Introduction
Supervisor: China Association for Science and Technology
Organizer: Chinese Nuclear Society
Publisher: Editorial Board of Nuclear Science and Engineering
Address: No. 43 Fucheng Road, Haidian District, Beijing, China
Tel:010-88828647/1576
E-mail: hkxygc_gz@aep.com.cn， hkxygc@aep.com.cn
Postal Issue No. 82-603
ISSN：0258—0918
CN：11—1861/TL
Approximate Solution of Flow Distribution for Parallel Tubes in Fuel Bundles
ZHOU Lei;ZHENG Jiantao;XU Jianjun;YAN Xiao;The annular and plate fuel assembly have been applied in new research reactors at home and abroad. Because of the compact structure, the power density and neutron flux can hence be effectively increased. The coolant passages are narrow channels that are parallel to each other, and the flow distribution between the channels is a key factor in determining the power distribution on both sides of the fuel element and the maximum temperature inside the fuels, as well as the focus of hydraulic experiments and the design of the fuel assembly structure. Based on the conservation equation of fluid mass and momentum, the approximate solution of the flow distribution result of the parallel narrow channel was given by reasonable simplification and mathematical derivation. Combining the geometric characteristics of the coolant channels, the simplified form of theoretical solution is further given for convenience. The research results can provide a reference for the optimal design of the fuel assembly of the relevant reactor, and can also provide theoretical guidance for the analysis of hydraulic experiments and verification of the numerical calculation program.
Numerical Research of Particles Deposition in Rod Bundle
WANG Suhao;GUO Liang;YUE Nina;LI Ying;XIAO Hui;LOU Ruifan;ZHUO Wenbin;Liquid lead-bismuth eutectic(LBE) will cause corrosion, abrasion and erosion in contact materials under the flowing state, resulting in insoluble particles produced in coolant. The fuel assembly is a typical area where insoluble particles are generated. In order to avoid the decrease of heat release ability caused by particles deposition, it is necessary to carry out a numerical analysis on the deposition characteristics of particles in the fuel assembly. The Fluent DPM model was used to simulate the deposition characteristics of particles in liquid LBE rod bundle channels. The effects of particle size, particle inlet volume fraction and liquid LBE flow rate on the deposition characteristics were obtained. The results show that the deposition ratio and deposition rate of particles in the rod bundle channel increase first and then decrease with the increase of particle size. The inlet volume fraction and inlet mass flow rate have little effect on the deposition ratio and deposition rate of small particles in the rod bundle channel. Small size particles are more difficult to deposit in the rod bundle channel, and the main deposition area is concentrated at the interface between the rod and the wire-wrapped, and with the increase of particle size, the amount of particle deposition at the rod bundle outlet also increases.
Study on the Thermal Resistance Model and Simulation Method of Sodium Heat Pipe Startup Process
ZHANG Jie;LIU Fudong;WANG Kunpeng;AN Jieru;ZUO Jiaxu;The startup process of high temperature heat pipe usually needs to be divided into multiple stages and studied by different simulation methods. In this paper, some simulation methods are combined to form a unified simulation scheme. The experimental test system of high temperature sodium heat pipe was built using simscape, a multiphysics modeling tool of MATLAB/Simulink. The heating process from room temperature to operating temperature(>800 K) was simulated. The startup stage of the heat pipe was calculated by the equation of the flat-front startup model, and the steady state operation stage was solved by a thermal resistance model. Comparing the calculation results with the reference data, it is found that, the two methods are in good agreement. It shows that the method introduced by this paper can simulate the steady state and transient processes of the heat pipe system to a certain extent. And it is the characteried by less resource occupation and shorter time consumption in the calculation process, which provides a new solution for the analysis of heat pipe system.
Calculation of Stress Intensity Factor Method in the Ligament of Perforated Head of Reactor Pressure Vessel
ZENG Xin;LIU Fangwei;TAN Jianping;LIU Changjun;LIU Pan;WANG Dasheng;LI Yue;ZHANG Ruikai;WANG Long;The ligament of perforated head of reactor pressure vessel head is prone to defects due to the complexity of its structure and load. In order to ensure the structural integrity of reactor pressure vessel during its service life, safety evaluation of defective parts needs to be carried out, in which the stress intensity factor is a key parameter required for evaluation. The finite element method was adopted to calculate the stress intensity factors of semi-circular cracks in the ligament of perforated head of reactor pressure vessel under pressure loading and thermal loading, respectively. The finite element results were compared with the influence function method proposed by Chapuliot. The results show that if the crack depths are small, the Chapuliot method can be conservatively used to calculate stress intensity factors. As the crack depths are larger than 0. 2 times of head thickness, the Chapuliot method is no longer suitable. In order to obtain a more accurate engineering calculation method for the stress intensity factors of cracks in the ligament of perforated head, based on the finite element solutions and the influence function method in the RSE-M standard, the influence coefficients of the central points and the surface points of the semi-circular cracks with different sizes are fitted under pressure loading and thermal loading.
Research On the Safety Classification Method of a Metal Fast Reactor
CHE Juan;As the fourth generation reactor, there is no technical standard of the safety classification for a metal fast reactor. The two main domestic safety classification methods(GB/HAD) are based on pressurized water reactors but the methodology itself is universal. In order to analyze the applicability and advantages and disadvantages of these two methods in a metal fast reactor, this paper carries out the research on the safety classification method of a metal fast reactor. Through the classification of normal operation system, special safety system and support system, the two methods are compared in the classification logic, classification process and technical specification selection. The research results show that the two methods are suitable for a metal fast reactor, but the GB method has clearer logic, more scientific and rigorous classification process, and the classification is justified which is more suitable for the forward safety classification of metal fast reactor. This paper also proposes a safety classification system for metal fast reactors based on the GB method, through which the classification work of a metal fast reactor can be carried out, which can reflect the design characteristics of a metal fast reactor, ensure that the safety important items are compatible with the importance of the safety functions they perform, and the results of the classification can guide the research and development design of a metal fast reactor.
Optimization Model of Measurement Uncertainty of Leakage Rate of Nuclear Power Plant Containment
LIU Mingmei;LI Jianfa;HUA Yongzhen;The NB/T 20018 standard does not provide a model for the Type B measurement uncertainty of the containment leakage rate. In order to perfect the theoretical model concerning the containment leakage rate within this standard and to further enhance its instructiveness in engineering experiments， a Type B uncertainty analysis model for the containment leakage rate has been proposed in this study. Firstly， based on the uncertainty theory in the JJF 1059 metrological standard， a theoretical model for Type B measurement uncertainty was developed for the containment leakage rate calculation formula specified in NB/T 20018. Subsequently， experiments were conducted with sensors deployed in a 1000m3 steel containment mock-up to validate the effectiveness of the model and to explore the factors influencing uncertainty. Furthermore， the uncertainty model proposed in this paper was utilized to analyze the pressure test data of the Hualong One containment， thereby verifying the accuracy of the theoretical model in engineering-scale tests. The research results indicate that the combined uncertainty calculated using the model proposed in this study is consistent with the combined uncertainty calculated using the standard method， with a relative deviation of only 6%. Type A uncertainty is found to be related to the gas state inside the containment and the duration of the test， while Type B uncertainty is associated with the uncertainty of the instruments and their quantity. The conclusions of this study can serve to provide theoretical support for the improvement of the NB/T 20018 standard and promote the independent development of containment sealing test technology.
[Downloads: 0 ] [Citations: 0 ] [Reads: 0 ] HTML PDF Cite this article
Analysis and Optimization of Load Linear Variation Test for a GEN-III Nuclear Power Plant
Song Fei;Liu Shuangjin;Guo Jichun;Qiao Jianfeng;The power variations test is an important test to evaluate the load tracking performance of nuclear power plants during the commissioning stage， and it is one of the key and challenging aspects of the reactor control system transient tests. During the power variations test in the first unit of the GEN-III nuclear power plant， due to the excessive downward insertion of the G-rod， the R-rod did not insert sufficiently in the load reduction process， and during the load increase process， the R-rod prematurely rose to the top of the reactor. When the nuclear power reached 94%FP， it triggered the predictive alarms of C22 ( over-cooling protection of the primary loop) and C21 (axial power deviation protection). Through data comparison and analysis across multiple power plants， the cause was identified， and measures were formulated to optimize the power control rod position calibration. During the start-up phase of the second unit of this nuclear power plant， the risk of over-cooling in the primary loop during the power variations test was effectively avoided， enhancing the load tracking performance of the plants. This provides a technical support for improving the load tracking performance of subsequent units of this reactor type.
[Downloads: 0 ] [Citations: 0 ] [Reads: 0 ] HTML PDF Cite this article
Uncertainty evaluation of americium dioxide impurity element determination based on ICP-AES
ZHANG Zhiqiang;LI Yanping;LI Yuwen;ZHANG Qinggui;ZHAO Jing;LIU Chun;The mathematical model was established， and the uncertainty sources of 14 impurity elements Al， B， Ba， Be， Cd， Cr， Cu， Fe， Li， Mg， Mn， Ni， Ti and Zn in americium dioxide solid samples measured by ICP-AES were analyzed by ' fishbone diagram '. The relative standard uncertainty of each source was calculated respectively. Through synthesis and expansion， the expanded uncertainty of the measured values of each element was calculated to be in the range of 0.0144-0.0344， and the uncertainty report was given. Through the calculation of uncertainty， the key steps of measuring the solid impurity content of americium dioxide by this method were identified as measurement， series standard configuration and standard curve drawing.
[Downloads: 2 ] [Citations: 0 ] [Reads: 0 ] HTML PDF Cite this article
Research on the Application of DFEM-SVAT in the Calculation of Neutron Diffusion Equations
ZHANG Binhang;WANG Jinyao;YUAN Xianbao;ZHANG Yonghong;XIONG Qingwen;HUANG Xiangping;ZHOU Jianjun;XIAO Yunlong;TANG Haibo;The Finite Element Method (FEM) is widely applied to the numerical solution of neutron diffusion equations due to its excellent geometric adaptability. However， with the continuous elevation of engineering accuracy requirements， the computational precision of traditional FEM is insufficient to meet the demands in solving such equations. To address this issue， this study develops a Dual Finite Element Method (DFEM). By introducing the calculation of dual variables， the superior computational accuracy of DFEM compared to traditional FEM is theoretically demonstrated based on its fundamental principles. Considering that DFEM incurs high computational costs when solving large-scale problems， to promote its engineering application， a Symmetrical Variational Acceleration Technique (SVAT) is further proposed in the DFEM solution process. This technique significantly improves computational efficiency while ensuring no loss of precision. To verify the effectiveness of the proposed methods， four international neutron diffusion benchmark problems (IAEA-2D， VVER-440-2D， IAEA-3D， and VVER-440-3D) are selected for systematic validation from two dimensions: computational accuracy(including power error and keff error) and efficiency (including computation time and number of iterations). The results fully confirm the high-precision characteristic of DFEM and the high efficiency of SVAT.
[Downloads: 22 ] [Citations: 0 ] [Reads: 1 ] HTML PDF Cite this article
A Human Reliability Analysis Method for Intelligent Nuclear Power Plants
XU Zhihui;GUO Beiyuan;XIE Yuhao;PENG Huaqing;ZHANG Jiajia;Intelligentization has become an inevitable trend in the nuclear power sector. No matter how intelligent nuclear power plants evolve， they will remain intelligent human-machine systems. In such systems， human beings continue to play an indispensable role， influencing the overall risk level of the system. However， existing HRA methods generally fail to capture the characteristics of such systems， making it difficult to conduct precise risk assessments. Thus， this paper proposes China's first HRA method dedicated to intelligent nuclear power plants. Firstly， this paper incorporates the concept of human-machine collaboration mode to characterise typical features of intelligent human-machine systems， ensuring encompassing all potential scenarios of intelligent interactions. Secondly， this paper develops human error model based on collaboration mode to describe human error characteristics in intelligent interactions. Additionally， Performance Shaping Factors centred on trust and reliance are incorporated to delineate task contexts under such interactions. Finally， this paper provides quantitative parameters for the proposed method through data collection efforts and conducts relevant case application.
[Downloads: 121 ] [Citations: 0 ] [Reads: 3 ] HTML PDF Cite this article
R&D status of China low activation martensitic steel
HUANG Qun-ying1,LI Chun-jing1,LI Yan-fen1,LIU Shao-jun1,WU Yi-can1,LI Jian-gang1,WAN Fa-rong2,JU Xin2,SHAN Yi-yin3,YU Jin-nan4, ZHU Sheng-yun4,ZHANG Pin-yuan4,YANG Jian-feng5,HAN Fu-sheng6, KONG Ming-guang6,LI He-qin7,T.Muroga8,T.Nagasaka8 (1.Institute of Plasma Physics,Chinese Academy of Sciences,Hefei of Anhui Prov.230031,China; 2.University of Science and Technology Beijing,Beijing 100083,China; 3.Institute of Metal Research,Chinese Academy of Sciences,Shenyang of Liaoning Prov.110016,China; 4.China Institute of Atomic Energy,Beijing 102413,China; 5.Xi'an Jiaotong University,Xi'an of Shaanxi Prov.710049,China; 6.Institute of Solid State Physics,Chinese Academy of Sciences,Hefei of Anhui Prov.230031,China; 7.Hefei University of Technology,Hefei of Anhui Prov.230000,China; 8.National Institute for Fusion Science,Toki,509-5292,Japan)The Reduced Activation Ferritic/Martensitic(RAFM) steel is considered as the primary candidate structural material for DEMO and the first fusion plant,and widely studied in the world.China low activation martensitic steel(CLAM) is being developed in Institute of Plasma Physics,Chinese Academy of Sciences,under wide collaboration with many other domestic and foreign institutes and universities.This paper summarized the main R&D progress on CLAM,which covered composition optimization of the CLAM,smelting and processing techniques,physical and mechanical property test and evaluation before and after irradiation,compatibility with liquid LiPb,welding techniques etc.Finally,further research and development,and the prospects on its application were stated.
[Downloads: 2,183 ] [Citations: 246 ] [Reads: 2 ] HTML PDF Cite this article
The development of low activation martensitic steels for fusion reactor
HUANG Qun-ying;YU Jin-nan;WAN Fa-rong;LI Jian-gang;Reduced Activation Ferritic/Martensitic steels (RAFMs) are considered as one of the candidate structural materials for the DEMO reactor and the first fusion power re- actor. Researches on RAFMs are carried out in Europe,Japan and USA in recent twenty years and have made some inspiring progness. This paper introduced the worldwide re- search progress and the key issues on RAFMs in details. Meanwhile, the necessity for designing special China Low Activation Martensitic steel (i. e. CLAM steel) was men- tioned. Studies on CLAM steel, such as its metallurgy technology and tests of its prop- erties etc., are underway and stated in the paper.
Conceptual Design of China Lead-based Research Reactor CLEAR-I
WU Yi-can;BAI Yun-qing;SONG Yong;HUANG Qun-ying;LIU Chao;WANG Ming-huang;ZHOU Tao;JIN Ming;WU Qing-sheng;WANG Jian-ye;JANG Jie-qiong;HU Li-qing;LI Chun-jing;GAO Sheng;LI Ya-zhou;LONG Peng-cheng;ZHAO Zhu-min;YU Jie;FDS Team;The conceptual design of 10MWt China Lead-based Research Reactor(CLEAR-I)had been finished,which has critical and subcritical dual-mode operation capability for validation of ADS transmutation system and the Generation-IV lead-cooled fast reactor technology.Lead-bismuth eutectic was selected as primary coolant.The mature fuel and material technology was preferred,and the fuel assembly can be loaded by remote refueling system.CLEAR-I has excellent safety reliability,technology feasibility,experiment flexibility and technical continuity.The overview of reference scenario design,safety characteristics analysis and R&D status of the CLEAR-I was presented.
FUSION RESEARCH IN THE WORLD
Qiu Lijian (Institute of Plasma Physics,Chinese Academy of Sciences P.O.Box 1126,Hefei 230031,China)The present status and future prospect of fusion research and the predicted energy demand in the world have been reviewed.A proposal on development of fusion fission hybrid reactors,including outlined schedule and predicted achievements,has been presented based on the discussion on the strategy and chance of development of fusion energy forward to the future and the advantages of fusion fission hybrid reactors.
Conceptual Design of China Lead-based Research Reactor CLEAR-I
WU Yi-can;BAI Yun-qing;SONG Yong;HUANG Qun-ying;LIU Chao;WANG Ming-huang;ZHOU Tao;JIN Ming;WU Qing-sheng;WANG Jian-ye;JANG Jie-qiong;HU Li-qing;LI Chun-jing;GAO Sheng;LI Ya-zhou;LONG Peng-cheng;ZHAO Zhu-min;YU Jie;FDS Team;The conceptual design of 10MWt China Lead-based Research Reactor(CLEAR-I)had been finished,which has critical and subcritical dual-mode operation capability for validation of ADS transmutation system and the Generation-IV lead-cooled fast reactor technology.Lead-bismuth eutectic was selected as primary coolant.The mature fuel and material technology was preferred,and the fuel assembly can be loaded by remote refueling system.CLEAR-I has excellent safety reliability,technology feasibility,experiment flexibility and technical continuity.The overview of reference scenario design,safety characteristics analysis and R&D status of the CLEAR-I was presented.
Current status of generation Ⅲ nuclear power and assessment of AP1000 developed by Westinghouse
ZANG Ming-chang (China Power International Holding Ltd, Beijing 100101, China)In order to make greater contributions to the environment, new nuclear power systems will be needed to meet the increase of electricity demand and to replace plants to be decommissioned. A series of new designs, so called Generation Ⅲ and Generation Ⅲ+, are being developed to ensure their deployment in a Near-Term Deployment Road-map in US by 2010 and in Europe by 2015. The AP1000, developed by Westinghouse, is a two-loop 1000MWe PWR with passive safety features and extensive simplifications to enhance its competitiveness in cost and tariff. It is the first Generation Ⅲ + plant receiving the Final Design Approval by the US NRC. This paper briefly describes AP1000 design features and technical specifications, and presents a more detailed design evaluation with reference to relevant literatures. Both the opportunity and challenges for nuclear power development in China during the first decade of the 21st century in a historic transition from Gen Ⅱ to Gen Ⅲ are analyzed. The key is to balance risks and benefits if the first AP1000 to be settled down in China.
Development Status and Prospects of Lead-based Reactors
WU Yi-can;WANG Ming-huang;HUANG Qun-ying;ZHAO Zhu-min;HU Li-qin;SONG Yong;JIANG Jie-qiong;LI Chun-jing;LONG Peng-cheng;BAI Yun-qing;LIU Chao;ZHOU Tao;JIN Ming;FDS Team;Lead-based reactor has become the most promising candidate for Generation IV nuclear system,accelerator driven sub-critical system and fusion reactor,due to its good neutronics,thermal-hydraulics and inherent safety properties,and most likely it will impact the development trend of nuclear energy.In this paper,the development status and prospects of lead-based reactors were introduced and the potential role in future nuclear energy development was summarized.
World Development of Nuclear Power System and High Temperature Gas-cooled Reactor
WU Zhongxin, ZHANG Zuoyi (Institute of Nuclear Energy Technology, Tsinghua University, Beijing, 100084, China)The major challenges of the nuclear power are economic competition, safety, profilication resistance and waste storage. The safety of the nuclear power plants has been enhanced continuously in order to improve the public acceptance. The de regulation of the electricity market in the United States and Europe encourages the competition between the different power generation technologies. The waste storage has been paid great attention. The ALWRs like ABWR, System 80+, EPR and AP 600 development from mid of 1980s are the major commercial products available in the market in the near future. When the time enter the 21 century, US DOE is planning to develop the Generation IV nuclear power system, in order to design one or more nuclear power systems that is market available before or in the year of 2020 and could be used to replace the current nuclear power plants. The modular pebble bed High Temperature Gas cooled Reactor (HTGR) is considered as a prefer candidate in the Generation IV nuclear power systems. The South Africa selected the modular HTGR as the reactor type to develop. The 10 MW test HTGR (HTR 10) designed and constructed in Tsinghua University Beijing is scheduled to reach critical in the year of 2000. Based on the HTR 10 project, China has established her preliminary capability of designing, constructing and manufacturing pebble bed modular HTGRs.
Analysis of Nuclear Energy Utilization under the Background of “Carbon Peaking and Carbon Neutrality”
XING Ji;GAO Li;HUO Xiaodong;WU Yuxiang;DONG Yemin;LUO Yibo;Under the goal of carbon peaking and carbon neutrality,China's energy system will speed up the transformation towards clean and low-carbon. Different from the high carbon emission of traditional fossil fuels,nuclear energy has advantages like low carbon emission,high energy density,resistance of extreme natural conditions and guarantee of safe and stable electricity,etc. It's an essential part of China's energy system transformation. By reviewing and summarizing the current situation of China's energy system,this paper analyzes the developing opportunities of nuclear energy in the area of electricity generation,heating,desalination,hydrogen production,etc. It puts forwards problems need to be paid attention and suggestions on the high quality development of nuclear energy,and provides an idea for developing nuclear energy in a positive and orderly way”.
Tracking the information about your manuscript
Communicate with the editorial office
Query manuscript payment status Editorial OfficeCollecting, editing, reviewing and other affairs offices
Managing manuscripts
Managing author information and external review Expert Information Expert ReviewOnline Review
Online Communication with the Editorial Department
Guide for Author
Page Views
Page visits today: 1