PHENIX U.S.-Japan Collaboration Investigation of Thermal and Mechanical Properties of Thermal Neutron–Shielded Irradiated Tungsten

Lauren M. Garrison, Yutai Katoh, Josina W. Geringer, Masafumi Akiyoshi, Xiang Chen, Makoto Fukuda, Akira Hasegawa, Tatsuya Hinoki, Xunxiang Hu, Takaaki Koyanagi, Eric Lang, Michael McAlister, Joel McDuffee, Takeshi Miyazawa, Chad Parish, Emily Proehl, Nathan Reid, Janet Robertson, Hsin Wang

Research output: Contribution to journalArticlepeer-review

22 Citations (Scopus)


The United States and Japan have collaborated on fusion materials research in a series of agreements reaching back to 1981. The PHENIX collaboration is the latest U.S.-Japan project which spans 2013 to 2019 and has the goal of assessing technical feasibility of tungsten-based, helium-cooled plasma-facing component concepts for a demonstration fusion power reactor (DEMO). Task 2 within the PHENIX project is focused on evaluating the neutron irradiation effects in tungsten. For tungsten, the transmutation to Re and Os is at least as important to determining its properties after irradiation as the displacement damage, and the transmutation rate depends on the energy spectrum of the reactor. A large-scale, instrumented irradiation capsule with thermal neutron shielding to better mimic fusion conditions was irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. The tungsten specimens were irradiated in different temperature zones between 500°C and 1200°C to doses of ~0.2 to 0.7 displacements per atom. More than 20 varieties of pure tungsten and tungsten alloys were included in the irradiation, and they were evaluated in the 3025E hot-cell facility and at the Low Activation Materials Development and Analysis Laboratory. The elevated temperature tensile, fracture toughness, hardness, thermal conductivity, electrical resistivity, density, elemental composition, and microstructure properties of the irradiated materials are being collected. This paper overviews the experimental design, specimen matrix, and the initial results of postirradiation examinations.

Original languageEnglish
Pages (from-to)499-509
Number of pages11
JournalFusion Science and Technology
Issue number6
Publication statusPublished - 2019 Aug 18


  • Neutron irradiation
  • divertor
  • fusion materials
  • mechanical properties
  • tungsten

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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