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Thermal Hydraulics
The division provides a forum for focused technical dialogue on thermal hydraulic technology in the nuclear industry. Specifically, this will include heat transfer and fluid mechanics involved in the utilization of nuclear energy. It is intended to attract the highest quality of theoretical and experimental work to ANS, including research on basic phenomena and application to nuclear system design.
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ANS Student Conference 2025
April 3–5, 2025
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The Standards Committee is responsible for the development and maintenance of voluntary consensus standards that address the design, analysis, and operation of components, systems, and facilities related to the application of nuclear science and technology. Find out What’s New, check out the Standards Store, or Get Involved today!
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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.
Andrea Bucalossi, Alessandro Petruzzi, Marian Kristof, Francesco D'Auria
Nuclear Technology | Volume 172 | Number 1 | October 2010 | Pages 29-47
Technical Paper | Reactor Safety | doi.org/10.13182/NT172-29
Articles are hosted by Taylor and Francis Online.
Computational reactor safety analysis is trending to replace conservative evaluation model calculations with best-estimate analysis complemented by uncertainty evaluation of the code results. In such cases, the evaluation of the margin to acceptance criteria (e.g., the maximum fuel rod clad temperature) is based on the upper limit of the calculated uncertainty range. Uncertainty analysis is compulsory if relevant conclusions are to be obtained from best-estimate thermal-hydraulic code calculations in order to avoid presenting single values of unknown accuracy for comparison with regulatory acceptance limits.This paper, after a thorough introduction of conservative and best-estimate methods and characterization of the main sources of uncertainties affecting best-estimate system codes, applies a best-estimate-plus-uncertainty (BEPU) method to three cases having as reference different nuclear power plants and different types of transients. Finally, the results from the BEPU approach is compared with a conservative approach and a combined approach.
