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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.
P. Gierszewski1, G. Williams2, J. Blevins1, H. Brunnader1, P. Cumyn3, B. Dean4, J. Galambos5, C. Holloway6, R. Kelly7, A. Natalizio1, S. Smith4
Fusion Science and Technology | Volume 26 | Number 3 | November 1994 | Pages 1146-1150
Fusion Power Reactor, Economic, and Alternate Concept | Proceedings of the Eleventh Topical Meeting on the Technology of Fusion Energy New Orleans, Louisiana June 19-23, 1994 | doi.org/10.13182/FST94-A40308
Articles are hosted by Taylor and Francis Online.
The CFFTP Pilot plant concept is a driven, steady-state H-mode tokamak with ion cyclotron current drive. The fuel cycle uses low-tritium-inventory technologies, including compact toroid fuelling. The mechanical design is based on helium cooling, radial blanket maintenance, ceramic pebble breeder blanket, and demountable copper magnets. It is expected to operate for 1 full-power-year with 0.25 MW/m2 average (0.4 MW/m2 peak) neutron wall load. The machine would produce 20 MW of fusion power with 40 MW of auxiliary power. The 2.7 MA plasma current is ramped up inductively, and then sustained by the bootstrap current (32%) and fast wave current drive (68%). The plasma would be roughly the size of the TFTR plasma, but elongated with double-null divertors and an aspect ratio of 5. The total electric power consumption would be around 450 MWe. The tritium supply requirements, given partial breeding, would be only 0.8 kg. The on-site tritium inventory would be about 0.3 kg. The direct construction cost is estimated at 1.1 B$Cdn, with a total project cost of 2.3 B$Cdn (1992). CFFTP Pilot would provide steady-state integrated nuclear testing at a fluence and neutron wall load of about one-quarter ITER CDA, for a cost of about one-third ITER CDA. The blanket test area would be similar to the ITER CDA blanket test port area.
