The saturation rates of the spent fuel (SF) wet storage at the Kori Nuclear Power Plant (NPP), Hanbit, and Hanul are 83.3%, 74.2%, and 80.8% as of the fourth quarter of 2021. The storages of Kori NPP and Hanbit NPP are expected to be saturated in 2031, and Hanul is expected to be saturated in 2032. Therefore, the construction of an interim storage facility to store the SF temporarily stored in the NPP was planned, and preparations for the safe transport of the SF are required. In this paper, radiological preliminary assessment using NRC-RADTRAN in the process of sea transport of SF from the wet storage or ISFSI of the Hanbit NPP to the optional interim storage facility was performed. Since domestic SF transport vessels are not currently in operation, the specifications of the UK Pacific Grebe vessel which can carry up to 20 casks were used. The transport cask used the specifications of KORAD-21, a transport container developed in Korea. Because it can carry more SF assemblies than the existing KN-18. In addition, a land transport safety test was conducted in 2020 and a sea transport test is planned. The sea transport route was entered by referring to the transport route of domestic low and intermediate level waste. The accidents rate was calculated using statistics on maritime accidents from 2017 to 2021. The probability accidents along the transportation route were evaluated as 3.152E -10. When transporting to an interim storage facility, the SF expected to be the main transport target was selected as WH 17X17, combustion 45,000 MWD/MTU, and concentration of 4.5%. The source term was calculated and entered according to this data and the release fraction was entered with reference to the DOE report. In the case of normal transport without accident, the individual dose of the crew member and public residents were estimated to be 0.0525% and 0.000492% of the annual limit of 1 mSv/yr for the general public. Under the accident conditions, the annual individual doses of residents were 0.0011%, 0.0023%, 0.0034%, and 0.0046% of the annual limit of 1 mSv/yr when carrying 5, 10, 15, and 20 casks. Currently, the site of the interim storage facility has not been precisely determined, but a preliminary radiation assessment through sea transport resulted in a significantly lower than the limit. Combined scenario sea transport followed by land transport will be carried out in the next stage of study.

As an alternative technology for the efficient disposal of spent nuclear fuel, various process flows can be selected based on the recovered and separated radioactive nuclide group. This is to examine the efficiency of the disposal area of spent nuclear fuel when various disposal technologies and several treatment processes are applied to spent nuclear fuel, compared to the deep geological disposal of burying the entire spent fuel in the ground. Above all, the biggest advantage of the optional treatment processes is that it can be applied to various disposal methods (deep borehole disposal, deep geological disposal) because it can process spent fuel in various sizes and separate into some groups according to the properties of radionuclides. These optional processes are not new technology and currently available as of today, and the level is classified based on the stepwise separation of high heat emission nuclides and long half-life nuclides. This is to increase the efficiency of the disposal of spent nuclear fuel by separating and managing high-risk radionuclides separately. Relatively various optional processes are possible depending on the level, and characteristic analysis is performed on wastes treated with alternative technologies. The mass balance for each option process is completed, and the amount of waste is also calculated accordingly. These are used as basic data for waste disposal area and economic evaluation. Besides it is easy to process spent fuel of various sizes suitable for deep geological disposal or deep borehole disposal technology when an optional treatment technology is applied to spent fuel. However, since this selective process is based on the process structure constructed in a broad framework, it is considered that additional follow-up studies are needed not only on detailed technology but also on the flow and amount of waste.

In case a spent nuclear fuel transport cask is lost in the sea due to an accident during maritime transport, it is necessary to evaluate the critical depth by which the pressure resistance of the cask is maintained. A licensed type B package should maintain the integrity of containment boundary under water up to 200 m of depth. However, if the cask is damaged during accidents of severity excessing those of design basis accidents, or it is submerged in a sea deeper than 200 m, detailed analyses should be performed to evaluated the condition of the cask and possible scenarios for the release of radioactive contents contained in the cask. In this work, models to evaluate pressure resistance of an undamaged cask in the deep sea are developed and coded into a computer module. To ensure the reliability of the models and to maintain enough flexibility to account for a variety of input conditions, models in three different fidelities are utilized. A very sophisticated finite element analysis model is constructed to provide accurate response of containment boundary against external pressure. A simplified finite element model which can be easily generated with parameters derived from the dimensions and material properties of the cask. Lastly, mathematical formulas based on the shell theory are utilized to evaluate the stress and strain of cask body, lid and the bolts. The models in mathematical formula will be coded into computer model once they show good agreement with the other two model with much higher fidelity. The evaluation of the cask was largely divided into the lid, body, and bottom, bolts of the cask. It was confirmed that the internal stress of the cask was increased in accordance with the hydrostatic pressure. In particular, the lid and bottom have a circular plate shape and showed a similar deformation pattern with deflection at the center. The maximum stress occurred where the lid was in the center and the bottom was in contact with the body. Because the body was simplified and evaluated as a cylinder, only simple compression without torsion and bending was observed. The maximum stress occurred in the tangential direction from the inner side of the cylinder. The bolt connecting the lid and the body was subjected to both bending and tension at the same time, and the maximum stress was evaluated considering both tension and bending loads. In general, the results calculated by the formulas were evaluated to have higher maximum stresses than the analysis results of the simplified model. The results of the maximum stress evaluation in this study confirms that the mathematical models provide conservative results than the finite element models and can be used in the computer module.

This study is to investigate fuel cladding temperature in a transport system for the purpose of developing a methodology for evaluating the thermal performance of spent fuel. Detailed temperature analysis in the transport system is important because the degradation mechanism of the fuel cladding is generally sensitive to temperature and temperature history. In such a system, the magnitude of the temperature change is determined by examining the temperature sensitivity of fuel assemblies and system components including fuel cladding temperature, considering the material properties, component specifications, component aging mechanism, and heat transfer mechanism. The sensitivity analysis is performed using heat transfer models by computational fluid dynamics for the horizontal transport system. The heat transfer within the system by convection, conduction and thermal radiation is calculated by thermal-hydraulic analysis code FLUENT. The calculation region is divided into a basket cell and a transport cask. The thermal analysis of the basket cell is for predicting the fuel cladding temperature. And the reason for analyzing the transport cask is to provide the boundary condition for the basket cell by reflecting the external environmental conditions. Here, the basket cell containing the spent fuel assembly is modeled on the homogeneous effective thermal conductivity. The purpose of this analysis is to evaluate fuel cladding temperatures for the following four main items. That is the effect of surface emissivity changes in basket due to the oxide layer of the fuel cladding, the effect of degradation of the canister backfill helium gas, the effect of fuel assembly position in basket cell on fuel cladding and basket temperatures in canister, and the effect of using the homogeneous effective thermal conductivity model instead of the fuel assembly in basket cell. As a result of the analysis, the maximum temperatures in basket cells are evaluated for the above four items. Thermal margins for each item are investigated for thermal performance requirements (e.g., peak clad temperature below 400oC).

In ROK, when designing a spent nuclear fuel (SNF) storage facility and cask, criticality safety analysis is performed assuming that the SNF is a fresh fuel in order to ensure conservatism. Storage and transportation capacity can be increased by more than 30% by applying the burnup credit, but it has not been applied to the management of SNF. On the other hand, currently in criticality safety analysis, average burnup value is applied to axial burnup profiles, and it is not conservative because burnup of the middle of SNF is greater than average value. Thus, measuring burnup of SNF with high accuracy contributes to the economics and safety of the management of SNF. In this paper, nondestructive burnup evaluation methods for SNF are reviewed in order to study how to measure burnup more accurately. Gamma ray spectrometry and neutron counting have been used as non-destructive burnup evaluation methods of SNF. Gamma spectrum analysis uses the ratio of Cs-134/Cs-137 or Eu-154/Cs-137. The ratio of Cs-134/Cs-137 is used to SNF with cooling time less than 20 years, and the ratio of Eu- 154/Cs-137 is used to SNF with cooling time more than 20 years due to their half-life. In spectrum analysis, detector sensors with high efficiency and energy resolution are needed to clarify each spectrum. High-purity germanium (HPGe) detector has high energy resolution. However, it is not suitable for the analysis of the SNF in the spent fuel pool because it requires separate cooling system and large volume. Thus, CdZnTe (CZT) detector, which has medium energy resolution, is used as a detector of gamma ray spectrometry for the analysis of the SNF in the spent fuel pool. Recently, LaBr3 detector has been commercialized. Although it is difficult to compare clearly due to different conditions such as detector volume and crystal size, LaBr3 detector showed better resolution than CZT in the entire energy region. Neutron counting method has a large error compared to gamma spectrometry because the neutron flux is lower than gamma ray, and neutron absorption reaction, induced fission, and pool environment have to be considered. Large quantity of gamma energy is deposited in the detector by the fission fragments near the SNF. Therefore, fission chambers, which have the highest insensitivity to gamma rays, must be used as neutron detector in order to avoid noise from gamma rays.

Maintaining fuel sheath integrity during dry storage is important. Intact sheath acts as the primary containment barrier for both fuel pellets and fission products over the dry storage periods and during subsequent fuel handling operations. 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, sheath 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. 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. 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 (SCC), delayed hydride cracking (DHC), and sheath splitting due to UO2 oxidation for a defective fuel. The failure by creep rupture, SCC or DHC is in the form of small cracks or punctures. The failure by sheath oxidation or sheath splitting due to UO2 oxidation results in a gross sheath rupture. The second step was to examine the technical bases of all modules of the in-house code, identify and extend the ranges of all modules to required operating ranges. This step assessed the degradation mechanisms for the fuel integrity. The objective of this assessment is to predict the probability of sheath through-wall failure by a degradation mechanisms as a function of the sheath temperature during dry storage. 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 inhouse 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.

Currently, as the saturation capacity of wet storage pool for spent nuclear fuel (SNF) of PWR in Korea has reached approximately 75%, Dry Storage Facilities (DSF) are necessary for sustainable operation of nuclear power plants. It is necessary to develop acceptance requirements for the delivery of SNF from reactor storage site to Centralized DSF. To do this end, the mechanical integrity of SNF is directly related to its repacking, retrieving, and transporting/handling performances. And also, this integrity is a key factor associated with the criticality safety that is connected to the damaged status of SNF. According to the NUREG/CR-6835, the NRC expects that the potential for nuclear fuel failures will increase because of the increase of the fuel discharge burnup and the degradation of fuel and clad material properties. Due to such damages and/or degradation, the fuel rods in the fuel assembly may be extracted and empty for following treatments (transportation, storage, handling etc). This condition can have a detrimental effect on the criticality safety of SNF. Thus, this study investigated whether extracted and empty of damaged SNF rod affects criticality safety. In this analysis, it is assumed that up to four fuel rods are missed. As a result of the analysis, As the number of fuel rods miss up to a certain number, the value of multiplication factor value of the fuel assembly increases. In addition, since the fuel rods located at the outermost layer contained relatively less fissile material than the fuel rods located center of the lattice, and neutrons were lost by the absorption material, the effective multiplication factor value gradually decreased. Nevertheless, the criticality safety was assessed to be maintained.

Thermal analysis and safety assessment of spent fuel transport cask are mainly conducted using commercial Computational Fluid Dynamics (CFD) codes based on Finite Volume Method (FVM). The reliability and predictability of CFD codes have greatly been improved by the development in the computer systems, and are widely used to calculate heat flow in complex structures that cannot be analyzed theoretically. In the field of thermal analysis using the CFD code, it is important to clearly reflect the physical model of the transport cask, and a grid configuration suitable for the physical model is essential for accurate analysis. However, since there are no clear standard and guidelines for grid configuration and size, it is highly dependent on the user’s insight. Spatial discretization errors result from the use of finite-width grids and the approximation of the differential terms in the model equations by difference operators. Since the user usually cannot change the truncation error order of a given discretization scheme, spatial discretization errors can only be influenced by the provision of optimal grids. Therefore, it is necessary to quantify the spatial discretization errors caused by the grid. In the case of Orano TN’s NUHOMS® MP197 transport cask, considering four grids for two sets, the temperature uncertainty of the neutron shield, which has the lowest margin at the limit temperature among transport cask components, was quantified by applying 5-step procedure of the Grid Convergence Index (GCI) method for the uncertainty estimation presented in ASME V&V 20-2009. In the case of domestic spent nuclear fuel transport cask (KORAD21), neutron shield among the transport cask components has the lowest margin at the limited temperature. Accordingly, in this study, the temperature uncertainty of the neutron shield was quantified by applying GCI to three sets considering seven grids. As a result of the calculation, the uncertainty was less than ± 1°C, and the temperature of the neutron shield including the uncertainty was evaluated to be maintained below the limit temperature of 148°C.

For the transport of spent nuclear fuel, it is necessary to evaluate the amount of radioactivity for each assembly and the total amount of radioactivity for each cask. Currently, KHNP is evaluating the radioactivity using the Express mode of the OrigenArp program in the SCALE6.1 code. Express mode is a method to evaluate the radioactivity assuming that it has been burned with the same power per cycle, and Detail mode is a method to evaluate the actual combustion history such as power and cooling time for each cycle. For a total of 3,795 assemblies, including 1,391 assembliess for Kori Unit 1, 1,427 assemblies for Hanbit Unit 2, and 977 assemblies for Hanul Unit 3, the radioactivity was evaluated in Express mode and Detail mode, respectively, and the results were compared. As a result of the evaluation, it was confirmed that the results of the Express mode were evaluated more conservatively by 2.5~12.9% than that of the Detail mode. Accordingly, KHNP established a plan to change the evaluation method from Express mode to Detail mode in order to improve the accuracy of the radioactivity assessment results and eliminate conservatism.

Numerous spent nuclear fuels are generated every year in Korea. To solve the spent nuclear fuel problem within saturated temporary storage, the authorities are readying to build an interim storage and a permanent disposal facility in the country. At the same time, the authorities are readying to establish a management procedure for spent nuclear fuel. In the future, the authorities need to make and apply the Database of spent nuclear fuel to practice the management procedure. However, the structure of a traditional database is not reasonable for information management because it has a problem with listing data and identifying data features due to its structure. In addition, the traditional database always exists human error from working in Excel program by a human. Therefore, this research proposes a new standard information management model based on Semantic Web technique. Semantic Web uses a data structure named ontology. By using the ontology in the information database of the spent nuclear fuel, users, such as institutions related to management, could more easily recognize and understand the Database. Furthermore, since this task proceeds in the ontology construction program, the human error in the new model reduces rather than an environment of the traditional database.

The skeleton of fuel assembly is composed of top nozzle, bottom nozzle, grids, and guide tubes. In the reactor core, all the parts of the fuel assembly suffer degradations due to the condition of high temperature, pressure and water environment. Therefore, many material properties of high temperature mechanical strength, corrosion and irradiation resistance have been considered to choose the material for fuel assembly parts in the fuel development stage. The guide tubes have important roles to connect each parts and support the load of fuel assembly while the fuel is lifted. In Westinghouse 14×14 standard fuel assembly, Zircaloy-4 was used for the material of the guide tubes. Zircaloy-4 has a resistance to water corrosion and maintain good mechanical properties after the discharge from the core, so this alloy is also utilized for a fuel rod cladding material although the microstructure is slightly different due to the heat treatment difference. Thus, it is expected that there is no issue regarding the guide tube integrity after the discharge and during the storage in the pool, especially in case of low burn-up. However, the surface oxidation and resultant hydrogen pick-up can affect to the embrittlement to the Zr alloy. So, it is needed to know the actual status of spent fuel assembly by performing post-irradiation examination. In this study, the degradation level of the guide Tubes in low burn-up spent fuel assembly was investigated using the KAERI PIE facility in order to make some data which can be utilized to the baseline for evaluating the integrity of the spent fuel skeleton.

This paper mainly focuses on the maximum decay heat estimation generated from spent fuel assemblies in the spent fuel pool of Kori units 3&4 at the beginning decommissioning. It is assumed that the spent fuel pool is fully occupied with 2,260 spent fuel assemblies, same as its design capacity. In addition, equally 56.5 spent fuel assemblies have been generated per year. The minimum cooling time is five years considering the transition phase between the permanent shutdown and the amendment of Operating License for decommissioning. Sending and receiving of spent fuel assemblies to/from other units are neglected. Seven representative spent fuel assembly groups are established based on the burnup rate and cooling time. Conservatively high values for the burnup rates and low values for the cooling times are applied. Calculation of the decay heat of each representative group has been performed by using ORIGEN decay solver of SCALE. Then, total decay heat has been calculated based on this. Group 1, 2, and 3 contain comparatively old spent fuel assemblies with 45 GWd/tU burnup rate and 20~30 cooling years. The calculation shows 489~586 watts of decay heat per assembly. Group 4, 5, 6, and 7 contain comparatively new spent fuel assemblies with 55 GWd/tU burnup rate and 5~20 cooling years. The calculation shows 741~1,483 watts of decay heat per assembly. The total maximum decay heat therefore is estimated as 1,609,459 watts.

For safe management of spent nuclear fuels, they should be delivered to repository or waste disposal site. As the amount of spent nuclear fuel transportation is expected to increase in the future due to the provision of an intermediate storage facility, the necessity to secure transportation cask is emerging. In order to secure the spent nuclear fuel transportation cask, it is necessary to analyze the regulatory processes for domestic and foreign spent nuclear fuel transportation cask. In this study, the regulatory processes for domestic and foreign spent nuclear fuel transportation cask was analyzed. In this study, the IAEA, US, and Korea spent nuclear fuel transportation cask regulatory processes were analyzed. The domestic and foreign spent nuclear fuel transportation cask regulatory processes consist of design phase, manufacturing phase, and operation phase. In the design stage, the transport requirements are designed in accordance with the safety requirements of international organizations and countries. The application to be submitted when applying for approval should include a safety analysis report, evidence proving compliance with safety requirements et al. In the manufacturing stage, it is a stage to check whether the safety requirements are satisfied before the first use after manufacturing the transportation cask. Inspections include welding inspection, leakage inspection, shielding inspection, and thermal inspection. In the operation stage, it is a stage of periodically performing inspections for continuous maintenance of the package when the transportation cask is used. The inspection items to be performed are similar to the manufacturing stage and typically include performance inspection of components and leakage inspection. In this study, domestic and foreign spent nuclear fuel transportation cask regulatory processes were analyzed. It was found that the domestic and foreign spent nuclear fuel transportation cask regulatory processes consist of the design phase, the manufacturing phase, and the operation phase. The results of this study can be used as basic data for policy decision-making for the spent nuclear fuel cask.

A rod internal pressure increased by fission gas release is major factor that causes degradation during dry storage of spent fuel. Because rod internal pressure is greatly affected by fuel design, operation power history, it is essential to perform complex calculation using performance code to accurately predict rod internal pressure as function of burnup. However, because it is difficult to apply a complex method into dry storage design and to determine rod internal pressure based on conservative way this study presents a simple correlation that can predict an approximate rod internal pressure as function of burnup For the development of simple correlation, rod internal pressure and fuel rod void volume data measured through about 400 PIE (Post Irradiation Examination) data were used. The developed simple correlation can cover various fuel rod arrays, discharged fuel average burnup, operation history, cladding type, burnup range, and information on Westinghouse type fuel rods such as Spain ENUSA, USA EPRI/ANL/ORNL/PNNL, WEC, etc. In this paper, the data of simple correlation determination is briefly introduced, and the data analysis process and results are summarized. Two correlations that can conservatively determine rod internal pressure and free void volume in fuel rod according to fuel rod average burnup were presented, and the effect of initial He fill pressure was evaluated. In particular, the results of Post Irradiation Examination for 46 fuel rods conducted in Korea are also included, so it is expected that newly presented correlations can be used easily in various ways in the domestic research, industry, and academia.

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

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

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

The IAEA states that in the event of sabotage, nuclear material and equipment in quantities that can cause high radiological consequences (HRC), as well as the minimum systems and devices necessary to prevent HRC, must be located within one or more vital areas. Accordingly, in Article 2 of the ACT ON PHYSICAL PROTECTION AND RADIOLOGICAL EMERGENCY, the definition of the vital area is specified, and a nuclear facility operator submits a draft to the Nuclear Safety and Security Commission to establish vital areas and must obtain approval from Nuclear Safety and Security Commission. Since the spent fuel pool and new fuel storage area are areas where nuclear material is used and stored, they can be candidates for vital areas as direct targets of sabotage. The spent fuel pool is a wet spent fuel storage facility currently operated by most power plants in Korea to cool and store spent nuclear fuel. Considering the HRC against sabotage, it is necessary to review whether sepnt fuel pool needs to establish a vital area. In addition, depending on the status of plant operation during the spent fuel management cycle, the operation status of safety systems to mitigate accidents and power system change, so vital areas in fuel handling building (including spent fuel pool) also need to be adjusted flexibly. This study compares the results of the review on whether the essential consideration factors are reflected in the identification of essential safety systems and devices to minimize HRC caused by sabotage in the spent fuel storage system with the procedure for identifying the vital area in nuclear power plants. It was reviewed from the following viewpoints: Necessity to identify necessary devices to minimize the radiation effects against sabotage on the spent fuel pool, Review of necessary elements when identifying vital areas to minimize the radiation effects of spent fuel pool against sabotage, Necessity to adjust vital areas according to the spent fuel management cycle. The main assumptions used in the analysis of the vital area of the power plant need to be equally reflected when identifying vital areas in spent fuel pool. And, the results of this study are for the purpose of minimizing the radiological consequences against sabotage on the spent fuel storage system including the spent fuel pool and used to establish regulatory standards in the spent fuel storage stage.

Operating and decommissioning nuclear power plants generates radioactive waste. This radioactive waste can be categorized into several different levels, for example, low, intermediate, and high, according to the regulations. Currently, low and intermediate-level waste are stored in conventional 200-liter drums to be disposed. However, in Korea, the disposal of intermediate-level radioactive waste is virtually impossible as there are no available facilities. Furthermore, large-sized intermediate- level radioactive waste, such as reactor internals from decommissioning, need to be segmented into smaller sizes so they can be adequately stored in the conventional drums. This segmentation process requires additional costs and also produces secondary waste. Therefore, this paper suggests repurposing the no-longer-used spent nuclear fuel casks. The casks are larger in size than the conventional drums, thus requiring less segmentation of waste. Furthermore, the safety requirements of the spent nuclear fuel casks are severer than those of the drums. Hence, repurposed spent nuclear fuel casks could better address potential risks such as dropping, submerging, or a fire. In addition, the spent nuclear fuel casks need to be disposed in compliance with the regulations for low level radioactive waste. This cost may be avoided by repurposing the casks.

Thermodynamically, TRUOx, REOx, and SrOx can be chlorinated using ammonium chloride (NH4Cl) as a chlorinating agent, whereas uranium oxides (U3O8 and UO2) remain in the oxide form. In the preliminary experiments of this study, U3O8 and CeO2 are reacted separately with NH4Cl at 623 K in a sealed reactor. CeO2 is highly reactive with NH4Cl and becomes chlorinated into CeCl3. The chlorination yield ranges from 96% to 100%. By contrast, U3O8 remains as UO2 even after chlorination. We produced U/REOx- and U/SrOx-simulated fuels to understand the chlorination characteristics of the oxide compounds. Each simulated fuel is chlorinated with NH4Cl, and the products are dissolved in LiCl-KCl salt to separate the oxide compounds from the chloride salt. The oxide compounds precipitate at the bottom. The precipitate and salt phases are sampled and analyzed via X-ray diffraction, scanning electron microscope-energy dispersive spectroscopy, and inductively coupled plasma-optical emission spectroscopy. The analysis results indicate that REOx and SrOx can be easily chlorinated from the simulated fuels; however, only a few of U oxide phases is chlorinated, particularly from the U/SrOx-simulated fuels.