Assessment of heat transfer correlations in the sub-channels of proposed rod bundle geometry for supercritical water reactor

Many heat-transfer correlations exist for bare tubes cooled with SuperCritical Water (SCW). However, there is very few correlations that describe SCW heat transfer in bundles. Due to the lack of extensive data on bundles, a limited dataset on heat transfer in a SCW-cooled bundle was studied and analyzed using existing bare-tube correlations to find the best-fit correlation. This dataset was obtained by Razumovskiy et al. (National Technical University of Ukraine “KPI”) in SCW flowing upward in a vertical annular channel (1-rod channel) and tight 3-rod bundle consisting of tubes of 5.2-mm outside diameter and 485-mm heated length. The heat-transfer data were obtained at pressures of 22.5, 24.5, and 27.5 MPa, mass flux within a range from 800 to 3000 kg/m2s, inlet temperature from 125 to 352°C, outlet temperature up to 372°C and heat flux up to 4.6 MW/m2. The objective of this study is to compare bare-tube SCW heat-transfer correlations with the data on 1- and 3-rod bundles. This work is in support of SuperCritical Water-cooled Reactors (SCWRs) as one of the six concepts of Generation-IV nuclear systems. SCWRs will operate at pressures of ∼25MPa and inlet temperatures of 350°C.

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Coupled Monte Carlo-CFD analysis of heat transfer phenomena in a supercritical water reactor fuel assembly

Annals of Nuclear Energy ◽

10.1016/j.anucene.2020.107312 ◽

2020 ◽

Vol 141 ◽

pp. 107312 ◽

Cited By ~ 2

Author(s):

Landy Castro ◽

Juan-Luis François ◽

Carlos García

Keyword(s):

Heat Transfer ◽

Monte Carlo ◽

Fuel Assembly ◽

Supercritical Water ◽

Reactor Fuel ◽

Cfd Analysis ◽

Water Reactor ◽

Reactor Fuel Assembly ◽

Supercritical Water Reactor

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CFD analysis of flow and heat transfer in Canadian supercritical water reactor bundle

Annals of Nuclear Energy ◽

10.1016/j.anucene.2014.07.039 ◽

2015 ◽

Vol 75 ◽

pp. 1-10 ◽

Cited By ~ 16

Author(s):

K. Podila ◽

Y.F. Rao

Keyword(s):

Heat Transfer ◽

Supercritical Water ◽

Cfd Analysis ◽

Water Reactor ◽

Flow And Heat Transfer ◽

Supercritical Water Reactor

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Numerical Analysis of Flow and Heat transfer in Sub-channel of Supercritical Water Reactor

Procedia Engineering ◽

10.1016/j.proeng.2013.09.125 ◽

2013 ◽

Vol 64 ◽

pp. 507-517

Author(s):

Ajoy Debbarma ◽

K.M. Pandey

Keyword(s):

Heat Transfer ◽

Numerical Analysis ◽

Supercritical Water ◽

Water Reactor ◽

Flow And Heat Transfer ◽

Supercritical Water Reactor

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Transient Three-Dimensional Stability Analysis of Supercritical Water Reactor Rod Bundle Subchannels by a Computatonal Fluid Dynamics Code

Journal of Engineering for Gas Turbines and Power ◽

10.1115/1.3032437 ◽

2009 ◽

Vol 131 (2) ◽

Cited By ~ 13

Author(s):

M. Sharabi ◽

W. Ambrosini ◽

S. He ◽

Pei-Xue Jiang ◽

Chen-Ru Zhao

Keyword(s):

Fluid Dynamics ◽

Supercritical Water ◽

Density Wave ◽

Three Dimensional ◽

Heating Rates ◽

Rod Bundles ◽

Water Reactor ◽

Rod Bundle ◽

Computational Fluid Dynamics Cfd ◽

Supercritical Water Reactor

The paper describes the application of computational fluid dynamics (CFD) in simulating density wave oscillations in triangular and square pitch rod bundles. The FLUENT code is used for this purpose, addressing typical conditions proposed for supercritical water reactor (SCWR) conceptual design. The RELAP5 code and an in-house 1D linear stability code are also adopted to compare the results for instability thresholds obtained by different techniques. Transient analyses are performed both by the CFD code and RELAP5, with increasing heating rates and constant pressure drop across the channel, up to the occurrence of unstable behavior. The obtained results confirm that the density wave mechanism is similar in rod bundle and in axisymmetric configurations.

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Coupled Computational Heat Transfer and Reactor Physics for SCWR

Volume 4: Heat and Mass Transfer Under Extreme Conditions; Environmental Heat Transfer; Computational Heat Transfer; Visualization of Heat Transfer; Heat Transfer Education and Future Directions in Heat Transfer; Nuclear Energy ◽

10.1115/ht2013-17376 ◽

2013 ◽

Cited By ~ 1

Author(s):

Christopher R. Hughes ◽

DuWayne Schubring ◽

Kelly A. Jordan ◽

Dominik Rätz

Keyword(s):

Heat Transfer ◽

Heat Generation ◽

Supercritical Water ◽

Thermal Model ◽

Computational Cost ◽

Radial Dependence ◽

Radial Temperature ◽

Water Reactor ◽

Geometry Model ◽

Supercritical Water Reactor

To fully model the physics present within the proposed supercritical water reactor (SCWR), the thermal hydraulics calculations (yielding temperatures and densities in each material as a function of space) must be coupled to the neutronic calculations (yielding reactivity and neutron flux shapes). To enable this full coupling, a 3D model of a supercritical water reactor is being implemented in the CFD software OpenFOAM with the same geometry as a 3D MCNP (neutronics) model. Coupling will performed through result exchange between the two codes — densities and temperatures from OpenFOAM to MCNP, with heat generation returned from the neutronics calculations. Use of a reduced-geometry model is advisable due to the high computational cost of each OpenFOAM/MCNP coupling iteration. In the present work, a 1.5D thermal model of a single fuel pin was coupled with a 3D MCNP model. The thermal model includes single channel analysis (cladding/coolant heat transfer) as well as heat transfer within the cladding, helium gas gap, and uranium dioxide fuel itself. These heat transfer zones provide specific data points of the radial temperature profile. Because no radial mesh is considered, full radial dependence is not possible. The model provides limited radial dependence unlike what a 1D code could provide; thus, 1.5D is used to indicate the incomplete radial dependence that is included in the model. Iteration between the two codes is performed until heat generation is converged to within 10% between successful MCNP results. A discussion of these 3D/1.5D coupled results and the path forward to fully 3D coupling is provided.

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Numerical Analysis to Investigate the Effects of Thermal-Hydraulic Instabilities on Deterioration Heat Transfer and Wall Temperature in the CANDU Supercritical Water Reactor

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4030199 ◽

2015 ◽

Vol 1 (4) ◽

Cited By ~ 2

Author(s):

Goutam Dutta ◽

Chao Zhang ◽

Jin Jiang

Keyword(s):

Heat Transfer ◽

Supercritical Water ◽

Wall Temperature ◽

Stability Boundary ◽

Density Wave ◽

Operating Regime ◽

Maximum Temperature ◽

Water Reactor ◽

Heating Section ◽

Supercritical Water Reactor

The present work analyzes the thermal-hydraulic behavior of the CANDU supercritical water reactor (SCWR) using a 1-D numerical model. The possibility of a static instability, the Ledinegg excursion, is investigated, which reveals it can occur only in a hypothetical condition, far from the proposed operating regime of the CANDU SCWR. The investigation demonstrates the possibility of density wave oscillations (DWOs), a dynamic instability, in the operating regime of the CANDU SCWR and its marginal stability boundary (MSB) is obtained. The phenomenon of the deterioration in heat transfer is observed, and the related investigation shows that the strong buoyancy effect is responsible for its appearance inside the heating section of the channel of the CANDU SCWR core. The MSB is found to be inadequate in determining the safe operating zone of the reactor because the wall temperature can exceed the allowable limit from metallurgical consideration. The investigations also determine the safe as well as stable zone where the CANDU SCWR should operate in order to avoid the maximum temperature limit and DWOs.

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