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Kudo, Tamotsu
Proceedings of 3rd Korea-Japan Joint Summer School (JSS-3) for Students and Young Researchers, p.203 - 210, 2007/08
After the TMI-2 accident, a number of experimental data on radionuclide release from fuel have been obtained in the world. However, these data were obtained under only conditions with atmospheric pressure, fuel temperature below 2900 K. And data for radionuclide release from MOX, and interactive effect of fuel oxidation and dissolution were limited. Then, the VEGA program was conducted at JAEA at pressure of 0.1 and 1.0 MPa, temperature to about 3130 K, using UO
and MOX fuels, under inert and oxidized atmosphere. The VEGA program showed that the Cs release at 1.0 MPa was smaller by about 30 % than that at atmospheric pressure, and acceleration of Cs release rate was observed above 2800 K due to foaming and melting of fuel. This article mainly describes the radionuclide release from UO and MOX fuels, and effects of fuel oxidation and dissolution.
Sawa, Kazuhiro
Proceedings of 3rd Korea-Japan Joint Summer School (JSS-3) for Students and Young Researchers, p.147 - 153, 2007/08
The high-temperature gas-cooled reactor (HTGR), which is graphite-moderated and helium-cooled, is attractive due to its unique capability of producing high temperature helium gas and its fully inherent reactor safety. In particular, hydrogen production using the nuclear heat from HTGR offers one of the most promising technological solutions to curb the rising level of CO emission. In the field of fuels and materials for HTGR, the JAEA precedes the several important research and developments. For upgrading of HTGR technologies, the JAEA has developed an extended burnup TRISO-coated fuel particle and ZrC coated particles. The JAEA also proceeds development of new graphite, C/C composite material. In parallel, in order to extend in-core life of the core components, the JAEA started to develop non-destructive evaluation method of irradiation damage by ultrasonic wave propagation and micro-indentation. This test introduces the present status of research and development for HTGR fuels and materials.
Ogawa, Toru
no journal, ,
Research and developments on the materials for Generation-IV reactors are reviewed, focusing particularly on SFR, VHTR and SCWR. The materials issues under environments specific to each reactor combined with higher temperature and increasing neutron fluences are discussed. The needs of modeling development are also briefly discussed.
Nemoto, Yoshiyuki
no journal, ,
Basic of stress corrosion cracking (SCC) on boiling water reactor (BWR) will be presented for the students and the young researchers in Japan and Korea. In the middle 70's and after 2000, there was lowering of plant utilization rate. Lowering in the 70's was caused by SCC on structural materials. After that period, application of low carbon austenitic stainless steel increased the plant utilization rate however it was lowered again after 2000. This lowering was caused by intergranular SCC (IGSCC) of low carbon austenitic stainless steels of core shrouds and primary loop recirculation (PLR) piping in BWRs. Thus the investigation on SCC is one of the most important issues to maintain the integrity of structures and components in BWR. The basic and instances of SCC, and topics of relevant researches at the Japan Atomic Energy Agency (JAEA) will be presented.
Tobita, Toru
no journal, ,
To maintain the structural integrity of reactor pressure vessels (RPVs) difficult to replace is one of the key safety issues in a nuclear plant operation. RPVs are operated in the ductile regime to prevent the non-ductile fracture of RPV that may lead to a catastrophic accident. Since RPV steel is subjected to neutron from the reactor core which causes a shift in ductile-to-brittle transition temperature, a surveillance test is performed to monitor the fracture toughness decrease due to neutron irradiation. The present paper gives an overview on how to perform the surveillance test and assess the structural integrity of RPVs in a long-term operation.