核动力工程

Experimental Study on Natural Circulation Characteristics of Passive Residual Heat Removal System

Xi Zhao, Xiong Wanyu, Xie Feng, Gong Houjun, Zhuo Wenbin, Li Pengzhou

2015, 36(4): 1-3. doi: 10.13832/j.jnpe.2015.04.0001

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Abstract:
Experiments on the natural circulation performance of the secondary passive residual heat removal system（PRS） for Hua Long Yi Hao have been performed by using the high temperature high pressure thermal-hydraulic test facility ESPRIT（Emergency Secondary Passive Residual Heat Removal System Integral Test Facility）. The experimental results show that the heat transfer of PRS and the condensation heat exchanger meet the design parameters of 0.5%FP. The result also shows that the core heat can be sufficiently removed during 72 h after shutdown in station black-out（SBO） accident.

Flow Patterns Transition in Vertical and Upwardly Inclined Two-Phase Flow

Xie Tianzhou, Chen Bingde, Xu Jianjun, Bao Wei

2015, 36(4): 4-7. doi: 10.13832/j.jnpe.2015.04.0004

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An experimental study on the flow patterns in vertical and upwardly inclined two-phase flow is carried out using de-ionized water and air as the working media. Five major flow regimes, i.e, dispersed bubble, bubble, slug, churn and annular flow, are recognized. By mechanism analysis, flow pattern transitions from dispersed bubble to bubble/slug, bubble to slug, slug to churn, and churn to annular, are derived and agree well with the experimental data.

Theoretical Study on Effect of Flow Channel Local Inclination on Critical Heat Flux

Liu Wenxing, Peng Jingfeng, Xu Jianjun, Huang Yanping, Yang Zumao

2015, 36(4): 8-11. doi: 10.13832/j.jnpe.2015.04.0008

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The present work aims at the curved structure of coolant flow channel and corresponding operation conditions in the subcritical energy blanket of fusion-fission hybrid reactor. The critical heat flux（CHF） model is established for the local inclined channel and corresponding code is developed. Then the model is verified with experimental data under both vertical and inclined conditions. The results show that the present model has good prediction accuracy in a wide parameter range. After that, the local inclination effect on CHF is investigated using the developed code. The results show that with the increasing of the local inclination angle and worsening thermal-hydraulic conditions, the CHF ratio of local-inclined to vertical flow channel decreases.

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

Du Daiquan, Zhou Huihui, Xu Jianjun, Yang Zumao, Huang Yanping

2015, 36(4): 12-16. doi: 10.13832/j.jnpe.2015.04.0012

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Within the range of Reynolds number from 467 to 3350, and heat flux from 50 to 150 k W/m², the experimental investigation was carried out on the flow and heat transfer characteristics of subcooling water（single phase） for equivalent model of sphere bed. The effects of thermal parameters on the flow and heat transfer characteristics of equivalent model were analyzed, and the experimental correlations of the flow resistance coefficient and heat transfer coefficient were obtained. The experimental results show that the pressure drop is increased quadratically with the increasing of superficial velocity, and the pressure drop increases with the decreasing of inlet temperature. The convection heat transfer coefficient increases with the increasing of heat flux and superficial velocity.

Effect of Liquid Viscosity on Film Interfacial Instability on the Wall of Cyclone Separator

Huang Zhen, Xiao Zejun, Yan Xiao, Zan Yuanfeng, Li Yong, Yuan Dewen

2015, 36(4): 17-22. doi: 10.13832/j.jnpe.2015.04.0017

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A theory study on film interface stability on the wall of cyclone separator in viscous condition was performed in this paper. The dynamics boundary condition and kinematic boundary condition in the rotational flow field were obtained based on the stress analysis of film in viscous condition. Then the momentum equations and continuity equations of gas-liquid fluid were linearized by substitution of potential function. The dispersion relation of film interface wave in viscous condition was established by the combination of boundary conditions and the linearized equations. The judgment criteria for film interface stability was obtained. A code was developed according to the model of film motion and dispersion relation. The effect of liquid viscosity on the film interface stability in the cyclone separator was obtained based on the analysis of film interface stability in viscous condition.

Effect of the Structure of Steam Flow Limiter on Its Resistance and Flow Field Details

Yang Xuelong, Feng Jing, Zhang Qian, Wang Wei, Wang Xianyuan

2015, 36(4): 23-27. doi: 10.13832/j.jnpe.2015.04.0023

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Three dimensional numerical simulations and flow analysis were carried out to study the resistance and flow field of the steam flow limiter adopted in a steam generator. The realizable k-ε model with standard wall function was utilized to calculate the three-dimensional flow field of the flow limiter. The effect of the structure on the flow details and resistance was studied through analyzing the simulation results. The resistance of the flow limiter increases with the inlet mass flow rate while the resistance coefficient maintains constant, about 4.95 and 3.05 respectively. For the 7-nozzle flow limiter, the major pressure loss, about 71.4%, is caused by the backflow region, while for the 19-nozzle flow limiter, the major pressure loss is caused by the Venturi nozzles and the backflow region, about 50.5% and 44.5% respectively. The pressure, velocity and turbulent kinetic energy in the flow fields of the 19-nozzle flow limiter reach steady more rapidly than that of the 7-nozzle flow limiter. For the 7-nozzle flow limiter, the mass flow rate of the centric nozzle is 1.04 times that of the outer nozzle, while for the 19-nozzle flow limiter, the mass flow rate of the centric nozzle is 1.18 times that of the outer nozzle. The outer nozzle of the 19-nozzle flow limiter has backflow region in its diffuser, hence the performance of the 19-nozzle flow limiter could be improved through optimizing its structure.

Analysis and Improvement of Throttling Orifice of NPP RCV System

Zhao Jingxiong, Liu Zhangliang

2015, 36(4): 28-31. doi: 10.13832/j.jnpe.2015.04.0028

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We analyzed and simulated the throttling orifice RCV001/002/003 DI of M310 NPP’s RCV System. The defect in the original design of M310 has been found and optimized by using the step-orifice to reduce vibration and noise. This design change has been applied and verified by several NPPs which are either under construction or operation.

Preliminary Research on Design of Traveling Wave Reactor

Yan Mingyu, Chen Bin, Feng Linna, Zhang Yong

2015, 36(4): 32-36. doi: 10.13832/j.jnpe.2015.04.0032

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An engineering feasible conceptual core design of large scale（e.g. 1000 MWe output） TWR is proposed with investigation and qualitative optimization on the proper design of fuel element structure, fuel pellet, liquid metal filling gap, fuel assembly structure, core reflector and shielding and shutdown control rods. The optimized design presents a flatten radial neutron flux with a better equivalent state distribution, which means the long term burning state could be defined by initial core design and further corrected by the travelling wave progress. The optimized fuel structure improves the flow distribution between the central, parallel and corner channels. Furthermore, the power control of TWR could be implemented by the adjusting of coolant pump rotation speed as the change of coolant flow. Though the load rejection and power control between 15% to 100% nominal power could not be fulfilled by flow control without the participation of bank A control rods.

Fuel Management for VVER-1000 using PC Fuel

Xu Min, Wang Hongxia, huo Xiaodong, Yi Xuan, Yu Yang

2015, 36(4): 37-40. doi: 10.13832/j.jnpe.2015.04.0037

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When the nuclear power plants use the fuel assemblies produced from reprocessing plant, the Closed Loop of nuclear fuel cycle is realized. It can increase the fuel utilization and mitigate the spent fuel storage problems. The PC fuel is made from the fuel reprocessing products by enriching and processing. It has been used over 20 years in the Russian VVER-1000 plants. The owner has intended to use PC grade fuel from Cycle 10 of Unit 1. The research and design about VVER-1000 using TVS-2M fuel assembly made of PC uranium hexafluoride is carried out using KASKAD program package. An optimized fuel management scheme is obtained through the feasibility analysis.

Development and Verification of Three Dimensional Code System for SCWR Core Steady State Analysis

Wang Lianjie, Zhao Wenbo, Yang Ping, Ma Yongqiang, Lu Di, Sun Wei

2015, 36(4): 41-44. doi: 10.13832/j.jnpe.2015.04.0041

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A coupled neutronics/thermal-hydraulics three dimensional code system SNTA is developed for SCWR core steady state analysis by modular coupling the improved neutronics nodal method code and SCWR thermal-hydraulic sub-channel code. The problem of CSR1000 core is studied to verify SNTA. The results calculated by SNTA are agreed well with those by CASIR and SRAC. SNTA is more efficient than CASIR and SRAC which neutronics modules are based on the Finite Difference Method. The numeric results show that SNTA can be well applied to SCWR core steady state analysis and core concept design.

Bai Wenting, Feng Guozhong

2015, 36(4): 45-48. doi: 10.13832/j.jnpe.2015.04.0045

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In order to compensate for a lack of strong earthquake records when Code for Seismic Design of the Nuclear Power Plant（GB50267-97） was drafted at that time, the records of strong earthquakes and great earthquakes are chosen as a supplementary for revising the standard response spectrum in GB50267-97. Referring to the way of not distinguishing between the hard field and rock field in RG1.60 and modified RG1.60, the response spectrum of hard field and rock field are integrated, and the suggested corrections of response spectrum are obtained at intermediate frequency and high frequency in GB with the method of probability statistics.

Structural Integrity Assessment for RPV under Typical Event Transients

Zhu Guangqiang, Tian Xianglu, Wei Wenbin

2015, 36(4): 49-53. doi: 10.13832/j.jnpe.2015.04.0049

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The beltline region and inlet nozzle of the reactor pressure vessel（RPV） is investigated in this paper. Fracture mechanics finite element model is built, the transient temperature filed and stress filed are analyzed using the detailed thermal engineering analysis result of typical event transients as the input conditions. Combining with the irradiation embrittlement assessment result, the structure integrity of RPV under events is analyzed and assessed adopting the analysis method for the deterministic fracture mechanics. The analysis result shows that brittle rupture would not happen in the interesting region of RPV during the life time of forty years, but attention should be paid to the transients with great changing rate of coolant temperature.

Analysis of Optimization Requirements of Impact Test in Welding Procedure Qualification for Mechanical Components of Nuclear Islands

Wang Heng, Dong An

2015, 36(4): 54-56. doi: 10.13832/j.jnpe.2015.04.0054

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There is a requirement in standard NB/T 20002.3, for class 1 components in Mn Mo Ni quenched and tempered steels, additional test specimens shall taken with a notch at 4mm from the fusion line. The paper analyzes it from three aspects of large impact test data, relevant standards and requirements, and organizations and performance characteristics in the HAZ. The results indicate that the optimization requirements of impact test in the welding procedure qualification for mechanical components of nuclear islands are reasonable.

Test Study on Effect of Blocking Mass on Finite Plate Vibration Attenuation

Li Pengzhou, Lu Jun, Sun Lei

2015, 36(4): 57-60. doi: 10.13832/j.jnpe.2015.04.0057

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The test model of finite plate stiffened by blocking mass used to attenuate the structural vibration is established. The vibration frequencies, the modal shapes of finite plates affected by blocking mass and its capacity to attenuate the plate vibration are measured. The test results are compared with the theoretical calculations. The attenuation trend of the vibration transmission from the exciting plate to the receiving plate is concluded for plates with three different sizes. The results show that the insertion loss resulting from test and calculation have little difference. There are "block-frequency-range" and "pass-frequency-range" of vibration transmission, their comprehensive effects result to effectively attenuate vibration transmission of finite plate and the attenuation effect gets better with the increasing of the mass radio.

Load Analysis and Factors Determined Method of Special Lifting Device for NPP Nuclear Island Major Components

Weng Songfeng, Dong Zhengping

2015, 36(4): 61-64. doi: 10.13832/j.jnpe.2015.04.0061

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At present, no standards is appropriative to the design of the special lifting device for the nuclear island major components, which is used to hoist the major components of reactor and main loop. This paper analyzes the specified rated load, safety factor and trial load of the special lifting device for NPP nuclear island major components, and gives the method to determine the factors according to the standards and experience. Furthermore, design experience and feedback are discussed in this paper.

Development of Computer Code on Pipe Crack Leakage

Wu Wanjun, Xie Hai, Lan Bin, Huang Xuan, YE Xianhui

2015, 36(4): 65-68. doi: 10.13832/j.jnpe.2015.04.0065

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The crack leakage is a key parameter to determine whether the candidate pipe meets the requirements of leak before break application. Study on the theory of crack leakage calculation was performed, and a computer code named PICLES was developed. For more convenience usage, the ability of calculating leak crack length directly was also implemented. The software is verified by comparing the calculation with the target software. The results show that PICLES has the same result with the target software for same inputs and is more efficient. PICLES can be used for the evaluation of leak before break of pipes..

Optimal Design of Pressure Control System of PWR Pressurizer Based on Digital Regulators

Qian Hong, Zhou Lei, Mao Lei

2015, 36(4): 69-73. doi: 10.13832/j.jnpe.2015.04.0069

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Based on the pressure control system of the pressurizer with digital regulator, and according to different dynamic characteristics of the pressurizer pressure rising or reducing, a control system using a digital regulator with two sets of parameters used in control pressure of pressurizer is proposed in this paper. The design of SAMA Diagram of Control Strategy with automatic switching function was given and two sets of PI parameters of digital regulators were tuned in SIMULINK. The results verify that the design could better recover the pressure to the set point, thus the design improves the security of operation in nuclear power plants.

Application of GASFLOW and COM3D in Analysis of Hydrogen Behavior of Nuclear Reactors

Zhang Ruidong, Sun Ximing, Dong Yujie

2015, 36(4): 74-78. doi: 10.13832/j.jnpe.2015.04.0074

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The risk of hydrogen combustion and detonation is always one of the most important nuclear safety issues. In this paper, the hydrogen transport and distribution in the containment is simulated with GASFLOW code when a LOCA accident happened. Then by coupling the GASFLOW data to COM3 D which is developed by KIT to simulate the hydrogen combustion and detonation, the pressure and temperature development in the containment with different combustion locations is analyzed.

Strategies of Factory Tests of Safety Digital Instrumentation and Control System in Nuclear Power Plants

Wang Zhongqiu, Wu Qi, Zhang Yunbo, Liu Le

2015, 36(4): 79-82. doi: 10.13832/j.jnpe.2015.04.0079

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Combining the domestic practice and the experiences in the nuclear safety review of the factory tests（FT） of the safety digital instrument and control system（DCS）, DCS FT is identified as several stages according to nuclear safety regulations and standards. The results show that the safety DCS function and performance should be tested. FOAK test strategy should be used discreetly, and not suggested to be used for the first system.

Corrosion Fatigue Cracking Behavior of Inconel 690（TT） in Secondary Water of Pressurized Water Reactors

Xiao Jun, Chen Luyao, Fu Zhenghong, Qiu Shaoyu, Chen Yong, Lin Zhenxia

2015, 36(4): 83-85. doi: 10.13832/j.jnpe.2015.04.0083

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Inconel 690（TT） is one of the key materials for tubes of steam generators for pressurized water reactors, where it is susceptible to corrosion fatigue cracking. In this paper, the corrosion fatigue cracking behavior of Inconel 690（TT） was investigated under small scale yielding conditions, in the simulated secondary water of pressurized water reactor. It was observed that the fatigue crack growth rate was accelerated by a maximum factor up to 3 in the simulated secondary water, comparing to that in room temperature air. In addition, it was found that the accelerating effect was influenced by out-of-plane cracking of corrosion fatigue cracks and also correlated with stress intensity factor range, maximum stress intensity factor and stress ratio.

Effect of Coating Temperature on Density of Porous Pyrolytic Carbon

Tang Mingguo, Wu Shihong, Li Yingjie, Yang Jing

2015, 36(4): 86-89. doi: 10.13832/j.jnpe.2015.04.0086

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The porous pyrolytic carbon was made by chemical vapor deposition in the temperature range from 1100~1550℃, using acetylene as the carbon source gas. Density of porous pyrolytic carbon was measured by image analysis technique. The microstructure of porous pyrolytic carbon was analyzed by SEM. The results show that: with the increasing of coating temperature at 1100~1250℃, the density of porous pyrolytic carbon increases; with the increasing of coating temperature at 1250~1350℃, the density of porous pyrolytic carbon reduces, and the density of porous pyrolytic carbon tends to be stable at＞1350℃ range; when the coating temperature equals to（1450±50）℃, the eligible porous pyrolytic carbon can be made.

Study on Compatibility of U-Mo Alloy and Zirconium Alloy

Liu Yunming, Chen Jiangang, Liu Chaohong, Sun Zhanglong, Pang Xiaoxuan, Wang Luquan, Yin Changgeng

2015, 36(4): 90-94. doi: 10.13832/j.jnpe.2015.04.0090

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The compatibility of U-Mo alloy and Zirconium alloy was studied in this paper. The diffusion-couple was prepared by coated hot rolling and annealed at 750℃ and 850℃ for 10 and 50 hours. The result shows that the diffusion-layer has delaminated between U-Mo alloy and Zirconium alloy, and the diffusion-layer has a complicated composition with the Mo₂Zr second precipitate. XRD result shows that the composition of diffusion-layer is the Mo₂Zr, UZr₂ and U. The diffused mechanism of U-Mo and Zirconium is that the atom of U and Mo diffuse into Zirconium alloy, then Mo and Zr form Mo₂Zr firstly and the U continuously through out Mo₂Zr to form γ-（U,Zr） solid-solution. The U-Mo alloy has good compatibility with Zirconium alloy.

Development of ZCU Water Levels Search Function of CANDU Reactor

Ju Haitao, Wu Hongchun

2015, 36(4): 95-97. doi: 10.13832/j.jnpe.2015.04.0095

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The Zone Controllers Unit（ZCU） levels Search method of Canada Deuterium Uranium（CANDU） reactor is researched in this paper. The equations about the ZCU critical levels search function and target power distribution search function are presented. Both are implemented in the code DONJON. The corresponding computation and research on the combination of union search functions are carried out in Qinshan-III reactor and the results prove that the method is successful.

Experimental Investigation on Entrainment of Fourth Stage Automatic Depressurization System in AP1000

Xiang Yan, Sun Doucheng, Liu Jianchang, Wu Yingwei, Zhang Peng, Qiu Suizheng, Su Guanghui

2015, 36(4): 98-102. doi: 10.13832/j.jnpe.2015.04.0098

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In order to investigate the entrainment process of the fourth stage automatic depressurization system（ADS-4）, ADS-4 Depressurization and Entrainment Test Loop（ADETEL） scaled after AP1000 was constructed. The phenomenon of the entrainment was recorded by a high speed camera and analyzed particularly. The data was compared with the existing data and correlations. It discovers that the entrainment rate declined rapidly with the decrease of liquid level in hot leg when the liquid level is low. The onset of entrainment is more likely to happen in the condition of small ratio of branch-head line. The ADS-4 entrainment rate in AP1000 is lower than that in AP600 at the same relative hot leg liquid level.

Design Improvement of CRDM Nozzle for RPV

Wang Xiaobin, Li Yuguang, Luo Ying, Fang Caishun, Chen Haibo

2015, 36(4): 103-106. doi: 10.13832/j.jnpe.2015.04.0103

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Based on the design structure and manufacture techniques of CRDM（Control Rod Drive Mechanism） nozzle, this paper analyzes and determines the root causes for the changes of inside diameter and perpendicularity for CRDM nozzle in terms of weld structure design, characteristics of stainless steel, design strength and implementation of pressure tests of the nozzle, and proposes the measures to control the key parameters.

Consideration of ALARA Principle in Design of Steam Generator

Cui Suwen, Ren Hongbing, Gao Xipei

2015, 36(4): 107-110. doi: 10.13832/j.jnpe.2015.04.0107

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The principle of ALARA（As Low as Reasonably Achievable） is one of the basic considerations in radiation protection. As one of the most important primary equipment, during the design of steam generators, the principle of ALARA shall be considered to limit or reduce the radiation exposure of plant personnel. Through analyzing the design process which might affect the irradiation dose, it is concluded that SG will further fulfill the ALARA principle if measures are taken to optimize the material used, to drain the primary coolant completely and quickly, to improve the in-service methods, to strengthen radiation shielding and to optimizing the water chemistry. Based on the conclusions, some improvements are proposed to make the SG of Generation 2⁺ more complied with the ALARA principle.

Analysis of Reactor Core Positioning Method for PWR Manipulator Crane

Liu Yiqing, Ma Lei, Yan Tingyu

2015, 36(4): 111-114. doi: 10.13832/j.jnpe.2015.04.0111

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Manipulator crane is one of the key equipments which belong to the fuel handling and storage system（PMC system） in PWRs. In-core positioning is a prerequisite for the first fuel loading, and the security of reactor refueling. Single point positioning method is the prevailing method of reactor core positioning in the domestic PWRs. During the core positioning test with the single point positioning method, the problem that the dummy would bump against the lower internals appeared in a nuclear power plant. Through the innovative core positioning method—four points centre and average step length method, this problem was solved successfully. This paper analyzes the advantages and disadvantages of the single point positioning method and the four points centre and average step length method.

Study on Improvement of Operation Parameters at Cement Solidification System

Yan Wenchao, Huang Wentao, Ceng Bin, Zhang Jingsong, Chen Yunming, Li Xingyi, Hong Yongxia

2015, 36(4): 115-117. doi: 10.13832/j.jnpe.2015.04.0115

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Cement solidification is one of the most mature methods to deal with the radioactive wastes. Its production efficiency is subject to the solidification curing time. The optimization of parameter test and solidification performance test are carried out to determine the best optimum operation parameters at cement solidification system. The production capacity was increased from 5 barrels per day to 10 barrels per day on the basis of solidification performance meeting the national standard requirements in China.

Calculations and Simulation of Ultrasonic Inaccessible Zones for Outlet to Shell Weld in RPV

Hong Maocheng, Yu Zhe, Lin Ge, Xiao Xuezhu, Ma Guanbing

2015, 36(4): 118-121. doi: 10.13832/j.jnpe.2015.04.0118

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Based on the saddle-shaped surface feature of Outlet to Shell weld in RPV, this paper firstly constructs the universal equations for space curves within the ultrasonic needs by ASME and RSE-M codes. Afterwards, it deduces the left-side and right-side boundary calculations of inaccessible zones, which depending on the beam incident angle. Using the Pro/E software to build the 3D models, it can evaluate the quantitative properties of inaccessible zones by incident angle and section angle.

Research on Judgement Method for Fuel Element Damages in HFETR

Chen Qibing, Li Ziyan, Yu Dejun

2015, 36(4): 122-124. doi: 10.13832/j.jnpe.2015.04.0122

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HFETR operation experience shows that, defects exist when the nuclide trend analysis method is adopted in the analysis of fuel element damage, and the release of nuclear fuel cladding affects the accuracy of the nuclide trend analysis method. Based on the operation experiences and taking 131 I as an example, the relationship of the normal corrosion products and fission products for the HFETR sleeve-type fuel elements is analyzed, and the damage trend for the fuel element is studied. The trend K value method and radionuclide area determination method are proposed, and the accuracy of the two analytical methods is defined based on operational data, to ensure the safe operation of the reactor.

Maintenance Strategy Optimization Analysis for Circulating Water Pump in Daya Bay Nuclear Power Plant

Zhang Sheng, Wu Tao, Mo Chunni

2015, 36(4): 125-129. doi: 10.13832/j.jnpe.2015.04.0125

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This paper uses Reliability-centered Maintenance（RCM） method to optimize the maintenance strategy of circulating water pump in Daya Bay nuclear power plant. The overhaul cycle of circulating water pump is given in the paper, and the cycle is far more than the proposed cycle（ 8 years） of vendor and the cycle of similar equipment at home and abroad, not only greatly reducing the maintenance costs, but also decreasing maintenance-induced failures, thus further improving the equipment reliability and availability.

Improvement of Replacement Scheme for RCP No.1 Cartridge Flange Screw in CPR1000 Nuclear Power Station

Wang Yuxu, Yue Kai, Su Bin

2015, 36(4): 130-132. doi: 10.13832/j.jnpe.2015.04.0130

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This paper briefly introduces the replacement scheme of 16 flange bolts in No.1 sealed chamber of CPR1000 nuclear power plant reactor coolant pump（RCP）, which should be replaced by spare parts during the installation and operation of the construction for some reason. The innovation scheme of 4×4 replacement is used for the disassembly and inspection of 16 flange bolts, instead of the traditional whole replacement of 16 bolts. The innovation scheme reduces the work load in narrow space, optimizes the time schedule and avoids risks.

Analysis and Solution of Unable Tripping Fault of Refueling Machine Gripper

Peng Feng

2015, 36(4): 133-135. doi: 10.13832/j.jnpe.2015.04.0133

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Fuel assembly gripper driving circuit is a key part of the refueling machine. Its reliability has great significance for safety and stability of refueling operation. A number of tripping faults have occurred since the operation of the refueling machine No.1 and No.2 of Qinshan nuclear power plant II, and it has a certain impact on the refueling operation. With gripper driving circuit as the research object, considering the fault phenomenon and the maintenance experiences, the reasons of key components defect were analyzed in detail, the troubleshooting methods and solutions were summarized, and the measures to solve the problem was provided.

Safety Analysis of Spent Fuel Pool in Case of LOCA

Wang Haitao, Dan Jianqiang, Gou Junli, Zhang Bo

2015, 36(4): 136-139. doi: 10.13832/j.jnpe.2015.04.0136

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The Daya Bay spent fuel pool was chosen as the analysis object and the best estimate system thermal hydraulic code RELAP5/MOD3.3 was employed to investigate the processes in the spent fuel pool in case of LOCA. The results of calculations showed that, the increasing of water temperature in the pool from 52.6℃ up to 100℃ took 6.1 hours, and the evaporation of water volume above the SFAs took 16.6 hours. During this period, if some remedial measures were taken, such as taking the appropriate measures to cool the spent fuel or timely suppling water to the pool, it would be possible to avoid the uncovering of the SFAs.

Research on the Effect of Coupled Motions on Flow Instability Boundary

Ma Yingying, Qian Libo, Tian Wenxi, Su Guanghui, Qiu Suizheng, Huang Yanping, Yan Xiao

2015, 36(4): 140-144. doi: 10.13832/j.jnpe.2015.0140

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The effect of ocean conditions on coolant flow can be attributed to the additional force in the momentum equation. Based on the non-inertial reference frame momentum equation, the unstable boundaries in parallel channels under four kinds of coupled motions were obtained. It can be concluded that coupled motions has no effect on the flow instability boundary in parallel channels under forced circulation.

Study on Drop Performance Test of AP1000 Control Rod Drive Line

Gu Hanyang, Zhang Chaozhu, Chen Yuqing, Zhou Xiaojia, Liu Gang

2015, 36(4): 145-148. doi: 10.13832/j.jnpe.2015.04.0145

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A full scale of dive line system of AP1000 is employed to study the drop performance. The effects of mass flux both in transverse and longitudinal directions and dislocation or deformation in the components of dive line system on the dropping time are experimental studied. The time-travel curve is obtained using the high-speed photography technology. The experiment results show that the mass flux in transverse direction affects the dropping time significantly while the effects of other factors are negligible.

Safety Analysis of Loss of Cooling Accident for Typical Spent Fuel Pool

Zhang Zhongwei, Liang Guoxing

2015, 36(4): 149-153. doi: 10.13832/j.jnpe.2015.04.0149

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A model of spent fuel pool in nuclear power plants is developed by using the system code RELAP5/MOD3 and it simulates the hot water behavior of the whole spent fuel pool in the present paper. The nodalization in the model is based on the cycles of spent fuels. Based on the model, normal operation of the spent fuel pool and its cooling system is simulated. The spent fuel pool loss of cooling accident is simulated after that. The effects of radiation heat transfer mode, spray before leakage of radioactive material and the nodalization of hot channel and bypass outside of rack on the temperature response are analyzed. It is shown in the calculation that the model is able to simulate the normal operation of the spent fuel pool and its cooling system and succeeds in establishing the mechanism of natural convection cooling. The fuel uncover time is 17.87 days and it takes 19.14 days for fuel PCT to exceed 1204℃. The 12.6 kg/s spray cooling takes 2.4 hour to reduce the temperature of the fuels from 726.9℃ to 100℃.

Study on Engineering Development for Monte Carlo Burnup Code

Qiang Shenglong, Yin Qiang, Liu Cong, Yao Dong, Song Danrong, Lu Wei, Liu Dong

2015, 36(4): 154-157. doi: 10.13832/j.jnpe.2015.04.0154

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This paper analyzes the basic needs of the burnup codes in reactor design, and points out that the related functions in the traditional theory of implementation does not apply to Monte Carlo burnup code. Therefore, according to the characteristics of exact combination geometry and complex depletion chain, mixed burnup mode used in fuel and burnable poison depletion, more accurate body movement used in critical rod position searching and restart function under complex depletion chain have been developed for the Monte Carlo burnup code MOI, which would be preliminary feasibility for engineering application.

Numerical Simulation of Mixing Turbulent Flow in Rod Bundle with Spacer Grids

Wei Zonglan, Zhang Yu, Liu Songtao, Ma Qiang

2015, 36(4): 158-162. doi: 10.13832/j.jnpe.2015.04.0158

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Abstract:
The purpose of the present paper is to evaluate the validity of CFD, especially Large Eddy Simulation（LES）, to simulate the OECD/NEA benchmark experiment MATi S-H. The 5×5 rod bundle with a split-type spacer grid is simulated by both steady and transient simulation and the numerical results of time averaged and root-mean-square velocity field are compared to the experimental data. It is demonstrated that LES can better describe the mixing turbulent flow velocity field than the RANS method. The mesh sensitivity analysis further reveals that LES with fine mesh is more appropriate for analyzing the turbulent flow field in the rod bundle with spacer grids. LES results help understand the mixing turbulence characteristics with the mixing effect evaluation in transient. The simulated flow field data can also be of value for mechanical evaluations such as flow-induced vibration and grid-to-rod-fretting phenomenon.

2015 Vol. 36, No. 4