• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.2
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• pp.193-207
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• 2022

Thermodynamically, TRUO_x, REO_x, and SrO_x can be chlorinated using ammonium chloride (NH₄Cl) as a chlorinating agent, whereas uranium oxides (U₃O₈ and UO₂) remain in the oxide form. In the preliminary experiments of this study, U₃O₈ and CeO₂ are reacted separately with NH₄Cl at 623 K in a sealed reactor. CeO₂ is highly reactive with NH₄Cl and becomes chlorinated into CeCl₃. The chlorination yield ranges from 96% to 100%. By contrast, U₃O₈ remains as UO₂ even after chlorination. We produced U/REO_x- and U/SrO_x-simulated fuels to understand the chlorination characteristics of the oxide compounds. Each simulated fuel is chlorinated with NH₄Cl, and the products are dissolved in LiCl-KCl salt to separate the oxide compounds from the chloride salt. The oxide compounds precipitate at the bottom. The precipitate and salt phases are sampled and analyzed via X-ray diffraction, scanning electron microscope-energy dispersive spectroscopy, and inductively coupled plasma-optical emission spectroscopy. The analysis results indicate that REO_x and SrO_x can be easily chlorinated from the simulated fuels; however, only a few of U oxide phases is chlorinated, particularly from the U/SrO_x-simulated fuels.

• Applied Biological Chemistry
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• v.41 no.5
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• pp.331-336
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• 1998

The possibility of using waste brine from kimchi factory as a substrate for the production of the single cell protein was investigated. The growth of Pichia guilliermondii A9 isolated from waste brine was not inhibited by the NaCl up to 10% (w/v). BOD of the waste brine was reduced to one tenth after 24 hours of yeast culture. The addition of ammonium salt, phosphate, and micronutrients to the waste brine did not enhance the growth of P. guilliermondii A9. However, when the brine was enriched with juice from waste cabbage, the final cell mass increased proportionally with the amount of added organic material, suggesting a practical application for the treatment of two different types of waste produced during kimchi manufacturing.

• Nuclear Engineering and Technology
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• v.9 no.2
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• pp.75-81
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• 1977

This experiment has been carried out to determine the pH dependent cation exchange capacity concerning the sorption phenomenon of long-lived radionuclides contained in low-level liquid radioactive waste on various clay minerals. The pH dependent cation exchange capacity determined by Sawhney's method are used to the analysis of sorption phenomenon. About 70 percent of the total cation exchange capacity is contributed by the pH dependent CEC due to the negative charge originated naturally in clays in case of clinoptilolite, vermiculite and sodalite. It is sugested in this test that the high neutral salt CEC, that is, highly charged clays would show good fixation yield. The removal of radionuclides at the pH range more than pH 9 is considered the hydroxide precipitation of metal ion rather than the cation exchange. The Na-clay prepared by the method of successive isomorphic substitution with electrolyte showed a considerable improvement in removal efficiency for the decontamination.

• Journal of Radiation Protection and Research
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• v.27 no.2
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• pp.95-100
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• 2002

Conventional and modified electrolytic decontamination experiment were performed in the 1.7 M solution of sodium sulfate and sodium nitrate tot decontamination of carbon steel as the simulated metal wastes which have been produced in large amounts from nuclear power plants. Anode ant cathode were used as inconel and titanium respective. The reaction time and temperature were 1 hr and $25^{\circ}C$ The analyses were performed of the characteristics such as weight loss arid thickness change of metal waste. suspended solid in electrolyte and SEM observation. In modified electrolyte decontamination system with increased current density ranged from 0.1 to $0.6A/cm^2$, the metal waste showed thickness changes of $0.48{\pm}0.005$ to $67.7{\pm}0.02{\mu}m$ in 1.7 M sodium sulfate and those of $0.06{\pm}0.005$ to $17.7{\pm}0.05{\mu}m$ in sodium nitrate. Metal waste in modified electrolyte decontamination system showed the thickness change of $9.8{\pm}0.01{\mu}m$ while it reacted up to $3.7{\pm}0.03{\mu}m$ in conventional system with $0.3 A/cm^2$ of current density and 1.7 M sodium sulfate. Decontamination efficiencies of modified electrolytic process ate much hither than that of conventional electrolytic process when both are applied to metal waste.

• Journal of the Korean Society of Food Science and Nutrition
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• v.41 no.1
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• pp.33-40
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• 2012

This study investigated the lipid profiles for the possible improving activity of insoluble dietary fiber extracted from cabbage waste discarded during kimchi manufacture and market distribution. Enzyme-treated Insoluble Dietary Fiber (EIDF) is the remnants of crude dietary fiber (CDF) treated by ${\alpha}$-amylase, protease, and amyloglucosidase after removal of soluble materials from cabbage waste. The insoluble dietary fiber of CDF and EIDF in cabbage waste was $65.33{\pm}0.33%$ and $73.57{\pm}0.09%$, respectively. To examine the effects of EIDF on plasma cholesterol concentration, two group of rats were fed either a high fat diet (HFD) or a HFD containing 0.1~1.0% EIDF for 4 weeks. The body weight of all groups was not significantly different (p<0.05) but the body weight of EIDF+HFD groups was less than that of the HFD group (p<0.1). Compared with the HFD group, EIDF also lowered serum levels of total triglycerides to 11.2~23.3% and cholesterol to 26.8~28.5%. In EIDF+HFD groups, the HDL-cholesterol level increased by 7.2~26.1%, while LDL-cholesterol especially decreased by 51.0~61.4% and VLDL-cholesterol by 16.9~26.4%. The atherogenic index of EIDF+HFD groups was also reduced twice that of the HFD group. From these results, EIDF from cabbage waste could be a potential effective food ingredient for improving lipid profiles.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.4
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• pp.213-218
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• 2009

Electrowinning process recovers uranium with actinide elements from spent fuels and is a key step in the Pyroprocessing because of proliferation resistance. An analysis of heat transfer of the Electrowinning cell was conducted to develop basic tool for designing engineering-scale Electrowinner. For the calculation of the heat transfer, ANSYS CFX commercial code was adapted. As a result of the calculation, the vertical Heating Zone length had great effect upon temperature of LiCl-KCl eutectic salt. To maintain constant temperature in the salt, the Heating Zone length should be three times longer than the height of the salt. However, the argon and salt temperatures were barely affected by the Cooling Zone length. The temperature under the Cell cover was mainly influenced by the number of the cooling plates. When the cooling plates were installed more than the number of 5, temperature under the cover was maintained below $250^{\circ}C$. These temperature properties had similar tendency toward the temperature of the Cell which was measured from experiments, Simulated heat transfer information from this study could be used to design engineering-scale Electrowinner.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.1
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• pp.59-64
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• 2018

In case of high contents of rare earths in the LiCl-KCl salt, it is not easy to recover U and TRU metals as a usable resource form from LiCl-KCl eutectic salts generated from the pyroprocessing of spent nuclear fuel. In this study, a conversion of $UCl_3$ into an oxide form using $K_2CO_3$ and an electrodeposition of $NdCl_3$ into a metal form in $LiCl-KCl-UCl_3-NdCl_3$ system were conducted to resolve the problem. Before conducting the conversion, experimental conditions for the conversion were determined by performing a thermodynamic equilibrium calculation. In this study, almost all of $UCl_3$ disappeared in the LiCl-KCl salt when the injection of $K_2CO_3$ reached theoretical equivalent for the conversion, and then $NdCl_3$ was effectively electrodeposited as a metal form using liquid zinc cathode. After that, the LiCl-KCl salt became transparent, and uranium oxides were precipitated to the bottom of the LiCl-KCl salt. These results will be utilized in designing a process to separate U and rare earths in LiCl-KCl salt.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.1
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• pp.49-56
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• 2010

In the pyroprocessing of spent nuclear fuels, LiCl-KCl waste salt containing radioactive rare earth chlorides are generated. The radioactive rare earth oxides are recovered by co-oxidative precipitation of rare earth elements. The powder phase of rare eath oxide waste must be immobilized to produce a monolithic wasteform suitable for storage and ultimate disposal. The immobilization of these waste developed in this study involves a solid state sintering of the waste with host borosilicate glass and zinc titanate based ceramic matrix(ZIT). And the rare-earth monazite which synthesised by reaction of ammonium di-hydrogen phosphate with the rare earth oxides waste, were immobilzed with the borosilicate glass. It is shown that the developed ZIT ceramic wasteform is highly resistant the leaching process, high density and thermal conductivity.

• Journal of Nuclear Fuel Cycle and Waste Technology
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• v.1 no.1
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• pp.29-35
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• 2013

The United States Department of Energy (US DOE) is conducting research and development (R&D) activities under the Used Fuel Disposition Campaign (UFDC) to support storage, transportation, and disposal of used nuclear fuel (UNF) and wastes generated by existing and future nuclear fuel cycles. R&D activities are ongoing at nine national laboratories, and are divided into storage, transportation and disposal. Storage R&D focuses on closing technical gaps related to extended storage of UNF. Transportation R&D focuses on ensuring transportability of UNF following extended storage, and addressing data gaps regarding nuclear fuel integrity, retrievability, and demonstration of subcriticality. Disposal R&D focuses on identifying geologic disposal options and addressing technical challenges for generic disposal concepts in mined repositories in salt, clay/shale, and granitic rocks, and deep borehole disposal. UFDC R&D goals include increasing confidence in the robustness of generic disposal concepts, reducing generic sources of uncertainty that may impact the viability of disposal concepts, and developing science and engineering tools to support the selection, characterization, and licensing of a repository. The US DOE has also initiated activities in the Nuclear Fuel Storage and Transportation (NFST) Planning Project to facilitate the development of an interim storage facility and to support transportation infrastructure in the near term.

• Korean Journal of Materials Research
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• v.27 no.12
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• pp.672-678
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• 2017

In the production of zirconium cladding tube, a pickling acid solution is used to remove surface contaminants, which generates tons of pickling acid waste. The waste pickling solution is a valuable resource of Hf-free Zr. Many studies have investigated separating the Hf-free Zr source from the waste pickling acid. The results showed that $Ba_2ZrF_8$ precipitates prepared from the waste pickling acid were useful as an electrolyte for the electrorefining of Zr in molten salt. In the present work, electrorefining was performed in a $Ba_2ZrF_8-LiF$ binary electrolyte to recover Zr from a Hf-free CuZr ingot anode prepared by electroreduction. Before electrorefining, two pretreatments are performed. First, electrolyte melting was carried out to determine the eutectic temperature, and second, the electrolyte was treated to eliminate impurities, mainly hydride. After electrorefining, the cathode deposits were analyzed by $O_2$ gas analyzer and SEM-EDX to explore the possibility of recovering nuclear-grade Zr metal. Moreover, the anode was analyzed by SEM-EDX to determine the Zr dissolution depth.