The main purpose of the Bilateral Nuclear Cooperation Agreement is to obtain the prior consent of suppliers in the case of peaceful use of items covered by the agreement, application of IAEA safeguards, reprocessing, enrichment or transfer to a third country. Reports on inventory changes and status for mutually transferred obligated items should be exchanged annually. According to the Agreement, items subject to bilateral agreement information must be exchanged with each other prior to direct or indirect transfer of controlled items. And the importing country proceeds with prior confirmation. After that, upon receipt of the target item, shipment notification and shipment confirmation are made, and an annual report on the target item is made. Such as the Korea Atomic Energy Research Institute (KAERI), annual reporting and management of obligated items are made centered on institutions that use a lot of nuclear materials. But there are cases of delays in the agreement work due to the implementation, and discrepancies in data are occurring in the process of checking inventory details of obligated items. In addition, it was difficult to check the inventory of items subject to the agreement and the status of Export and Import status online, making it impossible for managers to monitor all aspects of bilateral agreements. Currently, there is generated to inconsistent in information between the annual report and the international transfer report in terms of Export and Import control. To solve these problems, KAERI is aim of promoting transparency in the international nuclear power sector and enhancing national reliability. And It is planning to establish an Export and Import management system for items subject to bilateral. In order to ensure the accuracy, it is going to enhance the efficiency of management methods such as new registration for new institutions when exporting and Importing items. This has the ultimate purpose of improving the efficiency of the implementation of the agreement items through the systemization of the database of agreement items and the management of the implementation of the agreement based on the sincere and timely implementation of the agreement.

The detector response was simulated to design a fork detection system for verifying the characteristics of spent fuel. The fork detection system currently used consists of two fission chamber and an ion chamber, and it is nuclear safeguard equipment that measures the gross neutrons and gross gamma rays emitted from the spent fuel assembly to identify the characteristics of the spent fuel and verify the authenticity of the operation history. In order to improve the current fork detection system, we are developing a system that applies CZT, a room temperature semiconductor detector, and a stilbene detector, which is an organic scintillator. Depletion calculations were performed using the ORIGEN code to determine the radiological characteristics emitted from spent nuclear fuel assembly. The flux of radiation emitted from the spent nuclear fuel assembly was calculated by changing the conditions such as initial enrichment, burnup, and cooling time, which are major variables of spent fuel assembly. The calculated result is used as the source term of the particle transport code. Considering the general operating conditions of the pressurized light water reactor, the conditions were changed in the range of 3-5% for initial enrichment and 30-72 GWD/MTU for burnup, and the cooling time was given within 10 years. MCNP 6.2, a Monte Carlo simulation code, was used to simulate the detector response to radiation emitted from spent nuclear fuel assembly. According to the shape, size, and position of the CZT detector, the gamma counts incident on the detector were calculated and derived the initial design of our fork detection system.

The nuclide management technology for separating high-heat generating/high-mobility/long-lived nuclides from high-level wastes based on the chemical reactions is under development. In order to secure the reliability of nuclear non-proliferation and to implement the effective safeguards, it is necessary to consider the safeguards from the conceptual design phase of the novel technologies. However, there was no experience and research on safeguards for the chemical reaction based nuclide management technology. In order to development the available monitoring techniques for the safeguards of nuclide management technology, the possible diversion scenarios were developed and the material flows of major nuclear materials were analyzed according to the various diversion strategies for each unit process in this study. The diversion strategies in this study is limited to the diversion of nuclear materials according to the change of operational parameters (temperature, chemical reagents, pressures, etc). The nuclear material distribution behaviors under the abnormal conditions were analyzed and compared with normal conditions using the HSC Chemistry. The results will be used to determine the proper signals and feasible techniques to monitor the abnormal operations.

In addition to Korea, various countries such as the United States, the United Kingdom, France, and China are designing small module-type reactors. In particular, a small modular reactor is the power of 300 MWe or less, in which the main equipment constituting the nuclear reactor is integrated into a single container. Depending on the purpose, small modular reactors are being developed to help daily life such as power, heating supply, and seawater desalination, or for power supply such as icebreakers, nuclear submarines, and spacecraft propellants. Small modular reactors are classified according to form. It can be classified into light-water reactors/ pressurized light-water reactors based on technology proven in commercial reactors, and non-lightwater reactors based on fuel and coolant type such as Sodium-cooled Fast Reactor, High temperature gas-cooled reactor, Very high temperature reactor and Moltenn salt reactor. SMRs, which are designed for various purposes, have the biggest difference from commercial nuclear reactors. The size of SMRs is as small as 1/5 of that of the commercial reactors. Several modules may be installed to generate the same power as commercial reactors. Because of the individually operation for each module, load follow is possible. Also, The reactor can be cooled by natural convection because the size is small enough. It is manufactured as a module, the construction period can be reduced. Depending on the characteristics of these SMRs, application for safeguards is considered. There are many things to consider in terms of safeguards. Therefore, it is IAEA inspection or other approaches for SMRs installed and remotely operated in isolated areas, data integrity for remote monitoring equipment to prevent the diversion of nuclear materials, verification method and material accountancy and control for new fuel types and reactors. Since SMR is more compact and technical intensive, safeguards should be considered at the design stage so that safeguards can be efficiently and effectively implemented, which is called the Safeguards by design (SBD) in the IAEA. In this paper, according to the characteristics of SMR, we will analyze the advantages/disadvantages from the point of view of safeguards and explain what should be considered.

Material balance evaluation is an important measure to determine whether or not nuclear material is diverted. A prototype code to evaluate material balance has been developed for uranium fuel fabrication facility. However, it is difficult to analyze the code’s functionality and performance because the utilization of real facility data related to material balance evaluation is very limited. It is also restricted to deliberately implement various abnormal situations based on real facility data, such as nuclear diversion condition. In this study, process flow simulator of uranium fuel fabrication facility has been developed to produce various process data required for material balance evaluation. The process flow simulator was developed on the basis of the Simulink-SimEvents framework of the MathWorks. This framework is suitable for batch-based process modeling like uranium fuel fabrication facility. It dynamically simulates the movement of nuclear material according to the time function and provides process data such as nuclear material amount at inputs, outputs, and inventories required for Material Unaccounted For (MUF) and MUF uncertainty calculation. The process flow simulator code provides these data to the material balance evaluation code. And then the material balance evaluation code calculates MUF and MUF uncertainty to evaluate whether or not nuclear material is diverted. The process flow simulator code can simulate the movement of nuclear material for any abnormal situation which is difficult to implement with real process data. This code is expected to contribute to checking and improving the functionality and performance of the prototype code of material balance evaluation by simulating process data for various operation scenarios.

In response to the increase in international terrorism threats and demands for terrorism prevention and response activities, the Act on Counter-Terrorism for the Protection of Citizens and Public Security was enacted in 2016, and the need for research to strengthen protection against explosive threats was raised. In the same manner, the Design Basis Threats, which become the standard for the design and evaluation of physical protection systems for nuclear facilities, have been developed and it includes explosive threats. However, the regulatory standards for physical barriers against explosive threats are still not established. Therefore, it is first required to set up a performance database of physical barriers subject to blast loading in order to prepare the regulatory standards. In this study, the pressure with the trinitrotoluene (TNT) charge weights of 0.5-2 kg as a function of time was calculated using Ansys Autodyn software by assuming that the TNT is used for malicious purposes and is attached to a reinforced concrete (RC) corridor wall. The shape of the corridor was the 3×3×6 m cuboid with a rectangular hole of 1.78×1.78×6 m. The RC walls, which make up the corridor, contained the reinforcing bars with a spacing of 0.229 m and a diameter of 0.036 m. The spherical charge of a TNT was placed 0.2 m away from a RC wall in the middle of the corridor. To measure the reflected pressure after the internal explosion with a TNT, three pressure gauges were installed on the three sides of the RC walls in the middle of the corridor, respectively. The results showed that the peak reflected pressure on a RC wall with the standoff distance of 0.2 m was about ten times higher than the opposite RC wall with the standoff distance of 1.58 m in the same condition of TNT charge weight. Thus, it was verified that blast loads are highly affected by standoff distance. It seems that preventing the explosive detonation close to a physical barrier is strategically important to maintain the integrity of the physical barrier.

According to Article 3(2) of the NPT and NSG the Guidelines, the exporting country should be guaranteed the import country’s willingness to implement nuclear non-proliferation and the level of implementation before the transfer of Trigger List Items. Also, unknown or new end-users could be officially identified through GA procedure. Accordingly, Korea government requests the importing country a formal Governmental Assurance (GA), before issuing an export license. This study summarizes GA items and characteristics. First, GA is The NSG guidelines suggests four items that should be assured by importing countries: peaceful use of export items, full-scope safeguards application, physical protection measures, and retransfer control. Therefore, these four items are generally requested based on the NSG guidelines. However, if they are already guaranteed by bilateral nuclear cooperation agreement, the GA could be based on the nuclear cooperation agreement. The GA procedure could be omitted in the case of concluding an administrative agreement that imposes another implementation procedure. The levels of requested GA requested vary with the countries, since the NSG guidelines are just recommendations that are not legally enforceable. Korea requests the level based on the NSG guidelines. Among the four GA items, peaceful use of export items is the most fundamental obligation, and levels of safeguards and physical protection of the importing country could be verified in advance by reviewing the conclusion status of international agreement with the IAEA. Thus, the important thing for the licensee to consider is to decide the level of retransfer control. The NSG guidelines suggest two levels of retransfer control, taking into account the sensitivity of the export items and the level of safeguards in the end-user country, which is either to receive GA from the third receiving country at an equivalent level with that originally requested from exporting country, or to get a prior consent by the exporting country. The latter should be approached more carefully, as requiring a prior consent is not only to have authority, but also to have responsibility when problem occurs. In addition, the level of retransfer control must be decided through sufficient reviewing on the transaction characteristics, it may affect the domestic export industry.

Trades are classified as a direct trade, in which an exporter and an importer directly conclude a contract to execute a transaction, and an indirect trade, in which a transaction is conducted through a third party. A license issued by NSSC is required for the indirect trade of Trigger List Items although the items do not cross the Korea’s borders. This study would summarize characteristics for each type of the indirect trade, and suggest things considered from the perspective of regulations. Sensitive items such as weapons of mass destruction could be illegally transferred through black trade, and the role of brokers is important. Therefore, Korea controls indirect trade of Trigger List Items in a more conservative approach, although NSG guidelines does not include it as a control scope. Advanced countries in export control field such as the US, UK, and Canada also controls on indirect trade. One of the indirect trade types is an intermediate trade in which the goods are imported from a foreign country and exported to another foreign country without being brought into the trader’s country. Another type is a merchandising trade which is a form of brokerage between exporter from a foreign country and importer from another foreign country. Both types have one thing in common that the goods are not crossed the trader’s country. The difference is that an intermediate trader directly participates in the contract and earn a difference between import and export amount, whereas a merchandising trader just arranges the transaction and earns a brokerage fee. The profits from the intermediate trade are considered as export records, while merchandising trade profits are not considered export records. In other words, only the intermediary trade is considered as an export. Also, the license types are different for each of them. An export license should be issued for the intermediate trade of Trigger List Items, whereas a brokerage license should be issued for the merchandising trade of Trigger List Items. The definition of export in the Foreign Trade Act includes intermediate trade for only goods, but technology is missing, although the sub-regulation specifies the intermediate trade including both. The technology need to be added as it can be the subject of intermediate trade in spite of intangible characteristics. Also, outreach activities are more needed as nuclear industry awareness on export control for both trades is low.

Nuclear power plants (NPPs) are designed in consideration of redundancy, diversity, and independence to prevent leakage of radioactive materials from safety of view, and a contingency plan is established in case of DBA (Design Basis Accident) occurrence. In addition, NPPs have established contingency plans for physical attacks, including terrorist intrusions and bomb attacks. However, the level of contingency plan caused by cyberattacks is quite insufficient compared to the contingency plan in terms of safety and physical protection. The purpose of this paper is to present the problems of cyberattack contingency plan and methods to supplement it. The first problem with cyberattack contingency plan is that the initiating event for implementing the contingency plan is undecided. In terms of safety, the DBA is identified as an initial event, and each contingency plan is based on the initial events specified in the DBA such as Loss of Coolant Accident and Loss of Offsite Power. In terms of physical protection, each has a contingency plan by identifying bomb attacks and terrorist intrusions in Protected Area and Vital Area as initial events. On the other hand, in the contingency plan related to a cyberattack, an initial event caused by a cyberattack is not identified. For this, it is necessary to classify the attack results that may occur when the CDA is compromised based on the attack technique described in Design Basis Threat. Based on this, an initiating event should be selected and a contingency plan according to each initiating event should be established. The second problem is that there is no responsibility matrix according to the occurrence of the initiating event. From a safety point of view, when a DBA occurs, the organization’s mission according to each initial event is described in the contingency plan, and related countermeasures are defined in case of an accident through Emergency Operation Procedure. In the case of physical protection, referring to IAEA’s Regulatory Guide 5.54, the organization’s responsibility is defined in matrix form when an initial event such as a bomb attack occurs. In this way, the responsibility matrix to be carried out in case of initiating events based on the defined initial event should be described in the contingency plan. In this paper, the problems of the cyberattack contingency plan are presented, and for this purpose, the definition of the initial event and the need for a responsibility matrix when the initial event occurs are presented.

Concerns about North Korea’s 7th nuclear test have been rising recently, and it is a significant threat to the situation around the Korean Peninsula. Amidst these threats, the Korean government also shows a strong will for denuclearizing the Korean Peninsula, referring to the “Audacious Initiative.” For denuclearization negotiations with North Korea, it is essential first to understand North Korea’s nuclear capabilities. However, since access to information is complicated and contains many uncertainties, many studies have been conducted to estimate it. Among them, Von Hippel surveyed to estimate the total amount of uranium ore based on information on uranium mining, which is relatively widely known throughout North Korea’s nuclear fuel cycle, and the amounts of HEU and Pu suggested by many experts. KINAC has conducted a study on a methodology that can narrow the estimation range and improve reliability through the Bayesian Network based on Von Hippel’s research results. However, in this study, the probability distribution is assumed to be the simplest form of uniform distribution, and the estimation formula for the amount of Pu produced compared to the amount of uranium loaded in the core is used as it is, which is an error in Von Hippel’s study. Improvement is needed. This study proposes a more reliable BN model by supplementing this and attempts to estimate the amount of uranium ore that North Korea produces or possesses. Of course, the data used as the basic structure of the model is insufficient, and the estimation formula is straightforward, so it is somewhat unreliable to trust the estimate for uranium ore. However, it is expected to be a suitable methodology that can narrow the scope of North Korea’s nuclear material production estimate or compensate for the uncertainty of the nuclear material production estimation model being developed at KINAC.

The spent fuel safety information delivered from the consignor to the disposal facility operator directly affects the operation and safety of the disposal facility. Therefore, the operator of a disposal facility must perform data quality management to increase data reliability, and anomaly detection is a representative method among quality control methods. We propose a quality control method to detect anomalies using XGBoost, known for its excellent performance, prevention of overfitting, and fast training speed. First, we select significant variables such as release burnup, enrichment, and amount U from the spent fuel safety information and train models for each variable using only normal data. A model trained using only normal data generates a small error for a normal pattern and a large error for an abnormal pattern. Then, when the data error exceeds a set threshold, the data is determined as an anomaly. In this paper, we implement the XGBoost models using virtual spent fuel information and optimize the hyperparameter of XGBoost using a simulated annealing method for high accuracy. The optimized XGBoost models show high accuracy in a normal input and provide a stable prediction value even in an abnormal input. In addition, we perform anomaly detection by including defect input in the data to validate the presented method. The proposed method shows the result of effectively classifying normal values and anomalies.

One of the promising candidates for heat transfer fluid is molten chloride salts. They have been studied in various fields such as the electrolyte of pyroprocessing, the molten salt reactor coolant, and the energy storage system media. Main considerations for utilizing molten chloride salts are the compatibility of salts with structural materials. The corrosion behavior of structural materials in molten chloride salts must be understood to identify suitable materials against the corrosive environment. In this study, the corrosion behavior of a candidate structural material, Hastelloy N, in molten LiCl- KCl salt at 500°C were investigated by the electrochemical impedance spectroscopy (EIS) method. The sheet type of Hastelloy N was utilized as the working electrode in LiCl-KCl to measure the EIS data for 100 hours with 5 hours of time intervals. The EIS data were measured in the frequency range from 104 Hz to 10-2 Hz with the AC signal (amplitude = 20 mV) at open circuit potential. The capacitance semicircle observed in Nyquist plots for all periods indicates that charge-transfer controlled reactions occur. As the immersion time progresses, the radius of the semicircle in Nyquist plots and the impedance and phase angle in Bode plots decrease. These behaviors suggest a decreasing reaction resistance and the corrosion reactions are accelerated with the immersion time. The EIS data were fitted using the equivalent circuit to achieve quantitative results. Two capacitor-resistor components were considered due to the overlapped shape of two valleys in phase angle. The depressed shape of the semicircle in Nyquist plots led to the use of the constant phase element(Q) instead of the capacitor. Therefore, R(Q(R(QR))) circuit was selected to fit the EIS data. Fitting results show that the charge transfer resistance decreases dramatically within 1 day and then converges. The film resistance shows no clear trends, but the increase of the film admittance value indicates the decreased film thickness. Consequently, the film appears to exist like the oxide layer but it does not act as a protective layer. The real-time EIS data were measured in molten salt and provides the corrosion behavior over time. The corrosion mitigation strategy should consider that the corrosion of Hastelloy N accelerates over time and its intrinsic film cannot act as the protective layer. The next steps of this study are to evaluate other candidate structural materials and to demonstrate the presence of the film.

Separation of high heat generating-radioactive isotopes from spent nuclear fuel is an important issue because it can reduce the final disposal area. As one of the technologies that can selectively separate only high heat generating-radioactive isotopes without dissolving spent fuel, the methods using molten salt have recently attracted attention. Although studies on chemical changes of Sr oxides in molten salts have been reported, they have limitation in that alternative oxide reagents rather than oxide fuel were used. In this study, the separation behaviors of Sr from simulated oxide fuel using various molten salts were investigated. A powder type containing 95.7wt% of U and 0.123wt% of Sr was used as the simulated oxide fuel. LiCl, LiCl-CaCl2, MgCl2, LiCl-KCl-MgCl2 and NaCl-MgCl2 were used as molten chloride salts. The separation of Sr from the simulated oxide fuel was conducted by loading it in porous alumina basket and immersing it in a salt. The concentration of Sr in the salt was measured by ICP analysis after sampling the salt outside the basket using dip-stick technique. The separation efficiencies of Sr from simulated oxide fuel using the salts were compared. Furthermore, the causes of their separation efficiency were systematically investigated.

Hydride reorientation is one of the major concerns for cladding integrity during dry storage. In this study, mechanical property of post-reorientation cladding was investigated according to the morphology and amount of the hydrides. Cladding peak temperature limit 400°C was suggested by U.S. NRC in concern of cladding creep and hydride reorientation. In line with this regulatory limit, hydride reorientation was conducted during cool-down process from the maximum temperature of 400°C, using constant internal pressurization method. The specimens were charged for hydrogen from 100 to 1,000 wppm, and various pressures range of 7.5-18.5 MPa were applied. The morphology was examined by optical microscopy. Radial hydride fraction (RHF) and radial hydride continuous path (RHCP) were calculated using image analysis software PROPHET. Finally, strain energy density (SED) was investigated via ring compress tests and the hydrogen concentration was analyzed. The result shows that when RHF is higher than 5%, SED exponentially decreases with RHF. For RHF less than 5%, SED was primarily affected by the total amount of hydrogen. Shortened length of radial hydrides with the presence of circumferential hydrides may block the radial propagation of crack. The result implies that lower burnup spent fuel with lower hydrogen concentration may be more vulnerable in terms of radial hydride compared to higher burnup fuel.

B4C/Al composite is mainly used for neutron absorbing materials, which is one of the components of equipment that manages spent nuclear fuel. There are various processes for manufacturing neutron absorbing materials, but most of them are based on the powder metallurgy. In this study, B4C/Al composite in which the reinforcement was uniformly dispersed was manufactured by using the stir casting process. The microstructure, thermal neutron absorption rate, mechanical properties and dispersibility of the reinforcement of the prepared B4C/Al composite were analyzed.

To minimize the short-term thermal load on the repository facility, heat generating nuclides such as Cs-137 and Sr-90 should be separated from the spent nuclear fuel for efficiency of repository facility. In particular, Sr-90 must be separated because it generates high heat during the decay process. Recently, Korea Atomic Energy Research Institute (KEARI) is developing a waste burden minimization technology to reduce the environmental burden caused by the disposal of spent nuclear fuel and maximize the utilization of the disposal facility. The technology includes a nuclide management process that can maximize disposal efficiency by selectively separating and collecting major nuclides such as Cs, Sr, I, TRU/RE, and Tc/Se from spent nuclear fuel. Among the major nuclides, Sr nuclides dissolve in chloride phase during the chlorination process of spent nuclear fuel and recovered in the form of carbonate or oxide via reactive distillation. In this process, Ba nuclides are also recovered along with Sr nuclides due to their chemical similarity. In this study, we prepared group II nuclide ceramic waste form, Ba(x)Sr(1-x)TiO3 (x=0, 0.25, 0.5, 0.75, 1), using the solid-state reaction method by considering the various ratio of Sr/Ba nuclides generated from nuclide management process. The established waste form fabrication process was able to produce a stable waste form regardless of the ratio of Sr/Ba nuclides. To evaluate the stability of group II waste form, physicochemical properties such as leaching and thermal properties were evaluated. Also, the radiological properties of the Ba(x)Sr(1-x)TiO3 waste forms with various Sr/Ba ratios were evaluated, and the estimation of centerline temperature was carried out using the experimental thermal property data. These results provided fundamental data for long-term storage and management of group II nuclides waste form.

The management before disposal of spent nuclear fuel is an essential process for safe management. It is important to determine the amount of nuclide inventory in order to ensure the integrity of spent nuclear fuel, as radiation generated from the nuclides is generated along with residual heat in the spent nuclear fuel. Based on the data on the characteristics of spent nuclear fuel generated in Korea, the correlation equation between burnup and enrichment was derived by referring to overseas cases (Sweden). Source term analysis was performed using the SCALE ORIGEN ARP code by securing the burnup history of nuclear fuel. Calculation was performed by inputting the combustion history of the fuel WH14×14 and WH17×17 as a reference for CE16×16 spent fuel. Through this study, the relationship was identified using the burnup, enrichment, and cooling time factors that influence the characteristics of spent nuclear fuel. In addition, the total source and spectrum data from neutrons and gamma sources were used to find out the characteristics of fuel.

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

Since SMR’s reduced reactor radius results in higher neutron leakage, SMR operates at a relatively lower discharge burnup level than traditional Light Water Reactors (LWRs). It may result in larger spent fuel amounts for SMRs. Furthermore, recent studies demonstrated that NuScale reactor will generate a significantly higher volume of low- and intermediate-level waste owing to components located near the active core including the core barrel and the neutron reflector. For spent nuclear fuel simulation, FRAPCON-4.0 was updated. Major modifications were made for fission and decay gas release, pellet swelling, cladding creep, axial temperature distribution, corrosion, and extended simulation time covering from steady-state to dry storage. In this study, typical 17×17 PWR fuel (60 MWd/kgU) and NuScale Power Module (36 MWd/kgU) was compared. NuFuel-HTP2™ fuel assembly, which has a half-length of proven LWR fuel, was employed. Owing to the lower discharge burnup and operating temperature, the maximum hydrogen pickup was 73 wppm and the maximum hoop stress was ~25 MPa. Therefore, hydride reorientation issue is irrelevant to SMR spent fuel. In this context, the current regulatory limit for dry storage (i.e. 400°C and 90 MPa) can be significantly alleviated for LWR-based SMRs. The increased safety margin for SMR spent fuel may compensate high spent fuel management cost of SMRs incurred by an increased amount. The comprehensive analysis on SMR spent fuel management implications are discussed based on simulated SMR fuel characteristics.

This study reassess safety margin of the current Peak Cladding Temperature (PCT) limit of dry storage in terms of hydrogen migration by predicting axial hydrogen diffusion throughout dry storage with respect to wet storage time and average burnup. Applying the hydride nucleation, growth, and dissolution model, an axial finite difference method code for thermal diffusion of hydrogen in zirconium alloy was developed and validated against past experiments. The developed model has been implemented in GIFT – a nuclear fuel analysis code developed by Seoul National University. Various discharge burnups and wet storage time relevant to spent fuel characteristics of Korea were simulated. The result shows that that the amount of hydrogen migrated towards the axial end during dry storage for reference PWR spent fuel is limited to ~50 wppm. This result demonstrates that the current PCT margin is sufficient in terms of hydrogen migration.