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.

Given the domestic situation, all nuclear power plants are located at the seaside, where interim storage sites are also likely to be located and maritime transportation is considered inevitable. Currently, Korea does not have an independently developed maritime transportation risk assessment code, and no research has been conducted to evaluate the release rate of radioactive waste from a submerged transportation cask in the sea. Therefore, secure technology is necessary to assess the impact of immersion accidents and establish a regulatory framework to assess, mitigate, and prevent maritime transportation accidents causing serious radiological consequences. The flow rate through a gap in a containment boundary should be calculated to determine the accurate release rate of radionuclides. The fluid flow through the micro-scale gap can be evaluated by combining the flow inside and outside the transportation cask. In this study, detailed computational fluid dynamic and simplified models are constructed to evaluate the internal flow in a transportation cask and to capture the flow and heat transfer around the transportation cask in the sea, respectively. In the future, fluid flow through the gap will be evaluated by coupling the models developed in this study.

The decommissioning of a nuclear power plant is a project that consists of several stages, and various technologies are applied when performing various tasks at each stage. And it is essential to secure safety and economic feasibility. As the paradigm has changed due to digital transformation in various industries, digitalization is applied to the life cycle of nuclear power plant from construction, operation and decommissioning project. Element technologies are being developed for decommissioning plan establishment, process design, econtamination method, decommissioning work process, waste management, environmental monitoring and radiation dose simulation. The utilization of digital twin in the decommissioning stage is classified into three categories. ① Process Monitoring (decommissioning work procedure, work progress (plan/actual), real-time work status and etc.) ② Facility Monitoring (real-time sensing and video data monitoring, decommissioning SSCs information, work alarm and etc.) ③ Safety Monitoring (work safety, radiation exposure, fire monitoring, work risk and etc.) A system suitable for the decommissioning stage and work should be developed in consideration of the target of use, development function, and when to create data according to the purpose of the system. Simulation module according to user purpose should be provided. In addition, data-base management should be performed according to the decommissioning characteristics in consideration of the data associated with the existing operating system. The system to be developed should support the project management to comply with the domestic standards and regulations to be determined in the future. This will improve the competitiveness of domestic and foreign markets.

During the decommissioning of nuclear facilities, 3D digital model that precisely describes the work environment can expedite the accomplishment of the work. Thus, the workers’ exposure to radiation is minimized and the safety risk to the workers is reduced, while precluding inadvertent effects on the environment. However, it is common that the 3D model does not exist for legacy nuclear facilities as most of the initial design drawings are 2D drawings and even some of the 2D drawings are missing. Even in the case that all of the 2D drawings are intact, these initial design drawings need to be updated using asbuilt data because facilities get modified through years of operation. In those cases, 3D scanning can be a good option to quickly and accurately generate a structure’s actual 3D geometric information. 3D scanning is a technique used to capture the shape of an object in the form of point cloud. Point cloud is a collection of large number of points on the external surfaces of objects measured by 3D scanners. The conversion of point cloud to 3D digital model is a labor-intensive process as a human worker needs to recognize objects in the point cloud and convert the objects into 3D model, even though some of the conversion process can be automated by using commercial software packages. With the aim of full automation of scan-to-3D-model process, deep learning techniques that take point cloud as input and generate corresponding 3D model have been studies recently. This paper introduces an efficient scan simulation method. The simulator generates synthetic point cloud data used to train deep learning models for classifying reactor parts in robotic nuclear decommissioning system. The simulator is built by implementing a ray-casting mechanism using a python library called ‘Pycaster’. In order to improve the speed of simulation, multiprocessing is applied. This paper describes the ray casting simulation mechanism and compares the in-house scan simulator with an open source sensor simulation package called Blensor.

Recently, it is being carried out the project to evaluate the properties of materials harvested from nuclear reactor after the decommissioning of Kori Unit 1. However, it is not sufficient adequate machining equipment and remote machining technique to perform the projects for evaluation of materials harvested from nuclear reactor. Thus, it is required to develop the remote machining technique in hotcell to evaluate the mechanical properties of nuclear reactor materials. The machining technique should be performed inside a hotcell to evaluate mechanical properties of materials harvested from nuclear reactor and is essential to prevent radiation exposure of workers. Also, it is essential to design the apparatus and develop the machining process so that it can be operated with a manipulator and minimize contamination in hotcell. In this research, development of remote specimen machining technique in hotcell such as machining apparatus, technique and process for compact tension specimens of material harvested from nuclear reactor are described. Remote machining technique will be useful in specimen machining to evaluate changes in mechanical properties of materials harvested in high-radioactive reactor. Also, it is expected that various types of specimens can be machining by applying the developed machining technique in the future.

In 2017, Kori unit 1 nuclear power plant was permanently shut down at the end of its life. Currently, Historical Site Assessment (HSA) for MARSSIM characteristics evaluation is being conducted according to the NUREG-1575 procedure, this is conducted through comprehensive details such as radiological characteristics preliminary investigation and on-site interview. Thus, the decommissioning of nuclear power plant must consider safety and economic feasibility of structures and sites. For this purpose, the establishment of optimal work plan is required which simulations in various fields. This study aims to establish procedure that can form a basis for a rational decommissioning plan using the virtual nuclear power plant model. The mapping procedure for 3D platform implementation consisted of three steps. First, scan the inside and outside of the nuclear power plant for decommissioning structure analysis, 3D modeling is performed based on the data. After that, a platform is designed to directly measure the radiation dose rate and mapped the derived to the program. Finally, mapping the radiation dose rate for each point in 3D using the radiation dose rate calculation factor according to the time change the measured value created on the 3D mapping platform. When the mapping is completed, it is possible to manage the exposure dose of workers according to the ALARA principle through the charge of radiation dose rate over time because of visualization of the color difference to the radiation dose rate at each point. For addition, the exposure dose evaluation considering the movement route and economic feasibility can be considered using developed program. As the interest in safety accidents for workers increases, the importance of minimum radiation dose and optimal work plan for workers is becoming increasingly important. Through this mapping procedure, it will be possible to contribute to the establishment of reasonable process for dismantling nuclear power plant in the future.

In Korea, research on the development of safety case, including the safety assessment of disposal facility for the spent nuclear fuel, is being conducted for long-term management planning. The safety assessment procedure on disposal facility for the spent nuclear fuel heavily involves creating scenarios in which radioactive materials from the repository reach the human biosphere by combining Features, Events and Processes (FEP) that describe processes or events occurring around the disposal area. Meanwhile, the general guidelines provided by the IAEA or top-tier regulatory requirements addressed by each country do not mention detailed methods of ‘how to develop scenarios by combining individual FEPs’. For this reason, the overall frameworks of developing scenarios are almost similar, but their details are quite different depending on situation. Therefore, in order to follow up and clearly analyze the methods of how to develop scenarios, it is necessary to understand and compare case studies performed by each institution. In the previous companion paper entitled ‘Research Status and Trends’, the characteristics and advantages/disadvantages of representative scenario development methods were described. In this paper, which is a next series of the companion papers, we investigate and review with a focus on details of scenario development methods officially documented. In particular, we summarize some cases for the most commonly utilized methods, which are categorized as the ‘systematic method’, and this method is addressed by Process Influence Diagram (PID) and Rock Engineering System (RES). The lessons-learned and insight of these approaches can be used to develop the scenarios for enhanced Korean disposal facility for the spent nuclear fuel in the future.

In case a spent nuclear fuel transport cask is lost in the sea due to an accident during maritime transport, it is necessary to evaluate the critical depth by which the pressure resistance of the cask is maintained. A licensed type B package should maintain the integrity of containment boundary under water up to 200 m of depth. However, if the cask is damaged during accidents of severity excessing those of design basis accidents, or it is submerged in a sea deeper than 200 m, detailed analyses should be performed to evaluated the condition of the cask and possible scenarios for the release of radioactive contents contained in the cask. In this work, models to evaluate pressure resistance of an undamaged cask in the deep sea are developed and coded into a computer module. To ensure the reliability of the models and to maintain enough flexibility to account for a variety of input conditions, models in three different fidelities are utilized. A very sophisticated finite element analysis model is constructed to provide accurate response of containment boundary against external pressure. A simplified finite element model which can be easily generated with parameters derived from the dimensions and material properties of the cask. Lastly, mathematical formulas based on the shell theory are utilized to evaluate the stress and strain of cask body, lid and the bolts. The models in mathematical formula will be coded into computer model once they show good agreement with the other two model with much higher fidelity. The evaluation of the cask was largely divided into the lid, body, and bottom, bolts of the cask. It was confirmed that the internal stress of the cask was increased in accordance with the hydrostatic pressure. In particular, the lid and bottom have a circular plate shape and showed a similar deformation pattern with deflection at the center. The maximum stress occurred where the lid was in the center and the bottom was in contact with the body. Because the body was simplified and evaluated as a cylinder, only simple compression without torsion and bending was observed. The maximum stress occurred in the tangential direction from the inner side of the cylinder. The bolt connecting the lid and the body was subjected to both bending and tension at the same time, and the maximum stress was evaluated considering both tension and bending loads. In general, the results calculated by the formulas were evaluated to have higher maximum stresses than the analysis results of the simplified model. The results of the maximum stress evaluation in this study confirms that the mathematical models provide conservative results than the finite element models and can be used in the computer module.

As the saturation rate of temporary storage facilities for spent nuclear fuel increases, regulatory demands such as interim storage and permanent disposal of spent nuclear fuel are expected to begin in earnest. Considering the domestic situation where all nuclear power plants are located on the waterfront site, the interim storage site is also likely to be located on the waterfront site, and maritime transportation is one of the essential management stages. Currently, there are no independently developed maritime transportation risk assessment code in Korea, and no research has been conducted to evaluate the release of radioactive waste due to the sinking of transport container. Therefore, it is necessary to secure technology to properly reflect the domestic maritime transportation environment and to assess the impact of the sinking accident and to carry out safety regulations. To accurately calculate the releaser rate of radionuclides contained in a cask with breached containment boundary, the flow rate through the gap generated in the containment boundary should be calculated. The fluid flow through this gap which is probably in micro scale in most situations should be evaluated combining the fluid flow inside and outside the cask. In this study, a detailed computational fluid dynamics model to evaluate the internal fluid flow in the cask and a simplified model to capture the fluid flow and the heat transfer around the cask in the sea are constructed. The results for the large scale model are compared with the analytic formula for verification of heat transfer coefficient and they showed good agreements. The heat transfer coefficient thus found can be used in the detailed model to provide more realistic data than those obtained from assumed heat transfer coefficient around the surface of the cask. In the future, fluid flow through the gap between the lid and the body of the cask will be evaluated coupling the models developed in this work.

For countering nuclear proliferation, satellite imagery is being used to monitor suspicious nuclear activities in inaccessible countries or regions. Monitoring such activities involves detecting changes over time in nuclear facilities and their surroundings, and interpreting them based on prior knowledge in terms of nuclear proliferation or weaponisation. Therefore, analysts need to acquire and analyze satellite images periodically and have an understanding of nuclear fuel cycle as well as expertise in remote sensing. Meanwhile, as accessibility of satellite information has been increasing and accordingly a large amount of high-resolution satellite images is available, a lack of experts with expertise in both fields to perform satellite imagery analysis is being concerned. In this regard, the Institute of Korea Nonproliferation and Control (KINAC) has developed a prototype of semi-automatic satellite imagery analysis system that can support monitoring of potential nuclear activities to overcome the limitations of professionals and increase analysis efficiency. The system provides a satellite imagery database that can manage acquired images, and the users can load images from the database and analyze them in stages. The system includes a preprocessing module capable of resizing, correcting and matching images, a change detection module equipped with a pixel-object-based change detection algorithm for multi-temporal images, and a module that automatically generates reports with relevant information. In particular, this system continuously updates open-source information database related to potential nuclear activities and provides users with an integrated analytics platform that can support their interpretation by linking related images and textual information together. As such, the system could save time and cost in processing and interpreting satellite images by providing semi-automated analytic workflows for monitoring potential nuclear activities.

This study is about the production of radiation sources of simulated concrete and soil reference materials to verify the validity of the quality establishment and measurement of the detector (HPGe) of the radioactive soil and concrete waste classification system, which is being developed to quickly and accurately classify nuclear decommissioning waste. Specific activity of gamma nucleus among radioactive wastes is evaluated using gamma spectroscopy. At this time, in order to verify the validity and reliability of measuring equipment, it shall be a standardized substance of the same medium as nuclear decommissioning waste (chemical ingredients, particles, density, etc.) in order to correct the energy and efficiency of gamma nuclide analysis equipment. The CRM used for the detector’s energy correction used a 1 L Marinelli beaker standard correctional radiation source consisting of 10 radioactive isotopes. In order to correct efficiency, in accordance with the production and certification process of the Korea Standards and Research Institute, it has produced artificial simulated radioactive concrete similar to nuclear decommissioning waste (30% for cement, 60% for regulation and 10% for bentonite). The radioactive homogeneity of the simulated concrete reference materials was evaluated using dispersion analysis (ANOVA) in accordance with ISO Guide 35, while 137Cs and 60Co of concrete reference materials were able to obtain homogeneous measurements both in and between bottles. The self-absorption rate of the simulated concrete reference material was determined by the MCNP computer simulation measurement method, and the self-absorption correction coefficients of 137Cs and 60Co were assessed at 0.995 and 0.996, respectively, and the standard value for the radiation of the simulated concrete reference material was calculated on the weighted average of the measurements of 20 samples. The uncertainty about the reference value was calculated by combining measurement uncertainty (Type B evaluation), bottle to bottle standard deviation, and uncertainty within bottle by modifying the formula suggested in ISO Guide 35. The concentration of 137Cs and 60Co of reference materials was divided into high-speed measurement mode and precision measurement mode in consideration of the self-disposal standard. The reference value and uncertainty of expansion among reference materials for high-speed measurement mode were rated at 1,032.7 ± 64.0 Bq·kg−1and 1,083.7 Bq·kg−1, respectively. The standard value and expansion uncertainty for 137Cs and 60Co among reference materials for precision measurement mode were rated at 113.7 ± 10.0 Bq·kg−1 and 122.3 ± 10.3 Bq·kg−1, respectively.

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

As Kori-1 permanently shut down in Korea, it is expected that a large amount of radioactive waste will be generated during decommissioning of nuclear power plants. Radioactive concrete waste is contaminated up to depth of 100 mm with radionuclides such as 137Cs and 60Co. The radioactive waste should be accurately classified to reduce the cost of disposing of radioactive waste. Therefore, the specific radioactivity of waste must be precisely evaluated by gamma-ray measurements emitted from the radionuclides. In general, the effectiveness of the radioactivity measurement and process is confirmed using certified reference material (CRM) composed of water or agar. However, the decommissioning waste differs from this CRM in apparent density and chemical elements, so the specific radioactivity is underestimated or overestimated. Therefore, reference material composed of the same apparent density and chemical elements as the sample is required to improve the quality of radioactivity measurement. The purpose of this study is to develop a concrete reference material for the nuclear decommissioning waste. The concrete reference material composed of SiO2, CaO, and Al2O3 were manufactured in compliance with ISO Guide 35. 10 bottles were randomly selected for homogeneity test, and 2 samples for analysis were taken from each bottle. The specific radioactivity was measured using an HPGe detector with an efficiency of 30%. The results of the homogeneity test of 137Cs and 60Co satisfied the requirements of ISO Guide 35. Coincidence summing and selfabsorption effects were corrected using the Monte Carlo efficiency transfer code and Monte Carlo NParticle transport code. The reference values of 137Cs and 60Co in the concrete reference material were evaluated in the range of 1,000–1,100 Bq·kg−1 and extended uncertainty was around 7%.

Today, the domestic and international nuclear power industry is experiencing an acceleration in the scale of the nuclear facility decommissioning market. This phenomenon is also due to policy changes in some countries, but the main reason is the rapid increase in the proportion of old nuclear power plants in the world, mainly in countries that introduced nuclear power plants in the early stages. Decontamination is essential in the process of decommissioning nuclear facilities. Among various decontamination targets, radionuclides are adsorbed between pores in the soil, making physical decontamination quite difficult. Therefore, various chemical decontamination technologies are used for contaminated soil decontamination, and the current decontamination technologies have a problem of generating a large amount of secondary wastes. In this study, soil decontamination technology using supercritical carbon dioxide is proposed and aimed to make it into a process. This technology applies cleaning technology using supercritical fluids to decontamination of radioactive waste, it has important technical characteristics that do not fundamentally generate secondary wastes during radioactive waste treatment. Supercritical carbon dioxide is harmless and is a very useful fluid with advantages such as high dissolution, high diffusion coefficient, and low surface tension. However, since carbon dioxide, a non-polar material, shows limitations in removing polar and ionic metal wastes, a chelating ligand was introduced as an additive. In this study, a ligand material that can be dissolved in supercritical carbon dioxide and has high binding ability with polar metal ions was selected. In addition, in order to increase the decontamination efficiency, an experiment was conducted by adding an auxiliary ligand material and ultrasonic waves as additives. In this study, the possibility of liquefaction of chelating ligands and auxiliary ligands was tested for process continuity and efficiency, and the decontamination efficiency was compared by applying it to the actual soil classified according to the particle size. The decontamination efficiency was derived by measuring the concentration of target nuclides in the soil before and after decontamination through ICP-MS. As a result of the experiment, it was confirmed that the liquefaction of the additive had a positive effect on the decontamination efficiency, and a difference in the decontamination efficiency was confirmed according to the actual particle size of the soil. Through this study, it is expected that economic value can be created in addition to the social value of the technology by ensuring the continuity of the decontamination process using supercritical carbon dioxide.

An accumulation of spent nuclear fuel (SNF) has brought a considerable interest due to its energy and environmental issue. To effectively manage SNF, a pyroprocessing is introduced to separate useful resources from the spent fuels and to manufacture suitable fuels. In head-end process of pyroprocessing, spent fuels are thermally treated to prepare UO2 pellets, where various radioactive gases from SNFs are released during thermal treatment. Within these gases, C-14 as CO2 form is a radioactive fission product which had a long half-life of 5,730 years and emits beta radiation of 0.156 MeV. Generally, current CO2 capturing technologies include adsorption by solid materials, absorption by aqueous solutions, and membrane separation. Among these methods, absorption is an effective approach which traps CO2 effectively and and it is easy to operate at room temperature. In addition, it is highly recommended as immobilizing 14CO2 as CaCO3 formation due to the high thermal and chemical stability, and the relatively low solubility in water. Generally, a double alkali method has been proposed to capture low concentrated 14CO2 from the stream. This method for CO2 capture includes absorption process with NaOH solution and causticization using Ca(OH)2. In this study, CO2 emitted from SNF is captured using double alkali method, and the effects of operating conditions on capturing efficiency were investigated. Furthermore, considering the two-film theory, the effects of trapping conditions on the CO2 absorption performance were examined. The recovered CaCO3 from causticization was collected from the absorbing solution and analyzed.

The amount of temporarily stored spent nuclear fuel in South Korea will be reaching saturation in a near future. Therefore, it is an urgent issue to construct a spent nuclear fuel storage system. In order to construct the storage system, some coastal environmental characteristics such as temperature, pH, and chemical composition of sea water in South Korea have to be evaluated and predicted because they can affect in deterioration of the storage system. However, in South Korea, the coastal environmental characteristics of area where the storage system is likely to be built are not well established until now. In this study, a time-series deep-learning algorithm is developed using the Long-Short Term Memory (LSTM) algorithm to predict and evaluate the coastal environmental characteristics based on the wellestablished data from Korea Meteorological Administration (KMA) and Ministry of Oceans and Fisheries (MOF). As a result, by developing the predictive model to evaluate the coastal environmental characteristics, we intend to apply it for site evaluation to construct the spent nuclear fuel storage system or many other applications related to the nuclear as well.

In decommissioning a nuclear power plant, numerous concrete structures need to be demolished and decontaminated. Although concrete decontamination technologies have been developed globally, concrete cutting remains problematic due to the secondary waste production and dispersion risk from concrete scabbling. To minimize workers’ radiation exposure and secondary waste in dismantling and decontaminating concrete structures, the following conceptual designs were developed. A micro-blast type scabbling technology using explosive materials and a multi-dimensional contamination measurement and artificial intelligence (AI) mapping technology capable of identifying the contamination status of concrete surfaces. Trials revealed that this technology has several merits, including nuclide identification of more than 5 nuclides, radioactivity measurement capability of 0.1–107 Bq·g−1, 1.5 kg robot weight for easy handling, 10 cm robot self-running capability, 100% detonator performance, decontamination factor (DF) of 100 and 8,000 cm2·hr−1 decontamination speed, better than that of TWI (7,500 cm2·hr−1). Hence, the micro-blast type scabbling technology is a suitable method for concrete decontamination. As the Korean explosives industry is well developed and robot and mapping systems are supported by government research and development, this scabbling technology can efficiently aid the Korean decommissioning industry.

본 연구의 목적은 북한의 전술핵무기 개발 경과와 한국에 미치는 함의를 규명하는 것이다. 북한은 2021년 1월 8차 당대회를 통해 핵능력 고도화와 전술핵무기 개발을 공표하였다. 그리고 지속적으로 핵능력과 투발수단을 발전시키고 있다. 핵능력에서 가장 기본적이고 중요한 것은 무기급 플루토늄(WGPu)과 고농축우라늄(HEU), 삼중수소 등 핵물질과 다양한 투발수단이다. 북한은 원자로 가동 후 재처리를 통해 획득할 수 있는 WGPu보다는 은밀한 지하시설에서 농축이 가능한 HEU 위주로 핵물질을 증산하고 있고, 삼중수소의 생산을 위해 영변 핵시설을 이용 할 수 밖에 없을 것으로 판단된다. 그리고 투발수단의 능력향상을 위해 KN-23, KN-24, KN-25와 이동식발사대(TEL), 열차이동발사 등 다양 한 투발수단의 실험을 지속하면서 전술핵무기능력을 고도화하고 있다. 북한의 핵개발 단계는 사실상의 핵보유국을 넘어서 실제 핵을 운용할 수 있는 핵전력 작전운용과 현대화 단계로 진입한 것으로 판단된다. 북한은 이를 통해 공세적 핵태세를 추구하고, 고강도 국지도발 감행 등 현상변경을 위한 강압으로 확증보복전략뿐만 아니라 한반도에서 적극적 전쟁수행전략을 구사할 것으로 예상된다. 이에 대한 새로운 대응전략이 요구되고 있다.

본 연구의 목적은 미국의 확장억제 실행력과 신뢰성 제고를 위한 미국의 저위력 핵무기 개발 및 함의를 규명하는 것이다. 사실상의 핵보유국으로 부상한 북한의 핵위협에 대한 실질적인 대책마련이 시급하다. 이에 북한의 핵능력이 지속적으로 고도화되는 상황에서 한국의 북핵 대응전략은 북핵 위협 억제‧대응을 위한 ‘핵‧WMD 대응체계’ 구축을 위한 핵심전력을 조기에 확보하기 위해 노력하고 있다. 하지만 재래식 전력으로 북핵을 대응하는 것에도 한계가 있고, 핵금기에 따른 확장억제의 신뢰성도 의심받고 있는 실정이다. 이러한 미국의 확장억제 실행력과 신뢰성 제고 측면에서 미국이 개발하고 있는 저위력 핵무기 개발과 함의를 규명하고, 한국의 억제‧대응능력 강화를 위한 방안을 적극적으로 모색할 필요가 있다. 이를 위해 한‧미는 맞춤형 억제전략 발전과 미국의 확장억제 공약의 실행력을 제고해야 한다. 한‧미 연합 억제‧대응 능력을 강화하기 위해 나토의 핵공유와 같이 핵동맹으로 격상시키고, 확장억제전략협의체(EDSCG)보다 긴밀한 고위급 핵계획그룹(NPG) 발전 등이 있어야 할 것이다.