• 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.

• 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.54 no.7
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• pp.2702-2713
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• 2022

A steam generator tube rupture accompanying a core damage may cause the fission product to be released to environment bypassing the containment. In such an accident, the steam generator is the major path of the radioactive aerosol release. AEOLUS facility, the scaled-down model of Korean type steam generator, was built to examine the aerosol removal in the steam generator during the steam generator tube rupture accident. Integral and separate effect tests were performed with the facility for the dry and flooded conditions, and the decontamination factors were presented for different tube configurations and submergences. The dry test results were compared with the existing test results and with the analyses to investigate the aerosol retention physics by the tube bundle, with respect to the particle size and the bundle geometry. In the flooded tests, the effect of submergence were shown and the retention in the jet injection region were presented with respect to the Stokes number. The test results are planned to be used to constitute the aerosol retention model, specifically applicable for the analysis of the steam generator tube rupture accident in Korean nuclear power plants to evaluate realistic fission product behavior.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4296-4309
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• 2022

Mesh and turbulence model sensitivity analyses have been performed on computational fluid dynamics simulations executed with Hydra and ANSYS Fluent for a single CANadian Deuterium Uranium (CANDU) 37M nuclear fuel bundle placed within a standard pressure tube. The goal of this work was to perform a methodical analysis to objectively determine an appropriate mesh and to gauge the sensitivity of different turbulence models for CANDU subchannel flow under isothermal conditions. The boundary conditions and material properties are representative of normal operating conditions in a high-powered channel of the Darlington Nuclear Generating Station. Four meshes were generated with ANSYS Workbench Meshing, ranging from 22 to 84 million cells, and analyzed here to determine an appropriate level of mesh resolution and quality. Five turbulence models were compared in the turbulence model sensitivity analysis: standard k - ε, RNG k - ε, realizable k - ε, SST k - ω, and the Reynolds Stress Model. The intent of this work was to gain confidence in mesh generation and turbulence model selection of a single bundle to inform the decision making of subsequent investigations of an entire fuel channel containing a string of twelve bundles.

• Nuclear Engineering and Technology
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• v.41 no.3
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• pp.295-306
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• 2009

The subchannel grade void distributions in the NUPEC (Nuclear Power Engineering Corporation) BFBT (BWR Full-Size Fine-Mesh Bundle Tests) facility were evaluated with the subchannel analysis code MATRA and the system code MARS. Fifteen test series from five different test bundles were selected for an analysis of the steady-state subchannel void distributions. Two transient cases, a turbine trip without a bypass as a typical power transient and a re-circulation pump trip as a flow transient, were also chosen for this analysis. It was found that the steady-state void distributions calculated by both the MATRA and MARS codes coincided well with the measured data in the range of thermodynamic qualities from 5% to 25%. The results of the transient calculations were also similar and were highly feasible. However, the computational aspects of the two codes were clearly different.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1118-1124
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• 2023

Experimental study on the fluidelastic instability (FEI) of a curved tube bundle in single phase downward cross flow is investigated for the design qualification and analysis input preparation of helical coiled steam generator tubing. A 6×9 normal square curved tube array with equal and different vertical/horizontal pitch-to-diameter ratio was under-tested up to 6 m/s in term of gap flow velocity to measure the critical velocity for FEI. The critical velocity for FEI was measured at the turning point from the vibration amplitude plot along the gap flow velocity. Our test results were compared with straight tube results and published data in the design guideline. The applicability of the current design guidelines to a curved tube bundle is also assessed. We found that introducing frequency difference in a curved tube array increases the critical velocity for fluidelastic instability.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.886-892
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• 2024

The cladding temperatures and axial mass distribution computed by MAAP5 were compared with their measured values in the test bundle of the Phebus Test FPT0. The computed cladding temperatures were in good agreed with the measured values in the pre-transient phase. In the transient heat-up phase, the computed temperatures were overestimated by the Baker-Just correlation in MAAP5, but the computed temperatures could simulate the subsequently measured values. The computed mass distribution in the axial direction was in qualitative agreement with the measured one for post-test fuel damage observations. The calculated flow areas of inner and outer regions in the test bundle were compared with the photographic observations. MAAP5 computed them at the height of 0.2 m where the molten pool formed was in qualitative agreement with the photographic observations. It was found that the remaining steam flow paths might be caused by the gas-liquid two-phase flow counter-current flow limitation.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1914-1928
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• 2022

To improve the heat transfer efficiency of the reactor fuel assembly, it is necessary to accurately calculate the two-phase flow boiling characteristics and the critical heat flux (CHF) in the fuel assembly. In this paper, a Eulerian two-fluid model combined with the extended wall boiling model was used to numerically simulate the 5 × 5 fuel rod bundle with spacer grids (four sets of mixing vane grids and four sets of simple support grids without mixing vanes). We calculated and analyzed 11 experimental conditions under different pressure, inlet temperature, and mass flux. After comparing the CHF and the location of departure from the nucleate boiling obtained by the numerical simulation with the experimental results, we confirmed the reliability of computational fluid dynamic analysis for the prediction of the CHF of the rod bundle and the boiling characteristics of the two-phase flow. Subsequently, we analyzed the influence of the spacer grid and mixing vanes on the void fraction, liquid temperature, and secondary flow distribution. The research in this article provides theoretical support for the design of fuel assemblies.

• Journal of the Institute of Electronics Engineers of Korea SD
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• v.42 no.11
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• pp.23-32
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• 2005

The automatic generation of circuit diagrams has been practically used in the HDL based design for decades. Nevertheless, the diagrams became too complicated for the designers to identify the signal flows in the RTL and system level designs. In this paper, we propose four techniques to enhance the roadability of the complicated diagrams. They include i) the transformation of repetitive instances and terminals into vector forms, ii) an improved loop breaking algorithm, iii) a flat tap which simplifies the two level bus ripping structure that is required for the connection of a bundle net to multiple buses, and iv) the identification of block strings, and alignment of the corresponding blocks. Towards validating the proposed techniques, the diagrams of an industrial strength design m generated. The complexity of the diagrams has been reduced by up to $90\%$ in terms of the number of wires, the aggregate wire length, and the area.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1109-1119
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• 2016

Although, the direct use of spent pressurized water reactor (PWR) fuel in CANda Deuterium Uranium (CANDU) reactors (DUPIC) cycle is still under investigation, DUPIC cycle is a promising method for uranium utilization improvement, for reduction of high level nuclear waste, and for high degree of proliferation resistance. This paper focuses on the effect of DUPIC cycle on CANDU reactor safety parameters. MCNP6 was used for lattice cell simulation of a typical 3,411 MWth PWR fueled by $UO_2$ enriched to 4.5w/o U-235 to calculate the spent fuel inventories after a burnup of 51.7 MWd/kgU. The code was also used to simulate the lattice cell of CANDU-6 reactor fueled with spent fuel after its fabrication into the standard 37-element fuel bundle. It is assumed a 5-year cooling time between the spent fuel discharges from the PWR to the loading into the CANDU-6. The simulation was carried out to calculate the burnup and the effect of DUPIC fuel on: (1) the power distribution amongst the fuel elements of the bundle; (2) the coolant void reactivity; and (3) the reactor point-kinetics parameters.