Best-estimate analysis of a loss-of-coolant accident in a four-loop US PWR using TRAC-PD2

Conference · Fri Jan 01 00:00:00 EST 1982

OSTI ID:5338200

Ireland, J R

A 200-percent double-ended cold-leg break loss-of-coolant accident (LOCA) in a typical US pressurized water reactor (PWR) was simulated using the Transient Reactor Analysis Code (TRAC-PD2). The reactor system modeled represented a typical US PWR with four loops (three intact, one broken) and cold-leg emergency-core-cooling systems (ECCS). The finely noded TRAC model employed 440 three dimensional (r, THETA, z) vessel cells along with approximately 300 one-dimensional cells that modeled the primary system loops. The calculated peak-clad temperature of 950/sup 0/K occurred during blowdown and the clad temperature excursion was terminated at 175 s, when complete core quenching occurred. Accumulator flows were initiated at 10 s, when the system pressure reached 4.08 MPa, and the refill phase ended at 36 s when the lower plenum refilled. During reflood, both bottom and falling film quench fronts were calculated.

• Research Organization: Los Alamos National Laboratory (LANL), Los Alamos, NM (United States)
• DOE Contract Number: W-7405-ENG-36
• OSTI ID: 5338200
• Report Number(s): LA-UR-81-2899; CONF-820901-1; ON: DE82000726
• Resource Relation: Conference: International heat transfer conference, Munchen, F.R. Germany, 6 Sep 1982
• 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
HEAT TRANSFER
HYDRAULICS
TEMPERATURE GRADIENTS
PWR TYPE REACTORS
COMPUTER CALCULATIONS
ECCS
REACTOR SAFETY
ACCIDENTS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
FLUID MECHANICS
MECHANICS
REACTOR ACCIDENTS
REACTOR PROTECTION SYSTEMS
REACTORS
SAFETY
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled