核动力工程

Numerical Simulation of URANS and LES on Flow Field in 5×5 Rod Bundles with Split-Type Spacers

Wang Ningbo, Xiao Zejun, Zhou Lei, Zan Yuanfeng, Yan Xiao

2018, 39(6): 1-4. doi: 10.13832/j.jnpe.2018.06.0001

Abstract(227)

Abstract:
Unsteady Reyolds Average Navier-Stokes Simulation (URANS) and Large Eddy Simulation (LES) are used to study the flow field of 5×5 rod bundles with split-type spacers. An entire geometry of spacers, including grid straps and split-type vane, is simulated, and the results are compared to the experimental results of MATiS-H. The results show that URANS and LES can predict the averaged value of components of velocity, though there is a little difference between simulations and experiment. For RMS value of components of velocity, Unsteady SST k-(URANS) solution cannot reproduce the pulsation of components of velocity and the pulsation of components of velocity is under predicted by Unsteady RSM (URANS). The result of pulsation of components of velocity can be reproduced by LES while it is under predicted.

Numerical Study on Boiling Critical Characteristics of Rod Beam Channel Positioning Lattice

Dong Xiaomeng, Zhang Zhijian, Liu Dong, Tian Zhaofei, Chen Guangliang

2018, 39(6): 5-10. doi: 10.13832/j.jnpe.2018.06.0005

Abstract(475)

Abstract:
In order to investigate the influence of the spacer grids and mixing vanes on boiling crisis in nuclear reactor, the effects of number, location of spacer grid and bending angle of mixing vane on the flow and heat transfer in rod bundle are studied by using the computational fluid dynamics (CFD) analysis method. The results show that the spacer grid will cause resistance to the mainstream flow, and the peak temperature value is increased along with the number of spacer grid. However, the peak value of temperature can be restrained to a certain degree when the spacer grid is arranged in the position where boiling crisis occurs. . The mixing vane can decrease the void fraction in near wall region to improve the heat transfer effectively. But the mixing vane with too large bending angle will bring the boiling crisis in advance.

Working and Verification of ACE Libray Photonuclear Cross Section of Partial Nuclein Based on ENDF/B-VII

Li Wenqiang, Liu Bin, Han Jinsheng, Lyu Xuefeng

2018, 39(6): 11-14. doi: 10.13832/j.jnpe.2018.06.0011

Abstract(249)

Abstract:
MCNP5 program can be used to model the electronic accelerator driven sub critical system, the cross section database which used by MCNP5 that lacks photonuclear data of some radionuclides. We use the NJOY program to process seven nuclei photonuclear data from ENDFB-VII in the original photonuclear data into MCNP5 using ACE format, and use MCNP5 to build a model to calculate the photoneutron cross section, to get the curve of the photonetron cross section with the photon energy, and compare with the cross section curve of each nuclear data center. The results show that the manufacturing process of nuclide cross section is correct and most of the data processed by the nuclide are correct, but the photoneutron reaction cross sections of ⁵²Cr, ⁵⁸Ni and ⁹¹Zr were different in different databases.

Study on Flow Pattern and Flow Regime Transition Criterion in a Heating Rectangular Mini-Channel

Tian Ye, Huang Wei, Luo Hanyu, Wang Haisong, Li Pengfei, Cao Simin

2018, 39(6): 15-18. doi: 10.13832/j.jnpe.2018.06.0015

Abstract(198)

Abstract:
A visual experimental system is designed to study the flow patterns in a rectangular mini-channel. The forces on gas-liquid two phases in critical state are analyzed. A new flow regime transition criterion based on mechanical model is proposed, which is compared with the experimental data. The results show that the prediction accuracy of bubbly flow-confined bubble flow transition criterion is 93.94% and the prediction accuracy of confined bubble flow-annular flow transition criterion is 94.07%. The prediction result is in agreement with the experimental data.

Design and Research on Anti-Hang-up of Grid Based on Hydraulic Force

Chen Jie, Peng Yuan, Lei Tao, Chen Ping, Li Quan, Huang Yongzhong

2018, 39(6): 19-22. doi: 10.13832/j.jnpe.2018.06.0019

Abstract(159)

Abstract:
Grid is an important part of fuel assembly, making up of the contour of fuel assembly together with up nozzle and bottom nozzle. In order to reduce the hang-up risk of fuel assemblies during handling, the guide vanes have been set on outer strap, which is an important advance in the structural design of grids. Focused on the hydraulic environment of fuel assemblies in the core, considering the possible transmutation of outer strap caused by the lateral hydraulic force chronically acting on grid during operation, the hydraulic force on outer strap of grid is studied with CFD method in this paper.Further research and scheme design are also conducted to reduce the force on outer strap. The results show that the pressure difference between two sides of outer strap will be reduced effectively with the holes set in outer strap, and the effect will be more prominent as the holes move downstream. Meanwhile, increasing the size of holes will also further reduce the pressure difference. Moving the holes to dimples will further balance the pressure between two sides of outer strap, that is, the design of dimples with holes will lower the hydraulic force on outer strap in a greater extent. It is positive to reduce the transmutation of outer strap caused by hydraulic force with the designs above. Thereby, the hang-up risk of fuel assemblies will be reduced during refueling.

Experimental Study of Gas-Liquid Countercurrent Flow Characteristics in Horizontal Pipes with Rectangular Cross-section

Ma Youfu, Zhang Yuyan, Yue Rong, Lyv Junfu, Peng Jiewei

2018, 39(6): 23-28. doi: 10.13832/j.jnpe.2018.06.0023

Abstract(150)

Abstract:
To clarify the effects of the cross-sectional sizes and water depths on the gas-liquid countercurrent flow characteristics (CCFC) in horizontal rectangular pipes, a visualization CCFC experiment was carried out through an underwater exhaust experimental system using air and water as the gas and liquid phase. In the experiment, a rectangular plexiglass pipe in a length of 2 m and inner cross-section scale of 106×60 mm was tested under the transversal and upright layout, respectively, and each layout was tested under the water depth of 1 m and 3 m. Conclusions can be drawn as follows: (1) The height of flow channel has a significant effect on the horizontal CCFC in pipes. The liquid phase backflow can be greatly enhanced if using a higher channel height under a given cross-sectional area of pipes, which is beneficial to condensate water reflux in hot legs thus ensuring the cooling of reactor cores under the loss of coolant accident. (2) The dimensionless CCFC curves of the transversal and upright layout, which curves are expressed by the Wallis parameters of the gas and liquid phase, were coincided when the channel height was used as the characteristic scale of Wallis parameters, meaning that the CCFC of the horizontal pipes in different cross-sectional sizes could be well correlated using such a definition of Wallis parameters. (3) The influence of changing water depth on the dimensionless CCFC in horizontal rectangular pipes is insignificant in the range of water depth 1-3 m. Finally, an experimental correlation was proposed for predicting the CCFC in horizontal rectangular pipes.

Study on Heat Transfer Model of Supercitical Water Based on Acceleration Effect

Zeng Xiaokang, Li Yongliang, Yan Xiao, Huang Zhigang, Huang Yanping

2018, 39(6): 29-33. doi: 10.13832/j.jnpe.2018.06.0029

Abstract(200)

Abstract:
Supercritical-water cooled reactor is above the thermal critical point (22.1 MPa,374℃). The coolant is supercritical water which physical property changes sharply with the change of temperature. The characteristic of heat transfer deterioration of supercritical water is that the wall temperature increases slowly via heat power, which is different from CHF of pressured-water reactors. So, forecasting the cladding wall temperature of the fuel rod for heat transfer deterioration is very important for the safety analysis of SCWR. The paper infers the modified term of acceleration effect based on the boundary layer equations. The paper achieves the fitting of the correlative factor of the modified term by the test data and the semi-empirical relationship is gained. By comparing the test data, the relationship is capable to forecast the wall temperature for normal heat transfer and heat transfer deterioration.

Theoretical Analysis of Effect of Buoyancy and Flow Acceleration on Heat Transfer of Supercritical Carbon Dioxide

Liu Guangxu, Huang Yanping, Wang Junfeng, Liu Shenghui, Zan Yuanfeng, Lang Xuemei

2018, 39(6): 34-38. doi: 10.13832/j.jnpe.2018.06.0034

Abstract(268)

Abstract:
Based on the basic behaviors of the boundary layer, this paper analyzed the effects of buoyancy and flow acceleration resulted from the significant changes of thermal properties near the pseudo-critical region on the heat transfer of supercritical fluids near the wall. The theoretical thresholds of heat transfer deterioration, which resulted from buoyancy force and flow acceleration, were developed. Results showed that, for the heat transfer in the vertical flow, both buoyancy and flow acceleration would decrease the shear stress near the wall, and finally retrain the production and diffusion of turbulence. The thresholds of heat transfer deterioration resulted from buoyancy force and flow acceleration were Bu=1.16×10^-5 and Ac=2.91×10^-6, respectively. These thresholds agreed well with experimental data.

Study on Energies Selection Principle of Multi-Energies Gamma-Ray Transmission Technique

Luo Ji, ong, Xu Guiping, Wang Xuequan, Liu Xiaojun

2018, 39(6): 39-42. doi: 10.13832/j.jnpe.2018.06.0039

Abstract(541)

Abstract:
Up to now the energies selection principle for multi-energies gamma-ray transmission technique was ambiguous, and the research object of this paper was to extract a energies selection principle for it. The relationship between measurement precision and ill-equations has been demonstrated in this paper, and a energy selection principle has been got based on these analysis, which is to chose the energy parameters by determining the condition number of the attenuation coefficient matrix, and the condition number for engineering application proposal is less than 100. Meanwhile, a series experiments based on U and Zr mixture have been used to verify the principle. The results show that smaller condition number of the attenuation coefficient matrix lead to high measurement accuracy, and the relative measurement accuracy can be controlled within 5% based on the precondition of condition number smaller than 100.

Design and Experimental Study of Irradiation Temperature in Fuel Specimen

Yang Wenhua, Zhang Liang, Si Junping, Nie Liangbing, Tong Mingyan

2018, 39(6): 43-48. doi: 10.13832/j.jnpe.2018.06.0043

Abstract(602)

Abstract:
An investigation on the irradiation temperature and irradiation test of fuel specimens was performed in the High Flux Engineering Test Reactor（HFETR）, for conducting fuel irradiation test in the worst design condition. The test results show that the axial factor of heat flux density on fuel specimen surface can be decreased by adjusting the fuel specimen locations in the irradiation rig with its uranium content. A high irradiation temperature level of fuel meat and fuel cladding in the fuel specimens was obtained by coating the fuel specimens with a layer of liquid lead-bismuth alloy and a stainless steel container, while the specimens were cooled with a rather low temperature coolant of HFETR. The test results show that the stainless steel surface temperature is lower than the limit of maximum cladding surface temperature of the HFETR fuel element under the steady state condition or a short-term transient operation condition, satisfying the requirements of reactor operation and fuel irradiation test. To decrease the peak factor of fuel specimen temperature in the irradiation channel and then to increase its irradiation temperature under the steady state operation condition, the short-term transient power variation of fuel specimens caused by the reactivity disturbance in the irradiation rig should be avoided or reduced by optimizing the core arrangement design and reactor operation plan

Seismic Analysis of Nuclear Auxiliary Equipment Subjected to Multi-Nozzle Loads

Du Kun, Ding Menglong, Wang Xiaofeng

2018, 39(6): 49-52. doi: 10.13832/j.jnpe.2018.06.0049

Abstract(223)

Abstract:
A method was presented to effectively obtain the maximum calculation stress for the multiformity of nuclear auxiliary equipment nozzle loads. Based on ANSYS platform and APDL language, the special calculation modules were developed for seismic analysis and stress evaluation of nuclear auxiliary equipment subjected to mulit-nozzle loads, and the convenient GUI was customized. An engineer example confirmed that the calculation modules were reasonable and effective.

Sensitivity Analysis for Seismic Fragilities of Lateral Supports of RPV for High Temperature Gas-Cooled Reacto

Jiang Zhuoer, Wang Haitao, Zhao Jun, Shi Li

2018, 39(6): 53-58. doi: 10.13832/j.jnpe.2018.06.0053

Abstract(203)

Abstract:
The lateral supports of reactor pressure vessel (RPV) are expected as a major contributor to the seismic risk of high temperature gas-cooled reactor (HTR) and are crucial to the safety of HTR RPV under seismic events. In this paper, we focus on the identification of seismic fragility variables, the analysis of reasonable factor values of fragility variables, the calculation of High Confidence Low Probabilistic Failure (HCLPF) Capacity and seismic fragility Curves, the selection of the key parameters among fragility variables, and the exploration into the sensitivity of HCLPF to these key parameters of seismic fragility analysis. The results show that the HCLPF of lateral supports is much higher than the design basis earthquake level and that it is insensitive to the variability of the key parameters.

Application of Rapid Fatigue Analysis Method in Fatigue Monitoring System of Nuclear Power Plants

Chen Rong, Liu Xin, Zhang Guihe

2018, 39(6): 59-63. doi: 10.13832/j.jnpe.2018.06.0059

Abstract(286)

Abstract:
The fatigue monitoring system collects the operation parameters of the primary loop and equipment prone to heat fatigue, and fast fatigue analysis method was used to conduct the real-time fatigue calculation of the pipe and equipment which were monitored, to obtain the real fatigue damage. The fast fatigue analysis method was based on the Green function method, and the fast calculation of thermal stress and fatigue coefficient is conducted by compiling the calculation program. By comparison with the result of finite element analysis, the fast fatigue analysis method is proved to be efficient, fast and accurate.

Optimization Analysis about Pre-Tightening Technology of Stud for Reactor Pressure Vessels

Wang Xiaobing, Long Tao, Fan Yijun, Wen Xiaojun

2018, 39(6): 64-68. doi: 10.13832/j.jnpe.2018.06.0064

Abstract(267)

Abstract:
The pre-tightening process of the stud for reactor pressure vessels with three steps used for stretcher which are orthogonal has been studied by software ANSYS, and the optimization analysis have been carried out. The numerical results showed as follows: The studs which were loaded symmetrically at the first step, at interval at the second step and sequentially at the last step have small non-uniformity and dispersion. The non-uniformity and dispersion can be decreased obviously when the studs were with variable loads at the last step, thus to strengthen the sealing of the reactor pressure vessel.

Research on Automatic Control Method and Process Optimization for Second Circuit Dosing System of High Temperature Gas-Cooled Reactor

Zhang Ruixiang, Zhang Lintao, Meng Yingqi, Wang Hongyu, Gao Jinghui, ZhangYafu

2018, 39(6): 69-73. doi: 10.13832/j.jnpe.2018.06.0069

Abstract(183)

Abstract:
There is no relevant criterion for the second circuit water quality of the high temperature gas-cooled reactor (HTGR) and for the dosing control indexes. Based on the research of the material and operating conditions of the second circuit, referring to the operation experience of PWRs and Once-through Boiler, when the pH of the second circuit water was set between 9.5~9.8 and the dosing hydrazine was controlled in 80~120 μg/L, better corrosion prevention effect can be achieved in the second circuit. Based on the design status of the dosing system for the second circuit of high temperature reactors, this paper put forward the optimization and modification scheme of dosing system design, using the hydrazine meter instead of dissolved oxygen meter, calculating pH by conductivity, thus to avoid the dissolved oxygen and pH meter lag and instability. Finally, by improving the control method, the automatic and precise dosing of ammonia and hydrazine in the second circuit of HTGR can be realized.

Study on Risk Analysis and Mitigation Measure for Loss of Heat Sink Accident

Pei Liang, Zhou Shengwen

2018, 39(6): 74-78. doi: 10.13832/j.jnpe.2018.06.0074

Abstract(184)

Abstract:
This paper analyzes and evaluates the risk of total loss of heat sink accident of a Chinese second-generation pressurized water reactor (PWR). Possible weak links during the accident mitigation are identified and corresponding suggestions of improvement are provided in order to mitigate the risk of total loss of heat sink for this NPP.

Causal Analysis and Resolution of Main Feedwater Pump Frequency-Doubled Vibration of Tianwan NPP Units 3 and 4

Zhao Di, Chen Xiaomeng, Wang Xiufeng, Sun Zhongzhi, Zhou Zhengping, Zhou Zhijun, Hu Guanghui

2018, 39(6): 79-80. doi: 10.13832/j.jnpe.2018.06.0079

Abstract(130)

Abstract:
The main feedwater pump frequency-doubled vibration occurred in Tianwan NPP units 3 and 4 in the process of commissioning. The frequency-doubled is approximately equal to the three times frequency of the pump structure according to the result of modal test. The three times frequency of the pump structure is increased by reinforcing the pedestal of the main feedwater pump. According to the test result, the vibration of the bearing pedestal is reduced obviously. The reinforcement results in good effect.

Development and Research of Severe Accident Simulator Based on Core Model Coupling Transition Method

Cao Ying, Zhao Xiumei, Zhang Yu, Lin Meng

2018, 39(6): 81-85. doi: 10.13832/j.jnpe.2018.06.0081

Abstract(131)

Abstract:
This paper presented a codes coupling method of system analysis code and integrated severe accident (SA) code to develop a new SA simulator. It uses system analysis code to calculate early stage of an accident, when satisfying coupling condition, system analysis code stops calculation and switches to SA code to simulate mid-late-term accident. In order to achieve smooth transition of parameters, this paper took full scope simulator general code RELAP5 and SA code MAAP4 as example to analyze their repetitive thermal-hydraulic models, especially the core model. Core fuel temperature, cladding temperature and core power were selected as transition parameters. Based on a same small break loss of coolant accident (SBLOCA) of generic million kilowatts of Pressurized Water Reactor (PWR), coupling transition calculation and SA code independent calculation were compared to test the method. The simulation results showed that this selection of transition parameters is correct. This method can not only smoothly transmit the parameters but also ensure accuracy of subsequent calculation.

Shield Building with Inter-Story Isolation for AP1000 Nuclear Power Units and Its Seismic Performance Study

Hou Gangling, Zhao Can, Wang Xiaodong, Sun Hai, Zheng Gang, Chen Yaodong, Shen Feng

2018, 39(6): 86-91. doi: 10.13832/j.jnpe.2018.06.0086

Abstract(225)

Abstract:
Based on the features of the shield building for AP1000 nuclear power units, the optimal model of the shield building for stiffness and damper is given, and the new structural system with inter-story isolation is established by modifying the connection between different parts of the shield building. By comparing with the seismic responses of the traditional model and the base isolation model, the different vibration mechanism of the inter-story isolation model is shown obviously, and its earthquake resistant behaviors (i.e., reduction effect stability, anti-seismic robust, and adaptive site) are presented.

Development of Dry Circular Mechanical Cutting Device

Hu Dongmei, Peng Jing, Zhang Bin, Liu Xiaoqiong

2018, 39(6): 92-95. doi: 10.13832/j.jnpe.2018.06.0092

Abstract(117)

Abstract:
The spent fuel that encapsulated in the spent fuel storage canister shall be withdrawn in its reprocessing. Based on the circumstance of hot cell and the structure characteristics of spent fuel storage canister, as the radiation resistant design, the fixation of the storage canister, cutting feed, turning tool and its replacement, and rad-waste minimization shall be considered in the design. A dry circular mechanical cutting device is developed to open the spent fuel storage canister in the hot cell. Function tests verify that the cutting device meets the requirements of its design and application.

Assessment on Wear of In-Core Flux Thimble in a Typical Pressurized Water Reactor

Zhang Mingqian, Huang Meiliang, Fu Yueming, Chu Qianqian

2018, 39(6): 96-100. doi: 10.13832/j.jnpe.2018.06.0096

Abstract(309)

Abstract:
The wear damage of the in-core flux thimble was often reported from the advanced pressurized water reactor （1000 MW）in China. In order to improve our understanding of the wear phenomena and obtain the factors which influence the result of wear, 36 groups of wear data for the in-core flux thimble are collected from the nuclear power plants in service, which are measured by eddy current examination during the plant outage. The main conclusions obtained on the basis of the analysis about the wear data are as follows: The wear rate of the in-core flux thimble is directly related with the type of the support column and the tie plate. In-core flux thimbles located closely at the area of the zero azimuth in the reactor could be more subjected to worse wear. The wear defect-growing trend increases significantly at the initial stage of the plant operation, and moves toward stabilization with it. The wear damage reported from the plants constructed in recent years is gradually deteriorating.

Theoretical Calculation and Modeling Analysis of Pump Comprehensive Performance Test of Auxiliary Feed Water System

2018, 39(6): 101-103. doi: 10.13832/j.jnpe.2018.06.0101

Abstract(116)

Abstract:
When the results of the pump comprehensive performance test of the auxiliary feedwater system (ASG) of CPR1000 do not satisfy the supervision requirements, the throttle elements in the pipeline system need to be adjusted. This paper gives a brief description of the adjustment principle of the throttling element. The methods of the theoretical calculation of engineering fluid mechanics and computational fluid dynamics (CFD) provide two ways to obtain the adjustment of the throttling element results. The comparison between two analytical methods shows that the theoretical calculation results are consistent with the modeling analysis results. According to the respective characteristics of theoretical calculation and modeling analysis, the most convenient method can be selected according to the actual needs.

Study on Condition Detection of Major Equipment in Nuclear Power Plants Based on Fuzzy Synthetic Assessment

Shen Jiangfei, Pan Tiancheng, Mao Xiaoming, Wu Tianhao, Gu Fang

2018, 39(6): 104-110. doi: 10.13832/j.jnpe.2018.06.0104

Abstract(194)

Abstract:
Focusing on the major equipment management detailed rules, equipment operation characteristics and requirements of nuclear power plants, a comprehensive evaluation method for major equipment health status of nuclear power plants is proposed. According to the characteristics of equipment monitoring of nuclear power plants, a multi-level indicator system model based on monitoring tasks is established. Based on the potential failure modes of equipment components, the fault phenomena and the equipment monitoring tasks are analyzed, and the monitoring tasks membership function models are constructed. Summarizing the practical assessment experience of the experts, this paper presents that indexes weight are shared by the multiple monitoring tasks, the weights are inherited by the most severely degraded monitoring task, and the indicator status of the equipment is obtained. Through the improved Analytic Hierarchy Process, the initial weights of each indicator are assigned, and the hierarchical variable weight theory models based on the indicator status level are proposed with a balanced consideration of the key indicators deterioration. The established evaluation method is applied to the health status assessment of the nuclear island main pump shaft seal system. The results show that the method is reliable and practical, and can effectively characterize the actual operational health status of major equipment.

Study on On-Site Calibration Technology for Fixed Ambient Dose Equivalent Radiation Ratemeters

Gao Fei, Xu Yang, Xiao Xuefu, Ni Ning, Hou Jinbing

2018, 39(6): 111-115. doi: 10.13832/j.jnpe.2018.06.0111

Abstract(204)

Abstract:
Fixed ambient dose equivalent radiation ratemeter is widely distributed in the nuclear power plants, for conventional continuous monitoring or nuclear accident emergency monitoring. It is not easy to disassemble the fixed installation and send it to the metrology laboratory for calibration. In order to ensure the accuracy of fixed ambient dose equivalent radiation ratemeter, the Monte Carlo method has been used for the design of ambient dose equivalent secondary standard ionization chamber and portable gamma ray irradiation device, and on-site calibration experiment has been carried out. The on-site calibration factors have been compared to the calibration factors obtained in laboratory, and the results showed that on-site calibration problems can be solved using ambient dose equivalent secondary standard ionization chamber and portable gamma ray irradiation device.

Research of Variance Reduction Techniques for Large Space Dose Calculation

Wang Song, Yang Yong xin, Lu Chang bing, Chen Ying feng

2018, 39(6): 116-121. doi: 10.13832/j.jnpe.2018.06.0116

Abstract(143)

Abstract:
In large space dose calculation such as nuclear material stockpile, the calculation using Monte Carlo method without variance reduction techniques requires long calculation time and with large error.. A variety of variance reduction techniques such as Weight-Window, Exponential Transform, Forced Collision, DXTRAN Sphere are tested and compared. Then an optimum combination of variance reduction techniques is proposed for the point calculation based on the detector method and for monolithic calculation based on mesh method in large space. Results indicate that the combination of weight windows used forced collision and Exponential Transform with Truncation can greatly improve the efficiency, especially for those points away from the radioactive source.

A Long Distance Monitoring Method for Strong Radiation Based on Wide Range Detection Technology

Liu Zhiqiang, Ma Yan, Liu Zhiyong, Gao Jing

2018, 39(6): 122-125. doi: 10.13832/j.jnpe.2018.06.0122

Abstract(158)

Abstract:
Aiming at the characteristics that the electronic components are easy damaged in strong radiation environment, a long distance detection method for strong radiation based on wide range detection technology is proposed. In this method, the GM counter is placed in the strong radiation environment, the measuring circuit is placed outside of the strong radiation environment, and a long cable is applied to connect the counter and the circuit. Theoretical analysis and experimental results show that it can improve the measuring range and service life of the counter, and can avoid the damage of circuit due to radiation. Therefore the method can be applied to the long distance monitoring in strong radiation environment.

Research on Crew Dose Assessment System for Large Nuclear Power Ships

Yu Hong, Li Lan, Cheng Shisi, Yang Shuqi

2018, 39(6): 126-131. doi: 10.13832/j.jnpe.2018.06.0126

Abstract(168)

Abstract:
Establishing a crew dose assessment system is one of the key techniques to be solved urgently in the development planning of the nuclear power mechant ships in China. Research was carried out on large nuclear-powered ships with the largest number of crew members, the most complex staffing and the widest coverage of radiation protection. A dose assessment system for crew memebers on large nuclear power ship is given in this paper. The dose assessment system is constructed according to the methods recommended by International Commission on Radiological Protection and International Atomic EnergAgency, including basic dose quantities, dose limits and reference levels, and dose calculation and measurement, covering planned exposure and potential exposure, which can meet the demand of crew dose assessment in forecast dose, actual dose and review. Further more, the detailed measures used to estimate the crew dose is presented in the dose assessment system, such as the radiation sources and exposure pathways need to be considered, the model, quantity and values should be applied.

Research on Cyber Security Technique of Safety DCS Gateway in Nuclear Power Plants

Liu Mingxing, Ma Yu, Zhao Xinyan, Jiang Wei, Huang Jun

2018, 39(6): 132-136. doi: 10.13832/j.jnpe.2018.06.0132

Abstract(340)

Abstract:
In view of the information security problems that Modbus/TCP protocol is facing, this paper discusses 5 kinds of information security services which are lacked in the protocol, and describes the threats on the information safety brought by the lack of the services. To solve the problems, this paper designs a practical measure in Modbus/TCP information security defense for cyber security in Nuclear Advanced Safety Platform Instrumentation and Control (NASPIC), including access control, integrity check, encryption certification and safety alarm.

Study on Factors and Countermeasures in Application of Pressure Transmitter and Differential Pressure Transmitter in Research Reactors

Li Linhong, Ge Yuan, Lin Jianhua

2018, 39(6): 137-140. doi: 10.13832/j.jnpe.2018.06.0137

Abstract(156)

Abstract:
Pressure transmitter and differential pressure transmitter are widely used in the parameter measurement in research reactors, however the improper application will have negative effect on the safety of the process system and the safety of personnel and equipment. In addition, if the measurement signal is unreliable or the equipment fails, it may cause serious safety accidents and economic losses to the research reactors. In this paper, the factors and countermeasures that should be considered in the application of pressure transmitter and differential pressure transmitter are studied. It is proposed to design the primary instrument room and the reverse flushing system, and use the disassembly free method for instrument calibration in order to achieve reliable measurement and efficient maintains, thus to make sure the operation and maintenance of transmitter meet the requirement of safe operation and effective utilization of reactor.

Discussion on Reviewing of Reliability Assurance Program for Nuclear Power Plants

Zhang Wenguang, Yang Chenggang, Sun Zhan, Zhang Yue

2018, 39(6): 141-145. doi: 10.13832/j.jnpe.2018.06.0141

Abstract(157)

Abstract:
Reliability assurance program for the nuclear power plant plays an important role in promoting the reliability, availability, maintainability and economy of equipment. During the design process, the feasible reliability performance indicators of SSCs can be defined through the analysis and categorization of risk-important SSCs, and the accumulated reliabilities of SSCs can be highly improved by the strict quality control means during the following phases, thus to improve the safety and economical performances of nuclear power plants. By analyzing the requirements of the reliability laws and regulations at home and abroad, and embracing the latest research results and technical insights, the key elements of reliability assurance program for designing were given out, the relevant reviewing problems were analyzed and discussed, in addition some suggestions were put out for designing units and regulation agencies.

Design for Reconstructionof A Sealed Radiation Source Workshop

Liu Huping, Guo Shenghui, Wang Yueyong, Li Xiu, Gao Pengjie

2018, 39(6): 146-150. doi: 10.13832/j.jnpe.2018.06.0146

Abstract(80)

Abstract:
Some potential hazards occur in a sealed radioactive source workshop, because of the inadequate consideration in the original design, the improper construction of supporting facilities and insufficient maintenance. The design principles and key factors are studied from the factors such as process layout, radiation protection and monitoring, radioactive liquid collection and utility systems, and a suitable solution is put forward, including optimizing personnel traffic routes and radioactive material transport routes, enhancing the shielding and monitoring of hot cells, and reinforcing the classification and collection of radioactive liquid waste. In addition, the ventilation, decoration, automatic control and fire protection systems are improved. With these measures above, the potential safety hazards are removed, exposure risks are reduced, applicable statutory and regulatory requirements and follow-up production demands are satisfied, to protect the public and the environment properly.

Research on Application of Maintenance Rule in Domestic Nuclear Power Plants

Cheng Bin, Chen Yu, Zhang Sheng

2018, 39(6): 151-155. doi: 10.13832/j.jnpe.2018.06.0151

Abstract(230)

Abstract:
In order to meet the urgent need of domestic nuclear power plants for maintenance effectiveness evaluation system, the core content and specific implementation process of US Maintenance Rule (MR) are studied and analyzed. Combined with the actual situation of nuclear power plants in China, it is suggested to continuously improve and innovate, and develop a maintenance effectiveness evaluation system suitable for the characteristics of domestic nuclear power plants.

Application of Ant Colony Optimization Least Squares Support Vector Machine in Measurement Data Fitting

Jiang Botao, Hines J. Wesley, Zhao Fuyu

2018, 39(6): 156-160. doi: 10.13832/j.jnpe.2018.06.0156

Abstract(356)

Abstract:
Aiming at the disadvantages of traditional data fitting methods, such as relying on the user’s experience and needing to predetermine the estimated fitting function, a data fitting method based on Ant colony least squares support vector regression(ACO-LSSVR) is proposed. The method uses ant colony optimization (ACO) to optimize the parameters of least squares support vector regression machine (LSSVR) and obtain the optimal parameters to establish a data fitting model. This method is used to fit the measured data of nuclear engineering with the traditional regression fitting method. The core power curve and the melt droplet movement characteristic curve in coolant are obtained. The fitting results of the two curves are compared. Results show that ACO-LSSVR has high fitting accuracy and does not need to determine the fitting function of data segments.

Research of Characteristics of 14Cr17Ni2 Martensitic Stainless Steel Forging Used for Reactor Internals

Wang Qingtian, Luo Ying, Du Hua, Duan Chunhui, Wang Liubing, Hu Chaowei, Wang Zhonghui, Chen Xin

2018, 39(6): 161-166. doi: 10.13832/j.jnpe.2018.06.0161

Abstract(150)

Abstract:
The paper analyzes the reasons of cracks occurred twice in the forging of 14Cr17Ni2 Martensitic Stainless Steel forging used for the reactor internals. Through the calculation of phase diagram, and the experimental research of the influence on the content and pattern of δ ferrite in different heating temperature, different holding time and forging temperature, the paper concludes that the higher the heating temperature, the longer the holding time, the lower the forging temperature, the more easy the cracks appearance. Results show that no crack appears in the forging after the improvement of the heating temperature, holding time and finish-forging temperature.

Effect of 3-Layer Corium Pool Configurations on Heat Flux Distribution of RPV Outside Wall for AP1000

Liu Lili, Yu Hongxing, Chen Liang, Deng Jian, Deng Chunrui, Xiang Qing’an, Zou Zhiqiang

2018, 39(6): 167-171. doi: 10.13832/j.jnpe.2018.06.0167

Abstract(199)

Abstract:
The corium in the lower plenum mainly containing U-Zr-O-Fe may separate into two immiscible liquids. However, there are significant differences between existing 3-layer models. The configurations of the pool in the lower plenum for AP1000 is assessed by three well-accepted 3-layer models developed by Esmaili & Khatib-Rahbar, Seiler, Salay & Fichot respectively. In addition, the heat transfer in the pool for these configurations is calculated in present study. The results show that the configuration which has the higher heat flux on the vessel contacted with the light metallic layer is obtained by the model developed by Seiler compared with Esmaili & Khatib-Rahbar model. The 3-layer model implemented in MAAP5 code is based on the Salay & Fichot model to treat the compositions in oxide and heavy metallic layers, but assumes that the light metallic layer forms automatically after the corium falls down, which should be developed further.

Study on CHF Mechanistic Model in Upflow Boiling Vertical Round Tube under High Pressure

Liu Wei, Peng Shinian, Jiang Guangming, Liu Yu, Shan Jianqiang

2018, 39(6): 172-177. doi: 10.13832/j.jnpe.2018.06.0172

Abstract(217)

Abstract:
According to the characteristics of DNB type CHF under high pressure, the constitutive correlations of Weisman & Pei model are proposed. The prediction results of three entrainment and deposition correlations of Kataoka, Celata and Hewitt corresponding to the Dry-out type CHF are analyzed. Based on the two improved models above, a comprehensive CHF mechanistic model under high pressure condition combined the DNB and Dry-out type boiling crisis is established. The verification based on the experiment database of upflow boiling in vertical round tube and the parametric trends analysis of CHF with thermal-hydraulic and geometric parameters are carried out.

Vibration Fault Diagnosis and Analysis of Fan In Nuclear Fuel Workshop

Yang Ji

2018, 39(6): 178-181. doi: 10.13832/j.jnpe.2018.06.0178

Abstract(150)

Abstract:
The abnormal vibration phenomenon was discovered by monitoring the state of the 2# fan located in the nuclear workshop in Unit 1 of Fangchenggang Nuclear Power Plant. In order to capture the specific fault frequency signal, it is necessary to optimize the vibration data collection definition according to the fan structure style. The spectral analysis and failure diagnosis found that the existence of beat vibration is the reason that the motor vibration exceeds the standard. The several potential equipment defects were identified successfully. According to the diagnostic results, an exact treatment scheme was provided as a preventive maintenance. It effectively reduced the vibration level, and ensured the healthy and stable operation of the equipment.

Design of Charge Signal Generator Based on DDS Technology

Wang Lei, Li Xiang, Hu Jianrong

2018, 39(6): 182-185. doi: 10.13832/j.jnpe.2018.06.0182

Abstract(178)

Abstract:
The charge signal generator was designed by using Field Programmable Logic Gate Array(FPGA), Digital to Analog Conversion (DAC) circuit and charge conversion circuit. The Direct Digital Frequency Synthesis (DDS) was used to realize the frequency modulation, amplitude modulation and phase modulation of the signal. At the same time, the operating principle of the DDS technology and the design ideas of the analog circuit were described in detail. Under the Modelsim software platform, the function simulation of the program was carried out, and the real-time signal was captured and displayed by the SignalTap II logic analyzer. The correctness of the program was proved by the simulation results and the actual waveform. The voltage signal was converted into the charge signal, and joint debugging with the charge converter and the conditioning amplifier. Tests showed that the output waveform can meet the design specification and this proves the validity and reliability of the charge signal generator. The charge signal generator can be used as a standard signal source for the installation and commissioning of loose parts and vibration monitoring system, and it can be used for overhaul inspection of nuclear power stations.

Analysis and Optimization of Reactor Transient Fault Caused by Protection System Switch Fault

Guo Jiaxu

2018, 39(6): 186-188. doi: 10.13832/j.jnpe.2018.06.0186

Abstract(150)

Abstract:
When the analog channel test of reactor protection system(T1 test) in Qinshan Phase Ⅱ Nuclear Power Plant, the transient fault of the unit was caused by the failure of the switch. In this paper, based on the transient fault caused by the switch fault during the test, the switch body is analyzed, and the method of switch switching verification is proposed. Combined with the work practice, the optimized maintenance strategy is put forward, which effectively ensures the safe and stable operation of the unit.

Typical False Alarm Analysis and Processing of Loose Parts Monitoring System in Qinshan Nuclear Power Plant

Zhou Xing, Du Congbo

2018, 39(6): 189-193. doi: 10.13832/j.jnpe.2018.06.0189

Abstract(301)

Abstract:
Loose parts monitoring system（LPMS） is a basic security tool for monitoring loose parts in primary loop. One of the problems of detecting loose parts is false alarm. In order to identify false alarm events, this paper analyzed the alarm data and working condition and got the conclusion: if the event happened during the process of reactor coolant pump start and stop; event caused by lightning storm; event caused by the switch of fan on top of the reactor vessel; event happened during the temperature and pressure increasing of the reactor, these alarms are false alarms. The solution for these kinds of typical false alarms is given.

2018 Vol. 39, No. 6