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Colin Judge: Testing structural materials in Idaho’s newest hot cell facility
Idaho National Laboratory’s newest facility—the Sample Preparation Laboratory (SPL)—sits across the road from the Hot Fuel Examination Facility (HFEF), which started operating in 1975. SPL will host the first new hot cells at INL’s Materials and Fuels Complex (MFC) in 50 years, giving INL researchers and partners new flexibility to test the structural properties of irradiated materials fresh from the Advanced Test Reactor (ATR) or from a partner’s facility.
Materials meant to withstand extreme conditions in fission or fusion power plants must be tested under similar conditions and pushed past their breaking points so performance and limitations can be understood and improved. Once irradiated, materials samples can be cut down to size in SPL and packaged for testing in other facilities at INL or other national laboratories, commercial labs, or universities. But they can also be subjected to extreme thermal or corrosive conditions and mechanical testing right in SPL, explains Colin Judge, who, as INL’s division director for nuclear materials performance, oversees SPL and other facilities at the MFC.
SPL won’t go “hot” until January 2026, but Judge spoke with NN staff writer Susan Gallier about its capabilities as his team was moving instruments into the new facility.
Donald L. Hagrman, Joy L. Rempe
Nuclear Technology | Volume 133 | Number 2 | February 2001 | Pages 194-212
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT01-A3169
Articles are hosted by Taylor and Francis Online.
A mechanistic model, based on a quasi-equilibrium analysis of oxidation reactions, is proposed for predicting high-temperature corium oxidation. The analysis suggests that oxide forming on the surface of corium containing uranium, zirconium, and iron is similar to the oxides formed on zirconium and uranium as long as there is a small percentage of unoxidized zirconium or uranium in the metallic phase. This is because of the higher affinity of zirconium and uranium for oxygen. Hence, oxidation rates and heat production rates are similar to (U,Zr) compounds until nearly all the uranium and zirconium in the corium oxidizes. Oxidation rates after this point are predicted to be similar to those implied by the oxide thickness present when the forming oxide ceases to be protective, and heat generation rates should be similar to those implied by iron oxidation, i.e., ~4% of the zirconium oxidation heating rate.The maximum atomic ratio of unoxidized iron to unoxidized liquid zirconium plus uranium for the formation of a solid protective oxide below 2800 K is estimated for a temperature, T (in Kelvin), as follows:(unoxidized iron)/(unoxidized zirconium + turanium) = (1/28){5.7/exp[-(147 061 + 12.08T log(T) - 61.03T - 0.000555T2/1.986T)]}1/2.As long as this limit is not exceeded, either zirconium or uranium metal oxidation rates and heating describe the corium oxidation rate. If this limit is exceeded, diffusion of steam to the corium surface will limit the oxidation rate, and linear time-dependent growth of a nonprotective, mostly FeO, layer will occur below the protective (Zr,U) O2 scale. When this happens, the oxidation should be at the constant rate given by the thickness of the protective layer. Heat generation should be similar to that of iron oxidation.