CHF considerations for highly moderated 100% MOX fuels PWRs

Conference · Fri Sep 01 04:00:00 EDT 1995

OSTI ID:111421

Saphier, D; Raymond, P ^[1]

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1. CEA Saclay, DMT/SERMA/LETR, Gif-sur-Yvette (France)

A feasibility study on using 100% MOX fuel in a PWR with increased moderating ratio, RMA, was initiated. In the proposed design all the parameters were chosen identical to the French 1450MW PWR, except the fuel pin diameter which was reduced to achieve higher moderating ratios, V{sub M}/V{sub F}, where V{sub M} and V{sub F} are the moderator and fuel volume respectively. Moderating ratios from 2 to 4 were considered. In the present study the thermal-hydraulic feasibility of using fuel assemblies with smaller diameter fuel pins was investigated. The major design constrain in this study was the critical heat flux (CHF). In order to maintain the fuel pin integrity under nominal operating and transient conditions, the minimum DNBR, (Departure from Nucleate Boiling Ratio given by CHF/q{close_quotes}{sub local}, where q{close_quotes}{sub local} is the local heat flux), has to be above a given value. The limitations of the existing CHF correlations for the present study are outlined. Two designs based on the conventional 17x17 fuel assembly and on the advanced 19x19 assembly meeting the MDNBR criteria and satisfying the control margin requirements, are proposed.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Technology; American Nuclear Society, La Grange Park, IL (United States); American Inst. of Chemical Engineers, New York, NY (United States); American Society of Mechanical Engineers, New York, NY (United States); Canadian Nuclear Society, Toronto, ON (Canada); European Nuclear Society (ENS), Bern (Switzerland); Atomic Energy Society of Japan, Tokyo (Japan); Japan Society of Multiphase Flow, Kyoto (Japan)
• OSTI ID: 111421
• Report Number(s): NUREG/CP--0142-Vol.4; CONF-950904--Vol.4; ON: TI95017080
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
CRITICAL HEAT FLUX
DESIGN
DIMENSIONS
FEASIBILITY STUDIES
FUEL ASSEMBLIES
HEAT TRANSFER
HYDRAULICS
MATHEMATICAL MODELS
PWR TYPE REACTORS