核动力工程

Technical Improvement of ATE System of Ling’ao Nuclear Power Plant PhaseⅡ

ZHU Xing-bao, XIONG Jing-chuan, LIANG Qiao-hong

2009, 30(S2): 1-5.

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Abstract:
In order to solve the problem that the content of SO42-in Steam Generator significantly increased beyond the criteria after the use of the condensate treatment （ATE） system in Daya Bay Nuclear Power Plant and Ling’ao Nuclear Power Plant PhaseⅠ, technical improvement have been conducted on the sizes of the fore cation bed and the mixed bed, water distributing devices,ion exchange resins and separation facility. The effectiveness for the ion exchange of the mixed bed is improved, the resolved substance of cation resin is decreased; it is more impossible for fragments and powder which would lead high SO42-content in Steam Generator. Finally,the quality of the steam-water could be improved and ensured.

Casting Technique for Primary Coolant Piping for Pressurized Water Reactor Power Plant and Its Localization

LI Yuan-tai, ZHANG Chun-lai, LEI Zhong-li

2009, 30(S2): 6-10.

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The required chemical content range of Primary Coolant Piping （Static cast elbow, centrifugally cast pipe） by Design and Construction Rules for Mechanical Components of PWR Nuclear Islands （RCCM）（French Norm） is wide. The result of manufacturing qualification test, which was performed by YANTAI TAIHAI MANOIR NUCLEAR EQUIPMENT CO. LTD. （THM）, shows that qualification pieces satisfied the chemical requirement of RCCM may not satisfy its mechanical requirement. CNPEC/ECE together with THM analyzed the anticorrosion and strengthening mechanism of Material Z3CN2009M （French Brand） for Cast piece of 1000 MW PWR （CRP1000） Primary Coolant Piping through metallography methods. Strictly internal chemical control range of ladle analysis was established and it makes the mechanical property of cast piece of CPR1000 PWR Primary coolant piping satisfy the RCCM requirement without losing the anti-corrosion property and weld ability. The mechanical test value obtained has rich margin, little degree of separation and exhibits stable quality and advanced overall properties. This success solved the preliminary requirement for domestic making of Primary Coolant Piping of NINGDE and YANGJIANG Nuclear Power Plant and finally realized the domestic manufacturing of Primary Coolant Piping cast piece for CRP1000 Nuclear Power Plant.

Analysis of Differences between Residual Heat Removal Systems in AP1000 and M310

WANG Jian-wei

2009, 30(S2): 11-14,67.

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This paper introduces the design characteristics of the residual heat removal systems in AP1000 which is one kind of the third generation of reactor designed by Westing House Electric Company and M310 designed by China, and analyzes their main differences. Through the comparison of these two systems, M310 RRA system is improved in terms of process system, and thus its reliability and safety are improved.

Comparison Capability between STORK Deaerator and Tray-Spray Steam Deaerator

LIU Xing, LI Ping-yang, JIANG Cheng-ren

2009, 30(S2): 15-18,74.

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This article briefly describes the deaeration principle and the main functions of the deaerator, introduces the detailed structure and the deaeration principle of STORK deaerator used in LAII （Ling’ao Nuclear power station phase II） and deaerator used in HYH and ND power station. After analyzing the different influence by these two types deaerators on the safety operation and the system design of the nuclear power station, the optimized system design proposal of the deaerator of HYH and ND power station was presented with examples. The conclusion is the safety and the economy of improved Tray-spray steam deaerator is more higher than the other one.

Nuclear Power Station Deaerator Transient Study

JIANG Cheng-ren, ZHANG Shi-jun

2009, 30(S2): 19-22.

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Studying how it develops about the thermodynamics parameters of feedwater systems in case of the load rejection conditions about nuclear power station,building mathematical control models,a program software which can simulate the conditions is built up,and the deaerator pressure, feedwater temperature, Net Positive Suction Head Available（NPSHA） for feedpump are worked out.The results can help supply a better data about the height of deaerator and a efficient control measure that keep the feedpump safe under the transient condition.

Analysis of Turbine Fire Protection System of Ling’ao Nuclear Power Plant Phase II

PEI Yan, YAN Li-jing, JIANG Cheng-ren

2009, 30(S2): 23-25.

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This paper mainly analyzes the design and existent problems for the fire protection system used in the turbine bearing of Ling’ao nuclear power plant Phase I, and introduces the design scheme and startup fashion of that in Phase II. The comparison and analysis results for the fire protection systems of Phase I and Phase II indicate that the design of the fire protection system of Phase II meets the requirements of all relevant specifications, and thus the design is safe and reliable.

RETRAN Modeling for CPR1000 Designed Transient Calculation

ZHOU Jing

2009, 30(S2): 26-30.

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This paper discusses the method about RETRAN modeling for designed transient calculation of CPR1000. Firstly, the geometric parameters of the main equipments in the nuclear island are calculated. Secondly, according to the needs of stress calculation, we simulate the main equipments in a detailed way during which a large amount of volume and heat structures are modeled. As a result, the model basically meets the requirement of transient analysis and addition in stress analysis.

Analysis of Thermal Stratification for Pressurizer Surge Line

LI Shu, CAO Xiao-wei

2009, 30(S2): 31-34.

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Thermal stratification is an asymmetrical temperature distributing phenomena which is due to the hot liquid and the cold liquid which have low velocity are not mixed. There are thermal stratification phenomena in the pressurizer surge line. It’s of great importance for the safety operation of nuclear power plants to research the thermal stratification in the pressurizer surge line. This paper theoretically analyzed the origin for thermal stratification, and numerically simulated the thermal stratification phenomena. The stratification transients are completed in different cross-sections of the surge line.

Technical Plan for Improving Reliability of Main Control Room Communication System in Nuclear Power Plants

JIANG Shang-yue, CAO Yu, ZHAI Shou-yang, DENG Xiao-fei

2009, 30(S2): 35-38.

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This paper introduces the present situation of main control room communication system of Daya Bay and Ling’ao nuclear power stations. According to the design principle of nuclear power plant communication system, it thoroughly researches the reliability of main control room communication system in nuclear power plant, and puts forward an optimized technical plan for main control room communication system which combines all sub-communication systems together.

Nuclear Power Station Condenser Unavailable Calculation and Analysis

JIANG Cheng-ren, DING Jia-peng

2009, 30(S2): 39-44.

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If the pressure of condenser is higher than the set-point value defined as "condenser unavailable", it will be forbidden to dump any steam into the condenser. As a requirement of CPR1000 nuclear power station, bypass steam must be discharged to condenser at least 10-12 seconds incessantly after the reactor has tripped, otherwise the pipes of the main steam system will be overpressure. As the pressure of the condenser will rise rapidly during some transient conditions, a transient calculation about condenser is needed to meet the requirement of safe dumping steam for nuclear island. A transient calculation in this paper will simulate the pressure variety of condenser related with turbine trip and reactor trip, and the technical solution about avoiding main steam system overpressure is analyzed for safe dumping steam from the nuclear island.

Comparison of CPR1000 and AP1000 Rod Position Indication Systems

LEI Qing

2009, 30(S2): 45-48.

Abstract(11) PDF(1)

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This paper introduces the structure, the function, the digital detection principle of reactor control rod position and monitoring systems in CPR1000 and AP1000, comparing with the characteristics of the system design. The results show that the operation mode and function of AP1000 Rod position indication system are similar to that of CPR1000, but AP1000 rod position system provides higher reliability, and reduces the numbers of containment electrical penetrations and is with better characteristics than that of CPR1000, since it incorporated the redundancy design and data communication.

Implementation of Nuclear Power Plant Simulation in Start-up Commissioning of Reactor Control System

YANG Zong-wei, HUANG Tie-ming, FENG Guang-yu, LUAN Zhen-hua, LIN Meng, ZHU Li-zhi

2009, 30(S2): 49-53,59.

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Based on the nuclear power thermal-hydraulic model, Labview graphical programming language and virtual instrument data acquisition technology, this paper describes a dedicate test platform to solve the problem that the reactor control system （RRC） can not be evaluated and analyzed far before the actual startup of the unit. By connecting the test platform to the nuclear Digital Control System （DCS）, the step-by-step closed-looped test and global function test of RRC system were performed, the dynamic validation and logical function demonstration for RRC were realized, and a lot of configuration mistakes of RRC and nonconformity were solved. The test for unit 3 of Ling’ao phase II has proved that the implementation of nuclear power plant simulation in the start-up commissioning of RRC can greatly reduce the risk of normal power operation and great transient tests, with which the term of startup for overall unit test can be greatly shortened.

Fatigue Analysis Method of RCC-M Class 1 Pressure-Retaining Component

ZHANG Gui-he

2009, 30(S2): 54-59.

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In this paper, a fatigue analysis method suitable for RCC-M class 1 component was presented based on Miner linear cumulative damage theory, rain-flow counting method and RCC-M requirement. it also provided an optimized fatigue analysis method-Group Combination method by considering the actual operation condition. One simulated example was induced to present this optimized method. The result shows that Group Combination method allows to perform the analysis with more reality to the real situation and allows to obtain the result with more accuracy.

Study on Localization Technology of Bolt Fasteners for CPR1000 Internals

SHI Yao-xin, ZHANG Ming-qian

2009, 30(S2): 60-62.

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To ensure the reliability and safety of the screw connection in reactor operation, the screw tightening torque of CPR1000 internals is as a start for the study. The step and method for replacing the inch screw by meter screw in the transformation design of CPR1000 internals is discussed, and screw structural elements which have influence on screw tightening torque are given in this paper.

Critical Path Analysis in Early Stage of Nuclear Power Project

XIE A-hai

2009, 30(S2): 63-67.

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The technical program and contract model as well as project management system and preliminary design defined in the early stage of nuclear power project are the key condition impact on the quality, schedule and cost of the nuclear power project. This paper, taking the CPR1000 coastal nuclear power station as an example, analyzes the critical path in the early stage of nuclear power project for five fields, i.e. licensing, design and procurement, site preparation, tender of construction contracts and construction preparation, and organization.

Evaluation of Foreign Currency Payment Ability of China Nuclear Power Engineering Corporate

JIANG Zhi-qiong, LU Gang, ZHANG Qi-bo, WEN Sui-ru, WU Wei-wei

2009, 30(S2): 68-74.

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Through this paper, after making a detailed research into the current foreign currency policy of China and the experience of China Nuclear Power Engineering Corporate（CNPEC） during LA2 project, the author evaluates the current ability of foreign currency settlement ability and defines the applicable process in CNPEC, in order to meet the future needs of CNPEC to make foreign currency payment for the multiple nuclear power projects. To ensure the foreign payment can be settled successfully, CNPEC should pay more attention to the import duty, foreign currency loan, clearing method, remittance after verification, as well as the financial risk management of foreign currency loan. On the premise that CNPEC can also get entitlement of import duty and value added tax preference, the author makes the point about how to enhance the capacity of foreign currency payment of CNPEC.

Optimization of Management and Implementation on Nuclear Equipment Quality Surveillance Work

GUO De-peng, WANG Yong-jiao

2009, 30(S2): 75-78,87.

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Nuclear equipment manufacturing quality surveillance is an important part of the quality management work in the nuclear power plant construction. China Nuclear Power Engineering Co. LTD optimizes the management and implementation of surveillance work based on the cooperation with other top-ranking surveillance teams and the experience of nuclear power stations, such as Daya Bay and Ling’ao Nuclear Power Projects. The optimization effect can be showed as following items: the organization acclimatizes itself to the need of project; the surveillance doctrine becomes more clear, and the surveillance management becomes intensive from extensive.

Software Requirements Management Based on Use Cases

XIAO Jin

2009, 30(S2): 79-83,99.

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In this paper, the requirements management based on use cases is theoretically explored, and a multi-layer use-case model is introduced, which combined with three levels of use cases and a single use-case refinement model. Through the practice in a software project, the multi-layer use-case model provides a good solution on how to control the requirements scope and change, and provides the balance of work assignment between customer departments, information management departments and software development outsourcing team.

Application Exploration of Primavera6.0 for Design-Schedule in Nuclear Power Station

ZHANG Xiao-ping

2009, 30(S2): 84-87.

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The ideas and flow of plan establishment during the nuclear power station design is discussed in this paper, including the main steps in the plan establishment, and the design schedule and the management of documents and interfaces are improved. An integral design management innovation is proposed for the management of "schedule, interfaces and drawings and files".

Study on Neutron Transfer Matrix

XUE Bin, FAN Zhi-guo, LI Jing, LI Junde

2009, 30(S2): 88-91.

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The physical principle of the neutron transfer matrix is presented on the basis of simplified hypothesis, and then a simple mathematical model is induced. According to this model, the conjugate gradient theorem of matrices is introduced to explore a method of calibration this matrix from the signals by RIC system and RPN system. This method indicates that the model induced in this paper is reasonable and practicable. The neutron transfer matrix calculated by the method explored in this paper, can be implanted into the LOCA surveillance system （LSS）, and the axial power distribution and the axial power density of the reactor processed by the LSS can be displayed on real time.

Recalculation of RPS Probabilistic Safety Assessment Results of Ling’Ao Nuclear Power Plant Units 3＆4

LIU Jing-jing, HAN Pin-lin

2009, 30(S2): 92-99.

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Since Ling’Ao Nuclear Power Plant Units 3＆4 （L2）, the CPR1000 Nuclear Power Plants （NPP） use Digital Control System （DCS） as their control means. And more advanced third generation NPP also implements DCS. The reliability of Reactor Protection System （RPS） is very important to plant safety. However, because of many redundancy equipments are used in DCS part of the RPS, it’s complicated to model RPS in detail. Until now, there have been few research in-depth. To evaluate the effect of DCS to plant safety, one reactor trip fault tree and one engineered safety features actuation fault tree are built, referring to SIEMENS’ research on RPS of L2. Both input data provided by SIEMENS and calculated by using failure rates, testing periods, and β factors are used to recalculate the RPS Probabilistic Safety Assessment （PSA） results. And more actual results, such as dominant cut sets and failure probability are obtained. According to the recalculated results, the failure probability of reactor trip is 5.5×10^-8 when considering double redundancy. This value satisfies the reliability target, which is 1.0×10^-7. And the failure probability of auxiliary feedwater motor-driven pump actuation signal is 5.21×10^-6, or 8.32×10^-6, which also satisfies its reliability target （1.0×10^-5）. The DCS instrumentation and control system of NPP fulfills its reliability target.

Development of Block-Based Logic Diagram and Analog Diagram Design Platform

XIA Zu-guo, JIANG Guo-jin, XU Xiao-zhen, ZHANG Fan, DENG Tian, YE Lin

2009, 30(S2): 100-104.

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LD/AD design platform, which is based on AutoCAD and applies a block-based design method, has been developed to fulfill the requirements of I＆C （Instrument and Control） and Electrical design and V＆V simulation platform in nuclear plant projects. The LD/AD design platform has been also successfully applied in all design phases for CPR1000 projects. As a result, the design platform can keep consistency of data, relieve the design difficulty and improve the efficiency.

Research on Application of Kalina-Cycle in Second-Loop System of Nuclear Power Plants

ZHANG Yi-min, WANG Xue-feng, XIONG Xing-cai

2009, 30(S2): 105-108.

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On the basis of Rankine-Cycle, this paper introduces the characteristics and innovative advantages of Kalina-Cycle, and briefs the development status in other countries. It mainly analyzes and describes the feasibility of applying Kalina-Cycle in second-loop of nuclear power plants academically. The conclusion indicates that by applying Kalina-Cycle, the thermal efficiency of nuclear power plants can be increased by above 10% and the costs of investment in circulating cool-water system and reheating system can also be decreased considerably. However, the change of medium in second-loop of nuclear power plants will change the heat energy system accordingly, in addition, the change of thermal system will lead to the re-R＆D of the whole thermal system facilities.Whether Kalina-Cycle can be used in the second-loop system of nuclear power plant, there are still many problems to be reviewed.

2009 Vol. 30, No. S2