The type of accidents associated with the operation of a melting facility for radioactive metal waste is assumed to only marginally differ from those associated with similar activities in the conventional metal casting industry or the current waste melting facility. However, the radiological consequences from a mishap or a technical failure differ widely. Three critical and at the same time possible accidents were identified: (1) activity release due to vapor explosion, (2) activity release due to ladle breakthrough, (3) consequences of failure in the hot-cell or furnace chamber not possible to remedy using remote equipment.

The remote dismantling system proposed in this paper is a system that performs the actual dismantling process using the process and program predefined in the digital manufacturing system. The key to the successful applying this remote dismantling system is how to overcome the problem of the difference between the digital mockup and the actual dismantling site. In the case of nuclear facility decommissioning, compensation between the virtual world and the real world is difficult due to harsh environments such as unsophisticated dismantling sites, radiation, and underwater, while offline programming can be proposed as a solution for other industries due to its sophisticated and controllable environment. In this paper, the problem caused by the difference in the digital mockup is overcome through three steps of acquisition of 3D point cloud in radiation and underwater environment, refraction correction, and 3D registration. The 3D point cloud is acquired with a 3D scanner originally developed in our laboratory to achieve 1 kGy of radiation resistance and water resistance. Refraction correction processes the 3D point cloud acquired underwater so that the processed 3D point cloud represents the actual position of the scanned object. 3D registration creates a transformation matrix that can transform a digital mockup of the virtual world into the actual location of a scanned object at the dismantling site. The proposed remote dismantling system is verified through various cutting experiments. In the experiments, the cutting test object has a shape similar to the reactor upper internals and is made of the same material as the reactor upper internals. The 105 successful experiments demonstrate that the proposed remote dismantling system successfully solved the key problem presented in this paper.

In this work, we introduce a 100 kW class mobile plasma melting system designed for non-combustible radioactive wastes treatment. To ensure mobility, the designed system consists of two 24-ft commercial containers, each in charge of the plasma utilities and melting process. In the container for plasma utilities, a 100 kW class DC power supply is installed together with a chiller and gas supply system whereas the container for melting process has a transferred type arc melter as well as off-gas treatment system consisting of a heat exchanger, filtrations, scrubber and NOx removal system. As a heat source for a transferred type arc melter, we adopted a hollow electrode plasma torch with reverse polarity discharge structure. Detailed design for a 100 kW class mobile plasma melting system will be presented together with the main specifications of the components. In addition, the basic performance data of the melting system is also presented and discussed.

The mechanical safety of the container designed according to the IP-2 type technology standard was analyzed for the temporary storage and transportation of Very-Low-Level-Waste (VLLW) for liquid occurring at the nuclear facilities decommissioning site. The container was designed and manufactured as a composite shielding container with the effect of storing and shielding liquid radioactive waste using High Density Polyethylene (HDPE) and eco-friendly shielding material (BaSO4) with corrosion and chemical resistance. The main material of the composite shielding container is HDPE and BaSO4, the material of the cover, cage and pallet is SUS304, and the angle guard is elastic rubber. The test and analysis requirements were analyzed for structural analysis of container drop and lamination test. As test requirements for IP-2 type transport containers should be verified by performing drop and lamination tests. There should be no loss or dispersion of contents through the 1.2 m high free-fall drop and lamination test for a load five times the amount of transported material. ABAQUS/Explicit, a commercial finite element analysis program, was used for structural analysis of the drop and lamination test of the transport and storage container. (Drop test) It was confirmed that the container was most affected when it falls from a 45-degree slope. Although plastic deformation was observed at the edge axis of the cover, it was evaluated that the range of plastic deformation was limited to the cover and cage, and stress within the elastic limit occurred in the inner container. In the analysis results for other falling direction conditions, it was evaluated that stress within the elastic limit was generated in the inner container except for minor plastic deformation. In the case of on-site simulation evaluation, deformation of the inner container and frame due to the drop impact occurred, but leakage and loss of contents, which are major evaluation indicators, did not occur. (Lamination test) The maximum stress was calculated to be 19.9 MPa under the lamination condition for a load 5 times the container weight, and the maximum stress point appeared at the corner axis of the pallet. The calculated value for the maximum stress is about 10%, assuming the conservative yield strength of SUS304 is 200 MPa. It was evaluated that stress within the limit occurred. In the case of on-site simulation evaluation, it was confirmed that there was no container deformation or loss of contents due to the load.

As the plan for the nuclear dismantlement due to the permanent shutdown of Kori-1 and Wolseong- 1 nuclear power plants has been concretized, a “movable radionuclide analysis system” is being developed that can quickly and accurately analyze large amounts of radioactive waste generated on the sites during dismantling. This system has various advantages from the perspective of strict regulations on the radioactive waste movement and social acceptability, such as preventing unexpected accidents while moving on the national highway or expressway, reducing various documents and immediate response to dismantling plans. Currently the system is being developed to be equipped with previously developed sample pretreatment and radioactivity measuring equipment and automated volatile and nonvolatile nuclide separation equipments, but to ensure mobile stability, it needs to analyze factors and establish stability standards. In the KS Q ISO/IEC 17025:2017 standard, the requirements for “facilities and environmental conditions” are a very important factor in building reliability for consumers as part of the quality guarantee for this facility. In order to meet the requirements, the technical standards of various test equipment to be installed in this facility were investigated. The physical, chemical, and radiological hazards that could affect the safety of the equipment and workers in the process of moving the equipment between nuclear power plants or between nuclear dismantling sites were derived from vibrations, rapid changes in temperature and humidity, and the spread of contamination from radioactive waste samples. Therefore, the scope of application of the law, which is the basis for securing stability during movement, was classified into two situations: movement from facility manufacturer to installation site (non-contaminated) and movement from primary to secondary use (contaminated). And in order to investigate the Nuclear Safety Act, enforcement ordinances, and radiation safety management, and to establish standards for packaging and transportation of radioactive materials, the results of transportation tests and transport details were compared and analyzed. Finally, the air suspension systems and the automatic temperature and humidity control devices were analyzed to establish standards for securing stability against the vibration and the sharp changes in the temperature and humidity, and countermeasures such as accident measures in accordance with the Enforcement Decree of the Nuclear Safety Act were also investigated.

With the development of the nuclear industry and the increase in the use of radioactive materials, the generation of radioactive waste is increasing. As the generation of radioactive waste increases, the occurrence of related safety accidents is also increasing, and it is necessary to develop a radioactive waste monitoring technology to prevent such accidents in advance and efficiently manage radioactive waste. In Information and Communication Technology (ICT), various ICT technologies such as Internet of Things (IoT), Augmented Reality (AR), and Virtual Reality (VR) that can help with the safety management of these radioactive wastes are being developed. In this study, a radioactive waste monitoring technology was developed using ICT technology, such as management of the entire cycle history of waste using Quick Response (QR) codes, and development of AR visualization technology for small packages of radioactive waste. In addition, by using IoT technology to collect desired data from sensors and store the results, after the waste drum is loaded in the waste storage, a technology was developed to track and monitor the history and movement of the waste drum from repackaging to transfer to the storage. The data required for monitoring the radioactive waste drum includes location information, whether the drum is open or closed, temperature and humidity, etc. To collect this information, a drum monitoring technology was built with a 2.4 G wireless router, an anchor constituting a virtual zone, a tag to be mounted on the drum container, and a WNT server that collects sensor data. The network tool provided by WirePas was used for network configuration, and the status of gateways and nodes can be monitored by interworking with the WNT server. The configured IoT sensor technology were tested in a waste storage environment. Four anchors were installed and linked to the network to match the virtual zone and the real storage zone, and it was confirmed whether the movement of the tag was recorded on the network while moving the tag including the IoT sensor for analyzing location information. Based on these research results, it can contribute to the safety management of radioactive waste and establishment of Waste Acceptance Criteria (WCP) by and managing the history and monitoring the waste in the entire cycle from repackaging to disposal.

The buffer material plays a role in preventing the excessive rise in temperature generated from the high-level radioactive waste by dissipating the decay heat to the rock. For this reason, the buffer material must have thermal properties to ensure the performance of the deep geological repository. This study measured the thermal conductivity of sand-bentonite according to the mixing ratio to improve the thermal properties. The compacted buffer was manufactured with a sand-bentonite mixing ratio of 6:4, 7:3, and 8:2 with 9 to 12% water content. As a result, the thermal conductivity increases as the ratio of sand increases. As a further study, it is necessary to experiment on whether sand-bentonite’s hydraulic, mechanical, and chemical performance is suitable for the stable operation of a repository.

In order to monitor the long-term condition of structures in nuclear waste disposal system and evaluate the degree of damage, it is necessary to secure quantitative monitoring, diagnosis, and prediction technology. However, at present, only simple monitoring or deterioration evaluation of the structure is being performed. Recently, there is a trend to develop monitoring systems using artificial intelligence algorithms, such as to introduce artificial intelligence-based failure diagnosis technology in nuclear power plant facilities. An artificial intelligence algorithm was applied to distinguish the noise signal and the destructive signal collected in the field. This can minimize false alarms in the monitoring system. However, it is difficult to apply artificial intelligence to industrial sites only by learning through laboratory data. Therefore, a database of noise signals and destructive signals was constructed through laboratory data, and signals effective for quantitative soundness determination of structures were separated and learned. In addition, an adaptive artificial intelligence algorithm was developed to enable additional learning and adaptive learning using field data, and its performance was verified through experiments.

Dry head end process is developing for pyro-processing at KAERI (Korea Atomic Energy Research Institute). Dry processes, which include disassembly, mechanical decladding, vol-oxidation, blending, compaction, and sintering shall be performed in advance as the head-end process of pyroprocessing. An important goal of the head-end process is the fabrication of a proper feed material for the subsequent electrolytic reduction process. In the vol-oxidation process, the pellet type-SFs are pulverized by an oxidation under an air-blowing condition, and some volatile fission products are removed from the produced powders by using an air flow. After blending, the U3O8 powders are moved to a compactor of compaction process to obtain U3O8 porous pellets. In the fine powders removal system connected with compactor, for the improved performance of oxide reduction process coupled to dry head-end process, the removal/recovery system for fine powders potentially attached to the surface of oxide reduction raw material was developed and applied to the removal of fine powders from green pellets fabricated in dry head-end process. The removal efficiency of fine powders was also verified using porous U3O8 pellets in the fine powders removal system.

For economic and safe management of Spent Nuclear Fuel (SNF), it is very important to maintain the structural integrity of SNF and to keep the fuel undamaged and handleable. The cladding surrounding nuclear fuel must be protected from physical and mechanical deterioration. The structural evaluation of SNF is very complicated and numerically demanding and it is essential to develop a simplified model for the fuel rod. In this study, a simplified model was developed using a new cladding failure criterion. The simplified model was developed considering only the horizontal or lateral static load utilizing the cladding material properties of irradiated Zirclaoy-4, and applicability in horizontal and vertical drop impacts was investigated. When a fuel rod is subject to bending, a very complicated 3D stress state is generated within the vicinity of the pellet–pellet interface. A very localized stress concentration is observed in the area where the edges of the pellets contact the cladding. If the failure strain criteria obtained from the uniaxial tension test or biaxial tube test is applied, failure is predicted at the beginning stage of loading with premature through-thickness stress or strain development. The localized contact stress or strain is self-limiting and is not a good candidate for the cladding failure criteria. In this work, a new cladding failure criterion is proposed, which can account for the localized stress concentration and the through-thickness stress development. The failure of the cladding is determined by the membrane plus bending stress generated through the thickness of the cladding, which can be calculated by a process called stress linearization along the stress classification line. The failure criterion for SNF was selected as the membrane plus bending stress through stress linearization in the cross-sections through the thickness of the cladding. Because the stress concentration in the cladding around the vicinity of the pellet–pellet interface cannot be simulated in a simplified beam model, a stress correction factor is derived through a comparison of the simplified model and detailed model. The applicability of the developed simplified model is checked through horizontal and vertical drop impact simulations. It is shown that the stress correction factor derived considering static bending loading can be effectively applied to the dynamic impact analyses in both horizontal and vertical orientations.

In this study, for thermal neutron absorption, an aluminum metal composite in which B4C particles were uniformly dispersed was prepared using stirring casting and hot rolling processes. The microstructure, thermal neutron absorption rate, mechanical properties and dispersibility of the reinforcement of the prepared B4C/Al composite were analyzed. The composite in which the 40 μm sized B4C particles were uniformly dispersed increased the tensile strength as the volume ratio of the reinforcement increased.

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.

Molten salt immersion technique has been tested with several Sr oxides, SrZrO3, SrMoO4 and U2SrOy, and MgCl2 based molten salts for the Sr nuclide separation. Reaction time, temperature, and salt composition were varied to effectively separate Sr in chloride forms. ICP-OES, XRD, and SEM analysis were conducted for the conversion efficiency and structure and morphology analysis. It is confirmed that all experiments of SrZrO3 with MgCl2 at 800°C for reaction time 5, 10, 20 hours showed higher conversion efficiency than 99% and in LiCl-KCl-MgCl2 and NaCl-MgCl2 molten salts at 500°C or 600°C, conversion efficiency higher than 97% was obtained. SrMoO4 in MgCl2 immersion experiments for 10 hours showed higher conversion efficiency than 99% when the molar ratio of salt/oxide powder is 7. U2SrOy was also tested with MgCl2 molten salt at 800°C and higher efficiency than 99% and mainly MgUO4 were produced as a reaction product.

Facing the problem of saturation of spent nuclear fuel (SNF) stored in temporary storage facilities on sites, interest in the treatment of SNF is increasing, and it is recognized as a task that needs to be solved promptly. Although direct disposal is a general method for dealing with SNF, the entire fuel assembly is classified as high-level waste; thus, the burden of disposal is high. In order to minimize the disposal burden with enhancing safety for long term storage, it is necessary to develop SNF treatment technologies and continuous efforts are required from a national policy perspective. The present study focused on minimizing the volume of high level waste from light water reactor fuel by separation of uranium, which accounts for most of SNF. The chlorination characteristics of uranium (U), rare earth (RE) oxides were confirmed through lab-scale experiments, and the possibility of uranium separation from U-RE simulated fuel was evaluated using NH4Cl chlorinating agent. The detailed results will be posted and discussed.

Prior to the investigations on fuel degradation it is necessary to describe the reference characteristics of the spent fuel. It establishes the initial condition of the reference fuel bundle at the start of dry storage. In a few technology areas, CANDU fuels have not yet developed comprehensive analysis tools anywhere near the levels in the LWR industry. This requires significantly improved computer codes for CANDU fuel design. In KNF, in-house fuel performance code was developed to predict the overall behavior of a fuel rod under normal operating conditions. It includes the analysis modules to predict temperature, pellet cracking and deformation, clad stress and strain at the mid-plane of the pellet and pellet-pellet interfaces, fission gas release and internal gas pressure. The main focus of the code is to provide information on initial conditions prior to dry storage, such as fission gas inventory and its distribution within the fuel pellet, initial volumes of storage spaces and their locations, radial profile of heat generation within the pellet, etc. Potential degradation mechanisms that may affect sheath integrity of CANDU spent fuel during dry storage are: creep rupture under internal gas pressure, sheath oxidation in air environment, stress corrosion cracking, delayed hydride cracking, and sheath splitting due to UO2 oxidation for a defective fuel. To upgrade the developed code that address all the damage mechanisms, the first step was a review of the available technical information on phenomena relevant to fuel integrity. The second step was an examination of the technical bases of all modules of the in-house code, identify and extend the ranges of all modules to required operating ranges. Further improvements being considered include upgrades of the analysis module to achieve sufficient accuracy in key output parameters. The emphasis in the near future will be on validation of the in-house code according to a rigorous and formal methodology. The developed models provide a platform for research and industrial applications, including the design of fuel behavior experiments and prediction of safe operating margins for CANDU spent fuel.

To estimate the removal efficiency of TRU and rare earth elements in an oxide spent fuel, basic dissolution experiments were performed for the reaction of rare earth elements from the prepared simfuel with chlorination reagents in LiCl-KCl molten salt. Based on the literature survey, NH4Cl, UCl3, and ZrCl4 were selected as chlorination reagent. CeO2 and Gd2O3 powders were mixed with uranium oxide as a representative material of rare earth elements. Simfuel pellets were prepared through molding and sintering processes, and mechanically pulverized to a powder form. The experiments for the reaction of the simfuel powder and chlorination reagents were carried out in a LiCl-KCl molten salt at 500°C. To observe the dissolution behavior of rare earth elements, molten salt samples were collected before and after the reactions, and concentration analysis was performed using ICP. After the reaction completed, the remaining oxide was washed with water and separated from the molten salt, and XRD was used for structural analysis. As a result of salt concentration analysis, the dissolution performance of rare earth elements was confirmed in the reaction experiments of all chlorination reagents. In an experiment using NH4Cl and ZrCl4, the uranium concentration in the molten salt was also measured. In other words, it seemed that not only rare elements but also uranium oxide, which is a main component of simfuel, was dissolved. Therefore, it is thought that the dissolution of rare earth elements is also possible due to the collapse of the uranium oxide structure of the solid powder and the reaction with the oxide of rare earth elements exposed to molten salt. As a result of analyzing the concentration changes of Simfuel before and after each reaction, there was little loss of uranium and rare earth elements (Ce/Gd) in the NH4Cl experiment, but a significant amount of rare earth elements were found to be reduced in the UCl3 experiment, and a large amount of rare earth elements were reduced in the ZrCl4 reaction.

Strontium-90 is a high heat-generating nuclide in spent nuclear fuel. The removal of the nuclide separation is indispensable to reduce the burden of storage and disposal of high-level radioactive waste. Korea Atomic Energy Research Institute has developed the molten salt immersion technique to separate the strontium by the chlorination of the strontium oxide in molten salt. It is needed to separate the salt for the recovery of strontium from the salt solution after the chlorination reaction. In this study, it was investigated on the recovery of the strontium from the salt. Vacuum distillation was used for the separation of strontium from the molten salt. The vapor pressures of the candidate salts were calculated by HSC chemistry and the apparent evaporation rates (AER) were measured at 830°C to evaluate the salts for strontium recovery. The candidate salts were LiCl, KCl, MgCl2, NaCl and CaCl2. The AERs of MgCl2 and NaCl were 1.9 and 1.3 g/cm2-h, respectively. Those two salts can be separated from the strontium compound even though the AER values are much lower than those of LiCl-KCl (~ 8 g/cm2-h). CaCl2 salt was rarely evaporated (AER < 0.03 g/cm2-h) and it is not suitable to use as a strontium recovery salt. Therefore, MgCl2, NaCl, LiCl and KCl can be regarded as candidates for a strontium recovery salt.

The evaluation of the damage ratio of spent nuclear fuel is a very important intermediate variable for dry storage risk assessment, which requires an interdisciplinary and comprehensive investigation. It is known that the pinch load applied to the cladding can leaded to Mode-3 failure and the cladding becomes more vulnerable to this failure mode with the existence of radial hydrides and other forms of mechanical defects. In this study, the failure resistance of Zircaloy-4 cladding against the pinch load is investigated using numerical simulations assuming the existence of radial hydrides. The simulation model is based on the microscopic images of cladding. A pixel-based finite element model was created by separating the Zircaloy-4 and hydride using the image segmentation method. The image segmentation method uses a morphology operation basis, which is a preprocessing method through erosion operation after image expansion to enable normal segmentation by emphasizing pixels corresponding to hydrides. The segmented images are converted into a finite element model by assigning node and element numbers together with corresponding material properties. Using the generated hydride cladding finite element model, several numerical methods are investigated to simulate crack propagation and cladding failure under pinch load. Using extended finite element (XFEM) models the initiation and propagation of a discrete crack along an arbitrary, solution-dependent path can be simulated without the requirement of remeshing. The applicability of fracture mechanical parameters such as stress intensity, J-integral was also investigated.

In accordance with the Enforcement Decree of the Act on Physical Protection and Radiological Emergency, operators of Nuclear Power Plants (NPP)s must conduct full cyber security exercise once a year and partial exercise at least once every half year. Nuclear operators need to conduct exercise on systems with high attack attractiveness in order to respond to the unauthorized removal of nuclear or other radioactive material and sabotage of nuclear facilities. Nuclear facilities identify digital assets that perform SSEP (Safety, Security, and Emergency Preparedness) functions as CDA (Critical Digital Assets), and nuclear operators select exercise target systems from the CDA list and perform the exercise. However, digital assets that have an indirect impact (providing access, support, and protection) from cyber attacks are also identified as CDAs, and these CDAs are relatively less attractive to attack. Therefore, guidelines are needed to select the exercise target system in the case of unauthorized removal of nuclear or other radioactive material and sabotage response exercise. In the case of unauthorized removal of nuclear or other radioactive material, these situations cannot occur with cyber attacks and external factors such as terrorists must be taken into consideration. Therefore, it is necessary to identify the list of CDAs that terrorists can use for cyber attacks among CDAs located in the path of stealing and transporting nuclear material and conduct intensive exercise on these CDAs. A typical example is a security system that can delay detection when terrorists attack facilities. In the case of sabotage exercise, a safety-related system that causes an initiating event by a cyber attack or failure to mitigate an accident in a DBA (Design Basis Accident) situation should be selected as an exercise target. It is difficult for sabotage to occur through a single cyber attack because a nuclear facility has several safety concepts such as redundancy, diversity. Therefore, it can be considered to select an exercise target system under the premise of not only a cyber attack but also a physical attack. In the case of NPPs, it is assumed that LOOP (Loss of Offsite Power) has occurred, and CDA relationships to accident mitigation can be selected as an exercise target. Through exercise on the CDA, which is more associated with unauthorized removal of nuclear or other radioactive material and sabotage of nuclear facilities, it is expected to review the continuity plan and check systematic response capabilities in emergencies caused by cyber attacks.

Efforts for nuclear non-proliferation have continued since the development of nuclear weapons and the conclusion of the NPT Treaty. Nuclear proliferation requires materials, facilities, and human resources to make nuclear weapons, and it takes a medium to long-term time. There are many restrictions in the current system to obtain nuclear materials and facilities, so it is often done through illegal means, black markets, or confidential transactions. Methods have been developed to evaluate the nuclear non-proliferation regime to strengthen the non-proliferation and solve the problems. The IAEA and the United States DOE initiated the proliferation resistance evaluation in 1980. The DOE conducted the assessment in three main evaluation categories: materials, technical characteristics of facilities, and institutional barriers. In another nuclear non-proliferation evaluation study, some researchers evaluated three main types: current capacity, political situation, and international situation. Detailed indicators include economic capacity, industrial capacity, nuclear capacity, leader’s intentions, political structure, competitive relations, alliances, and international norms. Most of these evaluations are based on the situation at the time of assessment at the national level. Historical examples of nuclear proliferation are rare, and verification is also challenging. The Bayesian probability is widely used when the data is small, experiments are impossible, and the causal relationship is unclear. A Bayesian network is a combination of Bayesian probability and graphics. It is used throughout the industry because it can easily derive results according to causal relationships and weights of various variables, evaluate the risk for decision-making, and obtain changed results through data updates. In particular, to evaluate the proliferation of nuclear weapons, Freeman developed the Freeman network in 2008 and the Freeman-Mella network in 2014. Freeman explained in detail only the process of deriving variables, correlations, and probabilities of factors related to factors such as motivation, intention, and resources. It isn’t easy to view as an objective result value because it does not describe the academic background for path selection, motivation list, intention, and resource variable selection. However, the research was meaningful because he first used the Bayesian network for nuclear proliferation. Although some studies have been done at the macro level, there is no case of applying it in export controls, which is the beginning of the actual spread. Also, there is no quantitative value for factors for risk assessment. There is little data, and verification of causality is difficult, so if the Bayesian network is applied to export control and applied to actual implementation, it will help make decisions such as export license or export denial.