The evaluation of the damage ratio of spent nuclear fuel is a very important intermediate variable for dry storage risk assessment, which requires an interdisciplinary and comprehensive investigation. It is known that the pinch load applied to the cladding can leaded to Mode-3 failure and the cladding becomes more vulnerable to this failure mode with the existence of radial hydrides and other forms of mechanical defects. In this study, the failure resistance of Zircaloy-4 cladding against the pinch load is investigated using numerical simulations assuming the existence of radial hydrides. The simulation model is based on the microscopic images of cladding. A pixel-based finite element model was created by separating the Zircaloy-4 and hydride using the image segmentation method. The image segmentation method uses a morphology operation basis, which is a preprocessing method through erosion operation after image expansion to enable normal segmentation by emphasizing pixels corresponding to hydrides. The segmented images are converted into a finite element model by assigning node and element numbers together with corresponding material properties. Using the generated hydride cladding finite element model, several numerical methods are investigated to simulate crack propagation and cladding failure under pinch load. Using extended finite element (XFEM) models the initiation and propagation of a discrete crack along an arbitrary, solution-dependent path can be simulated without the requirement of remeshing. The applicability of fracture mechanical parameters such as stress intensity, J-integral was also investigated.

The fabrication of waste forms with high thermal and structural stability is an essential technology for the safe management and disposal of radioactive wastes. In particular, the thermal characteristics of waste forms containing high heat-generating nuclides such as Cs and Sr can be used for the optimized design of the waste form to secure its thermal safety, and they also provide basic design data for the safe management of canisters, storage systems, and repositories. The Korea Atomic Energy Research Institute is actively developing processes and equipment for fabricating various types of high-level wastes into a stable glass or ceramic waste form. In previous research related to the thermal analysis of the waste form, a relatively simple analysis was performed by using the analytic solution of the one-dimensional steady-state heat conduction equation considering the decay heat properties of the waste. As a specific application study, the optimized diameter of the cylindrical glass waste form was proposed by evaluating the centerline temperature of the waste form. In this study, we extended previous research by introducing a more complicated model, and the main results are summarized as follows. First, an analytical solution was derived by applying the temperaturedependent thermal conductivity expressed in the general form of polynomial function to the onedimensional heat conduction problem previously studied. Second, the two-dimensional axisymmetric steady-state heat conduction problem with a more realistic cylinder model with finite length was modeled and solved by using the finite element method via Matlab’s PDE (partial differential equation) toolbox. Third, thermal analysis was performed on the SrTiO3 waste form, selected as a stable form of strontium nuclide, using the developed analytical and numerical methods. The differences in the temperature distribution and computation time were evaluated through a comparative study of both solutions. Although the problem considered in this study could easily be solved by using commercial CFD software such as ANSYS or SolidWorks, a code-based program was developed to facilitate parametric design study in conjunction with optimization algorithms. The analysis results could be used to evaluate the thermal stability of waste form and to optimize the shape and size of the waste form in consideration of the design constraints of storage systems or repositories.

Strontium-90 is a high heat-generating nuclide in spent nuclear fuel. The removal of the nuclide separation is indispensable to reduce the burden of storage and disposal of high-level radioactive waste. Korea Atomic Energy Research Institute has developed the molten salt immersion technique to separate the strontium by the chlorination of the strontium oxide in molten salt. It is needed to separate the salt for the recovery of strontium from the salt solution after the chlorination reaction. In this study, it was investigated on the recovery of the strontium from the salt. Vacuum distillation was used for the separation of strontium from the molten salt. The vapor pressures of the candidate salts were calculated by HSC chemistry and the apparent evaporation rates (AER) were measured at 830°C to evaluate the salts for strontium recovery. The candidate salts were LiCl, KCl, MgCl2, NaCl and CaCl2. The AERs of MgCl2 and NaCl were 1.9 and 1.3 g/cm2-h, respectively. Those two salts can be separated from the strontium compound even though the AER values are much lower than those of LiCl-KCl (~ 8 g/cm2-h). CaCl2 salt was rarely evaporated (AER < 0.03 g/cm2-h) and it is not suitable to use as a strontium recovery salt. Therefore, MgCl2, NaCl, LiCl and KCl can be regarded as candidates for a strontium recovery salt.

The origin of Fe oxide deposition on zirconium oxide with UV irradiation has been investigated in this study. After 7 day corrosion in the flowing autoclave, Fe based oxide is formed on the zirconium oxidewith UV irradiation at 260°C, 6 MPa DI water. Zircaloy-4 coupon is irradiated with a 200 mW·cm−2 UV, and the dissolved oxygen level is maintained below 100 ppb, and dissolved hydrogen concentration is maintained as 2.5 ppm. Zircaloy-4 coupon supplied from Westinghouse is used for this study. MULTEQ version 4.0 developed by EPRI is adopted to simulate how ions dissolved in water can generate deposits on the zirconium oxide with UV irradiation. ICP-OES data after 30 d corrosion in the flowing loop experiment is used for input file for MULTEQ simulation. The system temperature is set as 260°C, and 2,592 L of water is considered the total amount water into the autoclave (0.06 mL·min−1, 30 d). Total numbers of simulation run is set as 8, and the system pH at 260°C is 6.06. Oxidation potential after run #8 is −0.44 V. From MULTEQ simulation, most Fe is existed as Fe(OH)3 and Fe(OH)2, and Fe ions can also exist, but no Fe metal observed. 5.09 × 10−6 ppm (9.73 ppb) of Fe2+, 2.81 × 10−6 ppm FeOH+, and 3.77 × 10−9 ppm Fe(OH)3are in the system. It can be concluded Fe is existed as ion or hydroxide form in the solution. Two precipitates are found from MULTEQ simulation, First, NiO(s) = 5.21 × 10−5 g (52.1 μg), NiFe2O4 = 8.06 × 10−5 g (80.6 μg), and still they are negligible amount. The total concentration of Fe in the electrolyte is the summation of each Fe species concentration and it is equal to 2.69×10−4 ppm. This value is equivalent to 0.269 μg·kg−1 in the solution. The total water volume of the 30 d experiment is 2,592 L (considering water flow from high-pressure pump), so the amount of Fe from ICP-OES data and MULTEQ results in 2,592 L electrolyte is 697.2 μg. This value is order of magnitudes higher than the mass of Fe from the deposits, which was already an upper estimate based on the assumptions. This clearly shows that Fe ions dissolved in the electrolyte can be the source of Fe3O4 on Zr oxide during corrosion with UV irradiation.

In general, if a nuclear fuel cladding tube is damaged during reactor operation, it is called fuel failure. If the cladding tube is damaged, the function of sealing the nuclear fuel material is lost, and the fission products accumulated inside the nuclear fuel rod may leak into the coolant. The causes are the most damage caused by foreign substances in a coolant such as small iron wires, and GTRF (Gridto- Rod Wear) due to a grid, end-plug welding defect, PCMI (pellet cladding mechanical interaction), and oxidation corrosion damage. In this study, a device of simulating friction damage and debris induced damage between grid-fuel rods, which are the main causes of cladding tube damage, was developed. An air vibrator was installed as a function to induce vibration of the nuclear fuel rod. Sandpaper was installed between the grid and the fuel rod to induce friction between the grid-fuel rods. Saw teeth were installed on the grid to induce damage to foreign substances. It is believed that the simulated damaged nuclear fuel rod can be manufactured through on-study to provide the simulated damaged nuclear fuel rod necessary for the stabilization study of the damaged nuclear fuel rod.

NFDC (Nuclear Fuel and materials Data Center) developed standard reference data for oxidation of HANA-6 cladding material. Thermo-gravimetric analyzer (TGA) was used to measure oxidation, and the measuring device was self-calibrated using standard materials. The oxidation amount of the HANA6 cladding was measured in an oxidizing atmosphere in the temperature range of 400 to 700°C. Through this, oxidation data, oxidation rate model equation, and graph were developed. The uncertainty factors were analyzed from the oxidation model. The expanded uncertainty of oxidation data was calculated by evaluating the uncertainty for each uncertainty factor. The oxidation data produced in this study was self-rated through deliberation by a specialized committee of NFDC and third experts. It was finally registered as a reference standard through the technical committee of the National Reference Standards Center. It is believed that the standard reference data developed in this study will be helpful for increasing reliability and stability evaluation of nuclear fuel and spent fuel.

National Standard Reference Center from Ministry of Trade Industry and Energy at Dec.30 2008. The fields of designation were nuclear fuel and energy materials. NFDC produces standard reference data of nuclear fuel and materials. To ensure reliability of experimental data uncertainty should be estimated. There are two kinds of uncertainty: A-type uncertainty from tester and B-type uncertainty from experimental equipments. To reduce the former, the measurement should be repeated for sufficient amount of times, and to reduce the latter type uncertainty all equipment have to be calibrated. In this study the uncertainty evaluation process of thermo-gravimetric analyzer (TGA) data was developed. The self calibration was performed using the standard mass and correction factor was obtained. The measurement model of oxidation was established, factors affected to uncertainty was analyzed, uncertainty of each factor using sensitivity coefficient was evaluated, combined uncertainty was calculated, and expanded uncertainty using coverage factor was calculated. It is believed that the uncertainty evaluation process of TGA data developed in this study will be helpful for increasing reliability and stability evaluation of nuclear fuel and spent fuel.

Currently, the HI-STAR 63 transport cask, developed to transport CANDU spent nuclear fuel from the wet storage pool to the dry storage facility which is called the MACSTOR/KN-400, has a transport capacity of 120 bundles, which is unfavorable when considering transportation costs and other related aspects. According to the ‘Basic Plan for High-Level Radioactive Waste Management (draft)’, the total amount of CANDU spent nuclear fuel is expected to be approximately 660,000 bundles. To safely and efficiently transport this amount to interim storage facilities, it is essential to develop a large-capacity transport cask. Therefore, we have been developing a large-capacity PHWR spent nuclear fuel transport cask, called the KTC-360 transport cask. According to the transport-cask related regulations, the KTC-360 transport cask was classified as a Type B package, and such packages need to maintain integrity under the normal transport and accident conditions described in these regulations. To prove the thermal integrity of this cask under the normal transport and accident conditions, high-temperature and fire tests were performed using a one-third slice model of an actual KTC-360 cask. The results revealed that the surface temperature of the cask was 62°C, indicating that such casks need to be transported exclusively. The highest temperature of the CANDU spent nuclear fuel was predicted to be lower than the melting temperature of Zircaloy-4, which was the sheath material used. Therefore, if normal operating conditions are applied, the thermal integrity of a KTC- 360 cask could be maintained under normal transport conditions. The fire test revealed that the maximum temperatures of the structural materials, stainless steel, and carbon steel, were 446°C lower than the permitted maximum temperatures, proving the thermal integrity of the cask under fireaccident conditions.

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.

One of the options for spent fuel dry storage systems is to store them in canisters using metal or concrete casks close to shore. The interaction between the austenitic stainless steel and the chloride atmosphere generated from the sea creates detrimental conditions leading to chloride induced stress corrosion cracking (CISCC) in the canister. The corrosion integrity of the canister in the concrete cask is very important because the canister is sealed and used for a long period of time. A canister made of austenitic stainless steel has several welding lines on the wall and lid, which are generated during the welding process and have high residual tensile stress. The interaction between the austenitic stainless steel and the chloride atmosphere generated from the sea creates detrimental conditions leading to chloride induced stress corrosion cracking (CISCC) in the canister. The corrosion integrity of the canister in the concrete cask is very important because the canister is sealed and used for a long period of time. In order to evaluate such soundness, an accelerated test capable of simulating the CISCC crack propagation phenomenon of the canister weld is required. In this study, the current status of CISCC simulation tests performed around the world to build a test equipment for the CISCC simulation accelerated test is investigated, and based on this, the test conditions suitable for the simulation test and specimen specifications are selected to establish the test equipment. The settings were performed. In consideration of the set device requirements, the essential limiting conditions for device manufacturing were derived, and detailed design was performed to satisfy them, and it was used to build a CISCC simulation test device for welding parts. The CISCC simulation test equipment requires performance to maintain the test temperature range of room temperature to 80°C and humidity 35 to 95%. In addition, it should be manufactured in consideration of humidity and temperature maintenance in the chamber of the complex corrosion tester, measures to prevent leakage of the connection part between the chamber and the salt water tank of the complex corrosion tester, and measures to supply stable salt water and maintain temperature in the salt water tank. Based on these contents, detailed specifications and design contents of the chloride stress corrosion cracking simulation test apparatus were presented in this study.

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.

Concrete structures of spent nuclear fuel interim storage facility should maintain their shielding ability 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 its unique functions (shielding 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 of the concrete are required to evaluate the structural integrity of the concrete structures. Unlike the calculated damage results for the static deformation of the concrete structure, it is very difficult to accurately estimate the damage values of the degraded concrete structures where an aircraft collides at a high strain rate. Therefore, the present authors have a plan to establish a database of the dynamic material properties of deteriorated concrete and implement to a Finite Element Analysis model. Prior to that, dynamic increase factors described in a few technical specifications were investigated. The dynamic increase factor represents the ratio of the dynamic to static strength and is normally reported as function of strain rate. In ACI-349, only the strain rate is used as a variable in the empirical formula obtained from the test results of specified concrete strengths of 28 to 42 MPa. The maximum value of dynamic increase factor is limited to 1.25 in the axial direction and 1.10 in the shear direction. On the other hand, in the case of the CEB model, static strength is included as variables in addition to the strain rate, and a constitutive equation in which the slope changes from the strain rate of 30 /s is proposed. As plotting the two dynamic increase factor models, in the case of ACI, it is drawn as a single line, but in the case of CEB, it is plotted as multiple lines depending on the static strength. The test methods and specimen sizes of the previously performed tests, which measured the concrete dynamic properties, were also investigated. When the strain rate is less than 10 /s, hydraulic or drop hammer machines were generally used and the length of the specimens was more than twice the diameter in most cases. However, in the case of Split Hopkinson Pressure Bar tests, the small size specimens are preferred to minimize the inertia effect, so the specimens were small and the length was less than twice the diameter. We will construct the dynamic properties DB with our planned deteriorate concrete specimen test, and also include the dynamic property data already built in the previous studies.

To estimate the removal efficiency of TRU and rare earth elements in an oxide spent fuel, basic dissolution experiments were performed for the reaction of rare earth elements from the prepared simfuel with chlorination reagents in LiCl-KCl molten salt. Based on the literature survey, NH4Cl, UCl3, and ZrCl4 were selected as chlorination reagent. CeO2 and Gd2O3 powders were mixed with uranium oxide as a representative material of rare earth elements. Simfuel pellets were prepared through molding and sintering processes, and mechanically pulverized to a powder form. The experiments for the reaction of the simfuel powder and chlorination reagents were carried out in a LiCl-KCl molten salt at 500°C. To observe the dissolution behavior of rare earth elements, molten salt samples were collected before and after the reactions, and concentration analysis was performed using ICP. After the reaction completed, the remaining oxide was washed with water and separated from the molten salt, and XRD was used for structural analysis. As a result of salt concentration analysis, the dissolution performance of rare earth elements was confirmed in the reaction experiments of all chlorination reagents. In an experiment using NH4Cl and ZrCl4, the uranium concentration in the molten salt was also measured. In other words, it seemed that not only rare elements but also uranium oxide, which is a main component of simfuel, was dissolved. Therefore, it is thought that the dissolution of rare earth elements is also possible due to the collapse of the uranium oxide structure of the solid powder and the reaction with the oxide of rare earth elements exposed to molten salt. As a result of analyzing the concentration changes of Simfuel before and after each reaction, there was little loss of uranium and rare earth elements (Ce/Gd) in the NH4Cl experiment, but a significant amount of rare earth elements were found to be reduced in the UCl3 experiment, and a large amount of rare earth elements were reduced in the ZrCl4 reaction.

A long-term cooling effect on hydride reorientation of a cladding tube can affect the integrity of spent nuclear fuel transportation and long-term storage. In this study, experimental setup for investigating the degree of radial reorientation of hydrides in the circumferential direction during the long-term cooling was established. The experimental setup was designed to be simplified since the long-term evaluation requires a long term period such as 12, 18 and 24 months when the cladding tube specimen is gradually cooled down from 400°C to 100°C. For the test, hydrogen-charged specimens of 100 ppm, 200 ppm, and 500 ppm were prepared. The specimen was sealed with fixtures and check valve, and was pressurized up to 90 Mpa. To heat the specimen, a box-type furnace was used while the temperature of the specimen was measured from thermocouples attached to the specimen. After the heat treatment, the long-term cooling was performed by developing temperature control program to investigate several cooling rate conditions of the specimen. As a reference case, microstructure and brittle property of the hydrogen-charged specimens of 100 ppm, 200 ppm, and 500 ppm without the long-term cooling was observed. In the case of the hydrogen content, it was uniformly distributed in circumferential direction although it was non-uniform in the axial direction. In the case of the brittle property, a compression test was performed. For the future work, the microstructure and brittle property of the hydrogencharged specimens after the several long-cooling conditions were investigated. Then, the degree of radial reorientation of hydrides in the circumferential direction during the long-term cooling was studied.

Prior to the investigations on fuel degradation it is necessary to describe the reference characteristics of the spent fuel. It establishes the initial condition of the reference fuel bundle at the start of dry storage. In a few technology areas, CANDU fuels have not yet developed comprehensive analysis tools anywhere near the levels in the LWR industry. This requires significantly improved computer codes for CANDU fuel design. In KNF, in-house fuel performance code was developed to predict the overall behavior of a fuel rod under normal operating conditions. It includes the analysis modules to predict temperature, pellet cracking and deformation, clad stress and strain at the mid-plane of the pellet and pellet-pellet interfaces, fission gas release and internal gas pressure. The main focus of the code is to provide information on initial conditions prior to dry storage, such as fission gas inventory and its distribution within the fuel pellet, initial volumes of storage spaces and their locations, radial profile of heat generation within the pellet, etc. Potential degradation mechanisms that may affect sheath integrity of CANDU spent 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 code 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 of the in-house code, identify and extend the ranges of all modules to required operating ranges. Further improvements being considered include upgrades of the analysis module to achieve sufficient accuracy in key output parameters. The emphasis in the near future will be on validation of the in-house code according to a rigorous and formal methodology. The developed models provide a platform for research and industrial applications, including the design of fuel behavior experiments and prediction of safe operating margins for CANDU spent fuel.

The spent fuel dry storage canister is generally made of austenitic stainless-steel and has the role of an important barrier to encapsulate spent fuels and radioactive materials. The canister on the dry storage system has several welding lines in the wall and lid, which have high residual tensile stresses after welding procedure. Interaction between stainless steel and chloride environment from a sea results in an aged-related degradation phenomenon causing chloride induced stress corrosion cracking (CISCC) in the dry storage system. A pending issue to the interim storage of spent fuel awaiting repository disposal is their susceptibility to CISCC of stainless steel canisters. The available mitigation technology should be studied sufficiently to prevent the degradation phenomenon. This paper assesses stress-based mitigation to control residual tensile stress practically applicable to the atmospheric CISCC for the aging management of the stainless steel canisters. There are major components, that is, elevated tensile stress, susceptible material and corrosive environment that must be simultaneously present for CISCC degradation to occur. Surface stress improvement can effectively mitigate the potential for CISCC of the canister external surfaces. The potential deleterious effect of the additional work is negated by the presence of compressive residual stress, which removes the tensile stress needed for CISCC to occur. Surface stress improvement methods such as shock-based peening, shot peening and low plasticity burnishing can be applied for surface stress improvement of the canisters. Stress relaxation processes and advanced welding methods such as laser beam welding and friction stir welding can be also available to mitigate the susceptibility to CISCC. As the result assessing the stress-based mitigation technologies, promising candidate methods could be selected to reduce the residual tensile stresses and to control an aged-related degradation condition causing CISCC in the spent fuel dry storage canister.

Dry storage is a predominantly used method as a spent nuclear fuel storage system after spent nuclear fuel is cooled in the spent fuel pool. Spent nuclear fuel is highly radioactive and it generates heat called decay heat originated by fission products and radiation. Therefore, temperature of spent nuclear fuel should be predicted whether its cladding temperature is maintained under 400°C, which is the allowable temperature limit of cladding in a dry storage. ANSYS Fluent and COBRA-SFS are predominantly used computational method to investigate the temperature of spent nuclear fuels in a dry storage. Herein, thermal analysis results with the methods were compared based on a Single Assembly Heat Transfer Test, which is a heat test with an electrically heated model of a single PWR fuel assembly in a dry cask performed at the Pacific Northwest Laboratory. Decay heat was 1kW and backfill gas was air. Fix temperature boundary condition is applied to inner shell according to measured temperature. In case of peak cladding temperature (PCT), Fluent predicted 240–284°C, while COBRA-SFS gave 243–292°C. The discrepancy between the codes is under 2.5%. The location where PCT took place was 3.65 m from the bottom of the assembly in both results. However, temperature difference between the upper and lower region of the assembly based on the Fluent was smaller than the temperature difference based on the COBRA-SFS. It means the heat was well transferred in an axial direction with Fluent compared to COBRA-SFS. In lower plenum region where air was naturally circulated, COBRASFS had disadvantages compared to Fluent because it modeled the lower plenum by single node, so it was hard to simulate convection heat transfer by natural circulation. natural circulation speed of air in a center region of the assembly was 0.07–0.1 m·s−1 in both cases.

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.

The design of nuclear fuel storage and handling area includes the activities related to the storage and inspection before fuel loading, transfer into the reactor, removal of irradiated fuel to the spent fuel storage rack, underwater handling and storage, and handling into a shipping cask. The purpose of this study is to provide the design requirements for the spent fuel pool to be prevented from the loss of cooling water and for heavy load control to prevent any load drop resulting in damage to safetyrelated systems during heavy load handling in accordance with the regulatory guidelines. And another purpose is to review the sizing of minimum wet storage capacity in the spent fuel pool based on the maximum refueling batch from the core during refueling plus a full core off-load of fuel assemblies and the minimum discharge burnup spent fuel storage during the design life of plant requested by the utility. As the results of this study, the current general arrangement for the spent fuel storage and handling area and the minimum storage capacity are evaluated. These can be good recommendations to enhance more safe and efficient if implemented to the new nuclear power plants.

Facing the problem of saturation of spent nuclear fuel (SNF) stored in temporary storage facilities on sites, interest in the treatment of SNF is increasing, and it is recognized as a task that needs to be solved promptly. Although direct disposal is a general method for dealing with SNF, the entire fuel assembly is classified as high-level waste; thus, the burden of disposal is high. In order to minimize the disposal burden with enhancing safety for long term storage, it is necessary to develop SNF treatment technologies and continuous efforts are required from a national policy perspective. The present study focused on minimizing the volume of high level waste from light water reactor fuel by separation of uranium, which accounts for most of SNF. The chlorination characteristics of uranium (U), rare earth (RE) oxides were confirmed through lab-scale experiments, and the possibility of uranium separation from U-RE simulated fuel was evaluated using NH4Cl chlorinating agent. The detailed results will be posted and discussed.