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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.
Y. Nobuta et al.
Fusion Science and Technology | Volume 60 | Number 4 | November 2011 | Pages 1535-1538
Interaction with Materials | Proceedings of the Ninth International Conference on Tritium Science and Technology (Part 2) | doi.org/10.13182/FST11-A12725
Articles are hosted by Taylor and Francis Online.
Tritium retention in plasma facing materials is a primary issue for ITER and next step fusion devices, since it greatly affects its safety and operational schedule. In the ITER, carbon and tungsten are used as divertor materials. In the present study, co-deposited carbon film, tungsten and isotropic graphite were exposed to tritium gas, and then the amount of absorbed tritium was investigated. During the tritium exposure, the partial pressure of tritium gas was kept at 10 Pa. The sample temperature was kept a constant in the range from RT to 573 K. The amounts of absorbed tritium were evaluated by -ray-induced X-ray spectrometry (BIXS). The amounts of absorbed tritium in co-deposited carbon films were one or two orders of magnitude larger than that of polycrystalline tungsten and isotropic graphite. The amount of absorbed tritium for co-deposited carbon film with a high volume density (1.53 g/cm3) was several times larger than that of the film with a low volume density (1.13 g/cm3). The amount of absorbed tritium increased with the temperature. These results indicate that co-deposited carbon films can absorb much larger amount of tritium than tungsten and graphite, and carbon film density affects the amount of absorbed tritium.
