K-DOSE60, a off-site dose calculation program currently used by khnp, is performing evaluation based on the gaseous effluent evaluation methodology of NRC Reg. Guide 1.109. In particular, H-3 and C-14, which are the major nuclides of gaseous effluent, are evaluated using a ratio activity model. Among them, H-3 is additionally evaluating the dose to OBT (Organically Bound Tritium) and HT as well as HTO (Triated water). However, NRC Reg. Guide 1.109 is a methodology developed in the 1970s, and verification was performed by applying the evaluation methodology of H-3 and C-13 presented by IAEA TRS-472 in 2010 to the current K-DOSE60. The IAEA TRS-472 methodology also includes OBT and HT for H-3. In order to apply the ratio radioactivity model presented in IAEA TRS-472, the absolute and relative humidity were calculated using the weather tower of the nuclear site and used for H-3 evaluation. For the dose evaluation of HT, the previously used Canada Chalk River Lab. (CNL) conversion factor was used. For atmospheric carbon concentration, the carbon concentration presented in IAEA TRS-472 was used, not the carbon concentration in the 1970s of NRC Reg. Guide 1.109. It was confirmed that the K-DOSE60, which applied the changed input data and methodology, was satisfied by performing comparative verification with the numerical calculation value.

In 2022 and 2023, the Korea Institute of Nuclear Safety (KINS), a regulatory body, revised the regulatory guidelines for off-site dose evaluation to residents, marine characteristics surveys around nuclear facilities, and environmental radiation surveys and evaluation around nuclear facilities. In addition, the NRC, a US regulatory body, has revised regulatory guide 1.21 (MEASURING, EVALUATING, AND REPORTING RADIOACTIVE MATERIAL IN LIQUID AND GASEOUS EFFLUENTS AND SOLID WASTE) to change environmental programs for nuclear facilities. The domestic regulatory guidelines were revised and added to reflect the experience of site dose evaluation for multiple units during the operation license review of nuclear facilities, the resident exposure dose age group was modified to conform to ICRP-72, and the environmental monitoring plan was clarified. In the case of the US, the recommended guidelines for updating the long-term average atmospheric diffusion factor and deposition factor, the clarification of the I-131 environmental monitoring guidelines for drinking water, and the clarification of the procedures described in the technical guidelines when changing environmental programs have been revised and added. Through such regulatory trend review, it is necessary to preemptively respond to changes in the regulatory environment in the future.

The ability to both assay the presence of, and to selectively remove ions in a solution is an important tool for waste water treatment in many industrial sectors, especially the nuclear industry. Nuclear waste streams contain high concentrations of heavy metals ions and radionuclides, which are extremely toxic and harmful to the environment, wildlife and humans. For the UK nuclear industry alone, it is estimated that there will be 4.9 million metric tonnes of radioactive waste by 2125, which contains a significant number of toxic radionuclides and heavy metals. This is exacerbated further by increased international growth of nuclear new build and decommissioning. Efforts to remove radionuclides have been focused on the development and optimisation of current separation and sequestering techniques as well as new technologies. Due to the large volumes of waste the techniques must be economical, simple to use and highly efficient in application. Magnetic nanoparticles (MNPs) offer a powerful enhancement of normal ion exchange materials in that they can be navigated to specific places using external magnetic fields and hence can be used to investigate challenges such as, pipework in preparation of decommissioning projects. They also have the potential to be fine-tuned to extract a variety of other radionuclides and toxic heavy metals. It has been demonstrated that with the right functional groups these particles become very strongly selective to radionuclides, such as Uranium. However, this new technology also has the potential to effectively aid nuclear waste remediation at a low cost for the separation of both radionuclides and heavy metals. In this work, we investigate the origin of the selectivity of superparamagnetic iron oxide nanoparticles (SPIONs) to Uranium by making systematic changes to the existing surface chemistry and determining how these changes influence the selectivity. Identifying the mechanism by which selected common nuclear related metals, such as Na(I), K(I), Cs(I), Ca(II), Cu(II), Co(II), Ni(II), Cd(II), Mg(II), Sr(II), Pb(II), Al(III), Mn(II), Eu(III) and Fe(III), are sorbed will allow for specific NP-target (nanoparticle) ion interactions to be revealed. Ultimately this understanding will provide guidance in the design of new targeted NP-ligand constructs for other environmental systems.

After the Fukushima accident, significant amount of radioactively contaminated waste has been generated with 50~250 m3/day and stored in tanks of the Fukushima Daiichi nuclear power plant site. The contaminated water is treated by various treatment facility such as KURION, SARRY, Reverse Osmosis, and ALPS to remove 62 radioactive nuclides except H-3. For the contaminated water treatment process, massive secondary wastes such as sludge, spent adsorbent, and so on as by-product are being generated by the facilities. In Japan, to treat the secondary wastes, melting technologies such as GeoMelt, In-can vitrification and Cold Crucible Induction Melting vitrification are considered as a candidate technologies. In this study, the technologies were reviewed, and the advantage and disadvantage of each technology were evaluated as the candidate technologies for treatment of the secondary wastes.

KHNP’s vitrification technology introduced a commercialized vitrification facility to the Hanul nuclear power site after a commercialization test through a lab test and a pilot plant at KHNP-CRI. France’s ANADEC (consortium with CEA, Orano, ECM Technologies and Andra) conducted a feasibility evaluation from FY2018 to FY2021 to apply In-Can vitrification, which was developed to treat Fukushima Effluent Treatment Waste (FETW) such as carbonate slurry and ferric slurry generated from ALPS (Advanced Liquid Processing System-Multi Radionuclides Removal) facilities for waste treatment in Fukushima, Japan. For commercialization, the following method was used. First, through the Laboratory scale studies, the possibility of high waste loading (60wt% in dry mass) of slurry on borosilicate matrix was tested. In addition, the volatility of radionuclide was evaluated through radionuclides surrogates with a Bench-scale mockup and glass discharge (100 kg) was evaluated through In-Can vitrification process verification. The feeding system was improved through a pilot scale test, and finally, glass discharge (300 kg) was evaluated after large amount of waste was treated through an industrial prototype (Fullscale) at the CEA Marcoule site (France).

A vitrification facility control area is formed to control and monitor the vitrification facility process, and the control system is designed to manage the vitrification facility more safely and effectively. The control system is largely composed of a process control system and an off-gas monitoring system. The process control system is operated so that operation variables can be maintained in a normal state even in normal and transient conditions, and is designed so that the vitrification facility can be stably maintained in the event of an abnormality in the facility. The process control system consists of Programmable Logic Controller (PLC) and Local Control Panel (LCP), which controls and monitors each unit device. In addition, operation variables are provided to the operator so that the operator can manage operation variables during process control in a centralized manner for the operation of the vitrification facility. The off-gas monitoring system is operated to monitor whether the off-gas discharged to the environment is stably maintained within the standard level, and the off-gas is monitored through an independent monitoring system.

After melting glass at a high temperature of about 1,100 degrees in the Cold Crucible Induction Melter (CCIM) of the vitrification facility, radioactive waste is fed into the CCIM to vitrify radioactive waste. Accordingly, since the metal sector of the CCIM contacts the high-temperature molten glass, cooling water is supplied to continuously cool the metal sector. The cooling system is divided into primary and secondary cooling water systems. The primary cooling water flows inside the metal sector of the CCIM to maintain the metal sector within normal temperature, thereby forming a glass layer between the metal sector and the high-temperature melting glass. The secondary cooling system is a system that cools the primary cooling water that cools the metal sector, and removes heat generated from the primary cooling system. In addition, it is designed to stably supply cooling water to the secondary cooling water system through an emergency cooling water system so that cooling water can be stably supplied to the secondary cooling water system in the event of secondary cooling water loss. Therefore, it is designed to maintain the facility stably in the event of loss of cooling water for the CCIM of the vitrification facility.

In order to reduce the area of the high-level radioactive waste (HLW) repository, a buffer material with high thermal conductivity is required. This is because if the thermal conductivity of the buffer material is high, the distance between the disposal tunnels and the deposition holes can be reduced. Sand, which is a natural material and has higher thermal conductivity than bentonite, is added to bentonite to develop an enhanced buffer material. For the sand-bentonite mixture, it is important which sand to use and how much to add because an enhanced buffer material should satisfy both hydraulic (H) and mechanical (M) performance criteria while improving thermal conductivity (T). In this study, we would like to show what type of sand and how much sand should be added to develop an enhanced buffer material by adding sand to Gyeongju bentonite, a representative bentonite in Korea. For this purpose, the thermal conductivity, hydraulic conductivity, and swelling pressure of the sand-Gyeongju bentonite mixture according to the sand addition rate were measured. It is more efficient to use silica sand with smaller particles than Jumunjin sand which is a representative sand in Korea as an additive for an enhanced buffer material than using the Jumunjin sand. In order for the sand-Gyeongju bentonite buffer material to satisfy both the hydraulic and mechanical performance criteria as a buffer material while increasing the thermal conductivity, it is judged that the optimum dry density is 1.7 g/cm3 at least and the optimum sand addition rate is 10% at most.

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.