The study on the fatigue damage estimation method for the buckled reactor vessel under earthquake loading

Book · Sat Dec 30 23:00:00 EST 1995

OSTI ID:170346

Ogiso, S; Sasaki, T; Ooka, Y ^[1]; Nakamura, H ^[2]

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1. Kawasaki Heavy Industries, Ltd., Tokyo (Japan). Nuclear Systems Division
2. Central Research Institute of Electric Power Industry, Chiba (Japan)

An estimation of fatigue damage derived from buckling deformation is one of the important factor to evaluate a safety margin of seismic design criteria for thin walled fast breeder reactor components, in addition to estimate the critical buckling strength itself. Post-buckling waves obtained by seismic response analyses of SDOF (Single Degree of Freedom) models were analyzed as displacement range-frequency histogram. Cyclic post buckling fatigue tests were performed with 304 S.S. cylinders to obtain basic fracture data and to simulate the seismic response characteristic with the static approach. Fatigue damage estimation methods were discussed with the correlation to the maximum global displacement response.

• OSTI ID: 170346
• Report Number(s): CONF-950740--; ISBN 0-7918-1343-6
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
DEFORMATION
EXPERIMENTAL DATA
FAILURE MODE ANALYSIS
FBR TYPE REACTORS
FRACTURE PROPERTIES
MATHEMATICAL MODELS
MECHANICAL TESTS
REACTOR VESSELS
RESPONSE FUNCTIONS
SEISMIC EFFECTS
STAINLESS STEEL-304