After the major radioactivation structures (RPV, Core, SG, etc.) due to neutron irradiation from the nuclear fuel in the reactor are permanently shut down, numerous nuclides that emit alpha-rays, beta-rays, gamma-rays, etc. exist within the radioactive structures. In this study, nuclides were selected to evaluate the source term for worker exposure management (external exposure) at the time of decommissioning. The selection of nuclides was derived by sequentially considering the four steps. In the first stage, the classification of isotopes of major nuclides generated from the radiation of fission products, neutron-radiated products, coolant-induced corrosion products, and other impurities was considered as a step to select evaluation nuclides in major primary system structures. As a second step, in order to select the major radionuclides to be considered at the time of decommissioning, it is necessary to select the nuclides considering their half-life. Considering this, nuclides that were less than 5 years after permanent suspension were excluded. As a third step, since the purpose of reducing worker exposure during decommissioning is significant, nuclides that emit gamma rays when decaying were selected. As a final step, it is a material made by radiation from the fuel rod of the reactor and is often a fission product found in the event of a Severe accident at a nuclear power plant, and is excluded from the nuclide for evaluation at the time of decommissioning is excluded. The final selected Co-60 is a nuclide that emits high-energy gamma rays and was classified as a major nuclide that affects the reduction of radiation exposure to decommissioning workers. In the future, based on the nuclide selection results derived from this study, we plan to study the evaluation of worker radiation exposure from crud to decommissioning workers by deriving evaluation results of crud and radioactive source terms within the reactor core.

In Natural Analogue Study, Concrete is one of the important engineering barrier components in the Multi-thin wall concept of radioactive waste disposal and plays the most important role in disposal sites. The concrete barrier at the disposal site loses its role as a barrier due to various deterioration phenomena such as settlement, earthquake, and ground movement, causing the disposed waste to leak into the natural ecosystem. Some of the key factor is deterioration triggered by sulfate attack. Concrete deterioration induced by sulfate is commonly manifested in an extensive scale when a concrete structure makes contact with soil or water, aggravating its performance. In this study, an accelerated concrete deterioration evaluation experiment was performed using a total of three experimental methods to evaluate the reaction between concrete and water. The first experiment was a deterioration evaluation using Demi. Water, the second was a deterioration evaluation using KURT groundwater after extraction, and the last experiment was a concrete deterioration evaluation using KURT groundwater and sodium sulfate. For all of these experiments, accelerated concrete deterioration experiments were conducted after immersion for a total of 365 days, and specimens were taken out at 30-day intervals and characterization analysis such as SEM and EDS was performed. Experimental analyzes have shown that various chemical species are generated or destroyed over time. In the future, we plan to continue to conduct a total of three concrete deterioration evaluation experiments above, and additionally evaluate the chemical reaction between bentonite and concrete.

Radioactive waste generated during nuclear power plant decommissioning is classified as radioactive waste before the concentration is identified, but more than 90% of the amount generated is at a level that can be by clearance. However, due to a problem in the analysis procedure, the analysis is not carried out at the place of on-site but is transported to an external institution to identify concentration, which implies a problem of human error because 100% manual. As a way to solve this problem, research is underway to develop a mobile radioactive waste nuclide analysis facility. The mobile radionuclide analysis facility consists of a preparation room, a sample storage room, a measurement room, a pretreatment room, and a waste storage room, and is connected to an external ventilation facility. In addition, since the automation module is built-in from the sample pre-threatening step to the separation step, safety can be improved and rapid analysis can be performed by being located in the decommissioning site. As an initial study for the introduction of a mobile nuclide analysis facility, Visiplan, a preliminary external exposure evaluation code, was used to derive the analysis workload by a single process and evaluate the exposure dose of workers. Based on this, as a follow-up study, the amount of analysis work according to the continuous process and the exposure dose of workers were evaluated. As a result of the evaluation, the Regulatory dose limit was satisfied, and in future studies, internal and external exposure doses were evaluated in consideration of the route of movement, and it is intended to be used as basic data in the field introduction process.

Radioactive waste must be stored for at least 300 years and must bear astronomical costs. In addition, unexpected potential risk factors are also a considerable burden. In the case of low-level radioactive waste, combustible and liquid low-contamination radioactive waste can be treated relatively easily through high-temperature plasma which the volume can be reduced by 1/250 and the weight by 1/30. It is possible to permanently dispose of the ash leached after plasma treatment in a more stable manner compared to the conventional methods. Types of low-level combustible radioactive waste, including paper, vinyl, clothing, filters, and resins, account for more than 30% of the total waste volume. Furthermore, high-temperature plasma treatment of low-level radioactive waste from petrochemical plants and medical institutions have many advantages, namely astronomical cost savings, securing free space in existing storage facilities, and improving the image of nuclear energy. Korea is preparing to decommission the Kori No. 1 nuclear power plant, and small and mediumsized enterprises and related organizations are conducting various studies to incinerate radioactive waste. In foreign countries, Britain began incineration technology in the 1970s, and Plasma Energy Group, LLC, headquartered in Florida, USA, physically changed the molecular structure of the material by combining plasma chambers and plasma arcs and obtained a patent application in 1992. Germany was approved for operation in 2002, and Switzerland completed a trial run of a plasma technologybased facility in 2004. Important radionuclides in terms of radioactive gas waste treatment include inert gases, radioiodine, and radioactive suspended particles. Gas waste is compressed in a compressor through a surge tank in the gas waste treatment system and filters at each stage. after that, the shortlife nuclide is naturally collapsed for 30 to 60 days in the storage or activated carbon adsorbent in the attenuation tank and released through HEPA filters. The radioactive concentration at discharge is monitored and managed using continuous monitoring equipment, and the oxygen concentration is managed in the gas waste treatment system to prevent explosion risk. The problem of radioactive waste disposal is not only a problem for people living in the present era, but also a big social issue that brings a burden to future generations While interest in plasma treatment is increasing from the decommissioning of the Kori Unit 1. in Korea, it is showed that there is a lack of systematic management and research especially in the radioactive volatile gases fields, that’s why I propose some ideas as follows. First, the government and related institutions should invest to the continuous radioactive monitoring system to produce and distribute continuous radioactive monitoring facilities with an affordable price. Second, it is recommended that radioactive waste incineration would be connected to the GRS system of the plant’s gas radwaste treatment system, and radioactive volatile materials should be monitored through continuous monitoring system. Third, radioactive volatile materials generated according to the temperatures and times during plasma incineration treatment are different. Therefore, prior classification of each expected radioactive volatile substance must be performed before incineration.

n Korea, the decommissioning of nuclear power plants is being prepared, and a large amount of radioactive waste is expected to be generated. In particular, clearance level waste, which accounts for more than 90%, requires prevention of cross-contamination and prompt classification. In this study, the possible exposure route and the derivation of exposure dose for worker exposure management in a movable analysis system that can be analyzed onsite were studied. The movable radionuclide analysis system is divided into a preparatory room, a sample storage room, a radioanalysis room, a laboratory, and a waste storage room. It consists of one radioanalysis worker and one pre-treatment worker, and the main radiation exposure is expected to occur in the movement path in the sample storage room, radioanalysis room, and laboratory. The source term for the exposure evaluation, the annual usage dose presented in the radiation safety report in the movable radionuclide analysis system was used. The input data for the evaluation of the external exposure dose under normal circumstances (exposure situation, working hours, distance, etc.) is referenced at facility specifications. The internal exposure dose evaluation was assumed to be acute exposure (1 hour) assumed as internal pollution due to the drop in liquid sample during the pretreatment work. As an evaluation method, a method using a calculation formula and a method using an evaluation code was performed. For the evaluation of exposure dose using the calculation formula, a preliminary evaluation was performed using the point source method, the point kernel method, and intake and dose conversion factors. In addition, VISIPLAN and IMBA codes were used to evaluate exposure dose using the evaluation code, and the input data were supplemented for evaluation. As a result of the evaluation, the annual exposure dose limit of 20 mSv was satisfied for both normal and non-normal situations. In future research, it is planned to derive the evaluation results by particular scenarios for the detailed movement route and evaluation time according to the work process in the mobile radionuclide analysis.

Radiological characterization, one of the key factors for any successful decommissioning project for a nuclear facility, is defined as a systematic identification of the types, quantities, forms, and locations of radioactive contamination within a facility. This characterization is an essential early step in the development of a decommissioning plan, in particular during transition period after permanent shutdown of the facility, and also to be used for classification of decommissioned radioactive wastes so that their disposal criteria can be met. Therefore, the characterization should be well planned and performed. In the transition period, the characterization information developed during the operational phase is usually reexamined with respect to the applied assumptions, the actual status of the facility after shutdown, the accuracy of the required measurements and changes in its radiological properties to support the development of the final decommissioning plan. Based on some national (Korean, USA’s and Japanese) laws including the related regulations, and some related documents published by OECD/NEA, IAEA, and ASTM, key elements of radiological characterization, which should be developed in the transition period, could be proposed as the followings. The key elements might be an operational history including facility operation history and contamination by events and/or accidents, radiological inventory of the facility and site area, characterization survey including in-situ survey and/or sampling and analyses, radiological mapping (which is able to identify radiological contamination levels of SSCs, and the facility area and, if contamination may be suspected, the surroundings) with tabulating, residual radioactivity (or derived concentration guideline levels) of selected major radionuclides for remediation of the site, (retainable and retrievable) recording, and quality control and quality assurance. In review process of the operational history, interviews of current or former long-tenured knowledgeable employees of the facility should be conducted to identify conditions that may have been missing from the records.