Hydrogen permeation characteristics of some Fe-Cr-Al alloys

Deventer, E.H. Van; Maroni, V.A.

doi:10.1016/0022-3115(83)90167-8

“…It appears that the hydrogen permeability of FeCrAl alloys is five times greater than that of 304SS at 350°C and three times higher at 650°C because of the difference in crystal structrue. Compared with pure iron (brown dashed line) [26], the hydrogen permeabilities of the FeCrAl alloys are one to two orders of magnitude lower, which is consistent with the results reported in Van Deventer and Maroni [22]. The black dashed line in Fig.…”

Section: Hydrogen Permeability Of Fecral Alloyssupporting

confidence: 88%

“…The behavior of hydrogen isotopes in austenitic [13][14][15][16][17][18][19][20][21] and ferritic-martensitic steels [12,[22][23][24][25][26][27] has been studied both experimentally and theoretically at considerable depth since the 1960s. These studies aimed to measure hydrogen isotope permeability, diffusivity, and solubility and to investigate the rate-limiting mechanisms of hydrogen isotope transport through these alloys.…”

Section: Introductionmentioning

confidence: 99%

Hydrogen permeation in FeCrAl alloys for LWR cladding application

Hu

¹

,

Terrani

²

,

Wirth

³

et al. 2015

Journal of Nuclear Materials

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FeCrAl, 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 this 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. 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.

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“…It appears that the hydrogen permeability of FeCrAl alloys is five times greater than that of 304SS at 350°C and three times higher at 650°C because of the difference in crystal structrue. Compared with pure iron (brown dashed line) [26], the hydrogen permeabilities of the FeCrAl alloys are one to two orders of magnitude lower, which is consistent with the results reported in Van Deventer and Maroni [22]. The black dashed line in Fig.…”

supporting

confidence: 89%

“…The behavior of hydrogen isotopes in austenitic [13][14][15][16][17][18][19][20][21] and ferritic-martensitic steels [12,[22][23][24][25][26][27] has been studied both experimentally and theoretically at considerable depth since the 1960s. These studies aimed to measure hydrogen isotope permeability, diffusivity, and solubility and to investigate the rate-limiting mechanisms of hydrogen isotope transport through these alloys.…”

Section: Introductionmentioning

confidence: 99%

Defect evolution in single crystalline tungsten following low temperature and low dose neutron irradiation

Hu

¹

,

Koyanagi

²

,

Fukuda

³

et al. 2016

Journal of Nuclear Materials

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FeCrAl, 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 this 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. 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.

show abstract

“…4 and 9) in the ID of the cladding will significantly reduce the hydrogen permeation from the fuel to the coolant [25]. It has been shown that not only alumina would reduce permeation of tritium from the fuel to the coolant, other oxides will also reduce hydrogen or tritium permeation [26,27].…”

Section: Mitigation Measures To Neutron Absorption and Tritium Releasementioning

confidence: 99%

Iron-chrome-aluminum alloy cladding for increasing safety in nuclear power plants

Rebak

¹

2017

EPJ Nuclear Sci. Technol.

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Abstract. After a tsunami caused plant black out at Fukushima, followed by hydrogen explosions, the US Department of Energy partnered with fuel vendors to study safer alternatives to the current UO 2 -zirconium alloy system. This accident tolerant fuel alternative should better tolerate loss of cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. General electric, Oak ridge national laboratory, and their partners are proposing to replace zirconium alloy cladding in current commercial light water power reactors with an iron-chromium-aluminum (FeCrAl) cladding such as APMT or C26M. Extensive testing and evaluation is being conducted to determine the suitability of FeCrAl under normal operation conditions and under severe accident conditions. Results show that FeCrAl has excellent corrosion resistance under normal operation conditions and FeCrAl is several orders of magnitude more resistant than zirconium alloys to degradation by superheated steam under accident conditions, generating less heat of oxidation and lower amount of combustible hydrogen gas. Higher neutron absorption and tritium release effects can be minimized by design changes. The implementation of FeCrAl cladding is a near term solution to enhance the safety of the current fleet of commercial light water power reactors.

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Hydrogen permeation characteristics of some Fe-Cr-Al alloys

Cited by 31 publications

References 9 publications

Hydrogen permeation in FeCrAl alloys for LWR cladding application

Hydrogen permeation in FeCrAl alloys for LWR cladding application

Defect evolution in single crystalline tungsten following low temperature and low dose neutron irradiation

Iron-chrome-aluminum alloy cladding for increasing safety in nuclear power plants

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