Transition Boiling Heat Transfer in the Semiscale Mod-3 Core During Reflood

The semi three-dimensional numerical analysis model of the FDR-ITER to simulate the in-vessel LOCA or the ICE was constructed using the modified TRAC-BF1 code in which the vacuum vessel, the pressure suppression tank and the relief pipe header are modeled using one VESSEL component. Analytical results obtained from the modified TRAC-BF1 code were compared with those from the MELCOR code concerning pressure in the vacuum vessel when a cooling pipe of 0.6 m 2 flow area breaks in it. The mass flow rate of the injected water into the vacuum vessel and the initial wall temperature in the modified TRAC-BF1 input data were the same as those calculated in the MELCOR analysis. The maximum pressure in the vacuum vessel obtained from the modified TRAC-BF1 code is 10% higher than that from the MELCOR code, but it is still below the design pressure 0.5 MPa of the FDR-ITER vacuum vessel. The semi three-dimensional analysis using the modified TRAC-BF1 obtained that the maximum delay in opening time among the four rupture-disks is within 3 ms. The maximum pressure in the vacuum vessel does not exceed the design pressure, even if one of rupture-disks is not opened.

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Numerical Analysis on Ingress of Coolant Event in Vacuum Vessel of Fusion Reactor Using Modified TRAC-BF1.

Kurihara

Ajima

Ueda³

et al. 2001

J. Nucl. Sci. Technol.

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Transition Boiling Heat Transfer in the Semiscale Mod-3 Core During Reflood

Cited by 3 publications

References 2 publications

Low quality transition and inverted annular flow film boiling of water: An updated review

Low quality transition and inverted annular flow film boiling of water: An updated review

Numerical Analysis on Ingress of Coolant Event in Vacuum Vessel of Fusion Reactor Using Modified TRAC-BF1

Numerical Analysis on Ingress of Coolant Event in Vacuum Vessel of Fusion Reactor Using Modified TRAC-BF1.

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