• Proceedings of the Korean Nuclear Society Conference
• /
• 1997.05a
• /
• pp.316-321
• /
• 1997

The mixing vane effect on the Critical Heat Flux (CHF) is discussed with focus on the vortex now effect. In the subchannel approach, this effect is not quantified by the calculation model, but directly taken into account by the CHF correlation itself through data analysis. The vortex now effect is identified the two Westinghouse correlations, and then the CHF margin issue given rise to by the Vantage-5H design change is evaluated and discussed. It is noted that deficiency about CHF dependency on the vortex flow effect could induce an error in the Departure from Nucleate Boiling Ratio (DNBR) sensitivity Calculation.

• Nuclear Engineering and Technology
• /
• v.29 no.5
• /
• pp.406-416
• /
• 1997

A computational method is developed for predicting the steady-state temperature field in an LMR core. Detailed core-wide coolant temperature profiles are efficiently calculated using the simplified energy equation mixing model[1] and the subchannel analysis method. The $\theta$-method is employed for discretizing the energy equations in the axial direction. The interassembly coupling is achieved by interassembly gap flow. Cladding and fuel temperatures are calculated with the one-dimensional conduction model and temperature integrals of conductivities. The accuracy of the method is tested by performing several benchmark calculations for too LMR problems. The results indicate that the accuracy is comparable to the other methods based on ENERGY model. It is also shown that the implicit scheme for the axial discretization is more efficient than the explicit scheme.

• Proceedings of the IEEK Conference
• /
• 2003.11c
• /
• pp.3-6
• /
• 2003

Crosstalk among telephone lines in the same bundles is a major impairment in current VDSL systems. In this paper, we analysis the crosstalk effect of DSL using different frequency band upon ANSI standard VDSL systems. Simulation results show that data rate of ANSI VDSL decrease by crosstalk from another VDSL using different frequency. This study has the potential implication and benefit for ANSI standard DMT VDSL energy distribution in subchannel

• Nuclear Engineering and Technology
• /
• v.52 no.9
• /
• pp.1945-1954
• /
• 2020

Subchannel code is one of the effective simulation tools for thermal-hydraulic analysis in nuclear reactor core. In order to reduce the computational cost and improve the calculation efficiency, empirical correlation of turbulent mixing coefficient is employed to calculate the lateral mixing velocity between adjacent subchannels. However, correlations utilized currently are often fitted from data achieved in central channel of fuel assembly, which would simply neglect the wall effects. In this paper, the CFD approach based on spectral element method is employed to predict turbulent mixing phenomena through gaps in 3 × 3 bare tight lattice rod bundle and investigate the flow pulsation through gaps in different positions. Re = 5000,10000,20500 and P/D = 1.03 and 1.06 have been covered in the simulation cases. With a well verified mesh, lateral velocities at gap center between corner channel and wall channel (W-Co), wall channel and wall channel (W-W), wall channel and center channel (W-C) as well as center channel and center channel (C-C) are collected and compared with each other. The obvious turbulent mixing distributions are presented in the different channels of rod bundle. The peak frequency values at W-Co channel could have about 40%-50% reduction comparing with the C-C channel value and the turbulent mixing coefficient β could decrease around 25%. corrections for β should be performed in subchannel code at wall channel and corner channel for a reasonable prediction result. A preliminary analysis on fluctuation at channel gap has also performed. Eddy cascade should be considered carefully in detailed analysis for fluctuating in rod bundle.

• Journal of Energy Engineering
• /
• v.13 no.2
• /
• pp.101-111
• /
• 2004

The main purpose of a liquid metal reactor core thermal-hydraulic design is to efficiently extract the core thermal power by distributing the appropriate sodium coolant flow according to the power distribution in the core. The thermal-hydraulic design procedure consists of the coolant flow distribution to the sub-assemblies, the coolant/fuel temperature calculations and detailed subchannel analysis. This paper describes the LMR core thermal-hydraulic design methodology and summarizes the major design and analysis results of KALIMER breeder and breakeven cores and subassemblies. KALIMER is a 150 MWe rated (392 MWth) heterogeneous core with U-TRU-Zr ternary alloy fuel and sodium coolant.

• Nuclear Engineering and Technology
• /
• v.54 no.8
• /
• pp.3154-3165
• /
• 2022

During a loss-of-coolant accident (LOCA) in the pressurized water reactor (PWR), there is a possibility that high temperature and internal pressure of the fuel rods lead to ballooning of the cladding, which causes a partial blockage of flow area in a subchannel. Such flow blockage would influence the core coolant flow, thus affecting the core heat transfer during a reflooding phase and subsequent severe accident. However, most of the system analysis codes simulate the accident process based on the assumed channel blockage ratio, resulting in the fact that the simulation results are not consistent with the actual situation. This paper integrates the developed core Fuel Rod Thermal-Mechanical Behavior analysis (FRTMB) module into the self-developed severe accident analysis code ISAA. At the same time, the existing flow blockage model is improved to make it possible to simulate the change of flow distribution due to fuel rod deformation. Finally, the ISAA-FRTMB is used to simulate the QUENCH-LOCA-0 experiment to verify the correctness and effectiveness of the improved flow blockage model, and then the effect of clad ballooning on core heat transfer and subsequent parts of core degradation is analyzed.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
• /
• v.29 no.10
• /
• pp.538-550
• /
• 2017

In a PWR (Pressurized Water Reactor), the appropriate heat removal from the surface of fuel rod bundle is important for ensuring thermal margins and safety. Although many CFD (Computational Fluid Dynamics) software have been used to predict complex flows inside fuel assemblies with mixing vanes, there is no domestic regulatory guideline for the comprehensive evaluation of CFD software. Therefore, from the nuclear regulatory perspective, it is necessary to perform the systematic assessment and prepare the domestic regulatory guideline for checking whether valid CFD software is used for nuclear safety problems. In this study, to provide systematic evaluation and guidance on the applicability of CFD software to the domestic nuclear safety area, the results of the sensitivity analysis for the effect of the discretization scheme accuracy for the convection terms and turbulence models, which are main factors that contribute to the uncertainty in the calculation of the nuclear safety problems, on the prediction performance for the turbulent flow distribution inside the fuel assembly with split-type mixing vanes were explained.

• Nuclear Engineering and Technology
• /
• v.24 no.2
• /
• pp.168-177
• /
• 1992

The importance of the study of two phase flow phenomena has increased for both fuel performance and safety analysis of nuclear power plants. In the analysis of two phase flow system, an accurate prediction of local void fractions is very important. In this study, a vertical rectangular subchannel having 4 electrically heated rods is constructed for the measurement of local void fraction under two phase flow. The measurement has been conducted by electrical conductivity probes and signal processing circuit which are known to be adequate to measuring local void fraction. Also experiments are performed with varying the inlet flow rate to search for radial void fraction profile accordingly to the different flow rate even with the same averaged void fraction. From the result of experiments, the validity of electrical conductivity probe and electrical circuit is confirmed.

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

• Proceedings of the Korean Nuclear Society Conference
• /
• 1997.10a
• /
• pp.534-539
• /
• 1997

A CHF table method has been assessed in this study for rod bundle CHF predictions. At the conceptual design stage for a new reactor, a general critical heat flux (CHF) prediction method with a wide applicable range and reasonable accuracy is essential to the thermal-hydraulic design and safety analysis. In many aspects, a CHF table method (i.e., the use of a round tube CHF table with appropriate bundle correction factors) can be a promising way to fulfill this need. So the assessment of the CHF table method has been performed with the bundle CHF data relevant to pressurized water reactors (PWRs). For comparison purposes, W-3R and EPRI-1 were also applied to the same data base. Data analysis has been conducted with the subchannel code COBRA-IV-I. The CHF table method shows the best predictions based on the direct substitution method. Improvements of the bundle correction factors, especially for the spacer grid and cold wall effects, are desirable for better predictions. Though the present assessment is somewhat limited in both fuel geometries and operating conditions, the CHF table method clearly shows potential to be a general CHF predictor.