Spent nuclear fuels should be safely stored until being disposed and dry storage system is predominantly used to retain the fuels. Thermal analysis to estimate temperatures of spent nuclear fuel and the storage system should be performed to evaluate whether the temperatures exceed safety limit. Recently, thermal hydraulic analysis with CFD codes is widely used to investigate the temperature of spent nuclear fuel in dry storage. COBRA-SFS is a legacy code based on subchannel analysis code, and its fidelity is verified for evaluating the thermal analysis for licensing a dry cask system. Herein, thermal analysis result based on CFD and COBRA-SFS codes is compared and the Dry Cask Simulator (DCS) is assessed as a benchmark experiment in this study. Extended Storage Collaborating Program (ESCP) led by the Electric Power Research Institute (EPRI) is organized to address the degradation effects of spent nuclear fuel during long-term dry storage, and DCS is the first phase of the program. The dry storage system, containing a single BWR assembly in a canister, was designed to produce validation-quality data for thermal analysis model. ANSYS FLUENT was used to simulate DCS. Simulations were conducted in various decay heat and helium pressure inside the canister. In realistic conditions of decay heat and helium pressure of actual dry cask system, CFD and COBRA-SFS analysis results gave good agreement with experimental measurement. Peak temperatures of channel can, basket, canister and shell predicted by CFD simulation also showed good prediction and the discrepancies were less than 7 K while measurements uncertainty was 7 K. In high decay heat and high pressure condition, however, CFD and COBRA-SFS underestimated peak cladding temperature than experimental results.

South Korea has been storing UNF in spent fuel pool dry storage facility within Nuclear Power Plants. The dry storage facility of used nuclear fuel (UNF) is essential to sustain safety and sustain stable operation of a nuclear power plant. Most abroad countries have attempted to develop a variety of dry storage facility for used nuclear fuel in order to retain the safe restoration. Many studies have been conducting to safety evaluation for the dry storage facility. However, there is not a ventilation evaluation in the wake of fire event that could influence of the thermal effect on the dry storage facility, even though it will likely to occur fire events such as wildfire, air craft crash. In practice, it happened to catastrophic disaster due to the wild fire adjacent to ul-jin mountain. Also, it happened to fire accident near to the Japonia NPP in Ukraine territory caused of military air plane missile. It has not mostly been studied on the ventilation evaluation considered to thermal safety in the dry storage facility excepted for some researches. It could need the mechanical ventilation systems such as HVAC system in the dry storage system, so that thermal effect can be reduced. In this study, we conducted to the ventilation control modelling by using fire modelling tool (Fire Dynamic Simulator v.6.7). The ventilation scenarios made up for 3 case that can compare flowrate variation with ventilation control. As a result of modelling, there is no differentiation between ventilation control using performance curve with not using performance curve even though the pressure fluctuation would be increased, compared with the case of considering performance curve. Second, it evaluated that the mode for fraction control would occur to pressure rise in the state of controlling the ventilation system flowrate. However, sensitivity of flowrate control was more decreased below less than 5 seconds. Third, in the case of on/off control system revealed more higher resolution than other cases caused by flowrate variation. These results could be considered as the design guidelines for the development dry storage facility to improve the thermal performance that can reduce thermal risk. Furthermore, the study results would expect HVAC system installed in dry storage to help automatic ventilation control relevant to dry storage safety increased.

Research on the safety of nuclear spent fuel has been heavily experimented and modelled from a mechanical perspective. The issues of corrosion, irradiation creep, hydride and hydrogen embrittlement have been addressed more than two decades since the early 2000s. Among these degradation behavior, hydrogen embrittlement and hydride reorientation have been the most important topics for establishing the integrity of nuclear spent fuel and have been studied in depth. In order to assess the safety of spent nuclear fuel, firstly, it is necessary to establish the safety criteria in all nuclear cycle, i.e., the failure criteria guidelines for nuclear fuel assemblies and nuclear fuel rods, and then examine the safety analysis. The contents of U.S.NRC Regulations, Title 10 General, Chapter 1 Code of Federal Regulation (CFR), Part 50, 71 and 72, describe the safety criteria for the safety assessment of nuclear fuel assemblies and nuclear fuel rods. In this study, technically important points in safety analysis on nuclear fuel are checked through the reference of those NRC regulation. As result, we confirmed that the safety assessment of nuclear fuel after 20 years of interim storage is now being tested by ORNL and PNNL. There are not quantitative criteria related to material safety. However qualitative criteria which is dependent on environmentally condition describe the safety analysis. There is some literature study about DBTT, yield stress, ultimate tensile strength, flexural rigidity data. In FRAPCON code Modelling of yield strength and creep had been established, but radial hydride or hydride reorientation has not considered.

Concrete structures of spent nuclear fuel interim storage facility should maintain their ability to shield and structural integrity during normal, off-normal and accident conditions. The concrete structures may deteriorate if the interim storage facility operates for more than several decades. Even if deterioration occurs, the concrete structures must maintain their own functions such as radiation shielding protection and structural integrity. Therefore, it is necessary to establish an analysis methodology that can evaluate whether the deteriorated concrete structure maintains its integrity under not only normal or off-normal condition but also accident condition. In accident conditions such as tip over and aircraft collision, both static material properties and dynamic properties are needed to evaluate the structural integrity of the concrete structures. Especially, it has been known to be difficult to estimate the resulted damage precisely where an aircraft collides with the degraded concrete structures at a high strain rate. In this study, damage evaluation of concrete overpack due to aircraft collisions was conducted. First, in order to verify the impact analysis methodology, the aircraft impact analysis of plane concrete overpack was performed and compared with the test results previously conducted by our research team. Then, the impact analysis for the overpack of KORAD21C was performed. In the future, the radiation shielding analysis will be performed under the conditions to evaluate whether or not the radiation shielding ability is maintained.

In concrete structures exposed to chloride environments such as seashore structures, chloride ions penetrate into the concrete. Chlorine ions in concrete react with cement hydrates to form Friedel’s salt and change the microstructure. Changes in the microstructure of concrete affect the mechanical performance, and the effect varies depending on the concentration of chloride ions that have penetrated. However, research on the mechanical performance of concrete by chloride ion penetration is lacking. In this study, the effect of chloride ion penetration on the mechanical performance of dry cask concrete exposed to the marine environment was investigated. The mixture proportion of self-compacting concrete is used to produce concrete specimens. CaCl2 was used to add chlorine ions, and 0, 1, 2, and 4% of the binder in weight were added. To evaluate the mechanical performance of concrete, a compressive strength test, and a splitting tensile strength test were performed. The compressive strength test was conducted through displacement control to obtain a stress-strain curve, and the loading speed was set to 10 με/sec, which is the speed of the quasi-static level. The splitting tensile strength test was performed according to KS F 2423. As a result of the experiment, the compressive strength increased when the chloride ion concentration was 1%, and the compressive strength decreased when the chlorine ion concentration was 4%. The effect of the chloride ion concentration on the peak strain was not shown. In order to present a stress-strain curve model according to the chloride ion concentration, the existing concrete compressive stress-strain models were reviewed, and it was confirmed that the experimental results could be simulated through the Popovics model.

The dry storage of spent fuel has become an increasingly important issue in the field of nuclear energy. Square-gridded baskets have been widely used for the storage of spent fuel because of their superior heat transfer and structural integrity. In this paper, we review the fabrication process of square-gridded baskets for dry storage of spent fuel. The review includes the design considerations, material selection, manufacturing methods, and quality control measures. We also discuss the challenges and opportunities for further improvement in the fabrication of square-gridded baskets. The fabrication of square-gridded baskets is a critical process for the safe and reliable dry storage of spent fuel. The review of the fabrication process highlights the importance of design considerations, material selection, manufacturing methods, and quality control measures. Continued efforts to improve the fabrication process will help to ensure the safe and secure storage of spent fuel.

Since the time to consider when evaluating leakage of spent fuel dry storage systems is very long, assumptions that continue to leak at the initial leakage rate are too conservative. Therefore, this study developed a dynamic methodology to calculate the change in leakage rate using time-varying variables and apply it to calculate the amount of radioactive leakage during the evaluation period. The developed dynamic methodology was then applied to calculate the leakage radiation source term for a hypothetical dry storage system and used to perform a public dose assessment. When applying the developed dynamic leakage rate evaluation methodology for more accurate confinement evaluation in case of containment damage of dry storage system, it was found that the change of leak rate with time is very insignificant if the hole diameter is small enough, and the leak rate decreases rapidly with time when a hole with a certain diameter or larger occurs. In the case of the accident condition, except when the hole is very large, it corresponds to the chocked flow condition, and the leak rate decreases rapidly as soon as the internal pressure is sufficiently lowered to enter the molecular and continuum flow region. In the case of a small hole diameter, the leakage volume is very small, so even if the dynamic methodology is applied, the evaluation results are not different from the case where the initial leakage rate continues, and when the hole diameter exceeds a certain value, the internal pressure drops according to the leakage volume, and the leakage rate decreases significantly. As a result of evaluating the dose to residents by applying the calculated radiation source term, it was confirmed that the dose criteria was exceeded when a hole with a diameter of about 4 μm occurred under off-normal conditions, and the dose standard was exceeded under accident conditions when a chocked flow occurred between the diameter of the hole and 2-3 μm, resulting in a rapid increase in the dose. The results of this study are expected to contribute to a more accurate evaluation of the confinement performance of storage systems, which will contribute to the design of optimal dry storage systems.

In Korea, the construction of dry storage facilities for spent nuclear fuel is being promoted through the 2nd basic plan for high-level radioactive waste management. When operating dry storage facilities, exposure dose assessment for workers should be performed, and for this, exposure scenarios based on work procedures should be derived prior. However, the dry storage method has not yet been sufficiently established in Korea, so the work procedure has not been established. Therefore, research is needed to apply it domestically based on the analysis of spent nuclear fuel management methods in major overseas leading countries. In this study, the procedure for receiving and storing spent nuclear fuel in a concrete overpack-based storage facility was analyzed. Among the various spent nuclear fuel management systems, the metal overpack-based HI-STAR 100 system and the concrete overpackbased HI-STORM 100 system are quite common methods in the United States. Therefore, in this study, work procedures were analyzed based on each final safety analysis report. First, the HI-STAR 100 overpack enters the facility and is placed in the transfer area. Remove the impact limiter of the overpack and install the alignment device on the top of the overpack. Place the HI-TRAC, an on-site transfer device, on top of the alignment unit and remove the lids of the two devices to insert the canister into the HI-TRAC. When the canister transfer is complete, reseat the lid to seal it, and disconnect the HI-TRAC from the HI-STAR 100. Raise the canister-loaded HI-TRAC over the alignment device on the top of the HI-STORM 100 overpack and remove the lids of the two devices that are in contact. Insert the canister into the HI-STORM 100 and reseat the lid. The HI-STORM 100 loaded with spent nuclear fuel is transferred to the designated storage area. In this study, the procedure for receiving and storing spent nuclear fuel in a concrete overpack-based storage facility was analyzed. The main procedure was the transfer of canisters between overpacks, and it was confirmed that HI-TRAC was used in the work procedure. The results of this study can be used as basic data for evaluating the exposure dose of operating workers for the construction of dry storage facilities in Korea.

Once systems, structures and components (SSCs) of dry storage systems are classified with respect to safety function or safety significance (i.e., safety classification), appropriate engineering rules can be applied to ensure that they are designed, manufactured, maintained, managed (e.g. aging management) etc. In Unites States, the systems, structures and components (SSCs) consisting DSSs are classified into two or several grades (i.e., class A, B and C or not important to safety, and important to safety (ITS) or not important to safety (NITS)) with respect to intended safety function and safety significance. This classification methods were based on Regulatory Guide 7.10 (i.e., guidance for use in developing quality assurance programs for packaging). Also, in Korea, SSCs of DSSs should be classified into ITS and NITS in much the same as method based on Regulatory Guide 7.10. In that guidance, for providing graded approach to manage the SSCs of packaging, they were trying to classifying SSCs in accordance with radiological consequences. But there was limitations that the provided classification criteria was still qualitative, so that it was not enough for managing the SSCs according to graded approach. On the other hand, in some other nuclear facilities (i.e., nuclear power plant, radioactive waste management facility and disposal facility etc.), quantitative criteria relevant to radiological consequence (i.e., radiation doses to workers or to the public) or inventory of radioactivity are existed so that it can be applied for classifying safety classes. In summary, the study on the application safety classification that applied quantitative criteria to perform safety classification of SSCs in DSS is inadequate or insufficient. The purpose of this study is proposing the preliminary framework for estimating safety significance of SSCs in DSS which can be utilized in our further advanced studies. In this study, a framework was established to estimate the safety significance of SSCs related to radiation shielding and confinement using MCNP® 6.2 and Microsoft Excel. Referring to the methodology of IAEA Specific Safety Guide 30, we assumed severity for failures of components that could lead to degradation of the SSC’s performance. The safety class of SSC was decided based on the impact of SSC’s failure on consequences.

Spent fuel from the Wolsong CANDU reactor has been stored in above-ground dry storage canisters. Wolsong concrete dry storage canisters (silos) are around 6 m high, 3 m in outside diameter, and have shielding comprised of around 1 m of concrete and 10 mm of steel liner. The storage configuration is such that a number of fuel bundles are placed inside a cylindrical steel container known as a Fuel Basket. The canisters hold up to 9 baskets each that are 304 L stainless steel, around 42” in diameter, 22” in height, and hold 60 fuel bundles each. The operating license for the dry storage canisters needs to be extended. It is desired to perform in-situ inspections of the fuel baskets to very their condition is suitable for retrieval (if necessary) and that the temperature within the fuel baskets is as predicted in the canister’s design basis. KHNP-CNL (Canadian Nuclear Lab.) has set-up the design requirements to perform the in-situ inspections in the dry storage canisters. This Design Requirements applies to the design of the dry storage canister inspection system.

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.

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.

Owing to the increase in saturation rate of the spent fuel storage pond in the Kori nuclear power plant, the interim spent fuel dry storage facility is scheduled to be constructed at the Kori site. To implement safeguards in the new dry storage facility effectively, the concept of “Safeguards-by- Design” (SBD) should be applied to reflect nuclear safeguard provisions in the earliest design stages. Detailed design information pertaining to dry storage facilities has not been determined; however, the design information related to safeguards have been inferred using case studies and interviews with nuclear power plant operators worldwide. On the basis of the results of the case studies on spent fuel dry storage facilities for light water reactors, most countries apply the metal cask method in containment buildings considering safety. Furthermore, Korean operators are also considering the same method owing to tight licensing schedules and safety issues. Using the Facility Safeguardability Assessment (FSA) methodology (one of the safeguard evaluation methodologies), the difference in design between the heavy water reactor spent fuel dry storage facility, an established IAEA safeguards approach reference nuclear facility, and the light water reactor spent fuel dry storage facility (the new nuclear facility) were analyzed. Two major differences were noted as issues pertaining to potential safeguards. First, the difference in design and transport method in terms of the difference in size and weight of the spent nuclear fuel is important; light water reactor fuel is 20 times heavier than heavy water reactor that needs partial defect inspection in assemblies. Second, the difference in safeguard approach owing to the difference between the modular storage method in heavy water reactor and the container type storage method in light water reactor must be considered; movable storage cask renders the IAEA surveillance approach difficult. The results of this study can be used to identify the safeguards requirements in advance, enabling the operator to design new dry storage facilities resulting in timely and cost-effective implementation.

Spent nuclear fuels should be safely stored until being disposed and dry storage system is predominantly used to retain the fuels. During long-term storage, there are several mechanisms that could result in the degradation of spent nuclear fuels, and the temperature is the most important parameter to predict and estimate the degradation behaviors. Therefore, thermal analysis to estimate temperatures of spent nuclear fuel and the storage system should be performed to evaluate whether the temperatures exceed safety limit. Recently, thermal hydraulic analysis with CFD codes is widely used to investigate the temperature of spent nuclear fuel in dry storage. Herein, Explicit CFD analysis model is introduced and validated by estimating the thermal hydraulic response of the dry storage system that is Dry Cask Simulator (DCS). Extended Storage Collaborating Program (ESCP) led by the Electric Power Research Institute (EPRI) is organized to assess degradation effects of spent nuclear fuel during long-term dry storage, and DCS is the first phase of the program. The dry storage system, containing a single BWR assembly in a canister, was designed to produce validation-quality data for thermal analysis model. ANSYS FLUENT is used to simulate DCS, and the test condition of 0.5 kW decay heat and 100 kPa helium pressure was investigated in this study. In case of peak cladding temperature (PCT), PCT from the experiment was 376 K while that of CFD was 374 K. It implies CFD simulation gives good agreement with experimental measurement. Peak temperatures of channel can, basket, canister and shell predicted by CFD simulation also show good prediction and the discrepancies were less than 7 K while measurements uncertainty was 7 K.

As the amount of on-site Spent Nuclear Fuel (SNF) in storage increases due to the continued operation of Nuclear Power Plants (NPPs) in Korea, the on-site wet storage pool is expected to become saturated. Therefore, a facility for safely storing the spent nuclear fuel is required so that there is no problem with operation of the NPP until permanent disposal of SNF. Prior to the construction of such a facility, the safety analysis of the interim storage facility and verification of the safety of the spent fuel storage system (e.g. cask, silo) to be used are required according to Article 63 of the Nuclear Safety Act. In this process, analysis of the Structures, Systems, and Components (SSCs) of the storage system is needed. Based on the analysis, it is necessary to efficiently classify SSCs that are important to safety in order to differentiate management that more thoroughly manages those important to safety. In Korea, according to the notice of the Nuclear Safety and Security Commission, the components performing essential safety functions for the safe storage of spent fuel storage system are to be classified as “important safety equipment”. 10 CFR Part 72, a federal regulation related to interim storage facilities in the United States, also requires the identification of SSCs that fall under “Important to Safety (ITS)”, which is like domestic case. In addition, it has been confirmed that there are cases in which detailed classification according to Reg Guide 7.10 and NUREG-CR/6407 is added in Safety Analysis Report. However, these existing classification methods are not only classified as a single grade except for the method according to the Reg guide, but all are classified according to a qualitative standard. Qualitative criteria may cause ambiguity in judgment, resulting in subjective judgment of the person who proceeds in the classification process. Therefore, in this study, a new classification method is proposed to solve the problem according to the qualitative classification method. Assessing the level of radiological harm to the general public due to the assumption of failure of SSC in the spent fuel storage system is used as a quantitative evaluation standard.

Some Spent Fuel Pools (SFPs) will be full of Spent Nuclear Fuels (SNFs) within several years. Because of this reason, building interim storage facilities or permanent disposal facilities should be considered. These storage facilities are divided into wet storage facilities and dry storage facilities. Wet storage facility is a method of storing SNF in SFP to cool decay heat and shielding radiation, and dry storage facility is a method of storing SNF in a cask and placing on the ground or storage building. However, wet storage facilities have disadvantages in that operating costs are higher than that of dry storage facilities, and additional capacity expansion is difficult. Dry storage facilities have relatively low operating costs and are relatively easy to increase capacity when additional SNFs need to be stored. For this reason, since the 1990s, the number of cases of applying dry storage facilities has been increasing even abroad. Dry storage facilities are divided into indoor storage facilities and outdoor storage facilities, and outdoor storage facilities are mostly used to take advantage of dry storage facilities. In the case of outdoor storage facilities, the cask in which SNFs are stored is placed on a designed concrete pad. During this storage, the boring heat generated by SNFs cools into natural convection and the cask shields the radiation that SNFs generates. However, if an accident such as an earthquake occurs and the cask overturns during storage, there may be a risk of radiation leakage. Such a tip-over accident may be caused by the cask slipping due to the vibration of an earthquake, or by not supporting the cask properly due to a problem in the concrete pad. Therefore, in the case of outdoor dry storage facilities, it is necessary to evaluate the seismic safety of concrete pads. In this paper, various soil conditions were applied in the seismic analysis. Soil conditions were classified according to the shear wave velocity, and the shear wave velocity was classified according to the ground classification criteria according to the general seismic design (KDS 17 10 00). The concrete pad was designed with a size that 8 casks can be arranged at regular intervals, and 11# reinforcing bars were used for the design of the internal reinforcement of the concrete pad according to literature research. The cask was designed as a rigid body to shorten the analysis time. The soil to which the elastic model was applied was designed under the concrete pad, and infinite elements were applied to the sides and bottom of the soil. The effect on the concrete pad and cask by applying a seismic wave conforming to RG 1.60 to the bottom of the soil was analyzed with a finite element model.

Due to the saturation of the on-site storage capacity of spent nuclear fuel within a few years, dry storage facility should be introduced. However, it is unclear when to start operating the dry storage facility, so in case of Kori Unit 1, which is being decommissioning, the spent fuel must be stored in the spent fuel pool of another power plant. In addition, in the case of damaged fuel, it is impossible to transfer and store it with general handling methods. Therefore, a damaged fuel canister (DFC) should be able to handle damaged or failed fuel as intact fuel, and both wet and dry storage should be possible. The canister developed by Korea Hydro & Nuclear Power is designed to satisfy criticality, shielding, cooling performance, and structural integrity in accordance with NUREG-1536 and 2215. In addition, it can be handled as existing fuel handling devices rather than new handling tools. Fastening of the DFC lid and body in the spent fuel pool is possible with a hexagonal socket wrench, one of the fuel repair tools. And it is designed to facilitate visual identification of whether it is fastenedor not. The lifting method for transferring DFC to another facility is the same as the nuclear fuel lifting method. And a unique sealing and mesh structure of the lid and body is devised to completely block leakage of nuclear fuel fragments of 0.2 mm or more during vacuum drying for dry storage. The usability of DFC has been verified through test operation of the prototype, and it will be manufactured before discharging spent fuel for the decommissioning of Kori Unit 1.

For Dry Storage of Spent Nuclear Fuel (SNF), all moisture must be removed from the dry storage canister through subjected to a drying process to ensure the long-term integrity. In NUREG-1536, the evacuation of most water contained within the canister is recommended a pressure of 0.4 kPa (3 torr) to be held in the canister for at least 30 minutes while isolated from active vacuum pumping as a measure of sufficient dryness in the canister. In the existing drying process, the determination of drying end point was determined using a dew point sensor indirectly. Various methods are being studied to quantify the moisture content remaining inside the canister. We presented a moisture quantification method using the drying process variables, like as temperature, pressure, and relative humidity operation data. During the drying process, it exists in the form of a mixed gas of water vapor and air inside the canister. At this time, if the density of water vapor in the mixed gas discharged out of the canister by the vacuum pump is known, the mass of water removed by vacuum drying can be calculated. The canister is equipped with a pressure gauge, thermometer and dew point sensor. The density of water vapor is calculated using the pressure, temperature and relative humidity of the gas obtained from these sensors. First, calculate the saturated water vapor pressure, and then calculate the humidity ratio. The humidity ratio refers to the ratio of water vapor mass to the dry air mass. After calculating the density of dry gas, multiply the density by the humidity ratio to calculate the density of water vapor (kg/m3). Multiply the water vapor density by the volume flow (m3/s) to obtain the mass value of water (kg). The calculated mass value is the mass value obtained per second since it is calculated through the flow data obtained every second, and the amount of water removed can be obtained by summing all the mass values. By comparing this value with the initial moisture content, the amount of moisture remaining inside the canister can be estimated. The validity of the calculations will be verified through an experimental test in the near future. We plan to conduct various research and development to quantify residual water, which is important to ensure the safety of the drying process for dry storage.

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.

In Korea, research on the introduction of dry storage facility is being conducted as an alternative to saturation of temporary storage facilities for spent nuclear fuel. The introduction of dry storage facilities requires a radiological impact assessment on the workers of the facility, and for this, an appropriate exposure scenario must be derived through work procedure analysis. In this study, the procedure for storing spent nuclear fuel in dry storage facilities was analyzed based on the case of evaluating the radiological impact of workers in dry storage facilities abroad. We investigated cases of radiological impact assessment on workers at on-site dry storage facilities by PNNL, Dominion, and P. F. Weck. PNNL and Dominion analyzed the storage work procedure of the VSC (Vertical Storage Cask) method using CASTOR V/21, TN-32, respectively, and conducted a radiological impact assessment. P. F. Weck analyzed the storage work procedure of various spent nuclear fuel casks for VSC and HSM (Horizontal Storage Module), conducted a radiological impact assessment. As a result of comparing the procedure for storing spent nuclear fuel by case, it was found that the storage procedure was determined by the storage method and the cask type. In the case of VSC method, canister-type casks and basket-type casks are used, and the storage procedure are partially different according to each. Canister-type cask requires repackaging from transfer overpack to storage overpack, but basket-type cask doesn’t require that procedure. In the case of the HSM method, only the canister type cask was found to be used. However, the storage procedure was different depending on the type of HSM system. Depending on the type of HSM system, the necessity of cask for on-site transport was different. In this study, we investigated and analyzed the work procedure according to the storage method of dry storage facilities abroad. It was found that the dry storage procedure of spent nuclear fuel different according to the storage method and type of cask. The results of this study can be used as basic when deriving the exposure scenario for spent nuclear fuel dry storage workers suitable for the domestic situation.