核动力工程

Issues for Special Attention in Design of PWR Structure

DUAN Yuan-gang, XU Chuan, TANG Chuan-bao

2007, 28(5): 1-4,9.

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Based on the engineering practices,comprehensive assessment,which include overall layout,functional requirements,interface control,thermal and hydraulic behavior,performance of drive line,structure integrity,function principles for reactor internals and impact on fuel assembly,is necessary during the design of reactor structure,to ensure the general assembly,function and safety of reactor.

Analytic Basis Function Expansion Method for Neutron Diffusion Calculation in Two-Dimensional Triangular Geometry

LU Hao-liang, WU Hong-chun

2007, 28(5): 5-9.

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A new nodal method for directly solving the two-group neutron diffusion equation in the triangular geometry was proposed.The neutron flux distributions within a node were expanded in a series of analytic basis functions for each group.Nodes were coupled each other with both the zero-and first-order partial neutron current moments simultaneously.With a new sweeping scheme developed for triangular geometry,the response matrix technique was used to solve the nodal diffusion equations iteratively.Based on the proposed model,the code ABFEM-T was developed.The numerical results for a series of benchmark problems show that the core multiplication factor and nodal powers are predicted accurately using this model for unstructured neutron diffusion problems.

Establishment of Sodium Boiling Model and Validation on BE+1 Experiment for Fast Reactors

ZHAO Shu-feng, LUO Rui, WANG Zhou, SHI Xiao-bo, YANG Xian-yong

2007, 28(5): 10-13,35.

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Under total instantaneous blockage of single subassembly accidental condition in the fast re-actors,the sodium will be boiling soon,which results in the relocation of material and pressure,and has great effect on the development of the accident.To present an integral analysis for this accident,the mechanism of sodium boiling has to be understood firstly.The two-fluid and six-equation model is adopted,and the sub-channel approach was used to discretize the computational volume,at last the model is solved by the semi-implicit numerical method developed by D.R.Liles et al.After validated on the BE+1 experiment,the thermal-hydraulic behavior of sodium under TIB accident is interpreted in detail.

Large Eddy Simulation of Turbulent Buffet Forces in Flow Induced Vibration

XI Zhi-de, CHEN Bing-de, LI Peng-zhou

2007, 28(5): 14-17.

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The pressure pulse filter and Smagorinsky subgrid stress model of the LES（Large Eddy Simulation） are introduced.The fluid field in the annular plenum between the pressure vessel and the core barrel of the Qingshan II phase model test are simulated,and the distribution of the total pressure in the space and time are obtained.The results show that the PSD（Power Spectrum Density） of LES from the calculation and the test keeps the same quantity order.Thus,the pressure of LES can be a load to stimulate the barrel vibration.

Steady-State Thermal-Hydraulic Analysis of SCWR Assembly

LIU Xiao-jing, CHENG Xu

2007, 28(5): 18-21,58.

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Among the six GEN-IV reactor concepts recommended by the Gen-IV International Forum（GIF）,supercritical water-cooled reactor（SCWR） is the only reactor type with water as coolant.Due to its high outlet temperature,it achieves a high thermal efficiency and subsequently has economic advantages over the existing reactors.The present paper performs the thermal-hydraulic analysis of the SCWR assembly using the modified COBRA-IV code.Two approaches to reduce the hot channel factor are investigated: decreasing the moderator mass flow or increasing the heat resistant between moderator channel and its adjacent sub-channels,and the latter also can get a better moderation.According to this paper,the heat transfer deterioration phenomenon is inevitable in the SCWR design.Thus,it is necessary to calculate the cladding temperature accurately,to preserve the fuel rod cladding integrity under heat transfer deterioration.

Modification and Verification of Discharge and CHF Model in LOCA Evaluation Program Development Based on RELAP5 Platform

CHEN Bing-sheng, KUANG Bo, LU Lu

2007, 28(5): 22-25.

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In order to develop the licensing LOCA evaluation program,a discharge model and CHF model modification and verification to RELAP5/MOD3.3 in accordance with Appendix K of 10CFR 50.46 is carried out.With the accomplishment of typical separated tests modeling,namely,Marviken Test-22 and ORNL THTF,input modeling,comparative analysis for the modification are presented.Verification to the conservatism of the employed Moody two-phase critical flow model and CHF correlations,as per related Appendix K requirements,is demonstrated and discussed,which lays a preliminary foundation for further development of LOCA licensing safety evaluation program.

Numerical Simulation of Natural Convective Heat Transfer in Pebble Bed Modular Reactor

SUN Qiao-qun, HE Yu-rong, LIU Guo-dong, LU Hui-lin

2007, 28(5): 26-30,123.

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A numerical simulations of single phase fluid flow and natural convective heat transfer is local friction coefficient of wall,and Nusselt number are predicted at the different Rayleigh numbers.Simulated results show that the natural circulation of fluid is formed between the heating wall with up-flow performed presented in pebble bed modular reactor（PBMR） with the consideration of the inertial and viscous forces.Flow behavior and heat transfer caused by inner heat source in a vertical annulus with saturated fuel particles is numerically investigated at the conditions of shutdown.Distributions of velocity,temperature,and the cooling wall with down-flow.The center of the vortex moves up with the increase of Reyleigh number.The extremum of local Nusselt number and friction factor of wall is existed,and the position moves up along axial direction with the increase of Rayleigh number.Comparing with nitrogen as a cooling medium,Helium fluid will give a more uniform distribution of temperature in the reactor core.

Study on Hydrodynamic Characteristics of Safety Valve Based on CFD Caluculation

FENG Jin, ZHANG Man-lai, HUANG Tian-cheng

2007, 28(5): 31-35.

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The Computational Fluid Dynamics（CFD） method is employed to numerically simulate the working process and the hydrodynamic characteristics of a safety valve.When the inlet pressure is larger than the cracking pressure,the safety valve begins to vend gas for decompression through the moving of valve clack,and the flow zone in the safety is changed.In order to make the computed model adaptive,relevant dynamic meshes which are created or suspended according to the move of valve clack are adopted to update the CFD model in every time step.As an example,the performance of the safety valve with spring direct in compression is studied,and the relation of different static inlet pressures to the spool stroke、flux and the axial force acting on the valve clack are attained.In addition,data about flow field in safety valve is obtained through the unsteady CFD simulation,which indicates that CFD can be a strong tool for us to study the hydrodynamic characteristics of safety valve.

Experimental Investigation on Movement Characteristics of Molten Melt Drops in Coolant

LI Liang-xing, LI Hui-xiong, CHEN Ting-kuan

2007, 28(5): 36-41.

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Experimental facility of Molten Fuel Coolant interactions（MFCI） was designed and set up in this paper.The falling course of molten melt drops was recorded by a high-speed camera and moving curves were obtained.Emphasis was placed on the effect of the drops temperature and the coolant temperature on the course of MFCI.Results showed that falling velocity of the metal drop decreased suddenly and rapidly at first then increased again after the metal drop impacting the surface of water.And then,falling velocity dropped slowly to a comparatively steady velocity.When coolant temperature was constant,falling velocity increased with the rising of drop temperature and when drop temperature was constant,falling velocity increased with the rising of coolant temperature.

Effect of Rolling Motion on Flow Instability of Natural Circulation

TAN Si-chao, GAO Wen-jie, GAO Pu-zhen, SU Guang-hui

2007, 28(5): 42-45.

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Flow instability of natural circulation under rolling motion condition was studied experimentally in this paper.The results showed that the rolling motion might cause the early occurrence of the OFI of natural circulation and changed the types of flow instability.Flow oscillation caused by the rolling motion might overlap with the density wave oscillation.The system stability decreased with the increasing of the rolling amplitude and frequency.There were two stable regions under rolling motion condition,between which there is an unstable region.

Study on Synthesis Process of Uranium Nitride Powder

YI Wei, DAI Sheng-ping, SHEN Bao-luo, ZUO Guo-ping, TAN Xian-qiu, PENG Qian, WANG Ying

2007, 28(5): 46-49,118.

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One of the most important procedures for fabrication of Uranium Nitride fuel is the synthesis of uranium nitride powders with high purity and good sintering ability.The carbothermic synthesis process of uranium nitride powder by reaction of UO₂ powder with Carbon and nitrogen has been studied,and the effect of mole ratio of carbon to uranium,atmosphere controlling,reaction temperature and time on the content of uranium nitride powder has been discussed in this paper.The results show that uranium nitride powder with higher purity can be fabricated by using suitable mole ratio of carbon to uranium（2.3-2.4） and reaction procedures.

Oxidation Performance of Graphite Material Used in Reactor

LUO Xiao-wei, YU Xin-li, YU Su-yuan

2007, 28(5): 50-53,98.

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In high temperature gas-cooled reactor（HTGR）,graphite is used as structural materials and moderator.During reactor operating,graphite oxidation which influences the safety and operation of reactor occurs due to the impurities in coolant or water ingress accident and air ingress accident which may be take place.In this paper,the oxidation process of graphite was introduced firstly,and then the factors which influence the graphite oxidation were analyzed and discussed.Some new questions for further studing were given.

Temperature Effect on Microstructure and Wear Property of 316L Stainless Steel by High-Voltage DC Plasma Glow-Discharge Nitriding

LI Gui-jiang, PENG Qian, LI Cong, WANG Ying, GAO Jian, WANG Jun, SHEN Bao-luo

2007, 28(5): 54-58.

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By applying the different high-voltage DC plasma glow-discharge nitriding（DCPN） temperatures,the microstructure and wear resistance of the surfacenitrided 316L stainless steel were investigated.The phase composition,thickness,micro-hardness and element profile were examined by X-ray diffractometer,Optical microscopy,Hardness tester and EDS（Energy dispersive spectrometer）,respectively.The results showed that a homogenous layer S-phase was produced between 350℃and 400℃ and a single layer CrN at 480℃.The nitrided layer depth was approximately 5~9μm.Both the thickness and the surface-hardness of the nitrided layer increased with the temperature increasing.It also revealed that the wear resistance of the DCPN stainless steel can be improved by one times or more.Wear of the untreated 316L was severe and characterized by strong adhesion,wear,whilst wear of the DCPN-treated 316L was mild and dominated by oxidation wear.

Study on Micro-Structure of a QPQ Complex Salt Bath Heat-Treated 17-4 PH Stainless Steel

LI Gui-jiang, PENG Qian, LI Cong, WANG Ying, CHEN Shu-yuan, WANG Jun, SHEN Bao-luo

2007, 28(5): 59-62.

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The micro-structure of the treated layer of 17-4PH stainless steel（SS） by the QPQ（Quench-Polish-Quench） complex salt bath heat-treatment was investigated,using X-ray diffractometer（XRD） for microstructure analysis and the scanning electronic microscopy（SEM） with EDS（energy dispersive spectrum） for the profile of carbon,nitrogen and oxygen of the treated layer.The results show that the QPQ heat-treated layer is approximately 60μm,of which the surface zone was the oxide Fe₃O₄;the subsurface zone is the compound ξ-Fe₂（N,C） and the inner zone is the mixture of CrN and αN.The compound nitride γ′-Fe₄N exists value at subsurface.After the oxidation layer was removed,the micro-hardness of the treated layer is ob-served to be first increasing and then declining with the depth increasing.between 15μm and 25μm.The interface between the oxide（magnetite） and the compound γ′-Fe₄N was sup-posed to be the place of the ξ-Fe₂N（N,C）_1-x nuclei.The oxygen concentration reveals the highest value at approximately 3μm from the surface,while the nitrogen and carbon concentration orderly reaches the high peak

Using Petri Nets Model to Improve the Algorithm of MCS Solution

ZHANG Yong-fa, CAI Qi, ZHAO Xin-wen

2007, 28(5): 63-68.

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This paper uses the Petri nets to analyze and calculate system fault tree model,and an improved algorithm of MCS（Minimal Cut Set） solution is advanced,which is based on dual and re-absorption.Using this algorithm to analyze the reliability of High Pressure Safety Injection system of NPP,the result showed that the calculation magnitude is largely reduced,the minimal cut set and minimal path set can be obtained at the same time,and the algorithm is conveniently implemented by computer.

Sensitivity Analysis of Hydrogen Source during Severe Accident Induced by Large LOCA for NPP

GUO Lian-cheng, CAO Xue-wu

2007, 28(5): 69-74,108.

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The hydrogen source,which generates through oxidation of Zircaloy by steam and molten core concrete interaction in cavity,during the severe accident induced by large LOCA for the 600MW NPP was analyzed in this paper by using MELCOR code,including the sensitivity impacts of the different break sizes and locations.The results are summarized as follows: during the large LOCA severe accident,it has no substantial effect on hydrogen source with a different break size;when there is the same break size,the hydrogen generation rate peaks when the break locates in a hot leg;and the total cumulative hydrogen production peaks when the break locates in a cold leg.

Project for Implementation of Maintenance Rule Based on Risk-Informed Safety Classification

QIAN Yong-bo, TONG Jie-juan, ZHANG Zuo-yi

2007, 28(5): 75-78.

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Recently,US nuclear power plants（NPPs） have gained excellent achievement and keep ahead of the world.To a great extent,the Maintenance Rule（MR） has played an important role in this result.In this paper,the requirements of MR and implementation method of US NPPs are analyzed.Finally,based on the risk-informed safety classification of the structures,systems and components（SSCs）,an MR implementation project which is applicable to NPPs in China are put forward.

Decision-Making Support System Based on MAS for Emergency Response to Nuclear Accidents of Marine Pressurized-Water Reactor

CHEN Deng-ke, ZHANG Da-fa, JIANG Wei, CHEN Yong-hong

2007, 28(5): 79-82,108.

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Emergency decision-making to Marine Pressurized-Water Reactor（MPWR） was severely restricted by the complex environment.To enhance the emergency decision-making ability of MPWR,reducing（EVA） and Environment Agent（ENA） were designed.The MAS were with the characteristics of autonomy,reactivity and initiative,which were fully used in the system to make effective decision for emergencies.the effect of emergencies,an emergency Decision-making Support System（DSS） which based on Multi-agent System（MAS） was presented.In the system,the HLA/RTI was used as the support environment,and the structure and the Control Agent（SCA）,Analyse Agent（AA）,Countermeasure Agent（CA）,Evaluation Agent

Experiment Research on Sludge Collector for Steam Generator

LIU Hong-yun, WANG Wei, CHENG Hui-ping, WANG Xian-yuan, XU Liang-jun, SHEN Yong-bo, XIONG Chang-huai

2007, 28(5): 83-86,90.

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Sludge collector provides an important active management technique for the sludge deposited ers sludge grains with an extend diameter spectrum.at steam generator（SG） secondary side.With cold experiment,the structure,performance and effectiveness of flowrate of the sludge collector for 1000MWt PWR SG are studied.The experiment results show that the tested sludge collector has a collecting efficiency of over 50%,and has good collecting capability which covers sludge grains with an extend diameter spectrum.

Ageing Analysis and Assessment Methods for Reactor Pressure Vessel

YANG Yu

2007, 28(5): 87-90.

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The ageing mechanisms of PWR RPV include the irradiation embrittlement,thermal Ageing,temper embrittlement,fatigue,corrosion,and mechanical fretting/wear.Aging analysis of PWR RPV is a complicated process,which includes the understanding of aging mechanisms,aging related information,methods of aging data collection,methods of analysis and assessment.Based on recent researches on RPV aging analysis and assessment methods,the process and methods,are presented through tabular in this paper.This paper only concludes the aging susceptibility analysis methods,not involves time-limited aging analysis methods.

Performance Improvements of Radioactive Liquid Waste Handling System for Qinshan Phase III Nuclear Power Plant

WEI Jun-ming, XUE Jun-feng, YU Jian-ming, YAO Guang-yun

2007, 28(5): 91-94.

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This paper described the design principle of radioactive liquid waste handling system for Qinshan Phase III Nuclear Power Plant,and analyzed the problems which occurred during normal operation and the detail schemes of all modifications.Base on these experiences,some system design performances are suggested,which are helpful for the system design of new nuclear power plant and modification of operating plants.

Design and Performance Test of Pneumatic Cut-off Valve for Control Rod Drive System of CARR

ZHANG Ji-ge, YAN Hui-jie, WU Yuan-qiang, WU Xin-xin

2007, 28(5): 95-98.

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The cut-off valve installed in the control rod drive system for CARR is one of the important equipment,which guarantees the safety and reduces the maintenance cost.The function of the cut-off valve is to cut off the passage between the control rod drive system and primary loop,and to prevent the loss of water coolant during maintenance.This paper designed a pneumatic cut-off valve to assist the refueling.With this valve,it is not necessary to empty the water coolant maintenance.Moreover,according to the requirement of ASME Section Ⅲ code,the demonstration test for performance of the cut-off valve was carried out in special test system.The test result showed that the performance of the cut-off valve can meet the design requirement and relevant nuclear codes.

Numerical Studies on Time-Averaged Fundamental Performance of Pulsed Liquid Jet Pump

GAO Chuan-chang, SHANG Hua, NING Feng, QIN Hai-xia

2007, 28(5): 99-103.

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Based on the momentum and energy equations of the pulsed liquid,the time-averaged fun-experimental results.damental performance equation and mathematical expressions of the coefficients are derived,the time-averaged fundamental performance,the momentum correction coefficients and the inlet function of throat pipe in the pulsed liquid jet pump are calculated.The calculated results are in good agreement with the

Simulation Research and Optimal Design for Digital Power Regulating System of China Advanced Research Reactor

DONG Hua-ping, ZHANG Jian-min, CENG Hai, JIN Hua-jin

2007, 28(5): 104-108.

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Based on SimPort simulation platform of nuclear power plant,a simulation model for Digital Power Regulating System（DPRS） of China Advanced Research Reactor（CARR） was established.The transient state of DPRS was simulation studied using this model.According to the characteristics of the driving mechanism of the control rods,the effects of the driving precision of the control rod and its displacement delay upon the system stability were analyzed.Considering the process requirements of CARR and the function characteristic of DRPS,the adjusting parameters for the digital PID controller and the stability limits of the driving mechanism of the control rods were obtained.The sampling period of the digital PID controller is 100ms and its proportion gain is 300.The stability limit of the driving precision of the control rod is 0.4mm.The stability limit of displacement delay between electromagnetic coil and armature is 6.0mm.

Localization Explore of Electrical and I＆C Equipments in Daya Bay Nuclear Power Station

HUANG Wei-gang, DAI Zhong-hua, ZHANG Rui-qiong, WANG Dong, ZHANG Le-fu

2007, 28(5): 109-113.

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The supply of electronic instrumentation and controlling（I＆C） equipments in Daya Bay Nuclear Power Station depends on importation up to now,however,reliable supply of spare parts is presently threatened by many factors.Consequently,Daya Bay Nuclear Power Station tries to localize these equipments.According to a survey,China has possessed the ability to undertake the research,development and production of the I＆C system in nuclear power independently.We propose that localization should firstly starts with the production of replacement board and parts for analogical I＆C system,gradually leads to the substitution of system cabinet units,and finally realize the independent research,development and production of advanced digital main control system.A series of prominent problems of producing nuclear grade I＆C equipments are focused on the designing,manufacturing,testing and reliability evaluation.In this paper,technical procedures,regulations,standards and management experiences related to the localization of nu-clear power electrical and I＆C are introduced.

Supplier Selection Based on Improved MOGA and Its Applicationin Nuclear Power Equipment Procurement

YAN Zhao-jun, ZHOU Lei, WANG De-zhong

2007, 28(5): 114-118.

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Considering the fact that there are few objective and available methods supporting the supplier selection in nuclear power equipment purchasing process,a supplier selection method based on improved multi-objective genetic algorithm（MOGA） is proposed.The simulation results demonstrate the effectiveness and efficiency of this method for the supplier selection in nuclear power equipment procurement process.

Control Measures for Prevention of Mass Concrete Cracks in Construction of Chashma Nuclear Power Project Unit 2

LU Hongzao

2007, 28(5): 119-123.

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The pour of mass concrete,which requires an effective control of possible cracks,is a key technical element in the civil construction of nuclear power plants.This paper offers a detailed description of the main measures and the effectiveness for the control of mass concrete cracks of Chashma Nuclear Power Plant Unit 2.

Methods for the Design of Shielding Concrete Mix Ratio

WU Chong-ming, DING De-xin, ZHANG Hui-chi

2007, 28(5): 124-127.

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Guided by general concrete mix principles,we made a comprehensive study on methods for the design of shielding concrete mix ratio as well as its related factors by means of orthogonal design experiments and regression analysis method.Then we put forward the calculating formulae and steps for the design of shielding concrete mix ratio which combined the weight-holding method with the volume method.A series of tests and practical application show that this method of mix design is accurate,efficient and reliable.

2007 Vol. 28, No. 5