Primary coolant pump effects on core thermal response during large break LOCA transients in a commercial pressurized water reactor
A computer study was made of the thermal-hydraulic response of a full-scale pressurized water reactor (PWR) to a hypothesized, loss-of-coolant accident (LOCA). This study was intended to determine whether system conditions could exist which would prevent an early fuel-rod quench during a 200%, double-ended, cold leg (DECL) break LOCA. The method of investigation was a study of system conditions which could lead to early core-flow stagnation persisting through the blowdown phase of the transient. This study provides a reference for the planning of loss-of-fluid test (LOFT) experiments L2-5 and L2-6. The PWR considered in this study is the Zion plant, which is similar in design to the LOFT facility. The system conditions found to cause the most severe thermal transient on the fuel rods during a 200%, DECL break LOCA were one intact-loop, primary coolant pump shaft broken and the remaining primary coolant pumps unpowered. 8 refs., 36 figs., 14 tabs.
- Research Organization:
- EG and G Idaho, Inc., Idaho Falls (USA)
- DOE Contract Number:
- AC07-76ID01570
- OSTI ID:
- 5624036
- Report Number(s):
- EGG-LOFT-5505; ON: TI85014377
- Resource Relation:
- Other Information: Includes 5 sheets of 48x reduction microfiche
- Country of Publication:
- United States
- Language:
- English
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Numerical simulation of PWR response to a small break LOCA (loss-of-coolant accident) with reactor coolant pumps operating
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
ZION-1 REACTOR
LOSS OF COOLANT
PRIMARY COOLANT CIRCUITS
BENCH-SCALE EXPERIMENTS
BLOWDOWN
COMPUTER CODES
COMPUTERIZED SIMULATION
FUEL DENSIFICATION
FUEL RODS
HEAT TRANSFER
HYDRAULICS
LEAKS
PUMPS
REACTOR SAFETY
RELIABILITY
U CODES
VALIDATION
ACCIDENTS
COOLING SYSTEMS
ENERGY SYSTEMS
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
FLUID MECHANICS
FUEL ELEMENTS
MECHANICS
POWER REACTORS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTORS
SAFETY
SIMULATION
TESTING
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled