核动力工程

Development Status and Application Prospect of Supercritical-Pressure Light Water Cooled Reactor

LI Man-chang, WANG Ming-li

2006, 27(2): 1-4,44.

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The Supercritical-pressure Light Water Cooled Reactor（SCWR）is selected by the Generation IV International Forum（GIF）as one of the six Generation IV nuclear systems that will be developed in the future,and it is an innovative design based on the existing technologies used in LWR and supercritical coal-fired plants.Technically,SCWR may be based on the design,construction and operation experiences in existing PWR and supercritical coal-fired plants,which means that there is no insolvable technology difficult-ties.Since PWR technology will be adopted in the near term and medium term projects in China,and considering the sustainable development of the technology,it is an inevitable choice to research and develop the nuclear system of supercritical light water cooled reactor.

Qinshan NPP In-Core Fuel Management Improvement

KONG De-ping, LIAO Ze-jun, WU Xi-feng, WEI Wen-bin, WANG Yong-ming, LI Hua

2006, 27(2): 5-8.

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In the 10-year operation of Qinshan Nuclear Power Plant,the initial designed reloading strategy has been improved step by step based on the operation experiences and the advanced domestic and international fuel management methods.Higher burnup has been achieved and more economic operation gained through the loading pattern improvement and the fuel enrichment increased.This article introduces the in-core fuel management strategy improvement of Qinshan Nuclear Power Plant in its 10-year operation.

Study on Reactor Power Change and Ambiguous Control of Third Qinshan Nuclear Power Plant

WANG Gong-zhan, GU Jun

2006, 27(2): 9-13.

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The phenomenon of the average power reduction during long term full power operating in Third Qinshan nuclear power plant is analyzed in this paper.According to the basic conclusions of reactor power fluctuating derived by probability statistic and calculation the corresponding ambiguous control project is proposed.The operating performance could be achieved by the present controlling project is predicted additionally.

An Uniform Formula of Neutron Multiplication Calculation from Sub-Criticality to Prompt-Criticality with Step Change of Reactivity

CAI Zhang-sheng, GUI Xue-wen, YU Lei

2006, 27(2): 14-16,25.

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An uniform formula of neutron multiplication calculation with step change of reactivity is derived from the point reactor kinetics equation.The model is applicable to calculate neutron multiplication in reactor sub-critical state,delay-critical state and prompt-critical state.The results show that the uniform formula can be used in quantitative analysis and calculation with a high operational precision.

Comparison betweenTechnical Requirements in Different Stan-dards on Synthesis of Design Ground Motion History for Nuclear Power Plants

LI Zhong-cheng, ZHAO Feng-xin

2006, 27(2): 17-21.

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The main technical requirements in several domestic and American Standards,Codes and Guides involved in the seismic analysis and design activities of nuclear power plants（NPPs）in China are construed in detail in this paper.Based on better understanding of the technical backgrounds and resultant requirements on the synthesis of design ground motion history for NPPs,some discussions and viewpoints in application are conducted together with the engineering in practice.Some conclusions can be provided for reference to conduct the seismic safety evaluation,seismic analysis and design of NPPs.And the suggestions in this paper can be employed in the revision and refinement of the relevant Standards,Codes and Guides in China.

Equivalent K Conversion and Equivalent δ Conversion in Defect Assessment of Pressure Vessel

ZOU Guang-ping, WU Guo-hui

2006, 27(2): 22-25.

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Up till now,in the defect assessment standard of pressure vessel,a semi-experience method,equivalent K（stress intensity factor）conversion,usually used to deal with beried and surface three dimen-sional crack.In this paper,the method using elastic-plastic fracture parameter δ（crack tip opening displace-ment）to replace K（stress intensity factor）is first proposed,and equivalent K conversion in usual pressure vessel assessment is replaced by equivalent δ conversion,and equivalent K conversion and equivalent δ conversion is demonstrated in penny shaped crack.It is observed that the equivalent K conversion is conser-vative,and the equivalent δ conversion is more reasonable than the equivalent K conversion.

Comparison on Approaches Evaluating Irradiation Embrittlement of Pressure Vessel Steel

ZHANG Xiao-zhong

2006, 27(2): 26-29.

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Although efforts have been made to establish the cleavage fracture model in the past for more than 30 years,uncertainties remain in predicting the cleavage fracture toughness behavior of pressure vessel steels in the transition region.In this paper,the traditional approach and Master Curve approach of describing the cleavage fracture toughness of pressure vessel steels in the transition region are discussed and compared.The need for focusing on the fundamental principles and criteria of cleavage fracture is addressed.

Stochastic Seismic Response Analysis for Nuclear Power PlantStructure Considering Parameter Uncertainties of Soil

LI Zhong-xian, LI Zhong-cheng, LIANG Wan-shun

2006, 27(2): 30-35.

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In this paper,a stochastic seismic response analysis for a nuclear power plant structure is carried out considering SSI effect,in order to discuss the influences of the uncertainty of soil on the nuclear power plant structural seismic responses. The soil-structure dynamic interaction is modeled by setting the spring and damping elements on the boundary using the finite element analytical program ANSYS. The un-certainty of the dynamical parameter of soil can be achieved by adjusting the values of these parameters re-lated to the aforementioned elements. Finally,a reactor building in a 1000MW class pressurized water reactor nuclear station is numerically analyzed. The comparison of the numerical results of stochastic method with those of deterministic analyses indicates that this method is practical,and it is a useful balance for the deter-ministic analytical approach. The combination of these two methods can yield much more reasonable results for evaluating the effects of the property uncertainties of soil on the seismic responses. The conclusions pro-vide some references for the evaluation of the sensitivity of seismic nuclear power plant structural response to uncertain parameters.

Investigation of Flow Induced Vibration of Core Rods in Nuclear Reactor Using Vortex Lattice CFD Method

ZHAO Jun, SHANG Zhi, CHEN Shuo

2006, 27(2): 36-39.

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In this paper,the vortex lattice CFD（Computational Fluid Dynamics）method is employed to investigate the dynamic response of core fuel and control rods on the fluid flow in nuclear reactors.The vor-tex lattice CFD method can represent the mechanism of flow-induced vibration through studying the vortex convention and vortex diffusion.Based on the computational results,compared with other published experi-mental and numerical data,the variation of magnitude and direction of the force is found to be the basic rea-son for flow induced vibration.

Drift Flux Model for Vertical Upward Two-Phase Flow

SUN Qi, ZHAO Hua, YANG Rui-chang, ZHANG Hong-yan

2006, 27(2): 40-44.

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Drift flux model is prevailing for the prediction of void fraction in two-phase flow,whereas distribution parameter C₀ and drift velocity u_gj of the model are an issue argued by different researches until now.This subject is experimentally and theoretically studied in the paper and the limit-ing conditions of distribution parameter and drift velocity for upward two-phase flow are proposed.Fi-nally,the recommended correlation for the void fraction prediction is also presented based on the theo-retical analyses and data comparisons.

Experimental Investigation on the Flow Instability in Evaporator of Separated Type Heat Pipes

ZHU Yu-qin, BI Qin-cheng, CAO Zi-dong, CHEN Ting-kuan, WANG Wei-shu, DENG Zhi-an

2006, 27(2): 45-49.

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An experimental investigation on the working fluid flow characteristics and the flow instability in the evaporator of Separated Type Heat Pipe（STHP）was presented.Flow patterns were observed and analyzed by visualization experiment.It was found that the bubbly flow,slug flow,and wavy foam flow occurred orderly in the evaporator with a small inclination angle when the heat flux increased.Two types of flow instabilities occurred,i.e.,the flow pattern transformation instability and the density wave instability.The effects of pressure,mass velocity,inlet subcooling,heat flux,and exit throttle on the flow instability were determined in the modeling experiment.The thresholds of the flow pattern transformation instability and that of the density wave instability were obtained.Non-dimensional correlations for predicting the flow instability in STHP were given by means of one-dimension single-fluid model.The results may be used for the engineering design of largescale heat exchangers of STHP with a small inclination angle in evaporator.

Control of Homogeneity of U-Mo Alloy

LIU Chao-hong, JIANG Ming-zhong, YIN Chang-geng

2006, 27(2): 50-53,63.

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The coating of graphite crucible and its preparation technology,and homogeneity controlling for U-Mo alloy are studied when we melt U-Mo alloy in the vacuum induction furnace.The experimental re-sults show that the material is calcium oxide stabilized by 5wt%~10wt% titanium oxide and the coating is sintered in the vacuum induction furnace,thus the stability of the coating will be improved.Molybdenum is loaded on the bottom of graphite crucible,which is helpful to control the homogeneity of U-Mo alloy.When the melting temperature is about 1480℃,and melting time is about 8 minutes,the general efficiency of the homogeneity controlling for Mo component and the carbon impurity decreasing of U-Mo alloy is good.The crystal grain is basically the equiaxial crystal,and the grade is 6.4（about 35 μm）.The content of Mo in the center of crystal grain is higher than that of grain boundary;The alloy which is not cast has effect on the phase composition of alloy,and the body isα-U.In order to keep metastable γ as much as possible,the alloy must be cast and cooled quickly.

Unsymmetrical Crack-Tip Stress-Strain Field and Its Engineering Treatment Method of Mixity Parameters of Welded Joint

ZHANG Min, DING Fang, CHENG Zu-hai

2006, 27(2): 54-58,63.

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Considering the stress-strain field distribution rules of welded joint containing a crack on fusion line,the calculation procedures of mixity parameter M^P is discussed on unsymmetrical strain-stress yield.And furthermore,the fracture parameter COD is decomposed into two parts of modeⅠand ModeⅡ components.By the finite element method,the mixity angle φ of mixed-mode is calculated and the result is analyzed on the condition of different lording levels.In the end,for the engineering treatment,the calculating method of the mixity angle φ,which can be used for the mixed mode of the asymmetric cracked stress-strain field of welded joints,are given also.

Experimental Study on Non-proportionally Multiaxial Time-dependent Ratcheting of SS304 Stainless Steel at Room Temperature

KAN Qian-hua, KANG Guo-zheng, ZHANG Juan, SUN Ya-fang

2006, 27(2): 59-63.

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The time-dependent ratcheting behavior of SS304 stainless steel was experimental studied at room temperature under non-proportionally multiaxial cyclic loading.The effects of stressing rate,holding time and loading path on ratcheting were discussed in the paper.The results indicate that the ratcheting of the material presents remarkable time-dependence,i.e.,the ratcheting depends not only greatly on loading rate,but also apparently on hold time even at room temperature.Further more the ratcheting also presents greatly dependence on loading path.Some significant conclusions are achieved to construct a time-dependent consti-tutive model of ratcheting.

Preliminary Studies on Hydrogen Deflagration Mitigation Measures during Severe Accident

XIAO Jian-jun, ZHOU Zhi-wei, JING Ying-qing

2006, 27(2): 64-67,77.

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Largescale hydrogen combustion and detonation in the containment building may threaten the integrity of the containment when a postulated severe accident takes place in a light water reactor nuclear power plant.Hydrogen recombiners and igniters are two promising hydrogen mitigation equipments under severe accident.Three mitigation measures,including using igniters only,employing hydrogen recombiners only and adopting the combination of recombiners and igniters,were investigated in this paper.The analysis results indicate that combination of hydrogen recombiners and igniters can mitigate the risk of hydrogen combustion in a safe,continuous,and effective way.

Study of Failure Mechanisms of Hexcan-Wall in Severe Accident Condition in Fast Reactors

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

2006, 27(2): 68-71,96.

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In the frame work of liquid metal fast breeder reactor safety,the failure location and time of stainless steel hexcan wall is an important issue,because it may greatly influence the courses and consequences of many severe accidents.For severe local accidents,the possibility of hexcan-wall failure is almost the same as the possibility of propagation of the accidents to neighboring subassemblies.According to the SCARABEE-N and SIMBATH experiments,a new failure mechanism was proposed.The melt-through of a well-cooled hexcan-wall could be caused by the burn out of sodium resulting from the local thermal erosion on the steel wall.Accordingly,for the Chinese experimental fast reactor,under the total instantaneous single subassembly blockage accident,the failure of the neighboring hexcan-wall was predicted to occur at 7.2~8 s after blockage formation.Subsequently,the molten mixture in the boiling pool begins to penetrate into the neighboring bundles,and then the accident propagates to the neighboring subassemblies.

Test and Study of Cold Condition for Dryer in Steam Generator for 1000MW PWR Nuclear Power Plant

CHEN Jun-liang, CHENG Hui-ping, XUE Yun-kui, WANG Xian-yuan, LIU Hong-yun, BA Zhang-xi, ZUO Chao-ping

2006, 27(2): 72-77.

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Screening test and study of air-water cold condition for dryer in steam generator for 1000MW PWR nuclear power plant was conducted.The dryer with optimized design features was obtained by the tests.Separation component of the dryer employed the double-hooked chevrons.Space between hook and chevron decreased gradually from the inlet to the outlet of the dryer;Baffle perforated asymmetrically was set in the inlet of the dryer;Diameter of hole on the baffle decreased from the bottom to the top.The tests indicated that this type dryer possessed higher critical velocity.

Optimization of Feedwater Pump Configuration for the M310 Nuclear Power Station

OUYANG Zhong-hua, HU Jing-song

2006, 27(2): 78-82.

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After reviewing the selection of the feed water pump configuration for Ling’ao I Nuclear Power Station and its operation,and based on the study results for several feedwater pump configurations which have been used in domestic and foreign nuclear power stations,several typical feed water pump con-figurations are compared and analyzed technically and economically.Considering the economics and reliabil-ity,we think the configuration of 2×50% turbine driven feedwater pump plus 2×25% electrical driven fe-edwater pump is better than that of 2×75% turbine driven feedwater pump plus 2×50% electrical driven feedwater pump.If the capability of one motor of feedwater pump can be increased from 10MW to 14～ 15MW and the voltage can be increased from 6.6kV to 11kV,the capacity of one electrical driven feedwater pump can reach 75% of the total capacity,then the configuration of 3×50% electrical driven feedwater pump is also accepted.

Study on Measurement Methods of Relative Internal Efficiency of Nuclear Steam Turbine

LI Yong, LIU Yu-duo, ZHANG Yi

2006, 27(2): 83-86.

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The relative internal efficiency for steam turbine is one of the main economic indexes in evaluating the flow path conditions of steam turbines.The steam turbine used in nuclear power plant works in wet steam area.The relative internal efficiency may not be obtained accurately through measuring of the wetness of regenerative extraction steam and exhaust steam.In this paper,by analyzing and comparing the enthalpy dropping and power types relative internal efficiency,it is found that when the operation conditions of the flow path in steam turbines change,both the two kinds of relative internal efficiency can reflect the operation conditions of the flow path in steam turbines.Also,when the operation conditions of regenerative system change,the enthalpy dropping and power types relative internal efficiency are equivalent.So either one can be used as the economic index for nuclear steam turbine.

Control and Protect System of the Zero Power Device Based on Programmable Controller

LU Yi, SU Min, YANG Cheng-de, LI Meng, XIANG Wei-ling

2006, 27(2): 87-90.

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In order to study the zero power device,the programmable controller is used to construct the control and protect system of the zero power device in this paper.The ability of logical process and I/O module are utilized and the system is modularized and redundant.The function of this system is tested and verified by simulation and it can be proved that the programmable controller could be used in the control and protect system.

Research on Nuclear Island Engineering ＆ Procurement Schedule Control

FAN Kai, XU Shun

2006, 27(2): 91-96.

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Firstly,“Basic Logic Diagram of NI Design ＆ Procurement” is presented in this paper,to illustrate the basic principle for the development of the design procurement plan,which is categorized by the system design,construction design and equipment procurement.Typical logic diagram for the schedule of each category is proposed followed by the detailed analysis of schedule preparation solutions.

2006 Vol. 27, No. 2