The Comprehensive Analyzer of Real Estimation for spent fuel POOL (CAREPOOL) has been developed for evaluating the thermal safety of a spent nuclear fuel pool (SFP) during the normal and accident conditions. The management of spent nuclear fuel function provides a management tool for spent nuclear fuel in the SFP. The fuel assemblies both in SFP and reactor side can be shown graphically in the screen. The loading sequence into transfer cask can be checked respectively in the CAREPOOL. A basic heat balance equation was used to estimate the SFP temperature using the heat load calculated in the previous step. The characteristics of typical SFPs and associated cooling systems at reactor sites in the Korea were applied. Accident simulation like station black out leading to loss of SFP cooling or inventory is possible. Emergency cooling water injection pipe installed subsequent to the events at Fukushima 2011 is also modeled in this system. The CAREPOOL provides four main functions- management of spent nuclear fuel, decay heat calculation by ORIGEN-S code, estimation of the time to boil/fuel uncovering by thermal-hydraulics calculations, fuel selection for periodic spent fuel transferring campaign. All of these are integrated into the GUI based CAREPOOL system. The CAREPOOL would be very beneficial to nuclear power plant operator and trainee who have responsibility for the SFP operation.

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

There have been a variety of issues related to spent nuclear fuel in Korea recently. Most of the issues are related to intermediate storage and disposal of spent nuclear fuel. However, recently, various studies have been started in advanced nuclear countries such as the United States to reduce spent nuclear fuel, focusing on measures to reduce spent nuclear fuel. In this study, a simple preliminary assessment of the thermal part was performed for the consolidation storage method which separates fuel rods from spent nuclear fuel and stores them. The preliminary thermal evaluation was analyzed separately for storing the spent fuel in fuel assembly state and separating the fuel rods and storing them. The consolidation storage method in separating the fuel rods was advantageous in terms of thermal conductivity. However, detailed evaluation should be performed considering heat transfer by convection and vessel shape when storing multiple fuel bundles simultaneously.

After spent fuel is stored in a dry storage container, it becomes difficult to obtain information on the fuel’s characteristics. As a result, it is necessary to identify the characteristics of spent nuclear fuel in advance and secure the information necessary to establish delivery acceptance requirements for interim storage and disposal in the future. Therefore, it is necessary to evaluate the characteristics of spent fuel before loading dry storage casks. In order to prepare for the dry storage of spent fuel, information on the basic characteristics of the fuel is required. As part of this information, it is also necessary to establish calculation criteria for spent fuel burnup. Spent fuel burnup can be classified into three categories. The first is burnup evaluated using design codes (design burnup), the second is burnup measured by furnace instruments during power plant operation (actual burnup), and the third is burnup measured through measurement equipment (measured burnup). This paper describes a comparative evaluation of design burnup, actual burnup, and measured burnup for specific fuels (40 bundles).

The spent fuel is classified based on the arrangement of fuel rods, which is considered the primary characteristic data for selecting nuclear fuel. The reason for prioritizing the classification by fuel rod arrangement is that it has the greatest physical impact on the production, supply, operation, reactor type, rack size within the containment vessel, and specifications for the basket in the future dry storage system. Additionally, as mentioned earlier, various meanings of nuclear fuel types are distinguished according to the arrangement of fuel rod. The burnup and cooling period ranges are also important factors in the characterization analysis for the selection of spent fuel, the burnup range was set for both low and high burnup ranges and the cooling period is necessary to consider the reliability during handling of nuclear fuel thermal distribution within the storage system

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

Currently, in the United States, Spent Nuclear Fuel (SNF) is stored at the Independent Spent Fuel Storage Installations (ISFSIs) at 73 Nuclear Power Plants (NPPs). The SNF inventory stored on-site either in pools or dry storage was 84,500 MTU in 2020. The inventory stored in on-site dry storage facilities was 39,207 MTU (46% of the total), and it is growing at a rate of approximately 3,500 MTUs per year. However, because a site for geologic repository for permanent disposal of SNF has not been constructed in the U.S., the SNF will need to be stored in dry storage facilities across the U.S. for a much longer period of time than originally planned. During this time, the dry storage facilities could experience earthquakes of a different magnitude than the one for which they were originally designed. However, there is little data on the response of SNF inside dry storage systems to seismic loads in the U.S., and the various gaps and nonlinearities between storage containers, canisters, baskets, aggregates, and fuel make it very difficult to evaluate by analytical methods. Therefore, a full-scale shake table test is being planned as an international joint research project led by Sandia National Laboratories (SNL) in the U.S. In Korea, KNF decided to participate in this seismic test through the project of SNF integrity evaluation under road and sea normal transportation conditions organized by KNF and conducted by KORAD, KAERI, and Kyung-Hee University, and has provided the KNF 17ACE7 and PLUS7 test assemblies for the tests to SNL. The test will be conducted at the LHPOST6 shake table test facility operated by University of California in San Diego (UCSD) from 2023 to 2024, with the participation of KNF, CRI, and KAERI in Korea. The test units consist of a NUHOMS 32 PTH2 canister, a mockup of a generic vertical cask, a mockup of a generic horizontal storage module, 4 surrogate fuel assemblies, and 28 dummy assemblies. The seismic inputs for the tests will consist of ground motions (acceleration time histories) representative of hard rock, soft rock, and soil sites and seismic conditions in moderately tectonically active Central and Eastern US and highly tectonically active Western US. Ground accelerations for soft rock and soil conditions will be developed taking in account soil-structure interaction. Not only is this test almost impossible to conduct independently in Korea in terms of scale, facilities and costs, but it is also considered an essential test for those of us who are preparing for dry storage of spent nuclear fuel, given the increasing social concern about earthquakes due to the recent earthquake in Turkey.

The electrochemical behavior was investigated during the electrolysis of nickel oxide in LiCl-Li2O salt mixture at 650℃ by changing several components. The focus of this work is to improve anode design and shroud design to increase current densities. The tested components were ceramic anode shroud porosity, porosity size, anode geometry, anode material, and metallic porous anode shroud. The goal of these experiments was to optimize and improve the reduction process. The highest contributors to higher current densities were anode shroud porosity and anode geometry.

We investigated the chemical composition of the planetary host halo star HD47536 via high-resolution spectral observations recorded using a 1.5 meter Cerro Tololo Inter-American Observatory (CTIO) telescope (Chile). Furthermore, we determined the abundances of 38 chemical elements. Both light and heavy elements were overabundant compared to the iron group elements. The abundance pattern of HD47536 was similar to that of halo-type stars, with an enrichment of heavy elements. We analyzed the relationships between the relative abundances of chemical elements and their second ionization potentials and condensation temperatures. We demonstrated that the interplay of charge-exchange reactions owing to the accretion of interstellar matter and the gas-dust separation mechanism can influence the initial abundances and can be used to qualitatively explain the abundance patterns in the atmosphere of HD47536.

Niobium (Nb) is present in Ni-based alloys and stainless steels used in nuclear reactors as structural materials. Nb-93 is a naturally occurring and stable isotope of niobium and Nb-94 (half-life = 20,000 years) is produced by neutron activation of Nb-93. Nb-94 can be present in waste streams from dismantling of nuclear power plants and treatment of the primary coolant circuit. Hence, the radioactive wastes containing active Nb-94 are disposed of in the repositories for low- and intermediate-level waste (LILW). Nb predominantly exhibits a pentavalent oxidation state (i.e., +V) within the stability field of water. Cementitious materials (concrete, mortar, and grout) are extensively utilized in LILW disposal systems as structural components and chemical agents for the stabilization of waste. Solubility defines the source term (i.e., upper concentration limit) in the repository system. However, the solubility behavior of Nb in cementitious systems at high pH remains ill-defined, and information available on the Nb solid phases controlling the solubility is scarce and often ambiguous. Sorption on cementbased materials is one of the main mechanisms controlling the retention of niobium(V) in a LILW repository, and distribution coefficients (Rd) are necessary to evaluate the retention capacity by sorption in the safety assessment of disposal systems. Available sorption data of Nb(V) on cement showed a large discrepancy in Rd, moreover, no sorption data is available for Nb(V) under conditions characterizing the first degradation stage of cement (young cement condition) at pH 13 – 13.5. In this context, the solubility of Nb was extensively investigated in porewater conditions representative of the cement degradation stage I, as well as in CaCl2-Ca(OH)2 systems. Special focus was given to the accurate characterization of the solubility-controlling solid niobium phases. We also studied the sorption of Nb(V) by hardened cement pastes (HCP) and calcium silicate hydrates (CSH, major hydrate of HCP). This work provides the results on Rd, sorption isotherm and sorption mechanisms of Nb(V). Besides, the impact of ISA (polyhydroxycarboxylic acid generated by the degradation of cellulose) on Nb(V) sorption and the dissolution of cement materials was investigated.

In accordance with the notification of the Nuclear Safety and Security Commission (NSSC), environmental impact assessments around nuclear power plants are conducted annually and the results are disclosed to the public. The effects of direct radiation exposure from nuclear power plants as well as liquid effluents and gaseous effluents are taken into consideration in the evaluation of dose calculation for residents. In the United States, regulatory guidelines on direct radiation exposure are described in Reg. Guide 4.1, and the effects of direct radiation are evaluated through regulatory guidelines in Korea. We are going to review optimal evaluation method by reviewing the direct radiation exposure evaluation method currently being conducted in domestic nuclear power plants and the direct radiation exposure evaluation method in overseas nuclear power plants such as in the United States.

In 2022, new regulatory guidelines were announced in relation to the off-site dose calculation (ODC), and accordingly, measures to improve the off-site does calculation program (ODCP), kdose60, were reviewed. The main consideration is, first, that if multiple nuclear facilities are operated on the same site, the boundaries of the restricted areas shall be set as the overlapping outer boundaries of the restricted areas determined by calculation for each nuclear facility. Second, the external exposure caused by direct radiation from a number of nuclear facilities in the same site must be partially or fully applied depending on the facility and site characteristics. Third, the dose conversion coefficient should be evaluated by checking whether the effect of the daughter nuclides is properly reflected. Fourth, the soil contamination period is a factor to consider that radioactive substances deposited on the surface, such as particulate nuclides, affect residents over a long period of time. Fifth, due to the recent construction of Shin-Kori Units 5 and 6, there is a change in the site boundary of the Kori/Saeul site, so as the site boundary is expanded, it is required to add an exposure dose assessment point due to gas effluents and change the exposure dose assessment point according to crop intake. Therefore, through this study, the direction for improving the ODCP will be prepared by reviewing the recent revision of the regulatory guidelines.

The buffer block, which is one of the main components of the engineering barrier system, plays an essential role in mitigating groundwater infiltration and radionuclide transport in a high-level nuclear waste repository. To achieve those purposes, the compacted buffer block must satisfy the functional safety criteria for dry density, water content, and many other components. In this study, the compation curves of the compacted bentonite-sand mixtures were evaluated to identify the relationship between the dry density and the water content of the buffer material. The floating die press at 10 MPa and the cold isostatic press at 40 MPa were applied to compaction of a buffer block with a diameter of 100 mm and a thickness of 10 mm. The condition of a bentonite-sand mixing ratio was 6:4, 7:3, 8:2, and 9:1 with 9 to 21% water content. As a result, the maximum dry density increases, the optimum moisture content decreases as the sand content of buffer material increases. This study can provide the conditions for manufacturing the compacted bentonite-sand buffer block.

In Korea, borated stainless steel (BSS) is used as spent fuel pool (SFP) storage rack to maintain nuclear criticality of spent fuels. As number of nuclear power plants and corresponding number of spent fuels increased, density in SFP storage rack also increased. In this regard, maintain subcriticality of spent nuclear fuels was raised as an issue and BSS was selected as structural material and neutron absorber for high density storage rack. Because it is difficult to replace storage rack, corrosion resistance and neutron absorbency are required for long period. BSS is based on stainless steel 304 and it is specified in the ASTM A887-89 standard depending on the boron concentration from 304B (0.20-0.29% B) to 304B7 (1.75-2.25% B). Due to low solubility of boron in austenitic stainless steel, metallic borides such as (Fe, Cr)2B are formed as secondary phase metallic borides could make Cr depletion near it which could decrease the corrosion resistance of material. In this paper, long-term corrosion behavior of BSS and its oxide microstructures are investigated through accelerated corrosion experiment in simulated SFP condition. Because corrosion rate of austenitic stainless steel is known to be dependent on the Arrhenius equation, a function of temperature, corrosion experiment is conducted by increasing the experimental temperature. Detail microstructural analysis was conducted with scanning electron microscope, transmission electron microscope and energy dispersive spectrometer. After oxidation, hematite structure oxide film is formed and pitting corrosions occur on the surface of specimens. Most of pitting corrosions are found at the substrate surface because corrosion resistance of substrate, which has low Cr content, is relatively low. Also, oxidation reaction of B in the secondary phase has the lowest Gibbs free energy compared to other elements. Furthermore, oxidation of Cr has low Gibbs free energy which means that oxidation of B and Cr could be faster than other elements. Thus, the long-term corrosion might affect to boron content and the neutron absorption ability of the material.

Even though it is emphasized to apply safeguards-by-design (SBD) concept in the early phase of the design of a new nuclear facilities, there is no clear guideline or tools for the practical SBD implementation. Generally known approach is trying to review whether there is any conflicts or shortcomings on a conceptual safeguards components in a design information. This study tries to build a systematic tools which can be easily applied to safeguards analysis. In evaluating the safeguards system or performance in a facility, it is essential to analyze the diversion path for nuclear materials. Diversion paths, however, can be either extremely simplified or complicated depending on the level of knowledge and purpose of specific person who do analyze in the field. In the context, this study discusses the applicability of an event tree and fault tree method to generating diversion paths systematically. The essential components constituting the diversion path were reviewed and the logical flow for systematically creating the diversion path was developed. The path generation algorithm based on the facility design components and logical flow as well as the initial information of the nuclear materials and material flows was test using event tree and fault tree analysis tools. The usage and limitation of the applicability of this two logic methods are discussed and idea to incorporate the logic algorithm into the practical program tools is suggested.The results will be used to develop a program module which can systematically generate diversion paths using the event tree and fault tree method.

Molten salt consisting primarily of eutectic LiCl-KCl is currently being used in electrorefiners in the Fuel Conditioning Facility at Idaho National Laboratory. Options are currently being evaluated for storing this salt outside of the argon atmosphere hot cell. The hygroscopic nature of eutectic LiCl-KCl makes is susceptible to deliquescence in air followed by extreme corrosion of metallic cannisters. In this study, the effect of occluding the salt into a zeolite on water sorption/desorption was tested. Two zeolites were investigated: Na-Y and zeolite 4A. Na-Y was ineffective at occluding a high percentage of the salt at either 10 or 20wt% loading. Zeolite-4A was effective at occluding the salt with high efficiency at both loading levels. Weight gain in salt occluded zeolite-4A (SOZ) from water sorption at 20% relative humidity and 40℃ was 17wt% for 10% SOZ and 10wt% for 20% SOZ. In both cases, neither deliquescence nor corrosion occurred over a period of 31 days. After hydration, most of the water could be driven off by heating the hydrated salt occluded zeolite to 530℃. However, some HCl forms during dehydration due to salt hydrolysis. Over a wide range of temperatures (320–700℃) and ramp rates (5, 10, and 20℃ min−1), HCl formation was no more than 0.6% of the Cl− in the original salt.

The Far-UltraViolet (FUV) imager onboard the Ionospheric Connection Explorer (ICON) spacecraft provides two-dimensional limb images of oxygen airglow in the nightside low-latitude ionosphere that are used to determine the oxygen ion density. As yet, no FUV limb imager has been used for climatological analyses of Equatorial Plasma Bubbles (EPBs). To examine the potential of ICON/FUV for this purpose, we statistically investigate small-scale (~180 km) fluctuations of oxygen ion density in its limb images. The seasonal-longitudinal variations of the fluctuation level reasonably conform to the EPB statistics in existing literature. To further validate the ICON/FUV data quality, we also inspect climatology of the ambient (unfiltered) nightside oxygen ion density. The ambient density exhibits (1) the well-known zonal wavenumber-4 signatures in the Equatorial Ionization Anomaly (EIA) and (2) off-equatorial enhancement above the Caribbean, both of which agree with previous studies. Merits of ICON/FUV observations over other conventional data sets are discussed in this paper. Furthermore, we suggest possible directions of future work, e.g., synergy between ICON/FUV and the Global-scale Observations of the Limb and Disk (GOLD) mission.

A compacted bentonite buffer is a major component of engineered barrier systems, which are designed for the disposal of high-level radioactive waste. In most countries, the target temperature required to maintain safe functioning is below 100°C. If the target temperature of the compacted bentonite buffer can be increased above 100°C, the disposal area can be dramatically reduced. To increase the target temperature of the buffer, it is necessary to investigate its properties at temperatures above 100°C. Although some studies have investigated thermal-hydraulic properties above 100°C, few have evaluated the water suction of compacted bentonite. This study addresses that knowledge gap by evaluating the water suction variation for compacted Korean bentonite in the 25–150°C range, with initial saturations of 0 and 0.22 under constant saturation conditions. We found that water suction decreased by 5–20% for a temperature increase of 100–150°C.

The elements that impact the dynamics and collaborations of waves and particles in the magnetosphere of planets have been considered here. Saturn’s internal magnetosphere is determined by substantiated instabilities and discovered to be an exceptional zone of wave activity. Interchanged instability is found to be one of the responsible events in view of temperature anisotropy and energization processes of magnetospheric species. The generated active ions alongside electrons that constitute the populations of highly magnetized planets like Saturn’s ring electron current are taken into consideration in the current framework. The previous and similar method of characteristics and the perturbed distribution function have been used to derive dispersion relation. In incorporating this investigation, the characteristics of electromagnetic ion cyclotron wave (EMIC) waves are determined by the composition of ions in plasmas through which the waves propagate. The effect of ring distribution illustrates non-monotonous description on growth rate (GR) depending upon plasma parameters picked out. Observations made by Cassini found appropriate for modern study, have been applied to the Kronian magnetosphere. Using Maxwellian ring distribution function of ions and detailed mathematical formulation, an expression for dispersion relation as well as GR and real frequency (RF) are evaluated. Analysis of plasma parameters shows that, proliferating EMIC waves are not developed much when propagation is parallelly aligned with magnetosphere as compared to waves propagating in oblique direction. GR for the oblique case, is influenced by temperature anisotropy as well as by alternating current (AC) frequency, whereas it is much affected only by AC frequency for parallel propagating waves.

During their respective missions, the spacecraft Voyager and Cassini measured several Saturn magnetosphere parameters at different radial distances. As a result of information gathered throughout the journey, Voyager 1 discovered hot and cold electron distribution components, number density, and energy in the 6–18 Rs range. Observations made by Voyager of intensity fluctuations in the 20–30 keV range show electrons are situated in the resonance spectrum’s high energy tail. Plasma waves in the magnetosphere can be used to locate Saturn’s inner magnetosphere’s plasma clusters, which are controlled by Saturn’s spin. Electromagnetic electron cyclotron (EMEC) wave ring distribution function has been investigated. Kinetic and linear approaches have been used to study electromagnetic cyclotron (EMEC) wave propagation. EMEC waves’ stability can be assessed by analyzing the dispersion relation’s effect on the ring distribution function. The primary goal of this study is to determine the impact of the magnetosphere parameters which is observed by Cassini. The magnetosphere of Saturn has also been observed. When the plasma parameters are increased as the distribution index, the growth/damping rate increases until the magnetic field model affects the magnetic field at equator, as can be seen in the graphs. We discuss the outputs of our model in the context of measurements made in situ by the Cassini spacecraft.