Hydrogen permeation in FeCrAl alloys for LWR cladding application

Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; Snead, Lance L.

doi:10.1016/j.jnucmat.2015.02.040

Title: Hydrogen permeation in FeCrAl alloys for LWR cladding application

Journal Article · Thu Mar 19 00:00:00 EDT 2015 · Journal of Nuclear Materials

DOI:https://doi.org/10.1016/j.jnucmat.2015.02.040· OSTI ID:1185774

Hu, Xunxiang ^[1]; Terrani, Kurt A. ^[2]; Wirth, Brian D. ^[1]; Snead, Lance L. ^[2]

Univ. of Tennessee, Knoxville, TN (United States); Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

FeCrAl is an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In our study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. Also, the total tritium inventory inside the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.

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Research Organization:: Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

Sponsoring Organization:: USDOE Office of Nuclear Energy (NE)

Grant/Contract Number:: AC05-00OR22725

OSTI ID:: 1185774

Alternate ID(s):: OSTI ID: 1246642

Journal Information:: Journal of Nuclear Materials, Vol. 461, Issue C; ISSN 0022-3115

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

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Cited by: 36 works

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