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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.
H. Takenaga, H. Kawashima, S. Nishio, K. Tobita
Fusion Science and Technology | Volume 57 | Number 1 | January 2010 | Pages 94-102
Technical Paper | doi.org/10.13182/FST10-A9270
Articles are hosted by Taylor and Francis Online.
A fueling scenario in a fusion reactor has been investigated, where tritium is fueled in the main plasma and deuterium is fueled in both the main plasma and the edge plasma. The tritium fueling in the main plasma minimizes the tritium fueling rate necessary for sustaining the high tritium density in the main plasma, resulting in the minimum tritium recycling level at the fixed pumping fraction. The deuterium fueling in the main plasma sustains the high deuterium density in the main plasma, and the deuterium fueling in the edge plasma enhances the deuterium recycling level for reducing the divertor temperature. Based on this scenario, particle balance was quantitatively investigated using the SlimCS design parameters at 2.95-GW fusion output with consideration of confinement times separately estimated for the particles fueled in the main plasma and the edge plasma. The fueling rates in the main plasma were evaluated to be 2.5 × 1022/s for tritium and 1.4 × 1022/s for deuterium when the confinement times for the particles fueled in the main and edge plasmas were assumed to be 2 s and 2 ms, respectively, and the divertor pumping fraction was assumed to be 3% of the particle flux to the divertor plates. For enhancement of the recycling level, the additional deuterium fueling in the edge plasma of 3.6 × 1023/s was required in this case. In order to satisfy the tritium balance, it was necessary to suppress the tritium retention rate to <0.01% of the tritium recycling rate and the tritium loss in the tritium cycle system to below 0.2% of the tritium fueling rate with the tritium breeding ratio of 1.05.