• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.7 no.4
• /
• pp.51-56
• /
• 2011

Steam generator is one of the main equipments that affect safety and long term operation in nuclear power plants. Fluid flows inside and outside of the steam generator tubes and induces vibration. To prevent the vibration the tubes are supported by AVB (anti vibration bar). When the steam generator tube contact to AVB, it is damaged by the accumulation of wear and corrosion. Therefore studies are required to determine the effects of the gap between the steam generator tube and AVB. In order to obtain the stress and the displacement distributions of the steam generator tube, three dimensional finite element analyses were performed by using the commercial program ANSYS. Using the calculated the stress and the displacement distributions, the static residual strength of the steam generator tube can be evaluated. The results show that the stress and displacement of the steam generator tube increase significantly compared with the results from a zero-gap model.

• Nuclear Engineering and Technology
• /
• v.36 no.3
• /
• pp.248-262
• /
• 2004

Turbulence models are separately assessed for a three dimensional thermal-hydraulic analysis of the integral reactor SMART. Seven models (mixing length, k-l, standard $k-{\epsilon},\;k-{\epsilon}-f{\mu},\;k-{\epsilon}-v2$, RRSM, and ERRSM) are investigated for flat plate channel flow, rotating channel flow, and square sectioned U-bend duct flow. The results of these models are compared to the DNS data and experiment data. The results are assessed in terms of many aspects such as economical efficiency, accuracy, theorization, and applicability. The standard $k-{\epsilon}$ model (high Reynolds model), the $k-{\epsilon}-v2$ model, and the ERRSM (low Reynolds models) are selected from the assessment results. The standard $k-{\epsilon}$ model using small grid numbers predicts the channel flow with higher accuracy in comparison with the other eddy viscosity models in the logarithmic layer. The elliptic-relaxation type models, $k-{\epsilon}-v2$, and ERRSM have the advantage of application to complex geometries and show good prediction for near wall flows.

• Nuclear Engineering and Technology
• /
• v.36 no.6
• /
• pp.559-570
• /
• 2004

A computational fluid dynamics model for predicting moderator circulation inside the Canada deuterium uranium (CANDU) reactor vessel has been developed to estimate the local subcooling of the moderator in the vicinity of the calandria tubes. The buoyancy effect induced by the internal heating is accounted for by the Boussinesq approximation. The standard $k-{\varepsilon}$ turbulence model with logarithmic wall treatment is applied to predict the turbulent jet flows from the inlet nozzles. The matrix of the calandria tubes in the core region is simplified to a porous media in which the anisotropic hydraulic impedance is modeled using an empirical correlation of pressure loss. The governing equations are solved by DFX-4.4, a commercial CFD code developed by AEA technology. The resultant flow patterns of the constant-z slices containing the inlet nozzles and the outlet port are "mined-type", as observed in the former 2-dimensional experimental investigations. With 103% full power for conservatism, the maximum temperature of the moderator is $82.9^{\circ}C$ at the top of the core region. Considering the hydrostatic pressure change, the minimum subcooling is $24.8^{\circ}C$.

• Proceedings of the KIEE Conference
• /
• 2007.11a
• /
• pp.210-211
• /
• 2007

Once high current flows through the vacuum interrupter, eddy current occurs due to the time-varying axial magnetic field caused by the current(AC) and it causes a decrease in axial magnetic field generated by current flowing through coil electrode. but if there are slots on contact electrode it is possible to increase the amplitude of axial magnetic field by reducing the influence of eddy current. there has been many studies about the number of slot of the contact electrode[1][2][3]. In this paper, in addition to these previous results we deal with the influence of the length and width of the slots on axial magnetic field at the mid-gap plane in 4 segment coil type vacuum interrupter by using 3D finite element method software.

• Nuclear Engineering and Technology
• /
• v.49 no.3
• /
• pp.466-475
• /
• 2017

Coolant mixing under natural circulation flow regime constitutes a key parameter that may play a role in the course of an accidental transient in a nuclear pressurized water reactor. This issue has motivated some experimental investigations carried out within the OECD/NEA PKL projects. The aim was to assess the coolant mixing phenomenon in the reactor pressure vessel downcomer and the core lower plenum under several asymmetric steady and unsteady flow conditions, and to provide experimental data for code validations. Former studies addressed the mixing phenomenon using, on the one hand, one-dimensional computational approaches with cross flows that are not fully validated under transient conditions and, on the other hand, expensive computational fluid dynamic tools that are not always justified for large-scale macroscopic phenomena. In the current framework, an unsteady coolant mixing experiment carried out in the Rossendorf coolant mixing test facility is simulated using the three-dimensional porous media capabilities of the thermal-hydraulic system CATHARE code. The current study allows highlighting the current capabilities of these codes and their suitability for reproducing the main phenomena occurring during asymmetric transient natural circulation mixing conditions.

• Journal of the Korean Society of Visualization
• /
• v.19 no.3
• /
• pp.92-98
• /
• 2021

Large Eddy Simulation (LES) has been popularly applied and used in the last several decades to simulate turbulent boundary layer in the numerical domain. A fully developed turbulent boundary layer has also been applied to predict the complicated wake flow behind bluff bodies. In this study we aimed to generate an artificial turbulent boundary layer, which is based on an exponential correlation function, and generates a series of realistic three-dimensional velocity data in two-dimensional inlet section which are correlated both in space and in time. The results suggest its excellent capability for high Reynolds number flows. To make an effective generation, a hexahedral mesh has been used and Cholesky decomposition was applied to possess suitable turbulent statistics such as the randomness and correlation of turbulent flow. As a result, the flow characteristics in the domain and fluctuating pressure near the wall are very close to those of fully developed turbulent boundary layers.

• Nuclear Engineering and Technology
• /
• v.53 no.3
• /
• pp.752-762
• /
• 2021

The Smoothed Particle Hydrodynamics is one of the most widely used mesh-free numerical method for thermo-fluid dynamics. Due to its Lagrangian nature and simplicity, it is recently gaining popularity in simulating complex physics with large deformations. In this study, the 3D single/two-phase numerical simulations are performed on the Liquid Metal Reactor (LMR) centralized sloshing benchmark experiment using the SPH parallelized using a GPU. In order to capture multi-phase flows with a large density ratio more effectively, the original SPH density and continuity equations are re-formulated in terms of the normalized-density. Based upon this approach, maximum sloshing height and arrival time in various experimental cases are calculated by using both single-phase and multi-phase SPH framework and the results are compared with the benchmark results. Overall, the results of SPH simulations show excellent agreement with all the benchmark experiments both in qualitative and quantitative manners. According to the sensitivity study of the particle-size, the prediction accuracy is gradually increasing with decreasing the particle-size leading to a higher resolution. In addition, it is found that the multi-phase SPH model considering both liquid and air provides a better prediction on the experimental results and the reality.

• Journal of Energy Engineering
• /
• v.13 no.1
• /
• pp.51-59
• /
• 2004

The NSSS (Nuclear Steam Supply System) thermal-hydraulic programs adopted in the domestic full-scope power plant simulators were provided in early 1980s by foreign vendors. Because of limited compulsational capability at that time, they usually used very simplified physical models for a real-time simulation of NSSS thermal-hydraulic transients, which entails inaccurate results and, thus, the possibility of so-called "negative training", especially for complicated two-phase flows in the reactor coolant system. In resolve the problem, KEPRI developed a realistic NSSS T/H program ARTS which was based on the RETRAN-3D code for the improvement of the Nuclear Power Plant full-scope simulator. The ARTS (based on the RETRAN-3D code) guarantees the real-time calculations of almost all transients and ensures the robustness of simulations. However, there is some possibility of failing to calculate in the case of large break loss of coolant accident (LBLOCA) and low-pressure low-flow transient. In this case, the backup calculation system cover automatically the ARTS. The backup calculation system was expected to provide substantially more accurate predictions in the analysis of the system transients involving LBLOCA. The results were reasonable in terms of accuracy, real-time simulation, robustness and education of operators, complying with FSAR and the AMSI/ANS-3.5-1998 simulator software performance criteria.

• Nuclear Engineering and Technology
• /
• v.45 no.5
• /
• pp.613-624
• /
• 2013

Multi-dimensional two-phase phenomena occur in many industrial applications, particularly in a nuclear reactor during steady operation or a transient period. Appropriate modeling of complicated behavior induced by a multi-dimensional flow is important for the reactor safety analysis results. SPACE, a safety analysis code for thermal hydraulic systems which is currently being developed, was designed to have the capacity of multi-dimensional two-phase thermo-dynamic phenomena induced in the various phases of a nuclear system. To validate the performance of SPACE, a two-dimensional two-phase flow test was performed with slab geometry of the test section having a scale of $1.43m{\times}1.43m{\times}0.11m$. The test section has three inlet and three outlet nozzles on the bottom and top gap walls, respectively, and two outlet nozzles installed directly on the surface of the slab. Various kinds of two-dimensional air/water flows were simulated by selecting combinations of the inlet and outlet nozzles. In this study, two-dimensional two-phase void fraction profiles were quantified by measuring the local gap impedance at 225 points. The flow conditions cover various flow regimes by controlling the flow rate at the inlet boundary. For each selected inlet and outlet nozzle combination, the water flow rate ranged from 2 to 20 kg/s, and the air flow rate ranged from 2.0 to 20 g/s, which corresponds to 0.4 to 4 m/s and 0.2 to 2.3 m/s of the superficial liquid and gas velocities based on the inlet port area, respectively.

• Journal of Korea Water Resources Association
• /
• v.30 no.1
• /
• pp.23-34
• /
• 1997

The Nanji-Do ladnfill is an ill-conditioned reclaimed land without pollution intercepting facilities, and has high ground water table and deep stratum. The purpose of this study is to analyze the solute transport in steady-state groundwater flow and to predict the solute dispersion in Nanji-Do landfill using HST-3D model. As results, the groundwater flows radially outward from the center of No. 1 and No. 2 landfills, and large amount of runoff is moved into Han River. The predicted relative concentration of total dissolved solute(TDS) at two years later was 0.25 in the weathering zone, 0.26 in the lower alluvium, and 0.28 in the upper alluvium. Thus, the further pollution to bottom rock and Han River was predicted by comparing the corresponding present values of 0.29, 0.32 and 0.35.