The bioreduction process from soluble U(VI) to insoluble U(IV) has been extensively studied in the field of radionuclides migration. Since acetic acid (AcOH) is widely used as an electron donor for bioreduction of U(VI), it is necessary to understand the effect of U(VI)-AcOH complexes that exist in different species depending on pH on this process. Changes in samples before and after bioreduction can be compared using time-resolved laser luminescence spectroscopy (TRLLS), which measures the characteristic luminescence spectra of different U(VI) species. Although luminescence properties of U(VI)-AcOH species were reported, experiments were conducted under conditions below pH 4.5. In this study, spectrophotometry and TRLLS for U(VI)-AcOH species (10−100 μM U(VI) and 20 mM AcOH) were performed in pH ranges extending to neutral and alkaline pH regions similar to groundwater conditions as well as acidic pH region. Two different complexes (UO2(AcO)+, UO2(AcO)2 with U(VI) and acetate ratios of 1:1, 1:2) were observed in the acidic pH region. The 1:1 complex, which appears as the pH increases, has no luminescence properties, but its presence can be confirmed because it serves to reduce the luminescence intensity of UO2 2+. In contrast, the 1:2 complex exhibits distinct luminescence properties that distinguish it from UO2 2+. The 1:3 complex (UO2(AcO)3 -) expected to appear with increasing pH was not observed. Two different complexes ((UO2)3(OH)5 +, (UO2)3(OH)7 - with U(VI) and OH ratios of 3:5, 3:7) were observed as the major species in the neutral pH region, but their luminescence lifetimes are remarkably short compared those in the absence of AcOH. Solid U(VI) particles were observed in the alkaline pH region, and they also had completely different luminescence properties from the aforementioned U(VI)-AcOH and U(VI)-hydrolysis species. Based on these results, the effect of pH in the presence of AcOH on the bioreduction process from U(VI) to U(IV) will be discussed.

Bacterial metabolisms influence the behavior of uranium (U) in deep geological repository (DGR) system because bacteria are ubiquitous in the natural environment. Nevertheless, most studies for the U(VI) bioreduction have focused on a few model bacterium, such as Shewanella putrefaciens, Desulfovibrio desulfuricans, and Geobacter sulfurreducens. In this study, the potential of aqueous U(VI) ((U(VI)aq) reduction by indigenous bacteria was examined under anaerobic conditions with addition of 20 mM sodium acetate for 24 weeks. Three different indigenous bacterial communities obtained from granitic groundwater at depths of 44–60 m (S1), 92–116 m (S2), and 234–244 m (S3) were applied for U(VI)aq reduction experiments. The S2 groundwater contained the highest U concentration of 885.4 μg/L among three groundwater samples, where U mainly existed in the form of Ca2UO2(CO3)3(aq). The S2 groundwater amended 20 mM of sodium acetate was used for the U(VI)aq bioreduction experiment. Variations in the U(VI)aq concentration and redox potential were monitored for 24 weeks to compare U(VI)aq removal efficiency in response to indigenous bacteria. The U(VI)aq removal efficiencies varied among three indigenous bacteria: 57.8% (S3), 43.1% (S2), and 37.7% (S1). The presence of the thermodynamically stable uranyl carbonate complex resulted in the incomplete U(VI)aq removal. Significant shifts in indigenous bacterial communities were observed through highthroughput 16S rRNA gene sequencing analysis. Two SRB species, Thermodesulfovibrio yellowstonii and Desulfatirhabdium butyrativorans, were dominant in the S3 sample after the anaerobic reaction, which enhanced the bioreduction of U(VI)aq. The precipitates produced by bacterial activity were determined to be U(IV)-silicate nanoparticles by a transmission electron microscope (TEM)-energy dispersive spectroscope (EDS) analysis. These results demonstrated that considerable U immobilization is possible by stimulating the activity of indigenous bacteria in the DGR environment.

The 2007 Recommendation of the International Commission on Radiological Protection recommended the application of dose constraints to optimize radiation protection to resolve the inequity of exposure among radiation workers. The average annual occupational doses in Korean nuclear power plants (NPPs) are 0.3-0.8 mSv. These doses are much lower than the annual effective dose limit of 50 mSv for radiation workers stipulated by the Nuclear Safety Act. In addition, most NPP workers received less than 0.1 mSv per year. These doses are lower than the average annual occupational doses of 0.3- 0.8 mSv. Korean regulatory body conducted the study to legislate the dose constraints in the Korean regulatory system and determine dose constraints (draft) for radiation workers. The legislation of dose constraints would not greatly affect the radiation protection programs in Korean NPPs because most workers received very low doses. However, some workers received relatively higher doses than others. This study analyzed the occupational exposure conditions, such as exposure type and situation, in Korean NPPs. This study investigated the internal and external radiation doses and the radiation doses depending on the NPP operating conditions, including normal operation, planned maintenance, and intermediate maintenance, for the last ten years (2012-2021). As a result, most NPP workers received external exposure rather than internal exposure. Furthermore, most radiation exposures occurred during the planned maintenance period. The results of this study can be used for optimizing occupational doses in Korean NPPs.

The phosphate industry is classified by IAEA as one of the Naturally Occurring Radioactive Materials (NORM) industry sectors most likely to require regulatory consideration. The production of phosphorous fertilizers constitutes the major activity in the industry, which can give rise to exposures of workers and the public through the handling and usage of phosphate rock and residues associated with processing. During the production process, when phosphate rock is digested with acid to produce phosphoric acid, some radionuclides, particularly Radium, become concentrated in residues, such as the scale that tends to form inside pipes and vessels. The registered radioactivity of phosphate rock in South Korea is less than 1.7 Bq/g for U-238, but according to the IAEA SRS No. 49, the radioactivity of phosphate scale can be up to 1,000 times higher than the raw mineral. Therefore, this study evaluated the potential for worker exposure during maintenance related to the removal of scales at a fertilizer manufacturing facility producing phosphoric acid in Korea.

Radiation workers who handle radioisotopes, radioactive waste, nuclear material etc. may be contaminated with radioactive material due to inhalation, resulting in internal radiation exposure. For preventing radiation damage and monitoring the exposure of workers, KAERI operates a Body Radiation Measurement Laboratory. According to Article 5 of the Nuclear Safety and Security Commission (NSSC) Notice No. 2017-77, “Regulation on Measurement and Calculation of Internal Radiation Dose,” The nuclear energy-related business operator with workers etc. shall establish and operate procedures and methods including the following Subparagraphs to secure the reliability of measurement of the internal radiation dose : operation and calibration of measuring instrument, inspection procedures, uncertainty of measurement, lower limit of detection and geometric configuration used for measurement. In accordance with the provision, Whole Body Counter utilized in the Body radiation Measurement Laboratory has periodic calibration / QA procedures to ensure reliability. This paper performed reliability validation of the measurement system of the Body Radiation Measurement Laboratory in the KAERI based on the performance criteria for radio-bioassay criteria presented in ISO 28218 and ANSI HPS N13.30-2011(R2017). The first criteria is MTL (Minimum Testing Level). ISO 28218 provides MTLs for each measurement category, type and nuclide. For reliable results, it is recommended to use calibration sources with higher radioactivity than the values given. The MTL for fission products in total body counting is 3 kBq and for the last 3 years the laboratory has been using sources of 6-7 kBq (Co-60, Cs-137 etc.). The second criteria is RMSE (Root Mean Square Error). It is a measure of total error defined as the square root of the sum of the square of the relative precision (SB) and the square of the relative bias (Br). The RMSE shall be lower than or equal to 0.25. The largest RMSE in the last 3 years is 0.12, and average value is 0.065, which meets the criteria. In this study, we verified the reliability of the radioactivity measurement system (WBC) based on the radio-bioassay standards presented in ISO 28218 and ANSI HPS N13.30-2011(R2017). The values were obtained using 3 years of calibration count data, and it was found that both MTL, RMSE for each nuclide met the standards with a large margin of error and were in good operating condition. This study can be applied to the maintenance, performance check, and reliability verification of similar in vivo radio-bioassay methods.

As an initial part of Kori-1 & Wolsung-1 Unit decommissioning planning, a characterization plan is developed to define the nature, extent and location of contaminants, determine sampling locations and protocols, determine quality assurance objectives for characterization, and define documentation requirements. The actual characterization of a facility is an iterative process that involves initial sampling according to the characterization plan, field management (such as labeling, packaging, storing, and transport) of the samples, laboratory analysis, conformance to the data quality objectives (DQOs), and then identifying any additional sampling required, refining the DQOs, and modifying the characterization plan accordingly. The final product of the facility characterization is a document that describes the type, amount, and location of contaminants that will require consideration and removal during the decommissioning operations sufficient to prepare a decommissioning plan. In this study, implementing a characterization plan, developed in accordance with this standard, will result in obtaining or deriving the above information.

Kori Nuclear Power Plant Unit 1, which began operating in 1978, is Korea’s oldest commercial nuclear reactor. The reactor was permanently shut down in June 2017, and now the decommissioning process has begun. The decommissioning process will generate a significant amount of waste that requires appropriate management to minimize the impact on the environment and human health. And the waste routing, i.e. the activities and logistics for managing the material generated, is a key point in a decommissioning project. It determines the routes from the material inventory to the envisaged material end states. In this study, we review on several factors for the selection of the waste routes in a decommissioning project. In terms of sustainability, the ‘waste hierarchy’ should be applied to routing materials from nuclear facilities. According to the waste hierarchy, the preferred end state is reuse or recycling of the waste as material or, more preferably, the avoidance of waste generation. In addition, treatments (such as decontamination and thermal treatment) that can reduce the volumes requiring disposal as radioactive waste should be considered. Another important parameter is the need to secure availability and capacity of waste routes. Short-term bottlenecks or any delay in the removal of the waste from the site often has an impact on other site activities. If possible, at least two alternative waste routes should be identified for the main categories of waste and kept available throughout the decommissioning project. All routes should be direct to the material end state if possible, but it is more important that waste is removed from the site so that other site operations are not impeded.

KHNP is carrying out international technical cooperation and joint research projects to decommission Wolsong unit 1 reactor. Construction data of the reactor structures, experience data on the pressure tube replacement projects, and the operation history were reviewed, and the amount of dismantled waste was calculated and waste was classified through activation analysis. By reviewing COG (CANDU owners Group) technical cooperation and experience in refurbishment projects, KHNP’s unique Wolsong unit 1 reactor decommissioning process was established, and basic design of a number of decommissioning equipment was carried out. Based on this, a study is being conducted to estimate the worker dose of dismantling workers. In order to evaluate the dose of external exposure of dismantling workers, detailed preparation and dismantling processes and radiation field evaluation of activated structures are required. The preparation process can be divided into dismantlement of existing facilities that interfere with the reactor dismantling work and construction of various facilities for the dismantlement process. Through process details, the work time, manpower, and location required for each process will be calculated. Radiation field evaluation takes into account changes in the shape of structures by process and calculates millions of areas by process, so integrated scripts are developed and utilized to integrate input text data. If the radiation field evaluation confirms that the radiation risk of workers is high, mutual feedback will be exchanged so that the process can be improved, such as the installation of temporary shields. The results of this study will be used as basic data for the final decommissioning plan for Wolsong unit 1. By reasonably estimating the dose of workers through computer analysis, safety will be the top priority when decommissioning.

Prevention of radiation hazards to workers and the environment in the event of decommissioning nuclear power plants is a top priority. To this end, it is essential to continuously perform radiation characterization before and during decommissioning. In operating nuclear power plants, various detectors are used depending on the purpose of measurement. Portable detectors used in power plants have excellent portability, but there is a limit to the use of a single measuring device alone to quantify radioactive contamination, nuclide analysis, and ensure representation of measurement results. In foreign countries, gamma-ray visualization detectors are being actively used for operating and decommissioning nuclear power plants. KHNP is also conducting research on the development of gamma-ray visualization detectors for multipurpose field measurement at decommissioning nuclear power plants. It aims to develop detectors capable of visualizing radioactive contamination, analyzing nuclides, estimating radioactivity, and estimating dose rates. To this end, we are developing related software according to the development process by purchasing sensors from H3D, which account for more than 75% of the US gamma-ray visualization detector market. In addition, field tests are planned in the order of Wolsong Unit 1 and Kori Unit 1 with Research reactor in Gongneung-dong in accordance with the progress of development. The detector will be optimized by analyzing the test results according to various gamma radiation field environments. The development detector will be used for various measurement purposes for Kori unit 1 and Wolsong

Spent nuclear fuel temporary storage in South Korea is approximately 70% of total storage capacity as of the 4th quarter of 2022 amount is stored. In addition, according to the analysis of the Korean Radioactive Waste Society, saturation of nuclear power plant temporary storage is expected sequentially from 2031, and accordingly, the need for high-level radioactive waste disposal facilities has emerged. Globally, after the conclusion of the EU Taxonomy, for nuclear energy in order to become an ecofriendly energy, it is necessary to have a high-level radioactive waste disposal site and submit a detailed operation plan for high-level radioactive waste disposal site by 2050. Finland and Sweden have already received permission for the construction of high-level radioactive waste disposal facilities, and other countries, such as Switzerland, Japan, the United States, and Canada, are in the process of licensing disposal facilities. In order to establish a repository for high-level radioactive waste, the performance and safety analysis of the repository must be conducted in compliance with regulatory requirements. For safety analysis, it needs a collection of arguments and evidence. and IAEA defined it as ‘Safety case’. The Systematic method, which derives scenarios by systematizing and combining possible phenomena around the repository, is widely used for developing Safety case. Systematic methods make use of the concept of Features, Events and Processes (FEP). FEP identifies features that affect repository performance, events that can affect a short period of time, and processes that can have an impact over a long period of time. Since it is a characteristic of the Systematic method to compose a scenario by combining these FEP, the Systematic method is the basic premise for the development of FEP. Completeness is important for FEP, and comprehensiveness is important for scenarios. However, combining all the FEP into one scenario is time-consuming and difficult to ascertain the comprehensiveness of the scenario. Therefore, an Integrated FEP list is being developed to facilitate tracking between FEP and scenarios by integrating similar FEP. In this study, during the integrated FEP development process, a method for utilizing experts that can be used for difficult parts of quantitative evaluation and a quantitative evaluation process through the method were presented.

In Korea, borated stainless steel (BSS) is used as a storage rack in spent fuel pools (SFP) to maintain the nuclear criticality of spent fuels. As the number of nuclear power plants and the corresponding amount of spent fuels increased, the density in SFP storage rack also increased. In this regard, maintaining subcriticality of spent nuclear fuels became an issue and BSS was selected as the structural material and neutron absorber for high density storage rack. Since it is difficult to replace the storage rack, corrosion resistance and neutron absorbency are required for long period. BSS is based on stainless steel 304 and 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 the low solubility of boron in austenitic stainless steel, metallic borides such as (Fe, Cr) 2B are formed as a secondary phase. Metallic borides could cause Cr depletion near it, which could decrease the corrosion resistance of the material. In this paper, the long-term corrosion behavior of BSS and its oxide microstructures are investigated through accelerated corrosion experiment in simulated SFP conditions. Because the corrosion rate of austenitic stainless steel is known to be dependent on the Arrhenius equation, a function of temperature, the corrosion experiment is conducted by increasing the experimental temperature. Detail microstructural analysis is conducted using a scanning electron microscope, transmission electron microscope and energy dispersive spectrometer. After oxidation, a hematite structure oxide film is formed, and pitting corrosion occurs on the surface of specimens. Most of the pitting corrosion is found at the substrate surface because the corrosion resistance of the substrate, which has low Cr content, is relatively low. Also, the 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 the boron content and the neutron absorption ability of the material. Using boron’s high cross-section for neutrons, the neutron absorption performance of BSS was evaluated through neutron transmission tests. The effect of the corrosion behavior of BSS on its neutron absorption performance was investigated. Samples simulated to undergo up to 60 years of degradation before corrosion through accelerated corrosion testing did not show significant changes in the neutron shielding ability before and after corrosion. This can be explained in relation to the corrosion behavior of BSS. Boron was only leached out from the secondary phase exposed on the surface, and this oxidized secondary phase corresponds to about 0.17% of the volume of the total secondary phase. This can be seen as a very small proportion compared to the total boron content and is not expected to have a significant impact on neutron absorption performance.

Molten Salt Reactor (MSR) is one of Generation-IV nuclear reactors that uses molten salts as a fuel and coolant in liquid forms at high temperatures. The advantages of MSR, such as safety, economic feasibility, and scalability, are attributed from the fact that the molten salt fuel in a liquid state is chemically stable and has excellent thermo-physical properties. MSR combines the fuel and coolant by dissolving the actinides (U, Th, TRU, etc.) in the molten salt coolant, eliminating the possibility of a core meltdown accident due to loss of coolant (LOCA). Even if the molten salt fuel leaks, the radioactive fission products dissolved in the molten salt will solidify with the fuel salt at room temperature, preventing potential leakage to the outside. MSR was first demonstrated at ORNL starting with the Aircraft Reactor Experiment (ARE) in 1954 and was extended to the 7.4 MWth MSRE developed in 1964 and operated for 5 years. Recently, various start-ups, including TerraPower, Terrestrial Energy, Moltex Energy, and Seaborg, have been conducting research and development on various types of MSR, particularly focusing on its inherent safety and simplicity. While in the past, fluoride-based molten salt fuels were used for thermal neutron reactors, recently, a chlorine-based molten salt fuel with a relatively high solubility for actinides and advantageous for the transmutation of spent nuclear fuel and online reprocessing has been developing for fast neutron spectrum MSRs. This paper describes the development status of the process and equipment for producing highpurity UCl3, a fuel material for the chlorine-based molten salt fuel, and the development status of the gas fission product capturing technologies to remove the gaseous fission products generated during MSR operation. In addition, the results of the corrosion property evaluation of structural materials using a natural circulation molten salt loop will also be included.

To investigate the effect of the catalyst and metal–support interaction on the methane decomposition behavior and physical properties of the produced carbon, catalytic decomposition of methane (CDM) was studied using Ni/SiO2 catalysts with different metal–support interactions (synthesized based on the presence or absence of urea). During catalyst synthesis, the addition of urea led to uniform and stable precipitation of the Ni metal precursor on the SiO2 support to produce Ni-phyllosilicates that enhanced the metal–support interaction. The resulting catalyst upon reduction showed the formation of uniform Ni0 particles (< 10 nm) that were smaller than those of a catalyst prepared using a conventional impregnation method (~ 80 nm). The growth mechanisms of methane-decomposition-derived carbon nanotubes was base growth or tip growth according to the metal–support interaction of the catalysts synthesized with and without urea, respectively. As a result, the catalyst with Ni-phyllosilicates resulting from the addition of urea induced highly dispersed and strongly interacting Ni0 active sites and produced carbon nanotubes with a small and uniform diameter via the base-growth mechanism. Considering the results, such a Ni-phyllosilicate-based catalyst are expected to be suitable for industrial base grown carbon nanotube production and application since as-synthesized carbon nanotubes can be easily harvested and the catalyst can be regenerated without being consumed during carbon nanotube extraction process.

Coking coal is an important raw material for coke production. In this study, in an inert atmosphere, two Chinese coking coal samples were, respectively, heated gradually to 1200 °C to release volatile and form char and coke in succession, then cooled naturally to close room temperature to age the coke. The whole heating and cooling process on carbonization were monitored in situ by simultaneous small and wide-angle X-ray scattering (SAXS-WAXS) technique based on a synchrotron radiation platform. The simultaneous structural changes of pore and skeleton in coal during carbonization are revealed for the first time. The two raw coal samples, with similar carbon content and slightly different coalification degree, undergone a carbonization process similar in whole and different in parts. The carbonization presents approximately three stages during heating process and one stage during cooling process. The coal structure changes wavily during heating and monotonously during cooling. The corresponding structural change mechanism is analyzed.