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

doi:10.1080/15361055.2017.1333868

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

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Abstract

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.

Original language	English
Pages (from-to)	222-232
Number of pages	11
Journal	Fusion Science and Technology
Volume	72
Issue number	3
DOIs	https://doi.org/10.1080/15361055.2017.1333868
State	Published - Oct 2017

Funding

energy research and development. These projects are defined and approved by the Coordinating Committee on Fusion Energy under the “Agreement between the Government of the United States of America and the Government of Japan on Cooperation in Research and Development in Energy and Related Fields” signed in 1979. Since the first project, RTNS-II, that started in 1981, the core research subjects have evolved from This manuscript was authored by UT-Battelle, LLC, under contract number DE-AC05-00OR22725 with the U.S. Department of Energy. The work presented was performed as a part of the U.S.–Japan PHENIX Cooperation Project on Technological Assessment of Plasma Facing Components for DEMO Reactors, supported by the U.S. Department of 1. T. MUROGA et al., “Midterm Summary of Japan-US Fusion Cooperation Program TITAN,” Fusion Sci. Technol., 60, 321 (2011); https://doi.org/10.13182/FST11-A12373. Energy, Office of Science, Fusion Energy Sciences and Ministry of Education, Culture, Sports, Science and Technology, Japan. The authors thank Xunxiang Hu for a technical review and Deborah Counce for an editorial review of the manuscript.

Keywords

Helium-cooled divertor
Plasma facing components
Tungsten armor

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10.1080/15361055.2017.1333868

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Katoh, Y., Clark, D., Ueda, Y., Hatano, Y., Yoda, M., Sabau, A. S., Yokomine, T., Garrison, L. M., Geringer, J. W., Hasegawa, A., Hinoki, T., Shimada, M., Buchenauer, D., Oya, Y., & Muroga, T. (2017). Progress in the U.S./Japan PHENIX project for the technological assessment of plasma facing components for DEMO reactors. Fusion Science and Technology, 72(3), 222-232. https://doi.org/10.1080/15361055.2017.1333868

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abstract = "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.",

keywords = "Helium-cooled divertor, Plasma facing components, Tungsten armor",

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

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doi = "10.1080/15361055.2017.1333868",

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Katoh, Y, Clark, D, Ueda, Y, Hatano, Y, Yoda, M, Sabau, AS, Yokomine, T, Garrison, LM, Geringer, JW, Hasegawa, A, Hinoki, T, Shimada, M, Buchenauer, D, Oya, Y & Muroga, T 2017, 'Progress in the U.S./Japan PHENIX project for the technological assessment of plasma facing components for DEMO reactors', Fusion Science and Technology, vol. 72, no. 3, pp. 222-232. https://doi.org/10.1080/15361055.2017.1333868

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N2 - 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.

AB - 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.

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

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