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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.
Sung Joong Kim, Lin-Wen Hu, Floyd Dunn
Nuclear Technology | Volume 182 | Number 3 | June 2013 | Pages 315-334
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT12-81
Articles are hosted by Taylor and Francis Online.
The Massachusetts Institute of Technology Research Reactor (MITR) is evaluating a transitional core conversion strategy for converting from high-enrichment uranium (HEU) to low-enrichment uranium (LEU) fuel. The objective of this study is to analyze steady-state operational safety margins and loss of primary flow (LOF) accidents for the postulated HEU-LEU transitional core configurations. The thermal-hydraulic calculation was performed using the RELAP5 MOD 3.3 code based on 7.40-MW reactor power, which is the limiting safety system settings of the current licensed reactor power of 6 MW. A lumped average and a single hot channel were modeled in each core configuration with radial peaking factors of 2.0 and 1.76 for HEU and LEU fuel elements, respectively. Four natural convection valves and two antisiphon valves were modeled for natural convective heat removal during the LOF transient. Two different hot-channel configurations and full- and side-channel geometries were evaluated because the unique design of the MITR fuel element can form these two types of geometries. RELAP5 calculation results suggest that the transitional core conversion strategy is feasible and that sufficient thermal-hydraulic safety margins can be maintained.
