• Journal of computational fluids engineering
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• v.21 no.1
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• pp.19-29
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• 2016

A Prototype Gen-IV Sodium-cooled Fast Reactor which is one of the $4^{th}$ generation nuclear reactors is in development by Korea Atomic Energy Research Institute. The reactor is composed of four main fluid systems which are categorized by its functions, i.e., Primary Heat Transport System, Intermediate Heat Transport System, Decay Heat Removal System and Sodium-Water Reaction Pressure Relief System. The coolant of the reactor is liquid sodium and sodium-to-sodium heat exchangers are installed at the interfaces between two fluid systems, Intermediate Heat Exchangers between the Primary Heat Transport System and the Intermediate Heat Transport System and Decay Heat Exchangers between the Primary Heat Transport System and the Decay Heat Removal System. For the design and performance analysis of the Intermediate Heat Exchanger and the Decay Heat Exchanger, a computer code was written during previous step of research. In this work, the computer code named "SHXSA" has been validated preliminarily by computational fluid dynamics simulations.

• Proceedings of the KSME Conference
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• 2002.08a
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• pp.55-58
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• 2002

Parallel finite element code has been recently developed for the analysis of the incompressible Wavier-Stokes equations using domain decomposition method. Metis and MPI libraries are used for the domain partitioning of an unstructured mesh and the data communication between sub-domains, respectively. For unsteady computation of the incompressible Navier-Stokes equations, 4-step splitting method is combined with P1P1 finite element formulation. Smagorinsky and dynamic model are implemented for the simulation of turbulent flows. For the validation performance-estimation of the developed parallel code, three-dimensional Laplace equation has been solved. It has been found that the speed-up of 40 has been obtained from the present parallel code fir the bench mark problem. Lastly, the turbulent flows around the MIRA model and Tiburon model have been solved using 32 processors on IBM SMP cluster and unstructured mesh. The computed drag coefficient agrees better with the existing experiment as the mesh resolution of the region increases, where the variation of pressure is severe.

• Nuclear Engineering and Technology
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• v.31 no.4
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• pp.401-407
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• 1999

We have performed two experiments in the fast critical assembly BFS to measure the effective delayed neutron fraction $\beta$$_{eff}$ values and compared the results to validate the $\beta$$_{eff}$ computation code, BETA-K. Measurements of $\beta$$_{eff}$ were carried out in a metallic plutonium core and a metallic uranium core with Cf$^{252}$ source pseudo-reactivity method. Fission integrals and correction factors, which were used to obtain the experimental $\beta$$_{eff}$ values, were calculated by using the LMR core design computation code system of KAERI. BETA-K has been developed consistently with the hexagonal Nodal Expansion Method (NEM) and it used delayed neutron data of ENDF/B-VI. By comparing the computed $\beta$$_{eff}$ values with the measured ones, we found that the results from BETA-K agreed with the experimental values within the experimental error bound.ror bound.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.709-718
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• 2012

The pressurized water reactor APR1400 adopts DVI (Direct Vessel Injection) for the emergency cooling water in the upper downcomer annulus. The International Standard Problem number 50 (ISP-50) was launched with the aim to investigate thermal hydraulic phenomena during a 50% DVI line break scenario with best estimate codes making use of the experimental data available from the ATLAS facility located at KAERI. The present work describes the calculation results obtained for the ISP-50 using the RELAP5/MOD3.3 system code. The work aims at validation and assessment of the code to reproduce the observed phenomena and investigate about its limitations to predict complicated mixing phenomena between the subcooled emergency cooling water and the two-phase flow in the downcomer. The obtained results show that the overall trends of the main test variables are well reproduced by the calculations. In particular, the pressure in the primary system show excellent agreement with the experiment. The loop seal clearance phenomenon was observed in the calculation and it was found to have an important influence on the transient progression. Moreover, the collapsed water levels in the core are accurately reproduced in the simulations. However, the drop in the downcomer level before the activation of the DVI from safety injection tanks was underestimated due to multi-dimensional phenomena in the downcomer that are not properly captured by one-dimensional simulations.

• Journal of computational fluids engineering
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• v.11 no.1 s.32
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• pp.57-66
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• 2006

An optimal shape design approach is presented for a subsonic S-shaped intake using aerodynamic sensitivity analysis. Two-equation turbulence model is employed to capture strong counter vortices in the S-shaped duct more precisely. Sensitivity analysis is performed for the three-dimensional Navier-Stokes equations coupled with two-equation turbulence models using a discrete adjoint method For code validation, the result of the flow solver is compared with experiment data and other computational results of bench marking test. To study the influence oj turbulence models and grid refinement on the duct flow analysis, the results from several turbulence models are compared with one another and the minimum number of grid points, which can yield an accurate solution is investigated The adjoint variable code is validated by comparing the complex step derivative results. To realize a sufficient and flexible design space, NURBS equations are introduced as a geometric representation and a new grid modification technique, Least Square NURBS Grid Approximation is applied With the verified flow solver, the sensitivity analysis code and the geometric modification technique, the optimization of S-shaped intake is carried out and the enhancement of overall intake performance is achieved The designed S-shaped duct is tested in several off-design conditions to confirm the robustness of the current design approach. As a result, the capability and the efficiency of the present design tools are successfully demonstrated in three-dimensional highly turbulent internal flow design and off-design conditions.

• 한국전산유체공학회:학술대회논문집
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• 2008.03b
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• pp.268-271
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• 2008

If a reactor vessel is failed to retain a molten corium in a postulated severe accident, the molten corium is released outside the reactor vessel into a reactor cavity. The molten corium would attack the concrete wall and basemat of the reactor cavity, which may lead to inevitable concrete decompositions and possible radiological releases. In the OECD/MCCI project, a series of tests were performed to secure the data for cooling the molten corium spread out at the reactor cavity and for the long-term CCI (Core Concrete Interaction). Also, a MCCI (Molten Core Concrete Interaction) analysis code, CORQUENCH was upgraded at Argonne National Laboratory with embedding the new models developed for the tests. This paper deals with analyses of MCCI at plant scale under the conditions of top flooding using the upgraded CORQUENCH code. The modeling approach is briefly summarized first, followed by presentation of a validation calculation that illustrates the predicative capability of the modeling tool. With this background in place, the model is then used to carry out a parametric set of scoping calculations that define approximate coolability envelopes for the LCS (Limestone Common Sand) concrete that has been evaluated in the OECD/MCCI project.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.31 no.4
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• pp.320-329
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• 2007

Electrokinesis and dielectrophoresis are important transport phenomena produced by external electric field applied to a microchannel containing a conductive fluid. We developed a CFD code to predict electrokinetic and dielectrophoretic flows in a microchannel with a uniform circular post array. Using the code, we calculated particle velocities driven by electrokinesis and dielectrophoresis, and conducted Monte Carlo simulations to visualize the particle motions. The code was validated by comparing the results with those from previous studies in literature. At a low electric field, electrokinesis and diffusion is the dominant transport mechanism. At a moderate electric field, dielectrophoresis is balanced with electrokinesis and diffusion, resulting in flowing filaments of particles in the microchannels. However, dielectrophoresis overwhelms the flow at a high electric field and traps particles locally. These results provide useful insight for optimizing design parameters of a microfluidic chip for biochemical analysis, especially for development of on-chip sample pretreatment techniques using electrokinetic and dielectrophoretic effects.

• The Journal of Korean Institute of Communications and Information Sciences
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• v.38C no.5
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• pp.420-427
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• 2013

Recently, for secure software development, static analysis tools have been used mostly to analyze the source code of the software and identify software weaknesses caused of vulnerabilities. In order to select the optimal static analysis tool, both weaknesses rules and analysis capabilities of the tool are important factors. Therefore, in this paper we propose the test codes developed for evaluating the rules and analysis capabilities of the tools. The test codes to involve 43 weaknesses such as SQL injection etc. can be used to evaluate the adequacy of the rules and analysis capabilities of the tools.

• Electronics and Telecommunications Trends
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• v.32 no.6
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• pp.105-115
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• 2017

This paper describes a project - based quality management model that identifies development technologies and codes while at the same time verifying the ability to implement processes that are essential in R & D projects. In order to verify the process implemented in the R & D project implementation, there are review, checking points, and evaluating methods the process performance levels for five processes such as defining requirements which is the beginning stage of system development, testing process which is the completion stage of system development, and project management and peer review process for project management and support in the proposed project-based quality management model. For development technology and code validation, the model included the documented test cases for each requirement by the developer in the requirements definition stage, debugging and testing in the design and implementation stages, static analysis and open source licence verification procedure, and system environment. After applying the model in SW development R&D project for evaluating the process performance, and verifying the development technology and the code, the developers responded that the improvement in the development technique and the code, and upgrade of process performance level for project are more than 10%.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2830-2838
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• 2021

This work deals with the assessment of the burnup capabilities of the Serpent Monte Carlo code to predict spent nuclear fuel (SNF) isotopic concentrations for low-enriched uranium (LEU) fuel at different burnup levels up to 47 MWd/kgU. The irradiation of six UO₂ experimental samples in three different VVER-1000 reactor units has been simulated and the predicted concentrations of actinides up to ²⁴⁴Cm have been compared with the corresponding measured values. The results show a global good agreement between calculated and experimental concentrations, in several cases within the margins of the nuclear data uncertainties and in a few cases even within the reported experimental uncertainties. The differences in the performances of the JEFF3.1.1, ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries (NDLs) have also been assessed and the use of the newly released ENDF/B-VIII.0 library has shown an increased accuracy in the prediction of the C/E's for some of the actinides considered, particularly for the plutonium isotopes. This work represents a step forward towards the validation of advanced simulation tools against post irradiation experimental data and the obtained results provide an evidence of the capabilities of the Serpent Monte-Carlo code with the associated modern NDLs to accurately compute SNF nuclide inventory concentrations for VVER-1000 type reactors.