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Research Papers

Assessment of Computational Tools in Support of Heat-Transfer Correlation Development for Fuel Assembly of Canadian Supercritical Water-Cooled Reactor

[+] Author and Article Information
Laurence K. H. Leung

Canadian Nuclear Laboratories,
Chalk River, ON K0J 1J0, Canada
e-mail: Laurence.Leung@cnl.ca

Yanfei Rao

Canadian Nuclear Laboratories,
Chalk River, ON K0J 1J0, Canada
e-mail: Yanfei.Rao@cnl.ca

Krishna Podila

Canadian Nuclear Laboratories,
Chalk River, ON K0J 1J0, Canada
e-mail: Krishna.Podila@cnl.ca

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 2(1), 011006 (Dec 09, 2015) (9 pages) Paper No: 15-1094; doi: 10.1115/1.4031283 History: Received May 25, 2015; Accepted August 04, 2015

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.

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References

Figures

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Fig. 1

Schematic diagram of the Canadian SCWR fuel-assembly concept inside the fuel channel

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Fig. 2

Rod and subchannel numbering systems for ASSERT subchannel code applications

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Fig. 3

2×2 Bundle test-section configuration

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Fig. 4

Thermocouple measuring points and cross-sectional geometry of flow channel

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Fig. 5

Circumferential wall-temperature distributions obtained from the Phase-I Test

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Fig. 6

Circumferential wall-temperature distributions obtained from the Phase-II Test

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Fig. 14

STAR-CCM+ predictions of circumferential wall-temperature distributions near pseudocritical temperature

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Fig. 13

STAR-CCM+ predictions of circumferential wall-temperature distributions at supercritical temperature

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Fig. 12

Mesh sensitivity analysis for the 2×2 rod bundle assembly at the sub- and supercritical test conditions

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Fig. 11

Computational domain and mesh on a cross section used for the STAR-CCM+ simulations

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Fig. 10

Cross section of the CAD model for the 2×2 fuel rod bundle. Dark gray and bisque shaded regions indicate solid and fluid domains, respectively

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Fig. 9

ASSERT predictions of circumferential wall-temperature distributions at subcritical temperature

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Fig. 8

ASSERT predictions of circumferential wall-temperature distributions near pseudocritical temperature

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Fig. 7

ASSERT predictions of circumferential wall-temperature distributions at supercritical temperature

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Fig. 15

STAR-CCM+ predictions of circumferential wall-temperature distributions at subcritical temperature

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