• 한국인터넷방송통신학회논문지
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• 제17권6호
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• pp.145-150
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• 2017

소프트웨어 유지보수는 소프트웨어 생명주기에서 전체 비용의 많은 부분을 차지하고 있다. 소프트웨어를 유지보수하기 위해서는 코드를 읽는 것이 필수적이고, 이는 유지보수 활동에서 가장 많은 시간이 소요되는 활동이다. 코드의 가독성은 사람이 소스코드를 이해하는데 드는 어려움의 정도를 측정하는 메트릭이다. 코드의 가독성이 좋을수록 사람이 소스코드를 이해하기 쉬워진다. 본 논문에서는 소스코드의 가독성을 이분법적으로 판단하는 기존의 연구보다 향상된 기법을 사용하여, 개발 중인 소스코드의 가독성 정도를 정량적으로 측정할 수 있는 새로운 소스코드 가독성 메트릭을 제안한다. 이를 평가하기 위해, 우리는 설문조사를 수행하고 가독성을 가장 잘 설명하는 척도를 찾을 수 있도록 회귀 분석 기법을 사용해 그 결과를 분석할 것이다.

• 한국IT서비스학회지
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• 제10권2호
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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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• 제37권6호
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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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• 제51권6호
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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.

• 한국소프트웨어감정평가학회 논문지
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• 제17권1호
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• pp.25-30
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• 2021

임베디드 시스템 소스코드 감정목적물의 유사성을 판단하는 경우, 제공된 소스코드가 컴파일이 가능한 상태인지, 실행이 정상적으로 되는 것인지, 하드웨어와의 인터페이스가 일치하는지 등의 확인이 되지 않은 상태로 제공되는 경우가 지속적으로 발생하고 있다. 분쟁을 제기하는 측에서는 동작의 특성이 유사하고, 나타나는 기능의 효과가 유사한 이유로 소스코드의 많은 부분이 유사할 것으로 판단하고, 감정을 요청하게 되지만, 위의 여러 가지 상황으로 유사성 분석 결과가 기대와 다르게 나타나는 가능성이 우려된다. 본 연구에서는 감정사례를 통해 소스코드의 분석 과정과 검증되지 않은 소스코드의 유사성 도출의 개선방향을 제시한다.

• Nuclear Engineering and Technology
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• 제55권9호
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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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• 제42권3호
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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.

• 한국철도학회:학술대회논문집
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• 한국철도학회 2002년도 추계학술대회 논문집(II)
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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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• 제46권5호
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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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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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.