TRAC-PF1/MOD1 analysis of a 200% cold-leg break in a US/Japanese PWR with four loops and 15 x 15 fuel

Conference · Wed Jan 01 00:00:00 EST 1986

OSTI ID:6050062

Spore, J W; Cappiello, M W

This report presents the results of a TRAC-PF1/MOD1 calculation that simulated a 200% double-ended cold-leg-break loss-of-coolant accident in a generic US/Japanese pressurized water reactor. This is a best-estimate analysis using conservative boundary conditions and minimum safeguards. The calculation shows that the peak cladding temperature (PCT) occurs during blowdown and that the core reheat is minimal during reflood. The results also show that for an evaluation-model peak rod linear power of 15.85 kW/ft, a PCT of 1084 K is reached at 3.5 s into the blowdown transient, which is approx.394 K below the design basis limit of 1478 K. 10 figs.

• Research Organization: Los Alamos National Laboratory (LANL), Los Alamos, NM (United States)
• DOE Contract Number: W-7405-ENG-36
• OSTI ID: 6050062
• Report Number(s): LA-UR-85-3959; CONF-860454-2; ON: TI86003680
• Resource Relation: Conference: 2. international topical meeting on nuclear power plant thermal hydraulics and operations, Tokyo, Japan, 15 Apr 1986
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
JAPAN
PWR TYPE REACTORS
LOSS OF COOLANT
HEAT TRANSFER
HYDRAULICS
REACTOR COOLING SYSTEMS
USA
BLOWDOWN
COMPUTERIZED SIMULATION
VERY HIGH TEMPERATURE
ACCIDENTS
ASIA
COOLING SYSTEMS
ENERGY SYSTEMS
ENERGY TRANSFER
FLUID MECHANICS
MECHANICS
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REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
SIMULATION
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled