According to the “Law on protection and response measures for nuclear facilities and radiation”, Nuclear Power Plant (NPP) licensees should conduct periodic exercises based on hypothetical cyberattack scenarios, and there is a need to select significant and probable ones in a systematic manner. Since cyber-attacks are carried out intentionally, it is difficult to statistically specify the sequences, and it is not easy to systematically establish exercise scenarios because existing engineering safety facilities can be forcibly disabled. To deal with the above situation, this paper suggests a procedure using the Probabilistic Safety Assessment (PSA) model to develop a cybersecurity exercise scenario. The process for creating cyber security exercise scenarios consists of (i) selecting cyber-attack-causing initiating events, (ii) identifying digital systems, (iii) assigning cyber-attack vectors to a digital system, (iv) determining and adding type for operator’s response, (v) modifying a baseline PSA model, and (vi) extracting top-ranked minimal cut sets, and (vii) selecting a representative scenario. This procedure is described in detail through a case study, an expected cyber-attack scenario General Transient-Anticipated Transient Without Scram (GTRN-ATWS). It refers to an accident scenario for ATWS induced by GTRN. Since ATWS is targeted for cyber training in some NPPs, and GTRN is one of the most common accidents occurring in NPPs, GTRN-ATWS was chosen as an example. As for the cyber-attack vector, portable media and mobile devices were selected as examples based on expert judgment. In this paper, only brief examples of GTRN-ATWS events have been presented, but future studies will be conducted on an analysis of all initiating events in which cyber-attacks can occur.

The characterization of nuclear materials is crucial for global nuclear safeguards efforts, as these materials can potentially be used for illicit purposes. In this study, we evaluated the applicability and performance of the In-Situ Object Counting System (ISOCS) equipment for the characterization and quantification of uranium, including uranium pellets and radioactive wastes. Our methodology involved using ISOCS to measure samples with different enrichments and total amounts of uranium, and to analyze the results in order to evaluate the ISOCS’s effectiveness in accurately characterizing the various uranium samples. To this end, we compared the ISOCS results with those of the Multi-Group Analysis for Uranium (MGAU) system, which is currently used in the field of international safeguards. The results of this study showed that the ISOCS was sensitive enough to analyze small amounts of uranium pellet, with %differences ranging from -0.7% to 19%. However, when analyzing shielded nuclear materials like in concrete waste, the uncertainty was relatively high, with %differences ranging from 11% to 67%. On the other hand, the MGAU system was unable to analyze uranium for the same spectrum, indicating the superiority of the ISOCS in terms of usability. The ISOCS instrument was also found to be effective in analyzing uranium in various types of samples without the need of standard sources. Overall, the findings of this study have important implications for the development of more effective safeguards strategies for the characterization of nuclear materials. The ISOCS instrument could be a reliable tool for analyzing nuclear materials, contributing to global safeguards efforts to reduce the risk of nuclear proliferation.

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.

Nuclear fusion energy is considered as a future energy source due to its higher power density and no emission of greenhouse gas. Therefore, various researches on nuclear fusion is being conducted. One of the key materials for the nuclear fusion research is tritium because the D-T reaction is the main reaction in the nuclear fusion system. However, that tritium can also be used for non-peaceful purposes such as hydrogen bombs. Therefore, it is necessary to establish the safeguards system for tritium. In that regards, this study analyzed the possibility of applying safeguards to tritium. To achieve this objective, the tritium production capacity through the light water reactor was analyzed. Tritium Production Burnable Absorber Rod (TPBAR) was modeled through the MCNP code, and tritium production was analyzed. The TPBAR is composed of a cylindrical tube with a double coating of aluminum and zirconium, filled with a sintered lithium aluminate (LiAlO2) pellet to prevent the release of tritium. Tritium is produced by the reaction of Li-6 in the TPBAR with neutrons, and it is extracted and stored at the Tritium Extraction Facility (TEF). As a result, the tritium production increased as the burnup and Li-6 mass increased. In addition, when the tritium produced in this way was transferred to TEF and extracted through the process, the application of safeguards measures was analyzed. To this end, various safeguards measures were devised, such as setting an Material Balance Area (MBA) for TEF and analyzing Material Balance Period (MBP). As there is no designated Significant Quantity (SQ) for tritium, cases were classified based on the type and form of nuclear weapons to estimate the Sigma MUF (Material Unaccounted For) of the TEF. Finally, the comprehensive application of safeguards to tritium was discussed. This research is expected to contribute to the establishment of IAEA safeguards standards related to tritium by applying the findings to actual facilities.

The configuration management system was implemented on the basis of a document management system that secured stable understanding, scalability, document security, and convenience in small modular reactor. To reduce the cost and risk of errors, configuration management is implemented to maintain a balance between design requirements, physical configuration, facility configuration information. In the initial stages, configuration change review procedures was developed with the main purpose of change management such as classification system management, configuration control committee management, configuration change review preparation, configuration control committee operation, followup measures, current status and tracking management. The preparation of the configuration change review consisted of preparation, distribution approval, designation of reviewers, review, collection of review opinions, and preparation of resolution results. In the operation of the configuration control committee, it was conducted by designating review members, reviewing members, collecting operation, and approval them. The next step is to supplement and develop the requirements of IEEE Std 828-2012, such as configuration management planning, configuration management control, configuration identification, configuration change control, configuration status monitoring, configuration audit, interface management, and release management. Through this, issue raising, action management, and baseline management will be implemented.

To evaluate the safeguards system or performance in a facility, it is crucial to analyze the diversion path for nuclear materials. However, diversion paths can range from the extremely simplified to the complicated depending on the level of knowledge and the specific person conducting the analysis. This study developed the diversion path analysis tools using an event tree and fault tree method to generating diversion paths systematically. The essential components of the diversion path were reviewed, and a logical flow was developed for systematically creating the diversion path. An algorithm was created based on the facility design components and logical flow, as well as the initial information of the nuclear materials and material flows. The event tree and fault tree analysis tools were used to test the path generation algorithm. The usage and limitations of these two logic methods are discussed, and ideas to incorporate the logic algorithm into practical program tools are suggested. The tests were analyzed on a typical light water reactor as an example, including automatic generation of dedicated pathways, configuration of safeguards measures, and analyzing paths with strategies for avoiding safeguard systems. The results led to the development of a draft pathway analyzer program that can be applied to general nuclear systems. The results of this study will be used to develop a program module that can systematically generate diversion paths using the event tree and fault tree method. It can help to guide and provide practical tools for implementing SBD.

The global nuclear nonproliferation regime has developed over the past 50 years based on the Treaty on the Non-Proliferation of Nuclear Weapons (NPT) with three pillars: disarmament, nonproliferation and peaceful use of nuclear energy. Due to climate change and energy security in recent years, nuclear energy has been in the spotlight as an electricity generation source, and many countries are paying attention to introducing nuclear power plants (NPP). Whereas exporters pursue profit by selling their NPP, international organisations and member states that seek nuclear nonproliferation are concerned with potential proliferation risks by expanding the nuclear power industry worldwide. Simultaneously, the member states’ right to peaceful use of nuclear energy has to be guaranteed as specified in NPT Article IV. Accordingly, the trade of nuclear power between the member states taking full responsibility is desirable from the nonproliferation perspective. This paper investigates whether the countries capable of exporting their nuclear power have complied with the global nuclear nonproliferation regime, deriving the role and position that South Korea is faced with, accordingly, has to take. The dynamics of exporters’ competitiveness are discussed, emphasising that compliance with the regime must be considered a qualification when exporting NPP. The achievement that South Korea has attained, fulfilling its role and responsibility under the regime, is highlighted. Since South Korea has developed the nuclear power industry in cooperation with the United States under the NPT and the ROK-US Agreement for Peaceful Nuclear Cooperation, the status quo of the two countries in the nuclear nonproliferation and industrial landscape is discussed. Among the newcomers who have officially announced the plan to introduce NPP, Saudi Arabia is put in a crucial position to aggregate or alleviate nuclear nonproliferation. To this end, the rationale for the ROK-US cooperation is proposed, evaluating the value of nuclear nonproliferation in support of exporting nuclear power.

Domestic nuclear power plants have developed radiological emergency plans based on the USNRC’s NUREG-0654/FEMA-REP-Rev.1 report and the Korea Institute of Nuclear Safety’s (KINS) research report on radiation emergency criteria for power reactors (KINS/RR-12). NUREG-0654 is a US emergency planning guide for nuclear power plants and provides detailed technical requirements for the content of radiological emergency plans. The document classifies radiological emergencies into three levels: Alert, Site Area Emergency, and General Emergency, which correspond to the white, blue, and red emergency levels used in domestic nuclear power plants. KINS/RR-12 is a technical guidance document published by the Korea Institute of Nuclear Safety in 2012, which divides radiological emergency criteria into criteria for pressurized water reactors (PWRs) and criteria for boiling water reactors (BWRs), and describes in detail the regulatory position and implementation of radiological emergency criteria for domestic PWRs and BWRs. The physical protection-related radiation emergency criteria included in the radiological emergency plan are specified in the radiological emergency criteria guidelines. There are two items each related to white and blue emergencies and one item related to red emergencies. Standard order of emergency plan lists the physical protection-related radiological emergency criteria for domestic PWRs and BWRs, which are identical according to the radiological emergency criteria guidelines. To enhance the physical protection regulation, the legal and regulatory basis for target set identification and vital area identification need to be established by considering radiological and physical protection emergency plan.

The domestic representative nuclear fuel cycle facilities are post-irradiation examination facility (PIEF) and Irradiated Examination Facility (IMEF) at KAERI. They have regularly operated since 1991 and 1993, respectively. Due to the long period of use, the facilities are ageing, and maintenance costs are increasing every year. The maintenance methods have mainly been breakdown maintenance (BM) and partially preventive maintenance (PM). They involve replacing components that have problems through periodic inspections by on-site inspectors. However, these methods are not only uncertain in terms of replacement cycles due to worker’s deviation on the inspection results, but also make it difficult to respond accidents developed through failures on the critical equipment that confines radioactive material. Therefore, an advanced operation and maintenance studied in 2022 through all of nuclear facilities operated at KAERI. Advancement strategy in four categories (safety, sustainability, performance, innovativeness) was analyzed and their priorities according to a facility environment were determined so a roadmap for advanced operation and maintenance could be developed. The safety and sustainability are higher importance than the performance and innovativeness because facilities at KAERI has an emphasis on research and development rather than industrial production. Thus, strategy for advancement has focused even more on strengthening the safety and sustainability. To enhance safety, it has been identified that immediate improvement of aged structures, systems, and components (SSCs) through large-scale replacement is necessary, while consideration of implementing an ageing management program (AMP) in the medium to long term is also required. Facility sustainability requires strengthening operation expertise through training, education, and cultivation of specialized personnel for each system, and addressing outstanding regulatory issues such as approval of radiation environment report on the nuclear fuel processing facilities and improvement work according to fire hazard analysis. One of the safety enhancement methods, AMP, is a new maintenance approach that has not been previously applied, so it had to be thoroughly examined. In this study, an analysis was conducted on the procedure and method for introducing an AMP. An AMP for nuclear fuel cycle facilities was developed by analyzing the AMP applied to the BR2 research reactor in Belgium and modifying it for application to nuclear fuel cycle facilities. The ageing management for BR2 has the objective to maintain safety, availability and cost efficiency and three-step process. The first step is the classification of SSCs into four classes to apply graded approach. Secondly, ageing risk is assessed to identify critical failure modes, their frequency and precursors. Final step involves defining measures to reduce the ageing risk to an acceptable level in order to integrate the physical and economic aspects of ageing into a strategy for inspection, repair, and replacement. Similar approach was applied to the nuclear fuel cycle facility. Firstly, the SSCs of nuclear fuel cycle facilities have been classified according to their safety and quality classifications, as well as whether they are part of the confinement boundary. The SSCs involved in the confinement boundary were given more weight in the classification process, even if they are not classified as safety-class. A risk index for ageing was introduced to determine which prevention and mitigation measure should be chosen. By multiplying the health index and the impact index, the ageing risk matrix provides a numerical score that represents guidance on the prevention and mitigation of ageing effect. The health index is determined by combining the likelihood of failure and engineering evaluation of the current condition of SSCs, whereas the impact index is calculated by taking into account the severity of consequences and the duration of downtime resulting from a failure. This ageing management has to be thoroughly reviewed and modified to suit each facility before being applied to nuclear fuel cycle facilities.

Owing to the increase in saturation rate of the spent fuel storage pond in the Kori nuclear power plant, the interim spent fuel dry storage facility is scheduled to be constructed at the Kori site. To implement safeguards in the new dry storage facility effectively, the concept of “Safeguards-by- Design” (SBD) should be applied to reflect nuclear safeguard provisions in the earliest design stages. Detailed design information pertaining to dry storage facilities has not been determined; however, the design information related to safeguards have been inferred using case studies and interviews with nuclear power plant operators worldwide. On the basis of the results of the case studies on spent fuel dry storage facilities for light water reactors, most countries apply the metal cask method in containment buildings considering safety. Furthermore, Korean operators are also considering the same method owing to tight licensing schedules and safety issues. Using the Facility Safeguardability Assessment (FSA) methodology (one of the safeguard evaluation methodologies), the difference in design between the heavy water reactor spent fuel dry storage facility, an established IAEA safeguards approach reference nuclear facility, and the light water reactor spent fuel dry storage facility (the new nuclear facility) were analyzed. Two major differences were noted as issues pertaining to potential safeguards. First, the difference in design and transport method in terms of the difference in size and weight of the spent nuclear fuel is important; light water reactor fuel is 20 times heavier than heavy water reactor that needs partial defect inspection in assemblies. Second, the difference in safeguard approach owing to the difference between the modular storage method in heavy water reactor and the container type storage method in light water reactor must be considered; movable storage cask renders the IAEA surveillance approach difficult. The results of this study can be used to identify the safeguards requirements in advance, enabling the operator to design new dry storage facilities resulting in timely and cost-effective implementation.

An administrative agreement (AA) was signed between NSSC and UAE FANR in January 2023 under Article 5 of the ROK-UAE Nuclear Cooperation Agreement. The AA aims to enhance regulatory efficiency in safeguards and export control. This study reviewed the export control measures for the items subject to the agreement (ISA) and implementation procedures under ROK-UAE AA by comparing them with other countries cases. First of all, the ROK-UAE AA distinguishes between ISA and the inventory management target items. Technology is divided into two categories, one requiring consent for retransfer and the other, considering the characteristics of technology that is free to be copied and deleted, and thus less useful for inventory management. Only the former is included in the annual report, which differs from the ROK-Canada or ROK-Japan NCA, which includes all technologies subject to the agreements in the annual report. When ROK notifies export information, it is mandatory to specify whether the technology requires consent for retransfer. Furthermore, some technologies should be controlled as strategic information, even if excluded from the annual report, so efforts to prevent confusion are required. Secondly, the ROK-UAE AA covers all items in INFCIRC/254/rev.9/part1, unlike the ROK-U.S. and ROK-Canada NCA, which listed equipment subject to them. This is significant because it clarifies the criteria for regulation by increasing the consistency between the trigger list items in the domestic law and the ISA. However, the expanded ISA scope could result in some changes in export control procedures. For example, when importing nuclear material (NM) from the US, only uranium was controlled as ISA, and the packages were not considered. In contrast, when exporting fuel assemblies (FA) for UAE, both uranium and zirconium cladding should be treated as ISA. To this end, NEPS was improved to implement the features of the ROK-UAE AA. Consideration of the criteria and methods for imposing obligations under the agreement is essential because this is the first case of Korea concluded AA under exporting NPP and as a supplier of FA. Generally, the obligations for NM are imposed by the country of origin, conversion, and enrichment countries. Canada and EU recognize the fuel fabrication process as a substantial transformation and impose customs origin where the process takes place. Hence, NM fabricated from Canadian equipment is also subject to the same obligations as NM of Canadian origin. From this perspective, it would be appropriate to ensure ROK acts as a supplier and controls when exporting domestically manufactured FA. Moreover, a proper national obligation code system will be required to specify Korea’s control rights.

KINAC began dispatching the resident inspector in 2012 to strengthen on-site Wolsong nuclear power plants (NPPs) regulations. The dispatched resident inspector is a member of the regional office of the Nuclear Safety and Security Commission (NSSC) and is in charge of technical support, on-site regulation of safeguards, and physical protection for the Wolsong regional office of NSSC. As the number of nuclear facilities in the ROK increased, the resident inspectors began to be dispatched to other regional offices. The resident inspectors were assigned to Hanul in November 2015, Kori in March 2017, Hanbit in March 2015, Saeul in March 2022, and Wolsong in March 2023. Accordingly, this paper intends to reflect on the increasing role of resident inspectors and predict on-site regulatory work in the field of nuclear control. The role of the resident inspectors is described in detail in the internal regulations of KINAC. Among the tasks in the common field is technical support at regional offices for the most critical areas of nuclear control implementation, and on-site verification of the matters requested by the director of each implementation division shall be carried out. Tasks in the field of safeguards include an on-site check of facility regulation review, implementation of national inspections, technical support for IAEA inspections, and information management. Among them, technical support work for Unannounced inspections should be the top priority. These days, in particular, the importance of reviewing the results of checking advanced information and containment and surveillance equipment by facility operators is emerging. Among the tasks performed by the resident inspectors, more than 80% of the functions related to physical protection account for. The resident inspectors check the status of the physical protection system by weekly/monthly/quarter, implement physical protection regulation review and inspection, conduct exercise evaluation, and perform technical support for special assessments. Recently, regulatory activities related to radioactive terrorism and the emergence of illegal drones have been strengthened. In the field of cybersecurity, where its role has recently been increasing, the resident inspectors are performing basic field regulation tasks. Similar to the area of physical protection, the resident inspectors check the cybersecurity system for weekly, monthly, and quarterly readiness, and on-site inspections of cybersecurity review and inspection technical support, exercise evaluation, and other requests are mainly performed. The role of the resident inspectors is expected to expand further in the future due to the increase in terrorist risks at home and abroad and changes in the regulatory environment. However, there is a limit to performing an increasing number of tasks, with the human resources of the resident inspectors limited to one to two for each site. If the resident inspectors are dispatched for each field of safety measures, physical protection, and cybersecurity, they can perform their duties more efficiently, but problems may arise in the operation of our personnel. Therefore, the proper and precise allocation of work while maintaining the current system is an essential part. The roles and prospects of the resident inspectors analyzed in this paper can be used to deploy the headquarters and field regulation personnel and set the direction of work in the future.

According to the ROK-IAEA Comprehensive Safeguards Agreement (CSA), the ROK submits inventory change reports (ICRs), physical inventory lists (PILs), and material balance reports (MBRs). Suppose inventory changes occur in each material balance area (MBA). In that case, the facility operators prepare ICRs monthly, conduct physical inventory taking (PIT) every 12 to 18 months, and submit PILs and MBRs to KINAC. KINAC reviews ICR presented by the facility operators, submits it to the IAEA, and reports it to the Nuclear Safety and Security Commission (NSSC). Various methods have been prepared and implemented to minimize errors in reviewing the accounting reports submitted by the facility operators. Accordingly, this paper analyzes the mistakes in the accounting reports that occurred over the past two years and proposed methods to improve them. The basis for carrying out the accounting reports is stipulated mainly in the CSA and the Nuclear Safety Act. First, Article 63 of the CSA describes the rationale for submitting the accounting reports, and the details are described in detail in the subsidiary arrangement. Article 98 of the Nuclear Safety Act stipulates information related to accounting reports, and details are described in the regulations on reporting internationally regulated materials, etc., of the NSSC Notice No. 2017-84. Among the accounting reports submitted in 2021, a total of 36 errors were confirmed. There were ten errors related to inventory changes, followed by six errors in the material balance period (MBP) in the header information. There were four cases of spacing, weight mismatch, and overdue errors, and the rest were related to grammar errors. There were a total of 30 errors in the accounting reports identified in 2022. MBP errors of header information, which occurred the second most in 2021, was the highest with nine, followed by six inventory change errors and five weight mismatch and overdue errors, respectively. Compared to 2021, the total number of errors has decreased by about six, which is interpreted as the result of outreach activities through accounting reporting workshops and nuclear control education conducted by KINAC. Accounting reporting is the most critical part of the Nuclear Material Accounting and Control (NMAC) system. Efforts to check errors in accounting reports and improve report quality through outreach activities could be confirmed by the statistics of the two years analyzed earlier. In the future, if the reporting program used by the facility operators is improved to minimize errors and manage the accounting reporting system through continuous maintenance work, the quality of the accounting reports will be upgraded to the next level.

Nuclear Safety and Security Commission (NSSC) and KINAC review a Cyber Security Plan (CSP) by「ACT ON PHYSICAL PROTECTION AND RADIOLOGICAL EMERGENCY」. The CSP contains cyber security implementation plans for the licensee’s nuclear power plant, and it shall meet the requirements of KINAC/RS-015, a regulatory standard. The KINAC/RS-015 provides more detailed information on the legal requirements, so if licensees implement cyber security under the approved CSP, they can meet the law. To protect nuclear facilities from cyber-attacks, licensees should identify their essential digital assets, so-called “Critical Digital Assets” (CDAs). Then, they apply cyber security controls (countermeasures for cyber-attacks) on CDAs consisting of technical, operational, and management security controls. However, it is hard to apply cyber security controls on CDAs because of the large amounts of CDAs and security controls in contrast to the shortage of human resources. So, licensees in the USA developed a methodology to solve this problem and documented it by NEI 13-10, and US NRC endorsed this document. The main idea of this methodology is, by classifying CDAs according to their importance, applying small amounts of security controls on less important CDAs, so-called non-direct CDAs. In the case of non-direct CDAs, only basic cyber security controls are applied, that is, baseline cyber security controls. The baseline cyber security controls are a minimum set of cyber security controls; they consist of control a) from control g) a total of 7 controls. Although non-direct CDAs are less critical than other CDAs (direct CDAs), they are still essential to protect them from cyber-attacks. This paper aims to suggest a cyber security enhancement method for non-direct CDAs by analyzing the baseline cyber security controls. In this paper, baseline cyber security controls were analyzed respectively and relatively and then concluded how to apply small amounts of cyber security controls on non-direct CDAs rather than direct CDAs without scarifying cyber security.

Licensees are required to protect critical digital assets (CDAs) in nuclear facilities against cyber-attacks, up to and including design basis threat (DBT), according to「ACT ON PHYSICAL PROTECTION AND RADIOLOGICAL EMERGENCY」. However, CDAs may be excluded from cyber security regulations at nuclear power plant decommissioning, and this may lead to severe consequences if the excluded CDAs contain sensitive information such as the number and location of nuclear fuels and information on security officers. In that case, that information could be leaked to the adversary without adequately removing the information before discarding the CDAs. It can be potentially abused to threaten nuclear facilities inducing radiological sabotage and nuclear material theft. So, controls of sensitive information are needed. This study aims to derive regulatory improvements related to discarding CDAs that have sensitive information by analyzing foreign cases such as IAEA and U.S. NRC. The sensitive information in the IAEA guide is the following: (1) details of physical protection systems and any other security measures in place for nuclear material, other radioactive material, associated facilities, and activities; (2) information relating to the quantity and form of nuclear material or other radioactive material in use or storage; (3) information relating to the quantity and form of nuclear material or other radioactive material in transport; (4) details of computer systems; (5) contingency and response plans for nuclear security events; (6) personal information; (7) threat assessments and security alerting information; (8) details of sensitive technology; (9) details of vulnerabilities or weaknesses that relate to the above topics; (10) historical information on any of the above topics. In the case of the U.S. NRC, they categorize sensitive information into three groups: (1) classified information, (2) safeguard information (SGI), (3) sensitive unclassified non-safeguards information (SUNSI). Classified information is information whose compromise would cause damage to national security or assist in manufacturing nuclear weapons. The SGI concerns the physical protection of operating power reactors, spent fuel shipments, strategic special nuclear material, or other radioactive material. Finally, SUNSI is generally not publicly available information such as personnel privacy, attorney-client privilege, and a confidential source. IAEA recommends protecting the above sensitive information in accordance with NSS No.23-G (Security of Nuclear Information), and NRC protects classified information, SGI, and SUNSI under relative laws. In the case of ROK, if security control measures are enhanced CDAs that possess sensitive information, the risk of information leakage will be decreased when those CDAs are discarded.

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.

The purpose of this study is to detect future signals and changes in nuclear-related research to apply safeguards by design to new nuclear facilities or to determine nuclear fuel cycle-related research and development (R&D) activities. First, a total of 2,029 scientific articles published between 2015 and 2022 in the journal of “Nuclear Engineering and Technology” by the Korean Nuclear Society were collected. The authors of the scientific article used their expertise and knowledge to select keywords that can properly represent the article. Therefore, in this study, the keywords of each scientific article were analyzed using the technique of text mining. We then calculated the “word frequency” and “term frequency-inverse document (TF-IDF)” values of the keywords. Consequently, significant words such as “reactor,” “nuclear,” and “fuel” were extracted, which were represented as word clouds. Furthermore, keywords extracted through text mining were quantitatively classified into weak or strong signals using a keyword emergence map (KEM). The KEM is a tool that can explore future signals because essential keywords have a high frequency of appearance, and newer keywords are more important than older keywords. The KEM results showed no keywords in the strong-signal area in the field of nuclear academia. However, keywords such as “deep learning,” “earthquake,” “zircaloy,” and “CFD” were confirmed to be distributed in the weak signal area. A weak signal indicates the most probable topic that could become a strong signal in the near future. The weak signal methodology can be applied to predict future nuclear scientific trends in the rapidly changing world. Based on the results of the study, changes in the subject of nuclear-related scientific articles over the past eight years and future signals were interpreted. The results confirmed that this method can be applied to safeguards measures of new nuclear facilities in the design stage and can be used to detect R&D activities related to the nuclear fuel cycle in advance.

The concept of catch-all controls is not new, having been developed in the US and EU over 30 years ago and legislated about 20 years ago. It is a system that controls what is not in an item based on the end-user and end-use rather than the strategic items defined by the existing multilateral export control regimes. Most importantly, the controls are based on end-user or end-use. Catch-all controls are the best tool to solve the problems of traditional list based controls. Technological advances, emerging technologies, and globalization have made standard export controls insufficient. Catch-all controls are optimized to control non-compliant, under-specified, partial, and intangible technology transfers. Catch-all controls have been adopted and implemented in the four multilateral export control regimes (MECR) and are also reflected in legislation in the United States and the European Union. The NSG has a catch-all provision in INFCIRC/254/Part2 Paragraph 5, the AG has a catch-all condition in the Guidelines, and the MTCR has a catch-all provision in the Guidelines Paragraph 7. Korea has also implemented catch-all controls reflected in the Foreign Trade Act of 2004. However, there are still many challenges of catch-all controls. First, the problem with catch-all controls is that there are different cases in different countries for list controls, and there are too many cases in other countries to control what is not on the list. Second, there is very little international legal enforcement for failing or missing catch-all controls. Finally, information about these catch-all cases is not uniformly available. For catch-all control, Korea currently consolidates the list of traders of concern including UN sanction list and denial list of MECR, provides a correlation table of significant items and countries requiring catch-all control for stakeholders, and organizes and operates a coordination committee for traders of concerns. In the future, Korea should strengthen the monitoring of other countries’ catch-all control cases and participate in international cooperation meetings to disseminate Korea’s best practices.

Both ISO 21001 and ISO 9001 are standards developed by the International Organization for Standardization (ISO) for quality management systems. However, while ISO 9001 focuses on the general requirements for Quality Management Systems (QMS), ISO 21001 is specifically designed for educational institutions. ISO 9001 is a widely recognized standard for QMS applicable to almost industries, including manufacturing and services. It defines the requirements for establishing, implementing, maintaining, and continuing improvement of QMS to improve customer satisfaction by meeting customer requirements and improving overall performance. Meanwhile, ISO 21001 focuses specifically on educational institutions and is designed to develop and improve the curriculum efficiency by meeting trainees needs. It provides a system in which educational institutions can build, implement, maintain, and continuously improve the Education Management System (EMS) for the purpose of improving the satisfaction of trainees and other stakeholders. ISO 21001 covers a wide range of educational organizations, including schools, universities, and education providers. KINAC/INSA, the Center of Excellence in Korea, is an educational institution in the field of nuclear control. So It has been developed and operated various international and domestic curriculum. KINAC/INSA obtained ISO 9001 certification in November 2016 and has been certified so far. However, in the scope of ISO 9001 certification, curriculum development process is not included so KINAC/ INSA needs to obtain additional ISO certification specialized in education to improve the education quality. That is why KINAC/INSA is developing the ISO 21001 system, and aims to acquire certification in November 2023. This paper explains the necessity for educational institutions to obtain ISO 21001 in comparison with ISO 9001. It also introduces the process of developing ISO 21001 system of KINAC/INSA. By implementing EMS based on ISO 21001, KINAC/INSA can expect to improve the educational satisfaction of trainees and other stakeholders through effective curriculum development and educational operation.

A person who performs or plans to conduct a physical protection inspection as stipulated by the law, the act on physical protection and radiological emergency, should obtain an inspector’s ID card certified and authorized by Nuclear Safety and Security Commission Order No.137 (referred to as Order 137). In addition, according to Order 137, KINAC has been operating some training courses for those with the inspector’s ID card or intending to acquire it. Also, strenuous efforts have been put to incrementally elevate their inspection related expertise. Since Republic of Korea has to import uranium enriched less than 20% in order to manufacture fuels of nuclear reactors in domestic and abroad, the physical protection for categorization III nuclear material in transit is significantly important along with an increase in transport. The expertise of inspectors should be constantly needed to strengthen as the increase in transport leads to an increase in inspection of nuclear material in transit. We have suggested a special way to improve the inspector’s capacities through Virtual Reality technology (VR). A 3-Dimensional virtual space was designed and developed using a 3-axis simulator and VR equipment for practical training. HP’s Reverb G2 product, which was developed in collaboration with VALVE Corporation and MicroSoft, was used as VR equipment, and the 3-axis motion simulator was developed by M-line STUDIO corp. in Korea for the purpose of realizing virtual reality. The training scenarios of transport inspection consist of three parts: preparation at the shipping point, transport in route including stops and handover at the receiving point. At the departure point, scenario of the transport preparation is composed with the contents of checking the transport-related documents which should be carried by shipper and/or carrier during transport and confirming who the shipper and/or carrier is. Second, scenario is designed for inspector to experience how carrier and/or shipper protect the nuclear material during transport or stops for rests or contingency and how they communicate with each other during transport. Lastly, scenario is developed focusing on key check items during handover of responsibilities to the facility operator at the destination. Those training scenarios can be adopted to strengthen the capabilities of those with inspector’s ID card of physical protection in accordance with Order 137 and to help new inspectors acquire inspectionrelated expertise. In addition, they can be used for domestic education to promote understanding of nuclear security, or may be used for education for people overseas for the purpose of export of nuclear facilities.