• 제목/요약/키워드: Main Feedwater System

검색결과 21건 처리시간 0.025초

Piping Failure Frequency Analysis for the Main Feedwater System in Domestic Nuclear Power Plants

  • Choi Sun Yeong;Choi Young Hwan
    • Nuclear Engineering and Technology
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    • 제36권1호
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    • pp.112-120
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    • 2004
  • The purpose of this paper is to analyze the piping failure frequency for the main feedwater system in domestic nuclear power plants(NPPs) for the application to an in-service inspection(ISI), leak before break(LBB) concept, aging management program(AMP), and probabilistic safety analysis(PSA). First, a database was developed for piping failure events in domestic NPPs, and 23 domestic piping failure events were collected. Among the 23 events, 12 locations of wall thinning due to flow accelerated corrosion(FAC) were identified in the main feedwater system in 4 domestic WH 3-loop NPPs. Two types of the piping failure frequency such as the damage frequency and rupture frequency were considered in this study. The damage frequency was calculated from both the plant population data and damage(s) including crack, wall thinning, leak, and/or rupture, while the rupture frequency was estimated by using both the well-known Jeffreys method and a new method considering the degradation due to FAC. The results showed that the damage frequencies based on the number of the base metal piping susceptible to FAC ranged from $1.26{\times}10^{-3}/cr.yr\;to\;3.91{\times}10^{-3}/cr.yr$ for the main feedwater system of domestic WH 3-loop NPPs. The rupture frequencies obtained from the Jeffreys method for the main feedwater system were $1.01{\times}10^{-2}/cr.yr\;and\;4.54{\times}10^{-3}/cr.yr$ for the domestic WH 3-loop NPPs and all the other domestic PWR NPPs respectively, while those from the new method considering the degradation were higher than those from the Jeffreys method by about an order of one.

DES를 이용한 SIR의 주급수계통의 고장진단 (Failure Diagnosis of Main Feedwater System for SIR using DES)

  • 박준효;김희표;김철수;이석
    • 한국정밀공학회:학술대회논문집
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    • 한국정밀공학회 2001년도 춘계학술대회 논문집
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    • pp.570-573
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    • 2001
  • Safety is very important to operate nuclear power plant. To have the safety, nuclear power plant should be run without trouble. This paper presents the application of a failure diagnosis approach based on discrete event system theory to the Main Feedwater System for Safe Integral Reactor.

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웨스팅하우스형 원전의 보조급수계통 설계변경 영향 평가 (A Safety Improvement for the Design Change of Westinghouse 2 Loop Auxiliary Feedwater System)

  • 나장환;배연경;이은찬
    • 한국압력기기공학회 논문집
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    • 제9권1호
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    • pp.15-19
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    • 2013
  • The auxiliary feedwater is an important to remove the heat from the reactor core when the main feedwater system is unavailable. In most initiating events in Probabilistic Safety Assessment(PSA), the operaton of this system is required to mitigate the accidents. For one of domestic nuclear power plants, a design change of a turbine-driven auxiliary feedwater pump(TD-AFWP), pipe, and valves in the auxiliary system is implemented due to the aging related deterioration by long time operation. This change includes the replacement of the TD-AFWP, the relocation of some valves for improving the system availability, a new cross-tie line, and the installation of manual valves for maintenance. The design modification affects the PSA because the system is critical to mitigate the accidents. In this paper, the safety effect of the change of the auxiliary feedwater system is assessed with regard to the PSA view point. The results demonstrate that this change can supply the auxiliary feedwater from the TD-AFWP in the accident with the motor-driven auxiliary feedwater pump(MD-AFWP) unavailable due to test or maintenance. In addition, the change of MOV's normal position from "close" to "open" can deliver the water to steam generator in the loss of offsite power(LOOP) event. Therefore, it is confirmed that the design change of the auxiliary feedwater system reduces the total core damage frequency(CDF).

AN EVALUATION OF THE APERIODIC AND FLUCTUATING INSTABILITIES FOR THE PASSIVE RESIDUAL HEAT REMOVAL SYSTEM OF AN INTEGRAL REACTOR

  • Kang Han-Ok;Lee Yong-Ho;Yoon Ju-Hyeon
    • Nuclear Engineering and Technology
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    • 제38권4호
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    • pp.343-352
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    • 2006
  • Convenient analytical tools for evaluation of the aperiodic and the fluctuating instabilities of the passive residual heat removal system (PRHRS) of an integral reactor are developed and results are discussed from the viewpoint of the system design. First, a static model for the aperiodic instability using the system hydraulic loss relation and the downcomer feedwater heating equations is developed. The calculated hydraulic relation between the pressure drop and the feedwater flow rate shows that several static states can exist with various numbers of water-mode feedwater module pipes. It is shown that the most probable state can exist by basic physical reasoning, that there is no flow rate through the steam-mode feedwater module pipes. Second, a dynamic model for the fluctuating instability due to steam generation retardation in the steam generator and the dynamic interaction of two compressible volumes, that is, the steam volume of the main steam pipe lines and the gas volume of the compensating tank is formulated and the D-decomposition method is applied after linearization of the governing equations. The results show that the PRHRS becomes stabilized with a smaller volume compensating tank, a larger volume steam space and higher hydraulic resistance of the path $a_{ct}$. Increasing the operating steam pressure has a stabilizing effect. The analytical model and the results obtained from this study will be utilized for PRHRS performance improvement.

화력 발전소 드럼형 보일러 시스템의 고장 진단을 위한 퍼지 전문가 시스템의 개발 (Development of Fuzzy Expert System for Fault Diagnosis in a Drum-type Boiler System of Fossil Power Plant)

  • 변승현;박세화
    • 전자공학회논문지B
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    • 제31B권10호
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    • pp.53-66
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    • 1994
  • In this paper, a fuzzy expert system is developed for fault diagnoisis of a drum-type boiler system in fossil power plants. The develped fuzzy espert system is composed of knowledge base, fuzzification module, knowledge base process module, knowledge base management module, inference module, and linguistic approximation module. The main objective of the fuzzy expert system is to check the states of the system including the drum level and detect faults such as the feedwater flow sensor fault, feedwater flow control valve fault, and water wall bube rupture. The fuzzy expert system diagnoses faults using process values, manipulated values, and knowledge base which is built via interviews and questionaries with the experts on the plant operations. Finally, the validity of the developed fuzzy expert system is shown via experiments using the digital simulator for boiler system is Seoul Power Plant Unit 4.

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중수로 원전 가상의 mSGTR과 SBO 다중 사건에 대한 MARS-KS 코드 분석 (Analysis on Hypothetical Multiple Events of mSGTR and SBO at CANDU-6 Plants Using MARS-KS Code)

  • 유선오;이경원;백경록;김만웅
    • 한국압력기기공학회 논문집
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    • 제17권1호
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    • pp.18-27
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    • 2021
  • This study aims to develop an improved evaluation technology for assessing CANDU-6 safety. For this purpose, the multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout (SBO) in a CANDU-6 plant was selected as a hypothetical event scenario and the analysis model to evaluate the plant responses was envisioned into the MARS-KS input model. The model includes logic models for controlling the pressure and inventory of the primary heat transport system (PHTS) decreasing due to the u-tubes' rupture, as well as the main features of PHTS with a simplified model for the horizontal fuel channels, the secondary heat transport system including the shell side of steam generators, feedwater and main steam line, and moderator system. A steady state condition was successfully achieved to confirm the stable convergence of the key parameters. Until the turbine trip, the fuel channels were adequately cooled by forced circulation of coolant and supply of main feedwater. However, due to the continuous reduction of PHTS pressure and inventory, the reactor and turbine were shut down and the thermal-hydraulic behaviors between intact and broken loops got asymmetric. Furthermore, as the conditions of low-flow coolant and high void fraction in the broken loop persisted, leading to degradation of decay heat removal, it was evaluated that the peak cladding temperature (PCT) exceeded the limit criteria for ensuring nuclear fuel integrity. This study is expected to provide the technical bases to the accident management strategy for transient conditions with multiple events.

Immune Based Intelligent Tuning of the 2-DOF PID Controller for Thermal Power Plant

  • Kim, Dong-Hwa
    • 제어로봇시스템학회:학술대회논문집
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    • 제어로봇시스템학회 2002년도 ICCAS
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    • pp.101.3-101
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    • 2002
  • Contents 1 Abstract- In the thermal power plant, there are six manipulated variables; main steam flow, feedwater flow, air flow, spray flow, fuel flow, and gas recirculation flow. Therefore, the thermal power plant control system is a multi-input and output system. In the control system, the main steam temperature typically is regulated by the fuel flow rate and the spray flow rate, and the reheater steam temperature is regulated by the gas recirculation flow rate. Up to the present time, the PID controller has been used to operate this system. This paper focuses on the characteristic comparison of the PID controller, the modified 2-DOF PID Controller on the DCS, in order to design an optimal...

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신형경수로(APR1400)의 터빈 싸이클 열성능 분석 (Turbine Cycle Thermal Performance Analysis of Advanced Power Reactor 1400)

  • 정대율;임혁순;정대욱;허균영
    • 대한기계학회:학술대회논문집
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    • 대한기계학회 2001년도 춘계학술대회논문집D
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    • pp.343-347
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    • 2001
  • Advanced Pressurized Reactor 1400(APR-1400), which is a standard evolutionary advanced light water reactor(ALWR), has been developed from 1992 as one of long-term Government Project(G-7). The APR-1400 is designed to operate at the rated output of 4000MWt to produce an electric power output of around 1450MWe. The balance of plant (BOP) for the secondary system consists of main steam, feedwater, condensate, turbine generator and auxiliary system. In this paper, we describe the major design features of secondary component, balance of plant configuration, and then the turbine cycle thermal performance evaluation using PEPSE code.

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국내 안전등급 배관에 대한 손상사례 분석 (Piping Failure Analysis In Domestic Nuclear Safety Piping System)

  • 최선영;최영환
    • 대한기계학회:학술대회논문집
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    • 대한기계학회 2003년도 춘계학술대회
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    • pp.617-621
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    • 2003
  • The purpose of this paper is to analyze piping failure trend of safety pipings In domestic nuclear power plants. First, database for the piping failure was constructed with 105 data fields. The database includes plant population data, event data, and service history data. 7 kinds of piping failures in domestic NPPs were investigated. Among the 7 cases, detailed root causes were investigated for 3 cases. The first one is pipe wall thinning in main feedwater pipings of Westinghouse 3 loop type plants. The root cause of the wall thinning was flow accelerated corrosion near welding area. The next one is leak event in chemical and volume control system(CVCS) due to vibration. Some cracks occurred in socket welding area. The events showed that the integrity or socket weld is very vulnerable to vibration. The last one is also a leak event in primary sampling line in Korean standard reactor due to thermal fatigue. Although the structural integrity was not maintained by the events, there was no effect on nuclear safety in the above 3 piping failure eases.

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울진1,2호기 출력최적화 및 증기발생기 교체가 주급수 제어계통 안정도에 미치는 영향연구 (Research on a Stability of Feedwater Control System after Stretched Power Uprate and Replacement Steam Generator for Ulchin Units 1&2)

  • 윤덕주;김인환;이재용
    • 한국압력기기공학회 논문집
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    • 제8권2호
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    • pp.14-20
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    • 2012
  • Full load rejection capability of nuclear power plant depends primarily on steam dump capacity (SDCAP) and steam generator level control capability. Recently, Ulchin Units 1&2 have performed stretched power uprate (SPU) and replacement steam generator (RSG) projects, which increase the power by 4.5 percent. They change major design or operating parameters and especially reduces steam dump capacity at full power due to increase of the steam flow. The reduction of SDC after SPU results in degradation of heat removal capability in full load rejection transients. Therefore, we should perform evaluation to determine whether reactor trips occur in large load rejection transients. Uchin Units 1&2 have experienced full load rejection (FLR) three times from 2004 to 2010. Operating data from the plant occurrence of FLR at Ulchin Units 1&2 showed that steam generator (SG) level transients were limiting in point of reactor trip. However the plant had never reached reactor trip in the FLR and successfully continued in house load operation. The parameters and setpoints for the SG will be changed if the SG is replaced. Therefore, we evaluated the appropriateness of steam dump, main feedwater and steam generator water level control system preventing the plant from reactor trip in case of FLR by the parameter sensitivity study whether SG water level operated smoothly after SPU and RSG projects.