Comparison of computer codes (CE-THERM, FRAP-T5, GT3-FLECHT, and TRUMP-FLECHT) with data from the NRU-LOCA thermal hydraulic tests

Conference · Wed Jul 01 00:00:00 EDT 1981

OSTI ID:6204333

Mohr, C L; Rausch, W N; Hesson, G M

The LOCA Simulation Program in the NRU reactor is the first set of experiments to provide data on the behavior of full-length, nuclear-heated PWR fuel bundles during the heatup, reflood, and quench phases of a loss-of-coolant accident (LOCA). This paper compares the temperature time histories of 4 experimental test cases with 4 computer codes: CE-THERM, FRAP-T5, GT3-FLECHT, and TRUMP-FLECHT. The preliminary comparisons between prediction and experiment show that the state-of-the art fuel codes have large uncertainties and are not necessarily conservative in predicting peak temperatures, turn around times, and bundle quench times.

• Research Organization: Battelle Pacific Northwest Labs., Richland, WA (USA)
• DOE Contract Number: AC06-76RL01830
• OSTI ID: 6204333
• Report Number(s): PNL-SA-9328; CONF-810801-56; ON: DE81028136; TRN: 81-015162
• Resource Relation: Conference: 6. international conference on structural mechanics in reactor technology, Paris, France, 17 Aug 1981
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
LOSS OF COOLANT
COMPUTERIZED SIMULATION
PWR TYPE REACTORS
C CODES
COMPARATIVE EVALUATIONS
COMPUTER CODES
F CODES
FUEL ELEMENT CLUSTERS
G CODES
NRU REACTOR
T CODES
ACCIDENTS
FUEL ASSEMBLIES
HEAVY WATER COOLED REACTORS
HEAVY WATER MODERATED REACTORS
IRRADIATION REACTORS
ISOTOPE PRODUCTION REACTORS
NATURAL URANIUM REACTORS
REACTOR ACCIDENTS
REACTORS
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SIMULATION
TANK TYPE REACTORS
TEST REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled