The Derived Concentration Guideline Level (DCGL) is required to release the facility from the nuclear safety act at the stage of site restoration of the decommissioning nuclear power plant. In order to evaluate DCGL, there are various requirements, and among them, the selection of input parameters based on the application scenario is the main task. Especially, it is important to select input parameters that reflect site characteristics, and at this time, a single deterministic value or a probabilistic distribution can be applied. If it is inappropriate to apply a particular single value, it may be reasonable to apply various distributions, and the RESRAD code provides for evaluation using probabilistic methods. Therefore, this study aims to analyze the difference between the application of the deterministic method and the application of the probabilistic method to the area and thickness of the contaminated zone among the site characteristics data. This study analyzed the thickness and area of the contaminated zone, and in the case of thickness, the deterministic method was applied by changing the thickness at regular intervals from the minimum depth considered by MARSSIM to the thickness of the unsaturated zone identified in previous research data. In addition, a probabilistic analysis was performed by applying a distribution to the thickness of contaminated zone. Second, for the area of the contaminated zone, the dose was evaluated for each area in consideration of the areas to be considered when deriving Area Factor (AF), and the resulting change in DCGL was observed. As a result, the DCGL tends to decrease as the thickness increases, and it seems to be saturated when the thickness exceeds a certain thickness. Therefore, It was confirmed that the level of saturated values is similar to that of entering a probabilistic distribution, and in the case of a parameter that is reasonable to enter as a distribution rather than as a single value, it is sufficiently conservative to perform a probabilistic evaluation. In the case of area change, the DCGL evaluation result showed that the DCGL increased as the scale decreased. The magnitude of the change varies depending on the characteristics of each radionuclide, and in the case of radionuclides where external exposure gamma rays have a major exposure effect, the change is relatively small. It can be seen that the change in DCGL according to the area has the same tendency as the AF applicable to the survey unit for small survey units applied in the final status survey.

The decommissioning of domestic Nuclear Power Plants (NPPs) in Korea is expected to begin with the Kori-1, which was permanently shutdown in 2017. In addition, Wolsong-1 has been also permanently shutdown, and another type will be the decommissioning project following Kori-1. KHNP is promoting operation and decommissioning projects as the owner of NPPs, and the Central Research Institute (CRI) has been developing a Final Decommissioning Plan (FDP) for the decommissioning license document. The FDP consists of 11 major chapters in the order of overview of the project, characteristic evaluation, safety assessment, radiation protection, decontamination & dismantlement activities, waste management, etc. The contents described in each chapter are individual chapters, but there are also parts that consider the connection with other chapters. The CRI, which develops the FDP for the first decommissioning project in Korea, has spent a lot of time and effort considering this and has been proceeding through trial and error until the present stage. Therefore, this study aims to explain the current status of FDP, a license document for domestic decommissioning projects, and the link between major input data in major chapters. It can be said that System, Structure, and Components (SSCs) subject to dismantling are considered as the scope of FDP. Chapters that perform estimations on these dismantling targets may include safety assessments, exposure dose assessments for workers and residents, and waste inventory assessments. Therefore, an important part of performing the estimation works is to consider the entire scope of decommissioning activities, and as a way, it can start from data based on the inventory data. After generating the inventory data, the waste treatment classification for the inventory is designated by reflecting the results of the characterization. In addition, for cost estimation, the cost of decommissioning project is predicted by inputting some data (i.e., UCF) such as work process, number of workers, and time required for each item with data reflected in quantity and characterization. After that, based on these inventory, characterization, and UCF data, accident scenarios and industrial safety evaluation are performed for the safety assessment. The worker exposure dose is estimated by considering the dose rate of the workspace with these data. In the case of the amount of waste, the final amount of waste is estimated by considering the factors of reduction and decontamination. In summary, the main estimation contents of FDP are evaluated by adding elements required for the purpose of each chapter from data combined with inventory, characterization, and UCF, so the contents of these chapters are based on the logic of considering the entire scope of decommissioning in common.

The effects of an individual effective dose from radioactive contamination that will remain during site reuse after the decommissioning of nuclear facilities is generally assessed using the RESRAD code. The calculated results should meet the site reuse criteria presented by regulators, 0.25 mSv/yr in the United States and 0.1 mSv/yr in Korea. After completion of decommissioning, the dose is not subject to measurement, resulting in Derived Concentration Guideline Level (DCGL) remaining at the site that is practically consistent with the dose criteria. In order to assess dose using the RESRAD code, various requirements will need to be considered and determined, where the selection of input parameters is one of the important factors in the dose assessment. In addition, appropriate selection of site-specific parameters is important to reflect the site characteristics of each decommissioned Nuclear Power Plant (NPP). Therefore, this study intends to analyze the impact of site-specific parameters by referring to the cases of overseas decommissioned NPPs. In order to evaluate doses using RESRAD code, a site reuse scenario must first be selected. In general, in the case of unrestricted reuse, the resident farmer scenario can be applied, so the resident farmer scenario was also selected in this study. In addition, once a resident farmer scenario is selected, input parameters are selected according to the scenario, and the input parameter inputs a single value or distribution according to the deterministic or probabilistic evaluation method. Therefore, since this study is to evaluate the effect on site-specific parameters, a single value was applied as a deterministic evaluation method. For the 10 site-specific parameters considered in overseas cases, the difference was set twice using the F9 function key in the RESRAD code and the results were analyzed. In this study, we used prior research data targeting domestic nuclear facility for sensitivity analysis. Related parameters include the category of contamination layer, soil, water transport, ingestion, and occupancy. The parameters that appeared as the greatest influence among the 10 parameters were different in radionuclide on the contaminated zone. We showed the changes according to the difference in input parameters was presented using the graph provided by the RESRAD code. As a result, in the evaluation for Co-60 in this study, no significant change was observed. However, in case of H-3, several parameters values were changed, indicating that the effect on dose will be different depending on the site characteristics of the nuclear facilities.

The Derived Concentration Guideline Level (DCGL) using RESRAD code is generally obtained for the reuse of the site and remaining buildings of the decommissioning of nuclear facilities. At this time, the evaluation first considers wide DCGL assuming homogenous contamination for the entire target site. The DCGL derived through this will be compared with the actual contamination measured at the Final Status Survey (FSS) stage to determine whether the site is compliance with criteria. Guidelines for Survey units are presented in MARSSIM and suggested in Class 1 through 3. Therefore, DCGL for the survey unit of a certain smaller area is established by applying a correction factor from wide DCGL, which is define as an Area Factor (AF). Therefore, this study reviewed the AF applied in overseas cases, reviewed the necessary factors for derivation, and compared them by applying factors to the preliminary experimental target area for domestic nuclear installations. The AF is the ratio of the dose from the base-case contaminated area to the dose from a smaller contaminated area with the same radioactive concentration. To this end, an unrestricted resident farmer scenario was applied as the site reuse scenario, which deals with all exposure pathways considered in the RESRAD. The potential exposure pathways considered in resident farmer scenarios are largely divided into external and internal exposures, which are based on NUREG/CR-5512. In addition, in order to calculate the AF, a change in the contaminated area occurs, and accordingly, a variable that varies according to the area, i.e., length parallel to aquifer flow (LCZPAQ), the contaminated fraction of plant food ingested (FPLANT), the contaminated fraction of meat and milk (FMEAT and FMILK), is accompanied. As the contamination area decreases, these variables decrease, and the criteria for reduction were reflected through overseas cases. In this study, three nuclides (C-14, Co-60, and Cs-137) were assumed as representative nuclides, and the area of the contaminated site was selected as 50,000 m2 and reduced at a certain rate. As a result, each nuclide showed different characteristics, but in general, AF increases as the area decreases. Compared to the area of this study, AF values were calculated to be smaller than those of overseas cases, but it was confirmed that the area of the values showed similar patterns. In addition, in the case of C-14, the slope of AF increased rapidly as the area decreased, while Co-60 and Cs-137 showed similar slopes.

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.

For licensees who face the decommissioning project for the first time, even if they can utilize their experience in operation, they should be well prepared and assessed for the risks of dismantling activities reflecting the characteristics of decommissioning. This can be included in the risk management of the decommissioning project, but what we want to discuss in this study is the evaluation of the industrial risk of the actual work before the dismantling work is carried out. We would like to focus more on the review of dismantling activities subject to industrial risk assessment and a series of processes for risk assessment. The dismantling work plan will need to obtain approval from the supervisory department before work on the Systems, Structures, and Components (SSCs) can be carried out. At this time, risk assessment may be included among many safety-related required documents, which are divided into radiological and non-radiological risks. The target activities at Level 1 level can include preparation for dismantling and maintenance of facilities, dismantling big components, removing the contamination of concrete structures, managing radioactive waste, etc. In addition, it can be composed of preparation work, removal of connections, lifting/installation, cutting, radiation/radioactivity measurement, and withdrawal as detailed work stages of each item’s activities. For domestic nuclear decommissioning projects, two major performance organizations, licensees and contractors, must be considered. Regarding risk assessment, the licensee will have a supervisory department controlling decommissioning activities and an HSE department at the site, and a process will need to be established in consideration of the contractor’s work organization. Therefore, activities in the risk assessment process may be established. In this study, risk assessment was reviewed as safety-related matters to be considered when carrying out the dismantling work. Safety-related risk assessment is a necessary procedure for performing practical dismantling activities, and this should be considered well in advance. Therefore, work activities and criteria were established for risk assessment, and the performance process was assumed to apply them. In terms of the performance organization and the responsibilities and roles of the processes to be performed by each organization were constructed, and this can be referred to in the process of preparing for the decommissioning project.