• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1995.05a
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• pp.113-116
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• 1995

Simplified approach has been adopted for the prediction of the thermal behavior of CANDU reactor core during power transients. Based on the assumption that the ratio of mass flow rate for each core channel does not vary during the transient, quasy-steady state analysis technique is applied with predicted core inlet boundary conditions(total mass flow rate and specific enthalpy). For restricted transient case, the presented method shows functionally reasonable estimation of core thermal behavior which could be implemented in the fast running reactor simulation program.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.77-82
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• 1997

The CANDU-6 transition core refuelled from 37-element fuel to CANFLEX-NU fuel has been evaluated by an 100full power day time-dependent fuel-management simulation to find the core compatibility with the CANFLEX fuel loading. The simulation calculations for the transition core were carried out with the RFSP code, provided by the cell averaged fuel properties obtained from the POWDERPUFS-V code. The simulation results were compared with those of the current 37-element fuel loading only. The results show that the CANFLEX-NU fuel bundles will be compatible with the CANDU-6 reactor because the core physics characteristics of CANFLEX-NU fuel are very similar to those of the 37-element fuel bundle.

• Proceedings of the KIPE Conference
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• 2015.11a
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• pp.9-10
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• 2015

EI core inductor in power electronic circuit simulation is usually assumed as linear by using matrix model. However, nonlinear magnetic characteristics such as B-H characteristic are also important for the accurate simulation of the circuit behavior. To model nonlinear magnetic characteristics of EI core inductor with only DC bias table, this paper presents a method in PLECS simulation tool which is a commercially available simulation tool for power electronics circuit analysis. Comparing with ideal matrix model, the simplification and accuracy are improved by this modeling method. Also, compared to analysis by FEM, it is much simpler, faster and easier to simulate with power electronics circuit. Validation of the proposed model was verified by simulation and experiment results.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.62 no.2
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• pp.252-256
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• 2013

In order to overcome the complexity and performance limit problems of superscalar processors, the multi-core architecture has been prevalent recently. Usually, the number of cores mostly used for the multi-core processor architecture ranges from 2 to 16. However in the near future, more than 32-cores are likely to be utilized, which is called as many-core processor architecture. Using SPEC 2000 benchmarks as input, the trace-driven simulation has been performed for the 32 to 1024 many-core architectures extensively. For 1024-cores, the average performance scores 15.7 IPC, but the performance increase rate is saturated.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.3491-3496
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• 2007

If a molten core is released from a reactor vessel into a reactor cavity during a severe accident, an important safety issue of coolability of the molten core from top-flooding and concrete ablation due to a molten core concrete interaction (MCCI) is still unresolved. The released molten core debris would attack the concrete wall and basemat of the reactor cavity, which will lead to inevitable concrete decompositions and possible radiological releases. In a OECD/MCCI project scheduled for 4 years from 2002. 1 to 2005. 12, a series of tests were performed to secure the data for cooling the molten core spread out at the reactor cavity and for the 2-D long-term core concrete interaction (CCI). The tests included not only separate effect tests such as a melt eruption, water ingression, and crust failure tests with a prototypic material but also 2-D CCI tests with a prototypic material under dry and flooded cavity conditions. The paper deals with the transient simulations on the CCI-2 test by using a severe accident analysis code, CORQUENCH, which was developed at Argonne National Laboratory (ANL). Similar simulations had been already per for me d by using MELCOR 1.8.5 code. Unlike the MELCOR 1.8.5, the CORQUENCH includes a melt eruption mode I and a newly developed water ingression model based on the water ingression tests under the OECD/MCCI project. In order to adjust the geometrical differences between the CCI-2 test (rectangular geometry) and the simulations (cylindrical geometry), the same scaling methodology as used in the MELCOR simulation was applied. For the direct comparison of the simulation results, the same inputs for the MELCOR simulation were used. The simulation results were compared with the previous results by using MELCOR 1.8.5.

• Geomechanics and Engineering
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• v.10 no.6
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• pp.709-726
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• 2016

Corrugated-core sandwich panels are prevalent for many applications in industries. The researches performed with the aim of optimization of such structures in the literature have considered a deterministic approach. However, it is believed that deterministic optimum points may lead to high-risk designs instead of optimum ones. In this paper, an effort has been made to provide a reliable and robust design of corrugated-core sandwich structures through stochastic and probabilistic multi-objective optimization approach. The optimization is performed using a coupling between genetic algorithm (GA), Monte Carlo simulation (MCS) and finite element method (FEM). To this aim, Prob. Design module in ANSYS is employed and using a coupling between optimization codes in MATLAB and ANSYS, a connection has been made between numerical results and optimization process. Results in both cases of deterministic and probabilistic multi-objective optimizations are illustrated and compared together to gain a better understanding of the best sandwich panel design by taking into account reliability and robustness. Comparison of results with a similar deterministic optimization study demonstrated better reliability and robustness of optimum point of this study.

• Journal of KIISE
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• v.44 no.6
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• pp.566-572
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• 2017

As NAND flash storage is being widely used, its simulation methodologies have been studied in various aspects such as performance, reliability, and endurance. As a result, there have been advances in NAND flash storage simulation for both functional modeling and timing modeling. However, in addition to these advances, there is a need to drastically reduce the long simulation time that is required to evaluate the aging effect on flash storage. This paper proposes a so-called multi-core scalable real-time flash storage simulation method, which can control the simulation speed according to the user's preference. According to this method, it is possible to speed up the simulation in proportion to the number of CPU cores arbitrarily given while guaranteeing the correctness of the simulation result. Using our simulator implemented in the form of the Linux kernel module, we demonstrate the multi-core scalability and correctness of the proposed method.

• Nuclear Engineering and Technology
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• v.37 no.1
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• pp.79-90
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• 2005

In this paper are reviewed the current status of nuclear fuel management and reactor operational aid tools. In addition, we indicate deficiencies in current capabilities and what future research is judged warranted. For the nuclear fuel management review the focus is on light water reactors and the utilization of stochastic optimization methods applied to the lattice, fuel bundle, core loading pattern, and for BWRs the control rod pattern/core flow design decision making problems. Significant progress in addressing separately each of these design problems on a single cycle basis is noted; however, the outstanding challenge of addressing the integrated design problem over multiple cycles under conditions of uncertainty remains to be addressed. For the reactor operational aid tools review the focus is on core simulators, used to both process core instrumentation signals and as an operator aid to predict future core behaviors under various operational strategies. After briefly reviewing the current status of capabilities, a more in depth review of adaptive core simulation capabilities, where core simulator input data are adjusted within their known uncertainties to improved agreement between prediction and measurement, is presented. This is done in support of the belief that further development of adaptive core simulation capabilities is required to further significantly advance the utility of core simulators in support of reactor operational aid tools.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.152-157
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• 1997

To test the viability of CANFLEX-SEU bundles in an existing CANDU 6 reactor, core follow-up simulation has been carried out using the reactor fueling simulation program of the CANDU 6, RFSP computer code, and a lattice physics code, WIMS-AECL. During the core follow-up, bundle and channel powers and zone levels have been checked against their operating limits at each simulation. It is observed from the simulation results that an equilibrium core loaded with 0.9 w/o CANFLEX-SEU bundles could be refueled ,and maintained for 550 FPD without any significant violations in the channel and bundle power limits and the permissible operating range of the liquid zone controllers.

• 전기의세계
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• v.13 no.4
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• pp.16-24
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• 1964

The dynamic equations of the void fraction and the water velocity in boiling region of the BWR reactor core are derived. And these equations are approximated to be able to set on an PACE analog computor. The transient analysis and the frequency response obtained by analog computer are compared with other by digital computor.