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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.
T. M. Conboy, T. J. McKrell, M. S. Kazimi
Nuclear Technology | Volume 182 | Number 3 | June 2013 | Pages 259-273
Technical Paper | Fuel Cycle and Management/Thermal Hydraulics | doi.org/10.13182/NT12-58
Articles are hosted by Taylor and Francis Online.
In order to significantly increase the power density of light water reactors (LWRs), the authors propose the helical-cruciform (HC) fuel rod assembly as an alternative to traditional fuel geometry. The HC fuel rod assembly is a self-supporting nuclear fuel configuration consisting of four-petalled, axially twisted fuel rods closely packed against one another in a square array. Within the LWR core, HC fuel would possess several advantages over traditional fuel, potentially allowing for operation at a higher power density. Chief among these advantages are a larger surface-to-volume ratio, improved radial mixing characteristics of the coolant, and a shorter radial heat conduction path in the fuel pellet. In adapting helical rod geometry to the LWR core, the authors identified a shortage of correlations for fluid flow in twisted geometry flow channels, causing uncertainty in modeling studies. This gap was addressed by constructing an experimental facility for the measurement of hydraulic resistance and assembly mixing within a mock bundle of HC fuel rods. The rods were manufactured and tested in 4 × 4 square arrays at twist pitches of 200, 100, and 50 cm. Hydraulic resistance was evaluated by measuring frictional pressure drop over a 1-m length of the assembly. Results showed a higher pressure drop for the HC rods in comparison to bare cylindrical rods with no spacers, at a given mass flux, but no apparent dependency on twist pitch. However, data indicated that the HC-rod effective hydraulic diameter was only 90% of the expected value given its wetted perimeter and flow area, suggesting a shift from the traditional definition of Dh for this unique shape. Mixing tests used the technique of a hot water tracer injection into the central subchannel of the assembly of room-temperature water. Downstream temperature measurements were used to judge the rate of lateral cross flow within the HC rod bundle. Over 300 tests were analyzed, yielding a best-fit correlation for use with any twist pitch, rod length, or coolant mass flux. Compared to a traditional rod bundle, this correlation implies an enhancement in the intensity of turbulent interchange of 40% brought about by the HC geometry and a 1.6% forced diversion of axial flow per subchannel, per quarter-turn along the rod length. The correlations for hydraulics and cross-flow mixing presented here should reduce the uncertainty in future analysis of this fuel type for high-power-density LWRs.
