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Study of Thermal Hydraulic Behavior of Supercritical Water Flowing Through Fuel Rod Bundles
Indian Institute of Technology Guwahati.ORCID iD: 0000-0001-8001-9323
Indian Institute of Technology Guwahati.
Indian Institute of Technology Guwahati.
2009 (English)In: The 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13), 2009Conference paper, Published paper (Refereed)
Abstract [en]

Investigations on thermal-hydraulic behavior in Supercritical Water Reactor (SCWR) fuel assembly have obtained a significant attention in the international SCWR community because of its potential to obtain high thermal efficiency and compact design. Present work deals with CFD analysis to study the flow and heat transfer behavior of supercritical water in 4 metre long 7-pin fuel bundle using commercial CFD package ANSYS CFX for single phase steady state conditions. Considering the symmetric conditions, 1/12th part of the fuel rod bundle is taken as a domain of analysis. RNG K-epsilon model with scalable wall functions is used for modeling the turbulence behavior. Constant heat flux boundary condition is applied at the fuel rod surface. IAPWS equations of state are used to compute thermo-physical properties of supercritical water. Sharp variations in its thermo-physical properties (specific heat, density) are observed near the pseudo-critical temperature causing sharp change in heat transfer coefficient. The pseudo-critical point initially appears in the gaps among heated fuel rods, and then spreads radially outward reaching the adiabatic wall as the flow goes downstream. The enthalpy gain in the centre of the channel is much higher than that in the wall region. Non-uniformity in the circumferential distribution of surface temperature and heat transfer coefficient is observed which is in agreement with published literature. Heat transfer coefficient is high on the rod surface near the tight region and decreases as the distance between rod surfaces increases.

Place, publisher, year, edition, pages
2009.
Keyword [en]
supercritical water reactor
National Category
Energy Engineering
Identifiers
URN: urn:nbn:se:kth:diva-62277OAI: oai:DiVA.org:kth-62277DiVA: diva2:480070
Conference
The 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13). Kanazawa City, Japan, September 27-October 2, 2009
Note
QC 20120119Available from: 2012-01-18 Created: 2012-01-18 Last updated: 2012-01-19Bibliographically approved

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Thakre, Sachin

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CiteExportLink to record
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  • apa
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