• Nuclear Engineering and Technology
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• v.13 no.1
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• pp.22-35
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• 1981

The U$_3$O$_{8}$ requirements are estimated for the high, intermediate, and low growth projections of nuclear power in Korea. To each projection, four illustrative reactor-mix strategies and four fuel cycle options are applied for estimating the requirements. The reactor types considered are PWR, PHWR. and FBR. The fuel cycles considered are once-through cycle, U/Pu recycle, and improved once-through cycle. Also the amount of Pu-fissile recovered from U recycle is estimated. The maximum cumulative (to the year 2000) requirements of U$_3$O$_{8}$ occupy about 4 to 5 percent of the WOCA requirements and are about 23 times larger than the U$_3$O$_{8}$ resources in Korea. For the high nuclear power growth projection, the cumulative amount of Pu-fissile recovered from U recycle is sufficient for the startup of 2 units of 1200 MWe fast reactors by the year 2000. 2000.

• Nuclear Engineering and Technology
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• v.31 no.5
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• pp.455-464
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• 1999

The effect of TiO$_2$ on the sintering behavior of mixed UO$_2$-U$_3$O$_{8}$ Powder compacts has been investigated using the U$_{3}$O$_{8}$ powder made tv oxidation of defective UO$_{2}$ pellets. Without TiO$_2$, UO$_2$ pellet density is inversely proportional to U$_3$O$_{8}$ content and is below 94 %TD in the U$_3$O$_{8}$ range above 15 wt%. Using more than 0.1 wt % TiO$_2$, however, the density decreases slightly with U$_3$O$_{8}$ content and thus is higher than about 94% TD in the whole range of U$_3$O$_{8}$ content. The grain sizes of UO$_2$ pellets with more than 0.1 wt % TiO$_2$are larger than about 30${\mu}{\textrm}{m}$. Therefore, the U$_3$O$_{8}$ Powder can be reused without any restriction on its amount in UO$_2$ pellet fabrication by sintering the mixed UO$_2$-U$_3$O$_{8}$ compact with the aid of TiO$_2$. Mechanisms for densification and grain growth are proposed and discussed, based on a dilatometry study and an examination of microstructure. microstructure.

• Journal of Microbiology and Biotechnology
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• v.31 no.3
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• pp.419-428
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• 2021

To efficiently recycle GH78 thermostable rhamnosidase (TpeRha) and easily separate it from the reaction mixture and furtherly improve the enzyme properties, the magnetic particle Fe3O4-SiO2-NH2-Cellu-ZIF8 (FSNcZ8) was prepared by modifying Fe3O4-NH2 with tetraethyl silicate (TEOS), microcrystalline cellulose and zinc nitrate hexahydrate. FSNcZ8 displayed better magnetic stability and higher-temperature stability than unmodified Fe3O4-NH2 (FN), and it was used to adsorb and immobilize TpeRha from Thermotoga petrophilea 13995. As for properties, FSNcZ8-TpeRha showed optimal reaction temperature and pH of 90℃ and 5.0, while its highest activity approached 714 U/g. In addition, FSNcZ8-TpeRha had better higher-temperature stability than FN. After incubation at 80℃ for 3 h, the residual enzyme activities of FSNcZ8-TpeRha, FN-TpeRha and free enzyme were 93.5%, 63.32%, and 62.77%, respectively. The organic solvent tolerance and the monosaccharides tolerance of FSNcZ8-TpeRha, compared with free TpeRha, were greatly improved. Using naringin (1 mmol/l) as the substrate, the optimal conversion conditions were as follows: FSNcZ8-TpeRha concentration was 6 U/ml; induction temperature was 80℃; the pH was 5.5; induction time was 30 min, and the yield of products was the same as free enzyme. After repeating the reaction 10 times, the conversion of naringin remained above 80%, showing great improvement of the catalytic efficiency and repeated utilization of the immobilized α-L-rhamnosidase.

• Applied Chemistry for Engineering
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• v.7 no.6
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• pp.1164-1173
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• 1996

Treating methods and characteristics of waste from a nuclear fuel powder conversion plant were studied. To recovery or treat a trace uranium in liquid waste, the ammonium uranyl carbonate(AUC) filtrate must be heated for $CO_2$ expelling, essentially. Uranium content of final treated waste solution from fuel powder processes for a heavy water reactor(HWR) could be lowered to 1 ppm by the lime treatment after the ammonium di-uranate(ADU) precipitation by simple heating. Otherwise, in case of the waste from fuel powder processes for a pressurized light water reactor(PWR), it is result in 0.8 ppm as a form of uranium peroxide such as $UO_4{\cdot}2NH_4F$ compounds. Optimum condition was found at $101^{\circ}C$ by the simple heating method in case of HWR powder process waste. And in case of PWR powder process waste, optimum condition could be obtained by precipitating with adding hydrogen peroxide and adjusting at pH 9.5 with ammonia gas at $60^{\circ}C$ after heating the waste In order to expelling $CO_2$. As the characteristics of recovered uranium compounds, median particle size of ADU was increased with pH increasing in case of HWP waste. Also, in case of uranium proxide compound recovered from PWR waste, the property of $U_3O_8$ power obtained after thermal treatment in air atmosphere was similar to that of the powder prepared from AUC conversion plant.