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

Abstract

A 200% 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 and cold-leg emergency-core-cooling systems (ECCS). The calculated peak cladding temperature of 950 K occurred during blowdown and the cladding 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. Top quenching was caused by entrainment from the lower plenum and lower core regions. The entrained liquid was sufficient to form a small, saturated pool (0.3 m deep) above the upper core support plate. Also, some of the entrained liquid was carried out the hot legs and vaporized in the steam generators. Strong multidimensional effects were calculated in the reactor vessel, particularly with respect to rod quenching.

Authors:
Publication Date:
Research Org.:
Los Alamos National Lab., NM (USA)
OSTI Identifier:
6869762
Report Number(s):
LA-UR-82-1026; CONF-821101-2
ON: DE82014082
DOE Contract Number:  
W-7405-ENG-36
Resource Type:
Conference
Resource Relation:
Conference: ASME winter annual meeting, Pheonix, AZ, USA, 14 Nov 1982; Other Information: Portions of document are illegible
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; LOSS OF COOLANT; HEAT TRANSFER; HYDRAULICS; PWR TYPE REACTORS; COMPUTER CALCULATIONS; ECCS; REACTOR SAFETY; TEMPERATURE GRADIENTS; 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

Citation Formats

Ireland, J R. Best-estimate analysis of a loss-of-coolant accident in a four-loop US PWR using TRAC-PD2. United States: N. p., 1982. Web.
Ireland, J R. Best-estimate analysis of a loss-of-coolant accident in a four-loop US PWR using TRAC-PD2. United States.
Ireland, J R. Fri . "Best-estimate analysis of a loss-of-coolant accident in a four-loop US PWR using TRAC-PD2". United States.
@article{osti_6869762,
title = {Best-estimate analysis of a loss-of-coolant accident in a four-loop US PWR using TRAC-PD2},
author = {Ireland, J R},
abstractNote = {A 200% 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 and cold-leg emergency-core-cooling systems (ECCS). The calculated peak cladding temperature of 950 K occurred during blowdown and the cladding 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. Top quenching was caused by entrainment from the lower plenum and lower core regions. The entrained liquid was sufficient to form a small, saturated pool (0.3 m deep) above the upper core support plate. Also, some of the entrained liquid was carried out the hot legs and vaporized in the steam generators. Strong multidimensional effects were calculated in the reactor vessel, particularly with respect to rod quenching.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {1982},
month = {1}
}

Conference:
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