Improvements to a probabilistic fracture mechanics code for evaluating the integrity of a RPV under transient loading-Reference-Cited by-同舟云学术

Improvements to a probabilistic fracture mechanics code for evaluating the integrity of a RPV under transient loading

Published:2001-04 Issue:4 Volume:78 Page:271-282
ISSN:0308-0161
Container-title:International Journal of Pressure Vessels and Piping
language:en
Short-container-title:International Journal of Pressure Vessels and Piping

Author:

Li Yinsheng,Kato Daisuke,Shibata Katsuyuki,Onizawa Kunio

Publisher

Elsevier BV

Subject

Mechanical Engineering,Mechanics of Materials,General Materials Science

Reference24 articles.

1. Simonen FA, Jonshon KI, et al. VISA-II — a computer code for predicting the probability of reactor pressure vessel failure. NUREG/CR-4486, 1986.

2. Cheverton RD, Ball DG. OCA-P, A deterministic and probabilistic fracture-mechanics code for application to pressure vessels. NUREG/CR-3618, 1984.

3. An overview of FAVOR: a fracture analysis computer code for nuclear reactor pressure vessels;Dickson;SMiRT,1995

4. Persoz M, Hugonnard-Bruyere S, et al. Deterministic and probabilistic assessment of the reactor pressure vessel structural integrity with a user-friendly software. ASME 2000, PVP-vol. 403, 2000;137–44.

5. Harris DO, Lim EY, et al. Probability of pipe fracture in the primary coolant loop of a PWR plant. NUREG/CR-2189, 1981.

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