined by reliability statistical methods when the confidence level was 0.9 or 0.95. The differences and revised methods between the fatigue life of the specimens and the spindle were analyzed. Using the revised specimen data, the reliability assessment of the spindle fatigue failure weak links were carried out. The results show that the reliability of the spindle without fatigue failure during the lifetime exceeds 0.9999 when the confidence level is 0.9 or 0.95. In this paper, a more accurate reliability assessment of spindle fatigue failure of reactor main pump is achieved.

Test and Verification for Digital Nuclear Instrumentation System Prototype on Reactor

Wang Yinli, He Zhengxi, Bao Chao, Gao Zhiyu, Wu Wenchao, Luo Tingfang, Yu Heng, Luo Wei

2022, 43(2): 226-231. doi: 10.13832/j.jnpe.2022.02.0226

Abstract(283) HTML (85) PDF(36)

Abstract:
In view of the current situation of nuclear instrumentation system equipment mainly relying on import in the current domestic nuclear power plant, a set of digital nuclear instrumentation system prototype was designed and developed. The system prototype mainly includes neutron detector assembly, signal conditioning and processing prototype and signal monitoring equipment. By introducing the installation and test of the prototype on the commercial reactor, the test data during reactor startup, power increase, full power and power reduction are analyzed in detail. The test results show that the neutron detector cooperates well with the signal conditioning and processing prototype, and the whole system prototype runs stably and reliably.

Test Method on Starting Drag Torque of the Thrust Bearing of the Reactor Main Pump in Nuclear Power Plant

Zhang Jianxin, Gu Jipin, Chen Shuming, Wang Mingzheng, Liu Xiaojun

2022, 43(2): 232-236. doi: 10.13832/j.jnpe.2022.02.0232

Abstract(268) HTML (209) PDF(30)

Abstract:
In order to obtain the ultimate starting drag torque of the thrust bearing of the reactor main pump in nuclear power plant during its lifetime, and ensure that the auxiliary motor performing the accident residual heat removal function can start the main pump under extreme conditions, a test method of the starting drag torque (which refers to the drag torque at the moment of starting) of the thrust bearing was proposed and a test device was designed. The orthogonal test method was used to study the three influencing factors (roughness, specific pressure and lubricating oil temperature) that affect the starting drag torque of the thrust bearing. The single factor method was used to test the influence of different downtime (which refers to the static loading time) on the starting drag torque of the thrust bearing. The results show that the starting drag torque changes little when the three factors change within the specified control range, but the downtime has a great impact on the starting drag torque of the thrust bearing. The design of the auxiliary motor was carried out based on the ultimate starting drag torque determined by the test, and has been verified by repeated start-stop tests of the thrust bearing prototype and the main pump prototype. The research in this paper can provide accurate and reliable input for the design of the starting drag torque of the auxiliary motor.

Calibration and Verification of AFD Used for PWR Nuclear Power Plant Control System with the Mechanical Shim

Wei Guangjun

2022, 43(2): 237-241. doi: 10.13832/j.jnpe.2022.02.0237

Abstract(300) HTML (93) PDF(46)

Abstract:
The rod shadow effect of the mechanical shim of PWR nuclear power plant causes the axial flux deviation (AFD) indicated by the protection system to become the weighted value of peripheral components rather than the average value of the core. Therefore, the control system based on the average AFD of the core cannot directly use the output value of the protection system like the traditional power plant. Based on this problem, several linear indication ways of core average AFD of control system are explored, the effects of different control rod positions and xenon oscillation transient on linear relationship are studied, and a core average AFD calibration method with “deviation correction” is proposed. Through the verification of the unit peak shaving process, the results show that the method can eliminate the influence of the control rod position, the unit power change and the xenon concentration change on the AFD indication of the control system, and can meet the requirements of the system indication accuracy. Therefore, this method can be used for the calibration of the mechanical shim control system AFD.

Research on Pressurizer Level Measurement Based on Sub-regional Density Compensation

Liu Zhilong, Zhan Li, Nie Changhua, Liu Jie, Tang Zhangchun, Li Tongxi

2022, 43(2): 242-245. doi: 10.13832/j.jnpe.2022.02.0242

Abstract(320) HTML (63) PDF(44)

Abstract:
In view of the large temperature difference between the upper part and the bottom of the pressurizer when the electric heating element heats the pressurizer, which leads to the large measurement error of the liquid level measurement of the traditional pressurizer differential pressure method, a pressurizer liquid level measurement method based on sub-regional density compensation is proposed. Firstly, according to the actual situation, the pressurizer is divided into saturated area and unsaturated area. The saturated area is the area where the saturated steam is located. The measured temperature is used to compensate the saturated steam density; the unsaturated area is the area where the medium water is located, and the average temperature of the unsaturated area is used to compensate the medium water density. Secondly, a polynomial fitting model based on the least square method is established in the saturated and unsaturated areas of the pressurizer to compensate the density variable, and then the liquid level is calculated by combining the density of the cold water section. Finally, the experiment is carried out on the experimental device and compared with the reference liquid level. The experiment shows that the pressurizer liquid level measurement method proposed in this paper can obtain reliable measurement results, so this method can be widely applied to pressure vessel liquid level measurement in industrial fields including nuclear industry.

Condition Prediction of Reactor Coolant Pump in Nuclear Power Plants based on the Combination of ARIMA and LSTM

Zhu Shaomin, Xia Hong, Lyu Xinzhi, Lu Chuan, Zhang Jiyu, Wang Zhichao, Yin Wenzhe

2022, 43(2): 246-253. doi: 10.13832/j.jnpe.2022.02.0246

Abstract(491) HTML (104) PDF(70)

Abstract:
To monitor and track the operation process and improve the early warning of the reactor coolant pump (RCP) in nuclear power plants (NPPs), a hybrid RCP condition prediction approach based on autoregressive integrated moving average (ARIMA) model and long short-term memory (LSTM) neural network is proposed in this paper. This method is used to predict the thrust bearing temperature and controllable leakage flow of the RCP of a nuclear power plant in one step and multiple steps, and the prediction accuracy is evaluated with the root mean square error (RMSE) as the index. The results show that the combination model of ARIMA and LSTM neural network can accurately predict and track the state of the RCP, and the prediction accuracy of the combination model is better than that of single ARIMA or LSTM model, especially in the multi-step prediction, the advantage of the combination model is more obvious.

Ultrasonic Inspection Process Design for CRDM Pressure Housing Weld in Nuclear Power Plant

Tang Jianbang, Yu Zhe, Wang Weiqiang, Sun Jiawei, Lyu Tianming

2022, 43(2): 254-258. doi: 10.13832/j.jnpe.2022.02.0254

Abstract(377) HTML (109) PDF(49)

Abstract:
The pressure housing of the control rod drive mechanism (CRDM) belongs to the main circuit of the nuclear power plant, and its joint weld is the weak link of the pressure boundary of the whole radioactive circuit. Its safety and reliability directly affect the safe operation state of the reactor. In view of its narrow space, thin wall thickness and poor accessibility, a special flat bicrystal focused ultrasonic probe and inspection process are designed by using simulation technology. The test results meet the requirements of the regulations. The supervision difficulties of in-service inspection of nuclear power plant are solved, and the process design and verification method of similar weld inspection of class I components of main circuit of nuclear power plant are obtained.

2022 Vol. 43, No. 2