核动力工程

Model Analysis of Monte-Carlo Method for First Criticality Physics Calculation in Pebble Bed High Temperature Gas-Cooled Reactor

CHANG Hong, YANG Yong-wei, JING Ying-qing

2005, 26(5): 419-424.

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The present study compared and analyzed the geometry treatments used for 10 MW High Temperature Gas-Cooled Reactors（HTR-10） initial criticality calculation,and TRIPOLI-4.3 and MCNP4B codes based on Monte-Carlo method were employed to describe the random arrangement of Coated Fuel Par-ticles（CFPs） in the fuel pebbles and the random distribution of the fuel and dummy pebbles in the core.HTR-10 initial criticality was calculated with TRIPOLI-4.3 code and compared with the previous results from MCNP4B.The present study showed that Monte Carlo codes TRIPOLI-4.3 as well as MCNP4B could be used in the first criticality physics calculations for pebble bed High Temperature Gas-Cooled Reactors by adopting appropriate geometry modellings.

Flux Expansion Nodal Method for Solving Neutron Diffusion Equations in Hexagonal-z Geometry

XIA Bang-yang, XIE Zhong-sheng

2005, 26(5): 425-430.

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A new flux expansion nodal method is developed to solve the neutron diffusion equations in hexagonal-z geometry.The intra-nodal flux distribution is expanded in a series of analytic basis functions for each group.In order to improve the nodal coupling relations,a new type of nodal boundary conditions is proposed,which simultaneously requires the continuity of both the zero-and first-order moments of partial current across the nodal surfaces.The response matrix technique is used,which gives a fast-running scheme for the iterative solution of the nodal diffusion equations.Based on the proposed model,the code FEMHEX has been developed and tested against 2-D and 3-D benchmark problems for the VVER-type reactors.The numerical results show that FEMHEX can predict accurately the multiplication factor and nodal powers.

Numerical Calculation of Static Holding Flow of Hydraulic Stepped Cylinder

WANG Jin-hua, BO Han-liang, JIANG Sheng-yao, ZHENG Wen-xiang

2005, 26(5): 431-435.

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This study used the Computational Fluid Dynamics（CFD） program CFX5 to calculate the flow in the chamfer holes,exhaust hole and maze to achieve the static holding flow of the hydraulic stepped cylinder.The results indicated that the flow through the chamfer holes was the main part of the static holding flow,which determined the changing rule of the static holding flow with the change of the outer tube position.The flow through the exhaust hole and maze was relatively small,and the effect on the static holding flow was also small.The maximum and minimum static holding flow increased with the increase of temperature,which agreed well with the experimental results.

Research on Steady Characteristic of Passive Residual Heat Removal System of Chinese Advanced PWR

XIAO Ze-jun, ZHUO Wen-bin, CHEN Bing-de, JIA Dou-nan, ZHOU Lian-bang

2005, 26(5): 436-442.

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Research on the steady characteristic of passive residual heat removal system（PRHRS） for Chinese advanced PWR has been performed in a systematic way.A total of 237 sets of test data at steady state have been obtained and the main influence factors,for example,system pressure,height between heat resource and heat sink,system effective resistance coefficient,and wind velocity,on the two-phase natural circulation flow rate and residual heat removal capability were identified.On the basis of theory analysis,steady correlation of two-phase natural circulation has been obtained.Relative errors of 95% test data were less than ±16%.There is a considerable effect of system state parameter（such as system pressure and valve form resistance coefficient） and system boundary condition（such as environmental temperature and chimney height） on the threshold of elevation between heat resource and heat sink,and its correlation of two-phase natural circulation system has been obtained.Steady characteristic research shows that the PRHRS has the capability of removing the core decay power（2% full power） through natural circulation.

Investigation of Enhanced Vapor Condensation Heat Transfer outside Horizontal B30 Circularly-Grooved Tube

LIU Jun-yan, HUANG Wei-tang, Zhaorigetu

2005, 26(5): 443-447.

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Investigation of heat transfer and friction characteristics had been conducted at atmosphere pressure on B30 horizontal circularly-grooved tube.The experimental result indicates that the circularly-grooved tube can effectively enhance the heat transfer.Within the experimental scope,the total heat transfer coefficient of the circularly-grooved tube can be increased more than 48% compared with that of the smooth tube.Through regression analysis on the experimental data,the experimental correlations for the inside heat transfer coefficient,the condensation heat transfer coefficient on film condensation and the friction coefficient were achieved.

Investigation of Enhanced Condensation Heat Transfer outside Vertical Titanium Circularly-Grooved Tube

Zhaorigetu, HUANG Wei-tang, LU Xiang-bo, LIU Feng

2005, 26(5): 448-451,460.

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The investigation of enhanced condensation heat transfer had been conducted on the outside vertical Titanium circularly-grooved tube.The experimental result indicates that the Titanium circularly-grooved tube is fairly efficient in enhancing the heat transfer.Within the experimental scope,the total heat transfer coefficient of the optimum circularly-grooved tube is 1.12 to 1.36 times of that of the Titanium smooth tube.Through regression analysis on the experimental data,the experimental correlations for the in-side heat transfer coefficient,the condensation heat transfer coefficient on film condensation and the friction coefficient were achieved.

Experimental Study of Heat Transfer Enhancement ofIntegral Pin-fin Tube

DING Ming, YAN Chang-qi, MOU Hong-jian, SUN Li-cheng

2005, 26(5): 452-455,470.

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Under the condition of lubricating-oil vertically flowing through the surface of integral pin-fin tube,the experimental investigations of heat transfer characteristics were carried out.Effects of pitch and height of fins,machining direction,inlet temperature of lubricating-oil and velocity of cooling water on heat transfer characteristic were analyzed.In these experiments,total heat transfer coefficient of integral pin-fin tube reached 200~1470 W/m²ˇK,which increased 1~4 times as compared with the smooth tube.Experi-mental results showed that the integral pin-fin tube was a new kind of enhanced tube fitting for high-viscosity fluid.

Study of Hydrogen Induced Embrittlement of Ti-Al-Zr Alloys

LIU Yan-zhang, HUANG Xin-quan, QIU Shao-yu, ZU Xiao-tao, KANG Peng

2005, 26(5): 456-460.

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The tensile and impact tests of the cathodic charging Ti-Al-Zr alloy were measured with the hydrogen concentration up to 1200 μg/g in order to study the behavior of hydrogen in titanium alloys.The results showed that the strength of the alloy had increased and this was the result of solid solution strengthen-ing.The total elongation exhibited a negative dependence on the hydrogen and decreased with the hydrogen monotonically from ~30% to 22% in the range.The absorbed energy also exhibited a negative dependence on the hydrogen and decreased by 80% at ~300 μg/g,which reflected that this alloy had much sensitivity to em-brittlement by hydrogen even in the slight increase of hydrogen content.The fracture characteristics for the tensile and impact specimens were analyzed,and it showed that there existed a lot of large secondary cracks in the impact specimens,and the fracture mode was changed from ductile to cleavage as the hydrogen content increases.However,there only had some small secondary cracks in the tensile specimens.

Non-Saturated Cyclic Hardening of 304 Stainless Steel and Its Constitutive Modeling

KANG Guo-zheng

2005, 26(5): 461-465,474.

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A visco-plastic constitutive model was proposed to describe the strain-range-dependence and non-saturated property of cyclic hardening observed by experiments.In the proposed model,a new variable was employed to characterize the cyclic hardening.In the evolution rule of the variable,a critical state was introduced to describe the strain-range-dependence of cyclic hardening.Furthermore,the cyclic hardening variable was divided into two parts with different evolution rules so that its non-saturated property presented at relatively large strain amplitude could be described.The cyclic hardening variable is incorporated into the model by considering its effect on kinematic and isotropic hardening evolutions simultaneously.The reasonability of proposed model is verified by comparing the simulations with the corresponding experimental results of 304 stainless steel.

Research and Improvement of Monitoring System for Gaseous Effluent in Qinshan Nuclear Power Plant

ZHANG Yong

2005, 26(5): 466-470.

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The long horizontal and vertical sampling pipelines are used in the gaseous radioactive ef-small corrective room at the bottom of the stack,therefore,85.4 meters long horizontal sampling pipe is can-celled,only 70.6 meter long vertical sampling pipe is reserved.The test and assessment result indicated that the corrective factor of the radioactive iodine sampling is 1.65.For radioactive particulate,according to con-servative estimate,the corrective factor is less than 1.65 under accident condition.During normal operation,fluent monitoring system in Qinshan Nuclear Power Plant,so there is design deficiency for the monitoring of gaseous effluent.The long sampling pipelines,especially horizontal pipe,will result in the losing of sampling particulate and iodine.By means of modification,the sampler for particulate and iodine is moved into the the particulate is consisted of tiny particles,and the corrective factor is 1.1.

Aseismatic Strength Analysis of Main Nuclear Reactor Pump Based on Real Three Dimension Model

SHENG Xuan-yu, LUO Xiao-wei, FU Ji-yang

2005, 26(5): 471-474.

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When the finite element method is used in the pump strength analysis,the shell element is often used,which has the same thickness for the whole pump.3 dimension main pump model that has the same size and shape as the real main pump is built on the software CATIA.Tetrahedron element,instead of shell element,is used to build the finite element grid that may overcome the same thickness approximation.The calculation is more reliable by the tetrahedron element.The strength of auxiliary reactor pump is safe and fulfills the requirement of the standard document.calculated under the seismic load and temperature field.The maxim Von Mises equivalent stress is 29.9MPa,which is smaller than the allowable stress strength 132.825 MPa in ASME-III ND3400.The main pump is

Validation of RIMPUFF Model in RODOS

ZOU Jing, QU Jing-yuan, CAO Jian-zhu

2005, 26(5): 475-479,501.

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This paper evaluated the RIMPUFF model in RODOS by the statistical method and Kincaid dataset of MVK tools,and compared it with other four atmospheric dispersion models.The results indicated that the predicted data of RIMPUFF was reasonable in a 20-km scale with the comparison of other models.

Study of Problems in the International Law of Radioactive Waste Management and Its Solution

LI Qi-wei, PENG Ben-li

2005, 26(5): 480-483.

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nternational Law on Radioactive Waste Management is the important part of the international environmental laws,which guarantees the sustainable development in nuclear safety.Considering that more and more nuclear crisis occurred,this paper introduces and analyzes the insufficiency of International Law of Radioactive Waste Management,and proposes solutions for the completion of existent legislation.

Real-time Simulation of Ex-core Nuclear Instrumentation System

ZHAO Qiang, ZHANG Zhi-jian, CAO Xin-rong

2005, 26(5): 484-487.

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Real-time simulation of ex-core nuclear instrumentation system is an indispensable part of nuclear power plant（NPP） full-scope training simulator.The simulation method,which is based upon the theory of measurement,is introduced in the paper.The fitting formula between the measured data and the three-dimensional neutron flux distribution in the core is established.The fitting parameter is adjusted according to the reactor physical calculation or the experiment of power calibration.The simulation result shows that the method can simulate the ex-core neutron instrumentation system accurately in real-time and meets the needs of NPP full-scope training simulator.

Simulation Software of 3-D Two-Neutron Energy Groups for Ship Reactor with Hexagonal Fuel Subassembly

ZHANG Fan, CAI Zhang-sheng, GUI Xue-wen

2005, 26(5): 488-491.

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A core simulation software for 3-D two-neutron energy groups is developed.This software is used to simulate the ship reactor with hexagonal fuel subassembly after 10、150 and 200 burnup days,considering the hydraulic and thermal feedback.It shows that this software is applicable in the real time simulation of nuclear reactor core physics.It accurately simulates the characteristics of the fast and thermal neutrons and the detailed power distribution in a reactor under normal and abnormal operation condition.

Real-Time Simulation Calculation of Reactor Coolant System

LI Hua, YAN Chang-qi

2005, 26(5): 492-495.

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The real-time simulation of reactor coolant system is studied in the paper.Under the real-time simulation support system ASCA,a reactor coolant system is modeled,programmed and computed.The in-fluence of uneven flow between the gas and liquid is calculated with the drift-flux model.The four quadrant homologous curves in term of the dimensionless similarity parameters are used to characterize the pump per-formance in the program.The dynamic trend of system pressure,SG pressure and water level are obtained under the decreasing power in the simulation system.The results show that the program can be used for train-ing simulator and safety analysis of coolant system.

Numerical Simulation of Heat Transfer in Rod-Support Heat Exchanger

WANG Ding-biao, WEI Xin-li, XIANG Sa, WU Jin-xing, GUO Cha-xiu

2005, 26(5): 496-501.

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The numerical simulation method is put forward to research the rod-support heat exchanger,in order to overcome the shortcoming of the main method of experiment.Numerical simulation model is established by using theory of similarity,and the various structure parameters that affect the characteristics of heat transfer and fluid flow in rod-support heat exchanger are studied.The modulus formulas of the heat transfer and flow resistance in the laminar flow have been derived by using the least-squares method.A set ofvariable structure parameter experimental equipment for the rod-support heat exchanger is designed and manufactured.100 experiment data are obtained under the conditions of the rod-array between single-row tubes and the rod-array between double-row tubes and different structure parameter.The results show a well agreement between the computed and experimented values.It is favorable to use the numerical simulation method in the research and development of heat exchangers.

Numerical Simulation of Pump-Stopping Water Hammer in Tertiary Circulating Water System of PWR Nuclear Power Plant

LIU Zhi-yong, LIU Guang-lin, SU Feng-jie

2005, 26(5): 502-505,518.

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Based on the basic theory of water hammer,the mathematical models of boundary condition of each component of tertiary circulating water system of PWR nuclear power plant,including pump unit,pump discharge valve,condenser and siphon outlet sump,were developed in this paper.Also,the solving method for the functions,namely characteristic-line method,was introduced.Combining with the calculation example,the author pointed out that the close schedule of pump discharge valve had notable influence on the water hammer pressure,while the butterfly valve closed in two stages could reduce the water hammer pressure effectively,but its optimized closing schedule should be determined based on the numerical simulation of fluid transient.

Research of Integrated Management Model for Nuclear Installation Decommissioning Project

LIU Ben-yu, LU Ruo-yu, ZHANG Hong-qi

2005, 26(5): 506-509.

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With the increasing of numbers of nuclear installations entering decommissioning period,it is more and more important to optimize the management of nuclear installation decommissioning project.This paper puts forward the conception of integrated management model.The authors design the construction of integrated management model for nuclear installation decommissioning project from several aspects,such as the construction of target,range,principle and the content.In the end,the authors design and explain from the aspect of organization.

Application of Contemporary Integrated Manufacture Systems to Nuclear Power Plants Management

ZHOU Gang, WANG Lu-shuai, TANG Yao-yang

2005, 26(5): 510-513.

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In order to improve the safety,economy and reliability of the operation of a nuclear power plant（NPP）,a novel integrated management method is proposed based on the "integration" concept of the computer and contemporary integrated manufacture systems（CIMS） in this paper.The design of integrated management system for NPP is studied.In the design of this system,information integration method based on the database and product data management（PDM） technology is adopted.In order to design an integrated management system satisfying the needs of NPP management,all activities of NPP are divided into different categories according to its characteristics.There are subsystems under the general management system to conduct the management work of different catego-ries.All subsystems are interrelated in the environment of CIMS,but relatively independent.The application of CIMS to NPP provides a new way for scientific management of NPP,and makes the best of human,material and information resources.

Analysis of Critical Paths for Schedule Control of Nuclear Power Projects

SHI Liang-min, MA Li-min, FAN Kai

2005, 26(5): 514-518.

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Study on the critical paths for the schedule control of nuclear power projects has been performed for reactor M310,based on the experiences from Daya Bay project and Ling Ao project.The study shows that the critical paths for the nuclear power project are from NI civil work,NI erection,commissioning of single systems directly serving the CFT to the joint-test.For NI civil work,the critical path is the main civil work of the reactor building,pre-stressing,handover of room s for important areas,and key CW-erection interfaces;there are four critical paths for NI erection;For startup,two stages can be identified: commissioning of 16 single systems directly serving the CFT and joint-test.

Safety Classification of Items in Tianwan Nuclear Power Plant

SUN Yong-bin

2005, 26(5): 519-522,527.

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The principle of integrality,moderation and equilibrium should be considered in the safety classification of items in nuclear power plant.The basic ways for safety classification of items is to classify the safety function based on the effect of the outside enclosure damage of the items（parts） on the safety.Tianwan Nuclear Power Plant adopts Russian VVER-1000/428 type reactor,its safety classification mainly refers to Russian Guidelines and standards.The safety classification of the electric equipment refers to IEEE-308（80） standard,including 1E and Non 1E classification.The safety classification of the instrumentation and control equipment refers to GB/T 15474-1995 standard,including safety 1E,safety-related SR and NC non-safety classification.The safety classification of Tianwan Nuclear Power Plant has to be approved by NNSA and satisfy Chinese Nuclear Safety Guidelines.

Study of Expert System of Fault Diagnosis for Nuclear Power Plant

CHEN Zhi-hui, XIA Hong, LIU Miao

2005, 26(5): 523-527.

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Based on the fault features of Nuclear Power Plant,the ES（expert system） of fault diagnosis has been programmed.The knowledge in the ES adopts the production systems,which can express the certain and uncertain knowledge.For certain knowledge,the simple reasoning mechanism of prepositional logic is adopted.For the uncertain knowledge,CF（certain factor） is used to express the uncertain,thus to set up the reasoning mechanism.In order to solve the "bottleneck" problem for knowledge acquisition,rough set theory is incorporated into the fault diagnose system and the reduction algorithm based on the discernibility matrix is improved.In the improved algorithm,the measure of attribute importance first calculate the attribute which have the same value in the same decision-sort,then calculate the degrees of attribute in the discernibility ma-trix.Several different faults have been diagnosed on some emulator with this expert system.

Substitution of Asbestos Gasket and Selection of Non-Asbestos Gasket in Daya Bay Nuclear Power Station

GUO Man-hua

2005, 26(5): 528-530.

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This paper gives a brief introduction of several types of non-asbestos gaskets,and provides several methods for the selection and setting of those gaskets.

2005 Vol. 26, No. 5