• Title, Summary, Keyword: ASME Sec. Xl

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압력격리밸브 누설시험 절차 및 방법 개선 방안

  • 조종철;조두연
    • Proceedings of the Korean Nuclear Society Conference
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    • pp.725-730
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    • 1998
  • 가동중 원자력 발전소들에서는 압력격리밸브들에 대한 기술지침서 감시시험요건과 가동중 시험 규제요건을 충족시키기 위하여 누설시험을 일정 주기로 수행하고 있다. 동 주기시험은 ASME Ba&PV Code Sec. Xl IWV-3420 또는 ASME ON Code ISTC(Part 10) 4.2.2절의 운전에 부합되는 방법과 절차에 따라 이루어지도록 규정되어 있다. 이러한 주기시험의 근본 목적과 시험방법 및 절차요건에 대한 기술적 근거의 이해는 동 시험활동의 성과를 높이는데 큰 도움이 될 것임에 틀림없다. 따라서, 본 논문에서는 압력격리 밸브들에 대한 누설시험 목적 및 시험요건의 기술적 근거를 소개하였으며, 잠재적 문제점들을 도출하여 분석 검토하고 적절한 대처 방안을 제시하였다

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Development of a RVIES Syetem for Reactor Vessel Integrity Evaluation (원자로용기 건전성평가를 위한 RVIES 시스템의 개발)

  • Lee, Taek-Jin;Choe, Jae-Bung;Kim, Yeong-Jin;Park, Yun-Won;Jeong, Myeong-Jo
    • Transactions of the Korean Society of Mechanical Engineers A
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    • v.24 no.8
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    • pp.2083-2090
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    • 2000
  • In order to manage nuclear power plants safely and cost effectively, it is necessary to develop integrity evaluation methodologies for the main components. Recently, the integrity evaluation techniques were broadly studied regarding the license renewal of nuclear power plants which were approaching their design lives. Since the integrity evaluation process requires special knowledges and complicated calculation procedures, it has been allowed only to experts in the specified area. In this paper, an integrity evaluation system for reactor pressure vessel was developed. RVIES(Reactor Vessel Integrity Evaluation System) provides four specific integrity evaluation procedures covering PTS(Pressurized Thermal Shock) analysis, P-T(Pressure-Temperature) limit curve generation, USE(Upper Shelf Energy) analysis and Fatigue analysis. Each module was verified by comparing with published results.

Stress Intensity factor Calculation for the Axial Semi-Elliptical Surface Flaws on the Thin-Wall Cylinder Using Influence Coefficients (영향계수를 이용한 원통용기 축방향 표면결함의 응력확대계수의 계산)

  • Jang, Chang-Heui;Moon, Ho-Rim;Jeong, Ill-Seok;Kim, Tae-Ryong
    • Transactions of the Korean Society of Mechanical Engineers A
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    • v.26 no.11
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    • pp.2390-2398
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    • 2002
  • For integrity analysis of nuclear reactor pressure vessel, including the Pressurized thermal shock analysis, the fast and accurate calculation of the stress intensity factor at the crack tip is needed. For this, a simple approximation scheme is developed and the resulting stress intensity factors for axial semi-elliptical cracks in cylindrical vessel under various loading conditions are compared with those of the finite element method and other approximation methods, such as Raju-Newman's equation and ASME Sec. Xl approach. For these, three-dimensional finite-element analyses are performed to obtain the stress intensity factors for various surface cracks with t/R = 0.1. The approximation methods, incorporated in VINTIN (Vessel INTegrity analysis-INner flaws), utilizes the influence coefficients to calculate the stress intensity factor at the crack tip. This method has been compared with other solution methods including 3-D finite clement analysis for internal pressure, cooldown, and pressurized thermal shock loading conditions. The approximation solutions are within $\pm$2.5% of the those of FEA using symmetric model of one-forth of a vessel under pressure loading, and 1-3% higher under pressurized thermal shock condition. The analysis results confirm that the VINTIN method provides sufficiently accurate stress intensity factor values for axial semi-elliptical flaws on the surface of the reactor pressure vessel.