In environments where buffer materials are exposed to increased temperature due to the decay heat emitted by radioactive waste, it is crucial to assess the performance of the buffer material in relation to temperature effects. In this study, we conducted experiments using Bentonil-WRK, a calcium-type bentonite, compacted to a dry density of 1.65 g/cm3 and an initial water content of 15%. The experimental temperature conditions were set to 30, 60, 90, 110, and 130°C. We observed that the swelling pressure of the compacted bentonite buffer decreased as the temperature increased. The findings from this study can provide valuable guidance for the design of high-level waste repository in Korea.

After the Fukushima disaster, overseas nuclear power plants have established conditions for issuing a red alert in the event of fuel damage within the spent fuel pool and they have already implemented conditions for issuing a blue alert when fuel is exposed above the water surface. In South Korean nuclear power plants, a real-time monitoring system is in place to oversee the exposure of spent fuel to the surface within the spent fuel pool. To achieve this, a water level indicator gauge is installed within the spent fuel pool, allowing for continuous real-time monitoring. This paper conducted a comparative assessment of radiation levels from water level monitoring system in two units’ spent fuel pools based on the low water levels (1 feet from the storage rack), utilizing the radiation analysis code (MCNP).

More than 20,000 bundles of spent nuclear fuel are stored in the spent nuclear fuel storage pool of domestic nuclear power plants, and the dry storage facility project in the nuclear power plant site is being promoted as the saturation of the wet storage pool is imminent. Since bending or twisting of spent nuclear fuel is an important item in order to load spent nuclear fuel into a dry storage cask, PSE (Pool Side Examination) was performed to verify this. This paper describes whether it can be safely loaded into a dry storage cask based on the measurement results of bending or twisting of spent nuclear fuel. The nuclear fuel assembly is designed to prevent excessive assembly bending and twisting because it can cause interference during dry storage and handling due to factors such as differences in depletion of nuclear fuel rods, irradiation growth, and coolant flow during reactor operation. The bending of the nuclear fuel assembly is measured by establishing a Plumb Line to photograph the nuclear fuel assembly based on it, and calculating a pixel that images the distance between the support grid and the Plumb Line. The twisting of the nuclear fuel assembly is measured by forming a virtual vertical plane with two Plumb Lines, and based on this, the twisting angle of the lower fixed compared to the upper fixed. As a result of the measurement, the bending of spent nuclear fuel was about 0.0-10.2 mm, much lower than the reactor loading criteria of 15.0 mm, and in the case of twisting, about 0.0~2.2° much lower than the reactor loading criteria of 5.0°. Therefore, it was confirmed that spent nuclear fuel at domestic nuclear power plants was not affected by bending and twisting when loading into dry storage cask.

South It is necessary to develop the future technologies to improve the sustainability and acceptability of nuclear power plants generation. Currently, our company is preparing to build the dry storage facility on-site in accordance with the basic plan for managing high-level radioactive waste announced by the government in 2021. However, studies on technologies for the volume reduction of spent nuclear fuel to increase the efficiency of on-site spent fuel dry storage facilities are very not enough. Accordingly, in this study, the storage efficiency and appropriateness for the SF volume reduction processing technologies such as SF oxide processing technology and consolidation technology are evaluated. Finally, the goal is to develop the optimized technologies to improve the storage efficiency of spent nuclear fuel. As a result in this study is followings. [Safety] After removing volatile fission products (Xe, Kr, I, etc.), Xe, Kr, etc. are removed during storage of the sintered structures. UO2 has a high melting point of approximately 1,000°C after cesium (Cs) has been removed, and heat can be removed by natural convection. [Economy]1999 DUPIC unit facility unit price reference, 2020 standard 328 $/kg estimated. A Comprehensive Approach Considering the Whole System is needed. Benefit from replacement and continuous operation of metal storage containers. Changes in economic efficiency obtained in conjunction with fluctuations in electricity prices and disposal. [Waste filter] A separated solidification facility high-level waste filter is required, and overseas outsourcing must be considered. [Waste cladding]. Cannot be accommodated in low-level disposal site. This reason is why the Ni nuclides occur to be in bulk. [Metal structural material] It is possible to reduce the initial volume by 7.6% or more when compressed or melted, but the technology needs to be advanced. [Oxide blocks] Larger size and density are expected to improve storage and disposal efficiency. [Facilities operation waste] Expected to be able to be disposed of at mid-to-low level decommissioning sites in Gyeongju city. [Solidified volatile nuclides and activated metals] Expected to improve storage efficiency when used volume is reduced and stored, such as outsourced reprocessing. [Oxide block] Radioactivity and decay heat are estimated to be reduced by half during oxide treatment. 75% reduction in volume and 40% reduction in storage area compared to used nuclear fuel before treatment. [Merits/Shortages] Improvement of storage and disposal efficiency empirical research such as large-capacity [real-scale] oxide block production is required. Oxide processing facilities are likely to be classified as post-use nuclear fuel processing facilities. It is determined that additional documents such as a Radiation Environmental Report (RER) must be submitted. Existence of possible external leaks of glass, highly mobile radionuclides from the point of view of nuclear criticality and heat removal. Acceptancy requirements of citizens in the process of creating additional sites for oxide treatment facilities. Considering social public opinion, it is necessary to secure the acceptability such as residents’ opinions convergence. Characteristics of high nuclear non-propagation compared to other processing technologies involving chemical processing. Also, Expectation of volume reduction effect for spent nuclear fuel itself. Volume reduction methods for solid waste and gaseous waste are required.

Neutron resonance transmission technique was applied for assaying isotopic fissile materials produced in the pyro-process. In each process of the pyro-process, a different composition of the fissile material is produced. Simulation was basically performed on 235U and 239Pu assay for TRU-RE product, hull waste, and uranium addition. The resonance energies were evaluated for uranium and plutonium in the simulation, and the linearity in the detection response was examined on the fissile content variation. The linear resonance energies were determined for the analysis of 235U and 239Pu on the different fissile materials. For enriched TRU-RE assay, the sample condition was suggested; The sample density, content, and thickness are the key factors to obtain accurate fissile content. The detection signal is discriminated for uranium and plutonium in neutron resonance technique. The transmitted signal for fissile resonance has a direct relation with the content of fissile. The simulation results indicated that the neutron resonance technique is promising to analyze 235U and 239Pu for various types of the pyro-process material. An accurate fissile assay will contribute toward safeguarding the pyro-processing system.

The US NRC developed a program called NRCDose3 to evaluates the environmental impact of radiation around nuclear facilities. The NRCDose3 code is a software suite that integrates the functionality of three individual LADTAP II, GASPAR II, and XOQDOQ Fortran codes that were developed by the NRC in the 1980’s and have been in use by the nuclear industry and the NRC staff for assessments of liquid effluent and gaseous effluent, and meteorological transport and dispersion, respectively. Through the integrated program, it is possible to conduct safety assessment and environmental impact assessment from liquid and gaseous effluent when operating permits are granted. In addition to a more user-friendly graphic user interface (GUI) for inputting data, significant changes have been made to the data management and operation to support expanded capabilities. The basic calculation methods of the LADTAP II, GASPAR II, and XOQDOQ have not been changed with this update to the NRCDose3 code. Several features have been added. The previous program used only ICRP-2 dose conversion factor, but the new program can additionally use dose conversion factor of ICRP-30 and ICRP-72. In the previous program, 4 age groups (infant, child, teen, and adult) were evaluated during dose evaluation, but when ICRP-72 was selected, 6 age groups (infant, 1-year, 5-year, 10-year, 15-year, and adult) could be evaluated. In addition, when selecting ICRP-72, many user-modifiable parameters such as food intake and exposure time were added. It will be referred to E-DOSE60, a program currently under development.

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.