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JAEA Reports

Mechanical properties database of reactor pressure vessel steels related to fracture toughness evaluation

Tobita, Toru; Nishiyama, Yutaka; Onizawa, Kunio

JAEA-Data/Code 2018-013, 60 Pages, 2018/11

JAEA-Data-Code-2018-013.pdf:1.67MB

Mechanical properties of materials including fracture toughness are extremely important for evaluating the structural integrity of reactor pressure vessels (RPVs). In this report, the published data of mechanical properties of nuclear RPVs steels, including neutron irradiated materials, acquired by the Japan Atomic Energy Agency (JAEA), specifically tensile test data, Charpy impact test data, drop-weight test data, and fracture toughness test data, are summarized. There are five types of RPVs steels with different toughness levels equivalent to JIS SQV2A (ASTM A533B Class 1) containing impurities in the range corresponding to the early plant to the latest plant. In addition to the base material of RPVs, the mechanical property data of the two types of stainless overlay cladding materials used as the lining of the RPV are summarized as well. These mechanical property data are organized graphically for each material and listed in tabular form to facilitate easy utilization of data.

Journal Articles

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

Li, Y.*; Kato, Daisuke*; Shibata, Katsuyuki; Onizawa, Kunio

International Journal of Pressure Vessels and Piping, 78(4), p.271 - 282, 2001/04

 Times Cited Count:5 Percentile:41.36(Engineering, Multidisciplinary)

no abstracts in English

JAEA Reports

Development of probabilistic fracture mechanics code PASCAL and user's manual

Shibata, Katsuyuki; Onizawa, Kunio; Li, Y.*; Kato, Daisuke*

JAERI-Data/Code 2001-011, 233 Pages, 2001/03

JAERI-Data-Code-2001-011.pdf:7.42MB

no abstracts in English

Journal Articles

A Method for estimating peak temperature reached of the TMI-2 vessel lower head by microstructural examination of 308 stainless steel overlay

Tsukada, Takashi; Suzuki, Masahide; Kawasaki, Satoru

Proceedings of Three Mile Island Reactor Pressure Vessel Investigation Project Open Forum, p.151 - 163, 1994/00

no abstracts in English

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