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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.
Jae Jun Jeong, Dae Hyun Hwang, Bub Dong Chung
Nuclear Technology | Volume 156 | Number 3 | December 2006 | Pages 360-368
Technical Note | Thermal Hydraulics | doi.org/10.13182/NT06-A3797
Articles are hosted by Taylor and Francis Online.
MARS is a best-estimate system analysis code that is based on the RELAP5/MOD3 and COBRA-TF codes. The COBRA-TF code was adapted as a three-dimensional thermal-hydraulic module in MARS. It uses a two-fluid, three-field model for two-phase flows and has a subchannel flow mixing model. The subchannel flow mixing model of the MARS three-dimensional module was assessed by using the ISPRA 16-rod bundle test and the GE 9-rod bundle test data. These tests represent typical pressurized water reactor and boiling water reactor core thermal-hydraulic conditions, respectively. Two interconnected subchannel tests that were performed under atmospheric pressure conditions were also used for the assessment. From the results of the assessments, a simple modification of the subchannel flow mixing model was suggested to take into account the effects of the system pressure on the void drift phenomena.
