• The Journal of the Institute of Internet, Broadcasting and Communication
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• v.17 no.6
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• pp.145-150
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• 2017

Software maintenance accounts for a large portion of the software life cycle cost. In the software maintenance phase, comprehending the legacy source code is inevitable, which takes most of the time. Source code readability is a metric of the extent of code readers' difficulty of code comprehension based on the source code itself. The better the code is readable, the easier it is for code readers to comprehend the source code. This paper proposes novel source code readability metric to quantitative measure the extent of current source code under development, which is more enhanced measurement method than previous research that dichotomously judges whether the source code was readable or not. As an evaluation, we carried out a survey and analyzed them with Regression Analysis to find best parameters of the metric.

• Journal of Information Technology Services
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• v.10 no.2
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• pp.247-257
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• 2011

In this paper, we propose a multiple-disseminators determining mechanism for Efficient Code Dissemination with low-delay(ECoDi) for wireless sensor networks (WSN). Code dissemination is in the spotlight as an important research issue since sensor nodes are necessary for updating new software remotely or fixing bugs dynamically. In particular, the time factor for code dissemination is the most important factor in order that the normal operation of nodes can be continuously performed as soon as finishing the dissemination. For this factor, ECoDi determines the set of disseminators through regression analysis based on the size of distributed code and the time of past unicasts and broadcasts. Then it transmits the entire code as a unicast to multiple disseminators, and the disseminators broadcast the code to the remaining neighbor nodes. Performance results on a testbed show that ECoDi reduces dissemination time significantly compared to a conventional scheme.

• Nuclear Engineering and Technology
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• v.37 no.6
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• pp.587-594
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• 2005

In an effort to develop a thermal-hydraulic (TH) safety analysis code for Gas-cooled Reactors (GCRs), the MARS code, which was primarily developed for TH analysis of water reactor systems, has been extended here for application to GCRs. The modeling requirements of the system code were derived from a review of major processes and phenomena that are expected to occur during normal and accident conditions of GCRs. Models fur code improvement were then identified through a review of existing MARS code capability. Among these, the following priority models necessary fur the analysis of limiting high and low pressure conduction cooling events were evaluated and incorporated in MARS-GCR/V1 : 1) Helium (He) and Carbon Dioxide ($CO_2$) as main system fluids, 2) gas convection heat transfer, 3) radiation heat transfer, and 4) contact heat transfer models. Each model has been assessed using various conceptual problems for code-to-code benchmarks and it was demonstrated that MARS-GCR/V1 is capable of capturing the relevant phenomena. This paper describes the models implemented in MARS-GCR/V1 and their verification and validation results.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1596-1609
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• 2019

The fuel rod performance has great importance for the safety and economy of an operating reactor. The fuel rod performance analysis code, which considers the thermal-mechanical response and irradiation effects of fuel rod, is usually developed in order to predict fuel rod performance accurately. The FRIPAC (${\underline{F}}uel$ ${\underline{R}}od$ ${\underline{I}}ntegral$ ${\underline{P}}erformance$ ${\underline{A}}nalysis$ ${\underline{C}}ode$) is such a fuel rod performance analysis code that has been developed recently by China Nuclear Power Technology Research Institute Co. Ltd. The code aims at the computational simulation of the Pressurized Water Reactor fuel rod behavior for both steady-state and power ramp condition. A brief overview of FRIPAC is presented including the computational framework and the main behavioral models. Validation of the code is also presented and it focuses on the fuel rod behavior including fuel center temperature, fission gas release, rod internal pressure/internal void volume, cladding outer diameter and cladding corrosion thickness. The validation is based on experimental data from several international projects. The validation results indicate that FRIPAC is an accurate and reliable fuel rod performance analysis code because of the satisfactory comparison results between the experimental measurements and the code predictions.

• Journal of Software Assessment and Valuation
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• v.17 no.1
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• pp.25-30
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• 2021

In case of evaluating the similarity of the source code analysis material in the embedded system, the provided source code must be confirmed to be executable. However, it is currently being in which compilation and interface matching with hardware are provided in an unconfirmed materials. The complainant assumes that many parts of the source code are similar because the characteristics of the operation are similar and the expression of the function is similar. As for the analysis result, the analysis result may appear different than expected due to these unidentified objects. In this study, the improvement direction is sugested through the case study by the analysis process of the source code and the similarity of the unverified source code.

• 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.

• Nuclear Engineering and Technology
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• v.42 no.3
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• pp.279-296
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• 2010

For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal hydraulics code, named CUPID, has been developed. The CUPID code adopts a two-fluid, three-field model for two-phase flows, and the governing equations were solved over unstructured grids, which are very useful for the analysis of flows in complicated geometries. To obtain numerical solutions, the semi-implicit numerical method for the REALP5 code was modified for an application to unstructured grids, and it has been further improved for enhanced accuracy and fast running. For the verification of the CUPID code, a set of conceptual problems and experiments were simulated. This paper presents the flow model, the numerical solution method, and the results of the preliminary assessment.

• Proceedings of the KSR Conference
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• 2002.10b
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• pp.884-891
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• 2002

This report is the result performed the structural analysis and the static and fatigue load test of bogie frame for the purpose of designing and verifying the bogie frame which satisfy the load condition required in the UIC code. This investigation is proposed the efficient draft of the design to satisfy the load condition required in the UIC code. And It is performed the structural analysis to evaluate the static strength and the fatigue life of the patient material and the welded part. Also, This is proposed the efficient draft of the test to satisfy the method of the static and fatigue test required in the UC code. And it is carried out the static and the fatigue load test to verify it. We can designed the bogie frame in compliance with UIC 515-4 and 615-4 code.

• Nuclear Engineering and Technology
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• v.46 no.5
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• pp.655-666
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• 2014

The CUPID code has been developed at KAERI for a transient, three-dimensional analysis of a two-phase flow in light water nuclear reactor components. It can provide both a component-scale and a CFD-scale simulation by using a porous media or an open media model for a two-phase flow. In this paper, recent advances in the CUPID code are presented in three sections. First, the domain decomposition parallel method implemented in the CUPID code is described with the parallel efficiency test for multiple processors. Then, the coupling of CUPID-MARS via heat structure is introduced, where CUPID has been coupled with a system-scale thermal-hydraulics code, MARS, through the heat structure. The coupled code has been applied to a multi-scale thermal-hydraulic analysis of a pool mixing test. Finally, CUPID-SG is developed for analyzing two-phase flows in PWR steam generators. Physical models and validation results of CUPID-SG are discussed.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.64-69
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• 1997

The code WIMS-AECL has been used for the lattice analysis of DUPIC fuel. The lattice parameters calculated by the code is sensitive to the choice of number of parameters, such as the number of tracking lines, number of condensed groups, mesh spacing in the moderator region, other parameters vital to the calculation of probabilities and burnup analysis. We have studied this sensitivity with respect to these parameters and recommend their proper values which are necessary for carrying out the lattice analysis of DUPIC fuel.