核动力工程

Method to Study Water Hammer with Fluid-Structure Interaction in Spatial Pipe

XI Zhi-de, MA Jian-zhong, SUN Lei

2013, 34(2): 1-4,8.

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Abstract:
The theory of coupling 4-function models and its solution approach are first introduced in this paper,and the method of CFD to calculate fluid-structure interaction is also introduced.Finally,the model in related reference is applied with this method to simulate the process of water hammer.By CFD calculation for the classical water hammer,the numerical scheme and grid are selected,and the results of CFD are compared with reference.The results show that the method in this paper can be used in more complex pipe system to simulate the water hammer effect.

Experimental Investigation of Hot Block Rewetting Process during Nuclear Reactor Emergency Cooling

LIU Bin, CHEN Deqi, PAN Liang-ming

2013, 34(2): 5-8.

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An experimental investigation was carried out to simulate the process of emergency cooling rewetting of the hot block such as molten material and pressure vessel wall in nuclear reactor core under the serious accident.According to the present experimental study,the liquid spatters can pre-cool the hot block;the heat transfer in the y direction has been enhanced with the increasing liquid level and the central temperature cooling speed is very high.The experiments reveal the nonlinear relationship between the center temperature dropping speed and the liquid level increasing speed,and it shows an U-shape trend which suggests that a minimum center temperature dropping speed exists during rewetting with a certain liquid level increasing speed which should be avoid.With higher initial temperature,the temperature dropping speed is affected by the initial temperature mildly.

Experimental Study on Flow Characteristics and Parameters Influences of Geysering Phenomena in Heating System

QI Zhan-fei, CHEN Jin-bo, TONG Li-li, CAO Xue-wu

2013, 34(2): 9-15.

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The geysering phenomena is studied experimentally under several cases with different height of upper plenum,diameter of vertical column,initial subcooling,coolant inventory and heating power to investigate their influences on the geysering phenomenon.According to the analysis of experimental phenomenon and data,it is found that the liquid periodically boils and various flow patterns alternately occur in the heating system.The refilling of cooler liquid causes significant change of temperature and pressure in the system.The geysering phenomena can be identified by the ratio of pressure difference and temperature difference of vertical column and is more obvious under greater height of upper plenum,less diameter of vertical column and higher heating power.

Model Development on Cooling of Debris Formed of Molten Material in Lower Head

YU Hong-xing, LU Qing, HE Xiao-qiang, SU Guang-hui

2013, 34(2): 16-19,39.

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Based on the experimental phenomena,an improved lumped parameters mathematic model has been set up and programmed for the analysis of the cooling mechanism of the debris formed of moltenmaterial,which dumped into the lower head of the vessel during a severe accident.The proposed mode has been verified by simulating the results of ALPHA tests.Results indicated that the proposed model can well predict the cooling mechanism of the debris formed of molten material,which located in the lower head of the vessel during a severe accident.

Real Time Thermal Hydraulic Model for High Temperature Gas-cooled Reactor Core

SUI Zhe, ZHANG Rui-peng, SUN Jun, MA Yuan-le

2013, 34(2): 20-24.

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A real-time thermal hydraulic model of the reactor core was described and integrated into the simulation system for the high temperature gas-cooled pebble bed reactor nuclear power plant,which was developed in the vPower platform,a new simulation environment for nuclear and fossil power plants.In the thermal hydraulic model,the helium flow paths were established by the flow network tools in order to obtain the flow rates and pressure distributions.Meanwhile,the heat structures,representing all the solid heat transfer elements in the pebble bed,graphite reflectors and carbon bricks,were connected by the heat transfer network in order to solve the temperature distributions in the reactor core.The flow network and heat transfer network were coupled and calculated in real time.Two steady states(100% and 50% full power) and two transients(inlet temperature step and flow step) were tested that the quantitative comparisons of the steady results with design data and qualitative analysis of the transients showed the good applicability of the present thermal hydraulic model.

Scaling Effect on Flow Characteristics of a Structured Packed Bed Using CFD Method

LI Jian, SONG Xiao-ming, LU Jian-chao, LI Zhong-chun

2013, 34(2): 25-29.

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Numerical simulations were taken on single phase flow in a structured packed bed on different scaling using CFD modeling.The point contact between adjacent spheres was simplified with a new grid-generation method,and over 11 layers of spheres were calculated,141 particles in all,which represents flow characteristics of a large scale packed bed in the real system.The scaling effect on the flow characteristics and pressure drop was investigated,and the resistance formulas on different scales were obtained.The mechanism of boundary effect on flow characteristics of a small channel was studied and the scope of the inlet/outlet effect on resistance characteristics is evaluated by the pressure drop on axial direction.

Numerical Simulation of Steam Migration in Containment of Large-Sized Advanced PWRs

ZHANG Shi, WANG De-zhong, MA Yuan-wei

2013, 34(2): 30-33.

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The migration of steam in the containment after the break of primary circulation pipelines of large-sized advanced PWR is studied by CFD method with Uchida correlations to simulate steam condensation analyzing the steam distribution of different coolant leakage rates.The result suggests that the distribution trend of steam is basically the same at various coolant leakage rates,and the steam concentration near the containment wall increases fluctuant for which the higher position has the smaller fluctuation.

CFD Analysis of Effect of Helix Rib on Flow and Heat Transfer of Supercritical Water in Square Annular Channel

ZHU Hai-yan, YAN Xiao, CENG Xiao-kang, LI Yong-liang, HUANG Yan-ping, XIAO Ze-jun

2013, 34(2): 34-39.

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This paper presents a primary numerical simulation on the heat transfer behavior of supercritical water in smooth square annular channel and square annular channel with helix rib.The simulation results indicated that the heat transfer behavior of supercritical in smooth square channel is greatly non-uniformity in circumference.The helix rib can enhance the heat transfer in the square annular channel and reduce the circumferential non-uniformity of heat transfer.In addition,the helix rib will increase the flow resistance in the square annular channel,which will add more works to the main pump.So,there is need to consider the heat transfer,flow mixed and flow resistance comprehensively in the design of the helix rib of SCWR.

Thermal-Hydraulic Analysis and Design Improvement for Coolant Channel of ITER Shield Block

ZHAO Ling, LI Hua-qi, ZHENG Jian-tao, YI Jing-wei, KANG Wei-shan, CHEN Ji-ming

2013, 34(2): 40-43.

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As an important part for ITER,shield block is used to shield the neutron heat.The structure design of shield block,especially the inner coolant channel design will influence its cooling effect and safety significantly.In this study,the thermal-hydraulic analysis for shield block has been performed by the computational fluid dynamics software,some optimization suggestions have been proposed and thermal-hydraulic characteristics of the improved model has been analyzed again.The analysis results for improved model show that pressure drop through flow path near the inlet and outlet region of the shield block has been reduced,and the total pressure drop in cooling path has been reduced too;the uniformity of the mass flowrate distribution and the velocity distribution have been improved in main cooling branches;the local highest temperature of solid domain reduced considerably,which could avoid thermal stress becoming too large because of coolant effect unevenly.

Fully Developed Laminar Flow in a Rolling Narrow Retangular Duct

MA Jian, HUANG Yan-ping, LIU Xiao-zhong, LI Long-jian

2013, 34(2): 44-50.

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To obtain the fully developed laminar flow law in a rolling narrow rectangular duct,the forces on fluid particle within a moving coordinates frame was analyzed firstly,and the momentum equation under rolling condition was solved to get fully developed laminar velocity distribution and friction factor.Then,the isothermal laminar flow experiments with 900 ≤ Re ≤ 2600 under rolling condition with amplitude of ±15° and period of 8s were conducted.The research results show that the most difference between rolling and stationary conditions is that the pressure gradient varies periodically with each body force,thus hydro-dynamical structural relationship readjusts while the shear stress keeps constant;the pressure drop wave induced by each body force will results in the total pressure drop wave while the friction pressure drop and mass flow rate keep constant;the laminar friction factor under rolling condition also keep constant,and the theoretical prediction deviates from-1.1% to +4.3% in comparison with the experimental data,therefore both results are with a good agreement.

Effect of Rolling Motion on Forced Circular Boiling Flow Characteristics

WANG Chang, GAO Pu-zhen, TAN Si-chao, HUANG Yan-ping

2013, 34(2): 51-55.

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The boiling flow characteristics under rolling motion condition is experimentally studied.The results show that the variation of spatial location under rolling motion condition leads to the clogging of bubble and induces the mass flow rate fluctuate periodical.The marginal stability boundary(MSB) is affected by the flow flux,rolling amplitude,rolling period,inlet subcooling and inlet pressure.The effect of rolling amplitude and rolling period on the MSB decreases with the increasing of flow flux.In addition,increasing the mass flow rate,inlet subcooling or inlet pressure tends to stabilize the boiling system.

Development of Digital Centralized Data Acquisition System for Reactors

DAI Hang-yang, DENG Sheng, CUI Can

2013, 34(2): 56-59.

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By analyzing the reactor operation data and experiment requirements,the digital centralized data acquisition system based on PCI bus is designed.The system synthetically makes the use of bus technology,electrical isolation,virtual instrument,Labwindows/CVI platform and ACCESS database,and achieves automatic data acquisition,failure alarm,data analysis,remote monitoring of radiation dose,and so on;Software design based on multi-layer architecture implements the interface display of operational status and the fault diagnosis,and improves the data acquisition for more visual and reliability.It is proved that the system is stable,reliable and secure.On the other hand,the practical operation is simple and easy to maintain.Furthermore,the digital centralized data acquisition system has enhanced the digitization and automation of data acquisition of the reactor,and meets the design requirements.

Theoretical Calculation and Analysis of Emergency Ventilation to Reduce Nuclear Accident Consequence

LIN Xiao-ling, YANG Yong-xin

2013, 34(2): 60-61,93.

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The computational model of radionuclide activities in reactor compartment was set up for the nuclear accident of nuclear ships.The effect on the reducing of radioactive concentration by emergency ventilator was calculated quantitatively.It showed that the emergency Ventilation had a significant effect on the reducing of the radiological consequence,the reduction effect depends on the ratio of ventilation flux to the net volume of the reactor compartment,and the greater the ratio,the faster the reduction.Within half an hour after the emergency exhaust operated,the activity in the reactor compartment can be reduced by 90%.

Application of Wavelet Analysis inSignal Singularity Detection

CHEN Zhi-hui, XIA Hong, WU Zhi-sheng, DENG Li-ping, HUANG Wei, PENG Min-jun, HUANG Hua

2013, 34(2): 62-67.

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The wavelet transformation can be used to determine the location of mutation of signals.However,further research discovered that the single sub-band algorithm improved from wavelet transformation could recognize the singular signal of the first type effectively,but it could not recognize the singular signal of second type.The reason of resulting this problem is that the improved algorithm did not transfer the signal smoothly.This paper proposed the improved single sub-band reconstruction algorithm,and resolved the issue of recognizing the transient signal of the second type by introducing transition functions.The effectiveness of this method has been proved by related experiments.

Effects of Loading Reactivity at Dynamic State on Wave of Neutrons in Burst Reactor

GAO Hui, LIU Xiao-bo, FAN Xiao-qiang

2013, 34(2): 68-71.

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Based on the point reactor model,the program for simulating the burst of reactors,including delay neutron,thermal feedback and reactivity of rod,was developed.The program proves to be suitable to burst reactor by experimental data.The program can describe the process of neutron-intensity change in burst reactors.With the program,the parameters of burst(wave of burst,power of peak and reactivity of reactor) under the condition of dynamic reactivity can be calculated.The calculated result demonstrates that the later the burst is initiated,the greater its power of peak and yield are and that the maximum yield coordinates with the yield under static state.

Optimization of Data Processing for ORIGEN-S Calculations

XIN Feng, LIU Yuan-yuan, ZHENG Peng, ZHANG Chun-ming

2013, 34(2): 72-74.

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As an upgraded version of ORIGEN-2,ORIGEN-S has been widely utilized in computing the compositions and radioactivity in reactors.ORIGEN-S has a high computational efficiency,but data processing of the results is rather cumbersome,as extraction and integration of useful data consume a lot of manual operations,which costs much time.To solve this problem,a data processing optimization program named OSDO based on MATLAB platform is developed.The results demonstrate that,under the same circumstance,data processing time of OSDO is greatly reduced to 1/70 compared with the conventional method.Besides,this study can be easily extended to other cases where the output files are in txt format.

Study on ⁶LiD Convertor as 14 MeV Neutron Source in HFETR

SUN Shouhua, PENG Feng, XU Taozhong

2013, 34(2): 75-78.

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A brief theoretical principle of the production of 14 MeV neutrons from thermal neutrons has been given.The ⁶LiD convertors rely on the absorption of a thermal neutron resulting in a triton which interacts with deuterium producing a 14 MeV neutron.The calculation and analysis model of 14 MeV neutron flux in the convertor have considered the neutron flux disturbance of convertor to irradiation channel and the leakage of T in ⁶LiD converter.The calculation and analysis model of effective yield of 14 MeV neutron,which generated from thermal neutron,has been established.The results show that the optimal thickness of ⁶LiD is 0.85 mm,and effective yield(Y) of thermal neutron converting 14 MeV neutron is 3.18×10^-4.

Analysis of Effects on Radioactive Fission Product Release from Fuel Pellet to Coolant

JING Fu-ting, CHEN Bing-de, YANG Hong-run, LU Huan-wen

2013, 34(2): 79-82.

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PROFIP5 was used to analyze the effects of linear heat rate,decay constant and defect size on fission product fraction release.Conclusions about fission product release were drawn:at temperature below 1000℃,the release fractions are independent of temperature;at temperature above 1000℃,the release fractions increase with the temperature;For the same chemical species,the release fractions decrease as the decay constants increase,and this phenomenon is more obvious as the linear heat rate increases.

High Temperature Performance of Cobalt-Free Nickel-Based Cladder on Nuclear Valve Sealing Surface

YU Jia-li, FU Ge-yan, LIU Shuang, WU Yun-xia

2013, 34(2): 83-85.

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For the purpose of the high temperature wear resistance and corrosion resistance close to or over the cobalt-based alloy,one kind of cobalt-free nickel-based alloy Ni-3 is designed.And in the temperature about 350℃,high temperature friction and wear and high temperature corrosion experiments of Ni-3,Norem02 and Stellite06 are tested.Analysis showed that the Ni-3 is an ideal cobalt-free alloy.It is more excellent than Norem02 in the high temperature wear resistance and corrosion resistance;compared with the cobalt-based alloys,the high temperature wear resistance of Ni-3 is close to them,but the high temperature corrosion resistance of Ni-3 is slightly less than them.Therefore,within a certain rang,the Ni-3 can replace the surfacing that is commonly used in the valve sealing surface of the nuclear valves,chemical and so on.

Study on Creep Damage Behaviors of Ni-Based Alloy C276

MAO Xue-ping, GUO Qi, ZHANG Sheng-yuan, HU Su-yang, LU Dao-gang, XU Hong

2013, 34(2): 86-89.

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High temperature creep tests were carried out for Ni-based alloy C276 at 650℃,700℃and 750℃,which is one of the candidate materials for the fuel cladding of the supercritical water reactor.Methods of damage mechanics were adopted to calculate and analyze these data.Damage factors calculated by Kachanov formula and Norton formula based on θ projection method were compared.The results show that the damage factors about the material are similar at the three temperatures according to Kachanov formula.The predicted creep curves calculated by θ projection method have a close agreement with the experimental data.The damages calculated by Norton formula start at about 0.3~0.4 lifetime,and the damage factors calculated by Kachanov formula are relatively conservative.

Numerical Investigation of Residual Heat Removal Heat Exchange Capacity in Pressurized Water Reactor Plants

QIU Jin-meng, LI Jun, WANG Xiao-jiang, WANG Zhi-gang

2013, 34(2): 90-93.

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Mathematical model is established for shell and tube heat exchanger of PWR residual heat removal system.Comparison with the calculated results from the design software HTRI for the heat exchanger is conducted,and the results show that outlet temperatures of shell and tube side are precisely predicted.The calculation of overall heat exchanger coefficient sensibility shows that this parameter varies with the flow rates of tube and shell side and it is not advisable to keep it as constant value in the simulation.The overall heat exchanger coefficient shall be coupled with the heat exchanger flow and heat exchanging conditions during calculation.The overall coefficient is extremely dependent on the flowrates which is less than 1000 m³/h while it seems to be independent with the higher flowrates.The model can be easily employed to predict other system conditions elsewhere and provide the convenient way for the plant test,design verification and high efficient operation of nuclear power plants.

Scaling Study for Experimental Test of Advanced Reactor Analysis of Passive Feature of Containment and Its Simulation Hierarchy

LI Shengqiang, LI Weihua, JIANG Shengyao

2013, 34(2): 94-98.

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Different containment safety features are analyzed.Combined with the accident process analysis and physical model classification,a containment system analysis hierarchy,which can be applied to advanced containment system simulation and experimental verification,is built.Different physical models,experimental methodology,scaling criteria,PIRT,mathematical models and test process logic are included.It attempts to provide theoretical support and design reference for future containment system characteristics verification experiments.It is also applicable for the system simulation,including complex coupling of physical phenomena and processes.

Dynamic Detection System for Deuterium Pellet Row Length

GUO Yong-cai, ZHOU Sen, GAO Chao, WANG Jun, ZHANG Jie

2013, 34(2): 99-103.

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In view of the characteristics of deuterium pellet rows,a dynamic pellet row measurement model is established,and the automatic detection system for deuterium pellet row is designed.Setting up measurement platform and using HMI for controlling the 18 rows of pellets loading,compaction,relaxation,transportation and other actions,and 2-D dynamic measure the length through laser triangulation between the ring end plane of moving pellet row and the datum plane by non-contract way.The system achieved pellet row length dynamic detection.It obtains the resolution over 10μm,the accuracy over 100μm,and the speed 18 rows/sec;moreover,it also has such function as data display in real time,system state detection,security alarm,and emergency shutdown.The actual operation results show that the system poses security,rapid,stable and reliable characteristics,and well meets the needs of pellet control and detection in industrial automation.

Design of Automatic Welding Device for Nuclear Vessel Head

KONG Jing-song, MENG Kai

2013, 34(2): 104-106.

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Nuclear materials container is strongly radioactive due to the radioactive content inside,and before the storage and transportation,the vessel head and body shall be tightly welded.To improve the welding quality and reduce the radiological dose that workers expose,an automatic welding device is developed based on MIG welding and OTC robot system.This device is safe and with high stability,and automatic welding on complex surface can be achieved with remote programming control.

Operation Research on Steam Generator Tube Rupture Accident in PWR NPPs

GUO Cheng

2013, 34(2): 107-110.

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This paper comprehensively analyzes PWR steam generator heat transfer tube rupture accident(SGTR),and summarizes the accident processing key strategies in the terms of detection means and event nuclear safety analysis.It analyzes the accident processing difficulty and key risk.Taking the America Indian point2 nuclear power plant SGTR as an example,the events de tailed process is described and the corresponding operation experience is given.

Analysis of Origin for Thermal Power Fluctuation in Lose of Coolant Accident Monitoring System

HU Ru-ping, LI Zhi-jun, ZHOU Xiao-ling, PENG Song

2013, 34(2): 111-113.

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Based on data investigation and theory analysis,the paper thoroughly analyzed the fluctuation of LSS thermal power.According to the analysis on the dynamic item and the static item,the fluctuation of dynamic item is the main contributor for the fluctuation of LSS thermal power,and the change of temperature with time is the main contributor for the fluctuation of dynamic item.

Effect of Steam Distribution Modes on Performance of Last Stage in Steam Turbine

GAO Yi-qiu, LI Yi-xing, ZHOU Zhen-dong

2013, 34(2): 114-117.

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Considering that the power level drops significantly and steam flow separates from the outlet of rotor blades in the last stage of steam turbine at low load conditions,the last stage in whole working conditions was calculated and the height of flow separation in different volume flow was obtained.Different conditions of turbine under the same flow path with nozzle governing and throttle governing were calculated respectively.The results show that the flow separation bubble and separation height increased dramatically when volume flow decreased;the throttle governing at low load conditions increased the steam rate of turbine and the enthalpy drop of pressure stage,while reduced the changes of power distribution under the influence of steam parameters changing;in the same load,the throttle governing increased the volume flow in last stage,and reduced the separation height of rotor blades accordingly;it was propitious to improve the flow of the last stage at low load conditions,and postpone the emergence of zero power conditions.

A General Model Based on Graph Theory for Quantitative Analysis of PWR Thermodynamic System

RAN Peng, WANG Ya-se

2013, 34(2): 118-122.

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Based on the analysis of the structure feature of PWR nuclear power plants,graph theory are introduced in the thermal economy analysis fields.According to the abstraction rule of the thermal system in PWR nuclear power plants,the boundary delimitation of a power plant thermal system is determined,and the thermal system of PWR nuclear power plants is expressed as the form of graph theory.A new unified rules for analyzing the thermal system are established.Combined with the first thermodynamics law and mass conservation law,weighted diagraph adjacency matrix is deducted.An example is given to illustrate the validity of the method.

Wall Thinning Analysis of Carbon Steel Pipes in Nuclear Power Plants

MA Na, YIN Chang-geng, QIN Jin-guang

2013, 34(2): 123-125.

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The wall-thinning problem of carbon steel pipes causes great attention at present because of the accidents happened in nuclear power plants.This paper introduces the wall thinning research of carbon steel pipes in the conventional island of a nuclear power plant.Operation condition analysis,chemical composition analysis,metallic phase analysis,micrographic examination,and XRD analysis are performed.According to the analysis above mentioned,the reasons of carbon steel pipes wall-thinning in this nuclear power plant are erosion-corrosion,cavitation,and flow accelerated corrosion,respectively.

A Scheme for Establishing of Criterions for Reactor Safe Operation in Condition of Fuel Clad Failure

LIN Xiao-ling

2013, 34(2): 126-128.

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Operation criterions are used to decide if the reactor can continue to work when the fuel clad failure.The method for establish the limits is presented.The tolerated maximum of failure fuel rods for the reactor safety should be calculated by risk analysis.The parameters are determined which can not only reflect the quantity but also be measured directly.The relationship is set up between the amounts with the parameters.The data calculated corresponding to maximum of failure fuel element which the reactor safety can stand are technical limits used to decide if the reactor can work continually.

Analysis and Management of Accelerated Aging of Relay in NPPs

ZHANG Sheng, MA Yi-jin, JIANG Hong, CHEN Shi-jun, HUANG Li-jun, MO Chun-ni

2013, 34(2): 129-132.

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This paper reveals the main factors of accelerated the aging of relay through statistical analysis of aging and failure data of intermediate relay and output relay for MV distribution board in Daya Bay nuclear power plant and Ling'ao nuclear power plant.The factors of accelerating relay aging are operation context including humidity,temperature,vibration impact,and etc.,and aging mechanism including oxidation corrosion,mechanical wear,electrical wear,organic gas corrosion,insulation aging,electromagnetic interference,and etc.The suggested measures are the classified management,relay replacement,periodic check,and periodic replacement or maintenance.

Heavy Water Reactor Pressure Tube Elongation Measurement and Data Analysis

SHANG Jun-min, ZOU Lian-lie, YUAN Jian-zhong, ZHOU Yi, ZHAO Wei-dong

2013, 34(2): 133-136.

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In this paper,mathematical statics method is used to analyze pressure tubes' elongation data acquired from two methods:refueling machine's encoder in Z direction and in-service inspection(ultrasonic).The result show that the data acquired from two methods is valid,and predict that the pressure tubes' service life are satisfied with design values from the elongation view,even the service life can prolong.

Study on Application of Ethanolamine in NPP Secondary Water Treatment

WANG Lin, XIE Yang, CUI Huai-ming

2013, 34(2): 137-140.

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Ethanolamine(ETA) is an organic amine which has stronger alkalinity and lower volatility than ammonia.The paper analyzes the influence of ETA used in the secondary system and calculates its distributing status at the different nodes of the secondary system.The study shows that ETA can enhance the pH value in liquid phase of vapor-liquid two phase regions,effectively inhibit the flow-accelerated corrosion,reduce the corrosion rate of the secondary system materials and improve the economy of the in-service nuclear power plants.It is concluded that ETA as a pH additive to the secondary side of pressurized water reactor is completely viable.

Effects of Seismic Wave Inputting Interface on Designed Surface Ground Motion in AP1000 Nuclear Island Structure

HOU Chun-lin, LI Xiao-jun

2013, 34(2): 141-146,152.

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In the process of seismic hazard assessment of nuclear power plant,there are some differences on bedrock faces that were defined according to shear wave velocity between China and America code.We select different bedrock faces,based on AP1000 Nuclear Power Plant design ground parameter model,that corresponding to Shear wave velocity of 700 m/s,1100 m/s and 2438 m/s,and using two methods of Chinese and American seismic responses for soil lays,we analyze the different seismic input interface on one analysis model and one seismic input,and obtain the peak acceleration of ground motion and response spectrum.The calculation results indicate that different seismic interface in one soil model and one seismic input could result in quite different results,and AP1000 Nuclear Power Plant design surface ground motion and spectrum characteristics change greatly,and the difference of peak acceleration of ground motion can be 2.25 times.So we suggest that we should use shear wave velocity of 2438m/s as the seismic input interface in AP1000 nuclear power plant site seismic safety analysis.The results of the paper could give some guidance for the later AP1000 Nuclear Power Plant construction and research.

Auto Type-Selection of Constant Supporting in Nuclear Power Stations

LIU Hu, WANG Fu-jun, LIU Wei, LI Zhao-qing

2013, 34(2): 147-152.

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To solve the type-selection of constant supporting in nuclear power stations,combining the characteristics of constant supporting which can adjust in the certain scope and the rules of load-displacement,the requirements and process for the type-selection of constant supporting is proposed,and the process of type-selection is optimized by Visual Basic.After inputting of the known parameters,the process can automatically select the most economical and reasonable constant supporting by array and function.

Feasibility Study on Development of Small Nuclear Power Reactors in China

CHEN Wen-jun, JIANG Sheng-yao

2013, 34(2): 153-156.

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Small nuclear power reactors have many advantages like low site requirements,flexible applications,low nuclear safety risks,easy operation,short construction period and low one-time investment.From the perspective of social and economic development in China,this paper analyzes the market requirements of small nuclear power reactors in energy saving and emission reduction,ocean development,and overseas export,and then analyzes the feasibility in technology and economy by combining with the development situation of small nuclear power reactors in China.

Research on Rejection Performance of Reverse Osmosis to Nickel in Simulated Radioactive Wastewater

KONG Jing-song, WANG Xiao-wei

2013, 34(2): 157-159,163.

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In order to reveal the rejection performance of the reverse osmosis applied in the radioactive wastewater treatment,treatment experiments were carried out on a pilot reverse osmosis equipment using wastewater containing nickel nuclide.Results showed that the rejection ratio of reverse osmosis to nickel was almost not affected by the operation pressure and the ratio of reclaiming,and had no direct relation with the salt rejection ratio.The ratio of nickel rejection reached 95% in the experiment condition and could meet the requirement on the disposal of radioactive wastewater produced by nuclear powered installations.

Removal of U（Ⅵ） from Aqueous Solution by Nanoscale Zero-Valent Iron

LI Xiao-yan, LIU Yi-bao, HUA Ming, GAO Bo

2013, 34(2): 160-163.

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Nanoscale zero-valent iron(NZVI) was synthesized in aqueous solutions by reduction of Fe³⁺with KBH₄.Removal of U(Ⅵ) by NZVI was investigated to understand the effect of dosages of NZVI,pH value of solution,initial concentrations of U(Ⅵ) and contact time.The results showed that:NZVI has very good removal effect on U(Ⅵ),and when pH value of solution is 5.5,the initial concentration of U(Ⅵ) is 45 mg/L,the dosage of NZVI is 1.0 g/L,the contact time is 2.5 h,the removal rate and adsorption capacity reached 98.98% and 27.22 mg·g^-1 respectively.

2013 Vol. 34, No. 2