The aim of this study is to ensure the structural integrity of a canister to be used in a dry storage system currently being developed in Korea. Based on burnup and cooling periods, the canister is designed with 24 bundles of spent nuclear fuel stored inside it. It is a cylindrical structure with a height of 4,890 mm, an internal diameter of 1,708 mm, and an inner length of 4,590 mm. The canister lid is fixed with multiple seals and welds to maintain its confinement boundary to prevent the leakage of radioactive waste. The canister is evaluated under different loads that may be generated under normal, off-normal, and accident conditions, and combinations of these loads are compared against the allowable stress thresholds to assess its structural integrity in accordance with NUREG-2215. The evaluation result shows that the stress intensities applied on the canister under normal, off-normal, and accident conditions are below the allowable stress thresholds, thus confirming its structural integrity.

The Wolsong Nuclear Power Plant (NPP) operates an on-site spent fuel dry storage facility using concrete silo and vertical module systems. This facility must be safely maintained until the spent nuclear fuel (SNF) is transferred to an external interim or final disposal facility, aligning with national policies on spent nuclear fuel management. The concrete silo system, operational since 1992, requires an aging management review for its long-term operation and potential license renewal. This involves comparing aging management programs of different dry storage systems against the U.S. NRC’s guidelines for license renewal of spent nuclear fuel dry storage facilities and the U.S. DOE’s program for long-term storage. Based on this comparison, a specific aging management program for the silo system was developed. Furthermore, the facility’s current practices—periodic checks of surface dose rate, contamination, weld integrity, leakage, surface and groundwater, cumulative dose, and concrete structure—were evaluated for their suitability in managing the silo system’s aging. Based on this review, several improvements were proposed.

A transfer cask serves as the container for transporting and handling canisters loaded with spent nuclear fuels from light water reactors. This study focuses on a cylindrical transfer cask, standing at 5,300 mm with an external diameter of 2,170 mm, featuring impact limiters on the top and bottom sides. The base of the cask body has an openable/closable lid for loading canisters with storage modules. The transfer cask houses a canister containing spent nuclear fuels from lightweight reactors, serving as the confinement boundary while the cask itself lacks the confinement structure. The objective of this study was to conduct a structural analysis evaluation of the transfer cask, currently under development in Korea, ensuring its safety. This evaluation encompasses analyses of loads under normal, off-normal, and accident conditions, adhering to NUREG-2215. Structural integrity was assessed by comparing combined results for each load against stress limits. The results confirm that the transfer cask meets stress limits across normal, off-normal, and accident conditions, establishing its structural safety.

In our previous study, we developed a CFD thermal analysis model for a CANDU spent fuel dry storage silo. The purpose of this model is to reasonably predict the thermal behavior within the silo, particularly Peak Cladding Temperature (PCT), from a safety perspective. The model was developed via two steps, considering optimal thermal analysis and computational efficiency. In the first step, we simplified the complex geometry of the storage basket, which stored 2,220 fuel rods, by replacing it with an equivalent heat conductor with effective thermal conductivity. Detailed CFD analysis results were utilized during this step. In the second step, we derived a thermal analysis model that realistically considered the design and heat transfer mechanisms within the silo. We developed an uncertainty quantification method rooted in the widely adopted Best Estimate Plus Uncertainty (BEPU) method in the nuclear industry. The primary objective of this method is to derive the 95/95 tolerance limits of uncertainty for critical analysis outcomes. We initiated by assessing the uncertainty associated with the CFD input mesh and the physical model applied in thermal analysis. And then, we identified key parameters related to the heat transfer mechanism in the silo, such as thermal conductivity, surface emissivity, viscosity, etc., and determined their mean values and Probability Density Functions (PDFs). Using these derived parameters, we generated CFD inputs for uncertainty quantification, following the principles of the 3rd order Wilks’ formula. By calculating inputs, A database could be constructed based on the results. And this comprehensive database allowed us not only to quantify uncertainty, but also to evaluate the most conservative estimates and assess the influence of parameters. Through the aforementioned method, we quantified the uncertainty and evaluated the most conservative estimates for both PCT and MCT. Additionally, we conducted a quantitative evaluation of parameter influences on both. The entire process from input generation to data analysis took a relatively short period of time, approximately 5 days, which shows that the developed method is efficient. In conclusion, our developed method is effective and efficient tool for quantifying uncertainty and gaining insights into the behavior of silo temperatures under various conditions.

The objective of this study is development of graphite-boron composite material as a replacement for metal canisters to Improve the heat dissipation and radiation shielding performance of dry spent nuclear fuel storage system and reduce the volume of waste storage system. KEARI research team plan to use the graphite matrix manufacturing technology to pelletize the graphite matrix and adjust the content of phenolic resin binder to minimize pore formation. Specifically, we plan to adjust the ratio of natural and synthetic graphite powder and use uniaxial pressing technology to manufacture black graphite matrix with extremely high radial thermal conductivity. After optimizing the thermal conductivity of the graphite matrix, we plan to mix it with selected boron compounds, shape it, and perform sintering and purification heat treatments at high temperatures to manufacture standard composite materials.

Dry storage of nuclear fuel is compromised by threats to the cladding integrity, such as creep and hydride reorientation. To predict these phenomena, spent fuel simulation codes have been developed. In spent fuel simulation, temperature information is the most influential factor for creep and hydride formation. Traditional fuel simulation codes required a user-defined temperature history input which is given by separate thermal analysis. Moreover, geometric changes in nuclear fuel, such as creep, can alter the cask’s internal subchannels, thereby changing the thermal analysis. This necessitates the development of a coupled thermal and nuclear fuel analysis code. In this study, we integrated the 2D FDM nuclear fuel code GIFT developed at SNU with COBRA -SFS. Using this, we analyzed spent nuclear stored in TN-24P dry storage cask over several decades and identified conditions posing threats due to phenomena like creep and hydrogen reorientation, represented by the burnup and peak cladding temperature at the start of dry storage. We also investigated the safety zone of spent nuclear fuel based on burnup and wet storage duration using decay heat.

It is very important that the confinement of a spent fuel storage systems is maintained because if the confinement is damaged, the gaseous radioactive material inside the storage cask can leak out and have a radiological impact on the surrounding public. For this reason, leakage rate tests using helium are required for certificate of compliance (CoC) and fabrication inspections of spent fuel storage cask. For transport cask, the allowable leakage rate can be calculated according to the standardized scenario presented by the IAEA. However, for storage cask, the allowable leakage rate is determined by the canister, facility, and site specific information, so it is difficult to establish a standardized leakage rate criterion. Therefore, this study aims to establish a system that can derive system-specific leakage test criteria that can be used for leakage test of actual storage systems. First, the variables that can affect the allowable leakage rate for normal and accident conditions were derived. Unlike transportation systems, for storage systems, the dose from the shielding analysis and the dose from the confinement analysis are summed up to determine whether the dose standard is satisfied, and even the dose from the existing nuclear facilities is summed up during normal operation condition. For this reason, the target dose is used as an input variable when calculating the allowable leakage rate for the storage system. In addition, the main variables are the distance from the boundary of the exclusive area, the number of cask, the inventory of nuclide material in the cask, the free volume, and the internal and external pressure. Utilizing domestic and US NRC guidelines, we derived basic recommended values for the selected variables. The GASPARII computer code that can evaluate the dose to the public under normal operating conditions was utilized. Using the above variables, the allowable leakage rate is calculated and converted to the allowable criteria for helium leakage rate test. The developed system was used to calculate the allowable leakage rate for normal and accident conditions for a hypothetical storage system. The leakage rate criteria calculation system developed in this study can be useful for CoC and fabrication inspections of storage systems in the future, and a GUI-based program will be built for user convenience.

To investigate the mechanical integrity of spent nuclear fuel, the failure behavior of the cladding tube was examined under accident conditions. According to the SNL report, the failure behavior of cladding can be broadly classified into two types. The first is failure due to bending load caused by falling. The second is failure due to pinch load caused by space grid. In this study, mechanical integrity was evaluated through the stress intensity factor applied to the crack in failure behavior due to bending load. Since the exact value of the impact load due to fall was unknown, the load was applied by increasing the value up to 200 G in 20 G increments. The size of the crack is an important input variable, and 300 um was given by referring to the EPRI report, and the elastic modulus, a material property that determines the stress field, was given 75.22 GPa by referring to the FRAPCON code. Since the relationship between the direction of stress and the direction of the crack is also a major variable, simulations were conducted for both cracks perpendicular to and parallel to the stress direction. It was confirmed that at a load of 200 G, when the crack was parallel to the stress direction, stress concentration did not occur and had a very low stress intensity factor 0.01 􀜯􀜲􀜽√􀝉. When perpendicular to the direction of stress, the stress intensity factor showed a value of 1 􀜯􀜲􀜽√􀝉. However, considering that the critical value of the stress intensity factor due to hydride is 5 􀜯􀜲􀜽√􀝉, it can be seen that perpendicular result also ensures the mechanical integrity of the cladding.

Notice of the NSSC No.2021-14 defines the term ‘Neutron Absorber’ as a material with a high neutron absorption cross section, which is used to prevent criticality during nuclear fission reactions and includes neutron absorbers as target items for manufacture inspection. U.S.NRC report of the NUREG-2214 states that the subcriticality of spent nuclear fuel (SNF) in Dry Storage Systems (DSSs) may be maintained, in part, by the placement of neutron absorbers, or poison plates, around the fuel assemblies. This report mentions the need for Time-Limited Aging Analysis (TLAA) on depletion of Boron (10B) in neutron absorbers for HI-STORM 100 and HISTAR 100. Also, this report mentions that 10B depletion occurs during neutron irradiation of neutron absorbers, but only 0.02% of the available 10B is to be depleted through conservative assumptions regarding the neutron flux or accumulated fluence during irradiation, which supports the continued use of the neutron absorbers in the SNF dry storage cask even after 60 years of evaluated period. There are several types of commercially available neutron absorbers, broadly classified into Boron Carbide Cermets (e.g., Boral®), Metal Matrix Composites (MMC) (e.g., METAMIC), Borated Stainless Steel (BSS), and Borated Al alloy. While irradiation tests for neutron absorbers are primarily conducted during wet storage systems, there are also some prior studies available on irradiation tests for neutron absorbers during dry storage systems. For examples, there is an analysis of previous research on high-temperature irradiation test of metallic materials and identification of limitations in existing methodologies were conducted. Furthermore, an improvement plan for simulating the high-temperature irradiation damage of neutron absorbers was developed. In report published by corrosion society summarizes the evaluation results of the degradation mechanisms for Stainless Steel- and Al-based neutron absorbers used in SNF dry storage systems.

International Atomic Energy Agency defines the term “Poison” as a substance used to reduce reactivity, by virtue of its high neutron absorption cross-section, in IAEA glossary. Poison material is generally used in the reactor core, but it is also used in dry storage systems to maintain the subcriticality of spent fuel. Most neutron poison materials for dry storage systems are boron-based materials such as Al-B Carbide Cermet (e.g., Boral®), Al-B Carbide MMC (e.g., METAMIC), Borated Stainless Steel, Borated Al alloy. These materials help maintain subcriticality as a part of the basket. U.S.NRC report NUREG-2214 provides a general assessment of aging mechanisms that may impair the ability of SSCs of dry storage systems to perform their safety functions during longterm storage periods. Boron depletion is an aging mechanism of neutron poison evaluated in that report. Although that report concludes that boron depletion is not considered to be a credible aging mechanism, the report says analysis of boron depletion is needed in original design bases for providing long-term safety of DSS. Therefore, this study aimed to simulate the composition change of neutron poison material in the KORAD-21 system during cooling time considering spent fuel that can be stored. The neutron source term of spent fuel was calculated by ORIGEN-ARP. Using that source term, neutron transport calculation for counting neutrons that reach neutron poison material was carried out by MCNP®-6.2. Then, the composition change of neutron poison material by neutron-induced reaction was simulated by FISPACT-II. The boron-10 concentration change of neutron poison material was analyzed at the end. This study is expected to be the preliminary study for the aging analysis of neutron poison material about boron depletion.

In nuclear fuel development research, consideration of the back-end cycle is essential. In particular, a review of an in-reactor performance of nuclear fuel related to the various degradation phenomena that can occur during spent fuel dry storage is an important area. The important factors affecting the degradation of zirconium-based cladding during dry storage are the cladding’s hydrogen concentration and rod internal pressure after irradiation. In this study, a preliminary analysis of the in-reactor behavior of the HANA cladding, which has been developed and is currently undergoing licensing review, was performed, and based on this result, a comparative analysis between nuclear fuel with HANA cladding and current commercial fuel under storage conditions was performed. The results show that the rod internal pressure of nuclear fuel with HANA cladding is not significantly different from that of commercial cladding, and the hydrogen concentration in the cladding tends to reduce due to the increased corrosion resistance, so fuel integrity in a dry storage conditions is not expected to be a major problem. Although the lack of cladding creep data under dry storage conditions, the results from the Halden research reactor test comparing in-reactor creep behavior with Zircaloy-4 showed that there is sufficient margin for degradation due to creep during storage.

Since the September 11 terrorist attacks in the United States, concerns about intentional aircraft crashes into nationally critical facilities have soared in countries around the world. The United States government advised nuclear utilities to strengthen the security of nuclear power plants against aircraft crashes and stipulated aircraft crash assessment for new nuclear facilities. Interest in military missile attacks on nuclear facilities has grown after Russia attacked Ukraine’s Zaporizhzhia nuclear power plant, where spent nuclear power dry storage facility is operated. Spent nuclear fuel dry storage facilities in nuclear power plant sites should also strengthen security in preparation for such aircraft crashes. Most, but not all, spent nuclear fuel dry storage facilities in Europe, Japan and Canada are operated within buildings, while the United States and Korea operate dry storage facilities outdoors. Since all of Korea’s dry storage systems are concrete structures vulnerable to crash loads and are exposed to the outside, it is more necessary to prepare for aircraft crash terrorist attacks due to the Korea’s military situation. Residents near nuclear power plants are also demanding assessment and protective measures against such aircraft crashes. However, nuclear power plants, including spent nuclear fuel dry storage facilities, are strong structures and have very high security, so they are unlikely to be selected as targets of terrorism, and spent nuclear fuel dry storage systems are so small that aircraft cannot hit them accurately. Collected opinions on the assessment of aircraft crash accidents at spent nuclear fuel dry storage facilities in nuclear power plant sites were reviewed. In addition, the current laws and regulatory requirements related to strengthening the security of new and existing nuclear power plants against intentional aircraft crashes are summarized. Such strengthening of security can not only ensure the safety of on-site spent nuclear fuel dry storage facilities, but also contribute to the continuous operation of nuclear power plants by increasing resident acceptance.

Regulatory agencies require burn-up verification to ensure that dry storage casks using burn-up credit are not loaded with fuel with a reactivity greater than the allowable standard. Accordingly, in preparation for dry storage of SF, the reliability of the burnup was verified and action plans for fuel with confirmed errors were reviewed. Reliability verification was performed by comparing the actual burnup calculated with combustion calculation code (TOTE, ISOTIN) used in NPP and the design burnup calculated with the nuclear design code (ANC). As a result of comparing the differences between actual burnup and design burnup for 7,414 assemblies of SF generated from CE-type NPPs, the average deviation was confirmed to be 0.79% and 220 MWD/MTU. In the CE-type NPPs, no fuel showing large deviations was identified, and it was confirmed that reliability was secured. As a result of comparing the differences in 11,082 assemblies of SF generated from WH-type NPPs, the differences were not large, averaging 1.16% or 422 MWD/MTU. However, fuels showing significant differences were identified, and cause analysis was performed for those fuels. The cause analysis used a method of comparing the burnup of symmetrically loaded fuels in the reactor. For fuels that were not symmetrically loaded, a method was used to compare them with fuels with similar combustion histories. As a result of the review, it was confirmed that the fuel was under- or over-burned compared to symmetrically loaded fuel. For fuels for which clear errors have been identified, we are considering replacing them with the design burnup, and for fuels whose causes cannot be confirmed, we are considering ways to maintain the actual burnup.

In Korea, most temporary storage facilities for spent nuclear fuel are nearing saturation. As an alternative to this, the 2nd basic plan for high-level radioactive waste management specified the operation plan of dry interim storage facility. Meanwhile, the NSSC No. 2021-19 stipulates that it is necessary to evaluate the possibility and potential effect of accident before operating interim storage facility. Therefore, this study analyzed the categories of accident scenarios that may occur in dry storage facility as part of prior research on this. We investigated the case of categorization of dry storage facility accident scenarios of IAEA, NRC, KAREI, and KINS. The IAEA presented accident scenarios that could occur in on-site dry storage facility operated with silo and cask method. NRC has classified accident scenarios in dry storage facility and estimated the probability of accidents for each. KAERI and KINS selected major accident scenarios and analyzed the processes for each, in preparation for the introduction of dry storage facility in Korea in the future. Overall, a total of 10 accident scenarios were considered, and the scenarios considered by each institution were different. Among 10 scenarios, cask drop and aircraft collision were included in the categorization of most institutions. The results of this study can be used as basic data for cataloging accidents subject to safety evaluation when introducing dry interim storage facility in Korea in the future.

South It is necessary to develop the future technologies to improve the sustainability and acceptability of nuclear power plants generation. Currently, our company is preparing to build the dry storage facility on-site in accordance with the basic plan for managing high-level radioactive waste announced by the government in 2021. However, studies on technologies for the volume reduction of spent nuclear fuel to increase the efficiency of on-site spent fuel dry storage facilities are very not enough. Accordingly, in this study, the storage efficiency and appropriateness for the SF volume reduction processing technologies such as SF oxide processing technology and consolidation technology are evaluated. Finally, the goal is to develop the optimized technologies to improve the storage efficiency of spent nuclear fuel. As a result in this study is followings. [Safety] After removing volatile fission products (Xe, Kr, I, etc.), Xe, Kr, etc. are removed during storage of the sintered structures. UO2 has a high melting point of approximately 1,000°C after cesium (Cs) has been removed, and heat can be removed by natural convection. [Economy]1999 DUPIC unit facility unit price reference, 2020 standard 328 $/kg estimated. A Comprehensive Approach Considering the Whole System is needed. Benefit from replacement and continuous operation of metal storage containers. Changes in economic efficiency obtained in conjunction with fluctuations in electricity prices and disposal. [Waste filter] A separated solidification facility high-level waste filter is required, and overseas outsourcing must be considered. [Waste cladding]. Cannot be accommodated in low-level disposal site. This reason is why the Ni nuclides occur to be in bulk. [Metal structural material] It is possible to reduce the initial volume by 7.6% or more when compressed or melted, but the technology needs to be advanced. [Oxide blocks] Larger size and density are expected to improve storage and disposal efficiency. [Facilities operation waste] Expected to be able to be disposed of at mid-to-low level decommissioning sites in Gyeongju city. [Solidified volatile nuclides and activated metals] Expected to improve storage efficiency when used volume is reduced and stored, such as outsourced reprocessing. [Oxide block] Radioactivity and decay heat are estimated to be reduced by half during oxide treatment. 75% reduction in volume and 40% reduction in storage area compared to used nuclear fuel before treatment. [Merits/Shortages] Improvement of storage and disposal efficiency empirical research such as large-capacity [real-scale] oxide block production is required. Oxide processing facilities are likely to be classified as post-use nuclear fuel processing facilities. It is determined that additional documents such as a Radiation Environmental Report (RER) must be submitted. Existence of possible external leaks of glass, highly mobile radionuclides from the point of view of nuclear criticality and heat removal. Acceptancy requirements of citizens in the process of creating additional sites for oxide treatment facilities. Considering social public opinion, it is necessary to secure the acceptability such as residents’ opinions convergence. Characteristics of high nuclear non-propagation compared to other processing technologies involving chemical processing. Also, Expectation of volume reduction effect for spent nuclear fuel itself. Volume reduction methods for solid waste and gaseous waste are required.

In KNF, fuel performance analysis modules were developed to predict the overall behavior of a fuel rod under normal operating conditions. Their main focus is to provide information on initial conditions prior to dry storage. Potential degradation mechanisms that may affect sheath integrity of spent CANDU 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 modules 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, identify and extend the ranges of all modules to required operating ranges. The 300°C spent CANDU fuel sheath temperature metric for dry storage ensures spent CANDU fuel element integrity from the failure mechanisms of creep rupture, oxidation and stress corrosion cracking at a failure probability of 2×10-5 for a dry storage time of 100 years. The 300°C sheath temperature metric for dry storage has relatively a lower failure rate than the target criteria for dry storage of spent LWR fuel. Although different modes of failure were treated separately for simplicity, ignoring possible synergistic effects, these results are conservative because of the conservative assumptions that have been made for evaluating spent fuel element conditions, and because of the inherent conservatism of the applied models. Additional conservatism of the model comes from the fact that isothermal conditions do not prevail in actual storage conditions. Further R&D being considered includes acquisition of new functional models to implement overall fuel behavior evaluation and cover spent CANDU fuel in dry storage, and upgrades of the analysis module to achieve sufficient accuracy in key output parameters. The developed modules provide a platform for research and industrial applications, including the design of fuel behavior experiments and prediction of safe operating margins for spent CANDU fuel.

Korea Hydro & Nuclear Power (KHNP) is currently developing a vertical concrete dry storage module for the dry storage of used nuclear fuel within nuclear power plants. This module is designed with a structure consisting of cylinders, which can block the ingress of external air, thereby preventing Chloride-Induced Stress Corrosion Cracking (CISCC). However, due to the presence of these cylinder structures, unlike conventional dry storage systems, it cannot directly dissipate heat to the external atmosphere, making thermal evaluation an important issue. The SF dry storage module being developed by KHNP is a massive concrete structure of approximately 20 m × 10 m × 7 m in size, employing a vertical storage system. To demonstrate the safety of such a large structure, there is no alternative to conducting experiments with scaled-down models. Furthermore, according to NUREG-2215 Section 5.5.4, it is explicitly mentioned that design-verification testing can be performed using scaled-down models. In this paper, a 1/4 scaled-down model was constructed to perform thermal performance verification experiments, and the effectiveness of this model was analyzed using Computational Fluid Dynamics (CFD) methods. The analysis results indicated that there was not a significant difference in terms of maximum concrete temperature and air outlet temperature. However, a considerable difference was observed in the canister surface temperature. Therefore, it is concluded that careful consideration of natural convection heat transfer is necessary for the full application of the scaled-down model.

In the case of dry storage facilities, slipping of the cask or tip-over are dangerous phenomena. For this reason, in dry storage facilities, measures against slipping and tip-over or related safety evaluations are important. Accidental conditions that can cause cask slippage and tip-over in dry storage facilities include natural phenomena such as floods, tornadoes, tsunamis, typhoons, earthquakes, and artificial phenomena such as airplane crashes. However, among natural phenomena, earthquakes are the most important natural phenomenon that causes tip-over. Also, many people had the stereotype that Korea is an earthquake-safe zone before 2016. However, earthquakes become a major disaster in Korea due to the 2016 Gyeongju earthquake and the 2017 Pohang earthquake, followed by the Goesan earthquake in October 2022. In this paper, seismic analysis was performed based on dry storage facilities including multiple casks. Design variables for the construction of an analysis model for dry storage facilities were investigated, and seismic analysis was performed. To evaluate tip-over accident during earthquake, seismic load was used from 0.2 g PGA to 0.8 g PGA and these earthquakes were followed Design Response Spectrum (DRS) in RG 1.60. The friction coefficient of concrete pad was used from 0.2 to 1.0. As a result of the analysis, tip-over accident could not find in the analysis from 0.2 g to 0.6 g. However, tip-over was appeared at friction coefficients of 0.8 and 1.0 at 0.8 g PGA. Tip-over angular velocity of cask was derived by seismic analysis and was compared with formula and tip-over analysis results. As a result, a generalized dry storage facility analysis model was proposed, and dry storage facility safety evaluation was conducted through seismic analysis. Also, tip-over angular velocity was derived using seismic analysis for tip-over analysis.

In nuclear facilities, a graded approach is applied to achieve safety effectively and efficiently. It means that the structures, systems, and components (SSCs) that are important to safety should be assured to be high quality. Accordingly, SSCs that consist of nuclear facilities should be classified with respect to their safety importance as several classes, so that the requirements of quality assurance relevant to the designing, manufacturing, testing, maintenance, etc. can be applied. Guidance for the safety classification of SSCs consisting of nuclear power plants and radioactive waste management facilities was developed by U.S.NRC and IAEA. Especially, in guidance for nuclear power plants, safety significance can be evaluated as following details. The single SSC that mitigates or/and prevents the radiological consequence or hazard was assumed to be failure or malfunction as the initiating event/accident occurred and the following radiological consequence was evaluated. Considering both the consequence and frequency of the occurrence of the initiating event/accident, the safety significance of each SSC can be evaluated. Based on the evaluated safety significance, a safety class can be assigned. The guidance for the safety classification of the spent nuclear fuel dry storage systems (DSS) was also developed in the United States (NUREG/CR-6407) and the U.S.NRC acknowledges the application of it to the safety classification of DSS in the United States. Also, worldwide including the KOREA, that guidance has been applied to several DSSs. However, the guidance does not include the methodology for classifying the safety or the evaluated safety significance of each SSC, and the classification criteria are not based on quantitative safety significance but are expressed somewhat qualitatively. Vendors of DSS may have difficulties to apply this guidance appropriately due to the different design characteristics of DSSs. Therefore, the purpose of this study is to evaluate the safety significance of representative SSCs in DSS. A framework was established to evaluate the safety significance of SSCs performing safety functions related to radiation shielding and confinement of radioactive materials. Furthermore, the framework was applied to the test case.

A seal is one of the primary means of safeguards along with surveillance. The International Atomic Energy Agency (IAEA) uses various types of seals to verify the diversion of nuclear materials and is developing new seals according to the development of technology. Independent of the IAEA, ROK uses national safeguards seals for state-level regulation. A national safeguards inspector binds the nuclear material storage by combining a seal with a metal wire and checks the serial number of the RFID chip inserted in the seal with a reader. The Wolsong spent fuel dry storage facility has 14 modules, each with 24 seals, and thus a maximum of 336 national seals will be installed. Although dependent on the sealing method, it takes about 5 minutes to verify one seal. As such, a considerable workforce is required for verification, and both the IAEA and the ROK are currently conducting random inspections. In addition, there are cases where verification is impossible because old seals are damaged due to harsh environments and long exchange periods. Therefore, in this study, we analyzed cases in areas where sealing technology has been developed to improve the problems of the existing national safeguards seals. And we proposed a method for improving national seals by finding requirements of seals considering spent fuel dry storage facility characteristics. In international logistics, sealing is essential in product transport verification, terrorism prevention, and tariff imposition. Accordingly, the field of container sealing has been extensively developed, and the International Organization for Standardization (ISO) has regulated the mechanical requirements of the seal as ISO 17712 and the electronic requirements as ISO 18185. Mechanical seals include metal and plastic seals and metal seals include bolt seals, ball seals, and cable seals. In addition, there are various electronic seals, such as radio frequency identification (RFID), near field communication (NFC), infrared (IR). Recently, there has been a trend to use active seals that have a built-in battery and can implement various additional functions. Among the various seals, the main requirements for selecting seals suitable for dry storage facilities are as follows. First, use of a sealing tube longer than 10m should be possible. Second, it should have corrosion resistance so that it can be used for more than five years in the coastal area. Third, it must be a passive seal without a power supply. Fourth, it should not be overly costly. Finally, the seal verification time should be short. As a seal that satisfies these requirements, an electronic seal with application of the passive RFID method to the mechanical form of a metal cable seal is suitable. Since it is not an active seal, it is difficult to determine the time of breakage. Therefore, designing the seal such that the RFID is also damaged when the metal seal is broken will be helpful for verification. In this study, the requirements for national safeguards seals in dry storage facilities were defined, and measures to improve the existing national seals were studied. Field applicability will be evaluated through future sealing device design and demonstration tests.