Study of Two-Phase Natural Circulation Cooling of Core Catcher System Using Scaled Model

Author:

Revankar Shripad T.1,Song Kiwon2,Rhee B. W.3,Park R. J.3,Ha K. S.3,Song J. H.3

Affiliation:

1. Fellow ASME School of Nuclear Engineering, Purdue University and Pohang University of Science and Technology, 400 Central Drive, West Lafayette, IN 47906 e-mail:

2. DANE, Pohang University of Science and Technology, Pohang, Gyeongbuk 790-784, South Korea e-mail:

3. Korean Atomic Energy Research Institute, Daejeon, Yuseong-gu 305-353, South Korea e-mail:

Abstract

A two-phase natural circulation cooling has been proposed to remove melted core decay heat by external core catcher cooling system during sever accident scenario. In this paper, two types of the core catcher cooling loops, one with heated loop and the other adiabatic loop simulated with air water system are analytically studied. First, a scaling analysis was carried out for natural circulation flow in a closed loop. Based on the scaling analyses, simulation of two-phase natural circulation is carried out both for air–water and steam–water system in an inclined rectangular channel. The heat flux corresponding to the decay heat is simulated with steam generation rate or air flux into the test section to produce equivalent flow quality and void fraction. Design calculations were carried out for typical core catcher design to estimate the expected natural circulation rates. The natural circulation flow rate and two-phase pressure drop were obtained for different heat inputs or equivalent air injection rates expressed as void fraction for a select downcomer pipe size. These results can be used to scale a steam water system using scaling consideration presented. The results indicate that the air–water and steam water system show similar flow and pressure drop behavior.

Publisher

ASME International

Subject

Fluid Flow and Transfer Processes,General Engineering,Condensed Matter Physics,General Materials Science

Reference18 articles.

1. Song, J. H., Kim, S. B., and Kim, H. D., 2000, “An Analysis of Natural Circulation Cooling of a Reactor Vessel During Severe Accidents by RELAP5/MOD3 Computer Code,” Proceedings of the Korean Nuclear Society Spring Meeting, Kori, Korea, pp. 1–10.

2. Analysis of External Cooling of the Reactor Vessel During Severe Accidents;Nucl. Technol.,2002

3. In-Vessel Retention Strategy for High Power Reactors,2003

4. A Non-Heating Experimental Study on the Two-Phase Natural Circulation Through the Annular Gap Between Reactor Vessel and Insulation Under External Vessel Cooling,2004

5. Corium Coolability Under Ex-Vessel Accident Conditions for LWRs;Nucl. Eng. Technol.,2009

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