核动力工程

Numerical Study on Characteristics of Single Droplet Impacting on Single Dry Flat Wire Mesh under Airflow Disturbance

Chen Bowen, Li Jingsong, Tian Ruifeng, Mao Feng, Lu Chuan, Wen Jiming

2020, 41(2): 1-5.

Abstract(347) PDF(239)

Abstract:
Mesh separators are widely used in industry. Based on the dynamic process of droplet impact on the mesh in this paper, the CLSVOF method is used to solve the problem of single droplet impact on the dry flat wire mesh. After reasonable simplification, a two-dimensional mathematical model of the single droplet impacting mesh surface is established and analyzed. The effects of droplet impact angle and impact position on the behavior of droplet impacting were studied. The numerical results show that the droplets impacting on the dry mesh are divided into two processes: spreading and splashing. The smaller the impact angle, the bigger the front spreading radius, and the smaller the back spreading radius, the larger the volume of the daughter droplets; the closer the impact point is to the end of the mesh, the more likely to produce daughter droplets, the more the total volume of daughter droplets.

Automatic Image Recognition and Analysis of Bubble Behavior in Rectangular Narrow Channel

He Xueqiang, Liu Hanzhou, Chen Deqi, Lu Qi

2020, 41(2): 6-10.

Abstract(359)

Abstract:
Aiming at the bubble dynamic behaviors generated during the boiling flow in the rectangular narrow channel, the high-speed camera was used to capture the bubble behaviors, and then the image processing technology was used to bubble recognition, and finally the characteristic parameters are extracted for analysis. Through the improved image processing method, the problem that some unclosed bubble edges exist in the binarized image because of the light spots was solved, and a good gas-liquid segmentation effect was achieved. Then, the bubble projection area ratio of processed image was extracted by automatic processing, and the bubble projection area ratio by the manual recognition was used for verification, and the difference between the two is within 10%. Finally, the bubble volume calculation model was established according to the contact angle, and the change trend of void fraction during the bubble growth was obtained.

Error Analysis of LPD On-line Monitoring System in HPR 1000

Zhang Zhizhu, Liao Hongkuan, Li Qing, Gong Helin, Chen Zhang, Li Xiangyang, Liu Qiwei

2020, 41(2): 11-15.

Abstract(385)

Abstract:
LPD on-line monitoring system is adopted in nuclear power plants to calculate and display the real monitoring parameters such as linear power density (LPD) and hot spot factor, which can present the state of the reactor core accurately and in time, thus to improve the economy and safety of the nuclear power plant. The overall error of LPD should be considered in the alarm limit settings of LPD, with certain margin. The analysis method for overall uncertainty of LPD on-line monitoring in HPR1000 is studied, and combining errors of various parts obtained by statistical method, the overall error limit is obtained. The results show that the error of LPD on-line monitoring system in HPR1000 meets the engineering requirement.

Analysis of Thorium-Loaded S&B Fuel Blocks Using RGPu and WGPu as Driver Fuels

Wang Jincheng, Huang Jie, Ding Ming

2020, 41(2): 16-21.

Abstract(182)

Abstract:

First-Principle Study of H Atom Adsorption on Zr(0001) Surface

Chen Wu, Zhang Hengquan, Ye Xiaofeng, Zeng Jing, Xiao Hongxing, Zhou Mengbing

2020, 41(2): 22-26.

Abstract(371)

Abstract:
The adsorption sites and mechanism of H atom on α-Zr(0001) surface were calculated and analyzed from microstructure, adsorption probability, adsorption energy, Mulliken charge population and density of state, and etc. based on the integration of Monte Carlo (MC) simulation and first-principle density functional theory (DFT) method. The results indicated that the H atom firstly generated physical adsorption on the Zr(0001) surface and then changes to chemical adsorption. The charge continuously transferred from the surface Zr(0001) atoms to the H atom throughout the entire process, and finally stabilized. Furthermore, the H atom directly bonded with the most surface Zr(0001) atoms after stable adsorption, and the major contribution of Zr-H bond was made by partial density of state of H(s), Zr(s) and Zr(d) orbitals. Comprehensive analysis shows that the priority order of the adsorption sites of H atoms on the Zr(0001) surface is hexagonal close packed gap (hcp)> face centered cubic gap (fcc)>bridge, and the top site is the impossible adsorption site.

Effect of Nano-Y2O3 on Microstructure and Properties of ODS-HT9 Steel

Wu Kaixia, Zha Wusheng, Zhao Jiancheng, Tang Rui

2020, 41(2): 27-31.

Abstract(359)

Abstract:
By power metallurgy method, the samples of ODS-HT9 steel, which were strengthened by dispersion oxide of 0.1%~0.9% Y₂O₃, were prepared in order to study the effect of nano-Y₂O₃ on the microstructure and mechanical properties of HT9 steel. The HV, tensile strength, and elongation of the samples were measured. The distribution, the shape, and the phase structure of Nano-Y₂O₃ oxide particles in the sample were observed by transmission electron microscope (TEM), and the fracture microscope was by scanning electron(SEM). The results show that Y₂O₃ particles were homogenously distributed on the HT9 steel, and their shape and phase structure have no change after ball milling and hot-pressing. Due to the strengthened dispersion of nano-Y₂O₃ particles, the tensile strength and Vickers hardness of ODS-HT9 steel increase, and the elongation decreases, significantly, with the increasing of Y₂O₃ addition. If the addition of Y₂O₃ is less than 0.7%, the fracture of samples is ductile. But, if the addition of Y₂O₃ increases furthermore, the fracture will change to brittle. The ODS-HT9 steel samples with 0.3%~0.5%(mass fraction) of Y₂O₃ have a better properties. The tensile strength and elongation reach to 913~936 MPa and 10.7%~11.2%, respectively. The results will be helpful to the study of ODS-HT9 high temperature performance and its practical application in reactors.

Application of Compressed Air Agitation in Preparation of ADUN Solution

Peng Jiancai, Li Jia, Yuan Bo, Yin Rongcai, Zhang Jie

2020, 41(2): 32-35.

Abstract(182)

Abstract:

R&D Progress of Accident-Tolerant UO2-Based Composite Fuel Pellets

Mo Huajun, Zhang Wei, Wu Lu, Luo Hao, He Wen, Pan Rongjian, Wang Zhen, Wu Xiaoyong, Wen Bang

2020, 41(2): 36-39.

Abstract(471)

Abstract:
In this paper, the research background for the advanced accident-tolerant fuel (ATF) pellets was summarized, focusing on the R&D status of accident-tolerant UO2-based composite fuel pellets at home and abroad. It is concluded that the UN, U3Si2 and ThO2 are doped phases with the most development potential for accident-tolerant UO2-based composite fuel pellets. However, the optimal adding quantity and distribution of these doped phases should be investigated further by multi-scale modeling combined with experiments.

Study on Ultrasonic Inspection of Baffle Bolts in Nuclear Power Plants

Wang Weiqiang, Ma Guanbing, Wang Bin, Tang Jianbang, Zeng Chenming, Xu Junlong

2020, 41(2): 40-44.

Abstract(380)

Abstract:
Baffle bolts are the critical joint parts of the internal structures in nuclear power plants, and it is necessary to perform non-destructive testing on their structural integrity in order to detect the defects such as irradiation assisted stress corrosion cracking (IASCC) which may occur during a long-term service. In this study, the structural characteristics of the baffle bolts and the in-service inspection conditions were analyzed, and a combined-crystal ultrasonic technique (UT) was developed for the external hexagon bolts. The probe design and defect assessment were reviewed to ensure the good sound field coverage, and the cracks in different parts of the bolt can be detected. The practical trials on the flawed samples were performed to verify the feasibility of the ultrasonic process, and the result showed that currently technique is able to detect the flaw on the order of 30% of the shank cross-sectional area and the signal to noise ratio (SNR) can reach more than 12dB, which meet the requirements for the in-service inspection.

Application of Master Curve in Reactor Pressure Vessle Ageing

Yu Xiaohuan, Du Juan, Shao Xuejiao, Yang Yu, Liu Zhenyu, Tian Jun, Yang Lingfang

2020, 41(2): 45-48.

Abstract(306)

Abstract:
The pressure to temperature (P-T) curve is significant to ensure the vessel integrity and deal with the reactor pressure vessel(RPV) ageing. The traditional method using the material near static facture limit value (K_Ic) by t-R_TNDT curve to plot the P-T curve cannot obtain the appropriate R_TNDTfor an irradiated material and is overly conservative. With the surveillance data of a reactor pressure vessel, this paper calculates the Master Curve index temperature, R_TT0, at the ending of the RPV 50 years life. The ASME Code Case N629 effort to utilize the Master Curve Method is used to plot the P-T curve , compared with results of traditional method, the P-T operating window is extended and the RPV economy is improved.

Study on Functional Capability of Piping System Evaluation Criteria

Tang Feng, Li Qiang, Peng Jian, Wang Mingyu

2020, 41(2): 49-53.

Abstract(292)

Abstract:
The functional capability of piping systems is different from the pressure boundary integrity of piping system, and United States Nuclear Regulatory Commission (NRC) specified two evaluation criteria for the piping functional capability. To explore the origin and the application of functional capability evaluation criteria, and based on the research of classical literatures about the content of functional evaluation criteria, this paper expounds the origin and the basis of two kinds of evaluation criteria, analyzes the characteristics of two kinds of evaluation criteria, and proposed notes on the application of two kinds of evaluation criteria. At the end of this paper, through an example of piping functional capability assessment, two strategies for the application of functional assessment criteria in different nuclear plant design stages are proposed. For newly built nuclear power plants, C level stress limit shall be used as possible to ensure the functional capability of piping systems. For the nuclear power plants that had been built, D level stress limit and five additional conditions can be applied to ensure the functional capability of piping systems.

Buffering Effect Analysis for Secondary Supports in Reactor Vessel Internals under Assumption Accident of Core Drop

Fang Jian, Duan Yuangang, Ran Xiaobing, Ma Ruoqun

2020, 41(2): 54-58.

Abstract(249)

Abstract:
To research the buffering effect of secondary supports in reactor vessel internals under the assumption accident of core drop (core barrel broken down), a buffering effect analysis of secondary supports is performed, based on the technology of water buffering and mechanical buffering. In this paper, the process for core drop is numerically modeled, and the core barrel movement mechanism under different vertical gap, coolant temperature and initial flow rate is analyzed based on Fluent. The impact process in the end of core drop is analyzed and the water buffering effect of the secondary support is studied based on LS-DYNA. The analysis results show that, using the technology of water buffering and mechanical buffering, the secondary supports can bearing the impact energy of the core drop, and the coolant can buffer most of the impact energy. The buffering effect of secondary supports in this paper is better than that of the traditional support structure, to ensure the integrity of the reactor pressure vessel (RPV).

Nonlinear Dynamic Analysis of Soil-Pile Interactions of Nuclear Island with Nonlithological Foundation

Cao Liang, Shi Lin, Wu Xianmin, Huang Jianxue, Chen Bingbing, Xie Jigao, Zheng Sanlong

2020, 41(2): 59-65.

Abstract(229)

Abstract:
In the simulated water environment, the stiffness test of the shrinkage specimen of Hold Down Spring of The Reactor Internals was carried out, and compared with the analysis results of the finite element simulation, theoretical model based on small disturbance and the large disturbance follow-up model. The results show that when the friction coefficient is 0.189 according to the experimental value of the literature, the finite element simulation, the stiffness value calculated based on the small disturbance theory model and the large disturbance follow-up model are similar to those obtained by the test; The stiffness of the unloading stable section is significantly smaller than that of the loading stable section, which is about 0.6 times that of the loading. The finite element simulation analysis further clarifies that during the deformation process of Hold Down Spring, the section of Hold Down Spring has a rotation, and the contact point between it and the pad is not fixed, and there is a radial displacement back and forth in the process of loading and unloading. Moreover, the direction of the frictional force on the contact surface of the Hold Down Spring is reversed, so that there is a large difference in the stiffness of the Hold Down Spring during the loading and unloading process. To compare with theoretical model based on small disturbance, the results obtained by considering the large disturbance follow-up model of the Hold Down Spring section rotation and the radial displacement of the contact point are closer to the finite element simulation.

Research on Stiffness Test and Calculation Model of Hold Down Spring under Simulated Water Environment

Cao Liang, Shi Lin, Wu Xianmin, Huang Jianxue, Chen Bingbing, Xie Jigao, Zheng Sanlong

2020, 41(2): 66-71.

Abstract(309)

Abstract:
In the simulated water environment, the stiffness test of the shrinkage specimen of Hold Down Spring of The Reactor Internals was carried out, and compared with the analysis results of the finite element simulation, theoretical model based on small disturbance and the large disturbance follow-up model. The results show that when the friction coefficient is 0.189 according to the experimental value of the literature, the finite element simulation, the stiffness value calculated based on the small disturbance theory model and the large disturbance follow-up model are similar to those obtained by the test; The stiffness of the unloading stable section is significantly smaller than that of the loading stable section, which is about 0.6 times that of the loading. The finite element simulation analysis further clarifies that during the deformation process of Hold Down Spring, the section of Hold Down Spring has a rotation, and the contact point between it and the pad is not fixed, and there is a radial displacement back and forth in the process of loading and unloading. Moreover, the direction of the frictional force on the contact surface of the Hold Down Spring is reversed, so that there is a large difference in the stiffness of the Hold Down Spring during the loading and unloading process. To compare with theoretical model based on small disturbance, the results obtained by considering the large disturbance follow-up model of the Hold Down Spring section rotation and the radial displacement of the contact point are closer to the finite element simulation.

Calculation of Radionuclide Migration in Coastal Waters Based on Lagrangian Method

Li Zichao, Zhou Tao, Si Guangcheng, Qin Xuemeng

2020, 41(2): 72-77.

Abstract(265)

Abstract:
Based on the real-time meteorological data, the hydrodynamic model in the coastal waters of the nuclear power plant was established and the model reliability was verified. Then a radionuclide diffusion model in the coastal waters of the nuclear power plant was established based on the Lagrange method. The hydrodynamic characteristics and radionuclide migration in the coastal waters of nuclear power plants were analyzed. The results show that the simulation hydrodynamic results in the coastal waters of the nuclear power plant can well depict the semi-diurnal tides in the coastal waters of the nuclear power plant. The basic radionuclide migration direction is from the nuclear power plant, along the coast, to the northeast direction. Influenced by the tidal current, the radionuclides advance in rotation. The surface radionuclides migrate along the coast with the shortest distance, and the middle radionuclides migrate with the farthest distance, and the bottom radionuclides migrate with the distance between surface and bottom radionuclides.

Research of Functional Reliability Evaluation Method for Passive Systems Based on Data Mining Technology

Wang Baosheng, Tang Xiuhuan, Bao Lihong, Zhu Lei, Sun Peiwei

2020, 41(2): 78-83.

Abstract(176)

Abstract:
In order to solve the problem of multi-dimensional uncertainty parameters and small probability of functional failure, an innovative functional reliability estimation method named Data Mining Technology was presented. In the presented method, with the combination of the bootstrap response surface model and optimization line sampling design, the functional failure probability can be evaluated with high efficiency. This method was applied in the natural circulation cooling in Xi’an Pulsed Reactor (XAPR). Combined with Medium Break Loss of Coolant Accident (MLOCA), the uncertainties related to the input parameters and the model were considered. And then the probability of functional failure was estimated with the presented method. The numerical results show that the bootstrap response surface model has a high degree of fitting, and the optimized line sampling technique has a high computational efficiency and an excellent computational accuracy. In addition, the evaluation method in this paper has strong adaptability to the implicit nonlinear functional failure analysis of the passive systems.

Study on Initial Passivation Process of Primary Loop of PWR Nuclear Power Plants

Ji Dapeng, Zhang Yeliang

2020, 41(2): 84-88.

Abstract(293)

Abstract:
The development of the passive oxide film on the primary loop during hot functional testing of PWR nuclear power plants is very important in the minimizing corrosion of the materials and corrosion product release. In this paper, based on the theoretical research and engineering practice, it is concluded that the electrochemical reaction and the chemical reaction exist in the formation of the passive oxide film during the hot functional testing, and the growth mechanism of the double-layer film is explained. The reasonableness of the analysis of passivation test by electrochemical measurement method is demonstrated. The functional relationship between passivation temperature and passivation film reaction rate is deduced. When the temperature increases, the reaction rate will increase and the passivation time shortens. The theoretical limit of passivation temperature should not be lower than 260℃.

Effect of Wall Radiation on Turbulent Natural Convection Heat Transfer Characteristics in an Enclosed Cavity with Built-in Fins

Wang Ye, Zhao Xingjie, Ma Bingshan, Guan Guoxiang

2020, 41(2): 89-95.

Abstract(253)

Abstract:
In order to study the effects of wall emissivity on natural convection heat transfer characteristics of an air-filled enclosure with built-in fins, the RNG k-ε model was adopted to numerically analyze the temperature field, the flow field and the Nusselt numbers on the vertical wall of the cavity with aspect ratio of 1. The results show that the thickness of the vertical thermal boundary layer and the vertical velocity boundary layer are both increasing due to the combined effect of the built-in fin and the wall radiation. And the horizontal velocity of the top and bottom regions of the cavity fluctuates to a certain extent. When considering wall radiation, the effect of double fins structure on the local heat transfer capacity of the hot wall is similar to that of the single fin structure. When the inside wall emissivity is set to 0.3, 0.6 and 0.9, respectively, correspondingly, the average Nusselt number on the hot wall of single fin cavity will increase by 39.95%, 88.55% and 144.97% compared with that under the condition of no radiation. Meanwhile, the average Nusselt number on the hot wall of double fins cavity are increased by 41.09%, 87.32% and 141.23% compared with that of no radiation, respectively. For the enclosed cavity of the double-fin structure, high wall emissivity is disadvantageous to the convection heat transfer.

Application of Generalized Predictive Control in Reactor Core Variable Power Control

Pan Yuekai, Qian Hong, Jiang Cheng, Liu Xiaojing

2020, 41(2): 96-101.

Abstract(194)

Abstract:
Aiming at the nonlinear problem of the core power model caused by steady-state neutron density at different power levels, generalized predictive control (GPC) is applied to the core power control to realize the automatic control of the core power under variable working conditions. This paper firstly establishes a core power model based on zero-power core model and temperature feedback model. The prediction time domain is designed based on the order of the model, and the model parameters at different power levels are identified online by the least square method with forgetting factor in the GPC correction link according to the input and output data of the system. In order to verify the robustness of the controller, the reactive disturbance is added at full power smooth operation. The performance of the controller is verified by simulation based on MATLAB platform, and the results show that the GPC designed in this paper can quickly and accurately track the set value when the core is in variable working condition, and can identify the core model parameters of different power levels on-line, and has certain anti-interference ability.

Reason Analysis for Unsatisfactory of Accuracy and Improvement of Accuracy of On-line Boron Meter in a Nuclear Power Plant

Zheng Junwei, Liu Yang, Deng Sheng, Ma Shu, Wang Canhui, Du Wenlong, Cui Yongle

2020, 41(2): 102-108.

Abstract(403)

Abstract:
In order to find out the reason for the unsatisfactory of measurement accuracy of the online type boron meter (BCMS) supplied by AREVA NP GmbH, the scheme to improve the BCMS measurement accuracy was studied. According to the working principle of BCMS, the factors affecting the measurement accuracy of BCMS were analyzed in terms of the neutron measurement and total boron concentration calculation, and the measurement data of BCMS is statistically analyzed; the offline type boron meter(OFBM) with measurement accuracy higher than that of BCMS was selected as the comparison object, and the comparison of the measurement accuracy between BCMS and OFBM was made under the same statistical error analysis condition; the online type boron meter (ONBM) developed by Nuclear Power Institute of China was selected as the reference, and the quantitative analysis of the effect of the measured pipe size on the BCMS neutron count rate was made. The results show that: the unsatisfactory of measurement accuracy existed both in the manufacture factory test and in the nuclear power plant operation of BCMS, which was caused by the extremely low value of the cumulative neutron count rate measured by its neutron measurement device within a fixed counting time; and the loss neutron by polyethylene exists between the neutron source and the neutron detector of BCMS. A modification scheme to improve the BCMS measurement accuracy was proposed, i.e., to cut the polyethylene shielding layer and replace with the higher thermal neutron sensitivity detector .

Core Power Control of Liquid Molten Salt Reactor Based on Variable Universe Fuzzy-PID

Jiang Qingfeng, Zeng Wenjie, Yu Tao, Xie Jinsen, Chen Lezhi

2020, 41(2): 109-113.

Abstract(217)

Abstract:
The core system of liquid molten salt reactor is with the characteristics of non-linearity and time-varying. The initial domain of the fuzzy PID control technology cannot change following the change of error, so the control accuracy of the system is deteriorated. Therefore, the core power control strategy based on variable universe fuzzy-PID controller is designed. Taking MSBR core as an example, the simulation results of PID control, fuzzy-PID control and variable universe fuzzy-PID control are compared by using Matlab/Simulink under the disturbance of core inlet temperature or reactivity. The results show that the core power control system based on variable universe fuzzy-PID controller has faster response speed, smaller overshoot and better control effect.

Research on Separation Characteristics of Two Stage Cyclone Steam Separator

Liu Yan, Ke Bingzheng, Wang Xianyuan, Yang Xuelong, Tian Ruifeng

2020, 41(2): 114-119.

Abstract(384)

Abstract:
A special two-stage whirlwind steam separator was designed based on the research of relevant domestic and foreign data, and the detailed numerical simulation research was conducted in terms of its separation performance. According to the structure of the two-stage cyclone separator, a computational analysis model was established, and the separation performance of the separator was calculated by numerical simulation. The effect of different inlet velocity and different humidity on the separation characteristics of the separator was studied, and the air-water cooling test circuit was built to verify the model. The results show that the numerical simulation results are consistent with the cold test results, and the deviation of separation efficiency is small. In the design condition, the overall separation efficiency of the separator is better than 99.5%. The primary separator is suitable for coarse separation, and its separation efficiency decreases with the increasing of inlet humidity and velocity. The secondary separator is suitable for the separation of small droplets, and its separation efficiency is positively correlated with the inlet velocity and non-linear with the inlet humidity.

Failure Analysis and Design Improvement of Thin-Walled Gears for Mechanical Transmission Mechanism of Air-Lock in Nuclear Power Plants

Liu Shengyong, He Yingyong, Xie Honghu, Huang Yili, Zhang Feng, Zhang Shipeng

2020, 41(2): 120-124.

Abstract(233)

Abstract:
Research work is performed on the failure of thin-walled gears used in mechanical transmission mechanism of air-lock in nuclear power plants, based on the failure mechanism of thin-walled gears and its loading condition. The essential causes of failure of thin-walled gears founded, and they are: a. Material selection of thin-walled gears is unreasonable. Raw material is ANSI 1340, and composition analysis results show that S element (content is 0.11%) is far beyond the standard requirement. The gears contain a large amount of MnS inclusions, which deteriorates the material properties, and resulting in tooth undercut fracture, b. Structure design of thin-walled gears is unreasonable. A special structural displacement gear with a number of 12 teeth is adopted, and the displacement of the gear exceeded its permissible limit, resulting in the fracture of the tooth. Finally, the optimization and improvement measures for thin-walled gears are proposed from aspects such as the material selection, the manufacturing process control and the structural design.

Analysis of Effect of Primary Piping Diameter on Performance of Main Pump

Wang Yan, Cui Huaiming, Guo Yanlei, Mao Yuanfan, Duan Yongqiang, Li Lei, Su Xianshun

2020, 41(2): 125-129.

Abstract(280)

Abstract:
To the nuclear primary pump of PWR and its two different primary piping cold leg diameter configuration schemes, a three-dimensional model is established by combining the primary pump and the primary pipe, and the hexahedron structured mesh is used to perform the model partition and calculate the unsteady flow characteristics of the whole flow region. The unsteady pressure fluctuation characteristics in the pump and the pipe under different cold leg configuration schemes of the primary pipe are obtained. The results show that the primary pump efficiency decreases with the increasing of the cold section diameter, but the hydraulic loss decreases with the increasing of the cold section diameter, and the efficiency of the whole system increases by 1.3%; the pressure fluctuation amplitude of transition section can be significantly reduced with the increasing of the cold section diameter; the difference of pressure fluctuation amplitude between the inlet position of guide vane and the pressure chamber in the two schemes of cold section diameter is small, and the amplitude of the pressure fluctuation in the cold section is also small, and all of them show the characteristics of no cycle and no rule; the configuration of large diameter cold section will slightly reduce the amplitude of the axial force fluctuation.

Research on Crack Failure of Auxiliary Feed Water Pump Impeller Based on CFD

Li Yuanzheng, Liu Zheng, Zheng Hongen

2020, 41(2): 130-134.

Abstract(246)

Abstract:
In order to study the crack failure characteristics of the auxiliary feed pump impeller, the pressure pulsation response law of the auxiliary feed pump casing measuring point under different impeller crack lengths is explored based on CFD, and the CFD results are verified by the vibration monitoring data of the auxiliary pump feed pump in the presence or absence of impeller cracks. The research shows that the pressure pulsation response of the pump casing can better reflect the crack failure characteristics of the auxiliary feed pump impeller; when there is a crack in the auxiliary feed pump impeller, there will be a clear sideband with a width of rotating frequency of the blade passing frequency in the vibration spectrum of the pump casing measuring point.

Study on Modification for Auxiliary Feed Water System Temperature Beyond the Limit in Operation Technical Specification of Nuclear Power Units

Wang Shuqiang

2020, 41(2): 135-139.

Abstract(311)

Abstract:
For the unit fallback case in which the water temperature of Auxiliary Water Supply System exceeds the temperature limit in the Operation Technical Specification due to the minimum flowrate operation in Summer, the modification of installing a cooler on the line of the circulation pump for ASG tank are proposed, and the process design, changes in instrument and control system and operation control are investigated. The results show that with the modification, the water level and temperature can be maintained within the limits of OTS, and the units can operate in a safe and economic mode. The research of this paper is with a reference value for nuclear safety improvement and outage optimization.

Modification of First Electrical Penetration in Domestic Nuclear Power Plants

Yi Feifan, Qian Houjun, Chen Ziming, Chen Fujie, Zeng Shu, Li Shuping, Chen Qihao

2020, 41(2): 140-144.

Abstract(303)

Abstract:
As the key equipment on the containment, the electrical penetration of nuclear power plants undertakes the important functions of various power and signal transmission inside and outside the nuclear island and ensuring the integrity of the containment pressure boundary. Through the implementation of domestic DDG-1 type electrical penetration replacement and transformation project during the 18th refueling outage of 300000 units in Qinshan Nuclear Power Plant Phase I project, the status and the necessity of the transformation of the electrical penetration equipment in Qinshan Phase I project are analyzed. At the same time, in view of the limited short time available for the replacement and modification of the penetrations in the nuclear power plants in service and the difficulty in verifying the sealing performance of the penetration, by optimizing the inspection process and adopting a special inspection tool for, the time required for the penetration modification is shorten, and the sealing performance of the penetration can be verified.

Simulation Research of Ultrasonic Inspection of Primary Tube Seat Weld in Nuclear Power Plants

Hu Chenxu

2020, 41(2): 145-149.

Abstract(233)

Abstract:
The small-size branch joint (BOSS) weld is the weak point of the primary pressure boundary in nuclear power plants, and its monitoring is critical but difficulty in the daily and in-service inspection. This paper designs and verifies the ultrasonic phased array inspection technology for BOSS welds, by using the simulation technology, technical test and application verification, and also solves the difficulties of monitoring BOSS welds in daily and in-service overhaul of nuclear power plants. The method of design and verification can be used in similar ultrasonic phased array inspection process.

Design Methodology for Standard Commissioning Guidelines for Native Advanced PWRs

Shang Chen, Tian Qiwei, Mao Huan, Liu Yong

2020, 41(2): 150-154.

Abstract(266)

Abstract:
As one of the basic technical guidance documents for nuclear power plant commissioning, the standard commissioning guidelines are designed to perform commissioning test with the common method for a same type of equipments and components, or a given type of test. Based on the requirements of domestic and foreign regulations and standards, by analyzing the design characteristics and commissioning task requirements of native advanced pressurized water reactor(PWR)nuclear power plant, a common design methodology of standard commissioning guidelines for native advanced PWR is given. Combined with the commissioning experience of the second generation PWR, a system of standard commissioning guidelines with independent intellectual property is designed for native advanced PWR and a new form of document classification and coding is determined, to reduce the risk of misquotes and misuses, and to reduce the workload of commissioning personnel, while facilitating the documentations use, management and archive.

Analysis and Research on Key Design Parameters of Leakage Detection System for Steel Liner of Radioactive Storage Pool in Nuclear Power Plants

Chen Chuyuan, Xie Honghu, Chen Zhao, Liu Xiaohua

2020, 41(2): 155-159.

Abstract(562)

Abstract:
Leakage detection system is one of the most important parts of the steel liner for long-term radioactive storage pool. In order to confirm the key design parameters of the leakage detection system, the analysis work is performed by risk management analysis and hydrodynamics calculation methods. Based on the analytical conclusion, the detail design of the leakage system for the spent fuel storage pool liner of 3rd generation nuclear power plants is cited as an example. Moreover, it shows that, there is sufficient design margin for the spent fuel pool system design by comparing the water replenishment capacity and liquid level alarm design of PTR system with the obtained data.

Diagnosis of Leakage Degree of Steam Generator Tube Based on Time Series Neural Network

Qian Hong, Jiang Cheng, Pan Yuekai, Wei Yingchen, Liu Xiaojing

2020, 41(2): 160-167.

Abstract(326) PDF(126)

Abstract:
Aiming at the leakage of the steam generator U-shaped tube, a method based on time series neural network is proposed to diagnoze the leakage degree of steam generator tube. Firstly, the leakage mechanism of the steam generator U-shaped tube of the nuclear power plant is analyzed, its mathematical model is constructed, the direct characteristic parameters of leakage are extracted, and then the indirect characteristic parameters are extracted according to Fisher score method. Secondly, data samples are generated from the pretreated time series data by sliding time window method, which is used as the input of time series neural network. Based on Back propagation (BP) algorithm, the five-layer neural network system is trained to get the time series neural network model of the leakage of steam generator U-shaped tube. Finally, the time series test data of the leakage of steam generator U-shaped tube during nuclear power operation are simulated. The simulation shows that the time series neural network is with better effectiveness and generalization ability in dealing with evolutionary events, which is of a reference value for fault diagnosis research.

Study on Production Capacity of 63Ni by HFETR Irradiation

Zhou Chunlin, Li Haitao, Zhang Jiangyun, Chen Liang, Li Ketian, Lai Lisi

2020, 41(2): 168-172.

Abstract(385)

Abstract:
To complete the production of radioisotope ⁶³Ni based on self-reliance, the reactor physical calculation program is used to establish a balanced core model of HFETR. The process of irradiating natural nickel and high-purity ⁶²Ni to produce ⁶³Ni at four typical in-pile locations is simulated. The results show that high purity ⁶²Ni is the most suitable target material. The central channel of K14 fuel element is the best irradiation position. The high specific activity ⁶³Ni (≥5.55×10¹¹Bq/g) is produced by high purity ⁶²Ni target with irradiation time about 60 cycles; The radioactive isotope ⁶³Ni produced by irradiation has low impurity nuclide content; The target at K14 position during the irradiation has the highest heat release rate of 9.73 W/g. Furthermore, a continuous batch production plan for segmented loading and segmentation is proposed, which can provide technical support for the subsequent engineering production line design.

Research on Interconnection Security Protection System of Nuclear Reactor Industrial Control System and Enterprise Information System

Qin Lihua, Wang Dan, Wang Daqiu

2020, 41(2): 173-177.

Abstract(232)

Abstract:
In view of the scattered and isolated security protection strategies of the nuclear reactor industrial control system and enterprise information system interconnection projects, this paper conducts a detailed analysis of the security and confidentiality risks at the management level and the technical level with the consideration of the security and confidentiality requirements of the system interconnection. On the basis of the analyzed results, this paper proposes a security interconnection protection system based on the management and technology dual defense, which provides a guide for the construction and implementation of the interconnection of a nuclear reactor industrial control system and a enterprise information system.

Study on Pattern and Key Issues of Contactless Power Supply for Hot Cell Transfer System

Zhuo Yong, Wang Dingchao, Ye Senmao

2020, 41(2): 178-183.

Abstract(225)

Abstract:
With the analysis of the shortages of traditional power supply for hot cell transfer system, a novel contactless power supply system is proposed based on the study of magnetic coupling structure, which realizes the reliable contactless power supply for the hot cell transfer system when it moves randomly. Contactless power supply system can overcome the motion limitation compared with traditional hot cell transfer system, and improve the flexibility of hot cell operation. The feasibility of contactless power supply for hot cell transfer system is verified by theoretical analysis and computer simulation.

Analysis of Accuracy and Efficiency of Two Node Method for Solving Fine Mesh SP3 Neutron Transport Equations

Zhao Wenbo, Yu Yingrui, Chai Xiaoming, Ning Zhonghao

2020, 41(2): 184-188.

Abstract(393)

Abstract:
A two node method is developed to solve the fine mesh SP3 equations. The Laplace operator on the 0th moment flux is treated by nodal method. The transverse integrated 0th moment flux is expanded to a second order polynomial, and the transverse leakage term is approximated with flat distribution. A prototype code named CORCA-PIN is developed for the fine mesh whole core transport calculation. And fine mesh finite difference method is implemented in CORCA-PIN. KAIST 3A benchmark problem and extended 3D problem are tested. Numerical results show that an acceptable accuracy is achieved using the two node method with 1by1 mesh per cell. Meanwhile, the time is 12% of the fine mesh finite difference method with comparable accuracy. The two node method proposed in this paper is suitable as the solver of fine mesh SP3 neutron transport equations.

Study on Coupling Characteristics of Small Break LOCA in Advanced PWR

Jin Yuan, Jiang Xiaowei, Deng Jian, Liu Yu, Bi Shumao, Zhu Dahuan, Yang Fan

2020, 41(2): 189-192.

Abstract(358)

Abstract:
In order to understand the transient response characteristics of the passive containment cooling system, the passive core cooling system and the passive residual heat removal system after the small break LOCA in an advanced pressurized water reactor, the coupling response characteristics of the reactor coolant system and the containment in SBLOCA are studied. The analysis results show that in the SBLOCA, the characteristics of the passive residual heat removal system, the passive core cooling system and the passive containment cooling system in the coupling analysis are quite different from the independent calculations. The containment peak pressure obtained by the coupling analysis in the SBLOCA is less than that by the independent calculation.

Effect of Main Feedwater Isolation Modes on Mass and Energy Release and Containment Response for MSLB

Guan Zhonghua, Qiu Zhifang, Jiang Xiaowei, Duan Yongqiang, Shen Yunhai, Fang Haoyu

2020, 41(2): 193-197.

Abstract(342)

Abstract:
Taking the 3&4 units in Qinshan Second Nuclear Power Plant as an example, the effect of two modes of feedwater isolation on the mass and energy release during a large main steam line break (MSLB) accident, by valve isolation and shutdown of the main feedwater pump, is analyzed by THEMIS code. The thermal-hydraulic phenomenon in the containment is analyzed by PAREO9 code. It is concluded from the results that the isolation mode by valve isolation is more effective to reduce the mass and energy release in MSLB and the increasing of boron concentration in the boron injection tank can also reduce the mass and energy release in MSLB.

Case Analysis of Loose Parts Alarm in Nuclear Power Plants

Jian Jie, Luo Ting, Liu Caixue, Wang Guangjin, Hu Jianrong, Yang Taibo

2020, 41(2): 198-202.

Abstract(268)

Abstract:
In view of the phenomenon that several alarms of loose parts occurred in the thermal test phase of a nuclear power plant during the overhaul, a probe was temporarily installed at a key location for field test, and the event location of the collected test data was conducted by analyzing the time difference of impact signal array waves reaching different sensors. The results show that the origin of the alarm event of the loose part is related to the change of thermal test-related state of the nuclear power unit. The event may be due to the spontaneous or indirect conduction of the swing limiter at that location. The successful diagnosis of this alarm event not only solves the actual engineering problem, but also is with great practical value to improve the detection interval of the existing monitoring system for loose parts.

A Remote Operation and Maintenance Data Management System for Nuclear Reactors Based on Integration of Big Data and Relational Data

Bai Yi, Qin Lihua, Wang Sishi

2020, 41(2): 202-206.

Abstract(335)

Abstract:
Based on the process of the nuclear power operation and maintenance, this paper overviews various operation and maintenance data in detail, analyzes the data types, data characteristics and magnitude, and then puts forward a data management framework between big data and relational data; Based on the business characteristics of the remote operation and maintenance, the data management requirements of nuclear power operation and maintenance are further analyzed; With the B/S framework, a data management system based on the integration of big data and relational data is developed, and the management mode of the nuclear power operation and maintenance data is innovated; The system has been effectively applied in the operation and maintenance service, which greatly improving the efficiency of the nuclear power operation and maintenance, and also providing the foundation for the follow-up research of the intelligent operation and maintenance.

Design and Application of a Remote Online Self-Checking Unit Based on Charge Signal

Li Xiang, Wang Lei, Li Hai, Zou Bohao, Fu Guoen, Deng Sheng

2020, 41(2): 207-213.

Abstract(210)

Abstract:
A remote online self-checking unit based on charge signal is designed to provide a fast and effective remote online self-checking method for the primary instrument of the Loose Parts Monitoring System (LPMS) installed in the reactor building of the nuclear power plant. The design principle and the method of the remote online self-checking unit, as well as the hardware design and software flow design are introduced. A laboratory test is conducted on the remote online self-checking unit, and the test result shows that the remote self-checking unit can meet the requirements of 250 m under the condition of no-load output 5 V voltage. Meanwhile, the domestic LPMS prototype integrated with the remote online self-checking unit is verified to meet the engineering application environment through the qualification test. Two sets of the domestic LPMS products integrated with the remote online self-checking units have been successfully applied in the C-3/C-4 unit of the Chashma Nuclear Power Plant in Pakistan. With low cost and high efficiency, the online self-checking function of LPMS primary instrument during the reactor loading operation is realized.

2020 Vol. 41, No. 2