核动力工程

The Research and Development Progress of SCWR

Zang Jinguang, Huang Yanping

2021, 42(6): 1-4. doi: 10.13832/j.jnpe.2021.06.0001

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Abstract:
Supercritical water cooled reactor (SCWR) is the only water-cooled reactor among the six types of reactors identified by the Generation IV International Forum. This paper describes the technical characteristics of SCWR, reviews the research and development (R&D) process of SCWR in China, and briefly reviews the latest R&D developments of SCWR in Canada, EU, Japan and other countries. Finally, this paper summarizes the technical advantages, technical challenges and development opportunities of SCWR.

Selection of Burnable Poison Loadings of Long-Life Plate-Shaped PWR Assemblies

Xu Shikun, Yu Tao, Xie Jinsen, Li Mancang, Xia Yi, Yao Lei

2021, 42(6): 5-11. doi: 10.13832/j.jnpe.2021.06.0005

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Plate-shaped PWRs have good application prospects in long-life reactors. According to the needs of long-life plate-shaped PWRs, the four burnable poisons with better neutronics performance (¹⁵⁷Gd₂O₃, ¹⁶⁷Er₂O₃, ²³¹Pa₂O₃ and PACS-J) were studied on the selection of burnable poison loading forms, and the loading form with better neutronics performance was screened out. The results show that: Different loading forms for different burnable poisons can better meet the comprehensive requirements of long-life plate-shaped PWRs; For burnable poison with better neutronics performance, ¹⁵⁷Gd₂O₃ and ¹⁶⁷Er₂O₃ can be loaded with fuel evenly mixed; ²³¹Pa₂O₃can be loaded with burnable poison mixed in the cladding; PACS-J can be loaded in pellet form mixed with fuel.

Characteristic Analysis on Distribution of Void Fraction during Stable Flow Flashing Process under Low Pressure

Zhang Yongfa, Du Keyue, Cao Xiaxin, Fang Yuliang, Liu Xiaoya, Jiang Lizhi

2021, 42(6): 12-16. doi: 10.13832/j.jnpe.2021.06.0012

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Abstract:
Taking the flashing phenomenon in the low pressure natural circulation system as the research object, the change law of void fraction in the stage of stable two-phase natural circulation flow driven by flashing is analyzed. Through the analysis, the following law is found that the inlet fluid temperature of the rising section and the liquid level height of the water tank will cause the flow to affect the flashing vaporization process, which makes the radial and axial distribution of the void fraction different. Through the analysis, it is concluded that the main factor affecting the vaporization process is the fluid superheat degree. It is known that reducing the fluid degree of subcooling at the inlet of the rising section, the flashing starting point will move down, and the flashing two-phase section will become longer. Along with the continuous flashing vaporization, the fluid degree of subcooling decreases gradually, the axial void fraction distribution increases rapidly first and then gradually slows down, and the radial void fraction changes from “wall peak” type to “nuclear peak” type. Then, based on the change of local fluid degree of subcooling, fitting gives the calculation formula of axial void fraction under different cases. By comparing with the experimental data, it is found that the fitting is good, and the relative error is within ±15%.

Experimental Study of Bottom Reflooding in a Narrow Rectangular Channel

Deng Yonghao, Xu Wei, Liu Xiaojing, Guo Jiuyuan, Wu Dan

2021, 42(6): 17-23. doi: 10.13832/j.jnpe.2021.06.0017

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Abstract:
In order to investigate thermal hydraulic characteristics during bottom reflooding when LOCA occurs in a narrow rectangular channel, the reflooding experiments were carried out under different conditions. The narrow rectangular channel was made by welding two Inconel alloy plates. This study analyzes the bottom reflooding process according to the temperature variation curve, calculates and compares the advancing speed of the quench front (quench speed) under different experimental conditions, and studies the influence of experimental parameters on the reflooding process. Experimental results indicated that quench velocity under bottom reflooding is increasing with increased system pressure, increased inlet velocity, reduced initial wall surface temperature and increase in the degree of subcooling of the coolant. A comparative analysis of the bottom and combined reflooding case indicates that quench velocity under bottom reflooding is higher than that during combined reflooding with the same flow rate. This study lays the foundation for the research of accident prevention and mitigation of plate fuel element reactors.

Fine Calculation of Decay Heat Power in Xi’an Pulsed Reactor

Yang Ning, Li Huaqi, Zhang Xinyi, Tian Xiaoyan, Zhang Qiang

2021, 42(6): 24-31. doi: 10.13832/j.jnpe.2021.06.0024

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Abstract:
In order to provide a more precise and reliable energy source term for the numerical calculation of waste heat export of pulsed reactor, the calculation method XAPRDH and the program with the same name are developed by coupling MCNP program and ORIGEN2 program. In terms of method realization, firstly, the fuel part of each fuel element (including the control rod follower) is divided into 30 independent units in the form of 10 equal parts in the axial direction and 3 parts in the radial direction, forming a total of 3180 units in the whole reactor; Then, the material composition, nuclear reaction cross section, neutron flux density and fission power of each unit are obtained by flexibly calling MCNP program and ORIGEN2 program, and finally the decay heat of each unit is calculated and tracked independently. The analysis shows that the burnup evaluation data in this paper are in good agreement with the literature values, and are consistent with the experimental values within the measurement error range. The decay heat calculation results of the whole reactor are also in line with the industry standards, indicating that the fine calculation method of decay heat established in this paper is feasible and the calculation results are reliable.

Dynamic Viscosity Measurement of Helium-Xenon Mixture Gas

Hu Wenzhen, Li Zhongchun, Liu Xiaojing, Deng Jian, Qu Wenhai

2021, 42(6): 32-37. doi: 10.13832/j.jnpe.2021.06.0032

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Abstract:
In view of the lack of research on the thermal and physical properties of helium - xenon gas mixture, the viscosity of helium-xenon gas mixture was studied. The experimental apparatus was designed based on the dual capillary method and the correction term was considered. The dynamic viscosities of two kinds of helium-xenon mixtures (15 and 40 g/mol) at temperatures of 298.15~548.15 K and pressures of 0.1~2.5 MPa were measured and evaluated after the calibration of the experimental apparatus with argon gas. In order to obtain the viscosity of He-xenon mixture at high temperature, the viscosity fitting value was extrapolated to 1273 K by using the method of fitting viscosity relation. The results show that the experimental results are in good agreement with the literature values. The standard uncertainty of synthesis measured by the experimental equipment is 3.88%. Compared with the experimental and calculated values in the literature, the deviation between the fitted values and the calculated values is small. This study provides the basic thermal and physical parameters for the design and optimization of space gas cooled reactor.

Measurements of Large Bubble Volume Based on 2-D Images Processing Applying Convolutional Neural Network

Hu Ningning, Dang Zhuoran, Zhang Muhao, Tang Ke, Mamoru Ishii

2021, 42(6): 38-43. doi: 10.13832/j.jnpe.2021.06.0038

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Abstract:
Large bubbles (500<Re<2000) in static flow field have the irregular geometrical shape due to the effects of surface tension and inertia force, which brings the enormous error to spherical or ellipsoid equivalent method used in 2D image processing to obtain 3D volume. Besides, due to the refraction and reflection on the irregular surface, the bubble boundary in 2D image is blurred and unrecognizable. The present research applied high-speed video camera to obtain 2D gray scale images of large bubbles in static flow field and used the images as the input of the convolutional neural network (CNN). The bubble 2D projected area and volume obtained in experiments were applied to train the CNN. Finally well-trained CNN was applied to predict the bubble volume. In experiment, actual bubble volume was obtained by small bubble superposition and compared with the predicted value of CNN. The results show that comparing with the traditional image processing method, the proposed method needs no assumption for bubble shape and improves the applicability for large bubble.

Analysis of Disturbance Resisting Ability of Dual-Loop Natural Circulation System under Asymmetrical Conditions

Deng Shengwen, Zhu Enping, Zhao Pengcheng, Zhai Pengdi, Liu Zijing, Yu Qingyuan

2021, 42(6): 44-49. doi: 10.13832/j.jnpe.2021.06.0044

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Abstract:
Natural circulation lead-cooled fast reactor heat exchangers operate under the harsh environment of high temperature, high pressure difference, high density and high corrosion, which can easily induce heat transfer tube rupture and blockage accidents, resulting in asymmetric thermal load or asymmetric resistance operation of the reactor. It has an important impact on the safe and stable operation of the reactor. This paper takes the lead-cooled dual-loop natural circulation system as the research object, and uses a non-dimensional analysis method to derive the theoretical solution of the natural circulation flow of the dual-loop system; then carries out the disturbance characteristics analysis of the natural circulation system under different thermal load differences or resistance differences, using the fitting approximation method to establish the natural circulation characteristic parameters that characterize the anti-disturbance ability, and get the best anti-disturbance interval. The research results show that when the system introduces some thermal load and resistance disturbance, the loop flow does not change greatly, and the system has strong anti-disturbance ability.

Analysis of Primary Loop System of High-Order Fully-Implicit Nuclear Reactor Based on MOOSE Platform

Niu Yuhang, He Yanan, Wu Yingwei, Xiang Fengrui, Deng Chaoqun, Yu Yang, Su Guanghui, Qiu Suizheng, Tian Wenxi, Lu Tianyu

2021, 42(6): 50-57. doi: 10.13832/j.jnpe.2021.06.0050

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Abstract:
Based on the multi-physics coupling platform MOOSE (Multiphysics Object Oriented Simulation Environment), the modular system safety analysis code, ZEBRA, was developed. The high-order and fully-implicit discrete format was presented to establish the nuclear reactor primary loop model. And the coupling calculation of nuclear reactor system including neutron diffusion, two-dimensional solid heat conduction, and one-dimensional fluid were also implemented. For the problem of flow and heat transfer in a single pipe, the ZEBRA code is coupled and verified, and the solution accuracy of first-order and second-order spatial discrete formats under steady-state conditions and Implicit-Euler, Crank-Nicolson and BDF2 schemes under transient conditions are compared, and the steady-state and power-reducing transient conditions of PWR loop system are simulated and analyzed. The results show that the high-order spatial discrete format has a higher accuracy, and the BDF2 time discrete format is in best agreement with the theoretical solution. The temperature, velocity, and pressure distribution of the PWR loop system are reasonable. The steady-state and transient calculation results are in good agreement with those of RELAP5.

Evaluation of the Impact of Core Power Distribution on Reflooding Phenomenon in CCTF Test

Zhao Ningning, Yuan Hongsheng, Wen Qinglong, Ruan Shenhui

2021, 42(6): 58-64. doi: 10.13832/j.jnpe.2021.06.0058

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Abstract:
The cylindrical core test facility (CCTF) is taken as the research object, optimal estimation program for transient behavior of LWR cooling system under accident condition RELAP5 and autonomous core design and safety analysis program LOCUST are used to evaluate the impact of core power distribution on the reflooding phenomenon under CCTF C2-SH2 (Run54) test conditions. The research shows: ① The calculations, such as the pressure drop in the descending section and core and the steam mass flowrate at the outlet of the code, provide good agreement with the test results; ② For the calculation of the average channel cladding peak temperature at 1.015 m of the core, the cladding peak temperatures calculated by RELAP5 and LOCUST programs are 816 K and 813 K respectively, the test results are 898 K, the calculated value is about 82 K lower than the test value, and the average channel cladding temperature finally stabilized at about 400 K. The calculated results are consistent with the test results. Therefore, the results of this study show that the LOCUST program can better simulate the transient process of the reflooding phase in the large-break loss-of-coolant accident (LBLOCA).

Assessment of RELAP5 Code for Predicting Unstable Boundary of Type-I Density Wave with Experiment

Teng Chen, Xie Heng, Jia Haijun

2021, 42(6): 65-71. doi: 10.13832/j.jnpe.2021.06.0065

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In order to estimate the unstable (Type-I DWO) boundary of Type-I density wave of low temperature nuclear heating reactor and determine the parameter interval of its slight-boiling mode, the RELAP5 numerical model of NHR200 similarity experimental loop HRTL200 of low temperature nuclear heating reactor is established in this paper. By comparing the simulation results with the experimental results, the general characteristics of RELAP5/MOD3.2 program for simulating Type-I DWO and the ability to predict the unstable boundary are evaluated, and the effects of inlet and outlet resistance coefficients and interphase friction on the simulation results are analyzed. The results show that the general characteristics of Type-I DWO simulated by RELAP5 program are in good agreement with the experiment; When the operating pressure is not higher than 25 bar (1 bar=10⁵Pa), the deviation between the degree of subcooling boundary value of the unstable boundary calculated by the program and the experimental value is within 3 K; when the operating pressure is greater than 30 bar, the accurate interphase friction relationship can improve the prediction results. Therefore, RELAP5 program can be used to simulate and predict Type-I DWO after selecting the interphase friction relationship matching the circuit.

Supercritical Water-Cooled Small Modular Reactor R&D

Zang Jinguang, Huang Yanping

2021, 42(6): 72-76. doi: 10.13832/j.jnpe.2021.06.0072

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Supercritical Water-cooled Small Modular Reactors (SCW-SMR) is a fusion of Supercritical Water-cooled Reactor (SCWR) and Small Modular Reactor (SMR). SCW-SMR could have both the advantages of SCWR and SMR, which have independent market demands and could provide basis for the engineering practice of the million-kilowatt SCWR. This article introduces SCW-SMR's R&D background, main international R&D developments, technical characteristics and advantages, and puts forward some thoughts on the R&D process, including overall design principles, main design requirements, specific design considerations, and R&D stage recommendations for subsequent R&D reference.

Simplification of Critical Criteria for Reverse Flow Determination Based on Reverse Flow Theory Model of U-Tube Steam Generator

Liu Hao, Ma Zaiyong, Jiang Zhangrui, Tang Yu, Zhang Luteng, Xu Jianjun, Sun Wan, Yuan Dewen, Pan Liangming, Zhou Wenxiong

2021, 42(6): 77-81. doi: 10.13832/j.jnpe.2021.06.0077

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In the problem of U-tube reverse flow, accurate determination of the critical point of reverse flow is quite important. In this paper, a theoretical analysis model of U-tube reverse flow characteristics was established, and the criticality criterion of reverse flow based on the theoretical analysis model was simplified. The calculation results of the theoretical analysis model and the simplified reverse flow critical criterion were compared with the experimental results. The results showed that the calculation results of the reverse flow theoretical analysis model and the simplified reverse flow critical criterion were in good agreement with the experimental results, and the absolute values of the average relative errors were 3.4% and 3.7%, respectively. It showed that the theoretical analysis model of reverse flow and the simplified reverse flow critical criterion were more accurate for the prediction results of reverse flow point.

Experimental Study on Pressurizer Surge Line Double-Ended Break of Small Reactor

Huang Zhigang, Zhang Yan, Peng Chuanxin, Zan Yuanfeng, Zhuo Wenbin, Yan Xiao

2021, 42(6): 82-86. doi: 10.13832/j.jnpe.2021.06.0082

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Abstract:
Accident response characteristics of the passive safety system and the parameter variation of the primary loop system were obtained through the experimental study of double-ended break of modular small reactor pressurizer surge line . The results show that under the extreme condition of double-ended break of the pressurizer surge line, the MP accumulator can provide a large stable safety injection flow in a short time to replenish the water capacity of the system in time; The operation process of the HP safety injection system is relatively complex, and the safety injection flow rate is closely related to the medium state of the pressure balance pipeline of the core makeup tank and the operation state of the MP safety injection system, and it is in the state of intermittent injection within 1.7 hours. During the whole accident process, the core has been submerged. The modular small reactor passive safety system can ensure the core safety under the extreme condition of double-ended break of pressurizer surge line.

Design and Optimization of Typical Cells of Solid Core for Heat Pipe Reactor

Huang Yongzhong, Li Yuanming, Li Wenjie, Li Quan, Chai Xiaoming, Zhao Bo, Tang Changbing

2021, 42(6): 87-92. doi: 10.13832/j.jnpe.2021.06.0087

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The solid reactor core design is a key problem in heat pipe reactor, which can affect the reactor heat transfer performance and structural integrity. To avoid the over high temperature and excessive stress caused by the gap thermal resistance, the three-dimensional thermodynamics models for typical cells of four reactor core designs have been established for optimal calculation with different fillings under the key parameters about gap size and cell section size. The result showed that, the fuel cladding and pellet temperature decrease effectively with filling the assembly clearance with liquid sodium which has high thermal conductivity, but the thermal stress increases instead. The thermodynamic performance of tubes with liquid sodium is the best; Among the solid core solutions, the thermodynamic performance of hexagonal tubes filling with helium is best in solid reactor core designs.

Corrosion Behavior Study of Fe-22Cr-25Ni Austenitic Heat-Resistant Steel under Supercritical CO₂ Condition

Guo Tingshan, Liang Zhiyuan, Gui Yong, Zhao Qinxin

2021, 42(6): 93-99. doi: 10.13832/j.jnpe.2021.06.0093

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The high temperature corrosion behavior of austenitic heat-resistant steel Fe-22Cr-25Ni under 600℃/700℃ and 15 MPa supercritical CO₂ condition was investigated. The composition, content and element distribution of corrosion products were characterized by Raman spectroscopy, Glow Discharge Spectroscopy, SEM and EDS. The results show that the corrosion kinetics of Fe-22Cr-25Ni at 600℃/700℃ follow the parabola-like law, and the change of corrosion mass gain increases with increasing temperature. By observing the characterization results and thermodynamic calculations, it is concluded that the corrosion product composition is mainly Cr₂O_3, specifically, from the gas side to the substrate side are the outermost Mn oxide, the internal Cr₂O₃ and Mn-Cr oxide, the SiO₂ layer at the oxide layer / substrate interface, and the carbide and internal oxide in the substrate; C is mainly deposited on the surface of corrosion products, the width and depth of the Cr-depleted zone increase with increasing time. At the same time, according to the mass ratio of O and C and the results of thermodynamic calculation, it is suggested that C is very likely to be diffused in ionic state.

Study on Relationship between Microinhomogeneity and Measurement Precision of γ-Ray Transmission Technique

Qu Xiaolong, Luo Jiandong, Chen Jie, Peng Sitong

2021, 42(6): 100-104. doi: 10.13832/j.jnpe.2021.06.0100

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In this paper, the relationship between the measurement accuracy of monoenergetic γ-ray transmission method and the microscopic uniformity of the sample is studied. By establishing a microscopic distribution model parallel to and perpendicular to the transmission direction of γ-ray, and carrying out theoretical analysis of different models, it is concluded that the microscopic inhomogeneity in the direction parallel to the transmission of γ-ray will increase the transmittance of the γ-ray, which in turn leads to low mass thickness measurement results; the microscopic inhomogeneity in the direction perpendicular to the transmission of γ-ray has no effect on the measurement results. At the same time, the microscopic inhomogeneity models of different mass thickness distributions are established, and these models are used to analyze the law that the measurement accuracy is affected by the microscopic inhomogeneity. The inhomogeneity of the model is described by the coefficient of variation. The results show that under the same experimental conditions, the greater the coefficient of variation, the greater the influence on the transmission intensity of γ-ray, and there is a linear increasing relationship between the relative change of γ-ray transmittance and the coefficient of variation of the sample.

Analysis and Research on Magnetic and Eccentric Characteristics of CRDM Motor

Deng Qiang, Peng Hang, Yu Tianda, Zhang Zhiqiang, Liu Yanting, Zhou Xu

2021, 42(6): 105-108. doi: 10.13832/j.jnpe.2021.06.0105

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Control Rod Drive Mechanism (CRDM) of the nuclear power plant reactor uses synchronous motor as the key component for electromechanical energy conversion. The unilateral radial magnetic pull of the motor would lead to deformation of motor shaft system and aggravate the bearing wear, which has an important influence on the life of CRDM and the reliability of nuclear reactor operation. In this paper, the mechanism of radial static eccentricity, radial dynamic eccentricity and inclined eccentricity of CRDM motor is analyzed, and the mathematical models for calculating radial and axial magnetic pull are established. The variation rules of radial and axial magnetic pull and their relationship with eccentricity are obtained. The result shows that radial magnetic pull is significantly larger than axial magnetic force for CRDM motor, and is linearly proportional to the eccentricity of the rotor center and nonlinearly proportional to the maximum tilt eccentricity at both ends of the rotor. The conclusion can provide guidance and basis for the optimization and improvement of CRDM motor and shaft system structure design.

Evaluation Method and Application of Seismic Acceleration Response Amplification Factor of Line-Mounted Equipment in Nuclear Power Plants

Tan Xiaohui, Lu Jun, Zhang Pan, Sun Lei

2021, 42(6): 109-113. doi: 10.13832/j.jnpe.2021.06.0109

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In order to fast evaluate the seismic load on line-mounted equipment of nuclear power plant, based on the spring-mass models, and the relationship between the piping and line-mounted rigid equipment, a method of obtaining the amplification factor of the seismic acceleration response of equipment and its applications in obtaining seismic input for anti-seismic qualification of equipment are proposed. And furthermore, the method and its applications are verified by using finite element analysis. The result shows this evaluation method provides a feasible application method which is easy to operate without being too conservative and reducing economy in determining the seismic input for anti-seismic qualification of line-mounted equipment.

Research on Architecture Design of Reactor Protection System Based on NASPIC Platform

Liu Mingming, Jin Gang, Hu Qingren, Wu Youguang, Wu Zhiqiang, Xiao Peng

2021, 42(6): 114-119. doi: 10.13832/j.jnpe.2021.06.0114

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Reactor protection system is a very important safety system in nuclear power plant. It is mainly used to protect the safety of reactors, the environment and personnel. It is the Class 1E instrumentation & control equipment of nuclear power plants. Its own reliability and safety play an important role in the normal operation of nuclear power plants. The architecture of the reactor protection system plays a key role in the reliability, availability and maintainability of the entire system. Based on the NASPIC platform and according to the functional requirements and design criteria of the reactor protection system, this paper proposes a relatively complete and reasonable reactor protection system architecture that meets the requirements of the functions and design criteria. At the same time, a scientific research prototype of “Hualong No. 1” was built according to the architecture design, and based on the FTA/Markov reliability analysis method, the function test and reliability analysis and calculation of the scientific research prototype of the protection system built were carried out. It proves that the architecture design of the reactor protection system meets the design requirements and provides a reference for the system architecture design of subsequent projects.

Research on Probabilistic Safety Margin Analysis Method of SGTR Based on Raven

Kong Huanjun, Liu Ziyin, Xu Anqi, Wang He

2021, 42(6): 120-127. doi: 10.13832/j.jnpe.2021.06.0120

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A Risk-informed Safety Margin Characterization (RISMC) analysis method based on RAVEN software and Monte Carlo (MC) sampling is introduced in the paper. The influence of thermal parameters and personnel action time uncertainty on the safety margin of steam generator tube rupture (SGTR) accident is comprehensively analyzed. The calculation results are compared with the traditional safety assessment methods. Aiming at the key influence parameters of the accident, based on MC sampling to quantify the key parameter samples affecting the safety margin, the system simulation model of SGTR accident is established by RELAP5 program, coupled calculation and analysis are carried out by RAVEN software. And finally the probabilistic safety margin of SGTR accident and its sensitivity to various influencing parameters are obtained under the condition of auxiliary water supply system failure.

Analysis of Impact of Medium Pressure Safety Injection on Severe Accident Process under RRA Connected Mode

Yu Chengxin, Hao Bin, Deng Lingling

2021, 42(6): 128-134. doi: 10.13832/j.jnpe.2021.06.0128

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Aiming at the loss of coolant accident (LOCA) of China's improved million kilowatt PWR (CPR1000) Nuclear Power Unit losing high and low pressure safety injection and spray in reactor residual heat removal system (RRA) connected mode of intermediate shutdown, the reactor core, reactor coolant system and containment system of the reference unit are simulated and calculated by using MAAP5 program, at the same time the impact of medium pressure safety injection on the severe accident process is analyzed by combining with the calculation results, and the mitigation effect on the accident is studied. The results show that the timely injection of medium pressure safety injection can effectively limit the temperature rising behavior of the core, and play an important role in mitigating the process of severe accidents during the process of core exposure and temperature rise caused by LOCA under RRA connected mode, and even it is possible to avoid core integrity damage when high and low pressure safety injection and spray is lost under accident conditions. Finally, according to the analysis results, the improvement suggestions are put forward for the current operation regulations of nuclear power units. For the administrative isolation behavior of the medium pressure safety injection tank, only the corresponding isolation operation is performed for its electrical switch, while the local part of the valves in the containment building is provided with a tag warning, and the on-site padlock operation is not performed, which can not only avoid the misinjection of the medium pressure safety injection tank under normal operation conditions, the integrity of the reactor core can be also effectively guaranteed after the occurrence of LOCA under RRA connected mode. The normal and safe operation of the power plant can be ensured, and the mitigation ability of severe accidents of the unit under this mode can be improved.

Prediction of Cavitation Characteristics of Throttle Orifice Plate Based on Improved BP Neural Network

Zhang Yu, Sun Lei, He Chao, Yuan Shaobo

2021, 42(6): 135-140. doi: 10.13832/j.jnpe.2021.06.0135

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In order to obtain the cavitation characteristics near throttle orifice plate in nuclear pipeline efficiently, a reliable improved backpropagation (BP) neural network prediction model was constructed. Firstly, the geometric feature parameters of the throttle orifice plate were extracted, and the sample data base of these parameters was generated by using the Latin Hypercube Sampling approach. Then, the minimum cavitation number corresponding to each sample is obtained by computational fluid dynamics (CFD) method, and the dimensionless parameter is used as the output response. Finally, in view of the deficiency of the original BP neural network prediction model, an improved prediction model of throttle orifice plate cavitation characteristics is established by combining genetic algorithm. The results show that the orifice diameter and front opening angle have strong global sensitivity to the minimum cavitation number; the prediction accuracy of the BP neural network prediction model optimized by genetic algorithm has been greatly improved, and the root mean square error is reduced by about 36.4%.

Fatigue Reliability Evaluation Method for Hold Down Spring of Reactor Vessel Internals Considering Stress Relaxation and Irradiation

Zhang Yi, Li Yan, Sun Bo, Cao Qifeng, Pan Junlin, Yang Xi, Yang Dezhen, Ren Yi

2021, 42(6): 141-147. doi: 10.13832/j.jnpe.2021.06.0141

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For the fatigue failure modes of the holddown spring of reactor internals, the reliability evaluation was carried out based on the simulation method under the condition of considering the stress relaxation and the influence of radiation. Firstly, combining the fatigue model and the stress relaxation Landgraf model, considering the influence of irradiation on the fatigue parameters, the fatigue life model of the holddown spring was established. Based on the holddown spring fatigue life model, the reliability of holddown spring was defined according to the generalized stress-strength interference model and the sensitivity analysis was carried out. Taking the holddown spring of AP1000 passive pressurized water reactor as an example, the fatigue life with a reliability of 95% and 50% at 95% confidence level were calculated. The results show that the total fatigue life of the holddown spring decreases by 88.3% without considering the stress relaxation, and the prediction results are conservative from the perspective of economy. The results of sensitivity analysis show that the elastic modulus and fatigue strength coefficient have great influence on reliability. The holddown spring can be optimized by adjusting the design variables under certain reliability requirements.

Analysis and Improvement of Safety Class DCS Network Structure in a Nuclear Power Plant

Zhao Jian, Liu Dongliang, Ma Zhixin

2021, 42(6): 148-154. doi: 10.13832/j.jnpe.2021.06.0148

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Based on the design basis accident of degraded operation of ACPR1000 nuclear power unit caused by instrument and control system fault, five kinds of specific common cause faults are proposed. According to the connection robustness indicator, the improved method of the arrangement of the ring network nodes in the safety level DCS is summarized, and the common cause failure table is made for the network nodes with common cause faults, the adjacent sequencing is carried out according to the probability of common cause faults and the ring network structure is arranged to improve the connection robustness indicator of the network. The method is used to adjust the distribution of safety level DCS network in nuclear power plant, which improves the robustness of the network structure and contributes to the safety and reliability of nuclear power unit operation.

Reliability Analysis of Digital Emergency Shutdown System Based on Markov/CCMT Method

Chen Ling, Chen Han, Zhang Yongfa

2021, 42(6): 155-160. doi: 10.13832/j.jnpe.2021.06.0155

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Traditional reliability analysis methods are more restricted in the reliability application of digital I&C systems, and new reliability analysis methods need to be explored. Taking the digital emergency shutdown system as the research object, Markov/CCMT is selected to complete the failure modeling of the system, and the qualitative analysis result of the emergency shutdown system is obtained. On this basis, quantitative analysis is performed to obtain the failure rate of the system. The results show that the failure probability of the digital emergency shutdown system analyzed by the Markov/CCMT method is 1.3×10⁻⁶ h⁻¹, and the Marko/CCMT method has certain applicability in the reliability analysis of the digital I&C system. Compared with the traditional reliability analysis method, Markov/CCMT method is more comprehensive in the expression of the dynamic characteristics of the system.

Study on IVR Capability Margin of Pressurized Water Reactor Based on Decay Heat Uncertainty

Song Jian, Yu Hongxing, Deng Jian, Xiang Qingan, Zhu Dahuan, Xu Youyou, Luo Yuejian

2021, 42(6): 161-166. doi: 10.13832/j.jnpe.2021.06.0161

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Abstract:
In order to determine the influence of decay heat on the IVR (In-Vessel Retention) capability margin of HPWR, an evaluation method combining significance level evaluation and sampling failure rate was used to evaluate IVR capability margin. Using CISER for IVR to carry out the sampling calculation, the ratio of the peak heat flux on the lower head wall to the local critical heat flux (CHF) under different power levels and decay heat distribution parameters of nuclear power plants was obtained to carry out the significance level estimation and failure rate calculation of the heat flux ratio to judge whether the IVR measure is effective based on the penetration criterion of lower head (less than local CHF) to obtain the IVR capability margin. The results show that IVR alone is not recommended as a serious accident mitigation measure for PWR plants with an electrical power of more than 1400 MW without any optimization of the heat transfer composition inside and outside the lower head.

Equipment Design and Research of the HPR1000 In-Containment Refueling Water Storage Tank Strainers Based on Single Variable Solution Method

Gong Zhao, Zhu Jingmei, Zhang Wei, Huang Ruolin, Zhu Minghua

2021, 42(6): 167-173. doi: 10.13832/j.jnpe.2021.06.0167

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Abstract:
In order to solve the pressure drop solution problem in the equipment design of HPR1000 In-containment Refueling Water Storage Tank (IRWST) strainers after accident, this paper proposed a method of single variable solution for the pressure drop of IRWST strainers: a resistance element was added between the strainer module and the collector pipe, thus the multiple groups of variables in the IRWST strainers pressure drop solution were converted into the single group of variables (the flow area of resistance elements), and the pressure drop solution of IRWST strainers was performed. The results showed that the debris amount and flowrate on each strainer module can be equal by using the single variable solution method, and the preliminary filtering area can be determined by calculating the pressure drop of IRWST strainers; the preliminary filtering area was verified by the debris pressure drop test, and the results satisfied the design requirements of safety systems.

Research of Margin Matching between Steam Generator and Steam Turbine in Nuclear Power Plant

Zhao Qingsen, Wang Shiyong

2021, 42(6): 174-178. doi: 10.13832/j.jnpe.2021.06.0174

Abstract(237) HTML (50) PDF(24)

Abstract:
Some of the nuclear power units of PWR newly put into operation in China have suffered from sharp shaking of the main steam control valve, slow pressure drop of new steam and other problems, which have certain impacts on the safety and economy of the unit. Through the comparison of the design parameters and the test data, and using the heat transfer calculation model to analyze, this study found that the design margin of 55/19B steam generator was lower than the similar 60F steam generator, and was also lower than the usual steam pressure margin of 1.6×10⁵ Pa. The new steam pressure of the steam generator, the main steam control valve and the flow area of the steam turbine governing stage shall be comprehensively matched at the beginning of the design, especially the valves with linear regulation characteristics in the opening range of more than 70% shall be selected, so as to ensure the reliability and economy of the nuclear power unit throughout its lifetime.

Acceptance Criteria Optimization of Stationary Gripper Opening Time of Control Rod Drive Mechanism

Zhang Hengkai, Zhao Yuntao, Shi Mengchen, Yang Wenqing, Wang Yu, Liu Jikun, Wei Qiao

2021, 42(6): 179-182. doi: 10.13832/j.jnpe.2021.06.0179

Abstract(95) HTML (50) PDF(19)

Abstract:
Based on the power supply timing sequences of control rod drive mechanism (CRDM), the action characteristics of stationary gripper in CPR1000 nuclear power unit and the test results in 10 units, the effect law of primary loop liquid fluid temperature, pressure and with or without CRDM on the opening time of the CRDM stationary gripper during unit commissioning start-up and normal operation is found, and based on this, the acceptance criteria range of CRDM stationary gripper opening time under different working conditions is optimized. The conclusion can not only be used as a reference for the acceptance criteria of CRDM stationary gripper opening time in the other units with the same type during commissioning start-up and normal operation, but also can be used as a reference for the analysis of the theoretical force model of the opening process of CRDM stationary gripper.

Research on Influence of Heating Power on Two-Phase Flow Characteristics in Natural Circulation System with Low Height Difference at Low Pressure

Sun Jianchuang, Li Feng, Ding Ming, Ran Xu, Yang Fan

2021, 42(6): 183-189. doi: 10.13832/j.jnpe.2021.06.0183

Abstract(128) HTML (53) PDF(28)

Abstract:
Two-phase flow characteristics of a kind of natural circulation system with low height difference and inclined heat pipe section in floating nuclear power plants are experimentally studied at low pressure. The influences of heating power on the two-phase flow characteristics are analyzed. The results show that there are two flow modes in the system under different power conditions: two-phase stable condensation and two-phase oscillation along with steam condensation induced water hammer. The backflow of subcooled water in the heat pipe section and the direct contact condensation of steam and low-temperature subcooled water are the internal mechanisms resulting in the two flow modes. In addition, the occurrence of steam condensation induced water hammer will generate large pressure pulses and cause a significant increase in the backflow length of subcooled water, which further aggravates flow instability. Further research shows that the gas content at the outlet of the heating section can be used as a basis for judging the flow instability.

Study on the Influence of Eccentricity Ratio on the Sealing Excitation Force Caused by Impeller Wear-ring of a Reactor Coolant Pump

Li Yibin, Pang Minchao, Wang Yan, Wang Xiuyong

2021, 42(6): 190-198. doi: 10.13832/j.jnpe.2021.06.0190

Abstract(279) HTML (109) PDF(39)

Abstract:
In order to explore the influence of rotor eccentricity ratio on the sealing excitation force caused by rotor of the reactor coolant pump, based on the Reynolds-averaged Navier-Stokes equations and RNG k-ε turbulence model, 3 types of wear-ring sealing schemes of plane seal, labyrinth seal and spiral seal are selected to numerically calculate the internal flow inside the wear-ring clearance of the reactor coolant pump, and the distribution laws of pressure, leakage and sealing excitation force inside the clearance of wear-ring are obtained. The results show the predicted performance of the model pump is in good agreement with the experimental value, and the maximum error of head is 4.78%. When the rotor has no eccentricity, compared with the plane seal, the leakage of the wear-ring by the spiral seal scheme can be significantly reduced by 93.1%, and the sealing excitation force can increase by 63%. When the eccentricity ratio is 10%, the pressure distribution of the wear-ring is more uniform along the circumferential direction; When the eccentricity ratio is 30%, there is a band pressure sudden rise zone in the clearance of wear-ring near the eccentric position in the circumferential direction. Compared with the non-eccentric scheme, the leakage of the plane seal is significantly reduced by 43.6%, and the sealing excitation force is increased by 4.4 times, labyrinth seal and spiral seal can significantly reduce the sealing excitation force caused by eccentric rotor, and labyrinth seal can significantly reduce by 55%; When the eccentricity ratio is 50%, the band pressure sudden rise zone within the wear-ring clearance tends to the high-pressure side; The numerical prediction method provides a theoretical basis for revealing the influencing factors of eccentric rotor on the sealing excitation force of the reactor coolant pump.

Study on the Influence of Pressure Pipe Deformation on Life Extension of QINSHAN CANDU-6 HWR

Xu Tongxi, Zhou Jianwei

2021, 42(6): 199-202. doi: 10.13832/j.jnpe.2021.06.0199

Abstract(167) HTML (85) PDF(20)

Abstract:
In order to prove the feasibility of the second stage life extension of pressure pipe in Qinshan CANDU-6 heavy water reactor (HWR), the influence of the size change (axial extension, radial expansion and wall thickness thinning) in the second stage life extension of pressure pipe is analyzed comprehensively by means of Tang deformation equation calculation and data fitting, and the evaluation criteria and evaluation conclusions are given for the factors of the change of pressure pipe size. The results show that the radial expansion of the pressure pipe may be the restraining factor for the life extension of the pressure pipe. In view of this restraining factor, this paper gives the related influence analysis and countermeasures. This study lays the foundation for the second stage life extension of the pressure pipe of Qinshan No. 3 Nuclear Power Plant.

Research on Fault Pattern Recognition Model of Nuclear Power Plant Water Pump Based on Frequency-Domain Data Attention Mechanism

Liu Ziming, Luo Neng, Ai Qiong

2021, 42(6): 203-208. doi: 10.13832/j.jnpe.2021.06.0203

Abstract(128) HTML (80) PDF(30)

Abstract:
In view of the common fault modes of nuclear power plant pump, such as abnormal vibration, friction and abrasion of rotor parts, etc., this paper uses frequency domain data of the acceleration signal on pump shell which is easiest to be obtained as input, proposes a new method for frequency-domain data attention mechanism which combines convolutional neural network and attention network, and establishes the recognition model of fault mode of nuclear power plant water pump. The results show that: Compared with the traditional methods, the water pump fault pattern recognition model based on frequency domain data as input and based on frequency domain data attention network algorithm has a shorter input data length and can effectively improve the efficiency of model training. The fault pattern recognition accuracy of the fault pattern recognition model on the test set is 100%, which is better than other fault diagnosis models based on deep learning algorithm, which proves the advantages of the method proposed in this paper.

Parameter Autoregression Algorithm-Based Early Warning Method for Critical Equipment in Nuclear Power Plants

Zhao Qingbing, Wei Shiyuan, Zhai Xiaofei, Lyu Yuanliang, Wang Zihu, Pan Fan, Zhao Tong

2021, 42(6): 209-214. doi: 10.13832/j.jnpe.2021.06.0209

Abstract(214) HTML (31) PDF(43)

Abstract:
A methodology based on the parameter autoregression algorithm is designed and developed for the early warning of critical equipment in nuclear power plants. The method innovatively introduces a parameter autoregression algorithm based on multi-dimensional time sequence data, which estimates the parameters under normal operation of the equipment and extracts the residual characteristics by comparing them with the measured values, thus realizing a dynamic threshold-based equipment condition monitoring mechanism. In addition, combined with the equipment mechanism, this study proposes and adopts the key concept of measurement point importance, and through the modeling of the core components of the equipment, the identification of the equipment operating state, the early warning of abnormal signs, the identification of faulty components and the generation of relevant alarm events are achieved. This study tests and validates the designed and developed method on the reactor coolant pump (hereinafter referred to as the main pump), the core equipment of AP1000 nuclear power unit. Through the analysis of the actual operation data and abnormal events of the main pump, compared with the existing equipment condition monitoring methods, the newly established early warning method for critical equipment can detect abnormal signs of relevant equipment at an early stage, produce early warning, and provide key information to assist engineers in fault analysis and localization, thus significantly shortening the time for fault diagnosis and troubleshooting, and greatly reducing the manpower input for critical equipment monitoring.

Analysis and Treatment on Control Rod Measured Position Fluctuation of A Nuclear Power Plant

Yang Wenqing, Liu Jikun, Huang Chuhao, Cheng Xiongwei, Wei Qiao

2021, 42(6): 215-218. doi: 10.13832/j.jnpe.2021.06.0215

Abstract(311) HTML (84) PDF(27)

Abstract:
Core reactivity control and shutdown margin are provided by control rods. The actual physical position of control rod is monitored by rod position measurement system. After equipment type change of rod control and rod position measurement system in a nuclear power plant, an issue that fluctuation of measured position exists even with a stable physical position occurred during the system commissioning. Based on the study of rod position measurement principle and post-processing circuit characteristics, combined with the analysis of existing data, this paper determines that the root cause of the rod position fluctuation is that the power supply of the detector's primary coil is different in frequency and in phase, and the parameter setting method of rod position detector is also given. By data acquisition on site and development of calibration data processing software for rod position detectors, the rod position fluctuation problem of the power plant was solved. At the same time, the hardware-level optimization suggestions are put forward for the root cause of measured position fluctuation.

Study on Helium Leak Detection System for Heat Transfer Tube of Steam Generator in Nuclear Power Plant

Yang Jiong, Lei Chunhui, Ma Xianhong

2021, 42(6): 219-224. doi: 10.13832/j.jnpe.2021.06.0219

Abstract(388) HTML (62) PDF(32)

Abstract:
This study introduces the principle and system composition of the in-service helium leakage detection system of the heat transfer tube of steam generator in a nuclear power plant, and simulates the leakage detection test of the heat transfer tube during the in-service overhaul of the steam generator in a nuclear power plant. The test results show that the optimal parameters can be set as follows: the helium concentration share of the secondary side of the steam generator is 30%; the pumping rate is 20 L/min; the secondary side pressure of the steam generator is 0.6 MPa; the system leaking point location error is within 0.5 m. The helium leakage detection system of heat transfer tube of steam generator studied in this paper can provide reliable technical guarantee for the safe and stable operation of domestic nuclear power plants.

Risk-Informed Analysis of Nuclear Power Plant Emergency Diesel Generator Returning to Factory for Overhaul

Wang Yiming, Gu Xiaohui, Li Youyi, Ma Chao, Deng Wei

2021, 42(6): 225-229. doi: 10.13832/j.jnpe.2021.06.0225

Abstract(169) HTML (58) PDF(23)

Abstract:
In order to support an emergency diesel generator of Tianwan Nuclear Power Plant Units 1 and 2 for returning to the factory for overhaul, this paper proposes an optimization plan for the unavailability recovery time of an emergency diesel generator, and adopts a combination of determinism and probability theory. The risk-informed approach analyzes the feasibility of the optimization plan. The analysis results show that the unavailability recovery time optimization of an emergency diesel generator meets the requirements of relevant regulations and guidelines and traditional engineering analysis, and the risk to the power plant is small and acceptable. An emergency diesel generator can be returned to the factory for overhaul.

Research on Influence of Air Gap and Contact Thermal Resistance on Thermal Safety of Container for Spent Fuel Dry Transfer

Zhu Linglin, Tang Qionghui, Chen Liutong

2021, 42(6): 230-236. doi: 10.13832/j.jnpe.2021.06.0230

Abstract(228) HTML (76) PDF(20)

Abstract:
The gap between the structural shell and the lead layer is one of the important paths for the transfer container to discharge decay heat, and the heat transfer between the two is affected by the contact thermal resistance. Based on the thermal safety assessment of the transfer container, the air gap with different thickness is set for the contact thermal resistance between the lead layer and the structural shell produced during the lead filling process, and the transient numerical simulation during horizontal transfer is carried out by using FLUENT software. The results show the contact thermal resistance generated by the air gap layer between the lead layer and the structural shell causes a significant temperature difference between the two. The temperature difference increases with the thickness of the air layer. Excessive temperature difference can easily cause the lead layer to overheat and lose the shielding safety function; In the design and manufacturing process of the transfer container, the optimization of the lead filling process shall aim to reduce the gap thickness between the lead layer and the structural shell, and enhance the fit degree between the two layers, so as to improve the thermal safety performance of the transfer container.

Study on Electrochemical Decontamination of ⁶⁰Co Contamination on the Stainless Steel Surface

Lu Yunyun, Cao Qi, Chen Yunming, Yang Yu, Dai Shuang, Xiong Wei, Wang Zhen

2021, 42(6): 237-243. doi: 10.13832/j.jnpe.2021.06.0237

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Abstract:
During the decommissioning of nuclear facilities and the ‘three wastes’ treatment, a large number of stainless steel metal parts will be produced, and the working environment and personnel will face the problems of potential radioactive contamination and exposure dose. Aiming at the problem of ⁶⁰Co contamination on the surface of stainless steel, a portable electrochemical test device was designed to remove the contamination on the surface of stainless steel effectively. The orthogonal experiment method was used to optimize electrochemical process parameters such as electrolyte concentration, current density and electrode spacing, and to verify the decontamination of ⁶⁰Co contamination on the surface of stainless steel. The results show that the electrochemical in-situ decontamination method established in this paper has the advantages of short decontamination time and high efficiency; Under the conditions of electrolysis time of 30 s, electrolyte of 10 mol / L nitric acid, current density of 0.3 A/cm² and electrode spacing of 0.4 cm, the decontamination efficiency of this method for ⁶⁰Co contamination on stainless steel surface can reach more than 99.9%, and the corrosion depth is more than 10 μm, which can reduce the contamination to the environmental background radiation level.

Validation and Evaluation of TRANTH Based on Pressurizer Safety Valve Bank Flow Test

Xu Qinglan, Qiu Zhifang, Yu Na, Zhou Ke, Chen Hongxia, Wu Peng, Chen Guo, Wu Guanghao, Yuan Peng

2021, 42(6): 244-247. doi: 10.13832/j.jnpe.2021.06.0244

Abstract(199) HTML (49) PDF(19)

Abstract:
Nuclear power autonomous transient calculation software TRANTH is used to design and analyze nuclear power plant safety. Relevant verification needs to be done after the completion of software development. Energy conservation and mass conservation in gas region and liquid region are considered in pressurizer model (one of the key models). Safety valve, relief valve, electric heater, spray and related simulation are simulated. Combined with the flow test data and software simulation results of pressurizer safety valve of Fangjiashan Nuclear Power Plant 1# unit, the pressurizer model is verified. The results show that the simulation results are in good agreement with the field test data, and the accuracy of the model meets the engineering design requirements.

Study on the Evolution of Dislocation Loop under Zirconium Alloy In-Situ Ion Irradiation

Pang Hua, Li Yipeng, Lyv Liangliang, Zhang Xiang, Zhao Yanli, Zhang Jibin, Peng Hang, Zhang Hongzhi, Sun Zhipeng, Chen Jie

2021, 42(6): 248-253. doi: 10.13832/j.jnpe.2021.06.0248

Abstract(298) HTML (42) PDF(38)

Abstract:
Dislocation loop evolution is one of the major features of the microstructure evolution of nuclear-grade zirconium alloy exposed to irradiation, which determines the mechanical properties (strength, plasticity, etc.) of the alloy after irradiation. To date, experimental researches on the irradiation dislocation loop evolution of zirconium alloy are mainly based on ex-situ neutron or ion irradiation, without a direct observation of the dislocation loop evolution process. To obtain an in-depth understanding of microstructure evolution of zirconium alloy under irradiation, the present work utilizes the advanced in-situ ion irradiation technique for a real-time observation of Zr-2 alloy dislocation loop evolution. As a result, the rule of irradiation dose and temperature dependence of the evolution process has been revealed, and the irradiation hardening effect has been evaluated by applying the dispersed barrier hardening model, the feasibility and advancement of the in-situ ion irradiation technique on studying dislocation loop evolution and mechanical property evaluation of zirconium alloy cladding materials after irradiation is demonstrated.

Research on In-Pile Thermo-Mechanical Performance for U-10Mo/Zr Monolithic Fuel Element under Steady Condition

Guo Zixuan, Jian Xiaobin, Li Wenjie, Zhang Kun, Wang Peng, Wang Yanpei

2021, 42(6): 254-260. doi: 10.13832/j.jnpe.2021.06.0254

Abstract(252) HTML (91) PDF(66)

Abstract:
In the paper, the models of irradiation performance and thermo-mechanical constitutive relations of U-10Mo/Zr monolithic fuel element were established. With the finite element method, numerical simulation of fuel element thermo-mechanical performance under steady heterogeneous irradiation condition was conducted, the distribution and evolution characteristics of temperature, strain and stress in U-10Mo/Zr monolithic fuel element were acquired and analyzed. The results showed that the thickness increment of the fuel pellet becomes largest near the interface between fuel pellet and cladding, predominantly affected by fuel irradiation creep. Under the low burn-up conditions, the simulation result of fuel pellet swelling at high temperature condition equals the irradiation test results at low temperature. There are stress concentrations in the corner area of the fuel pellet and the outer area of the cladding end surface.

2021 Vol. 42, No. 6