核动力工程

Overall Study of Dock-based Floating Nuclear Power Plant

Wang Donghui, Li Qing, Song Danrong, Qin Dong, Liu Jia

2023, 44(4): 1-8. doi: 10.13832/j.jnpe.2023.04.0001

Abstract(2223) HTML (250) PDF(262)

Abstract:
Considering the domestic and overseas development trend of floating nuclear power plant (FNPP), a dock-based ACP100S FNPP is proposed in this paper to promote the construction of FNPP in China. A preliminary evaluation is conducted in the external event, nuclear reactor design, hull design, economic analysis, water-electric-gas cogeneration, emergency safety, and plant deployment. A two-step construction mode is suggested from “offshore FNPP” to “open sea FNPP”. The dock-based ACP100S FNPP provides a meaningful solution for the early engineering construction in our country.

High-fidelity Full Core Neutronics Calculation Method for Fuel Assembly Bowing and Its Application

Li Fan, Liu Zhouyu, Wang Xining, Cao Liangzhi, Wu Hongchun

2023, 44(4): 9-16. doi: 10.13832/j.jnpe.2023.04.0009

Abstract(393) HTML (28) PDF(51)

Abstract:
In order to study the influence of bowing deformation of fuel assemblies on core power distribution, this paper proposes an equivalent method for simulating the bowing of fuel assemblies in PWR Core. That is, according to the principle of conservation of atomic number of the water gap material around the fuel assembly before and after bowing, the nuclear density of all nuclides in the water gap after bowing is changed by keeping the width of the water gap unchanged before and after bowing, which approximates the change of the water gap around the equivalent fuel assembly after bowing. Its correctness is verified by Monte-Carlo code NECP-MCX and deterministic numerical code NECP-X. Based on the deterministic numerical code NECP-X, the fuel assembly bowing condition of the EPR core is simulated and analyzed. The calculation results show that due to the change of local moderating effect, the small bending of fuel assembly has a relatively great influence on the core power distribution. When the maximum offset in the whole core problem is about 2mm, the relative change of the assembly power can reach about 5%.

Selection of Shadow Shielding Structure and Material for Space Reactor

Huang Qianming, Li Lan, Chai Xiaoming, Liu Bin, Ying Dongchuan

2023, 44(4): 17-24. doi: 10.13832/j.jnpe.2023.04.0017

Abstract(267) HTML (44) PDF(77)

Abstract:
The space reactor has strict requirements on the size and weight of radiation shielding. In order to find out a suitable shielding scheme, it is necessary to select the shielding material and structure. In this paper, firstly, the research progress of shielding targets and limits of space reactors at home and abroad is introduced. Based on the design principle of reactor shielding, flat model and spherical model are abstracted for different application scenarios, and the shielding performance of different materials is analyzed under different design objectives. Based on the analysis results, automatic optimization tools are used to select shielding schemes, and the advantages and disadvantages of each scheme are analyzed. The results show that the energy spectrum of the source, the size of the source, the distance between the shield and the source, and different shielding design objectives will affect the selection of shielding materials and structures, which needs to be carried out according to the application requirements. Boron carbide, lithium hydride and tungsten are good shielding materials for space reactors, and the use of automatic optimization tools for layered layout of the shield can achieve effective weight reduction.

Research and Application of Single-Point Calibration Method for Ex-core Nuclear Instrument System of PWRs

Bai Jiahe, Yang Haozhe, Wan Chenghui, Pan Zefei, Li Zaipeng, Li Yunzhao, Wu Hongchun

2023, 44(4): 25-32. doi: 10.13832/j.jnpe.2023.04.0025

Abstract(190) HTML (56) PDF(45)

Abstract:
The ex-core detector is used to indicate the reactor power and axial-power difference in commercial PWRs, and it needs to be calibrated regularly to ensure the accuracy. The conventional multi-point calibration method needs to move the control rods and measure the flux-mapping several times, which influences the economy and safety of power plants. Therefore, the single-point calibration method for the ex-core nuclear instrument system has been proposed in this paper, which combines both the theoretical calculation and measurement data. The axial offset correction factor and the ex-core detector sensitivity are introduced in this method. And the core-analysis code is utilized to simulate the movement of control rods. Eventually, the calibration coefficients could be determined. Based on the single-point calibration method, the function development has been achieved in the core-analysis code SPARK, and it has been verified by the multiple cycles’ measurement data from M310 reactor. The verification results indicate that the single-point calibration method proposed in this research has higher calibration accuracy. Compared with the measured power level and axial power difference, the average biases of calibration results are 0.31% and 0.16%, repectively. Therefore, the single-point calibration method is capable to determine the calibration coefficients accurately, and save calibration time without control rod movement. This method could improve the economy and safety of power plants and has the engineering application and popularization value.

Research on Direct Transport Calculation Method Based on Numerical Nuclear Reactor Physics Code SHARK

Zhao Chen, Zhao Wenbo, Zhang Hongbo, Wang Bo, Chen Zhang, Peng Xingjie, Gong Zhaohu, Zeng Wei, Li Qing

2023, 44(4): 33-40. doi: 10.13832/j.jnpe.2023.04.0033

Abstract(1274) HTML (42) PDF(107)

Abstract:
In order to establish the next-generation reactor physics calculation method based on the numerical nuclear reactor technology and realize high-fidelity modeling, high-resolution and high-precision calculation, the research of direct transport method was conducted based on the numerical nuclear reactor physics code SHARK, and 2D/1D and quasi-3D MOC methods were built and compared. Based on the large-scale parallel acceleration technology of domain decomposition and coarse mesh finite difference (CMFD), the whole reactor direct transport calculation was realized for the pin-type and plate-type cores. Compared with Monte-Carlo reference results, eigenvalue differences were less than 100pcm (1pcm=10^–5) and maximum pin/plate powers were less than 3%. Numerical results showed the good accuracy of SHARK in the direct transport calculation, and can be apllied in multi-application scenarios of pin-type and plate-type cores.

Research on Simulation of Neutron Transport with Thick Diffusion Limit in Curved Meshes

Wang Xinyu, Zhang Bin, Chen Yixue

2023, 44(4): 41-48. doi: 10.13832/j.jnpe.2023.04.0041

Abstract(145) HTML (23) PDF(26)

Abstract:
Discrete ordinate method is one of the main numerical methods for solving the problem of neutron transport with thick diffusion limit. Its commonly used spatial discrete schemes, such as finite difference scheme, are easy to cause numerical diffusion in optical thick media, and the application of discrete ordinate method has certain limitations due to the lack of coarse mesh accuracy and difficulty in adapting to complex geometry. In this paper, Galerkin method is used to derive discrete ordinate equations in weak form or variational form. Based on the idea of discontinuous finite element, the Lagrangian finite element basis function in the higher-order curved meshes is constructed to establish the higher-order finite element discrete scheme of the transport equation. Two manufactured solution examples, the IAEA EIR-2 benchmark problem and the thick diffusion limit example, are selected for modeling and transport calculation, and the calculation accuracy and convergence of the spatial discrete scheme are tested and verified, and its thick diffusion limit characteristics are analyzed. The numerical results show that the relative error between the calculated results of high-order discontinuous finite element scheme and the reference value is less than 1%, and it also has high calculation accuracy and convergence rate in curved meshes. The discrete scheme can effectively solve the problem of neutron transport with thick diffusion limit in curved meshes. It has good numerical characteristics, and has the asymptotic preservation property of diffusion limit under optical thick diffusion limit.

Research on Automatic Modeling Method of TORT Program Based on CAD Model

Xu Fangyuan, Yang Chao, Yu Tao, Chen Zhenping, Huang Guocai, Li Leiming, Li Yukun, Xian Xirui, Du Hua

2023, 44(4): 49-54. doi: 10.13832/j.jnpe.2023.04.0049

Abstract(219) HTML (55) PDF(34)

Abstract:
Aiming at the problems of complex geometry of reactor shielding structure, limited geometric processing ability, low efficiency and error prone of traditional manual modeling, based on the multi-objective modeling and simulation platform for radiation transport (MOSRT), the volume weight homogenization method of discrete mesh materials is used to homogenize the discrete meshes of multi-materials, and the fine transformation from CAD model to three-dimensional discrete ordinate method (S_N) calculation program TORT shielding calculation model is realized. Based on the Kobayashi and NUREG-CR-6115 benchmark models, the automatic modeling method is verified. The results show that for the Kobayashi benchmark problem, the automatic modeling method is completely consistent with the manual modeling results. For the NUREG-CR-6115 benchmark problem, the maximum error between the automatic modeling method and the reference solution is 12.2%. The verification results show the effectiveness and correctness of the automatic modeling method.

Numerical Study on Flow and Heat Transfer of High-pressure Sub-cooled Water Injection into High-temperature Lead-bismuth Alloy under Lead-bismuth Cooled Fast Reactor SGTR Accident

Liu Li, Yuan Junjie, Gu Hanyang, Bao Ruiqi, Liu Maolong, Wang Ke

2023, 44(4): 55-64. doi: 10.13832/j.jnpe.2023.04.0055

Abstract(425) HTML (88) PDF(94)

Abstract:
There are high-pressure sub-cooled water and high-temperature lead-bismuth coolant on both sides of the heat transfer tube of the steam generator in the lead-bismuth fast reactor. The large pressure difference and temperature difference on both sides of the heat transfer tube and the corrosion effect of Lead-bismuth eutectic (LBE) may cause the steam generator heat transfer tube rupture (SGTR) accident. It is of great academic significance and engineering application value to deeply study the characteristics of jet boiling and phase change product steam diffusion of high-pressure sub-cooled water impacting LBE after the accident. In order to reveal the heat and mass transfer mechanism of the interaction between LBE and water under accident conditions, this paper establishes a three-dimensional numerical calculation model of water/steam-liquid lead-bismuth multiphase flow and heat transfer based on the volume of fluid (VOF) method, combined with LES turbulence model and Lee phase change model. The phase change and heat transfer process occurred during the high-pressure sub-cooled water injection into the high-temperature LBE is systematically studied. Combined with the factors such as injection pressure and sub-cooled water temperature, the effects of different conditions on the jet shape, migration depth and boiling behavior during the jet boiling process are analyzed. The research results can provide guidance for the prediction of core safety under SGTR accident conditions.

Derivation and Evaluation of Carbon Dioxide Partial Derivative Property by Implicit Solution of Brayton Cycle System

Wen Shuang, Wen Qinglong, Hu Wenjun, Xu Shijia

2023, 44(4): 65-71. doi: 10.13832/j.jnpe.2023.04.0065

Abstract(371) HTML (52) PDF(24)

Abstract:
To improve the accuracy of solving the Brayton Cycle equation of supercritical carbon dioxide (S-CO₂), the fully implicit or semi-implicit difference scheme is used to solve the fluid conservation equation discretely, and the partial derivative property is indispensable for the implicit solution. This study will evaluate the accuracy of the most typical state equation of carbon dioxide gas in the full parameter range. On this basis, the Maxwell equation is used to derive the partial derivative correlation of carbon dioxide gas. The results show that:①the SW equation has the highest accuracy in subcritical and supercritical regions, and the error is kept within 3%; ②based on the SW equation and Maxwell equation, the correlations of (∂h/∂ρ)_p and (∂h/∂p)_ρ of carbon dioxide gas at 216~1100 K and 0 ~800 MPa are derived; ③ the error of most data points of (∂h/∂ρ)_p and (∂h/∂p)_ρ remains within ±0.01%, and the error increases slightly near the critical point. The maximum error of (∂h/∂ρ)_p is 0.373%, and the maximum error of (∂h/∂p)_ρ is −0.798%.

Development of Dimensionless Rod-bundle CHF Correlation Based on Stepwise Regression and Determination of DNBR Limit

Yin Yuan, Feng Simin, Pang Bo, Xi Yanyan, Zhang Yuxiang, Fu Xiangang

2023, 44(4): 72-78. doi: 10.13832/j.jnpe.2023.04.0072

Abstract(1038) HTML (72) PDF(41)

Abstract:
At present, the empirical correlations of critical heat flux (CHF) of advanced PWR rod-bundles at home and abroad generally have the common problems of complex mathematical form, numerous independent variable coefficients and lack of physical significance. In this study, based on 485 rod-bundle CHF data points of 5×5 PWR rod-bundles selected from the rod bundle CHF database of American Electric Power Research Institute (EPRI), a new dimensionless CHF correlation is developed with stepwise regression analysis. Considering the cold wall effect and axial non-uniform heating effect of the guide tube, the average value of the ratio M/P between the measured CHF and the predicted CHF is 0.998, the root mean square error is 0.0546, and the standard deviation is 0.0546. Based on the grouping method, the limit of the 95/95 departure from nucleate boiling ratio (DNBR) of the correlation is determined to be 1.16.

Study on the Effect of Inclination Angle on the Natural Convection of Molten Salt in Heat Pipe Cooled Molten Salt Reactor Core

Chen Zehan, Chen Xingwei, Dai Ye, Zou Yang

2023, 44(4): 79-87. doi: 10.13832/j.jnpe.2023.04.0079

Abstract(255) HTML (55) PDF(52)

Abstract:
The inclination angle of heat pipe-cooled molten salt reactor (MSR) core has an important influence on the core temperature distribution and local hot spots. In order to obtain the natural convection heat transfer characteristics of molten salt in the core at different inclination angles, optimize the core design and improve the system safety, the three-dimensional modeling of the core is carried out, and the temperature field and flow field of the natural convection of molten salt in the core are analyzed for both horizontal and vertical placement by numerical simulation with the software Fluent. At the same time, the influence of core inclination angle change on the core temperature field and local hot spots is discussed. The results show that the local hot spots always appear in the upper part of the core, and the temperature field and flow field of the core are more unstable when the core is placed horizontally than vertically. When the inclination angle is in the range of 5°-10°, the local hot spot temperature is the highest, and the hot spot temperature is the lowest when the core is vertical. The simulation results show the natural convection characteristics of molten salt in the core, and provide a reference for the thermal design of the heat pipe-cooled MSR.

Effect of the Arrangement for Grid Spacers with Mixing Vanes on the Thermal Hydraulic Characteristics of a 5×5 Rod Bundle via CFD Analysis

Su Qianhua, Fan Guanhua, Lyu Lulu, Lu Donghua, Yang Ping, Gan Fujun, Yan Binghuo, Wang Chengyue

2023, 44(4): 88-94. doi: 10.13832/j.jnpe.2023.04.0088

Abstract(240) HTML (61) PDF(59)

Abstract:
In order to optimize the arrangement of the grid spacer for a rod bundle, the computational fluid dynamics (CFD) method is employed to analyze the flow and temperature fields for a 5×5 rod bundle equipped with three grid spacers. The effect of the axial separation between two neighboring grid spacers and the relative rotation angle of the grid spacer is investigated, and the reasons for different results are analyzed from the perspective of convective heat transfer. It is found that the variation of axial separation brings negligible change on the drag and heat-transfer characteristics of the rod bundle. However, rotating the middle grid by 90° relative to the grids at both ends in the cross section can effectively reduce the non-uniformity of temperature distribution in the cross section and reduce the maximum temperature in the cross section.

Research on Fatigue Performance of Newly Developed Supporting Structure of Spacer Grid

Guo Xiaoming, Ren Quanyao, Chen Jie, Ren Yi

2023, 44(4): 95-99. doi: 10.13832/j.jnpe.2023.04.0095

Abstract(217) HTML (76) PDF(60)

Abstract:
Aiming at the innovative supporting structure and its fatigue performance, this paper evaluates the stress state under its operating conditions by means of finite element analysis, and carries out the spring fatigue test to analyze the fatigue failure characteristics and crack initiation locations. The results show that the spring fatigue failure of the supporting structure roughly goes through the stages of load reduction and rapid failure. Due to the support legs of the spring, the spring still maintains a certain load after failure, which is about 0.42 times of the nominal load. The fatigue and crack locations are not only determined by the stress value, but also by the transferring feature of load. The finite element model is an effective method for the design and optimization of the supporting structure of spacer grid, and can preliminarily identify the location of fatigue failure.

Study on Refined Calculation Method for Added Mass of Special-shaped Structures in Complex Fluid Domain

Tan Ximing, Gao Fuhai, Qi Min, Wang Yueying, Liu Zhaoyang

2023, 44(4): 100-106. doi: 10.13832/j.jnpe.2023.04.0100

Abstract(155) HTML (61) PDF(30)

Abstract:
The engineering design of reactor internals usually uses the added mass method to simulate the dynamic effects of fluids on the structure. Taking a special-shaped pressure pipe immersed in a complex fluid domain as an example, a structural added mass iterative calculation method is proposed, which uses the wet mode obtained from the acoustic-structure coupling analysis as the benchmark, comprehensively considers the fluid density and bulk modulus to determine the structural added mass, and discusses the applicability of this method. The accuracy of the added mass calculation has been demonstrated by comparing the measured and simulated results of the 1∶1 vibration test of the pressure pipe. The results show that the method proposed in this paper can accurately obtain the main frequency and vibration mode of the structure, and has a fast convergence speed, which can be used as a reference for similar engineering simulation analysis.

Experimental Study on Mechanical Performance of Nuclear Containment Truncated Cone Region

Lan Tianyun, Wu Yuzheng, Xiao Dan, Zhou Chuanbo, Dong Zhanfa, Guo Junying, Xiong Meng

2023, 44(4): 107-115. doi: 10.13832/j.jnpe.2023.04.0107

Abstract(105) HTML (36) PDF(23)

Abstract:
The truncated cone area of nuclear containment (the junction of shell and raft foundation) has irregular shape and complex stress. Studying the stress mechanism of this area is important to master the structural performance of the whole containment. In this study, the truncated cone area was taken as the object of study. On the basis of the scale model of truncated cone area obtained by ABAQUS, two specimens with scale ratio of 1∶3 were made to carry out static tests under two load conditions: with and without internal pressure. The test results show that the failure mode of specimen 1 (without internal pressure) is bending shear failure with horizontal displacement of 27.7mm, while the failure mode of specimen 2 (with internal pressure) is shear failure with horizontal displacement of 32.6mm. The main crack zone of the two specimens is the junction of the inner bottom plate and the truncated cone. The yield position of the reinforcement mainly appears at the upper transverse reinforcement, while the vertical reinforcement on the outside of the cone is not fully functional. The results can guide the optimization of structural reinforcement.

Multi-objective Optimization Design of the Support in Flow Induced Vibration Test of Reactor Internals

Zhang Yu, Li Pengzhou, Qiao Hongwei, Miao Yuhan, Gao Lixia, Yu Danping, Sun Lei

2023, 44(4): 116-120. doi: 10.13832/j.jnpe.2023.04.0116

Abstract(190) HTML (63) PDF(29)

Abstract:
In order to fully improve the material utilization rate and isolate the fundamental frequency of the reactor internals, a parameter optimization study based on multi-objective optimization algorithm was carried out for the support structure in the flow induced vibration (FIV) test. The finite element model of support was established first, and the support quality, maximum stress intensity and fundamental frequency were regarded as the optimization goals. Then, the strength pareto evolutionary algorithm (SPEA-Ⅱ) was utilized to obtain global optimal solutions. Finally, the global sensitivity between the input and output parameters and pareto solution set were obtained, and a feasible support design scheme was given. By carrying out multi-objective optimization for the support, the manufacturing cost was significantly reduced while the structural strength was guaranteed. The related work can provide reference for similar support design.

Analysis of Thermal Stress and Fatigue Induced by Dryout Oscillation in Once Through Steam Generator

Chen Ling, Wang Xinming, Zhang Yongfa, Zhang Liming, Jiang Lizhi, Jiao Meng, Liu Xiaoya

2023, 44(4): 121-127. doi: 10.13832/j.jnpe.2023.04.0121

Abstract(128) HTML (26) PDF(24)

Abstract:
In order to study the damage of the heat transfer tube caused by dryout oscillation, the once through steam generator designed by Babcock&Wilcox company is taken as a prototype. Firstly, the relevant thermal and hydraulic parameters are obtained by using the method of primary and secondary side coupled heat transfer. By comparing the radial temperature distribution of dryout point at different fluctuation frequencies, the influence of fluctuation frequency was determined, and the stress distribution of heat transfer tubes was obtained by finite element analysis. Finally, the fatigue of heat transfer tubes was evaluated according to S-N curve, and the influence of related factors was discussed. The results show that when the fluctuation frequency of the drying point is low, the radial temperature distribution is similar to the steady state. The outer wall in contact with the secondary side is more prone to fatigue damage, although the alternating stress is less than the limit value, there is a certain operating risk in the reactor environment. The increase of temperature fluctuation frequency will lead to the obvious decrease of the life of heat transfer tubes, and the use of elastic constraint is helpful to alleviate the fatigue caused by the oscillation of steam drying point. This study provides a reference for the life prediction and safe operation of the heat transfer tubes of once-through steam generator under the condition of fluctuating evaporating point.

Reaction Thermodynamics and Kinetics Analysis of Dry Extraction of Ba¹⁴CO₃ by Irradiated AlN

Cao Qi, Chen Yunming, Zhang Jingsong, Luo Ning, Dai Shuang, Lu Yunyun, Yang Yu

2023, 44(4): 128-132. doi: 10.13832/j.jnpe.2023.04.0128

Abstract(142) HTML (55) PDF(15)

Abstract:
The reaction thermodynamics, kinetics analysis and demonstration test of a high temperature oxidation reaction under complex irradiated AlN system was carried out by thermodynamic calculation software (HSC). The results show that it is necessary to ensure sufficient O₂ during the reaction to improve the yield of ¹⁴CO₂. The reaction rate is controlled by the diffusion of O₂ adsorption and ¹⁴CO₂desorption. To improve the reaction rate, the process should be improved from the aspects of reactor selection, contact mode, contact time and reaction temperature of gas-solid reactants. It provides important technical support for batch preparation of high specific activity Ba¹⁴CO₃.

Probabilistic Safety Analysis Framework for Internal Events of Aqueous Homogeneous Reactor

Wang Zhe, Zhang Dan, Zou Zhiqiang, Wang Ningning, Yang Weidong, Du Yu

2023, 44(4): 133-137. doi: 10.13832/j.jnpe.2023.04.0133

Abstract(153) HTML (26) PDF(33)

Abstract:
There are significant differences in the safety design and operation characteristics between liquid fuel reactor (aqueous homogeneous reactor) and traditional solid fuel reactor, therefore, it is impossible to carry out the safety design of aqueous homogeneous reactor only using the existing safety design methods based on deterministic theories, and probabilistic safety analysis (PSA) technique must be adopted at the beginning of the design period. Due to the differences in fuel forms, safety barriers and mitigation systems, the traditional PSA technique for reactor, which is based on the core damage, cannot be directly applied to aqueous homogeneous reactors. After investigating the requirements and analysis of traditional research reactors, aqueous homogeneous reactors and spent fuel reprocessing facilities at home and abroad, taking the medical isotope test reactor being developed in China as the object, the PSA safety objectives of aqueous homogeneous reactor is put forward, and the PSA technical framework is established, which lays the foundation for the PSA development and safety review of this type of reactor.

Study on Control Strategy of Natural Circulation Lead-cooled Fast Reactor Coupled with S-CO₂ Brayton Cycle

Liu Guixiu, Yi Jingwei, Li Gen, Liang Tiebo, Fang Huawei, Chen Weixiong

2023, 44(4): 138-147. doi: 10.13832/j.jnpe.2023.04.0138

Abstract(281) HTML (77) PDF(59)

Abstract:
The coupled power generation system of natural circulation lead cooled fast reactor with supercritical carbon dioxide (S-CO₂) Brayton cycle is the development trend of advanced nuclear energy systems in the future. Based on the software Apros, a dynamic model of the coupled power generation system was built, and two reactor control schemes were designed, one of which was the conventional control scheme of the reference core power control system of pressurized water reactor, and the other was the compensation control scheme with rod position limit of control rods. The research results showed that under a small variable load rate of 3% FP/min (FP is short for full power), the dynamic deviation of load following under both control schemes was between −2% and 1%, however, for the stability of core outlet coolant temperature, the compensation control scheme was superior to the conventional control scheme; at a large variable load rate of 6%FP/min-18%FP/min, the variation range of core outlet temperature under conventional control was −40℃-0℃, while the variation range of the core outlet temperature under compensation control was −5℃-2℃. Therefore, the compensation control scheme can be used as an effective means for the control of natural circulation lead cooled fast reactor.

Uncertain Information Modeling and Processing in SPAR-H Method under Group Decision Making

Guan Xuefan, Wang Danyu, Su Xiaoyan, Xu Zhihui, Qian Hong

2023, 44(4): 148-153. doi: 10.13832/j.jnpe.2023.04.0148

Abstract(176) HTML (51) PDF(21)

Abstract:
Considering the uncertain information in Standardized Plant Analysis of Risk-Human Reliability Analysis (SPAR-H) method and the inability to handle the assessment with multiple experts, this paper proposes a method of uncertain information modeling and processing in SPAR-H method under group decision making environment based on analytic hierarchy process (AHP) and Dempster-Shafer (D-S) evidence theory. Firstly, the weight of each expert is calculated based on AHP. Then, based on D-S evidence theory, uncertain information in expert opinions is described and basic belief assignment (BBA) is generated. Next, the weighted BBA is determined by considering experts’ weights. The combined BBA of evaluations is fused through Dempster combination rule, and the values of eight performance shaping factors (PSF) are generated. Finally, the final human error probability (HEP) is acquired based on the SPAR-H method. In this paper, the effectiveness of the proposed method is illustrated by taking a digital nuclear power plant under shutdown and low power conditions as an example.

Research on Cascade Control Method of Electric Power of NUSTER-100

Pu Songmao, Huang Jiajun, Sun Peiwei, Wei Xinyu

2023, 44(4): 154-162. doi: 10.13832/j.jnpe.2023.04.0154

Abstract(123) HTML (59) PDF(31)

Abstract:
The heat pipe cooled reactor (hereinafter referred to as heat pipe reactor) has the design concept of solid-state reactor, and the heat is passively transferred out of the core through heat pipes. It has the advantages of simple structure, high safety, low noise, compact structure and long working time. In this paper, a 100 kW silent heat pipe reactor (NUSTER-100) is taken as the research object, and the nonlinear dynamic model is built based on MATLAB/Simulink platform. The transfer function model is obtained by linearization based on perturbation theory. Based on the analysis of dynamic characteristics, a cascade control method of electric power is proposed, in which the inner loop is the core power regulation and the outer loop is the electric power control. Based on the transfer function between reactivity and nuclear power, and the transfer function between nuclear power and electric power, an electric power cascade control system is designed, and its nonlinear problem is solved by gain scheduling. The simulation results show that the cascade control system can meet the requirements of control performance and realize the safe and reliable operation of nuclear reactor.

An Approach for Dynamic Reliability Assessment of Reactor Trip System of HPR1000

Li Kunxiang, Sui Yang, Dai Tao, Yu Tao

2023, 44(4): 163-169. doi: 10.13832/j.jnpe.2023.04.0163

Abstract(171) HTML (29) PDF(40)

Abstract:
The structure of reactor trip system (RTS) is complex, which leads to its dynamic interaction, time dependence and probability uncertainty. However, the traditional static reliability assessment methods are difficult to characterize these three characteristics. To solve this problem, a novel approach for dynamic reliability assessment of the RTS of Hualong Pressurized Reactor 1000 (HPR1000) was proposed. Firstly, the dynamic fault tree (DFT) was used to establish the DFT model for the characterization of RTS dynamic interaction. Then, dynamic Bayesian network (DBN) and fuzzy set theory (FST) were used to establish fuzzy DBN model based on the established DFT model for the characterization of dynamic interaction, time dependence and probability uncertainty of the RTS. Finally, Latin hypercube sampling (LHS) was used to define a new fuzzy Bayesian inference algorithm. The defined algorithm was used to carry out the fuzzy Bayesian forward inference and backward inference, the dynamic reliability of RTS was calculated, and the weak points of RTS were identified. Comparing the defined fuzzy Bayesian inference algorithm with the traditional fuzzy Bayesian inference algorithm, the accuracy and precision of the algorithm defined in this paper were verified. The obtained research results provide a scientific basis for further improving the reliability of the RTS of HPR1000.

Optimization of Feedwater Control for Casing Steam Generator Based on Apros

Liu Haipeng, Wang Changshuo, Ye Zhu, Tian Peiyu

2023, 44(4): 170-178. doi: 10.13832/j.jnpe.2023.04.0170

Abstract(446) HTML (62) PDF(36)

Abstract:
In allusion to the problem of water supply control caused by the strong coupling of casing steam generator, the steam-water circulation system of commercial modular small reactor with casing steam generator is taken as the research object, and the simulation model of steam-water circulation system is established based on the software APROS. The steady-state simulation results show that the simulation model has high simulation accuracy and meets the requirements of simulation analysis. Through the transient simulation test of power load increasing and decreasing, the transient operation characteristics of casing steam generator are studied. The research results show that steam and feedwater flows have good load following characteristics when the traditional control scheme is adopted, but relatively large steam pressure fluctuation can be observed, and the steam dump will be triggered when 80%FP drops to 50%FP. To solve this problem, an optimization scheme of feed water control is proposed. The simulation analysis results show that the fluctuation range of steam pressure is obviously reduced after optimization, and the steam dump action is not triggered, and the safety and stability of the system are effectively improved.

Research on Model-driven Top-level Architecture Design Method of Nuclear Engineering

Pan Xinxin, Zhuang Yaping, Song Chunjing, Lin Chao

2023, 44(4): 179-184. doi: 10.13832/j.jnpe.2023.04.0179

Abstract(839) HTML (59) PDF(65)

Abstract:
An iterative optimization modeling process of nuclear engineering top-level functional architecture and logical architecture based on MagicGrid methodology is proposed. Starting from normal conditions, functional analysis and logic construction are gradually carried out. Based on the fault mode of logical structure, abnormal conditions are derived, and the next design iteration is started driven by the abnormal conditions. After covering all the conditions, the functional architecture and logical architecture of nuclear engineering are formed. The feasibility of the method is verified in a heating reactor project, and it can be used to guide the modeling of Model-based System Engineering (MBSE) for the top-level function of nuclear engineering.

Research on Ultrasonic Technology of Accurate Measurement of Main Feedwater Flow in Nuclear Power Plant

Bai Tian, Wang Fengning, Liu Yan, Wei Huatong, Cui Xiwei, Guo Lin, Liu Hai, Liu Li

2023, 44(4): 185-191. doi: 10.13832/j.jnpe.2023.04.0185

Abstract(189) HTML (22) PDF(45)

Abstract:
A high-precision ultrasonic flow meter with a design measurement uncertainty of 0.3% has been investigated and developed for the measurement of the main feedwater flow in the secondary loop. This flowmeter could be used to realize the small power enhancement of PWR nuclear power units. The influences of flow measurement under high temperature and high pressure were sequentially analyzed through the method of uncertainty decomposition. The influence of the flow condition variation caused by high temperature on the measurement can be partially verified by the real flow calibration of the calibration coefficient under normal temperature and pressure and medium temperature. The change of the tube structure size could be obtained by finite element simulation and measurement on a static experimental equipment which could create a high temperature and high pressure environment. The error introduced by the flow integral algorithm could be obtained by CFD simulation and verified by real flow calibration. The results show that each component can be quantified and traced back by using the flow standard device under the existing domestic conditions. Therefore, the measurement uncertainty of the main feedwater flowmeter developed in this thesis can finally achieve the design target.

Prediction of RPV Irradiation Embrittlement Performance and Life Evaluation of Extended Operation in a Nuclear Power Plant

Fang Yonggang, Tong Zhenfeng, Chu Qibao, Zeng Zhen, Shen Yingcai

2023, 44(4): 192-197. doi: 10.13832/j.jnpe.2023.04.0192

Abstract(194) HTML (57) PDF(37)

Abstract:
A domestic self-designed and built nuclear power plant has entered the stage of extended operation. A foreign prediction model was adopted for the irradiation embrittlement evaluation of the reactor pressure vessel, however, the irradiation data on which the foreign prediction model was based cannot effectively represent the irradiation embrittlement performance of domestic RPV materials, especially for the extended operation stage. Based on the radiation embrittlement prediction models of RPV at home and abroad and their development mechanisms, an independent low-Cu RPV radiation embrittlement prediction model suitable for domestic engineering applications has been formed, which considers the key factors of irradiation embrittlement such as stable matrix defects and alloy element precipitation. According to the irradiation embrittlement data of domestic low-Cu RPV materials, the standard deviation and margin analysis of the independent model were carried out, and the results indicated a high level of confidence in prediction. Based on the model, the irradiation performance of the RPV was evaluated, and the feasibility of extended operation to 60 equivalent full power years (EFPY) was demonstrated.

Study on Issues of Transformation to Chinese Technical Specifications

Li Huwei, Zhang Yangcheng, Qian Xiaoming, Zhang Chi

2023, 44(4): 198-202. doi: 10.13832/j.jnpe.2023.04.0198

Abstract(175) HTML (38) PDF(29)

Abstract:
During the transformation of French technical specifications (TSs) to Chinese TSs for M310 NPPs such as Daya Bay Nuclear Power Plant with reference to the Standard Technical Specifications - Westinghouse Nuclear Power Plant (NUREG-1431, Rev. 4), it was found that there is no effective management of the unavailability of multiple safety functions or the unavailability of both safety-related systems and equipment in NUREG-1431 (Rev. 4). To this end, this paper fully studies the management status of the unavailability of multiple safety functions or both safety-related systems and equipment in M310 NPPs, compares the risk impact caused by the tranformation from French TSs to Chinese TSs, summarizes the safety problems caused by the change of management methods, and finally puts forward targeted work suggestions. This study has referential significance to the transformation from French TSs to Chinese TSs for M310 NPPs in China.

Research on Automatic Calibration Algorithm of Reactor Fuel Rods

Li Xiangdong, Jiang Hesong, Wang Xueyuan, Xu Xuejin, He Xiaochun

2023, 44(4): 203-208. doi: 10.13832/j.jnpe.2023.04.0203

Abstract(115) HTML (38) PDF(21)

Abstract:
Since the nuclear reactors need frequent replacement of fuel rods, it is necessary to determine the type and installation position of the core fuel rods accurately to ensure the safe operation of the reactor. Herein, the global and local virtual two-dimensional coordinate mapping models have been established in terms of the distribution relationship of fuel rod installation positions. The local sequence pictures of each viewpoint are taken to identify the central position of the fuel rods in the local pictures, and the local virtual two-dimensional coordinate mapping model is calibrated. Then, the Euclidean distance between the central position of the fuel rods and the position in the calibrated local mapping model is measured to realize type refactor, and the core panoramic mosaic is further obtained to assist calibration. The simulation results show that the algorithm can effectively detect the type and installation position of fuel rods, the recognition rate is higher than 98%, the accuracy rate reaches 100%, and the panoramic mosaic results are stable and reliable. It has great application potential in the calibration of core fuel rods.

Design of Control System for Inspection Robot of In-Containment Refueling Water Storage Tank

Guan Chaopeng, Wu DongDong, Gui Liang

2023, 44(4): 209-213. doi: 10.13832/j.jnpe.2023.04.0209

Abstract(129) HTML (59) PDF(23)

Abstract:
Aiming at the underwater inspection requirements of in-containment refueling water storage tank (IRWST) during the in-service overhaul of nuclear power unit and minimizing personnel radiation dose, a remote-controlled inspection robot is developed. Through the analysis of the IRWST inspection requirements, the function of the robot is defined, and the overall structure of the robot control system is designed. In order to meet the requirements of miniaturization of robot, the robot driver board, video coding board and decoding board based on embedded system are designed, and the upper and lower computer software of humanized operation is developed. The results show that the developed inspection robot can quickly and stably complete the video inspection and foreign object salvage tasks of various components of IRWST, which further improves the intelligent operation and maintenance level of robot in the nuclear industry.

Study on Optimization of Allowable Outage Time for Essential Service Water System in HPR1000

Chen Guocai, Yang Yun, Tong Jiejuan

2023, 44(4): 214-219. doi: 10.13832/j.jnpe.2023.04.0214

Abstract(113) HTML (48) PDF(20)

Abstract:
The risk-informed integrated decision-making method combining deterministic and probability theories was adopted for the optimization analysis and demonstration of the allowed outage time (AOT) in the terms of operational technical specifications for the essential service water system (WES) of Hualong One (HPR1000), including the determination of proposed changes, defense-in-depth analysis, safety margin analysis and probabilistic safety assessment (PSA). The results show that it is acceptable to extend the AOT from 72 h to 96 h for the unavailable train of the WES of HPR1000; the principle of defense-in-depth and safety margin requirements of HPR1000 are met; the increased risk meets the acceptance guidelines of the U.S. Nuclear Regulatory Commission (NRC) regulatory guidelines RG1.174 and RG1.177. On the premise of risk control, the flexibility of power plant operation can be further improved.

Experimental Study on Flow Instability in Parallel Channels with Supercritical Carbon Dioxide

Huang Jiajian, Zhou Yuan, Huang Yanping, Luo Qiao, Hu Wei

2023, 44(4): 220-225. doi: 10.13832/j.jnpe.2023.04.0220

Abstract(394) HTML (37) PDF(36)

Abstract:
Carbon dioxide has unique physical and chemical properties near the quasi-critical point. The Bretton cycle system using supercritical carbon dioxide (S-CO₂) as heat exchange medium has considerable system thermal efficiency, but the drastic change of physical properties may lead to the problem of flow instability. In this work, the experimental study of S-CO₂ flow instability in two-channel is carried out, and the experimental data of flow instability are obtained. The results show that the flow instability occurs in both the positive and negative slope regions of the natural cycle power-flow curve. The flow instability in the first section is systemic oscillation with a long period, and it is concluded as pressure drop oscillation, while the flow instability in the second section is inter-channel high-frequency oscillation. The onset power of the instability increases linearly with the increasing of system pressure and inlet mass flow rate. Therefore, increasing the system pressure and inlet mass flow rate will enhance the stability.

Development and Verification of 3D Code for Steam Generator Tube Rupture Accident of LBE-cooled Reactor

Gu Zhixing, Yu Hongxing, Huang Daishun, Yan Mingyu, Shen Yaou, Feng Wenpei, Gong Zhengyu

2023, 44(4): 226-233. doi: 10.13832/j.jnpe.2023.04.0226

Abstract(205) HTML (32) PDF(36)

Abstract:
Steam Generator Tube Rupture (SGTR) accident is one of the significant safety problems that must be considered in the design of Lead-Bismuth-Eutectic (LBE) cooled reactor. With respect to the SGTR in LBE-cooled reactor, and to cope with the challenges of 3D propagation of pressure waves and 3D migration of water steam under the complex geometric structures, the three-dimensional numerical model and algorithm of the interaction between LBE and water were studied based on the Euler hydrodynamic theory of multiphase flow, and a special code was developed. The code was verified by means of experimental comparison and code-to-code comparison, and the verification results were in good agreement. It is demonstrated that the numerical theories and models used in this paper are suitable for describing the "LBE-water" interaction during SGTR accidents in LBE-cooled reactors. And the 3D code developed in this paper has important potential application values in coping with the 3D evolution processes of SGTR accidents in LBE-cooled reactors under the complex geometric structures, including the pressure wave propagation and steam migration. The research achievements in this paper are expected to provide strong supports for the SGTR accident analyses of LBE-cooled reactors in China.

Study on Evaluation Method for Creep Performance of New Zirconium Alloy Cladding

Xing Shuo, Pu Zengping, Zhang Kun, Jiao Yongjun, Dai Xun, He Liang

2023, 44(4): 234-239. doi: 10.13832/j.jnpe.2023.04.0234

Abstract(278) HTML (82) PDF(57)

Abstract:
In order to establish a creep model for new zirconium alloy, based on the data from creep test, the creep behavior of the new zirconium alloy cladding tubes at the temperature of 593-673K and the stress of 60-160 MPa was investigated. The classical zirconium alloy creep model was used to predict the creep behavior of the new zirconium alloy clad tube. Based on the constitutive equation of creep behavior, a preliminary evaluation method of creep of new zirconium alloy cladding tube was studied and established, and the preliminary verification was carried out. The verification results show that the predicted values are in good agreement with the measured values, and the maximum relative deviation is whithin 25%. The model is in agreement with the change of creep behavior of zirconium alloy in reactor. In this study, a new zirconium alloy cladding creep model is established, and the model is in agreement with the change of the in-core creep behavior of the zirconium alloy, which provides support for the prediction of the new zirconium alloy's in-reactor behavior.

Experimental Study on Prototype Melt under Severe Accident

Li Yang, Gong Houjun, Guo Kerong, Hu Yuwen, Yang Shengxing, Zan Yuanfeng, Yang Zumao, Huang Yanping

2023, 44(4): 240-246. doi: 10.13832/j.jnpe.2023.04.0240

Abstract(605) HTML (112) PDF(37)

Abstract:
In order to study the stratification morphology of the molten pool of the lower head of the pressure vessel under severe accident conditions, it is necessary to melt the prototype melt into liquid for experiments. In this study, CESEF experimental device is used, and the prototype melt is melted with the electromagnetic cold crucible technology. The maximum charge capacity is 5000 g, and the maximum temperature is 3000°C. The supporting high-frequency power supply is of 400 kW and a frequency of 100 kHz. An experimental study is carried out on the composition of Hualong One (HPR 1000) core melt, and it is found that the molten pool is divided into two layers: the metal layer and the oxide layer. By sampling different positions of the metal layer and oxide layer, it is found that the metal layer is mainly composed of stainless steel with some U and Zr, and the oxide layer is mainly composed of U, Zr and O in suboxide state, with few other contents.

Experimental Study on Cooling Characteristics of Mixed Particle Size Debris Bed under Different Water Injection Methods

Yang Shengxing, Gong Houjun, Fang Yu, Li Yang, Hu Yuwen, Zan Yuanfeng, Yang Zumao, Zhuo Wenbin

2023, 44(4): 247-252. doi: 10.13832/j.jnpe.2023.04.0247

Abstract(445) HTML (69) PDF(26)

Abstract:
The liquid core melt interacts with the coolant and breaks into a particle bed. Effective cooling of the particle bed can realize the retention of the melt and stop the accident process. In this paper, based on the particle size distribution and porosity of the fragments after the prototype molten FCI experiment, a mixed particle size sand debris bed with internal heat source was constructed, and the dryout characteristics of the debris bed under different enhanced heat removal measures (submerged pool at the top, water injection at the bottom driven by natural circulation, and circumferential water inlet) were studied. The results show that, under the condition of submerged pool at the top, the bubble clogging zone appears first in the middle and upper part of the debris bed, and then the liquid deficiency dryout zone appears in the middle and lower part of the debris bed. Under the condition of water injection at the bottom driven by natural circulation, the fluid deficiency at the bottom of the debris bed was greatly improved, and the dryout heat flux (DHF) increased by more than 2.5 times, the dryout area was located in the middle and upper part of the debris bed. Under the circumferential water inlet, DHF also increased by more than 2.5 times.

Experimental Research on Mechanism of Aerosol Re-entrainment Behavior in Containment under Severe Accident Condition

Hu Zhen, Chen Linlin, Ji Songtao, Wei Yansong, Shi Xiaolei, Zheng Guangzong

2023, 44(4): 253-258. doi: 10.13832/j.jnpe.2023.04.0253

Abstract(414) HTML (46) PDF(34)

Abstract:
Based on the Integral test facility of aerosol re-suspension and re-entrainment, the experimental study on aerosol re-entrainment behavior in passive containment was carried out. By measuring the mass concentration, quantity concentration and particle size distribution of particulate matter, the motion and distribution of aerosol in the test vessel during the re-entrainment stage were analyzed. By changing the pool size, pool surface tension, particle material and concentration, the change of aerosol re-entrainment rate under different water vapor evaporation rate on unit area, different pool surface tension, different particle solubility and concentration was studied. The results show that, during the pool boiling, the re-entrainment rate of aerosol is affected by the surface tension and water vapor evaporation rate on unit area, and the particle size distribution is related to the particle material.he concentration of particulate matter solubility. The results show that, during the pool boiling, the re-entrainment rate of aerosol is affected by the surface tension and water vapor evaporation rate on unit area, and the particle size distribution is related to the particle material.

Development and Verification of Ph Calculation Model of in-Containment Refueling Water Storage Tank under Severe Accidents

Dai Wei, Jiang Pingting, Chen Peng, He Dongyu

2023, 44(4): 259-266. doi: 10.13832/j.jnpe.2023.04.0259

Abstract(409) HTML (53) PDF(23)

Abstract:
In order to solve the problem of lacking tools for calculating the pH value of in-containment refueling water storage tank in nuclear power plants after the accident, this paper develops a direct modeling and in-time analysis model for pH value calculation. Based on Newton-Raphson method, by establishing the database of physical properties of key items and reactions, and by building a gas-liquid two-phase chemical equilibrium model, a pH value calculation software CalcpH with a complete database, the capability of high radiation calculation and the capability of coupling calculation with accident evolution is developed. The results of CalcpH are compared with the results calculated by ASTEC (a commonly used severe accident analysis software) and PHREEQC (a commonly used chemical equilibrium analysis software). For non-irradiation reactions, the difference between the calculation results of CalcpH and the results of PHREEQC is within 1.3%. For irradiation reactions, the difference between the results of CalcpH and the results of ASTEC is within 2.7%. While comparing the calculation results of CalcpH with experimental results, the difference is within 1%. The reliability of the calculation results is fully proved by software comparison and experimental comparison. Therefore, the numerical calculation model established by CalcpH can be used to predict the pH value of in-containment refueling water storage tank after the accident..

2023 Vol. 44, No. 4