In-reactor corrosion behavior of stainless steel in high temperature sodium

Sodium-cladding chemical interaction has been studied on fuel pins clad with 20 percent cold worked type 316 stainless steel and irradiated in the EBR-II at temperatures up to 7050C and for exposures to 5300 hours. The measured corrosion rate of the cladding surface immediately above the top of the fuel column was 12.5 μm per year at 6900C. The loss of Ni at 7000C resulted in the formation of a ferrite layer approximately 5 μm thick. A zone depleted in Ni and Cr extends into the austenite from the ferrite-austenite interface an additional distance of approximately 15 μm. No large changes in volumetric average carbon or nitrogen were observed. The results suggest a slight increase in carbon during the first 2300 hours followed by a slow decrease for longer exposure times. A 25 to 30 percent decrease in bulk nitrogen was observed in cladding with an initial nitrogen content of 225 ppM. Cladding with a lower initial nitrogen content of 63 ppM showed essentially no change. Fuel pin cladding corrosion behavior agrees well with that predicted using out-of-reactor Na loop data, thus providing confidence that out-of-reactor corrosion data can be reliably used in predicting in-reactor fuel cladding performance