[en] In this study, the proper methods to treat the following four kinds of wastewater containing mercury were investigated through the experiment. These wastewater were generated from the processes to manufacture the special battery. - Amalgamation wastewater - Amalgamation rinsing wastewater - Alkali wastewater containing mercury - Alkali rinsing wastewater

[en] Radioactive organic wastes containing acetone, alcohol, and particularly TriButyl Phosphate (TBP)/dodecane contaminated with uranium are extracted from the PUREX process and the decontamination of related equipment. An evaporation method that utilizes existing DU oxidation apparatuses and ventilation systems and a typical muffle furnace installed with an aspirating system are adopted. A separation method using phosphoric acid especially for the TBP/dodecane waste is also studied and evaluated. The results show that a simple evaporation process is utilizable for wastes containing acetone or alcohol with a lower boiling point. A modified muffle furnace is more appropriate to dispose directly of organic wastes having a higher boiling point, such as TBP/dodecane, without generating a condensed waste solution. It is recommended that, when the uranium concentration of TBP/dodecane waste is much higher than stipulated levels, separation technology should be applied to remove uranium from the mixture. Each type of solvent after separation can then be considered disposable below the regulatory limit in the modified furnace discussed in this study

[en] In this report, research and development results on the bituminization of radioactive wastes are closely reviewed, especially those regarding waste treatment technologies, waste solidifying procedures and the characteristics of asphalt and solidified forms. A new concept of the bituminization method is suggested in this report which can improve the characteristics of solidified forms. Stable solid forms with high leach resistance, high thermal resistance and good compression strength were produced by the suggested bituminization method, in which spent polyethylene from agricultural farms was added. This report can help further research and development of improved bituminized forms of radioactive wastes that will maintain long term stabilities in disposal sites. (author). 59 refs., 19 tabs., 18 figs

[en] This report is contained research status of foam separation technique and thus theoretical backgrounds, the kinds of surfactants using for the foam separation technique, their characteristics, general structure and role, surfactant absorption mechanism at liquid/vapor/solid interfaces, effectiveness and efficiency were interpreted with well known models. Ion flotation and precipitate flotation which are applicable for the treatment of very low radioactive liquid wastes were analyzed on the effect of pH, foreign ions, initial concentration of metal ion through the recent presented papers. As the result of technical analysis of foam separation technique, foam separation technique seems to be applicable for the treatment of very low radioactive liquid wastes such as laundry and shower waste. (author). 42 refs., 2 tabs., 37 figs

[en] Chemical wastes containing small amounts of uranium can not be disposed of them after treatment as an industrial waste, because the uranium concentration in the final dry cake exceeds the exemption level. Especially for the removal of uranium in this study, the method for immobilizing Diphosil powder within alginate beads is adopted to make a bead form from a powdered resin. Sodium alginate bead itself showed a capability to uptake uranium to above 60%, but the value was decreased to below 30% after equilibrium. The adsorption rate of uranium increased with the increasing content of Diphosil in the sodium alginate bead. Diphosil resin itself showed very fast uptake of uranium from early stages, and then the rates were leveled off. Diphosil bead showed an improved capability to uptake uranium considering the pure Diphosil content in the composite bead, and provide a considerable potential for further applications of a continuous process by using Diphosil as a bead form.

[en] The KIJANG Research Reactor (KJRR) is a radioisotope production reactor and one of its main purposes is producing fission ⁹⁹Mo, which is the most widely used medical radioisotope. The waste generation from the fission ⁹⁹Mo production process is inevitable and disposal plan for the radioactive wastes should be set up for mass production of the fission ⁹⁹Mo. KAERI has developed the ⁹⁹Mo production process and the production of ⁹⁹Mo will be followed soon after the criticality of the KJRR. The ILLW in the production process has been a critical issue, but the treatment process of the ILLW is not developed yet. This paper introduces the treatment plan of the ILLW. KAERI has selected the conventional cementation method and development study was started last year. The development plan shows that the treatment facility could be operable within 4⁓10 years after start of the production.

[en] Fenton's reagent is applied to directly decompose the ion-exchange resins, IRN-78 and the mixed resin with IRN-77. The newly applied procedures is to dry the resin first and the catalyst solution is completely absorbed into the resin, then a limited dose of H₂0₂ is introduced for an effective reaction between the reagents within the resin. As a characteristic on the decomposition of IRN-78, the resin mixture should be heated to 40 .deg. C to induce the initial reaction and lag time is also needed for about 20 minutes until the main reaction occurs. The effectiveness of the decomposition is investigated using CuSo₄, Cu(NO₃)₂ and FeSo₄ as a catalyst and the decomposition rate is compared depending on the concentration of each catalyst and the amount of H₂O₂. The most effective catalyst was found to be FeSO₄ for IRN-78 alone and the mixed resin with IRN-77, and FeSO₄ showed a special effect that the reaction was initiated without heating and a lag time. Furthermore, the optimum concentration of the catalyst for each resin and the mixed one is suggested in the view point of the amount of H₂O₂ needed and the stability of the decomposition reaction

[en] Fenton's Reagent is applied to directly dissolve the cation-exchange resin, IRN-77. The characteristics of the experimental procedure is to dry the resin first and FeSO₄ solution is completely absorbed into the resin, and then H₂O₂ is introduced later for an effective reaction between the reagents within the resin. As a characteristic of the dissolution, the lag time is needed for about 1 hour until the main reaction is occurred, which was more affected with the less concentration of FeSO₄ and the less initial dose of H₂O₂. The dose of F₂O₂ was equally divided into the early stage and the later stage after the initial reaction to provide an effective and safe reaction condition. The optimum conditions is appeared that the concentration of FeSO₄ is 0.9 M and the dose of 15% H₂O₂ solution is 6-7 volume for the dissolution of unit weight of IRN-77. The effect of the heating on the lag time was checked and the time could be reduced within 5 minutes at 50 .deg. C, which is a relatively low temperature. The large amount of the resin, 5g and 10g, was also completely decomposed by increasing the dose of H₂O₂ to 9-10 volume ratio

[en] A pilot size of electrochemical flushing equipment will be manufactured suitable to the contamination characteristics of radioactive soil and concrete stored in KAERI radioactive waste storage. An optimal reagent and an optimal decontamination conditions should be decided through many experiments. - Contamination characterises analysis of TRIGA radioactive soil and concrete - Manufacture of pilot-scale electrochemical flushing equipment - Manufacture and improvement of suitable electrochemical flushing equipment for contamination characteristics in pilot size - Decontamination experiments of electrochemical flushing equipment in a pilot scale

[en] Chemical wastes are generated from nuclear facilities and R and D laboratories, but the uranium concentration in the final dried cake is evaluated into 11.2 Bq/g, which exceeds the exemption level of 10 Bq/g for each U isotopes, so the cake is categorized into a radioactive waste. Acid dissolution was applied to extract uranium from the waste sludge, and uranium adsorption on the dissolved solution was experimented by using IRN-77 and Diphosil bead. A large amount of resin was required to get above 80% of uranium removal, which was found to be due to a large amount of metal ions simultaneously dissolved from the precipitates with uranium. As an alternative method, acid dissolution is applied to the dewatered wet cake of the sludge, and the natural evaporation method is adopted for the dissolved solution. The uranium concentration of the dissolved solution was estimated to be 6.97 E-01 Bq/ml, and the specific activity of the final waste sheets is evaluated to be 4.3 Bq/g. These results lead to the suggestion that the application of acid dissolution to the wet cake and the natural evaporation for the dissolved solution is an effective treatment method for chemical wastes containing uranium.