核动力工程

Research and Implementation of Stretch-Out Operation in Daya Bay Nuclear Power Station

XIAO Min, ZHU Min-hong, LI Xian-feng, XU Shu-wen, TANG Xiao-qing

2006, 27(1): 1-4,50.

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Abstract:
Stretch-out operation mode can deepen the reactor burnup when the boron concentration is near 0 mg/L,in which the additional reactivity is introduced by the reducing of the moderator temperature and the decreasing of the load.Stretch-out is used in many nuclear power plants all over the world.The first stretch-out operation has been used for the first time in China..As a specific operation mode,which outruns the original reactor core design,the related and specialized design argument and safety analysis is required.As a consequence of the continuous or stepwise reduction of load and moderator temperature,the neurotic measurement system and the reactor control and protection system parameters should be modified specially.
Based on the schedule of the electricity production,the first stretch-out operation had been carried out from March 12 to March 21 2003.It successfully avoided the overlapping between 209 and 109 inspection shutdowns in Daya Bay Nuclear power station,and additional 30 GWH energies were produced.

Comparative Study of WIMS-AECL and MCNP ＆ MCBurn

MAN Xiao-yu, WANG Kan, YU Gang-lin

2006, 27(1): 5-8,79.

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This work,using WIMS-AECL,MCNP-4B and MCBurn codes,has calculated the criticality and burnup problems of a series of benchmarks and advanced CANDU lattices,including all-uranium and thorium-based fuels.WIMS-AECL uses ENDF/B-V and ENDF/B-VI nuclear data library separately.Comparison manifests that for benchmark,both B-V and B-VI can lead to good results and B-V is better;for all-uranium fuel of ACR and thorium-based fuel of ACR,WIMS-AECL with B-V is a better choice.

Three-Dimensional Numerical Simulation of Temperature Field and Helium Flow Field in ITER China TBM

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

2006, 27(1): 9-13.

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Three dimensional temperature field and helium flow field of TBM are simulated using the general purpose computational fluid dynamics（CFD） code FLUENT.The temperature distribution of Be Armor,Be Pebble Bed,Li₄SiO₄ Pebble Bed,Structure Material of TBM,and helium flow field in the cooling pipe are presented.The research indicates that the work temperature of each material is under the material temperature allowed except some places where high temperature should be excluded in the design.The re-sults will provide references for further optimized thermal hydraulic design of ITER China TBM.

Numerical Analysis of Heat-Transfer Coefficient of Different Diameter Thermo-Tube and Disposal Mode

DONG Qi-wu, LI Jing, LIU Min-shan, LIU Qian

2006, 27(1): 14-17.

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For the shell-tube heat exchanger a novel tube arrangement of diameter reducing is presented in the paper.The effect of flow reduction on heat transfer in the inner side of the tubes is studied by CFD.The results show that flow flux disperses in tubes with different diameters and the difference is related to the length of the tube.The heat transfer deterioration in a thin tube is not obviously as that in a big one with a certain velocity range.

Experimental Research on Subcooled Boiling Flow Instability of Natural Circulation

TAN Si-chao, PANG Feng-ge, GAO Pu-zhen

2006, 27(1): 18-21,93.

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Experimental research was conducted in the view of subcooled boiling high frequency oscillations.The results showed that the high frequency oscillations were acoustic oscillations.Through analysis,three types and the boundary of the acoustic oscillations were determined.The influences of system parameters on the onset of the acoustic oscillations were investigated.The analysis showed that the onset of FDB had direct relationship with the onset and disappearance of acoustic oscillations.Based on this,an empirical correlation of the onset of the acoustic oscillations was obtained.

Simulation Device for Two-Phase Flow in Steam Generator

LI Qiang, SHEN Ming-qi, SONG Wei

2006, 27(1): 22-24.

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In this paper,a simulation device,which can be used to simulate the thermodynamic pa-rameters and flow parameters in steam generator of nuclear power station（mainly for Qinshan Nuclear Power Station）,is introduced;computer can control its parameters.By utilizing this device,we can get a simulation environment of steam generator for the dynamic test of the SG corrosion inhibitor and corrosion test of the metal materials.

Experimental Study on Neutron Noise Induced by Core Barrel Vibration of Nuclear Reactor

LIU Cai-xue, WEI Dong, FANG Cheng-chun, HE Shao-qun, HU Jing-hua

2006, 27(1): 25-29.

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This paper deals with the experiment of neutron noise induced by the core barrel vibration of the critical installation.Three kinds of stimulation including random signal,narrow band sweep frequency signal and single frequency signal are adopted to stimulate the core barrel vibration.At biggest stimulation force,the biggest vibration amplitude is 21μm,generally in 3 to 10μm.The amplitude of the frequency spectrum of background noise presents an exponential reduction with the frequency,at 10Hz frequency,and the amplitude reduces 10 times or so.To this reactor,the frequency spectrum of the neutron noise induced by the core barrel vibration distributes mainly over tens Hz,and may be identified easily.A high-pass digital filter is adopted to improve the the observation and identification of the characteristic spectrum.The characteristic spectrum of neutron noise in the lower and middle frequency range is brought on by the core barrel vibration stimulated by random force.As stimulated by the narrow band sweep frequency force,the of frequency peak at about 125Hz is brought on by the core barrel vibration stimulated by sweep frequency range from 110 to 180Hz.The characteristic spectrum including several frequency peaks is brought on by the core barrel vibration stimulated by single frequency range from 30 to 150Hz.The mode frequencies of the core barrel vibration in normal temperature water are 76.2,125.0,181,215 and 239.7Hz。

Application Study on Core Barrel Vibration Monitoring Using Neutron Noise

LIU Cai-xue, WEI Dong, FANG Cheng-chun, HE Shao-qun

2006, 27(1): 30-33.

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The Power Spectrum Density（PSD） of neutron noise induced by the Core barrel vibration of has been carried out.This work verifies that the vibration frequencies and amplitudes can be obtained by analyzing the neutron noise,and indicates that the validity of vibration monitoring to internals using neutron noise.critical installation is analyzed,and the analysis demonstrates the possibility of obtaining the vibration char-acteristic of core barrel from PSD of neutron noise.Calculation methods for the scale factor between PSD of neutron detectors and the vibration amplitude of core barrel are introduced and the corresponding calculation

Cracking Induced by Oxidation-Hydriding in Welding Joints of Zircaloy-4 Plates

ZHOU Bang-xin, YAO Mei-yi, MIAO Zhi, LI Qiang, LIU Wen-qing

2006, 27(1): 34-36.

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Abstract:
The welding joints of three plates of Zircaloy-4 obtained by hot rolling diffusion welding at 800℃ in vacuum were cracked during autoclave tests at 400℃ superheated steam after exposure longer than 150 days.The section of specimens was examined by optical microscopy and the composition of mi-cro-area at the welding seams and the matrix was analyzed and compared.The result shows that the con-tamination of carbon at the welding surface and the combination of oxidation and hydriding induced cracking is responsible for the cracking of welding seams during autoclave testing.

Modeling of Accident Scenario Caused by Total Blockage of Single Subassembly in a Fast Reactor and its Validation

SHI Xiao-bo, LUO Rui, WANG Zhou

2006, 27(1): 37-42.

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In order to predict the accident scenario caused by total inlet blockage of single subassembly at full power in a fast reactor,a calculational model was developed according to SCARABEE-N subassembly melting and propagation tests.In the model,sodium thermal-hydraulics was described by a two-fluid sub-channel model.The motion of melting cladding and melting fuel was dealt with very simple approach similar to that of SURFASS code.The behavior of UO₂-steel mixed boiling pool was determined by a one dimen-sional semi-empirical model,in which the distribution of void fraction was calculated by drift flux model,heat transfer coefficient between the boiling pool and the wall was determined by modified Greene correla-tion,the temperature field and thickness of solidified UO₂ crust were obtained by solving enthalpy model.Calculation for SCARABEE BE+1 experiment was performed to validate the model,the results approxi-mately agree with the experimental observation.

Research on Identification Method Based on Neural Network for Nuclear Steam Generator Water Level

PENG Wei, ZHANG Da-fa, ZHOU Gang

2006, 27(1): 43-46.

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Abstract:
The false water level,which caused by the change of the number of steam bubbles which follows the change of local pressure due to the change of steam flow rate,make the nuclear steam generator water level difficult to identify.In order to improve the effect of identification,the identification method based on neural networks for nuclear steam generator water level is researched in this paper.The series-parallel model is applied to the identification and the back propagation algorithm of Levenberg— Marququardt type（LMBP） is employed to train the network.The proposed method with a good performance of identification is demonstrated by simulation results.

Calculation of Effective Doses for External Photons to Human Body with Monte Carlo Method

WANG Feng-chun, YANG Zhong-qin, TIAN Xin-shan

2006, 27(1): 47-50.

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This paper describes the calculation of effective doses for external photons to human body with Monte Carlo method（MCNP4C code）.In order to compare with ICRP No.74 Publication,the calcula-tion has considered 20 incident energies between 0.02MeV and 10MeV.This calculation has also considered two incident photon directions: antero-posterior（AP） and postero-anterior（PA）.Comparison of the calculated results show that the results calculated by MCNP4C considerably coincide that from ICRP No.74 Publica-tion.

Calculation of Neutron Flux in SPRR-300 Reactor Irradiation Channel with MCNP Code

DOU Hai-feng, DAI Jun-long

2006, 27(1): 51-54.

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The issue predigests geometric structure of SPRR-300 reactor core with repeated structure geometry feature of MCNP code and sets up the irradiation channel mathematical model to calculate the neutron flux.The calculated results accorded well with the experimental data shows that the mathematical model in this issue is reasonable and feasible.

Study of Maximal Load Capacity of Moveable Coil Electromagnetic Drive for Reactor Control Rod

MA Cang, BO Han-liang, JIANG Sheng-yao, ZHANG Hong-chao, SUN Chang-long

2006, 27(1): 55-57.

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The study of the maximal load capacity of the moveable coil electromagnetic drive for reactor control rod has been conducted at Institute of Nuclear and New Energy,Tsinghua University.The experiment finds that the load value of the control rod drive mechanism（CRDM） is linearly increasing following the current value of the coil,and the maximal load is 2276 N when the current value is 5A.The maximal load of CRDM declines clearly at running condition,and the decline scope is about 12%.The CRDM carrying capability is linearly increasing as the coil current,and relates with the running speed of the coil.

Design of Passive Heat-Switch between Pressure Tube and Calandria Tube Based on the Principle of Metal Expand on Heating and Contract on Cooling of TACR

XU Liang-wang, JIA Bao-shan, YU Ji-yang

2006, 27(1): 58-61,65.

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The paper presents a design of passive heat-switch between pressure tube and calandria tube of Thorium-based Advanced CANDU Reactor（TACR）.The heat switch is designed on the principle of heated metal expanding and contract on cooling.The valve driven by a metal stick controls the type of heat conductor between pressure tube and calandria tube.The design meets the requirement of high heat transfer resistance in normal operation and low heat transfer resistance to remove the residual heat in accident.The design is reliable and effective.

Design of Fast Neutron Experiment Device at Pulsed Reactor

JIANG Xin-biao, LIU Shu-huan, ZHONG Yun-hong, ZHANG Wen-shou, CHEN Wei, WANG Wei

2006, 27(1): 62-65.

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In this paper,the fast neutron radiation experiment device,which has been developed and satisfies the studying on radiation effects of electronics apparatus,is designed by Monte Carlo and discrete ordinates method.The parameters of the fast neutron radiation devices are measured and analyzed by using the SAND-II multiple-foil activation method and the thermo luminescence dosimetry foils.The designed parameters are agreeable to the experiments.

Principle and Opening-Closing Character Analysis of DC Check Valve

HAN Xu, ZHOU Yu

2006, 27(1): 66-69.

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Abstract:
when the behaviour of main pump in PWR power plant change,as a result of arresting fluid countercurrent current by check valve,water hammer phenomena met occur more or less in loop.Serious water hammer not only met brought incident of over pressure,imperil pressure boundary,but also may en-gender check valve lapse.DC check valve is a kind of new theory check valve,is designed to avoid serious water hammer phenomena at tradition check valve closing.the analyses and experiment indicate that DC check valve can availably solve water hammer problem in the loop,and be able to reliably prevent fluid countercurrent.

Measurement for Moisture Separator Efficiency

HUANG Yun-feng, ZHANG Zhen-zhong, JIANG Feng, YE Sui-sheng

2006, 27(1): 70-72,89.

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The efficiency measurement is very important in the development of moisture separator.In this paper one moisture separator efficiency measurement system has been set up on the basis of the re-searches on both the foreign MSA type moisture separator and the innovative domestic one.The weight fil-tration efficiency is obtained by means of direct water weight measurement.The count filtration efficiency is measured by two Cassella Impactors which are used to sample the droplets in the downstream and upstream of the separator and one microscope which is used to observe the number of the droplets.The experiment result shows this efficiency measurement system is accurate and dependable.

Experimental Study on the Scram of Electromagnetic Movable Coil Control Rod Drive Mechanism

SUN Chang-long, BO Han-liang, JIANG Sheng-yao, ZHANG Hong-chao, MA Cang, WANG Jin-hua, QIN Ben-ke

2006, 27(1): 73-75.

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Abstract:
Electromagnetic movable coil control rod drive mechanism is a new type drive mechanism.The drive mechanism is experimentally studied to gain the characteristic of scram time.Further more,the reason of the different scram phenomena is analyzed and the disciplinarian of scram is also summarized.On the base of series experiments it can be concluded that scram time of AC break is longer than that of DC break and the residual current of coil’s can distinctly influence the scram time.The scram time of AC break is 300~700 ms longer than that of DC break.

Numerical Simulation for Two-Phase Flow in First-Stage Moisture Separator of Steam Generator

HUANG Wei, CHEN Wu-xing, ZHANG Wen-qi, WANG Hai-song, HE Jing-song

2006, 27(1): 76-79.

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This paper presents the numerical simulation for the two-phase flow in the moisture separator by using CFD method.Unstructured grid and multi-block grid are used for the computational region.The detail flow characteristics are shown.Outlet steam quality is compared with the calculation result by using CFD method and special computer code for the steam generator hydraulic calculation,and a good agreement is obtained.

Research on the Simulation Method of Control and Protection System in Nuclear Power Station

HU Rui, YANG Yan-hua, LIN Meng, ZHANG Rong-hua

2006, 27(1): 80-84.

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Simulating of control,protection and auxiliary system in nuclear power plant（NPP） is necessary for improving precision of the NPP simulation.This paper analyzes the models of those systems in DAYABAY NPP.Based on the DAYABAY NPP model,we realize the simulation code through object-oriented programming concepts.The results of transient test events show that the simulation successfully fulfills control and protection function of primary and secondary loop in NPP.The code is more general,portable and low-cost.

Experimental Research on Steam-Turbine Rotor Vibration Fault Diagnosis Based on the Fractal Box Counting Dimension

FAN Fu-mei, LIANG Ping, WU Geng-shen

2006, 27(1): 85-89.

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The rotor is the main component during turbine fault forecasting and diagnosing.Vibration signal and state parameter are usually selected as the monitored signals to equipment diagnosis.The steam-turbine vibrating is experimentally simulated on the Bently facility.Using box counting dimension method the different type of turbine fault with fractal characteristic is numerically researched.The result shows that the axis track under mass unbalance is smooth and its box counting dimension reaches the small-est.The axis track under friction presents disorder and its box counting dimension is the largest.The box counting dimension of un-countershaft and loosing are in the middle.As a result,the box counting dimension possesses better distinguish during steam-turbine rotor fault type diagnosing.

Quality Assurance Grading of Conventional Equipments at Nuclear Power Station

LI Ping, LI Shi-chang

2006, 27(1): 90-93.

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Equipment QA grading with the systematic and standardized approach will benefit the concerned organizations by effective allocating of limited resources to guarantee the quality of essential equipments.This paper presents a new quality assurance grading system for the convention systems/equipments of nuclear power station,which is operative and at the same time could help the owner to allot resource reasonably through the analysis of the purpose of grading and the experience and lessons of LINGAO Phase I project.

Study of Cesium Recovery with Zirconyl Phosphate-Amonium Molybdophosphate Complex Ion Exchanger

DENG Qi-min, LI Mao-liang, CHENG Zuo-yong

2006, 27(1): 94-96.

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The paper describes the synthesise of Zirconyl Phosphate-Amonium Molybdophosphate（ZrP-AMP） complex ion exchanger by soak method,and the absorption and desorption of cesium with ZrP-AMP also be investigated.The results indicate that ZrP-AMP granules can be synthesized with the in-creasing of hydrochloride acid in reactant,and these granules can be used for column operation.Exchanger capacity of ZrP-AMP for cesium is 59mg/g,and recovery rate is 80%.

2006 Vol. 27, No. 1