核动力工程

Practice and Learning in the Field of Structural Integrity during the HPR1000 International Regulatory and Certified Review

Mao Qing, Zhang Tao, Xu Xiao, Cen Peng, Wang Guofeng

2024, 45(2): 1-9. doi: 10.13832/j.jnpe.2024.02.0001

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Abstract:
In order to promote the international market development, China General Nuclear Power Group (CGN) has carried out the Generic Design Assessment (GDA) of UK and the European Utility Requirements (EUR) assessment. The structural integrity of the structures, systems and components in nuclear power plant is the focus of international review and certification. In order to meet the requirements of British nuclear safety supervision and EUR, the project team adopted the design concept of analytical method, and completed the structural integrity analysis, evaluation and design improvement of British and European versions of HPR1000 with mechanical analysis, test and demonstration as the key means, which fully met the requirements of British nuclear safety supervision and EUR. Through the HPR1000 international regulatory and certified review, the relevant technical requirements in the field of structural integrity are studied, and the practice and learning in the review process are summarized, which can support the continuous improvement and innovation of HPR1000 technology.

Research on the Source Convergence Diagnose Method of Monte Carlo Critical Calculation for Loosely Coupled System

Zhang Yin, Cheng Yuting, Zhou Qi, Zhu Qingfu, Xia Zhaodong, Ning Tong, Zhang Zhenyang

2024, 45(2): 10-18. doi: 10.13832/j.jnpe.2024.02.0010

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In order to improve the reliability and accuracy of Monte Carlo code in the critical safety calculation of loosely coupled systems, it is necessary to determine the source convergence of the results. In this paper, an improved Shannon entropy diagnose method considering the distribution of fission sources and the statistical deviation weight factor is proposed. The relative deviation of fission sources distribution between adjacent generations and the standard deviation of fission source distribution during source iteration are used as the weight of fission source distribution. Finally, an improved Shannon entropy convergence index is established, which makes up for the shortcomings of traditional Shannon entropy for lacking the details of local fission source. The improved Shannon entropy convergence index is applied to the benchmark problem: spent fuel rod and loosely coupled solution slabs published by OECD/NEA. The results show that compared with the traditional Shannon entropy index, the improved Shannon entropy is more sensitive to the iterative convergence process of fission source, and can more intuitively and accurately determine the convergence of fission source distribution accompanied by source iteration. For a typical case with slow convergence speed, the conclusion of pseudo-convergence is given with the traditional Shannon entropy. The improved Shannon entropy can accurately determine the number of iterations when the source iteration reaches convergence.

Research on Burnup Calculation Method Based on Splicing Fission Matrix

Gao Ruishuang, Liu Xiaojing, He Donghao, Pan Qingquan

2024, 45(2): 19-23. doi: 10.13832/j.jnpe.2024.02.0019

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In order to realize efficient transport-burnup coupling calculation, a new calculation method based on splicing fission matrix theory is proposed in this paper. The combined fission matrix method enables to obtain the system fission matrix from a pre-calculated database, and then the database matrices are combined according to the actual conditions of the calculation model and the properties of the terminal area. The fission matrix database is obtained by Monte Carlo fixed source calculation, and the core calculation does not need Monte Carlo simulation, so the time-consuming critical calculation can be avoided. In this paper, the effective multiplication factor and fission source distribution of a typical two-assembly model with a burnup up to 600EFPD are obtained. The burnup-enrichment composite correction ratio can reduce the root-mean-square error of the fission source to less than 0.7%, thereby confirming the feasibility of this algorithm.

Verification and Validation of NECP-Bamboo Based on Measurement Data from Nuclear Power Plants

Liang Yilin, Li Yunzhao, Zhou Yuancheng, Li Yisong, Zhang Hengrui, Zhou Shilong, Wang Weiguo, Ou Yuxiang, Wang Songzhe, Qin Junwei, Shao Ruizhi

2024, 45(2): 24-34. doi: 10.13832/j.jnpe.2024.02.0024

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The verification and validation (V&V) are important aspects in the life cycle of engineering software, and reflect the process of software development to practical application. They are also significant symbols of software from "bookshelf" to "goods shelf". In this paper, V&V for the software NECP-Bamboo were carried out based on the actual measurement data of five types of commercial pressurized water reactors, including CNP300, M310, CNP650, BEAVRS and HPR1000, for a total of 48 operation cycles. The results show that the errors between the calculated values and the measured values of the key reactor parameters, such as control rod worth, temperature coefficient, critical boron concentration and assembly power distribution, obtained by using NECP-Bamboo can meet the requirements of industrial limits. The 95% confidence interval range of the corresponding key parameter errors of various types of reactors is summarized as follows: [−37.80, 35.39]ppm for critical boron concentration, [−6.18%, 3.68%] for control rod worth, [−3.27, 2.99] pcm/K for the reactivity coefficient of temperature, [−0.64%, −0.12%] for the assembly power distribution when the relative power is greater than 0.9, and [1.18%, 2.94%] for that when the relative power is less than 0.9.

Study on New Burnable Poison Material Combinations Based on Long-life Pressurized Water Reactor

Tong Ji, Qu Lingyi, Xie Jinsen, Xu Shikun, Yu Tao

2024, 45(2): 35-41. doi: 10.13832/j.jnpe.2024.02.0035

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Long-life pressurized water reactors using a single burnable poison assembly design is not the optimal choice. In order to meet the comprehensive requirements of burnable poisons for reactivity control in long-life pressurized water reactors, this paper conducts a combined study on new burnable poisons with good neutronics characteristics: ²³¹Pa₂O₃, PACS-J, PACS-L, ¹⁶⁷Er₂O₃, and ¹⁵⁷Gd₂O₃. The results show that the combination of "fast burnup" and "slow burnup" burnable poisons can achieve better results. For fuel assemblies with "low enrichment", the combination of ²³¹Pa₂O₃ with PACS-J, PACS-L with ¹⁶⁷Er₂O₃ can be selected; for fuel assemblies with "high enrichment", the combination of ²³¹Pa₂O₃ with PACS-J, ¹⁵⁷Gd₂O₃ with ¹⁶⁷Er₂O₃ can be selected. Therefore, this paper realizes the initial excess reactivity control, smooth release of reactivity, and increased burnup level of long-life pressurized water reactors through the combined design of "fast burnup" and "slow burnup" burnable poisons.

Analytical Solution Study on Transient Neutron Transport in Homogeneous Bare Reactor Based on Neutron Telegraph Equation

Wang Weiguo, Li Yunzhao, Qin Junwei, Cao Liangzhi

2024, 45(2): 42-46. doi: 10.13832/j.jnpe.2024.02.0042

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The Fick's law, as the foundation of the diffusion theory, further neglects the time derivative term of the neutron current density in the P₁ approximation equation (telegraph equation) of the neutron transport equation. Therefore, it is difficult to accurately describe the actual neutron kinetic behavior in the transient process. In this paper, based on the monoenergetic neutron telegraph equation, an analytical solution is derived for the one-dimensional infinite slab bare reactor neutron kinetics problem using the method of separation of variables, and it is compared and analyzed with the analytical solution of the neutron diffusion equation. The study reveals that during transient changes, the spatial term of the telegraph equation still maintains the form of a cosine function compared to the diffusion equation's solution, but the variation of the temporal term is more complex. Firstly, the combination form of the temporal term's orders is influenced by the geometry and materials of the problem. Secondly, higher-order harmonics exhibit oscillatory changes with time. These research findings can provide references and foundations for subsequent numerical theoretical studies based on the neutron telegraph equation.

Research on Evaluation Method of Actinide Nuclide Activities in Primary Coolant System of Pressurized Water Reactor Nuclear Power Plant

Xiong Jun, Lyu Weifeng, Guo Runchun, Gao Yaoyi, Jiang Zhenyu

2024, 45(2): 47-52. doi: 10.13832/j.jnpe.2024.02.0047

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In order to evaluate the radiation risk of workers and the damage degree of fuel cladding in PWR nuclear power plant, it is necessary to improve the evaluation approach of actinide nuclide activities in the primary coolant systems based on characteristic physical quantities. Based on the generation and migration mechanism of actinide nuclides, the balance equations of actinide nuclides in primary coolant system are established, and three easy-to-monitor characteristic physical quantities are selected to evaluate the release and distribution of actinide nuclides to the primary coolant system. Hence, the evaluation approach of actinide nuclide activities in primary coolant systems is established. The evaluation approach is validated by the measurement data without and with fuel cladding damage from in-service nuclear power plant in China, respectively. The validation results show that the established approach is applicable for the evaluation of actinide nuclide activities in the primary coolant system without and with fuel cladding damage, and the evaluation results are in line with expectations. The results of this paper can provide guidance for the assessment of actinide nuclides and their distribution in the primary coolant system during the operation of PWR nuclear power plants, so as to optimize the back-end staff protection measures and reduce the radiation risk.

Study on Neutron Diffusion Calculation Method Based on hp-VPINN

Zeng Fulin, Zhang Xiaolong, Zhao Pengcheng

2024, 45(2): 53-62. doi: 10.13832/j.jnpe.2024.02.0053

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Advanced reactor simulations require the inversion of key parameters of the whole reactor based on less actual detection data. To meet this need, this paper constructs a computational model based on variational residual physics-informed neural network (hp-VPINN) with high-order polynomial domain decomposition function, which is used to solve the neutron diffusion equation forward and backward. The model uses neural network as trial function, and substitutes it into the neutron diffusion equation to form variational residual as loss function for gradient descent. In order to improve the accuracy and efficiency of the solution, this paper also proposes some innovative key technologies such as effective multiplication factor intelligent search and inversion based on the physical characteristics of neutron diffusion equation, and realizes the self-optimization of neural network hyperparameters based on whale optimization algorithm (WOA). Finally, several examples are verified, and the results show that the method has high accuracy and low dependence on training data, providing a solution with less training data and higher accuracy output for advanced reactor simulations.

Study on Load-Following Characteristics of a Thermionic Space Reactor Power System

Han Xufan, Ouyang Zeyu, Wang Zhao, Shan Jianqiang

2024, 45(2): 63-71. doi: 10.13832/j.jnpe.2024.02.0063

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In order to analyze the power load-following characteristics of thermionic space reactor power system, TOPAZ-II thermionic space reactor system code is established by using the simulation language of reusable hierarchical component model. The influence of Cs steam pressure and electrode gap on the output power is analyzed. The load tracking operation characteristics of thermionic space reactor power supply in orbit under different load changes are analyzed by using the method of core reactivity feedback and external load resistance collaborative control. The results show that the moderator temperature does not change significantly when the steady-state electric power is 5.5 kW and the electric power varies between 0.95 kW and 7.25 kW, and the reactor has self-stability. Beyond this range, the reactor loses self-stability, which is closely related to the positive temperature reactivity feedback of the moderator.

Experimental Study on Heat Transfer Characteristics of Pure Steam with Incomplete Condensation in Vertical Tube

Liu Jiabao, Cao Xiaxin, Yang Peixun

2024, 45(2): 72-81. doi: 10.13832/j.jnpe.2024.02.0072

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In order to study the heat transfer characteristics of pure steam with incomplete condensation in a vertical tube, an experiment was carried out using a heat exchange tube with an inner diameter of 25 mm, with an inlet pressure of 0.1~0.3 MPa and a steam mass flux of 12~70 kg/(m²·s). The effects of inlet pressure, mass flux and mass quality on the average and local condensation heat transfer coefficients in the tube were investigated. The liquid film flow pattern in the condensation process was identified, and the effects of liquid film turbulence and droplet entrainment on the condensation heat transfer in the tube were analyzed. It is shown that the condensation heat transfer coefficient increases with the increase of mass flux and mass quality. However, the condensation heat transfer coefficient of vertical tube decreases with the increase of inlet pressure. The liquid film flow pattern in the experiment is mainly in the transition flow region, and the occurrence of droplet entrainment increases the local condensation heat transfer coefficient. Four annular flow condensation heat transfer equations are compared. The results show that the basic deviation of Shah's empirical equation is within ±30%, and the mean absolute deviation (MAD) was 18.91%. An empirical correlation based on the experimental data is developed, and the basic deviation between the calculated value and the experimental value is within ±10%.

Numerical Simulation of Pressure and Temperature Fields in Rectangular Narrow Channel under Blistering Condition

Wang Xuejian, Zhao Yanan, Yu Tao

2024, 45(2): 82-87. doi: 10.13832/j.jnpe.2024.02.0082

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When the rectangular narrow channel is deformed due to radiation blistering, the flow heat transfer characteristics in the channel will be changed. In this study, the fluid pressure field and temperature field in the rectangular narrow channel under blistering condition is analyzed by numerical calculation method. The results show that the blocking effect of blistering on the fluid leads to a high pressure point before the first blistering. After each blistering, the local pressure drop suddenly rises first and then suddenly drops, causing the fluid temperature at the blistering to rise and the wall temperature to decrease. The influence of blistering conditions on the pressure and temperature fields is obtained through model calculation and analysis.

Simulation Study on Transport Characteristics of Radioactive Aerosol in Containment

Tian Jiaming, Wang Yueshe, Li Biao, Zou Lin

2024, 45(2): 88-95. doi: 10.13832/j.jnpe.2024.02.0088

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In order to find out the transport characteristics of radioactive aerosol in a serious accident under the actual containment size, the spatial distribution of radioactive aerosol in a serious accident was simulated by using the coupling of computational fluid dynamics and particle swarm equilibrium equation, and the effects of different coalescence and deposition mechanisms on the aerosol transport process are quantitatively analyzed. The results show that the interaction among aerosol particulates with a particle size smaller than 0.1 μm are mainly driven by Brownian coalescence, while those larger than 10 μm are dependent mainly on turbulent inertial coalescence, and those between the two sizes are dominated by both of Brownian coalescence and turbulent coalescence (turbulent inertial coalescence and turbulent shear coalescence). For the deposition phenomenon, the aerosols with particle size less than 0.1 μm rely mainly on Brownian deposition, while those larger than 0.1 μm are mainly affected by gravity deposition. It is found that the average coalescence velocity of turbulent coalescence is 2.99 times that of Brownian coalescence, and the average deposition rate of Brownian deposition is 1.38 times that of gravitational deposition. This study provides a solution for the selection of radioactive aerosol removal technology under the actual containment size.

Experimental Study on Core Flow Instability Caused by Vortex Flow in Reactor Lower Plenum

Meng Yang, Liao Hengji, Jiang Lin, Fang Ying, Zhang Jiaqi, Li Yong, Yang Zumao

2024, 45(2): 96-102. doi: 10.13832/j.jnpe.2024.02.0096

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Strong turbulent mixing of coolant in each loop of the PWR occurs in the lower plenum, and a large number of vertex are generated in the lower plenum, which can lead to random fluctuation of the fuel assembly inlet flow and trigger transient flow instability in the core, thus affecting the reactor thermal safety, structural safety or heat transfer performance. In this paper, the flow characteristics of the fuel assembly area in the reactor are studied, and the flow pulsation data of the fuel assembly inlet with and without flow distribution skirt in the lower plenum are obtained by hydraulic test under various operating conditions. The experimental results show that the lower plenum flow distribution skirt has obvious effect on restraining vortex, and the core flow pulsation is obviously reduced under the action of vortex breaking rectification. With the decrease of the number of operating loops, the symmetry of the flow field in the lower plenum decreases, the vertex flow increases, and the flow pulsation in the core increases obviously. The vortex in the lower chamber will also adversely affect the uniformity of flow distribution at the inlet of the core, and the flow distribution factor corresponding to the area with large flow pulsation is obviously small.

Analysis of Steady and Transient Characteristics of Once-through Steam Generator for Lead Bismuth Fast Reactor

Huang Zhe, Liang Tiebo, Yang Wen, Lu Chuan, Li Yang, He Zhonghai, Shen Xin

2024, 45(2): 103-109. doi: 10.13832/j.jnpe.2024.02.0103

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The safe and stable operation of lead bismuth fast reactor is closely related to the heat dissipation performance between the primary and secondary sides of the heat exchanger. In this study, the steady-state and transient coupled distribution parameter model of once-through steam generator (OTSG) based on lead bismuth fast reactor is established. The distribution difference of thermal and hydraulic characteristics in OTSG under different load conditions is analyzed and compared, and the influence of primary enthalpy and flow disturbance of lead bismuth fast reactor on the dynamic heat dissipation performance of heat exchanger is further revealed. The steady-state results show that the temperature drop on the primary side of the lead bismuth fast reactor was mainly concentrated in the subcooled boiling and nucleate boiling regions, and the decrease in the secondary side load can cause the temperature jump on the tube wall. The dynamic results show that the primary side inlet enthalpy only decreases by 5% under design conditions, which may lead to the lead bismuth reactor cycle entering accident conditions after 90 seconds. The relevant results provide valuable suggestions for the study of OTSG flow and heat transfer characteristics and structural design optimization of the lead bismuth fast reactor.

Study on Capillary Characteristics of High Temperature Liquid Sodium in Stainless Steel Wire Mesh Wick

Zhu Yiru, Ma Yugao, Zhang Luteng, Xi Zhiguo, Tang Simiao, Ma Zaiyong, Pan Liangming, Zhang Zhuohua, Ding Shuhua

2024, 45(2): 110-115. doi: 10.13832/j.jnpe.2024.02.0110

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The capillary characteristics of the alkali metal high temperature heat pipe wick are of great significance to the normal operation of the heat pipe. In this paper, the capillary rise method and the weighing method were used to carry out the experimental study on the capillary characteristics of the high temperature liquid metal sodium in the 304 stainless steel wire mesh wick with different mesh numbers in the glove box. The variation of the capillary performance of the wick at different temperature stages was obtained. The results show that the improvement of the capillary performance of the wick is limited below 460℃, which is due to the influence of Cr₂O₃ oxide film on the wettability of liquid sodium on the stainless steel surface. When the temperature is higher than 460℃, the liquid sodium reacts with Cr₂O₃, and the capillary performance of the wick is obviously enhanced. The damage of liquid sodium to the oxide film on the surface of stainless steel is irreversible, and the best wettability will be maintained in the temperature change stage of the subsequent heating cycle. The capillary characteristics of wick are mainly affected by surface tension and physical properties, showing a changing trend that the mass of wick decreases with the increase of temperature. According to the capillary theory and the structure of the wick, a capillary suction model of the wick is established. The error between the predicted sodium absorption mass and the experimental value in the cooling stage is within 12%.

Study on the Influence of Spacer Grid on Temperature Field Distribution at Rod Bundle Channel Outlet

Qiu Zicheng, Xie Shijie, Lang Xuemei, Li Pengzhou, Zhuo Wenbin

2024, 45(2): 116-122. doi: 10.13832/j.jnpe.2024.02.0116

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To analyze the mixing ability of the spacer grid and further optimize the thermal design method of the spacer grid, this article investigates the influence of spacer grid with different structures on the temperature field distribution at the outlet of rod bundle channel through experiments. In the experiments the system pressure is 7.0~16.5 MPa, the mass velocity is 900~4500 kg/m²s, and the temperature difference between the inlet and outlet of the experimental section is 30~128℃. The experimental results show that, within the range of PWR operating parameters, the ratio of the difference between the maximum and minimum temperatures of the sub-channel at the bundle channel outlet section to the temperature difference at the inlet and outlet has no obvious change trend under different thermal parameters such as pressure, inlet and outlet temperature difference and mass flow rate for the same spacer grid structure. The mixing wing dominates the mixing effect of the spacer grid, and Type II mixing wing has stronger mixing effect. For the D-type spacer grid, the asymmetric strip structure will lead to a certain deflection of the whole flow field, which will reduce the mixing effect of the spacer grid to some extent.

Effect of Baffle on Flow Induced Vibration of Heat Exchange Tubes

Wang Shuaiquan, Zhang Kai, Xiong Zhenqin, Shi Linpeng

2024, 45(2): 123-129. doi: 10.13832/j.jnpe.2024.02.0123

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The wear of tubes caused by flow induced vibration is one of the most significant factors that cause performance attenuation and malfunction of the heat exchanger. In this paper, aiming at the vibration caused by the non-uniform cross flow formed by the heat exchanger baffle, a visualized three-span 7×12 heat transfer tube bundle flow-induced vibration test facility is established. The acceleration data of the vibration in the first row of midspan tubes and the change data of the gap between the tubes and the support plate were obtained by the acceleration sensor in the tube and the visual vibration measurement by the high-speed camera, respectively, and the flow field between the tubes was obtained by numerical simulation. By comparing the experimental and simulation results of cross-flow-induced vibration of tubes under non-baffle mode and 2/3 gap baffle mode, it is shown that the velocity difference between the two structures after the third row of tubes is small, and the fluid elastic instability (FEI) critical velocity is similar, and the critical velocity of 2/3-gap baffle is slightly higher than that of the structure without baffle. By comparing five classical critical velocity relationships, it is found that Chen’s correlation can conservatively predict the occurrence of instability. The visual test of the gap between the tube and the support plate shows that the two kinds of tubes with incoming flow structure are relatively stable against the hole on one side of the support plate at very low flow velocity. At a moderate flow rate, the tube reciprocates in the hole of the support plate with a large amplitude under the non-baffle mode, and slightly knocks, while in the non-uniform inflow structure, the tube still slides on the inner edge of the plum blossom hole on the support plate with a small amplitude. In the case of FEI, the non-baffle mode is mainly knocking, while the 2/3-gap baffle will obviously slide and knock. The 2/3-gap baffle is more prone to sliding wear between the heat transfer tube and the support plate, which will threaten the integrity of the heat transfer tube.

Numerical Simulation Method of Boiling Heat Transfer and Its Application Characteristics under Multi-tube Coupled Heat Transfer

Zhang Zhenguo, Tan Sichao, Li Xiaochang, Liu Sichao, Feng Yi, Huang Yujian, Tian Ruifeng

2024, 45(2): 130-138. doi: 10.13832/j.jnpe.2024.02.0130

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Aiming at the problems of high difficulty in simulation, low computational efficiency, and large uncertainties when applying the method of computational fluid dynamics (CFD) to the analysis of coupled heat transfer on primary and secondary sides and full-regime flow boiling heat transfer in once-through steam generator (OTSG), based on the Eulerian two-fluid multiphase model and the critical heat flux density (CHF) wall boiling model, a numerical analysis model of the full-regime flow boiling heat transfer in the tube is established, and the effectiveness of the model is verified. Based on the verified model, the application characteristics of the numerical model under multi-tube coupling heat transfer are studied, the reliability of the numerical simulation method under multi-tube coupling is clarified, and the sensitivity of the calculation results of temperature and phase distribution to the interphase force model is numerically analyzed. The results show that based on Euler's two-fluid multiphase flow model and CHF wall boiling model, the full-regime flow boiling heat transfer process of water in the tube from supercooling to overheating can be predicted accurately. The location of dry-out point and the peak temperature of the wall are in good agreement with the experimental values, with a maximum error of less than 10%. The numerical method based on Euler two-fluid multiphase flow model and CHF wall boiling model has good applicability to multi-tube coupling conditions, and the calculated secondary side temperature is in good agreement with the experimental results. The interphase drag force has obvious influence on the calculation results of wall temperature and cavitation share, but the non-drag force has little influence on wall temperature. Therefore, for large-scale engineering application calculation, the influence of some interphase non-drag forces may not be considered in the analysis. The results of this paper can provide useful reference for the model selection of OSTG's three-dimensional refined numerical analysis.

Optimization and Thermal Safety Analysis of CFETR Advanced Small Sample Irradiation Capsule

Liu Chang, Liu Wenbin, Dai Yong, Sun Shouhua, Zhang Ping, Kang Changhu, Song Jiyang, Wang Kaimin, Peng Lei, Zheng Pengfei

2024, 45(2): 139-146. doi: 10.13832/j.jnpe.2024.02.0139

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The capsule structure of the advanced material irradiation test sample in China Fusion Engineering Experimental Reactor (CFETR) is relatively complicated, and the capsule is filled with helium. The size and location of the capsule rib and the material of the filling inside the capsule have great influence on the sample temperature. Based on the STAR-CCM+ code, a full-scale capsule model was established in the CFETR advanced small sample irradiation device, and the rib and filler materials of the capsule were adjusted according to the target temperature of the sample. For the problem that the heat release rate of the whole sample in the capsule is low, using tungsten material with higher heat release rate as filling material can significantly increase the overall sample temperature. For the case that the heat release rate of local samples is quite different, the temperature between samples can be well controlled by adjusting the size and position of the local rib, so that the calculated temperature of the sample can meet the target temperature range. The results show that after optimization with the above method, the sample center temperature can meet the target temperature range and meet the thermal safety of irradiation in the High Flux Engineering Test Reactor (HFETR), ensuring the smooth development of the entire irradiation task.

A Porous Media Model for Thermal-hydraulic Analysis of Wire-wrapped Fuel Assembly in Sodium Cooled Fast Reactor

Wang Xinan, Zhang Dalin, Wang Ting, Qiu Suizheng, Su Guanghui

2024, 45(2): 147-153. doi: 10.13832/j.jnpe.2024.02.0147

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In order to accurately predict the distribution of three-dimensional thermal-hydraulic parameters in the core of sodium-cooled fast reactor and reduce the demand for computing resources, a three-dimensional porous medium model for the wire-wrapped fuel assembly was established based on the concept of representative element volume. The interaction between the coolant and the solid surface is decomposed into distributed resistance force according to the geometry characteristics of the assembly. An effective heat transfer coefficient model including turbulent mixing heat transfer, fluid heat transfer and fuel rod heat transfer was introduced to describe the radial heat transfer of the assembly. Numerical analysis was performed for the liquid sodium cooled 37-pin wire wrapped fuel assembly experiment conducted by the Nuclear Energy Engineering Laboratory of Toshiba Corporation. Compared with the experimental results, the numerical calculation results show that the porous medium model proposed in this paper can well reproduce the experimental results under various conditions. Therefore, the porous medium model proposed in this study can be used to predict the distribution of three-dimensional thermal-hydraulic parameters of the wire-wrapped fuel assembly of sodium-cooled fast reactor.

Model Analysis of Fuel Assembly Grid Spring Stiffness

Jin Yuan, Zhou Sai, Chen Wei, Li Weicai, Zhang Yuxiang

2024, 45(2): 154-159. doi: 10.13832/j.jnpe.2024.02.0154

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Grid spring is a critical component of PWR fuel assembly, which provides clamp function for the fuel rod. Stiffness is a critical characteristic of grid spring, which is related to the in-pile operational performance of the fuel rod. According to the grid spring structure and mechanical characteristics, this paper proposes a mechanical analysis model of the grid spring, and the theoretical stiffness calculation formula is obtained through deduction, and the grid spring theoretical stiffness model is also obtained. Furthermore, the finite element stiffness models of grid spring with various sizes are established in the commercial finite element code ABAQUS, and the deformation and stiffness curves of grid spring are obtained by calculation. By comparison between the FEM analysis results and the theoretical calculation ones, the rationality of the theoretical stiffness model is proved, and then the advantages and shortcomings of the theoretical stiffness model are analyzed. The theoretical stiffness model of grid spring proposed for the first time in this paper can be used to replace the finite element iterative process during the design of grid scheme, and the main design dimensions of the optimized scheme can be obtained quickly. However, the theoretical method in this paper cannot replace the experiments, and tests are still needed to determine the stiffness of the grid spring after the scheme is solidified. The grid spring theoretical model in this paper provides a new idea to quickly optimize the fuel assembly grid spring design.

Improvement of Radiolytic Gas Model for Criticality Accident Analysis of Nuclear Fuel Solution System

Sheng Huimin, He Junyi, Gou Junli, Shan Jianqiang, Liu Guoming

2024, 45(2): 160-165. doi: 10.13832/j.jnpe.2024.02.0160

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The accurate simulation of transient criticality accident is a key factor in critical safety assessment of nuclear fuel solution system. However, the existing radiolysis gas models contain many empirical parameters, which result in significant deviations in the prediction of the power characteristics. To improve the simulation accuracy and avoid relying on the empirical parameters in the model, the radiolytic gas model needs to be improved. Based on the analysis of radiolysis gas behavior in solution and simplified assumptions, a conservation model including radiolytic gas concentration, mass per unit volume of radiolytic bubbles and number of density bubbles was established. This model was coupled with a point reactor kinetics model and a two-dimensional heat conduction model to develop a two-dimensional transient analysis code for solution systems. The code was verified with the Japanese TRACY experiment. The results show that the simulated values of the code are in good agreement with the experimental data, and the improved model can accurately simulate the power change of the solution system during critical accidents.

Analysis of Spent Fuel Cask Dropping Accident and Research on Relevant Improvement Measures

Tan Jingyao, Chen Yao, Ji Wenying, He Yingyong, Zhang Feng, Xia Yan, Ning Xinxian

2024, 45(2): 166-170. doi: 10.13832/j.jnpe.2024.02.0166

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"Cask immersion" is generally used in domestic nuclear power plants to deliver spent fuel, thus the risk of spent fuel cask dropping accident cannot be completely eliminated. The radioactive consequences of spent fuel cask dropping accident are analyzed and calculated, and how to ensure the integrity of the cask after dropping or how to contain radioactive materials after the cask integrity is broken is analyzed and studied. A series of improvement measures are put forward to alleviate the consequences of spent fuel ccask dropping, so as to reduce the risk of over-limit release of radioactive materials, which have positive guiding significance for improving the safety of spent fuel hoisting operation in nuclear power plants.

Research on Inter-unit Common Cause Failure of Multi-unit PSA

Yang Chunju, Wang Ming, Lin Modi, Zhang Bing, Wang Jinkai

2024, 45(2): 171-177. doi: 10.13832/j.jnpe.2024.02.0171

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Abstract:
How to reasonably estimate the contribution of inter-unit common cause failure (CCF) to the safety risk of power plant is an important technical problem that needs to be solved in multi-unit probabilistic safety analysis (PSA) modeling. In this study, analysis methods of inter-unit CCF group selection, modeling and parameter estimation are presented, and the loss of offsite power (LOOP) event of a four-unit pressurized water reactor (PWR) nuclear power plant is taken as an analysis case to quantitatively evaluate the change of core damage frequency (CDF) of multi-unit PSA model before and after considering CCF. The results show that after considering the inter-unit CCF, the frequency of core damage (CD) occurrence in only single unit and the frequency of CD occurrence in multiple units will increase. It can be seen that inter-unit CCF has a certain effect on multi-unit PSA results.

Simulation Study on Steam Dump Control System of Small Pressurized Water Reactor Nuclear Power Plant

Wang Linna, Chen Chuqi, Zeng Wenjie, Chen Lekang, Li Ruokun

2024, 45(2): 178-182. doi: 10.13832/j.jnpe.2024.02.0178

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Abstract:
Small pressurized water reactor nuclear power plant usually uses compact once-through steam generator, which has dramatically increased steam pressure under load rejection condition. In order to prevent damage to the system equipment caused by excessive steam pressure, based on the establishment of the mathematical model and the control system for the primary and secondary equipment of SPWR, this paper designs the steam dump control system based on pressure mode, temperature-pressure mode and power-pressure mode respectively, and carries out simulation research of SPWR system under load rejection condition. The results show that under load rejection condition, the temperature-pressure mode and power-pressure mode can effectively reduce the steam pressure overshoot compared with the pressure mode.

Design and Experiment of Feed-water Control System for CEFR Steam Generator

Liu Yong, Duan Tianying, Feng Weiwei, Zhang Weiying

2024, 45(2): 183-186. doi: 10.13832/j.jnpe.2024.02.0183

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Abstract:
The feed-water control system of steam generator is an important control system of China Experimental Fast Reactor (CEFR), which needs to be automatically controlled. According to the control requirements of the system, a set of feed-water control scheme is designed, and the control scheme is verified under the on-reactor conditions of CEFR. The results of the on-reactor experiment indicate that this control scheme can maintain the stability of the sodium temperature at the outlet of steam generator under various design conditions, and has the ability to suppress interference from other parameters of the reactor. Through the design and experiment of the feed-water control system of CEFR steam generator, the automatic control of the feed-water control system of CEFR steam generator has been achieved.

Research and Practice on Transformation Scheme of Hand Operator in Main Control Room of Nuclear Power Plant Based on Digital System

Xu Ying, Yang Wu, Liu Shengzhi, Jiang Xiaolong, Zhao Ke, Yang Jin

2024, 45(2): 187-192. doi: 10.13832/j.jnpe.2024.02.0187

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Abstract:
The instrument and control system of Daya Bay Nuclear Power Station is based on analog electronic technology, and it is planned to be digitized during the 30-year overhaul, during which the first-floor analog system will be completely digitized, but the man-machine interface equipment on the second floor will still be controlled by the hard hand operator in the main control room. According to the market research, four transformation schemes of hand operator based on digital system are preliminarily determined. Finally, the overall transformation scheme of developing new hand operator through standard I/O board interface is determined by considering many factors such as functional consistency, product reliability, technical feasibility, development cost, input/output (I/O) signal distribution efficiency and application feedback. This scheme can realize the optimal interface between the hand operator and the transfomed digital system, and effectively improve the utilization efficiency of I/O board channels. At the same time, through the rich software algorithm block function and digital system fault diagnosis function, the improvement measures such as automatic undisturbed switching of actuator and forced manual switching of process signal quality level are added, which effectively improves the reliability and stability of unit equipment control and provides an important reference scheme for similar digital transformation projects.

Flow Field and Blowdown Test of Secondary Side Lower Part of Steam Generator

Huang Jun, Ying Bingbin, Zheng Mingguang

2024, 45(2): 193-198. doi: 10.13832/j.jnpe.2024.02.0193

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Abstract:
The lower blowdown structure located in the tubesheet of steam generator (SG) in nuclear power plant can suck out the sludge on the tubesheet secondary side surface and discharge it. In order to reasonably design this structure and improve the blowdown efficiency, based on the SG prototype structure of passive large-scale advanced pressurized water reactor (CAP1400), a blowdown test piece was designed according to the ratio of 1:4 to simulate the tubesheet, heat transfer tube and other components at the lower part of SG. Computational fluid dynamics (CFD) calculation was carried out for the lower flow field, and the results were compared with those of the blowdown test. The fluid flow characteristics near the tube sheet surface were obtained. This test provides a basis for reducing sludge accumulation and improving the design by studying the fluid flow characteristics and sludge distribution law near SG tube plate, measuring the pressure drop of each part of the test piece and comparing the design of SG single-side and double-side blowdown structures. It is found that the performance of single/double-side blowdown structure is basically the same, and the single-side blowdown structure can effectively discharge the sludge particles in the test piece.

Research of Ultimate Load Characteristics of Hoisting Structure Carrier for Equipment Hatch

Hou Jiru, Liu Shifeng

2024, 45(2): 199-202. doi: 10.13832/j.jnpe.2024.02.0199

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Abstract:
Instability of the equipment hatch’s hoisting structure carrier occurs under the combined action of permanent and variable loads, and the accuracy of the ultimate load calculation results has an important influence on the design performance of equipment hatch. In engineering calculations, the traditional algorithm of directly solving eigenvalues without distinguishing load types resulted in overly conservative results. In this paper, on the basis of the original algorithm using ANSYS for buckling analysis, the loads that should not be amplified by the eigenvalues are distinguished. The eigenvalues are taken as the objective function, the reasonable ultimate loads are obtained through optimization iteration, and the correction coefficient is introduced in the iterative process to speed up the convergence. The accuracy of the improved engineering calculation method was verified through nonlinear analysis, and a suitable ultimate load algorithm for equipment hatch’s hoisting structure carrier was found for engineering calculations.

Primary Zinc Injection Technology Application and Zinc Injection and Dosing Equipment Design

Zou Wei, Lin Genxian, Sun Yun

2024, 45(2): 203-206. doi: 10.13832/j.jnpe.2024.02.0203

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Abstract:
The application of primary zinc injection technology makes the oxide film generated on the surface of the primary equipment more stable and plays a role in reducing the collective dose of the unit. By analyzing the effect of zinc injection on the unit, and combining the target concentration of zinc injection in a domestic PWR unit and the conditions of the on-site equipment, a set of zinc injection and dosing equipment is designed, and its design scheme, equipment component parameters and working flow are defined. The equipment is composed of a solution mixing unit, a filtration unit, a dosing unit and a control unit, which can be used to add zinc acetate during operation and hydrogen peroxide during overhaul.

Research on Historical Anomaly Data Detection Technology for Nuclear Power Plant Based on Deep Auto-Encoder

Yang Jihong, Chen Ling, Wang Xiaolong, Zhang Yongfa, Gao Ming

2024, 45(2): 207-213. doi: 10.13832/j.jnpe.2024.02.0207

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Abstract:
In order to solve the problem of new anomaly identification difficulties in the detection of nuclear power historical anomaly data, according to the idea of reconstruction error, an anomaly detection model based on deep auto-encoder is proposed. The model takes the normal historical data under steady-state operating condition as the learning object, trains itself by minimizing the reconstruction error of the normal data, and judges whether the unknown data is abnormal according to the size of the reconstruction error. The research results show that the deep autoencoder has better ability to reconstruct normal data but insufficient ability to reconstruct abnormal data. Thus, by comparing the reconstruction error size, the deep autoencoder can effectively detect the historical abnormal data of nuclear power plant, and its performance is better than that of one-class support vector machine, which can provide relevant basis for the state evaluation of nuclear power plants.

Design and Verification of Fuel Assembly Repair System

Li Chengye, Wang Wanjin, Wang Yajun, Liu Hao, Zheng Haichuan

2024, 45(2): 214-217. doi: 10.13832/j.jnpe.2024.02.0214

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Abstract:
In order to make full use of the post-irradiation fuel assembly, a fuel assembly repair process was proposed, and a repair system for post-irradiation fuel assembly was designed based on the dense and lightweight characteristics of the fuel assembly. Main parts were designed in detail and checked by seismic analysis using ANSYS. The results show that the stress and deformation of the main frame of the fuel assembly repair system meet the requirements. It is verified by the engineering test prototype that the system could dismantle and assemble the post-irradiation fuel assembly safely and reliably, and created conditions for the return of the fuel assembly.

Fault Risk Analysis Method of Instrument and Control System in Nuclear Power Plant

Luo Hui, Li Jianwei, Wang Xiangyu

2024, 45(2): 218-224. doi: 10.13832/j.jnpe.2024.02.0218

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Abstract:
In order to meet the requirements of rapid accurate fault location and comprehensive reliable risk analysis of instrument and control system in nuclear power plant, a fault risk analysis method of instrument and control system based on automatic analysis model is proposed. Firstly, the traditional system embedded self-diagnostic analysis, FMEA and automatic analysis of logic signals are integrated to build an automatic risk analysis model. Then, an automatic analysis fusion platform for fault risk is developed based on the automatic risk analysis model. Finally, the effectiveness of the proposed method is verified by example analysis. Practice shows that this method can effectively improve the efficiency of accurate fault location, reduce the fault risk analysis time, and reduce the risk of human error.

Numerical Simulation of Displacement Process of Liquid Scintillator and Ultrapure Water in a Large Diameter Spherical Tank

Zhao Hongwei, Dong Rui, Li Junjie, Ouyang Zhengrong, Heng Yuekun

2024, 45(2): 225-230. doi: 10.13832/j.jnpe.2024.02.0225

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Abstract:
On the one hand, the Neutrino experimental facility needs to shield a lot of cosmic rays and natural radioactivity from rocks, air and dust. On the other hand, it is necessary to reduce the radioactive background of the experimental facility as much as possible. In this paper, the displacement process of liquid scinulator (liquid flash) and ultrapure water in the central detector (large-diameter spherical tank) is studied by CFD simulation. The situation of radioactive impurities in the water entering the liquid scinulator is analyzed, and the effect of natural convection on the liquid flash background during the displacement is studied. By simulating the two conditions of "no natural convection" and "with natural convection", the phase and flow distribution under the two different conditions was obtained, and the changes in the phase content with time are also obtained. The vertical upward velocity at the phase interface was extracted, and the total upward flow of the section was calculated to judge the influence of natural convection on the displacement process. The results show that the upward average volume flow of the liquid scintillator in the natural convection is less than that under the condition of no natural convection, which shows that natural convection can inhibit radioactive impurities from entering the liquid scintillator.

Research on Airtightness Test of Habitable Zone in Main Control Room of HPR1000

Xu Funan, Wang Guangjun, Wu Ying

2024, 45(2): 231-234. doi: 10.13832/j.jnpe.2024.02.0231

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Abstract:
Based on the constant flow method, research was conducted on the airtightness test of the habitable area in the main control room from multiple dimensions such as test method selection, tracer gas injection, sampling method, equilibrium concentration, and calculation of test results. Improvements and applications were made to the K2/K3 unit of the Karachi nuclear power project. The results showed that by studying the factors affecting the uniformity of tracer gas mixing, the relationship between positive pressure and internal leakage was determined, and the ideal positive pressure value for conducting the experiment was approximately 50 Pa; By studying the relationship between positive pressure and air volume difference, it was found that the ideal positive pressure range for conducting the experiment is 35~50 Pa; By analyzing the relationship between positive pressure and internal leakage, the ideal positive pressure value for conducting the experiment was obtained; On this basis, through comparative analysis of tracer gas concentration data, the fluctuation amplitude and pattern of concentration were analyzed, and effective measures to promote tracer balance were proposed. That is, adding mixed flow equipment can make the tracer gas concentration fully uniform and reduce the fluctuation amplitude of concentration.

Application of High Precision Laser Measurement Technology Based on Spherical Coordinate Method in Nuclear Power Plant Reactor Body Maintenance

Wang Xuefang, Lu Shaowei

2024, 45(2): 235-240. doi: 10.13832/j.jnpe.2024.02.0235

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Abstract:
The maintenance of reactor body is an important work that must be done in refueling overhaul of nuclear power units. Due to the complex structure, large size, tight maintenance period and high radiation dose rate of the equipment in the reactor body, it is very difficult to measure and inspect defects and install and locate underwater maintenance tools, which cannot be measured and inspected with conventional techniques. In this study, advanced high-precision laser measurement technology based on spherical coordinate method combined with flexible tooling can realize accurate underwater long-distance measurement in large space. The trend of the measured data can be analyzed, and the reasons for the deviation are obtained, which can provide reliable data support for the overhaul of the unit, thus ensuring the smooth execution of underwater maintenance work in high radiation environment and ensuring the safe, reliable and stable operation of the nuclear power plant.

Analysis on Water Ingress Accident of a Gas Cooled Reactor

Ma Yugao, Cao Zhongbin, Wang Jinyu, Deng Jian, Bao Hui, Ding Shuhua, Cheng Kun, Hu Wenzhen

2024, 45(2): 241-247. doi: 10.13832/j.jnpe.2024.02.0241

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Abstract:
The gas cooled reactor is featured with high inherent safety, small size, and a simple start-up process. However, the water ingress accident might occur due to the influence of the working environment or operating status. Based on the design scheme of the Submersion-Subcritical Safe Space(S⁴) reactor, this work simulated and analyzed the water ingress accident caused by the rupture of the heat transfer tube of the condenser under normal operating conditions, and studied the accident consequences such as the introduction of positive reactivity, the overpressure of the Brayton cycle. This work calculated the influence of spectral shift materials on reactivity introduction during the water ingress process with the Reactor Monte Carlo code RMC. And the temperature and Brayton cycle pressure were calculated during the water ingress process with the gas cooled reactor transient analysis code, HXRTRAN. The results show that when a water ingress accident occurs, 0.5 kg water ingress causes the pressure of the Brayton cycle to exceed 10 MPa, which may lead to larger damage to the condenser pipeline and secondary seawater injection. Meanwhile, water ingress may lead to a large amount of positive reactivity introduction. If spectral shift absorbers, Ir, are added to the fuel surface in the reactor, the core may reduce power or even subcritical shutdown spontaneously in the water ingress accident. When the amount of water vapor exceeds 5 kg, the core power quickly decreases to about 2.2% FP and gradually approaches shutdown. Therefore, the spectral shift materials, Ir, have a significant inhibition effect on the introduction of positive reactivity caused by water ingress of the core.

Development and Validation of DNBR On-line Monitoring System for HPR1000 Reactor

Chen Xi, Wu Qing, Deng Jian, Liu Yu, Ren Chunming, Wang Xiaoyu, Peng Huanhuan

2024, 45(2): 248-253. doi: 10.13832/j.jnpe.2024.02.0248

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Abstract:
In the traditional protection system of generation II reactor, over-temperature protection signal is usually used. This protection method is conservative and indirect, however, the Departure from Nucleate Boiling Ratio (DNBR) on-line monitoring system can directly monitor the changes of safety parameters. In order to directly monitor the safety margin of Hualong Pressurized Reactor (HPR1000) and improve the flexibility of reactor safety operation, this paper puts forward a DNBR on-line monitoring calculation model with both calculation speed and solution accuracy based on the basic model of thermal design of PWR core, and develops a set of DNBR on-line monitoring system for HPR1000. The calculation model of on-line monitoring system is fully verified from three aspects: independent verification of core, verification of analog signal and verification of actual operation data of HPR1000. The results demonstrate that the precision of the on-line monitoring system is very high, and equvalent with the sub-channel code. In conclusion, the DNBR on-line monitoring system could satisfy the needs of HPR1000 engineering design.

Current Issue 2024, Volume 45, Issue 2

Current Issue
2024, Volume 45, Issue 2