• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.973-974
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.73-73
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• 1998

• Journal of the Korean Society for Nondestructive Testing
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• v.21 no.5
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• pp.561-564
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• 2001

The cladding area of HANARO Research Reactor fuel rods should be checked not to have any defects larger than the size required at QA documents by using eddy torrent testing method doting fabrication process. To apply eddy current testing inspection to the fuel rods, encircling differential probes and standard specimen were designed and fabricated. The impedance of the fabricated probes was measured with impedance analyzer in order to cheek that the probe has a suitable impedance for the inspection frequency, and with this probe and MIZ-40A eddy current equipment, the detectability of this probes was investigated. The developed probes could detect artificial notch with 2mm length 10% depth of cladding thickness in cladding area. In addition, the probe was successfully applied to detect the defects in cladding area doting fabrication of the research reactor rods.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.423-428
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• 1996

The feasibility study on burning Am-241, 243 and Cm-244 nuclides in a conventional PWR (Pressurized Water Reactor) was carried out by using the TRIFON code that was developed by the Institute of Theoretical and Experimental Physics in Russia in 1992. TRIFON code uses updated ABBN Russian nuclear cross section library. The reference reactor is the Korea nuclear power plant unit 8 (YGN 2). The burning effect of Am-241, 243 and Cm-244 nuclides was studied with UO$_2$(3.5 w/o)fuel assembly and MOX (4.44 w/o) fuel assembly. The loaded mass ratio of Am-241, 243 and Cm-244 nuclides was obtained from the mass ratio of Am-241, 243 and Cm-244 nuclides in 10 year cooling spent fuel with average discharge burnup of 33 GWD/MTU. The effective transmutation rates of Am-241, 243 and Cm-244 nuclides in UO$_2$ fuel assembly were found to be higher than those in MOX fuel assembly. The result from TRIFON code was compared to that from CASMO-3/NEM-3D code system. For more reliable calculation of transmutation for MA(Minor Actinides) more sophisticated decay chain scheme of MA should be investigated and nuclear cross section library of MA should be considerably improved.

• Proceedings of the Korea Electromagnetic Engineering Society Conference
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• 2003.11a
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• pp.213-218
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• 2003

Because of the development of digital technology, modern digital instrumentation & control systems are being innovativly developed in industrial plants. Whereas, many analog systems are still being used in nuclear plants, because of the demerits of digital equipment. As known, the demerits of digital equipment are the uncertainty and weaknesses in ambient environments such as smoke & electromagnetic interference In an Integral Reactor, a digital I&C system will be composed of microprocessor, memory and network card. Designers will apply new technique for digital equipment. Thus, it is important for digital I&C systems to operate according to designed functions & performance in the ambient environments during a life cycle. Digital I&C systems should have tolerance in such environments and environment qualification should be concluded To acquire electromagnetic interference qualification of digital equipment, this paper suggests an EMI test requirement. Designers should consider the electromagnetic compatibility and test digital equipment according to each test procedure. This paper involves an EMI test requirement and the results analysis of EUT(Equipment Under Test). Test result analysis will be used as electromagnetic compatibility design guides for Integral Reactor I&C systems.

• Nuclear Engineering and Technology
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• v.51 no.8
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• pp.1916-1938
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• 2019

The objective of this paper is to discuss the modeling principles of phenomena governing core degradation/melting and in-vessel melt relocation during severe accidents in light water reactors. The proposed modeling approach has been applied in the development of a new accident simulation package, COMPASS (COre Meltdown Progression Accident Simulation Software). COMPASS can be used either as a stand-alone tool to simulate in-vessel meltdown progression up to and including RPV failure, or as a component of an integrated simulation package being developed in Korea for the APR1400 reactor. Interestingly, since the emphasis in the development of COMPASS modeling framework has been on capturing generic mechanistic aspects of accident progression in light water reactors, several parts of the overall model should be useful for future accident studies of other reactor designs, both PWRs and BWRs. The issues discussed in the paper include the overall structure of the model, the rationale behind the formulation of the governing equations and the associated simplifying assumptions, as well as the methodology used to verify both the physical and numerical consistencies of the overall solver. Furthermore, the results of COMPASS validation against two experimental data sets (CORA and PHEBUS) are shown, as well as of the predicted accident progression at TMI-2 reactor.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.901-902
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.195-196
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.107-108
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• 2005

• Nuclear Engineering and Technology
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• v.22 no.4
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• pp.351-358
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• 1990

A new fission-product burnup chain model was developed for use in research reactor analysis capable of predicting the burnup-dependent reactivity with high precision over a wide range of burnup. The new model consists of 63 nuclides treated explicitly and one fissile-independent pseudo-element. The effective absorption cross sections for the pseudo-element and the pseudo-element yield of actinide nuclides were evaluated in the this report. The model is capable of predicting the high burnup behavior of low-enriched uranium-fueled research reactors.