Zinc injection into the coolant system of nuclear power plants is an effective method for reducing corrosion and improving performance. The effectiveness of this method is influenced by various factors such as zinc concentration and injection rate. This paper provides an overview of the factors affecting the effectiveness of zinc injection in nuclear power plants, with a focus on zinc concentration and injection rate, and discusses various research results on the effects of these factors on corrosion reduction and coolant system performance. Zinc concentration is an important factor affecting the effectiveness of zinc injection. The research results show that gradual increases in zinc concentration are more effective for coolant system stability. However, the concentration should not exceed the recommended levels as high zinc concentrations can have negative effects on the system. Injection rate is also an important factor affecting the effectiveness of this method. The research results show that gradual increases in injection rate are more effective for coolant system stability. However, excessive injection rates can have negative effects on the system such as overload of the zinc injection facility and chemical shocks within the coolant system, and therefore, should be optimized. In conclusion, zinc concentration and injection rate are important factors affecting the effectiveness of zinc injection in nuclear power plants. The optimal concentration and injection rate should be determined based on specific reactor conditions and system requirements, and efforts should be made to maximize corrosion reduction and performance improvement.

The 2007 Recommendation of the International Commission on Radiological Protection recommended the application of dose constraints to optimize radiation protection to resolve the inequity of exposure among radiation workers. The average annual occupational doses in Korean nuclear power plants (NPPs) are 0.3-0.8 mSv. These doses are much lower than the annual effective dose limit of 50 mSv for radiation workers stipulated by the Nuclear Safety Act. In addition, most NPP workers received less than 0.1 mSv per year. These doses are lower than the average annual occupational doses of 0.3- 0.8 mSv. Korean regulatory body conducted the study to legislate the dose constraints in the Korean regulatory system and determine dose constraints (draft) for radiation workers. The legislation of dose constraints would not greatly affect the radiation protection programs in Korean NPPs because most workers received very low doses. However, some workers received relatively higher doses than others. This study analyzed the occupational exposure conditions, such as exposure type and situation, in Korean NPPs. This study investigated the internal and external radiation doses and the radiation doses depending on the NPP operating conditions, including normal operation, planned maintenance, and intermediate maintenance, for the last ten years (2012-2021). As a result, most NPP workers received external exposure rather than internal exposure. Furthermore, most radiation exposures occurred during the planned maintenance period. The results of this study can be used for optimizing occupational doses in Korean NPPs.

In this research, the dose rate was measured using a backpack-type scan survey device at 4 sites in sites around Nuclear Power Plants (Kori, Wolsong, Hanbit, Hanul), and the radioactivity ratio for each nuclide was evaluated using an high-purity germanium (HPGe) detector. Kori, Wolsong and Hanul power plants were measured within 2 km of the power plant, and Hanbit power plants were measured about 6.7 km from the power plant. As a result of measuring the dose rate with a backpacktype scan survey device, the average dose rate was the lowest in the measurement site 1 at 0.090 μSv/h, and the highest in the measurement site 4 at 0.145 μSv/h. All measurement points showed the domestic environmental dose rate level. The data obtained by the scan survey was visualized using the classed post and gridding functions of the surfer program. As a result of measurement with the HPGe detector, 137Cs was not detected, and only natural nuclides were detected. Among the detected natural nuclides, the radioactivity ratio was the highest for 40K with an average of 94.56%, and the lowest for 214Pb with an average of 0.26%. The results of this research can be used as basic data for radiation environment surveys around nuclear power plants. Further studies are needed to evaluate the radiation impacts by region and environment through periodic measurements.

According to attached Table 1 of the Enforcement Ordinance of the Nuclear Safety Act, the effective dose limit of transport workers shall not exceed 6 mSv per year. In addition, the enforcement ordinance defines a transport worker as a person who transports radioactive substances outside the radiation management area and does not correspond to a radiation worker. In the nuclear power plants (NPPs), substances in radiation management areas are frequently transported inside or outside the plant. During loading of substances in the radiation management area onto the vehicle, the transport workers (including driver) are located outside the radiation management area. And also the exposure dose of transport workers is managed by using Automatic Dose Reader (ADR). However, the exposure dose of transport workers managed by NPP licensee is limited to the exposure caused by the transport actions required by the plant. This means that radiation exposure caused by the transport of radioactive materials carried out separately by individual transport workers other than the plant requirements cannot be managed. Therefore, even if the NPP licensee manages the transport worker’s dose below 6 mSv, it is difficult to guarantee that the total annual exposure dose, including the transport worker’s individual transport behavior, is less than 6 mSv. Therefore, it would be appropriate to manage the dose of the transport worker by the transport worker’s agency rather than by the NPP licensee.

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.

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.

As a result of various generation, transmutation, and decay schemes, a wide variety of radionuclides exist in the reactor prior to accident occurrence. Considering all of the radionuclides as the accident source term in an offsite consequence analysis will inevitably take up excessive computer resources and time. Calculation time can be reduced with minimal impact on the accuracy of the results by considering only the nuclides that have a significant effect on the calculation among the potential radioactive sources that may be released into the environment. In earlier studies related to offsite consequence analysis, it is shown that the principal criteria for the radionuclide screening applied are as follows; radionuclide inventory in the reactor, radioactive half-life, radionuclide release fraction to the environment, relative dose contribution of nuclides within a specific group, and radiobiological importance. As a result, it is confirmed that 54, 60, and 69 nuclides are applied to the risk assessment performed in WASH-1400, NUREG-1150, and SOARCA (State-of-the-Art Reactor Consequence Analyses) project in the United States, respectively. In addition, in this study, the technical consultations with domestic and foreign experts were carried out to confirm details on criteria and process for screening out radionuclides in offsite consequence analysis. In this paper, based on the literature survey and technical consulting, we derived the screening process of selecting a list of radionuclides to be considered in the offsite consequence analysis. The first step is to eliminate radionuclides with little core inventory (less than specific threshold) or very short half-lives. However, important decay products of radionuclides that have short half-lives should not be excluded by this process. The next step is to further eliminate radionuclides by considering contribution to offsite impact, which is defined as a product of radioactivity released to the environment (i.e. ‘inventory in the reactor’ times ‘release fraction to offsite’) and comprehensive dose (or risk) coefficient taking into account all exposure pathways to be included. The final step is to delete isotopes that contribute less than certain threshold to any important dose metric through additional computer runs for each important source term. Even though it is presumed that this process is applicable to existing light water reactors and the set of accidents that would be considered in PSA, some of the assumptions or specific recommendations may need to be reconsidered for other reactor types or set of accident categories.

Gamma imaging devices that can accurately localize the radioactive contamination could be effectively used during nuclear decommissioning or radioactive waste management. While several hand-held devices have been proposed, their low efficiency due to small sensors have severely limited their application. To overcome this limitation, a high-speed gamma imaging system is under development which comprises two quad-type detectors and a tungsten coded aperture mask. Each quad-type detector consists of four rectangular NaI(Tl) crystals with dimensions of 146×146 mm2 and 72 square-type photomultiplier tubes (PMTs). The detectors are placed in front and back to serve as scatter and absorber, respectively, for Compton imaging. In addition, a coded aperture mask was fabricated in rank 19 modified uniformly redundant array pattern and placed in front of the scatter for coded aperture imaging. The system offers several advanced features including 1) high efficiency achieved by employing large-area NaI(Tl) crystals and 2) broad energy range of imaging by employing a hybrid imaging combining Compton and coded aperture imaging. The imaging performance of the system was evaluated through experiments in various conditions with different gamma energies and source positions. The imaging system provides clear images of the source locations for gamma energies ranging from as low as 59.5 keV (241Am) to as high as 1,330 keV (60Co). The imaging resolution was within the range of 7.5–9.4°, depending on gamma energies, when a hybrid maximum likelihood estimation maximization (MLEM) algorithm was used. The developed system showed high sensitivity, as the 137Cs source at distance, incurring dose rate lower than background level (0.03 μSv/h above background dose rate), could be imaged in approximately 2 seconds. Even under lower dose rate condition (i.e., 0.003 μSv/h above background dose rate), the system was able to image the source within 30 seconds. The system developed in the present study broadens the applicable conditions of the gamma ray imaging in terms of gamma ray energy, dose rate, and imaging speed. The performance demonstrated here suggests a new perspective on radiation imaging in the nuclear decontamination and radioactive waste management field.

After permanent shutdown, contamination existing in nuclear facilities must be removed according to decontamination and dismantling procedures to achieve the target end state. In Korea, Korea Research Reactor (KRR) Units 1, 2 are being decommissioned, and Kori Unit 1 is in the process of reviewing the final decommissioning plan for the start of decommissioning. In order to complete decommissioning of nuclear facilities, it is necessary to satisfy the dose criteria according to the residual radioactivity remaining in the site and buildings. In the United States, which has a lot of experience in decommissioning, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) is used as a procedure for measuring and analyzing residual radioactivity. In MARSSIM, survey units are classified according to the level of contamination, and the radiation survey procedure and effort can be determined according to the survey unit level. After the radiological analysis and statistical verification of the survey unit, it is decided whether to release the site. At this time, the geographical area to be used as the background level is called the reference area. Therefore, selection of an appropriate reference area is important for accurate residual radioactivity analysis and for the release of the site. In this study, reference area evaluation cases and domestic decommissioning procedures were analyzed to derive considerations for selecting an appropriate reference area. For example, Zion NPP in the US selected a place outside the boundary of the restricted area unaffected by nuclear power plant operation as a reference area by referring to the meteorological monitoring report. Among Korea’s decommissioning procedures, the appropriateness of the reference area can be confirmed through the final status report submitted upon completion of decommissioning. However, since the selection and application of the reference area needs to be reflected during decommissioning, relevant information must be updated through periodic communication between operator and regulatory agency. The results of this study will be used as considerations for selecting a reference area.

To evaluate the characteristics of radioactive waste from permanently shut down nuclear power plants for decommissioning, there is a method of directly analyzing samples and, on the other hand, a computerized evaluation method based on operation history. Even if the radioactivity of the structures or radioactive wastes in the nuclear power plant is analyzed by the computerized evaluation method, a method of directly analyzing the sample must be accompanied in order to more accurately know the characteristics of the nuclear power plant’s radioactive waste material. In order to obtain such samples, we need a way to collect materials from radioactive waste. However, in the case of a permanently shut down nuclear power plant with a long operating history, human access is limited due to radiation of the material. In this study, we propose a method of remotely collecting samples that guarantees radiation protection and worker safety at the site where radioactive waste is located.

As an initial part of Kori-1 & Wolsung-1 Unit decommissioning planning, a characterization plan is developed to define the nature, extent and location of contaminants, determine sampling locations and protocols, determine quality assurance objectives for characterization, and define documentation requirements. The actual characterization of a facility is an iterative process that involves initial sampling according to the characterization plan, field management (such as labeling, packaging, storing, and transport) of the samples, laboratory analysis, conformance to the data quality objectives (DQOs), and then identifying any additional sampling required, refining the DQOs, and modifying the characterization plan accordingly. The final product of the facility characterization is a document that describes the type, amount, and location of contaminants that will require consideration and removal during the decommissioning operations sufficient to prepare a decommissioning plan. In this study, implementing a characterization plan, developed in accordance with this standard, will result in obtaining or deriving the above information.

In order to start decommissioning domestic nuclear facilities, the Final Decommissioning Plan (FDP) must be prepared and approved by the regulatory agency. The contents of domestic FDP consist of 12 chapters, and there is the decommissioning feasibility design that should be described in Chapter 5 as contents to be considered from the construction stage of nuclear facilities. The design of decommissioning feasibility for nuclear facilities seems to be largely divided into three items. In summary, there ae minimization of contaminations to facilities and the environment, easy of dismantling, and minimization of the radioactive waste generation. In addition, the design characteristics to which the ALARA principle is applied in terms of optimizing the exposure dose of workers and residents may also correspond to the decommissioning feasibility design. The design characteristics for decommissioning feasibility during the period leading up to the design, operation, and decommissioning of nuclear facilities can be listed as the main points as follows. Minimization of facility contamination will include contents related to the leakage of systems and components, minimization of effluents to the environment will involve gaseous and liquid effluents from systems and components to the environment, easy of dismantling will involves history and inspection records during operation, and minimization of radioactive waste generation can be the contents related to the radioactive waste management plans. The design characteristics of facilities and equipment to meet the ALARA principles can be listed as follows. It means taking into account the benefits and costs of the design improvement plan, and the elimination of unnecessary radiation exposure can be maintained at the exposure dose ALARA, which is in line with the decommissioning feasibility design. Among the requirements of licensing documents for decommissioning domestic nuclear facilities is the decommissioning feasibility design. This item relates to the design characteristics for decommissioning considered in the construction stage of the facility and should present the effectiveness of measures for them until operation and decommissioning. In this study, the regulatory requirements presented in the construction and operation stage and the contents presented in the U.S. case were reviewed, and it is hoped that it will be used as reference for the preparation of FDP.

Radioactive Oxide is formed on the surface of the coolant pipe of the nuclear power plant. In order to remove the oxide film that is formed on the surfaces of the coolant pipe, chemical and physical decontamination technologies are used. The disadvantage of traditional technologies is that they produce secondary radioactive wastes. Therefore, in this study, the short-pulsed laser eco-friendly technology was used in order to reduce the production of secondary radioactive wastes. It was also used to minimize the damage that was caused to the base material and to remove the contaminated oxide film. The study was carried out using a Stainless steel 304 specimen that was coated with nickel-ferrite particles. Additionally, a transport robot was 3D modeled and manufactured in order to efficiently remove the oxide film from the coolant pipe of the nuclear power plant. The transport robot has a fixed laser head to move inside the horizontal and vertical pipes. The rotating laser head removes the contaminated oxide film on the inner surface of the coolant pipe. In the future, as a condition of the 1064nm short-pulsed laser ablation technique determined by basic analysis, we plan to analyze whether the transport robot is applicable to the radiation contamination site of the nuclear power plant.

With the rapid growth of nuclear power in China, a large number of dry wastes, which mainly include the high efficiency particulate air filters (glass fiber), cotton, polyethylene, and absorbent paper with low-level radioactivity and high volume, will be produced during the operation and maintenance of the nuclear power plants. Thermal plasma treatment is a world acceptable technology to incinerate and immobilize radioactive wastes, owing to the high volume reduction factor and the excellent chemical durability of the vitrified waste form. China has developed thermal plasma technology for the treatment of dry wastes from nuclear power plants for more than 15 years and the pilot plant has been constructed. This work will concentrate on the formulation of waste glass fiber to adapt to the vitrification process. A three-component (glass fiber-CaO-Na2O) constrained-region mixture experiment was designed and their viscosity data was mainly studied. The quadratic Scheffé model was used to plot the component effect on melting temperature. The retentions of simulated nuclides, such as Co, Sr, and Cs in the glasses were analyzed. In addition, the glass fiber as a glass matrix to immobilize residual ashes from the thermal plasma gasification of cotton, polyethylene, and absorbent paper was investigated as well.

Kori-1 and Wolseong-1 nuclear power plants were permanently shut down in June 2017 and December 2019, and are currently in the preparation stage for decommissioning. In this regard, it is necessary to secure nuclear power plant decommissioning capacity in preparation for the domestic decommissioning marketplace. To address this, the Korea Research Institute of Decommissioning (KRID) was established to build a framework for the development of integrated nuclear decommissioning technology to support the nuclear decommissioning industry. The institute is currently under construction in the Busan-Ulsan border area, and a branch is planned to be established in the Gyeongju area. Recently, R&D projects have been launched to develop equipment for the demonstration and support verification of decommissioning technology. As part of the R&D project titled “Development and demonstration of the system for radioactivity measurement at the decommissioning site of a nuclear power plant”, we introduce the plan to develop a radioactivity measurement system at the decommissioning site and establish a demonstration system. The tasks include (1) measurement of soil radioactive contamination and classification system, (2) visualization system for massive dismantling of nuclear facilities, (3) automatic remote measurement equipment for surface contamination, and (4) bulk clearance verification equipment. The final goal is to develop a real-time measurement and classification system for contaminated soil at the decommissioning site, and to establish a demonstration system for nuclear power plant decommissioning. The KRID aims to contribute and support the technological independence and commercialization for domestic decommissioning sites remediation of nuclear power plant decommissioning site by establishing a field applicability evaluation system for the environmental remediation technology and equipment demonstration.

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.

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.