Fatigue Strength Evaluation of Welded Joint in Pressure Vessel by Probabilistic Fracture Mechanics.

The structural integrity assessment of a nuclear Reactor Pressure Vessel (RPV) during accidental conditions, such as loss-of-coolant accident (LOCA), is a major safety concern. Besides Conventional deterministic calculations to justify the RPV integrity, Electricite´ de France (EDF) carries out probabilistic analyses. Since in the USA the probabilistic fracture mechanics analyses are accepted by the Nuclear Regulatory Commission (NRC), a benchmark has been realized between EDF and Oak Ridge Structural Assessments, Inc. (ORSA) to compare the models and the computational methodologies used in respective deterministic and probabilistic fracture mechanics analyses. Six cases involving two distinct transients imposed on RPVs containing specific flaw configurations (two axial subclad, two circumferential surface-breaking, and two axial surface-braking flaw configurations) were defined for a French vessel. In two separate phases, deterministic and probabilistic, fracture mechanics analyses were performed for these six cases.

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2038 Application of Mesh Superposition method to Fatigue strength evaluation of welded joint

The proceedings of the JSME annual meeting ◽

10.1299/jsmemecjo.2006.6.0_77 ◽

2006 ◽

Vol 2006.6 (0) ◽

pp. 77-78

Author(s):

Masanari OKUDA ◽

Masashi Mohri ◽

Katsuyuki SUZUKI

Keyword(s):

Fatigue Strength ◽

Welded Joint ◽

Superposition Method ◽

Strength Evaluation

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Probabilistic Fracture Mechanics Evaluation of Embrittlement Recovery by Annealing of Reactor Pressure Vessel

Proceedings of the 1992 Annual Meeting of JSME/MMD ◽

10.1299/jsmezairiki.2000.0_435 ◽

2000 ◽

Vol 2000 (0) ◽

pp. 435-436

Author(s):

Katsuyuki SHIBATA ◽

Kunio ONIZAWA ◽

Yinsheng LI ◽

Daisuke KATO

Keyword(s):

Fracture Mechanics ◽

Pressure Vessel ◽

Reactor Pressure Vessel ◽

Probabilistic Fracture Mechanics ◽

Reactor Pressure

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Development of Probabilistic Fracture Mechanics Analysis Codes for Reactor Pressure Vessels Considering Recent Embrittlement Model and Calculation Method of SIF: Progress of the Work

Volume 6A: Materials and Fabrication ◽

10.1115/pvp2016-63128 ◽

2016 ◽

Author(s):

Jongmin Kim ◽

Bongsang Lee ◽

Taehyun Kim ◽

Yoonsuk Chang

Keyword(s):

Stress Intensity ◽

Fracture Mechanics ◽

Pressure Vessel ◽

Calculation Method ◽

Reactor Pressure Vessel ◽

Radiation Embrittlement ◽

Calculation Methods ◽

Probabilistic Fracture Mechanics ◽

Mechanics Analysis ◽

Reactor Pressure

It is widely recognized that the state of knowledge and data for the probabilistic calculations which had been proposed in the early 1980s made a conservative treatment of several key factors and models. Recently, applications of some new radiation embrittlement model, material database, calculation method of stress intensity factors and others which can improve fracture mechanics assessment of reactor pressure vessel (RPV) are introduced. This improvement on the accuracy and reliability of the probabilistic fracture mechanics (PFM) analysis necessitated changes in PFM analysis procedures and calculations. Modification and application of newly developed models and calculation methods are the main target of developing a probabilistic fracture mechanics analysis code based on the structure of existing R-PIE and VISA computer code to reflect the latest technical basis. Failure probabilities of reactor pressure vessel under pressurized thermal shock (PTS) conditions were calculated through finite difference method (FDM) and Monte Carlo simulation techniques with user friendly graphic interface. Moreover, various radiation embrittlement models and calculation methods of stress intensity factor at crack tip based on AFCEN code are applied and verified in the present work.

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