核动力工程

Japanese Small Nuclear Reactors and Their Technical Characteristics

CHEN Bing-de

2004, 25(3): 193-197,202.

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Abstract:
Various small reactors, including the integral-type marine reactor （MRX）, have been designed and developed by Japanese Atomic Energy Research Institute （JAERI）. MRX adopts the INV-CRDM, a water-filled containment and passive residual heat removal system. MR-100G and MR-1G, the power supply system for district heating and cooling, are a self-pressurized, nature circulated integral PWR. MR-100G and MR-1C are with characteristics of low radioactive discharge and modular structure, and can be built in a city, even in the basement of an office building. Submersible compact reactor （SCR） for under-sea research vessel has similar design concept with MRX but the primary circuit is of nature circulation. The basic consideration to develop the advanced marine reactor and to enlarge their applications to other fields shows the strategic foresight of Japan. To exploit the experiences and technology in the large-scale nuclear power development in other applications in China, such as a small nuclear electricity source, the experiences obtained in Japanese small reactors can be used as references.

Theoretical Study in the Fluid-Induced Vibration Analysis of Reactor Core Barrel

MAO Qing, ZHANG Jing-hui

2004, 25(3): 198-202.

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Abstract:
In the design of Qinshan Ⅱ nuclear power plant, Fluid-Induced Vibration of the reactor internals has been studied with the combination method of experimental and theoretical analysis. For the theo- retical analysis of core barrel, a special methodology has been adopted. This paper describes the analysis process of this method, deduces the theoretical basis, puts forward the restriction conditions for usage, and studies the applicability in Qinshan Ⅱ. It is considered that this method has definite theoretical foundation and it could be used in the engineering design if the restriction conditions are satisfied. The engineering practice of Qinshan Ⅱ may have some limitation in satisfying the restriction conditions simultaneously, especially for the characteristics of excitation spectrum, damping, and nonlinear effects. Farther theoretical and engineer application study should be devoted in this field.

Fatigue Analysis of Welded Joint at Tube-plate of Steam Generator in HTR-10

HE Shu-yan, LI Xiao-tian

2004, 25(3): 203-206.

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Abstract:
When 10MW high temperature gas cooling reactor restarts in short time after hot shutdown, the reactor could start and rise the power quickly and save much time. However, the temperature of steam generator keeps high （about 430℃） after hot shutdown , and the temperature of the secondary loop is about 100℃. The large temperature differences will generate high thermal stress and thus lead fatigue damage at the weld joint of the tube and tube-plate of steam generator. This paper will analyzis the temperature differences, calculate the stress level and evaluate the fatigue damage of weld joint of steam generator in HTR-10.

Numerical Simulation Analysis of the Effect of Crud on Flow-rate Distribution of Hot Channel in Plate-type Fuel Core

HU Jun, ZHAO Hua, JIANG Xu-lun, CHEN Jun

2004, 25(3): 207-209,221.

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Abstract:
With CFX code, the effect of crud on flow-rate distribution of hot channel in plate-type core is analyzed. The research shows that the flow-rate distribution of hot channel is decreased by 18.2%, on account of the clearance decrease 10% due to crud on the surfaces of both hot channel fuel plates in plate-type fuel elements core.

Parallel Algorithm and Software Development for Structure Seismic Response Analysis

LI Li-jun, JIN Xian-long, LI Yuan-yin

2004, 25(3): 210-213.

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The seismic response analysis of massively structure is more important. The research is algorithm for synthesizing seismic acceleration time histories and structural seismic response analysis, and a new type of analysis system based both on great commercial finite element code and massively parallel computer is presented. The client interface for the integrated application system has been developed. The system couples the advantages of serial structural analysis and high performance computing, and can improve the precision and efficiency of seismic analysis.

Dynamic Responses of Control Rods of Nuclear Ship under Impact Load

LUO Xiao-wei, WU Hong-lin, YU Su-yuan

2004, 25(3): 214-217.

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Abstract:
When the nuclear-powered ship encounters an impact load, especially in the vertical direction, the control rods of the reactor would be displayed to some extent, and the reactor activity and the output power of the reactor would be changed, and thus the normal work of reactor would be affected. This can greatly influence the safety of ship navigation even to broken-down ship. In this paper, a reasonable calculation model has been established to study the responses of the control rods including displacement and acceleration to a vertical impact.The effect of the fundamental frequency of the reactor and the actuation duration of the impact load on the displacement responses of the control rods has been also studied.

Visualization Test for Flow Field of Bundles with Grid Spacers

XIONG Wan-yu, CHEN Bing-de, XIAO Ze-jun, WANG Xiao-jun

2004, 25(3): 218-221.

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Abstract:
This paper studied the effect of AFA-2G 5x5 grid spacers on flow used in visualization test. The dye method was used to study the mixture effect of mixing vanes. Tracing particle method was used to observe 3D-flow field in rod bundles. 2D Laser Doppler Velocimeter （LDV） measured axial and lateral velocities of up-flow and down-flow of grid spacers. Pressure drops were also measured to estimate loss coefficients for the grid spacer and friction factors for rod bundles, and obtained formulation for loss coefficients and Reynolds and formulation for friction factors and Reynolds.

Experimental Study of Critical Heat Flux during Natural Convective Boiling in Inclined Annular Tubes Submerged in Saturated Liquids

LIU Zhen-hua, ZHANG Tong

2004, 25(3): 222-225.

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Abstract:
An experimental study was carried out to investigate the critical heat flux （CHF） during natural convective boiling in the inclined annular tube with a heated inner tube submerged in both saturated liquids of water and R-11 under atmospheric pressure. The experimental results show that the CHF decreases with increasing the ratio of the effective heated length to the tube gap, while the CHF decreases with decreasing the inclination angles. A semi-theoretical correlation using for the CHF during natural convective boiling in the vertical annular tube was modified. The modified correlation can reasonably well predict the CHF for the inclined annular tube. The correlation was also compared with the CHF data of the inclined annular tube from other literature.

An Experimental Investigation of Drag Characteristic of A Hot Particle Moving in Coolant Liquid

CHEN Dong-hai, CAO Xue-wu

2004, 25(3): 226-229,240.

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Abstract:
The movement characteristic of a hot particle moving in coolant liquid is experimentally studied in this paper, it shows the movement rules of the hot particle with such specific configuration in coolant liquid, in the coarse premixing phase of interaction between hot molten particle and volatile coolant liquid. The results are significant to the development of multi-phase multi-component analysis procedure and the research of FCIs in severe accidents of reactors.

Improvements of Evaporation Drag Model

LI Xiao-yan, YANG Yan-hua, XU Ji-yun

2004, 25(3): 230-232,245.

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Abstract:
special observable experiment facility has been established, and a series of experiments have been carried out on this facility by pouring one or several high-temperature particles into a water pool. The experiment has verified the evaporation drag model, which believe the non-symmetric profile of the local evaporation rate and the local density of the vapor would bring about a resultant force on the hot particle so as to resist its motion. However, in Yang’s evaporation drag model, radiation heat transfer is taken as the only way to transfer heat from hot particle to the vapor-liquid interface and all of the radiation energy is deposited on the vapor-liquid interface, thus contributing to the vaporization rate and mass balance of the vapor film. So, the heat conduction and the heat convection are taken into account in improved model. At the same time, the improved model given by this paper presented calculations of the effect of hot particles temperature on the radiation absorption behavior of water.

Surface Modification of a Ti-2AI-2.5Zr Alloy by N Ion Implantation

ZU Xiao-tao, FENG Xiang-dong, DENG Jiong, QIU Shao-yu, ZHOU Ji-meng, WANG Zhi-guo

2004, 25(3): 233-235,251.

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Abstract:
Implantation-enhanced hardening of a Ti-2Al-2.5Zr alloy by 75keV nitrogen ion implantation with the dose of 3×10¹⁷ N⁺ cm^-2 and 8×10¹⁷ N⁺ cm^-2 is investigated. The surface characterization of samples was studied by XRD and Vickers hardness tests. The nitrogen implantation increases the surface hardness up 340% and 260 for the dose of 8×10¹⁷ and 3×10¹⁷ N⁺ cm^-2, respectively. The results of XRD analysis show that TiN is formed in the surface region. Nitrogen-ion-implantation-enhanced hardening is mainly due to the formation of TiN.

Thermal Expansion of Zr, Zr-4 and New Zirconium Alloys

XUE Shu-juan, WANG Yun-hui, ZHAO Wen-jin, YING Shi-hao

2004, 25(3): 236-240.

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In this paper, the expansion behavior of rod and plate of Zr, Zr-4 and new zirconium alloys was studied from room temperature to 800℃ by using quartz dilatometer. The results showed that there had little difference among N18, N36 and Zr-4 alloys. The relationship between linear thermal expansion and microstructure of N18 plate was studied and from the results it could be concluded that there lied anisotropy for the cold-rolling of Zr, Zr-4 and new zirconium alloys

Synthetic Analysis for the Probabilistic Models of Cyclic Strain-life Relations of 16MnR Steel

YANG Bing, ZHAO Yong-xiang, WU Ping-bo, CENG Jing

2004, 25(3): 241-245.

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Abstract:
A synthetic analysis is made for the probabilistic models of cyclic strain-life relations of 16MnR steel weld joint. Purpose aims to employee synthetically the test data under different test conditions, which could reflect the production environments, to provide the reasonable curves for fatigue reliability design and analysis of the material and structures. The models are constructed with the probabilistic three-parameter exponent relations. They consist of the survival probability-based strain-life curves, the confidence-based strain-life curves, and the survival probability and confidence-based strain-life curves. These curves could provide a wide advice for the fatigue probabilistic design and analysis of the material and structures. Parameters of parent metal are also provided to make a comparison between weld joint and parent metal.

Thermo-electric Performance Investigation of Two Layer Inhomogeneous Doped Graded Material FeSi₂

LIU Xiao-zhen, Z. A. Munir, ZOU Cong-pei

2004, 25(3): 246-248,259.

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The two layer inhomogeneous graded P phase FeSi₂ was preparated by Field-activitated method. The dopant content in the interface was analyzed by WDS. The Seebeck coefficient and resistivity of Mn-doped and Co-doped inhomogeneous doped samples were measured. Power factor of samples were calculated. Compared with homogeneous material, this kind of inhomogeneous dopant structure can adjust the relation between the thermo-electric performance and temperature.

Improvement of Fabrication and Installation Technology for Reactor Pressure Vessel and Steam Generator of Ling’ao Nuclear Power Station

CHEN Zhen-wei

2004, 25(3): 249-251.

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Abstract:
The fabrication and installation technology improvement for primary equipment Reactor Pressure Vessel and Steam Generators on Ling’ao Nuclear Power Station is given. The advantage of these improvement and impacts on the construction of Ling’ao Nuclear Power Station is analyzed.

Aging Analysis of Reactor Pressure Vessel for Unit 2 of DAYA Bay Nuclear Power Station

WAN Li-hang, LIU Peng, TAO Yu-chun

2004, 25(3): 252-254.

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Aging in nuclear power plants shall be managed effectively to ensure that the design function remains available throughout the service life of the plant. This paper analyzes the aging mechanism, which affects the function of reactor vessel, and evaluates the present situation of unit 2 of DAYA Bay nuclear power stationt.

Modeling and Steady-state Analysis for a Horizontal Steam Generator

DONG Yong-sheng, ZHANG Jian-min

2004, 25(3): 255-259.

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Abstract:
This paper studied the mathematical model in the steady state for the horizontal steam generator, and based on this study, the thermal-hydraulics analysis code HSG-S for the HSG had been developed, and the steady state calculation had been preformed. The results were correct and fit well with RELAP5 results.

Design of Passive Heat-switch between Pressure Tube and Calandria Tube of Thorium-based Advanced Nuclear Energy System

YOU Song-bo, YU Ji-yang, YANG Gao-sheng

2004, 25(3): 260-263.

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Abstract:
The paper presents a design of passive heat-switch between pressure tube and calandria tube of Thorium-based Advanced Nuclear Energy System. The heat-switch is based on shape memory alloys （SMA）. When the temperature of the pressure tube reaches 340℃, the SMA in the heat-switch will change its shape to drive heat conductors to connect the pressure tube, and calandria tube results in heat transferred from pressure tube to the calandria tube and the moderator. The passive moderator cooling system will bring the heat to the environment finally. A heat-switch with 24 slices can satisfy the requirement by theory calculation.

Abrasion and Blockage Mechanism of Incore Flux Thimble and Preventive Maintenance Strategy

LIU Zheng-jun

2004, 25(3): 264-266,269.

Abstract(17) PDF(1)

Abstract:
This paper describes the abrasion and blockage mechanism of incore flux thimble and the corrective measures for the thimble in service inspection and preventive maintenance taked by Qinshan Nuclear Power Company. It is valuable for the operation and maintenance of Incore Flux Measurement System.

Discussion on Replacement of Steam Generator

LING Xing, HUANG Su-yi

2004, 25(3): 267-269.

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The replacement of steam generators requires 35 to 60 days in general. The design demonstration and safety review for the building storing the replaced steam generators shall be also carried out based on the related national codes on nuclear safety, while the demonstration and review of the replacement scheme are conducted. All prerequisites and conditions shall be available prior to the implementation of the SG replacement. The feasibility of the processes to be adopted shall be demonstrated.

Discussion about Design of Front Guide Impeller of Axial Flow Pump

LI Tong-zhuo, CHEN Jian, LU Hong-qi, XIANG Qing-jiang

2004, 25(3): 270-274,283.

Abstract(8) PDF(0)

Abstract:
In order to mitigate the cavitations of the axial flow pump and improve the flow state of the main impeller, this paper gives a primary study on the design of the front guide impeller of the axial flow pump and its configuration. This paper address the determination and analysis of the parameters for the design of the front guide impeller, the determination of the guide impeller dimension and the calculation and diagraph of the shape of the guide impeller vane and so on. The guide impeller is designed to effectviely reduce the water vibration and noise of the axial flow pum

Preliminary Study on the Effect of Spray Model on Hydrogen Explosion

LIN Ji-ming, JIA Bao-shan, LIU Bao-ting

2004, 25(3): 275-278,283.

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Abstract:
The effects of different spray models on the accident progress after the power recovery at blackout accident in Daya Bay Nuclear Power Plant were compared by MELCOR code. The result showed that the hydrogen combustion can be avoided or mitigated substantially by shorter spray duration and proper start time to activate the spray, and thus the time for the pressure in the containment reaching the limit can be retarded.

Study on the Progression of Severe Core Damage Induced by SGTR and Mitigation Measures for QINSHAN NPP Unit 1

XU Yi-quan, SU Yun, CAO Xue-wu

2004, 25(3): 279-283.

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Abstract:
The progression of core damage induced by SGTR in Qinshan NPP Unit 1 is analyzed by a NPP severe accident simulator developed in this laboratory, based on SCDAP/RELAP5/ MOD3.1 and PROSYS. By employing the results of USA SAN ONOFRE NPP’s IPE and SURRY’s PSA, to end the core damage progression and mitigate the consequences of SGTR, the measures for the management of the severe accident induced by SGTR are selected, such as feed-and-bleed and depressurization, which are verified through the calculation by using the simulator. The results suggest that the implementing of feed-and-bleed and depressurization could be an available and effective way to arrest the SGTR sequences in Qinshan NPP unit 1.

Management Self-assessment in Construction Phase of Nuclear Power Station

SHAO Bi-xiu

2004, 25(3): 284-288.

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Management self-assessment is one of the quality management methods to achieve the organization’s continuous improvement. Through the investigation of the methods, the author hopes to give guideline to the management at all levels involved in the construction of nuclear power station in the implementation of management self-assessment, and also to make the management self-assessment more normalized, systematical and effective.

2004 Vol. 25, No. 3