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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.
Gabriel F. Cuevas Vivas, Yassin A. Hassan
Nuclear Technology | Volume 127 | Number 3 | September 1999 | Pages 287-300
Technical Paper | Reactor Safety | doi.org/10.13182/NT99-A3002
Articles are hosted by Taylor and Francis Online.
An analysis of the simplified boiling water reactor (SBWR) is carried out using the reactor analysis computer program RAMONA-4B in an operational transient scenario, a turbine trip with failure of all the bypass valves. The SBWR model represents the vessel's internal components, such as flow areas, diameters, and volumes. The one-quarter-core neutron parameters are calculated with the CASMO-3 transport theory lattice physics computer program. The three-dimensional representation of the reactor core uses some standard fuel design parameters, such as a wide central water rod, 8 x 8 lattice, gadolinium rods, etc. The thermal-hydraulic equations are solved with the RAMONA-4B computer program in a closed loop inside the reactor vessel and in 184 parallel channels (including bypass) in the core.Finally, the two-phase coolant and neutronic parameters are calculated in steady state and during the turbine trip transient. The results obtained compare favorably with the standard safety analysis report data.
