Simulating fuel assemblies with low resolution CFD approaches

Roelofs, F.; Gopala, Vinay R.; Chandra, Laltu; Viellieber, Mathias; Class, Andreas G.

doi:10.1016/j.nucengdes.2012.05.029

Cited by 28 publications

(8 citation statements)

References 8 publications

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“…In this approach a momentum source model was developed to mimic the effects of the wire-wraps without explicitly modeling the geometrical representation of the wire. Similar low grid resolution approaches of rod bundle simulation have been reported by Roelofs et al (2012) and Viellieber and Class (2015). Such low-resolution approaches are required to simulate the overall behavior of the coolant flow in a complete fuel assembly or even multiple fuel assemblies interacting with each other.…”

Section: Introductionsupporting

confidence: 77%

A computationally efficient method for full-core conjugate heat transfer modeling of sodium fast reactors

2016

Nuclear Engineering and Design

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For efficient and accurate temperature predictions of sodium fast reactor structures, a 3-D full-core conjugate heat transfer modeling capability is developed for an advanced system analysis tool, SAM. The hexagon lattice core is modeled with 1-D parallel channels representing the subassembly flow, and 2-D duct walls and inter-assembly gaps. The six sides of the hexagon duct wall and near-wall coolant region are modeled separately to account for different temperatures and heat transfer between coolant flow and each side of the duct wall. The Jacobian Free Newton Krylov (JFNK) solution method is applied to solve the fluid and solid field simultaneously in a fully coupled fashion. The 3-D full-core conjugate heat transfer modeling capability in SAM has been demonstrated by a verification test problem with 7 fuel assemblies in a hexagon lattice layout. Additionally, the SAM simulation results are compared with RANS-based CFD simulations. Very good agreements have been achieved between the results of the two approaches.

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Section: Introductionsupporting

confidence: 77%

A computationally efficient method for full-core conjugate heat transfer modeling of sodium fast reactors

2016

Nuclear Engineering and Design

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“…The local temperature increase due to the blockage is about 150°C. As evidenced in other computational and theoretical studies (Di (Kirsch, 1975) (Roelofs, 2012), the reason for the local temperature peak is the vortex generated downstream by the obstacle represented by the blockage; in the central, stagnation point of the vortex heat is exchanged only by conduction and the temperature increases. This hydrodynamic effect is shown in Figure 9, where temperature field and secondary flow field (left) are shown at the cross-sections at z=10, 20, 30, 45 mm from the blockage (located at z=0 mm at the beginning of the active region, first grid) for case 1 (sector blockage).…”

Section: Numerical Models and Methodsmentioning

confidence: 63%

Pre-test CFD simulations of the NACIE-UP BFPS test section

Marinari

Piazza

Forgione

et al. 2017

Annals of Nuclear Energy

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The present paper is focused on the CFD pre-test analysis and design of the new experimental facility Blocked Fuel Pin bundle Simulator (BFPS) that will be installed into the NACIE-UP (NAtural CIrculation Experiment-UPgrade) facility located at the ENEA Brasimone Research Center (Italy). The BFPS test section will carry out suitable experiments to fully investigate different flow blockage regimes in a 19 fuel pin bundle providing experimental data in support of the development of the ALFRED (Advanced Lead-cooled Fast Reactor European Demonstrator) LFR DEMO. The geometrical domain of the fuel pin bundle simulator was designed to reproduce the geometrical features of ALFRED, e.g. the external wrapper in the active region and the spacer grids. Pre-tests calculations were carried out by applying accurate boundary conditions; the conjugate heat transfer in the clad is also considered. The blockages investigated are internal blockages of different extensions and in different locations: central sub-channel blockage, corner sub-channel blockage, edge sub-channel blockage, one sector blockage, and two-sector blockage. RANS simulations were carried out adopting the ANSYS CFX commercial code with the laminar sublayer resolved by the mesh resolution. The loci of the peak temperatures and their width as predicted by the CFD simulations are used for determining the location of the pin bundle instrumentation. The CFD pre-test analysis allowed also investigating the temperature distribution in the clad to operate the test section safely

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“…Roelofs et al [43] discuss the shortcomings of subchannel code models with respect to the effects which can be captured, and propose low resolution CFD simulations to capture the additional details. The authors discuss the need for models which can be applied to subchannel codes and are "as efficient as subchannel codes and … independent from experimental or empirical data".…”

Section: Numerical Studies On Spacer Grid Effects On Flow Patterns Anmentioning

confidence: 99%

Development and Implementation of CFD-Informed Models for the Advanced Subchannel Code CTF

Blyth¹

2017

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The research described in this PhD thesis contributes to the development of efficient methods for utilization of high-fidelity models and codes to inform low-fidelity models and codes in the area of nuclear reactor core thermal-hydraulics. The objective is to increase the accuracy of predictions of quantities of interests using high-fidelity CFD models while preserving the efficiency of low-fidelity subchannel core calculations. An original methodology named Physicsbased Approach for High-to-Low Model Information has been further developed and tested. The overall physical phenomena and corresponding localized effects, which are introduced by the presence of spacer grids in light water reactor (LWR) cores, are dissected in corresponding four building basic processes, and corresponding models are informed using high-fidelity CFD codes. These models are a spacer grid-directed cross-flow model, a grid-enhanced turbulent mixing model, a heat transfer enhancement model, and a spacer grid pressure loss model. The localized CFD-models are developed and tested using the CFD code STAR-CCM+, and the corresponding global model development and testing in sub-channel formulation is performed in the thermalhydraulic subchannel code CTF. The improved CTF simulations utilize data-files derived from CFD STAR-CCM+ simulation results covering the spacer grid design desired for inclusion in the CTF calculation. The current implementation of these models is examined and possibilities for improvement and further development are suggested. The validation experimental database is extended by including the OECD/NRC PSBT benchmark data. The outcome is an enhanced accuracy of CTF predictions while preserving the computational efficiency of a low-fidelity subchannel code. v

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Simulating fuel assemblies with low resolution CFD approaches

Cited by 28 publications

References 8 publications

A computationally efficient method for full-core conjugate heat transfer modeling of sodium fast reactors

A computationally efficient method for full-core conjugate heat transfer modeling of sodium fast reactors

Pre-test CFD simulations of the NACIE-UP BFPS test section

Development and Implementation of CFD-Informed Models for the Advanced Subchannel Code CTF

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