18 July 2012, Volume 30 Issue 20

    

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  Characterization of Porosity Development in Oxidized Nuclear Graphite

  WANG Peng;WANG Yao;YU Suyuan

  . 2012, 30(20): 19-24. https://doi.org/10.3981/j.issn.1000-7857.2012.20.001

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  The High Temperature Gas-cooled Reactor (HTGR) is one of the candidates for the next generation nuclear power plants. A huge amount of nuclear graphite is used to serve as the neutron moderator, the reflector and the structural materials in the HTGR. However the graphite's thermal and mechanical properties will be degraded in the oxidation process caused either by the oxidizing impurities in the coolant or the air ingress accident. The change of the pore structure inside the graphite materials is believed to be the key reason for this degradation. Parameters for characterizing the oxidized graphite's pore structure include the porosity, the burn-off degree, the pore size distribution and the BET surface area. The methods and applications in the characterization of the porosity structural development in the oxidized nuclear graphite are systematically discussed, including the direct-method and the indirect-method. The former includes the mass-volume method, the Hg porosimetry, the gas adsorption method, etc.; while the latter includes the optical image method, the X-ray radiography, the micro CT, the ultra-sonic microscopy, the electron microscopy, etc.. The limitations, the advantages and disadvantages of these methods are discussed. With the development of the microscopy and the computer techniques, the computer-aided microscopy will become an effective and powerful tool to characterize the pore development of the oxidized nuclear graphite.
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  Error Analysis and Experimental Research of HTR-10 Burn-up Measurement System Based on MCNP Modeling

  MA Tao;XIA Bing;WANG Junling;CHEN Xiaoming;JIANG Erdong

  . 2012, 30(20): 25-28. https://doi.org/10.3981/j.issn.1000-7857.2012.20.002

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  With the burn-up measurement system for the 10MW high temperature gas-cooled reactor, the burn-up of the spherical fuel element is obtained indirectly by measuring the gamma-ray from the fission product ¹³⁷Cs. The accuracy of results will affect the security and the economy of the reactor. In this paper, on the basis of the existing equipment conditions of HTR-10, the diversion experiment of the elevator is designed and implemented. By making the spherical fuel element gradually deviate the running measurement place, the relative position between the center of the spherical fuel element and the axis of the collimator is changed. and the relationship between the angle of the diversion and the count rate is obtained and is then used to determine the offset in the circumferential direction between the center of the spherical fuel element and the collimator axis. The model of the HTR-10 burn-up measurement system, established by the MCNP program, can be used to simulate the gamma photon transporting process from the spherical fuel element, through the elevator, the sealed flange and the collimator, to the HPGe detector. The MCNP model can be used to simulate the experiment in different radial deviations. By analyzing the experiment results and calculating results of the MCNP model, the radial diversion from the core of the spherical fuel element to the axis of the collimator can be estimated.
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  Analysis of Containment Aerosol Removal Induced by Passive Containment Cooling Mechanism

  TONG Lili;CAO Xuewu

  . 2012, 30(20): 29-32. https://doi.org/10.3981/j.issn.1000-7857.2012.20.003

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  Based on an integrated severe accident analysis code, a severe accident analysis model is established to include the Reactor Coolant System (RCS), the Engineered Safeguard Facility (ESF) and the containment system, coupled with a thermo-hydraulic analysis, the containment response and the fission products behavior. A passive containment cooling system model is also built with consideration of the air convection cooling and the liquid film heat transfer. Three typical severe accidents induced by small break loss of the coolant accident (SB-LOCA), large break loss of the coolant accident (LB-LOCA) and loss of the feed water accident (LOFW) are considered with the containment atmosphere decontamination as a result of the passive containment cooling mechanism. For every sequence, three different cases are considered, which are the case without cooling, the case with air convection, and the case with air convection and liquid film. The aerosol behaviors for different cases are studied focusing on the volatile fission products and non-volatile fission products. The results show that the passive containment cooling mechanism can enhance the removal effect of diffusion phoresis and thermo-phoresis.
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  Operating Characteristic Analysis of Passive Residual Heat Removal System of HTR-PM

  WANG Dengying;HAO Chen;LI Fu

  . 2012, 30(20): 33-38. https://doi.org/10.3981/j.issn.1000-7857.2012.20.004

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  The passive residual heat removal system (RHRS) is an important safety system for the HTR-PM reactor. Due to the thermo-dynamic coupling between the reactor core and the passive RHRS, an overlapping domain decomposition coupling method is employed for the analysis of the operating characteristics of the passive RHRS of the HTR-PM. The software TINTE-RHRS is developed based on this methodology and is used for the coupling calculation, and an integration solution can be obtained for both the primary system and the passive RHRS. The system models are also presented in this paper. The thermal-hydraulic characteristics of the passive RHRS in the pressurized loss of a forced cooling (PLOFC) accident are simulated using the TINTE-RHRS code. The importance of the coupling calculation between the primary system and the passive RHRS is shown; and the effects of the number of operating systems and the air temperature on the operating performance for the passive RHRS are also addressed.
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  Core Modeling and Neutron Flux Calculation for Supercritical Water Reactor Using MCNP

  TANG Xiaobin;XIE Qin;GENG Changran;CHEN Da;

  . 2012, 30(20): 39-43. https://doi.org/10.3981/j.issn.1000-7857.2012.20.005

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  The supercritical water reactor is one of the six reactors recommended by the Generation IV International Forum. Compared with existing light water reactors,it enjoys advantages of high thermal efficiency, simplified system structure and low cost. The physical model of the supercritical water reactor is established with the MCNP program in this paper, with consideration of the variations of the coolant density along the axial and the intricate geometry of the fuel assembly. Based on the core model of the supercritical water reactor, the radial neutron flux density distribution of the core is calculated and a design scheme of the flattening power distribution is proposed. The axial neutron flux density distribution of the core is calculated, and the effect of the control rod on the peak shift of the axial neutron flux density is discussed. It is suggested that the control rod of the supercritical water reactor should be inserted up from the bottom. This paper provides some food of thought for the construction and the security analysis of the supercritical water reactor, as well as for the application and the development of the supercritical water reactor.
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  Application of MCNP Code in the Calculation of CARR Nuclear Commissioning

  LU Zheng;SUN Zhiyong;XIAO Shigang;LI Jianlong;HUA Xiao

  . 2012, 30(20): 44-47. https://doi.org/10.3981/j.issn.1000-7857.2012.20.006

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  There is no reference reactor for CARR and no zero power experiment carried out, the nuclear commissioning of CARR will be run based totally on the theoretical analysis, as the first case in the China's domestic research reactor start-up. Because of the complicated structure and the large variations of the load of the core, the simulation calculation of the experiment can not be carried out with most neutronics codes based on the diffusion theory. In this paper, the MCNP code is used to run the calculation because of its real geometry function. The calculation results provide some important reference for the CARR start-up. The results of the experiment show that the calculation results are accurate and reliable and the application of the code is successful.
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  Calculation of IHNI-1 Reactor Core's Neutronics Parameters

  ZHAO Zhumin;ZHANG Liang;JIANG Xinbiao;CHEN Lixin;ZHU Yangni;ZHOU Yongmao

  . 2012, 30(20): 48-51. https://doi.org/10.3981/j.issn.1000-7857.2012.20.007

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  Using the WIMS & CITATION program, a neutronic parameter calculating model for the IHNI-I (In-Hospital Neutron Irradiator Mark I) reactor is presented in this paper. In the calculations of the cell group parameters, the bundle model is adopted. The control rod, the top Be reflector, the bottom Be reflector, the side Be reflector and each circle fuel rod of the core are taken as different cell types. In the whole core calculation, the R-z model is adopted by using the CITATION code. The power distribution, the reactivity worth of the control rod and the top beryllium, the temperature coefficient and the burnup are calculated. It is shown that the results agree with the values in literature, and the method is appropriate for the physical calculation of the IHIN-I reactor.
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  Burnup Calculation Methods for Xi'an Pulsed Reactor Base on Coupled Code of WIMS and MCNP

  GUO Hewei;JIANG Xinbiao;ZHAO Zhumin;CHEN Lixin;ZHANG Xinyi

  . 2012, 30(20): 52-55. https://doi.org/10.3981/j.issn.1000-7857.2012.20.008

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  Based on the multigroup calculation feature of the MCNP code, the multigroup function is extended. The coupled code is used to calculate the burnup with the lattice homogenous code WIMS. The 69-group resonance and self-shield macroscopic neutron cross-section, is generated by the WIMS code to simulate the fuel cell and the fuel assembly and the comparative experiment is carried out. The calculation results are in good agreement with the results of other methods and experiments. It indicates that the coupled code is correct and rational. Finally, the paper uses the coupled code to compute and analyze the burnup of the Xi'an pulsed reactor.
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  Reactor Core Scheme for Small Nuclear Power Plant

  XIE Jiachun;LIU Tiancai

  . 2012, 30(20): 56-60. https://doi.org/10.3981/j.issn.1000-7857.2012.20.009

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  The small nuclear power plant enjoys advantages of long life and passive safety and is an important choice in the future nuclear power development. A conceptual core is designed for the small nuclear power plant. It is a pool-type fast reactor with sodium as coolant, the movable reflector and the fixed absorber as the reactivity control system for long-life. Further calculation results show that the life of the reactor could be as long as 30 years, with a reasonable power distribution, all the reactivity coefficients negative, enough reactivity control worth, and all parameters satisfy the design requirements.
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  Experiments on External Power Grid Failure of Research Reactor

  WANG Mingshan;XIA Ming;ZHANG Yang;LIU Chuang;SHEN Shulong;CHEN Yushan;KANG Huanggang

  . 2012, 30(20): 61-64. https://doi.org/10.3981/j.issn.1000-7857.2012.20.010

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  The special process is adopted for the design basis accident of the plant power failure in the design of a research reactor. The external power grid failure test is carried out in the course of the research reactor rated power operation in order to verify the emergency and standby power supply systems. The results of the experiments show that the electrical control and the power can satisfy the requirements of the emergency shutdown, the building isolation, the core cooling, the residual removal, the dose monitoring, the control and protection and the emergency ventilation.
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  Localization of Nuclear Power Plant Equipment in China

  WANG Yuanlong;

  . 2012, 30(20): 65-70. https://doi.org/10.3981/j.issn.1000-7857.2012.20.011

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  This paper reviews the localization of the current nuclear power plant equipment in China, based on the open literature related with research, development, and manufacture of the nuclear power plant equipment in nuclear power plant items under construction in China, including mainly three main manufacture bases of nuclear power plant equipment in China, which are Dongfang Electric Corporation, Shanghai Electric Corporation, and Harbin Electric Corporation. The main focus is the large key installations. They are the containment, the reactor, the pressurizer, the steam generator, the reactor coolant pump in the nuclear island, and the turbine, the condenser, deaerator, the feedwater pump, the dryer reheater in the turbine island, etc. The review shows the current situation of the localization in Chinese nuclear power plant equipment, the manufacture capability of Chinese nuclear power plant equipment and the development direction of the nuclear power plant equipment manufacture in China.
Articles
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  ENDGAME Guidance Law of Air-to-air Missile

  ZHAN Jianchao;GENG Guanglong;MA Xiaoxiao;RONG Penghui

  . 2012, 30(20): 71-74. https://doi.org/10.3981/j.issn.1000-7857.2012.20.012

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  For the ENDGAME section of air-to-air missiles, it is very important to select an appropriate guidance law for missile to hit targets accurately. This paper analyzes three different guidance laws, that is, the proportional guidance law, the differential game guidance law and the variable structure guidance law. The direct force control is introduced, and then the performance of three different guidance laws under various attack conditions is analyzed by the computer simulation. It is shown that under the engaged condition, the variable structure and the differential game guidance laws are better than the proportional navigation law. For a large motor and sustained high speed targets, the guidance performance will be better with the direct force control.
Reviews
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  Friction Stir Welding Technology

  YUAN Kaifeng

  . 2012, 30(20): 75-79. https://doi.org/10.3981/j.issn.1000-7857.2012.20.013

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  This paper studies the friction stir welding technology, with a review on its theoretical and application background. Some quantitative methods in the experiments are used., The friction stir welding is also used for the welding of a variety of light metal alloys and non-metallic materials. The welding materials, the structure of weld joints, the mechanical properties of friction stir welds, the welding procedures and the welding tools are discussed.