Analyses of Passive Plasma Shutdown during Ex-vessel Loss of Coolant Accident in the First Wall Shield Blanket of Fusion Reactor.

• HONDA Takuro

  Hitachi Research Laboratory, Hitachi Ltd.

• BARTELS H-W

  ITER Joint Central Team, San Diego

• UCKAN N.A.

  Oak Ridge National Laboratory

• OKAZAKI Takashi

  Hitachi Research Laboratory, Hitachi Ltd.

• SEKI Yasushi

  Naka Fusion Research Establishment, Japan Atomic Energy Research Institute

Bibliographic Information

Other Title

• Analyses of Passive Plasma Shutdown dur

Abstract

An ex-vessel loss of coolant accident (LOCA) in the first wall/shield blanket of a fusion reactor has been analyzed by a hybrid code consisting of plasma dynamics and heat transfer analysis of in-vessel components. We investigated possibility of passive plasma shutdown scenario during the accident in International Thermonuclear Experimental Reactor (ITER). The safety analysis code which we developed can treat impurity concentration from the first wall and the divertor with a transport probability into the main plasma and a time delay given as input. It was found that the plasma is passively shutdown by a density limit disruption due to beryllium release from heated first wall surfaces about 168 seconds after the LOCA, when the transport probability of beryllium from the first wall into the plasma and the time delay were assumed to be 10^-2and the energy confinement time, respectively. At that time, the surface temperature of the outboard center (plasma facing component (PFC) with beryllium) and the temperature of the coolant tube in the first wall (stainless steel 316) reach about 1, 120°C and about 1, 080°C, respectively. Although the coolant tube does not melt, the copper heat sink between the PFC and the coolant tube melts before the passive shutdown. The heat sink of copper in the outboard baffle also melts before the passive shutdown, though the PFC surface of tungsten does not melt. Consequently, we have a possibility of passive plasma shutdown before the cooling tubes melt during the ex-LOCA of the first wall/shield blanket in ITER, however, further studies are needed on the effects on plasma burn control, impurity release and emission of implanted D-T fuel.

Journal

• Journal of Nuclear Science and Technology

  Journal of Nuclear Science and Technology 34 (6), 538-543, 1997

  Atomic Energy Society of Japan

Keywords

• ITER
• safety analysis
• passive plasma shutdown
• loss of coolant
• plasma dynamics
• thermal analysis
• impurities
• disruption
• density limit
• heat transfer

Details

• CRID: 1390001204093930112
• NII Article ID: 10002077302
• NII Book ID: AA00703720
• DOI: 10.3327/jnst.34.538
• COI: 1:CAS:528:DyaK2sXkvVCiu78%3D
• ISSN: 18811248
• ISSN: 00223131
• NDL BIB ID: 4240100
• Web Site: https://ndlsearch.ndl.go.jp/books/R000000004-I4240100
• Text Lang: en
• Data Source:
  • JaLC
  • NDL
  • Crossref
  • CiNii Articles
• Abstract License Flag: Disallowed