核动力工程

Study on the Strategy of Advanced Fuel Management for Ling'ao Nuclear Power Station

GAO Ligang, HAN Qinghao, XIAO Min, LI Xiaoming

2005, 26(S1): 1-6.

Abstract(3) PDF(0)

Abstract:
This paper introduces the strategy and consideration in Advanced Fuel Management for Ling’ao Nuclear Power Station,and the differences of advanced fuel management between Ling’ao and Daya Bay Nuclear Power Stations.The overall transition mode of Ling’ao fuel management has been analyzed,especially the transition strategy and steps for the advanced fuel management after the mixed core（from cy-cle 2-cycle 5）.It has been expatiated what would be the most suitable specific advanced fuel management strategy（yearly quarter core fuel management,or 18 month fuel management） from cycle 6 for Ling’ao Nuclear Power station.The feasibility of Advanced Fuel Management Strategy has been demonstrated from the view points of economy,advantage and the adopted advanced technology in the yearly quarter core fuel management project.It is concluded,through the comparison of different advanced fuel management strategies（yearly 1/4 fuel management,18 month fuel management and 24 month fuel management）,that the implementation of yearly 1/4 fuel management in Ling’ao Nuclear Power Station can carry forward the advantages and avoid the shortage in the fuel management strategy,and it is very economical and suitable for the non-summer outage planning.Although 18-month fuel cycle has many advantages in general,the market and pricing for Ling’ao Nuclear Power Station are quite different from that of Daya Bay.The power grid has more rigorous limitation on Ling’Ao Nuclear Power Station.The seeking for more power generation than planned is not the best way for Ling’ao Nuclear Power Station.

Study & Implementation of Mixied Core & Increased Enrichment in Ling'ao Nuclear Power Station

XIAO Min, ZHANG Hong, HUANG Wenyou, ZHANG Shishun

2005, 26(S1): 7-10,18.

Abstract(4) PDF(0)

Abstract:
The first cycles of Ling’ao and Daya Bay Nuclear Power Stations are the same.It is the challenge for Ling’ao NPS to design the fuel management mode and how to utilize the successful experiences from Daya Bay NPS.The fuel management design for the second cycle is rather difficult in normal conditions.There must be clear answers to the questions that whether or not to change the fuel type from the second cycle and which type of fuel to be used for the change（AFA 3G with or without MSMG）.If the same fuel type of AFA 3G with MSMG in Daya Bay NPS would be used in Ling’ao second cycle,it would be the mixed core and there should be special safety analysis and licensing issues.This paper introduces the overall optimization and design from cycle 2 to 5.

Primary Study on Core Concept Design and Safety Margin of CPR1000

XIAO Min, HAO Sixiong, HAN Qinghao, LI Xianfeng, LIU Daohe

2005, 26(S1): 11-18.

Abstract(8) PDF(1)

Abstract:
CRP1000 is the China PWR design based on step-by-step improvement and innovation by China Guangdong Nuclear Power Company（CGNPC） over 20 years.The reference design of CPR10000 is Ling’ao II plus some improvements.CPR1000 will be the main PWR to be built in China by CGNPC within the next 10-15 years.The prerequisite is to define the fuel loading pattern and the fuel management mode for the CPR1000 initial core design,which is the base for CPR1000 nuclear island design,safety analysis and licensing.This paper summarizes the CPR1000 initial core concept design,fuel management design and the safety evaluation based on Daya Bay and Ling’ao operation and fuel management experiences and international technology development.This paper proposes the fuel type to be used for CPR1000 initial core design,analyze the 18-month fuel cycle fuel,the management strategy and design,and conduct the preliminary safety evaluation for CPR1000.

Rod Drop Accident Analysis for Mixed Core in Ling'ao Nuclear Power Station

ZHANG Shishun, ZHOU Zhou

2005, 26(S1): 19-23.

Abstract(3) PDF(0)

Abstract:
AFA 2G assemblies in Ling’ao nuclear power station were replaced gradually by AFA 3G assemblies from cycle 2 and subsequent cycles.The enrichment of the fuels was increased from 3.2% to 3.7% from cycle 3 in Ling’ao.Therefore,the study of Ling’ao mixed core and increased enrichment was launched in 2001.Rod drop accident for Ling’ao Nuclear Power Station was reanalyzed in the frame of Framatome ANP analytical methodology.The analytical codes used in the accident analysis includes SCIENCE,ESPADON,CINEMA,CANTAL and FLICA III.For detected dropped rod configurations,the negative flux rate protection system actuates a reactor trip.For the non-detected dropped rod configurations,the minimum DNBR values have been evaluated with the conservative analysis methodology and assumptions,and the DNBR fuel design limit is respected.

Analysis of Effect of Steam-Line Break Accident on Core of Ling'ao Unit 1 Cycle 3

ZHANG Shishun, SUN Jiliang, XIAO Min

2005, 26(S1): 24-29.

Abstract(3) PDF(0)

Abstract:
In the specific reload safety evaluations in Daya Bay and Ling’ao nuclear power stations,the steam line break（SLB） accident often resulted in the core parameters exceeding the safety limits during recent years.This leads to the reanalysis of SLB accident.Therefore,the analysis of effects of SLB on core is very important in the reload design in Daya Bay and Ling’ao nuclear power stations.This paper presents the analysis of the effect of SLB on core for Ling’ao unit 1 cycle 3,based on the Framatome ANP core design procedure manual.The analytical results show that,the core designed by DMNC for Ling’ao unit 1 cycle 3 is safe,during the SLB accident.

Revision of Emergency Operation Procedure and Site Emergency Plan on Implementation of SAMG in Daya Bay Nuclear Power Station

HUANG Huizhang, YANG Ling, SUN Jiliang, LIN Shumou, XIAO Min

2005, 26(S1): 30-32,44.

Abstract(4) PDF(0)

Abstract:
The report describes the definition of interface between EOP and Severe accident management guidance in GNPS,and also introduces the revised content of GNPS Site Emergency Plan in the implementation of SAMG.

Review and Validation of Severe Accident Management Guideline in Daya Bay Nuclear Power Station

XIAO Min, SUN Jiliang, ZHANG Shishun, YAN Shu, AN Junjing, LU Xianghui

2005, 26(S1): 33-34,51.

Abstract(3) PDF(0)

Abstract:
The report introduced the validation process of Severe Accident Management Guideline（SAMG） in Daya Bay Nuclear Power Station,mainly the validation plan and organization.The action of review and validation is required according to the arrangement of the project plan to develop SAMG in Daya Bay Nuclear Power Station.

Safety Analysis Focused on Fuel Reloading in Nuclear Power Station in Guangdong Province

LI Xianfeng, LIANG Wei, LI Lei, XIAO Min

2005, 26(S1): 35-38.

Abstract(5) PDF(0)

Abstract:
Reloading error took place at Dampierre nuclear power Station unit 4 in April 2001 in French,which lowered the sub-critical margin.Although “the snake” reloading process is adopted in nuclear power Station in Guangdong Province,it is still necessary to perform the safety analysis focused on the similar accident during the fuel reloading,in order to verify the safety of fuel reloading process in nuclear power Station in Guangdong province.Based on the conservative calculation considering the disadvantageous fuel cycle of Ling’ao and Daya Bay reactors,the results show that the similar reloading error will result in the arrangement of new fuel assemblies in one place,but the sub-criticality required by FSAR can be satisfied and safety can be guaranted.

Strategy for Core Control when PWR Returning to Full Power Operation after Shutdown

LIAN Zhikun, LI Xianfeng, PENG Song

2005, 26(S1): 39-44.

Abstract(3) PDF(0)

Abstract:
There are difficulties in the control of reactor core when the reactor returning the full power operation after shutdown during end of cycle.The physics factors and process that with effects on the control have been analyzed,and a strategy,which based on the selection of recriticality parameters with using of power bank withdraw and boron concentration dilution,is applied.The test results demonstrate the validity of the strategy,which could be used in the same condition of other PWR nuclear power units.

Fuel Management Study on 18-Months Fuel Cycle for Ling'ao Phase-II Nuclear Power Station

LI Lei, XIAO Min, LIANG Wei

2005, 26(S1): 45-47.

Abstract(1) PDF(0)

Abstract:
A typical 18-months fuel cycle strategy is introduced in this paper.It deals with the transition towards 18-months reload from the first cycle and reaching equilibrium cycle through the minimum cycles at Ling’ao Phase-II Nuclear Power Station.Cycle length,gadolinium use,loading pattern,as well as main key safety parameters are given.The fulfillment of the study demonstrates the feasibility of launching 18-month reload at Ling’ao phase-II and offers theoretical basis for determining the final fuel management strategy.

Application of Three-Dimensional Theoretical Calibration Curve at Grey Rod-Turbine Load Test

LI Lei, LU Xianghui, LI Xianfeng

2005, 26(S1): 48-51.

Abstract(1) PDF(0)

Abstract:
At Daya Bay and Ling’ao nuclear power stations,the present periodical test procedure stipulates: Grey rod-turbine load test is to be performed to verify the accuracy of calibration curve by quickly reducing the power.As Daya Bay nuclear power station has entered the 18-month fuel cycle,the bias of calibration curve from 1D code is growing larger.In this paper,a new solution of 3D methodology has been presented and the efficiency has been demonstrated through the comparison with the site test.

Analysis of NIS PARs Installation in Daya Bay Nuclear Power Station

ZHANG Shishun, SUN Jiliang, XIAO Min

2005, 26(S1): 52-54.

Abstract(4) PDF(0)

Abstract:
This paper describes the hydrogen production and release,and the effect of NIS PARs on thehydrogen concentration in the containment in Daya Bay nuclear power station.The variation of hydrogenconcentrations in the containment with the assumed number of NIS PARs inside the containment is studied.The results show that the peak hydrogen concentration could be controlled at about 10% with around 50 NISPARs inside the containment.

Study of Station Blackout Caused Severe Accident Process in Daya Bay Nuclear Power Station

QI Panjin, XIAO Min, SUN Jiliang, ZHANG Shishun

2005, 26(S1): 55-57,61.

Abstract(3) PDF(0)

Abstract:
Based on the severe accident calculation results of Daya Bay Nuclear Power Station,this paper analyzes a certain of accident sequence and the mitigating measures of severe accident.The results show that the top of core will be uncovered about 1~2 hours later,and the reactor pressure vessel will fail after 5~7 hours.The containment will fail after about 100 hours.The cavity basement can be penetrated about 5.5m after 8.7 days.The results also show that the pit flooding can prevent the melt-through of the basement and reduce hydrogen mass from interaction of molten corium with concrete.

Analysis of Uncontrolled Boron Dilution Accident at Ling'ao Nuclear Power Station

LI Lei, ZHANG Hong, XIAO Min

2005, 26(S1): 58-61.

Abstract(3) PDF(0)

Abstract:
Boron Dilution Accident（UBDA） analysis plays a significant role in the reactivity accidents during the justification of Ling’ao Mixed Core and Increased Enrichment project.As the conclusion of the accident analysis will define the standard control rod position at both cold and hot shutdown conditions,of which the control rod position of hot shutdown serves as the premise of Rod Withdrawal from Subcriticality（RWS） accident analysis,the rod position of standard hot shutdown can be adjusted through UBDA analysis in an effort to provide the margin for RWS accident analysis.A detailed analysis method and conclusion of BDA at Ling’ao Nuclear Power Station are described in this paper.

Study on Core Power Capability during Stretch-Out for Ling'ao Nuclear Power Station

LI Jinggang, ZHANG Hong, LIANG Wei, LI Lei, MAO Guoping

2005, 26(S1): 62-64.

Abstract(4) PDF(0)

Abstract:
The core power capability during the stretch out for Ling’ao nuclear power station is studied margins increase during the stretch out compared with that in natural cycle.in this paper.The possible power distributions during the stretch out in condition I and condition II are simulated with several neutronic codes.The linear power density and DNBR margins in condition I and condition II,which reflects the core safety,have been justified.It also verified that the core power capability

Exploration & Practice of Reliability Management System for Nuclear Power Station Equipment

LU Wenyue, LI Xiaoming, HAN Qinghao, DAI Zhonghua, HONG Zhenmin, CHEN Shijun

2005, 26(S1): 65-72.

Abstract(5) PDF(0)

Abstract:
Equipment reliability is one of the key factors that maintains the nuclear power plant（NPP） agement and operation of equipments suitable for NPPs.Step by step,the equipment reliability management system is shaped up,which covers the normal operations and outage operations,with the participation of all functional departments of DNMC.This system involves the equipment reliability analysis and maintenance strategy optimization,equipment status monitoring and defects management.

Application and Innovation in Nuclear Power Stations of Reliability-Centered Maintenance

LI Xiaoming, CHEN Shijun, WU Tao, ZHOU Shiliang, DAI Zhonghua, LU Wenyue

2005, 26(S1): 73-77.

Abstract(6) PDF(0)

Abstract:
Reliability-Centered Maintenance（RCM） is a method for maintenance optimization,which is used worldwide.In this paper,RCM method is briefly introduced,the implement of RCM in Daya day nuclear power station and the modification of the RCM decision logic diagram are addressed,and finally the effects of the implement after the modification is presented.

Technical-Analysis Centered Maintenance of Instrumentation and Control Equipment

LU Wenyue, HONG Zhenmin, ZHANG Ruiqiong

2005, 26(S1): 78-81.

Abstract(3) PDF(0)

Abstract:
In order to increase the reliability of instrumentation and control（I＆C） equipments,the method of Technical-analysis Centered Maintenance（TCM） of I＆C Equipment was developed in Daya Bay nuclear power station.This paper explained the characteristics of technical study and its structure and the system for prediction of the equipment status and trend analysis,and introduced the applications of the maintenance method on I＆C equipment.

Application of Root Cause Analysis on Equipment Failure to Improve the Safety Performance of Key Equipments in Nuclear Power Station

GAO Ligang, LU Qunxian

2005, 26(S1): 82-86.

Abstract(3) PDF(0)

Abstract:
This paper describes briefly the methodology of Root Cause Analysis（RCA） on equipment failure,and introduces the improvement made on RCA personnel,training,job organization,and correct action management in Guangdong Nuclear Power Station.The examples and results show that RCA is a very useful tool to improve the safety performance of equipment.The suggestions to further enhance RCA in Nu-clear Power Plants are also mentioned.

Aging and Life Management of Daya Bay Nuclear Power Station

DAI Zhonghua, LIU Peng, LU Wenyue, CHEN Shijun

2005, 26(S1): 87-89,115.

Abstract(5) PDF(0)

Abstract:
This report presents the aging management method recommended by IAEA and the proposed methods for the ageing and lifetime management of Daya Bay Nuclear Power Operation and Management Co.Ltd.（DNMC）.Firstly,the aging and life management focuses on the key equipment of the nuclear power plants.With the accumulated of experiences and aging data,the emphasis of the aging and lifetime management will be the understanding of the aging mechanism,the selecting of components susceptible to the aging,and then the analysis of ageing trend and evaluation of the lifetime of the key equipments.

Key Point Analysis for the Maintenance of Turbine

CHEN Zengxing

2005, 26(S1): 90-92,119.

Abstract(2) PDF(0)

Abstract:
This paper presents the key points for the maintenance of large turbines,and suggests that in the maintenance process,the clearance of components,alignment of shaft,bearings and sliding pin of the turbines shall be considered,and at the same time related data shall be carefully studied,and the analysis and evaluation shall be based on all the related factors.In this way,rational solutions may be proposed to ensure the effective maintenance and the safe and effective operation of the turbine unit.

Thermal Aging and Aging Management of Cast Stainless Steel in LWR Nuclear Power Station

LIU Peng, XUE Fei, DAI Zhonghua, CHEN Shijun, ZHU Wenbin, WANG Xiaolong, TI Wenxin

2005, 26(S1): 93-96.

Abstract(4) PDF(0)

Abstract:
Austenitic Stainless Steel Castings in LWR nuclear power plants are susceptible to thermal aging at operating temperature.This article describes the mechanism of thermal aging for CASS components.It gives a general procedure for thermal aging management in nuclear power plants（three steps: susceptibility screening,assessment of degree of material embrittlement and ISI programme renew） on the basis of the aging management practices.Then it gives some advices about thermal aging management of nuclear power plants in China.

Establishment of Alarm Value for Aging Susceptibility Point of Valve Subassembly and Countermeasures for Aging Mitigation

LIANG Hansheng, JIANG Dinggong, ZHANG Chaowen, LU Wenyue, ZHANG Hongquan

2005, 26(S1): 97-102.

Abstract(4) PDF(0)

Abstract:
The article analyzes the aging mechanism and the possible failure modes of safety class 1 Motor-valve subassembly with dead pipe phenomenon in Daya Bay NPP;The most important aging susceptibility point is screened and discussed,and then the alarm value is established,Finally,some methods for aging mitigation adapted in Daya Bay NPP are brought forward and summarized.

Effect of Operation Conditions of Primary Coolant System of PWR NPP on Mechanic Performance of Primary Coolant Pump

WANG Qinhu, LI Shekun, LU Wenyue, YU Haifeng, CAO Zhipeng

2005, 26(S1): 103-108.

Abstract(3) PDF(0)

Abstract:
This paper analyzed the effect of the change of operation conditions of the primary coolant system of 1000MW pressurized water reactor nuclear power plant on the main mechanic parameters of the primary coolant pump,and provided the theoretical evidences to prevent the primary coolant pump from the influence of change of operation conditions and improve the mechanical performance of the pump through the adjustment of system parameters.

Root Cause Analysis on Cracks of Pistons of Emergency Diesels of Ling'ao Nuclear Power Station

HU Dingjin, LI Weijun

2005, 26(S1): 109-115.

Abstract(5) PDF(0)

Abstract:
As a middle voltage auxiliary power source,when nuclear power plant loses all of its outer power supplies,the emergency diesel generation sets will provide emergency power supply to the important safety systems in the plant,to ensure that the reactors may shutdown safely and to protect the important equipments from being damaged.At the end of 2003 and the beginning of 2004,after the disassembling of all 8 diesels in Ling’ao plant,plenty of remarkable cracks were found on 29 of all their 96 pistons.Those serious defects will probably violate the availability of those emergence diesel unitsand endanger the plant safety.This article introduces the analysis process on causes and mechanism how those cracks produced,and a risk analysis on cracks of pistons.

Cause Analysis of the Acetylene Produced in Main Transformer Phase B in LNPS

GUAN Jianjun, HAN Qinghao, YANG Jicheng

2005, 26(S1): 116-119.

Abstract(3) PDF(0)

Abstract:
In the analysis of main transformer phase B of unit 2 in Ling’ao Nuclear Power Station（LNPS） by chromatography,acetylene was found.In order to carry out the inside check,the main transformer had been stopped and restored after necessary maintenances.But shortly,0.86 μL·L^-1 acetylene was measured,so the transformer had to be replaced for the safe operation in a new cycle.The paper analyzed the most possible causes based on the check results after lifting the tank of transformer,and the new corrective measurements and the considerations were provided to improve the transformer reliability.

Treatment of High Pressure Difference of RRI/SEC Heat Exchanger in Daya Bay Nuclear Power Station

CAO Zhipeng, ZHANG Chaowen, DAI Zhonghua, WANG Qinhu

2005, 26(S1): 120-122.

Abstract(5) PDF(0)

Abstract:
This paper have analyzed the reason of high pressure difference of RRI/SEC heat exchanger inDaya Bay nuclear power station,by adopting the Root Cause Analysis method（RCA） and the analysis matrixof failure mode.After few experiments have been done on the SEC system,we confirm the reason of theexchanger failure.Through modifying the method of adding chemicals to SEC system,the exchanger failureis solved satisfactorily.Two years of operation indicate that the 8 exchangers ever with failure can operatesatisfactorily,and the pressure difference is in the normal range.

Root Cause Analysis for Micro-Governor Valve Module DC-DC Converter Failure in Daya Bay Nuclear Power Station

XIA Hongwei, LU Wenyue, QU Yueqi

2005, 26(S1): 123-125.

Abstract(4) PDF(0)

Abstract:
Root cause analysis about DC-DC converter failure in Micro-governor valve module in unit 2 of Daya Bay Nuclear Power Station revealed that poor quality of the transistor material inside the converter,poor thermal design of the valve modules and insufficient ventilation to the cabinet resulted in the aging of converter after long-term operation.Followed correction action had been suggested and implemented on site.

2005 Vol. 26, No. S1