The new Code Case to substitute the allowable crack angle for a circumferential surface crack in “Rules on Fitness-for-Service (FFS)” published from The Japan Society of Mechanical Engineers (JSME). Until now, th e angle of a circumferential surface crack was limited below 60°. This is the limitation with consideration to the stability of the crack if the deep crack penetrates wall of the pipe. Therefore, a long crack (su ch like an SCC) was obliged to repair or replace even if it was shallow enough. Weld ...
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Code Case to Substitute the Allowable Crack Angle in Rules on Fitness-for-Service Codes of JSME

原子力発電設備の配管を対象として、複数回の継続検査や亀裂寸法測定を実施した場合の破壊評価の信頼度をモンテカルロシミュレーション（MCS）によって評価し、検査や亀裂寸法測定を繰返し実施する効果を定量化するとともに、耐圧バウンダリとして要求される信頼度を満足する検査条件を提案する。...
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Reliability-Based Optimization of Number of Flaw Inspections and Measurements

This study was performed to clarify the effect of defect sizing performance on the reliability of the piping applied “Rules on Fitness for Service.' Probabilistic fracture mechanics analyses with parameters of defect sizing error of ultrasonic test were performed, and the rel ation between the reliability of piping and the defect sizing error was evaluated. The reliability of piping applied “Rules on Fitness for Service’ is equal to that of which the all of defects are repaired, if the sizing error is cons...
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Study on Reliability of Nuclear Piping Considering Defect Sizing Performance

Nuclear piping has various kinds of thermal fatigue failure modes. Main causes of thermal loads are structural responses to fluid temperature changes during plant operation. These phenomena have complex mechanisms and so many patterns, that their problems still occur even though well-known issues. To prevent thermal fatigue due to above thermal loads, the JSME guideline is adopted. Both thermal load and fatigue failure mechanism have been investigated and summarized into the knowledgebase. Based on abo...
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Development of thermal fatigue evaluation methods of piping systems

An evaluation method of the occurrence probability of a through-wall crack in a reactor vessel of a fast breeder reactor due to fatigue-creep interaction has been proposed. Input data were p repared for a trial evaluation and the proposed evaluation method was applied. The result was c ompared with the allowable occurrence probability derived from the safety requirements for FBR....
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Development of a Reliability Evaluation Method for Fatigue-Creep Interaction Failure of an FBR Reactor Vessel 髙屋 茂 Shigeru TAKAYA

This paper describes the outline of the guidelines on structural reliability evaluation for the passive components of the fast breeder reactor (FBR). The guidelines are now being prepared by the working group for the system based code in the Japan society of mechanical engineers. They consist of five chapters, which are “General rules”, “Reliability evaluation”, “Failure scenario setting”, “Modeling”, and “Failure probability calculation”, respectively. In the chapter of “Reliability evaluation”, the g...
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Development of Guidelines for Structural Reliability Evaluation for FBR

Abstract This paper describes a procedure of structural reliability evaluation of passive components of fast breeder reactors for a failure due fatigue-creep interaction damage, and also reports results of benchmark evaluation conducted by using two independently developed codes. The s tructural integrity of a reactor vessel was evaluated deterministically and probabilistically. The results estimated by two codes agree well for oth of deterministic and probabilistic evaluations, which shows that these cod...
英字タイトル:
Benchmark Evaluation of Structural Reliability Evaluation Codes for FBR