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

Conference · Sun Mar 01 00:00:00 EST 1992

OSTI ID:10123460

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 Laboratory (LANL), Los Alamos, NM (United States)
• Sponsoring Organization: Nuclear Regulatory Commission, Washington, DC (United States)
• DOE Contract Number: W-7405-ENG-36
• OSTI ID: 10123460
• Report Number(s): LA-UR-92-203; CONF-920804-4; ON: DE92007435
• Resource Relation: Conference: 28. national heat transfer conference,San Diego, CA (United States),9-12 Aug 1992; Other Information: PBD: [1992]
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
99 GENERAL AND MISCELLANEOUS//MATHEMATICS, COMPUTING, AND INFORMATION SCIENCE
PWR TYPE REACTORS
LOSS OF COOLANT
HEAT TRANSFER
REACTOR SAFETY
HYDRAULICS
T CODES
COMPUTERIZED SIMULATION
220900
210200
990200
POWER REACTORS
NONBREEDING
LIGHT-WATER MODERATED
NONBOILING WATER COOLED
MATHEMATICS AND COMPUTERS