Hydrogen Isotopes Permeation in Clean or Unoxidized FeCrAl Alloys: A Review

The US Department of Energy is working with fuel vendors to develop accident tolerant fuels (ATF) for the current fleet of light water reactors (LWRs). The ATF should be more resilient to loss of coolant accident scenarios and help extending the life of the operating LWRs. One of the proposed ATF concepts is to use iron-chromium-aluminum (FeCrAl) alloys for the cladding of the fuel. A concern in using ferritic FeCrAl is that this type of cladding may result in an increase in the concentration of tritium in the coolant. The objective of the current critical review is to collect and assess information from the literature regarding diffusion or permeation of hydrogen (H) and its isotopes deuterium (D) and Tritium (T) across industrial alloys (including FeCrAl) used or intended for the nuclear industry. Over a hundred years of data reviewed shows that the solubility of hydrogen in ferritic alloys is lower than in austenitic alloys but hydrogen permeates faster through a ferritic material than through austenitic materials. The tritium permeation rates in FeCrAl alloys are between those in austenitic stainless steels and in ferritic FeCr steels. The activation energy for hydrogen permeation is approximately 30 pct higher in the austenitic alloys compared with the ferritic (typically ∼ 50 kJ/mol in ferritic vs. ∼ 65 kJ/mol in the austenitic). None of the major elements in FeCrAl alloys react with hydrogen to form detrimental hydride phases. The effect of surface oxides on FeCrAl delaying hydrogen entrance into FeCrAl alloy is not part of this review.

Acid dissolution behavior of ferritic FeCrAl tubes candidates for nuclear fuel cladding.

The international materials community is engaged in finding safer alternatives to zirconium alloys for the cladding of fuel in light water reactors. One solution is to replace the zirconium cladding using ferritic iron-chromium-aluminum -FeCrAl- alloys, which offer extraordinary resistance to high temperature reaction with air or steam due to the formation of a protective alumina layer on the external surface. It is important to characterize the behavior of FeCrAl not only during accident conditions but in the entire fuel cycle, which may include reprocessing of the used fuel after it is removed from the power reactors. The reprocessing may involve the dissolution of the fuel rods in mineral acids. Little or nothing is known on the dissolution of FeCrAl alloys in common mineral acids, therefore the objective of this research was to study the dissolution of typical cladding tubing having two compositions of FeCrAl (APMT & C26M) in three acids (H2SO4, HNO3 & HCl) as a function of the temperature using both standard ASTM immersion tests as well as electrochemical tests. The dissolution behavior of the FeCrAl alloys is compared to the dissolution capability of other traditional nuclear materials such as austenitic stainless steels (304SS & 316SS) and austenitic nickel alloys (Alloy 600 and Hastelloy C-276). Results show that both C26M and APMT have a higher dissolution capability in the studied mineral acids, which will be beneficial for reprocessing procedures.