• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.376-385
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• 2016

A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic design. The iso-kinetic sampling method has been adopted to measure the flow rate at subchannels, and newly designed sampling probes which preserve the flow area of subchannels have been devised. Experimental tests have been performed at 20-115% of the nominal flow rate and $60^{\circ}C$ (equivalent to Re ~ 37,100) at the inlet of the test rig. The pressure loss data in three measured subchannels were almost identical regardless of the subchannel locations. The flow rate at each type of subchannel was identified and the flow split factors were evaluated from the measured data. The predicted correlations and the computational fluid dynamics results agreed reasonably with the experimental data.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.128-128
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• 2001

• The KSFM Journal of Fluid Machinery
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• v.17 no.6
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• pp.89-94
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• 2014

The 4th generation SFR is being designed with a milestone of construction by 2028. It is important to understand the subchannel flow characteristics in fuel assembly through the experimental investigations and to estimate the calculation uncertainties for insuring the confidence of the design code calculation results. The friction coefficient and the mixing coefficient are selected as primary parameters. The two parameters are related to the flow distribution and diffusion. To identify the flow distribution, an iso-kinetic method was developed based on the previous study. For the mixing parameters, a wire mesh system and a laser induced fluorescence methods were developed in parallel. The measuring systems were adopted on 37 rod bundle test geometry, which was developed based on the Euler number scaling. A scaling method for a design of experimental facility and the experimental identification techniques for the flow distribution and mixing parameters were developed based on the measurement requirement.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3133-3150
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• 2021

This work deals with the validation of a high-fidelity multiphysics system coupling the Serpent 2 Monte Carlo neutron transport code with SUBCHANFLOW, a subchannel thermalhydraulics code, and TRANSURANUS, a fuel-performance analysis code. The results for a full-core pin-by-pin burnup calculation for the ninth operating cycle of the Temelín II VVER-1000 plant, which starts from a fresh core, are presented and assessed using experimental data. A good agreement is found comparing the critical boron concentration and a set of pin-level neutron flux profiles against measurements. In addition, the calculated axial and radial power distributions match closely the values reported by the core monitoring system. To demonstrate the modeling capabilities of the three-code coupling, pin-level neutronic, thermalhydraulic and thermomechanic results are shown as well. These studies are encompassed in the final phase of the EU Horizon 2020 McSAFE project, during which the Serpent-SUBCHANFLOW-TRANSURANUS system was developed.

• Journal of computational fluids engineering
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• v.8 no.2
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• pp.8-15
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• 2003

Three nonlinear κ-ε models with the wall function method are applied to the fully developed turbulent flow in a triangular subchannel of a bare rod bundle. Typical predicted quantities such as axial and secondary velocities, turbulent kinetic energy and wall shear stress are compared in details both qualitatively and quantitatively with both each other and experimental data. The nonlinear κ-ε models by Speziale[1] and Myong and Kasagi[2] are found to be capable of predicting accurately noncircular duct flows involving turbulence-driven secondary motion. The nonlinear κ-ε model by Shih et aL.[3] adopted in a commercial code is found to be unable to predict accurately noncircular flows with the prediction level of secondary flows one order less than that of the experiment.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.403-408
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• 1996

The subchannel analysis was performed to verify the behavior of hot channel characteristics and obtain the information to support the core thermal-hydraulic behavior at post-trip steam line break with low flow condition. During this postulated accident, buoyancy-induced cross flow occurs, and the coupled nuclear and thermal-hydraulic interactions become important. The code predictions with TORC are in good agreement with the test data. Under such conditions, the mass flow increase in the hot channel by buoyancy-induced cross flow depends on the parameter $GR^{*}\;/\;Re^2$, and buoyancy effect becomes more noticeable as $GR^{*}\;/\;Re^2$ increases.

• 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.

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1326-1338
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• 2017

The continuous energy Monte Carlo neutron transport code, MC21, was coupled to the CTF subchannel thermal-hydraulics code using a combination of Consortium for Advanced Simulation of Light Water Reactors (CASL) tools and in-house Python scripts. An MC21/CTF solution for VERA Core Physics Benchmark Progression Problem 6 demonstrated good agreement with MC21/COBRA-IE and VERA solutions. The MC21/CTF solution for VERA Core Physics Benchmark Progression Problem 7, Watts Bar Unit 1 at beginning of cycle hot full power equilibrium xenon conditions, is the first published coupled Monte Carlo neutronics/subchannel T-H solution for this problem. MC21/CTF predicted a critical boron concentration of 854.5 ppm, yielding a critical eigenvalue of $0.99994{\pm}6.8E-6$ (95% confidence interval). Excellent agreement with a VERA solution of Problem 7 was also demonstrated for integral and local power and temperature parameters.

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1339-1345
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• 2017

In the framework of the EU funded project NURESAFE, the subchannel code CTF and the neutronics code DYN3D were integrated and coupled on the NURESIM platform. The developments achieved during this 3-year project include assembly-level and pin-by-pin multiphysics thermal hydraulics/neutron kinetics coupling. In order to test this coupling, a PWR rod ejection transient was simulated on a MOX/UOX minicore. The transient is simulated using two different models of the minicore. In the first simulation, both codes model the core with an assembly-wise resolution. In the second simulation, a pin-by-pin fuel-centered model is used in CTF for the central assembly, and a pin power reconstruction method is applied in DYN3D. The analysis shows the influence of the different models on global parameters, such as the power and the average fuel temperature, but also on local parameters such as the maximum fuel temperature.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.148.1-148
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• 2001