The first commercial operation of Kori-1, which commenced in April 1978, was permanently shut down in June 2017, with plans for immediate dismantling. The decommissioning process of nuclear power plants generates a substantial amount of radioactive waste and poses significant radiation exposure risks to workers. Radioactivity is widely distributed throughout the primary coolant system of the reactor, including the reactor pressure vessel (RPV), steam generator (SG), and pressurizer. In particular, the SG has a considerable size and complex geometry, weighing approximately 326 tons and having a volume of 400 m3. The SG tubes are known to contain high levels of radioactivity, leading to significant radiation exposure to workers during the dismantling process. Therefore, this study aims to evaluate the workers’ radiation exposure during the cutting of SG tubes, which account for approximately 95% of the total radiation dose in the SG. Firstly, the CRUDTRAN code, developed to predict the behavior of soluble and particulate corrosion products in a PWR primary coolant system, is used to estimate the radioactive inventory in the SG tubes. Based on decontamination factors (DF) obtained in the SG tubes through overseas experience, the expected reduction in radioactivity during the Kori-1 reactor’s full-system decontamination (FSD) process is considered in the CRUDTRAN results. These results are then processed to estimate the radioactivity in both the straight and bent sections of the tubes. Subsequently, these radioactivity values are used as inputs for the MicroShield code to calculate the worker radiation exposure during the cutting of both straight and bent sections of the tubes. The cutting process assumes that each SG tube section is cut in a separate, shielded area, and the radiation exposure is assessed, taking into account factors such as cutting equipment, cutting length, working hours, and working distance. This study evaluates the worker radiation exposure during the cutting of SG tubes, which are expected to have a significantly high radioactivity due to chalk river unidentified deposit (CRUD). This assessment also considers the reduction in radioactivity within the steam generator tubes resulting from the FSD process. Consequently, it enables an examination of how worker radiation exposure varies based on the extent of FSD. This study may provide valuable insights for determining the scope and extent of the FSD process and the development of shielding methods during the dismantling of SG tubes in the future.

In South Korea, the replacement of steam generators began with Kori Unit 1 in 1995, and to date, 20 steam generators have been replaced and are currently stored in intermediate storage facilities. In the future, additional decommissioned steam generators may arise due to measures like the extension of the lifespan of nuclear power plants. In Korea, technological development for dismantling steam generators is underway, and there is no track record of actual dismantling. Although the replaced decommissioned steam generators are stored in intermediate facilities, for site recycling purposes, steam generators, which have relatively lower radiation doses compared to reactor heads and other primary equipment, should be prioritized for dismantling. While there are various specifications for steam generators, those dismantled and stored domestically are of the Recirculation Type. They can be classified into three types: the Westinghouse type WH-51 used in Kori Unit 1, the Fra-51B used in Han-ul Units 1 and 2, and the OPR-1000 used in Han-ul Units 3 and 4. The quantity of U-Tubes varies depending on the specification, but the radiation is concentrated in the primary side components, the U-Tube and Chamber. Since the parts related to the secondary side are not contaminated, they can be disposed of independently after classification. To dismantle a steam generator, it is of utmost importance to first create a scenario regarding where and how the dismantling will take place. Through the analysis of the advantages and disadvantages of each scenario, the optimal timing, location, and cutting method for dismantling should be researched. Furthermore, based on those findings, the best scenario should be derived through an analysis of worker radiation exposure and dismantling costs. To achieve this, a 3D simulation software developed by Cyclelife Digital Solutions under the French EDF was utilized to conduct simulations based on different dismantling schedules and methods. As a result, the optimal scenario for dismantling the steam generator was derived.

The phosphate industry is classified by IAEA as one of the Naturally Occurring Radioactive Materials (NORM) industry sectors most likely to require regulatory consideration. The production of phosphorous fertilizers constitutes the major activity in the industry, which can give rise to exposures of workers and the public through the handling and usage of phosphate rock and residues associated with processing. During the production process, when phosphate rock is digested with acid to produce phosphoric acid, some radionuclides, particularly Radium, become concentrated in residues, such as the scale that tends to form inside pipes and vessels. The registered radioactivity of phosphate rock in South Korea is less than 1.7 Bq/g for U-238, but according to the IAEA SRS No. 49, the radioactivity of phosphate scale can be up to 1,000 times higher than the raw mineral. Therefore, this study evaluated the potential for worker exposure during maintenance related to the removal of scales at a fertilizer manufacturing facility producing phosphoric acid in Korea.

Safety-related items in the decommissioning Nuclear Power Plants (NPPs) can largely consider safety for workers and residents. At this time, the effects of radioactive contamination on the Systems, Structures, and Components (SSCs) are caused by the performance of work related to Decontamination and Dismantlement (D&D) activities. Classification according to dismantling activities will be important, and the decay factor of radionuclides and the impact of contaminations due to plant characteristic (thermal and electrical capacity) in estimation of exposure dose from such activities will be considered compared to other overseas NPPs. Therefore, this study will consider some factors to consider for comparison with overseas cases in estimating worker exposure dose. To assess worker exposure doses, the classification of decommissioning activities must first be made. It should be classified including large components that can be generally considered, and the contents should be similar to compare with overseas cases. In case of decommissioned NPPs with prior experience, it is possible to predict worker’s exposure with respect to plant capacity, but this does not seem to have a specific correlation when reviewing the related data. Depending on the plant capacity, the occurrence of contamination of radioactive materials may have some correlation, but it cannot be determined that it has causality with the worker’s dose when dismantling. In addition, it is expected that the effects of workers’ exposure doses will vary depending on when the highly contaminated SSCs will be dismantled from permanent shut down. Therefore, the decay correlation coefficient for this high radiation dose works should be considered. If the high radiation dose work is performed before the base year, a correlation coefficient larger than 1 value will be applied, and in the opposite case, a value less than 1 will be applied. Whether or not to perform Full System Decontamination (FSD) is also an important consideration that affects worker dose, and correlation factors should be applied. In this study, the matters to be considered when estimating worker dose for dismantling NPPs were reviewed. This suggests factors to be reflected in the work classification and dose results for comparison with overseas NPP experiences. Therefore, when doing the workers’ dose estimation, it is necessary to derive a normalized doses considering each correlation factor when comparing with overseas cases along with dose estimation for the dismantling activities.

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.

This study is for evaluation and optimization of workers’ radiation exposure for dismantling Reactor Vessel (RV) at Kori unit 1 in connection with its decommissioning process for the purpose of establishing Radiation Safety Management Plan. This is because the safety of workers in a radiation environment is an important issue. The basis of radiological conditions of this evaluation is supposed to be those of 10 years after the permanent shutdown of Kori unit 1 when dismantling work of Reactor Vessel would suppose to be started. Dose rates of work areas were evaluated on the basis of spatial dose rate derived from activation level calculated by MCNP (Monte Carlo N-Particle Transport) and ORIGEN-S code. RV are radiated by neutrons during operation, creating an environment in which it is difficult for operators to access and work. Therefore, the RV must be dismantled remotely. However, due to work such as installing devices or dismantling surrounding structures, it is not possible to completely block the access of workers. Accordingly, the exposure of workers to the RV dismantling process should be assessed and safety management carried out. The dismantling process of Kori unit 1 RV was developed based on in-situ execution in atmospheric environment using the oxigen-propane cutting technology as the following steps of Preparation, Dismantling of Peripheral Structures, Dismantling of RV and Finishing Work. For evaluation of exposure of RV dismantling work, those processes of each steps are correlated with spatial dose rates of each work areas where the jobs being done. Results of the evaluation show that workers’ collective dose for RV dismantling work would be in the range of 536–778 man- mSv. The most critical process would be dismantling of upper connecting parts of RV with 170–256 manmSv while among the working groups, the expert group performing dismantling of ICI (In-core instrumentation) nozzles and handling & packaging of cut-off pieces is evaluated as the most significantly affected group with 37.5–39.4 man- mSv. Based on the evaluation, improvement plan for better working conditions of the most critical process and the most affected workers in terms of radiation safety were suggested.

본 연구에서는 부산지역의 컴퓨터단층촬영검사실의 근무자를 대상으로 소아 두부 CT 검사를 시행함에 있어 방 사선방어에 대한 지식정도와 행위에 대한 인식도 및 의식도를 설문조사하여 기관별로 분류하여 상급종합병원, 종 합병원, 병원간의 인식도와 의식도 점수를 비교하여 영향을 미치는 요인이 무엇인지 알아보고자 하였다. 연구결과 기관별 인식도 평균점수는 상급종합병원이 42.29, 종합병원 38.43, 병원 34.06으로 상급종합병원이 가장 높게 나타 났으며 종합병원, 병원 순으로 나타났다. 기관별 의식도 평균점수는 상급종합병원이 21.37, 종합병원 24.68, 병원 29. 19로 병원이 가장 높은 것으로 나타났으며 종합병원, 상급종합병원 순으로 나타났다. 따라서 종합병원이나 병원의 CT 근무자들의 인식도를 높이기 위해서는 보수교육 및 학회 등을 통해 방사선에 대한 인식을 함양시키려는 노력이 필요할 것으로 사료되며 또한 상급종합병원의 CT 근무자들의 의식도를 높이기 위해서 방사선 방어의 최적화를 모색하고 환자에 대한 방사선 피폭선량 감소에 대해 노력을 기울여야 할 것으로 판단된다.

Indoor and outdoor nitrogen dioxide (NO2) concentrations were measured and compared with measurements of personal exposures of 95 persons in Seoul, Korea and 57 persons in Brisbane, Australia, respectively. Time activity diary was used to determine the impact on NO2 exposure assessment and microenvironmental model to estimate the personal NO2 exposure. Most people both Seoul and Brisbane spent their times more than 90% of indoor and more than 50% in home, respectively. Personal NO2 exposures were significantly associated with indoor NO2 levels with Pearson coefficient of 0.70 (p<0.01) and outdoor NO2 levels with Pearson coefficient of 0.66 (p<0.01) in Seoul and of 0.51 (p<0.01) and of 0.33 (p<0.05) in Brisbane, respectively. Using microenvironmental model by time weighted average model, personal NO2 exposures were estimated with NO2 measurements in indoor home, indoor office and outdoor home. Estimated NO2 measurements were significantly correlated with measured personal exposures (r = 0.69, p<0.001) in Seoul and in Brisbane (r = 0.66, p<0.001), respectively. Difference between measured and estimated NO2 exposures by multiple regression analysis was explained that NO2 levels in near workplace and other outdoors in Seoul (p = 0.023), and in transportation in Brisbane (p = 0.019) affected the personal NO2 exposures.