Currently, KHNP has 24 operating nuclear power plant units with a toal combined capacity of about 23 GWe and two units are under construction. However, permanent stop of Kori unit 1 nuclear power plant was decided in 2017. Accordingly, interest in how to dispose of waste stored inside a permanently stopped nuclear power plant and waste generated as decommissioning process is increasing. KHNP CRI is conducting research on the advancement of plasma torch melting facilities for waste treatment generated during the plant decommissioning and operation period. Plasma torch melting facility is composed of various equipment such as a melting furnace (Melting chamber, Pyrolsis chamber), a torch, an exhaust system facility, a waste supply device, and other equipment. In demonstration test, concrete waste was put in a 200 L drum to check whether it can be pyrolyzed using a plasma torch melting facility. Reproducibility for waste treatment in the form of a 200 L drum and discharge of molten slag could be confirmed, the amount of concrete waste in 200 L Drum that could be treated according to power of plasma torch was confirmed. This demonstration test confirmed the field applicability and stability of plasma torch melting facility, and improved expectations for long-term operation.

Plasma torch melting technology can pyrolyze and melt waste with high-temperature heat (about 1,600°C) using electric arc phenomena such as lightning. Waste that may be treated in a plasma torch melting facility is injected in solid (combustible, non-combustible) and liquid form depending on facility capacity. The 200 L drum type, screw supply type, and nozzle type liquid injection device are applied to MW plasma facilities, and the push rod type and screw supply type are applied to smallcapacity plasma facilities. In consideration of the characteristics of radioactive waste generated from operating and dismantling nuclear power plants, a waste input device suitable for plasma torch facilities was developed and verified through tests. In the future, facility soundness will be confirmed through long-term performance tests, and stability will be secured through continuous improvement.

During the operation or decommission of nuclear facilities, a large amount of dry active waste and cable waste with various shape and material is generated. Most of these wastes have almost no radioactive contamination and can be disposed of by incineration, landfill, recycling, etc. under clearance regulation. For clearance of radioactive waste, it is necessary to verify the characteristics of radiological contamination and prove that it meets the criteria for clearance regulation. According to the domestic clearance regulation, if it is difficult to measure radioactivity of wastes due to their surface condition using direct or indirect measurement methods, representative samples should be collected and analyzed for radioactivity. When sampling, it is desirable to collect samples of about 1 kg that can represent waste contamination per 200 kg or per 1 m2, and the homogeneity of the samples also should be demonstrated. However, in the case of dry active wastes, it is very difficult to prove the homogeneity of the samples because of surface shapes and conditions of the wastes. In particular, considering cable waste generated during the decommission, it is hardly capable to prove the representativeness of the sample, even though the inner shell of the covering material and the copper wire are almost uncontaminated. In this study, we show the development of a treatment system that makes it easy to prove the representativeness of samples when disposing of dry active waste or cable waste generated in nuclear facilities. The treatment device is designed in such a way that it has different storage unit and cutting unit suitable for the material characteristics of each waste type (soft, hard and cable), and therefore optimizes the efficiency of the shredding or cutting process. In addition, it is expected that the work efficiency in the radioactive treatment site with a narrow area can also be improved by providing a moving part on the device.

In Korea, it is expected that the decommissioning of nuclear reactors will increase due to the license termination of reactors constructed in the 1960s to the 80s. According to the investigation of KORAD, VLLW accounts for 67.10% of decommissioning wastes and amounts to about 413,336 drums. Due to their huge amount, it is necessary to create an appropriate decommissioning waste management plan even though VLLW is disposed at the second-phase disposal facility of the Gyeongju repository. For efficient reduction in decommissioning wastes, it is required to actively use a clearance of metallic and concrete radioactive wastes. Regulations of nuclear safety and security commission notice that the radioactive waste can be reused or recycled if it meets the clearance criterion, 10 μSv·y−1 for individual dose. Therefore, it is important to develop a computational code which calculate individual doses for each scenario, and determine whether the clearance criterion is satisfied. However, in the case of metallic waste, RESRAD-RECYCLE used in dose assessment for the clearance has no longer been maintained or updated since 2005 and there is no code for recycling of concrete waste. For this reason, a dose assessment code RUCAS (Recycle-Underlying Computational dose Assessment System) has been developed by Ulsan National Institute of Science and Technology (UNIST). A point kernel method is adopted into external dose assessment model to calculate more realistic options, which are various geometries of source, and shielding effect. In the case of internal radiation, equations of internal dose from IAEA are used. This research conducts a verification of dose assessment model for recycling of metallic radioactive waste. RESRAD-RECYCLE is the comparison object and results from RESRAD-RECYCLE validation report are referenced. Targets are 14 recycling scenarios composed up to the smelting metal step of four steps, which are arising scrap metal, smelting scrap metal, and fabrication of metal product, and reusing/recycling of product. Seven isotopes, which are Ac-227, Am-241, Co-60, Cs-137, Pu-239, Sr- 90, and Zn-65, are selected for calculation. Validation results for external dose vary by isotopes, but show acceptable differences. It seems to be caused by difference in the calculation method. In the case of internal dose using same calculation formula, results are exactly matched to RESRAD-RECYCLE for all isotopes. Consequently, RUCAS can conduct functions supported by RESRAD-RECYCLE well and future work will be conducted related to domestic recycling scenarios considering public acceptance, and verification with radiation shielding codes for various geometries of source.

High-intensity focused ultrasonic (HIFU) decontamination technology to decontaminate complex metal radioactive waste was developed and verified. Ultrasonic decontamination technology is a method widely used in this field, but its energy strength is weak, so it cannot be applied to fixed contamination. The HIFU developed in this study can eliminate a wide range of fixed contamination due to the advantage of maintaining a high frequency while having hundreds of times the energy intensity compared to conventional general ultrasonic method. In addition, there is a merit in that there is no work that generates a lot of secondary wastes such as chemical decontamination method or threatens the safety of workers. In particular, high ultrasonic energy is transmitted to curved parts and inside pipes that cannot be decontaminated with blasting method, so various types of metal wastes can be treated with the HIFU method. In this study, the performance of the HIFU was verified for zirconium chips, and the radioactivity after decontamination was reduced to less than MDA in all subjects.

The disposing method of the low-intermediate-level radioactive waste, near-surface disposal facilities are generally used. This disposal method refers to a method of constructing a concrete structure on the surface of the ground, putting radioactive waste in it, and covering it with an engineered barrier to isolate human life. Among these, engineered barriers mean covering multiple layers of heterogeneous materials such as sand, clay, and gravel. Engineering barriers have the purpose of delaying the release of radioactive materials into the natural environment as much as possible, and maintaining the isolation of radioactive waste and human life for as long as possible. In this study, the design and construction method of the facility to demonstrate the performance of the engineered barrier that isolates the surface disposal facility from nature was described. In addition, the design and construction method of monitoring technology that can monitor the safety of engineered barriers by measuring information such as moisture, temperature, and slope safety in real time was also explained.

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.

During decommissioning of a nuclear power plant, a large amount of radioactive waste is produced, and it is known to cost more than 300 billion won to dispose the waste. To reduce the disposal cost, it is essential to minimize the number of radioactive waste drums, which can be achieved by detecting and removing hotspot contaminations in the radioactive waste drums. Therefore, a Compton CT system for radioactive waste monitoring is under development, which provides the images of both the internal structure of the drum and the radioactive hotspot(s) in the drum. Based on the acquired information, the activity of hotspots can be estimated. The performance of the system is affected by various geometry factors. Therefore, it is essential to determine optimal configuration by evaluating the effects of the factors on the performance of the system. In the present study, we determined the optimum value of the factors and then predicted the performance of the optimized system by using a simulator based on the Geant4 Monte Carlo simulation. For optimization, the factors were evaluated in terms of structural similarity index measure (SSIM) and measurement time. The considered factors were the activity of the CT source, source to object distance (SOD), object to detector distance (ODD), and projection angle. The simulation result showed that the activities of the CT sources were determined as 23 mCi for 137Cs and 9.6 mCi for 60Co. The optimal SOD and ODD were 180 cm and 40 cm, respectively. The optimal projection angle was evaluated as 4° since it achieves the SSIM of 0.95 faster than other projection angles. With the optimized parameters, the performance of the system was evaluated using the IAEA gamma CT standard phantom containing a hotspot of 137Cs (7.02 μCi). The Compton image was reconstructed using the back-projection algorithm, and the CT image was reconstructed using the filtered back-projection algorithm. The result showed that the location of the hotspot in the Compton image was well identified at the true position. The acquired CT image also well represented the internal structure of the phantom, and the estimated mean linear attenuation coefficient value (μ= 0.0789 cm−1) of the phantom was close to the true value (μ= 0.0752 cm−1). In addition, the hotspot activity estimated by combining the information of the Compton image and CT image was 8.06 μCi. Hence, it was found that the Compton CT system provides essential information for radioactive waste drums.

Safety evaluation of high-level radioactive waste disposal facilities including spent nuclear fuel is a very urgent and critical issue, and in order to do so, it is very important to develop a safety case that includes Feature, Event, Process (FEP) analysis, scenario development, and scenario uncertainty evaluation. In the case of Korea, the disposal of spent nuclear fuel is recognized as an unavoidable option, and in the end, Korea’s specific FEP (SFEP) development and safety evaluation according to the scenario should be conducted. Because each country’s situation and environment are different, it is necessary to develop an SFEP based on a generic FEP (International FEP). To this end, an understanding of IFEP is essential. In this study, about 1,000 major terms appearing in the OECD/NEA IFEP are classified to where each of them belongs among F, E, and P, and which FEP each word belongs to, and the correlation between the frequency of occurrence and each term is analyzed. This result will serve as a reference for the results of SFEP analysis such as POSIVA and SKB, which our research team will analyze later. In addition, each term belongs to which academic field, and the most appropriate translation for translating each term into Korean is also described.

The safety assessment of a geological disposal system is performed over a period of hundreds of thousands of years, during which the activity of radionuclides in spent nuclear fuel decreases to natural radioactivity levels. During this period, the biosphere also experiences the long-term evolution of the surface environment including climate, terrain, and ecosystem changes. These changes cause changes in the water balance, which in turn change the pathways of radionuclides in the subsurface. Therefore, it is essential to consider these long-term changes in the surface environment for a reasonable biosphere safety assessment. For this purpose, this study developed the biosphere assessment module considering the long-term evolution of the surface environment, as a sub-module of APro (Adaptive process-based total system performance assessment framework). As a preceding study, the biosphere assessment module was previously developed using COMSOL for hydraulic and radionuclide transport processes, to simulate the pathway of radionuclides traveling from the shallow aquifer to the surface water body and soil. To consider the long-term evolution of the surface environment, the previous module needed to be improved to apply different water balances as boundary conditions of the module at each snapshot, which is a sub-time period divided based on the surface evolution data. To this end, this study utilized SWAT (Soil and Water Assessment Tool) which calculates the water balance using the surface environmental data including climate, terrain, land cover, and soil type. Conceptually, SWAT calculated annual water balance considering surface environmental changes, and certain components (i.e., groundwater recharge and hydraulic head of water bodies) of water balance were transferred to COMSOL as external data to simulate the pathway of radionuclide transport and spatio-temporal variability of radionuclides. At the current stage, the biosphere computational module has been developed to correspond to its conceptual model, and we plan to further test the applicability of the module using different surface environmental data.

In recent years, the importance of the thermo-hydraulic-mechanical-chemical coupled processes is increasing in the performance assessment (PA) of the high-level radioactive waste repository. In the case of mechanical behavior, it is very important because it can affect fluid flow and radionuclide transport by changing the porosity and permeability of the medium. In particular, Excavation Damaged Zone (EDZ) should be considered essential in PA because the migration of radionuclide is affected by the enhanced hydraulic transmissivity and altered geomechanical behavior of EDZ. Furthermore, due to various thermo-hydraulic behaviors such as decay heat generated from radioactive waste, pore water pressure increase, and swelling pressure of bentonite buffer material, mechanical evolution is occurred which may change the size and physical properties of EDZ. Therefore, to solve this problem, analysis of coupled thermal-hydraulic-mechanical (THM) processes with the effect of long-term evolution of EDZ due to the mechanical behavior should be accompanied. In this study, numerical model for the long-term evolution due to mechanical behavior considering EDZ using the Adaptive Process-based total system performance analysis framework for a geological disposal system (APro) proposed by the Korea Atomic Energy Research Institute (KAERI). In the case of EDZ, the concept of Mazars’ damage evolution model was applied to simulate the behavior using the continuum model, and the change in hydraulic properties according to the degree of damage was considered. To investigate the importance of mechanical behavior in PA, the results were compared by performing numerical analysis according to the presence or absence of mechanical analysis. Finally, numerical analysis considering the mechanical evolution of EDZ was conducted using the model developed in this study to investigate the effect of EDZ.

Many countries plan to dispose of spent nuclear fuel through deep geological disposal system. In Korea, a plan is being established for the construction of a deep disposal facility to dispose of highlevel radioactive waste (or spent nuclear fuel). For construction of a deep geological repository, the NSSC (Nuclear Safety and Security Commission) stipulate that detailed technical standards for location, structure, and disposal system of deep geological repository are determined and announced by the Nuclear Safety and Security Commission Notification. Therefore, the regulatory body should carry out the process of regulatory review whether the technical standards developed by the implementer are suitable for the IAEA’s recommendations and guidelines and domestic conditions. In this process, there are many difficulties and uncertainties in terms of time and cost to independently develop safety factors in Korea by referring to the IAEA reports. So, this study intends to investigate and analyze regulatory cases for important safety factors through cases of overseas leading countries in deep geological disposal project. There are two regulatory cases intensively investigated in this study. The first is a regulatory case of regulatory bodies and external experts on the safety case, and the second is a regulatory review case in the process of site selection factor selection. In case of regulatory review of safety case, Sweden and France were selected as the representative target countries. In Sweden, safety cases such as SR-97, SR-Can, and SR-Site have been developed and there are cases of active regulatory review by regulatory agencies in the RD&D process. In France, several safety cases based on sedimentary rocks were developed and the OECD/NEA IRT (International Review Team) was inquired for review for each safety case. The site selection process is divided into a preliminary site selection stage, a site investigation stage, and a site selection and application stage. In each stage, evaluation to select a safe site is carried out using allocated siting factors of that stage. The IAEA SSG-14 report describes aspects that implementers consider in the site selection process and, with this reference, many countries are developing various siting factors and assessment methodologies in consideration of their domestic bedrock condition and geological positions. As a representative example, in Japan which is highly affected by earthquakes and igneous activities, the siting factor is classified into EF (Evaluation Factors) and FF (Favoulable Factors). So, site assessment is conducted preferentially using EF related to earthquakes and igneous activity.

Deep geologic repositories (DGR) are designed to store spent nuclear fuel and to isolate it from the biosphere for an extended period of time as long as millions of years. The long-term performance of the DGR replies on the performance of the natural geologic barriers after the end of the lifetime for the engineered barrier systems. Typically, multiple analytical and numerical models are used to analyze and ensure the safety of the repositories along both engineered and natural barrier systems. Despite the immense advancement in computing power and modeling techniques over the last few decades, a series of models and their linkage often require many simplifying assumptions in this safety assessment. The degree of the reliability and confidence of the safety analysis is thus highly dependent on the validity of those tools used. Considering the significance of the DGR performance and public attention, the highest level of quality control is necessary for the models employed in the assessment. The performance of the ultimate long-term geologic barrier is determined by the expected travel time of the radioactive species of interest, the level of their dilution or radioactivity at compliance areas, and the uncertainty associated with them. As the species of interest can be carried away from the repository location by groundwater flow, the travel time is determined by groundwater velocity along the flow path from source to biosphere while the dilution is a function of the decay and production rates as well as the diffusion and dispersion. Due to the time scale and the complexity of the physicochemical processes and geologic media involved, the models used for safety evaluation will need to become more and more comprehensive, robust, and efficient which is difficult to achieve in principle. They will also need to be transparent and flexible to satisfy the regulatory quality control requirements. This study thus attempts to develop an accessible, transparent, and extensible integrated hydrologic models (IHM) which can be widely accepted by the regulators as well as scientific community and thus suitable for current and future safety assessment of the DGR systems. The IHM can be considered as a tool and a framework at the same time when it is designed to easily accommodate additional processes and requirements for the future as it is necessary. The IHM is capable of handling the atmospheric, land surface, and subsurface processes for simultaneously analyzing the regional groundwater driving force and deep subsurface flow, and repository scale safety features, providing an ultimate basis for seamless safety assessment in the DGR program. The applicability of the IHM to the DGR safety assessment is demonstrated using simple illustrative examples.

Disposal facilities for radioactive waste shall be sited to provide isolation from the accessible biosphere. The features shall aim to provide this isolation for tens of thousands to a million years after closure. For the safety assessments of repository, the long-term natural evolution and possible events of the site, that can cause disturbances to the facility over the period of interest, should be considered. Geological development processes that the site have been experienced can contribute to understanding and descripting the present-day conditions. Moreover, knowledge of the past is necessary to predict the future evolution of the site. With regard to disposal site, understanding past geological evolution history allows to access the possibility of hazardous events of the site that can cause disturbances to the facility over the period of interest, and to verify the change in the geological environment is within the safe performance range even after the period of interest. In addition, certain parameters that change with the geological evolution can affect the hydrological and geochemical characteristics which are essential to disposal performance. There are various factors in the evolution of the geological environment, but not all are related to disposal safety. The objective of this research is to develop a geological reconstruction method considering factors that should be derived preferentially for the geological characteristics of the disposal site and the evaluation of the long-term safety. As a preliminary study on this, we investigated case studies related to geological reconstruction of overseas disposal research institutes, and reviewed which factors are suitable for the domestic granitoid distribution environment. It is expected that systematic and consistent results will be possible in the future through this methodology.

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.

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.

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.

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.

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.

Molten Salt Reactor (MSR) is one of the generation-IV advanced nuclear reactors in which hightemperature molten salt mixture is used as the primary coolant, or even the fuel itself unlike most nuclear reactors that adopt solid fuels. The MSR has received a great attention because of its excellent thermal efficiency, high power density, and structural simplicity. In particular, since the MSR uses molten salts with boiling points higher than the exit temperature of the reactor core, there is no severe accident such as a core melt-down which leads to a hydrogen explosion. In addition, it is possible to remove the residual heat through a completely passive way and when the fuel salt leaks to the outside, it solidifies at room-temperature without releasing radioactive fission products such as cesium, which make the MSR inherently safe. Both fluoride and chloride mixtures can be used as liquid fuel salts by adding actinide halides for MSRs. However, the MSRs using chloride-based salt fuels can be operated for a long time without adding nuclear fuel or online reprocessing because the actinide solubility in chloride salts is about six times higher than that in fluoride salts. Therefore, the chloride-based MSRs are more effective for the transmutation of long-lived radionuclides such as transuranic elements than the fluoride-based MSRs, which is beneficial to resolve the high radioactive spent nuclear fuel generated from light water reactors (LWRs). This paper examines liquid fuel fabrication using an improved U chlorination process for the chloride-based MSRs and presents the strategy for the management of gaseous fission products generated during the operation of MSR.