TRAC-PF1 code verification with data from the OTIS test facility. [Once-Through Intergral System]

Conference · Tue Jan 01 00:00:00 EST 1985

OSTI ID:5166963

Childerson, M T; Fujita, R K

A computer code (TRAC-PF1/MOD1) developed for predicting transient thermal and hydraulic integral nuclear steam supply system (NSSS) response was benchmarked. Post-small break loss-of-coolant accident (LOCA) data from a scaled, experimental facility, designated the One-Through Integral System (OTIS), were obtained for the Babcock and Wilcox NSSS and compared to TRAC predictions. The OTIS tests provided a challenging small break LOCA data set for TRAC verification. The major phases of a small break LOCA observed in the OTIS tests included pressurizer draining and loop saturation, intermittent reactor coolant system circulation, boiler-condenser mode, and the initial stages of refill. The TRAC code was successful in predicting OTIS loop conditions (system pressures and temperatures) after modification of the steam generator model. In particular, the code predicted both pool and auxiliary-feedwater initiated boiler-condenser mode heat transfer.

• Research Organization: Los Alamos National Laboratory (LANL), Los Alamos, NM (United States)
• DOE Contract Number: W-7405-ENG-36
• OSTI ID: 5166963
• Report Number(s): LA-UR-85-1265; CONF-851125-1; ON: DE85010709
• Resource Relation: Conference: American Society of Mechanical Engineers winter annual meeting, Miami, FL, USA, 17 Nov 1985
• 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
PWR TYPE REACTORS
COMPUTER CODES
STEAM GENERATORS
T CODES
TEST FACILITIES
ACCIDENTS
BOILERS
REACTOR ACCIDENTS
REACTORS
VAPOR GENERATORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled