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Park J.W.;Kim C.L.;Park J.B.;Lee E.Y.;Lee Y.M.;Kang C.H.;Zhou W.;Kozak M.W.
- Nuclear Engineering and Technology
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- v.35 no.6
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- pp.556-565
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- 2003
A quality assurance safety assessment database, called QUARK (QUality Assurance Program for Radioactive Waste Management in Korea), has been developed to manage both analysis information and parameter database for safety assessment of low- and intermediate-level radioactive waste (LILW) disposal facility in Korea. QUARK is such a tool that serves QA purposes for managing safety assessment information properly and securely. In QUARK, the information is organized and linked to maximize the integrity of information and traceability. QUARK provides guidance to conduct safety assessment analysis, from scenario generation to result analysis, and provides a window to inspect and trace previous safety assessment analysis and parameter values. QUARK also provides default database for safety assessment staff who construct input data files using SAGE(Safety Assessment Groundwater Evaluation), a safety assessment computer code.
PDF KSCI

Mahaffy, John
- Nuclear Engineering and Technology
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- v.42 no.4
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- pp.377-381
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- 2010
In 2007 the Nuclear Energy Agency's Committee on the Safety of Nuclear Installations published Best Practice Guidelines for the use of CFD in Nuclear Reactor Safety. This paper provides an overview of the document' contents and highlights a few of its recommendations. The document covers the full extent of a CFD analysis from initial problem definition and selection of an appropriate tool for the analysis, through final documentation of results. It provides advice on selection of appropriate simulation software, mesh construction, and selection of physical models. In addition it contains extensive discussion of the verification and validation process that should accompany any high-quality CFD analysis.
https://doi.org/10.5516/NET.2010.42.4.377 인용 PDF KSCI

Baek, Won-Pil;Yang, Joon-Eon;Ha, Jae-Joo
- Nuclear Engineering and Technology
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- v.41 no.4
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- pp.391-402
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- 2009
This paper introduces the development of safety assessment technology in Korea, focusing on the activities of the Korea Atomic Energy Research Institute in the areas of system thermal hydraulics, severe accidents and probabilistic safety assessment. In the 1970s and 1980s, safety analysis codes and methodologies were introduced from the United States, France, Canada and other developed countries along with technology related to the construction and operation of nuclear power plants. The main focus was on understanding and utilizing computer codes that were sourced from abroad up to the early 1990s, when efforts to develop domestic safety analysis codes and methodologies became active. Remarkable achievements have been made over the last 15 years in the development and application of safety analysis technologies. In addition, significant experimental work has been performed to verify the safety characteristics of reactors and fuels as well as to support the development and validation of analysis methods.
https://doi.org/10.5516/NET.2009.41.4.391 인용 PDF KSCI

LIU, MEIRU;XIA, HONG;SUN, LIN;LI, BIN;YANG, YANG
- Nuclear Engineering and Technology
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- v.47 no.2
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- pp.219-225
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- 2015
In this paper, a three-dimensional model for the dynamic analysis of a flywheel based on the finite element method is presented. The static structure analysis for the model provides stress and strain distribution cloud charts. The modal analysis provides the basis of dynamic analysis due to its ability to obtain the natural frequencies and the vibration-made vectors of the first 10 orders. The results show the main faults are attrition and cracks, while also indicating the locations and patterns of faults. The harmonic response simulation was performed to gain the vibration response of the flywheel under operation. In this paper, we present a Hilbert-Huang transform (HHT) algorithm for flywheel vibration analysis. The simulation indicated that the proposed flywheel vibration signal analysis method performs well, which means that the method can lay the foundation for the detection and diagnosis in a reactor main coolant pump.
https://doi.org/10.1016/j.net.2014.12.010 인용 KSCI