核动力工程

An Exploration of CNPEC’s Nuclear AE Organization, Characteristics and Its Industrial Value

ZHAO Jianguang, KUANG Wei

2013, 34(S2): 1-4,13.

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Abstract:
The paper studies and analyzes the CNPEC’s AE organizational operation model and its characteristics in details to explore its value and contribution to the reform of the Chinese state-owned enterprises. By building the Design and Construction Integration Platform, CNPEC integrates the resources of the nuclear industry chain to effectively ensure the whole performance, the safety and high quality of the NPPs under construction; by establishing the Total Quality Partnership which focuses on the cross-border quality management and control, CNPEC enhances the quality management level of the enterprises in the nuclear industry chain; by promoting the technology development cooperation, CNPEC pushed the technological advancement of the whole nuclear industry chain.

Development of Welding Material for Primary Coolant Pipe Automatic Welding in PWRs

ZHU Decai, LI Yuwei, LI Fuliang

2013, 34(S2): 5-8.

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Abstract:
Because of the narrow groove and welding process of one pass per layer are used in the automatic welding of primary coolant pipe, the welding material which more purity and suited to the welding process is to be developed in order to guarantee the quality of the weld. We had tested the weld able and the steady of new material by experiment; also we had evaluated the performance of the welding join by fatigue experiments. The result testify that the new welding material is suited to the welding process and can get excellent performance of the weld.

Simulation Test Research of Self-Compacting Concrete Used in Containment Cone of Yangjiang Nuclear Power Plant

LU Guangye, WANG Huicheng, YU Bing, CHEN Lisheng, YANG Hongbo

2013, 34(S2): 9-13.

Abstract(21) PDF(0)

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Details the simulation test model and test content of containment cone in nuclear power plant. After selecting the raw materials, the reasonable self-compacting concrete mix through experimental studies, can meet the complex technical requirements of concrete of the containment cone. The simulation test works smoothly with excellent performance of concrete, effective monitoring data, good apparent quality of concrete after form removal and uniform compacting entities.

Study on Scattering Dose Rate Calculation in Radiation Measurement Laboratory of Nuclear Power Stations

CHEN Xiaoqiang, ZHOU Wei, TANG Shaohua

2013, 34(S2): 14-17.

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In this paper, using Monte Carlo method for calculating the dose rate of scattering which rooted in calibration trolley, dial, around wall and collimator of the γ radiation facility in the Radiation Measurement Laboratory of one nuclear plant. The results showed that the maximum contribution of scattering dose rate at reference point is the scattered ray of collimator, and that the dose rate of scattered ray is less than 3% of total radiation dose rate. The result satisfies the regulations and requirements of ISO4037-1-1996 and GB/T12162-1-2001.

Analysis of Welding Procedure Qualification to T and Cross

LI Lei

2013, 34(S2): 18-22,34.

Abstract(26) PDF(0)

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On the basis of practical experience, the passing rate of T and cross WPQ is very low. After having done a thorough study on the failed projects, combining with the domestic and foreign standards of WPQ, this paper gives some measures to make sure that the T and cross welding process qualification can be solved smoothly.

Research on Release Calculation of Fission Product from Defect Fuel Rod to Reactor Coolant

LU Weifeng, XIONG Jun, TANG Shaohua

2013, 34(S2): 23-26.

Abstract(18) PDF(0)

Abstract:
The calculation model was established according to the research on the release mechanism of fission product from defect fuel rod to reactor coolant. The sensitivity analysis was carried out for fuel rod defect rate, fuel rod defect size and burnup base on the design date of CPR1000 nuclear power plant, and the calculated equivalent escape rate was compared with the escape rate described in AP1000 design control document. The sensitivity analysis and comparison results show that the calculation model is sensitive for fuel rod defect size but not sensitive for fuel rod defect rate and burn up. The calculated equivalent escape rate of part nuclides is similar as the escape rate of AP1000 design control document base on the calculation model and the 34μm fuel rod defect size.

Bumpless Switching of PID-controller for CRP1000 Nuclear Power Station

TAN Lei, LUAN Zhenhua, YANG Zongwei, LI Xianmin, TANG Bihui

2013, 34(S2): 27-30.

Abstract(11) PDF(0)

Abstract:
Bumpless auto/manual switching methods were developed forPID control loop. These methods were used in typical control systems in DCS control of CPR1000 NPP, achieving a good result.

Heat Load Operation Range of Plate Heat Exchanger of Component Cooling System for NPP at Extra-Low Temperature Site

YANG Ting, ZHU Min, HU Jian, XIAO Wei

2013, 34(S2): 31-34.

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Abstract:
The temperature of sea water at Hongyanhe nuclear power plant is too low that the temperature of Component Cooling System（RRI） might be lower than the limit, even the water might be frozen in the heat exchangers. It would cause the lost of final heat trap. Heat load subarea operation was performed to provide RRI from the low temperature. The edge of single subarea was calculated and the summation of total subareas were reckoned, based on the design limit, by the analysis of the input and boundary of both Essential Water System（SEC） hydrodynamic and RRI/SEC heat exchanging calculation. Designing of heat load subarea was the most important academic base of RRI heat exchanging design at Hongyanhe. This method informed a valuable example for the RRI heat exchanging design at low temperature sites.

Modification of Seismic Support Scheme for Reactor Coolant System

LU Zhi, WU Yingxi, MAO Qing, CAO Leisheng, LI Qiang

2013, 34(S2): 35-37,42.

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This paper studied the M310 reactor coolant system seismic response, and found that the steam generator vibrated greatly in the seismic, which was resulted from the weakness of the steam generator support structure. Based on the research, this paper gives out a new reactor coolant system support structure. Seismic calculation result proved that this new support structure could reinforce the steam generator support, and improve the capability of the reactor coolant system in seismic.

DC Voltage Endurance and Current Leakage Test for First Domestic 1000MW Nuclear Power Generator

LI Gang, ZHOU Li, LIU Jianyi, HAN Shuyin

2013, 34(S2): 38-42.

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This paper introduces the process of DC voltage endurance and current leakage test for the first domestic 1000MW nuclear power inner water cooling generator, analyzes the problems during the test; describes the ways to deal with abnormal current leakage of this kind of generator, and discusses the standards about test.

Research on Inrush Current for Nuclear Power Station Transformer by EMTP

LIU Sen, ZHANG Qingliang, LIU Qiang, JIANG Tao, YI Ning

2013, 34(S2): 43-46,50.

Abstract(16) PDF(0)

Abstract:
Based on the principle of the inrush current, this paper gives a EMTP simulation test system model for the energizing unload auxiliary transformer of nuclear power station. The inrush current in different flux linkage and energizing angle are simulated by setting the close and open time of the breakers. The harmonic of the inrush current is analyzed, which can give the site engineers useful instructions as well. Finally, the inrush current control strategy with EMTP/ATP simulation is discussed.

Study on Primary Coolant pH Regimes of Nuclear Power Plants

CHEN Chao, WANG Zhengguang, XIE Enfei

2013, 34(S2): 47-50.

Abstract(20) PDF(0)

Abstract:
Analyzing the factors having effects on the choice of the primary coolant pH, the optimal pH range is obtained, and the chemical regimes applied in the different nuclear power plants worldwide in history is summarized. Meanwhile, based on the comparison of the pH conditioning of CPR and EPR, together with the experiences both from home and abroad, the preliminary pH regime for the new research projects is put forward.

Research on Application of Quantities and Workload Model in Nuclear Power AE Company

ZHOU Guangshu, HUANG Xiao, TANG Xiaoming

2013, 34(S2): 51-54.

Abstract(19) PDF(0)

Abstract:
The Quantities and Workload Model is established by improving the earned value method. The model tends to simplify, standardize, quantify and informationize the complex process and cost data of building nuclear power project, providing the nuclear power AE company more standardized, delicate, personalized information, which can be used in management activities such as multi-project management, and delicate management and project accounting. This paper describes the process of establisheding the model and its characteristics, and makes a deep research on the model’s application.

Research of Software Parametric Filtering and Pattern Analysis for Generator Partial Discharge Monitoring System

YUAN Jin, LIU Guoqiang, CHEN Xiaoyi, LI Xiang

2013, 34(S2): 55-57.

Abstract(16) PDF(0)

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The paper was based on the 1000pF capacitance coupling generator partial discharge monitoring system used at Ling’aO II, Hongyanhe and Ningde nuclear power stations. The application status of the mainstream wide band generator partial discharge on-line monitoring system at present was introduced. By means of the research of off-line calibration and software-hardware design, the general method of software parametric filtering according to the commissioning and startup practice of recent several newly-built nuclear power turbine generator was obtained. Then lots of real typical partial discharge pattern data from home and board were collected and studied. The induction and summarization of typical fault analysis and location method for the 2D-partial discharge pattern was completed. This paper also supplied a reference and guidance at the area of insulation judging and fault diagnosis for newly-built and on-line unit startup and operation. we have mastered parameter adjustment of local online monitor system of electric motor, high frequency filtering and photograph analysis, process of measured photograph.

Research of Test Scheme of Verification and Validation of Reactor Protection System Based on TXS Platform

WANG Qiang, LI Guomin

2013, 34(S2): 58-61.

Abstract(21) PDF(0)

Abstract:
Based on IEC60880 standard for software used in nuclear power plant safety systems, the test scheme of verification and validation（V＆V） test for reactor protection system（RPS） of Ling’ao nuclear power plant（NPP） based on safety digital TXS platform is introduced. Furthermore, the completeness and correctness of V＆V test is analyzed accordingly. The successful commercial operation of Ling’ao NPP fully proves the importance and necessity of RPS V＆V test, what is more important, the research provides good experiences and references for RPS V＆V test in future nuclear power plants based on other I＆C platform.

Research and Improvement of Deaerators in Turbine Bypass Condition

QIAO Piye, LIU Xi

2013, 34(S2): 62-64,69.

Abstract(14) PDF(0)

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The requirement of deaerators about turbine bypass system in nuclear power station has been described in this paper. The pressure change curves of spray deaerator and tray deaerator on turbine bypass condition has also been expounded according to the calculation. The pressure of tray deaerator will increase rapidly and may be over the design pressure on transient operating condition, and some measures about system optimization or equipment improvement can resolve these problems.

Start-Up Characteristic Data Processing and Start-Up Acceptance Test Research for 1E Level Motor

YUAN Jin, LIU Guoqiang, SUN Yi

2013, 34(S2): 65-69.

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Based on the acceptance criterion discussion for the 1E motor full voltage start- up characteristics used in nuclear power stations, this paper firstly simulated the switching transient response for the motor start-up LR circuit. The SIMULINK tool-FFT was used to process the current data and confirm the switching transient low-frequency modulation effect in the motor start-up process from the frequency domain. Then, the motor equivalent decomposition and superposition model was proposed. Though the equivalent decomposition, the relationship between full voltage start- up current and block running current was clear. Then based on much measured data, using Boltzmann model computer fitting, the conclusion was proved right. According to this paper, the acceptance criterion was unified at home and abroad basically.

Research on Software Online Downloading Technology Based on MELTAC 1E-DCS System

XU Jianfei, LUO Hao, CHEN Tong, YANG Jiazhu

2013, 34(S2): 70-72,86.

Abstract(21) PDF(0)

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Based on the Mitsubishi MELTAC digital control system（DCS） platform and its own characteristics, to meet the risk control requirements and download quality requirements, we used the risk-list method to verify the risks one by one. Through verification test and downloading experience, DCS online download technology scheme was put forward. The use in Hongyanhe, Ningde, Yangjiang nuclear power project shows that, the scheme solved the problems of long duration, high risk, wide influence, and the online download reliability rate reached 100%.

Root Reason and Solution Research of ASG TDFP Steam Inlet Isolating Valve Failure to Open

BAO Yongke, TIAN Kuo, WANG Langlang

2013, 34(S2): 73-76.

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Taking the phenomena of CPR1000 nuclear power plant steam inlet isolating valve of ASG TDFP fail to open as an example, combined with on-site commissioning process, this paper analyzed the fault causes base on the stress analysis and the work environment of the valve, and proposed the simple processing methods in the commissioning process and the reconstruction of the valve when nuclear power plants in service.

Modifications for Improving Safety Feature of Spent Fuel Pool in CPR1000 Nuclear Power Units

WANG Yaodong, GONG Aicheng, XU Jie, ZENG Jianli

2013, 34(S2): 77-80.

Abstract(19) PDF(0)

Abstract:
According to the result of safety review and inspection on Chinese existing and under construction nuclear power plants post Fukushima accident, it is found that there are some weak points in the CPR1000 nuclear power units in the accident of SSE（safety shutdown earthquake） with long term SBO（station black out） and LUHS（loss of ultimate heat sink）.Due to the lack of emergency cooling systems, the spent fuel pool could be damaged by these accidents. So it is necessary to improve its safety feature for CPR1000 units. From the point view of defense in depth, assumed the same kind of Fukushima accident, the possible accidents combination is established on the basis of confidential methodology. For these accidents, the safety functions or strategy for prevention or mitigation are given. And the new modification solutions, like spent fuel pool monitoring, make up and steam exhausting, are raised for CPR1000 units under different construction situation. They are proved to be feasible and available for these extreme accidents and the CPR1000’s safety and reliability are improved largely.

Analysis of Electromagnetic Shield for LV Motor Electric Leakage Protection

ZHOU Wenbin, QI Shuyun, ZHU Zengpei

2013, 34(S2): 81-82.

Abstract(17) PDF(0)

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By comparison and analysis of the original equivalence magnetic field and circuit of the zero sequence CT for electric leakage protection with which with a electromagnetic shield, this paper shows how the electromagnetic shield works, finds that the CT with electromagnetic shield is with different characteristics from the original one. It is concluded that the electric leakage protection will not pickup when certain fault occurs.

Numerical Analysis of Flow Skirt Influence on Core Inlet Flow Distribution

TANG Mao, ZHANG Mingqian, CHEN Liang, YU Xiaolei

2013, 34(S2): 83-86.

Abstract(15) PDF(0)

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The core inlet flow mal-distribution could lead to a local peak temperature point which is not allowed for the reactor safe operation. The flow skirt was used for a better core inlet flow distribution. In order to study the effect of the flow skirt on the flow distribution, the dimension "d" on the top of the flow skirt and the length "t" of the support plate were changed for the simulation. The results showed that "d" had a little effect on the flow distribution, but in the centre area, there was a better distribution with the decreased "d", if the "t" was smaller, the distribution was more uniform.

Research and Application of Module Parameters of Half-speed Turbine Speed Governing System in CPR Nuclear Power Plant

YANG Liping

2013, 34(S2): 87-89,96.

Abstract(17) PDF(0)

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CPR1000+ technology is applied into Ling’ao Phase Ⅱ1000MW-class nuclear power plant for the first time in China. It is important of security and steady operation for nuclear power plants and grid to get the module parameters of turbine governing system by the simulation tests. Characteristics of ALSTOM half-speed turbine governing system are analyzed. Simulation test of half-speed turbine governing system is put forward. Results of simulation are analyzed and the parameters are identified. Module parameters of ALSTOM half-speed turbine governing system are acquired. It is highly important for Ling’ao Phase and Ⅱ CPR nuclear power plants to ensure safe and steady operation of long-term.

Optimization of Anti-Dilution Protection Realization Optimize for CPR1000

CHEN Liang

2013, 34(S2): 90-92.

Abstract(13) PDF(0)

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This chapter introduces the realization process of Anti-Dilution Protection logic in detail and it is theoretically analyzed the various kinds of sequence charts based on the pragmatic condition. The all-around investigation for ADP logic realization will be introduced on the basis of MELTAC safety DCS platform. It’s to present the engineering configuration progress and the exposed problem in detail, and also the corresponding correction measure is provided.

Analysis of Choked Flow in Control Valves

YE Zhangliu, LIU Xichao, XIE Enfei

2013, 34(S2): 93-96.

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There is an experience feedback that a control valve choked flow problem in the steam generator blowdown system of Ling’ ao phase II nuclear power plant resulted in a frequency system trip. The problem is studied by using a method to calculate the choked flow in the control valve and orifice and some practical solutions are suggested to prevent the damage of choked flow. The control valve recovery facter FL and the pressure drop of orifice have a great influence to choked flow. To reduce the damage of choked flow,it is recommended to select a higher recovery facter FL control valve and install an orifice with bigger pressure drop downstream the control valve to decrease the pressure drop of control valve and increase the pressure downstream the control valve.

2013 Vol. 34, No. S2