Post-test analysis of semiscale tests S-UT-6 and S-UT-7 using TRAC PF1

Conference · Sat Jan 01 00:00:00 EST 1983

OSTI ID:6652273

Boyack, B E

A posttest study of Semiscale Tests S-UT-6 and S-UT-7 has been completed to assess TRAC-PFl predictions of pressurized water-reactor (PWR) small-break transients. The comparisons of the TRAC calculations and experimental results show that the correct qualitative influence of upper-head injection (UHI) was predicted. The major phenomenological difference predicted was the mode of core voiding. The data show a slow boiloff from the top of the core resulting in a dryout near the top of the core only. TRAC predicted a more extensive voiding with fluid forced from the bottom of the core by a pressure increase in the upper vessel plenum. The pressure increase was the primary consequence of a failure to predict a complete clearance of the seal in the intact-loop pump-suction upflow leg. Further review of the interphasic drag correlations, entrainment correlations, and critical-flow model is recommended. 20 figures.

• Research Organization: Los Alamos National Laboratory (LANL), Los Alamos, NM (United States)
• DOE Contract Number: W-7405-ENG-36
• OSTI ID: 6652273
• Report Number(s): LA-UR-82-1665; CONF-830103-7; ON: DE82018382
• Resource Relation: Conference: 2. international topical meeting on nuclear thermalhydraulics, Santa Barbara, CA, USA, 11 Jan 1982; Other Information: Portions of document are illegible
• 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
SIMULATION
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DRYOUT
REACTOR SAFETY EXPERIMENTS
TRANSIENTS
ACCIDENTS
REACTOR ACCIDENTS
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WATER COOLED REACTORS
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Light-Water Moderated
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