• 대한방사선방어학회:학술대회논문집
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• 2011.11a
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• pp.242-243
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• 2011

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.298-298
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• 2002

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.407-417
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• 2004

This study is to verify radiation detection method by using $\alpha$-spectroscopy and ${\gamma}$-spectroscopy for concretes and components which will be generated during the decommissioning of the uranium conversion plant. Components and inside walls of the building were contaminated with natural uranium materials. Some parts of the stainless steel pipes and concretes of the walls were sampled and analyzed their alpha and gamma activities respectively. Alpha and gamma activities are well matched each other in the range of high activity region to 0.01 Bq/g and gamma activities are over estimated comparing alpha activities corresponded in below 0.005 Bq/g region for the natural uranium of AUC sample. The $^{238}U$ originated from natural products of conversion process could be distinguished by measuring $^{214}Pb$ or $^{214}Bi$ and $^{234}Th$ or $^{234m}Pa$. Uranium contaminations mainly are in the wall surface of the plant. Decontamination process of generating wastes which can be reached tp background level gamma activities measured by gamma spectroscopy can also be used to conservative assessment data.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.322.2-322
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• 2002

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.689-695
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• 2003

Electrochemical decontamination process has been applied for recycle or self disposal with authorization of large amount of metallic wastes contaminated with uranium compounds such as $UO_2$, ammonium uranyl carbonate (AUC), ammonium di-uranate (ADU), and uranyl nitrate(UN) with tributylphosphate(TBP) and dodecane, which are generated by dismantling the contaminated system components and equipment of a retired uranium conversion plant in Korea Atomic Energy Research Institute (KAERI). Electrochemical decontamination for metallic wastes contaminated with uranium compounds was evaluated through the experiments on the electrolytic dissolution of stainless steel as the material of the system components in neutral salt electrolytes. The effects of type of neutral salt as the electrolyte, current density, and concentration of electrolyte on the dissolution of the materials were evaluated. Decontamination performance tests using the specimens taken from a uranium conversion plant were quite successful with the application electrochemical decontamination conditions obtained through the basic studies on the electrolytic dissolution of structural material of the system components.

• Nuclear Engineering and Technology
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• v.25 no.4
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• pp.548-551
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• 1993

A simple and new uranium analysis technique for raffinate solution of nuclear fuel conversion process was developed using a time-resolved laser-induced fluorimetry. The addition of 4 M-phosphoric acid more than 10 times in volume to the raffinate sample was found to be efficient for obtaining stable uranium fluorescence signal which was not influenced by many fluorescence quenchers. A calibration curve of a good linearity for the fluorescence intensity vs. the uranium concentration was obtained at the range of 3.0$\times$10$^{-6}$－6.0$\times$10$^{-5}$ M U $O_2$$^{2＋}$ in the raffinate samples.

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.613-623
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• 2016

Molybdenum-99 ($^{99}Mo$) is the most important isotope because its daughter isotope, technetium-99m ($^{99m}Tc$), has been the most widely used medical radioisotope for more than 50 years, accounting for > 80% of total nuclear diagnostics worldwide. In this review, radiochemical routes for the production of $^{99}Mo$, and the aspects for selecting a suitable process strategy are discussed from the historical viewpoint of $^{99}Mo$ technology developments. Most of the industrial-scale $^{99}Mo$ processes have been based on the fission of $^{235}U$. Recently, important issues have been raised for the conversion of fission $^{99}Mo$ targets from highly enriched uranium to low enriched uranium (LEU). The development of new LEU targets with higher density was requested to compensate for the loss of $^{99}Mo$ yield, caused by a significant reduction of $^{235}U$ enrichment, from the conversion. As the dramatic increment of intermediate level liquid waste is also expected from the conversion, an effective strategy to reduce the waste generation from the fission $^{99}Mo$ production is required. The mitigation of radioxenon emission from medical radioisotope production facilities is discussed in relation with the monitoring of nuclear explosions and comprehensive nuclear test ban. Lastly, the $^{99}Mo$ production process paired with the Korea Atomic Energy Research Institute's own LEU target is proposed as one of the most suitable processes for the LEU target.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.329.2-329
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• 2001

• Membrane and Water Treatment
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• v.3 no.4
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• pp.231-242
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• 2012

The performance of a nanofiltration membrane for treatment of a low-level radioactive liquid waste was investigated through static and dynamic tests. The liquid waste ("carbonated water") was obtained during conversion of $UF_6$ to $UO_2$. In the static tests membrane samples were immersed in the waste for 24, 48 or 72 h. The transport properties of the samples (hydraulic permeability, permeate flow, selectivity) were evaluated before and after immersion in the waste. In the dynamic tests the waste was permeated in a permeation flow front system under 0.5 MPa, to determine the selectivity of NF membranes to uranium. The surface layer of the membrane was characterized by zeta potential, field emission microscopy, atomic force spectroscopy and infrared spectroscopy. The static test showed that the pore size distribution of the selective layer was altered, but the membrane surface charge was not significantly changed. 99% of uranium was rejected after the dynamic tests.

• 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.