核动力工程

Study on 2D Arbitrary Geometry Coupling Resonance Method

He Lei, Wu Hongchun, Cao Liangzhi

2014, 35(6): 1-5. doi: 10.13832/j.jnpe.2014.06.0001

Abstract(9) PDF(0)

Abstract:
The paper firstly proposes a coupling resonance method in which subgroup method is employed in the serried peak energy region, and wavelet expansion method is employed in single peak energy region. The original subgroup model and wavelet expansion model are improved and coupled through the calculation of scattering source from subgroup to wavelet expansion, so that the self-shielding cross section in the whole energy region can be calculated accurately. To verify these theories and to prove the improvements,a PWR cell benchmark problem is calculated. It is demonstrated that, compared with other traditional multi-group resonance methods and continuous energy resonance method, this coupling resonance method has the ability to accurately calculate the whole energy region’s self-shielding cross section while keeping enough efficiency and finally has an ability to offer the accurate self-shielding parameters for latter transport calculation.

Research on High-order Perturbation Calculation Method with Perturbed Source Effects

Li Zeguang, Wang Kan, DenG Jingkang

2014, 35(6): 6-10. doi: 10.13832/j.jnpe.2014.06.0006

Abstract(9) PDF(0)

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Perturbation calculation can be used to estimate the changes of effective multiplication factor caused by perturbed system parameters, and is the fundamental of reactor sensitivity and uncertainty analysis.With the development of computer techniques and the increasing needs of new concept and complex reactoranalysis, Monte Carlo perturbation calculation is being used more and more widely. Traditional Monte Carlo perturbation calculation methods, which are limited to low-order perturbation and not considering of perturbed source effects, could cause large errors in perturbation calculations. To overcome the shortages of traditional methods, the high-order differential operator method with perturbed source effects is research in this paper, and this method is implemented in the code RMC. Numerical tests show that the perturbed source effects are very important in perturbation calculation, and large errors will be caused if the perturbed source effects are not considered in the calculation. Also, from the results, we can notice, with higher order, the perturbation results are more accurate.

Source Term of Xi’an Pulsed Reactor in Extreme Severe Accident

Yang Ning, Tang Xiuhuan, Zhang Wenshou, Yuan Jianxin

2014, 35(6): 11-16. doi: 10.13832/j.jnpe.2014.06.0011

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Inventories of radioactivity in Xi’an Pulsed Reactor（XAPR） are calculated, with the assumption that XAPR is continuously or intermittently operated at full power of 2MW until the fuel rods archive the design burnup. With a conservative releasing model, the biggest inventory 4.13×1016 Bq was selected to evaluate the source term of XAPR in an extreme severe accident, in which all fuel rods are supposed to be ruptured. The result shows that up to 41.0% of radioactive fission products is released in 1minute, the release lasts for about 5 minutes, and the total radioactivity released to environment is 4.54×1012Bq.

Experimental Investigation on Heat Transfer Characteristics of Subcritical Pressure Water Flowing in Subchannel with Square Distribution in SCWR

Xu Weihui, Wang Weishu, Liang Chengsheng, Lu Tong, Wang Han, Wang Linchuan, Bi Qincheng

2014, 35(6): 17-20. doi: 10.13832/j.jnpe.2014.06.0017

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Abstract:
Within the pressure range from 11 to 19 MPa, mass velocity from 700 to 1300 kg/（m²·s）, and inner wall heat flux from 200 to 1000 k W/m², experiments were performed to investigate the heat transfer characteristics of water flowing in subchannel with square distribution in SCWR. The subchannel was formed according to the fuel rods of the diameter 8 mm, and the pitch-to-diameter ratio was 1.2. The results show that the heat transfer characteristics of subcritical pressure water flowing in the subchannel with square distribution are significantly affected by the inner wall heat flux. Heat transfer deterioration more likely happens with higher inner wall heat flux. Heat transfer deterioration is more likely happened under low mass velocity, while it has little influence on the heat transfer characteristics at high mass velocity. The pressure has significant effects on the heat transfer characteristics. The higher pressure, the more likely the heat transfer deterioration occurred. The lower critical steam quality and the widely enthalpy region where heat transfer deterioration happen.

Experimental Study on Two-Phase Flow Characteristics in Pebble-Bed Porous Media Channel

Li Hua, Chen Ping, Qiu Suizheng, Su Guanghui, Tian Wenxi, Li Yun, Su Min

2014, 35(6): 21-25. doi: 10.13832/j.jnpe.2014.06.0021

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The experimental apparatus was designed and made up of isotropous particles（pebble） in porous media. All the pebbles were unordered accumulation. The experimental research on the flow characteristics of nitrogen-water two-phase flow in channel was conducted. The porous media channels were consisted of the polymethyl methacrylate（PMMA） circular pipe which diameters are 60, 70 mm and stainless steel ball which diameters are 2, 4, 6, and 8 mm. The experimental results show that when the liquid flow is constant, the pressure drop between the inlet and outlet of the test section will lower with the increasing of the inner diameter of the test pipe, but higher by decreasing the diameter of ball or accelerating the nitrogen flow.When the inner diameter of the test pipe is invariable, increasing the diameter of ball can decrease the pressure drop between inlet and outlet of the test section. The expressions of pressure drop were obtained under this experimental condition by fitting.

Verification and Analysis of CSR1000 Subchannel Code with Wire Wrap Model

Zang Jinguang, Du Daiquan, Yan Xiao, Huang Shanfang, Huang Yanping, Yu Junchong

2014, 35(6): 26-30. doi: 10.13832/j.jnpe.2014.06.0026

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Abstract:
Wire wrapped rod bundles were generally adopted by SCWR fuel assembly design. The wire wraps may have complicated impact on the thermal hydraulic behaviors of the rod bundle. In this paper, some improvements have been made on the geometry identification of the wire wrap model based on the subchannel code ATHAS and the results were compared with the CFD code. The improved subchannel code can generally reflect the thermal hydraulic behaviors of each subchannel and make reasonable response to the variation of geometry factors. The improved subchannel code is physically reliable, however, more work needs to be done to improve its accuracy.

Analysis of a Passive Containment Cooling System for NPPs

Huang Zheng

2014, 35(6): 31-36. doi: 10.13832/j.jnpe.2014.06.0031

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By using RELAP5 and MELCOR, transient behaviors of containment and PCS were investigated. The Influences of various parameters, such as pressure, condenser area and loop height were also studied. The results show that the heat in the containment can be effectively removed by PCS in a period of accident without spray. However the tank water should be made up and cooled for the long run. The heat transfer power of two-phase flow case is always higher, and varies more substantially with pressure, than that of one-phase flow case under the same boundary temperatures. However, flow instability may occur for two-phase flow case. By increasing the condenser area, the heat transfer is enhanced. The influence is less insignificant when the tank temperature is relatively high. It is also found that for both one-phase and two-phase flow, the change of loop height has little effect upon the heat transfer power under the given boundary condition in this study.

Calculation Method for Vapor Flow in Space Nuclear Reactor Heat Pipe

Li Huaqi, Jiang Xinbiao, Chen Lixin, Yang Ning, Hu Pan, Ma Tengyue, Zhang Liang

2014, 35(6): 37-40. doi: 10.13832/j.jnpe.2014.06.0037

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Abstract:
Heat pipe has been widely used to cool the space nuclear reactor core. Flow in the reactor heat pipe, which is characterized by the evaporation and condensation of its inner fluid, is remarkably different from the tube flow, thus the variable mass flow, axis velocity and radius velocity must be taken in account in its calculation. In this paper, a study was made on the calculation method of reactor heat pipe vapor flow’s pressure, temperature and velocity profile, and a computer code SNPS-HPD was developed to calculate the space nuclear reactor heat pipe vapor flow. The reliability of the code SNPS-HPD was verified by the data from sodium heat pipe experiment first. It is also proved to be well in a check calculation of, by comparing the result with that from reference, the vapor flow in the HP-STMCs Space nuclear reactor high temperature lithium heat pipe under various operating conditions. All that approved the employment of the SNPS-HPD code in the space nuclear reactor heat pipe design.

Effect of Sintering Temperature and Thorium Content on Density of （Th,U）O₂ Pellets

Long Dijun, Liu Jinhong, Zhu Jinhai, Li Jia, Wang Xinjie

2014, 35(6): 41-44. doi: 10.13832/j.jnpe.2014.06.0041

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This paper studies the effects of sintering temperature and thorium content On the density of（Th,U）O₂ pellets. The activation energy for sintering（Th,U）O₂ pellets were calculated, the movement and transplantation were analyzed by using Scanning Electron Microscope（SEM）. The results show that the density of（Th,U）O₂ pellets decrease as the proportion of thorium increase when sintering temperature is constant. The density of pellets increase as the sintering temperature increase. The activation energy for sintering（Th,U）O₂ pellets increase as the proportion of thorium increase. When the proportion of thorium were 20%, 50%,80%, the activation energy for sintering（Th,U）O₂ pellets were 277.65 k J/mol, 300.70 k J/mol,380.99 k J/mol respectively. The pores in the（Th,U）O₂ pellets which the proportion of thoriumwas 20% were mainly at the junction of grain boundary, and most of the pores were spherical.

Preparation of （Th,U）O_2+x Powder via Co-Precipitation Technique

Long Dijun, Lu Zhangxian, Zhang Yong, Liu Jinhong

2014, 35(6): 45-47. doi: 10.13832/j.jnpe.2014.06.0045

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The co-precipitate of Thorium and Uranium（AUTh C） was prepared by co-precipitation technique, the（Th,U）O₂+x powder was obtained by the decomposition and reduction of the AUTh C. The decomposition process of AUTh C was analyze by DTA-TG, the specific surface area of（Th,U）O₂+x powder was measured by BET method. The parameters which may effect the（Th,U）O₂+x powder properties were investigated, such as p H value, ammonia and ethanol. The result shows that the precipitation ratio was higher when the p H value was 5.5. Adding ammonia to make the p H value exceed 7.5 could improve the co-precipitation process. In the filtrate, the concentration of Uranium was 86 mg/L, the concentration of Thorium was 105 mg/L. Ethanol can prevent（Th,U）O₂+x powder grows hard-agglomeration in decomposition process, and the specific surface area of（Th,U）O₂+x could get up to 15 m²/g on the condition that the AUTh C decomposed at 600℃ for 2 h.

Master Curve Characterization of Fracture Toughness in Chinese RPV Steels Using PCVN and 1/2 PCVN Specimens

Peng Xiao, Wang Rongshan, Huang Ping, Li Chengliang, Yu Weiwei, Liu Xiangbing

2014, 35(6): 48-52. doi: 10.13832/j.jnpe.2014.06.0048

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Single temperature evaluation and multi-temperature analysis about the Pre-Cracked Charpy Specimens PCVN and 1/2 PCVN specimens of Chinese RPV steels were done. Both of the reference temperature T₀ of PCVN and 1/2 PCVN specimens were calculated. The results were corrected by the size criterion constant. The results showed that the size criterion constant equal 100 was appropriate for the application.

Analysis of Radiation Sources ＆ Verifying of Radiation Level for Irradiated Surveillance Specimens of WWER1000 Reactor Vessel

Chen Hao, Sun Kaibin, Wang Zhibing, Ding Zhanglong

2014, 35(6): 53-56. doi: 10.13832/j.jnpe.2014.06.0053

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The irradiated surveillance specimens are welted on the inner surface of WWER1000 reactor vessel. The specimens are withdrawed by melting and cutting the welding joint. The radiation sources and radiation level is analyzed in order to evaluate the radiation risks and choose the radiation protection measures.The radiation sources analytical method is verified, which is concise, good and safe in practice.

Key Code R＆D of LBB Design for Pipeline in Reactors

Li Pengzhou, Qiao Hongwei, Sun Lei

2014, 35(6): 57-60. doi: 10.13832/j.jnpe.2014.06.0057

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The Leak Before Break（LBB） concept, as an important feature of the third generation nuclear power technology, is widely used in the high-energy piping design in nuclear industry. However, for some historical reasons, the application of LBB concept in China lagged behind compared with that in the developed countries. Up to now, there are no credible codes developed by China which can be applied to an actual project, and the relevant design mainly relies on foreign companies. Hence, the development of an approved LBB design code has great theoretical and practical importance for China. In this paper, the background of the fracture mechanics analysis, Crack Opening Displacement（COD） calculation and leak rate calculation in LBB design are briefly introduced firstly, and then the R＆D situation of the key code in NPIC is also presented, and some examples from the approved codes and published papers are used to verify the self-developed code. The calculation results show that the accuracy of the code is in well consistence with the examples. This code can be used in the practical engineering after refinement and validation.

Flow Induced Vibration of RPV Upper Internals Support Columns

Xu Xiao, Ma Ruoqun, Zhang Guihe, Wang Dasheng

2014, 35(6): 61-65. doi: 10.13832/j.jnpe.2014.06.0061

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The Reactor Pressure Vessel（RPV） upper internals support columns are subjected to high speed liquid impact in the cross flow field during the operation conditions at nuclear power plants. With the cross flow impact, the vibratory behavior of the RPV upper internals support columns may occur. The vibratory phenomena studied are the vortex induced vibrations and the fluid elastic instability. According to ASME code, this paper studied the flow induced vibration（FIV） of RPV upper internals support columns, and verified the lock-in asynchronization, vibratory fatigue and fluid elastic instability. The study showed that compared with the full power operation condition, FIV is prone to occur in during the cold strart-up or cold shut-down transient.

Effect on Seismic Design of Primary System with Different Primary Reactor Coolant Pump

Ye Xianhui, Lan Bin, Zhang Yixiong, Liu Wenjin

2014, 35(6): 66-69. doi: 10.13832/j.jnpe.2014.06.0066

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The primary system seismic analysis was performed to study the influence on primary system seismic loads with different reactor coolant pump by using ANSYS software. The results show that the stresses evaluation of the primary pipe for two reactor coolant pump models are meet code requirement. Theaseismatic intensity of civil work at the position of the pump snubbers should be improved. The seismic loads based on the pump model in natural frequencies of system which are far away from the spectral peak are lower.

Model Analysis and Experimental Verification of Containment Spray Pump

Chu Qibao, Fang Yonggang, Xu Yu, Lu Yan

2014, 35(6): 70-72. doi: 10.13832/j.jnpe.2014.06.0070

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Modal analysis of the containment spray pump is carried out by using ANSYS program, and structural natural frequencies and modes of vibration can be obtained. After that, the dynamic characteristic test is performed for the containment spray pump. The comparison between modal analysis results and test results show that they agree well. It verifies the validity of the model and lays the foundation for the structural integrity analysis of the containment spray pump.

Analysis and Assessment of Reactor Beryllium Component Stress

Su Min, Feng Linna, Li Yuanming, Lei Tao, Gu Mingfei, Gao Lijun

2014, 35(6): 73-76. doi: 10.13832/j.jnpe.2014.06.0073

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Based on the mechanical and irradiation properties of beryllium, the irradiated beryllium component stress analysis was proposed, in which the mechanical stress, thermal stress and irradiation swelling induced stress and the load combination in the operation conditions are considered,, and also the maximum tensile stress theory was suggested to use in the assessment. In this paper, a stress analysis and assessment was made for the beryllium tube, which is the key component of HFETR beryllium assembly, by ABAOUS. The result showed that the maximum principle stress was much less than the tensile stress, and that the beryllium assembly can be still used for a long time from the perspective of stress rupture.

Analysis of Diversity and Independence for ATWT Mitigation System in Nuclear Power Plant

Zhang Yunbo, Zhang Mi, Huang Weijie, Mao Congji, Li Shixin, Yin Baojuan

2014, 35(6): 77-79. doi: 10.13832/j.jnpe.2014.06.0077

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It introduces the Anticipated Transisents Without Trip（ATWT） in this article,and analyzes the diversity and independence of ATWT mitigation system, and discusses the problems faced in the design of ATWT on the basis of relevant law and regulations.

Partition of Emergency Planning Zone in Marine Reactors

Wang Wei, Chen Lisheng, Zhang Fan, Zhao Xinwen

2014, 35(6): 80-83. doi: 10.13832/j.jnpe.2014.06.0080

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Making use of MELCOR program, the blackout accident of small reactor is simulated and analyzed in the paper, the result of which is provided to the atmospheric dispersion software MACCS, and the radioactive consequence is calculated and analyzed in a coastal area. Combining with the criteria for the partition of emergency planning zone in marine reactors, the size of the emergency planning zone which is suited for small reactor is captured through the calculation.

Bayesian Method of PSA Generic Data Processing Based on Jeffreys Prior

Shen Zhiyuan, Chen Wei, Yuan Jianxin, Tang Xiuhuan, Yang Jian

2014, 35(6): 84-87. doi: 10.13832/j.jnpe.2014.06.0084

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Jeffreys priors for gamma and beta distribution was derived after adopting Jeffreys prior theory.Furthermore, the expressions were given to estimate the hyper parameters of generic data distribution using B ayesian approach. Compared to the results using classical statistics theory, the data processing method based o n Jeffreys prior appeared much simpler and more concentrated, which reserved most sample information.

Research on dΦ/dt in Protection Channel of CPR1000 NPP

Hu Yousen, Zhou Sheng, Xi Yanyan, Zhang Haoyun

2014, 35(6): 88-91. doi: 10.13832/j.jnpe.2014.06.0088

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This paper describes the methodology and parameter definition of dΦ/dt correction channel of CPR1000. These correction parameters have been studied essentially by combining with real nuclear power plant design parameters and using best-estimated transient analysis code CATIA2. According to comparison with house-load test data from site, these correction parameters have been validated. This analysis process could also be used for the coming new CPR1000 projects.

Analysis of Minimum Gap for U-bends of Steam Generator Tube Bundle

Cui Suwen, Han Tongxing, Mo Shaojia

2014, 35(6): 92-95. doi: 10.13832/j.jnpe.2014.06.0092

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During the fabrication and tubing of the tube bundle for steam generator, some local deformations in the U-bend area might occur, which will result in a smaller gap or even contact between tubes. According to the design principles for tube bundle, taking CPR1000 steam generator（Type 55/19B） as the analysis model, the gap for the U-bends is analyzed and evaluated from the views of flow-induced vibration（FIV）, wear, dry-out and stress analysis. Based on the analysis results, it is concluded that the minimum gap for the U-bends should be larger than the turbulent amplitudes to avoid unacceptable flow-induced vibration, wear and dry-out which means enough safety margins should be considered during U-bends gap design.

Electrical and Control Design for Reactor Coolant Pumps in AP1000 Nuclear Power Plant

Han Yong, Liu Feiyang, Liu Wenjing, Gao Yong

2014, 35(6): 96-99. doi: 10.13832/j.jnpe.2014.06.0096

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The Reactor Coolant Pumps（RCP） of AP1000 nuclear power plant are the canned motor pump. Because of the restriction of the RCP motor design parameter and operation mode, the Variable Frequency Drive（VFD） shall be used to meet the operation and technical requirements. This paper studies the power supply mode, MV VFD technical and control logic of the RCP, so that the key technology of the AP1000 nuclear power plant can be totally caught and applied to the independent design of the third generation nuclear plants.

Design of Automatic Measurement System for Plate-Type Fuel Element Burn-up

Zhang Mao, He Chaoming, Cheng Peipei

2014, 35(6): 100-105. doi: 10.13832/j.jnpe.2014.06.0100

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In the production and use of radioactive fuel process, it is an important process to measure the burn-up level for fuel element in accurate and efficient way. At present, the automatic operation of burn-up measurement of plate-type fuel element is unimplemented, and it applies only to single-point measurement.According to the structure of plate-type fuel element and the measurement requirements, the design of a new system which the interactive measurement point planning, motion control system, the gamma-ray energy spectrometer control, data processing and result visualization were integrated. It can realize multi-point,segmentation, automatic, nondestructive burn-up measurement of plate-type fuel element. The operation results show that the system is with high reliability, and can improve the efficiency and accuracy of measurement, thus to reduce the workload of the users.

Analysis of Calculation Method for Reactor Control Rod Drop Time

Liu Yanwu, Huang Bingchen, Ran Xiaobing, Yu Xiaolei

2014, 35(6): 106-110. doi: 10.13832/j.jnpe.2014.06.0106

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The reactor control rod drop time is an important parameter for the safety analysis and significant to the power plant safe operation. In order to get the control rod drop time history, in this paper, the kinetics equation and calculation Method is established. The complicated fluid and mechanical friction forces during rod drop course are described and analyzed in detail, and the drop time is deduced. The comparison of the calculation results and relative validation tests showed that the calculation result is conservative and reliable.

A Preliminary Study on Qualification of Instrumentation and Control System for Nuclear Power Plants

Huang Weijie, Zhang Mi, Zhang Yunbo, YiN Baojuan, Mao Congji

2014, 35(6): 111-114. doi: 10.13832/j.jnpe.2014.06.0111

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The standard architecture and overall requirements of the equipment qualification for Class1 E digital instrumentation and control（I＆C） system are discussed. Based on the comparison of NUREG 0800 and RCC-E technical specification, the technique of equipment qualification for Class 1E digital I＆C system suitable to the condition in China is analyzed, and general qualification standard architecture and method are studied, and then the corresponding acceptance criteria are proposed.

Feasibility Study on Acoustic Emmission Technology Monitoring for New Pattern Steam Generator

Yang Lei, Li Shuliang, Yi Qiaoling, Xiong Jing

2014, 35(6): 115-119. doi: 10.13832/j.jnpe.2014.06.0115

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Aiming at the titanium alloy of new pattern steam generator unit, this paper proposed the use of AE technology to monitor the material performance during the manufactory. Using the AE monitoring system to monitor the whole process of the elongate test of standard tensile sample and analyze the obtained data, the AE characteristics of unit subassembly material and sensitivity of AE monitoring system is obtained.The validity of analytic result of standard sample is verified by the result from the tension test on tube-board tensile sample, and thus the feasibility of the use of AE technology to monitor steam generator unit subassembly on hydraulic pressure experimentation is affirmed.

Research on Defect Location of Steam Generator Bilateral Symmetry Welding

Yang Lei, Wang Zhe, WEi Wenchen

2014, 35(6): 120-121. doi: 10.13832/j.jnpe.2014.06.0120

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This research is aimed to defect location problem of X-Ray test technology on a new pattern steam generator unit bilateral welding, proposed a scheme that used vertical and tilt Radiographic inspection technology to locate disfigurement and affirmed validity of searching scheme by examination. The technology provided guarantee of product quality and improved production efficiency.

Research of ETA Water Chemistry Treatment Technology in Secondary Circuit of PWR Plant

Shen Jun

2014, 35(6): 122-125. doi: 10.13832/j.jnpe.2014.06.0122

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After using ETA as the p H adjustment agent in secondary Circuit of Qinshan 320 MWe PWR plant in place of ammonia, the p H of Moisture Separator Reheater（MSR） drainage and Stream Generator（SG）blowdown water are obviously increased, so that the corrosion conditions are improved, such as: the corrosion products of the equipment in the water phase decreased, flow accelerate corrosion is restrained and the transfer of corrosion products to the secondary side of the SG is reduced. As more period water output are produced from condensate mixed bed because of ETA applied in the secondary system, the regeneration times and acid, alkali and water consumption when regenerating resins are reduced, so the workload of operation workers can be alleviated and the environment pollution from the waste water are reduced either.

Functionality Recovering and Failure Analysis of Rod Cluster Assembly Protective Jacket in TNPS

Zhang Yuan, Zhou Li, ZhenG Haiquan

2014, 35(6): 126-129. doi: 10.13832/j.jnpe.2014.06.0126

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This paper introduces the function of TNPS refueling machine, the structure of rod cluster assembly protective jacket, and the condition of functionality recovering. It also describes some problems found when fuel assemblies and rod cluster assemblies are rearranged and replaced during refueling outage, as well as the analysis and processing of problems. In addition, it investigates the general causes of problems by disassembling inspection on rod cluster assembly protective jacket.

Analysis on Air Inlet Problem of Passive Containment Cooling Piping

Chen Shushan, Zhang Jiang, Zhu Linglin, Wang Liangliang, He Junshan

2014, 35(6): 130-134. doi: 10.13832/j.jnpe.2014.06.0130

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For analyzing the hydraulic performance and air inlet problem of passive containment cooling system（PCS）, the software FLOWMASTER is used to establish an one-dimensional hydraulic model, and using it, the PCCWST water supply flow rate for 72 hours after accident is calculated; The two-dimensional FLUENT model is used to analyze the air inlet phenomenon. The result indicates that the air will be drawn to the water supply piping after the highest vertical pipe uncovered by water, which creates disadvantage to the water supply piping. And the actual cooling water supply flow rate is slightly less than the flow rate without considering the air inlet.

Research of Grounding Capacitive Current of Neutral Non-Grounding Auxiliary System in Nuclear Power Plants

Yang Shan, Liu Li, Huang Xiaojing

2014, 35(6): 135-138. doi: 10.13832/j.jnpe.2014.06.0135

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In the domestic and abroad standards, the grounding capacitive current limitation in the non-grounding electric auxiliary system is less than 10 A. Limiting capacitive current in the standard aims to speed up the arc extinguishing, to reduce the duration of arc over-voltage, but not to prevent the arc producing. The arc over-voltage harm is related to the multiple, frequency and duration of the over-voltage.When the insulation vulnerabilities appear in the equipment, the arc over-voltage may result in insulation vulnerabilities of the electrical equipment breakdown, which leads to multiple short-circuit accidents. The cable connector, accessory and electromotor winding are all insulation vulnerabilities. Setting the arc suppression coil which can counteract the grounding capacitive current makes the arc vanish quickly. Using the casting bus which remarkably reduces the ground capacitance of the electric transmission line makes the equipment safer.

Application Study on Improved Wavelet Analysis Algorithm for Pump Rotor Fault Diagnosis

Chen Zhihui, Min Yuansheng, Li Yi, Xia Hong, Deng Liping, Huang Wei, Huang Hua

2014, 35(6): 139-143. doi: 10.13832/j.jnpe.2014.06.0139

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Wavelet Analysis is a new signal analysis theory, but in the way of the engineering application,the wavelet analysis has the problems such as frequency confusion and amplitude frequency offset distortion.This study makes the use of the frequency compensation, an improved algorithm of wavelet packet, and selects the test data of pump rotor typical faults, i.e., dynamic unbalance, rotor bow and dynamic unbalance. It has been proved that the measure of frequency compensation can recognize the fault of pump rotor effectively.This study shows that the improved algorithm can effectively solve the problems such as amplitude frequency,and confuse and distortion offset. This study established a certain foundation for application of wavelet analysis.

3-D Steady Analysis of Flow in CRDM Sewerage System

Sun Yan, Liang Tiebo, Chen Zhihui, Zhao Jing, Zhang Yulong

2014, 35(6): 144-147. doi: 10.13832/j.jnpe.2014.06.0144

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In order to obtain the flow state during sewer condition in Reactor and CRDM Sewerage system（RSE）, this paper analyzes the 3-D steady flow in RSE by using Computational Fluid Dynamics（CFD） method. In the premise that the pressure drop of the RSE is known, the mass flow rate, the velocity and the type of flow in the system is obtained with the inverse method, which is proposed and validated to be applicable in the paper. The result shows that in the sewerage conditions, the type of flow in the RSE is turbulence flow, which is helpful to sewer drain. The study results give an reference for the design of RSE.

Accessorial Software for Reliability Analysis of Reactor Protection System

Wang Chao, Leng Shan, Chang Qing, Chen Weihua, Jiang Hui, Sun Wei, Guo Zhiwu

2014, 35(6): 148-152. doi: 10.13832/j.jnpe.2014.06.0148

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Based on the reactor protection system of a certain type of digital I＆C system for a nuclear power plant, this paper analyzes its structure and functions, and puts forward the requirements of the graphic automatic modeling. It develops the dedicated accessorial software for reliability analysis based on the requirements. This software can realize the graphical configuration of reactor protection system in the reliability analysis process for safety functions,and it can create, modify and save the configuration figures. The software also has the functions of automatic analysis of configuration scheme and the automatic generation and output of fault tree information and etc.

Application of Lagrangian Particle Tracking Method in Study of Steam Generator Sludge Collector

Wu Ge, Cheng Xiang, Huang Wei, Sun Yan

2014, 35(6): 153-157. doi: 10.13832/j.jnpe.2014.06.0153

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The flow conditions and the tracks of sludge particles around the sludge collector in a typical U-tube recirculating steam generator are studied by a Computational Fluid Dynamics method. To evaluate the main factors affecting the deposition process of sludge particles, a Lagrangian Particle Tracking scheme is employed to account for the momentum transfer between the continuous phase of fluid and the discrete phase of particles. It shows that the performance of a sludge collector depends heavily on the transportation effects of flow field outside the sludge collector. The results for different particle diameters indicate that the particles with a larger diameter are more likely to be captured by the sludge collector. By increasing the inlet and outlet area on the top plate, the sludge collector performance can be obviously elevated with a limit of avoiding particle re-entrainment phenomenon.

Effect of Calculation Average Probability Failure on Common Cause Failure

He Li, Chen Jie, Zhou Jixiang, Xiao Peng

2014, 35(6): 158-161. doi: 10.13832/j.jnpe.2014.06.0158

Abstract(6) PDF(0)

Abstract:
In order to analyze the effect of common cause failure（CCF） on average probability of failure on demand（P_avg） calculation, fault tree models of a normal excitation 1oo2 system are established. The P_avg without regard to CCF and P_avg with regard to CCF（use β factor model） are both calculated. By comparing the calculation results, find that CCF has a great impact on P_avg calculation. Correlation analysis shows that when the system is safer（dangerous failure rate is lower, the detect rate of dangerous failure is higher）, the impact is greater.

Flow Regime Transition of Accelerated （Decelerated） Flow in Rectangular Channel

Yuan Hongsheng, Tan Sichao, Zhuang Nailiang, Tang Linghong, Zhang Chuan, Zhang Hong

2014, 35(6): 162-166. doi: 10.13832/j.jnpe.2014.06.0162

Abstract(8) PDF(0)

Abstract:
Flow regime transition under constant pressure drop rate in rectangular channel is investigated by theoretical and experimental study. Laminar velocity distribution and the method to determine the critical Reynolds number is obtained by theoretical study. The factors that affect the critical Reynolds is analyzed.The results show that the critical Reynolds number of accelerated（decelerated） flow is smaller（larger） than that of the steady flow. The pressure drop rate has little effect on the critical Reynolds number while the starting steady part pressure drop does. The critical Reynolds number approaches to the steady condition value while the steady part pressure drop approaches to the steady condition value. The pressure drop rate hardly affects the critical Reynolds number.The experimental data supports the smaller（larger） critical Reynolds number in accelerated（decelerated） flow, but is not consistent with the effect of pressure drop rate.

Research on Application Characteristics of Ethanolamine Used in Secondary System of Nuclear Power Plants

Zhao Yongfu, Wang Jinfang, MA Weigang, Shen Jun, Jiang E, Gong Bin, Liu Jinhua

2014, 35(6): 167-171. doi: 10.13832/j.jnpe.2014.06.0167

Abstract(15) PDF(0)

Abstract:
The characteristics including dissociation, vapor-liquid distribution, thermal decomposition,compatibility with materials of ethanolamine（ETA） used in secondary system of nuclear power plant（NPP）were investigated by the laboratory research and full-scale testing at Nuclear Station. The results show that ETA does not only have strong base strength, low vapor-liquid distribution and good thermal stability, but also have an excellent compatibility with the secondary system materials. The application of ETA in the secondary system of NPPs can effectively decrease the concentration of ions in feed water and the amounts of slugs in steam generator（SG）, and would make great significance to optimize the secondary water chemistry as well as reduce the corrosion rate of secondary system materials.

Analysis on Sedimentation Rate of Sodium Aerosol In Emergency Ventilation System

Sun Dajie, Zhang Donghui, Ren Lixia, Hu Wenjun

2014, 35(6): 172-175. doi: 10.13832/j.jnpe.2014.06.0172

Abstract(10) PDF(0)

Abstract:
In the paper, the chemical and physical characteristics of sodium aerosol particles as well as factors that influence the sedimentation rate of sodium aerosol during its transport are analyzed. Besides, a simplified model is given to calculate the sedimentation rate when aerosol getting through the emergency ventilation system. It has been found that the sedimentation rate is largely determined by gravity, turbulent coagulation and turbulent, other factors, such as brownian diffusion, could be ignored.

Stress Corrosion Cracking Analysis for Manual Valves and Seal Assemblies Wrongly Degreased with Tetrachloroethylene

Zhang Ping, Huang Bingchen, Ran Xiaobing, DAi Zhangnian, Liu Yanwu, Zhang Mingqian

2014, 35(6): 176-179. doi: 10.13832/j.jnpe.2014.06.0176

Abstract(8) PDF(0)

Abstract:
RIC manual valves and seal assemblies made of austenitic stainless steel were wrongly degreased with tetrachloroethylene by the manufacturer. Chloride ions（Cl^-） were produced during the pyrolysis of tetrachloroethylene,leading to the risk of stress corrosion cracking（SCC） due to the remaining Cl^-. The research shows that the Cr-Ni austenitic stainless steel becomes more sensitive to the stress corrosion cracking as the Cl^-content increases. Increasing temperature and oxygen（O） content can accelerate the SCC of austenitic stainless steel caused by Cl^-. After the analysis of the remaining Cl^- content in the manual valves and seal assemblies, operating temperature and oxygen content in the coolant, it is considered that the remaining Cl^- will not cause SCC during service of manual valves and seal assemblies. In addition, the pyrolysis of tetrachloroethylene under vacuum and abrasive blasting of surfaces can degrease Cl^- observably.

Design and Application of DORP Platform

Tu Yichun

2014, 35(6): 180-183. doi: 10.13832/j.jnpe.2014.06.0180

Abstract(9) PDF(0)

Abstract:
This paper gives a detailed introduction about the development demand, development method,design details and implement process of the management application information platform（DORP）. As some unique ideas such as container, reuse, combination are adopted in the design of DORP, the development of the information system of this project either rely on any development tools, or require professional staff to do development and management work, and it can modify the contents and layout of the windows and setup management process online without disturbing the users. The application of DORP indicates that the demands on the development and management staff are dramatically reduced, the maintenance of the platform is easy,and meanwhile, a huge amount of fees on software and hardware was cut off.

2014 Vol. 35, No. 6