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Division Spotlight
Mathematics & Computation
Division members promote the advancement of mathematical and computational methods for solving problems arising in all disciplines encompassed by the Society. They place particular emphasis on numerical techniques for efficient computer applications to aid in the dissemination, integration, and proper use of computer codes, including preparation of computational benchmark and development of standards for computing practices, and to encourage the development on new computer codes and broaden their use.
Meeting Spotlight
Conference on Nuclear Training and Education: A Biennial International Forum (CONTE 2023)
February 6–9, 2023
Amelia Island, FL|Omni Amelia Island Resort
Standards Program
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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Bringing 2022 ANS Standards Committee successes into the new year
By all accounts, 2022 brought many successes for the American Nuclear Society’s Standards Committee, including the initiation of five projects, reaffirmation of 11 current standards, and publication of seven new or revised standards. The entire collection of ANS current standards has been approved or reaffirmed (reapproved without change) by the American National Standards Institute (ANSI) within the past five years, keeping ANS in 100 percent compliance with ANSI’s requirement on maintaining current American National Standards. Also, the ANS standards program was reaccredited by ANSI on August 19, 2022, with the approval of a revised set of rules and procedures. ANS’s new rules and procedures take advantage of the opportunity to develop standards-related technical reports that may be registered with ANSI.
Simon A. Clément, Philippe M. Bardet
Nuclear Technology | Volume 199 | Number 2 | August 2017 | Pages 151-173
Technical Paper | doi.org/10.1080/00295450.2017.1327254
Articles are hosted by Taylor and Francis Online.
Because of the complexity of the flow within light water reactor (LWR) cores, numerous small-scale phenomena locally influence heat transfer and critical heat flux (CHF). They include development of viscous and thermal boundary layers, interchannel mixing, spacer grid mixing, rod vibrations, or confinement effects such as the proximity of the walls or the influence of the gap between adjacent fuel bundles. LWR prototypical conditions are particularly harsh environments and limit measurements to quantities such as pointwise pressure drop and temperature, the latter resulting in global heat transfer and CHF correlations. The local phenomena mentioned above are embedded in these correlations, leading to inherent empiricism (and therefore conservatism). Validated computational fluid dynamics (CFD) codes and models can predict these phenomena, thus providing modelization tools of greater accuracy. However, major requirements for validation campaigns include the matching of validation and application domains and the deployment of mature and high-resolution diagnostics. For the latter, many are available for single-phase flows due to their predominance in several research fields. Furthermore, in the lower part of LWR cores, flow is single phase, and only this regime is considered in this paper. To circumvent the challenges of deploying diagnostics in LWR conditions, surrogate fluids are commonly used, enabling the measurement of velocity, temperature, pressure, or wall shear stress. A large number of single-phase tests with resolution adequate to validate CFD models have been conducted with air, steam, and water at moderate temperature and pressure. However, to date, with these fluids, the application domain defined by the Reynolds and Prandtl numbers has not been reached.
Four surrogate gases are proposed to match application and validation domains while allowing the deployment of a broad range of diagnostics: pressurized sulfur hexafluoride, xenon, cryogenic nitrogen, and highly pressurized air. By controlling their operating temperature and pressure, they allow matching prototypical Reynolds and Prandtl numbers while preserving the length scale, velocity scale, and timescale. This is achieved by reproducing the kinematic viscosity and thermal diffusivity of several nuclear reactor coolants. Furthermore, for single-phase conjugate heat transfer, a complete scaling analysis is performed for one pressurized water reactor fuel rod within a bundle under normal operating conditions. Electrically heated rods made of magnesium oxide and Zircaloy, combined with the proposed surrogate fluids, provide a close matching of conjugate heat transfer. Additionally, the use of these surrogates offers a significant decrease of the heating and pumping powers. Single-phase heat transfer separate-effect tests can then be performed for the first time in a laboratory setup with one, or several, full-size fuel bundles at prototypical conditions, while allowing the deployment of a large range of diagnostics. Finally, existing test facilities for hydraulics and thermal hydraulics of prototypical fuel bundles can be utilized with minor retrofits, further facilitating test implementation.