Recently, Japan’s government has announced Tokyo Electric Power Company’s plan to discharge contaminated water stored from the tanks of the Fukushima Daiichi nuclear power plant site into the sea. The contaminated water is treated by advanced liquid processing system (ALPS) to remove 62 radionuclide containing cesium, strontium, iodine and etc. using co-precipitation (or precipitation) and adsorption for other nuclides (except for tritium and carbon-14). The total amount of the contaminated water generated by ALPS facility is 1,311,736 m3 (as of August 18, 2022). The amount of contaminated water is estimated same as Tokyo dome volume. Under the sea discharge plan, the contaminated water will be diluted in seawater more than 100 times, and tritium concentration lowered 1/7 of the drinking water standard set by the World Health Organization (10,000 Bq/liters). The diluted water will then move through an undersea tunnel and be discharged about 1 kilometer off the coast.

Lubricant oil waste contaminated with radioactive materials generated at nuclear facilities can be disposed of as industrial waste in accordance with self-disposal standards if only radioactive materials are removed. Lubricant oil used in nuclear facilities consists of oil of 75-85% and additives of 15-25%, and lubricant oil waste contains heavy metals, carbon, glycol, etc. In addition, lubricant oil waste from nuclear facilities contains metallic gamma-ray emission radionuclides including Co-60, Cs-137 and volatile beta-ray emission radionuclides such as C-14 and H-3, which are not present in lubricant oil waste from general industries and these radionuclides must be eliminated according to the Atomic Energy Act. In general industries, the wet treatment technologies such as acid-white soil treatment, ion purification, thin film distillation, high temperature pyrolysis, etc. are used as the refining technology of lubricant oil waste, but it is difficult to apply these technologies to nuclear industrial sites due to restrictions related with controlling the generation of secondary radioactive waste in sludge condition containing radionuclides of metal components, and limiting the concentration of volatile radioactive elements contained in refined oil to be below the legal threshold. In view of these characteristics, the refinement system capable of efficiently refining and treating lubricant oil waste contaminated with radioactive materials generated in nuclear facilities has been developed. The treatment process of this R&D system is as follows. First, the moisture in the radioactive lubricant oil waste pretreated through the preprocessing system is removed by the heated evaporating system, and the beta-emission radionuclides of H-3 and C-14 can be easily removed in this process. Second, the heated lubricant oil waste by the heated evaporating system is cooled through the heat exchanging system. Third, the particulate matters with gamma-ray emission radionuclides are removed through the electrostatic ionizing system. Forth, the lubricant oil waste is stored in the storage tank and the purified lubricant oil waste is discharged to the outside after sampling and checking from the upper, middle and lower positions of the lubricant oil waste stored in the storage tank. Using this R&D system, it is expected that the amount of radioactive waste can be reduced by efficiently refining and treating lubricant oil waste in the form of organic compounds contaminated with radioactive materials generated in nuclear facilities.

In general, dose assessment must be performed to obtain approval for clearance of radioactive waste. If the annual dose criteria through dose evaluation satisfies the clearance condition, radioactive waste can be disposed of. Various programs are used to perform dose assessment. NRCDOSE GASPAR is used as a program to assess the amount of radiation exposed to atmospheric emissions. Program is easy to use and results can be checked immediately after execution. GASPAR requires main input factors by exposure route such as site specifics, source term, special location, block data. Basically, program has default input values but user can easily modify it. The most important factor is that when entering a nuclide, the effect on progeny radionuclides is not automatically calculated. User should consider the dose contribution from progeny radionuclides. In this study, dose assessment was performed for combustible waste incineration using NRCDOSE GASPAR. And it was confirmed that exposure dose of individuals and groups criteria for clearance regulation.

There are various types of level gauging method such as using float, differential pressure, hypersonic, displacement and so on. In this study, among them, the method utilizing the differential pressure was reviewed. The strengths include: the differential pressure type level gauge can measure the level without direct contact of the sensor with media. That is to say, the level can be measured even if the sensor is far away from the tank. And regardless of the size of the tank, the level can be measured if the pneumatic pipes are installed. The weaknesses include: the sensor needs intermedium to recognize the level. The intermedium utilizes a fluid, which is compressed air. It is difficult to handle that compressed air has the properties of a gas. And to make compressed air needs compressor, tank and pneumatic pipes. So if you have many tanks, you need to install exponentially the pneumatic pipes. As well, level measurement range is limited to the points where the pneumatic pipes of the tank is installed. And if a compressed air that supplies to the sensor leaks, uncertainty will increase. A compressed air is colorless and odorless, so it’s difficult to pinpoint the leak. Finally, events like cracks and clogging can cause inaccurate measurement. Rather than using only differential pressure, it is better to use another measurement method according to the situation of the facility.

The number of dismantled nuclear facilities is increasing globally. Dismantling of nuclear facilities generates large amount of waste such as concrete, soil, and metal. Concrete waste accounts for 70% of the total amount of waste. Since hundreds of thousansds of tons of concrete waste generated, securing technology of reduction and recycling of waste is emerging as a very important issue. The objective of this study is to synthesize geopolymer using inorganic materials from cement fine powder in concrete waste. Dismantled concrete waste contains a large amount of calcium silicate hydrate(C-S-H), Ca(OH)2, SiO2, etc., which is an inorganic material required for the synthesis of geopolymer. SiO2 affects the compressive strength of the geopolymer and Ca(OH)2 affects the curing rate. A high concentration of alkali solution is used as an alkali activator, and alkali activator is necessary for the polymerzation reaction of metakaolinite. The experiment consists of three steps. The first step is to react with concrete waste and hydrochloric acid to extract ions. In the solid after filtration, SiO2 and Al2O3 are composed of 84.10%. It can be used instead of commercial SiO2 required for the synthesis of geopolymer. The second step is to add NaOH up to pH 10, impurities can be removed to extract Ca(OH)2 with high purity. The final step is to add NaOH up to pH 13, and Ca(OH)2 extraction. The alkali solution generated after the last reaction can be recycled into an alkali activator during the synthesis of the geopolymer. If dismantled concrete waste is recycled during geopolymer synthesized, the volume reduction rate of dismantled concrete waste is more than 50%. If you put the radioactive waste in the recycled solidification materials synthesis from concrete waste by dismantling of nuclear facilities, it is possible to reduce the amount of waste generated and disposal costs.

It is necessary to prepare for cutting and storing waste materials in the reactor vessel internals (RVI) for successful decommissioning of the nuclear power plant (NPP). Since RVI contain massive components and relatively highly activated, their decommissioning process should be conducted carefully in terms of radiological and industrial safety. To achieve efficient decommissioning waste management, this study presents radiation level of RVI and cutting optimization was performed for intermediate level waste. As a result of the radiation evaluation, a part of the core side and the upper part of RVI were evaluated as intermediate-level waste, and other components were evaluated as very low-level or lowlevel waste. For intermediate-level waste cutting, the minimum cutting method that can be put into a container was reviewed in consideration of the size, thickness, and cutting method of the interior product. The final segmentation parts are expected to be loaded into two storage containers.

For the disposal of radioactive waste generated from nuclear power plants, characterization of radioactive waste is essential. For characterization, samples of radioactive waste are directly collected or an indirect method is used through X-ray, etc. Through indirect analysis, which is a non-destructive method, the density, filling height, homogeneity and inter structure of the waste container can be analyzed. Currently, foreign institutions are in the process of developing a technology to perform characterization of radioactive waste through indirect analysis. In particular, research on improving internal image accuracy through image analysis techniques, improving measurement methods and enhancing portability for field application is ongoing. Through the review of such technology development trends, it will be utilized in the development of domestic radioactive waste disposal technolgy.

Glass fiber (GF) insulation is a non-combustible material, light, easy to transport/store, and has excellent thermal insulation performance, so it has been widely used in the piping of nuclear power plants. However, if the GF insulation is exposed to a high-temperature environment for a long period of time, there is a possibility that it may be crushed even with a small impact due to deterioration phenomenon and take the form of small particles. In fact, GF dust was generated in some of the insulation waste generated during the maintenance process. In the previous study, the disposal safety assessment of GF waste was performed under the abnormal condition of the disposal facility to calculate the radiation exposure dose of the public residing/ residents nearby facilities, and then the disposal safety of GF waste was verified by confirming that the exposure dose was less than the limit. However, the revised guidelines for safety assessment require the addition of exposure dose assessment of workers. Therefore, in this study, accident scenarios at disposal facilities were derived and the exposure dose to the workers during the accident was evaluated. The evaluation was carried out in the following order: (1) selection of accident scenario, (2) calculation of exposure dose, (3) comparison of evaluation results with dose limits, and confirmation of satisfaction. The representative accident scenarios with the highest risk among the facility accident were selected as; (a) the fire in the treatment facility, (b) the fire in the storage facility, and (c) fire after a collision of transport vehicles. The internal and external exposure doses of the worker by radioactive plume were calculated at 10m away from the accident point. In evaluation, the dose conversion factors ICRP-72 and FGR12 were used. As a result of the calculation, the exposure dose to workers was derived as about 0.08 mSv, 0.20 mSv, and 0.10 mSv, due to fire accidents (vehicle collision, storage facilities, treatment facilities). These were 0.2%, 0.4%, and 0.2% of the limit, and the radiation risk to workers was evaluated to be very low. The results of this study will be used as basic data to prove the safety of the disposal of GF waste. The sensitivity analysis will be performed by changing the radiation source and emission rate 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.

The integrity of the disposal repository structure must be guaranteed for few hundreds to few hundred thousand years until toxicity of radioactive waste is surely degraded. Acoustic emission (AE) method is widely utilized to evaluate the integrity of the structure because it can detect crack wave signals of the structures. It is well known that the cracking AE energy is proportional to the volume of the structure (Fractal theory). However, it is hard to destroy whole structures for obtaining AE energy. Therefore, the scaled specimens are prepared to obtain the relationship between volume of the structure and AE energy. The specimens are prepared with same of Wolsong Low and Intermediate Level Radioactive Waste Disposal Center (WLDC) silo concrete recipe. Their diameters are from 50 mm to 150 mm in each 10 mm and their heights are twice of the diameter. One set of 50 mm to 150 mm specimens (11 specimens in one set) are made in single mixers to maintain uniformity. Surface of the specimens are flatten with cement milk to prevent from applying load with eccentricity. The uniaxial compression test is performed by controlling displacement as 0.1 mm/min. The fractal constant is obtained using least square function from volume-cumulative AE energy relationship.