TRAC-PF1/MOD2 best-estimate analysis of a large-break LOCA in a 15 times 15 generic four-loop Westinghouse nuclear power plant

Conference · Wed Jan 01 04:00:00 EST 1992

OSTI ID:5244403

Spore, J W; Lin, J C; Schnurr, N M; White, J R; Cappiello, M C

Calculations of a large-break loss-of-coolant accident (LOCA) in a 15 {times} 15 generic four-loop Westinghouse nuclear power plant with both the TRAC-PF1/MOD1 and TRAC-PF1/MOD2 computer codes will be presented. The Transient Reactor Analysis Code (TRAC) has been developed by Los Alamos National Laboratory to provide advanced best-estimate simulations of real postulated transients in pressurized light-water reactors (LWRs) and for many related thermal-hydraulic facilities. The latest released version of TRAC is TRAC-PF1/MOD2. Significant improvements and enhancements over the MOD1 version were implemented in the MOD2 heat-transfer and constitutive models. One of the most significant improvements in the MOD2 code has been the implementation of the two-step numerics method in the three-dimensional components, which can significantly reduce run times for long, slow transients. A very important area of improvement has been in the reflood heat-transfer models. Developmental assessment results (i.e., code comparisons with experimental data) will be discussed for several separate-effects and integral test, including analysis of the Upper Plenum Test Facility (UPTF), the Cylindrical Core Test Facility (CCTF), and the Loss-of-Fluid Test Facility (LOFT). The assessment results provide information on the anticipated accuracy for the best-estimate models in the MOD2 computer code. The MOD1 to MOD2 comparison will provide an estimate for the effect of improved heat-transfer models on predicted peak cladding temperatures.

• Research Organization: Los Alamos National Lab., NM (United States)
• Sponsoring Organization: NRC; Nuclear Regulatory Commission, Washington, DC (United States)
• DOE Contract Number: W-7405-ENG-36
• OSTI ID: 5244403
• Report Number(s): LA-UR-92-203; CONF-920804--4; ON: DE92007435
• Country of Publication: United States
• Language: English

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99 GENERAL AND MISCELLANEOUS
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ACCIDENTS
COMPUTER CODES
COMPUTERIZED SIMULATION
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
MECHANICS
POWER REACTORS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
SIMULATION
T CODES
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS