核动力工程

Initiation and Development of Heat Pipe Cooled Reactor

Yu Hongxing, Ma Yugao, Zhang Zhuohua, Chai Xiaoming

2019, 40(4): 1-8.

Abstract(1852) PDF(600)

Abstract:
The heat pipe cooled reactor adopts the solid-state reactor design concept and passively transfer the heat out of the core through heat pipes. This paper summarizes the development history of the heat pipe cooled reactor, from the conceptual initiation, the active exploration and to the breakthrough. The technical characteristics of heat pipe cooled reactors are analyzed, including the key advantages, such as solid properties, inherent safety, simple operation, easy modularization and expansion, and transportability. In addition, this paper summarizes the technical status of heat pipe performance, material technology and energy conversion in heat pipe cooled reactors. The challenges in the further development of heat pipe cooled reactors are put forward, such as material technique, manufacturing, and operation maintainability. The future development trend of heat pipe cooled reactors is clarified, which provides a direction for the development and application of the innovative heat pipe cooled reactor technology. Overall, the heat pipe cooled reactor has broad application prospects in deep space exploration and propulsion, land-based nuclear power supply, sea exploration and other scenarios, which may become one of the most creative technologies to change the future nuclear power patterns.

Study on Non-Uniformity of Sub-Cooled Boiling Heat Transfer ofHigh Pressure Water in Fuel Assembly

Du Lipeng, Chen Xiaolong, Zhang Pengfei, Zhang Wenchao, Jin Guangyuan

2019, 40(4): 9-14.

Abstract(250)

Abstract:
Based on the VOF model in Fluent, a sub-cooled boiling model was established and was inserted by UDF code into the control equations. Sub-cooled boiling in the 5×5 rod bundles channel with spacer grid was simulated. Heat transfer characteristics in sub-channels were analyzed by investigating the void fraction distribution and the heat transfer characteristics of the fuel rods. The results show that the void fraction distribution is different even if the heating condition is the same, and the void fraction is larger in the edge channel than that in the angular channel. The analysis of the heat transfer characteristics around the rod shows that, the heat transfer of the direction of 0°, 90°, 180°, 270° of fuel rod is better, the x velocity is larger, and the boiling is stronger.

Numerical Study of Falling Film Flow and Wave Characteristics in Passive Containment

Qiu Qinggang, Long Huangxiang, Deng Jian, Qiu Zhifang, Meng Chuiju, Zhu Xiaojing, Shen Shengqiang

2019, 40(4): 15-20.

Abstract(246)

Abstract:
A simplified two-dimensional numerical model of 1:45 was established based on the passive safety system of the passive containment. Under the premise of keeping the dimensionless number unchanged, the Fluid Volume Function (VOF) model was used to capture the gas-liquid interface, and the liquid film spreading conditions under different Reynolds number Re conditions were analyzed. By comparing with the experimental results, the simulation results can better reflect the liquid film fluctuation in the experimental results, and the change of liquid film thickness is consistent with the Reynolds number. The process of vertical wall falling film is synthesized by a variety of forces, the fluid film can be divided into obvious laminar and isolated waves, the bottom flow velocity of laminar flow is low and the solitary wave moves faster along the flow direction, and the collision and fusion between the isolated waves intensifies the fluctuations of the liquid film with the spread of liquid film; With the increasing of Re, the thickness of the liquid film at the bottom layer and the amplitude of isolated wave increased, and the volatility increased.

Visualization Study on Boiling Flow by Image Processing under External Reactor Vessel Cooling

Xue Yanfang, Zhang Xiang, Tian Daogui

2019, 40(4): 21-24.

Abstract(236)

Abstract:
Based on the visualized data obtained by the scale-down test facility under external reactor vessel cooling, the image automatic resolution and interface capture program IMGPROCS1 was developed to analyze the observation of downward facing pool boiling phenomena under the hemispherical structure using the commercial software Matlab. Through this image processing program, two phase flow characteristics and morphology along the overheated surface such as interface evolution, vapor film thickness and boiling cycle were quantitatively acquired under different working conditions. The results showed that under nucleate boiling condition, the vapor thickness increased gradually with the increasing of heat flux. The boiling cycle was maintained constant.

Research on Fluid-Solid-Heat Coupling for Check Valves in Marine Nuclear Power Plants

Yu Hang, Zhao Xinwen, Fu Shengwei

2019, 40(4): 25-28.

Abstract(365)

Abstract:
Aiming at the leakage of check valves in marine nuclear power plants, the effect of rapid temperature change on the check valves is studied by the fluid-solid-heat coupling simulation method. The results show that the equivalent stress and deformation of check valves are reduced with the decreasing of temperature, the maximum equivalent stresses of the sealing gasket and the quarter ring are at the ends of the valve pipe, and the maximum deformations are at the front and rear of the valve. The maximum deformation and shrinkage of the sealing gasket are larger than that of the quarter ring. Because of high temperature and pressure, there is a clear gap between the sealing gasket and the valve cover, and it can easily lead to leakage. The gap increases with the decreasing of temperature, which may intensify the leakage.

Analysis of Reactivity Introducing Accident in ³He Irradiation Rig

Xu Taozhong, Deng Caiyu, Ma Liyong, Zhu Lei, Yang Bin, Kang Changhu, Zhang Ping, Liu Shuiqing, Yang Lingfang

2019, 40(4): 29-31.

Abstract(246)

Abstract:
In the high flux engineering test reactor (HFETR), the change of gas pressure in ³He loop would introduce reactivity to the reactor, and then affect the operation safety of the reactor. The reactivity change of the variable power rig was calculated by MCNP code in this paper, and the superimposed accidents of the loss of pressure in ³He tube and the withdrawing of the control rods were analyzed by RELAP5 code. The results showed the irradiation tests conducted in rig do not have effects on the normal operation of HFETR. Even when the loss of pressure in ³He tube and the withdrawing of the control rod happened at the same time, the safety of HFETR was guaranteed.

Computational Fluid Dynamic Simulation on Heat Transfer Characteristics of Water Flowing in Heated Tube during Depressurization Transient under Supercritical Pressure

Li Yongliang, Zeng Xiaokang, Wen Yan, Zang Jinguang, Yan Xiao, Xiao Zejun, Huang Yanping

2019, 40(4): 32-38.

Abstract(192)

Abstract:
The computational fluid dynamic simulation on the heat transfer characteristics during depressurization transient under supercritical pressure was performed in this paper by adopting Ansys Fluent 15.0 software based on available experimental data. The reliability of the calculation was verified by comparing the computational results with the experimental data, and the simulation method established in this paper was practical and applicable for transient heat transfer during depressurization under supercritical pressure. The calculation results indicated that the parameters at the outlet of the test section exceeded the pseudo-critical point but still stayed in the area where the thermal physical properties varied dramatically in simulated transient condition with thermal parameters nearby the critical point. The incompressible flow at the inlet of the test section was turned to compressible flow at outlet region. Furthermore, there was a phenomenon of peak around pseudo-critical point for both mass flow flux and pressure gradient. In addition, the classical steady-state supercritical heat transfer correlation–Jackson empirical correlation overestimated the Nusselt number obtained in the computation by about 20%~50% near the critical point region.

Scaled Experimental Study on Heat Transfer Characteristics of PCCS under DBA Conditions

Meng Xianke, Fei Likai, Gao Bin, Zhang Shengjun, He D

2019, 40(4): 39-43.

Abstract(147)

Abstract:
There is a passive containment cooling system (PCCS) in the AP / CAP series nuclear power plants, to cool the containment passively within 72 hours after accidents. However, after 72 hours, if the top tank is not able to replenish water in time, it is difficult to take all the residual heat away by the containment itself, and there exists potential overpressure for the containment. To solve the problem of the time limit in the residual heat removal of the containment in nuclear power plants, an experiment system for an innovative passive residual heat removal system for the containment is built, to study the heat transfer performance under different pressure, temperature and gas composition under design basis accident (DBA) conditions. The results show that the heat capacity of the system completely meets the design. Further, the correlation formula of the external tube condensation heat transfer coefficient with non-condensable gas for low subcooling condition is given.

Thermal-Hydraulic Characteristics of Filtered Containment Venting System for Long-Term Operation

Dong Shichang, Yang Jun, Deng Chengcheng, Sui Zengguang

2019, 40(4): 44-49.

Abstract(169)

Abstract:
Long-term operation characteristics of the filtered containment venting system (FCVS) is important for mitigating the severe accident. To research the long-term thermal-hydraulic characteristics of FCVS with periodically open-and-close-venting strategy, a typical FCVS under assumed severe accident was simulated by using the thermal-hydraulic code RELAP5. Further, the effects of initial fluid level, environment temperature and decay heat power on the operation characteristics were researched by sensitivity analysis. The results show that the FCVS can achieve stably the periodic operation up to 250 hours under the assumed accident. By sensitivity analysis, it is found that in order to ensure the normal operation of FCVS, the initial fluid level should be changed with different environment temperature, and the decay heat power should be controlled within a certain range. This research may provide guides for the operation and optimization of FCVS and the safety analysis of nuclear power plants.

Study on Processing Methods of HELIOS Format Multi-Group Library Applicable for ADS Assembly Calculation

Bao Lihong, Jiang Xinbiao, Zhang Xinyi, Tang Xiuhuan, Wang Lipeng, Xu Jialong

2019, 40(4): 50-55.

Abstract(201)

Abstract:
The HELIOS program built-in multi-group library does not include all the nuclides in the accelerator driven subcritical system(ADS) when it is applied in ADS assembly calculation. To solve this problem, the generating HELIOS format multi-group library method applicable for ADS assembly transportation calculation is studied. Based on ENDF/B VII.0, following the flow scheme of multi-group library generation, a new 45 groups library is generated for ¹⁶O that included in HELLIOS program built-in multi-group library, and examined at micro and macro levels. The results indicate that the method of generating multi-group library is correct. For nuclides that are not in the built-in multi-group library, HELLIOS program built-in 112 groups library is expanded and numerically validated by ADS cell. The results further indicate that the method of generating multi-group library is correct. Since the results calculated by HELIOS have large deviation from that by MCNP, the method to improve a multi-group library is proposed and numerically validated. The results indicate the relative deviation is obviously reduced after improvement compared to the original one.

Effect of Spectrum Variation on Activation Measurement for Neutron Fluence

Zou Peng, Cao Jiebao, Wang Yunbo, Tang Xiding, Kang Changhu, Yang Bin

2019, 40(4): 56-59.

Abstract(209)

Abstract:
In the reactor neutron fluence measurement, the activation detectors will suffer neutron radiation in several fuel cycles, with which neutron spectrum varies. Considering this factor, for the neutron fluence measurement was corrected for one batch of irradiation material for the domestic pressure vessel. Calculations show that the relative deviation for the weighted neutron fluence (E＞1.0 MeV) correction is 1.75%, reducing 3.73% compared to the uncorrected weighted neutron fluence (E＞1.0 MeV), and thus the effect of spectrum variation cannot be ignored.

Research and Validation of Burnup Library Processing for Lead Based Fast Reactor RBEC-M

Liu Jiayi, Ma Xubo, Qiu Ruomeng, Xu Qian, Chen Yixue

2019, 40(4): 60-64.

Abstract(249)

Abstract:
A new method to process the data for burnup calculation is proposed. Using NJOY is used to process ENDF-B-VII.1, and 33-group MATXS format library is generated. The multigroup cross section generating code (MGGC) can get the micro and macro cross section with composition information, using a new added module Triso to merge and output the data, and finally the ISOTXS format library for burnup calculation can be obtained. The fission product is expressed with the macroscopic cross section of lumped fission product, and others are in the form of microscopic cross section. The lead cooled fast reactor benchmark 900 MW RBEC-M was calculated using REBUS-3 burnup calculation module, and the results of effective neutron multiplication factor, power distribution and neutron spectrum were compared. The final results are consistent with the reference in the report, and the feasibility of this burnup library processing method is validated preliminarily.

Study on Properties of FeCrAl Coated ATF Zirconium Cladding Prepared by Plasma Spry

Li Rui, Liu Tong

2019, 40(4): 65-69.

Abstract(389)

Abstract:
This paper introduces the latest accident tolerant fuel (ATF) cladding achievements of China General Nuclear Power Corporation (CGN). The FeCrAl coating of zirconium with different technology were prepared by plasma spay. From the SEM and XRD results, the best technology for FeCrAl coating sample can be selected. The corrosion resistance of FeCrAl coating is tested by the high temperature steam oxidation test, and the element distribution in oxidated sample is analyzed by SEM and EDS, to study the corrosion mechanism of FeCrAl coating, and the improvement method is suggested.

Preliminary Analysis on Indoor Behaviors of Leaked UF₆ in Uranium Fuel Fabrication Facilities

Que Ji, He Wei, Zhang Min, Cao Fangfang

2019, 40(4): 70-75.

Abstract(243)

Abstract:
In order to analyze UF₆ leakage accident in detail, including the leakage flow changes and the behavior when it leaks into the room, based on the mass and energy balance in the UF₆ container and the mass balance in the room, a source term analysis model of UF₆ indoor release is established. This model is used to analyze the typical UF₆ leakage accident in the safety analysis report of domestic uranium fuel fabrication facilities. Some information are obtained, including phase change, leakage flow and the change of leakage state in the container, the concentration and deposition of hazardous materials in the room after leakage, and the concentration of hazardous materials eventually discharged into the environment, which can be used for the implementation of emergence plans and environmental impact assessments.

Study and Application of Commercial Grade Dedication in Digital Instrument Control System in Nuclear Power Plants

Sun Hongtao, Li Hongxia, Liu Jingbo, Zheng Weizhi

2019, 40(4): 76-80.

Abstract(289)

Abstract:
The progress of nuclear plant construction greatly intensify the market demand for indigenous digital instrument control system for nuclear power plants. However, it is hard to find the qualified suppliers who can provide numerous nuclear grade facilities for safety functions. We can only adopt non-nuclear grade facilities usually applicable to common industry standards from non-nuclear grade facilities suppliers, namely commercial grade items, which quality should be controlled and guaranteed as required by laws and regulations. With reference to current international experiences, we found that the Commercial Grade Dedication (CGD) has been proven to be a successful way. Basing on the designation and supplement of DCS for the generators No. 5 and No. 6in Tianwan Nuclear Power Plant, we studied related national and international standard and CGD methods and found a set of CGD system and methods suitable for NPPs in China. They have been proven to be successful in Tianwan Nuclear Power Plant, and should be a valuable reference for CGD for nuclear field in China, or even in the world.

Analysis of Effectiveness of Medium Voltage Mobile Power Supplies in M310 and Modified Nuclear Power Plants

Zhang Qi, Cao Guanghui, Geng Yan, Kong Jing, Liu Peng, Chen Zixi

2019, 40(4): 81-84.

Abstract(238)

Abstract:
This paper uses PSA tool to analyze the effectiveness of medium voltage mobile power supplies according to the General Technical Requirements for Nuclear Power Plant Improvement after Fukushima Nuclear Accident, referring to the design of mobile power supplies in Ling’ao Phase II Nuclear Power Plant. Through qualitative and quantitative analysis, this paper argues that the medium voltage mobile power supply has a significant mitigation effect on the station blackout accident under power running and shutdown conditions of nuclear power plants, and proposes suggestion on the improvement of General Technical Requirements for Nuclear Power Plant Improvement after Fukushima Nuclear Accident.

Analysis of Several Issues of Instrumentation and Control System of Nuclear Power Plant Design in UK GDA Process

Zheng Weizhi, Bai Xiangji, Sun Hongtao, Liu Jingbo

2019, 40(4): 85-90.

Abstract(197)

Abstract:
In order to undertake the construction project of the nuclear power plant in UK, Nuclear Safety Review and License Application for Nuclear Power Projects shall pass the GDA review conducted by UK ONR firstly. The GDA review is based on the SAPTAG guidelines that issued by ONR, and related IAEAWENRA guidelines and IEC standards. Through analyzing I&C design issues that are concerned by ONR and requirements of relevant guidelines and standards, based on the review experience of UK-EPR and UK-AP1000, the I&C overall architecture design strategies are proposed. It can be used as a reference for the follow-up GDA review project.

Characteristics of Operator Interface Management Tasks in Digital Main Control Room of Nuclear Power Plants

Zhang Li, Liu Jianqiao, Zou Yanhua, Qing Tao, Huang Rong

2019, 40(4): 91-95.

Abstract(292)

Abstract:
This paper firstly analyzes the key behaviors of operators in the interface management tasks by using Systematic Human Error Reduction and Prediction Approach (SHERPA), and then uses INTERACT9, a kind of software, to analyze the video of the operator operation in the interface management tasks on the full range simulator of one nuclear power plant in China. After the statistical analysis of the data collected by INTERACT9, the four general characteristics of the tasks are obtained: a. The most frequent operations are conducted on the menu bar, the monitoring target and the parameter interface by operators in the first and secondary loop; b.The operators prefer to enter different system interfaces through the menu bar; c. The interface management tasks under normal conditions and accident conditions for the operators in the primary loop are generally the same, but the interface management tasks under normal conditions are significantly less than that under accident conditions for the operators in the secondary loop; d. Under normal conditions, the interface management tasks for the operators in the primary loop are significantly more than that in the secondary loop. Under accident conditions, the interface management tasks for the operators in the primary loop are equivalent to that in the secondary loop.

Method of Scenario Selection for Integrated System Validation

Sun Qian, Chu Jiru, Wang Yuqi

2019, 40(4): 96-99.

Abstract(191)

Abstract:
Integrated system validation(ISV) is the performance-based assessment of integrated human-system interface (HIS)is prior to loading in nuclear power plants. Appropriate integrated system validation scenario selection method is to confirm that the activity is sufficient, necessary and orderly. Based on the HSI design of nuclear power plants, this paper proposes a set of systematic and implementable integrated system validation scenario selection method meeting the review requirements of Human Factors Engineering Program Review Model(NUREG 0711)on sampling of operational conditions, so that the integrated system validation work can be more reasonable and efficient.

Improvement of Online Tritium-in-Air Monitoring System in Heavy Water Reactor Power Plants

Su Guoquan

2019, 40(4): 100-103.

Abstract(86)

Abstract:
The analysis technology for the ion-chamber is used to measure the tritium concentration in heavy water reactor power plants. However, the online tritium-in-air monitoring system cannot measure tritium from multiple areas continuously. The system extraction technology has been modified to improve the system reliability. The modification results indicate that the system obtains accurate and real-time measurement readings, and that it can provide very important evidence for the locating of tritium leakage and the radiation protection of operators in nuclear power plants.

Study on Human Reliability of Operator in Digital Main Control Room

Chen Qingqing, Zhang Li, Hu Hong, Qing Tao, Dai Licao

2019, 40(4): 104-107.

Abstract(226)

Abstract:
At present, the classical and common human reliability analysis (HRA) methods are studied before the digitization of the main control room. Based on the training video of the simulator, this paper defines, collects and analyzes the operator's operation behavior in the digital main control room by using the behavioral method. The results show that the operation error probability of the operators is normal in the digital main control room, but the secondary task operation and its human error probability increases. Therefore, this study provides a basis for HRA of the operators in the digital main control room.

Roller Wear Degree Recognition of Control Rod Drive Mechanism Based on the Complexity of Singular Spectrum Entropy

Zhang Liming, Li Lin, Hong Liyang, Yang Xiaochen

2019, 40(4): 108-112.

Abstract(119)

Abstract:
Aiming at the problem of the wear condition monitoring of the roller in the control rod drive mechanism in reactors, the concept of complexity is introduced. As one of the complexity, the singular spectrum entropy is used to recognize the wear condition. After the reconstruction phase space of the observed data, the singular is calculated. Combined with the concept of entropy of information, the singular spectrum entropy is defined. The singular spectrum entropy is sensitive to the uncertainty of the observed data and shows the change of the components in the frequency domain which is related to the wear condition. By the experiment of the control rod drive mechanism in its whole life, it is found that the high-frequency components increase and the low-frequency components change little, with the running time increasing. The experiment proves that the method of singular spectrum entropy is quite valid in wear condition monitoring and recognition.

Study on Severe Accident Initiated by Surge Line Break for Small Modular Reactors

Yin Shasha, Luo Hanyu, Qiu Suizheng, Huang Wei, Chen Zhihui, Tian Ye, Fang Huawei

2019, 40(4): 113-116.

Abstract(271)

Abstract:
The potential surge line break should be fully considered in the research and analysis of severe accidents for SMR. Therefore, the severe accident analysis model was established based on the design characteristics of SMR. The severe accident process is simulated, and the thermal-hydraulic phenomena and the effects of different break sizes under the pressurizer surge line break accident are analyzed. he. The results suggested that the most serious accident process happened when the break area is 0.002m². Thus, this paper suggested to use this accident as the limit condition to provide reference for the formulation of the severe accident management guideline.

Research on Tank Sloshing of Pressure Suppression Pool in Foating Nuclear Power Plants

Zheng Yaxiong, Guo Jian, Fu Zhuangzhi, Liang Shuangling, Tan Mei

2019, 40(4): 117-122.

Abstract(269)

Abstract:
The operating environment of floating nuclear power plants is definitely different from that of land nuclear power plants, which should be taken into account in the design of special safety facilities. Especially in the design of the facilities involving liquid flow, the effect of marine environment adaptation under the ship’s motion exciting should be considered. Taking the pressure suppression pool as an example in this paper, finite volume method (FVM) is adopted to conduct simulation analysis for the motion process to study the water level change in extreme working environment and the differences with and without swash plate. Research results show that the lowest water level due to roll and pitch motion is lower than the initial water level. Resonance period of pressure suppression pool is not the same with the ship’s rocking period, thus the water level change mainly depends on the motion amplitude and structural members inside jointly.

Study on Behavior of Work Teams in Digital Main Control Room of Nuclear Power Plants

Chen Qingqing, Zhang Li, Hu Hong, Qing Tao, Dai Licao

2019, 40(4): 123-126.

Abstract(177)

Abstract:
When the main control room of the nuclear power plant is digitized, there are some changes in the exchange and cooperation of the work teams, but few studies have been conducted on work teams before, and mainly based on the empirical research instead of experimental research, thus it is difficult to properly quote or modify the Performance Shape Factors (PSFs) in the implementation of Human Reliability Analysis (HRA). In this paper, the behavior analysis software is used to study the behavior of a work team in the training of simulator. The behavioral characteristics and organizational structure of the digitalized main control room are studied for the first time from the perspective of behavioral analysis. Statistics and analysis of communication between teams in an accident training. The research results show that, the operator team in the digital main control room is a clear organizational structure with clear division of labor.

Research of Numerical Simulation Accuracy Based on Performance Prediction of Nuclear Main Pump

Hu Xiaodong, Wang Xiuyong, Liu Zailun, Zhang Xiaofei, Li Yibin

2019, 40(4): 127-133.

Abstract(402)

Abstract:
In order to improve the numerical simulation accuracy of the nuclear main pump at all the working points, three factors affecting the numerical simulation results, i.e., the near-wall mesh scale, turbulence model and flow state are comprehensively studied. The results show that the RNG k-ε turbulence model and the standard wall function method have higher calculation accuracy when y⁺=50, and the calculation accuracy is also higher than that of three kinds of turbulence models, i.e., the RNG k-ε enhanced wall function method, low Reynolds number k-ε and SST k-ω. However, the calculation results of the above-mentioned different grid scales and turbulence models have large errors. Under the unsteady state calculation, the calculation accuracy of the unsteady state calculation method is significantly improved compared with the steady state calculation method, which can reflect the hump phenomenon of the head curve near the dead point, and the calculation accuracy of efficiency is also improved. The unsteady calculation method is more suitable for performance prediction of nuclear main pump.

Experimental Research on Digital Processing Algorithm for RSPND Delay Signal

Xu Xiaoheng, Mo Huajun, Li Dongcang, Yang Lei, Zhu Zhaoyang, Yang Wenhua, Zhou Chunlin, Shao Jianxiong

2019, 40(4): 134-138.

Abstract(177)

Abstract:
The slow response characteristics of output current signal of rhodium self-powered neutron detectors (RSPND) has seriously affected the real-time response of the neutron flux rate measurement in the reactor, which is not conducive to real-time control and safety management of reactor. Inverse function calculation or various compensation methods are used to improve the response characteristics of RSPND, which is beneficial for the application of RSPND. This paper studies the forward differential transform method, backward differential transform method, step response method and bilinear transform method of four kinds of digital processing algorithms, and the response time for the output signal of the vanadium self-powered neutron detector is shortened effectively. The time constant is reduced to less than 5 seconds. The digital experimental system verifies the correctness of the algorithm, which provides the possibility for the rapid response of rhodium self-powered neutron detectors in the measurement the neutron flux rate in the reactor.

Research on Digital Prototype Technologies of Personnel Airlock for Nuclear Power Plants

Xie Honghu, Ma Wenqin, Zhang Feng, He Yingyong, Yang Jinchun

2019, 40(4): 139-144.

Abstract(103)

Abstract:
In order to improve the safety and reliability of the personnel airlock of nuclear power plants, this paper develops a digital prototype of personnel airlock and its key components. In addition, combined with structural reliability analysis with digital prototype simulation, the weak links that influence the reliability of the equipment are specifically analyzed. By the failure mode and effect analysis (FMEA) method, the weak links of the equipment are found. Weak link-lifting mechanism is analyzed by using dynamics simulation software ADAMS. And the finite element analysis carried out by ANSYS output the stress distribution and the deformation of these components in accident condition. The results show that the rationality of the structural design of personnel airlock and the correctness of the assembly relationship are verified.

Design and Research of Drilling and Draining Equipment for Canopy Seal Weld on CAP1400/AP1000 Canned Reactor Coolant Pump

Yan Guohua, Wen Zhong, Li Wei, Yu Zhaohui

2019, 40(4): 145-148.

Abstract(208)

Abstract:
The drilling equipment for the canopy seal weld was developed to drain the residual radioactive coolant from the steam generator during the disassembly of the canned reactor coolant pump in China Advanced Passive Power Plan t (CAP1400) and Advanced Passive Power Plant (AP1000). The results show that the expected sealing effect of the device is satisfied and there is no leakage of coolant during the whole process; the drill tool and machining parameters are optimized and the debris is small and fragmented, which can be drained with the coolant; the whole system is with compact structure and easy-control process, and can be remotely operated.

Development of Hydrogen Concentration On-line Monitoring Devices under Severe Accidents in NPPs

Wang Hongqing, Tang Min, Tang Yueming, Ma Weigang, Zheng Hua, Chu Li, Jiang E

2019, 40(4): 149-152.

Abstract(306)

Abstract:
The key technology problems such as the electrode coating and the structure design of the hydrogen concentration monitor are sovled in this paper. The on-line hydrogen concentration monitoring prototype was designed and manufactured successfully, and the measuring tests under different pressure, temperature and hydrogen concentration conditions, and the performance tests under normal conditions are conducted. The test results indicated that the above monitoring prototype could on-line monitor the hydrogen concentration with good selectivity, short response time, wide measuring range and high monitoring accuracy. The on-line hydrogen concentration monitoring devices can be applied to the HPR1000 and CAP1400 nuclear power plants.

Development of Feedthrough Assembly for Electrical Penetration Assembly Based on Third-Generation Nuclear Power Technology

Zhou Yuan, Wang Guangjin, Zhou Tian, Chen Qing, Zhao Yuheng, Qiu Xinyuan, Wang Jiangwu, Zhou Han

2019, 40(4): 153-156.

Abstract(194)

Abstract:
According to the special signal transmission requirements of the third-generation nuclear power plants, taking the tri-axial feedthrough as an example, and aiming at the special requirements of the tri-axial cables through the containment and its radio frequency signal transmission, this paper describes the design, manufacture, process test and type test of the special tri-axial feedthrough for the electrical penetration assembly. The results of the final type test show that the structure design of the tri-axial feedthrough is reasonable and the manufacturing process is feasible, which can meet the technical requirements of the electrical penetrator assembly in third-generation power plants, and its research results can be directly applied to the design of the electrical penetrator assembly for third-generation power plants.

Improvement of Irradiation Technology for Mono-Crystalline Silicon in MJTR

Yang Bin, Wang Hongyang, Wang Yunbo, Xiang Yuxin, Zhang Ping, Kang Changhu

2019, 40(4): 157-160.

Abstract(264)

Abstract:
In order to achieve the irradiation ability for 8 inch mono-crystalline silicon, also to improve the irradiation quality, this paper studies and improves the irradiation technology for mono-crystalline Silicon in Min Jiang Test Reactor (MJTR). Neutron screen is set inside the irradiation tube to reduce the irradiation uniformity, and the layout of the reflector is increased to achieve higher thermal neutron flux, in order to meet the requirement of 8inch mono-crystalline silicon irradiation. Meanwhile, this paper optimizes the irradiation process, and designs and improves the semi-automatic control process. After the improvement of irradiation technology, the operation time is shortened by about half, the number of irradiation tanks can be increased by 2%, and the efficiency of irradiation production is improved significantly.

Development and Validation of Critical Boron Concentration and Burnup Calculation of Software CORCA-3D

An Ping, Ma Yongqiang, Guo Fengchen, Sun Wei, Lu Wei, Liu Dong, Li Qing

2019, 40(4): 161-165.

Abstract(191)

Abstract:
Critical boron concentration and burn-up calculations are the basic function for the reactor nuclear design software. Three-dimensioned few-group neutron calculation software CORCA-3D was developed by NPIC on self-independence, with full intellectual property rights, and it can be applied to the nuclear design and analysis. The theory of critical boron concentration and burn-up calculations in CORCA-3D is introduced in this paper, and the functions are validated with benchmark problems, measured data and the SCIENCE system. Results show that CORCA-3D is with high accuracy.

Research on Integrated Reactor Power Control System Based on Intelligent Prediction

Zhao Mengwei, Chen Zhi, Liao Longtao, Li Yiliang, Huang Ke

2019, 40(4): 166-171.

Abstract(216)

Abstract:
In order to improve the control performance of proportional-integral(PI) control, an one-step prediction power controller is proposed based on online T-S fuzzy identification and particle swarm optimization algorithm. The simulation results show that the method is better than the original PI control in regulating response speed, overshoot and so on.

Structural Improvement of UMo-Zr Monolithic Fuel Plates

Yin Mingyang, Pang Hua, Tang Changbing, Li Yuanming, Zheng Lele, Yuan Pan, Zhao Yanli, Yue Huifang

2019, 40(4): 172-176.

Abstract(210)

Abstract:
In this study, the in-pile thermo-mechanical constitutive relation of fuel pellet and cladding was introduced into the numerical simulation based on ABAQUS, and the method of numerical simulation of the in-pile behavior of the UMo-Zr monolithic fuel plate was preliminarily established. Based on the method of numerical simulation, the effect of fuel pellet structure on the temperature field and stress field was analyzed by changing the length, the width, and the thickness of the fuel pellet and the shape of the edges. The research results indicated that the temperature field and stress field in the post-irradiation fuel pellet were sensitive to the thickness of the fuel pellet; the peak Mises stress in the post-irradiation fuel pellet was decreased by chamfering the fuel pellet.

Validation of CFD Method for Benchmark Experiment of Single-Phase 4×4 Rod Bundle Flow

Liu Luguo, Jiang Guangming, Li Songwei, Li Zhongchun, Chen Xi, Guo Chao, Yuan Hongsheng

2019, 40(4): 177-182.

Abstract(211)

Abstract:
In order to study the accuracy of the prediction of flow field distribution in bundle channels using Computational Fluid Dynamics(CFD) methods, STAR-CCM+ code is utilized to analyze the single-phase 4×4 bundle flow experiments conducted by Korea Atomic Energy Research Institute. Based on generating meshing scheme determined by meshing sensitivity study, standard k-ε(SKE), realized k-ε(RKE), standard k-ω(SKW) and SST turbulence model are adopted to simulate the bundle flow, and comparisons are conducted for the simulation results and experimental data for lateral and axial velocity. The results show that, four turbulence models can well predict velocity field distribution inside the bundle channels, the relative deviation for SKE and RKE is 19.6% to predict the lateral velocity, and SKE is better for simulating the lateral velocity analysis at zone near grids, otherwise RKE is better. For axial velocity prediction, SKE simulation is with the minimum relative deviation of 4.9%. All four models underestimate RMS velocities, but can predict RMS velocity distribution law inside the bundle channels, and RKE is suitable for near-grid zone, otherwise SST is suitable. The results provide references to the set-up of best practice guide for CFD analysis of single-phase bundle flow.

Mechanical Analysis and Experimental Research on Containment Spray Pump in Third Generation Nuclear Power Plant

Li Song, Xu Yu, Tang Huapeng, Wu Lin, Mu Keliang, Zhang Kai

2019, 40(4): 183-188.

Abstract(175)

Abstract:
In this paper, taking a containment spray pump in a third generation nuclear power plant as an example, the whole system is modeled and analyzed. The aseismic performance, modal and stress of each component, structural buckling and connection bolt stress of the system are analyzed and evaluated , and on-site vibration frequency is tested. The results show that the containment spray pump meets the requirements of the code, and the results are in good agreement with that from the equipment modal test, which provides research support for the follow-up safety assessment and equipment appraisal.

Application of Vibration Reduction Technologies for Central Air Conditioning in Cooling Water Pumps

Wang Chenguang, Tang Yanxiang

2019, 40(4): 189-192.

Abstract(139)

Abstract:
A marine cooling water pump is mainly used to deliver the cooling water, which is expected to operate with high reliability and low vibration noise. The compressors of the central air conditioning are with similar requirements and technologies. This paper mainly introduces the permanent magnet synchronous motor technology and the compressor vibration reduction method, and discuss the feasibility and prospect of its application in the marine cooling water pumps.

Study on Technology of Radiation Safety in Marine Nuclear Power Plants

Liu Shaoqiang, Zhang Hongyue, Tan Yi, Lyu Huanwen, Wang Shuang

2019, 40(4): 193-199.

Abstract(401)

Abstract:
Radiation safety technology is the fundamental guarantee for the radiation safety of marine nuclear power plants, and the radiation safety level need to be increased urgently in China. For the core technologies such as the technologies used in radioactive source item analysis and radiation shielding design, this paper surveyed the current research situation and development trend at home and abroad. From the perspective of development needs, the overall objectives, the key and difficulty and the development roadmaps of the technologies for radioactive source item analysis and radiation shielding design were proposed in this paper, which provided the direction and support for the development of radiation safety technology in marine nuclear power plants in China.

Experimental Research on Heat Transfer and Thermoelectric Characteristics of Small Nuclear Power Supply Facilities

Tang Simiao, Wang Chenglong, Su Guanghui, Tian Wenxi, Qiu Suizheng

2019, 40(4): 200-202.

Abstract(377)

Abstract:
Aiming at the shorthand of unmanned undersea vehicle (UUV) with poor endurance, combining with the technology of heat pipe reactor with silent thermoelectric conversion, a non-nuclear electric heating experiment platform is established to conduct the start -up experiment. Experiment results show that, the experimental start-up process is smooth, the isothermality is good after the start-up. The thermoelectric performance of the thermoelectric generator is good, with the efficiency of 6.8%. As the first electric heating experiment platform based on thermoelectric power generation and high temperature heat pipe heat transfer, the experimental data could provide a reference for the further design and application of the silent thermoelectric heat pipe reactor.

2019 Vol. 40, No. 4