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Conference Spotlight
Nuclear Energy Conference & Expo (NECX)
September 8–11, 2025
Atlanta, GA|Atlanta Marriott Marquis
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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Latest News
ANS joins others in seeking to discuss SNF/HLW impasse
The American Nuclear Society joined seven other organizations to send a letter to Energy Secretary Christopher Wright on July 8, asking to meet with him to discuss “the restoration of a highly functioning program to meet DOE’s legal responsibility to manage and dispose of the nation’s commercial and legacy defense spent nuclear fuel (SNF) and high-level radioactive waste (HLW).”
Satish Kumar Dhurandhar, S. L. Sinha, Shashi Kant Verma
Nuclear Technology | Volume 206 | Number 5 | May 2020 | Pages 663-696
Critical Review | doi.org/10.1080/00295450.2019.1698257
Articles are hosted by Taylor and Francis Online.
Flow and thermal performance analysis is an important area in the nuclear industry for the safe operation of reactors and the optimum design of fuel rod bundles with spacers in the reactor core. The spacer, an essential element in the nuclear fuel rod assembly, is used to support the fuel rods in the bundle and maintain suitable clearance between the rods. This paper represents a comprehensive review of the effects of spacer/mixing vane spacer on flow behavior and heat transfer performance in nuclear fuel rod bundles, annular channels, and circular tubes. This paper is based on the experimental and computational fluid dynamics (CFD) investigations performed by researchers in literature for both single-phase and two-phase fluid flow. In addition, the effects of some parameters, such as Reynolds number, hydraulic diameter, pitch-to-diameter ratio, blockage ratio, working fluid, and number of fuel rods in an assembly, on flow and thermal performance analysis have been reviewed. A detailed review has been carried out by considering both experimental and CFD methods used in the past for analysis of thermal-hydraulic performance in the fuel channels of nuclear reactors.