An experimental study of external reactor vessel cooling strategy on the critical heat flux using the graphene oxide nano-fluid

Conference · Sun Jul 01 00:00:00 EDT 2012

OSTI ID:22107804

Park, S. D.; Lee, S. W.; Kang, S.; Kim, S. M.; Seo, H.; Bang, I. C. ^[1]

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1. Ulsan National Inst. of Science and Technology UNIST, 100 Banyeon-ri, Eonyang-eup, Ulju-gun, Ulasn Metropolitan City 689-798 (Korea, Republic of)

External reactor vessel cooling (ERVC) for in-vessel retention (IVR) of corium as a key severe accident management strategy can be achieved by flooding the reactor cavity during a severe accident. In this accident mitigation strategy, the decay heat removal capability depends on whether the imposed heat flux exceeds critical heat flux (CHF). To provide sufficient cooling for high-power reactors such as APR1400, there have been some R and D efforts to use the reactor vessel with micro-porous coating and nano-fluids boiling-induced coating. The dispersion stability of graphene-oxide nano-fluid in the chemical conditions of flooding water that includes boric acid, lithium hydroxide (LiOH) and tri-sodium phosphate (TSP) was checked in terms of surface charge or zeta potential before the CHF experiments. Results showed that graphene-oxide nano-fluids were very stable under ERVC environment. The critical heat flux (CHF) on the reactor vessel external wall was measured using the small scale two-dimensional slide test section. The radius of the curvature is 0.1 m. The dimension of each part in the facility simulated the APR-1400. The heater was designed to produce the different heat flux. The magnitude of heat flux follows the one of the APR-1400 when the severe accident occurred. All tests were conducted under inlet subcooling 10 K. Graphene-oxide nano-fluids (concentration: 10 -4 V%) enhanced CHF limits up to about 20% at mass flux 50 kg/m{sup 2}s and 100 kg/m{sup 2}s in comparison with the results of the distilled water at same test condition. (authors)

• Research Organization: American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 22107804
• Resource Relation: Conference: ICAPP '12: 2012 International Congress on Advances in Nuclear Power Plants, Chicago, IL (United States), 24-28 Jun 2012; Other Information: Country of input: France; 14 refs.; Related Information: In: Proceedings of the 2012 International Congress on Advances in Nuclear Power Plants - ICAPP '12| 2799 p.
• Country of Publication: United States
• Language: English

Related Subjects

42 ENGINEERING
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
ACCIDENT MANAGEMENT
AFTER-HEAT REMOVAL
BOILING
BORIC ACID
COMPARATIVE EVALUATIONS
CRITICAL HEAT FLUX
FLUIDS
GRAPHENE
LITHIUM HYDROXIDES
OXIDES
POROUS MATERIALS
PWR TYPE REACTORS
REACTOR VESSELS
SIMULATION
SODIUM PHOSPHATES
SUBCOOLING
TWO-DIMENSIONAL CALCULATIONS
WATER