• Journal of Korean Society for Geospatial Information Science
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• v.13 no.1 s.31
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• pp.3-10
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• 2005

This study attempts to measure effectively three dimensional deformation in road slopes using digital close-range photographs. After the still video images were acquired respectively on the same multi-station geometric configuration in two epoches, photo-triangulation was rallied out respectively by conventional standard bundle adjustment and free-net bundle adjustment and the computed results were compared with those of geodetic method by total station. Three dimensional coordinates and deformation amounts were able to be derived with the RMSE of sub-millimeter and the relative accuracy of $1/30,000{\sim}1/35,000$. It was shown that free-net bundle adjustment is about 13% higher than standard bundle adjustment in the accuracy of photo-triangulation. It was ascertained that the free-net technique is able to promote fast and accurate deformation surveying without the necessity of geodetic control survey in complicated industrial sites.

• 한국지형공간정보학회:학술대회논문집
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• 2004.10a
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• pp.155-160
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• 2004

근거리 디지털 영상을 이용하여 도로 사면의 3차원 변형을 효율적으로 측정하고자 하였다. 광속조정법(bundle adjustment)에 근거한 사진삼각측량 방법을 적용함에 있어서 종래의 표준적인 기법과 Free-net 기법으로 각각 처리하여 사면의 약 1/35000의 정확도로 3차원 측정과 변형량을 도출하였으며, 토털스테이션에 의해 측정한 결과와 비교하였다. 연구를 통하여 Free-net 기법을 적용함으로써 복잡한 산업현장에서 기존의 측지학적 기준점측량을 행하지 않고도 신속하고 정확한 측정을 기대할 수 있음을 확인할 수 있었다.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.502-507
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• 1998

Two different types the CANDU fuel bundles hue been modeled for the ASSERT-IV code subchannel analysis. From calculated values of mixture enthalpy and void paction distributions in the fuel bundles, it is found that net buoyancy effect is pronounced in the central region of the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. It is also found that the central region of the DUPIC fuel bundle can be cooled more efficiently than that of the standard fuel bundle. From calculated mixture enthalpy distribution at the exit of fuel channel, it is found that the mixture enthalpy and void fraction can be highest in the peripheral region of the DUPIC fuel bundle. On the other hand, the enthalpy and the void fraction were found to be highest in the central region of the standard CANDU fuel bundle at the exit of the fuel channel. This study shows that the subchannel analysis is very useful assessing thermal behavior of the fuel bundle that could be used in CANDU reactors.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.410-418
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• 2024

Study of flow behavior within rod bundles has been an active topic. Surface modification technologies are important parts of the design of the fourth generation reactor, which can increase the strength of the secondary flow within the rod bundle lattices. Quasi-periodic large-scale vortex structure (QLVS) is introduced by arranging micro ribs on the surface of rod bundles, which enhanced the scale of the secondary flow between the rod bundle lattices. Using computational fluid dynamics (CFD) and water experiments, the flow field distribution and drag coefficient of the rod-bundle lattices are studied. The secondary flow between the micro-ribbed rod-bundle lattice is significantly enhanced compared to the standard rod-bundle lattice. The numerical simulation results agree well with the experimental results.

• Nuclear Engineering and Technology
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• v.41 no.7
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• pp.969-978
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• 2009

The ASSERT-PV subchannel code developed by AECL has been applied as a design-assist tool to the advanced $CANDU^{(R)1}$ reactor fuel bundle. Based primarily on the $CANFLEX^{(R)2}$ fuel bundle, several geometry changes (such as element sizes and pitch-circle diameters of various element rings) were examined to optimize the dryout power and pressure-drop performances of the new fuel bundle. An experiment was performed to obtain dryout power measurements for verification of the ASSERT-PV code predictions. It was carried out using an electrically heated, Refrigerant-134a cooled, fuel bundle string simulator. The axial power profile of the simulator was uniform, while the radial power profile of the element rings was varied simulating profiles in bundles with various fuel compositions and burn-ups. Dryout power measurements are predicted closely using the ASSERT-PV code, particularly at low flows and low pressures, but are overpredicted at high flows and high pressures. The majority of data shows that dryout powers are underpredicted at low inlet-fluid temperatures but overpredicted at high inlet-fluid temperatures.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3625-3634
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• 2021

The crossflow is the key phenomenon in turbulent flow inside rod bundles. In order to establish confidence on application of computational fluid dynamics (CFD) to simulate the crossflow in rod bundles, three Reynolds-Averaged Navier Stokes (RANS) models i.e. the realizable k-ε model, the k-ω SST model and the Reynolds stress model (RSM), and the Large Eddy simulations (LES) with the Wall-Adapting Local Eddy-viscosity (WALE) model are validated based on the Particle Image Velocimetry (PIV) flow measurement experiment in a 5 × 5 rod bundle. In order to investigate effects of periodic boundary condition in the gap, the numerical results obtained with four inner subchannels are compared with that obtained with the whole 5 × 5 rod bundle. The results show that periodic boundaries in the gaps produce strong errors far downstream of the spacer grid, and therefore the full 5 × 5 rod bundle should be simulated. Furthermore, it can be concluded, that the realizable k-ε model can only provide reasonable results very close to the spacer grid, while the other investigated models are in good agreement with the experimental data in the whole downstream flow in the rod bundle. The LES approach shows superiority to the RANS models.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.439-445
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• 2021

In view of the complex geometric structure of the rod bundle channel and the limitation of the current CHF visualization experiment technology, it is very difficult to obtain the rod bundle CHF mechanism directly through the phenomenon of the rod bundle CHF visualization experiment. In order to obtain the applicable CHF mechanism assumption for rod bundle channel, firstly, five most representative DNB type round tube CHF mechanistic models are obtained with evaluation and screening. Then these original round tube CHF mechanistic models based on inlet conditions are converted to local conditions and coupled with subchannel analysis code ATHAS. Based on 5 × 5 full-length rod bundle CHF experimental data independently developed by Nuclear Power Institute of China (NPIC), the applicability research of each model for CHF prediction performance in rod bundle channel is carried out, and the commonness and difference of each model are comparatively studied. The CHF mechanism assumption of superheated liquid layer depletion that is most likely to be applicable for the rod bundle channel is selected and two directions that need to be improved are given. This study provides a reference for the development of CHF mechanistic model in rod bundle channel.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1314-1321
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• 2019

In case of multiple failure scenario, such as LOCA with ECCS failure, the decay heat continues to raise the reactor core temperature, eventually leading to the core voiding. In such scenario the convective heat transfer becomes poor and the majority of the heat transfer from fuel bundle takes place by radiation mode. During this abnormal working condition, if the channel pressure is less than 1 MPa, the PT sags and come in contact with the CT. This results in high rate of heat transfer from contact location to moderator. The present paper aims to capture the temperature profile over a simulated nuclear channel during such scenario at a steady state temperature of $600^{\circ}C$ (Centre pin) at two different configurations of PT i.e. PT concentric with CT and PT contact with CT. The results showed that the bottom nodes of all the components (Fuel bundle, PT and CT) of the simulated channel was greatly influenced by the PT/CT contact. Moreover, higher temperature were observed at top nodes of the PT and outer pins of the fuel bundle. However, no significant variation in temperatures were obtained in fuel bundle and CT in concentric condition.

• Nuclear Engineering and Technology
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• v.46 no.6
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• pp.875-882
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• 2014

In terms of safety and the efficient management of spent fuel storage, detecting failed fuel is one of the most important tasks in a CANada Deuterium Uranium (CANDU) reactor operation. It has been successfully demonstrated that in a CANDU reactor, on-power failed fuel detection and location systems, along with alarm area gamma monitors, can detect and locate defective and suspect fuel bundles before discharging them from the reactor to the spent fuel storage bay. In the reception bay, however, only visual inspection has been used to identify suspect bundles. Gaseous fission product and delayed neutron monitoring systems cannot precisely distinguish failed fuel elements from each fuel bundle. This study reports the use of a sipping system in a CANDU reactor for the integrity assessment of spent fuel bundles. The integrity assessment of spent fuel bundles using this sipping system has shown promise as a nondestructive test for detecting a defective fuel bundle in a CANDU reactor.

• Nuclear Engineering and Technology
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• v.52 no.11
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• pp.2479-2490
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• 2020

We performed experiments to measure a single-phase upward flow in a 5 × 5 rod bundle with spacer grids using a particle image velocimetry, focusing on the crossflow. The Reynolds number based on the hydraulic diameter and the bulk velocity is 10,000. The ratio of pitch between rods and rod diameter is 1.4 and spacer grid is installed periodically. The turbulence in the rod bundle results from the combination of a forced mixing and natural mixing. The forced mixing by the spacer grid persists up to 10D_h from the spacer grid, while the natural mixing is attributed to the crossflow between adjacent subchannels. The combined effects contribute to a sinusoidal distribution of the time-averaged stream-wise velocity along the lateral direction, which is relatively weak right behind the spacer grid as well as in the gap. The streamwise and lateral turbulence intensities are stronger right behind the spacer grid and in the gap. Based on these findings, we newly defined a modified mixing coefficient as the ratio of the lateral turbulence intensity to the time-averaged streamwise velocity, which shows a spatial variation. Finally, we compared the developed model with the measured data, which shows a good agreement with each other.