For the commercialization of fast breeder reactors (FBR), it is essential to enhance the economic competitiveness by reduction of total cost by elongation of plant service period. In this point of view, it is important to establish the assessment method of integrity of aged weld joint and repair welding for the components of future long life FBR. In the present study, evolution of microstructure is evaluated for the 304SS-304SS weld joint which was used for 88,000h at 526-545°C in the French proto-typ...
英字タイトル:
Influence of FBR plant service and repair welding on microstructure and residual stress of austenitic stainless steel weld joint

１．緒言システム化規格（System Based Code, SBC）は、構造物に適用される複数の規格基準に含まれる技術項目の間で余裕を相互交換可能にすることにより、余裕の重複を避け、構造物の余裕を適正な水準に合理的に設定することを目標とする、柔軟な規格基準体系として我国で発案された(1)-(3)。発電用原子力設備の構造物には、一般に、材料規格、設計規格、溶接規格、維持規格等が適用されるが、規格の規定にはいずれも固有の余裕が含まれるため、すべての規格が適用される構造物は、結果として過剰な裕度を持つ傾......

For the commercialization of fast breeder reactors (FBR), it is essential to enhance the economic competitiveness by reduction of total cost by elongation of plant service period. For this reason, it is needed to establish the aged weld joint integrity assessment method for future FBR compon ents. Additionally, since it is expected to employ the repair weld for aged structural components, it is essential to investigate the effect of repair weld. To investigate the influence of high temperature-long service...
英字タイトル:
Basic investigation of aging weld joint and repair weld joint integrity assessmentusing neutron residual stress measurement method

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....
英字タイトル:
Development of a Reliability Evaluation Method for Fatigue-Creep Interaction Failure of an FBR Reactor Vessel 髙屋 茂 Shigeru TAKAYA

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