核动力工程

Correlation between Code Threshold/Criteria and Ultrasonic Testing Data Analysis Method in Reactor Vessel In-service Inspection

LI Su-jia, NIE Yong, XU Yuan-huan

2004, 25(1): 1-3.

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Abstract:
The relationships among recording threshold, threshold of characterization and acceptance criteria defined in RSEM Code, as well as that between the UT sizing techniques of 6dB drop and crack tip diffraction methods, are analyzed in detail. The corresponding relationships between Code defined thresholds or criteria and suitable data analysis and indication treatment methods in NPP RPV PSI and ISI are clearly determined in this paper. The principle adopted by this method can be applied to data analysis and indication acceptance of other components PSI and ISI in NPP primary loop.

Performance Monitoring and Fault Diagnosis about Flow Passage of Turbines in Nuclear Power Plant

JIANG Ning, LI Zheng

2004, 25(1): 4-7.

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Abstract:
A performance monitoring and fault diagnosis system for flow passage of turbines in nuclear power plant is introduced in this paper. This system uses the double-check method and efficiency correction method based on fault mechanism to partly solve the problems occurring in the heat calculation for wet steam turbine. A fault diagnosis system based on the change signal of the natural parameters is accomplished and proved to be useful.

New Idea for Thermal Parameters Fault Diagnosis of Nuclear Power Steam Turbine System Based on Large-scale Off-design Mathematic Model

CUI Da-long, LI Zheng

2004, 25(1): 8-12,26.

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Abstract:
In this article, based on the large-scale off-design mathematic model, a thermal parameter fault diagnosis system is set up for nuclear power steam turbine. Due to the description of the mechanism of thermal power system in the mathematic model, the training procedure using large amount of operation data in former diagnosis system is avoided, which makes this method more practical. Because the method introduced the concept of Characteristic Parameter and diagnosis the faults by the relationships between the Characteristic Parameter and its Dominant Factor, the difficulty of fault diagnosis based on phenomena is also avoided. Further more the fault diagnosis result is quantitative and could be an ideal guide for the operation and maintenance of the object system.

Transient Mixing Characteristic of Reactor Pressure Vessel under Pressurized Thermal Shock

WANG Hai-jun, CHEN Ting-kuan, LUO Yu-shan, LU Dong-hua, SUN Ying-xue

2004, 25(1): 13-17.

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The thermal/fluid characteristic of reactor pressure vessel under PTS is a critical issue related to the reactor safety. In this paper, an experiment is carried out to study the transient mixing characteristic un-der high pressure and temperature on a 1/10 scaled model, and the result is obtained under the following con-ditions: with and without loop flow, different temperature of work medium. The results shows at higher injec-tion flow rate the θ will drop at larger speed when there is no loop flow; with loop flow the θ will change relative slowly; and above certain temperature the θ in some measurement area will remarkably change. These results will be helpful to the further integrity analysis of reactor pressure vessel.

Shannon Entropy Characteristics of Two-phase Flow Density Wave Instability Experiments for 200MW Nuclear Heating Reactor

SHI Lei, ZHANG Zuo-yi, GAO Zu-ying

2004, 25(1): 18-21.

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Abstract:
Based on the information theory, the Shannon entropy characteristics of two-phase flow density wave instability has been studied on the research platform of Chinese 200MW nuclear heating reactor. Experimental data for the inlet pressure drop in the heating channel under 534 different conditions has been obtained by adjusting the heating power, inlet subcooling, operating pressure and other parameters. We find that tests with higher negative Shannon entropy （negentropy） are instable while those with lower negentropy are stable. Like energy used in many other fields, negentropy can be a factor to measure the system stability.

Experiments of Pool Boiling Heat Transfer Enhancement of Water-Based Magnetic Fluid

LIU Jun-hong, GU Jian-ming, LIAN Zhi-wei, YAN Zhi-meng, LIU Hui

2004, 25(1): 22-26.

Abstract(9) PDF(0)

Abstract:
A contrast experimental study of the pool boiling heat transfer of water and water-based magnetic fluid was conducted to investigate the effect of water-based magnetic fluid on pool boiling heat transfer, and mechanism of heat transfer enhancement was analyzed. The experimental results showed that the boiling heat transfer coefficient of magnetic fluid was enhanced at least 2 times than that of water at the same heat flux; When applied magnetic field, the boiling heat transfer of magnetic fluid can be enhanced further, which is over 5 times than that of water. The analysis of the effect of a magnetic field on bubbles showed that the applied magnetic field caused the departure diameter of bubble to decrease, the bubble growth rate to rise and the frequency of bubble departure to increase.

Research Progress on Passive Safety System of Advanced PWR

XIAO Ze-jun, ZHUO Wen-bin, CHEN Bing-de, BAI Xue-song, JIA Dou-nan

2004, 25(1): 27-31.

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This paper presents the research progress on the passive safety systems for advanced PWRs in both China and abroad, and suggests that the direction for passive safety systems research and development is the research of passive safety systems of new generation 1000MW PWR.

Technical Study of Real-time Simulation System for Digital I＆C System of Steam Generator in Nuclear Power Plant

SHI Ji, JIANG Ming-yu, MA Yun-qing

2004, 25(1): 32-36.

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Abstract:
The real-time simulation system, which forms a interactive closed circle together with the steam generator control system, has been developed using a dynamic mathematical model of steam generator in this paper. It can provide a simulation target for upgrades of digital Instrument ＆ Control system in Nuclear Power Plant （NPP） and is applicable for further research of control schemes. With this program, the authors have studied and analyzed the response of transient parameters to some different disturbance, the calculated results are in good agreement with those calculated by NPP simulator program. This will give a theoretical analysis for upgrades of digital I ＆ C system in nuclear power plant.

Study on Water Level Detector Composed of Encoded Thermal Resistance

ZHANG Hong-zhong, WANG Wen-ran, LIU Zhi-yong, DUAN Quan-sheng

2004, 25(1): 37-40,44.

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Abstract:
The water level in a nuclear reactor pressure vessel is an important operating parameter. Based on the differences of heat transfer coefficient between liquid phase and gas phase, a new level detector, which includes encoded thermal resistances, has been developed. The monitoring principle, theory analysis and experimental results under pressure of 0.1~3.0MPa were described. The analysis and experiments have shown that the new system is correct in principle, reliable and feasible in structure for monitoring the water level in NHR-200 reactor and other vessels.

Preparation of Micro Carbide Spheres by Gas Grinding

LONG Chong-sheng, QIU Bang-chen, YING Shi-hao

2004, 25(1): 41-44.

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Micro B₄C spheres with an average sphere factor of 0.95 has been prepared by gas grinding method. And the theoretical relationships between the product shape factor and gas velocity has been de-rived.

Random Cyclic Constitutive Models of 0Cr18Ni10Ti Pipe Steel

ZHAO Yong-xiang, YANG Bing, LI Peng-zhou

2004, 25(1): 45-49,63.

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Abstract:
Experimental study is performed on the random cyclic constitutive relations of a new pipe stainless steel, 0Cr18Ni10Ti, by an incremental strain-controlled fatigue test. In the test, it is verified that the random cyclic constitutive relations, like the wide recognized random cyclic strain-life relations, is an intrinsic fatigue phenomenon of engineering materials. Extrapolating the previous work by Zhao et al （Nucl. Eng. Des. 2000, 199（3）: 315-326）, probability-based constitutive models are constructed, respectively, on the bases of Ramberg-Osgood equation and its modified form. Scattering regularity and amount of the test data are taken into account. The models consist of the survival probability-strain-life curves, the confidence- strain-life curves, and the survival probability-confidence-strain-life curves. Availability and feasibility of the models have been indicated by analysis of the present test data.

Effect of 2MeV Proton Radiation on the Microstructure in Zircaloy 4

ZU Xiao-tao, ZHU Sha, WANG Lu-min, YOU Li-ping, WAN Fa-rong

2004, 25(1): 50-53.

Abstract(10) PDF(0)

Abstract:
2MeV proton radiation was performed at 350℃ at damage levels of 2, 5 and 7 displacement per atom. The resulted samples were investigated by the transmission electron microscopy. In the as-received samples, we observe precipitates of the commonly reported hexagonal close-packed Zr (Cr, Fe)₂, and a face-centered cubic alloy containing Zr and Fe, respectively. After radiation to 2dpa, 5dpa and 7dpa, dislocation loop are observed at a density of 7×10²¹ m^-3, 8×10²¹ m^-3 and 15×10²¹ m^-3, respectively. The corresponding loop size are 7nm、 11nm and 11nm, respectively. No amorphization happens for all precipitates.

Design, Test and Operation of Helium Circulator for HTR-10

ZHOU Hui-zhong, WANG Jie, TANG Quan-fa

2004, 25(1): 54-58.

Abstract(13) PDF(1)

Abstract:
The helium circulator is a key component for the 10MW High Temperature Gas-cooled Reactor （HTR-10）. The circulator operates at 250℃, 3.0MPa helium condition in order to transport the core heat to the steam generator. The design of the helium circulator involves the overall construction, the impeller type, the cooling system, the bearings, the instrumentation, the electrical penetrations, and the shutoff valve. Ex-factory test and cold and hot tests after the installation have been carried out for the helium circulator. According to the commissioning requirement of the reactor, initiative operation has been made for the helium circulator. The test and operation results show that the helium circulator meets the requirement of design and operation for HTR-10.

Study on the Impeller Hydraulic Performance for the Contra-Rotating Axial Flow Pump

WANG De-jun, ZHOU Hui-zhong, HUANG Zhi-yong

2004, 25(1): 59-63.

Abstract(9) PDF(0)

Abstract:
This paper discusses the design method and performance of the front and rear impellers of the contra-rotating axial flow pump. A definition of specific speed has been given. The design head of rear impeller is suggested to increase appropriately. By analyzing the inlet and outlet velocity triangles of the twin impellers in detail under design and off-design point, the paper gives a matching condition and formulas for working out the two triangles. Based on the velocity triangles and ‘lift method’, the geometric parameters of the twin impellers have been designed, the results have been analyzed qualitatively, and the hydraulic performance has been predicted. A practical design shows that at the same design head of the twin impellers, the relative velocity at the middle of inlet and outlet of rear impeller are larger than that of the front impeller, while the stagger angle is smaller and the head curve is much steeper.

Application of PLC in the Crane Control System of MJTR

LI Zi-qiang, JI Xiang-dong

2004, 25(1): 64-66,90.

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The paper describes the application of PLC （Programmable Controller） in the control system of the bridge crane. PLC is the essential part of the control system, which uses some equipment such as fre-quency transformer and photoelectric switch to implement remote, manual and automatic centering functions. This paper emphasizes the programming of the automatic hole centering.

Design of Harmonic Screw-driven Stop Valve for Nuclear Power Units

WANG Xiang-jiang

2004, 25(1): 67-69.

Abstract(12) PDF(1)

Abstract:
The harmonic screw type valve is driven by the double wave harmonic screw. It can transform circle movement to line movement for the closing and opening of the valve. The body and flange are airproofed by shell. The stern guiding device is used in the stop valve for guide. Radial distortion of the flexible component in the stop valve can be adjusted so as to ensure the joint depth between flexible screw thread trepanning circular groove and screw thread on the screw. Both flexible whorl cover and stern with screw are placed in flexible sealing shell. Flexible rolling bearing is composed of an outer ring and an inter ring. There is a raceway groove on the surface of the outer ring and the inter ring. The rolling elements in the groove are separated by cage, and the screw and stern are fixed by pin.

“Primary-Least Structure” Method as a Stress Categorization Tool for Design by Analysis

LIU Xiao-long, JIANG Jia-ling

2004, 25(1): 70-73.

Abstract(29) PDF(0)

Abstract:
A new valid method of stress categorization- “primary-least structure method”(PLSM) is proposed. The method is based on the stress-linearized technique and elastic finite element analyses (FEA). During the classification, the primary stresses (Pm + Pb) are evaluated in the “primary least structure”, while other stresses (PL+Q+F) in the original structure. By removing the “ruinous constrains” from the original structure, the lower primary stress is extracted, from which the greater allowable pressure approaching the lower-bound limit load can be obtained. An example is presented to demonstrate the application of the “primary-least structure” method as a stress categorization tool. The result shows that a higher allowable pressure can be obtained by this method, compared with the procedure without stress-linearized technique and the approach given by present Code.

Numerical Simulation of the Stress Distribution for Fast Burst Reactor Under Pulsed Operation

QIU Dong

2004, 25(1): 74-78,82.

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Abstract:
A two-dimension model based on China Fast Burst Reactor - II （CFBR-II） is built to analyze the stress responding of the components in fast burst reactor while the reactor is operated in burst model. The relative neutron flux distribution of the model is calculated by M.C （Monte Carlo） method. A thermal load function is driven to replace the solution of the kinetic equation. The results of experiment and neutron flux calculations are employed in the function. The stress distribution of main components of the reactor is calcu-lated with the finite element method in certain thermal load status. Because of the accurate geometry descrip-tion of the model and the introducing of experimental values, this method is more reasonable compared to the coupled method, especially for a complex model such as CFBR-Ⅱ.

Aging Management Methods for Nuclear Power Plant Components

WANG Lu-shuai, ZHOU Gang

2004, 25(1): 79-82.

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The component aging is unavoidable during the service of nuclear power plants and affects the safety, economic benefits and life of nuclear power equipments. The paper introduces the concept of the component aging and the aim of its management. In the meantime, the aging mechanism is analyzed and the management method is introduced in detail. It is very feasible in practice that the pro-maintenance control method is employed preferentially integrating with other maintenance strategies by combining investigation of aging mechanism and status supervision and fault diagnosis.

Application of Project Management Methodology in Design Management of Nuclear Safety Related Structure

CHEN Mao

2004, 25(1): 83-85.

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This paper focuses on the application of project management methodology in the design management of Nuclear Safety Related Structure （NSRS）, considering the design management features of its civil construction. Based on the experiences from the management of several projects, the project manage-ment triangle is proposed to be used in the management, to well treat the position of design interface in the project management. Some other management methods are also proposed.

Study of the Informatization of the Nuclear Power Equipment R & D System

ZHANG Jiong

2004, 25(1): 86-90.

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Based on the analysis of the composition of the nuclear power equipment R ＆ D system and the connotation of the informatization, this paper has studied the objectives, system architecture and items of the nuclear power equipment R ＆ D system, and discussed the engineering management problems involved in the informatization construction.

Irradiation Ability of Research and Test Reactors

PENG Feng

2004, 25(1): 91-92,96.

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Abstract:
Irradiation ability was suggested as a technical index and performance parameter. Irradiation ability could be used not only as an assessment index of core facility arrangement design, but also as a pa-rameter for the calculation of in-core irradiation costs. Relevant parameters of irradiation ability are: irradia-tion space volume, averaged total neutron flux, reactor power and operative time. Expressions for several definitions of irradiation ability were given, and their features and roles were compared. The High Flux En-gineering Test Reactor was selected as a main example to give the values of irradiation ability and their applications.

Neutron Spectra and Fluence of CFBR-II Reactor Measured by Foils Activation Technique

ZHENG Chun, WU Jian-hua, LI Jian-sheng, WANG Qiang, HE Zhao-zhong, HUANG Yi-chao, DAI Shao-feng

2004, 25(1): 93-96.

Abstract(15) PDF(0)

Abstract:
The neutron fluence and neutron spectra of CFBR-II （China Fast Burst Reactor II） has been measured by the foils activation technique. SAND-II method is used to unfold neutron spectra. The perturba-tion of the irradiation samples to the neutron spectra are studied. It shows that the irradiation sample can slow down the neutron energy. The neutron fluence measured by this method is coincident with the neutron flu-ence measured by ²³⁹Pu fission chamber.

2004 Vol. 25, No. 1