Boric acid-containing B-10 is used in a nuclear reactor as a coolant and absorbs thermal neutrons generated during nuclear fission in the primary circuit. Boron-containing coolant water waste is generated from maintenance, floor drain, decontamination, and reactor letdown flows. There are two options for aqueous solution waste of boric acid. One is recycling and discharge through filtration, ion exchange, and reverse osmosis. The other is immobilization after evaporation and crystallization processes. The dry powder of boric acid waste liquid can be immobilized by cement, polymer, etc. Before the mid-1990s, concentrated boric acid waste was solidified with a cement matrix. To overcome the disadvantage of low waste loading of cement waste form, a method of solidifying with paraffin was adopted. However, paraffin solids were insufficient to be disposed of as final waste. Paraffin is a kind of soft solidified material and has low compressive strength and poor leaching resistance. As a result, it was decided as an unsuitable form for disposal. In KOREA, paraffin waste form was adopted for boric acid waste treatment in the 1990s. A large amount of paraffin waste forms about 20,000 drums (200 l drum) were generated to treat boric acid waste and were stored in nuclear power sites without disposal. In this study, we want to obtain high-purity boric acid waste by oxidizing and decomposing solid paraffin waste form through a boric acid catalytic reaction. In this reaction, paraffin is separated in the form of various by-products, which can then be treated through a liquid waste treatment device or an exhaust gas treatment device. The proper temperature for sample decomposition during the catalytic reaction was set through TGA analysis. Compositions of by-products and residues generated at each stage of the reaction could be analyzed to determine the state during the reaction. Finally, the boric acid waste powder was perfectly separated from paraffin waste form with disposable products through this pyrolysis process.

In 2017, Kori unit 1 nuclear power plant was permanently shut down at the end of its life. Currently, Historical Site Assessment (HSA) for MARSSIM characteristics evaluation is being conducted according to the NUREG-1575 procedure, this is conducted through comprehensive details such as radiological characteristics preliminary investigation and on-site interview. Thus, the decommissioning of nuclear power plant must consider safety and economic feasibility of structures and sites. For this purpose, the establishment of optimal work plan is required which simulations in various fields. This study aims to establish procedure that can form a basis for a rational decommissioning plan using the virtual nuclear power plant model. The mapping procedure for 3D platform implementation consisted of three steps. First, scan the inside and outside of the nuclear power plant for decommissioning structure analysis, 3D modeling is performed based on the data. After that, a platform is designed to directly measure the radiation dose rate and mapped the derived to the program. Finally, mapping the radiation dose rate for each point in 3D using the radiation dose rate calculation factor according to the time change the measured value created on the 3D mapping platform. When the mapping is completed, it is possible to manage the exposure dose of workers according to the ALARA principle through the charge of radiation dose rate over time because of visualization of the color difference to the radiation dose rate at each point. For addition, the exposure dose evaluation considering the movement route and economic feasibility can be considered using developed program. As the interest in safety accidents for workers increases, the importance of minimum radiation dose and optimal work plan for workers is becoming increasingly important. Through this mapping procedure, it will be possible to contribute to the establishment of reasonable process for dismantling nuclear power plant in the future.

Maintaining fuel sheath integrity during dry storage is important. Intact sheath acts as the primary containment barrier for both fuel pellets and fission products over the dry storage periods and during subsequent fuel handling operations. In KNF, in-house fuel performance code was developed to predict the overall behavior of a fuel rod under normal operating conditions. It includes the analysis modules to predict temperature, pellet cracking and deformation, sheath stress and strain at the mid-plane of the pellet and pellet-pellet interfaces, fission gas release and internal gas pressure. The main focus of the code is to provide information on initial conditions prior to dry storage, such as fission gas inventory and its distribution within the fuel pellet, initial volumes of storage spaces and their locations, radial profile of heat generation within the pellet, etc. To upgrade the developed code that address all the damage mechanisms, the first step was a review of the available technical information on phenomena relevant to fuel integrity. Potential degradation mechanisms that may affect sheath integrity of CANDU spent fuel during dry storage are: creep rupture under internal gas pressure, sheath oxidation in air environment, stress corrosion cracking (SCC), delayed hydride cracking (DHC), and sheath splitting due to UO2 oxidation for a defective fuel. The failure by creep rupture, SCC or DHC is in the form of small cracks or punctures. The failure by sheath oxidation or sheath splitting due to UO2 oxidation results in a gross sheath rupture. The second step was to examine the technical bases of all modules of the in-house code, identify and extend the ranges of all modules to required operating ranges. This step assessed the degradation mechanisms for the fuel integrity. The objective of this assessment is to predict the probability of sheath through-wall failure by a degradation mechanisms as a function of the sheath temperature during dry storage. Further improvements being considered include upgrades of the analysis module to achieve sufficient accuracy in key output parameters. The emphasis in the near future will be on validation of the inhouse code according to a rigorous and formal methodology. The developed models provide a platform for research and industrial applications, including the design of fuel behavior experiments and prediction of safe operating margins for CANDU spent fuel.

Korea Research Reactor 1&2 (KRR-1&2), Korea’s first research reactor, began dismantling in 1997. As of 2022, the demolition of general areas such as offices has been completed, and contaminated areas such as reactor rooms remain. On the other hand, construction waste generated in contaminated areas of nuclear facilities cannot be disposed of as general industrial waste. It is predicted that about 5,000 tons of construction waste will be generated if the contaminated area of KRR-1&2 is demolished. In this study, the application plan for the demolition of contaminated area of KRR-1&2 was reviewed through a review of laws and cases related to domestic and overseas disposal. The only method for disposing of construction waste in contaminated areas that can be applied in Korea is clearance in accordance with Nuclear Safety Commission Notice No. 2020-06. In addition, there has been no case of demolishing large-scale nuclear facilities in Korea. Therefore, there are limitations in domestic laws and standards to be applied to the dismantling of contaminated areas of KRR-1&2. The IAEA and the United States specify comprehensive matters such as optimization of radiation protection and minimization of waste products. The EU recommends demolition after decontamination by removing contaminated areas before demolition of buildings. It also presents three options for reuse, recycling, and disposal of buildings and building waste. In particular, in the case of Germany, detailed radioactivity measurement methods for deregulation of buildings and building waste are presented in accordance with the EU’s guidelines. As a result of synthesizing this, it is judged that the EU and Germany building clearance plan will be suitable for domestic application.

The International Atomic Energy Agency recommends the deep geological disposal system as one of the disposal methods for high-level radioactive waste (HLW), such as spent nuclear fuel. The deep geological disposal system disposes of HLW in a deep and stable geological formation to isolate the HLW from the human biosphere and restrict the inflow of radionuclides into the ecosystem. It mainly consists of an engineered barrier and a natural barrier. Safety evaluation using a numerical model has been performed primarily to evaluate the buffer’s long-term stability. However, although the gas generation rate input for long-term stability evaluation is the critical factor that has the most significant influence on the long-term hydraulic-mechanical behavior of the buffer, in-depth research and experimental data are lacking. In this study, the gas generation rate on the interface between the disposal canister and the buffer material, a component of the engineered barrier, was mainly studied. Gas can be generated between the disposal canister and the buffer material due to various causes such as anaerobic corrosion of the disposal canister metal, organic matter decomposition, radiation decomposition, and steam generation due to high temperature. The generation of gas in such a disposal environment increases the pore gas pressure in the buffer and causes internal cracks. The occurred cracks increase the intrinsic permeability of the buffer, which leads to a decrease in the primary performance of the buffer. For this reason, it is essential to apply the appropriate gas generation rate according to the disposal condition and buffer material for accurate long-term stability analysis. Therefore, the theoretical models regarding the estimation of gas generation were summarized through a literature study. The amount of gas generated was estimated according to the disposal environment and material of the disposal canister. It is expected that estimated values might be used to estimate the long-term stability analysis of buffer performance according to the disposal condition.

Prior to the investigations on fuel degradation it is necessary to describe the reference characteristics of the spent fuel. It establishes the initial condition of the reference fuel bundle at the start of dry storage. In a few technology areas, CANDU fuels have not yet developed comprehensive analysis tools anywhere near the levels in the LWR industry. This requires significantly improved computer codes for CANDU fuel design. In KNF, in-house fuel performance code was developed to predict the overall behavior of a fuel rod under normal operating conditions. It includes the analysis modules to predict temperature, pellet cracking and deformation, clad stress and strain at the mid-plane of the pellet and pellet-pellet interfaces, fission gas release and internal gas pressure. The main focus of the code is to provide information on initial conditions prior to dry storage, such as fission gas inventory and its distribution within the fuel pellet, initial volumes of storage spaces and their locations, radial profile of heat generation within the pellet, etc. Potential degradation mechanisms that may affect sheath integrity of CANDU spent fuel during dry storage are: creep rupture under internal gas pressure, sheath oxidation in air environment, stress corrosion cracking, delayed hydride cracking, and sheath splitting due to UO2 oxidation for a defective fuel. To upgrade the developed code that address all the damage mechanisms, the first step was a review of the available technical information on phenomena relevant to fuel integrity. The second step was an examination of the technical bases of all modules of the in-house code, identify and extend the ranges of all modules to required operating ranges. Further improvements being considered include upgrades of the analysis module to achieve sufficient accuracy in key output parameters. The emphasis in the near future will be on validation of the in-house code according to a rigorous and formal methodology. The developed models provide a platform for research and industrial applications, including the design of fuel behavior experiments and prediction of safe operating margins for CANDU spent fuel.

The purpose of this study is to evaluate dogs' sociality toward human strangers in the absence of an owner by analyzing changes in dogs' behavior during a task of making eye contact with an experimenter to obtain snacks. A total of 17 dogs were divided into groups of high sociality (HS; n = 10, 4.4 ± 3.87 years) and low-sociality (LS; n = 7, 3.71 ± 2.06 years). A comparison of the average frequency of five behavioral types-fear-appeasement behaviors (P<0.001), sociability-related behaviors (P<0.001), stress-related behaviors (P<0.05), destruction (P < 0.001), and vocalization (P < 0.001)-between the groups showed a significant difference in all five categories. Together, these results suggest that dogs with high sociality are less exposed to various stresses and have a higher ability to adapt to new environments than dogs with low sociality. This can predict dogs' adaptability to a new environment and positive outcomes in their daily life with the owner.

This study aimed to evaluate the effects of four types of environmental enrichment on the improvement of companion dogs' behavioral problems due to separation anxiety. A total of 21 dogs of various breeds were included in the study. Data were collected to investigate the behaviors associated with anxiety in dogs, including vocalization, elimination, escape attempts, and destructiveness. A first stage, in which the dog and owner were together (P0), lasted 15 min, and a second stage, in which the dog and owner were separated (P1), lasted 15 min. After the dog and owner were separated (P1), the third stage (P2), during which the environment was enriched, lasted 20 min, and the fourth stage, following environment enrichment (P3), lasted 15 min. The results of the study indicated that compared to P0, the frequency of problematic behavior was highest during the 15 min following separation from the owner (P1). Following environmental enrichment, the average frequency of problematic behaviors in P2 decreased (P < 0.001) compared to P1. Environmental enrichment can also be used appropriately in the case of companion dogs, including shelter dogs or experimental dogs that use a limited kennel, and is a particularly effective means of improving the quality of life of dogs.

In this study, strawberry cultivation environment in a greenhouse located in Jeonju was monitored and internal environmental parameters were analyzed. Temperature, humidity, RAD, and PPF sensors were installed to monitor environmental conditions in the test greenhouse. Data were collected every 10 minutes during four winter months from sensors placed across the greenhouse to assess its permeability and environmental uniformity. Temperature and humidity inside the greenhouse were relatively uniform with negligible deviations among the center, south, and north; however, it was judged that further analysis of gradients of these parameters f rom the east to t he w est of t he g reenhouse w ould b e needed. Both R AD (Total solar radiation) a nd P PF (Photosynthetic photon flux) had high values on the south and were low on the north and the reduction rate of these parameters was 54% and 61%, respectively, indicating that a significant amount of light could not be transmitted. This implied a significant decrease in the amount of light entering the greenhouse during winter. Therefore, it is concluded that environmental control devices and auxiliary lighting are needed to achieve uniform greenhouse environment for efficient strawberry cultivation.

Red ginseng marc (RGM) has been used on primary industries using fertilizer or forage, and it mostly has been dumped. To improve utilization of RGM, the biological activities of RGM were examined. RGM was extracted and fractionated using various solvents and their biological activities were compared. The hexane fraction from the methanol extract of RGM (RGMMH) showed strong anti-cancer activity (58.56 ± 6.04% at 100 ㎍/mL) and anti-inflammatory effect (65.72 ± 1.33% at 100 ㎍/mL). But, oil extract of RGM extracted with hexane (RGMH) showed low activities (anti-cancer: 16.42 ± 3.33%, at 100 ㎍/mL, anti-inflammatory activity: 29.46 ± 2.10%, at 100 ㎍/mL). Their metabolites were analyzed using HPLC. Panaxydol known as anti-cancer compound of RGM was one of major compounds in RGMMH. Meanwhile, panaxydol was detected in trace amount in red ginseng marc oil (RGMH). In addition, RGMMH and RGMH showed big differences in HPLC profiling. This research suggests optimal extraction method of RGM oil.