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.

To rationalize the protection of spent nuclear fuel transport storage cask, we intend to investigate the status of domestic and foreign safety regulations and related technologies to develop sabotage scenarios and analyze the protection performance and radiation impact of transport storage cask. It is essential to conduct an aircraft collision safety evaluation on spent nuclear fuel transportation and storage casks in Korea due to changes in laws and regulations related to nuclear power plant design and demand for enhanced safety. Domestic and foreign research on the protection performance of spent nuclear fuel transport storage cask was based on 9.11 events, and the results of all studies show that the speed of the aircraft and leakage of nuclear materials are insignificant. The Sandia National Laboratory (SNL) calculates Aerosol emissions from spent fuel damage in the event of sabotage and calculates Source Term based on the Durbin-Luna model. In this paper, radiation sensitivity analysis was performed due to damage to the carrier according to the size of the accident, assuming that there was a hole enough to basket from the external shell among the collision scenarios identified for domestic cask models.

Since July 2021, the Korea Radioactive Waste Agency has been conducting a safety case development study for the Korean deep geological repository program. The safety case includes generating scenarios in which radioactive materials from spent nuclear fuel repository reach the human biosphere by combining selective FEPs (Features, Events, and Processes). This safety case should be able to transparently explain the process in which conclusions have been drawn not only to stakeholders but also to the public by presenting safety arguments. The scenario development stage consisting of FEP screening, scenario generation, and uncertainty analysis procedures should have a database management system. Database management system was performed in countries such as Sweden, which obtained approval for the construction of spent nuclear fuel repositories, and the United States, where various preliminary research was carried out. Korea Atomic Energy Research Institute also has experience in designing and operating its own database, which has conducted preliminary research on disposal of the spent nuclear fuel. Currently, the safety assessment of the Korean spent nuclear fuel repository is in the early stages of research, but it is necessary to set up a basic framework for database design while the collection of FEP data from domestic and international preliminary studies is under development, and it is advantageous for efficient database construction and operation. Therefore, this paper presents the current status of database design considering completeness and transparency from the FEP screening stage to the scenario development stage in the safety assessment process of the Korean spent nuclear fuel repository. In this process, the functional requirements that the database should provide, the database schema capable of implementing them, and simple examples are presented together. The objectives of this database design are flexible FEPs management, high integrity and consistency, and expandability for linking with the safety case database. The FEP data to be inputted into the database includes a list of major opened FEPs, including International FEPs from Nuclear Energy Agency, which were referred for PFEPs (Project-specific FEPs), and PFEPs applied to POSIVA's Olkiluoto repository. As an additional function, queries from the database are used to visually express the process of deriving scenarios through Rock Engineering System, a widely known scenario generation methodology.

PWR spent nuclear fuel generally showed an oxide film thickness of 100 um or more with a combustion rate of 45 MWD/MTU or higher, while CANDU spent nuclear fuel with an average combustion rate of about 7.8 MWD/MTU had few issues related to hydride corrosion. Even based on the actual power plant data, it is known that the thickness of the oxide film is 10 μm or less on the surface of the coating tube, and brittleness caused by hydride is shown from the thickness of the oxide film of about 80 μm, so it is not worth considering. However, since corrosion may be accelerated by lithium ions, lithium ions may be said to be a very important factor in controlling the hydro-chemical environment of heavy water. Lithium has a negative effect on the corrosion of zirconium alloys. However, since local below 5 ppb to prevent corrosion. maintained at a concentration between 0.35 and 0.55 ppm. Hydrogen is known to have a positive effect by suppressing radioactive decomposition of the coolant and suppressing cracks in nickelbased alloys. However, too much hydrogen can produce hydride in a pressure tube composed of Zr-2.5Nb, so DH (Disolved Hydrogen) maintains the range of 0.27–0.90 ppm. pH and conductivity are completely determined by lithium ions, and DH can be completely removed below 5 ppb to prevent corrosion. Therefore, for cladding corrosion simulation of the CANDU spent nuclear fuel, a hydrochemical of the equipment, not 310°C, and 14 uS·Cm−1 is targeted as conditions for corrosion acceleration. In addition, for acceleration, the temperature was set to 345°C (margin 10°C), which is the maximum accommodation range of the equipment, not 310°C.

Korea Atomic Energy Research Institute (KAERI) has investigated Pyroprocessing technology in order to decrease the burden of disposal system and increase availability of useful radionuclides in the spent nuclear fuel (SNF) for future. The treatment and the disposal of SNF, however, are very sensitive issues socially. In addition, under the energy transition policy phasing out nuclear energy gradually there have been demands for alternatives so far. Thus various alternatives should need to be investigated in preparation for unexpected situations. This study has been conducted roughly in effectiveness point of view of alternative pre-managements for SNF, not pyroprocessing technology, in disposal system, consisting of three stages according to the degree of burden in disposal system. Stage I is the case for making safety increase with removing highly-mobile radionuclides from SNF. Stage II is the case for eliminating high-heat radionuclides additionally, alleviating thermal risk in the disposal system. And Stage III is the case for recovering Uranium in addition to Stage II. These options of pre-management are thought to be able to provide an intuitive strategy for effective diversification of the disposal system. Because several types of waste form from pre-management make it possible to develop the effective, newly-composed waste disposal system according to the properties of radionuclides. And the processability of SNF through pre-management might be combination with available core-drilling technology, being able to design various disposal system as well. Even though the whole, detailed unit processes have not designed yet, mass balance and distributions of radionuclides are performed under the appropriate assumption of engineering processes. As a first step the alternative approaches for SNF pre-management for disposal system might be expected to be widely used in implementing SNF management policy in the future.

An accumulation of spent nuclear fuel (SNF) has brought a considerable interest due to its energy and environmental issue. To effectively manage SNF, a pyroprocessing is introduced to separate useful resources from the spent fuels and to manufacture suitable fuels. In head-end process of pyroprocessing, spent fuels are thermally treated to prepare UO2 pellets, where various radioactive gases from SNFs are released during thermal treatment. Within these gases, C-14 as CO2 form is a radioactive fission product which had a long half-life of 5,730 years and emits beta radiation of 0.156 MeV. Generally, current CO2 capturing technologies include adsorption by solid materials, absorption by aqueous solutions, and membrane separation. Among these methods, absorption is an effective approach which traps CO2 effectively and and it is easy to operate at room temperature. In addition, it is highly recommended as immobilizing 14CO2 as CaCO3 formation due to the high thermal and chemical stability, and the relatively low solubility in water. Generally, a double alkali method has been proposed to capture low concentrated 14CO2 from the stream. This method for CO2 capture includes absorption process with NaOH solution and causticization using Ca(OH)2. In this study, CO2 emitted from SNF is captured using double alkali method, and the effects of operating conditions on capturing efficiency were investigated. Furthermore, considering the two-film theory, the effects of trapping conditions on the CO2 absorption performance were examined. The recovered CaCO3 from causticization was collected from the absorbing solution and analyzed.

The manufactured nuclear fuel assembly is loaded into the nuclear reactor after the core design, and is finally discharged to the wet storage pool after depletion for 3 cycles. The discharged spent nuclear fuel is transported and stored in a dry storage system at the on-site of the nuclear power plant, which is cooled by natural convection, and undergoes final disposal or reprocessing through an intermediate dry storage facility. In this series of processes, the characteristics of the final product, the spent fuel, vary depending on the environmental conditions, so it is essential to manage each history data to verify the long-term integrity of the spent nuclear fuel. In this paper, safety information on spent nuclear fuel is described in order to establish technical requirements that should be considered in each stage of storage, transport, reprocessing, and disposal of spent nuclear fuel. Comprehensive safety information on spent nuclear fuel is basically calculated from basic information that considers characteristic information that can be obtained through the manufacture and design of nuclear fuel assemblies, operation history in a nuclear reactor, and location history in a wet storage pool. It can be divided into secondary production information (SF Burnup, Nuclide Inventory, etc.) and tertiary integrity-related information obtained through cladding inspection during spent fuel storage. KHNP produces this multi-layered information according to the production stage and manages it through the comprehensive management system of the spent nuclear fuel, and safety information with some errors is not only improved through re-verification but also continuously updated. In this paper, the spent nuclear fuel safety information was derived based on various information calculated in the entire process of being discharged and managed in a wet storage pool, including new fuel manufacturing information and depletion history. Such safety information will be used as basic data for long-term safe management of spent nuclear fuel, and will be continuously produced and managed. In the future, additional discussions will be held on the safety information of the spent nuclear fuel through consultation with KORAD and regulatory agencies.

Currently, the development of evaluation technology for vibration and shock loads transmitted to spent nuclear fuel and structural integrity of spent nuclear fuel under normal conditions of transport is progressing in Korea by the present authors. Road transportation tests using surrogate spent nuclear fuel were performed in September, 2020 using a test model of KORAD-21 transportation cask and sea transportation tests were conducted from September 30 to October 4, 2021. In order to investigate amplification or attenuation characteristics, according to the load transfer path, a number of accelerometers were attached on a ship cargo hold, cradle, cask, canister, disk assembly, basket, and surrogate fuel assemblies and to investigate the durability of spent nuclear fuel rods, strain gages were attached on surrogate fuel assemblies. A ship named “JW STELLA” which has similar deadweight (5,000 ton) of existing spent nuclear fuel transportation ships was used for the sea transportation tests. The ship is propelled by 1,825 hp two main engines with two 4-bladed propellers. There are two major vibration sources in the ship. One is the vibration from waves and the other is the vibration from the engine and propeller system. The sensor locations on the ship were determined considering the vibration sources. The sea transportation test was performed for 5 days, the test data were measured successfully. The ship with the test model was departed from Changwon and sailed to Uljin, sailed west to Yeonggwang and then returned to Changwon. In addition to sailing on a designated test route, circulation test, braking/acceleration test, depth of water test, and rolling test were conducted. As a result of the preliminary data analysis of the sea test, power spectral densities and shock response spectrums were obtained according to the different test conditions. The vibratory loads caused by the wave mainly occurred in the frequency range of 0.1 to 0.3 Hz. The vibratory loads caused by the propeller occurred near the n/rev rotating frequencies, such as 5, 10, 20 Hz etc. However, those frequencies are far from the natural frequencies of local mode of the fuel rods, so it is considered that the vibratory loads from the wave and the propeller do not have a significant influence on the structural integrity of the fuel rods. Among all the test cases, maximum strain occurred at SG31 near the bottom nozzle on the test; the magnitude was 73.62 micro strain. Based on the analyzed road and sea transportation test data, a few input spectra for the shaker table test will be obtained and the shaker table test will be conducted in 2022. It is expected that the detailed vibration characteristics of the assembly which were difficult to identify from the test results can be investigated.

The amount of temporarily stored spent nuclear fuel in South Korea will be reaching saturation in a near future. Therefore, it is an urgent issue to construct a spent nuclear fuel storage system. In order to construct the storage system, some coastal environmental characteristics such as temperature, pH, and chemical composition of sea water in South Korea have to be evaluated and predicted because they can affect in deterioration of the storage system. However, in South Korea, the coastal environmental characteristics of area where the storage system is likely to be built are not well established until now. In this study, a time-series deep-learning algorithm is developed using the Long-Short Term Memory (LSTM) algorithm to predict and evaluate the coastal environmental characteristics based on the wellestablished data from Korea Meteorological Administration (KMA) and Ministry of Oceans and Fisheries (MOF). As a result, by developing the predictive model to evaluate the coastal environmental characteristics, we intend to apply it for site evaluation to construct the spent nuclear fuel storage system or many other applications related to the nuclear as well.

This presentation summarizes recent research on estimating the mechanical loading environment of spent nuclear fuel (SNF) during normal storage and transportation scenarios sponsored by the US Department of Energy Spent Fuel and Waste Science and Technology (SFWST) program. Normal conditions of truck, ship, and railroad transportation of SNF were studied with testing and numerical modeling to determine that the shock and vibration loads applied to SNF during transportation are not expected to challenge SNF cladding integrity or the fatigue life of cladding. The 30 cm package drop scenario was studied with experiments and modeling to determine that mechanical loads during a 30 cm SNF package drop scenario are only expected to challenge SNF cladding integrity under worstcase conditions at elevated temperatures. The SFWST program is currently preparing seismic shake table testing to record SNF mechanical loads in a dry storage earthquake scenario. This presentation summarizes the findings of the transportation and package drop research and details the progress made on the current seismic test.

The purpose of this study is to develop the analysis procedures for the evaluation of the structural integrity of the spent fuel in normal condition of transport at sea. Spent nuclear fuel must be transported from the wet storage facility in the nuclear power plant to the intermediate storage facility, and the structural integrity must be maintained in vibration and shock loads during the transportation. In general, the transport of spent nuclear fuel is performed in three kinds of modes: road, rail, and sea. During transport, the spent nuclear fuel is subjected to repeated vibration and shock loads by road surfaces, railroad tracks, and waves of the sea. It should be evaluated whether the structural integrity of the spent fuel is maintained under these load conditions. All nuclear power plants in Korea are located in coastal sites, and the interim storage facility for spent nuclear fuel is highly likely to be decided as a coastal site as well. Therefore, the main mode of the spent nuclear fuel transport is expected to be maritime transport by ships. In this study, the analysis procedure was developed to evaluate the safety of spent fuel at maritime transport by ships, and the procedure for evaluating the integrity of spent fuel under normal conditions of maritime transport were proposed. CFD analysis using SeaFEM was performed for the vibration analysis of the ship by waves, and the structural vibration analysis of the transport system was simulated using the developed in-house codes. The fatigue durability of the cladding was also evaluated using the developed fatigue analysis program and the fatigue analysis used the strain data obtained from the structural analysis. It was concluded that the value of the fatigue damage on the spent fuel cladding during normal conditions of maritime transportation is close to “0” and the structural integrity of the spent fuel is maintained in the same condition.

Sandia National Laboratories is the lead laboratory for the United States Department of Energy for the research and development (R&D) efforts to support the technical basis for the long-term storage, subsequent transportation, and permanent disposal of commercial spent nuclear fuel and high-level waste. Sandia does not design nuclear facilities; Sandia performs R&D to help ensure facilities and the fuel cycle are safe, sustainable, and secure. This talk will focus on the spent fuel storage and transportation programs that contribute to this work. The goal in spent fuel storage and transportation R&D is to understand the mechanical integrity of the fuel, cladding, and storage system beyond interim storage and into disposal time frames. Our research is focused on understanding the high burn-up cladding integrity over time, understanding the thermal behavior during drying and storage, understanding potential cladding oxidation pathways, and quantifying in the external loads experienced during transportation, handling, and seismic events. Additionally, this work includes extensive work to understand the basic science of canister stress corrosion cracking and the potential consequences of a through wall canister crack.

Dry storage cask facilities are considered for temporary storage of spent nuclear fuels before their final disposal. According to relevant domestic laws and regulations, the integrity and gross defects of the PWR spent fuel must be inspected before they are transferred to the dry cask from a wet storage pool of a nuclear power plant. To meet nuclear safeguards requirements for a spent fuel transportation, the KINAC has been working to develop a simple and convenient Non-destructive Testing (NDT) equipment to verify the integrity and gross defects of the spent fuel assembly. This study was conducted in two processes. The first stage is to review the current NDT techniques conducted in the nuclear fuel manufacturing process. During the manufacturing process, the Ultrasonic testing (UT) and Eddy Current Testing (ECT) technique are used for detecting the cracks or foreign materials in a cladding of a fresh fuel. During an over-haul period after an end of one fuel cycle, the sipping test of the spent fuel is performed for detecting the failed fuel assemblies. If it is determined through the sipping test whether any fuel assembly contains a failed fuel rod, the failed fuel rod of lots of fuel rods in the assembly is found out using the UT instrument. The ECT is used for detecting the internal defects and oxide layer thickness of a fuel cladding. Because the UT and ECT are the wellknown technique and has already been employing for the spent fuel inspection, we adopted the UT and ECT technique for development of a new instrument for nuclear safeguards verification. The second stage is to design the UT and ECT equipment in consideration of nuclear safeguards activities in the spent fuel pool. For nuclear safeguards inspection, irradiated fuel or non-fuel items are distinguished. Thus, verification equipment newly designed using the UT and ECT should detect not only a failed rod, but also a false tube, or a false rod, or a different material from a cladding. New probe and signal processing methods are developed to achieve these goals. The design of UT and ECT probes are preferentially carried out according to technical requirements – the probe thickness including a damper material should be less than 1.0 mm - and the study on analyzing signal distortion caused from material difference will be conducted for development of the safeguards inspection equipment. Detailed results of our study will be discussed in this conference.

Since the commercial operation of Kori unit 1 in 1978, nuclear power has provided cheap, stable and clean electricity in South-Korea. For decades, the discussion about the spent fuel management has been dominated and the government is responsible for on-going research and development (R&D) related to long-term spent nuclear fuel management. The effective management of spent fuel should be applied from the early stage of the R&D process to licensing phases with the step-by-step evaluation system. As part of follow-up efforts after the Fukushima nuclear accident, the Nuclear Promotion Commission and Nuclear Safety Commission were divided in function as an independent agency for enhancing national nuclear safety and security, which aims to protect the public and environment from undue radiological hazard. The national spent fuel project must have a vibrant program for spent fuel management. Due to the nature of these projects, the establishment of a ‘conformity assessment’ system that collects the opinions from the licensing organisations on the results of research projects from the initial R&D stage should be applied advertently in order to efficiently conduct research projects and enhance public confidence. For the government-led project for spent nuclear fuel management, the adequacy and applicability of its technology R&D as well as its sustainability that includes financial, social and environmental performance measures should be evaluated in each stage. The institutionalisation arrangement, so called ‘conformity assessment system’ for the development of a national spent nuclear fuel management plan and related technology should be developed. This study aims to propose the basic principles for the introduction of the conformity assessment system: (1) national management responsibility, (2) spent fuel management project scope, (3) its management main principles, (4) project implementation system, (5) final management project scope and securing financial resources.

The ROK conducts several export procedures, communications in connection with transfers; exchange of information on export plan, shipments, and receipt of nuclear materials, in accordance with bilateral Nuclear Cooperation Agreements (NCA) and Administrative Arrangements (AA) signed with US, Canada, and Australia. Also, the inventory amount of items subject to NCA has reported annually. This study reviewed the export procedures and management methods for spent nuclear fuel subject to NCA. The re-transfer procedures start with obtaining consent from the original exporting country. It is impossible to retransfer nuclear material without consent, whether long-term or individual case-bycase. If the material has multiple obligations, prior consent from all of those countries is required. Therefore, it is necessary to clarify the foreign obligated materials correctly. In general, nuclear fuel is subject to multiple obligations of all countries through which the materials have passed during the front-end fuel cycle. Then the new obligations are imposed on those irradiated materials or their by-products after ‘used-in’ or ‘produced through the use of ’ equipment subject to NCA. For example, fuel assemblies manufactured under CANDU fuel fabrication equipment subject to ROK-Canada NCA or burned in nuclear reactors where US equipment is installed have obligations based on Canada or US agreements. In order to impose obligation to irradiated materials, the principle of proportionality is applied as stipulated in each Agreement. According to the AA between US and ROK, nuclear materials used in the equipment transferred under the Agreement and produced through them are differently controlled. After the cycle in the reactor with US-made equipment, uranium in the irradiated fuel is considered a material used in the equipment. So it would be appropriate to apply obligation proportionality according to its origin, regardless the US-made equipment. Meanwhile, the obligation under US NCA is given to the entire amount of produced plutonium in the irradiated fuel. Although the contribution to the production of fuel is to be discussed case-by-case basis in the case of Canadian obligation, applying a similar method is proper. Since the fuel is burned in the form of bundles or assemblies, it is impossible to separate the spent fuel into uranium and plutonium physically. However, as discussed above, to clarify the rights and obligations pursuant to Agreement and ensure accuracy in inventory management, the obligation codes should be imposed on irradiated fuel as not a single item but separated individual substance of materials. Moreover, when an obligation swap occurs for the irradiated fuel, its movement and combustion history should be considered to prevent confusion in confirming multiple obligations and implementing export procedure.

This paper aims to evaluate the mechanical integrity for Spent Nuclear Fuel (SNF) cladding under lateral loads during transportation. The evaluation process requires a conservative consideration of the degradation conditions of SNF cladding, especially the hydride effect, which reduces the ductility of the cladding. The dynamic forces occurring during the drop event are pinch force, axial force and bending moment. Among those forces, axial force and bending moment can induce transverse tearing of cladding. Our assessment of 14 × 14 PWR SNF was performed using finite element analysis considering SNF characteristics. We also considered the probabilistic procedures with a Monte Carlo method and a reliability evaluation. The evaluation results revealed that there was no probability of damage under normal conditions, and that under accident conditions the probability was small for transverse failure mode.