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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.
J. N. Brooks, D. M. Gruen, A. R. Krauss, R. F. Mattas, A. B. DeWald
Fusion Science and Technology | Volume 8 | Number 1 | July 1985 | Pages 1275-1280
Impurity Control and Vacuum Technology | Proceedings of the Sixth Topical Meeting on the Technology of Fusion Energy (San Francisco, California, March 3-7, 1985) | doi.org/10.13182/FST85-A39943
Articles are hosted by Taylor and Francis Online.
A new approach to impurity control involves the development of materials displaying both strong surface segregation of a low-Z component and high secondary ion fractions in the sputtering of that component. Key issues that have been studied with particular reference to copper-lithium alloys relate to the completeness of the overlayer, its rate of formation in a reactor environment, lowering of substrate sputtering and self-sputtering yields, durability of the overlayer, and depletion of the bulk alloy in the low-Z component. Other factors that must be considered in the materials selection process relate to response to disruptions, heat transfer, thermal stress, fabricability, radiation damage, activation, and tritium permeation. Copper-lithium alloys have been evaluated as a surface material for the impurity control system of the INTOR reactor. Both the medium-edge temperature limiter regime and the low-edge temperature divertor regime were examined. The analysis used TRIM code data to predict sputtering coefficients for copper-lithium with a 1.5 monolayer coverage of lithium. The REDEP code was used to evaluate the erosion performance for INTOR. Other properties such as fabrication and thermal performance were also briefly assessed. It was found from the standpoint of erosion that copper-lithium is a very good candidate material for the medium-edge temperature regime and also works well in the low-edge temperature regime. For the medium-edge temperature regime, the use of copper-lithium results in an almost negligible erosion rate over the entire surface.