• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1218-1230
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• 2019

A sub-channel solver, named ${\underline{S}}teady$ and ${\underline{T}}ransient$ ${\underline{A}}nalyzer$ for ${\underline{R}}eactor$ ${\underline{T}}hermal$ hydraulics (START), has been developed using the homogenous model for two-phase conditions of light water reactors. The code is developed as a fast and accurate TH-solver for coupled and multi-physics calculations. START has been validated against the NUPEC PWR Sub-channel and Bundle Test (PSBT) database. Tests like single-channel quality and void-fraction for steady state, outlet fluid temperature for steady state, rod-bundle quality and void-fraction for both steady state and transient conditions have been analyzed and compared with experimental values. Results reveal a good accuracy of solution for both steady state and transient scenarios. Axially different values for turbulent mixing coefficient are used based on different grid-spacer types. This provides better results as compared to using a single value of turbulent mixing coefficient. Code-to-code evaluation of PSBT results by the START code compares well with other industrial codes. The START code has been parallelized with the OpenMP algorithm and its numerical performance is evaluated with a large whole PWR core. Scaling study of START shows a good parallel performance.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4597-4606
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• 2023

Thermal analysis of the CANDU spent fuel dry storage canister is very important to ensure the integrity of the spent fuel. The analyses have been conducted using a conservative approach, with a particular focus on the peak cladding temperature (PCT) of the fuel rods in the canister. In this study, we have performed a realistic thermal analysis using a computational fluid dynamics (CFD) code. The canister contains 9 fuel bundle baskets. A detailed analysis of even a single basket requires significant computational resources. To overcome this challenge, we replaced each basket with an equivalent heat conductor (EHC), of which effective thermal conductivity (ETC) is developed from the results of detailed CFD calculations of a fuel bundle basket. Then, we investigated the effects of some conservative models, ultimately aiming at a realistic analysis. The results revealed: (i) The influence of convective heat transfer in the basket cannot be ignored, but it's less significant than expected. (ii) Modeling of the lifting rod is crucial, as it plays a decisive role in axial heat transfer at the center of the canister and significantly reduces the PCT. (iii) Convection within the canister is very important, as it not only reduces the PCT but also shifts its location upwards.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4561-4569
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• 2023

A correct understanding of vibration-based degradation is crucial from the standpoint of maintenance for Steam Generators (SG) as crucial mechanical equipment in nuclear power plants. This study has established a novel approach to developing a model for investigating tube bundle degradation according to crack growth caused by two-phase Flow-Induced Vibration (FIV). An important step in the approach is to calculate the two-phase flow field parameters between the SG tube bundles in various zones using the porous media model to determine the velocity and vapor volume fraction. Afterward, to determine the vibration properties of the tube bundles, the Fluid-Solid Interaction (FSI) analysis is performed in eighteen thermal-hydraulic zones. Tube bundle degradation based on crack growth using the sixteen most probable initial cracks and within each SG thermal-hydraulic zone is performed to calculate useful lifetime. Large Eddy Simulation (LES) model, Paris law, and Wiener process model are considered to model the turbulent crossflow around the tube bundles, simulation of elliptical crack growth due to the vibration characteristics, and estimation of SG tube bundles degradation, respectively. The analysis shows that the tube deforms most noticeably in the zone with the highest velocity. As a result, cracks propagate more quickly in the tube with a higher height. In all simulations based on different initial crack sizes, it was observed that zone 16 experiences the greatest deformation and, subsequently, the fastest degradation, with a velocity and vapor volume fraction of 0.5 m/s and 0.4, respectively.

• Nuclear Engineering and Technology
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• v.51 no.8
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• pp.1905-1915
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• 2019

This paper investigated the influence of simple support girds on flow, irrespective of having mixing vanes, in a 1 × 3 array rod bundle by using CFD methodology and the most accurate turbulence model which could reflect the actual physics of the flow was determined. In this context, a CFD model was created simulating the experimental studies on a single-phase flow [1] and the results were compared with the experimental data. In the first part of the study, influence of mesh was examined. Tetra, hybrid and poly type meshes were analyzed and convergence study was carried out on each in order to determine the most appropriate type and density. k - ε Standard and RSM LPS turbulence models were used in this section. In the second part of the study, the most appropriate turbulence model that could reflect the physics of the actual flow was investigated. RANS based turbulence models were examined using the mesh that was determined in the first part. Velocity and turbulence intensity results obtained on the upstream and downstream of the spacer grid at -3d_h, +3d_h and +40d_h locations were compared with the experimental data. In the last section of the study, the behavior of flow through the spacer grid was examined and its prominent aspects were highlighted on the most appropriate turbulence model determined in the second part. Results of the study revealed the importance of mesh type. Hybrid mesh having the largest number of structured elements performed remarkably better than the other two on results. While comparisons of numerical and experimental results showed an overall agreement within all turbulence models, RSM LPS presented better results than the others. Lastly, physical appearance of the flow through spacer grids revealed that springs has more influence on flow than dimples and induces transient flow behaviors. As a result, flow through a simple support grid was examined and the most appropriate turbulence model reflecting the actual physics of the flow was determined.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.36-48
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• 2022

As a valid numerical method to obtain a high-resolution result of a flow field, computational fluid dynamics (CFD) have been widely used to study coolant flow and heat transfer characteristics in fuel rod bundles. However, the time-consuming, iterative calculation of Navier-Stokes equations makes CFD unsuitable for the scenarios that require efficient simulation such as sensitivity analysis and uncertainty quantification. To solve this problem, a reduced-order model (ROM) based on proper orthogonal decomposition (POD) and machine learning (ML) is proposed to simulate the flow field efficiently. Firstly, a validated CFD model to output the flow field data set of the rod bundle is established. Secondly, based on the POD method, the modes and corresponding coefficients of the flow field were extracted. Then, an deep feed-forward neural network, due to its efficiency in approximating arbitrary functions and its ability to handle high-dimensional and strong nonlinear problems, is selected to build a model that maps the non-linear relationship between the mode coefficients and the boundary conditions. A trained surrogate model for modes coefficients prediction is obtained after a certain number of training iterations. Finally, the flow field is reconstructed by combining the product of the POD basis and coefficients. Based on the test dataset, an evaluation of the ROM is carried out. The evaluation results show that the proposed POD-ROM accurately describe the flow status of the fluid field in rod bundles with high resolution in only a few milliseconds.

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

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.721-733
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• 2020

In nuclear engineering, the occurrence of critical heat flux (CHF) is complicated for rod bundle, and it is much more difficult to predict the CHF when it is in natural circulation under motion condition. In this paper, the dryout-type CHF is investigated for the rod bundle in a natural circulation loop under rolling motion condition based on the coupled analysis of subchannel method, a one-dimensional system analysis method and a CHF mechanism model, namely the three-fluid model for annular flow. In order to consider the rolling effect of the natural circulation loop, the subchannel model is connected to the one-dimensional system code at the inlet and outlet of the rod bundle. The subchannel analysis provides the local thermal hydraulic parameters as input for the CHF mechanism model to calculate the occurrence of CHF. The rolling motion is modeled by additional motion forces in the momentum equation. First, the calculation methods of the natural circulation and CHF are validated by a published natural circulation experiment data and a CHF empirical correlation, respectively. Then, the CHF of the rod bundle in a natural circulation loop under both the stationary and rolling motion condition is predicted and analyzed. According to the calculation results, CHF under stationary condition is smaller than that under rolling motion condition. Besides, the CHF decreases with the increase of the rolling period and angular acceleration amplitude within the range of inlet subcooling and mass flux adopted in the current research. This paper can provide useful information for the prediction of CHF in natural circulation under motion condition, which is important for the nuclear reactor design improvement and safety analysis.

• Proceedings of KOSOMES biannual meeting
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• 2003.05a
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• pp.135-139
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• 2003

This paper describes on the development of Passive-type Radar Reflector for Fisheries (PRR-F) based on the newly revised 2000 SOLAS regulations. The purpose of PRR-F is to provide it as the protection devices of a fishing net and a fishing field. The PRR-F is composed of corner cluster bundle of light galvanized iron, and it is inserted into floating styrofoam. Performance tests for the PRR-F are carried out in an anechoic chamber. The test results show that the reflected radar signal from PRR-F is large enough for the purpose.

• Korean Journal of Materials Research
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• v.13 no.11
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• pp.732-736
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• 2003

A CP-GFRP(Carbon Powder-Glass Fiber Reinforced Plastic) sensor was fabricated for fracture detection. The electric resistance of the sensor was measured on condition of various composition of carbon powders and thickness of bundle of glass fibers. The resistance of the sensor was decreased as the increase of the content of carbon powders and the TEX of the glass fibers. In the case of loading on CP-GFRP, because inner crack was propagated, the part of percolation structures was disconnected. These observations show the following results. The conduction of CP-GFRP sensor is due to percolation structure of carbon powders and increase of resistance is due to expansion of cracks.

• Animal Systematics, Evolution and Diversity
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• v.28 no.2
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• pp.137-139
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• 2012

An aquatic oligochaete species, $Ripistes$ $parasita$ (Schmidt, 1847) collected from Ganghwado Island in Korea, is described and illustrated. Specimens inhabiting on aquatic vegetation at the edge of the streams were collected with a plankton hand net. Morphological features of present specimens such as the number of long hair chaetae per bundle and their length in VI-VIII, and shape and size of ventral chaetae are concordant with those of previous reports on this species. The genus $Ripistes$ contains one species, $R.$ $parasite$ which is distributed over, Europe, North America and eastern part of Asia including China and Japan. This is the first record of $R.$ $parasita$ in Korea.