Assessment of core damage models in SCDAP/RELAP5 during OECD LOFT LP-FP-2

Conference · Tue Dec 31 00:00:00 EST 1991

OSTI ID:10164710

Coryell, E W

The US Nuclear Regulatory Commission has sponsored a program to apply the SCDAP/RELAP5 code to analysis of the transient and reflood phases of the OECD LOFT LP-FP-2 Experiment. The principal objectives of the LP-FP-2 experiment were to determine the fission product release from the fuel during the early phases of a severe fuel damage scenario and to examine the phenomena controlling fission product transport in a vapor/aerosol environment. Calculations with the SCDAP/RELAP5 code, developed at the INEL with NRC support, have been performed to (1) examine the phenomena controlling the progression of both transient and reflood phases of the experiment, (2) enhance our understanding of the phenomena occurring during reflood and add credence to the postulated phenomenological sequence, (3) assess the ability of SCDAP/RELAP5 to examine severe fuel damage issues and phenomena, and (4) identify code strengths and deficiencies with the intent of prioritizing code improvements. Results indicate that the code is able to analyze the early phases of severe fuel damage reasonably well, with potential deficiencies in modelling interaction between molten control rod material and intact fuel.

• Research Organization: EG and G Idaho, Inc., Idaho Falls, ID (United States)
• Sponsoring Organization: Nuclear Regulatory Commission, Washington, DC (United States)
• DOE Contract Number: AC07-76ID01570
• OSTI ID: 10164710
• Report Number(s): EGG-M-91248; CONF-911107-78; ON: DE92017883
• Resource Relation: Conference: 1991 winter meeting of the American Nuclear Society (ANS),San Francisco, CA (United States),10-15 Nov 1991; Other Information: PBD: [1991]
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
FUEL ELEMENTS
FISSION PRODUCT RELEASE
LOFT REACTOR
TRANSIENTS
S CODES
R CODES
REACTOR CORES
DAMAGE
COMPUTER CALCULATIONS
RADIATION TRANSPORT
LOSS OF COOLANT
HYDRAULICS
HEAT TRANSFER
REACTOR SAFETY
TESTING
220900
220502
210200
RADIOACTIVE EFFLUENTS
POWER REACTORS
NONBREEDING
LIGHT-WATER MODERATED
NONBOILING WATER COOLED