Verification of probabilistic fracture mechanics analysis code PASCAL for reactor pressure vessel

Author:

LU Kai1,TAKAMIZAWA Hisashi1,LI Yinsheng1,MASAKI Koichi2,TAKAGOSHI Daiki3,NAGAI Masaki4,NANNICHI Takashi5,MURAKAMI Kenta6,KANTO Yasuhiro7,YASHIRODAI Kenji8,HAYASHI Takahiro9

Affiliation:

1. Japan Atomic Energy Agency (JAEA)

2. Mizuho Research & Technologies, Ltd.

3. Mitsubishi Heavy Industries, Ltd.

4. Central Research Institute of Electric Power Industry

5. IHI Corporation

6. The University of Tokyo

7. Ibaraki University

8. Hitachi, Ltd.

9. Toshiba Energy Systems & Solutions Corporation

Publisher

Japan Society of Mechanical Engineers

Subject

Process Chemistry and Technology,Economic Geology,Fuel Technology

Reference36 articles.

1. Chou, H.W. and Huang, C.C., Probabilistic structural integrity analysis of boiling water reactor pressure vessel under low temperature overpressure event, International Journal of Nuclear Energy, Vol. 2015 (2015), pp.785041:1-9.

2. Electric Power Research Institute, BWR reactor pressure vessel shell weld inspection requirements, EPRI-TR-105697 (1995).

3. EricksonKirk M., Junge, M., Arcieri, W., Bass, B., Beaton, R., Bessette, D., Chang, T.H., Dickson, T., et al., Technical basis for revision of the pressurized thermal shock (PTS) screening limit in the PTS Rule (10 CFR 50.61), NUREG-1806 (2006).

4. EricksonKirk M., Dickson, T., Mintz, T. and Simonen, F., Sensitivity studies of the probabilistic fracture mechanics model used in FAVOR, NUREG-1808 (2010).

5. Hirota, T., Sakamoto, H. and Ogawa, N., Proposal for update on evaluation procedure for reactor pressure vessels against pressurized thermal shock events in Japan, Proceedings of the ASME 2014 Pressure Vessels &

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