Improved Accident Tolerance of Austenitic Stainless Steel Cladding through Colossal Supersaturation with Interstitial Solutes

Presently, LWR (light-water reactor) fuel cladding (tube) is made from Zr-base alloys (e.g., Zircaloy). However, in steam at temperatures above 1000 K (727 °C), these alloys oxidize rapidly, stripping oxygen from water molecules and producing hydrogen, which may cause explosion. Moreover, Zr-base alloys are prone to GTRF (“grid-to-rod” fretting) failure, in which flowinduced vibration of fuel rods touching the grid of the nuclear fuel assembly causes structural damage that can lead to failure. One tangible alternative to Zr-based alloys are austenitic stainless steels, which have significantly better oxidation resistance under accident conditions. However, they are prone to SCC (stress-corrosion cracking). Our goal is to improve the properties of austenitic stainless steels such that they can be used for nuclear fuel cladding in LWRs and significantly excel on currently used alloys with regard to performance, safety, service life, and accident tolerance. The method we propose for this purpose is based on a new concept of surface engineering: “case” hardening (i.e., generating a hard shell) by CSS (colossal super-saturation) with interstitial solutes infused through the alloy surface [1,2,3]. CSS with carbon or nitrogen to levels that correspond to ≈ 100,000 times the equilibrium solubility limit while suppressing precipitation of undesired carbides or nitrides can be accomplished if the following conditions are fulfilled: (i) The alloy must contain metal atoms with a high affinity for carbon, e.g., Cr in AISI-316L. (ii) The processing temperature must be chosen such that the metal atoms (Fe, Ni, Cr), residing in substitutional sites of the alloy crystal structure, are practically immobile and therefore cannot precipitate in carbides or nitrides, whereas the carbon or nitrogen atoms, as they reside in interstitial sites, can still diffuse over technically useful distances (a few micrometers) within the processing time. CSS can significantly enhance the mechanical properties (hardness, wear resistance, and fatigue life) and the corrosion resistance of structural alloys. Currently, we investigate whether these improved properties can be used to improve accident tolerance of nuclear fuel cladding. Of particular interest is whether CSS also improves the resistance to stress-corrosion cracking, and possibly the resistance to microscopic structural damage caused by irradiation with electrons, neutrons, or ions. Of critical importance for success is (i) to optimize the process that “activates” the alloy surface for infusion of interstitial