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

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概要

• 論文の詳細を見る

• 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 m2 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.

• 社団法人 日本原子力学会の論文
• 2001-07-25

著者

• Seki Yasushi

  Nuclear Technology And Education Center Tokai Research Establishment Japan Atomic Energy Research In

• KURIHARA Ryoichi

  Fusion Reactor System Laboratory, Naka Research Establishment, Japan Atomic Energy Research Institut

• AJIMA Toshio

  Hitachi, Ltd.

• UEDA Shuzo

  JAERI Wien Office, Japan Atomic Energy Research Institute

• Ueda Shuzo

  Jaeri Wien Office Japan Atomic Energy Research Institute

• Kurihara Ryoichi

  Fusion Reactor System Laboratory Naka Research Establishment Japan Atomic Energy Research Institute

• Ajima Toshio

  Hitachi Ltd.

関連論文

• Transient Behaviors of Plasma and In-vessel Components when Considering Divertor Plasma State Transition in a Fusion Reactor
• Safety Analyses for Transient Behavior of Plasma and In-vessel Components during Plasma Abnormal Events in Fusion Reactor
• Analyses of Passive Plasma Shutdown during Ex-vessel Loss of Coolant Accident in the First Wall/Shield Blanket of Fusion Reactor
• Development of Time Dependent Safety Analysis Code for Plasma Anomaly Events in Fusion Reactors
• Analyses for Passive Safety of Fusion Reactor during Ex-Vessel Loss of Coolant Accident
• Numerical Analysis on Ingress of Coolant Event in Vacuum Vessel of Fusion Reactor Using Modified TRAC-BF1
• Development of Dust Removal System Using static Electricity for Fusion Experimental Reactors

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