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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.
H. Bonneville, L. Carenini, M. Barrachin
Nuclear Technology | Volume 196 | Number 3 | December 2016 | Pages 489-498
Technical Paper | doi.org/10.13182/NT16-27
Articles are hosted by Taylor and Francis Online.
The Accident Source Term Evaluation Code (ASTEC) is used to perform numerical simulations of the accidents at the Fukushima Daiichi nuclear power station in the frame of the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF) project. At present, simulations are available for Units 1, 2, and 3 of Fukushima Daiichi and for 6 days from the earthquake. A clear lesson from phase 1 of the project was that the uncertainties in the functioning of the safety systems and in accident progression are still large and there are many explanations for the measured thermohydraulic behavior. Rather than focusing on the thermohydraulic key parameters for which comparisons with measurements are available, this paper will address melt composition computation results that may provide insights relevant for the decommissioning process.
When molten corium relocates from the core down to the vessel lower head, the melt jets interact with water and may be totally or partially fragmented depending on the level of water. A U-Zr-O-Fe molten pool may form in the lower head, and because of chemical reactions, separation between nonmiscible metallic and oxide phases may occur. The models implemented in ASTEC enable the simulation of these phenomena. Up to five different axisymmetric corium layers in the vessel bottom head can be formed, which are, from bottom to top, a debris layer, a heavy metallic layer, an oxide layer, a light metallic layer, and another debris layer. An important process is the UO2 fuel reduction to metallic uranium by nonoxidized zirconium, which results in uranium transport to the dense metallic layer as demonstrated in the MAterial SCAling (MASCA) program.
Because of the large consensus on the accident progression of Fukushima Daiichi Unit 1, in this paper we present complex melt compositions before vessel failure for the current best-estimate cases for Unit 1. We do not present similar work performed for Units 2 and 3.
It should be underlined that in the case of vessel bottom failure, a part of this complex melt will relocate to the pedestal and molten core–concrete interaction will take place enhancing other complex physical phenomena with possible large consequences on the melt chemical composition and behavior.
