Icone19-43644 Temperature Profiles of a Vertical Bare 7-Element Bundle Cooled With Supercritical Freon-12

Richards, Graham; Shelegov, Aleksandr Sergeevich; Kirillov, P. L.; Pioro, Igor; Harvel, Glenn

doi:10.1299/jsmeicone.2011.19._icone1943_260

Cited by 3 publications

(3 citation statements)

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“…The experimental facility and cross-sectional view of the test section are shown in Figure 1. A detailed description of the experimental facility and test setups are discussed in Kirillov et al [6] and Richards et al [7]. Table 1 summarizes the test conditions of the four test cases (2, 3, 6, and 9; "low-massflux" tests) selected for this study.…”

Section: Experiments Used For Assessmentmentioning

confidence: 99%

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CFD Simulation of Vertical Seven-Rod Bundle Cooled With Supercritical Freon-12

HuangX.

PodilaK.

2014

AECL Nuclear Review

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In this paper, a seven-rod bare bundle was simulated using ANSYS Fluent 6.3.26 to accurately predict the fluid flow and heat transfer behaviour under supercritical flow conditions. Seven turbulence models were compared to identify the appropriate model to predict the experiments performed at the Institute of Physics and Power Engineering on a vertically oriented seven-rod bare bundle cooled with supercritical Freon-12. It was found that predictions of wall temperatures and heat transfer coefficients are sensitive to the choice of turbulence model as well as to the near-wall treatment. Overall, the CFD simulations were able to predict the measured sheath temperature profiles along the length of the bundle within reasonable accuracy.

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Section: Experiments Used For Assessmentmentioning

confidence: 99%

“…According to Kirillov et al [7], the material for the rod sheath is Russian stainless steel 1X9H10T, which is equivalent to US stainless steel 304L. Hence, the properties of stainless steel US 304L are implemented in the ANSYS Fluent case set up as listed in Table 5.…”

Section: Implementation Of the Materials Propertiesmentioning

confidence: 99%

CFD Simulation of Vertical Seven-Rod Bundle Cooled With Supercritical Freon-12

HuangX.

PodilaK.

2014

AECL Nuclear Review

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“…Naidin et al [168]. Fluid-to-fluid scaling has been widely studied due to the ease of building test loops that use working fluids with lower parameters to achieve a supercritical phase [169,170]. Comprehensive surveys were prepared by Pioro and…”

Section: Simplified Supercritical Water Reactor (Sscwr)mentioning

confidence: 99%

Computational fluid dynamics of advanced nuclear reactor cores and its contribution to safety analysis

Rowinski¹

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Analytical methods and Computational Fluid Dynamics (CFD) examined the safety and performance of advanced nuclear reactors. Thermal and pressure drop evaluations of an innovative annular fuel in the European lead-cooled System ELSY fast reactor were conducted. Annular fuel has superior cooling capacity, and for a square fuel lattice, the pressure drop can be lower than the reference value for standard solid fuel. Consequently, when annular fuel rods replace solid fuel rods of the same dimensions the power rating is upgraded. While the hexagonal lattice cools more effectively, the pressure drop is higher than the standard fuel assembly and incompatible with power augmentation. Numerical investigations of supercritical flow of both carbon dioxide and water in a vertical tube under non-uniform heat flux applied at the wall demonstrated that current correlations for the Heat Transfer Coefficient (HTC) at low enthalpy values are accurate to ±15%. At high enthalpy, only the general trend is replicated due to large property value differences near the wall and in the bulk flow. In these circumstances, the thermal conductivity, specific heat, density and viscosity are 5-8 times lower after the transition from sub-to super-critical conditions. Moreover, the efficiency of different HTC correlation models is coolant dependant, e.g. the Swenson* formula proves superior for carbon dioxide, whereas the Ornatsky model** provides better agreement with water, where a lower maximum wall temperature was obtained when a non-uniform heat flux was applied, as compared to the reference case of uniform heat flux. For supercritical water flow inside a 2  2 fuel rod bundle, non-uniform heat flux increases wall temperature beyond the reference uniform flux. Also, wall temperature peaks occur only at the gaps between the fuel rods.

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The flow resistance experiments of supercritical pressure water in 2 × 2 rod bundle

Zang

Yan

et al. 2020

International Journal of Heat and Mass Transfer

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Icone19-43644 Temperature Profiles of a Vertical Bare 7-Element Bundle Cooled With Supercritical Freon-12

Cited by 3 publications

References 0 publications

CFD Simulation of Vertical Seven-Rod Bundle Cooled With Supercritical Freon-12

CFD Simulation of Vertical Seven-Rod Bundle Cooled With Supercritical Freon-12

Computational fluid dynamics of advanced nuclear reactor cores and its contribution to safety analysis

The flow resistance experiments of supercritical pressure water in 2 × 2 rod bundle

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Icone19-43644 Temperature Profiles of a Vertical Bare 7-Element Bundle Cooled With Supercritical Freon-12

Cited by 3 publications

References 0 publications

CFD Simulation of Vertical Seven-Rod Bundle Cooled With Supercritical Freon-12

CFD Simulation of Vertical Seven-Rod Bundle Cooled With Supercritical Freon-12

Computational fluid dynamics of advanced nuclear reactor cores and its contribution to safety analysis

The flow resistance experiments of supercritical pressure water in 2 × 2 rod bundle

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Product

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The flow resistance experiments of supercritical pressure water in 2 × 2 rod bundle