核动力工程

Analysis of Effect of Design Transients and Fatigue Damage on RPV Design Life

SUN Ying-xue, DU Juan, LU Yue-chuan

2009, 30(3): 1-3,12.

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This paper presents the effect of design transient and fatigue damage on RPV design life by using the results of fatigue analysis, fatigue crack propagation analysis and fast fracture mechanics analysis. The result shows that 60-year design life of RPV satisfies the RCCM code criteria.

Seal Analysis Technology for Reactor Pressure Vessel

ZHENG Lian-gang, ZHANG Li-ping, YANG Yu, ZANG Feng-gang

2009, 30(3): 4-6.

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There is the coolant with radiation, high temperature and high pressure in the reactor pressure vessel（RPV）. It is closely correlated to RPV sealing capability whether the whole nuclear system work well or not. The aim of this paper is to study the seal analysis method and technology, such as the pre-tensioning of the bolt, elastoplastic contact and coupled technology of thermal and structure. The 3D elastoplastic seal analysis method really and generally consider the loads and model the contact problem with friction between the contact plates. This method is easier than the specialized seal program and used widely.And it is more really than the 2D seal analysis method. This 3D elastoplastic seal analysis method has been successfully used in the design and analysis of RPV.

Comparison and Analysis of Calculation Methods Used in Different Codes Concerning Impedance Function of Foundation for Nuclear Power Plants

HOU Chun-lin, LI Xiao-jun, PAN Rong

2009, 30(3): 7-12.

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In this paper, the impendence function of the containment structure of Nuclear Power Plants is calculated by two methods in different codes, and the difference between the two calculation methods for anti seismic characteristics of the structure of nuclear power plants is studied. The effects of each data for foundation on the results from two calculation methods are explored, under the effects of the different Shear Wave Velocity of surrounding rock, Poisson’s ration and Bulk Density of surrounding rock, and the exact different values between the two calculation methods is given. The study shows that the Shear Wave Velocity of surrounding rock is the key factor for the impendence function.

Floor Response Spectra Analysis of Nuclear Reactor Buildings with Additional Base-Isolation Measures Based on PSDF

CENG Ben, ZHOU Fu-lin, XU Zhong-gen

2009, 30(3): 13-16,21.

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Based on the reactor building of a 1 000 MW Pressurized Water Reactor（PWR） nuclear power station, the floor response spectra in the use of power spectrum density functions has been generated, the main factors（such as soil-structure interaction, second-primary structure interaction and damping ration of secondary system） which have great effect on FRS is investigated with the base-isolation devices added in reactor building, and the sensitivity of the floor acceleration response spectra of the reactor building to the change of this main factors is quantitatively analyzed. The result show that the peak acceleration of FRS is effectively decreased with additional base-isolation devices, and it mainly appears near the base-isolation frequency;in the same time, the base-isolation effect, SSI and second-primary structure interaction are with cross effect, and it is necessary to consider all the effects in the design of nuclear reactor buildings.

Equivalent Linearization Method for Limit Cycle Flutter Analysis of Plate-Type Structure in Axial Flow

LU Li, YANG Yi-ren

2009, 30(3): 17-21.

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The responses and limit cycle flutter of a plate-type structure with cubic stiffness in viscous flow were studied. The continuous system was dispersed by utilizing Galerkin Method. The equivalent linearization concept was performed to predict the ranges of limit cycle flutter velocities. The coupled map of flutter amplitude-equivalent linear stiffness-critical velocity was used to analyze the stability of limit cycle flutter. The theoretical results agree well with the results of numerical integration, which indicates that the equivalent linearization concept is available to the analysis of limit cycle flutter of plate-type structure.

Calculation Methods for Flow-Induced Vibration Response in Pipes with Orifices

MAO Qing, JIANG Nai-bin

2009, 30(3): 22-26.

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In this paper, based on the experimental results of power spectrum density of wall pressure fluctuations, the response of pipe vibration caused by orifice plate was calculated with random vibration analysis function of ANSYS code. The effect of correlation between wall pressure fluctuations on the response of pipe vibration was discussed.Furthermore, a simplified method was given, and the calculation result was compared with that of the detailed calculation. It was shown that the simplified method was convenient and efficient. The response of the pipe flow-induced vibration calculated with simplified method was conservative, so it can be used in the engineering evaluation.

Analytic Expression of Temperature Field of Cylindrical Fuel Element in Dynamic

ZHOU Lian-bang, YAO Qing-sheng, HAN Wei-shi, LI Hao-feng

2009, 30(3): 27-29,34.

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Based on the current form of differential equation of heat transfer, the analytic expression of temperature change with time of cylindrical fuel element was deduced in dynamic status.Analog computation was carried out by this expression. Then the correctness of this solution was validated.

Nondestructive Experimental Measurement of Fuel Element Burnup in Xi’an Pulse Reactor

ZHANG Wen-shou, A Jing-ye, CHEN Da, CHEN Wei, JIANG Xin-biao, WANG Kai, CHANG Yong-fu

2009, 30(3): 30-34.

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The fuel element burnup measurement theory, the equipment and experimental method are described in this paper. The key techniques for burnup measurement, such as neutron/Gamma self-absorb correction, ¹³⁷Cs activity correction and detect efficiency scale, are emphatically expatiated. The burnup of D5 and G14 are measured and the measurement results are compared with the theoretical calculation. The results show that the result of experimental measurement and theoretical calculation are accordant in uncertainty area, and the system is credible.

Calculation of PWR Neutron Pulse Count Rate and Analysis Method for Weak Source Area

LI Tong-lin, ZHOU Lian-bang, HAN Wei-shi

2009, 30(3): 35-37,46.

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The calculation formula of neutron pulse count rate was deduced for the weak source reactor using the model of point reactor.Method of blind area confirmation was discussed. The application shows that this method is consistent with the reality and useful for the improvement of the safety of physics star-up, which is applicable in the supervision of critical safety of nuclear activities, such as the maintenance and repair of reactor, loading and unloading of nuclear fuel and storage of fuel element.

Multiband Method for Resonance Self-Shielding Calculation of Fuel Assembly in Arbitrary Geometries

HUANG Wei-bing, WU Hong-chun, CAO Liang-zhi, YANG Wei-yan

2009, 30(3): 38-42.

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A formula to calculate the multiband parameters is derived based on the multiband method.Adopting the method combining two-band and three-band, and based on the WIMSD4-69 library, a code named RESCAL is developed. The validation shows that the results of RESCAL code are well in accordance with MCNP’s, and the numerical errors meet the practical requirement.Due to the limitation of WIMSD4 69-group library and the method adopted to calculate multiband parameters, the precision of RESCAL code is highly affected by the ratio of water to uranium.

Analysis of Influence of Decay-Cooling Time on on-Line Burnup Measurement of HTR-PM

ZHANG Li-guo, SHANG Ren-cheng

2009, 30(3): 43-46.

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High Temperature gas-cooled Reactor-Pebble bed Modular（HTR-PM） requires a online burnup assay system to check whether a fuel pebbe should be discharged from fuel circulation according to its burnup value.Each pebble that has flowed out of the core is going to undergo decay and cooling before the burnup measurement, so the interval of decay-cooling time influences the burnup measurement. The paper analyzes this effect using the isotopic depletion and decay analysis code of KORIGEN and particle transport code of MCNP. The result shows that the decay-cooling time of 50 hours can meet the HTR-PM requirement well with the choice of suitable γ spectrum analysis method.

Establishment of China Nuclear Fuel Assembly Database

CHEN Peng, JIN Yong-li, LU Hua-quan, CHEN Jian-xin, ZHANG Ying-chao

2009, 30(3): 47-50.

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China Nuclear Fuel Assembly Database（CNFAD） is developed based on Oracle system. It contains the information of fuel assemblies in the stages of its design, fabrication and post irradiation（PIE）. The structure of Browser Sever is adopted in the development of the software, which supports the HTTP protocol. It uses Java interface to transfer the codes from server to clients and make the sources of server and clients be utilized reasonably and sufficiently, so it can perform complicated tasks.Data in various stages of the fuel assemblies in Pressure Water Reactor（PWR）, such as the design, fabrication, operation, and post irradiation examination, can be stored in this database.Data can be shared by multi users and communicated within long distances. By using CNFAD, the problem of decentralization of fuel data in China nuclear power plants will be solved.

Constitutive Model for Non-Proportionally Multiaxial Time-Dependent Ratcheting of 304 Stainless Steel

KAN Qian-hua, SHI Zhi, KANG Guo-zheng, ZHANG Juan

2009, 30(3): 51-55,61.

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Based on the experimental results of non-proportionally multiaxial time-dependent ratcheting of 304 stainless steel at different stress rates and hold-times, a new non-proportionally multiaxial time-dependent constitutive model is proposed by extending Abdel-Karim-Ohno’s non-linear kinematic hardening rule. The model describes the non-proportionally multiaxial time-dependent ratcheting by adding non-proportional static recovery term into the kinematic hardening rule and employing Tanaka’s non-proportion-ality in the isotropic hardening rule.Finally, the capability of the model to describe the non-proportionally multiaxial time-dependent ratcheting is discussed by comparing the simulations with corresponding experi ments. It is shown that the developed model can describe the non-proportionally multiaxial time-dependent ratcheting of 304 stainless steel at room temperature and high temperature reasonably.

Feasibility Analysis for Application of Compact Power Reactor Fuelled with Uranium Zirconium Hydride

CHEN Bing-de, WANG Lian-jie, YING Shi-hao, ZHANG Yue, ZHANG Qiong

2009, 30(3): 56-61.

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This paper introduced the main properties and characteristics of the uranium zirconium hydride fuel element（especially its thermo-physical properties and in-core irradiation performance） as well as some work on the power reactor with uranium zirconium hydride element.And then, the feasibility of uranium zirconium hydride power reactor was studied. The basic conclusions were obtained: There is no severe technical problem in the uranium zirconium hydride fuel elements used for compact power reactor;and there is still a biggish prompt negative temperature coefficient in the compact power reactors with thin rod uranium-zirconium hydride element, in other words, the inherent safety can be retained on a certain extent.

Microstructure and Properties of Vacuum Diffusion Bonded CLAM Steel Joint

HAN Zeng-chan, LI Jing-long, ZHANG Fu-sheng, XIONG Jiang-tao, LI Wen-ya, HUANG Qun-ying

2009, 30(3): 62-65.

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In this study, CLAM steel was welded by vacuum diffusion bonding. The microstructures of the joints bonded under different temperatures and holding times were investigated. The mechanical properties of the joints were characterized by shear strength tests. The results indicate that the vacuum diffusion bonding can realize the reliable joining of CLAM under the appropriate bonding temperature and holding time（bonding ratio not less than 90%）. The shear strength of joints may be as high as 648 MPa.

Study on Salt Bath Nitrocarburizing of 17-4PH Stainless Steel

WANG Jun, XIONG Ji, PENG Qian, FAN Hong-yuan, WANG Ying, LI Gui-jiang, SHEN Bao-luo

2009, 30(3): 66-71,110.

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The effect of the salt bathing nitriding under different temperature on the microstructure of Martensite stainless steel and the change of hardness and wear ability under different treatment temperature are comparatively studied. The study results show that when 17-4PH stainless steel was subjected to the salt bathing nitriding, the main items in the nitrided layer are the expanded（nitride contended） martensite（α'）, Fe_2-3（N, C）, CrN, Fe₄N and Fe₃O₄. The amount of Fe₃O₄ and CrN was increased with the treatment temperature going up. The lattice constant of expanded martensite has the similar change. The activation energy of nitriding in this salt bath was 190.9kJ/mol. The depth of the nitrided layer was increased with the treatment temperature increasing.After the alloy nitriding at 580℃, the mass loss in the slide wear test was reduced from 21.1mg for H1100 condition to 1.0mg.

Investigation on Basic Equations and Scaling Criteria of PWR Natural Circulation

LU Donghua, XIAO Zejun, CHEN Bing-de

2009, 30(3): 72-84,94.

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With analyzing the characteristic of PWR natural circulation, a set of non-dimensional equations were obtained to describe natural circulation phenomenon for both single phase and two-phase independently from single and two phase one dimensional equations. Thus, a set of scaling criteria deduced from non-dimensional equations is able to be applied to scaling simulation for both single and two phase natural circulation. Then, the essence of scaling simulation of natural circulation and the physical meaning of criteria were discussed with initial parameters of simulation, flow instability simulation, heat transfer and critical heat transfer simulation.

Numerical Investigation on Gas-Liquid Flow and Heat Transfer Characteristics in a Liquid-Reactor

YANG Li-xin, NIE Hua-gang, SONG Xiao-ming, NIU Wen-hua

2009, 30(3): 85-90,130.

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Gas-liquid flow and heat transfer in a nuclear reactor were investigated by the numerical simulation, on the base of former numerical study of the experimental model in this paper. The Euler two phase flow model, the MUSIG model and the fluid-solid coupling model are applied to describe, respectively, the gas-liquid two phase flow, the bubble size distribution and interaction in the core and the heat transfer between cooling-pipe and liquid. The temperature distribution, velocity distribution, bubble diameter distribution in the nuclear reactor and heat transfer coefficient of the cooling coils were obtained by numerical method. The numerical results show that the temperature distribution in the core is more even when there is bubble disturbing, and the cooling pipe may achieve 85% heat transfer, which agrees with the test results. The study on the effect of different gas generation on the heat transfer of cooling pipe at nominal power shows that the heat transfer between the liquid and cooling pipe can be enhanced as the increase of bubble generation.

Numerical Study on Resistance Characteristics of Single-Phase Flow through Porous Media

ZHANG Zhen, LIU Xue-qiang, YAN Xiao, XIAO Ze-jun

2009, 30(3): 91-94.

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This paper studied the resistance characteristics of single-phase flow through porous media in numerical way. The flow field is modeled by the Brinkman-Darcy-Forchheimer model. The results obtained from this model agree well with the experiment results. The resistance loss coefficient K_loss of the model has been discussed. It is founded that K_loss dominated the resistance and the influence of Reynolds number on K_loss should be taken into account when Reynolds number at the high values.Numerical results also show that the resistance decreased with the increasing of dimensionless diameter D.

Thermal-Hydraulic Calculation and Test for Irradiation Target Assembly

HUANG Hong-wen, QIAN Da-zhi, YE Lin, XU Xian-qi, ZHOU Shan

2009, 30(3): 95-98.

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In the paper, the geometry and mathematics model of the irradiation Target Assembly are built.COBRAⅢ C/MIT cod is used to calculate the flow rate and pressure drop curve of the irradiation Target Assembly, and the error of theoretic result and hydraulic test is 1.32%. The maximum temperatures on the surface and core of Irradiation Target Assembly are 66.6℃ and 72.7℃, which are less than the design limited value. The MDNBR（minimum departure from nucleate boiling ratio） is also calculated in the paper, which is 5.72 and more than the design limited value. The calculation result has significance for the safety analysis of Irradiation Target Assembly.

Maintenance of Equipment and System during Installation and Commission of Nuclear Power Plants

ZHU Jian-jun

2009, 30(3): 99-101.

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This paper studied the necessity to carry out special maintenance of equipment and system during the installation and commission of nuclear power plants, and at the same time discussed how to organize the special maintenance of equipment and system.Some technical advices are proposed on the maintenance of the equipment and system during the installation and commission considering the characteristics of nuclear power plants, to ensure a high availability of the equipment and system of nuclear power plants during the normal operation.

Design of Fault Diagnosis System for Safety Isolation Valve

YANG Guo-feng, HU Shou-yin, LI Zhi-hui

2009, 30(3): 102-106.

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With the torque measurement method using Active Power, torque measurement theory and electrical measurement theory can be combined to measure the active power and the rotation speed of the drive shaft, instead of the measurement of valve torque. In this way, not only the torque can be effectively measured, but also the three-phase motor voltage, current, power, etc.of the driving motor can be monitored at the same time, and thus a more reasonable diagnosis result can be obtained based on comprehensive evaluation. By simulating the various valve fault forms, based on the collection of data and the analysis of signal features, fault database is developed by fuzzy clustering method. By pattern recognition, the real-time faults are analyzed, and the results are given. Tests show that this diagnosis method is simple, effective, and with detailed characterization for a variety of fault modes.

Design of Measurement System for Cross Power Spectrum Density in Reactors

CENG He-rong, HAN Wei-shi, LIU Yao-guang, QIAN Da-zhi, XU Xian-qi, DUAN Shi-lin

2009, 30(3): 107-110.

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This paper designs a measurement system based on PXI-1031 high-speed synchronization DAQ. By using Labview8.5 program, the data acquisition and data process and analysis program required by the measurement system is designed. By using Fluke-282 arbitrary waveform generator, two passages of sine signals with 150Hz and 270Hz are produced. By using the data acquisition and data process and analysis program, the cross power spectrum density is calculated for two passages of sine signals. The simulation test results show that the measurement system can meet the design requirement.

Nuclear Power Plant Real-Time Fault Diagnosis System Based on Genetic Algorithm

CAI Meng, ZHANG Da-fa, ZHANG Yu-sheng

2009, 30(3): 111-114,130.

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Genetic algorithm is used to optimize the neural networks based on the study on neural network fault diagnosis. Then, genetic optimized NN real-time fault diagnosis system of nuclear power plant is designed. The test results show that this fault diagnosis system can diagnose the set fault rapidly, exactly and reliably.

Concept Research on General Passive System

HAN Xu, ZHENG Ming-guang, YANG Yan-hua

2009, 30(3): 115-118,144.

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This paper summarized the current passive techniques used in nuclear power plants. Through classification and analysis, the functional characteristics and inherent identification of passive systems were elucidated. By improving and extending the concept of passive system, the general passive concept was proposed, and space and time relativity was discussed and assumption of general passive system were illustrated. The function of idealized general passive system is equivalent with the current passive system, but the design of idealized general passive system is more flexible.

Auxiliary Control System for Irradiation Specimen Automatic Transmission Based on Configuration Software

PENG Cui-yun, LIU Cai-xue, CHEN Zhi-mo, FANG Cheng-chun

2009, 30(3): 119-121,144.

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Auxiliary control system realizes sequential control and trace display and automatic transmission for irradiated specimen, which bases on configuration software（MCGS） and industrial control computer as the control platform. The system uses digital I/O cards to establish system state detection and output control arrays. It is showed that the structure posses stable, reliable and security characteristics and well meets the needs of specimen transmission and controlling in industrial automation.

Application of Grey Theory in Identification Model of Human Error Criticality

LI Peng-cheng, ZHANG Li, WANG Yi-qun

2009, 30(3): 122-125.

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The identification model for human error criticality is constructed on the basis of the principle of the Failure Mode and Effects Analysis. It consists of three decision-making factors, namely the scale of probability of occurrence of human error mode, the scale of probability of error-effect and the scale of error-consequence criticality. It is difficult to consider the weight of every factor, this paper employs the grey theory to identify the human error criticality, which provides a new viewpoint for the identification of the priority of error criticality and origin, which overcome the problem for the actual weight distribution.

Simulation of Dynamic Mathematical Modeling for PWR Nuclear Power Plant Core Based on PSASP

SHI Xi, LIU Di-chen, WU Ping, ZHAO Jie, XIONG Li, ZHANG Yuan-yuan, ZHAO Zun-lian

2009, 30(3): 126-130.

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Neutron dynamic model and fuel/coolant thermal output dynamic model were implemented in PSASP through a user-defined program. Based on the mathematical models of different orders, the dynamic behaviors of the NPP core under the input of step disturbance of reactivity and cool-line temperature were simulated in PSASP respectively. The simulation results demonstrate the self-stability of NPP core with temperature effect and poisoning effect, which is consistent with the real-world data.Moreover, the simulation validated the proposed core model, and it can be further used in dynamic calculation of the power system.

Numerical Calculation of Response-Time Influenced by Silver Plating on Fission Couple

FAN Xiao-qiang, LU Wei

2009, 30(3): 131-135.

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In this paper, the diathermanous process of fission couple was analyzed, and mathematics model was built, in which the fission-couple was separated into 3 parts of uranium, silver plating, and thermocouple leads. Two-dimension diathermanous equation, initialization and boundary conditions were given, finite difference method was used to do math disperse with the model. The influence lag time with silver plating is τ=123 ns, which was calculated using Turbo-C. This result indicates that there is no obvious effect of silver plating on the fission-couple.

Application of Shape Memory Alloys in Bolted Flanged Connections

ZHU Shi-chun, LU Xiao-feng

2009, 30(3): 136-140.

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The Shape Memory Effect（SME） and super elasticity of the Shape Memory Alloys（SMA） can make up the clamping force decreasing caused by the creep and relaxation behavior in Bolted Flanged Connections（BFC）, and improve the reliability of the BFC.Advances in the research of SMA in BFC home and abroad is summarized in this paper. The application prospects of Ti-Ni-Pd, Ti-Ni-Hf, Fe-Mn-Si, Cu-Al-Ni and Ni-Al-Mn in the BFC are also discussed. It is considered that the compressive characteristics of the parent phase of SMA should be studied further for the application of SMA to BFC besides the design of sealing structure.When more basic research data is accumulated, BFC with high sealing performance for the critical engineering applications can be developed based on the comprehensive consideration of the stability and reliability of the clamping force.

Performance Test on Shielding Concrete

WU Chong-ming, DING De-xin, ZHANG Hui-chi, CHEN Liang-zhu, YIN Qiang, FANG De-bin

2009, 30(3): 141-144.

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The cylinder of the shielding concrete is made from common Portland cement and home-made coarse or fine aggregates.Orthogonal design experiment and regression analysis are adopted to study the effects of the water content, sand percentage and water-cement ratio on the property of shielding concrete and the difference between them. The test shows that the tensile strength is in inverse proportion with water-cement ratio, and the influence is quite significant.Another factor is the type of aggregates. The effect of the age on its density is not obvious.Similarly, the concrete shielding γ rays shares the same influencing factors with that shielding neutron rays on density, slump and tensile strength.And both have the same change rules regarding to mechanical property.

2009 Vol. 30, No. 3