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Tanaka, Kosuke*; Maeda, Koji*; Katsuyama, Kozo*; Inoue, Masaki*; Iwai, Takashi; Arai, Yasuo
Journal of Nuclear Materials, 327(2-3), p.77 - 87, 2004/05
no abstracts in English
Inoue, Masaki*; Iwai, Takashi; Arai, Yasuo; Asaga, Takeo*
Proceedings of GLOBAL2003 Atoms for Prosperity; Updating Eisenhower's Global Vision for Nuclear Energy (CD-ROM), p.1694 - 1703, 2003/11
no abstracts in English
Arai, Yasuo; Iwai, Takashi; Nakajima, Kunihisa; Nagashima, Hisao; Nihei, Yasuo; Katsuyama, Kozo*; Inoue, Masaki*
Proceedings of GLOBAL2003 Atoms for Prosperity; Updating Eisenhower's Global Vision for Nuclear Energy (CD-ROM), p.1686 - 1693, 2003/00
no abstracts in English
Iwai, Takashi; Nakajima, Kunihisa; Kikuchi, Hironobu; Kimura, Yasuhiko; Nagashima, Hisao; Sekita, Noriaki; Arai, Yasuo
JAERI-Research 2000-010, p.110 - 0, 2000/03
JAERI-Research-2000-010.pdf:20.61MB
no abstracts in English
Arai, Yasuo; Iwai, Takashi; ; Okamoto, Yoshihiro; Nakajima, Kunihisa; Niimi, Motoji; ; Yamahara, Takeshi;
JAERI-Research 95-008, 92 Pages, 1995/02
JAERI-Research-95-008.pdf:5.04MB
no abstracts in English
Arai, Yasuo; Iwai, Takashi; ; Nakajima, Kunihisa; ; ;
JAERI-Research 94-027, 66 Pages, 1994/11
JAERI-Research-94-027.pdf:4.09MB
no abstracts in English
Arai, Yasuo; ; Iwai, Takashi; Maeda, Atsushi; ; Shiozawa, Kenichi; Omichi, Toshihiko
Journal of Nuclear Science and Technology, 30(8), p.824 - 830, 1993/08
Times Cited Count：7 Percentile：60.37(Nuclear Science & Technology)no abstracts in English
Arai, Yasuo; ; ; Iwai, Takashi; ; ; Niimi, Motoji; Omichi, Toshihiko
JAERI-M 91-191, 93 Pages, 1991/11
no abstracts in English
Ikusawa, Yoshihisa; Uwaba, Tomoyuki; Tanno, Takashi; Oka, Hiroshi; Kaito, Takeji; Nemoto, Junichi*
no journal, ,
In high burnup MOX fuel pins, cesium accumulates at the UO
- MOX pellet boundary and cladding tube diameter is locally increased around the position by FCMI, due to the formation of Cs-U-O compound. Based on the computation results obtained using an irradiation behavior analysis code "CEDAR", we evaluated the effect of this behavior on cladding creep damage. As the computation result, it was found that the cladding creep damage around the UO - MOX pellet boundary increases by the FCMI stress.
Tanno, Takashi; Oka, Hiroshi; Ikusawa, Yoshihisa; Uwaba, Tomoyuki; Kaito, Takeji
no journal, ,
Evaluation of diameter changes of fuel pellets by cesium (Cs) accumulation based on gamma-scan profiles was carried out in order to estimate the influence of axial migration and accumulation behavior of Cs for fuel-cladding mechanical interaction (FCMI). Local diameter change of the UO pellets facing MOX pellets and FCMI by the diameter change were predicted for EBR-II irradiation test. To prevent the FCMI, the radial gap should have been more 80
m larger than the initial radial gap as fabricated for EBR-II irradiation test.
Tanno, Takashi; Oka, Hiroshi; Ikusawa, Yoshihisa; Uwaba, Tomoyuki; Otsuka, Satoshi; Kaito, Takeji; Maeda, Seiichiro
no journal, ,
Cesium (Cs), which is a volatile fission product (FP), migrates axially from hot core region to cold top and/or bottom ones in fast reactor (FR) fuel pins. The Cs accumulated in cold region such as axial UO blanket fuel can form Cs-U-O compounds having lower density than that of the fuel pellet, causing fuel cladding mechanical interaction (FCMI). The severe FCMI would arouse concern about the integrity of fuel pins. This work aims to understand Cs axial migration behavior and its effect on FR fuel pins. Two MOX fuel pins irradiated in EBR-II were evaluated the Cs migration behavior. Pin diameter measurement and gamma-scanning were carried out, and calculations with the ORIGEN-2 code was also done to estimate FP inventory with burnup in the pins. It was found from the comparison between the calculated pellet swelling by the Cs-U-O compounds and the measured pin diameter increase that the localized pin diameter increases at the MOX fuel-blanket interfaces were due to the FCMI caused by the pellet swelling associated with the formation of the Cs-U-O compounds.
Ozawa, Takayuki; Hirooka, Shun; Kato, Masato; Smuin, T. J.*; Jensen, C. B.*; Woolstenhulme, N. E.*; Wachs, D. M.*
no journal, ,
The ARES-MOX transient program is planned in TREAT by using MOX fuels irradiated in SPA-2 irradiation tests in EBR-II, which was conducted in 1984-1994 under the international collaboration between US and Japan, and have been stored in the current INL. The MOX fuels irradiated up to the maximum burnup of about 130 GWd/t in EBR-II includes the solid FP content of about 10 wt.%. In this program, the objectives are to acquire not only valuable data to develop the FCMI threshold for high-burnup annular MOX fuels but also knowledge about irradiation behavior of FP at transient. The overview of ARES-MOX program, schedule and outcomes expected from fuel performance calculation for annular MOX fuels irradiated in EBR-II will be introduced here.
