Analysis of results from a loss-of-offsite-power-initiated ATWS experiment in the LOFT facility. [PWR]
An anticipated transient without scram (ATWS), initiated by loss-of-offsite power, was experimentally simulated in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR). Primary system pressure was controlled using a scaled safety relief valve (SRV) representative of those in a commercial PWR, while reactor power was reduced by moderator reactivity feedback in a natural circulation mode. The experiment showed that reactor power decreases more rapidly when the primary pumps are tripped in a loss-of-offsite-power ATWS than in a loss-of-feedwater induced ATWS when the primary pumps are left on. During the experiment, the SRV had sufficient relief capacity to control primary system pressure. Natural circulation was effective in removing core heat at high temperature, pressure, and core power. The system transient response predicted using the RELAP5/MOD1 computer code showed good agreement with the experimental data.
- Research Organization:
- EG and G Idaho, Inc., Idaho Falls (USA); Public Service Co. of New Hampshire, Seabrook (USA); Japan Atomic Energy Research Inst., Tokai, Ibaraki; Framatome, 92 - Paris La Defense (France)
- DOE Contract Number:
- AC07-76ID01570
- OSTI ID:
- 5624591
- Report Number(s):
- EGG-M-27682; CONF-830702-25; ON: DE84000560
- Resource Relation:
- Conference: 21. ASME/AIChE national heat transfer conference, Seattle, WA, USA, 24 Jul 1983
- Country of Publication:
- United States
- Language:
- English
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21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
ATWS
HEAT TRANSFER
HYDRAULICS
PWR TYPE REACTORS
COMPUTER CALCULATIONS
EXPERIMENTAL DATA
FAILURES
LOFT REACTOR
POWER SUPPLIES
PRESSURE GRADIENTS
REACTOR SAFETY
TEMPERATURE GRADIENTS
THEORETICAL DATA
ACCIDENTS
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ELECTRONIC EQUIPMENT
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NUMERICAL DATA
REACTOR ACCIDENTS
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RESEARCH AND TEST REACTORS
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TANK TYPE REACTORS
TEST REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
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