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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.
I. N. Sviatoslavsky, E. A. Mogahed, Y-K. M. Peng, B. E. Nelson, P. J. Fogarty, E. T. Cheng, R. J. Cerbone
Fusion Science and Technology | Volume 30 | Number 3 | December 1996 | Pages 1649-1653
Nonelectric Applications of Fusion | doi.org/10.13182/FST96-A11963187
Articles are hosted by Taylor and Francis Online.
Engineering design issues of a volumetric neutron source (VNS) based on a steady state low aspect ratio DT tokamak are presented. At the present the major radius is 0.8 m, the minor radius 0.6 m for an aspect ratio of 1.33, the plasma current is 10.1 MA, the toroidal field at the major radius is 1.8 T, the fusion power is 39 MW giving an average neutron wall loading of 1.0 MW/m2 on the outboard side with an available testing area of 10 m2. Two neutral beams delivering more than 20 MW are used to drive the steady state fusion plasma. A single turn unshielded water cooled dispersion strengthened (DS) Cu centerpost is used in conjunction with a conducting Cu bell jar which acts as a vacuum boundary and the return legs for the toroidal field (TF) coils. The centerpost is 9 m long, carries 7.2 MA and is specially shaped to minimize ohmic heating, which is calculated using temperature dependent DS Cu properties and increases in resistivity due to nuclear transmutations are accounted for. A naturally diverted plasma scrapeoff layer dominated by pressure-driven instabilities is assumed giving a peak heat flux of 5.2 MW/m2 on the diverter plates. Fabrication approaches for the centerpost and its replacement time lines have been estimated to be feasible and reasonable.
