Progress in the U.S./Japan PHENIX project for the technological assessment of plasma facing components for DEMO reactors

Yutai Katoh, Daniel Clark, Yoshio Ueda, Yuji Hatano, Minami Yoda, Adrian S. Sabau, Takehiko Yokomine, Lauren M. Garrison, J. Wilna Geringer, Akira Hasegawa, Tatsuya Hinoki, Masashi Shimada, Dean Buchenauer, Yasuhisa Oya, Takeo Muroga

研究成果: Article査読

10 被引用数 (Scopus)

抄録

The PHENIX Project is a 6-year U.S./Japan bilateral, multi-institutional collaboration program for the Technological Assessment of Plasma Facing Components for DEMO Reactors. The goal is to address the technical feasibility of helium-cooled divertor concepts using tungsten as the armor material in fusion power reactors. The project specifically attempts to (1) improve heat transfer modeling for helium-cooled divertor systems through experiments including steady-state and pulsed high-heat-load testing, (2) understand the thermomechanical properties of tungsten metals and alloys under divertor-relevant neutron irradiation conditions, and (3) determine the behavior of tritium in tungsten materials through high-flux plasma exposure experiments. The High Flux Isotope Reactor and the Plasma Arc Lamp facility at Oak Ridge National Laboratory, the Tritium Plasma Experiment facility at Idaho National Laboratory, and the helium loop at Georgia Institute of Technology are utilized for evaluation of the response to high heat loads and tritium interactions of irradiated and unirradiated materials and components. This paper provides an overview of the progress achieved during the first 3 years and discusses the plan for the remainder of the project.

本文言語English
ページ(範囲)222-232
ページ数11
ジャーナルFusion Science and Technology
72
3
DOI
出版ステータスPublished - 2017 10

ASJC Scopus subject areas

  • Civil and Structural Engineering
  • Nuclear and High Energy Physics
  • Nuclear Energy and Engineering
  • Materials Science(all)
  • Mechanical Engineering

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