核动力工程

Evalution and Application of Probabilistic Safety Assessment in Nuclear Energy Safety Analysis

Yu Hongxing, Wu Lingjun, Deng Chunrui, Deng Jian, Lu Yili, Zhang Hang, Peng Huanhuan, Wang Xiaoji

2020, 41(6): 1-7.

Abstract(675) PDF(382)

Abstract:
Probabilistic safety assessment (PSA) is one of the two major safety analysis methods in the field of nuclear energy safety analysis. Starting from the concept of PSA, this paper firstly compares the difference between deterministic safety assessment and probabilistic safety assessment from the theoretical basis and analytical perspective. Secondly, it combs the historical process of PSA in the field of nuclear energy safety analysis, and through reviewing the changes in technology and regulations of PSA, it shows the mutual promotion relationship between PSA and nuclear energy safety in the process of upgrading. Thirdly, the application of the PSA method in risk quantitative prediction, balance safety design, risk safety-making, and risk safety supervision is expounded, and the application of PSA in nuclear energy safety assessment is illustrated by HPR1000. Finally, the future development direction of PSA technology is predicted. It points out that the deterministic safety assessment and probabilistic safety assessment will deeply integrate, and the PSA analysis will shifts from security target to task target, from static to dynamic, and from cognitive to perceptual direction.

Numerical Simulation of Steady-State and Transient Characteristics of Once Through Steam Generators

Zhao Xiao, Bai Yufei, Zhang Zhen, Yang Xingtuan

2020, 41(6): 8-13.

Abstract(328)

Abstract:
In once through steam generators, the secondary fluid is heated by the primary fluid. Therefore, the axial heat flux distribution is related to the operational parameters of the fluids in both sides. The RELAP5/MOD4.0 code is adopted to perform the static and transient simulations of a once through steam generator. The temperature, thermal equilibrium quality and heat flux distributions under both heating modes are analyzed and compared. The oscillation behavior of the once through steam generator is discussed and the instantaneous axial profiles of secondary mass flow rate, heat flux as well as tube wall temperature corresponding to four representative times in an oscillation period are further investigated. Finally, the effect of increasing the secondary inlet throttling is studied. The results show that the shape of heat flux distribution under convective heating mainly depends on the temperature difference between primary fluid and secondary fluid. When density wave oscillations occur in the once through steam generators, the primary outlet temperature and total heating power oscillate intensely. The location of the dryout point moves greatly and the tube wall temperature fluctuates most violently in its vicinity. Flow oscillation could be suppressed by increasing the inlet throttling of the secondary side.

Development and Verification of Coolability Analysis Program for Debris Bed of Sodium-Cooled Fast Reactor in Long-Term Cooling Phase

He Lianren, Zhang Bin, Teng Chunming, Shan Jianqiang

2020, 41(6): 14-18.

Abstract(199)

Abstract:
In order to accurately analyze the cooling performance and temperature distribution of the sodium-cooled fast reactor debris bed in the boiling, drying and channel drying phases, and to improve the calculation efficiency, COMMEN-LT was developed based on the COMMEN program and the DEBRIS-HT program, combining their respective computational advantages. In order to verify the calculation results of the COMMEN-LT program, the ACRR-D10 experiment of Sandia Laboratory in the United States was selected for comparison. The results show that the COMMEN-LT program can well simulate the heat exchange mechanism and temperature distribution of the debris bed in the initial stage of heat transfer, boiling stage and channel drying.. The calculation time is reduced to the level of “second”, which saves huge calculation costs and greatly improves the calculation efficiency.

Study on Local Three Dimension Hydrogen Risk Using GOTHIC

Huang Gaofeng, Gong Yu, Fu Tingzao, Wang Jiayun, Zhang Kun, Fang Likai

2020, 41(6): 19-23.

Abstract(201)

Abstract:
Integrated severe accident code is used to analyze the hydrogen risk in current safety assessment. After Fukushima accident, higher requirements are placed on hydrogen risk analysis. In order to supplement the lumped parameter analysis, three dimension hydrogen risk analysis method using GOTHIC is studied. Local three dimension hydrogen risk model is constructed by GOTHIC. Based on model validation, typical severe accident cases are chosen to analyze the hydrogen distribution. The results show that, hydrogen and other gas are mixed well in the upper compartment of the containment, and hydrogen stratification phenomenon is not obvious. For DVI rupture accident in PXS-B, the lower area of the break is flooded, and the hydrogen concentration for the upper area of the break is large, however, the hydrogen risk is little.

Effect of PWR Cladding Materials in Different Nuclear Evaluated Data Libraries on Reactivity and Its Analysis

Xiao Xiang, Chen Yixue, Yang Tongrui, Wu Jun

2020, 41(6): 24-30.

Abstract(158)

Abstract:
Pressurized water reactor cladding material, Nuclear evaluated data library, NJOY2016

Heterogeneous Discontinuity Factors in Heterogeneous Variational Nodal Method to Eliminate Control Rod Cusping Effect

Liang Boning, Wu Hongchun, Li Yunzhao

2020, 41(6): 31-35.

Abstract(223)

Abstract:
To handle the control rod cusping effect in pressurized water reactor (PWR) fuel management calculation, the variational nodal method (VNM) in the fuel management calculation code system NECP-Bamboo has been extended to tread the heterogeneous cross section distribution by expanding the volumetric cross sections into piece-wise polynomials in the early work. However, the partially inserted control rods also introduces heterogeneous discontinuity factor (DF) on nodal interface. Thus, in this paper, an ultimate solution is proposed to fully handle this problem. Firstly, the surface integral in the VNM is modified to contain the discontinuity of neutron flux, incorporating a continuous discontinuity factor in that term. Secondly, the surface DF is expanded into the sum of pieces-wise orthogonal polynomials to construct the nodal response matrixes. Comparing with current representative re-homogenization methods, the application numerical results of the BEAVRS benchmark problem demonstrate the effectiveness of the heterogeneous VNM with heterogeneous DF. It can eliminate the cusping effect by providing more accurate differential control rode worth curves and pin power distributions.

Research on Multi-Physics Coupling Method of Pool Type Fast Reactor Based on CFD

Zhao Pengcheng Liu Zijing, Yu Tao, Liu Peiqi, Xie Jinsen, Chen Zhenping

2020, 41(6): 36-44.

Abstract(256) PDF(161)

Abstract:
Based on the critical/subcritical point kinetics model, the fuel pin heat transfer model, and the auxiliary thermal hydraulic models such as the heat exchanger model and the porous media model, a multi-physical coupling code CFD/PF was developed by means of the explicit iteration method, dynamic link library technique (DLL) and user-defined functions (UDF) of FLUENT. The CFD/PF was used to carry out the simulation of SNCLFR-100 unprotected transient of over power (UTOP) of a small natural circulation LBE cooled fast reactor, and the code-to-code comparison analysis was conducted with the renowned multi-physical coupling code SIMMER-III. The results indicated that the CFD/PF simulation results are in a good agreement with SIMMER-III calculation results, and the multi-physical analysis method and code development have been achieved initial success, which can be used to analyze the complex three-dimensional flow and heat transfer phenomena in pool-type fast reactors.

Poisoning in a Reactor Stimulated by Single Group Point Reactor Model and its Experimental Verification

He Ran, Zhang Kuan, You Haixiang, Zheng Xiaomin

2020, 41(6): 45-51.

Abstract(172)

Abstract:
To predict the poisoning quickly and precisely, a method to determine the parameters of the single group point reactor model with appropriate boundary conditions is developed, and the poisoning is predicted with the single group reactor model. And then, the evolution of Xenon poisoning and Samarium poisoning is simulated in several working conditions of the reactor of a M310 Nuclear Power Plant with the single group point reactor model. Xenon poisoning and Samarium poisoning calculated by the single group point reactor approximation fits well with the results of a more accurate 3- dimensional 2 group method. The accuracy of the point group model is examined by simulating the Xenon poisoning and Samarium poisoning in a real license operation event. The reactivity worth of Xenon poisoning and Samarium poisoning simulated by the model fits well with the reactivity worth obtained in the experiment. This work shows that the single group point reactor model can illustrate the Xenon poisoning and Samarium poisoning with significant accuracy with the parameters deduced by this method.

Experimental Study on Critical Heat Flux in Deformed Flow Channel between Pressure Vessel and Insulation

Liu Yusheng, Xue Yanfang, Wang Kunpeng, Wen Shuang, Zhang Zhengxin, Li Congxin

2020, 41(6): 52-57.

Abstract(137)

Abstract:
Aiming at the deformation issue of flow gap between reactor pressure vessel (RPV) and insulation in external reactor vessel cooling (ERVC), the effects of insulation deformation on the critical heat flux (CHF) of the bottom head were investigated with FIMR test facility within the same flow rate range. The influence laws of different factors including angle of bottom head wall, flow rate and insulation deformation on CHF of RPV wall were analyzed. The safety margin of ERVC under deformation condition was finally obtained. The results show that the CHF of the bottom head will increase as the angle of the bottom head wall increases or the flow rate increases. Compared to the CHF of the bottom head in the prototype channel, the varying amplitude of CHF under deformation condition is less than 7%. In a word, the effect of insulation deformation on CHF is not significant. What is more, the location, where the safety margin is smallest, at the bottom head wall in the deformed gap is the same with that in the normal gap, while the smallest safety margin in the deformed gap is slightly improved.

Analysis on Effect of Control Rod Worth in Extrapolation in Criticality Experiment

Song Jingkai, Wang Wencong, Yuan Wei, Huang Liyuan, Wang Yongchao

2020, 41(6): 58-61.

Abstract(239)

Abstract:
The application effect of the control rod worth in the criticality extrapolation experiment is studied. The criticality extrapolation experiment is conducted on a zero power reactor, and the analysis is carried out by 2 extrapolation methods, i.e., with and without the consideration of the control rod worth, based on the neutron count rate data. The results indicate that the control rod worth is with great effect on the extrapolation results in the first few extrapolation steps, and the extrapolation 1/2 adding control rod position would exceed the measured critical control rod position without taking consideration of the control rod worth. The extrapolation method taking the control rod worth into consideration is more precise, with better results.

Monte Carlo Neutron Transport Simulation for Dispersion Fuel Based on Chord Length Sampling Method

Chen Zhenping, Guo Qian, Yu Tao, Zhang Zhenyu, Ma Huiqiang, Xie Jinsen

2020, 41(6): 62-68.

Abstract(151)

Abstract:
The dispersion fuel is with the advantages of high burnup, strong ability of containing fission products and good thermal conductivity. It is widely used as an advanced fuel element in new types of nuclear reactors. However, the dispersion fuel element in which the fuel particles statistically distributed in the matrix material presents some new challenges for the conventional neutron transport simulation methods. In this paper, the Monte Carlo neutron transport simulation method based on the chord length sampling is developed. The method can realize the on-the-fly modeling of the dispersion fuel, in which the fuel particles are randomly distributed in the matrix material. The method can obtain neutron transport simulation results accurately and effectively. The method was verified with numerical benchmarks, which indicated the accuracy and reliability of the method in dealing with the dispersion fuels in criticality calculations.

Theoretical Research on Effective Thermal Conductivity of TRISO Particle

Qian Libo, Yu Hongxing, Sun Yufa, Deng Jian, Chen Wei, Liu Yu, Du Sijia, Shen Danhong

2020, 41(6): 69-74.

Abstract(318)

Abstract:
TRISO (tri-structural isotropic) fuel particle consists of a fuel kernel in the center coated with four layers, with good fission product retention capability. The effective thermal conductivity of TRISO fuel particle is an important basis for calculating the effective thermal conductivity of dispersed fuels. In the present work, the theoretical model of the effective thermal conductivity of TRISO particle is built based on the theory of the effective thermal conductivity in multiphase solids in the framework of spherical coordinate and then the effective thermal conductivity of metal matrix microencapsulated fuel (M3) is analyzed combined with the Chiew-Glandt model which is the effective thermal conductivity model for solid-solid binary composite. The results show that the present model provides an excellent prediction of the thermal conductivity of TRISO particle. Finally the effective thermal conductivity of fully encapsulated fuel (FCM) is presented.

Study on Heat Treatment Process for Improving the Toughness of 1Cr14Co14Mo5 Stainless Steel

Shu Ming, Wang Hao, Wu Songling, Liu Xiao, Wang Li, Kong Fanya, Xu Dianxin

2020, 41(6): 75-79.

Abstract(158)

Abstract:
In view of the brittle fracture failure of the end of ball screw caused by insufficient material toughness, this paper carries out the improvement research of martensitic precipitation hardening stainless steel 1Cr14Co14Mo5. Based on the local composition adjustment of 1Cr14Co14Mo5, the study adjusts the original solid solution + aging heat treatment process, that is, the microscopic mechanism of the austenite grains is refined by cyclic phase transformation treatment, to obtain a refined lath martensite structure, and thus to enhance the plasticity and toughness of the material. The microstructure and performance results show that increasing the solution temperature from 1000℃ to 1100℃ can effectively improve the impact toughness of 1Cr14Co14Mo5. 3~4 times of cycle phase transformation treatment can refine the grains and the average grain size is less than 10μm, which can improve the impact toughness KU2 of the material from 20.2 J to more than 60 J and can slightly increase the fracture toughness K1C value.

Research Progress of FeCrAl Cladding Materials Strengthened by Particles for ATF

Wan Haiyi, Wang Hui, Zha Wusheng, An Xuguang, Kong Qingquan, Chen Xiuli

2020, 41(6): 80-84.

Abstract(213)

Abstract:
Since the Fukushima Daiichi accident in Japan, the safety of zirconium alloy as a nuclear fuel cladding material has been questioned. Therefore, the development of accident tolerant fuel (ATF) has been proposed by many countries in the world. FeCrAl alloy has become one of the important materials in the development of advanced ATF cladding materials due to its excellent resistance to high temperature steam corrosion. In this paper, the research progress of particles reinforced FeCrAl cladding materials for ATF is summarized from the aspects of composition design, preparation method and selection of reinforcing particles. The problems of particles reinforced FeCrAl cladding materials are pointed out.

Effects of Thermal Creep on Blister Behavior in UMo/Zr Monolithic Fuel Plates

Yan Feng, Jian Xiaobin, Ding Shurong, Xin Yong, Tang Changbing, Li Yuanming

2020, 41(6): 85-91.

Abstract(325)

Abstract:
For a UMo/Zr monolithic fuel plate with a gas space, a method is developed to simulate the macroscale blister behavior considering the thermal creep effects of the cladding, in which the calculation of cladding deformation is coupled with the gas space pressure. Based on the developed simulation method, the effects of thermal creep strain of cladding and the internal fission gas atom number on the blister behavior are analyzed. The research results indicate that with the thermal creep of cladding considered, if the fission gas atom number is 4.0×1017, the predicted blister threshold temperature will be 100℃ lower than the case without considering the thermal creep of cladding, with the blister threshold temperature set as the temperature at which the blister height reaches 0.1 mm, with the fission gas atom number increasing from 2.5×1017 to 4.0×1017, the blister threshold temperature might decrease by 40℃. The blister threshold temperature of the fuel plates could be improved by using a cladding material with low thermal creep rate.

Research on Prediction Model of Irradiation Embrittlement of RPV Materials Based on Artificial Neural Network

Kang Jing, Sun Kai, Mi Xiaoxi, Wu Lu, Mao Jianjun, Zhang Shuo, Lei Yang, Pan Rongjian, Tang Aitao

2020, 41(6): 92-95.

Abstract(305)

Abstract:
Based on the analysis of a certain amount of on-site test samples, this paper constructs a high-precision artificial neural network model for the ductile-brittle transition temperature prediction of RPV materials. Then we use the model to explore the influence of neutron fluence and neutron fluence rate parameters on the ductile-brittle transition temperature of RPV materials. It is found that the ductile-brittle transition temperature increases linearly with the increasing of neutron fluence, and then rises slowly and finally saturates. The effect of neutron flux rate on the embrittlement of RPV materials is not obvious.

Optimization of Transmission Gear of Personnel Airlock for In-Service Nuclear Power Plant

He Yingyong, Li Qiang, Xie Honghu, Zhang Feng, Liu Xiaohua

2020, 41(6): 96-100.

Abstract(132)

Abstract:
ANSYS finite element analysis program was used to analyze the stresses of the transmission gear for an in-service nuclear power plant in China, and the causes for tooth rupture were found. The transmission torque was so large that the calculated stress exceeded the allowable stress limit of the material, which caused a root rupture accident of the transmission gear. Combined with the causes of tooth rupture for transmission gear, optimization was performed based on material selection and structural improving design. The optimized design scheme passed stress analysis and evaluation. The results show that the calculated stresses of the optimized transmission gear and the sector gear were smaller than the allowable stress limits of the material. The optimized scheme effectively reduced the stress level of the transmission gear, and obviously improved the safety and reliability of the transmission gear operation for personnel airlock.

Analysis of Flow-Induced Vibration of HP-Cooler Coil on Reactor Coolant Pump

Feng Xiaodong, Su Wentao, Ma Yu, Wang Lei, Li Xiaobin

2020, 41(6): 101-105.

Abstract(189)

Abstract:
To verify that the high-pressure cooler structure of the reactor coolant pump (reffered to as the main pump) can avoid the flow-induced vibration under normal operating conditions, this work analyzes the influence of the shell-side fluid on the vibration of the intermediate coil from three aspects, including the vortex shedding, the fluid-elastic instability and the turbulent excitation. The natural frequency of the helical tube equal to 1.877 Hz is determined by using the pre-stressed modal analysis for the comparison of subsequent evaluations. The vortex frequency is calculated for the maximum and the minimum flow areas, respectively, and the ratio of the natural frequency to the vortex frequency is less than 2. The flow velocity instability of the helical tube calculated by the Karman vortex frequency is higher than that of the shell side gap, indicating that the flow velocity of the inner shell side does not reach the critical velocity of the flow elastic instability of the helical tube. Besides, the center frequency of the turbulent excitation calculated by appropriate semi-empirical model of the helical tube bundle is 3.76 times the natural frequency of the helical tube. The results of these three calculations prove that the structural design of the HP-cooler is safe and reasonable, and can satisfy the requirements of nuclear power plants.

Research of Ultimate Bearing Characteristics for Equipment Hatch of Nuclear Power Plant under External Pressure

Du Kun, Zuo Yongde, Yuan Liang

2020, 41(6): 106-110.

Abstract(214)

Abstract:
The ultimate bearing characteristics of the equipment hatch under the external pressure was studied by the finite element analysis method. The parametric calculation model was established in ANSYS, and the external pressure ultimate load of the equipment gate head was obtained by nonlinear buckling analysis. The calculated results were similar to those of the empirical formula, which verified the reliability and effectiveness of the finite element analysis method. The sensitivity of the design parameters affecting the ultimate bearing characteristics of the external pressure was studied, and the changing law of the influence was summarized. The example of the optimization design showed the practicability of the research method and could provide the optimization scheme for the structural design of the equipment hatch.

Numerical Simulation Research on Coupling Vibration of Heat Transfer Tube Bundles under Lateral Action of Single-Phase Fluid

Bao Shiyi Zhu Hai, Tang Di, Yuan Wei, Huang Xipeng

2020, 41(6): 111-115.

Abstract(125)

Abstract:
The heat exchange tube bundle is an important part of the steam generator, and its reliability directly affects the safe operation of the nuclear power plant reactor. Based on the research in related fields, a computational fluid dynamics(CFD)/computational structural dynamics(CSD) coupling calculation method with higher accuracy is proposed and developed, and the numerical simulation research is carried out on the coupling vibration phenomenon between adjacent tube bundles. The vortex structure of the tube array and the vibration response law between adjacent tube bundles are analyzed in the time domain and frequency domain. The research results show that the vortex is generated and shed in the upstream of the tube bundle, and then gradually develops downstream. The interaction of a large number of shedding vortices between the tube bundles greatly enriches the vortex frequency in the flow field, and the vibration of the tube bundle is affected by the natural frequency and the vortex frequency of the tube bundle. The vibration of the surrounding adjacent tube bundles will have a significant impact on the fluid force fluctuation and frequency dominance of the tube bundle, and weaken the lift fluctuation to a certain extent, and when the adjacent vibrating tube bundles are in the same row, the impact on the vibration displacement of the tube bundle is more significant.

Study on Flow-Induced Vibration Damping Simulation of Heat Exchanger Tube in Non-Uniform Two-Phase Flow

Shen Pingchuan, Liu Qing, Qi Huanhuan, Huang Xuan, Liu Jian, Chen Guo

2020, 41(6): 116-119.

Abstract(216)

Abstract:
In the flow-induced vibration analysis of the heat exchanger tube of the stream generator, the damping of each position on the tube is different, since the secondary side of the tube is the two-phase flow (stream-water) and the void-fraction is gradually increased from bottom to top. It is necessary to study the simulation of the tube damping in non-uniform two-phase flow. The study is based on the Pettigrew's damping formula of tube in two-phase flow and the void-fraction distribution along the tube in the typical example. For the two-phase damping component of the tube damping, the disadvantage of damping overestimation of void-fraction processing method in common engineering software are analyzed. The reason is the nonlinearity of the void-fraction influence coefficient. The method of segmentation weighting void-fraction is developed. The effects of different segment lengths are studied, indicating that the section lengths should be minimized. For the subsequent segmentation damping weighting problem, the effects of different weighting factors in engineering methods and standards are compared, and the difference is small. The results of flow-induced vibration analysis with different damping inputs were compared to judge the applicability of Pettigrew's damping formula. From the above four researches, the recommended simulation method of the tube damping in non-uniform two-phase flow is given to more accurately carry out the flow-induced vibration analysis of heat exchanger tube.

Development of Symptom Based Emergency Operation Procedure for HPR1000

Ran Xu, Yu Na, Li Feng, Qian Libo, Chen Wei, Zhang Ming, Wu Qing, Liu Changwen, Leng Guijun

2020, 41(6): 121-125.

Abstract(380)

Abstract:
In order to compensate for the defects of event-oriented emergency procedure (EOP) and state-oriented emergency procedure (SOP), HPR1000 nuclear power technology takes the advantages of the two operation procedures. Considering probabilistic safety analysis (PSA), a new symptom based emergency operating procedures (SEOP) through a large number of operation analysis supporting calculations is established. As an example, the operator actions during steam line break accident guided by SEOP is studied and compared with EOP and SOP. The results show that SEOP can deal with the accident rapidly and directly and can defend multi-accidents. The accident identification and mitigation measures are reasonable and effective. It can make full use of HPR1000 active and passive safety systems to deal with accidents, give full play to the design advantages of the safety system, and enhance the safety level of HPR1000. The principle, methodology and technique of the development can be used in the procedure development for the similar plant and can be used as a reference to improve the procedures for nuclear power plants in service.

Application of T-S Fuzzy Switching Controller in Core Power Control

Jiang Qingfeng, Zeng Wenjie

2020, 41(6): 126-130.

Abstract(168)

Abstract:
The traditional PID controller is used to control the core power, which has the problems of large overshoot and long regulating time in the control process. In order to solve this problem, based on the core transfer function model, the PD controller, the PID controller and the fuzzy controller are weighted and switched by T-S fuzzy rules, and T-S fuzzy switching controller is designed. Taking the core power control of a lead cooled fast reactor as an example, a T-S fuzzy switching control system of the core power is established to simulate the relative power setpoint value step and the core inlet coolant temperature disturbance. The results show that the T-S fuzzy switching controller designed based on the core transfer function model can achieve a good control of the core power.

A New Uncertainty Analysis Method for Tolerance Limit Evaluation

Guo Jiafeng, Lu Chuan, Mao Huihui, Sun Zhongning, Wang Jianjun, Wang Xiaolie

2020, 41(6): 131-137.

Abstract(241)

Abstract:
Uncertainty analysis methods based on WILKS formula is most generally applied for its advantages in reducing the calculation amount. However, the high resolution calculation is required in the nuclear reactor design and safty evaluation , which demand the uncertainty analysis methods with higher efficiency. This paper provides a new method to evaluate the tolerance limit, which is based on the same mathmatical principal of WILKS formula method but with improvements. Compared to WILKS formula method, this new method is capable to decrease the minimum sample size for evaluating the tolerance limit.

Dynamic Reliability Analysis of AP1000 Equipment Cooling Water System Based on BDMP

Zhou Shiliang, Chen Xiyu, E Wanjiang, Zhang Lei

2020, 41(6): 138-142.

Abstract(228)

Abstract:
CCS is a kind of repairable systems with double redundancy. The factors such as the alternate operation of redundant equipment and the repair of faulty equipment have a great influence on the reliability analysis results. Due to the lack of the description of time factors in traditional fault tree analysis, the analysis assumption is too conservative for such dynamic time series problems. In view of the above limitations of fault tree analysis, the dynamic reliability analysis of the AP1000 CCS is carried out by using the BDMP method driven by Boolean logic. Then, based on BDMP model, the simplest fault combination is derived, the system failure probability is calculated, and the main contribution factors of system failure are analyzed. The results show that the failure probability of CCS is sensitive to the CCF of the cooling water pump. Reducing the CCF of the pump can improve the reliability of CCS.

Adaptive Predictive Control for Core Power Based on CPSO Rolling Optimization

Pan Yuekai, Qian Hong, Jiang cheng, Liu Xiaojing

2020, 41(6): 143-149.

Abstract(198)

Abstract:
In view of the nonlinear and reactive constraint of nuclear reactors in the process of variable power, this paper proposes an improved generalized predictive control (JGPC) for core power control. The JGPC calculates the predicted output value by predicting the model parameters and recursive relationships. At the same time, chaos particle swarm optimization (CPSO), which is improved by the sinusoidal chaos strategy and nonlinear inertia weight, is applied to the rolling optimization of JGPC. In the process of optimization, the reactive constraint are dealt with by setting optimization boundary and chaos strategy. The controlled auto-regressive integral moving average (CARIMA) model of core power is established as the JGPC prediction model, and the forgetting factor recursive least squares (FFRLS) method is used to identify the model parameters online. The JGPC controller is simulated and validated based on MATLAB platform. The results show that the controller can make the core power follow the set value quickly and steadily under the condition of satisfying the constraint, and has a certain anti-interference ability.

Development of Operation Procedure for Secondary Pipe Break Accident Integrating State Oriented and Event Oriented

Yu Na, Ran Xu, Xian Lin, Li Feng, Zhang Zhuohua, Wu Qing, Liu Changwen, Leng Guijun, Chen Wei, Fang Hongyu, Cheng Hongxia

2020, 41(6): 150-154.

Abstract(190)

Abstract:
HPR1000 adopts the symptom based emergency operating procedures (SEOP) to deal with accidents. In this paper, the related procedures for secondary pipe break accidents in SEOP are studied, including the development of procedures and supporting verification. In the development process of the procedures, a reasonable framework of procedures is constructed, and the operation procedures for different accidents are summarized. Based on the design characteristics of HPR1000, the main recovery strategies and relevant important setpoints are determined. In the supporting verification process, the typical secondary pipe break accidents are selected for analysis. The results show that the mitigation strategy provided by SEOP can effectively guide the nuclear power plant to the required safe and controllable state in time. In addition, by comparing different types of accident procedures, the advantages of SEOP in the coverage of accidents and the timeliness of recovery operation are proved. Through the research of this paper, a reasonable method is established for the development and verification of the operation procedure for HPR1000 secondary pipe break accidents.

Flow Rate Research of TXRIS007 Supervision Requirement Criterion

Qiu Yanfei, Wu Shungui, Yang Zijun, Lu Yang

2020, 41(6): 155-161.

Abstract(89)

Abstract:
In order to solve the problem that the check valve function verification tests (TXRIS007) of multi-unit accumulators in a nuclear power plant do not meet the acceptance criteria in most cases, the one-dimensional fluid simulation software (Flowmaster) was used to establish the test model to carry out the transient calculation of different configurations, and theoretical analysis was presented based on historical data. The research shows that while only the reasonable errors of the inner diameter of the test pipelines and the measuring instruments are considered, the initial flow rate range is between 2.37 m3/h and 5.70 m3/h under the normal condition of other test influencing factors; the difference of equipment performance such as the large change of valve resistance characteristics, the actual inner diameter and pipe wall roughness of the pipeline, and the serious blockage of the common loop will lead to the large change of flow; setting a uniform minimum flow may not recognize the change of the performance of valves on the loop with higher reference flow, so it is suggested that the reference flow for different units under the same test conditions shall be calibrated based on the calculation results, and the test results shall be recorded and summarized.

Development of Intelligent On-Line Monitoring Device for Fuel Cladding Defect in Nuclear Power Plants

Xiao Ming, Cheng Xiaoqiang, Wang Yangyi, Song Yun

2020, 41(6): 162-166.

Abstract(137)

Abstract:
An intelligent on-line monitoring device for fuel cladding defect has been developed. HPGe with anti-Compton scattering detection system is used to measure the activity of characteristic radionuclides in primary cooling water, and the multi-nuclide group coupled analysis method is used to diagnose the defect of the fuel cladding. The verification and calibration test shows that the absolute relative standard deviations of the measured typical nuclides 57Co, 137 Cs and 60Co are less than 3%, and the minimum detectable activity can reach 6.5 Bq.

Pre-Tightening Force Calibration Test Research for Stud without Measuring Rod of HTR-PM

Jin Dongjie, Wang Yin, Jin Gang, Geng Baojie, Guo Yunlong

2020, 41(6): 167-171.

Abstract(166)

Abstract:
In order to control the final pre-tightening force of the stud without measuring rod in the high temperature gas cooled reactor Pebble Module (HTR-PM) and ensure the sealing of the reactor primary circuit pressure boundary, it is necessary to calibrate the pre-tightening force of the stud. Taking the M56 stud without measuring rod as an example, the calibration test was carried out with a bolt tensioner, and the relation curve between the pre-tightening force and the final pre-tightening force of the stud was found. The final pre-tightening force on the stud was smaller than the pre-tightening force of the bolt tensioner when the nut was tightened. The reason of the force decreasing was analyzed. The relation formula between the pre-tightening force and the final pre-tightening force of the stud was established through the constitutive relation theory of material mechanics. It was shown that the final pre-tightening force and its relation between the pre-tightening force obtained from the test were close to the theoretical analysis results. Under the same output force of the bolt tensioner, the final pre-tightening force of the actual equipment nozzle flange studs will be larger than the calibrated value, but this is good for the sealing.

Rearserch of Localization Substitution of Reactor Pressure Vessel Seal Ring

Hu Wensheng, Hong Jun

2020, 41(6): 172-176.

Abstract(275)

Abstract:
C-ring is the core part for seal of the reactor pressure vessel top cover and cylinder, and the sealing performance is directly related to the safety and stably operation of the nuclear power plant. For a long time, the manufacturing technology of C-ring was monopolized by foreign company, with high price and long supply cycle. It is verified that the C-ring made in China fulfill the commercial application, through sealing properties test, helium leak detection test, water pressure test and thermal-cold cycle test. The functional impact analysis proved the equivalence of the domesticated C-ring. The installation quality of domestic C-ring was strictly controlled in nuclear power plant, and the performance of the domestic C-ring was verified by the in-service hydraulic test and the running test.

Study on Fluid Flow and Heat Transfer in Seal Chamber of Hydrodynamic Mechanical Seals in Reactor Coolant Pump

Xiang Xianbao, Yang Quanchao, Wen Xue, Zheng Jiarong

2020, 41(6): 177-181.

Abstract(230)

Abstract:
To study the distribution of the flow and temperature field in a type of hydrodynamic mechanical seal of a nuclear reactor coolant pump, a three-dimensional model of the mechanical seal and the seal chamber is established based on the software Pro/E. The N-S equations and energy equation coupling with the k-ε turbulence model are solved based on ANSYS Fluent. The heat generation between the mechanical seal rings and the heat transfer distribution in the sealing chamber is studied. The fluid flow and temperature field of the mechanical seals is analyzed. The results show that the pressure distribution of the mechanical seals is divided into the high-pressure zone and the low-pressure zone by the sealing end face. The liquid film pressure in the mechanical seal end faces gradually decreases from the outer radius to the inner radius. The highest temperature appears at the sealing face, and the temperature decreases gradually away from the sealing face. The viscous heat in the liquid film is transferred away through the seal rings and the heat convection of the fluid flow in the sealing chamber. The pumping ring of the mechanical seal strengths the heat convection of the end faces.

Analysis and Improvement of the Increasing of Sodium Ion in Secondary Loop Caused by Condensate Polishing Plant in Nuclear Power Plants

Cheng Zhenhua, Wang Guoliang

2020, 41(6): 182-186.

Abstract(274)

Abstract:
The content of sodium ion in one steam generator blowdown system cannot satisfy the WANO chemical indices. This paper analyzes the possible causes of sodium ions in the secondary loop of the nuclear power plant and the difference of water quality indicators before and after the operation of the condensate polishing plant. It is concluded that the resin regeneration effect of the condensate polishing plant directly determines the sodium ion content in the steam generator sewage. With the ATE regeneration, the removal of the oil contamination can be maximized and the sodium hydroxide consumption during the regeneration of the anion resin of mixed bed can be reduced, and the sodium ion content in the steam generator sewage can satisfy the WANO chemical indices when the ATE is operating.

Research on Evaluation Method of Nuclear Power Pipeline Fracture Size under Random Missing Data

Zhao Xin, Cai Qi, Zhang Liming, Zhao Xinwen, Wang Xiaolong, Li Haicui

2020, 41(6): 187-193.

Abstract(158)

Abstract:
The monitoring parameters of the nuclear power system are randomly lost due to noise interference,which affects the judgement of the operators onthe severity of the accident. A diagnosis model of fracture size with tolerance parameter loss is proposed. The multiple time series which fracture size is known is selected as the standard series, and several sampling sites are built on the standard series based on the accident mechanism. The sliding dynamic time warping algorithm is adopted to find the minimum cumulative distance between the diagnosed multivariate time series and the standard sampling site, and all the minimum cumulative distances obtained are taken as the characteristic values of the fracture diagnosis model. The support vector machine is used as the prediction model to predict the size of the fracture, and the ensemble learning strategy is used to optimize the diagnosis results. Taking the right main steam pipeline as an example for verification, the results show that this method does not have high requirements for the integrity of sequencing sequences, and the evaluation error of the fracture with random loss of parameters is within 10%, which makes it better for the auxiliary operator to conduct the evaluation of the fracture.

Numerical Simulation of Gas Injected Bubble Dynamics from Single Submerged Orifice

Shen Lanting, Chai Xiang, Cheng Xu

2020, 41(6): 194-197.

Abstract(403)

Abstract:
In severe accidents of a nuclear power plant, the released radioactive aerosols can be removed by pool scrubbing effect. Two-phase numerical simulation of the pool scrubbing process is necessary. The boundary conditions at the bubble injection point need to be determined before using the two-phase CFD program. Based on the framework of Integration of Pool scrubbing Research to Enhance Source-term Calculations (IPRESCA) project and the volume of fluid (VOF) method, a numerical simulation of gas injected bubble dynamics from single submerged orifice was carried out. Bubble size, shape, and detachment frequency at the orifice were captured. Sensitivity analysis of the influence of bubble injection speed on bubble detachment frequency was carried out. The bubble centroid height was obtained by DBSCAN clustering algorithm, and the bubble rising velocity at different heights was calculated. The distribution of the mean void fraction along the z-axis direction and the distribution of mean void fraction and mean mixture velocity along the horizontal lines and radius in central plane at different height are given.

Research and Application of Spent Fuel Storage Racks Positioning Test Method Based on Machine Vision

Cheng Wei, Wang Lei, Hu Jianhua, Wang Zhiming, Zhang Peng, Ji Dapeng

2020, 41(6): 198-201.

Abstract(214)

Abstract:
The positioning test of spent fuel racks is a key test in the commissioning of nuclear fuel handling and storage system. It directly affects the safety and efficiency of the receiving and storage of nuclear fuel assemblies. In this paper, the traditional manual positioning test method is studied, and an automatic positioning test method based on the machine vision technology is proposed. Engineering application shows that this method can greatly improve the efficiency of positioning test and reduce the need of manpower. Furthermore, it is also effective in the exclusion of foreign material and the protection of finished products. Therefore, it has considerable benefits in economic, safety and quality.

Effect of PWHT on Mechanical Properties and Microstructure of SA517 Gr.F Welded Joint

Li Juan, Wang Lulu, Yu Jie, Liu Hongpeng

2020, 41(6): 202-206.

Abstract(306)

Abstract:
In order to better understand the effect of PWHT (post weld heat treatment) on SA517 Gr.F quenched and tempered steel welded joint, the mechanical properties and microstructure distribution characteristics of SA517 Gr.F welded joint by SMAW (shielding metal arc welding) before and after PWHT were compared and analyzed. The analysis results show that, compared with that of as-weld welded joint, both the room temperature and 360℃ high temperature tensile properties of the welded joint decrease, and the peak micro-hardness of weld metal and HAZ also decrease after PWHT. The residual stress of the welded joint is decreased by PWHT, however, the impact properties of the weld and HAZ are not improved. It is suggested that PWHT can be exempted for SA517 Gr.F quenched and tempered steel if the requirements of ASME Code Case N-71-18 can be satisfied.

Research on Scaling Analysis Method for Natural Circulation Test Facility of Lead-Based Fast Reactor

Zhao Pengcheng, Zhu Enping, Yu Hongxing, Zhai Pengdi, Deng Shengwen, Xia Bangyang, Chen Baowen

2020, 41(6): 207-213.

Abstract(351)

Abstract:
Lead-based fast reactors have good natural circulation capabilities, and its natural circulation characteristics is of great value to improve the inherent safety of the reactor, and the scaling analysis method is the theoretical basis for establishing a reasonable and feasible lead-based fast reactor natural circulation test facility. In this paper, the main similarity groups could be determined by using dimensionless fluid governing equations of typical natural circulation lead-based fast reactor primary cooling system. Based on the constructed dimensionless similarity groups, the scaling analysis of small natural circulation lead-based fast reactor named SNCLFR-10 was carried out to obtain the geometric and thermal hydraulic design parameters of the dual-loop single-phase natural circulation experimental facility. The scaling method of the lead-based fast reactor natural circulation test facility was verified by comparing and analyzing the key thermal and hydraulic parameters of SNCLFR-10 and the scaled-down test facility under rated conditions. The research results show that the key thermal-hydraulic parameter ratios of SNCLFR-10 and the scaled-down facility are in good agreement with the theoretically deduced ratio, and the established lead-based fast reactor natural circulation experimental facility scaling analysis method is reasonable and feasible.

Study on Measurement Method of Subcritical Reactivity

Tang Xiao, Xiao Peng, Liu Tongxian, Liao Hongkuan, Huang Can, Zhao Dehua, Liu Mingquan, Lu Di, Li Mancang

2020, 41(6): 214-217.

Abstract(145)

Abstract:

Dynamics Numerical Calculation for Control Rod Drop

Zhang Jibin, Gao Xilong, He Hangxing, Gong Ruzhi, Ma Chao, Yue Ning

2020, 41(6): 218-223.

Abstract(330)

Abstract:
The dropping time and profile of control rods are important parameters during the safety evaluation of nuclear power plants. CFD method and dynamic mesh were used to study the dropping profile of the control rod and flow evolution. The changes of displacement, velocity and acceleration versus time during the dropping of the control rod are obtained. Meanwhile, the effect of the deformation of the control rod and channel on the control rod dropping is evaluated. The results of the work are of help to the design and optimization of the structure of control rods.

2020 Vol. 41, No. 6