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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, Shinichi Nakayama
Nuclear Technology | Volume 97 | Number 3 | March 1992 | Pages 323-335
Technical Paper | Radioactive Waste Management | doi.org/10.13182/NT92-A34640
Articles are hosted by Taylor and Francis Online.
Numerical results are presented for an analysis of diffusion of237Np, a redox-sensitive radionuclide, in engineered barriers consisting of overpack and bentonite-filled buffer regions, with oxidation-reduction kinetics of neptunium with iron and dissolved oxygen. Steady-state distributions of Fe(II) and dissolved oxygen are first obtained by considering the oxidation reaction of Fe(II) with oxygen. Based on these profiles, the neptunium diffusion models for pH 6.5 and 9.0 are established. Analytical or finite element solutions are obtained for the corresponding mathematical problems. At pH 9.0, even if dissolved oxygen intrudes on the buffer region from the rock/buffer interface, the penetration of oxygen into the overpack region and to the surface of the waste solid can be avoided. Thus, less soluble, strongly sorbing Np(IV) is released from the waste solid. At pH 6.5, dissolved oxygen can reach the waste surface. Weakly sorbing, soluble Np(V) is released from the waste solid. Although the released Np(V) is reduced to Np(IV) by Fe(II) in the overpack region, the Np(IV) is quickly oxidized by Fe(III) and dissolved oxygen at the overpack-buffer interface. Neglecting the existence of dissolved oxygen and assuming that the repository is kept under a reducing environment so that only Np(IV) migrates might lead to quite an optimistic estimate of the neptunium release rate from the engineered barriers.
