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ANS Student Conference 2025
April 3–5, 2025
Albuquerque, NM|The University of New Mexico
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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.
Wen-Shan Lin, Chien-Hsiung Lee, Bau-Shei Pei
Nuclear Technology | Volume 88 | Number 3 | December 1989 | Pages 294-306
Technical Paper | Heat Transfer and Fluid Flow | doi.org/10.13182/NT89-A34312
Articles are hosted by Taylor and Francis Online.
Based on the Helmholtz instability at the microlayer/vapor interface as a trigger condition for microlayer dryout, Lee and Mudawwar developed a mechanistic critical heat flux (CHF) model for subcooled flow boiling. An improved CHF model is implemented with more solid theoretical bases for subcooled and low-quality flow boiling under pressurized water reactor conditions. Comparisons between the predictions and experimental data show that the present model is more accurate than the well-known theoretical CHF model of Weisman and Pei and the empirical CHF correlations of W-3, Bowring, and Katto and Ohno for water flowing through uniformly heated round tubes within the applicable ranges. The applicability of the present model to rod bundles is also under investigation. Highly satisfactory results are obtained from the comparisons of predicted to observed bundle critical power.