Organic waste generated by small and medium-sized (S&M-sized) metal decontamination in NPP decommissioning. To lower the concentration of these organic substances for a level acceptable at the disposal site, the project of “Development of Treatment Process of Organic Decontamination Liquid Wastes from Decommissioning of Nuclear Power Plants” is being carried out. The conditioning and treatment process of organic liquid waste was designed. Also, the literature was investigated to make simulated organic liquid waste, and the composition of these waste was analyzed and compared. As the decontamination agent, organic acids such as EDTA, oxalic acid, citric acid are used. The sum of the concentrations of these organic materials was set to a maximum value of 1,000 ppm. The major metal ions of the decontamination liquid waste estimated are 59Fe, 51Cr, 54Mn, 63Ni, and the concentrations are respectively 527, 163, 161, 159 ppm. Additional major metal ions are 60Co, 58Co, 137Cs. 58Co is replaced by 60Co because it has the same chemical properties as 60Co. Unlike the HLW, the contamination level of S&M-sized metal in primary system was quite low, so 60Co is set to 2,000 Bq/g. Considering the contribution of fission and gamma ray dose constant, 137Cs was estimated to 360 Bq/g. Also, suspended solids of decontamination liquid waste were set at 500 ppm. Under these assumptions, the simulated organic liquid waste was made, and then organic substances and metal ions were analyzed with TOC analyzer and ICP-OES. The TOC analysis value was expected to 392 ppm in consideration of the equivalent organic quantity. the test result was 302 ppm. Some of organics appears to have been decomposed by acid. The values of metal ions (Fe3+, Cr3+, Mn2+, Ni2+) analyzed by ICP-OES are 139, 4, 152, 158 ppm, respectively. A large amount of Cr3+ and Fe3+ were expected to exist as ions, but they existed in the form of suspended solid. Mn2+ and Ni2+ came out similar to the expected values. The designed conditioning and treatment process is largely divided into pretreatment, conditioning, and decomposition processes. After collecting in the primary liquid waste storage tank, large particulate impurities and suspensions are removed through a pretreatment process. In the conditioning process, treated liquid waste passes through UF/RO membrane system, and pure water is discharged to the environment after monitoring. Concentrated water is decomposed in the electrochemical catalyst decomposition process, then this water secondarily passes through the RO membrane system and then discharged to the environment after monitoring. Through an additional experiment, the conditioning and treatment process will be verified.

Uranium-235, used in nuclear power generation, produces a lot of radioactive waste. Among radioactive waste nuclides, I-129 is problematic due to its long half-life (1.57×107 y) with high mobility in the environment. It should be captured and immobilized into a geological disposal environment through a stable waste form. In this study, various additives including Al, Bi, Pb, V, Mo and W were added to silver tellurite glass to prepare a matrix for immobilizing iodine, and its thermal and leaching properties were evaluated. To prepare glass, the glass precursor mixture was placed in alumina crucibles and heated at 800°C for 1 h. Except for aluminum, there was no significant loss of constituent elements. The loading of iodine in the matrix was approximately 11-15% by weigh, excluding oxygen. The normalized releases of all the elements obtained by PCT-A were below the order of 10-1 g/m2, which satisfies US regulation (2 g/m2). Differential scanning calorimetry was performed to evaluate the thermal properties of the glass samples. The glass transition temperature (Tg) increased by adding such as V2O5, MoO3, or WO3. The similar relative electrostatic field values of V2O5, MoO3, and WO3 could provide sufficient electro static field to the TeO2 interacting with the non-bridging oxygen forming Te-O-M (M = V, Mo, W) links. The addition of MoO3 or WO3 in the silver tellurite glass system increased glass transition temperature (Tg) and crystallization temperature (Tc) while maintaining the glass stability.

For safe management of spent nuclear fuels, they should be delivered to repository or waste disposal site. As the amount of spent nuclear fuel transportation is expected to increase in the future due to the provision of an intermediate storage facility, the necessity to secure transportation cask is emerging. In order to secure the spent nuclear fuel transportation cask, it is necessary to analyze the regulatory processes for domestic and foreign spent nuclear fuel transportation cask. In this study, the regulatory processes for domestic and foreign spent nuclear fuel transportation cask was analyzed. In this study, the IAEA, US, and Korea spent nuclear fuel transportation cask regulatory processes were analyzed. The domestic and foreign spent nuclear fuel transportation cask regulatory processes consist of design phase, manufacturing phase, and operation phase. In the design stage, the transport requirements are designed in accordance with the safety requirements of international organizations and countries. The application to be submitted when applying for approval should include a safety analysis report, evidence proving compliance with safety requirements et al. In the manufacturing stage, it is a stage to check whether the safety requirements are satisfied before the first use after manufacturing the transportation cask. Inspections include welding inspection, leakage inspection, shielding inspection, and thermal inspection. In the operation stage, it is a stage of periodically performing inspections for continuous maintenance of the package when the transportation cask is used. The inspection items to be performed are similar to the manufacturing stage and typically include performance inspection of components and leakage inspection. In this study, domestic and foreign spent nuclear fuel transportation cask regulatory processes were analyzed. It was found that the domestic and foreign spent nuclear fuel transportation cask regulatory processes consist of the design phase, the manufacturing phase, and the operation phase. The results of this study can be used as basic data for policy decision-making for the spent nuclear fuel cask.

The aim of this study was to investigate the effect of isolated lactic acid bacteria (LAB) on the quality of high moisture rye silage. Rye forage (Secale cereale L.) was harvested at the heading stage (27.3% of dry matter (DM)) and cut into approximately 3-5 cm lengths. Then, the forage divided into 4 treatments with different inoculants: 1) No additives (CON); 2) Lactobacillus brevis strain 100D8 at a 1.2 x 105 colony-forming unit (cfu)/g of fresh forage (LBR); 3) Leuconostoc holzapfelii strain 5H4 at a 1.0 x 105 cfu/g of fresh forage (LHO); and 4) Mixture of LBR and LHO (1:1 ratio) applied at a 1.0 x 105 cfu/g of fresh forage (MIX). About 3 kg of forage from each treatment was ensiled into a 20 L mini-bucket silo in quadruplicate for 100 days. After silo opening, silage was collected for analyses of chemical compositions, in vitro nutrient digestibilities, fermentation characteristics, and microbial enumerations. The CON silage had the highest concentrations of neutral detergent fiber and acid detergent fiber (p = 0.006; p = 0.008) and a lowest in vitro DM digestibility (p < 0.001). The pH was highest in CON silage, while lowest in LBR and MIX silages (p < 0.001). The concentrations of ammonia-N, lactate, and acetate were highest in LBR silage (p = 0.008; p < 0.001; p < 0.001). Propionate and butyrate concentrations were highest in CON silage (p = 0.004; p < 0.001). The LAB and yeast counts were higher in CON and LHO silages compare to LBR and MIX silages (p < 0.001). However, the mold did not detect in all treatments. Therefore, this study could conclude that L. brevis 100D8 and Leu. holzapfelii strain 5H4 can improve the digestibility and anti-fungal activity of high moisture rye silage.

Laser scabbling experiments were conducted with the aim of developing concrete decontamination technology. Laser scabbling system contains a 6 kW fiber laser (IPG YLS-6000, λ=1,070 nm) and optical head, which are connected with process fiber (core dia.: 600 μm, length: 20 m). Optical head consists of two lenses (f = 160 mm and 100 mm) to collimate and focus laser beam. The focused laser beam is passed through the small diameter of nozzle (throat dia.: 3 mm) to prevent the laser-produced debris into head. And then, the focused beam is directed toward concrete block as continuously diverging. The diverged laser beam was incident on the high-strength concrete with 300 mm (length) × 300 mm (height) × 80 mm (width) to induce explosive spalling on the concrete surface. The optical head was moved by X-Y-Z manipulate coupled with a computerized numerical control while scabbling. We have observed not only active spalling on the concrete surface but energetic scattering of laserproduced debris when scabbling on high-strength concretes. It indicates the need for a device capable of collecting the laser-produced debris. We newly designed and manufactured dust collector coupled with cylindrical tube to prevent scattering of laser-produced debris into ambient environment. The collecting system was evaluated by estimating the collecting efficiency for laser-produced debris while scabbling. The collecting efficiency was calculated on the basis of the information on the mass loss of concrete block after laser scabbling and the mass of collected debris in a container. The collecting efficiency was found to be over 85%.