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Microstructural Evolution of Self-Ion Irradiation HT9

[+] Author Affiliations
Elizabeth Beckett, Zhijie Jiao, Kai Sun, Gary S. Was

University of Michigan, Ann Arbor, MI

Micah Hackett

TerraPower, LLC, Bellevue, WA

Paper No. ICONE21-16595, pp. V001T02A043; 7 pages
doi:10.1115/ICONE21-16595
From:
  • 2013 21st International Conference on Nuclear Engineering
  • Volume 1: Plant Operations, Maintenance, Engineering, Modifications, Life Cycle and Balance of Plant; Nuclear Fuel and Materials; Radiation Protection and Nuclear Technology Applications
  • Chengdu, China, July 29–August 2, 2013
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 978-0-7918-5578-2
  • Copyright © 2013 by ASME

abstract

Ferritic/martensitic steels are candidates for fast reactors because of their sodium compatibility, superior resistance to corrosion and radiation damage, including swelling, and excellent thermal conductivity and thermal expansion coefficient. One significant limitation of any cladding material is its susceptibility to swelling at high doses. While HT9 has neutron irradiation performance data up to ∼200 dpa, dose requirements for the Traveling Wave Reactor (TWR) may be much higher. Obtaining higher-dose data will take many years, but in the interim, heavy ion irradiation could provide a useful tool toward predicting the swelling trends beyond 200 dpa. In this study, HT9 was irradiated from 440–480°C using 5 MeV Fe++ ions. The samples are compared to a portion of HT9 fuel assembly duct from FFTF, which was characterized after neutron irradiation at 440°C with an accumulated dose of 155 dpa. Comparisons are made of the void size and density using transmission electron microscopy (TEM). The increase in dose from 280 dpa to 375 dpa increased void size, number density and swelling at 440°C, while swelling was generally lower at 480°C for the same helium pre-implantation conditions. Helium generally enhanced the nucleation of voids, as measured by the void density.

Copyright © 2013 by ASME

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