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Latest News
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. Venkataramani, F. Ghezzi, G. Bonizzoni, W. T. Shmayda
Fusion Science and Technology | Volume 29 | Number 1 | January 1996 | Pages 91-104
Technical Paper | Tritium System | doi.org/10.13182/FST96-A30659
Articles are hosted by Taylor and Francis Online.
A follow-up is done to earlier work on the conversion of isotopic waters to hydrogen isotopes, and it involves the reaction behavior of water vapor with Zr(V0.5Fe0.5)2 getter alloy under water vapor flow conditions. The efficiency of the alloy, for the conversion of H2O and D2O to H2 and D2, respectively, has been measured at different reactor pressures in the range of 10 to 330 Pa for different alloy temperatures in the range of 150 to 400°C and with hydrogen and oxygen concentrations in the alloy ≤ 250 mmol/mol of alloy. The conversion efficiency was measured to be in the range of 25 to 35% at reactor pressures of ≈250 Pa for water vapor flow rates of ≈0.3 µmol/g of alloy per second, while it was found to be in the range of 70 to 80% at reactor pressures ≤20 Pa with flow rates of ≤0.02 µmol/g of alloy per second. These experiments demonstrate the feasibility of tritiated water vapor conversion to tritium using metallic getter alloys under quasi-steady-state conditions; this feasibility is very relevant to the fusion reactor fuel cycle.
