• Proceedings of the Korean Society of Soil and Groundwater Environment Conference
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• 2003.04a
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• pp.430-433
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• 2003

Nitrogen pollution in urban and rural groundwater is a common problem and poses a major threat to drinking water supplies based on groundwater. In this work, the kinetics of nitrification-denitrification coupled reactions are modeled and new reaction modules for the RT3D code describing the fate and transport of nitrogen species, dissolved oxygen, dissolved organic carbon, and biomass are developed and tested. The proposed nitrogen transformations and transport model showed very good match with the results of other public codes.

• Coupled systems mechanics
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• v.8 no.5
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• pp.393-414
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• 2019

This paper presents seismic analysis of concrete gravity dams considering soil-structure-fluid interaction. Displacement based plane strain finite element formulation is considered for the dam and foundation domain whereas pressure based finite element formulation is considered for the reservoir domain. A direct coupling method has been adopted to obtain the interaction effects among the dam, foundation and reservoir domain to obtain the dynamic responses of the dam. An efficient absorbing boundary condition has been implemented at the truncation surfaces of the foundation and reservoir domains. A parametric study has been carried out considering each domain separately and collectively based on natural frequencies, crest displacement and stress at the neck level of the dam body. The combined frequency of the entire coupled system is very less than that of the each individual sub-system. The crest displacement and neck level stresses of the dam shows prominent enhancement when coupling effect is taken into consideration. These outcomes suggest that a complete coupled analysis is necessary to obtain the actual responses of the concrete gravity dam. The developed methodology can easily be implemented in finite element code for analyzing the coupled problem to obtain the desired responses of the individual subdomains.

• Wind and Structures
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• v.32 no.3
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• pp.267-281
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• 2021

Bending-twisting coupling induced in big composite wind turbine blades is one of the passive control mechanisms which is exploited to mitigate loads incurred due to deformation of the blades. In the present study, flutter characteristics of bend-twist coupled blades, designed for load alleviation in wind turbine systems, are investigated by time-domain analysis. For this purpose, a baseline full GFRP blade, a bend-twist coupled full GFRP blade, and a hybrid GFRP and CFRP bend-twist coupled blade is designed for load reduction purpose for a 5 MW wind turbine model that is set up in the wind turbine multi-body dynamic code PHATAS. For the study of flutter characteristics of the blades, an over-speed analysis of the wind turbine system is performed without using any blade control and applying slowly increasing wind velocity. A detailed procedure of obtaining the flutter wind and rotational speeds from the time responses of the rotational speed of the rotor, flapwise and torsional deformation of the blade tip, and angle of attack and lift coefficient of the tip section of the blade is explained. Results show that flutter wind and rotational speeds of bend-twist coupled blades are lower than the flutter wind and rotational speeds of the baseline blade mainly due to the kinematic coupling between the bending and torsional deformation in bend-twist coupled blades.

• The Journal of Korean Institute of Communications and Information Sciences
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• v.39A no.5
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• pp.215-219
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• 2014

In holographic data storage (HDS), a high transmission rate is accomplished through the use of a charge coupled device array for reading two-dimensional (2D) pixel image data. Although HDS has many advantages in terms of storage capacity and data transmission rates, it also features problems, such as 2D intersymbol interference (ISI) by neighboring pixels and interpage interference (IPI) by multiple images stored in the same holographic volume. Modulation codes can be used to remove these problems. We introduce a 2D 8/9 error-correcting modulation code. The proposed modulation code exploits the trellis-coded modulation scheme, and the code rate is larger (about 0.889) than that of the conventional 6/8 balanced modulation code (an increase of approximately 13.9%). The performance of the bit error rate (BER) with the proposed scheme was improved compared with that of the 6/8 balanced modulation code and the simple 8/9 code without the trellis scheme.

• Coupled systems mechanics
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• v.5 no.2
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• pp.145-156
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• 2016

An efficient and general numerical strategy for fluid-structure interaction problems is presented where either the fluid or the structure part are represented by nonlinear models. This partitioned strategy is implemented under the form of code coupling that allows to (re)-use previous made developments in a more general multi-physics context. This strategy and its numerical implementation is verified on classical fluid-structure interaction benchmarks, and then applied to the impact of tsunamis waves on submerged structures.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1277-1288
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• 2018

In this work, the reactor kinetics capability is used to compute the design safety parameters in a PWR due to complete loss of coolant flow during protected and unprotected accidents. A thermal-hydraulic code coupled with a point reactor kinetic model are used for these calculations; where kinetics parameters have been developed from the neutronic SRAC code to provide inputs to RELAP5-3D code to calculate parameters related to safety and guarantee that they meet the regulatory requirements. In RELAP5-3D the reactivity feedback is computed by both separable and tabular models. The results show the importance of the reactivity feedback on calculating the power which is the key parameter that controls the clad and fuel temperatures to maintain them below their melting point and therefore prevent core melt. In addition, extending modeling capability from separable to tabular model has nonremarkable influence on calculated safety parameters.

• Nuclear Engineering and Technology
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• v.33 no.2
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• pp.209-224
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• 2001

The SSC-K system analysis code is under development at the Korea Atomic Energy Research Institute (KAERI) as a part of the KALIMER project. The SSC-K code is being used as the principal tool for analyzing a variety of off-normal conditions or accidents of the preliminary KALIMER design. The SSC-K code features a multiple-channel core representation coupled with a point kinetics model with reactivity feedback. It provides a detailed, one-dimensional thermal-hydraulic simulation of the primary and secondary sodium coolant circuits, as well as the balance-of-plant steam/water circuit. Recently a two-dimensional hot pool model was incorporated into SSC-K for analysis of thermal stratification phenomena in the hot pool. In addition, SSC-K contains detailed models for the passive decay heat removal system and a generalized plant control system. The SSC-K code has also been applied to the computational engine for an interactive simulation of the KALIMER plant. This paper presents an overview of the recent activities concerned with SSC-K code model development This paper focuses on both descriptions of the newly adopted thermal hydraulic and neutronic models, and applications to KALIMER analyses for typical anticipated transients without scram.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.455-465
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• 2021

The Molten Salt Reactor (MSR), one of the six advanced reactor types of the 4th generation nuclear energy systems, has many impressive features including economic advantages, inherent safety and nuclear non-proliferation. This paper introduces a system analysis code named TREND, which is developed and used for the steady and transient simulation of MSRs. The TREND code calculates the distributions of pressure, velocity and temperature of single-phase flows by solving the conservation equations of mass, momentum and energy, along with a fluid state equation. Heat structures coupled with the fluid dynamics model is sufficient to meet the demands of modeling MSR system-level thermal-hydraulics. The core power is based on the point reactor neutron kinetics model calculated by the typical Runge-Kutta method. An incremental PID controller is inserted to adjust the operation behaviors. The verification and validation of the TREND code have been carried out in two aspects: detailed code-to-code comparison with established thermal-hydraulic system codes such as RELAP5, and validation with the experimental data from MSRE and the CIET facility (the University of California, Berkeley's Compact Integral Effects Test facility).The results indicate that TREND can be used in analyzing the transient behaviors of MSRs and will be improved by validating with more experimental results with the support of SINAP.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1789-1803
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• 2022

RAST-F is a new full-core analysis code based on the two-step approach that couples a multi-group cross-section generation Monte-Carlo code MCS and a multi-group nodal diffusion solver. To demonstrate the feasibility of using MCS/RAST-F for fast reactor analysis, this paper presents the coupled nodal code verification results for the MET-1000 and CAR-3600 benchmark cores. Three different multi-group cross-section calculation schemes are employed to improve the agreement between the nodal and reference solutions. The reference solution is obtained by the MCS code using continuous-energy nuclear data. Additionally, the MCS/RAST-F nodal solution is verified with results based on cross-section generated by collision probability code TULIP. A good agreement between MCS/RAST-F and reference solution is observed with less than 120 pcm discrepancy in k_eff and less than 1.2% root-mean-square error in power distribution. This study confirms the two-step approach MCS/RAST-F as a reliable tool for the three-dimensional simulation of reactor cores with fast spectrum.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3388-3400
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• 2023

Detailed analysis of the neutron pathway through matter inside the nuclear reactor core is exceedingly needed for safety and economic considerations. Due to the constant development of high-performance computing technologies, neutronics analysis using computer codes became more effective and efficient to perform sophisticated neutronics calculations. In this work, a commercial pressurized water reactor (PWR) presented by Virtual Environment for Reactor Applications (VERA) Core Physics Benchmark are modeled and simulated using a high-fidelity simulation of OpenMC code in terms of criticality and fuel pin power distribution. Various problems have been selected from VERA benchmark ranging from a simple two-dimension (2D) pin cell problem to a complex three dimension (3D) full core problem. The development of the code capabilities for reactor physics methods has been implemented to investigate the accuracy and performance of the OpenMC code against VERA SCALE codes. The results of OpenMC code exhibit excellent agreement with VERA results with maximum Root Mean Square Error (RMSE) values of less than 0.04% and 1.3% for the criticality eigenvalues and pin power distributions, respectively. This demonstrates the successful utilization of the OpenMC code as a simulation tool for a whole core analysis. Further works are undergoing on the accuracy of OpenMC simulations for the impact of different fuel types and burnup levels and the analysis of the transient behavior and coupled thermal hydraulic feedback.