Comparison of steam-generator liquid holdup and core uncovery in two facilities of differing scale
This paper reports on Run SB-CL-05, a test similar to Semiscale Run S-UT-8. The test results show that the core was uncovered briefly during the accident and that the rods overheated at certain core locations. Liquid holdup on the upflow side of the steam-generator tubes was observed. After the loop seal cleared, the core refilled and the rods cooled. These behaviors were similar to those observed in the Semiscale run. The Large-Scale Test Facility (LSTF) Run SB-CL-06 is a counterpart test to Semiscale Run S-LH-01. The comparison of the results of both tests shows similar phenomena. The similarity of phenomena in these two facilities build confidence that these results can be expected to occur in a PWR. Similar holdup has now been observed in the 6 tubes of Semiscale and in the 141 tubes of LSTF. It is now more believable that holdup may occur in a full-scale steam generator with 3000 or more tubes. These results confirm the scaling of these phenomena from Semiscale (1/1705) to LSTF (1/48). The TRAC results for SB-CL-05 are in reasonable agreement with the test data. TRAC predicted the core uncovery and resulting rod heatup. The liquid holdup on the upflow side of the steam-generator tubes was also correctly predicted. The clearing of the loop seal allowed core recovery and cooled the overheated rods just as it had in the data. The TRAC analysis results of Run SB-CL-05 are similar to those from Semiscale Run S-UT-8. The ability of the TRAC code to calculate the phenomena equally well in the two experiments of different scales confirms the scalability of the many models in the code that are important in calculating this small break.
- 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:
- 6174305
- Report Number(s):
- LA-UR-87-1412; CONF-870816-14; ON: DE87009010
- Resource Relation:
- Conference: 24. national heat transfer conference and exhibition, Pittsburgh, PA, USA, 9 Aug 1987; Other Information: Portions of this document are illegible in microfiche products
- Country of Publication:
- United States
- Language:
- English
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PWR TYPE REACTORS
LOSS OF COOLANT
REACTOR SAFETY EXPERIMENTS
TRANSIENTS
INTERLABORATORY COMPARISONS
EXPERIMENTAL DATA
FAILURES
HEAT TRANSFER
HYDRAULICS
STEAM GENERATORS
T CODES
TEST FACILITIES
ACCIDENTS
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220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
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Light-Water Moderated
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