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

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.4 no.3
• /
• pp.209-216
• /
• 2006

Characteristics of heat release process, while the Wigner energy was drawn off the graphite during DSC(Differential Scanning Calorimenter) measurement as an example of annealing process which is one of release methods of Wigner energy that is contained in the irradiated graphite, was studied. Linear temperature rise method in DSC operation was selected to estimate the total Wigner energy content and the heat release rate of each graphite samples, which were located in several positions in the thermal column in KRR-2 research reactor. As an annealing process in DSC operation Wigner energy of the irradiated graphite samples were totally released by heat supplying to the graphite from room temperature to $500^{\circ}C$, in DSC. Characteristics of Wigner energy release from the graphite sample was well correlated with the various activation energy model of the kinetic equation.

• Applied Microscopy
• /
• v.35 no.3
• /
• pp.113-119
• /
• 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 the Korean Society for Nondestructive Testing
• /
• v.9 no.2
• /
• pp.8-17
• /
• 1989

Reconstruction of the radial two-dimensional of fission products using projections obtained by nondestructive transversal direction gamma scanning was evaluated, and then compared with the destructive sectional gamma scanning results. Although the nondestructive method has a little drawback compared to the destructive method in the accuracy, the former seemed to be sufficiently applicable.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2002.05a
• /
• pp.234-234
• /
• 2002

• Proceedings of the Korean Nuclear Society Conference
• /
• 2002.10a
• /
• pp.268-268
• /
• 2002

• Journal of Energy Engineering
• /
• v.18 no.3
• /
• pp.169-174
• /
• 2009

In the study, the effect of the ultrasonic energy in transportational diesel fuel on the engine performance and exhaust emission has been investigated for indirect injection diesel engine. It was tested to estimated change of engine performance and exhaust emission characteristics for the transportational diesel fuels and the reforming fuels which was irradiated by the ultrasonic energy. The results of the study may be concluded as follows; By the irradiation of ultrasonic energy on the diesel fuel, cylinder pressure, heat release rate and engine power were increased but bsfc, mass fraction burned, and smoke were reduced. Also, the combustion was more stabilized and became complete and NOx was increased.

• Nuclear Engineering and Technology
• /
• v.27 no.1
• /
• pp.67-73
• /
• 1995

To investigate the storage behavior of the defective LWR spent fuel, air-oxidation tests on non-irradiated and irradiated U $O_2$ were performed. The weight gains of non-irradiated U $O_2$ specimens are characterized by the S-shape curves at 250-40$0^{\circ}C$ temperature range. One hundred percent conversion of U $O_2$ to U$_3$ $O_{8}$ corresponds with about 4％ weight increase. The activation energies are 110 kJ/mol above 35$0^{\circ}C$ and 153 kJ/mol below 35$0^{\circ}C$. The irradiated U $O_2$ specimens with about 35 GWD/MTU burnup were oxidized at 300-40$0^{\circ}C$ in air. They show a rapid increase of weight gain at the initial stage and a slow increase at the later stage when compared with non-irradiated U $O_2$. The activation energy under these conditions is 95 kJ/mol. Burnup and aging effects of irradiated U $O_2$ were also investigated at 35$0^{\circ}C$ in air environment, but the specimens appears insensitive to these variables.s.

• Nuclear Engineering and Technology
• /
• v.21 no.4
• /
• pp.277-286
• /
• 1989

Destructive methods are used for the turnup determination of an irradiated PWR fuel. One of the methods includes U, Pu, Nd-148 and Nd-(145+146) determination by an isotope dilution mass spectrometry using triple spikes (U-233, Pu-242 and Nd-150). The method involves two sequential ion exchange resin separation procedures. Pu is eluted from the first anion exchange resin column (Dowex AG 1$\times$8) with 12 M HCl-0.1 M HI mixed solution, followed by U elution with 0.1 M HCl. Nd is isolated from other fission products on the second anion exchange resin column (Dowex AG 1$\times$4) with a nitric acid-methanol eluent. Each fraction is analysed by thermal ionization mass spectrometry. The difference between Nd-148 and Nd-(145+146) method is found with an average 2.07%. The results are compared with those by the heavy element method using U and Pu isotopes and by the destructive y-spectrometric measurement of Cs-137. The dependences of isotope composition of U and Pu on burn-up, and correlation between those isotopes are illustrated graphically.

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