International Atomic Energy Agency defines the term “Poison” as a substance used to reduce reactivity, by virtue of its high neutron absorption cross-section, in IAEA glossary. Poison material is generally used in the reactor core, but it is also used in dry storage systems to maintain the subcriticality of spent fuel. Most neutron poison materials for dry storage systems are boron-based materials such as Al-B Carbide Cermet (e.g., Boral®), Al-B Carbide MMC (e.g., METAMIC), Borated Stainless Steel, Borated Al alloy. These materials help maintain subcriticality as a part of the basket. U.S.NRC report NUREG-2214 provides a general assessment of aging mechanisms that may impair the ability of SSCs of dry storage systems to perform their safety functions during longterm storage periods. Boron depletion is an aging mechanism of neutron poison evaluated in that report. Although that report concludes that boron depletion is not considered to be a credible aging mechanism, the report says analysis of boron depletion is needed in original design bases for providing long-term safety of DSS. Therefore, this study aimed to simulate the composition change of neutron poison material in the KORAD-21 system during cooling time considering spent fuel that can be stored. The neutron source term of spent fuel was calculated by ORIGEN-ARP. Using that source term, neutron transport calculation for counting neutrons that reach neutron poison material was carried out by MCNP®-6.2. Then, the composition change of neutron poison material by neutron-induced reaction was simulated by FISPACT-II. The boron-10 concentration change of neutron poison material was analyzed at the end. This study is expected to be the preliminary study for the aging analysis of neutron poison material about boron depletion.

Due to the saturation of spent fuel pool of nuclear power plant in Korea, temporary storage for spent fuel will be installed, and spent fuel will be stored and managed in dry cask for a considerable period of time. Since spent nuclear fuel must withstand continuous decay heat, radiation and high internal pressure of the fuel rod in the cask, behavior of spent nuclear fuel is needed to be reviewed. Spent nuclear fuel used in Pressurized Water Reactor (PWR) in Korea is stored in a wet storage currently, but it is going to store a temporary dry-storage facility on Kori site. Therefore, it is very important and meaningful to evaluate the behavior of nuclear fuel with realistic modeling. Also, domestic PWR nuclear fuel has various burn-up. In the past, the burn-up of nuclear fuel in light water reactors was low, but in order to increase power generation efficiency, the concentration of uranium was increased and the number of new fuel was increased. Therefore, a large amount of nuclear fuel with burn-up of 45,000 MWD/MTU or higher, generally called high burn-up, is also stored in the spent fuel pool (SFP). Therefore, it is necessary to evaluate by dividing three different burn-up such as, low, medium, and high burn-up. Thus, this study will review the behavior of nuclear fuel at different burn-up during the temporary storage period with FALCON (EPRI), computational code and analyze the factors affecting the integrity of nuclear fuel, including when the temporary storage is extended its additional lifetime. And this evaluation will contribute developing the spent fuel management plan in Korea.

On a global scale, the storage of spent nuclear fuel (SNF) within nuclear power plants (NPP) has become an important research topic due to limited space caused by approaching capacity saturation. SNF have e been collected over decades of NPP operation, coming up to capacity limitation. In case of Korea, every reactor except Saeul 1 and 2 has reached a SNF storage saturation rate of over 75%. One of the most studied methods for enhancing storage capacity efficiency involves increasing storage density using racks with neutron absorbers. Neutron absorbers like borated stainless steel (BSS) are utilized to manage the reactivity of densely stored SNF. However, major challenges of applying BSS are manufacturing hardness from heterogenous microstructure and mechanical property degradation from helium bubble formation. This study suggests that innovative fabrication methods of 3D printing can be good candidate for easier fabrication and better structural integrity of BSS. Directed energy deposition (DED), one of the 3D printing methods have become major candidate method for various alloys. It deposits alloy powder on base melt surface by high intensity laser, similar with welding process. Powder manufacturing is already demonstrated superior performance compared to casting in ASTM-A887, such as increased mechanical properties, owing to its well distributed chemistry of alloy. Moreover, as its original microstructural property, the formation of micro-pores through DED could lead to long-term performance improvements by capturing helium generated from the neutron absorption of boron. The potential for fabricating complex structure is also among the advantages of DED-produced neutron absorbers. Expected challenge on DED application on BSS is lack of printing condition data, because the 3D printing process have to be kept very careful variables of thermal intensity, powder flux and etc. These processes may get through much of trial & error for initial condition approaching. Nonetheless, as a recommendation of improved neutron absorber for efficient SNF pool storage, the concept of 3D printed BSS stands out as an intriguing avenue for research.

In Korea, most temporary storage facilities for spent nuclear fuel are nearing saturation. As an alternative to this, the 2nd basic plan for high-level radioactive waste management specified the operation plan of dry interim storage facility. Meanwhile, the NSSC No. 2021-19 stipulates that it is necessary to evaluate the possibility and potential effect of accident before operating interim storage facility. Therefore, this study analyzed the categories of accident scenarios that may occur in dry storage facility as part of prior research on this. We investigated the case of categorization of dry storage facility accident scenarios of IAEA, NRC, KAREI, and KINS. The IAEA presented accident scenarios that could occur in on-site dry storage facility operated with silo and cask method. NRC has classified accident scenarios in dry storage facility and estimated the probability of accidents for each. KAERI and KINS selected major accident scenarios and analyzed the processes for each, in preparation for the introduction of dry storage facility in Korea in the future. Overall, a total of 10 accident scenarios were considered, and the scenarios considered by each institution were different. Among 10 scenarios, cask drop and aircraft collision were included in the categorization of most institutions. The results of this study can be used as basic data for cataloging accidents subject to safety evaluation when introducing dry interim storage facility in Korea in the future.

This study investigated the effectiveness of various chlorinating agents in partitioning light water reactor spent fuel, with the aim of optimizing the chlorination process. Through thermodynamic equilibrium calculations, the effects of using MgCl2, NH4Cl, and Cl2 as a single chlorinating agent or applying MgCl2, NH4Cl, and Cl2 sequentially for spent fuel chlorination were evaluated Furthermore, in this study, assuming the actual process operation situation, where only a part of the semi-volatile nuclides is removed during the heat treatment process, and including the process of precipitating the molten salt from the chlorination process with K3PO4 and K2CO3 precipitants, the percentage distribution of 50 nuclides in the light water reactor spent fuel into each process stream was quantitatively calculated using the simulation function of the HSC program and tabulated for intuitive viewing. Compared to a single chlorinator, sequential chlorination more effectively separated the heat and radioactivity of the spent fuel from the uranium-dominated product solids. Specifically, the sequential application of the chlorinating agents following heat treatment led to a final solid separation characterized by 93.1% mass retention, 5.1% radioactivity, and 15.4% decay heat, relative to the original spent fuel. The findings underscore that sequential chlorination can be an effective method for spent fuel partitioning, either as a standalone approach or in combination with other partitioning processes such as pyroprocessing.

In nuclear facilities, a graded approach is applied to achieve safety effectively and efficiently. It means that the structures, systems, and components (SSCs) that are important to safety should be assured to be high quality. Accordingly, SSCs that consist of nuclear facilities should be classified with respect to their safety importance as several classes, so that the requirements of quality assurance relevant to the designing, manufacturing, testing, maintenance, etc. can be applied. Guidance for the safety classification of SSCs consisting of nuclear power plants and radioactive waste management facilities was developed by U.S.NRC and IAEA. Especially, in guidance for nuclear power plants, safety significance can be evaluated as following details. The single SSC that mitigates or/and prevents the radiological consequence or hazard was assumed to be failure or malfunction as the initiating event/accident occurred and the following radiological consequence was evaluated. Considering both the consequence and frequency of the occurrence of the initiating event/accident, the safety significance of each SSC can be evaluated. Based on the evaluated safety significance, a safety class can be assigned. The guidance for the safety classification of the spent nuclear fuel dry storage systems (DSS) was also developed in the United States (NUREG/CR-6407) and the U.S.NRC acknowledges the application of it to the safety classification of DSS in the United States. Also, worldwide including the KOREA, that guidance has been applied to several DSSs. However, the guidance does not include the methodology for classifying the safety or the evaluated safety significance of each SSC, and the classification criteria are not based on quantitative safety significance but are expressed somewhat qualitatively. Vendors of DSS may have difficulties to apply this guidance appropriately due to the different design characteristics of DSSs. Therefore, the purpose of this study is to evaluate the safety significance of representative SSCs in DSS. A framework was established to evaluate the safety significance of SSCs performing safety functions related to radiation shielding and confinement of radioactive materials. Furthermore, the framework was applied to the test case.

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 types of fuel loaded and burned in domestic nuclear power plants are WH-type and OPR/ APR-type nuclear power plants, with a total of 19 types. In the case of spent nuclear fuel released in Korea, the low combustion level of 45,000 MWD/MTU or less accounts for about 75%. In terms of fuel type, WH 17×17 and CE 16×16 fuels account for about 85% of all spent nuclear fuels. The thickness of the oxide film of the fuel cladding can make the fuel rod vulnerable during reactor operation, directly affecting the integrity of the fuel rods. so, it is a very important design factor in design. Therefore, the fuel rod design code that predicts and evaluates this has also been developed to accurately predict fuel rod corrosion. And it’s being applied to the design. In this study, the ECT probe measured by inserting it between fuel rods. The thickness of the fuel cladding oxide film was measured for spent nuclear fuel. When reloading operational nuclear fuel, the IAEA recommends an oxide film thickness of up to 100 micrometers. In this study, it was confirmed that spent nuclear fuels keeping integrity burned for 2-3 cycles were sufficiently maintained within the limit. However, the difference could be confirmed according to the characteristics of the coating material, the combustion cycle, and the use of poison rods. For the reliability of the data, symmetrical to the quadrant fuels were selected, and the fuel burned at the same period was measured. The method of selecting the target fuel can produce meaningful results.

Nuclear safety, security, and safeguards (nuclear 3S) are essential components for establishing robust nuclear environments. Nuclear safety is to protect public and environments from radioactive contamination, which can be caused in accidents. Nuclear security is to protect nuclear facilities from terrorism or sabotage, which related to physical a ttacks or insider threats. And nuclear safeguards is to protect nuclear materials from extortion by a state with a purpose of weaponizing activities. When a new nuclear facility is introduced, it is possible to save abundant amount of resources by considering nuclear 3S in an early stage (design phases). Initially, the international atomic energy agency (IAEA) recommended safeguards-by-design (SBD) approach. The concept of SBD gradually expands to nuclear 3S-by-design (3SBD). Though there are differences in purpose and target subject, each nuclear ‘S’ is closely related with others. When introducing a certain technology or equipment in order to enhance one ‘S’, another ‘S’ also get affected. The effect can be synergies or conflicts. For instance, confidential information in nuclear security is required for a safeguards activity. On the contrary, inspection equipment for safeguards can be used for security. Pyroprocessing is a technology for managing used nuclear fuels. As pyroprocessing is a backend fuel cycle technology, a sensitive nuclear technology, safeguards has taken a large portion of nuclear 3S research in an effort to achieve international credibility and nuclear transparency. As mentioned, there are both synergies and conflicts in integrating nuclear 3S. In this study, we investigate potential challenges in applying nuclear 3S integration to pyroprocessing by addressing synergies and conflicts. This approach will suggest required supplementary methods to build the reliable pyroprocessing environment.

To mark the 70th anniversary of the alliance between South Korea and the United States, President Yoon Seok-youl of South Korea and President Joseph R. Biden of the United States convened at the White House, adopting the pivotal “Washington Declaration.” This significant act paved the way for the establishment and institutionalization of the ROK-US Nuclear Consultative Group (NCG). The NCG is envisioned as a mechanism to address North Korea’s nuclear threat, striving for nuclear sharing and a nuclear defense system, thereby alleviating concerns about nuclear security. The NCG is perceived as a crucial advancement in the realm of ‘tailored extended deterrence’ on the Korean Peninsula. However, its operational scope and efficacy remain subjects of debate within South Korea. A comparative analysis with other consultative entities, such as NATO’s Nuclear Planning Group (NPG) and Extended Deterrence Strategy Consultative Group (EDSCG), raises questions about NCG’s unique contributions and potential functional overlaps. Furthermore, the establishment of the NCG represents a notable progression in the strengthened ROK-US alliance. This progression coincides with the resumption of large-scale joint nuclear security military exercises under the new administrations of both nations. Anticipated future operations within the NCG framework encompass the continual deployment of strategic assets and the execution of nuclear simulation exercises. Such actions serve not merely as a deterrent message to North Korea but also aim to instill confidence in the US’s commitment to extended deterrence among the South Korean populace. This study aims to highlight the significance and implications of the ROK-US Nuclear Consultative Group (NCG) through an exhaustive comparative analysis of existing nuclear security consultative bodies and pertinent nuclear security policies. Moreover, this research emphasizes strategies to boost the NCG’s effectiveness, the necessity for policy enhancements to foster South Korea’s nuclear security autonomy, and the importance of raising nuclear security awareness among the general public.

As remote sensing measures, satellite imagery has played an essential role in verifying nuclear activities for decades. Starting with the first artificial satellite, Sputnik 1, in 1957, thousands of satellites are currently missioning in space. Since the 2000s, the level of detail in pixels of an image (spatial resolution) has been significantly improving, thereby identifying objects less than one meter, even tens of centimetres. The more things are identifiable, the wider regions become targets for observation. With the increasing number of satellites, computer vision technology is required to explore the applicability of algorithm-based automation. This paper aims to investigate the R&D publications worldwide from the 1990s to the present, which have tried to apply algorithms to verify any clandestine nuclear activities or detect anomalies at the site. The versatile open-source publications, including the IAEA, ESARDA, US-DOE national laboratories, and universities, are extensively reviewed from the perspective of nuclear nonproliferation (or counter-proliferation). Thus, target objects for applications are essentially located in nuclearrelated sites, and the source type of satellite sensors focuses on electro-optical images with high spatial resolution. The research trend over time by groups is discussed with limitations at the time in order to contemplate the role of algorithms in the field and to present recommendations on a way forward.

Any type of nuclear arms control or disarmament agreement requires some form of verification measure. Existing nuclear arms control treaties drew upon previous agreements such as the INF treaty, START, and IAEA nuclear safeguards inspections. However, previous treaties focused on limiting specific types of nuclear weapons and their delivery vehicles or reducing the total number of nuclear weapons rather than eliminating the nuclear enterprise as a whole. A potential nuclear disarmament verification treaty or agreement will depend on the geopolitical environment of the time as well as the national policies and priorities of each signatory state. Although research on the gradual reduction and eventual elimination of nuclear weapons is still ongoing, several states have cooperated to conduct experiments, exercises, and simulations on the procedures and technologies required for nuclear disarmament verification. Three of these efforts are the LETTERPRESS simulation conducted by the Quadrilateral Nuclear Verification Partnership (QUAD), NuDiVe Exercise conducted by the International Partnership for Nuclear Disarmament Verification (IPNDV), and the Menzingen experiment organized by the UNIDIR in partnership with the Swiss Armed Forces, Spiez Laboratory, Princeton University’s Program on Science and Global Security, and the Open Nuclear Network. These contain aspects for the development of a potential nuclear disarmament verification. The LETTERPRESS exercise conducted in 2017 tested potential activities and equipment inspectors might utilize in a nuclear weapon facility. The IPNDV NuDiVe exercises conducted in 2021 and 2022 tested the activities and equipment required for the verified dismantlement of a warhead within a dismantlement facility. Finally, the Menzingen experiment conducted in 2023 tested the practical procedures for the verification of a nuclear weapon’s absence at a storage site. This paper will analyze the three cases to offer considerations on the procedures and technologies future nuclear disarmament verification might include.

This study presents a method for analyzing the surface temperatures of specific facilities, such as the 5 MWe reactor within the Yongbyon Nuclear Complex, to explore its potential utility in monitoring suspected nuclear-related activities in North Korea using thermal infrared (TIR) satellite imagery (Landsat series). TIR band data is utilized to derive surface temperatures in the specified areas, and the temperatures are analyzed on a monthly basis to examine any patterns within these regions. This research provides a pattern-of-life on temperature variation for the target areas through multiple TIR image datasets, offering additional information to analyze facilities’ operational status in remote and inaccessible regions.

Safeguards systems and measures are determined through diversion scenario analysis based on the facility design information submitted to the IAEA when a new nuclear facility is introduced. While the concept of safeguards-by-design (SBD), which considers the safeguards from the design phase for a facility operator to minimize unplanned changes or disruption to facility operations as well as for the IAEA to increase the efficiency and effectiveness in safeguards implementation, has been emphasized for more than a decade, there is no practical tool or guidance on how to apply it. In this study, we develop a diversion path analysis tool and introduce how to apply SBD using it. A diversion path analysis tool was developed based on the elements that constitute diversion and the algorithm generated based on the initial information of facility and nuclear material flow. The results of utilizing the analysis tool depending on a different level of facility information and the safeguards set-ups were compared through examples. Taking a typical light water reactor as an example, the test analyzed the automatic generation of dedicated routes, configuration of safeguards measures, and diversion path analysis. Through this, the application and limitations of the analysis tool are discussed, and ideas for utilization according to the SBD concept and necessary regulatory guidance are proposed. The results of this study are expected to be directly utilized to domestic nuclear control during the regulation process for a construction of new nuclear power systems, and furthermore, to enhance national credibility in the engagement with the IAEA for implementation of safeguards.

The development of advanced nuclear facilities is progressing rapidly around the world. Newly designed facilities have differences in structure and operation from existing nuclear facilities, so Safeguards by Design (SBD), which applies safeguards at the design stage, is important. To this end, designers should consider the safeguardability of nuclear facilities when designing the system. Safeguardability represents a measure of the ease of safeguards, and representative evaluation methodologies are Facility Safeguardability Analysis (FSA) and Safeguardability Check-List (SCL). Those two have limitations in the quantification of safeguardability. Accordingly, in this study, the Safeguardability Evaluation Method (SEM), which has clear evaluation criteria based on engineering formulas, was developed. Nuclear Material Accountancy (NMA), a key element of Safeguards, requires the Material Balance Area (MBA) of the target facility and performs Material Balance Evaluation (MBE) based on the quantitative evaluation of nuclear materials entering or leaving the MBA. In this study, about 10 factors related to NMA were developed, including MBA, Key Measurement Point (KMP), Uncertainty of a detector, Radiation signatures, and MUF (Material Unaccounted For). For example, one of the factors, MUF is used in MBA to determine diversion through analysis of unquantified nuclear materials and refers to the difference between Book Inventory and Physical Inventory, as well as errors occurring during the process in bulk facilities, errors in measurement, or intentional use of nuclear materials. This occurs in situations such as attempted diversion, and accurate MUF evaluation is essential for solid Safeguards implementation. MUF can be evaluated using the following formula (MUF=(PB+X-Y)-PE). The IAEA’s Safeguards achievement conditions (MUF < SQ) should be met. Considering this, MUF-related factors were developed as follows. (􀜵􀜧􀜯 = 1 − 􀯆􀯎􀮿 􀯌􀯊 ) In this way, about 10 factors were developed and described in the text. This factors is expected to serve as an important factor in evaluating the safeguardability of NMA, and in the future, safeguardability factors related to Containment & Surveillance (C&S) and Design Information Verification (DIV) will be additionally developed to conduct a comprehensive safeguardability evaluation of the target facility. This methodology can significantly enhance safeguardability during the design stage of nuclear facilities.

When exporting nuclear-related items, export control is required from two perspectives: the control of “Trigger List Items” as controlled by Nuclear Supplier Groups (NSG) and the control of the “Items Subject to the Agreement” as specified in bilateral Nuclear Cooperation Agreements. While Trigger List Items and Items Subject to the Agreement are largely similar, there are some items where they do not overlap. Furthermore, national law for controlling each item is different. The Trigger List Items are governed by the Foreign Trade Act, and the Items Subject to the Agreement (Internationally Controlled Items) are governed by the Nuclear Safety Act. As a result, the detailed procedures and requirements for controlling each item are quite distinct. For the Trigger List Items, export license must be obtained in accordance with the Foreign Trade Act. The details such as responsible authority, the items subject to license, license requirements and procedures, penalties are specified in the Public Notice on Import and Export of Strategic Goods. For the Items Subject to the Agreement, the process and obligations set forth in bilateral agreements and related administrative agreements are fulfilled in accordance with the Nuclear Safety Act. However, in contrast to the Trigger List Items, the details for complying with the agreements are not specified legally. Since most of the Items Subject to the Agreement are fall within the category of the Trigger List Items, the obligations in accordance with the agreements are reviewed and implemented during the export license assessment process. However, if the Items Subject to the Agreement are not are fall within the category of the Trigger List Items, there is a risk of control omission. For example, this applies to cases of exporting tritium and tritium removal facilities, which are not the Trigger List Items, to Canada and Romania. Moreover, since subjects to the agreement and compliance procedures are respectively different for 29 bilateral Nuclear Cooperation Agreements signed with different countries, it is difficult for enterprise to recognize the appropriate procedures and obligations under the agreement by their own. The bilateral Nuclear Cooperation Agreements establish legal obligations between state parties while NSG are non-legally binding arrangements. Therefore, it could be even more necessary to comply strictly with the agreements. Consequently, legal improvements are required for effective implementations of Nuclear Cooperation Agreements. While it may be challenging to institutionalize details of 29 Nuclear Cooperation Agreements, it is essential to legally specify key elements such as the list of items subject to agreements, responsible authority, requirements and procedures for implement the agreement obligations, and penalties. Furthermore, domestic awareness on compliance with Nuclear Cooperation Agreements is lower compared to the system of export license for Trigger List Items. The continuous outreach is also necessary, along with institutional improvements.

The Nuclear Export and Import Control System (NEPS) is currently in operation for nuclear export and import control. To ensure consistent and efficient control, various computational systems are either already in place or being developed. With numerous scattered systems, it becomes crucial to integrate the databases from each to maximize their utility. In order to effectively utilize these scattered computer systems, it is necessary to integrate the databases of each system and develop an associated search system that can be used for integrated databases, so we investigated and analyzed the AI language model that can be applied to the associated search system. Language Models (LM) are primarily divided into two categories: understanding and generative. Understanding Language Models aim to precisely comprehend and analyze the provided text’s meaning. They consider the text’s bidirectional context to understand its deeper implications and are used in tasks such as text classification, sentiment analysis, question answering, and named entity recognition. In contrast, Generative Language Models focus on generating new text based on the given context. They produce new textual content continuously and are beneficial for text generation, machine translation, sentence completion, and storytelling. Given that the primary purpose of our associated search system is to comprehend user sentences or queries accurately, understanding language models are deemed more suitable. Among the understanding language models, we examined BERT and its derivatives, RoBERTa and DeBERTa. BERT (Bidirectional Encoder Representations from Transformers) uses a Bidirectional Transformer Encoder to understand the sentence context and engages in pre-training by predicting ‘MASKED’ segments. RoBERTa (A Robustly Optimized BERT Pre-training Approach) enhances BERT by optimizing its training methods and data processing. Although its core architecture is similar to BERT, it incorporates improvements such as eliminating the NSP (Next Sentence Prediction) task, introducing dynamic masking techniques, and refining training data volume, methodologies, and hyperparameters. DeBERTa (Decoding-enhanced BERT with disentangled attention) introduces a disentangled attention mechanism to the BERT architecture, calculating the relative importance score between word pairs to distribute attention more effectively and improve performance. In analyzing the three models, RoBERTa and DeBERTa demonstrated superior performance compared to BERT. However, considering factors like the acquisition and processing of training data, training time, and associated costs, these superior models may require additional efforts and resources. It’s therefore crucial to select a language model by evaluating the economic implications, objectives, training strategies, performance-assessing datasets, and hardware environments. Additionally, it was noted that by fine-tuning with methods from RoBERTa or DeBERTa based on pre-trained BERT models, the training speed could be significantly improved.

The Korea Institute of Nuclear Nonproliferation and Control (KINAC) conducts various outreach activities, such as publishing brochures and holding seminars and briefings, to make regulated parties aware of the importance and necessity of the export control regime. The outreach program aims to increase compliance rates by generating interest in the export control regime among recipients and to increase communication to support compliance. In order to explore the long-term development of outreach activities, we analyze how KINAC conducts outreach. KINAC conducts nuclear export control outreach to organizations that deal with trigger list items and related technologies. Educational institutions with nuclear energy-related departments, research institutes related to nuclear energy and materials, and industrial companies that handle equipment used in nuclear power plants or nuclear materials were selected for outreach. The outreach program provides information on the export control regime for trigger list items, the strategic technology control regime, and the Nuclear Cooperation Agreement. KINAC’s outreach programs can be categorized into education, exhibition, and publication. In the education program, we hold workshops and seminars for industrial companies, with customized content that considers the items handled by companies and the nature of technology transfer. We provide training for educational and research institutions focused on conducting research tasks and projects and transferring technology accordingly. As a result of the education program, there is a regret that the education for SMEs and educational institutions is not directly linked to the implementation of nuclear export control. The exhibition program operated a booth at nuclear-related exhibitions at least once a year. The booth distributed brochures or publications on the export control regime, conducted surveys to investigate awareness of the regime and conducted on-site consultations. The exhibition program effectively increased the understanding of the export control regime among the general public and potential regulated parties. However, it was only sometimes linked to the actual implementation of nuclear export control. The publication program produced promotional materials for use at education and exhibitions, as well as guidance materials on new and revised regulations. It used the agency’s online media to provide information on new and revised export control legislation and related issues. As a result of the publication program, various existing publications explaining the export control regime were consolidated into a single publication, increasing the efficiency and satisfaction of outreach.

Emerging technologies are innovative technologies currently under development or in the early stages of introduction. These technologies have the potential to impact a wide range of industries and sectors significantly and may, therefore, be subject to export controls. The list of emerging technologies subject to export controls varies from country to country and constantly changes as new technologies are developed. For example, the U.S., EU, and South Korea have responded to these changes by adding software and technologies related to artificial intelligence and machine learning to their export control lists. Nevertheless, export control of emerging technologies still presents challenges and limitations. The rapid pace of technological advancement makes it difficult for export control regulations to keep up. For export control purposes, international cooperation on information sharing and control methods is necessary for most countries to control similar items. Several new technologies in the nuclear field may be subject to export controls. These technologies include advanced reactors, nuclear fuel cycle technologies, and nuclear waste management technologies. Small modular reactors (SMRs) and fourth-generation reactors are being developed as advanced technologies, and new technologies are being developed to improve the nuclear fuel cycle. There is also active development of technologies for space applications utilizing nuclear reactors, such as the Nuclear Thermal Propulsion System and the Nuclear Electric Propulsion System. As these technologies may include new systems and items not in existing export control, they may pose a proliferation risk or may include software design know-how for advanced materials, it is necessary to consider whether and how they should be subject to export control to prevent nuclear proliferation. Overall, export controls are an essential issue in the emerging technology and nuclear energy sectors. Countries are moving toward strengthening regulations and international cooperation to overcome these challenges and ensure safe technology transfer, and South Korea should actively participate and lead this trend.

As the demand for nuclear power increases as a means to achieve carbon neutrality, concerns about nuclear proliferation have also grown. Consequently, significant researches have conducted to enhance nuclear non-proliferation resistance. Among these research, nuclear material attractiveness is a methodology used to evaluate how appealing a particular material is for potential use in nuclear weapons, based on the characteristics of that material. Existing nuclear material attractiveness assessments focused on materials like U, Pu, and TRU, which could be directly used in the production of nuclear weapons. However, these assessments did not consider how the properties of nuclear materials change throughout the nuclear fuel cycle, with each facility process. This study assumed a scenario of the nuclear fuel cycle of graphite reduction reactors and analyzed including enrichment facilities and PUREX. This study used the FOM (Figure-Of-Merit) method developed by LANL (Los Alamos National Laboratory) for evaluating the nuclear material attractiveness. The FOM formula consists of three parameters such as critical mass, heat content, and dose The critical mass of targe materials and the dose evaluation were conducted using the Monte Carlo N-Particle code. The heat content was calculated using the ORIGEN code embedded in the Scale code. In particular, if U-238 is dominant in the facility’s materials, such as mining and refining facilities, and critical mass evaluation is unpractical. Therefore, 1SQ (Significant Quantity) of that uranium was assumed as the critical mass value for the FOM evaluation, even though 1SQ is not identical to the critical mass As a result of this study, the attractiveness of Pu produced by PUREX among all nuclear fuel cycle facilities was 2.7616, which was the most attractive to be diverted to nuclear weapons. Through this study, it was shown that the proliferation risk of the nuclear facilities in the nuclear fuel cycle and risk of diversion among those facilities.