核动力工程

Reliability Analysis of Reactor Pressure Vessel Intensity

ZHENG Liangang, LU Yongbo

2012, 33(4): 1-4.

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This paper performs the reliability analysis of reactor pressure vessel（RPV） with ANSYS.The analysis method include direct Monte Carlo Simulation method,Latin Hypercube Sampling,central composite design and Box-Behnken Matrix design.The RPV integrity reliability under given input condition is proposed.The result shows that the effects on the RPV base material reliability are internal press,allowable basic stress and elasticity modulus of base material in descending order,and the effects on the bolt reliability are allowable basic stress of bolt material,preload of bolt and internal press in descending order.

Exploitation and Application of SUESGTP

WU Jiabin, ZHENG Xin, GAO Ming

2012, 33(4): 5-8.

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The calculation principle,configuration,validation process and result of CPR1000 nuclear reactor power calibration system are introduced.Based on IAPWS-IF97,which is a model of water and steam thermodynamic properties,and enthalpy difference-thermal power theory,CPR1000 nuclear reactor power calibration system is used for calculating the thermal power of nuclear steam supply system and the nuclear power of the reactor.The calibration system was used during the start-up of unit 3 of Ling’ao NPP,and on-line calibration for reactor power measuring and control system had been conducted successfully.It is proved that CPR1000 nuclear reactor power calibration system can be used to calibrate the reactor power measure and control systems reliably.

Influence on Injection Occasion to PWR Severe Accident Progression

YU Jian, XU Jin-liang, CHEN Wu-xing, TANG Yao-yang

2012, 33(4): 9-12.

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A typical three-loop pressurized water reactor (PWR) as reference planted is chose, and the PWR severe accident analysis models based on RELAP5/SCDAP/MOD3.2 code are developed. The best estimate calculation of hot leg large break LOCA of 18 cm without mitigating measures is performed. The mitigating capability of injection to the accident is studied when peak core surface temperature reaches 1100 K, 1300 K and 1500 K respectively. The results show that the injection occasion plays an important role in the accident progression. Injection during early stage will prevent core from melting effectively. However, injection during later stage may deteriorate the progression of severe accident and accelerate of core melt.

Accident Analysis of Station Blackout of Pool-Type Research Reactor during High Power Operation

HUANG Hongwen, LIU Hangang, QIAN Dazhi, XU Xianqi

2012, 33(4): 13-16.

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Based on the main event sequence,RETRAN-02 code is used to analyze the power supply station blackout accident and key thermal parameters of a pool-type research reactor and demonstrate the enduring capability of the accident.The result proves that the decay heat of the reactor could be removed through the idler wheel rotating,residual heat removal system supported by reliability power,and the natural circulation,which will prevent the nuclear safety accidents.The accessorial cooling supported by reliability power is an effective measure to relieve the accident,and the power can last for an hour at least.

New Interface Issues of Nuclear Reactor Neutron Flux Measurement System Using Unified Standard Signal

WANG Xuejie, TANG Fengping, ZHU Shilei, HUANG Wen, ZHONG Dingyong

2012, 33(4): 17-19.

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When nuclear reactor control and protection system uses a unified standard electrical analog signal（4~20 mA）as an exchange signal,there exist problems of synchronization between the virtual value and range of output signals during switching of Neutron Flux Measurement System（NFMS）,and thus false actions occur in the protection system,the alarm system and the power control system.The cause of these new interface issues were described in this paper,and one solution for this problem was given in detail.

Study on Application of T-S Fuzzy Neural Method in Once Through Steam Generator Feedwater Control

CHEN Zhi, LIAO Longtao, LIU Lixin, LI Wei

2012, 33(4): 20-23,33.

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Considering the problems of small water inventory and inferior heat accumulation capacity of the secondary side of the once through steam generator（OTSG）,and the need of rapid match of feedwater flowrate and the reactor power to ensure the steam quality,the T-S fuzzy neural method is incorporated into the OTSG feedwater control system to design a T-S fuzzy neural controller based on input and output data acquired by numerical simulation.The control results are verified by simulation method.The simulation results show that the control results under this control manner are better than those under the PID control manner and meet the control requirements satisfactorily.

Research on Homogenization via Monte Carlo Method and Its Application in Multi-group Monte Carlo Transport Calculation

ZHANG Peng, WANG Kan, LI Mancang

2012, 33(4): 24-28.

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A two-step calculation scheme may be one of the feasible ways if we intend to do core simulations with Monte Carlo method,i.e.,assembly level calculations with Monte Carlo method（MCNP） are done at first,the group constants are tallied at the same time,and then the core level simulations are performed using these group constants.In this paper,the group constants generation method via Monte Carlo is studied,and an equivalent homogenization theory has been applied to improve the results.The numerical calculation results illustrate that the group constants generation method via Monte Carlo is practicable,which can improve the efficiency of core simulations while ensuring the accuracy.

Particle Swarm Genetic Algorithm and Its Application

LIU Chengyang, YAN Changqi, WANG Jianjun, LIU Zhenhai

2012, 33(4): 29-33.

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To solve the problems of slow convergence speed and tendency to fall into the local optimum of the standard particle swarm optimization while dealing with nonlinear constraint optimization problem,a particle swarm genetic algorithm is designed.The proposed algorithm adopts feasibility principle handles constraint conditions and avoids the difficulty of penalty function method in selecting punishment factor,generates initial feasible group randomly,which accelerates particle swarm convergence speed,and introduces genetic algorithm crossover and mutation strategy to avoid particle swarm falls into the local optimum.Through the optimization calculation of the typical test functions,the results show that particle swarm genetic algorithm has better optimized performance.The algorithm is applied in nuclear power plant optimization,and the optimization results are significantly.

Calculation of Lead Shielding in Relative Distribution Measurement Systems of Neutron Flux in Reactors

LI Gaofeng, LIU Caixue, LAI Wanchang, MU Keliang

2012, 33(4): 34-36.

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The measurement system for relative distribution of multichannel neutron flux is designed to massively measure the radioactivity of irradiated foils.Every NaI detector measures one irradiated foil and the lead shield is used in detectors.To eliminate the effect of foils from other tracks on the measurement result of foil detector in its own track,this paper designs that the interference from the irradiated foils at the adjacent tracks on the measured tracks is less than 0.05%.This paper confirms the size of lead shield which meets the shield requirements by using the calculation method of point source integral and Monte Carlo verification method.

Analysis on Reason for Slender High U-Content U-ZrH_x-Er Fuel Rods Cracking during Hydrogenating

LIU Siwei, ZHANG Ruiqian, LAN Guangyou, PU Yongxing, YUAN Zhengchuan, HU Rui

2012, 33(4): 37-39,49.

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The microstructure of the alloy specimens which were melted with the impure uranium and pure uranium respectively,and their hydrided specimens were analyzed by OM and SEM,moreover,the composition of the hydrided specimens were studied by SEM-EDS.The results showed that the contents of impurity elements of Fe and C were both higher in the impure uranium-contained specimen than in the pure uranium-contained specimen,in which the elements were more homogenously distributed in the zirconium hydride matrix.There was great difference of expansion coefficient between the precipitate which consisted of the elements of U,C,a little Fe and zirconium matrix in the impure uranium-contained hydrided specimen,thus when the volume of zirconium matrix increased due to absorbing hydrogen,the precipitate as a secondary phase pinned in the matrix so that the stress was concentrated resulting in the hydrided specimen cracking.In the pure uranium-contained hydrided specimen,the elements of U and C which existed in the form of zirconium carbide relatively homogenously distributed in the zirconium hydride matrix,and the stress in the hydrided specimen did not concentrate so that the hydrided specimen did not cracked.

Wet Process of External Gelation of Uranium for Preparation of Uranium Dioxide Kernel of High Temperature Gas-cooled Reactors

ZHOU Xiangwen, HAO Shaochang, ZHAO Xingyu, MA Jingtao, WANG Yang, DENG Zhangsheng

2012, 33(4): 40-43.

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In order to prepare the dense uranium dioxide（UO₂） kernel for high temperature gas-cooled reactors（HTGR） fuel,the conventional sol-gel processes are optimized and modified.The wet process of modified external gelation of uranium（EGU） is introduced,which includes the dissolution of U₃O₈,i.e.the preparation of acid-deficiency uranyl nitrate,preparation of broth solution,casting and gelation of broth solution and aging,washing and drying of the gelled spheres,and etc.The mechanism of wet process of EGU is also discussed.With the optimized wet process,the ammonium diuranate（ADU） microspheres with good sphericity,uniform diameter and perfect porous structure are prepared.After the subsequent treatments of dry processes such as calcination,reduction and sintering,the eligible dense UO₂ kernels for HTGR fuel are manufactured.

A Cladding Oxidation Model Based on Diffusion Equations

HE Xiaoqiang, YU Hongxing, LI Feng, JIANG Guangming

2012, 33(4): 44-49.

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During severe accident in PWRs,the cladding oxidation with steam in the core is very important to the accident process.When oxidation time is long,or oxidation occurs in steam starvation conditions,the parabolic rate correlations based on experiments are restricted,which impacts the prediction of cladding failure,hydrogen production,and temperature.According to Fick’s laws,a cladding oxidation model in a wide temperature range based on diffusion equations is developed.The developed oxidation model has a wider applicability than those parabolic rate correlations,and can simulate long-term experiments well.The restricted assumptions of short-term oxidation time and enough steam environment in the core implemented by those parabolic rate correlations are removed in the model,therefore this model perfectly fit for long-term and steam starvation conditions which are more realistic during a severe accident.This model also can obtain detailed oxygen distribution in the cladding,which is helpful to simulate the cladding failure in detail and develop advanced cladding failure criteria.

Study on Resistance to Intergranular Corrosion of Heat Transfer Incoloy800H Tube

LI Jufeng, HAN Jiancheng, WU Zhijun, WANG Yi, SHI Zhangren, ZHANG Yanpeng, YU Huahe

2012, 33(4): 50-53,63.

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By using GB/T15260-94 B method,i.e.,copper-copper sulfate-16% sulfuric acid evaluates intergranular corrosion sensitivity of nickel-based alloys,the intergranular corrosion test results of domestic heat transfer Incoloy800H tube of High-Temperature Gas-cool Reactor（HTR） nuclear power plant demonstration project are studied under different test conditions,and compared to the intergranular corrosion performance of imported Incoloy800H alloy tube.The results show that the main factors to influence the alloy resistance to intergranular corrosion are C and Ti contents,and the sensitivity increases with increasing of C content.The effective measure to prevent intergranular corrosion is to add Ti element to reduce the sensitivity to intergranular corrosion.

Experimental Study on Absorption Spheres Pile after the Interruption and Restart of Pneumatic Conveying

LI Tianjin, CHEN Feng, HE Xuedong, CHEN Wensheng, SUN Bo, HUANG Zhiyong

2012, 33(4): 54-57.

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The experiments were conducted in a positive pressure system to investigate the absorption spheres pile after the pneumatic conveying interruption and the restart characteristics.The ambient air was used as the source gas.The absorption spheres were replaced by the glass spheres.The average pile height in the conveying pipe of the experiment system was 1172 mm,which was ~11% of the transport pipe height,as the pneumatic conveying interruption happened at the steady state stage.The continuing spheres supply from the feeder during the blower flow rate decreasing process should be the main factor for the pile up amount of spheres in the conveying pipe.The spheres amount in the steady state conveying line was only ~10% of the pile amount.The maximum and steady state pressure drop of the gas-solid conveying line in the restart process were approximately equal to those of the normal start process.

Experimental Study on Performance of Magnetic Levitation Control Rod Driving Line

ZHANG Zhihua, BO Hanliang, MI Xiangmiao, XU Xianqi, WANG Jiaying, ZHANG Xinrong, WU Xinxin

2012, 33(4): 58-63.

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To test the performance of magnetic levitation control rod driving line and verify the stability and reliability,the comprehensive study on drive line is conducted.The operational performance test gives out the static and dynamic parameters of the driving system and driving line.The operational life test proves the stability and reliability of the control rod driving line and verifies the comprehensive performance.

Technical Proposal for Polar Crane of AES-91 Nuclear Power Units

JIANG Baiwen, MA Yuandong

2012, 33(4): 64-66,90.

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The technical characteristics,function and standard system for the polar cranes of Russian AES-91 nuclear power unit and domestic M310 are compared and analyzed,and the differences are balanced based on the principle to ensure the safety function and cover the two standard system,and then the technical proposal for domestic manufacture of polar cranes for Tianwan NPP Units No.3 and No.4 is provided.

Welding Deformation Control of Bottom Mounted In-Core Instrumentation Thimble Guide Columns in Reactor Internals

WANG Qingtian, LI Yan, LI Na, XU Bin, JIANG Xingjun

2012, 33(4): 67-71.

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To the welding deformation of instrumentation thimble guide columns made of austenitic stainless steel,the paper put forward some measures to control the welding deformation,including optimizing the welding procedure and welding parameter.According to the improved welding procedure,no penetrant indications and radiographic defects appear after welding.Besides,welding deformations reduce greatly and no positional tolerance and dimension tolerance are out of the requirements.The improved welding procedure fully meets the design requirement.

Design of Control System for Test Platform of Nuclear Fuel Gripper

FANG Yu, ZHAO Apeng, WU Fengqi, LU Xiusheng

2012, 33(4): 72-75.

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As the key component in the fuel handing and storage system,the test on the function and durability of the fuel gripper is very important.Considering the actual work conditions of the gripper and using the programmable logic controller,the raising-lowering of the load and the locking and unlocking the gripper are implemented by the hydraulic drive and the pneumatic drive respectively.The design of the electrical hardware,PLC software and human machine interface software are also presented,and the test platform is established.The application shows that the control system of this test platform is with flexible control,convenient operation and high reliability.

Analysis of Flow Excursion Instabilities in Helically Coiled Tube Steam Generator of High Temperature Gas-Cooled Reactor

ZHU Hongye, YANG Xingtuan, JU Huaiming, JIANG Shengyao

2012, 33(4): 76-80.

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Flow excursion in helically coiled tube steam generator of high temperature gas-cooled reactor（SG） was analyzed by numerical simulation based on a one-dimensional model.Drift-flux model for the water/steam side and compressible flow model for helium side were used.Results of dynamical simulation show that flow excursion may occur under certain conditions,which results in remarkable differences of mass flow rate and temperature between the parallel tubes.The analysis of pressure-drop versus mass flow rate curve reveals that power input is the main parameter that affects the onset of flow excursion.With the increasing of power input,flow excursion occurs at larger range of mass flow rate.To increase the inlet resistance and sub-cool degree is helpful to avoid the onset of flow excursion.Stability maps were also presented.

Optimization Study on Flow and Heat Transfer in an Equivalent Model for Sphere Bed

ZHOU Huihui, YAN Xiao, CHEN Bingde, XIAO Zejun

2012, 33(4): 81-84.

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An equivalent model methodology for studying the thermal-hydraulics problems of spherical fuel assembly water cooled reactor is proposed.In this paper,3D geometry modeling,meshing and simulation calculation of flow and heat transfer for narrow-gap small bundle with annular fins are performed with computational fluid mechanics software,and the pressure drop and heat transfer coefficient with the velocity ranged from 0.076 to 0.334 m/s are obtained and compared with the existing experimental data for sphere bed.Based on comparison results,geometrical parameter for the equivalent models is optimized to get the most optimized equivalent model.The study result shows that the most optimized model is equivalent to sphere bed on characteristics of flow and heat transfer,and proves the feasibility to study the thermal-hydraulics problems of spherical fuel assembly water cooled reactor by the equivalent model methodology.

Numerical Simulation of Water Hammer Characteristics for Integral Reactor Primary Circuit

ZUO Qiaolin, QIU Suizheng, LU Wei, SU Guanghui, TIAN Wenxi, WANG Sheng, XIAO Zejun

2012, 33(4): 85-90.

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Based on the method of characteristic line,a set of closed mathematic expressions with reasonable boundary conditions of the nuclear reactor primary circuit has been established.WAHAP program has been developed,and numerical calculation is carried out on the check valve-induced water hammer phenomena in a four pumps paralleled feedwater system during the alternating startup process of parallel pumps.In this paper,we found that the valve disc stroke the valve seat at two different periods and there were two kinds of pressure vibration existed.

Numerical Simulations on Steady State Behavior of Supercritical Water Natural Circulation in a Rectangular Loop

LU Fa, HUANG Yanping, YAN Xiao, ZENG Xiaokang

2012, 33(4): 91-95.

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Numerical simulations on steady state behavior of supercritical water natural circulation in a rectangular loop were carried out.Different expressions of gravitational pressure drop were evaluated.The calculation results indicated that with the heat load variation,the mass flow rate of the steady-state natural circulation obtains a maximum value when the heater outlet temperature is near the pseudo-critical temperature,and the maximum value decreases with an increase in heater inlet temperature.The density difference and velocity difference between the heater inlet and outlet increase when the temperature of the whole heater moves towards the pseudo-critical area,while the pressure drop decreases in this condition.The evaluation of the gravitational pressure drop expressions indicated that when the expression proposed by Ornatskiy and Razumovskiy was used to calculate the gravitational pressure drop in a relatively long tube（2 meters）,the results were lower than those obtained from integral method which could be considered to be the real gravitational pressure drop,and the maximum error was near 30 percent.On the basis of Simpson integral method,a new method was proposed to calculate the gravitational pressure drop.The new method showed more satisfactory and could be adopted in experimental data analysis.

Experimental Study on Two-Phase Flow and Heat Transfer through Narrow Rectangular Channel under Middle-Low Pressure Condition

WANG Tao, WANG Jun, WANG Xiaojun

2012, 33(4): 96-101.

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Two-phase boiling experiments have been performed in narrow rectangular channel at middle-low pressure.Based on the measured pressure drop and temperature across the test section,the effect of pressure and mass flow rate has been analyzed.Saturated boiling heat transfer was studied.And a new correlation has been developed for saturated boiling heat transfer under middle-low pressure condition.Additionally,based on the predigested one-dimensioned model of two-phase boiling flow,a new correlation has been developed for two-phase flow pressure drop under middle-low pressure condition.Furthermore,F parameter has been defined to expresses proportion of moistening perimeter between steam-phase and wall in this new correlation.The calculation results of new correlation predict well the experimental results.The average saturated boiling heat transfer coefficient is great impacted by system pressure,mass flow and heat flux.

Modeling on Bubbly to Churn Flow Pattern Transition in Narrow Rectangular Channel

WANG Yanlin, CHEN Bingde, HUANG Yanping, WANG Junfeng

2012, 33(4): 102-105,110.

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A theoretical model based on some reasonable concepts was developed to predict the bubbly flow to churn flow pattern transition in vertical narrow rectangular channel under flow boiling condition.The maximum size of ideal bubble in narrow rectangular channel was calculated based on previous literature.The thermal hydraulics boundary condition of bubbly to churn flow pattern transition was exported from Helmholtz and maximum size of ideal bubble.The theoretical model was validated by existent experimental data.

Analysis of Single Phase Forced Circulation Heat Transfer Characteristics in Rectangular Narrow Channel

WANG Chang, GAO Puzhen, XU Chao

2012, 33(4): 106-110.

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The heat transfer characteristics in forced circulation in narrow rectangular channels under the operation condition of constant coolant average temperature have been experimentally and theoretically studied.The result shows that the heat transfer temperature difference between wall and water increases linearly with the Reynolds number in the laminar flow region,however,when the flow comes into the turbulent flow region,the heat transfer temperature difference nearly keeps constant with the increase of Reynolds number.The above phenomena were explained through the analysis of heat transfer mechanism.In addition,the larger the temperature differences between inlet and outlet,the smaller the Nusselt number.The reason that the temperature difference has effects on heat transfer characteristics is that the proportion of thermal resistance in the viscosity sub layer has been changed.

Experimental Investigation on Characteristics of Void Fraction for Bubbly Flow in an Inclined Circular Tube

XING Dianchuan, SUN Licheng, YAN Changqi, WANG Xin, SUN Bo

2012, 33(4): 111-115.

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Local void fraction of air-water bubbly flow in an inclined circular tube（ID 50mm） was obtained by optical fiber probe method.The results show that the averaged void fraction decreases with the increasing of the inclined angle,whereas its trend is moderated when the inclined angle larger than 5 degree.The void fraction nearly does not change with radius except in the near wall region where it has a peak value and falls to the minimal value in vertical condition and the typical saddle pattern profile is observed.Under inclined condition,bubbles move to the upper wall and congregate,therefore,the wall peak in the upper near wall region is enhanced nevertheless is weakened in the lower near wall region,even disappears with a larger inclined angle.

Numerical Analysis of Quenching Meshes on Laminar Flame Propagation

CHEN Bin, YU Hongxing, ZHANG Yu

2012, 33(4): 116-119.

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The impact of quenching meshes on laminar flame propagation is analyzed by computational fluid dynamics methodology based on simplified 2-dimension circumstances.The result of numerical analysis demonstrates that the propagation of the laminar flame is arrested when the chemical heat of hydrogen combustion is less than heat absorption by metal meshes and the flame front is quenched.

Design of Integrated Emergency Receiver Equipment for Radioactive Scrap Resin

LU Jingbin, KONG Jingsong, GUO Weiqun

2012, 33(4): 120-122.

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Based on the source item features and considering the requirement of safety,light-weight and interface support in radioactive scrap resin emergency receiver for PWR coolant circuit,the receiving process flow is proposed.The integrated receiving equipment is designed,and its corresponding experiment method and requirement is brought forward.

Experimental Study on Treatment of Simulated Boron Containing Radioactive Wastewater by Reverse Osmosis Membrane

WANG Xiaowei, YANG Kai, KONG Jingsong

2012, 33(4): 123-126.

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Experiments were carried out on a pilot equipment to study the performance of treatment of simulated boron containing radioactive wastewater by reverse osmosis membrane.Results showed that the rejection efficiency of boron,simulated nuclides and salts in simulated wastewater could be enhanced by increasing the feedwater pH,improving the applied pressure or reducing the recovery ratio.By adjusting the simulated wastewater pH to 9.0,the boron concentration in permeate could be lower than 5mg/L and two nuclides including cobalt and cesium could be removed to a satisfied extend.

Study on Treatment of Radioactive Wastewater Containing Activated Products by Micellar-Enhanced Ultrafiltration Process

KONG Jingsong, WANG Xiaowei, YAO Qingxu

2012, 33(4): 127-130.

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Experiments of removing nickel from simulated radioactive wastewater containing activated products by micellar-enhanced ultrafiltration which using odecylbenzene sulfonate（SDBS） and polyoxyethylene octyl phenyl ether（TritonX-100） as surfactants were carried out.Results showed that rejection efficiency of nickel was affected by the mole ratio of TritonX-100/SDBS,surfactants concentration and pH.Nickel rejection ratio could reach 95% when the mole ratio of TritonX-100/SDBS was 0.1 and feedwater pH was adjusted above 9.0.

Lessons from Fukushima Dai-ichi NPP Accident and Effect on NPP Development in China

PAN Jinzhao

2012, 33(4): 131-134.

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Although the Fukushima NPP accident caused no serious environmental problem in China,but it does have great effect on the future NPP development in China.The NPP development plan in China is so uniquely huge in the world energy field that great challenge has already arisen to the Chinese NPP constructors and operators.This paper analyzes the sequence of the Fukushima NPP,which can be referred to in the improvement of the NPP development plan and the NPP accident emergency plan.

Study on Natural Deposition of Fission Product Aerosol in Severe Accidents

HUANG Gaofeng, CAO Xuewu, TONG Lili

2012, 33(4): 135-138.

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Aerosol natural deposition model of gravitational sedimentation,diffusionphoresis,inertial impaction and thermophoresis are established based on integrated safety analysis model for 600 MW pressurized water reactor.Typical severe accidents are chosen,and natural deposition phenomenon of fission product aerosol is analyzed.Additionally,gravitational sedimentation model of MELOCR is coupled into integrated safety analysis model,and fraction of gravitational sedimentation is compared.The results show that gravitational sedimentation is the most important deposition mechanism,and deposition effect of gravitational sedimentation model in this paper is stronger than MELCOR.

Effect of Volume Ratio between Uranium and Water on Fusion-Fission Hybrid for Energy Production

SHI Xueming, PENG Xianjue

2012, 33(4): 139-142.

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The theoretic model of MCORGS code,which is used in this paper to evaluate the effect of the volume ratio between uranium and water（VR） on a fusion-fission hybrid for energy production,is introduced.It is showed that at natural uranium case,the VR of 2:1 is preferred,since it can reach a comprise between energy multiplication and tritium breeding ratio,and the neutronics performance of the system can maintain more than 100 years.At the case of spent fuel from PWR,there are more room for selection of VR.Both energy multiplication and tritium breeding ratio will improve,however the fissile material breeding capacity will degrade.

2012 Vol. 33, No. 4