Experimental data and correlations are not available for the fuel-assembly concept of the Canadian supercritical water-cooled reactor (SCWR). To facilitate the safety analyses, a strategy for developing a heat-transfer correlation has been established for the fuel-assembly concept at supercritical pressure conditions. It is based on an analytical approach using a computational fluid dynamics (CFD) tool and the ASSERT subchannel code to establish the heat transfer in supercritical pressure flow. Prior to the application, the CFD tool was assessed against experimental heat transfer data at the pseudocritical region obtained with bundle subassemblies to identify the appropriate turbulence model for use. Beyond the pseudocritical region, where the normal heat transfer behavior is anticipated, the ASSERT subchannel code also was assessed with appropriate closure relationships. Detailed information on the supporting experiments and the assessment results of the computational tools are presented.
Skip Nav Destination
Article navigation
January 2016
Research Papers
Assessment of Computational Tools in Support of Heat-Transfer Correlation Development for Fuel Assembly of Canadian Supercritical Water-Cooled Reactor
Krishna Podila
Krishna Podila
Search for other works by this author on:
Laurence K. H. Leung
Yanfei Rao
Krishna Podila
Manuscript received May 25, 2015; final manuscript received August 4, 2015; published online December 9, 2015. Assoc. Editor: Thomas Schulenberg.
ASME J of Nuclear Rad Sci. Jan 2016, 2(1): 011006 (9 pages)
Published Online: December 9, 2015
Article history
Received:
May 25, 2015
Revision Received:
August 4, 2015
Accepted:
August 4, 2015
Citation
Leung, L. K. H., Rao, Y., and Podila, K. (December 9, 2015). "Assessment of Computational Tools in Support of Heat-Transfer Correlation Development for Fuel Assembly of Canadian Supercritical Water-Cooled Reactor." ASME. ASME J of Nuclear Rad Sci. January 2016; 2(1): 011006. https://doi.org/10.1115/1.4031283
Download citation file:
Get Email Alerts
Cited By
Editorial
ASME J of Nuclear Rad Sci (January 2025)
Greeting From the President-Elect of JSME
ASME J of Nuclear Rad Sci (January 2025)
Operation Optimization Framework for Advanced Reactors Using a Data-Driven Digital Twin
ASME J of Nuclear Rad Sci (April 2025)
Numerical Analysis of Gas Generation and Migration in a Radioactive Waste Disposal Cell of a Deep Geological Repository
ASME J of Nuclear Rad Sci (April 2025)
Related Articles
Computational Fluid Dynamic Simulations of Heat Transfer From a 2 × 2 Wire-Wrapped Fuel Rod Bundle to Supercritical Pressure Water
ASME J of Nuclear Rad Sci (January,2018)
Studies of the Thermalhydraulics Subchannel Code ASSERT-PV 3.2-SC for Supercritical Applications
ASME J of Nuclear Rad Sci (April,2025)
Computational Fluid Dynamics Prediction of Heat Transfer in Rod Bundles With Water at Supercritical Pressure
ASME J of Nuclear Rad Sci (January,2016)
A Blind, Numerical Benchmark Study on Supercritical Water Heat Transfer Experiments in a 7-Rod Bundle
ASME J of Nuclear Rad Sci (April,2016)
Related Proceedings Papers
Related Chapters
Introduction
Heat Transfer & Hydraulic Resistance at Supercritical Pressures in Power Engineering Applications
Natural Gas Transmission
Pipeline Design & Construction: A Practical Approach, Third Edition
Compressive Deformation of Hot-Applied Rubberized Asphalt Waterproofing
Roofing Research and Standards Development: 10th Volume