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Nuclear Criticality Safety
NCSD provides communication among nuclear criticality safety professionals through the development of standards, the evolution of training methods and materials, the presentation of technical data and procedures, and the creation of specialty publications. In these ways, the division furthers the exchange of technical information on nuclear criticality safety with the ultimate goal of promoting the safe handling of fissionable materials outside reactors.
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International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2025)
April 27–30, 2025
Denver, CO|The Westin Denver Downtown
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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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Argonne’s METL gears up to test more sodium fast reactor components
Argonne National Laboratory has successfully swapped out an aging cold trap in the sodium test loop called METL (Mechanisms Engineering Test Loop), the Department of Energy announced April 23. The upgrade is the first of its kind in the United States in more than 30 years, according to the DOE, and will help test components and operations for the sodium-cooled fast reactors being developed now.
B. Beeny, R. Vaghetto, K. Vierow, Y. A. Hassan
Nuclear Technology | Volume 196 | Number 2 | November 2016 | Pages 292-302
Technical Paper | doi.org/10.13182/NT16-36
Articles are hosted by Taylor and Francis Online.
The thermal-hydraulic response of large dry pressurized water reactor containments under loss-of-coolant-accident conditions—particularly with respect to containment pressure and sump pool temperature—is crucial for risk-informed decision making about Generic Safety Issue 191. Texas A&M University has developed models with several computer codes including MELCOR and GOTHIC to model such scenarios.
MELCOR is a best-estimate thermal-hydraulic and severe accident code created and actively maintained by Sandia National Laboratories for the U.S. Nuclear Regulatory Commission. GOTHIC is a thermal-hydraulic software package meant for design, licensing, and safety calculations for, among other systems, nuclear power plant containments. It was developed and is maintained by Numerical Applications Inc. for the Electronic Power Research Institute.
The overarching goal of the analyses presented here is twofold: (1) produce best-estimate time profiles of sump pool temperature under double-ended guillotine-break conditions with MELCOR and GOTHIC and (2) investigate differences between the MELCOR and GOTHIC code results via a sensitivity study. The sump pool temperature was selected as a key parameter to compare because it has direct implications for sump pool chemistry, residual heat removal during recirculation, and pressure drop across sump screens.
Aspects of the MELCOR and GOTHIC modeling strategies are discussed, and best estimates of the containment thermal-hydraulic response are presented. There are significant disagreements between code predictions. Hypotheses to explain the differences are tested through a comparative code sensitivity study. In this context, “sensitivity” refers to how containment thermal hydraulics respond to differences in code inputs or code phenomenological models. Sensitivity calculations are performed to exclude, individually, the model effects on comparative thermal-hydraulic responses of containment fan coolers, containment sprays, thermal surface condensation/films, and break source definition. Calculations are also performed with multiple models excluded. Using containment sump pool temperature as an indicator, the most impactful physics in terms of code agreement are those of thermal surfaces (condensation, film phenomena) whereas fan cooler models have a minimal effect. Containment spray exclusion results in disagreement in parts of the event sequence, while break source definition and/or break effluent flashing models lead to disagreement.