Reflux cooling experiments on the NCSU scaled PWR facility

Conference · Fri Jan 01 04:00:00 EST 1993 · Transactions of the American Nuclear Society; (United States)

OSTI ID:6844108

Doster, J M; Giavedoni, E ^[1]

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1. North Carolina State Univ., Raleigh, NC (United States)

Under loss of forced circulation, coupled with the loss or reduction in primary side coolant inventory, horizontal stratified flows can develop in the hot and cold legs of pressurized water reactors (PWRs). Vapor produced in the reactor vessel is transported through the hot leg to the steam generator tubes where it condenses and flows back to the reactor vessel. Within the steam generator tubes, the flow regimes may range from countercurrent annular flow to single-phase convection. As a result, a number of heat transfer mechanisms are possible, depending on the loop configuration, total heat transfer rate, and the steam flow rate within the tubes. These include (but are not limited to) two-phase natural circulation, where the condensate flows concurrent to the vapor stream and is transported to the cold leg so that the entire reactor coolant loop is active, and reflux cooling, where the condensate flows back down the interior of the coolant tubes countercurrent to the vapor stream and is returned to the reactor vessel through the hot leg. While operating in the reflux cooling mode, the cold leg can effectively be inactive. Heat transfer can be further influenced by noncondensables in the vapor stream, which accumulate within the upper regions of the steam generator tube bundle. In addition to reducing the steam generator's effective heat transfer area, under these conditions operation under natural circulation may not be possible, and reflux cooling may be the only viable heat transfer mechanism. The scaled PWR (SPWR) facility in the nuclear engineering department at North Carolina State Univ. (NCSU) is being used to study the effectiveness of two-phase natural circulation and reflux cooling under conditions associated with loss of forced circulation, midloop coolant levels, and noncondensables in the primary coolant system.

• OSTI ID: 6844108
• Report Number(s): CONF-931160--
• Conference Information: Journal Name: Transactions of the American Nuclear Society; (United States) Journal Volume: 69
• Country of Publication: United States
• Language: English

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ACCIDENTS
COOLING SYSTEMS
COUNTER CURRENT
ENERGY SYSTEMS
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
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FLUID MECHANICS
HEAT TRANSFER
LOSS OF COOLANT
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POWER REACTORS
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PWR TYPE REACTORS
REACTOR ACCIDENTS
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REACTOR COOLING SYSTEMS
REACTOR SAFETY
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SAFETY
TEST REACTORS
THERMAL REACTORS
TWO-PHASE FLOW
WATER COOLED REACTORS
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