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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.
Hirokazu Ohta, Takanari Ogata, Takeshi Yokoo, Michel Ougier, Jean-Paul Glatz, Bruno Fontaine, Laurent Breton
Nuclear Technology | Volume 165 | Number 1 | January 2009 | Pages 96-110
Technical Paper | Materials for Nuclear Systems | doi.org/10.13182/NT09-A4063
Articles are hosted by Taylor and Francis Online.
Fast reactor metal fuels containing minor actinides (MAs) Np, Am, and Cm and/or rare earths (REs) have been irradiated in the fast reactor PHÉNIX to examine the effects of adding those elements on metal fuel irradiation behavior. In this experiment, two MA-containing metal fuel pins, in which the test alloys U-19Pu-10Zr-2MA-2RE and U-19Pu-10Zr-5MA/U-19Pu-10Zr-5MA-5RE (wt%) were loaded into part of a standard U-19Pu-10Zr alloy fuel stack, and a reference fuel pin of U-19Pu-10Zr alloy without MAs or REs was set in an irradiation capsule. Two other capsules with this same configuration are also irradiated. Postirradiation examinations are conducted at ~2.5, ~7, and ~11 at.% burnup. For the low-burnup fuel pins, nondestructive tests after irradiation have been performed, and the integrity of the pins was confirmed. The irradiation behavior of MA-containing metal fuels up to 2.5 at.% burnup was analyzed using the ALFUS code. The calculation results, such as the axial swelling distribution of a fuel slug or the extrusion behavior of bond sodium to the gas plenum, are consistent with the measurement data regardless of the addition of MAs and REs to the U-Pu-Zr alloy fuels. This observation result indicates that the macroscopic irradiation behavior of U-Pu-Zr fuels containing MAs and REs of 5 wt% or less is similar to that of U-Pu-Zr fuels up to ~2.5 at.% burnup.
