Final assessment of MOX fuel performance experiment with Japanese PWR specification fuel in the HBWR

Conference · Sun Jul 01 04:00:00 EDT 2007

OSTI ID:21229291

Fujii, Hajime; Teshima, Hideyuki; Kanasugi, Katsumasa ^[1]; Kosaka, Yuji ^[2]; Arakawa, Yasushi ^[3]

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1. Mitsubishi Heavy Industries, Ltd., 1-1, Wadasaki-cho 1-chome, Hyogo-ku, Kobe 652-8585 (Japan)
2. Nuclear Development Corporation, 622-12 Funaishikawa, Tokai-mura, Ibaraki 319-1111 (Japan)
3. The Kansai Electric Power Co., Inc., 8 Yokota, 13 Goichi, Mihama-cho, Mikata-gun, Fukui, 919-1141 (Japan)

In order to obtain high burn-up MOX fuel irradiation performance data, SBR and MIMAS MOX fuel rods with Pu-fissile enrichment of about 6 wt% had been irradiated in the HBWR from 1995 to 2006. The peak burn-up of MOX pellet achieved 72 GWd/tM. In this test, fuel centerline temperature, rod internal pressure, stack length and cladding length were measured for MOX fuel and UO{sub 2} fuel as reference. MOX fuel temperature is confirmed to have no significant difference in comparison with UO{sub 2}, taking into account of adequate thermal conductivity degradation due to PuO{sub 2} addition and burn-up development. And the measured fuel temperature agrees well with FINE code calculation up to high burn-up region. Fission gas release of MOX is possibly greater than UO{sub 2} based on temperature and pressure assessment. No significant difference is confirmed between SBR and MIMAS MOX on FGR behavior. MOX fuel swelling rate agrees well with solid swelling rate in the literature. Cladding elongation data shows onset of PCMI in high power region. (authors)

• Research Organization: American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 21229291
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
42 ENGINEERING
BURNUP
CLADDING
FISSION PRODUCT RELEASE
FUEL PELLETS
MIXED OXIDE FUELS
PERFORMANCE
PLUTONIUM OXIDES
PWR TYPE REACTORS
SPECIFICATIONS
SWELLING
THERMAL CONDUCTIVITY
URANIUM DIOXIDE