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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.
E. A. Mogahed, L. El-Guebaly, A. Abdou, P. Wilson, D. Henderson, ARIES Team
Fusion Science and Technology | Volume 39 | Number 2 | March 2001 | Pages 462-466
Advanced Designs | doi.org/10.13182/FST01-A11963279
Articles are hosted by Taylor and Francis Online.
Loss of coolant accident (LOCA) and loss of flow accident (LOFA) analysis is performed for ARIES-AT, an advanced fusion power plant design (1000 MWe). ARIES-AT employs a high performance, high temperature blanket system. It uses the high temperature SiC/SiC for structural material and LiPb for coolant-breeder. Due to the large difference between the time scale of plasma shutdown and the coolant or power loss, it is assumed that the plasma is immediately quenched at the onset of the LOCA/LOFA and the chamber components' temperature begins to rise due to the decay heat generated. A 2-D transient finite element model is established to examine the thermal behavior of the in-vessel components to determine the maximum temperature reached, the time, and duration of the peak. The model is axisymmetric in (r-z) around the reactor axis to show the details of temperature distribution in the vertical direction. The vacuum vessel is assumed adiabatic in the inboard side and radiates to the maintenance port located on the outboard side. The maximum temperature of steel in the reactor is about (600 °C - 700°C) after about 4 days from the onset of the accident. The highest temperature in the reactor is in the divertor region and it reaches ≈1050°C after about 2-3 hours. The analysis indicates that the reactor does not need any special scheme for decay heat removal.
