Experimental Investigation on the Heat Transfer Characteristics in a Vertical Upward Flow of Supercritical CO{sub 2}

Conference · Sat Jul 01 04:00:00 EDT 2006

OSTI ID:21016439

Kim, Hyungrae; Bae, Yoon Yeong; Kim, Hwan Yeol; Song, Jin Ho; Cho, Bong Hyun ^[1]

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1. Korea Atomic Energy Research Institute, 150 Deokjin, Yusung, Daejeon 305-353 (Korea, Republic of)

The SCWR (Supercritical Water-cooled Reactor) is one of the feasible options for the 4. generation nuclear power plant, which is being pursued by an international collaborative organization, the Gen IV International Forum (GIF). The major advantages of the SCWR include a high thermal efficiency and a maximum use of the existing technologies. In the SCWR, the coolant(water) of a supercritical pressure passes the pseudo-critical temperature as it flows upward through the sub-channels of the fuel assemblies. At certain conditions a heat transfer deterioration may occur near the pseudo-critical temperature and it may cause an excessive rise of the fuel surface temperature. Therefore, an accurate estimation of the heat transfer coefficient is necessary for the thermal-hydraulic design of a fuel pin, a fuel assembly, and the reactor core. A test facility, SPHINX, dedicated to produce heat transfer data and study flow characteristics, uses supercritical pressure CO{sub 2} as a medium to take advantage of the relatively low critical temperature and pressure; and similar physical properties with water. The produced data includes the temperature of the heating surface, the heat transfer coefficient, and the pressure drop at varying mass fluxes, heat fluxes, and operating pressures. The test section is a circular tube of ID 4.4 mm. The test range of the mass flux is 400 {approx} 1200 kg/m{sup 2}s and the maximum heat flux is 150 kW/m{sup 2}. The tests were performed for three different pressures, 7.75, 8.12, and 8.85 MPa. The test results are compared with the existing correlations of the heat transfer coefficient. In addition, the deterioration conditions observed in our test are compared against the criteria for a different fluid or a different tube size. (authors)

• Research Organization: American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 21016439
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
CARBON DIOXIDE
COOLANTS
CRITICAL TEMPERATURE
DESIGN
FLUIDS
FUEL ASSEMBLIES
FUEL PINS
HEAT FLUX
HEAT TRANSFER
NUCLEAR FUELS
NUCLEAR POWER PLANTS
PRESSURE DROP
REACTOR CORES
TEST FACILITIES
THERMAL EFFICIENCY
THERMAL HYDRAULICS
WATER
WATER COOLED REACTORS