Fabrication of low-O/M fast reactor MOX fuel and analysis on its oxygen potential behaviors

Hirooka, Shun; Vauchy, R.; Horii, Yuta; Sunaoshi, Takeo*; Saito, Kosuke

Proceedings of TopFuel 2025; Nuclear Reactor Fuel Performance Conference (Internet), 10 Pages, 2025/10

Reducing the oxygen-to-metal (O/M) ratio in MOX fuels plays an important role in suppressing the corrosion depth in the cladding due to fuel-cladding chemical interaction (FCCI), which is the key to determining the lifetime of fast reactor MOX fuels. Owing to a number of irradiation and post-irradiation experiments, a clear decreasing trend in the corrosion depth with lower O/M ratio in the as-fabricated MOX pellet was reported, whereas a significant redistribution of the O/M ratio in a pellet driven by the radial temperature gradient during irradiation could supply a higher oxygen potential near the pellet periphery where the FCCI should occur. The reduction of the O/M ratio in the MOX pellet fabrication processes can be achieved by heat treatment by taking high temperature, longer time, and lower oxygen partial pressure in the gas into account; the properties governing the reduction are not sufficiently studied. This study demonstrated the variation of O/M ratio in MOX pellets and the in-situ O/M ratio during the heat treatment was analyzed by using a thermogravimeter, which revealed a decreasing behavior during heating and dwell as well as an increasing behavior in the O/M ratio during the cooling step. Furthermore, the redistribution of O/M ratio, analyzed by Sari's model, was discussed to investigate the O/M ratio and the oxygen potential near the pellet periphery which is likely to have a more direct influence on the FCCI than the as-fabricated O/M ratio. By using the recent oxygen potential data on MOX, it is found that the oxygen potential profile in the radial direction is especially drastic near the pellet periphery and is sensitive to the as-fabricated O/M ratio.

Fission gas release and swelling in uranium-plutonium mixed nitride fuels

Tanaka, Kosuke*; Maeda, Koji*; Katsuyama, Kozo*; Inoue, Masaki*; Iwai, Takashi; Arai, Yasuo

Journal of Nuclear Materials, 327(2-3), p.77 - 87, 2004/05

https://doi.org/10.1016/j.jnucmat.2004.01.002

no abstracts in English

Behavior of uranium-plutonium mixed carbide fuel irradiated at JOYO

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

Behavior of actinides and fission products in mixed nitride fuels irradiated in JMTR; 88F-5A capsule

Iwai, Takashi; Nakajima, Kunihisa; Kikuchi, Hironobu; Kimura, Yasuhiko; Kanaitsuka, Fumio; Sekita, Noriaki; Arai, Yasuo

JAERI-Research 2000-009, p.36 - 0, 2000/03

JAERI-Research-2000-009.pdf:6.16MB

no abstracts in English

Post irradiation examinations of uranium-plutonium mixed nitride fuel irradiated in JMTR: 88F-5A capsule

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

Diffusion of iron in -iron telluride(FeTe) by Moessbauer spectroscopy and tracer method

; Tsuji, Toshihide*;

Journal of Nuclear Materials, 203, p.172 - 178, 1993/00

https://doi.org/10.1016/0022-3115(93)90054-3

no abstracts in English

Kinetics and mechanism of telluride scale growth on iron

; Tsuji, Toshihide*;

Journal of Nuclear Science and Technology, 29(1), p.68 - 77, 1992/01

https://doi.org/10.3327/jnst.29.68

no abstracts in English

High temperature corrosion of iron chromium alloys by tellurium

; Tsuji, Toshihide*;

Journal of Nuclear Science and Technology, 28(8), p.721 - 731, 1991/08

https://doi.org/10.3327/jnst.28.721

no abstracts in English

Chemical Interaction in Uranium-Plitnium Mixed Oxide Fuel Pins for LMFBR; Review

JAERI-M 8688, 44 Pages, 1980/02

JAERI-M-8688.pdf:1.76MB

no abstracts in English