核动力工程

Spatial Domain-Decomposed Parallel MOC Accelerated by CMFD

WU Wenbin, Li Qing, Wang Kan

2014, 35(1): 1-4.

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Spatial domain-decomposing is applicable for solving the neutron transport equation on massively parallel architectures. However, the more sub-domains there are, the slower the convergence is. Then CMFD is used to accelerate the spatial domain-decomposed parallel MOC in order to overcome the problem. The CMFD coarse diffusion equation is solved by ScaLAPACK, and the CMFD results are used to update both fine-mesh scalar fluxes and inner-interface angular fluxes. Numerical results on 1D slab example illustrate that CMFD is an efficient acceleration method for the spatial domain-decomposed parallel MOC, which improves the convergence remarkably.

Application of Diagonally Implicit Runge-Kutta Method for Solving Time-Dependent Convection-Diffusion Equation

Deng Zhihong, Sun Yuliang, Li Fu, Rizwan-uddin

2014, 35(1): 5-9.

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An efficient scheme for solving transient convection-diffusion equation was developed. Modified nodal expansion method（MNEM） was utilized for spatial discretization, while two kinds of diagonally implicit Runge-Kutta（DIRK） schemes—second-order DIRK and fourth-order DIRK were adopted for time discretization. The numerical results show that the numerical results of TDMNEM code agree with analytical solutions very well. MNEM has good ability in capturing sharp temperature variation. The efficiency of two time discretization methods depend on problem and the error criteria which been selected.

A New NMIS Characteristic Signature Acquisition Method Based on Time-Domain Fission Correlation Spectrum

Wei Biao, Yang Fan, Feng Peng, Ren Yong

2014, 35(1): 10-13.

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To deal with the disadvantages of the homogeneous signature of the nuclear material identification system（NMIS） and limited methods to extract the characteristic parameters of the nuclear materials, an enhanced method using the combination of the Time-of-Flight（TOF） and the Pulse Shape Discrimination（PSD） was introduced into the traditional characteristic parameters extraction and recognition system of the NMIS. With the help of the PSD, the γ signal and the neutron signal can be discriminated. Further based on the differences of the neutron-γ flight time of the detectors in various positions, a new timedomain signature reflecting the position information of unknown nuclear material was investigated. The simulation result showed that the algorithm is feasible and helpful to identify the relative position of unknown nuclear material.

Research of Core Fuel Management Using TVS-2M Fuel Assemblies in VVER

Wang Hongxia, XU Min

2014, 35(1): 14-18.

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Using KASKAD program package, the author make a research about the Tianwan nuclear power plant loading TVS-2M fuel assembly from the first cycle, also design the TVS-2M fuel assembly and on this basis, study fuel management, obtaining three fuel management cases, including year fuel cycle case and two long fuel cycle cases. In each program, the important parameters of the reactor core are analyzed and all the safety parameters meet the design requirements. In long fuel cycle program, TVS-2M is using from the first cycle and after the transition of only two cycles, the length of cycle reached the requirement of long period. The increased average annual capacity factor of the plant and the decreased times of overhaul during the core’s life which saving 30.8 percentage of the overhaul cost due to the long fuel cycle can largely improve the economic efficiency of the plant.

Research on Heat Transfer Correlation for Passive Containment Cooling System

Jiang Xiaoyu, YU Hongxing, Sun Yufa, YONG Jing

2014, 35(1): 19-22.

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In this paper, a CFD model in low Mach number regime was used to model the convection heat transfer tests in COPAIN facility, and validated using COPAIN experimental data. Using the CFD model to extend the experimental parameters in convection heat transfer to high temperature difference regime（40~80℃）, compute the Nusselt number for convection heat transfer, and a heat transfer correlation suitable for the PCCS design was derived. The condensation heat and mass transfer model closed by the new correlation was validated using the Wisconsin condensation tests data. Comparing with Dittus-Boelter correlation, the computed results are more close to the tests data.

Study on Neutronic-Thermohydraulic Coupling Calculation of SCWR

Liu Shichang, Cai Jiejin

2014, 35(1): 23-27,41.

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Based on the U.S. SCWR design, the thermohydraulic module is coupled with neutronic codes Dragon and Donjon directly, to develop a calculation codes for neutronic/thermohydraulic coupling, which can improve the accuracy and efficiency of coupling calculation. The relaxation factor is introduced to solve the problem of misconvergence of traditional successive iteration for new reactors such as SCWR. Key parameters have been calculated and analyzed to prove the feasibility and accuracy of the developed calculation codes. The study provides the basis for further study of stability and safety analysis for new SCWR.

Experimental Investigations on Heat Transfer of Supercritical CO₂ Flowing Through A Helically Coiled Tube

Wang Shuxiang, Niu Zhiyuan, Zhang Wei, XU Jinliang

2014, 35(1): 28-31.

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Heat transfer characteristics of supercritical CO₂ flowing upward through a uniform-heated helically- coiled tube were experimentally investigated under the conditions of q=0~25 kW/m², G=10~262 kg/（m²·s） and Pin =8~9 MPa. Effects of mass flux, heat flux and inlet pressure on the longitudinal heat transfer performance were analyzed. It was found that the heat transfer coefficients first increased and then decreased. The maximum values of heat transfer coefficients always occurred at the temperature range of Tb ＜Tpc ＜Tw. For the heat transfer coefficients ascent stage, the heat transfer capability enhancement per unit volume caused by the increased heat capacity, together with the thinner thermal boundary layer induced by the decreased viscosity, dominated the heat transfer performance at the near-wall region. While for the stage that CO₂ experienced the state transition from quasi-liquid to qusi-vapour, the sharp decreases of heat capacity and thermal conductivity became the dominant factor for the heat transfer deterioration. For the transcritical convective heat transfer with large physical property variations, the Nusselts number, standing for the ratio of heat convection to conduction, could not represent the real heat transfer capacity any more.

Numerical Computation of Sodium-Potassium Alloy Heat Pipe in Passive Residual Heat Removal System of Molten Salt Reactor

Wang Chenglong, Tian Wenxi, SU Guanghui, Zhang Dalin, WU Yingwei, Qiu Suizheng

2014, 35(1): 32-35.

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In the present paper, the transient performance of NaK heat pipe is numerical simulated in the case of MSR accident. The transient, two-dimensional conduction model is established in the region of heat pipe wall and wick by using Finite Element Method（FEM） and the quasi-steady, one-dimensional vapor model is employed in heat pipe vapor space. Finally, the distributions of temperature, velocity and pressure of entire NaK heat pipe are obtained. Numerical results show that NaK heat pipes have a rapid startup and remove the afterheat of fuel salt efficiently.

Analysis of Experiments for Vertical Out-Tube Steam Condensation in Presence of Non-Condensable Gases

SU Jiqiang, Sun Zhongning, Fan Guangming, Guo Zixuan

2014, 35(1): 36-41.

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In order to investigate the influence of various parameters in the steam condensation heat transfer process with non-condensable gas, and to get a more suitable empirical correlation, the wall under-cooling, pressure and the content of non-condensable gas were studied outside a vertical tube by experiments. The results showed that: at the same pressure, the relationship between wall sub-cooling and HTC is exponential, and helium stratification does not happen within the experimental range. Based on the analysis of various experimental variables, combined with a large number of experimental data, a wider scope of application of the empirical correlation associated is obtained with the experimental value of the error within ±20%.

Formation Mechanism of Radial Void Fraction Distribution of Bubbly Flow in A Vertical Circular Tube

Liu Guoqiang, Sun Licheng, Tian Daogui, Xing Dianchuan

2014, 35(1): 42-45,51.

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The formation mechanism of radial void fraction distribution of gas-liquid two phase flow in a vertical circular tube was investigated experimentally by using an optical fiber probe under ambient temperature and pressure. Experiments were conducted in a tube with inner diameter of 100 mm, and with the gas and liquid superficial velocity covering the ranges of 0.0042-0.053 m·s-1 and 0.071-0.213 m·s-1, respectively. The results show that local void fraction shows a ‘‘core’’ or a "wall" peak distribution along the radius with different gas and liquid flow rates. The analysis of the wall force and lift force acting on a bubble shows that the two forces are of great importance in leading to a "core" or "wall" peak distribution of the void fraction due to their key role in determining the transverse movement of a bubble.

Numerical Simulation Research of Subcooled Boiling Water in Vertical Concentric Annulus Pipe under Low Pressure

Li Songyu, Zhang Hong, Jiang Shengyao, YU Jiyang

2014, 35(1): 46-51.

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A numerical simulation of subcooled boiling in annulus pipe in low pressure conditions was performed on Computational Fluid Dynamics code CFX with user defined program. LEE’s subcooled boiling experiment results in 2002 and 2008 were used for calculations. Bubble departure diameter was used Tolubimsky modified correlation based on Unal bubble departure diameter model and bubble average diameter correlation was used Anglart correlation. The influences of different lift force models and turbulent dispersion force models in non drag models on the radial void fraction distribution were compared; a model use suggestion was given. Compared the calculation results and experiment, the simulation results indicate that the models are endowed with a preferable applicability in low pressure conditions

Experimental Evaluation of Critical Heat Flux Calculation Models of Rectangular Channel

Sheng Cheng, Zhou Tao, JU Zhongyun, Huang Yanping, Xiao Zejun

2014, 35(1): 52-55.

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Under the natural circulation conditions, the CHF occurring in rectangular channels is influenced by lots of factors, which have not understood completely yet. The experimental results obtained from current research were compared with the calculation results of Katto forced circulation model and Zhang natural circulation model. The applicability of the two models and effects of inlet velocity, outlet quality and pressure on the CHF under the experimental conditions were analyzed. Research indicated that the calculation results of Zhang model coincided better with the experimental values than that of Katto model. As the inlet velocity increased, the CHF of both natural and forced circulations would increase. As the outlet quality increased, the CHF of both circulations would decrease. As the system pressure increased, the CHF of both circulations would increase, and the increment speed of CHF would decrease under the higher pressure condition in natural circulation.

Study on Effect of Rolling on Two-Phase Frictional Pressure Drop in Narrow Channels

Zhang Zhen, Xiao Zejun, Yan Xiao, Qin Shengjie, Huang Yanping, Chen Bingde

2014, 35(1): 56-59.

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In the present paper, the effect of rolling motion on two-phase frictional pressure drop in rectangular channel was studied experimentally and theoretically. The result showed that, under rolling motion, the frictional pressure drop fluctuated similar to sine movement. And the frequency of the fluctuation was the same as the rolling frequency. Time averaged two-phase frictional pressure drop didn’t vary with rolling motion. The relative fluctuant friction pressure drop decreased with Full Liquid Reynolds number and mass quality increasing, and increased with rolling amplitude and rolling frequency increasing. The result also showed that, the relative fluctuant friction pressure drop did not vary with maximal angular acceleration. An empirical correlation to predict the relative fluctuant friction pressure drop under rolling motion was got.

Prediction about Chaotic Times Series of Natural Circulation Flow under Rolling Motion

Yuan Can, Cai Qi, Guo Li, Yan Feng

2014, 35(1): 60-63.

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The paper have proposed a chaotic time series prediction model, which combined phase space reconstruction with support vector machines. The model has been used to predict the coolant volume flow, in which a synchronous parameter optimization method was brought up based on particle swarm optimization algorithm, since the numerical value selection of related parameter was a key factor for the prediction precision. The average relative error of prediction values and actual observation values was 1.5% and relative precision was 0.9879. The result indicated that the model could apply for the natural circulation coolant volume flow prediction under rolling motion condition with high accuracy and robustness.

Experiment on Structural Stiffness of Energy Absorbing Device in Steam Generator Lower Horizontal Support

Xie Honghu, Zhou Peng, Ren Hongbing, Zhang Xinghui, Liang Xiaolong, Liu Xiaohua

2014, 35(1): 64-66.

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The 10 mm×10 mm energy absorbing device used for steam generator lower horizontal support, and its stiffness experiment, including material selection, experimental scheme, equipments and load exerting are briefly introduced. Then, the formulas of the load exerting vs displacement for the energy absorbing bar are formulated, and the experimental stiffness of the energy absorb device in the region of elastic and ductile are deduced from the formulas. Also, the curve of the load exerting vs residual displacement for the energy absorb device is obtained by the experimental data. All of which provide lots of conveniences for the energy absorb device used in steam generator lower horizontal support.

Analysis of Key Factors Related to Site Specific Design Ground Motion for NPPs

Li Xiaojun, He Qiumei, Hou Chunlin

2014, 35(1): 67-70,77.

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Based on 46 project data of evaluation of seismic safety for the Chinese nuclear power plant（NPP） siting and construction in the last ten years, the controlling impacts of the calculated results from different seismic hazard analysis methods are analyzed and discussed on the determination of the site specific design ground motion for nuclear power plants, and statistic analyses are made on the results calculated by the probabilistic seismic hazard analysis method and the results induced by tectonic earthquakes and diffuse earthquakes（diffuse seismicity） in the deterministic seismic hazard analysis method. The research result shows that: In the weaker seismicity area, the site specific design ground motion parameters are mainly controlled by the calculated results from the deterministic seismic hazard analysis method, especially the peak ground acceleration（PGA） and high frequency spectral accelerations are controlled by the results induced by the diffuse earthquake, and therefore, the NPP construction based on the site specific design ground motion in the area will have higher seismic safety margin; In relatively strong seismicity area, the site specific design ground motion parameters are more likely to controlled by the calculated results from the probabilistic method, and for some NPP sites, the ground motion parameters from probabilistic method, especially the low frequency spectral accelerations are much bigger than those from the deterministic method; The determination of site specific design ground motion parameters as a whole is very conservative for Chinese NPP sites.

Relationships between Response Spectra and PSD Functions for Simulation of Artificial Earthquakes for Nuclear Power Plant Design

Xing Hailing, Zhao Bin, LU Wensheng, Jiang Tong

2014, 35(1): 71-77.

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The requirements to envelope a target power spectral density（PSD） function compatible with the design response spectra are usually prescribed by seismic codes for generating artificial time series as seismic input. This paper contrasts several methodologies which are most widely used in engineering seismology for defining spectrum compatible PSD. Parameters of these methodologies were analyzed using the RG1.60 and AP1000 design response spectra of safe shutdown earthquake. The emphasis was then placed on the method proposed by Kaul and its applicability was illustrated. Artificial time histories, whose response spectra were in agreement with 2%-damped design response spectrum, were generated and their average PSD would be the minimum PSD requirements compatible with the spectrum. Numerical examples show the accuracy and applicability of those approaches.

Effect Analysis of Vibration Transmission of Finite Plate with Blocking Mass

Li Pengzhou, LU Jun, SUN Lei

2014, 35(1): 78-81,86.

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The analytical model of finite plate stiffened by blocking mass under a point force is established. The structural response is derived by a combination of the modal method and the traveling wave method, and used for studying effect of blocking mass on energy transmission from the exciting plate to the receiving plate taking mean square velocity as evaluation index. The effect by the parameters of the blocking mass such as the mass ratio on energy transmission is discussed. It is shown that in low frequencies range, blocking mass can effectively impede vibration energy transmission when the rigidity of the blocking mass is considered and will enhance vibration energy transmission when its rigidity is neglected. And it is also found that in higher frequencies range the attenuation effect gets better with the increase of the mass ratio in the higher frequencies.

Performance Assessment of Auxiliary Bearing in HTR-10 AMB Helium Circulator on the Event of Rotor Drop

Xiao Zhen, Yang Guojun, Li Yue, Shi Zhengang, YU Suyuan

2014, 35(1): 82-86.

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In this paper, a model for analyzing internal contact stress and external load of ball bearing from rotor displacement was developed based on the Hertz contact theory and applied to the analysis of the rotor drop test in HTR-10 helium circulator equipped with AMB（Active Magnetic Bearing） to gain a better understanding of auxiliary bearing performance at different stages after the rotor drop. It was shown that the auxiliary bearing can well resist axial impact produced by rotor drop, avoiding of internal severe plastic deformation and damage to the performance of the auxiliary bearing. Rotor’s rotary motion and the heat accumulation of the inner ring resulted from the initial acute acceleration are the main contributor of radial load during the rotor idling and may cause the failure of auxiliary bearing. This paper analyzed the influence of this load and confirmed that the auxiliary bearing can still work in its loading limits.

Study on Flow Induced Vibration Characteristics of Two Inline Tubes and Two Parallel Tubes

Feng Zhipeng, Zhang Yixiong, Zang Fenggang

2014, 35(1): 87-91.

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In order to investigate the fluid structure interaction problems occurring in tube bundles, a numerical model is presented. The unsteady three-dimensional Navier-Stokes equation and LES turbulence model are computed with finite volume approach on structured grids combined with the technique of dynamic mesh. The model presents a three dimensional fully coupled approach with solving the fluid flow and the structure vibration simultaneously, for the tube bundles in cross flows. The results show that, the presented model is validated with experimental data and compared with existing models in the literature. Trajectory of single tube is "8" shape at "locked in". The two parallel tubes oscillate in opposite phase and both have the same amplitude. For pitch higher than 1.5D, there is no mutual influence of the tubes and response is close to the case of single tube. The lift force and amplitude of downstream tube of two inline tubes are increasing as the pitch increases. The upstream tube is almost not influenced by the downstream one when the pitch is larger than 2D. The fluid force and response are close to the case of single tube.

A Model for Cracking of Ceramic Fuel Particles in Dispersion Fuel

LONG Chongsheng, Zhao Yi, Gao Wen, Xiao Hongxing, WEi Tianguo

2014, 35(1): 92-96,105.

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Based on the fracture strength of ceramic material, an analytical model for the cracking of ceramic fuel particle in nuclear dispersion fuel was proposed in the paper. The relationship between the cracking temperature and the burn-up of fuel particles has been calculated. The effects of metal matrix, external restraint, fuel particle size and content on the fuel particle cracking behavior have been investigated in detail, and the approaches to increase the cracking temperature were also discussed. The result shows that, the cracking temperature of fuel particles would decrease approximately in power law with the increasing of burn-up and would decrease linearly with the increasing of fuel particle content. Relative higher porosity and bigger pore size could be helpful to increase the cracking temperature.

Low Temperature Sintering Technology to UO₂ Fuel Pellets and Its Creep Properties in High Temperature

Li Rui

2014, 35(1): 97-100.

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In this paper, the low temperature sintering technology to UO2 pellets has been introduced, and we have studied the high temperature creep properties of the pellets which manufactured by low temperature sintering. The sintering temperatures are 1073 K, 1273 K, 1473 K and 1673 K, sintering time are 1 hour, 2 hours and 3 hours respectively. We obtained the highest sintering density of pellets at 1673K with 3 hours, which is 10.41g/cm³（94.98% theoretical density）. The grain size of pellets which prepared by low temperature sintering technology and traditional technology are 9.0μm and 23.8 μm respectively. So the high temperature creep properties of the two kinds of pellet must be studied. They were performed at 20-50 MPa, 1673 K and 1773 K respectively, under a nitrogen atmosphere to shorten the experimental time. According to the results, the creep rates of sintered UO2 with the grain sizes of 9.0 μm and 23.8 μm under the load of 10MPa are almost the same. The creep process is controlled by both Nabarro-Herring creep and Hamper-Dorn creep for uranium dioxide with grain size of 9.0 μm; while Hamper-Dorn creep is the dominant mechanism for uranium dioxide with grain size of 23.8 μm.

Research on Elastoplastic Strain Correction Factor of Titanium Alloy

DU Juan, Shao Xuejiao, Zhang Liping, Kan Qianhua, Guo Sujuan

2014, 35(1): 101-105.

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The ASME and RCC-M codes include a detailed fatigue evaluation based on elastically predicted stresses. A prerequisite for the fatigue analysis is that the primary-plus-secondary stress range does not exceed 3Sm. If this limit is exceeded, the code provides "simplified Elastoplatic Analysis" rules for the fatigue evaluation. A Ke penalty factor is the most important parameter of this method. The ASME and RCC-M codes provides the codified expressions of Ke and material parameters for general materials of nuclear components, but there is no expressions and material parameters for the Titanium Alloy. The determination of Ke expressions and parameters for Titanium Alloy is discussed. A exemplary verification calculations of general materials that have been performed and the conservative allowance is obtained. Based on the results, the proposed Ke factors of Titanium Alloy has been determined and deliver the same sufficiently conservative results.

Transient Identification in Nuclear Power Plants Based on Transient Division and Fuzzy Euclidean Distance

Chang Yuan, Huang Xiaojin, Li Chunwen, Hao Yi

2014, 35(1): 106-109.

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The transient identification techniques were recently developed to alert the operators about the faults in their early stages, therefore corrective actions can be taken in time to keep the safety of nuclear power plants. In this paper, the transients were split into two parts: the first part is identified by clustering method, while the other is identified by comparing three features of signals. The similarity between the on-line and the reference transients was described by Fuzzy Euclidean distance, which conforms to human understanding habits. The method was verified by simulator data of Pebble Bed Modular High Temperature Gas-cooled Reactor（HTR-PM） with Tsinghua University. It is shown that the transients can be correctly and quickly identified.

Research and Application of ARE Flow Measurement Device under Different Standards

Yang Dongfang

2014, 35(1): 110-112.

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The Ling’ao II PWR is a pirated and improvement from the Ling’ao I PWR, so the main water flow measurement device remained unchanged. But its acceptance criteria, British BS standard, is replaced by American ASME standard. This paper conducts the analysis and research on ARE flow measurement device selection, use, acceptance criteria and existing problems.

Discussion on Method of System I＆C Classification of CPR1000 Nuclear Power Plants

Si Hengyuan, Hu Jian, ZhONG Bin

2014, 35(1): 113-116.

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The item safety classification of nuclear power plants has to comply with the basic principles presented in relative rules and codes, while there is no uniform conclusion for the method of safety classification, so much the worse for that of system I＆C classification. The paper gives an implemention method for the system I＆C classification, namely, based on the system function classification and system operation analysis, identifying the safety signals of system in accident conditions, and distinguishing the control process of the relative equipments, then the system I＆C classification is determined based on the classification of system function which the signals and control process are performed. The implementation of the method on the I＆C design of essential service water system（SEC） in Hongyanhe nuclear power plant is reviewed.

Research on Measurement of Neutron Flux in Irradiation Channels of Research Reactor

Yin Zhitao, LÜ Zheng, Wang Yulin, Zheng Wuqin

2014, 35(1): 117-121.

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Relative distribution of thermal neutron flux in the irradiation channel is measured by classical activation foil method. After that, on a representative point in the irradiation channel, neutron temperature and absolute neutron flux are also measured. Cadmium ratio correction method is used to check the experiment result in the end. Comparative analysis shows that the results from two different methods are agreed pretty well, which adds the credibility of experiment results.

Defense-in-Depth and Diversity Design of Instrumentation and Control System in Nuclear Power Plant

Zhou Jixiang, ZhU Pan, Xiao Peng

2014, 35(1): 122-124.

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Nuclear power plant is designed with multiple level of defense for the function used to protect the core and limit the spread of radioactivity during an event. The design of instrumentation and control system supports this multiple level design philosophy. Defense-in-depth and diversity is considered in the design of instrumentation and control system to maintain the integrity and availability of protective barriers or means and defense potential common-cause failures. This can effectively limit the development of accident, mitigate accident consequence, prevent the spread of radioactive material to the environment and ensure safety of reactor and plant equipment and staff.

Studies on Design of Internal Hydraulic Control Rod Drive System Loop for Nuclear Reactors

Zhao Chenru, BO Hanliang

2014, 35(1): 125-128.

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Based on the characteristics of the internal hydraulic control rod drive technology, two kinds of drive system loop schemes, the centrifugal pump continuous operation scheme and the membrane pump intermittent operation scheme are proposed and compared from the perspective of vibration, noise and energy consumption. The membrane pump intermittent operation scheme is especially calculated and discussed for a typical reactor cold commissioning case. The effects of the hydraulic cylinder leak flow rate and the control rod action interval time on the pump intermittent time and pump operation time are analyzed. Results show that the centrifugal pump continuous operation scheme applies to cases in which the control rod number and hydraulic cylinder leak flow rate are relative large. When the leak flow rate is relative small, the membrane pump intermittent operation scheme is a better choice and the pump intermittent time could be elongated and pump operation time could be shortened by selecting proper pump flow rate and the ratio of the gas to water in the water tank, which is of great help in reducing vibration, noise and energy consumption.

Research of Optimum Design and Welding Procedure Improvement for Internals Reactor Core Barrel of NPPs

Wang Qingtian, Chen Xungang, XiA Xin

2014, 35(1): 129-133.

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The paper introduces the structure characteristics and design requirements of 1000 MWe nuclear power plant of improved second generation. Drawing welding and fabricating lessons of core barrel from domestic and overseas manufactory, combining the structure characteristics, stock option and design accuracy requirement, the paper put forward specific design and manufacture process measures to reduce welding deformation and control dimension precision on the base of theory analysis, requirement of ASME code and widespread experience in and aboard. Welding load and residual stress will be reduced largely and the manufacture difficulty and deformation risk also will be reduced through these measures.

Research on Design of Reactor Protection Equipment Checkout Console

Yang Yang, Han Wenxing, HE Li, WU Zhiqiang, MA Quan

2014, 35(1): 134-137.

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In order to improve the automatic testing level of maintainer and rapid test, fault detection for the unit of reactor protection assembly, this paper designs the reactor protection assembly checkout console. According to the method that pulse output circuit based on FPGA cooperates with fasting collection card sampling frequency and buffer data counting, the setpoint unit self-test interface test is conducted. The accurate mv signal simulation is realized according to the circuit of voltage feedback module and current feedback module. The fault locating of reactor protection assembly unit is realized also according to the fault locating expert system. After the testing and experimentation of Reactor Protection assembly checkout console, the result shows that the technical characters of Reactor Protection assembly checkout console satisfies the requirement of user.

Power Control Equipment for Control Rod Drive Mechanisms Based on IGBT in Nuclear Power Plant

Zheng Gao, Huang Kedong, YU Haitao, MA Quan, Jin Yuan, Tian Yu, Li Guoyong

2014, 35(1): 138-141.

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The closed-cycle control technology, Pulse-Width Modulation（PWM） technology and programmable logic controller（PLC） technology with the function of maturity are applied in control and power device for control rod drive mechanism in nuclear plant. The functions of the device are described as following: the control to the IGBT within control circuit is effective,the current of control object is adjustable linearity, and the discharging time is adjustable. Through development of this device, the inherent problems of native device and import device are solved. It is up to the level as control technology, performance and reliability of the import equipment.

Finite Element Analysis of Interference Fit of Impeller in HTR Primary Helium Circulator

Li Kai, ZhU Baoshan, Wang Hong

2014, 35(1): 142-146.

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In this paper, the contact elements in software ANSYS are used to simulate the interference fits for the impeller and the shaft in a high temperature reactor（HTR） primary helium circulator. Initial finite element magnitude of interference could be defined by adjusting these contact elements’ real constants, The simulations have been carried for different magnitude of interference, different friction coefficients and different impeller rotational speeds, contact pressure distributions on the fitting surface under these parameters are given. The results show that contact pressure changes along the circumferential direction periodically and it increase with the increase of magnitude of interference linearly. The higher rotational speed of impeller is, the lower contact pressure is on fitting surface close to the hub side and higher contact pressure is close to the shroud.

Test Interval for Engineered Safeguards Features of Nuclear Reactor

Shang Yanlong, CAi Qi, Chen Lisheng, Zhao Xi

2014, 35(1): 147-151.

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In this paper, based on the traditional average availability model of Engineered Safeguards Features（ESF） of the nuclear reactor, a component availability model has been established, which considers the states dependency in the standby, test and repair periods and uses less assumption. This model can be used in a more extensive field than the traditional one, and it is necessary to obtain the high-accuracy results for the large and complex system. The deduced model is applied for obtaining the TI of residual heat removal system as an example. The results can provide the guidance for the operation management and maintenance decision for ESF of the nuclear power plant.

Research on Characteristics of Passive Residual Heat Removal System for Modular Reactor

Fan Shuchun, LU Jianchao, Peng Shinian, Zhang Xianjun

2014, 35(1): 152-155.

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Based on the preliminary design passive residual heat removal system project of the modular reactor, and considering the structure and operation characteristics, the reasonable control volumes are divided and the mathematical models are established. Using numerical iteration method and in common currency program, the transient-state codes is adopted to analyze the thermo-hydraulic characteristics of the passive residual heat removal system. The results show that, after the reactor power failure, natural circulation of system can quickly be established; in the passive residual heat removal process, no flow instability happens; emergency cooler heat exchange area change of residual heat removal capability has no significant effect in a certain range.

Thermal Fatigue Phenomenon of High Pressure SafetyInjection Pipes and Modification Solution

Zhang Shoujie

2014, 35(1): 156-160,164.

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This paper introduce the thermal hydraulic characteristic of the liquid in the high pressure safety injection dead lines, and analyze the mechanism and harm of thermal fatigue which may occur on these pipes. A modification is proposed for Daya Bay and Ling’ao Phase I nuclear power plants high pressure saft injection deadline,and take fatigue analysis on pipes added by the modification, find that there is no pipes added by the modification become fatigue after various produce condition loaded on it,demonstrate the efficiency and safety of the well modification.

False Alarm Analysis of Loose Part Monitoring System at Tianwan NPP

Zhou Zhengping, OUYang Qin, Mao Qiuhua, YUan Shaobo

2014, 35(1): 161-164.

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The equipment component and system structure of Loose Part Monitoring System（LPMS） in Tianwan NPP are introduced, and the alarm logic of LPMS is described. According to years operation record of LPMS in Tianwan NPP, the typical cases of false alarm of LPMS are analyzed and diagnozed, and the signal characteristics and origin of loose part alarm are summarized.

Thermodynamic Analysis of High Temperature Gas-cooled Reactor He/Ammonia Combined Cycle

Luo Chending, Zhao Fuqiang, Zhou Jie, Zhang Na

2014, 35(1): 165-169.

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Thermal performances of a high temperature gas-cooled reactor（HTGR） He/ammonia combined cycle（HAC） have been analyzed in this paper. Research shows that compared with the HCC cycle, the thermal and exergy efficiencies of HAC increase by 6.0 %-points and 8.1 %-points respectively, and achieve 46.2% and 62.0% separately. The HAC cycle has effective nuclear energy utilization and good prospects for development.

Application Study on Wireless Acceleration Sensor in Experiment of Graphite Reactor in Seismic Core Structure

Ni Zhensong, Sun Libin, WU Xinxin

2014, 35(1): 170-173.

Abstract(6) PDF(0)

Abstract:
A method for wireless acceleration sensor for detecting the dynamic response of the graphite reactor internals under earthquake conditions is proposed, and it can make a clear assessment of integrity of the graphite structure under seismic conditions. Test results show the consistence of the detection result of wireless acceleration sensor and the cable acceleration sensor detection when all the conditions for the seismic experiments are satisfied. The wireless detection technology avoids the disadvantages of high installation cost, and poor reliability and safety of the graphite structure detection system based on the traditional wired network.

Feasibility Study on Application of Polyantimonic Acid in ⁹⁰Y Generator Preparation

Deng Qimin, Yin Bangshun, Cheng Zuoyong, Li Mingqi, Li Maoliang

2014, 35(1): 174-177.

Abstract(7) PDF(0)

Abstract:
The adsorption capacity for Sr and Y with Polyantimonic Acid（PAA） in 0.1mol/L nitric acid solution was 16.25 mg/mL and 42.1 μg/g respectively. Sr and Y can’t be separated with nitric acid solution of different concentration. More than 72% of Y can be eluted with 100 mL 0.02 mol/L EDTA as elution, but PAA was unstable in EDTA solution. 86.2% of Y can be desorbed with 20 mL 0.12mol/L DTPA at the temperature is 80 ℃. The results indicated that PAA may be used for 90Y generator preparation, but system of generator should be rational designed to meet requires of 90Y generator.

Estimation of Carbon-14 Production in HFETR

Liu Shuiqing, Sun Yu, MA Liyong, Yang Bin

2014, 35(1): 178-180.

Abstract(6) PDF(0)

Abstract:
The core reactivity change from AlN targets and ¹⁴C producing rate is estimated by calculating the using distribution of reactor neutron, while the fuel management code of HFETR is used to contrast with the estimation calculation. The calculation result shows that it causes about 250×10^-5 reduction of reactivity and core life will be cut down about 65 MW·d when all the 80 fuel assembly center holes are filled with 4000 gram AlN targets, while 1.0×10¹² Bq ¹⁴C could be produced per year. The estimation results accord well with HFETR reactor core design and operation status.

2014 Vol. 35, No. 1