Comparison of a TRAC calculation to the data from LSTF run SB-CL-05
Run SB-CL-05 is a 5% break in the side of the cold leg. The test results show that the core was uncovered briefly and that the rods overheated at certain core locations. Liquid holdup on the upflow side of the steam generator tubes was observed. When the loop seal cleared, the core refilled and the rods cooled. The TRAC results are in reasonable agreement with the test data, meaning that TRAC correctly predicted the major trends and phenomena. TRAC predicted the core uncovery, the resulting rod heatup, and the liquid holdup on the upflow side of the steam generator tubes correctly. The clearing of the loop seal allowed core recovery and cooled the overheated rods just as it had in the data, but TRAC predicted its occurrence 20 s late. The experimental and TRAC analysis results of run SB-CL-05 are similar to those for Semiscale Run S-UT-8. In both runs there was core uncovery, rod overheating, and steam generator liquid holdup. These results confirm scaling of these phenomena from Semiscale (1/1650) to LSTF (1/48).
- Research Organization:
- Los Alamos National Laboratory (LANL), Los Alamos, NM (United States); Idaho National Laboratory (INL), Idaho Falls, ID (United States)
- DOE Contract Number:
- W-7405-ENG-36
- OSTI ID:
- 7037566
- Report Number(s):
- LA-UR-86-3692; CONF-8610135-35; ON: DE87001985
- Resource Relation:
- Conference: 14. water reactor safety information meeting, Gaithersburg, MD, USA, 27 Oct 1986; Other Information: Portions of this document are illegible in microfiche products
- Country of Publication:
- United States
- Language:
- English
Similar Records
RELAP5 assessment using LSTF test data SB-CL-18
An investigation of core liquid level depression in small break loss-of-coolant accidents
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
HEAT TRANSFER
T CODES
HYDRAULICS
LOSS OF COOLANT
PWR TYPE REACTORS
BOUNDARY CONDITIONS
COMPARATIVE EVALUATIONS
TEST FACILITIES
TRANSIENTS
ACCIDENTS
COMPUTER CODES
ENERGY TRANSFER
FLUID MECHANICS
MECHANICS
REACTOR ACCIDENTS
REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled