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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.
Joonhong Ahn, Atsuyuki Suzuki
Nuclear Technology | Volume 101 | Number 1 | January 1993 | Pages 79-91
Technical Paper | Waste Management Special / Radioactive Waste Disposal | doi.org/10.13182/NT93-A34769
Articles are hosted by Taylor and Francis Online.
A mathematical analysis of the diffusion of the 241Am → 237Np decay chain in the artificial barrier of a high-level radioactive waste repository is presented. First, analytical solutions obtained are for the space- and time-dependent concentration of 241 Am in the artificial barrier and the time-dependent amount of americium precipitated at the surface of the waste glass, based on the assumption of the congruency of the radionuclides with solubility-limited dissolution of the glass matrix. The effects of solubility sharing with coexisting 243Am are considered. Transport and precipitation of 237Np in the artificial barrier are analyzed by dividing the time domain into a small time domain, where the 241Am concentration is so large that 237Np precipitation is dominant, and a large time domain, where the 241Am becomes negligible and the precipitation region shrinks by diffusion from the precipitation front. The equation for the movement of the precipitation front is obtained. As the overpack lifetime increases, the effect of neptunium precipitation becomes less significant. With a lifetime longer than ∼6000 yr, an earlier model, where neptunium is treated as a mother nuclide and the precipitation occurs only at the glass surface, can be used. With the solubility for Np(OH)4, the effect of neptunium precipitation is as small as a factor of 2 in terms of the maximum mass release rate at the outer boundary of the artificial barrier, and the earlier model can be used for safety assessment. With the solubility for NpO2, the current model gives a maximum mass release rate at the outer boundary that is one order of magnitude greater than the previous one.
