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.

One aspect of securing safety from the operation of Nuclear Power Plants (NPPs) is to evaluate the impact on residents at the facility’s exclusive area boundary to confirm that the radiological risk is below the allowable level. Normally, the risks from gaseous and liquid effluents are evaluated during the operation of facilities. Meanwhile, in order to be approved for the decommissioning plan, the environmental risks caused by activities during dismantling is also evaluated. Therefore, this study aims to investigate the exposure pathways considered in evaluating the risks to nearby residents from the operation and decommissioning of nuclear facilities and to examine the differences. The emission rate by radionuclide is calculated by evaluating the amount of leak from nuclear fuel during the operation of the facility through design data of the NPP. Each of the liquid and gaseous effluents is calculated, and the exposure dose received by nearby residents is calculated by considering the exposure pathways with these emission rates. In order to initiate the decommissioning of nuclear facilities, approval of the Final Decommissioning Plan (FDP) must be obtained. The FDP chapter shall describe the results of the environmental impact assessment of the decommissioning. It will not differ significantly in the exposure pathways during operation. However, the decommissioning of nuclear facilities is ultimately to remove Systems, Structures, and Components (SSCs) and to remove the regulation of the Nuclear Safety Act by ensuring that sites and remaining buildings meet the criteria for the license termination. In terms of release and reuse of nuclear facilities, the exposure dose to be considered in evaluating the dose can be considered for two main types: the site and the remaining building. The factors affecting the exposure pathways considered in assessing the environmental impacts considered in the operation and decommissioning of nuclear facilities are due to gaseous and liquid effluents. However, the difference should reflect the impact of NPP operations and decommissioning activities when evaluating the amount of radionuclides released by these effluents. Decommissioning should consider the impact after decommissioning, which is the effect of the receptor by radionuclides remaining on the site and in the remaining buildings. At this time, the effects of the source from the soil and the source from the surface of the building should be considered for the external and internal exposure pathways.

The decommissioning of the Nuclear Power Plant (NPP) is a long-term project of more than 15 years and will be carried out as a project, which will require project management skills accordingly. The risk of decommissioning project is a combination of many factors such as the decommissioning plan, the matters licensed by the regulatory agency, the design and implementation of dismantling, the dismantling plan and organization, and stakeholders. There will be some difficulties in risk management because key assumptions about many factors and the contents of major risks should be well considered. Risk management typically performs a series of processes ranging from identification and analysis to evaluation. In order to analyze and evaluate risks here, identification of potential risks is the first step, and in order to reasonably select potential risks, various factors mentioned should be considered. Therefore, the purpose of this study is to identify possible risks that should be considered for the decommissioning project in various aspects. The risk of the decommissioning project can be defined using the hazard keyword, and the risk family presented in the IAEA safety series can also be referred. It would be better to approach the radiological or non-radiological risks that may occur in the dismantling work with the hazard keyword, and if the characteristics of the decommissioning project are reflected, it would be a good idea to approach it on a risk family basis. There are 10 top risks in the risk family, 25 risks at the level 2 and 61 risks at the level 3 are presented. It may be complex to consider these hazards and risks recommended as risk families at the same time, so using the results of safety evaluation as input data for risk identification can be a reasonable approach. Therefore, this study intended to derive the possible risks of the decommissioning project based on the risk family structure. At this point, the reflection of the safety assessment results was intended to be materialized by considering the hazards checklist. As a result, this study defined and example of 38 possible risks for the decommissioning project, considering the 10 top risk family and lower level risk categories. This result is not finalized, and it will be necessary to further strengthened through expert workshops or HAZOP in the future.

The nuclear power plant decommissioning project inevitably considers time, cost, safety, document, etc. as major management areas according to the PMBOK technique. Among them, document management, like all projects, will be an area that must be systematically managed for the purpose of information delivery and record maintenance. In Korea, where there is no experience in the decommissioning project yet, data management is systematically managed and maintained during construction and operation. However, if the decommissioning project is to be launched soon, it is necessary to prepare in consideration of the system in operation, what difference will occur from it in terms of data management, and how it should be managed. As a document that can occur in the decommissioning project, this study was considered from the perspective of the licensee. Therefore, the types of documents that can be considered at Level 1 can be divided into (1) corresponding documents, (2) project documents, (3) internal documents, and (4) reference materials. Four document types are recommended based on Level 1 for the classification of documents to be managed in the decommissioning of nuclear facilities. In this study, documents to be managed in the decommissioning project of nuclear facilities were reviewed and the type was to be derived. Although it was preliminary, it was largely classified into major categories 1, middle categories 2, and 3 levels, and documents that could occur in each field were proposed. As a result, it could be largely classified into corresponding documents, project documents, internal documents, and reference materials, and subsequent classifications could be derived. Documents that may occur in the decommissioning project must be managed by distinguishing between types to reduce the time for duplication or search, and the capacity of the storage can be efficiently managed. Therefore, it is hoped that the document types considered in this study will be used as reference materials for the decommissioning project and develop into a more systematic structure.

In the case of decommissioning nuclear facilities in Korea, the dismantling activities will be initiated after obtaining approval from the regulatory agency for the Final Decommissioning Plan (FDP). The contents to be described in the FDP are presented in the notice of the Nuclear Safety and Security Commission, and among them, it is suggested to estimate and provide the basis for the decommissioning cost. The Work Breakdown Structure (WBS) is used for schedule management for the project, and the performance activities can be used as a cost management structure as well as schedule management. In order to easily manage the process and cost, the WBS structure can be normally used, and at this time, there might be a connection with ISDC if necessary. Therefore, this study aims to examine the link between activities from the WBS structure to ISDC in the decommissioning project. In general, the activities assumed as a WBS structure in this study in carrying out the decommissioning project were derived at the Level 1 and Level 2. Activities at Level 1 can be classified into project management costs, controlled area dismantling, conventional area dismantling, site remediation, waste treatment facility, construction/service, R&D, waste treatment and disposal, and characterization. For Level 2 activities, a cost activities embodied in Level 1 was derived. ISDC was developed by the OECD Nuclear Energy Agency (NEA) decommissioning cost estimation group, which improves ambiguous cost systems and presents common cost items for direct comparison between international decommissioning projects. The ISDC consists of Level 1, Level 2, and Level 3, where Level 1 represents the principal activity, Level 2 represents the activity group, and Level 3 represents the typical activity. The cost categories for typical activity at Level 3 consist of labour, investment, expenses, and contingency. In this study, the connection between WBS and ISDC was shown, and a comparison was made at Level 2. Directly, one-on-one matches have difficulties, and as much as possible, they were organized into similar items. We arranged the Level 2 linked to ISDC based on WBS. If there is a difficulty in one-on-one matches, it seems that the accurate cost calculation of ISDC items should consider the impact of additional cost distribution. Therefore, in order to calculate ISDC costs, it seems necessary to organize cost items of WBS in consideration of the ISDC.

The domestic Nuclear Power Plant (NPP) decommissioning project is expected to be carried out sequentially, starting with Kori Unit 1. As a license holder, in order to smoothly operate a new decommissioning project, a process in terms of project management must be well established. Therefore, this study will discuss what factors should be considered in establishing the process of decommissioning NPPs. Various standards have been proposed as project management tools on how to express the business process in writing and in what aspects to describe it. Representatively, PMBOK, ISO 21500, and PRICE 2 may be considered. It will be necessary to consider IAEA safety standards in the nuclear decommissioning project. GSR part 6 and part 2 can be considered as two major requirements. GSR part 6 presents a total of 15 requirements, including decommissioning plans, general safety requirements until execution and termination. GSR part 2 presents basic principles for securing the safety of nuclear facilities, and there are a total of 14 requirements. Domestic regulatory guidelines should be considered, and there will be largely laws and regulations related to the decommissioning of nuclear facilities, guidelines for regulatory agencies, and guidelines and regulations related to HSE. The Nuclear Safety Act, Enforcement Decree, Enforcement Rules, and NSSC should be considered in the applicable law for nuclear facilities. Since the construction and operation process has been established for domestic decommissioning project, there will be parts where existing procedures must be applied in terms of life cycle management of facilities and the same performance entity. As a management areas classification in the construction and operation stage, it seems that a classification similar to Level 1 and Level 2 should be applied to the decommissioning project. This study analyzed the factors to be considered in the management system in preparing for the first decommissioning project in Korea. Since it is project management, it is necessary to establish a system by referring to international standards, and it is suggested that domestic regulatory reflection, existing business procedures, and domestic business conditions should be considered.

The decommissioning of Korea’s nuclear power facilities is expected to take place starting with the Kori Unit 1 followed by the Wolsong Unit 1. In Korea, since there is no experience of decommissioning, considerations of site selection for the waste treatment facilities and reasonable selection methods will be needed. Only when factors to be considered for construction are properly selected and their effects are properly analyzed, it will be possible to operate a treatment facility suitable for future decommissioning projects. Therefore, this study aims to derive factors to be considered for the site selection of treatment facilities and present a reasonable selection methodology through evaluation of these factors. In order to select a site for waste treatment facilities, three virtual locations were applied in this study: warehouse 1 to warehouse 3. Such a virtual warehouse could be regarded as a site for construction warehouses, material warehouses, annexed building sites, and parking lots in nuclear facilities. If the selection of preliminary sites was made in the draft, then it is necessary to select the influencing factors for these sites. The site of the treatment facility shall be suitable for the transfer of the waste from the place where the dismantling waste is generated to the treatment facility. In addition, in order for construction to take place, interference with existing facilities and safety should not be affected, and it should not be complicated or narrow during construction. Considering the foundation and accessibility, the construction of the facility should be economical, and the final dismantling of the facility should also be easy. In order to determine one final preferred plan with three hypothetical locations and five influencing factors, there will be complex aspects and it will be difficult to maintain consistency as the evaluation between each factor progresses. Therefore, we introduce the Analytic Hierarchical Process (AHP) methodology to perform pairwise comparison between factors to derive an optimal plan. One optimal plan was selected by evaluating the three virtual places and five factors of consideration presented in this study. Given the complexity and consistency of multiple influencing factors present and prioritizing them, AHP tools help users make decisions easier by providing simple and useful features. Above all, it will be most important to secure sufficient grounds for pairwise comparison between influencing factors and conduct an evaluation based on this.

In nuclear decommissioning projects, past and present projects in the world, an important area to be managed is waste management. The management of waste should be done with various aspects of consideration in mind from the moment it occurs from the cutting and dismantling of Systems, Structures, and Components (SSCs). Therefore, this study aims to discuss the disposition considerations for the efficient management of low and very low level waste that is expected to be generated in large quantities and to examine its applicability to domestic nuclear facilities. As for waste management, radioactive wastes begin to be generated when SSCs are dismantled, so waste management should be carried out as a result of dismantling activities. In addition, the waste is stored in the final disposal container and transferred to the storage or disposable facilities. In order to store in the final disposal container or transfer container, it will have to be classified by radioactive level. From the perspective of waste classification, wastes below the low level can be divided into low levels, very low levels, and clearance in Korea. Therefore, as an important point of waste management, when SSCs are dismantled, the work process must be carried out until the final disposal in accordance with the disposition strategy based on the waste classification. As a disposition strategy, the process presented by the IAEA can be referred. The materials to be dismantled for the first time are largely divided into radioactive and suspected radioactive materials. After going through the dismantling process, three criteria are considered to satisfy the disposition option: unconditional release criterion, conditional criterion, and radioactive waste. The types of waste below the final low level are classified into two types as unconditional, two types as conditional, and low and very low levels. In this study, six disposition options are reviewed, including unconditioned reuse and disposal, conditional reuse and disposal, and disposal of VLLW and VLW. Options for radioactive waste may be subject to operational criteria and may need to be supplemented in terms of the acceptance criteria in the repository. In the case of the conditional option, the clearance criterion can be applied, but considering the decommissioning characteristics, it is an option that can be used for nuclear industry, and specific reuse scenarios should be supplemented through discussions with the regulatory agency. In addition, it seems that the unconditional option needs to establish a corresponding criterion.

Korea currently has two permanent shutdown Nuclear Power Plants (NPPs), and the decommissioning project is expected to begin soon, starting with the first commercial NPP. The decommissioning project will eventually be the disposal of radioactive waste in the final stage of the work, and in that respect, proper tracking and history management should be well established in the management of waste. This is in line with the guidelines that regulatory agencies should also properly manage radioactive waste. Therefore, this study intends to examine the factors that should be considered in terms of tracking and management of radioactive waste in decommissioning nuclear facilities. The starting and final point of tracking radioactive waste generated during decommissioning is the physical inventory of the current as-is state and the final container. In this respect, the tracking of waste starts from the beginning of the dismantling operation. Thus, at the stage of approval of the decommissioning work, it may begin with an ID scheme, such as the functional location in operation for the target System, Structure, and Components (SSCs). As the dismantling work progresses, SSCs will be classified by nature and radiological level, which will be placed in containers in small packaging units. At this time, the small package should be given an ID. After that, the dismantling work leads to the treatment of waste, which involves a series of operations such as cutting, decomposition, melting, and decontamination. Each step in which these tasks are performed will be placed in a container, and ID assignment is also required. Until now, the small packaging container is for transfer after each treatment, and it is placed in the storage container in the final stage, at which time the storage container also gives a unique ID. Considerations for follow-up management were reviewed assuming solid waste, which is the majority of dismantled radioactive waste considered in this study. The ID system should be prepared from the start of the dismantling work, ID generation of the small transporting container and ID generation of the final disposal container during the intermediate waste treatment process, and each ID generation of the previous stage should be linked to each generation stage. In addition, each ID must be generated, and the definition of the grant scheme and attributes is required.

The goal of the decommissioning of nuclear facilities is to remove the regulations from the Nuclear Safety Act. The media that can be considered at the time of remediation stage may usually include soils, buildings, and underground materials. In addition, underground materials may largely be the groundwater, buried pipes, and concrete structures. In fact, it can be seen that calculations of the Derived Concentration Guideline Level (DCGL) and ALARA action levels was conducted in the case of overseas decommissioning experiences of Nuclear Power Plants (NPPs). Therefore, the aim of this study is to review the remediation activities and scenarios applied for the calculation of ALARA action level from the overseas decommissioned nuclear power plants. Media that can be considered for DCGL calculation at the time of license termination may differ from site to site. If the DCGL for the target media was derived, whether additional remediation actions are required under the DCGL value from the ALARA perspective was identified by calculating the ALARA action levels in the case of the U.S. The activities to determine whether additional clean-up is justified under the regulatory criteria are remediation actions which is dependent on the material contaminated. Therefore, the typical materials that can be subjected to remediation are soils and structure basements in the overseas cases. Remediation actions involved in the decommissioning process on the structure surfaces can be typically considered to be scabbling, shaving, needle guns, chipping, sponge and abrasive blasting, pressure washing, washing and wiping, grit blasting, and removal of contaminated concrete. For the cost-benefit analysis of the media subject to DCGL calculation, it is necessary to assume a scenario for the remediation actions of the target media. The scenarios can be largely divided into two types. Those are basement fill and building occupancy scenario. In basement fill mode, buildings and structures on the site are removed, and the effect of receptors from the contamination of the remaining structures is considered. In the building occupancy mode, it is assumed that the standing building remains on the site after the remediation stage. It is a situation to evaluate how the effect of additional remediation actions changes as the receptors occupy inside of the contaminated building. Therefore, parameters such as population density, area being evaluated, monetary discount rate, numbers of years, etc. can be set and assessed according to the scenarios.

The establishment of processes for the decommissioning a Nuclear Power Plant (NPP) is one of the objects that must be prepared in carrying out the decommissioning project. In particular, in the domestic situation, where there is no experience of decommissioning commercial NPPs, it is necessary to organize the tasks and contents well in advance for the successful initiation of the project. Therefore, this study intends to present a guide-level approach to develop management for domestic decommissioning projects. As a documented template for recognizing a process, there may be a process map and description, and information such as the work structure and the relations between the activities should be indicated. In reality, activities will be managed through a set of computer system, so it would be better if the work content, activity flow, relation, management target information, computerization contents, etc. were materialized in the process. What is important here is to define the management areas and activities and draw the activity flow. Domestically, it has rich experience in construction of NPPs and has a track record of exporting NPPs to the UAE. From these experiences, we have established a framework for standardized work in construction management and construction processes, and are performing them through a computerized system. Since the work of decommissioning has a similar nature to that of construction, we will be able to benchmark the procedure for the decommissioning from the construction management procedures. Typically, in the case of schedule management, the concept and structure of the construction process will be applicable to the decommissioning. Meanwhile, the licensee of domestic decommissioning is the same as the licensee that performs the operation, and the members who will perform the decommissioning also have experience working in the operation period. Therefore, the decommissioning works are an extension of the task during operation. Representatively, there are some processes that can be applied as it is even when decommissioning, such as dismantling work and the safety management process of the radiation zone. Therefore, in carrying out the decommissioning of NPPs in Korea, processes and activities of the management area should be established from the construction processes with abundant experience and the processes during operation. Rather than making a completely new work process, this approach that properly reflects the existing work flow is expected to be an appropriate way to avoid the repulsion of employees and maladjustment to the new environment.