High Burnup Fuel Behavior Modeling

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

OSTI ID:21229335

Jahingir, M; Rand, R; Stachowski, R; Miles, B ^[1]; Kusagaya, K ^[2]

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1. Global Nuclear Fuel-Americas, PO Box 780, Wilmington, NC 28402 (United States)
2. Global Nuclear Fuel-Japan, 3-1, Uchikawa 2, Yokosuka, Kanagawa, 239-0836 (Japan)

This paper discusses the development and qualification of the PRIME03 code to address high burnup mechanisms and to improve uranium utilization in current and new reactor designs. Materials properties and behavioral models have been updated from previous thermal-mechanical codes to reflect the effects of burnup on fuel pellet thermal conductivity, Zircaloy creep, fuel pellet relocation, and fission gas release. These new models are based on results of in-pool and post irradiation examination (PIE) of commercial boiling water reactor (BWR) fuel rods at high burnup and results from international experimental programs. The new models incorporated into PRIME03 also address specific high burnup effects associated with formation of pellet rim porosity at high exposure. The PRIME03 code is qualified by comparison of predicted and measured fuel performance parameters for a large number of high, low, and moderate burnup test and commercial reactor rod. The extensive experimental qualification of the PRIME03 prediction capabilities confirms that it is a reliable best-estimate predictor of fuel rod thermal-mechanical performance over a wide range of design and operating conditions. (authors)

• Research Organization: American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 21229335
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
42 ENGINEERING
BURNUP
BWR TYPE REACTORS
CREEP
DESIGN
FISSION PRODUCT RELEASE
FUEL PELLETS
FUEL RODS
NUCLEAR FUELS
PERFORMANCE
POROSITY
POST-IRRADIATION EXAMINATION
SIMULATION
THERMAL CONDUCTIVITY
URANIUM
ZIRCALOY