The objective of this study is to investigate the safety awareness and effectiveness of the education and training for employees engaged in radiological emergency organization of the Korea Atomic Energy Research Institute (KAERI). In 2022, the questionnaire for the education satisfaction survey was revised to regulary evaluate the effect of edcation on perceptions of importance on emergency preparedness for nuclear research facilities. In line with, a standard questionnaire was created which covers 3 factors and 9 attributes, and the evaluation indicatior is based on a 5-point Likert scale. In 2023, the education on radiological emergency preparedness was conducted for 235 emergency staff. From May 24 to July 13, 2023, data was collected from a total of 235 emergency response personnels, including 28 new staffs and 207 maintenance staffs. Aa a result of response analysis, it was identified that education for radiological emergency response had a significant correlation with the promoting safety culture. It was found that senior emergency personnel with more years of experience are highly interested in radioactive disaster prevention and actively participate in and training. On the other hand, it was presented that new and less experienced groups tend to have a relatively high scored of the risk perception of nuclear research facilitites. Therefore, it is necessary to improve the practical curriculum in order to increase the participation of junior disaster prevention personnel in education and training, ensuring that they correctly recognize the risk of research facilities. This results are expected to be used to improve the quality of education and drills for radiological emergency response at KAERI.

In the nuclear fuel cycle (NFC) facilities, the failure of Heating Ventilation and Air Conditioning (HVAC) system starts with minor component failures and can escalate to affecting the entire system, ultimately resulting in radiological consequences to workers. In the field of air-conditioning and refrigerating engineering, the fault detection and diagnosis (FDD) of HVAC systems have been studied since faults occurring in improper routine operations and poor preventive maintenance of HVAC systems result in excessive energy consumption. This paper aims to provide a systematic review of existing FDD methods for HVAC systems therefore explore its potential application in nuclear field. For this goal, typical faults and FDD methods are investigated. The commonly occurring faults of HVAC are identified through various literature including publications from International Energy Agency (IEA) and American Society of Heating, Refrigerating and Air-Conditioning Engineers (ASHRAE). However, most literature does not explicitly addresses anomalies related to pressure, even though in nuclear facilities, abnormal pressure condition need to be carefully managed, particularly for maintaining radiological contamination differently within each zone. To build simulation model for FDD, the whole-building energy system modeling is needed because HVAC systems are major contributors to the whole building’s energy and thermal comfort, keeping the desired environment for occupants and other purposes. The whole-building energy modeling can be grouped into three categories: physics-based modeling (i.e., white-box models), hybrid modeling (i.e., grey-box models), and data-driven modeling (i.e., black-box models). To create a white-box FDD model, specialized tools such as EnergyPlus for modeling can be used. The EnergyPlus is open source program developed by US-DOE, and features heat balance calculation, enabling the dynamic simulation in transient state by heat balance calculation. The physics based modeling has the advantage of explaining clear cause-and-effect relationships between inputs and outputs based on heat and mass transfer equations, while creating accurate models requires time and effort. Creating a black-box FDD model requires a sufficient quantity and diverse types of operational data for machine learning. Since operation data for HVAC systems in existing nuclear cycle facilities are not fully available, so efforts to establish a monitoring system enabling the collection, storage, and management of sensor data indicating the status of HVAC systems and buildings should be prioritized. Once operational data are available, well-known machine learning methods such as linear regression, support vector machines, random forests, artificial neural networks, and recurrent neural networks (RNNs) can be used to classify and diagnose failures. The challenge with black-box models is the lack of access to failure data from operating facilities. To address this, one can consider developing black-box models using reference failure data provided by IEA or ASHRAE. Given the unavailability of operation data from the operating NFC facilities, there is a need for a short to medium-term plan for the development of a physics-based FDD model. Additionally, the development of a monitoring system to gather useful operation data is essential, which could serve both as a means to validate the physics-based model and as a potential foundation for building data-driven model in the long term.

Kori unit 1 and Wolsong unit 1 were permanently shut down in 2017 and 2019, respectively. Both plants were decided to demolish the building without reuse. Large structures must be demolished after removing systems and components in the building, and in the case of large structures, thorough planning is required because of the large scale of work. Therefore, in this study, important considerations in the phase of the demolition plan of large structures when decommissioning were analyzed. The demolition of large structures at nuclear facilities is major one phase of work within a broader decommissioning plan. Furthermore, the actual demolition of the structure (i.e., physical process) represents the last step in a process that begins with extensive planning and analysis. The National Demolition Association (NDA) has provided checklist items that should be considered before the start of a commercial demolition project and/or in the bid process. Important Considerations in the Phase of the demolition plan of large structures when decommissioning of nuclear facilities are Site knowledge and programs, Engineering survey/demolition plan, Hazardous and radioactive materials, Open air demolition, Financial and project management, Permits, Code adherence, and Special programs, Disposal pathway, Final site condition. The results of this study can be used as a basis for the Planning large structures demolition of the Kori unit 1 and Wolsong unit 1.

In Korea, Kori Unit 1 and Wolsong Unit 1, have been permanently shut down in 2017 and 2019, and more nuclear power plants are expected to be permanently shut down after continued operation successively. Spent fuel has been generated during operation and stored in spent fuel pools. Due to the expected saturation of spent fuel pools within the next several decades, transportation of a huge amount of spent fuel is anticipated to interim storage facilities or final disposal facilities, even though the specific location is not decided. The U.S. Nuclear Regulatory Commission (NRC) states that every environmental report prepared for the licensing stage of a Pressurized Water Reactor shall contain a statement concerning risk during the transportation of fuel and radioactive wastes to and from the reactor. Thus, the licensee should ensure that the radiological effects in accidents, as well as normal conditions in transport, do not exceed certain criteria or be small if cannot be numerically quantified. Specific conditions that a full description and detailed analysis of the environmental effects of transportation of fuel and wastes to and from the reactor are exempted are specified in 10 CFR Part 51. Since there are no official requirements for radiological dose assessment for workers and public during the transportation of spent fuel in Korea, the margin when applying the U.S. regulatory criteria to the environmental impact assessment during the transport of spent fuel generated from domestic nuclear power plants is evaluated. A different approach would be needed due to the difference in the characteristics of spent fuel and geographical features.

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.

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.

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.

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.

Dry active wastes (DAWs) are a type of combustible radioactive solid waste, which includes decontamination paper, protective clothing, filters, plastic bags, etc. generated from operating nuclear facilities and decommissioning projects. The volume of DAWs could be increased over time, disadvantage to higher disposal costs and space utilitization of disposal site. Additionally, incineration methods cannot be applied to DAWs, unlike general environmental waste, due to concerns about air pollution and the release of harmful chemicals with radioactive nuclides into the atmosphere. Recently, KAERI developed an alternative thermochemical process for reducing the volume of DAW, which involves a step-wise approach, including carbonization, chlorination, and solidification. The purpose of this process is to selectively separate the radioactive nuclides from carbonized DAWs that are less than clearance criteria, which can be disposed of as non-radioactive waste. In this research, we investigated the thermal decomposition characteristics of DAWs using nonisothermal thermogravimetric analysis, which was performed with different categorized wastes and heating conditions. As a result, the cellulose DAWs such as decontamination paper and cotton were thermally decomposed in three or four-step depending on the heating conditions. On the other hand, the hydrocarbon and rubber DAWs such as plastic bags and latex were thermally decomposed in one or two-step. Therefore, it could be suggested the thermochemical treatment conditions that minimize the decomposition of DAWs by controlling the reaction steps, and we will try to apply these results for cellulose type DAWs such as decontamination paper and cotton, which is generated majorly from the nuclear facilities in the future.

Working during decommissioning of nuclear facilities can subject workers to a number of industrial health and safety risks. Such facilities can contain hazardous processes and materials such as hot steam, harsh chemicals, electricity, pressurized fluids and mechanical hazards. Workers can be exposed to these and other hazards during normal duties (including slips, trips and falls, driving accidents and drowning). Industrial safety accidents, along with their direct impacts on the individuals involved, can negatively affect the image of nuclear facilities and their general acceptance by the public. Industrial safety is the condition of being protected from physical danger as a result of workplace conditions. Industrial safety program in a nuclear context are the policies and protections put in place to ensure nuclear facility workers are protected from hazards that could cause injury or illness. Preventive actions are those that detect, preclude or mitigate the degradation of a situation. They can be conducted regularly or periodically, one time in a planned manner, or in a predictive manner based on an observed condition. Corrective actions are those that restore a failed or degraded condition to its desired state based on observation of the failure or degradation. In industrial safety, the situations or conditions of interest are those observed via the performance monitoring, investigations, audits and management reviews. Preventive and corrective actions are those designed to place or return the system to its desired state. Preventive actions where possible are preferred as they eliminate the adverse condition prior to it occurring. When an accident or incident occurs, the primary focus is on obtaining appropriate treatment for injured people and securing the scene to prevent additional hazards or injuries. Once the injured personnel have been cared for and the scene has been secured, it is necessary to initiate a formal investigation to determine the extent of the damage, causal factors and corrective actions to be implemented. Certain tools may be needed to investigate such incidents and accidents. Initial identification of evidence immediately following the incident includes a list of people, equipment and materials involved and a recording of environmental factors such as weather, illumination, temperature, noise, ventilation and physical factors such as fatigue and age of the workers. The five Ws (what, who, when, where and why) are useful to remember in investigation of incidents and accidents.

Despite of careful planning of decommissioning projects, there are often surprises when facilities are opened for dismantling purposes, or when material is removed from hot cells, etc. Unexpected incidents and findings during the decommissioning of nuclear facilities have been referred to in the past as unknowns. However, many of the problems encountered during implementation of decommissioning are well known, it is simply that they were not expected to arise. In some other cases, the problem may not have been encountered in the decommissioning team’s experience, forcing the development of new techniques, tools and procedures to address the unexpected problem, with the attendant delays and cost overruns that this often involves. Unknowns in decommissioning cannot be eliminated, regardless of the efforts applied. This is especially the case in old facilities where documentation may have been lost or where modifications were carried out without updates to reports. As a result, when planning for decommissioning, it is prudent to assume that such problems will occur, and ensure that arrangements are in place to deal with them when they arise. This approach will not only improve the efficiency of the decommissioning project, but will also improve the safety of the operations. One of the most common root causes of unexpected events in decommissioning is the lack of detailed design information or missing records of modifications, maintenance issues and incidents during operation. It is therefore necessary to check the completeness of design information in existing plants and to ensure that configuration management techniques are applied at all stages of the lifetime of a plant. In the case of a new plant, archiving samples of materials can be a valuable source of information to support decommissioning planning. During the lifetime of plants, it is likely that modifications will be carried out involving the construction of new buildings. The opportunity should be taken in these circumstances to consider the layout, the physical size and other attributes of the plant to ensure that they do not make decommissioning of existing facilities more difficult and also to optimize the potential for reuse in support of the decommissioning of the whole site, later in the life of the facility. Characterization of all aspects of a plant is essential to reduce the number of unknowns and the likelihood of unexpected events. This characterization should be extensive, but there is a limit to the level of detail that should be sought as the cost versus benefit gain may reduce. Reducing unknowns by retrospectively obtaining physical data associated with a facility is a useful means of characterization, and there are many tools in existence that can be used to carry this out accurately and effectively. Regardless of the efforts that are employed in decommissioning planning, unexpected events should be anticipated and contingency plans prepared. Although the details of the event itself may not be anticipated, its impact may affect safety and environmental discharge, and may or may not involve radiological impacts. Regardless of more serious impacts, unexpected events are likely to result in modifications to the decommissioning plan, incur delays and cost money. Finally, regardless of the status of a facility, whether at the concept stage or at the decommissioning stage of its life cycle, it is never too early to begin thinking and planning for decommissioning.

Fault activity acts as the greatest risk factor in relation to the stability of the radioactive waste disposal facilities and nuclear power plant site, and for this reason, geological studies on areas with past fault activity history must precede site evaluation studies. This study aims to trace the fault activity history of large fault zones, including the Yangsan fault in the southeastern part of the Korean Peninsula, where two major earthquakes occurred, and to obtain fault activity direction information that is the basis for stability evaluation. The 3D-Shape Preferred Orientation (SPO) of particles in the fault rock created by the earthquake was investigated to analyze the direction of fault plane activity, and the age of fault activity was estimated through Illite Age Analysis (IAA) analysis. It is expected that the large-scale fault activity information in the southeastern part of the Korean Peninsula obtained through the SPO and IAA analysis can be used as basic data for safety evaluation of existing or future nuclear power plants and radioactive waste facilities.

In Korea, the construction of dry storage facilities for spent nuclear fuel is being promoted through the 2nd basic plan for high-level radioactive waste management. When operating dry storage facilities, exposure dose assessment for workers should be performed, and for this, exposure scenarios based on work procedures should be derived prior. However, the dry storage method has not yet been sufficiently established in Korea, so the work procedure has not been established. Therefore, research is needed to apply it domestically based on the analysis of spent nuclear fuel management methods in major overseas leading countries. In this study, the procedure for receiving and storing spent nuclear fuel in a concrete overpack-based storage facility was analyzed. Among the various spent nuclear fuel management systems, the metal overpack-based HI-STAR 100 system and the concrete overpackbased HI-STORM 100 system are quite common methods in the United States. Therefore, in this study, work procedures were analyzed based on each final safety analysis report. First, the HI-STAR 100 overpack enters the facility and is placed in the transfer area. Remove the impact limiter of the overpack and install the alignment device on the top of the overpack. Place the HI-TRAC, an on-site transfer device, on top of the alignment unit and remove the lids of the two devices to insert the canister into the HI-TRAC. When the canister transfer is complete, reseat the lid to seal it, and disconnect the HI-TRAC from the HI-STAR 100. Raise the canister-loaded HI-TRAC over the alignment device on the top of the HI-STORM 100 overpack and remove the lids of the two devices that are in contact. Insert the canister into the HI-STORM 100 and reseat the lid. The HI-STORM 100 loaded with spent nuclear fuel is transferred to the designated storage area. In this study, the procedure for receiving and storing spent nuclear fuel in a concrete overpack-based storage facility was analyzed. The main procedure was the transfer of canisters between overpacks, and it was confirmed that HI-TRAC was used in the work procedure. The results of this study can be used as basic data for evaluating the exposure dose of operating workers for the construction of dry storage facilities in Korea.

On April 28, 2022, a North Korean hacker (operator) recruited an active officer in exchange for virtual currency to steal military secrets and attempted to hack the battlefield network, and it was revealed that he tried to use PoisonTap during the investigation. Let’s analyze whether these events pose a threat to nuclear facilities. PoisonTap is a tool coded in the Node.js language to Raspberry Pi Zero and weaponized. When connected to a target PC with a USB or Thunderbolt port, the target PC can be occupied in about a minute. PoisonTap materials include Raspberry Pi Zero, a USB expansion port that can be connected to the target PC, a space where the code can operate (code about 12 MB), and Node.js (weaponizable code), which can be made without much difficulty. PoisonTap’s functionality allows cookies and sessions to be stolen through hijacking and allows remote access by exposing internal routers to the outside. Some of the reasons why PoisonTap occurred are that network devices connect directly to the computer without any conditions. And one of the big problems of this vulnerability is the design problem of the Internet itself, so it is difficult to block or defend technically. It is difficult to protect if it is simply a software problem because it is different from how to fix it through software code modification. According to the PoisonTap principle analysis, it connects the PoisonTap to the target PC based on the network’s characteristics (subnets of lower-priority network devices are given higher priority than gateways of higher-priority network devices). The HTML+Javascript generated while being connected becomes a backdoor that can be connected anytime. In other words, by creating a Websocket that can be connected to the web browser itself at any time, an attacker can connect to the target PC at any time. In such a threat, PoisonTap is used to break in and install a web backdoor on the target PC to make it continuously accessible and attack even if the PoisonTap is disconnected. This problem is believed to be an insider threat not only to military units but also to nuclear facilities that are closed networks. PoisonTap can be brought into major nuclear areas in cooperation with insiders with general maintenance of USB equipment. Ordinary workers often leave their laptops or leave them for a while by inserting a screen-saver password. In addition, because there is no communication with the outside, actions that do not seal USB ports and enter deep sleep mode (network connection) can be exposed as cyber threats to nuclear facilities using PoisonTap by malicious insiders.

In Korea, additional regulatory requirements are increasing due to the full-scale decommissioning of nuclear power plants following the permanent shutdown of Kori Unit 1 and Wolseong Unit 1. Accordingly, it is necessary to preemptively expand the scope of physical protection regulations from design, construction, and operation stage to back-end nuclear fuel cycle such as cessation of operation and decommissioning. According to Article 2, Paragraph 24 of the Nuclear Safety Act, the decommissioning of nuclear facilities is defined as all activities to exclude them from the application of the Nuclear Safety Act by permanently suspending the operation of nuclear facilities, demolishing the facilities and sites, or removing radioactive contamination. In other words, it refers to a series of technologies or activities to safely and efficiently dismantle nuclear power plant and remove radioactive contamination and restore them to their original state after permanently shut down of nuclear power plant. Security changes during decommissioning and decontamination since removing fuel from the reactor alters the plant’s safety status, some of the systems or components considered as vital equipment during plant operation will no longer be needed. The vital areas may be reduced as fewer buildings, equipment and systems need to be protected, which means access controls, surveillance and so on can be reduced. And also, decommissioning will probably require more workers than operation would, although this might not be the case at all times. From a security point of view, this might require more personnel or additional access points. Changing operating require changed security measures, to ensure that the required security level will be maintained while at the same time work proceeds efficiently. Once all of the fuel is removed from the plant, radiological release risk is much lower. The lower risk requires a lower level of security measures. Even during the removal of nuclear material and contaminated equipment from nuclear facilities, lower level of security measures should meet regulatory requirements based on a graded approach. Therefore, this study intends to examine the responsibilities and obligations of regulatory authorities, regulator, and nuclear operators in terms of protection after permanent shutdown and decommissioning.

Satellite imagery is an effective supplementary material for detecting and verifying nuclear activities and is helpful in areas where access and information are limited, such as nuclear facilities. This study aims to build training data using high-resolution KOMPSAT-3/3A satellite images to detect and identify key objects related to nuclear activities and facilities using a semantic segmentation algorithm. First, objects of interest, such as buildings, roads, and small objects, were selected, and the primary dataset was built by extracting them from the AI dataset provided by AIHub. In addition, to reflect the features of the area of interest (e.g., Yongbyon, Pyongsan), satellite images of the area were acquired, augmented, and annotated to construct an additional dataset (approximately 150,000). Finally, we conducted three stages of quality inspection to improve the accuracy of the training data. The training dataset of this study can be applied to semantic segmentation algorithms (e.g., U-Net) to detect objects of interest related to nuclear activities and facilities. Furthermore, it can be used for pixelbased object-of-interest change detection based on semantic segmentation results for multi-temporal images.

Milling facilities, which belong to the front end of the nuclear fuel cycle, are essential steps for utilizing uranium in nuclear power generation. These milling facilities currently provide the International Atomic Energy Agency (IAEA) with the location and annual production capacity of the facility through the Additional Protocol (AP, INFCIRC/540) and grant IAEA inspectors on-site sampling authority to apply safeguards to the facility. However, since milling facilities process a large amount of nuclear material and the product uranium ore concentrate (UOC) is bulk material, the absence of accounting for the facility could pose a potential risk of nuclear proliferation. Therefore, this study proposes technical approach that can be utilized for safeguards in milling facilities. Since the half-life of uranium isotopes is much longer than that of its daughter, they reach the secular equilibrium condition. However, after milling process, the fresh tailing showed the break of that secular equilibrium. As time goes on, they newly reach another secular equilibrium condition. Based on this observation, this study discussed the feasibility of the ratio method in safeguards purpose. The scenario applied in this study was 1% of uranium mill tailing. It was observed that the U-238/Th-234 and U- 238/Pa-234m ratios in fresh milling tails varied as a function of time after discharging, particularly during the first one year. This change can be worked as a significant signature in terms of safeguards. In conclusion, the ratio method in mill tails could be applicable for safeguards of nuclear milling facility.

The International Atomic Energy Agency (IAEA) entails independent decision-making for the safety supervision of civil nuclear facilities. To evaluate and review the safety of nuclear facilities, the national regulatory body usually consults independent institutions or external committees. Technical Support Organizations (TSOs) include national laboratories, research institutions, and consulting organizations. Support from professional organizations in other countries may also be required occasionally. Most of the world’s major nuclear power countries adopt an independent nuclear safety supervision model. Accordingly, China has continuously improved upon the construction of such a system by establishing the National Nuclear Safety Administration (NNSA) as the decision-making department for nuclear and radiation safety supervision, six regional safety supervision stations, the Nuclear and Radiation Safety Center (NSC), a nuclear safety expert committee, and the National Nuclear and Radiation Safety Supervision Technology R&D Base, which serves as the test, verification, and R&D platform for providing consultation and technical support. An R&D system, however, remains to be formed. Future endeavors must focus on improving the technical support capacity of these systems. As an enhancement from institutional independence to capability independence is necessary for ensuring the independence of China’s nuclear safety regulatory institution, its regulatory capacity must be improved in the future.