• Journal of the Korean Ceramic Society
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• v.51 no.5
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• pp.435-441
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• 2014

Silicon carbide-based ceramics and their composites have been studied for application to fusion and advanced fission energy systems. For fission reactors, $SiC_f$/SiC composites can be applied to core structural materials. Multilayered SiC composite fuel cladding, owing to its superior high temperature strength and low hydrogen generation under severe accident conditions, is a candidate for the replacement of zirconium alloy cladding. The SiC composite cladding has to retain its mechanical properties and original structure under the inner pressure caused by fission products; as such it can be applied as a cladding in fission reactor. A hoop strength test using an expandable polyurethane plug was designed in order to evaluate the mechanical properties of the fuel cladding. In this paper, a hoop strength test of the multilayered SiC composite tube for nuclear fuel cladding was simulated using FEA. The stress caused by the plug was distributed nonuniformly because of the friction coefficient difference between the inner surface of the tube and the plug. Hoop stress and shear stress at the tube was evaluated and the relationship between the concentrated stress at the inner layer of the tube and the fracture behavior of the tube was investigated.

• Nuclear Engineering and Technology
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• v.24 no.1
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• pp.40-51
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• 1992

One of the important irradiation performance characteristics of the silicide dispersion fuel element in research reactors is the diameteral increase resulting from fuel swelling. This paper, will attempt to develop a physical model for the fuel swelling, DFSWELL, by analyzing the basic irradiation behaviours and some experimental evidences. From the experimental evidences, it was shown that the volume changes in irradiated U$_3$Si-Al were strongly dependent on temperature and fission rate. The quantitative-amount of swelling for silicide fuel is estimated by considering temperature, fission rate, solid fission product build-up and gas bubble behavior. The swelling for the silicide fuel is comprised of three major components : i ) a volume change due to the formation of an interfacial layer between the fuel particle and matrix. ii ) a volume change due to the accumulation of gas bubble nucleation iii ) a volume change due to the accumulation of solid fission products The DFSWELL model which takes into account the above three major physical components predicts well the absolute magnitude of silicide fuel swelling in accordance with the power histories in comparison with the experimental data.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.975-981
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• 2016

The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.

• Analytical Science and Technology
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• v.13 no.4
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• pp.474-483
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• 2000

An inductively coupled plasma atomic emission spectrometer/shielding system was specially designed and built for the analysis of radioactive materials. Both of an inductively coupled plasma source and a sample transfer system to be contacted with radioactive materials was installed in a stainless steel glove box. In terms of analytical capability and radiation safety, characteristic feature of the system was investigated. Its applicability to the determination of fission products and corrosion products in the radioactive materials such as spent fuel dissolver solution and the primary coolant of nuclear power reactors was evaluated. In the concentration range $0.01-0.1mgL^{-1}$, the relative standard deviation was found to be less than 5%.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.335-343
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• 2005

The Methodology of burnup calculation with EPMA test set up in this study. The spent fuel from PWR nuclear power plant was used as specimen. This $UO_2$ fuel with $3.2\%$ of enrichment had been irradiated up to 35,000 MWd/MTU(reference data). The burnup is very important factor for nuclear fuel to estimate all fuel behaviors in reactor. To measure amounts of fission products and actinides for the burnup calcualation, chemical analysis (destructive method) has been used but it mattes long experimental time and second radio-wastes. In this study, EPMA test was available to measure amount of fission products. Neodymium is able to be detected and quantified. It can be compared with the results from chemical analysis and ORIGEN-2 code calculation. Concentration of Nd from EPMA test showed good agreement with result of ORIGEN-2 code in the same burnup.

• Applied Microscopy
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• v.35 no.3
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• pp.113-119
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• 2005

The Methodology of burnup calculation with EPMA test set up in this study. The spent fuel from PWR nuclear power plant was used as specimen. This $UO_2$ fuel with 3.2% of enrichment had been irradiated up to 35,000 MWd/MTU. The burnup is very important factor for nuclear fuel to estimate all fuel behaviors in reactor. To measure amounts of fission products and actinides for the burnup calcualation, destructive method analysis has been used but it makes long experimental time and second radio-wastes. In this study, EPMA test was available to measure amount of fission products. Neodymium is able to be detected and quantified. It can be compared with the results from chemical analysis and ORIGEN-2 code calculation. Concentration of Nd from EPMA test showed good agreement with result of ORIGEN-2 code in the same burnup.

• Journal of Energy Engineering
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• v.23 no.3
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• pp.186-190
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• 2014

Nuclear Power Plants require the control ability of chemical condition (pH) because pH control during transient accident such as LOCA makes an able the fission product removal capability to be maintained, stress corrosion cracking of stainless steel equipment to be prevented and the production of hydrogen by aluminum and zinc to be minimized. An NPP is designed to control the pH of containment spray and sump coolant using the spray additives 30% NaOH in the event of loss of coolant accident. In this paper, the pH of sump coolant of an NPP during LOCA was analyzed and the fission products removal constant and decontamination factor were calculated according to Standard Review Plan 6.5.2 related to spray chemical conditions of pH. The calculated pH value of recirculation mode using the computer code corresponds to 8.09~9.67, which meets the chemical environment regulation requirements. The fission product removal capability caused by containment spray system is performed to provide input to radiation analysis.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.1073-1074
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.879-880
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.379-380
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• 2005