Research programs on aging of reactor structural materials at Japan Atomic Energy Research Institute

Full text: A few of light water reactors (LWRs) in Japan began to operate in the early 1970's and have almost reached a 30-year operation. The LWRs will be the mainstream of nuclear power generation for a considerably long period. It is, therefore, necessary for the aged reactors to further operate when many aging phenomena are readily manageable. The domestic countermeasures designated to cope with the aging of LWRs have been already taken expecting a 60-year operation. Particularly important is to maintain the integrity of the safety-related structures and components subjected to aging during long-term operation. Reactor vessel, internal core components, primary coolant piping, recirculation pump, electrical cables, containment structures and concrete structures have been chosen as safety-related ones necessary for the management of aging in a report by Ministry of Economy, Trade and Industry (METI). The relevant research and development have been comprehensively carried out among the Japan Atomic Energy Research Institute (JAERI), national research organizations and nuclear energy industry under the domestic coordination of METI. At JAERI, the evaluation of aging degradation under irradiation has been investigated for structures and components which are important from the safety standpoint and are difficult in replacing and repairing, e.g. reactor pressure vessel (RPV), internal core structures in accordance with the national five-year safety research program issued by the Nuclear Safety Commission. For RPV, the following research and development have been conducted to precisely evaluate and predict irradiation embrittlement in the long-term operation; Fracture toughness evaluation using the Master Curve approach for Japanese RPV steels; Grain boundary embrittlement caused by irradiation-induced phosphorus segregation; Gamma-ray effect on irradiation hardening; and Evaluation of degradation on stainless overlay cladding. We have also developed a PFM code PASCAL (PFM Analysis of Structural Components in Aged LWR), and continue to investigate how this PFM method is applied to the domestic regulation under the extra-budgetary program from METI. The results of the above research will be used for new aging related code and/or standard, and also incorporated in PASCAL to analyze the integrity and reliability of aged RPV. For internal core structures, the main research item is irradiation assisted stress corrosion cracking (IASCC) which is caused by the synergistic effects of neutron and gamma-ray radiation, corrosion by high temperature water, and the residual and/or applied stresses. The following research and development programs have been conducted: Fundamental study for mechanistic understanding of IASCC; Development of IASCC evaluation technology for BWR plants based on postirradiation IASCC test data as a part of METI's national project; In-pile IASCC test for evaluation of the synergistic effects of irradiation, corrosion and stress; and Computational studies on the materials database and numerical simulation. For the programs of IASCC studies, a new high temperature water loop facility was designed and installed at the Japan Materials Testing Reactor (JMTR). This paper will describe the current status and typical results of the above-mentioned research programs. (author)