• Proceedings of the Korean Ceranic Society Conference
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• 2003.04a
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• pp.28.2-28
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• 2003

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.11a
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• pp.145-147
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• 2004

• Nuclear Engineering and Technology
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• v.29 no.4
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• pp.299-309
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• 1997

The neutronic feasibility of typical Korean three-loop 900MWe class PWR core loaded with mixed oxide fuels for both annual and 18-month cycle strategies has been investigated as a means for spent fuel management. For this study, a method of determining equivalent plutonium content was developed under the equivalence concept which gives the same cycle length as uranium fuel. Optimal plutonium zoning within the MOX assembly was also designed with the aim of minimizing the peak md power. Conceptual core designs hate hen developed for equilibrium cycle with the following variations: annual and 18-month cycle, 1/3 and full MOX loading schemes, and typical and high moderation lattice. The analysis of key core physics parameters shows that in all cases considered satisfactory core designs seem to be feasible, though addition of control rod system and change in Technical Specification for soluble boron concentration are required for full MOX loading in order to meet the current design requirements.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.363-367
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• 2003

Much attention has been given to an electrochemical reduction process for converting uranium oxide to uranium metal in molten salt. The process has the versatility of being adopted for reducing other actinide and rare-earth metals from their oxides. Using the metal oxide to be reduced as a integrated cathode designed originally and inert conductors as anodes, oxygen anions are removed from the cathode and oxidized at the surface of the anodes in a molten salt cell. However, the electrochemical properties of alkali and alkali-earth metal oxides in molten salt have not been investigated thoroughly, which made the process incomplete when it is considered as a unit process in a back-end fuel cycle. It is well known that cesium and strontium Isotopes in spent fuel are main contributors for head load. The properties of cesium, strontium, and barium oxides such as the dissolution rates and reduction potentials in molten LiC1 dissolving $Li_2O$ are examined.

• Nuclear Engineering and Technology
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• v.40 no.1
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• pp.21-36
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• 2008

High bum-up fuel technology has been developed through a national R&D program, which covers key technology areas such as claddings, $UO_2$ pellets, spacer grids, performance code, and fuel assembly tests. New cladding alloys were developed through alloy designs, tube fabrication, out-of-pile test and in-reactor test. The new Zr-Nb tubes are found to be much better in their corrosion resistance and creep strength than the Zircaloy-4 tube, owing to an optimized composition and heat treatment of the new Zr-Nb alloys. A new fabrication technology for large grain $UO_2$ pellets was developed using various uranium oxide seeds and a micro-doping of Al. The uranium oxide seeds, which were added to $UO_2$ powder, were prepared by oxidizing and heat-treating scrap $UO_2$ pellets. A $UO_2$ pellet containing tungsten channels was fabricated for a thermal conductivity enhancement. For the fuel performance analysis, new high burnup models were developed and implemented in a code. This code was verified by an international database and our own database. The developed spacer grid has two features of contoured contact spring and hybrid mixing vanes. Mechanical and hydraulic tests showed that the spacer grid is superior in its rodsupporting, wear resistance and CHF performance. Finally, fuel assembly test technology was also developed. Facilities for mechanical and thermal hydraulic tests were constructed and are now in operation. Several achievements are to be utilized soon by the Korea Nuclear Fuel and thereby contribute to the economy and safety of PWR fuel in Korea

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1799-1804
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• 2019

The concept of a new method, the CARBEX (CARBonate EXtraction) process, was proposed for reprocessing of spent uranium oxide fuel. The proposed process is based on use of water solutions of $Na_2CO_3$ or $(NH_4)_2CO_3$ and solvent extraction (SE) by the quaternary ammonium compounds for selective recovery and purification of U from the fission products (FPs). Applying of SE allows to reach high degree of purification of U from FPs. Carrying out the processes in poorly aggressive alkaline carbonate media leads to increasing safety of SNF's reprocessing and better selectivity of separation of lanthanides and actinides. Moreover carbonate reprocessing media allows to carry out a recycling and regeneration of reagents. We have been done laboratory scale experiments on the extraction components of simulated voloxidated spent fuel in the solutions of NaOH or $Na_2CO_3-H_2O_2$ and recovery of U from carbonate solutions by SE method using carbonate of methyltrioctylammonium in toluene. It was shown that the purification factors of U from impurities of simulated FPs reached values $10^3-10^5$. The received results support our opinion that CARBEX after the further development can become more safe, simple and profitable method of spent fuel reprocessing.

• Journal of Energy Engineering
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• v.5 no.2
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• pp.193-197
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• 1996

A study was conducted on the oxidation behavior of U-0.75 wt% Ti chips (Depleted Uranium, DU chips) using an XRD and a thermogravimetric analyzer in the temperature range from 250 to 500$^{\circ}C$ in air. At the temperature lower than 400$^{\circ}C$, DU chips were converted to UO$_2$, U$_3$O$\_$7/, and U$_3$O$\_$8/ whereas at the temperature higher than 400$^{\circ}C$, DU chips were completely converted to U$_3$O$\_$8/, the most stable form of uranium oxide. The activation energy for the oxidation of DU chips is found, 44.9 kJ/mol and the oxidation rate in terms of weight gain (%) can be expressed as; dW/dt8.4${\times}$10$^2$e(equation omitted) wt%/min (250$\leq$T($^{\circ}C$) $\leq$ 500) where W=weight gain (%), t=time and T=temperature.

• Journal of the Korean Electrochemical Society
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• v.16 no.3
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• pp.138-144
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• 2013

Electrochemical reductions of $UO_2$ at various anode-to-cathode distances (1.3, 2.3, 3.2, 3.7 and 5.8 cm) were carried out to investigate the effect of the anode-to-cathode distance on the electrochemical reduction rate. The geometry of the electrolysis cell in this study, apart from the anode-to-cathode distance, was identical for all of the electrolysis runs. Porous $UO_2$ pellets were electrolyzed by controlling a constant cell voltage in molten $Li_2O-LiCl$ at $650^{\circ}C$. A steel basket containing the porous $UO_2$ pellets and a platinum plate were used as the cathode and anode, respectively. The metallic products were characterized by means of a thermogravimetric analyzer, an X-ray diffractometer and a scanning electron microscope. The electrolysis runs conducted during this study revealed that a short anode-to-cathode distance is advantageous to achieve a high current density and accelerate the electrochemical reduction process.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2760-2770
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• 2020

A model for calculation of fuel temperature profile using binary gas mixture of Helium and Xenon for gap gas conductance is proposed here. In this model, the temperature profile of a fuel pencil from fuel centreline to fuel surface has been calculated by taking into account the dilution of Helium gas filled during fuel manufacturing due to accumulation of fission gas Xenon. In this model an explicit calculation of gap gas conductance of binary gas mixture of Helium and Xenon has been carried out. A computer code Fuel Characteristics Calculator (FCCAL) is developed for the model. The phenomena modelled by FCCAL takes into account heat conduction through the fuel pellet, heat transfer from pellet surface to the cladding through the gap gas and heat transfer from cladding to coolant. The binary noble gas mixture model used in FCCAL is an improvement over the parametric model of Lassmann and Pazdera. The results obtained from the code FCCAL is used for fuel temperature calculation in 3-D neutron diffusion solver for the coolant outlet temperature of the core at steady operation at full power. It is found that there is an improvement in calculation time without compromising accuracy with FCCAL.

• Nuclear Engineering and Technology
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• v.42 no.2
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• pp.183-192
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• 2010

An electrochemical reduction of a mixture of metal oxides was conducted in a LiCl molten salt containing 3 wt% $Li_2O$ at $650^{\circ}C$. The oxide reduction was carried out by applying a current to an electrolysis cell, and the $Li_2O$ concentration was analyzed during each run. The concentration of $Li_2O$ in the electrolyte bulk phase gradually decreases according to Faraday's law due to a slow diffusion of the $O^{2-}$ ions. A hindrance effect of the unreduced metal oxides was observed for the reduction of the uranium oxide. Cs, Sr, and Ba of high heat-load fission products were diffused into and accumulated in the salt phase as predicted with thermodynamic consideration.