• Proceedings of the KSME Conference
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• 2004.11a
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• pp.1592-1597
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• 2004

Experimental studies were carried out to confirm the turbulent enhancement of the cooling system of nuclear reactor by large scale vortex generation in nuclear fuel rod bundle. The large scale vortex motions were generated by rearranging the inclination angles of mixing vanes to the coordinate directions. Experimental studies were carried out at Reynolds Number 60,000 with hydraulic condition. Normal variations of mean velocity and turbulent intensity in the rod bundle subchannel were measured by the 2-color LDV measurement system. The turbulence generated by split mixing vanes has small length scales so that they maintain only about 10DH after the spacer grid. On the other hand, the turbulences generated by the large scale vortex continue more and remain up $25D_{H}$ after the spacer grid.

• Transactions of the Korean Society of Mechanical Engineers
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• v.18 no.11
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• pp.2986-2998
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• 1994

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.665-675
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• 2019

The wire-wrapped fuel bundle is an assembly design in a sodium-cooled fast reactor. A wire spacer is used to maintain a constant gap between rods and to enhance the mixing of coolants. The wire makes the flow complicated by creating a sweeping flow and vortex flow. The vortex affects the flow field and heat transfer inside the subchannels. However, studies on vortices in this geometry are limited. The purpose of this research is to investigate the vortex flow created in the wire-wrapped fuel bundle. For analysis, a RANS-based numerical analysis was conducted for a 37-pin geometry. The sensitivity study shows that simulation with the shear stress transport model is appropriate. For the case of Re of 37,100, the mechanisms of onset, periodicity, and rotational direction were analyzed. The vortex structures were reconstructed in a three-dimensional space. Vortices were periodically created in the interior subchannel three times for one wire rotation. In the edge subchannel, the largest vortex occurred. This large vortex structure blocked the swirl flow in the peripheral region. The small vortex formed in the corner subchannel was negligible. The results can help in understanding the flow field inside subchannels with sweeping flow and vortex structures.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1487-1503
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• 2019

The methods and performance of a pin-level nuclear reactor core thermal-hydraulics (T/H) code ESCOT employing the drift-flux model are presented. This code aims at providing an accurate yet fast core thermal-hydraulics solution capability to high-fidelity multiphysics core analysis systems targeting massively parallel computing platforms. The four equation drift-flux model is adopted for two-phase calculations, and numerical solutions are obtained by applying the Finite Volume Method (FVM) and the Semi-Implicit Method for Pressure-Linked Equation (SIMPLE)-like algorithm in a staggered grid system. Constitutive models involving turbulent mixing, pressure drop, and vapor generation are employed to simulate key phenomena in subchannel-scale analyses. ESCOT is parallelized by a domain decomposition scheme that involves both radial and axial decomposition to enable highly parallelized execution. The ESCOT solutions are validated through the applications to various experiments which include CNEN $4{\times}4$, Weiss et al. two assemblies, PNNL $2{\times}6$, RPI $2{\times}2$ air-water, and PSBT covering single/two-phase and unheated/heated conditions. The parameters of interest for validation include various flow characteristics such as turbulent mixing, spacer grid pressure drop, cross-flow, reverse flow, buoyancy effect, void drift, and bubble generation. For all the validation tests, ESCOT shows good agreements with measured data in the extent comparable to those of other subchannel-scale codes: COBRA-TF, MATRA and/or CUPID. The execution performance is examined with a mini-sized whole core consisting of 89 fuel assemblies and for an OPR1000 core. It turns out that it is about 1.5 times faster than a subchannel code based on the two-fluid three field model and the axial domain decomposition scheme works as well as the radial one yielding a steady-state solution for the OPR1000 core within 30 s with 104 processors.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.28 no.5
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• pp.186-194
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• 2016

As a turbulence-enhancing device, a mixing vane installed at a spacer grid of the fuel assembly plays a role in improving the convective heat transfer by generating either swirl flow in the subchannels or cross flow between fuel rod gaps. Therefore, both configuration and arrangement pattern of a mixing vane are important factors that determine the performance of a mixing vane. In this study, in order to examine the flow distribution features inside $5{\times}5$ fuel assembly with swirl-type mixing vanes used in benchmark calculation of OECD/NEA, simulations were conducted with commercial CFD software ANSYS CFX R.14. Predicted results were compared to data measured from MATiS-H (Measurement and Analysis of Turbulent Mixing in Subchannels-Horizontal) test facility. In addition, the effect of swirl-type mixing vanes on flow pattern inside the fuel assembly was described.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4357-4366
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• 2023

MARS-KS, a domestic regulatory confirmatory code of Republic of Korea, had been developed by integrating RELAP5/MOD2 and COBRA-TF. The integration of COBRA-TF allowed to extend the capability of MARS-KS, limited to one-dimensional analysis, to multi-dimensional analysis. The use of COBRA-TF was mainly focused on subchannel analyses for simulating multi-dimensional behavior within the reactor core. However, this feature has been remained as a legacy without ongoing maintenance. Meanwhile, MARS-KS also includes its own multidimensional component, namely MULTID, which is also feasible to simulate three-dimensional convection and diffusion. The MULTID is capable of modeling the turbulent diffusion using simple mixing length model. The implementation of the turbulent mixing is of importance for analyzing the reactor core where a disturbing cross-sectional structure of rod bundle makes the flow perturbation and corresponding mixing stronger. In addition, the presence of this turbulent behavior allows the secondary transports with net mass exchange between subchannels. However, a series of assessments performed in previous studies revealed that the turbulence model of the MULTID could not simulate the aforementioned effective mixing occurred in the subchannel-scale problems. This is obvious consequence since the physical models of the MULTID neglect the effect of mass transport and thereby, it cannot model the void drift effect and resulting phasic distribution within a bundle. Thus, in this study, the turbulence mixing model of the MULTID has been improved by means of the inter-channel mixing model, widely utilized in subchannel analysis, in order to extend the application of the MULTID to small-scale problems. A series of assessments has been performed against rod bundle experiments, namely GE 3X3 and PSBT, to evaluate the performance of the introduced mixing model. The assessment results revealed that the application of the inter-channel mixing model allowed to enhance the prediction of the MULTID in subchannel scale problems. In addition, it was indicated that the code could not predict appropriate phasic distribution in the rod bundle without the model. Considering that the proper prediction of the phasic distribution is important when considering pin-based and/or assembly-based expressions of the reactor core, the results of this study clearly indicate that the inter-channel mixing model is required for analyzing the rod bundle, appropriately.

• Nuclear Engineering and Technology
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• v.37 no.5
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• pp.469-478
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• 2005

The subchannel analysis capability of MARS, a multi-dimensional thermal-hydraulic system code, has been enhanced. In particular, the turbulent mixing and void drift models for the flow-mixing phenomena in rod bundles were improved. Then, the subchannel analysis feature was combined with the existing coupled system thermal-hydraulics (T/H) and 3D reactor kinetics calculation capability of MARS. These features allow for more realistic simulations of both the hot channel behavior and the global system T/H behavior. Using the coupled features of MARS, a coupled analysis of a main steam line break (MSLB) is carried out for demonstration purposes. The results of the calculations are very reasonable and promising.

• Proceedings of the KSME Conference
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• 2005.11a
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• pp.97.1-97.1
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.255-256
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• 2004

• Proceedings of the SAREK Conference
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• 2003.07a
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• pp.9-9
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• 2003