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2018 Vol. 39, No. 1

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Research on Three Dimensional Discrete Ordinate and Monte Carlo Hybrid Method
Zheng Zheng, Ding Qianxue, Zhou Yan
2018, 39(1): 1-5. doi: 10.13832/j.jnpe.2018.01.0001
Abstract(12) PDF(0)
Abstract:
For deep-penetration shielding calculation, Monte Carlo(MC) method requires the modeling of a great number of particles to obtain reliable results, thus huge computation time is the main problem of the MC method. Source biasing and weight window technique effectively decrease the tally error of deep penetration problem. This paper studies the 3D Discrete Ordinate(SN)-MC hybrid method, generates source biasing factors and weight window parameters for the MC method by using the adjoint fluence rates of the SN method, develops source sampling subroutine for MCNP, and verifies the hybrid method in the measurements in Qinshan Nuclear Power Plant I. The method is then applied in the calculation of CAP1400 reactor pressure vessel fast neutron fluence rate. Numerical results show that 3D SN-MC hybrid method increases the calculation efficiency by 1 to 2 orders for deep-penetration shielding calculation with high precision compared with unbiased MC method.
Temperature Analysis for Metal Reflector Based on Irradiation and Thermal-Fluid-Structure Coupling
Chen Shengjie, Fang Jian, Wu Xianmin, Shi Lin, Ran Xiaobing
2018, 39(1): 6-11. doi: 10.13832/j.jnpe.2018.01.0006
Abstract(15) PDF(0)
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The metal reflector is influenced by irradiation and heat flow in the reactor pressure vessel, thus its temperature analysis is a complicated thermal-fluid-structure coupling problem. In order to simulate the temperature distribution of the metal reflector more accurately, the heat generation rate caused by irradiation is taken into account. Then a method based on thermal-fluid coupling and thermal-structure coupling is proposed in this paper. The method utilizes an iteration of fluid steady-state thermal analysis and structure steady-state thermal analysis to compute the temperature distribution of the metal reflector and the natural convection heat transfer coefficient. Compared with the results of the existing method, the results of the method in this paper is more reliable for more boundary conditions are applied in the calculation.
Thermal-Hydraulics Analyses of Radial Flow Core of Molten Salt Pebble-Bed Reactor
Xue Chunhui, Dong Yujie
2018, 39(1): 12-16. doi: 10.13832/j.jnpe.2018.01.0012
Abstract(14) PDF(0)
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Nuclear Hot Spring is a novel design concept of molten salt pebble-bed reactor, featured by the full power natural circulation, with the primary coolant flowing radially through the reactor core. By using CFD software Fluent, the thermal-hydraulics of the radial flow reactor core of the NHS were obtained based on the local thermal non-equilibrium(LTNE) porous media model. Different opening ratios of the central and outer perforated plate were simulated. The simulation results indicated that the opening ratio of the central perforated plate has a great effect on the flow field distribution of the coolant; the maximum pebble center temperatures are far below the safety limit of the fuel, the core resistance is considerably less than the buoyant force, indicating that the natural circulation under full power operation is possible.
Optimization of Method for Finding Doppler Heating Starting Point in Qinshan Nuclear Power Plant Ⅱ
Liu Zhen, Yang Si, Wang Chenghan
2018, 39(1): 17-19. doi: 10.13832/j.jnpe.2018.01.0017
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Through analyzing the physical parameters and phenomenon during the tracing of Doppler heating starting point, the paper optimize the test method for finding Doppler heating starting point in Qinshan Nuclear Power Plant Ⅱ. The analysis found that when finding Doppler heating starting point, it is more suitable to use the signal of hot leg temperature of coolant temperature measuring bypass, and neutron flux rate logarithmic signal. In addition, it can avoid the misjudgment effectively in finding the Doppler heating starting point in the course of neutron ascending, and confirming once more in the course of neutron descending.
Analytic Method Study of Neutron-Kinetic Equations in Molten Salt Reactor
Liu Guocai, Wang Kaikai, Zhang Haiqian
2018, 39(1): 20-23. doi: 10.13832/j.jnpe.2018.01.0020
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Abstract:
The point kinetic equation of reactor can reflect the change of neutron density in reactor core with time, but the traditional point reactor kinetic equation is not suitable for molten salt reactor. This is due to the fluidity of the fuel in the molten salt reactor and the delayed neutron precursors generated by the core as the flow of molten salt affects the core neutron balance. In this paper, the modified point kinetic equation is used to derive the analytical solution of the equation. The numerical method is used to verify the analytical solution. The results show that the analytical solution and the numerical solution are basically accordant. According to the analytic solution, the expression of the cycle of the molten salt reactor is deduced and validated by the numerical solution. The results of analytical solution and numerical solution are basically in agreement.
Research of Next Event Estimation for Reflecting Surfaces
Wang Shuang, Shen Huayun, Zhong Bin, Pan Liujun
2018, 39(1): 24-27. doi: 10.13832/j.jnpe.2018.01.0024
Abstract(15) PDF(0)
Abstract:
This paper presents a modified next event estimation method, which could be used with reflecting surfaces. The point flux and the flux image by pinhole calculated by the Monte Carlo transport code NPTS are based on this new method. Several numerical cases with reflecting surfaces are chosen to verify the new method and the new functions of NPTS in this paper.
Thermal-Hydraulic Numerical Study of 3D Model Subcritical Energy Blanket
Wang Xi, Shi Xueming
2018, 39(1): 28-33. doi: 10.13832/j.jnpe.2018.01.0028
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Thermal hydraulic process without and with helium cooling first wall of 3D model light water cooled subcritical energy blanket driven by ITER is numerically studied with 3D distributions of fuel power density and power flattening via turbulence simulation with RNG k-ε model and fluid/solid heat transfer coupling method. The present results show that cooling the first wall with helium can significantly reduce the maximum temperatures of the solid material. The highest temperature distributions of fuel have the same variation trends as the fuel power density. Power flattening is beneficial to equally distribute the mass flow in the cooling channels. A thermal hydraulic design with enough thermal margin is obtained in this paper.
SIRIUS: A Code for Thermal Neutron Scattering Data Production for Solid Moderators
Wang Jia, Hu Zehua, Song Hongzhou, Ye Tao, Sun Weili
2018, 39(1): 34-37. doi: 10.13832/j.jnpe.2018.01.0034
Abstract(13) PDF(0)
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Based on the coherent elastic scattering, incoherent elastic scattering, coherent inelastic scattering and incoherent inelastic scattering, a code named SIRIUS is developed to generate the thermal neutron scattering data in ENDF-6 format for solid moderators, and the ab-initio calculation is used to calculate the phonon densities of states for the these solids. The TSL data for the metal 27Al as a FCC structure are generated by SIRIUS code, and are consistent with the data from ENDF/B-VII.1 library, which validate the correctness of the SIRIUS code.
Investigation on Effect of Cool/Hot Core Deviation on Natural Circulation Capacity in Natural Circulation Loop
Peng Yinbo, Zhang Yajun, Jia Haijun, Wu Lei, Liu Yang
2018, 39(1): 38-42. doi: 10.13832/j.jnpe.2018.01.0038
Abstract(11) PDF(0)
Abstract:
Based on the experimental phenomena, the effect of temperature distribution on natural circulation capacity is analyzed theoretically, and the expressions about the influence function of distribution are obtained. It is indicated that the heat transfer between upward section and downward section and the parabola temperature distribution in the heat source and heat sink have certain but not heavy influence on the natural circulation capacity in the natural circulation loop. The deviation of natural circulation caused by these two reasons is a function of geometric structure of natural circulation system, temperature increase in the upward section, the thermal expansion coefficient and the type of the parabola temperature distribution, and the effects of different reasons are separated. When the temperature increases or decreases in the upward section or downward section ΔT<1~20 ℃, the deviation of natural circulation capacity is limited in ε%<0.2%~4%.
Experimental Investigation on Critical Heat Flux in Horizontal Tube
Li Haoxiang, Peng Chuanxin, Zan Yuanfeng
2018, 39(1): 43-46. doi: 10.13832/j.jnpe.2018.01.0043
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The critical heat flux(CHF) in the horizontal tube under low flow condition is investigated in this paper. It is found that the boiling crises occur at the top of the tube. According to the exit quality and flow pattern, the category of boiling crisis in the horizontal tube low flow is considered to be dryout. The prediction results of Bowring correlation and Lookup table are much larger than the experimental data, because the Bowring correlation and Lookup table are proposed for the vertical round tube CHF prediction. The distribution of liquid film in the horizontal annular flow is asymmetrical due to gravity. It causes the dryout of liquid film at the top of tube early and the decreasing of CHF.
Experimental Study on Critical Heat Flux in Tight Arrangement Fuel Aassembly
Xie Feng, Xu Jianjun, Huang Yanping, Yang Zumao, Wang Hongtao
2018, 39(1): 47-50. doi: 10.13832/j.jnpe.2018.01.0047
Abstract(12) PDF(0)
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Based on the research of tight arrangement reactor, experiment investigation on Critical Heat Flux(CHF) for tight arrangement spiral-fin fuel assembly was performed. Results show that the CHF happens on the hot rod of fuel assembly, and when the CHF happens, the temperature of the heated wall will increase quickly, and the pressure will increase and the massflow rate will decrease at the same time. The pressure of the system, the mass flow rate, the quality and degree of subcooling have vital effects on CHF; The CHF relation of the tight arrangement spiral-fin fuel assembly was obtained based on the test data. The deviation between the proposed correlation and experimental data is less than 10%.
A Prediction Model for Liquid Film Thickness and Heat Transfer Coefficient of Annular Flow in a Narrow Channel with Uniform Heating
Wu Lianwei, Meng Qingzheng, Chen Chong, Wu Wei, Liu Dongmin
2018, 39(1): 51-55. doi: 10.13832/j.jnpe.2018.01.0051
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In order to research the heat transfer characteristics of annular flow in the rectangular narrow channel, a mathematical model of annular flow in rectangular narrow channel is developed based on the mass, momentum and energy conversation equations of liquid film, as well the momentum conversation equation of vapor core. Through numerically solving the closure equations, the boiling heat transfer coefficients of annular flow in the rectangular narrow channel is obtained, and the effects of heat flux, mass flow rate and rectangular channel size on the liquid film thickness are analyzed. The results show that the present model can well predict the two-phase heat transfer coefficient, and the deviation is within ±30%. The heat flux and rectangular channel size have a big effect on the liquid film thickness.
Evaluation and Correction of Low-Pressure Subcooled Flow Boiling Model for Athlet Code
Li Fei, Liu Xiaojing, Shen Feng
2018, 39(1): 56-60. doi: 10.13832/j.jnpe.2018.01.0056
Abstract(16) PDF(0)
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A new model for subcooled flow boiling at low pressure has been proposed. The model considers the closure relationships of one-dimensional thermal-hydraulic codes that are important for void fraction in the channel: wall evaporation model and drift-flux model. The new model was incorporated in the current version of the ATHLET code, MOD 2.1 A. The new modified code was validated against a number of published low-pressure subcooled boiling experiments, and in contrast to the current code, shows good agreement with the experimental data.
Study on Heat Exchange Experiment Characteristics of SPRHR Condenser for Nuclear Power Plants
Yang Jinchun
2018, 39(1): 61-63. doi: 10.13832/j.jnpe.2018.01.0061
Abstract:
In order to meet the safety requirement of nuclear power plants, the residual heat must be removed by the passive system under severe accidents. An experiment is conducted on the heat exchange characteristics of Secondary Passive Residual Heat Removal Condenser(SPRHR condenser) under different vapor pressure and temperature conditions. The results show that the heat exchange characteristics of SPRHR is stable and with enough safe margin. The heat exchange characteristics of SPRHR can be predicted well by Shah and Foster-Zuber correlations.
Assessment of CATHARE Code against DVI LOCA Experiment
Peng Chuanxin, Li Haoxiang, Zan Yuanfeng, Yan Xiao
2018, 39(1): 64-68. doi: 10.13832/j.jnpe.2018.01.0064
Abstract:
The direct vessel injection(DVI) line loss of coolant accident(LOCA) experiment is simulated by CATHARE code in this paper. The main processes and physical phenomena under DVI LOCA, such as system depressurization, passive injection and the cooling of core, are well simulated. The calculation results of primary system pressure, Core Makeup Tank(CMT) injection flowrate, Accumulator(ACC) injection flowrate, In-containment Refuelling Water Storage Tank(IRWST) injection flowrate and core fluid temperature agree well with the experimental data. It shows that the CATHARE code can be used to analyze the transient behaviors of passive injection system under LOCA.
Multi-Objective Design Optimization for A Conceptual Passive Containment Cooling System with Closed-Loop Configuration
Bai Jinhua, Zhao Bo
2018, 39(1): 69-74. doi: 10.13832/j.jnpe.2018.01.0069
Abstract(11) PDF(0)
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In this research, a set of models were set up and corresponding codes(PCCS-CL) were developed for the prediction of the performance of proposed conceptual passive containment cooling system(PCCS) based on one-dimensional homogeneous two phase flow model. An improved non-dominated genetic algorithm(INGA) was developed with the sorting algorithm, improved crowding-distance and optimum retention strategy. The multi-objective design optimization for the proposed PCCS was conducted by using developed codes of INGA. The sensitive study on key parameters show that the diameter of heat transfer tube both for in-containment heat exchanger and ex-containment heat exchanger plays a critical role for the heat removal capacity of PCCS. For the range of parameters in this paper, either reducing the inside diameter or increasing the length of heat transfer tube is helpful for improving the heat removal capacity of the system. The optimized scheme given in this study might provide references for the engineering design of PCCS.
Engineering Problems of EGU Process for HTGR Gel Microspheres
Hu Fengqi, Niu Xiaoping, Deng Changsheng, Ma Jingtao, Hao Shaochang
2018, 39(1): 75-78. doi: 10.13832/j.jnpe.2018.01.0075
Abstract:
China North Nuclear Fuel Co. Ltd(CNNC) built a HTGR nuclear fuel element production line. In this production line, the gel ball is mainly prepared by external gelation of uranium process(EGU). In the engineering research process, the problems such as the paste liquid deposition and the gel liquid flow can not be controlled accurately during Sol-gel, and gel ball cracks occur. In order to solve these engineering problems, the gel ball preparation process and equipment were optimized and modified, and 10 batches of UO2 kernels were tested and verified. The results show that the improved production line can achieve continuous and stable industrial production, and the qualified UO2 kernel products are over 80%.
Study on Fracture Properties of Nuclear Graphite
Chen Hongniao, Su Qiliang, Chen Jing, Xiao Jianchun
2018, 39(1): 79-83. doi: 10.13832/j.jnpe.2018.01.0079
Abstract:
Three-point bending tests were performed on the single-edge notched beams made of IG11 graphite to investigate the fracture properties of nuclear graphite. Electronic speckle pattern interferometry(ESPI) technique was used to measure the displacement fields of the graphite beams. Experimental results indicated that the initiation fracture load of graphite beams was 680~838 N and peak load was 845~974 N. At the peak, crack mouth opening displacement was 0.088~0.091 mm, crack tip opening displacement was 0.016~0.018 mm and the crack length was about 25 mm. Furthermore, the initiation fracture toughness and unstable fracture toughness according to the double-K fracture model of concrete were determined as 0.96~1.19 MPa·m1/2 and 1.61~1.85 MPa·m1/2 respectively. Elastic modulus was evaluated as 10.22 GPa according to linear elastic fracture mechanics. By analyzing the phase maps at different loading stages, fracture process zone of graphite was identified as square. At the pre-peak stage, the length of the FPZ did not exceed 3 mm; and at the post-peak stage, the length of the FPZ was within 5-8 mm.
Design and Application of Alarm System for Diversification Liquid Level of Radioactive Waste Storage Tank
Zhang Jiaheng, Zhang Jingsong, Li Wenyu, Li Zhengchen, Huang Wentao, Dai Jun, Li Hong, Liu Wei
2018, 39(1): 84-87. doi: 10.13832/j.jnpe.2018.01.0084
Abstract(11) PDF(0)
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In order to prevent the radioactive material leakage, which causing the risk of environment pollution, the liquid level measurement and alarm system is used. Through the introduction of incremental alarm technology, and combined with the traditional technology, a diversification liquid level alarm system is designed. Test data and practical results show that the alarm system is powerful, reliable and accurate, and it can automatically record the alarm data, and alert the operator to take appropriate measures at any liquid level of storage tank, which prevent the leakage of radioactive material and pollution of the environment.
Manufacture of Nuclear Safety Valve Mock-Up and Qualification Requirements
Shi Hong, Deng Dong, Xiong Dongqing, Lyu Yanxin, Zhao Libin
2018, 39(1): 88-92. doi: 10.13832/j.jnpe.2018.01.0088
Abstract(18) PDF(0)
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The nuclear safety class valves are very important in nuclear power plants, and the focused concern of National Nuclear Safety Administration. According to requirements of relative nuclear safety regulations, mock-ups must be made by applicants, at the same time, all corresponding qualification tests must be carried on. But no detailed requirements were given on how to choose the mock-up size and what qualification tests should be carried on. Combined with safety review practices, some basic requirements on how to choose the mock-up size, how to do the preparation work, how to control the process of manufacture and some mistakes may be made in qualification test are presented in this paper for applicants and nuclear safety technical reviewer to refer.
Improvement of Protection System T2 Tester in Ling’ao Nuclear Power Station
Xiong Guohua, Fang Yu
2018, 39(1): 93-96. doi: 10.13832/j.jnpe.2018.01.0093
Abstract:
Based on the principle of T2 test for reactor protection system, combined with the requirements of T2 test technology, using programmable logic controller and a friendly interactive interface, a new technical scheme of the tester was presented. The successful and practical application in Ling’ao nuclear power plant shows that the new T2 tester is agile in controlling, convenient in operation and high in reliability, and improves the safety and reliability of the reactor protection system.
Leak Problem Analysis and Detection Plan for Pressurizer Electric Heater
Hou Ye, Yu Ping, Zhou Yong, Huang Wei
2018, 39(1): 97-101. doi: 10.13832/j.jnpe.2018.01.0097
Abstract:
Electric heater is one of the important functional units of the pressurizer, and the electric heater sheath broken happened both in China and other countries. The broken is usually associated with leak. Leak will result in water penetration and electric heater will swell, thus the mechanical failure happens. It will cause the leak of the primary coolant circuit. A faulty insulation problem had been found during the hot functional test in a nuclear power project, and the leak flaw was found by test and analysis. In order to avoid the sheath broken, analysis and test have to be conducted to solve the problem. Researches show that only PT at welding is not enough, and it is risky. LT(He) is both important and RT is also needed at welding part. A new plan of none destructive test has been proposed. It will be favorable for the sealing performance of the pressurizer electric heater.
Research on New Conceptual Design of PWR Core Catcher
Han Xu, Jing Chunning, Zhu Chen, Wang Yiguang, Li Jun, Yuan Yidan
2018, 39(1): 102-105. doi: 10.13832/j.jnpe.2018.01.0102
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Based on the experience of Fukushima accident, this paper analyzed two typical core catcher designs, from EPR and WWER, and put forward three new conceptual schemes, i.e., the core grouping catcher, the Waffle core catcher, and the casting core catcher. The conclusion shows that the new schemes have obvious technical advantages.
Application of Fuzzy Control Method for Temperature Control of Molten Salt System
Ruan Jian, Zou Yang, Zhu Haihua, Zhang Jie, Li Minghai, Xu Hongjie
2018, 39(1): 106-111. doi: 10.13832/j.jnpe.2018.01.0106
Abstract:
Based on the nitrate salt experimental facility, the original PID controller of the facility is optimized and a fuzzy controller is studied to improve the temperature control performance of the experimental loop. The improved PID controller has shorter adjustment time and smaller overshoot. Besides, compared with PID control, the fuzzy control method can not only meet the desired control targets and demands, but also can be used to control the temperature in large inertia molten salt systems.
Design of Self-Diagnosis Scheme for ACPR1000 NPP RPS Based on Firm Sys
Qi Min, Mo Changyu, Xie Yiqin, Shi Guilian
2018, 39(1): 112-116. doi: 10.13832/j.jnpe.2018.01.0112
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This paper presents the design principle and scheme for ACPR1000 NPP RPS with a self-diagnosis function, based on Firm Sys, including Firm Sys platform fault diagnosing, diagnostic information transmitting and fault indicating. Test result showed that the self-diagnostic function can successfully transmit and indicate the fault information, as a result of that the self-diagnosis function can supply sufficient information for decision making on routine maintenance and emergency situation for ACPR1000 NPP, to satisfy the specifications, regulations and application requirements in nuclear power industry.
A Machine Learning Based System Performance Prediction Model for Small Reactors
Zeng Yuyun, Liu Jingquan, Yang Chunzhen, Sun Kaichao
2018, 39(1): 117-121. doi: 10.13832/j.jnpe.2018.01.0117
Abstract(23) PDF(0)
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To support the development of autonomous control system for the Transportable Fluoride-salt-cooled High-temperature Reactor(TFHR), a machine learning based reactor system performance prediction model is created. The prediction model consists of a reactor physics model and a thermal-hydraulic model, which are formulated using support vector machine(SVR) with training data generated by a RELAP model of TFHR primary system. A particle filtering method is used to estimate the model parameters with noisy instrument measurements. Verifications of the proposed models have been conducted using TFHR reactivity insertion events. Satisfactory performance in predicting the core behavior and estimating model parameters are concluded.
Numerical Study on Core Transient Heat Transfer and Melt Process after LOCA
Liu Yiqun, Zhang Xiaoying, Wang Biao, Xu Junying, Zhang Lei, Zhang Huiyong, Zhan Dekui
2018, 39(1): 122-127. doi: 10.13832/j.jnpe.2018.01.0122
Abstract(16) PDF(0)
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Relevant parameters of a pressurized water reactor are obtained to establish the 3D geometric model and two-dimensional heat conduction differential equation. To calculate the temperature distribution and melting process, the natural convection heat transfer model and radiation heat transfer with neighbor 8-rod-cavity model are applied. The study shows that with the accident processes, the water level in the core decreases, but the temperature increases, and the point with the highest temperature in the core goes down gradually. After 560 seconds, the control rods begin to melt, and after 1200 seconds, the stainless steel rods begin to melt; the fuel rods begin to melt after 2700 seconds. About 7000 seconds later, the proportion of core melt is more than 50%. When most of the core nodes melted, the shroud didn’t melt. The melt migrated directly down the lower plenum. Both the steam vapor and radiation can influence the melting time of the fuel rods, and the steam convection heat transfer is dominant. The effect of steam vapor cannot be neglected. Radiation heat transfer has the effect of flattening the temperature of the core.
Analysis and Improvement of Turbine Trip Acknowledge Signal Spurious Triggering Risk in Nuclear Power Plants
Shen Chao, Wang Yuan, Xiong Guohua, Wang Hongtao, He Wenkai, Zhao Youyou
2018, 39(1): 128-131. doi: 10.13832/j.jnpe.2018.01.0128
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Take the example of Ling’ao Nuclear Power Station II, the risk of turbine trip acknowledge signal produced by IP stop valves has been discussed. And the paper proposes three improved suggestions to put forward some proposals to the same problem for reference.
Research and Implementation of Radiation Field Visualization in Reactor Cabin and Absorbed Radiation Dose Calculation
Zhang Kaiyun, Qin Lihua, Hu Shiyuan, Kang Qihao, Wu Hong
2018, 39(1): 132-135. doi: 10.13832/j.jnpe.2018.01.0132
Abstract(14) PDF(0)
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The function of the visualization of radiation field in the reactor cabin and simulation calculation of human body absorbed radiation dose is developed based on the DELMIA software. The reactor cabin radiation intensity distribution is displayed visually by radiation field visualization, which can assist the work of personnel maintenance, reduce the radiation dose of the staff, and improve the work efficiency. Through the simulation calculation of the human body absorbed radiation dose, the human body damage degree can be understood, so as to provide the necessary radiation protection measures for the maintenance staff. Based on the analysis of statistical reports output, the maintenance process can be evaluated and optimized, and this plays an important role in the actual reactor maintenance engineering. The calculation result is fast and real-time, and the display effect is good.
Research on Condition Monitoring Technique of Sensors in NPPs Based on PCA
Li Wei, Peng Minjun, Liu Yongkuo
2018, 39(1): 136-139. doi: 10.13832/j.jnpe.2018.01.0136
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To solve the shortcomings of physical redundancy methods, Principal Component Analysis(PCA) is adopted. Firstly, PCA can be used to verify the monitoring results of physical redundancy method. Secondly, PCA method can detect the small drift of sensors which physical redundancy method can hardly deal with. Finally, PCA method can detect the common mode faults in the redundant sensors. At the end of this paper, sensor measurements from a real NPP are used to train the PCA model. Artificial failures are imposed to the original measurements. Simulation results show that the PCA method has good effects on the issues that have been raised above.
A Scheduling Algorithm for Information Navigation of Secondary Task of Digital Human-Computer Interface in a Nuclear Power Plant
Jiang Jianjun, Zhang Li, Wang Yiqun, Xie Tian, Li Min, Dai Licao, Peng Yuyuan
2018, 39(1): 140-145. doi: 10.13832/j.jnpe.2018.01.0140
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To relieve operator workload that is caused by information navigation picture configuration of digital human-computer interface secondary tasks in a nuclear power plant, this paper proposed an information navigation scheduling algorithm for secondary task in order to resolve the configuration of navigation information pictures. By establishing an executing process of scheduling algorithm, mathematic model, feature extraction sub-algorithm of information navigation and a sub-algorithm to sum the same keywords, the scheduling algorithm of navigation information picture configuration is obtained. The simulation experiment results show that the algorithm has good performance on average cycling time and correctness. It can be applied to allocate the navigation information pictures so that the cost to manually allocate pictures by operators can be decreased.
Research on Knowledge Base of Intelligent Diagnosis Expert Based on Tubing Leakage of High-Pressure Heater in Nuclear Power Plants
Zheng Miao, Qian Hong, Lin Siyun, Xiao Bole, Chu Xiaoping, Fei Minrui, Chen Kai
2018, 39(1): 146-151. doi: 10.13832/j.jnpe.2018.01.0146
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In order to improve the accuracy and timeliness of the tubing leakage of the high-pressure heater in nuclear power plants, the fault diagnosis is carried out with the terminal difference triggered by the heat economy of the unit. Through the mechanism modeling of the tubing leakage of the high-pressure heater, the set of the symptom parameters related to the fault is obtained. The expert knowledge base of the tubing leakage in the fault diagnosis system of the high-pressure heater is analyzed by using the mathematical statistics and the experience of the on-site experts. Through the insert of man-made fault in the 1000 MW nuclear power model, the intelligent diagnosis expert system of nuclear power plant based on terminal difference is used to fault diagnosis. The results show that the method can accurately diagnose the tubing leakage of the high-pressure heater at the beginning of the fault, which proves the validity and feasibility of the knowledge base.
Theoretical Analysis of High Temperature of Mechanical Seal Water in Residual Heat Removal Pump
Li Bo, Zhao Liang, Sun Min, Fan Ruibo, Xu Dezhong, Kong Lingjie, Niu Hongjun
2018, 39(1): 152-156. doi: 10.13832/j.jnpe.2018.01.0152
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On the hot standby condition of the residual heat removal pump in a nuclear power plant, the temperature at the measurement points of the mechanical seal water exceeds the alarm value. To deal with this issue, the root cause analysis(RCA) tree is established. Methods of CFD analysis and classical thermodynamics theory are implemented. The study found that in the radial position, the heat shield has greater effect on the temperature in the cooling chamber, which is ±0.458℃/mm. The cooling chamber size have the optimize value for temperature of mechanical seal water. And when the cooling water rate is less than 5 m3/h, it increase is obviously for this issue. Study found the sleeve clearance is the major factor, when the value change 0.1 mm, the temperature of mechanical seal water reduced 14℃. The ambient temperature increased from 16℃ to 45℃, the temperature rise 0.13℃.
Fuel Management Strategy Analysis of Small Modular Pressurized Reactor ACP100
Wang Liangzi, Ju Haitao, Qin Dong, Wang Lianjie, Yu Yingrui, Li Qing
2018, 39(1): 157-160. doi: 10.13832/j.jnpe.2018.01.0157
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In order to achieve 24-months cycle length for Advanced Small Modular Pressurized Reactor(ACP100), and keep a balance between good fuel utilization and flat power distribution to assure reactor safety, using the well-developed software applied in engineering practice, this paper construct cases with combination of different number of assemblies per batch, enrichment types, and radial loading patterns; characteristics analysis of these cases could be referred during the confirmation of fuel management strategy of ACP100 reactor, and a recommended strategy is attained: combination of 3 batch loading and 24 assemblies per batch, along with partial low leakage loading pattern, and a higher enrichment to improve fuel economics.
Engineering Applicability Strengthening Design and Practice of NESTOR Software Package
Lu Zongjian, Li Qing, Liu Dong, Chai Xiaoming, Fang Haoyu, Gong Zhaohu
2018, 39(1): 161-164. doi: 10.13832/j.jnpe.2018.01.0161
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The engineering applicability strengthening design work has been carried out from three aspects of software architecture, research and development process and calculation model for NESTOR-a software package for the nuclear power design and analysis for HPR1000 and other three generation nuclear power units. The practice proves that the design work can reduce the deviation of software development and engineering application demand, and improve the initial success rate of the development of engineering design and analysis software.
Simulation of SA508-3 Steel Cyclic Plastic Deformation Based on Damage-Coupled Constitutive Model
Zhang Liping, Tian Jun, Li Jian, Yang Yu, Kan Qianhua
2018, 39(1): 165-168. doi: 10.13832/j.jnpe.2018.01.0165
Abstract(12) PDF(0)
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This paper simulated the cyclic cumulative plastic deformation of SA508-3 steel based on damage-coupled constitutive model at 200℃ temperature. Firstly, a series of uniaxial tensile and cyclic loading experiments were conducted to determine the parameters of the damage-coupled constitutive model. Based on the damage-coupled constitutive model, monotonic loading and cyclic plastic deformation were simulated for SA508-3 steel. Compared to the simulation results by Chaboche model, the damage-coupled constitutive model can give a better simulation for monotonic tensile behavior and cyclic softening behavior of 508-3 steel under strain and stress control, agree well with the experimental results, and build the foundation for the cumulative plastic deformation simulation of nuclear equipments made with 508-3 steel.
Elimination of Rh SPND Detector Signal Noise in Low-Level Power
Li Kun, Han Wenxing, Yin Qiusheng, Weng Xiaohui
2018, 39(1): 169-172. doi: 10.13832/j.jnpe.2018.01.0169
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Abstract:
According to the frequency domain of Rh SPND, this paper designs an IIR digital filter method for eliminating the noise of Rh SPND. This method improves the signal-to-noise ratio of the Rh SPND output signal and advances the lower limit of the in-core measurement system.
Analysis of Core Boron Concentration Following LOCA Accident for Small Module Reactors
Ding Shuhua, Dang Gaojian, Li Zhe
2018, 39(1): 173-176. doi: 10.13832/j.jnpe.2018.01.0173
Abstract(11) PDF(0)
Abstract:
This paper is an evaluation of the potential buildup of boron in the core following the loss of coolant accident for small module reactors. The controlling equation for boron concentration is derived base on the mass conservation of the boron. The potential for boron buildup is evaluated in the short term before automatic depressurization system actuation as well as in the long term during containment recirculation. The results indicate that the passive safety injection systems could prevent the boron from building up significantly in the core and the core from returning to criticality during the long term containment recirculation.