The Korea Institute of Nuclear Nonproliferation and Control (KINAC) is developing a simulation model to estimate nuclear material production. This model is a foundational technology in interpretation and evaluation in preparation for denuclearization verification. Through this model, it is possible to estimate the amount of nuclear material that can be produced based on information on the activities of facilities related to the nuclear fuel cycle in the actual denuclearization verification stage. This model makes it possible to determine whether the declared amount of nuclear material is reliable. In addition, the reliability of the reported information can be confirmed through on-site inspection. However, there is a possibility that proliferation-related activities cannot be detected even through this inspection, and a normal state may be misdiagnosed as carrying out nuclear proliferation-related activities. Therefore, it is unreasonable to specify activities related to nuclear proliferation with only one inspection. Since each inspection method has its diagnosis rate and false diagnosis rate, measures such as repeating the same inspection method or combining different inspection methods are required to detect activities related to nuclear proliferation reliably. Therefore, a model capable of estimating the number of repetitions to obtain a reliable nuclear activity detection probability was developed by using each inspection method’s diagnosis rate and false diagnosis rate as input information through a Bayesian inference method. Through this model, it can be concluded that repetitive inspections increase the probability of detecting nuclear proliferation-related activities. This approach confirmed the possibility of repeatedly breaking away from the high-intensity inspection method that causes political and diplomatic resistance from the target country and substituting it with a more readily acceptable, low-intensity inspection method.

Given the limited terrestrial reserves of uranium (about 4.6 million tons), exploring alternative resources is essential to ensure the long-term supply and sustainability of nuclear energy. Uranium extraction from seawater (UES) is a potential solution to this issue since the amount of uranium dissolved in seawater (about 4.5 billion tons) is approximately 1000 times that of terrestrial reserves. However, the ultra-low concentration of uranium in seawater (about 3.3 ppb) makes it a challenging task to make UES economically feasible. This paper provides an overview of the current status of UES technology, which has evolved over the past seven decades. Starting from inorganic adsorbents such as hydrous titanium oxide in the 1960s, amidoxime-based fibrous adsorbents gained the most attention until the early 2010s due to their ease of deployment in actual seawater conditions and high affinity for uranium. Nowadays, research on organic adsorbents with microstructures is prevailing due to their ability to easily control surface area and compositions. In addition, this study identifies the key issues that need to be addressed to make UES technology economically viable.

Recently, more than 70 SMRs have been developed around the world due to their modularity, flexibility, and miniaturization. An innovative SMR (i-SMR) is also being developed in Korea, and operators are planning to apply for a Standard Design Approval (SDA) in 2026 after completing the standard design. Accordingly, regulatory organizations are conducting R&D on regulatory requirements and guidelines for systematic SMR standard design review by referring to IAEA and NRC cases. In terms of security, SMRs are expected to undergo many changes not only in terms of physical security through security systems, security areas, and vital equipments, but also in terms of cybersecurity through new digital technologies, remote monitoring, and automated operation. Accordingly, the IAEA Fundamental Safety Principles (SF-1) require operators to improve the safety of nuclear facilities by considering security requirements, access control requirements, and the results of operational impact assessments based on threats from the design and construction stages. Similarly, the U.S. nuclear regulatory body (NRC) has confirmed the status of security assessment and design considering design basis threats (DBTs) in the NuScale standard design review process, and the Canadian nuclear regulatory body (CNSC) has revised security regulatory guidelines and applied them to the SMR standard design review. Among these various activities related to SMR security, this paper analyzes the major changes in the cybersecurity regulatory guidelines for SMRs recently revised by the CNSC, the Canadian nuclear regulatory body. Compared to the previous guidelines, the Defensive Cybersecurity Architecture (DCSA), including external logical access control, security level and zone communication requirements, verification and validation (V&V) activities during development phases, and system & service acquisition security requirements have been added. Other changes, such as the cyber incident response program, will be analyzed and compared. Through the revised regulatory guidelines, the CNSC has divided cybersecurity levels into four (High, Moderate, Low, and Business), strictly prohibiting remote access to High and Moderate levels, and allowing remote access to Low levels only for maintenance purposes. In addition, the paper will analyze the detailed revisions, such as prohibiting access to the High level from lower levels and allowing only handshaking signals from the Low level to the Moderate level.

On April 28, 2022, a North Korean hacker (operator) recruited an active officer in exchange for virtual currency to steal military secrets and attempted to hack the battlefield network, and it was revealed that he tried to use PoisonTap during the investigation. Let’s analyze whether these events pose a threat to nuclear facilities. PoisonTap is a tool coded in the Node.js language to Raspberry Pi Zero and weaponized. When connected to a target PC with a USB or Thunderbolt port, the target PC can be occupied in about a minute. PoisonTap materials include Raspberry Pi Zero, a USB expansion port that can be connected to the target PC, a space where the code can operate (code about 12 MB), and Node.js (weaponizable code), which can be made without much difficulty. PoisonTap’s functionality allows cookies and sessions to be stolen through hijacking and allows remote access by exposing internal routers to the outside. Some of the reasons why PoisonTap occurred are that network devices connect directly to the computer without any conditions. And one of the big problems of this vulnerability is the design problem of the Internet itself, so it is difficult to block or defend technically. It is difficult to protect if it is simply a software problem because it is different from how to fix it through software code modification. According to the PoisonTap principle analysis, it connects the PoisonTap to the target PC based on the network’s characteristics (subnets of lower-priority network devices are given higher priority than gateways of higher-priority network devices). The HTML+Javascript generated while being connected becomes a backdoor that can be connected anytime. In other words, by creating a Websocket that can be connected to the web browser itself at any time, an attacker can connect to the target PC at any time. In such a threat, PoisonTap is used to break in and install a web backdoor on the target PC to make it continuously accessible and attack even if the PoisonTap is disconnected. This problem is believed to be an insider threat not only to military units but also to nuclear facilities that are closed networks. PoisonTap can be brought into major nuclear areas in cooperation with insiders with general maintenance of USB equipment. Ordinary workers often leave their laptops or leave them for a while by inserting a screen-saver password. In addition, because there is no communication with the outside, actions that do not seal USB ports and enter deep sleep mode (network connection) can be exposed as cyber threats to nuclear facilities using PoisonTap by malicious insiders.

As the use of nuclear energy has been expanded, issues in a spent nuclear fuel management are raised. Several methods have been proposed and developed to manage spent fuels safely and efficiently. One method is to reduce environmental burden in disposal of spent fuels by decreasing volume of high-level waste. A nuclides management process (NMP) is one example. Through this novel process, it is able to separate highly mobile nuclides (ex. iodine, krypton), high thermal emission nuclides (ex. strontium, barium), and optionally, uranium from spent fuels. Since the NMP is a back-end fuel cycle technology, a reliable safeguards system should be employed in the facility. As international atomic energy agency (IAEA) recommends safeguards-by-design (SBD), it is desirable to investigate an appropriate safeguards approach at a step of technology development. Process monitoring (PM) is a complemental safeguards technology for traditional safeguards technologies which based on mass balance. PM traces nuclear materials indirectly but consecutively by using process parameters such as temperature, pressure, and flow of fluid. These parameters are obtainable by installing appropriate sensors. In a respect of SBD, PM is a promising approach to achieve the safeguards goal, the timely detection of diversion of a nuclear material. However, it is necessary to classify useful process parameters from all available signals which provided from PM in order to properly utilize PM. In this study, we investigated application methods of the PM approach to NMP. NMP consists of several unit processes in series. Firstly, we inspected a principle and a feature of each unit process. Based on the results, we evaluated applicability of the PM approach to each unit process according to effectiveness in enhancing safeguardability. Several unit processes were expected that their safeguards are able to be enhanced by using certain process parameters from PM.

Long-term safe storage of spent nuclear fuel (SNF) determines sustainability of the current light water reactor (LWR) fleet. In the U.S., SNF is stored in stainless steel canister in dry cask storage system (DCSS) after spending several years in wet pool storage system while there is no DSCC in Republic of Korea. The SNF storage time in DSCC is expected to be multiple decades since no permanent geological repositories are identified in both countries. One limiting factor for extended storage of SNF in DSCC is chloride-induced stress corrosion cracking (CISCC) in the welded regions of the stainless steel canisters. The propensity for the occurrence of CISCC has warranted the development of the mitigation and repair technologies to ensure the safe and long-term storage for both present and new canister although no CISCC failure was reported yet. This study investigates cold spray deposition coatings of 304 L and 316 L stainless steels on prototypical stainless steel canisters such as sensitized flat and C-ring samples. The cold spray technology has been identified as the most promising approach by Extended Storage Collaboration Program (ESCP) driven by Electric Power Research Institute (EPRI). The talk includes microstructural characterization, adhesion strength measurement, residual stress evaluation, and corrosion behavior of the coated materials in boiling MgCl2 solution and electrochemical corrosion tests in NaCl solution. In addition, the capability of repair of cracks on the canister surface using the coating technology will be presented.

Uranium inventory in Boeun aquifer is facing the artificial reservoir that intended for supplying water to nearby cities (40-70 m apart) where, toxic radionuclides might mobile and enter the reservoir. In order to understand U mobility in the system, groundwater and fracture filling materials (FFMs) were analyzed for microbial signatures, C, O, Fe, S and U-series isotopes. The δ18O-H2O and 14C signatures suggested groundwater was originated from upland recharges dominantly and not affected by mixing with the surface water. However, the 234U/238U activity ratios (ARs) and 230Th/234U ARs in FFMs ranged from 0.93 to 1.67 and from 0.22 to 1.97, respectively, indicating that U was mobile along the fractures. In shallow FFMs, the U accumulations (~157 mg/kg) were found with Fe enrichments (~226798 mg/kg) and anomalies of δ56Fe and δ57Fe, implied U mobility in shallow depths was associated with Fe-rich environment. Also, in the shallow depths, Fe-oxidizers, Gallionella was prevailing in groundwater while Acidovorax was abundant near U ore depth. The Fe-rich environment in shallow depths was formed by pyrite dissolution, demonstrated using δ34S-SO4 and δ18O-SO4 distribution. Conclusively, the Fe-rich aquifer was capable of immobilizing the dissolved U through biotic and abiotic processes, without significant discharge into the nearby reservoir.

For Korean nuclear fuel cycle project, it is necessary to design and evaluate the integrity of spent fuel storage. For the design and evaluation of spent fuel storage, it is necessary to evaluate the properties of various materials used in spent fuel storage. The materials previously considered in the design of nuclear power plants were limited to static properties and were listed in design and manufacturing code and standards. However, for the evaluation of the storage containers in scenarios such as transportation and events, dynamic material property evaluations are required. Research on the dynamic properties of materials is generally conducted in the fields of automotive and aerospace, and most of the studies are on metal materials under sheet conditions. Since the structural materials of the storage containers for used nuclear fuel are mostly composed of thick materials, consideration should be given to property evaluation methodology and quantitative comparison. In this study, the mechanical properties of stainless steel material with canister application were evaluated according to the strain rate, and the crack resistance evaluation was also performed. It was confirmed the changes in strength and crack resistance according to the increase in strain rate and observed differences in microstructural hardening behavior.

In the event of a loss of a SNF (spent nuclear fuel) transport cask during maritime transportation, it is essential to evaluate the critical depth at which the integrity of the cask can be maintained under high water pressure. SNF transport casks are classified as Type B containers and the integrity of of the containment boundary must be maintained up to a depth of 200 meters unless the containment boundary was breached under beyond-design basis accidents. However, if an intact SNF cask is lost at a depth deeper than 200-meter, release of radioactive material may occur due to breach of containment boundary with over-pressure. In this study, we developed a code for the evaluation of the pressure limit of SNF transport cask, which can be evaluated by inputting the main dimensions and loading conditions of cask. The evaluation model was coded as a computer module for ease of use. In the previous study, models with three different fidelities were developed to ensure the reliability of the calculation and maintain sufficient flexibility to deal with various input conditions. Those three models consisted of a high-fidelity model that provided the most realistic response, a low-fidelity model with parameterized simplified geometry, and a mathematical model based on the shell theory. The maximum stress evaluation of the three models confirmed that the mathematical model provides the most conservative results than the other two models. The previous results demonstrate that mathematical models can be used in the code of computer modules. In this study, additional models of transport cask were created using parametric modeling techniques to improve the accuracy of the pressure limit assessment code for different cask and situations. The same boundary conditions and loading conditions were imposed as in the previous simplified model, and the maximum stress results considering the change in the shape of the transport container were derived and compared with the mathematical model. The comparison results showed that the mathematical model had more conservative values than the simplified model even under various input conditions. Accordingly, we applied the mathematical model to develop a transportation container pressure limit evaluation code that can be simulated in various situations such as shape change and various situations.

There is a need to develop a quantitative residual water measurement method to reduce the measurement uncertainty of the amount of residual water inside the canister after the end of vacuum drying. Therefore, a lab-scale vacuum drying apparatus was fabricated and its characteristics were evaluated by performing vacuum drying experiments based on the amount of residual water, vacuum drying experiments based on the surface area of residual water, and vacuum drying experiments based on the energy of residual water using the lab-scale vacuum drying apparatus. As a result of the vacuum drying experiments, if the surface area of water is the same, the greater the amount of water, the greater the energy of the water, so more energy is transferred to the surface of the water. Therefore, more water evaporated, and the average temperature of the remaining water was higher. The larger the surface area of the water, the more energy it takes to vaporize it, so the faster it dries and the faster the drying time. Before ice formed, energy was actively transferred by conduction heat transfer from the top, center, and bottom of the water to provide the energy needed for the water to evaporate from the surface. However, no energy was transferred from the water just before it turned into ice. When vacuum drying water, you can dry more water if you dry it slowly over a longer period of time. Therefore, by using a vacuum pump with a low flow rate, the pressure can be lowered slowly to prevent ice from freezing, thereby improving the drying quantity. It was evaluated that there was a good agreement between the energy used when water evaporated and the energy absorbed from the surroundings to within about 4%. Therefore, if the energy absorbed from the surroundings is known, it is possible to evaluate the amount of water evaporated in vacuum drying.

Currently, in the United States, Spent Nuclear Fuel (SNF) is stored at the Independent Spent Fuel Storage Installations (ISFSIs) at 73 Nuclear Power Plants (NPPs). The SNF inventory stored on-site either in pools or dry storage was 84,500 MTU in 2020. The inventory stored in on-site dry storage facilities was 39,207 MTU (46% of the total), and it is growing at a rate of approximately 3,500 MTUs per year. However, because a site for geologic repository for permanent disposal of SNF has not been constructed in the U.S., the SNF will need to be stored in dry storage facilities across the U.S. for a much longer period of time than originally planned. During this time, the dry storage facilities could experience earthquakes of a different magnitude than the one for which they were originally designed. However, there is little data on the response of SNF inside dry storage systems to seismic loads in the U.S., and the various gaps and nonlinearities between storage containers, canisters, baskets, aggregates, and fuel make it very difficult to evaluate by analytical methods. Therefore, a full-scale shake table test is being planned as an international joint research project led by Sandia National Laboratories (SNL) in the U.S. In Korea, KNF decided to participate in this seismic test through the project of SNF integrity evaluation under road and sea normal transportation conditions organized by KNF and conducted by KORAD, KAERI, and Kyung-Hee University, and has provided the KNF 17ACE7 and PLUS7 test assemblies for the tests to SNL. The test will be conducted at the LHPOST6 shake table test facility operated by University of California in San Diego (UCSD) from 2023 to 2024, with the participation of KNF, CRI, and KAERI in Korea. The test units consist of a NUHOMS 32 PTH2 canister, a mockup of a generic vertical cask, a mockup of a generic horizontal storage module, 4 surrogate fuel assemblies, and 28 dummy assemblies. The seismic inputs for the tests will consist of ground motions (acceleration time histories) representative of hard rock, soft rock, and soil sites and seismic conditions in moderately tectonically active Central and Eastern US and highly tectonically active Western US. Ground accelerations for soft rock and soil conditions will be developed taking in account soil-structure interaction. Not only is this test almost impossible to conduct independently in Korea in terms of scale, facilities and costs, but it is also considered an essential test for those of us who are preparing for dry storage of spent nuclear fuel, given the increasing social concern about earthquakes due to the recent earthquake in Turkey.

Al-B4C neutron absorbers are currently widely used to maintain the subcriticality of both wet and dry storage facilities of spent nuclear fuel (SNF), thus long-term and high-temperature material integrity of the absorbers has to be guaranteed for the expected operation periods of those facilities. Surface corrosion solely has been the main issue for the absorber performance and safety; however, the possibility of irradiation-assisted degradation has been recently suggested from an investigation on Al-B4C surveillance coupons used in a Korean spent nuclear fuel pool (SFP). Larger radiation damage than expectation was speculated to be induced from 10B(n, α)7Li reactions, which emit about a MeV α-particles and Li ions. In this study, we experimentally emulated the radiation damage accumulated in an Al-B4C neutron absorber utilizing heavy-ion accelerator. The absorber specimens were irradiated with He ions at various estimated system temperatures for a model SNF storage facility (room temperature, 150, 270, and 400°C). Through the in-situ heated ion irradiation, three exponentially increasing level of radiation damages (0.01, 0.1, and 1 dpa or displacement per atom) were achieved to compare differential gas bubble formation at near surface of the absorber, which could cause premature absorber corrosion and subsequential 10B loss in an SNF storage system. An extremely high radiation damage (10 dpa), which is unlikely achievable during a dry storage period, was also emulated through high temperature irradiation (350°C) to further test the radiation resistance of the absorber, conservatively. The irradiated specimens were characterized using HR-TEM and the average size and number density of radiation-induced He bubbles were measured from the obtained bright field (BF) TEM micrographs. Measured helium bubble sizes tend to increase with increasing system (or irradiation) temperature while decrease in their number density. Helium bubbles were found from even the lowest radiation damage specimens (0.01 dpa). Bubble coalescence was significant at grain boundaries and the irradiated specimen morphology was particularly similar with the bubble morphology observed at the interface between aluminum alloy matrix and B4C particle of the surveillance coupons. These characterized irradiated specimens will be used for the corrosion test with high-temperature humid gas to further study the irradiation-assisted degradation mechanism of the absorber in dry SNF storage system.

Pyroprocessing is a promising technique for the treatment of damaged fuel debris (corium) generated by severe nuclear accidents. The debris typically consists of (U, Zr)O2 originating from the UO2 fuel and Zr alloy-based cladding. By converting the corium to a metallic form, the principal components of the fuel can be recovered through subsequent electrorefining, allowing for long-term storage or final disposal. A study investigated the reduction of zirconium oxide compounds by Li metal as a reductant in molten LiCl salt. This research explored the feasibility of treating damaged nuclear fuel debris, which mainly consists of (U, Zr)O2. The results showed that ZrO2 was successfully reduced to Zr metal by Li metal in LiCl salt at 650C without the formation of Li2ZrO3. In particular, Zr metal was produced without the formation of Li2ZrO3 when LiCl salt containing a high concentration of Li metal was used. However, Zr metal was produced with Li2ZrO3 when LiCl salt containing both Li metal and Li2O was added. This suggests that the concentration of Li metal in the LiCl salt is an important factor in determining the formation of Li2ZrO3. The study also demonstrated that Li2ZrO3 was partially reduced to Zr metal by Li metal in LiCl salt. This finding suggests that Li metal may be effective in reducing other oxide compounds in molten LiCl salt, which could be useful in the treatment of corium. Overall, the research provides valuable insights into the feasibility of using pyroprocessing for the treatment of corium. The ability to recover and store the principal components of the fuel through electrorefining could have important implications for the long-term management of nuclear waste.

Integrity evaluation scheme for Spent Fuel (SF) dry storage has been developed under transportation failure modes. This method especially considered the degradation characteristics of Spent Fuel (SF) during dry storage such as radial and circumferential hydride content, hydride volume fraction, oxide thickness, etc. Hydride and zircaloy cladding are considered as material composite system, using correlation models related to material properties. Critical Strain Energy Density (CSED) is compared with Strain Energy Density (SED), to evaluate cladding integrity. CSED serves as material characteristics, while SED can be considered as boundary condition. To calculate the CSED of cladding in the lateral failure mode, circumferential hydride concentration is used. SED is calculated considering both the bending moment and axial load. On the other hand, in the longitudinal failure case, fuel rod temperature, internal pressure, hoop stress, radial hydride concentration is used to calculate CSED. And pinch force (contact) was considered to evaluate SED. Model validations were conducted by comparing hot cell SF test and existing validated evaluation results. To separately handle normal transportation conditions from hypothetical accident conditions, SED according to stress-strain analysis results was separated into elastic and plastic regions. As a result of applying this scheme for 14×14 SF, failure probability of normal condition was zero, which is the similar result with DOE and same with EPRI. Regarding accident condition, lateral case showed similar result, but longitudinal case showed different but reasonable result, which was due to the different analysis conditions. The proposed methodology which was indigenously developed through this study is named as K-method.

When a loss of coolant accident which causes a partial or a full drainage in the SFP would happen, Zircaloy-4 spent fuel cladding begin to react with high temperature air, and the heat generates by exothermic reaction between Zircaloy-4 cladding and surrounding air. Due to the heat, the ignition may occur in the surface of Zircaloy-4 cladding. If the Zr-fire phenomenon occurs during the accident in a SFP, the spent fuel cladding and pellets would be severely fragmented and powdered and it may bring about a massive release of radioactive source terms. Therefore, it is crucial to prevent the zirconium fire phenomenon for the spent fuel pool safety. However, a main cause to trigger the zirconium fire was not identified. In order to identify a possible mechanism of the Zr-fire phenomenon, OECD-NEA SFP Project I, II was initiated. In this paper, we reviewed the Zr-fire phenomenon which may occur in the spent fuel pool for complete loss of coolant accident scenario. The Spent Fuel Pool Project (hereinafter SFP project) is the experimental program to investigate the phenomena of spent fuel pool complete loss of coolant accident using a 17×17 PWR fuel assembly. In this section, the zirconium fire phenomenon which was observed from the SFP project is briefly investigated. This paper presented the fuel assembly temperature (i.e. zirconium alloy cladding temperature) and oxygen concentration profile of the SFP project phase-1 ignition test. At around 12.7 hour, the temperature abruptly increased and the oxygen concentration also dramatically decreased. This abrupt temperature escalation is the zirconium fire phenomenon. In order to investigate the mechanism of this zirconium fire phenomenon, behaviors of both temperature and oxygen concentration were fully compared. This paper reviewed the results of OECD-NEA SFP project experiment and then a mechanism of Zr-fire phenomenon was dscussed. It seems that the Zr-fire phenomenon might be a consequence of thermal mismatch between heat generation and dissipation. A large amount of heat might be generated by the air oxidation of Zircaloy-4 spent cladding immediately after the kinetic transition which is a breakaway phenomenon. This paper discussed the relationship between the breakaway phenomenon and the Zr-fire phenomenon in case of air oxidation of Zircaloy-4 spent cladding. This paper presents preliminary findings on the Zr-fire phenomenon from the open experiment data of the prototypic spent fuel severe accident scenario. These findings would enhance the understanding of Zircaloy-4 spent cladding air oxidation and severe accident scenario progression in a SFP.

Considering the domestic situation where all nuclear power plants are located on seaside, the interim storage site is also likely to be located on coastal site. Maritime transportation is inevitable and the its risk assessment is very important for safety. Currently, there is 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 immersion of transport cask. Previous studies show that the release rate of radionuclides contained in a submerged transport cask is significantly affected by the area of flow path generated at the breached containment boundary. Due to the robustness of a cask, the breach is the most likely generated between the lid and body of cask. CRIEPI investigated the effect of cask containment on the release rate of radioactive contents into the ocean and proposed a procedure to calculate the release rate considering the so-called barrier effect. However, the contribution of O-ring on the release rate was not considered in the work. In this study, test and analysis is performed to determine the equivalent flow path gap considering the influence of O-rings. These results will be implemented in the computational model to assess sea water flow through a breached containment boundary using CFD techniques to assess radionuclide release rates. The evaluation of release rate due to container lid gaps has been performed by CRIEPI and BAM. In CRIEPI, the gap of the flow path was calculated from the roughness of the container surface without a quantitative assessment of the severity of the accident. In this work, to evaluate the release rate as a function of lid displacement, a small containment vessel is engineered and a metal Oring of the Helicoflex HN type is installed, which is the most commonly used one in transport and storage casks. The lid of containment vessel is displaced in vertical and horizontal direction and the release rate of the vessel was quantified using the helium leak test and the pressure drop test. Through this work, the relationship between the vertical opening displacement and horizontal sliding displacement of the cask lid and the actual flow path area created is established. This will be implemented in the CFD model for flow rate calculation from a submerged transport cask in the deep sea.

Molten Salt Reactor (MSR) is one of Generation-IV nuclear reactors that uses molten salts as a fuel and coolant in liquid forms at high temperatures. The advantages of MSR, such as safety, economic feasibility, and scalability, are attributed from the fact that the molten salt fuel in a liquid state is chemically stable and has excellent thermo-physical properties. MSR combines the fuel and coolant by dissolving the actinides (U, Th, TRU, etc.) in the molten salt coolant, eliminating the possibility of a core meltdown accident due to loss of coolant (LOCA). Even if the molten salt fuel leaks, the radioactive fission products dissolved in the molten salt will solidify with the fuel salt at room temperature, preventing potential leakage to the outside. MSR was first demonstrated at ORNL starting with the Aircraft Reactor Experiment (ARE) in 1954 and was extended to the 7.4 MWth MSRE developed in 1964 and operated for 5 years. Recently, various start-ups, including TerraPower, Terrestrial Energy, Moltex Energy, and Seaborg, have been conducting research and development on various types of MSR, particularly focusing on its inherent safety and simplicity. While in the past, fluoride-based molten salt fuels were used for thermal neutron reactors, recently, a chlorine-based molten salt fuel with a relatively high solubility for actinides and advantageous for the transmutation of spent nuclear fuel and online reprocessing has been developing for fast neutron spectrum MSRs. This paper describes the development status of the process and equipment for producing highpurity UCl3, a fuel material for the chlorine-based molten salt fuel, and the development status of the gas fission product capturing technologies to remove the gaseous fission products generated during MSR operation. In addition, the results of the corrosion property evaluation of structural materials using a natural circulation molten salt loop will also be included.

The damage ratio of Spent Nuclear Fuel (SNF) is a very important intermediate variable for dry storage risk assessment which require an interdisciplinary and comprehensive investigation. It is known that the pinch load applied to the cladding can lead 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, a sensitivity analysis was performed to evaluate the importance of the damage parameters that need to be calibrated for the simulation of zircaloy-4 cladding failure using computational mechanics. The simulation model was generated from a microscopic image of the cladding with hydride. The image segmentation method was used to separate the Zircaloy-4, hydride, and hydride- Zircaloy matrix interfaces to create a pixel-based finite element model. The ring compression test (RCT) was simulated because the resistance of the cladding under pinch load can be evaluated by this test. It was assumed that the damage starts with the formation and growth of voids or small cracks in the material, which grow and combine to form larger cracks, eventually leading to the complete fracture of the material. Therefore, the ductile damage criterion was applied to all materials to simulate crack formation and propagation. The sensitivity analysis was performed based on the design of experiments using L8 orthogonal array. The effects of five factors on the fracture resistance of hydrided cladding were quantified, and they are the fracture strains describing the damage initiation in zircaloy-4 matrix, hydride, and hydride-zirconium matrix, and yield stress and Young’s modulus for hydride-zirconium matrix. Information on those parameters are hardly available in literature and experimental data which enable the estimation of those are also very rare. It is planned to build a computational model which can accurately simulate the fracture behavior of hydrided cladding by calibrating significant fracture parameters using reverse engineering. The results of this study will help to figure out those significant parameters.

Recently, the spent fuel pools withdrawn from nuclear power plants in Korea have been saturated. Therefore, specific regulations on the management of spent fuel pools, such as transportation and intermediate storage are needed. The burnup history is directly related to the management of spent nuclear fuel. This is because the decision to handle nuclear fuel may vary depending on the initial concentration of nuclear fuel, the degree to which nuclear fuel is irradiated and radioisotope nuclides are decayed, and the cooling state in the spent nuclear fuel storage tank. The purpose of this study is to determine the burnup of fuel based on the value obtained by scanning the surface of spent nuclear fuel through a neutron detector. Conversely, a database of neutron signals that scan bundles of spent nuclear fuel with an instrument with an already identified combustion history needs to be completed. First of all, the correlation between burnup history and nuclides was identified in previous studies. By setting the burnup history as the input value in the ORIGEN-ARP code, it was possible to identify the radioactive isotopes remaining in the bundle of nuclear fuel. Neutrons can finally be measured based on the amount of nuclide inventory that constitutes spent nuclear fuel. Through MCNP, the neutron detector was simulated and signals were measured to confirm how it correlates with the previously acquired burnup history database. In addition, the M (sub-critical multiplication) value, which is essential for neutron measurement, was checked to confirm the degree to which additional neutrons were generated in spent nuclear fuel in a subcritical state. The target nuclear fuel assembly was CE16×16, WH14×14, and WH17×17, which confirmed the correlation (1) between burnup, enrichment, and cooling time with the previous research topic, TNSI (Total neutron source intensity). 􀜤􀜷􁈺􀜩􀜹􀝀/􀜯􀜶􀜷􁈻 = 0.83􁈺􀜵􀯇􁈻􀬴.􀬶􀬷􀬼 ∙ 􁈺􀜫􀜧􁈻􀬴.􀬸􀬺􀬶􀬻 ∙ 􀝁􀬴.􀬴􀬴􀬼􀬷∙􀯧 􁈺1􁈻 A neutron signal will be obtained from the case according to each burnup history constituting this database. In particular, PAR=SF, a function that calculates the production amount of the fission product, was used. To confirm the computational logic of SF, it was confirmed whether a reasonable calculation was made by calculating with a nuclide spectrum.

One of the most important factors in the delivery and acceptance requirements for dry storage of spent fuel is the burnup of spent fuel. Here, burnup has a unit of MWD/MTU and is used as a measure of how much nuclear fuel is depleted in a nuclear reactor. In addition, since it is one of the most basic characteristic information for the soundness evaluation of spent nuclear fuel, it is a required item not only by regulatory agencies but also by KORAD, the acquiring agency. The burnup of spent nuclear fuel is the burnup calculated through flux mapping using signals measured from in-reactor instruments during nuclear power plant operation (hereinafter: actual burnup) and the burnup calculated using the core design code (hereinafter: design burnup). In this paper, the design burnup of spent nuclear fuel discharged from OPR100 NPPs (Nuclear Power Plants) in Korea was recalculated to confirm the reliability of the actual burnup currently managed at the nuclear power plant. Basically, since spent nuclear fuel must maintain subcriticality under wet storage or dry storage, a burnup error of about 5% is considered as a conservative approach when evaluating the criticality safety of wet storage tanks and dry storage systems. Therefore, in this paper, we tried to verify whether the difference between actual burnup and design burnup for all spent nuclear fuel released from domestic OPR100 type light water reactor nuclear power plants is within 5%. As a result of the evaluation, the largest deviation between actual burnup and design burnup was about 1,457 MWD/MTU, and when converted into a percentage, it was about 3.3%. Therefore, it was confirmed that the actual burnup managed by OPR1000 NPPs in Korea has sufficient reliability. In the future, we plan to check the reliability of the performance burnup managed in WH NPPs, and some of them will be verified through measurement.