Modeling and Validation of In-Vessel Debris Cooling during LWR Severe Accident
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概要
- 論文の詳細を見る
The authors proposed the heat transfer models to evaluate in-vessel debris cooling during a severe accident in a light water reactor (LWR), where the heat flux on the vessel inner surface was restricted by countercurrent flow limitation (CCFL) at the top end of the narrow gap between the overheated core debris and the reactor pressure vessel (RPV) or the local boiling heat flux, and derived correlations for CCFL and boiling heat fluxes using the existing data. In this paper, we improved the correlation of nucleate boiling heat fluxes using quenching data at the pressure of 0.1 MPa and the ALPHA test data at 1.3 MPa. Using the correlations, we calculated transient vessel temperatures in the LAVA experiment at 1.7 MPa. During the heating process of the vessel, the calculated average temperature was determined by CCFL and agreed well with the average of the measured temperatures. During its cooling process, the calculated local cooling rate of the vessel without the CCFL correlation was greatly affected by the correlation of nucleate boiling heat fluxes and the calculated results using the improved correlation agreed well with the measured values. As a result, applicability of the correlations for CCFL and boiling heat fluxes to in-vessel cooling was validated.
- 一般社団法人 日本原子力学会の論文
- 2005-04-25
著者
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Murase Michio
Institute Of Nuclear Safety System Inc.
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NAGAE Takashi
Institute of Nuclear Safety System, Incorporated
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Murase Michio
Institute Of Nuclear Safety System
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Nagae Takashi
Institute Of Nuclear Safety System Incorporated
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Nagae Takashi
Institute Of Nuclear Safety System
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OKANO Yukimitsu
Institute of Nuclear Safety System, Inc.
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Okano Yukimitsu
Institute Of Nuclear Safety System Incorporated
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Okano Yukimitsu
Institute Of Nuclear Safety System Inc.
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