In 2022, new regulatory guidelines were announced in relation to the off-site dose calculation (ODC), and accordingly, measures to improve the off-site does calculation program (ODCP), kdose60, were reviewed. The main consideration is, first, that if multiple nuclear facilities are operated on the same site, the boundaries of the restricted areas shall be set as the overlapping outer boundaries of the restricted areas determined by calculation for each nuclear facility. Second, the external exposure caused by direct radiation from a number of nuclear facilities in the same site must be partially or fully applied depending on the facility and site characteristics. Third, the dose conversion coefficient should be evaluated by checking whether the effect of the daughter nuclides is properly reflected. Fourth, the soil contamination period is a factor to consider that radioactive substances deposited on the surface, such as particulate nuclides, affect residents over a long period of time. Fifth, due to the recent construction of Shin-Kori Units 5 and 6, there is a change in the site boundary of the Kori/Saeul site, so as the site boundary is expanded, it is required to add an exposure dose assessment point due to gas effluents and change the exposure dose assessment point according to crop intake. Therefore, through this study, the direction for improving the ODCP will be prepared by reviewing the recent revision of the regulatory guidelines.

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.

In KAERI, Waste storage facility in the radiation management area has stored a large amount of wood waste. The amount of waste is approximately 27,000 kg, it accounts for 17% of the total waste in waste storage facility. Proper disposal of wood waste improves the fire resistance performance, secure storage space and reduce disposal costs. In order to self-disposal of wood waste, it is necessary to satisfy the self-disposal standards stipulated by the domestic Atomic Energy Act and the treatment standards of the Waste Management Act. The main factors of standards are surface contaminant, radionuclide activity and radiation dose effects. To confirm the contamination of wood waste, direct indirect measurement methods and gamma nuclide analysis were performed. To evaluate radiation dose, various computational programs were used. The results of the analysis were satisfied with domestic regulations on the classification and self-disposal of radioactive wastes. Based on this results, KAERI submitted the report on wood waste self-disposal plan to obtain approval. After final approval, wood waste is to be incinerated and incineration ash is to be buried in the designated place. The objective of this study is to provide total procedure of wood waste self-disposal and effective representative sampling method.

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.

Radioactive waste is classified into Intermediate level, low level, and very low potential based on the amount of radioactivity per unit gram, that is, the concentration limit. This method of classifying radioactivity per unit weight is not a problem if all packaged wastes are homogeneous. However, the reality is that not all waste is homogeneous. Relative hotspots may exist. Also, when several items are mixed, if one item has a relatively higher concentration than other items, it can become a relative hotspot. In Korea, even if all nuclides in a single radioactive waste package satisfy the low level concentration limit, if even one nuclide exceeds the concentration limit, the radioactive waste package becomes the intermediate level. In case of the United States, the US NRC provides regulations related to obtaining license as well as presents the technical position on the average waste concentration called Concentration Averaging and Encapsulation Branch Technical Position (CA BTP). CA BTP classifies waste into four types : Blendable Waste, Encapsulated items, Single Discrete Items, and Mixture of Discrete Items, and presents each approach to concentration averaging. In general, this is a method that suggests an acceptable ratio in case of the waste, which relatively high concentration waste is mixed. In order to apply this in Korea, we compare the classification standards for low and Intermediatelevel waste in Korea and the United States, related laws and backgrounds, and the application methods of CA BTP.

Following a radioactive waste criterion and clearance level radioactive waste Act Article 2. “The radioactive wastes confirmed by the Commission as having concentration by nuclide not exceeding the value determined by the Commission through incineration, reclamation, recycling, etc”. The combustible clearance level radioactive wastes like lumbers are incinerated and non-combustible wastes like concreted are buried. The metals clearance level radioactive wastes are recycled after being re-molded. However, the clearance level radioactive waste with keeping its original forms is not common. Due to the nature of KAERI, the equipment are brought into the radiation-controlled zone for experiments. Those equipment are conservatively considered contaminated and categorized with radioactive waste following nuclear safety acts. In this case, the spectroscopy device which is clearance level radioactive waste is self-disposed for use in non-controlled areas. The 4 devices are composed of 3 gamma-ray spectroscopy and 1 alpha, beta counting system. Those devices were used for clearance level radioactive waste’s radioisotope analysis in Radioactive Waste Form Test Facility which is used in a separated room for analysis. This room will be released in nonradiation controlled area, therefore those devices will be moved to non-controlled area and keep using. Last April self-disposal was reported to the regulatory body and got acceptance last May. Those devices were moved to non-controlled area last July. This case will be good example for reuse equipment which stop using in radiation controlled area but can keep used.

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.