• 한국세라믹학회:학술대회논문집
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• 한국세라믹학회 2003년도 춘계총회 및 연구발표회 초록집
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• pp.28.2-28
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• 2003

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2004년도 학술대회지
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• pp.145-147
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• 2004

• Nuclear Engineering and Technology
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• 제29권4호
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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.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2003년도 가을 학술논문집
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• pp.363-367
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• 2003

우라늄 산화물의 금속전환을 위해 고온 용융염 중에서 전기화학적 환원공정에 대한 관심이 고조되고 있다. 본 공정은 우라늄 산화물뿐만 아니라 다른 악틴족 원소 산화물 및 일부 희토류원소 산화물 역시 금속으로 환원되는 장점을 가지고 있다. 이러한 금속산화물들은 독창적으로 고안된 일체형 음극 및 불활성 양극을 이용하여 금속으로 환원되며, 음극에서 발생된 산소 이온은 양극으로 전달되어 산화됨으로서 산소기체를 발생시킨다. 용융염 중에서 알칼리 및 알칼리토류 산화물에 대한 전기화학적 거동은 아직 완전히 밝혀지지 않았으며, 후행핵연료주기의 단위공정으로서 개발중에 있다. 사용후핵연료의 열 부하는 주로 세슘 및 스트론슘에 의한 것으로, LiC1 용융염 중에서 세슘, 스트론슘 및 바륨 산화물에 대한 용해 속도 및 환원전위를 고찰하였다.

• Nuclear Engineering and Technology
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• 제40권1호
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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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• 제51권7호
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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.

• 에너지공학
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• 제5권2호
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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.

• 전기화학회지
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• 제16권3호
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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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• 제52권12호
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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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• 제42권2호
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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.