Natural convection cooling of a vertical channel

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

OSTI ID:6911840

Dougherty, T; Fighetti, C; Yang, B W ^[1]; McAssey, E Jr ^[2]; Smalec, L ^[3]; Fauske, H ^[4]; Grolmes, M ^[5]

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1. Columbia Univ., New York, NY (United States)
2. Villanova Univ., PA (United States)
3. Babcock Wilcox Co., Lynchburg, VA (United States)
4. Fauske Associates Inc., Burr Ridge, IL (United States)
5. Centaurus Technology Inc., Lisle, IL (United States)

An experimental program has been conducted to determine the feasibility of natural convection cooling of a reactor following a beyond-design-based accident. The particular application under consideration was the heavy-water new production reactor. The questions to be resolved include the verification that a natural convection cooling pattern would be established and the determination of the power limit for which convective cooling will occur for a significant period of time. In the experiment, the reactor configuration was simulated using small-diameter vertical heated tubes in parallel with a large-diameter bypass line. Following a loss-of-flow event, the flow in the bypass line will reverse direction and pass through the heated channel by means of natural convection. If, however, the channel power is too high, void formation will block the channel and prevent the reverse flow pattern from occurring. The test procedure involved the establishment of a set of steady-state conditions and then the simulation of flow loss. As voids form in the heated length, the flow in the bypass line reverses and cooling of the test section occurs. If the power input is too high for the test conditions (exit pressure and inlet temperature), the pressure drop caused by the voids would prevent the cooling fluid from entering the test section, and a temperature excursion would occur. To avoid burnout, thermocouples on the test section trip the power and prevent burnout. Tests were conducted using one and two heated sections.

• OSTI ID: 6911840
• 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

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210400 -- Power Reactors
Nonbreeding
Otherwise Moderated or Unmoderated
220600* -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
42 ENGINEERING
420400 -- Engineering-- Heat Transfer & Fluid Flow
ACCIDENTS
BURNOUT
CONVECTION
DESIGN BASIS ACCIDENTS
ENERGY TRANSFER
FLUID MECHANICS
HEAT TRANSFER
HEAVY WATER COOLED REACTORS
LOSS OF FLOW
MASS TRANSFER
MECHANICS
NATURAL CONVECTION
PRESSURE DROP
REACTOR ACCIDENTS
REACTOR CHANNELS
REACTOR COMPONENTS
REACTORS
TESTING
VOIDS