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Division Spotlight
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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2024 ANS Annual Conference
June 16–19, 2024
Las Vegas, NV|Mandalay Bay Resort and Casino
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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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Remembering Joseph M. Hendrie
Joseph M. Hendrie
To those of us who knew Joe, even prior to his appointment as chair of the Nuclear Regulatory Commission, it is an understatement to say that he was a larger-than-life member of the nuclear science and technology enterprise. He was best known to the broader community for two major accomplishments: the design and construction of the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory and the creation of the standard review plan (SRP) for the U.S. Atomic Energy Commission.
In addition to the products of these endeavors becoming major fundaments to their respective communities, they were uniquely Joe. The safety analysis report for the HFBR was written essentially single-handedly by him. This was true of the SRP as well, which became the key safety review document for the NRC as it performed safety reviews for the growing number of power reactor applications in the United States. His deep technical knowledge of nuclear engineering and his extraordinary management skills made this possible.
A. Epiney, S. Canepa, O. Zerkak, H. Ferroukhi
Nuclear Technology | Volume 196 | Number 2 | November 2016 | Pages 223-237
Technical Paper | doi.org/10.13182/NT16-47
Articles are hosted by Taylor and Francis Online.
The STARS project at the Paul Scherrer Institut (PSI) has adopted the TRACE thermal-hydraulic code. For analyses involving interactions between system and core, a coupling of TRACE with the SIMULATE-3K (S3K) light water reactor (LWR) core simulator has been developed. In this configuration, the codes and associated simulation models play a central role to achieve a comprehensive safety analysis capability. Therefore, efforts have now been undertaken to consolidate the validation strategy by implementing a more rigorous and structured assessment approach for TRACE applications. The principle is to systematically track the evolution of a given set of predicted physical quantities of interest (QoIs) over a multidimensional parametric space. If properly set up, such environment should provide code developers and code users with persistent (less affected by user effect) and quantified information (sensitivity of QoIs) on the applicability of a simulation scheme (codes, methodology, and input models) for steady-state and transient analysis of full LWR systems. Through this, for each given transient/accident, critical paths of the validation process can be identified that could then translate into defining reference schemes to be applied for downstream predictive simulations. To illustrate this approach, this validation strategy is applied to an inadvertent blowdown event that occurred in a Swiss BWR/6. The transient was initiated by the spurious actuation of the automatic depressurization system. Here, the validation approach progresses through a number of dimensions: (a) different versions of the TRACE code; (b) the methodology dimension—in this case imposed power and updated TRACE core models are investigated; and (c) the nodalization dimension, where changes to the input model are assessed. For each step in each validation dimension, a common set of QoIs is investigated. For the steady-state results, these include fuel temperature distributions. For the transient part of the present study, the evaluated QoIs include the system pressure evolution and water carryover into the steam line. It has been seen that the improvements to the model predictions resulted in a small impact on the system pressure gradient, thus confirming a persistency of the downstream mechanical stress estimate, whereas the water carryover could vary by up to 150% as a function of the adopted simulation methodology.