• 제목/요약/키워드: Main Steam Line

검색결과 69건 처리시간 0.031초

화력발전소에서 응축수 회수계통의 증기배관 개선에 의한 발전시스템의 효율 향상 (The improvement of the efficiency of power plant by the reformation of steam line in the return system)

  • 권영수;서정세
    • 대한기계학회:학술대회논문집
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    • 대한기계학회 2001년도 추계학술대회논문집B
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    • pp.867-871
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    • 2001

  • The main reason for reviewing the condensate water return system in the auxiliary steam system is to obtain the thermal high efficiency of the power plant and thus save the fossil energy in power plant. This study intends to analyze the thermal efficiency of the power plant and predict the increasing in the generator output by the return system reformation of auxiliary steam line in the thermal power plant.

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주증기계통 오리피스 후단 소구경 배관의 감육 및 누설 발생 (Cause Analysis for the Wall Thinning and Leakage of a Small Bore Piping Downstream of an Orifice)

  • 황경모
    • Corrosion Science and Technology
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    • 제12권5호
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    • pp.227-232
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    • 2013

  • A number of components installed in the secondary system of nuclear power plants are exposed to aging mechanisms such as FAC (Flow-Accelerated Corrosion), Cavitation, Flashing, and LDIE (Liquid Droplet Impingement Erosion). Those aging mechanisms can lead to thinning of the components. In April 2013, one (1) inch small bore piping branched from the main steam line experienced leakage resulting from wall thinning in a 1,000 MWe Korean PWR nuclear power plant. During the normal operation, extracted steam from the main steam line goes to condenser through the small bore piping. The leak occurred in the downstream of an orifice. A control valve with vertical flow path was placed on in front of the orifice. This paper deals with UT (Ultrasonic Test) thickness data, SEM images, and numerical simulation results in order to analyze the extent of damage and the cause of leakage in the small bore piping. As a result, it is concluded that the main cause of the small bore pipe wall thinning is liquid droplet impingement erosion. Moreover, it is observed that the leak occurred at the reattachment point of the vortex flow in the downstream side of the orifice.

  • https://doi.org/10.14773/cst.2013.12.5.227 인용 PDF KSCI

주증기관 파단사고에 대한 원자로 용기의 가압열충격 해석 (Pressurized Thermal Shock Analyses of Reactor Pressure Vessel for Main Steam Line Break)

  • 정명조;박윤원;장창희;정일석
    • 한국전산구조공학회논문집
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    • 제12권3호
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    • pp.271-279
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    • 1999

  • 본 연구에서는 국내에서 가장 취약할 것으로 예상되는 원자력 발전소에 가압열충격 사고를 유발할 수 있는 주증기관 파단사고를 가정하여 열수력 해석과 파괴역학 해석을 수행하였다. 원전수명관리연구의 일환으로 계통열수력 해석 및 혼합열유동 해석에 의하여 구한 냉각제의 온도와 압력의 이력 및 용기의 재질성분으로부터 용기의 응력확대계수와 파괴인성치를 계산하고 이들을 비교하여 균열의 진전여부를 판단하여 형상계수가 1/6인 표면균열이 견딜 수 있는 최대 기준무연성천이온도를 결정하였다.

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차세대 원자로 용기내 vessel 내면에서의 대류 열전달특성에 관한 수치해석적 연구 (A numerical study on convective heat transfer characteristics at the vessel surface of the Korean Next Generation Reactor)

  • 정삼두;김창녕
    • 대한기계학회:학술대회논문집
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    • 대한기계학회 2000년도 추계학술대회논문집B
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    • pp.228-233
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    • 2000

  • The Korean Next Generation Reactor(KNGR) is a Pressurized Water Reactor adopting direct vessel injection(DVI) to optimize the performance of emergency core cooling system(ECCS). In a certain accident, however, pressurized thermal shock(PTS) of the vessel due to the sudden contact with the injected cold water is expected. In this paper, an accident of Main Steam Line Break(MSLB) has been numerically investigated with direct vessel injections and an increased volume flow rate in some cold legs. Using FLUENT code, temperature distributions of the fluid in the downcomer and of reactor vessel including the core region have been calculated, together with the distribution of convective heat transfer coefficient(CHTC) at the cladding surface of the reactor vessel. The result shows that some parts of the core region of the reactor vessel have higher temperature gradient expressing higher thermal stress.

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