Analysis and experimental results on ingress of coolant event in vacuum vessel

Kurihara, Ryoichi; Ajima, Toshio; Kunugi, Tomoaki; Takase, Kazuyuki; Shibata, Mitsuhiko; Seki, Yasushi; Hosokai, I.; Ohmori, J.; Yamauchi, M.; Kasahara, Fumio

doi:10.1016/s0920-3796(97)00182-8

Cited by 9 publications

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“…At the Japan Atomic Energy Research Institute (JAERI), the ICE simulation experiment 2) has been carried out to obtain thermofluid data for verifying the safety analysis codes 3) to evaluate the safety margin at the ICE of the fusion reactor. This experiment simulates breaking of the cooling pipes for the blanket or the divertor.…”

Section: Introductionmentioning

confidence: 99%

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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

Journal of Nuclear Science and Technology

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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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Section: Introductionmentioning

confidence: 99%

“…3) On the other hand, the MELCOR code, 6) which was originally developed at the Sandia National Laboratory (SNL) to predict progression of severe accidents in nuclear fission reactor, has been modified to simulate accidents in fusion reactors at the INEEL.…”

Section: Introductionmentioning

confidence: 99%