RELAP5/MOD2 analysis of LOFT Experiment L9-3

Technical Report · Tue Mar 31 23:00:00 EST 1992

OSTI ID:10142073

Birchley, J C ^[1]

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1. UKAEA Atomic Energy Establishment, Winfrith (United Kingdom)

An analysis has been performed of LOFT Experiment L9-3, a loss-of-feedwater anticipated transient without trip, in order to support the validation of RELAP5/MOD2. Experiment L9-3 exhibited a rapid boildown of the steam generator, following the loss of feed, with the reactor remaining close to its initial power until the steam generator tubes became sufficiently uncovered for primary to secondary heat transfer to be significantly reduced. The ensuing heat up of the primary fluid resulted in a reduction in power induced by the moderator feedback. The primary system pressure increased to the safety relief valve setpoint, before the fall in reactor power allowed the mismatch between primary system heat input and heat removal via the steam generator to be accommodated by cycling of the pilot operated relief valve (PORV). Comparison between calculation and data shows generally good agreement, though with discrepancies in some areas. Weaknesses in the code`s treatment of interphase drag and in the representation of the pressuriser spray are indicated, although a shortage of definitive data, particularly in the steam generator, may also be a factor. The overprediction of interphase drag led to a tendency to underpredict the initial inventory in the steam generator and also, perhaps, to overpredict the steam generator heat transfer while the tubes were being uncovered. There is indication that the pressuriser vapour region conditions were close to equilibrium during spray operation. The point kinetics model in RELAP5/MOD2 proved a viable means of representing the power history for this transient.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research; UKAEA Atomic Energy Establishment, Winfrith (United Kingdom)
• Sponsoring Organization: Nuclear Regulatory Commission, Washington, DC (United States)
• OSTI ID: 10142073
• Report Number(s): NUREG/IA--0058; AEEW-R--2435; PWR/PKWG/P--88-390; ON: TI92013505
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
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22 GENERAL STUDIES OF NUCLEAR REACTORS
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99 GENERAL AND MISCELLANEOUS
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COMPUTER PROGRAM DOCUMENTATION
COMPUTERIZED SIMULATION
EXPERIMENTAL DATA
FLUID FLOW
HEAT TRANSFER
HYDRAULICS
LOFT REACTOR
LOSS OF COOLANT
MATHEMATICS AND COMPUTERS
POWER REACTORS
NONBREEDING
LIGHT-WATER MODERATED
NONBOILING WATER COOLED
PWR TYPE REACTORS
R CODES
REACTOR SAFETY
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