核动力工程

Numerical Calculation for Response of Core Barrel to Flow-Induced Vibration

XI Zhide, CHEN Bingde, LI Pengzhou

2008, 29(5): 1-4.

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The fluid load exciting structure vibration in weak coupled system is separated into two parts. One is related to the structure motion which is line with the inertia, damp and rigidity of structure, and is described by added mass, added damp, and added stress multipling the acceleration ,velocity, and displacement respectively.Normally, added damp and added stress can be neglected since is is rather small, and added mass is mainly considered. The other is independent of the structure motion which can be obtained by CFD. The response of core barrel of Qinshan II phase model induced by flow is calculated by this method .The result shows that under the wide-brand process, the response frequency is near the nature frequency of structure. For the bottom free boundary condition the core barrel vibrate under the first beam frequency. The calculation and experimental results are kept within 3.1 times. So the method is suitable for the preexperiment evaluation.

Impact Simulation of HTR-10 Vertical Magnetic Suspension Rotor Dropping into Catcher Bearings in Early Design Phase

TAN Zhongwen, WAN Li, HE Shuyan

2008, 29(5): 5-8.

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Catcher bearing is the auxiliary support structure for the magnetic-bearing and turbine rotor system, which is the safety guarantee for the system operation and the key and difficult technology in the design of the magnetic-bearing and turbine rotor system. This paper presents the analysis of impact simula-tion of the HTR-10 vertical magnetic suspension rotor in early design phase, by dropping into the catcher bearing with MSC.Dytran software. The results indicate that if the carbon fiber is designed for the buffer layer structure, the AMB may be damaged, so alternative material shall be considered. The rotary attitude of the vertical magnetic suspension rotor plays a vital role in the performance of the catcher bearing.

Application of Leak-Before-Break （LBB） Analysis Method for Liquid Metal Cooled Fast Breeder Reactors（LMFBR） in Pipeline Analysis

XIONG Dongqing, LU Daogang

2008, 29(5): 9-14.

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Based on the Leak-Before-Break （LBB） Analysis Method, this paper performs a fracture mechanics analysis on the intermediate loop of the decay heat removal system in China Experimental Fast Reactor（CEFR）. Assuming that before the operation of the reactor a throughwall defect had existed where the loads were most serious, according to J-integral tearing modulus inter-section method, the critical length of crack in the worst condition （namely synthesized enveloped operation conditions） was obtained. With Paris formula, the stability expansion of the crack was calculated under alternating loads. The open area of the critical crack as well as that of the life-end crack were obtained by a mechanical model established by CASTEM2000 code. It is proved that the biggest crack open area of the tube is less than DT/4 （D means pipeline diameter,and T tube wall thickness） at the end of its life.

Dynamic Propagation of Asymmetrical Mode Ⅲ Interface Crack

LU Nianchun, CHENG Yunhong, WANG Yuntao, CHENG Jin

2008, 29(5): 15-21.

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By the methods of the theory of complex functions, dynamic propagation problems concerning asymmetrical mode Ⅲ interface crack are studied. Asymmetrical dynamic propagation problems under the action of variable loads （Px/t） and （P³t/x²） located at the origin of the coordinates respectively, are resolved by the measures of self-similar functions, and the universal expressions of analytical solutions to stresses, displacements and stress intensity factors are gained. The variational rule of dynamic stress intensity factors is illustrated by the approaches of numerical computations, and the result obtained is coincident with relevant literatures.

Finite Element Contact Analysis of "O" Ring Sealing Structure in Horizontal Channel

ZHOU Dan, SHENG Xuanyu

2008, 29(5): 22-24,32.

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An elastic-plastic nonlinear analysis is conducted by ABAQUS software to study the sealing structure of flange connection with different material in horizontal channel. The processes for pretightening of primary bolts and pressurizing in the horizontal channel are simulated by finite element multi-step skill. The stress distribution and sealing performance of primary flange is studied. The analysis results show that the double O-ring sealing structure in horizontal channel can meet the requirements on strength and sealing dur-ing the pre-tightening and pressurizing processes.

Analysis of Prompt Negative Temperature Coefficient for UraniumZirconium Hydride Fuel Element in Power Reactor

WANG Lianjie, YAO Dong, CHEN Bingde

2008, 29(5): 25-28.

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The effects of weight percentage of uranium, fuel rod radius, fuel temperature, hydrogen content and additional poison on prompt negative temperature coefficient for uranium zirconium hydride element are calculated respectively in this paper, and then the results are analyzed. The study shows that the absolute value of prompt negative temperature coefficient decreases observably along with the increasing of weight percentage of uranium. Additional poison erbium makes prompt negative temperature coefficient much more negative.

Analysis of Reactivity Temperature Effect of CFBR-II

ZHAO Xiaolin, FAN Xiaoqiang

2008, 29(5): 29-32.

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The reactivity temperature effect of CFBR-II reactor is analyzed with single-group perturbation theory, the calculation formulas of the reactivity temperature coefficients of CFBR-II reactor under steady-state and pulse-state are obtained, and the factors that have influence on reactivity-tempera-ture-coefficient of CFBR-II reactor are discussed. These results show that the reactivity temperature coefficients of CFBR-II under steady-state and pulse-state are negative, which values are not equal. The reactivity temperature coefficient of CFBR-II under steady-state is depended on the reactor structure and materials, the thermal exchange conditions in reactor core and between reactor and surroundings, which value is independed on the reactor power. The reactivity temperature coefficient of CFBR-II reactor under pulse-state is an ap-proximate constant during pulse, which value is depended on the reactor structure and materials.

Angular Dependent Rebalance Acceleration Arithmetic for Neutron Transport Equation in Unstructured Mesh

JU Haitao, WU Hongchun

2008, 29(5): 33-37,42.

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The first-order neutron transport equation is solved by the least-squares finite element method based on the discrete ordinates discretization. The angular dependent rebalance （ADR） acceleration arithmetic and its extrapolate method are given. The numerical results of some benchmark problems demonstrate that the arithmetic can shorten the CPU time to 34%~50% and it is effective even for the strong scattering problem.

Study on Positron Annihilation Life for New U₃Si₂-Al Fuel Plates

ZHANG Zhihua, LIU Hangang, DENG Yue, QIAN Dazhi, ZHOU Wei, LIU Yaoguang, XU Xianqi

2008, 29(5): 38-42.

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Microstructure damage configuration and transformation of shell materials for U₃Si₂-Al fuel plates under different technical conditions were studied by positron annihilation technique. The study showed that micro-effects in the shell material for tempered fuel plates were mostly single vacancy point defects, that in cooled fuel plates were mostly couple vacancies and dislocation, and that in eroded fuel plates were mostly layer-dislocation and small vacancy groups. There were no large vacancy groups which had effect on the safety of fuel plates in all samples. Some technique such as temper and eroding would change microstructure damage of shell materials, and also alter mechanical capability of fuel plates.

Analysis of Flow Instabilities for Once-Through Steam Generator

HOU Suxia, ZHAO Fuyu, LIU Cheng, TAI Yun

2008, 29(5): 43-47.

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This paper analyzes the two-phase flow instabilities in parallel tubes using frequency domain control theory for steam generator in order to disclose the mechanism of flow instability. The mathematics expressions of heat transfer and flow of steam generator are proposed,and the eigenfunction of the close loop system is deduced using linearize perturbation theory and Laplace transformation. System stability is judged by Nyquist stability criterion of automatic control theory. On the above bases,the computer code is derived,and the stability analysis of steam generator is made by using this code. The results show that there is certain relations between the system stability and the channel number. The increasing of entrance resistance factor enhances the system stability; and the reducing of operation load impairs the system stability gradually.

Study on 3-D Flow Field in 5×5 Rod Bundles with Spacer Grids

TIAN Ruifeng, MAO Xiaohui, WANG Xiaojun

2008, 29(5): 48-51.

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This paper studied the 3-D single-phase flow field in 5×5 rod bundles with spacer grids by numerical method. By using hybrid grids technique, SST k-ω model and SIMPLEC algorithm, the Reynolds averaged mass conservation equation and momentum conservation equation were solved, and the pressure and velocity field were obtained. The results show that the spacer grid leads to intense lateral flow in rod bundles channel, and the lateral vilocity submits parabola distributing raw; axial velocities are distributed uniformly in the channel, and drag coefficient decreases as inlet Reynolds in the universe degrade. Calculation results agreed with experiment data well.

Experimental Study on Bubble Growth in Vertical Narrow Channel

CHEN Deqi, PAN Liangming, YUAN Dewen, DENG Jiewen, WANG Xiaojun

2008, 29(5): 52-55,59.

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A visual investigation of vapor bubble growth in subcooled upward forced convection boiling in vertical rectangular narrow channel （the gap is 2 mm） was performed. High speed camera was used to record the bubble growth with a speed of 5000 fps（frames per second）, and the picture size are 512×512 pixel. Experimental results about bubble growth in various work conditions （with variation heat flux and mass flux） were recorded; and based on the regression analysis, the exponential fitting relationship between the mean equivalent bubble diameter（D_e） and the bubble growth time was revealed. The results show that the exponential fitting can predict the mean bubble growth in narrow channel very well; but the coefficients of K and n, are quite different with that of conventional channel; and the K is less dependent on Ja number other than that of conventional channel, and the n has the value of wider range of 0.339～0.914.

Experimental Study on Pool Transition Boiling with Downward Facing Heated Surface

ZHAO Dawei, SU Guanghui, TIAN Wenxi, QIU Suizheng, K. Sugiyama

2008, 29(5): 56-59.

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Experimental study was conducted to examine the pool boiling occurring on a relative large downward facing surfaces confined space with width of 2.2, 2.6, 3.0, 3.2 and 13.0mm, respectively, at atmospheric pressure using water as the working fluid. The boiling curves of distilled water with downward facing heated surface were obtained. The heat flux on downward facing heated surface is less than one tenth of that on upward facing heated surface. The effects of wall superheat, ΔT_W, and gap size, δ, on transition boiling were also analyzed. The results show that Nu decreases with the increasing of ΔT_W under transition boiling condition. And increases generally with the increasing of gap size, δ, though there is a small peek on Nu-δ curve in narrow gap size area.

Technical Scheme for Test of Hydraulic Characteristics of Fission Target Irradiation Facility

QIAN Dazhi, FANG Ying, WANG Sheng, YANG Laisheng, MENG Yang, ZHANG Bo

2008, 29(5): 60-63.

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Considering the flow distribution test requirement of limited flow channel in fission target irradiation facility in nuclear reactors, the technical scheme for test of hydraulic characteristics for the facility is proposed in this paper. A subminiature pressure transducer is adopted to measure the flow pressure drop in canular channel. Additional demarcated experiment has been performed to achieve the relation between pressure drop and flow rate. The maximum of relative error of test and calculated value is less than ±2%. The relation between pressure drop and flow rate can be used in the research on the hydraulic characteristics of fission target irradiation facility in nuclear reactors.

Human Factor Analysis and Preventive Countermeasures of Maintenance in Nuclear Power Plant

ZHANG Li, HU Chao

2008, 29(5): 64-68.

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Based on the human error analysis theory and the characteristics of maintenance in a nuclear power plant, human factors of maintenance in NPP are divided into three different areas: human, technology, and organization, in which human refers to the individual factors, mainly including psychological quality, physiological characteristic, state of health, knowledge, skill level, and interpersonal relationship. Technology includes the maintenance technology, maintenance strategy, maintenance tool, maintenance interface, maintenance regulation, and work environment. Organization includes task arrangement, information communication, training, personnel external environment, team construction, and leadership. The analysis also reveals that the organization factors, which can indirectly influence personnel performance, are the primary initiators of human error. Based on these, some countermeasures are brought forward in order to reduce human errors.

Investigation on Scram Events Caused by I＆C Faults and Human Failures in Nuclear Power Stations

ZHOU Ping

2008, 29(5): 69-71.

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72 scram events in the nuclear power stations in China are classified in this paper, with focus on the analysis of the scram events caused by I＆C faults and human failures.The result demonstrates that the phantom signal, control unit malfunction, parameter unmerited and control valve malfunction are the first cause for instrument control equipment failure. Preventive solutions are proposed.

Study on Cooling Model for Debris in Lower Plenum and Countermeasures for Prevention of Focusing Effect

GUAN Zhonghua, YU Hongxing, JIANG Guangming

2008, 29(5): 72-76.

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From the basic energy conservation equations and experimental or empirical correlations, an intact model is constructed for the thermal calculation of the core debris in the lower plenum. For verification of this model, the results of two calculations for AP600 and AP1000 plants are compared with those presented in relevant literature. The analysis highlights on the impact of the decay heat power density and the focusing effect. In order to mitigate the focusing effect, it is proposed in this paper to change the lower head profile from hemisphere to parabola. The results show that this change of lower head profile can change the heat flux distribution of the debris, and mitigate the focusing effect.

Modeling for Reliability of Pressurizer Safety Valve in Nuclear Power Facilities

HUANG Yan, ZHOU Mi, LUO Zhiyuan, LI Xiaozhong, WANG Guangjin

2008, 29(5): 77-79,90.

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Based on the analysis of the function and mission profile of the pressurizer safety valve in nuclear power facilities, the basic reliability diagrams and the mission reliability diagrams were developed, and the reliability mathematical models were derived with the universal probabilistic method. The reliability modeling work is essential to the failure modes, effect and criticality analysis （FMECA） and reliability allocation.

Application of a New Type Fuzzy Controller in Power Control System for Nuclear Reactors

WU Hongxing, ZHAO Fuyu, ZHANG Zhongmin

2008, 29(5): 80-82.

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The paper introduces a new type general fuzzy controller, and its application in the design of a small research reactor power control system. The algorithm formula coveres all regulating means of the old-time fuzzy controller. The paper compares it with the conventional PID controller. Simulation results with a new type fuzzy controller and with a conventional PID controller show that the new type fuzzy control sys-tem has greatly improved the performance of reactor power control system, and its disturbance resistance is better than that of common PID controller.

Development Trend for Macroscopic Benchmark Test on Fast Neutron Critical Experiment

LI Jiansheng

2008, 29(5): 83-85,93.

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It is necessary to further improve the calculation accuracy for fast neutron critical assembly. In this paper, the origin, effect, method and measurement of benchmark test on fast neutron critical experiment is summarized.The research advances home and abroad are introduced. The suggestion is proposed with consideration of conditions in China.

Design of Reactor Alarm Instrument Based on SOPC

LI Meng, LU Yi, RONG Ru

2008, 29(5): 86-90.

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The design of embedded alarm instrument in reactors based on Nios II CPU is introduced in this paper. This design uses the SOPC technology based on the Cyclone series FPGA as a digital bench, and connects the MPU and drivers and interface of times,RS232, sdram,and etc. into a FPGA chip. It is proved that the system achieves the design goals in primary experimentation.

Measurement of ²⁵²Cf Fast Ionization Chamber Fission Neutron and γ Ray TOF Spectrum

ZHANG Yi, LI Jiansheng, JIN Yu, ZHOU Haojun

2008, 29(5): 91-93.

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This paper introduces the measurement method and result for the ²⁵²Cf fast ionization chamber fission neutron and γ ray TOF spectrum, and describes the composition and working principle of the measurement system in detail. The neutron and γ ray TOF spectrum measured by this method, with the FWHM of its γ ray peak being 0.9ns at best, can be applied to timing precision measurement for nuclear detection system and to check the effect of n-γ discrimination.

Performance Measurement of Aurum-Silicon Surface Barrier Detector in ⁶LiF Sandwich Semiconductor Spectrometer

JIANG Yong, LI Junjie, ZHENG Chun, XIAO Jianguo

2008, 29(5): 94-97.

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The Aurum-Silicon surface barrier detector is one of the key consituents of the ⁶LiF semicon-ductor sandwich spectrometer, which parameters will directly affect the performance of ⁶LiF semiconductor sandwich spectrometer. In this paper, leakage current, dead-thickness, the effect of air on α partical, the effect of bias on the peak position and resolution under different bias and largest energy peak position under 180V bias for Aurum-Silicon surface barrier detectors are experimentally measured. Based on the measurment re-sults, two groups of Aurum-Silicon surface barrier detectors with similar performance are choosed to form the effective ⁶LiF Semiconductor sandwich spectrometer and the background spectrometer.

Modular Design of Nuclear Pilot Operated Safety Valve

LUO Zhiyuan, WANG Sheng, LAI Xuping, WANG Taotao, LI Xiaozhong

2008, 29(5): 98-102.

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Modular Design is a new design method which is rapidly developed in the world. The basic concept of modular design was introduced in this paper. The modular construction of pressurizer pilot operated safety valve was analyzed and divided into several independed modules. The design proposal of the pilot operated safety valve localization production was studied based on the former modular analysis for the nuclear auxiliary system such as RRA and RCV.

Design and Experimental Study on Steam Injector with Pressure-Gain

ZHAO Liangju, LI Desheng, WANG Fei, LI Bin

2008, 29(5): 103-106.

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In the steam injector with pressure-gain, the water gains pressure by the power of steam. The outlet water pressure is higher than the inlet steam pressure. The characteristics of pressure-gain makes it possible to use the injector in the nuclear passive safety system. The study on the injector design is rarely reported. This paper presents the method to design the steam injector with pressure-gain. An experimental system was established according to the design. The experimental characteristic curve agrees well with the design curve, which proves the security and reliability of the design.

Electrical Penetration Assembly for Fast Critical Assembly

LI Maohui, YANG Chengde

2008, 29(5): 107-108,113.

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A electrical penetration assembly which does not require to cut cables and can be used in the hall for the fast critical assembly is introduced in this paper. The structure of the electrical penetration assembly is simple and facilitates the cable replacement. The results from the leakage test on the electrical penetration assembly and the hall indicate that the leakage ratio is less than the design ratio, and the overall seal performance of the hall satisfies the requirement on the normal operation and accident conditions for the fast critical assembly.

Development of Manipulator for In-Service Inspection and Maintenance of NPP Steam Generator

WANG Liquan, WU Jianrong, LIU Yulong

2008, 29(5): 109-113.

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Considering the specifically application environment for in-service inspection and maintenance of steam generators, an multifunctional manipulator system was developed first in China. This manipulator can be used in many occasions, such as eddy current inspection of tube, ultrasonic inspection, plugging, sleeving, and television inspection. The manipulator includes a base modular and six rotational joint modules. By modular design, not only the reliability and maintainability of the manipulator are improved, but also the requirements by different SGs are satisfied. The manipulator worked at a special two-base model, which improved the efficiency of services and greatly decreased the levels of radiation.

Design of Shielded Operation Cupboard for 10⁶ ²⁵²Cf Spontaneous Fission Neutron Source

JIN Yu, LI Jiansheng, ZHANG Zhen, ZHOU Haojun, JIANG Congming, SONG Lingli

2008, 29(5): 114-116.

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The paper describes the research and development of shielded operation cupboard for 10⁶ ²⁵²Cf spontaneous fission neutron source. The design principle, design idea,calculation of shielding material and choice of shielding material for shielded operation cupboard are introduced. The performance for the shielded operation cupboard is measured by using low radiation intensity source. Sheiding materiral outside/inside of the shielded operation cupboard is stainless steel shell （2 mm）, polyethylene （150 mm）, and stainless steel wall （20 mm）. The effect for neutron shielding reaches 9 times, and that for Gamma ray shielding reaches near 3 times at the operation position for operator. The results show that the shielded operation cupboard meets the needs for practical operation and dose limit value requirement on the operator.

Design of Time D-T Neutron Tube

ZENG Liheng, LI Jiansheng

2008, 29(5): 117-118,137.

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A new type of D-T neutron tube with time function has been designed,which will replace ²⁵²Cf source and be used as an interrogation source for the Nuclear Materials Identification System（NMIS） and the reactor noise analysis system. An alpha particle detector is installed in a D-T neutron tube, the time of the emitted neutron will be measured by detecting the alpha particles generated by the T（d,n）⁴He reaction.The design accords with the conditions of the reactor noise analysis system,which relies on the mature craft of neutron tube in China.

Real-Time Simulation of Nuclear Power Condensers

XUE Ruojun, TIAN Zhaofei, ZHAO Qiang

2008, 29(5): 119-123.

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A real-time model describing the dynamic behaviour of a nuclear power shell-and-tube condenser is presented, according the structure and working characteristics of the condensor. The model can be used under steady-state as well as transient conditions, and is also able to predict vapour and condensate flow rates, and the temperatures of the vapour, condensate, wall and coolant. The results of several simulations are analyzed after the simulation of normal and malfunction state. The results show that the model can be used for real-time simulation of nuclear power condensers.

Numerical Simulation of 3D Inclined Jet to Cross Flow

BU Lin, LI Huixiong, WANG Haijun, CHEN Tingkuan

2008, 29(5): 124-128.

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This paper is concerned with the thermal shock when the cold safety injection fluid is injected into the nuclear reactor. Based on a control volume method and k-ε turbulence model, numerical simulation was carried out to investigate the flow and the heat transfer of 3D inclined jet to cross flow at angles of 30°、60°、90°、120ånd 150° in the T-junction pipes. The pressure distribution, the temperature distribution and the flow structure were thoroughly studied. The paper analyzed the effect of geometric structure, the flow condition and the temperature variety on the flow structure, temperature distribution and thermal stress. The study found that temperature difference and velocity ratio had great effect on the thermal stress. On the upstream of the jet, thermal stress is more powerful.

Development of Supporting Software for Safety Analysis Simulator for Nuclear Reactor

LI Tonglin, YAO Qingsheng, HAN Weishi

2008, 29(5): 129-132.

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An investigation of nuclear reactor simulator was made on its physics model, thermal hydraulics model, modeling method for assistant system and numerical calculation technology. The simultaneous equations for each module and node of the steam supply system are solved by full implicit difference method. Then the supporting and calculation softwares for simulation was developed based on Windows system. Calculation and comparison have been done for small break LOCA by the safety analysis simulator developed in this paper and Relap5 code. The results show that the variation of primary parameters for break LOCA of this model accord with the calculation results from Relap5, and the simulator can perform real time calculation with well stability.

3-D Design Method for Welding Groove and Seal Weld of Reactor CRDM Adapter

MA Baiyong, WANG Xiaobin, ZHU Xiaoyong

2008, 29(5): 133-137.

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Based on the analysis of 2D and 3D shapes of welding groove and seal weld of reactor CRDM adapter, four intersecting curves are defined, and a method and gist to 3D design of adapter welding groove and seal weld is proposed. Parameterized design of adapter welding groove and seal weld has been realized using UG software, and the main factors which affect the welding section areas have been analyzed. Compared with the measurement, the error of weld section area of each adapter created by spline fittingmethod is less than 0.8%.

Discussion about Primary Water Chemistry Control Method in Qinshan Phase II

LIU Xinfu, ZHANG Lefu, GAO Minghua

2008, 29(5): 138-141.

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A 4-grade Boron-Lithium coordination solution to lower the risk of Reactor （inner） Structure corrosion by controlling lithium content is proposed based on the investigation of optimized primary loop water chemistry control method for the mixing core adopted in Qinshan Phase II. Validity of the optimized B-Li water chemistry control solution applied in the mixing core operation for anti-corrosion of structure material and reducing radiation in primary loop is testified by following up and analyzing the differential fuel cycle corrosion activation products.

Research on Polyantimonic Acid Synthesis and Adsorption for Strontium

DENG Qimin, LI Maoliang, CHENG Zuoyong

2008, 29(5): 142-144,14.

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The paper describes a new mothod for Polyantimonic Acid （PAA） synthesized and absorb capacticy of Sr²⁺ and UO₂²⁺ on PAA. PAA synthesized by oxidation process with SbCl₃ is with good mechanical strength and suitable for column operation. PAA has high affinity for Strontium in acid solution, and the adsorption capacity for Sr²⁺ is 42mg/g in 0.1mol/L nitrate acid solution. But the adsorption capacity of UO₂²⁺ on PAA is 27.7 mg/g. The results show that PAA can be used for ⁹⁰Sr removal from acid solution without UO₂²⁺.

2008 Vol. 29, No. 5