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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. Morita, C. A. Olson, Y. X. Sung, J. F. Connelley, Jr., E. H. Novendstern, S. Kapil, P. W. Rosenthal
Nuclear Technology | Volume 112 | Number 3 | December 1995 | Pages 401-411
Technical Paper | Heat Transfer and Fluid Flow | doi.org/10.13182/NT95-A35166
Articles are hosted by Taylor and Francis Online.
The AP600 reactor core approaches buoyancy-dominated flow at the departure from nucleate boiling (DNB)-limiting period of a postulated steam-line-break accident. The reactor core has a highly skewed power distribution at this time due to the conservative assumption of a withdrawn rod cluster control assembly (stuck rod). Under such conditions, strong buoyancy-induced core cross flow occurs, and coupled nuclear and thermal-hydraulic interactions become important. To analyze the transient, Westinghouse Electric Corporation has coupled THINC-IV with a neutronic code (ANC). Applicability of the THINC-IV subchannel code to the low-flow conditions with a steep radial power gradient is verified with existing rod bundle test results. The code predictions are in excellent agreement with the test data. The coupled codes provide a realistic three-dimensional simulation of core power by considering core flow distributions and the resultant enthalpy distributions in neutronic feedback. The safety analysis using the coupled code demonstrates that the DNB design basis is met during the postulated steam-line-break accident.
