[en] Desorption characteristics of C-14 adsorbed on spent resin as H¹⁴CO₃ ion type by applying various stripping solutions were analyzed, and some experiments for gasification of C-14 to CO₂ gas with were also performed. Based on these results, the process concept for spent resin treatment was suggested. Real spent resin was prepared from sampling in storage tank in site 1 of Wolseung Nuclear Power Plant. Desorption characteristics of C-14 and cations of Cs, Co from spent IRN-150 resin was evaluated. Desorption efficiency of C-14 from spent resin by using H₃PO₄ desorption solution was over 96% regardless of C-14 amount on initial spent resin when comparing a activity of C-14 on initial spent resin. Also, desorption percent of cation of Cs, Co from anion ion-exchange resin (IRN-77) showed that Co-60 was below 1%, Cs-134, 137 was in a range of 2 ∼ 5%. Fundamental studies include an development of adsorbent manufacturing technology and its performance evaluation for C-14 gas trapping, the adsorption process by adopting gas circulation method was suggested for the design of ¹⁴CO₂ gas treatment system generated from spent resin treatment process. In order to predict adsorbent performance of CO₂ trapping, modelling was carried out to verify the breakthrough curves of CO₂ trapping by using soda lime adsorbent. The effect of humidity on CO₂ trapping by using soda lime adsorbent was modelled via chemical reaction in porous media. Assessment of the state-of-the-arts on the solidification of the used adsorbent showed that the cement matrix would be the best-available binder from the view points of the matrix compatibility, properties of the final waste form, simplicity of the process and relatively low cost

[en] A new ion exchanger with porous silica as a supporting material and diphosphonic acid as a functional chelating group has been developed at ANL for the effective removal of transition metals and actinide ions from very acidic radioactive liquid wastes. The applicability of this resin for the treatment of low- and/or intermediate-level aqueous waste from nuclear power plants (NPP) has not been reported in scientific literature, but is under study now in Korea. The major radioisotopes in NPP radioactive liquid waste are Cs and Co in neutral pH ranges. This study on the thermal stabilization of metal-loaded waste resin has been carried out in parallel with the sorption experiment. Thermal treatment of metal (Co, Cs or U) loaded resin was accomplished to see the possibility of enhancing the safety and stability of the final product during transportation and disposal. In this paper, characteristics of the metal-loaded resins before and after heat treatment at three different thermal conditions were investigated and compared with each other to see the effectiveness of the thermal treatment method

[en] The oxygen-enriched combustion concept, which can minimize off-gas production, has been applied to the incineration of combustible uranium-containing wastes from a nuclear fuel fabrication facility. A simulation for oxygen combustion shows the off-gas production can be reduced by a factor of 6.7 theoretically, compared with conventional air combustion. The laboratory-scale oxygen enriched incineration (OEI) process with a thermal capacity of 350 MJ/h is composed of an oxygen feeding and control system, a combustion chamber, a quencher, a ceramic filter, an induced draft fan, a condenser, a stack, an off-gas recycle path, and a measurement and control system. Test burning with cleaning paper and office paper in this OEI process shows that the thermal capacity is about 320 MJ/h, 90 % of design value and the off-gas reduces by a factor of 3.5, compared with air combustion. The CO concentration for oxygen combustion is lower than that of air combustion, while the O2 concentration in off-gas is kept above 25 vol % for a simple incineration process without any grate. The NOx concentration in an off-gas stream does not reduce significantly due to air incoming by leakage, and the volume and weight reduction factors are not changed significantly, which suggests a need for an improvement in sealing

[en] Synroc (Synthetic Rock), a titanate-based ceramic originally proposed by Prof. A. Ringwood (ANU) and designed for the immobilization of high level nuclear waste (HLW), consists of three principal phases such as hollandite, zirconolite and perovskite. Nearly all the fission products and actinides in HLW can be incorporated as solid-solution in at least one of these phase. The preferred form of Synroc can be obtained up to 20 % of high level waste calcine to form dilute solid solution. The constituent minerals, or close structural analogues, have survived in a wide range of geochemical environments for periods of 20-2000 Myr while immobilizing the same elements present in nuclear waste. A dense, compact, and mechanically strong form of Synroc can be formed by hot pressing reactive precursor powders at about 1200 dg C and 20 MPa. In this state-of-the-art report, formulation method and characterization of Syroc with respect to the crystal structure, the consisting substances, types, etc. were reviewed. Additionally, a new promising powder process, ^Combustion Process^, was proposed and the properties of the combustion-synthesized powder were described. An international cooperative program between JAERI and ANSTO, and US patents for early Synroc research in Australia were also introduced. From the literatures review, Synroc is expected to have advantages in using as an immobilizer of HLW. Therefore, a systematic research to develop the Synroc is needed. (author). 53 refs., 2 tabs., 16 figs

[en] An advanced spent fuel management process based on Li reduction of the oxide spent fuel to a metallic form will generate a LiCl waste. Zeolite Na-A has been reported as a promising immobilization medium for waste salt with CsCl and SrCl₂. However, Sodium is accumulated as an ionic form (Na⁺-ion) in molten salt during ion exchange step between Na⁺-ion in the zeolite- A and Li⁺-ion in the molten salt. Therefore, zeolite Na-A need to be replaced by the Li-type zeolite for recycling the salt waste by removing the Cs and Sr ions. In this study, we prepared the bead type zeolite Li-A hydrothermally from a bead type zeolite 4A by P. Norby and M.C. Mascolo method, and its preparation characteristics of zeolite Li-A was investigated

[en] HEPA filters are widely used in the nuclear fields as a final off-gas cleaning unit. To assess the applicability of vitrification technology either to treat used filter media or to produce borosilicate glass medium for the solidification of alpha-contaminated wastes, various waste glasses of different compositions were fabricated by melting mixture of HEPA filter media and inorganic additives. Physicochemical properties such as microhardness, density, thermal expansion, and short-term leaching behavior were characterized. XRD analysis showed that amorphous glasses were formed for a wide range of mixing ratio. Leach resistances, measured by PCT-B leach tests, were superior to that of EA (Environmental Assessment) glass. Other properties were similar to those of glass media used for the vitrification of high-level radioactive wastes in foreign countries

[en] In case of zone freezing process, the concentration extent of Cs and Sr is increases with decreasing ascending velocity. When ascending velocity is 1.67 mm/hr, within about 10% of a solid salt, about 90% of Cs and Sr is present, which means Cs and Sr could be concentrated by zone freezing method. The concentration efficiency (or separation efficiency) in layer crystallization process is affected by various experimental conditions. In this study, the distribution coefficient K_diff, defined as the ratio of the Cs and Sr concentration of the formed solid layer and the initial feed concentration of the melt. The distribution coefficient is strongly affected by mass-growth rate of crystal, i.e. cooling rate. The high cooling rates caused the crystals to grow rapidly, thus leading to low separation efficiency

[en] In order to fabricate a monolithic waste form containing RE oxides, a vitrification at a high temperature or a ceramization by a HIP method is required. In this study, a series of monolithic wasteform with high waste loading were successfully produced at a mild condition, where the chemical structure was equivalent to the product by a high temperature process or a monolithic wasteform consisting of a durable ceramic host matrix for immobilizing RE elements

[en] The role of uranium chloride salt (UCl3) is to stabilize the initial cell voltage between electrodes in the electrorefining reactor. The process to produce a uranium chloride salt includes two steps: a reaction process of gaseous chlorine with liquid cadmium to form the CdCl2 occurring in a Cd layer, followed by a process to produce UCl3 by the reaction of U in the LiCl-KCl eutectic salt and CdCl2 The apparatus for producing UCl3 consists of a chlorine gas generator, a uranium chlorinator, a Cd distiller, the pelletizer, and a off-gas and a dry scrubber. The temperature of the reactants is maintained at about 600 .deg. C. After the reaction is completed in the uranium chlorinator, The salt products is transferred to the Cd distiller to decrease residual Cd concentration in the salts, and then salt is transferred to the mould of pelletizer by a transfer system to make pellet type salt. Making pellet type LiCl-KCl-UCl3 salt for electrorefining was carried out using the chlorinator, Cd distiller, and pelletizer. Salt transfer carried out by salt transfer equipment heated 500 .deg. C. The Cd concentration of final salt products distillated at 60 torr, 2 hrs, 600 .deg. C was 200 ppm from the ICP, XRD analysis. And pellet type salt products were fabricated by using the mould of pelletizer at 90∼130 .deg. C

[en] A fundamental study for the fabrication of target material as NaI chemical form in transmutation of LLFP ¹²⁹I was carried out. Iodine recovery experiment using the silver ion-exchanged zeolite, which is commonly used to trap radioiodine generated from high temperature treatment process of spent fuel, was performed using various extraction solutions. Extraction experiment of iodine reveals that about 100% recovery percent of iodine was obtained when Na₂S solution over 0.1M concentration and N₂H₄ over 1 M were applied. Radioiodine recovery percent using ¹³¹I iodine tracer solution represented a similar result obtained from recovery experiment using stable ¹²⁷I element. Evaporation method was applied to produce the NaI powder from iodide ion-contained solution. White-colored powder obtained from evaporation was identified NaI chemical form by XRD analysis