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Takahashi, Kenji; Shiina, Akira; Onizawa, Takahiro; Ibaki, Shoji; Yamaguchi, Toshihiko; Tagawa, Akihiro
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 9 Pages, 2009/07
Japanese prototype FBR Monju will restart its operation for the first time since the sodium leak accident in 1995. Concerning heat transfer tubes of the steam generator (SG) units, presumable deteriorations were studied and the integrity is confirmed when they have no lack of thickness by corrosion after the long-term suspension. It was impossible to test the tubes directly, then three tests were applied, which were ECT, VT (visual test) and leak test. ECT is to check for the lack of thickness caused by regional corrosion. VT checked for the global corrosion and the bottom of the tube. Leak test checked for pin-holes (penetrating holes). Total evaluation proved no significant lack of thickness and pin-holes in the tube.
Onizawa, Takahiro; Ito, Kenji; Osaki, Toshiyuki
no journal, ,
In the secondary sodium cooling system equipment of Monju, the leakage was ended at the early stage when sodium leaked, and the remodeling construction to control the influence on equipment by the sodium combustion etc. further more were done. These are the addition of drain piping, the enlargement of drain piping and the batch operation of emergency drain valve etc.. In the modification function tests of secondary sodium cooling system, these function and performance, etc. were confirmed.
Takahashi, Kenji; Yamaguchi, Toshihiko; Onizawa, Takahiro; Kurosawa, Norifumi; Shiina, Akira; Tagawa, Akihiro; Ibaki, Shoji
no journal, ,
no abstracts in English
Shiina, Akira; Yamaguchi, Toshihiko; Onizawa, Takahiro; Kurosawa, Norifumi; Takahashi, Kenji; Ibaki, Shoji
no journal, ,
no abstracts in English
Okuda, Takahiro; Yamashita, Hayato; Toyota, Kodai; Shimomura, Kenta; Onizawa, Takashi; Kato, Shoichi
no journal, ,
This study describes the setting of the material property equations of 316FR steel at an extremely high temperature which can be applied to severe accident conditions of generation IV fast reactors. 316FR steel will be applied to structural materials, e.g. reactor vessel, in the generation IV fast reactors. After the severe accident in Fukushima Daiichi Nuclear Power Plants, the evaluation of structural integrity was found to be very important severe accident condition. The development of the generation IV fast reactors requires the material properties of 316FR steel at the extremely high temperature. However, such data has not been acquired. Therefore, tensile and creep tests were carried out in the temperature range over 700
C for 316FR steel. Based on the acquired data from the tests, the equations that can evaluate the material properties of 316FR steel at the extremely high temperature were set up. They are an elasto-plastic stress-strain equation, a creep rupture equation and a creep strain equation.