In this study, four technologies were selected to treat river water, lake water, and groundwater that may be contaminated by tritium contaminated water and tritium outflow from nuclear power plants, performance evaluation was performed with a lab-scale device, and then a pilot-scale hybrid removal facility was designed. In the case of hybrid removal facilities, it consists of a pretreatment unit, a main treatment unit, and a post-treatment unit. After removing some ionic, particulate pollutants and tritium from the pretreatment unit consisting of UF, RO, EDI, and CDI, pure water (2 μS/cm) tritium contaminated water is sent to the main treatment process. In this treatment process, which is operated by combining four single process technologies using an inorganic adsorbent, a zeolite membrane, an electrochemical module and aluminumsupported ion exchange resin, the concentration of tritium can be reduced. At this time, the tritium treatment efficiency of this treatment process can be increased by improving the operation order of four single processes and the performance of inorganic adsorbents, zeolite membrane, electrochemical modules, and aluminum- supported ion exchange resins used in a single process. Therefore, in this study, as part of a study to increase the processing efficiency of the main treatment facility, the tritium removal efficiency according to the type of inorganic adsorbent was compared, and considerations were considered when operating the complex process.

Radioactive waste generated during decommissioning of nuclear power plants is classified according to the degree of radioactivity, of which concrete and soil are reclassified, some are discharged, and the rest is recycled. However, the management cost of large amounts of concrete and soil accounts for about 40% of the total waste management cost. In this study, a material that absorbs methyl iodine, a radioactive gas generated from nuclear power plants, was developed by materializing these concrete and soil, and performance evaluation was conducted. A ceramic filter was manufactured by forming and sintering mixed materials using waste concrete, waste soil, and by-products generated in steel mills, and TEDA was attached to the ceramic filter by 5wt% to 20wt% before adsorption performance test. During the deposition process, TEDA was vaporized at 95°C and attached to a ceramic filter, and the amount of TEDA deposition was analyzed using ICP-MS. The adsorption performance test device set experimental conditions based on ASTM-D3808. High purity nitrogen gas, nitrogen gas and methyl iodine mixed gas were used, the supply amount of methyl iodine was 1.75 ppm, the flow rate of gas was 12 m/min, and the supply of water was determined using the vapor pressure value of 30°C and the ideal gas equation to maintain 95%. Gas from the gas collector was sampled to analyze the removal efficiency of methyl iodine, and the amount of methyl iodine detected was measured using a methyl iodine detection tube.

In the field of 3H decontamination technology, the number of patent applications worldwide has been steadily increasing since 2012 after the Fukushima nuclear accident. In particular, Japan has a relatively large number of intellectual property rights in the field of 3H processing technology, and it seems to have entered a mature stage in which the growth rate of patent applications is slightly reduced. In Japan, tritium is being decontaminated through the Semi-Pilot-class complex process (ROSATOM, Russia) using vacuum distillation and hydrogen isotope exchange reaction, and the Combined Electrolysis Catalytic Exchange (CECE, Kurion, U.S.) process. However, it is not enough to handle the increasing number of HTOs every year, so the decision to release them to the sea has been made. Another commercial technology in foreign countries is the vapor phase catalyst exchange process (VPCE) in operation at the Darlington Nuclear Power Plant in Canada. This process is a case of applying tritium exchange technology using a catalyst in a high-temperature vapor state. The only commercially available tritium removal technology in Korea is the Wolseong Nuclear Power Plant’s Removal Facility (TRF). However, TRF is a process for removing HTO from D2O of pure water, so it is suitable only for heavy water with high tritium concentration, and is not suitable for seawater caused by Fukushima nuclear power plant’s serious accident, and surface water and groundwater contaminated by environmental outflow of tritium. Until now, such as low-temperature decompression distillation method, water-hydrogen isotope exchange method, gas hydrate method, acid and alkali treatment method, adsorption method using inorganic adsorbent (zeolite, activated carbon), separator method using electrolysis, ion exchange adsorption method using ion exchange resin, etc. have been studied as leading technologies for tritium decontamination. However, any single technology alone has problems such as energy efficiency and processing capacity in processing tritium, and needs to be supplemented. Therefore, in this study, four core technologies with potential for development were selected to select the elemental technology field of pilot facilities for treating tritium, and specialized research teams from four universities are conducting technology development. It was verified that, although each process has different operating conditions, tritium removal performance is up to 60% in the multi-stage zeolite membrane process, 30% in the metal oxide & electrochemical treatment process, 43% in the process using hydrophilic inorganic adsorbent, and 8% in the process using functional ion exchange resin. After that, in order to fuse with the pretreatment process technology for treating various water quality tritium contaminated water conducted in previous studies, the hybrid composite process was designed by reflecting the characteristics of each technology. The first goal is to create a Pilot hybrid tritium removal facility with 70% tritium removal efficiency and a flow rate of 10 L/hr, and eventually develop a 100 L/hr flow tritium removal system with 80% tritium removal efficiency through performance improvement and scale-up. It is also considering technology for the postprocessing process in the future.

In the case of decommissioning of a nuclear power plant, it is expected that a significant amount of VLLW and LLW that need to be disposed of are also expected. Conventional reduction technology is a method of extracting or removing radionuclides from waste, but this project is being carried out for the purpose of obtaining a reduction effect through the development of a material that treats another radioactive waste using radioactive waste. In this paper, the technology of impregnating LiOH capable of adsorbing radiocarbon to the gas filter material manufactured from concrete and soil waste as raw materials and the radiocarbon removal performance were reviewed. In this study, a raw material of ceramic filter was prepared by mixing concrete and soil waste with a powder of 40 m or less, and after sintering at 1,250°C, 5wt% to 40wt% of LiOH is impregnated with a filter capable of adsorbing carbon dioxide. was prepared. The prepared filter used ICP-OES and XRD to confirm the LiOH deposition result, and the concentration of carbon dioxide discharged through the carbon dioxide adsorption device was confirmed. It was possible to obtain the result that the amount of adsorption was changed depending on the flow rate of carbon dioxide supplied and the amount of material. Through this, it was possible to confirm the possibility of power generation in the adsorption performance of gas. In this study, after crushing waste concrete and waste soil, powders of 40 m or less were mixed with other additives to prepare raw materials for ceramic filters, and sintered at 1,250°C to manufacture filters. 5wt% to 40wt% of LiOH was impregnated on the prepared filter to give functionality to enable carbon dioxide adsorption. The results of LiOH deposition were confirmed using ICP-OES and XRD, and the change in the concentration of carbon dioxide emitted through a separately prepared adsorption device was confirmed. It was possible to obtain the result that the amount of adsorption was changed according to the flow rate of carbon dioxide supplied and the amount of material, and the possibility of developing a material for radioactive waste treatment using radioactive waste was confirmed when the porosity and specific surface area of the filter material were increased.

Regulations on the concentration of boron discharged from industrial facilities, including nuclear power plants, are increasingly being strengthened worldwide. Since boron exists as boric acid at pH 7 or lower, it is very difficult to remove it in the existing LRS (Liquid Radwaste System) using RO and ion exchange resin. As an alternative technology for removing boron emitted from nuclear power plants, the electrochemical boron removal technology, which has been experimentally applied at the Ringhal Power Plant in Sweden, was introduced in the last presentation. In this study, the internal structure of the electrochemical module was improved to reduce the boron concentration to 5 mg/L or less in the 50 mg/L level of boron-containing waste liquid. In addition, the applicability of the electrochemical boron removal technology was evaluated by increasing the capacity of the unit module to 1 m3/hr in consideration of the actual capacity of the monitor tank of the nuclear power plant. By applying various experimental conditions such as flow rate and pressure, the optimum boron removal conditions using electrochemical technology were confirmed, and various operating conditions necessary for actual operation were established by configuring a concentrated water recirculation system to minimize secondary waste generation. The optimal arrangement method of the 1 m3/hr unit module developed in this study was reviewed by performing mathematical modeling based on the actual capacity of monitor tank and discharge characteristics of nuclear power plant.