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

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.

Nuclear spent fuel (SNF) disposal in deep geological repositories is considered as one of sound options for the long-term and safe sequestration of radiotoxic SNF and the sustainable use of nuclear energy. The chemical behaviors of various radionuclides originated from SNF should be well understood to evaluate the migrational behaviors of radionuclides and their reactions and interactions with various geochemical components. Formation of secondary minerals, colloids, other insoluble precipitates is of interest since the concentrations of radionuclides in groundwaters can be limited by the solubility of those solid phases. Particularly when evaluating their solubility, the use of well-defined solid materials in terms of chemical composition and molecular structure is crucial to obtain reliable measurement results. In this study, a synthetic calcium uranyl silicate (Ca-U(VI)-silicate, or uranophane) was prepared and characterized by using various analytical methods including powder X-ray diffraction (pXRD), scanning electron microscopy/energy dispersive X-ray spectrometry (SEM/EDX), and vibrational (FTIR and Raman) spectroscopies. Uranyl silicate minerals are significant to the disposal of nuclear wastes. Our simulation demonstrates that uranophane (Ca[UO2SiO3OH]2·5H2O), one having a U:Si ratio of 1:1, can be a mineral species limiting U(VI) solubility under groundwater conditions in Korea. For the preparation of Ca-U(VI)-silicate, we applied a two-step hydrothermal synthetic procedure reported in literature with modification. Briefly, we conclude that the obtained mineral phase is the ‘α-uranophane’; our characterization results show that the structural and spectroscopic properties of the synthetic Ca-U(VI)-silicate agree well with those of α-uranophane. For instance, the pXRD patterns obtained from the solid show nearly identical diffraction peak positions with those from the reference XRD pattern. From IR and Raman spectroscopy it is noticed that the stretching modes of UO2 2+ and SiO4 4- ions result in strong absorption bands in a region of 700 ~ 1,100 cm-1. Elemental compositions of the synthetic solids were also estimated by using EDX analysis, which results in a Ca:U:Si ratio close to 1:2:2 on average. However, we found that it is difficult to obtain good crystallinity of uranophane, which can be observable by using SEM and its image analysis. We believe that this work serves as a model study to provide synthetic routes of radionuclide-related mineral phases and applicable solid phase characterization methods. In the presentation, the potential use of the U(VI)-silicate solid phase for the upcoming groundwater solubility measurements will be discussed. Keywords: Hexavalent Uranium, Silicate

Kr-85 has a half-life of 10.7 years and it stays in the atmosphere for a long time. However it does not accumulate as an noble gas but only emits beta particles. Therefore its contribution to environmental radiation dose is lower than any other radionuclides. Kr-85 is one of the main fission products produced by nuclear fission reaction and artificial radionuclide that does not exist in nature. For these reasons, monitoring Kr-85 from the atmosphere is meaningful so that the nuclear-related facilities are recommended to control and regulate environmental emissions. Post Irradiation Examination Facility (PIEF) which located in KAERI is a facility that conducts various material and chemical experiments using the irradiated nuclear fuels. Therefore, various radionuclides can present in gaseous effluent including Kr-85. To prevent the environmental hazards and guarantee the radiation safety of the public, nuclear facilities are recommended to be equipped with stack radiation/radioactivity monitoring system, so that the Kr-85 concentration in gaseous effluent is controlled within the regulatory criteria. Particularly, the Kr-85 concentration of gaseous effluent is commonly monitored by the stack monitoring system connected to the process ventilation system from the hot cell. The monitoring system supply the information such as beta count rate, dose rate and flow rate, etc. Due to the concentration of Kr-85 in gaseous effluent is subject to regulatory guide lines, a systemized procedure for calculating Kr-85 concentration of the stack exhaust is necessary. Furthermore, the emission should be monitored whether it satisfies the regulatory standard or does not. This paper performed discussion on the process of calculating the concentration of Kr-85 in the gaseous effluent of PIEF stack from the monitoring system (NGM209, MGP), and the amount of Kr- 85 over the last 2 years emissions was calculated. In addition to calculating effluent rate of radioactive Kr-85, the Minimum Detectable Concentration (MDC) and Decision Threshold (SD) were calculated. As a result, the calculated Kr-85 concentration was below the SD during the entire period. It is considered that there are no environmental emissions of Kr-85.

Surface contaminants may attach to surfaces or objects in the radiation controlled area to cause radiation exposure, or spread out to the general environment by person and object exiting the radiation workplace. Accordingly, in radiological safety control, surface contamination monitoring is one of the important factors in workplace monitoring. When obtaining the measurement results for the monitoring, the results are accompanied by uncertainty since measurements contain numerous errors. Accordingly, the International Organization for Standardization (ISO) has published the ISO 7503 series which is comprehensive and detailed guidelines on the measurement and evaluation of surface contamination. ISO 7503-3 especially presents a mathematical model for the contamination measurement and provide calculation guidelines on measurement uncertainty evaluation, decision threshold and detection limit. This paper is focused on reevaluating and comparing the surface contamination monitoring method applied to radiation safety management practice and its results based on the measurement and evaluation method set by the International Organization for Standardization. The evaluation was performed in accordance with ISO 7503, and the current reporting method for measurement results was compared with the method recommended in ISO 11929 publication.

Spent nuclear fuel (SNF) is the main source of high-level radioactive wastes (HLWs), which contains approximately 96% of uranium (U). For the safe disposal of the HLWs, the SNF is packed in canisters of cast iron and copper, and then is emplaced within 500 m of host rock surrounded by compacted bentonite clay buffer for at least 100,000 years. However, in case of the failure of the multi-barrier disposal system, U might be migrated through the rock fractures and groundwater, eventually, it could reach to the biosphere. Since the dissolved U interacts with indigenous bacteria under natural and engineered environments over the long storage periods of geologic disposal, it is important to understand the characteristics of U-microbe interactions under the geochemical conditions. In particular, a few of bacteria, i.e., sulfate-reducing bacteria (SRB), are able to reduce soluble U(VI) into insoluble U(IV) under anaerobic conditions by using their metabolisms, resulting in the immobilization of U. In this study, the aqueous U(VI) removal performance and change in bacterial community in response to the indigenous bacteria were investigated to understand the interactions of U-microbe under anaerobic conditions. Three different indigenous bacteria obtained from different depths of granitic groundwater (S1: 44–60 m, S2: 92–116 m, and S3: 234–244 m) were used for the reduction of U(VI)aq. After the anaerobic reaction of 24 weeks, the changes in bacterial community structure in response to the seeding indigenous bacteria were observed by high-throughput 16S rDNA gene sequencing analysis. The highest uranium removal efficiency of 57.8% was obtained in S3 sample, and followed by S2 (43.1%) and S1 (37.7%). Interestingly, SRB capable of reducing U(VI) greatly increased from 4.8% to 44.1% in S3 sample. Among the SRB identified, Thermodesulfovibrio yellowstonii played a key role on the removal of U(VI)aq. Transmission electron microscopy (TEM) analysis showed that the dspacing of precipitates observed in this study was identical with that of uraninite (UO2). This study presents the potential of U(VI)aq removal by indigenous bacteria under deep geological environment.

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.