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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.
N. Miya, M. Nemoto, N. Toyoshima
Fusion Science and Technology | Volume 26 | Number 3 | November 1994 | Pages 507-511
Fusion Material and Plasma-Facing Component | Proceedings of the Eleventh Topical Meeting on the Technology of Fusion Energy New Orleans, Louisiana June 19-23, 1994 | doi.org/10.13182/FST94-A40207
Articles are hosted by Taylor and Francis Online.
Tritium concentration measurements have been made of the tritium recovery from the JT-60U after high power neutral beam injected discharges. This work has been done to study tritium release characteristics and to assess the tritium inventory in the vacuum vessel. Cleanup operations by hydrogen divertor discharges and helium glow discharge cleanings were performed to deplete the tritium inside the wall prior to in-vessel maintenances. After the beginning of the divertor discharges the tritium release from walls increases gradually. The release rate depends on the surface temperature of divertor plates and it is enhanced by high-power neutral beam heated divertor discharges combined with high-temperature vessel baking. Helium glow discharges, furthermore, give strong release characteristics compared with divertor discharges in spite of their short duration within ∼1 hr. After the deuterium discharges and cleanup operations from July 1991 to July 1994, it is estimated that 70∼ 80% of the total tritium of 39 GBq generated in the plasma is retained inside the graphite first wall.
