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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.
Makoto Oyaidzu, Masayuki Ohta, Kentaro Ochiai, Atsushi Kasugai
Fusion Science and Technology | Volume 77 | Number 7 | November 2021 | Pages 842-847
Technical Paper | doi.org/10.1080/15361055.2021.1962119
Articles are hosted by Taylor and Francis Online.
In the Advanced Fusion Neutron Source (A-FNS), an accelerator-driven fusion-relevant neutron source that is planned for development in Japan, a few grams (3.5 g at full power operation) of tritium will be generated every year, mainly in the lithium target system. Since the generated tritium would migrate out of the lithium target system, it is necessary to estimate the tritium migration into and out of the lithium target system for the design of detritiation systems for the A-FNS. Therefore, a preliminary estimation is performed in the present study. As a result, it is found that almost all of the generated tritium in the lithium target system would be trapped in the impurity removal system, while less than 0.5% would migrate out. It is also indicated that the amount of tritium that would migrate out of the lithium target system would be able to be processed with the existing techniques so far.
