• Journal of the Korean Chemical Society
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• v.39 no.7
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• pp.538-542
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• 1995

The stability constant and the thermodynamic parameters of the hydrazide Schiff base ligands such as N,N'-oxalylbis(salicylaldehydehydrazone)(OBSH), N,N'-malonylbis(salicylaldehydehydrazone)(MBSH), and N,N'-succinylbis(salicylaldehydehydrazone)(SBSH) with dioxouranium(Ⅵ) ion have been determined by potentiometry in 95% DMF solution at various temperatures. The order of stability constants increased SBSH

• Analytical Science and Technology
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• v.17 no.1
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• pp.23-28
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• 2004

Trace amount of uranyl (II) has been determined spectrophotometrically by measuring the optical density of the light blue yellowish coloured solutions formed by reaction between the metal ion and nicotinohydroxamic acid (NHx) in presence of different secondary ligands in strong isoamyl alcohol alkaline medium. The absorption maxima for both aqueous and extracted systems measured at their respective optimum pH were found to be 360 and 559 nm (DETA), 375 and 358 nm (EDA), 369 and 362 nm (piperidine), 354 and 341 nm (pyridine) and 363 and 336 nm (3 piperidine), 354 and 341 nm (pyridine) and 363 and 336 nm (3 - picoline), respectively at which Beer's law was obeyed. Effect of pH, reagent concentration, order of addition of reagent, time, temperature and solvent media on the absorption spectra have also been studied. Among the different systems studied, the shortest concentration range of uranyl(II) adhering to Beer's Law was 2.4 - 10.5 ppm observed for $UO_2(II)$ - NHx - DETA system in aqueous medium and also for iso amyl alcohol(IAA) extracted $UO_2$ - NHx - pyridine system was 2.4 - 7.8.

• Nuclear Engineering and Technology
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• v.26 no.3
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• pp.374-380
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• 1994

A method to simulate the extent of interlinkage of grain boundary gas bubbles to the free surfaces of fuel pellet was developed. With the shape of UO$_2$gain treated as tetrakaidecahedron (TKD)), the interlinked fraction of fission gas bubbles to free surfaces at grain comers was calculated as a function of the radius of grain corner bubbles by the Monte Carlo technique. In spite of two dimensional analysis, the present method shooed reasonable agreement between predicted and measured fuel swelling at the moment that complete bubble interlinkage was achieved. However, for more realistic simulation of interlinkage, grain comer bubbles should be treated three dimensionally.

• 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.

• Nuclear Engineering and Technology
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• v.34 no.5
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• pp.482-493
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• 2002

The MOX fuel for LWR is fabricated either by direct mechanical blending of UO$_2$ and PuO$_2$ or by two stage mixing. Hence Pu-rich particles, whose Pu concentrations are higher than pellet average one and whose size distribution depends on a specific fabrication method, are inevitably dispersed in MOX pellet. Due to the inhomogeneous microstructure of MOX fuel, the thermal conductivity of LWR MOX fuel scatters from 80 to 100 % of UO$_2$ fuel. This paper describes a mechanistic thermal conductivity model for MOX fuel by considering this inhomogeneous microstructure and presents an explanation for the wide scattering of measured MOX fuel＇s thermal conductivity. The developed model has been incorporated into a KAERI＇s fuel performance code, COSMOS, and then evaluated using the measured in-pile data for MOX fuel. The database used for verification consists of homogeneous MOX fuel at beginning-of-life and inhomogeneous MOX fuel at high turnup. The COSMOS code predicts the thermal behavior of MOX fuel well except for the irradiation test accompanying substantial fission gas release. The over-prediction with substantial fission gas release seems to suggest the need for the introduction of a recovery factor to a term that considers the burnup effect on thermal conductivity.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.637-642
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• 1996

This paper presents the analysis results for the core degradation processes and the fission product release of the PHEBUS FPT0 experiment using MELCOR1.8.3. The objective of this study is to assess models associated with the core damage and fission product behavior in MELCOR. The calculation results were much improved through sensitivity studies. Thermal/hydraulic behavior in the core and the circuit was well predicted under the intact core geometry. In non-eutectic model case. the UO$_2$ dissolution model in the MELCOR always showed such a tendency that the resulting dissolved UO$_2$ mass was small at the highly oxidized condition due to the model logic. Total H$_2$ generation mass was underpredicted because the stiffner was not modeled and the liner in the shroud was not allowed to be oxidized in MELCOR. Some difficulties were found in modeling the activation product were solved by manipulating the RN input associated with the initial fission product inventory. These problem were occurred because there are no control rod model in MELCOR. Generally the fission product release ratio showed a similar trend compared with the measured data except the activation product. which have no model to simulate in MELCOR.

• Journal of Radiation Protection and Research
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• v.10 no.2
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• pp.144-147
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• 1985

The spontaneous fission rate of $^{238}U$ has been determined using a solid state track recorder that was a pre-etched mica. Counting the tracks revealed in mica made it possible to calculate the spontaneous fission rate. The mica remained in close contact with a $^{238}UO_2$ foil for about five years. The resulting fission rate was $5.21{\pm}0.33$ fissions/g-sec.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3213-3228
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• 2023

As an important part of the digital reactor, the pin-by-pin wise fine coupling calculation is a research hotspot in the field of nuclear engineering in recent years. It provides more precise and realistic simulation results for reactor design, operation and safety evaluation. CORCA-K a nodal code is redeveloped as a robust pin-by-pin wise neutronics and thermal-hydraulic coupled calculation code for pressurized water reactor (PWR) core. The nodal green's function method (NGFM) is used to solve the three-dimensional space-time neutron dynamics equation, and the single-phase single channel model and one-dimensional heat conduction model are used to solve the fluid field and fuel temperature field. The mesh scale of reactor core simulation is raised from the nodal-wise to the pin-wise. It is verified by two benchmarks: NEACRP 3D PWR and PWR MOX/UO₂. The results show that: 1) the pin-by-pin wise coupling calculation system has good accuracy and can accurately simulate the key parameters in steady-state and transient coupling conditions, which is in good agreement with the reference results; 2) Compared with the nodal-wise coupling calculation, the pin-by-pin wise coupling calculation improves the fuel peak temperature, the range of power distribution is expanded, and the lower limit is reduced more.

• Advances in concrete construction
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• v.16 no.1
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• pp.35-57
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• 2023

Majority of the plastic produced each year is being disposed in land after single-use, which becomes waste and takes up a lot of storage space. Therefore, there is an urgent need to find alternative solutions instead of disposal. Recycling and reusing the PET plastic waste as aggregate replacement and fiber in concrete production can be one of the eco- friendly methods as there is a great demand for concrete around the world, especially in developing countries by raising human awareness of the environment, the economy, and Carbon dioxide (CO2) emissions. Self-compacting concrete (SCC) is a key development in concrete technology that offers a number of attractive features over traditional concrete applications. Recently, in order to improve its durability and prevent such plastics from directly contacting the environment, various kinds of plastics have been added. This review article summarizes the latest evident on the performance of SCC containing recycled PET as eco-friendly aggregates and fiber. Moreover, it highlights the influence of substitution content, shape, length, and size on the fresh and properties of SCC incorporating PET plastic. Based on the findings of the articles that were reviewed for this study, it is observed that SCC made of PET plastic (PETSCC) can be employed in construction era owing to its acceptable mechanical and fresh properties. On the other hand, it is concluded that owing to the lightweight nature of plastic aggregate, Reusing PET waste in the construction application is an effective approach to reduces the earthquake risk of a building.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.897-898
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• 2004