核动力工程

Experimental Study on Integral Performance of Passive Containment Cooling System

Chang Huajian, Yang Xiang, Zhou Mingzheng, Sun Liuli, Zhao Ruichang

2016, 37(6): 1-5. doi: 10.13832/j.jnpe.2016.06.0001

Abstract(28) PDF(0)

Abstract:
It is necessary to verify the new developed passive containment system, and the small-scale test verification is an effective methodology. Methodology used to design the integral facility of passive containment system, test vessel, auxiliary system, test condition are presented briefly. Non-dimensional numbers of pressure condition, gas exist, and shell condensation heat transfer are analyzed, which indicates the non-dimensional numbers are in the acceptable range. This proves that the adopted methods in the integral test of passive containment system are reasonable and effective, and the test results show that the facility can exactly model the phenomenon of the prototype.

Numerical Simulation of Low Temperature Environmental Heat Dissipation Process of Passive Containment Cooling System Water Storage Tank

Song Daiyong, Han Xu, Zhang Li, Zhao Jingxiong, Chang Meng, Tang Huapeng

2016, 37(6): 6-10. doi: 10.13832/j.jnpe.2016.06.0006

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Abstract:
Passive Containment Cooling System（PCS） of AP1000 provides the safety-related ultimate heat sink for the plant with high probability. An uncertain risk of PCS is the possibility of PCSWST being frozen when system runs during cold winter. This paper estimated the above risk by multiphysical simulation, and presented suggestions for improving the PCSWST design.

Sensitivity Study on Secondary Side Passive Residual Heat Removal System Based on RELAP5

Gong Houjun, Xi Zhao, Zan Yuanfeng, Zhuo Wenbin, Huang Yanping

2016, 37(6): 11-14. doi: 10.13832/j.jnpe.2016.06.0011

Abstract(26) PDF(0)

Abstract:
In this paper, sensitivity study of secondary side passive residual heat removal system on starting mode, heat transfer area and flow resistance has been conducted by RELAP5. The results show that both water-injection starting mode and liquid-column starting mode have little effect on the transient characteristics of secondary side passive residual heat removal system. Compared to the standard condition, at the condition of 37.5% heat transfer area, the discharge flow rate of the makeup tank is relatively increased, and the drain time for the makeup tank is relatively decreased. At the condition of double resistance, the discharge flow rate of the make-up tank decreases, while the drain time is increased by 1000 s.

Error Sensitive Analysis for Circular Pipe Heat Transfer Experiments with Supercritical Water

Zang Jinguang, Yan Xiao, Li Yongliang, Huang Zhigang, Huang Yanping

2016, 37(6): 15-17. doi: 10.13832/j.jnpe.2016.06.0015

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Abstract:
This paper presented that the strong variation of fluid properties in the pseudocritical region would amplify the error of elemental T/H parameters. Firstly, the T/H sensitive analysis of heat transfer coefficient was performed with a typical heat transfer correlation. Then, the wall temperature deviation of two symmetrical thermal couples was discussed in the circular pipe supercritical water heat transfer experiments. It suggests that the pseudocritical fluid properties may alter the circumferential uniform characteristics of the circular pipe and could be the reason for the scatter of the heat transfer correlations.

Numerical Investigation of Buoyancy Effect in Forced Convective Heat Transfer to Supercritical Carbon Dioxide Flowing in a Tube

Liu Shenghui, Huang Yanping, Liu Guangxu, Wang Junfeng, Zan Yuanfeng, Lang Xuemei

2016, 37(6): 18-22. doi: 10.13832/j.jnpe.2016.06.0018

Abstract(26) PDF(0)

Abstract:
Numerical investigation of buoyancy effect in forced convective heat transfer to supercritical carbon dioxide flowing in a vertical tube was carried out. When the mass flux is low and the wall heat flux is high, the buoyancy effect is obvious, which might redistribute the radial and axial velocity, even M shaped distribution occurred in the radial direction. When the zero-velocity gradient zoon corresponding to the M shaped velocity distribution arises in the edge of viscous layer, the production and convection of eddy will be weakened, resulting in the heat transfer deteriorated. According to the extended calculation based on experimental data, reducing the wall heat flux, adding the mass flux or raising the inlet temperature can relieve the deterioration of heat transfer caused by buoyancy effect in different level.

Experimental Study on Critical Heat Flux of Chemical Water Boiling on a Downward Facing Curved Surface for IVR-ERVC Strategy

Yang Sheng, Hu Teng, Lu Wei, Chang Huajian

2016, 37(6): 23-27. doi: 10.13832/j.jnpe.2016.06.0023

Abstract(26) PDF(0)

Abstract:
Based on a large scale test facility which adopting the SA508Ⅲ steel as heating surface material and utilizing the chemical solution with addition of boric acid（BA:H₃BO₅） or/and trisodium phosphate（TSP: Na₃PO₄） as coolant, the CHF characteristics of chemical solution boiling on a downward facing curved real reactor pressure vessel material surface were investigated experimentally. Experimental results indicate that the CHF values show different changes in different chemical solution environment for the SA508 III steel material. The CHF of BA decreases with the concentration of 1000 to 3000 mg/L and is lower than that of deionized water. The CHF of TSP with lower concentration of 500 and 1000mg/L is enhanced, while weakened for a high concentration of 3500 mg/L. The CHF of mixed solution of BA and TSP increases firstly and then decreases with the increasing of TSP concentration.

Characteristics of Natural Circulation Flow and CHF under External Reactor Vessel Cooling within Surface Saturated Conditions

Hu Qiang, Yan Xiao, Huang Shanfang, Huang Yanping, Yu Junchong

2016, 37(6): 28-32. doi: 10.13832/j.jnpe.2016.06.0028

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Abstract:
Based on the one-dimensional conservation equations, the numerical calculation model has been established for the analysis of the characteristics of natural circulation flow on the outer surface of reactor pressure vessel utilizing the net vapor generation model in low-flow subcooled boiling and drift flux model. Combined with the SULTAN correlation, a comprehensive study of the effects of heating power, flow channel clearance, loss coefficient of inlet area and flooding levels on the outer cooling process of lower head under in-vessel condition has been conducted in this study.

Extension and Validation of RELAP/SCDAPSIM/MOD4.0 Code on FHR

Jiang Shuying, Cheng Maosong, Dai Zhimin, Chen Yushuang

2016, 37(6): 33-36. doi: 10.13832/j.jnpe.2016.06.0033

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Abstract:
On the basis of RELAP/SCDAPSIM/MOD4.0 code, adding and implementing the thermophysical properties and heat transfer coefficients of FLi Na K into the source code, the RELAP5-FHR code is developed, which is suitable for the analyzing of the FHR system. Then the RELAP5-FHR code is validated by the experimental data of the FLi Na K high temperature molten salt test loop. The results indicate that the simulation results from the RELAP5-FHR code are in close agreement with the experimental data, which validates the applicability of RELAP5-FHR code for FHR.

Simulation of Bubble Growth Process in Boiling Flow Using Lattice Boltzmann Method

Zeng Jianbang, Li Longjian, Ma Jian, Huang Yanping, Wu Nengyou

2016, 37(6): 37-40. doi: 10.13832/j.jnpe.2016.06.0037

Abstract(29) PDF(0)

Abstract:
A numerical model for liquid-vapor phase change, which is established by the lattice Boltzmann method, is employed to investigate the effect of horizontal acceleration on the growing process of the bubble in the boiling flow with constant gravitational acceleration. It can be found that the bubble departure diameter decreases exponentially and the release frequency increases exponentially with the horizontal acceleration. Before the detachment of the bubble, it is clearly shown the greater the horizontal acceleration, the greater the upstream contact angle and the smaller the downstream contact angle, and both the upstream and downstream contact angles reach constant as the horizontal acceleration increases to a certain degree. After the detachment of the bubble, the bubble motion trajectory is very close to the down boundary when the horizontal acceleration is large.

Investigation on Bubble Breakup in a Venturi-Tube Bubble Generator under Low Gas Fraction Condition

Mo Zhengyu, Du Min, Sun Licheng, Zhang Haiyan, Shao Ziyi, Wen Juan

2016, 37(6): 41-44. doi: 10.13832/j.jnpe.2016.06.0041

Abstract(25) PDF(0)

Abstract:
In order to illustrate the mechanism and process of bubble breakup in a venturi-type equipment for generating bubbles, the transportation and breakup process of bubbles in a venturi-type bubble generator is analyzed in detail with the help of image and video processing softwares of PFV and Image-Pro Plus. It is found that two distinguished stages of bubbles exist in the generator; Bubbles will be decelerated rapidly at the entrance of the expansion section, during which three different kinds of deformation may occur to the bubbles. It is believed that the deceleration of bubbles has a significant effect on the bubble breakup process. The rapid deceleration of bubbles in the expansion section results in the increasing of the velocity difference and intensifies the interaction between the gas and liquid, leading to the bubbles with relatively large diameter breaking up into a great many tiny bubbles.

Seismic Isolation Design and Parameter Effect Analysis of AP1000 Nuclear Island Structure

Zhuang Chuli, Zhang Yongshan, Wang Dayang

2016, 37(6): 45-49. doi: 10.13832/j.jnpe.2016.06.0045

Abstract(28) PDF(0)

Abstract:
The finite element model of AP1000 nuclear island structure was built by ANSYS and Midas Gen software platform, and its three-dimensional seismic response was analyzed through nonlinear time history method. The seismic response and seismic isolation control of the nuclear island structure under safe shutdown earthquake is investigated. In addition, the isolation period, stiffness ratio and flexion weight ratio is selected as the control parameters of the seismic isolation layer, to analyze the seismic isolation layer and superstructure seismic response with different seismic isolation layer parameters. The results show that, the base seismic isolation can significantly reduce the seismic response of the nuclear island structure. The superstructure top horizontal acceleration and shear of seismic isolation layer remarkably decreases with the increasing of the isolation period. Besides, the effect of flexion weight ratio and stiffness ratio on the seismic response of isolated nuclear island structure declines with the increasing of the isolation period.

Dynamic Impact Test and Simulation Analysis of Anti-Whip Device in Nuclear Power Plants

Wang Chunlin, Zhao Jicheng, Liu Chengyi, Shi Zuowei

2016, 37(6): 50-53. doi: 10.13832/j.jnpe.2016.06.0050

Abstract(25) PDF(0)

Abstract:
The dynamic impact test of anti-whip device was completed by vehicle impact test in a nuclear power plant. The time-history curve of acceleration, velocity and displacement were obtained in particular impact energy. Also with the maximum impact force and deformation, the simulation analysis of dynamic impact test was taken with the LS-DYNA software, and the simulation results were in good agreement with the experimental results. Dynamic impact test and simulation analysis verified the safety of the anti-whip device in particular impact energy.

Design and Calculation of HVAC Support in EPR NPP

Luo Yang, Hu Bei

2016, 37(6): 54-57. doi: 10.13832/j.jnpe.2016.06.0054

Abstract(27) PDF(0)

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This paper analyzes the design of the HVAC support in EPR NPP. Compared with the M310 NPP, the EPR NPP improves the support structure. Statics mechanics formula and ANSYS software are used to calculate the space between two supports. RSTAB software is used to calculate the strength of the profile and define the size of the profile and bolt. Based on the analysis, the standard support structure and distances between the supports of the EPR NPP is defined, and the support steel specification and loads is listed.

FEM Seismic Analysis Method of Large Storage Tank in Consideration of Liquid Sloshing

Du Kun, Chu Qibao, Liang Mingbang, Shi Zuowei

2016, 37(6): 58-61. doi: 10.13832/j.jnpe.2016.06.0058

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Abstract:
Based on ASCE-4-98, the 3D FEM model of large storage tank was established. The effect of liquid sloshing was simulated by the mass-spring model. The seismic analysis was processed by ANSYS software. The FEM calculated frequency of liquid sloshing was identical to the formula result of ASCE-4-98, shown that the 3D FEM model is reasonable and usable. It is a simple and suitable engineering method, from which, the stress distribution of the storage tank can be obtained directly from the 3D FEM model.

Analysis and Solutions for Effect of Change of Seismic Acceleration on Valve Qualification

Zhang Wei, Qu Changming

2016, 37(6): 62-65. doi: 10.13832/j.jnpe.2016.06.0062

Abstract(32) PDF(0)

Abstract:
The envelope seismic acceleration is normally used in the seismic requirements of valve design in nuclear power plants. The envelope seismic acceleration in the present design is increased for the change of the site condition or design requirements, which will change the cover scope of the parent valves which has been already qualified. The principle requirements on the analysis and tests for different suppliers are determined based on the engineering practice, codes and standards. The detail input parameters for the different seismic test methods are also provided based on the principle requirements in the code.

Fuzzy Adaptive PI Control of Steam Temperature in a Canadian SCWR

Dong Huaping, Lu Jianchao, Chen Peng, Sun Peiwei

2016, 37(6): 66-70. doi: 10.13832/j.jnpe.2016.06.0066

Abstract(28) PDF(0)

Abstract:
Supercritical Water-cooled Reactor（SCWR） is the only water-cooled reactor proposed for the Generation IV nuclear system. The dynamics of its steam temperature are strongly influenced by the reactor power and also with a high degree of nonlinearity. It is difficult to control using the traditional PI control only. In this paper, moving boundary method is adopted to construct the dynamic model of the steam temperature in a Canadian SCWR. The design of the feedforward controller and fuzzy adaptive PI controller is based on the dynamic characteristics of the steam temperature. Through numerical simulation, it is found that the feedforward control can reduce the effect of the reactor power on the steam temperature. The control parameters can be tuned online using fuzzy adaptive control and the control performance can be improved. Steam temperature can promptly stabilize and the variation is reduced using the designed fuzzy control system. It is concluded that the control performance is much better than that with PI controller and the control requirements of Canadian SCWR are satisfied.

Test Study and Optimization of Process Instrumentation System Electromagnetic Interference in Nuclear Power Plants

Chen Yongwei, Wang Renxiang, Chen Ke, Li Dong

2016, 37(6): 71-74. doi: 10.13832/j.jnpe.2016.06.0071

Abstract(24) PDF(0)

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Based on the abnormal trip of one domestic nuclear power plant reactor during the test of neutron flux measurement channels, Test scheme is established for the interference of the power source of the normal lighting system in the electrical building of the nuclear power station on the process instrumentation system, including the lights on/off interference test, the 220 V（AC） power supply box interference test, the safe distance between 220 V（AC） cable, process instrumentation system cabinet interference test and the cabinet ground check. Based on the test results, the causes of reactor trip is analyzed and the improvement program is proposed, including source term distance method, source term shielding method and grounding system correction. Some measures were verified that it can improve the ability of anti-electromagnetic interference.

Verification and Validation of HPD-Based Nuclear Power Plant Distributed Control System

Ding Yihang, Li Shixin

2016, 37(6): 75-79. doi: 10.13832/j.jnpe.2016.06.0075

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Abstract:
Verification and validation（V&V） process is one of the critical issues for the application and license of the safety class distributed control system（DCS） in nuclear power plants（NPPs）. The application of hardware description language programmable devices（HPDs） technology in NPP safety class DCS has introduced new regulatory issues. In this paper, preliminary analyses of overseas V&V regulations and IEEE standards are presented. Based on IEEE 1012-2012 standard, combined with the specific characteristics of HPDs, the whole life cycle V&V tasks and methods for HPDs system are detailed. The HPDs technology based safety DCS system regulation proposals are also provided.

System Development of Online Monitoring of Fuel Cladding Defect in PWR

Shan Chenyu, Jia Yuncang, Lyu Weifeng, Xiong Jun, Tang Shaohua, Pan Yuelong, Yang Linjun

2016, 37(6): 80-85. doi: 10.13832/j.jnpe.2016.06.0080

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Abstract:
This paper presents the diagnosis physical model based on quantitative analysis and the design of online monitoring device for the fuel cladding defect in PWR operation conditions. A complete solution for online monitoring of fuel defect is given. Meanwhile, the correctness of the system design is verified by measurement data and theoretical simulation, and the radiation source experiment for prototype as well, to solve the deficiencies of the existing means of fuel defect monitoring in CPR1000 units, and to improve the safety performance of pressurized water nuclear power unit operation.

Measurement Technology for Fission Gas of Irradiated HWR Fuel Elements

Kuang Liuwei, Jiang Linzhi, Ren Liang, Yu Feiyang, Guo Chengming

2016, 37(6): 86-89. doi: 10.13832/j.jnpe.2016.06.0086

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Abstract:
Based on the small cavity volume of fuel elements in heavy water reactors, this paper designed the measurement equipment to measure the fission gas released by fuel elements in the heavy water reactor, and studied the measurement technology of fission gas release, puncturing technology, measurement technique of fission gas and the pressure sampling technology of fission gas. The technological process and parameters were confirmed. By pressure test and accuracy measurement, this experiment verified the system sealing and volume measurement, established a technique for measuring the fission gas of fuel elements in heavy water reactor and realized the measurement of the fission gas of fuel elements in heavy water reactors.

Discussion on V＆V Activity Flow of Safety Software

He Peng, Yang Daibo, Zhu Jialiang, Li Hongxia, Yu Junhui, Zhu Biwei

2016, 37(6): 90-93. doi: 10.13832/j.jnpe.2016.06.0090

Abstract(27) PDF(0)

Abstract:
In order to insure the quality of safety software of the digital system inside the nuclear power plant, the thorough V&V activities need to be performed on the safety software. According to the standard systems at home and abroad, establishing the V&V activity flow which satisfies the standard requirements, and performing a series of strict and complete V&V activities, is a more viable way to improve the software quality. Based on IEEE 1012-2004, this paper discussed in detail the V&V flow in the development process of safety software, and analyzed the key problems during the V&V activities, so as to provide reference to the development of V&V activities for safety software.

Key Technology for Preparation of Aluminum-Based Boron Carbide Neutron Absorbing Material by Liquid Molding Method

Liu Yanzhang, Wang Xin, Luo Zhiyuan, Li Qiulin

2016, 37(6): 94-97. doi: 10.13832/j.jnpe.2016.06.0094

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In this study, 31 wt.% B₄C/Al plate is prepared successfully by liquid molding method, with the processes of powder surface treatment, melt modification, stirred and dispersed liquid molding and casting rolling. Comprehensive tests and analysis showed that the B₄C particles are dispersed uniformly in the matrix of the plate prepared by liquid molding method. Preferred orientation distribution of the boron carbide particles does not occur during rolling. A tight interface maintains between B₄C particles and the Al matrix. The average area density of ¹⁰B in the plate is 0.0372g/cm² and maintains uniformly. The high temperature accelerated corrosion test results which simulated the spent fuel pool showed that the plate corrosion weight change is less than 0.5% after 8000 h. The effect of radiation on the properties of the sheet is limited due to the low cumulative dose during service and the structure of the material.

Irradiation Technology for NPP Reactor Pressure Vessel Material in 49-2 SPR

Zhu Jie, Zhang Yadong, Tong Zhenfeng

2016, 37(6): 98-103. doi: 10.13832/j.jnpe.2016.06.0098

Abstract(29) PDF(1)

Abstract:
By setting up a model for the reactor core, calculation of the fast neutron energy spectrum of the position 25 cm above the bottom of the irradiation tube can be realized. Combining the results of the calculation and the multiple foil activation method measurement, the fast neutron fluence rate can be obtained by unfolding the spectrum with SANDII code. Adopting the relatively fast neutron fluence rate measurement, the distribution of the axial fast neutron fluence rate can be obtained, and, the hour-length and the scheme of the irradiation can be defined. The technique target fast neutron（E≥1 Me V） flux received by the sample reaches about 6×10¹⁹ cm^-2. For accomplishing the sample disassembly, firstly we obtained the radioactive source term of the sample by calculating with ORIGEN2 code. Then, according to the operating environment, a model was set up by MCNP code, and the gamma dose rate data in different Shielding layer thickness was obtained by calculating with the code. The comparison between the calculation result and the measurement result indicated that the shielding design was effective. This irradiation experiment completely satisfied the technique target.

Analysis of Fuel Handling Area Temperature and Pressure Response Following Loss of Spent Fuel Pool Cooling

Wang Guodong, Wang Zhe, Hu Benxue

2016, 37(6): 104-108. doi: 10.13832/j.jnpe.2016.06.0104

Abstract(35) PDF(0)

Abstract:
A normal-closed steam-relief panel is provided in the design of the spent fuel cooling system for CAP1000 nuclear power plant. Saturated steam is released from the spent fuel pool to fuel handling area under loss of cooling scenarios. The steam-relief panel will be opened in order to mitigate the pressurization process in the fuel handling area after the fusion of the fusible links, designed to yield at an appropriate temperature. A three-dimension temperature field and pressure response to the fuel handling area was performed using GOTHIC8.0 code in this paper. The results show that temperature stratification behavior is expected depending on different time sequence. At time of about 7500 s, the steam-relief panel is expected to be opened with distinct pressure decreasing in the fuel handling area and maintain a low pressure level after that.

Effect of Anti-Vibration Bar on Steam Generator Tube Integrity

Cui Suwen, Zhu Yong, Ren Hongbin

2016, 37(6): 109-112. doi: 10.13832/j.jnpe.2016.06.0109

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The anti-vibration bars are critical to the integrity of tubes for steam generator during operation. Taking the steam generator of Generation 2⁺ for example, the potential effects of anti-vibration bars on the integrity of SG tube is analyzed. It is concluded that the anti-vibration bar twist, the gap between anti-bibration bar and tube, and the insertion depth of the anti-vibration bar are very important to the tube integrity. Based on the conclusions, optimizations are proposed to make the tubes more reliable during operation.

Failure Analysis and Test for Motor-Driven Isolated Valve with Remote-Transmit Entity

Yang Zhang, Jiang Yanlong, Sun Chengbin, Wang Hexu

2016, 37(6): 113-116. doi: 10.13832/j.jnpe.2016.06.0113

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The failure analysis and treatment for motor-driven isolated valve with remote transmit entity was done in a pressurized water nuclear power station. It was found that the immediate cause of the abruption of clutch coupling brace was the impact loads by the repeated start and stop of the motor. The microscopic test and the valve performance curve test in field found that the root cause was the mismatch between the motor and the actuating mechanism. The feasibility of increasing the driving moment or decreasing the innocent resistance was studied theoretically and verified in field. It found that, when the output rotational speed of the driven motor decreased by 50%, the average friction force decreased by 30% during the isolated valve acting; when the gland tightening force decreased by 50%, the average friction force decreased by 50% during the isolated valve acting.

A Precision Micro-Current Source with Nanoamp Output

Jin Chuanxi, Guo Lifeng, Lu Gubing, Chen Panhui

2016, 37(6): 117-120. doi: 10.13832/j.jnpe.2016.06.0117

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A precision micro-current source is designed for the test and calibration of micro-ammeters. It is based on the digitally controlled structure of microcontroller system and uses the circuit design method for the micro-current situation to achieve a micro-current output with the range of 0~200.0nA. The temperature drift analysis and the software calibration method for the V/I convert circuit are discussed. The test shows that the calibrated micro-current source can work steadily with a high precision and the error less than ±0.01%FS.

Review of Heat Pipe Application in Advanced Nuclear Reactors

Liu Ye, Zhou Lei, Zan Yuanfeng, Huang Yanpin

2016, 37(6): 121-124. doi: 10.13832/j.jnpe.2016.06.0121

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After the Fukushima disaster, the passive safety for nuclear power system has received more and more attentions, leading to a broad application of heat pipes. This paper summarizes the overall application status of the heat pipe technology in nuclear power system design.

Parametric Modeling of Refueling Operations for Pressurized Water Reactor Nuclear Power Plants

Shao Changlei, Yin Junlian, He Xiaoming

2016, 37(6): 125-129. doi: 10.13832/j.jnpe.2016.06.0125

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In order to optimize the refueling operation and improve the efficiency of the pressurized water reactor nuclear power plants, kinematic models of the main refueling equipment such as refueling machine, underwater fuel transfer system and fuel handling machine are established, respectively. General formulas for total time of reactor refueling are derived based on the models. The parametric analysis method for refueling operations is developed. Finally the models and methods are proved to be proper and effective by the parameter calibration of a typical reactor.

Failure Analysis and Treatment of Temperature Detector in Main Loop for Qinshan Phase Ⅱ NPP

Wang Guoqing

2016, 37(6): 130-133. doi: 10.13832/j.jnpe.2016.06.0130

Abstract(21) PDF(1)

Abstract:
For the representative failure happened to the reactor coolant temperature detector in Qinshan Phase Ⅱ Nuclear Power Plant, we discovered the cause of failure and tried to solve this problem by analyzing the structure of material, the situations in other nuclear power plants, the vibration in laboratory and in field. The analysis shows that the reason is the collision happened between the protective well and the temperature detector. The temperature detector will be broken or ruptured. Adding the ring outside of the temperature detector is proved to work well.

Measurement and Formation Mechanism of Noise in Eddy Current Inspection of Heat Exchanger Tube of Steam Generator in Nuclear Power Plants

Zhang Jun, Gu Bo, Yang Hongbo, Pei Xibao, Song Tao, Wang Xiaogang, Kong Yuying

2016, 37(6): 134-137. doi: 10.13832/j.jnpe.2016.06.0134

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In the general eddy current（EC） inspection of the steam generator（SG） tube, the effect of the tube noise level on the flaws analysis is significance during the EC data analysis using bobbin coil data coil from the mock-up. In the theory of EC inspection, there are two methods to measure and calculate the noise level. One method is to use Root–Mean-Square（RMS） quantities to measure the noise level in the area of interest after sizing the tube signal by differentiation, and the real level of baseline noise can be obtained. The other method is to calculate the SNR of the detection system and combine the reference signal from the calibration tube. In this paper, the levels of baseline noise for several in-service inspection methods are quantitatively studied, and the mechanism causing the noise is discussed.

A periodic Test Scheme Design of ACPR1000 Reactor Protection System Based on Firm Sys

Shi Guilian, Xie Yiqin, Li Mingli, Sun Na

2016, 37(6): 138-142. doi: 10.13832/j.jnpe.2016.06.0138

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Firm Sys is the first and currently the only digital nuclear safety instrument and control system platform in China, which has been developed, qualified and applied. Based on ACPR1000, Yangjiang Nuclear Power Plant（units 5 and 6） is the first nuclear power plant to adopt Firm Sys platform to realize the function of reactor protection system. According to the requirement of related standards and regulations, the reactor protection system should be periodically tested. The periodic test should cover the whole protection system and it is closely associated with the diagnosis design and the structure of protection system. Additionally, the effect on the system safety functions need to be concerned. The periodic test design has been one of the key design technologies for digital protection systems. Based on safety analysis method, a complete set of periodic test solutions are presented in this paper, which meet the requirements of standards and regulations. Compared with the CPR1000 project, it simplifies the design of the protection system, optimizes the operation of test personnel, and improves the periodic test scheme of the operation program.

Ultrasonic Phased Array Inspection Technique for Valve Rod in a NPP

Qin Jinguang, Dong Jialong

2016, 37(6): 143-145. doi: 10.13832/j.jnpe.2016.06.0143

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The grinding undercut of 1117-600-Э-M3 valve rods had broken twice during the operation period in a NPP. In order to prevent the breaking happening again, it is necessary to carry out a complete inspection for the same type valve. The valve is very big, and the workload is also very heavy for the strip inspection. The technique of ultrasonic phased array inspection for the structure features of valve was developed, and an inspection was carried out during the outage. The inspection results showed that the ultrasonic phased array inspection technique can make an efficient inspection for the broken parts of valves and decrease the dismounting work at the same time.

Application of Improved Wiener Degradation Model in Reliability Assessment for Primary Loop Tube

Li Shaoshuai, Chen Ling, Cai Qi

2016, 37(6): 146-149. doi: 10.13832/j.jnpe.2016.06.0146

Abstract(36) PDF(0)

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As it can well solve the data volatility, Wiener degradation model is becoming a popular model for degradation data treatment. However, Wiener degradation model with drifting is aimed at linear degenerate, and even it demands product in same batch with fixed degenerate drift parameter, which limits the range of its application. In this study, Wiener model is adapted to non-linear problems, so that it could be used in more target of evaluation. Shift parameter of the model is supposed a variable changing with evaluation target, which is supplement and perfection for Wiener degradation model. At last, reliability of this model is analysed, basing on the data of pipe crack in nuclear reactor primary loop, to prove the rationality of the model.

FCM Fuel Thermal Performance Analysis

Liu Zhenhai, Li Wenjie, Chen Ping, Li Yuanming, Zhou Yi, Zhang Kun, Xing Shuo

2016, 37(6): 150-154. doi: 10.13832/j.jnpe.2016.06.0150

Abstract(33) PDF(0)

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In this paper, the method to analyze the FCM fuel thermal performance using finite element software ABAQUS is studied, and the comparative analysis of thermal performance of FCM pellet and UO₂ pellet is carried out. The results show that the temperature distribution trend of FCM pellet is similar with that of UO₂ pellet but with a strong non-uniformity. The temperature of FCM pellet is much lower than that of UO₂ pellet under typical PWR operating environment. The temperature of FCM is not sensitive to the burnup change at the same linear power density and the temperature increase rate of FCM pellet is much lower than that of UO₂ pellet along with the linear power density increment at the same fuel burnup.

Study of CF3 Fuel Assembly Major Irradiation Performance in Pile

Chen Ping, Jiao Yongjun, Zhou Yi, Liu Zhenhai, Zhang Kun, Lei Tao, Qin Mian

2016, 37(6): 155-158. doi: 10.13832/j.jnpe.2016.06.0155

Abstract(49) PDF(7)

Abstract:
The performance, including fuel rod corrosion, cladding growth and fuel assembly growth, of four group CF3 leading fuel assemblies in Qinshan-Ⅱ Nuclear Power Plant unit 2 from cycle 11 to cycle 13 are studied and compared to 1^st cycle irradiation of CF3 fuel assemblies. The results show that the irradiation performance of CF3 fuel assemblies are performing as expected and have more margin than the predicted value. The existing irradiation performance model can be used in the engineering design with further development.

Operation Coordination Analysis and Simulation of Integrated NPP with OTSG

Tai Yun, Tao Li, Yan Bing, Sun Jianhua

2016, 37(6): 159-163. doi: 10.13832/j.jnpe.2016.06.0159

Abstract(37) PDF(2)

Abstract:
This paper focuses on small nuclear power plant which adopts OTSG. Based on static mathematical models of some important parts such as reactor core, steam generator etc., combined with the control strategy of MRX, the paper simulates the dynamic feature during load following process of small NPP. Moreover, the paper analysis the operation coordination via heat transfer time and control time delaying. According to the Matlab/Simulink simulation results, the control strategy of MRX is valid and suitable for the power following of NPP. The operating time of the second-circuit system is relatively longer, and the response rate feed water flow is lower than that of the nuclear power changing. In practical application, it should be considered to adopt the fast response electric pump.

Phenomenological Investigation of Gas Bubble Behavior in the Vicinity of Mixing Vane of Spacer Grid

Zhang Junyi, Yan Xiao, Xu Jianjun, Huang Yanping

2016, 37(6): 164-167. doi: 10.13832/j.jnpe.2016.06.0164

Abstract(42) PDF(2)

Abstract:
Visualization technique was adopted to study the gas bubble behaviors in the vicinity of mixing vane of spacer grid in a 3×3 rod bundle under ambient temperature and atmospheric pressure. It was found that the bubble stagnation, which is dependent on liquid flow rate and volume fraction, occurs at the leeward of the mixing vane. The stagnation bubble interface oscillated with flow and inflow bubble coalescence. Smaller bubbles were generated by liquid entrainment at the end of the stagnation bubble, which is one of the main mechanisms of bubble size redistribution. The stagnation bubble size increased and break-up bubble size decreased with the increasing of the liquid flow rate under identical void fraction. The stagnation bubble size was independent upon void fraction under identical liquid flow rate, while the oscillation of stagnation bubble interface was strongly influenced by the void fraction, especially the inflow bubble size. The bubbles, passed through the mixing vane from windward and leeward mixing vane, led to different bubble flow directions and phase distribution at the downstream of the spacer grid.

Theoretical Calculation and Analysis of Critical Heat Flux Characteristics under Flow Oscillation Conditions

Liu Wenxing, Zhao Dawei, Su Guanghui, Huang Yanping

2016, 37(6): 168-172. doi: 10.13832/j.jnpe.2016.06.0168

Abstract(21) PDF(0)

Abstract:
The transient CHF calculation code was used to analyze the CHF characteristics under inlet flow oscillation conditions. The calculation results show that: Inlet flow oscillation would have significant effects on critical heat flux. Especially the CHF characteristics would be deteriorated when the oscillation magnitude and period are increased, the heated length is decreased or the outlet quality is decreased. The effect of inlet flow oscillation on CHF would have detrimental impact on the safety of the reactor system.

Study on Natural Circulation Density Wave Instability under Heaving Conditions

Gong Houjun, Huang Yanping, Zan Yuanfeng

2016, 37(6): 173-176. doi: 10.13832/j.jnpe.2016.06.0173

Abstract(24) PDF(0)

Abstract:
Two-phase natural circulation density wave instability under heaving conditions have been studied using PNCMC code（Program for natural circulation under motion condition）. Natural circulation density wave instability still occurred under heaving conditions, and the coupling of heaving motion and density wave instability induced much more complicated oscillation of mass flow rate. The period of complex flow oscillation was one heaving period under heaving with long period（13s, 23s）. However, there were two types of oscillation state under heaving with short period（3s）, for the cases of low heating power, the periodicity of flow oscillation was not significant; for the cases of high heating power, and the periodicity of flow oscillation was very significant. Heaving moved up stability boundary of natural circulation, the smaller the heaving period, the greater the amount of up.

2016 Vol. 37, No. 6