Derived Concentration Guideline Levels (DCGLs), which represent the residual radioactivity concentration limits, serve as the pivotal criteria for decontamination during decommissioning of nuclear power plants and are essential for license termination. The analysis of radionuclides in various media to check site-specific and radionuclide-specific DCGLs is a resource-intensive and time-consuming processes, and there are some radionuclides that are hard to analyze. In the decommissioning of the Rancho Seco nuclear power plant in the United States, a conservative approach was adopted. Potentially highly contaminated areas on the site were identified by collecting and analyzing soil samples, and radionuclides exceeding the Minimum Detectable Concentration (MDC) were selected as the potential Radionuclide of Concern (ROC), and surrogate DCGLs for hard-to-detect radionuclides were applied to soil samples. For soil samples in the Rancho Seco nuclear power plant, Cs-137 contributed more than 90% of the total radioactivity. DCGLs of the ROC were obtained using the scaling factors through analysis of Cs- 137 for a large amount of soil samples. In Korea, the scaling factor methodology has not been applied to the decommissioning of commercial nuclear power plants. An initial investigation was undertaken to assess the viability of implementing Surrogate Derived Concentration Guideline Levels (DCGLs) in the dismantling of Kori Unit 1, drawing insights from the U.S. nuclear power plant decommissioning experiences. To do this approach, the concentration ratio of radionuclides of interest to key radionuclide in contaminated soil should be known and consistent. But related information is not available at this time. So Surrogate DCGL for representative C-14, Fe-55, Ni-59, Ni-63, and Sr-90 was obtained using the scaling factors applied to radioactive waste data, specifically Decontaminated Aqueous Waste (DAW) and Spent Resin. In order to develop a reliable surrogate DCGLs the Kori Unit 1 site, it is important to analyze the radionuclides in the soil for the Kori Unit 1 decommissioning site to obtain consistent concentration ratio of the radionuclides of concern to the key radionuclides. When a the suitable DCGL is developed, it can be used for FSS planning and prior decision-making ensuring the safe and effective decommissioning of Kori Unit 1 and similar nuclear power plants.

The radioactive waste generated within radiation-controlled areas is classified and processed according to relevant laws and regulations based on contamination levels. In cases where such radioactive waste complies with the legally defined clearance concentration or dose criteria, it is disposed of as non-radioactive waste by means of incineration, reclamation, recycling, etc. Within radiation controlled areas, various consumables are periodically replaced to ensure the proper operation of the area. It is necessary to have appropriate disposal methods for these consumables. In particular, waste items such as fire extinguishers, fluorescent lamps, batteries, and pressure vessels (hereinafter referred to as “Special Waste Type”), which may contain hazardous substances within their internal components and contents, should be considered for appropriate disposal methods that comply with nuclear safety and environmental laws. In the present case, the specified special waste type do not come into direct contact with radiation sources, and they have impermeable surfaces, which significantly reduces the risk of external contamination infiltrating the interior. However, the current method of clearance is not suitable for these items (Typically, nuclear energy-related business operators are required to classify clearance target waste based on internal and external components and demonstrate compliance with the criteria. Nevertheless, for special waste type, it is difficult to separate and measure internal and external components within the radiation-controlled area). In this case, the Clearance Procedure for special waste type applied to Korea Atomic Energy Research Institute was introduced. Additionally, we have extracted considerations for future domestic clearance of the type.

To evaluate the inventory of radionuclides for the disposal of waste generated from nuclear power plants, indirect assessment methods such as the scaling factor method or average radioactivity concentration method can be applied. A scaling factor represents the average concentration ratio between key radionuclides and difficult-to-measure (DTM) radionuclides, while the average radioactivity concentration refers to the average concentration of DTM radionuclides, regardless of the concentration of key radionuclides or within specific ranges of key radionuclide concentrations. These indirect assessment methods can be statistically derived through the analysis of representative drums. This study will address how to apply these scaling factors and average radioactivity concentrations. Firstly, the concentration of gamma-emitting radionuclides will be analyzed using a drum radionuclide analyzer, and the concentration of DTM radionuclides will be determined by applying scaling factors specific to each DTM radionuclide. In the case of using the average radioactivity concentration method, the average concentration of DTM radionuclides will be applied independently of the concentration of gamma-emitting radionuclides. It is crucial to perform radioactive decay correction based on the date of generation or disposal when applying scaling factors or average radioactivity concentration. Additionally, for repackaged 320 L drums, determining which drum among the two 200 L drums inside should serve as the reference is of utmost importance

The radwaste repository consists of a multi-barrier, including natural and engineered barriers. The repository’s long-term safety is ensured by using the isolation and delay functions of the multi-barrier. Among them, natural barriers are difficult to artificially improve and have a long time scale. Therefore, in order to evaluate its performance, site characteristics should be investigated for a sufficient period using various analytical methods. Natural barriers are classified into lithological and structural characteristics and investigated. Structural factors such as fractures, faults, and joints are very important in a natural barrier because they can serve as a flow path for groundwater in performance evaluation. Considering the condition that the radioactive waste repository should be located in the deep part, the drill core is an important subject that can identify deep geological properties that could not be confirmed near the surface. However, in many previous studies, a unified method has not been used to define the boundaries of structural factors. Therefore, it is necessary to derive a method suitable for site characteristics by applying and comparing the boundary definition criteria of various structural factors to boreholes. This study utilized the 1,000 m deep AH-3 and DB-2 boreholes and the 500 m deep AH-1 and YS- 1 boreholes drilled around the KURT (KAERI Underground Research Tunnel) site. Methods applied to define the brittle structure boundary include comparing background levels of fracture and fracture density, excluding sections outside the zone of influence of deformation, and confining the zone to areas of concentrated deformation. All of these methods are analyzed along scanlines from the brittle structure. Deriving a site-specific method will contribute to reducing the uncertainties that may arise when analyzing the long-term evolution of brittle structures within natural barriers.

Advanced countries in the field of nuclear research and technology are currently examining the feasibility of deep geological disposal as the most appropriate method for the permanent management of high-level radioactive waste, with no intention of future retrieval. Deep geological disposal involves the placement of such waste deep underground within a stable geological formation, ensuring its permanent isolation from the human environment. To guarantee the enduring isolation and retardation of radionuclides with half-lives spanning tens of thousands to millions of years from the broader ecosystem, it is imperative to comprehend the long-term evolution of deep disposal systems, especially the role of natural barriers. These natural barriers, typically consisting of bedrock, encase the repository and undergo long-term evolutions due to tectonic movements and climate variations. For the effective disposal of high-level radioactive waste, a thorough assessment of the site’s long-term geological stability is essential. This necessitates a comprehensive understanding of its tectonic evolution and development characteristics, including susceptibility to seismic and magmatic events like earthquakes and intrusions. Furthermore, a detailed analysis of alterations in the hydrogeological and geochemical environment resulting from tectonic movements over extended time frames is required to assess the potential for the migration of radionuclides. In this paper, we have examined international evaluation methodologies employed to elucidate the predictive long-term evolution of natural barriers within disposal systems. We have extracted relevant methods from international case studies and applied a preliminary scenario illustrating the long-term evolution of the geological environment at the KURT (KAERI Underground Research Tunnel) site. Nevertheless, unlike international instances, the scarcity of quantitative data limits the depth of our interpretation. To present a dependable scenario in the future, it is imperative to develop predictive technologies aimed at comprehensively studying the geological evolution processes in the Korean peninsula, particularly within the context of radioactive waste disposal.

The high-level radioactive waste repository must ensure its performance for a long period of time enough to sufficiently reduce the potential risk of the waste, and for this purpose, multibarrier systems consisting of engineered and natural barrier systems are applied. If waste nuclides leak, the dominating mechanisms facilitating their movement toward human habitats include advection, dispersion and diffusion along groundwater flows. Therefore, it is of great importance to accurately assess the hydrogeological and geochemical characteristics of the host rock because it acts as a flow medium. Normally, borehole investigations were used to evaluate the characteristics and the use of multi-packer system is more efficient and economical compared to standpipes, as it divides a single borehole into multiple sections by installing multiple packers. For effective analyses and groundwater sampling, the entire system is designed by preselecting sections where groundwater flow is clearly remarkable. The selection is based on the analyses of various borehole and rock core logging data. Generally, sections with a high frequency of joints and evident water flow are chosen. Analyzing the logging data, which can be considered continuous, gives several local points where the results exhibit significant local changes. These clear deviations can be considered outliers within the data set, and machine learning algorithms have been frequently applied to classify them. The algorithms applied in this study include DBSCAN (density based spatial clustering of application with noise), OCSVM (one class support vector method), KNN (K nearest neighbor), and isolation forest, of which are widely used in many applications. This paper aims to evaluate the applicability of the aforementioned four algorithms to the design of multi-packer system. The data used for this evaluation were obtained from DB-2 borehole logging data, which is a deep borehole locates near KURT.

The objective of this study is development of graphite-boron composite material as a replacement for metal canisters to Improve the heat dissipation and radiation shielding performance of dry spent nuclear fuel storage system and reduce the volume of waste storage system. KEARI research team plan to use the graphite matrix manufacturing technology to pelletize the graphite matrix and adjust the content of phenolic resin binder to minimize pore formation. Specifically, we plan to adjust the ratio of natural and synthetic graphite powder and use uniaxial pressing technology to manufacture black graphite matrix with extremely high radial thermal conductivity. After optimizing the thermal conductivity of the graphite matrix, we plan to mix it with selected boron compounds, shape it, and perform sintering and purification heat treatments at high temperatures to manufacture standard composite materials.

Dry storage of nuclear fuel is compromised by threats to the cladding integrity, such as creep and hydride reorientation. To predict these phenomena, spent fuel simulation codes have been developed. In spent fuel simulation, temperature information is the most influential factor for creep and hydride formation. Traditional fuel simulation codes required a user-defined temperature history input which is given by separate thermal analysis. Moreover, geometric changes in nuclear fuel, such as creep, can alter the cask’s internal subchannels, thereby changing the thermal analysis. This necessitates the development of a coupled thermal and nuclear fuel analysis code. In this study, we integrated the 2D FDM nuclear fuel code GIFT developed at SNU with COBRA -SFS. Using this, we analyzed spent nuclear stored in TN-24P dry storage cask over several decades and identified conditions posing threats due to phenomena like creep and hydrogen reorientation, represented by the burnup and peak cladding temperature at the start of dry storage. We also investigated the safety zone of spent nuclear fuel based on burnup and wet storage duration using decay heat.

Considering the domestic situation where all nuclear power plants are located on seaside, the interim storage site is also likely to be located on coastal site. Maritime transportation is inevitable and the its risk assessment is very important for safety. Currently, there is no independently developed maritime transportation risk assessment code in Korea, and no research has been conducted to evaluate the release of radioactive waste due to the immersion of transport cask. Previous studies show that the release rate of radionuclides contained in a submerged transport cask is significantly affected by the area of flow path generated at the breached containment boundary. Due to the robustness of a cask, the breach is the most likely generated between the lid and body of cask. CRIEPI investigated the effect of cask containment on the release rate of radioactive contents into the ocean and proposed a procedure to calculate the release rate considering the socalled barrier effect. However, the contribution of O-ring on the release rate was not considered in the work. In this study, test and analysis is performed to determine the equivalent flow path gap considering the influence of O-rings. These results will be implemented in the computational model to assess sea water flow through a breached containment boundary using CFD techniques to assess radionuclide release rates. To evaluate the release rate as a function of lid displacement, a small containment vessel is engineered and a metal O-ring of the Helicoflex HN type is installed, which is the most commonly used one in transport and storage casks. The lid of containment vessel is displaced in vertical and horizontal direction and the release rate of the vessel was quantified using the helium leak test and the pressure drop test. Through this work, the relationship between the vertical opening displacement and horizontal sliding displacement of the cask lid and the actual flow path area created is established. This will be implemented in the CFD model for flow rate calculation from a submerged transport cask in the deep sea. In addition, the compression of the O-ring causes very small gaps, such as capillaries. In these cases, Poiseuille’s law is used to calculate the capillary flow rate.

An austenitic stainless steel canister functions as a containment barrier for spent nuclear fuel and radioactive materials. The canister on the spent fuel storage system near the coastal area has several welding lines in the wall and lid, which have high residual tensile stresses after welding procedure. Interaction between austenitic stainless steel and chloride environment from a sea forms a detrimental condition causing chloride induced stress corrosion cracking (CISCC) in the canister. The South Korea is concerned with the dry storage of high-level spent nuclear fuel and radioactive wastes to be built on the site of a nuclear power plant. The importance of aging management has recently emerged for mitigating CISCC of dry storage canisters. When a corrosive pit is created by a localized corrosion in a sea water atmosphere, it initiates and grows as CISCC crack. Surface stress improvement works by inducing plastic strain which results in elastic relaxation that generates residual compressive stress. Surface stress improvement methods such as roller burnishing process can effectively mitigate the potential for CISCC of the canister external surfaces. The generation of compressive stress layer can inhibit the transition to cracking initiation. In this study, a flat roller burnishing process was applied as a prevention technology to CISCC of stainless steel canisters. Roller burnishing process parameters have been selected for 1:3 scale canister model having a diameter of 600 mm, a length of 1,000 mm and a thickness of 10 mm on the basis of the burnishing conditions available to control residual tensile stress of austenitic stainless steel plate specimens. The surface roughness of the scaled canister model was investigated using a surface roughness measurement equipment after roller burnishing treatment. The surface residual stresses of the scaled canister model were measured by a hole drilling contour method attached with strain gauge. The burnishing test results showed that the surface roughness of the scaled canister model was considerably improved with flat rollers having the tip width of 4 mm. The surface of the scaled canister model had significant residual compressive stress after burnishing treatment. The roller burnished canister with good surface roughness could reduce the number of crack initiation sites and the residual compressive stress formed on the welded surface might prevent the crack initiation by reducing tensile residual stress in the weld zone, finally leads to CISCC resistance.

An Internal Compliance Program (ICP) is a system through which enterprise internally manage their own export control processes to ensure compliance with domestic export control laws. Around the world, ICPs are actively utilized as a means of export control for strategic items. However, they are not mostly applied to the Trigger List Items. However, advanced countries such as the United States and the Nuclear Suppliers Group (NSG) have been actively researching the potential application of ICPs to the Trigger List Items recently. This paper suggests additional considerations that should be taken into account when applying an ICP to the Trigger List Items. The key elements of classical ICP include Top-level management commitment to compliance; Risk analysis; Organizational structure/chain of responsibilities; Human and technical resources allocated to the management of exports; Workflow management and operational procedures; Record -keeping and documentation; Selection of staff; training and awareness-raising; Process-/Systemrelated controls (ICP audit)/Corrective Measures; Physical and technical security. An ICP for Trigger List Items must encompass all these core elements. Additionally, as the nuclear industry often involves collaborative projects participating with various companies, the effectiveness of the ICP could be enhanced through the operation of consultation groups among participating companies. Furthermore, enterprises must take into account the unique characteristics of Trigger List Items that differ from other strategic items, when making requirements of the ICP establishment. First, export requirements related to safety measures and physical protection should be reviewed to export the Trigger List Items. The procedure and obligations in aspects of internationally controlled items should also be reviewed. Moreover, active support from enterprises for GTGA procedures should also be included, since the Government to Government Assurance (GTGA) procedure is additionally required for the export of Trigger List Items, in contrast to other strategic items. Additionally, for materials categorized within Trigger List Items, such as deuterium and heavy water, should be controlled based on their end-use and cumulative quantity, which Government cannot effectively manage without enterprise supports. Therefore, enterprises must establish an internal material management system based on the end-use and cumulative quantity of these materials under ICP.

In compliance with the amended export control of strategic items and technology in Jan. 2014, KAERI should pay attention to the export control of ITT (Intangible Technology Transfer). To control an ITT (Intangible Technologies Transfer) effectively and efficiently, the Korean government encourages the R&D institute and universities obtaining the ICP (Internal Compliance Program) from the relevant authority, MOTIE. This means that the exporters can control the ITT by themselves, because the exporters know very well the counterparts of the trading and the exporting items and technologies. In fact, ICP is for export control of dual-use items and technology in Korea. However, KAERI has tried to obtain a license from the authority, MOTIE. In an effort to do so, KAERI completed enacting a new internal self–regulation for export controls in 2016, and proceeded to apply for an ‘AA’ license of ICP in 2017 and obtained the ICP license in 2018 and re-obtained the license in 2021 from the MOTIE. In light of KAERI’s case, to obtain the ‘AA’ license of ICP is one of the best methods to increase the ability of export controls. As of now, there is no R&D institutes sponsored by the Korean government to obtain the ‘AA’ license of ICP except KAERI. KAERI can provide the actual methods as a standard case to the R&D institutes in Korea for obtaining an ‘AA’ license of ICP. According to the internal regulation of KAERI for export control, KAERI implemented an inner self-audit for export control in Nov. 2022. This is the first real self-audit for export control at KAERI. The main purpose of the self-audit is to check the transfer management of ITT and the relationship of relevant office through the interview of the staffs in the ICP organization. KAERI self-audit planed specifically and implemented for the achievement of the basic principal of selfaudit. The specific contents of this self-audit is as follows - The interview of the relevant offices: physical protection office, manpower planning office, manpower management office, nuclear education and training center, technology transfer office and international cooperation office, nuclear control and management office - Building the self-audit checklist considering the characteristics of each office - The confirmation of the inner procedure and the status of management on the export controls Through the interview of the relevant office, KAERI checked the inner procedure and the status of management on the export controls and tried to provide the supplementary measures of each relevant offices. The followings are the main results of the inner self-audit implemented in Nov. 2022. - Generally, the staffs know the meaning and relevant regulation such as foreigner’s management and the intangible technology transfer - Each office reflects the necessities of export controls on the relevant regulation and procedures and make DB for the proper duty. However, there is no indication for export controls on the DB - In the case of foreigner’s temporary visit for simple work and site tour, there is a difficult situation not to be able to check all the visitors by checking the denial lists - If necessary, KAERI may build the TFT (Task force Team) for the efficiency of export controls - Others

Proton exchange membrane fuel cells (PEMFCs) are an auspicious energy conversion technology with the potential to address rising energy demands while reducing greenhouse gas emissions. The stack’s performance, durability, and economy scale are greatly influenced by the materials used for the PEMFC, viz., the membrane electrocatalyst assembly (MEA) and bipolar flow plates (BPPs). Despite extensive study, carbon-based materials have outstanding physicochemical, electrical, and structural attributes crucial to stack performance, making them an excellent choice for PEMFC manufacturers. Carbon materials substantially impact the cost, performance, and durability of PEMFCs since they are prevalently sought for and widely employed in the construction of BPPs and gas diffusion layers (GDLs)) and in electrocatalysts as a support material. Consequently, it is essential to assemble a review that centers on utilizing such material potential, focusing on its research development, applications, problems, and future possibilities. The prime focus of this assessment is to offer a clear understanding of the potential roles of carbon and its allotropes in PEMFC applications. Consequently, this article comprehensively evaluates the applicability, functionality, recent advancements, and ambiguous concerns associated with carbonbased materials in PEMFCs.

The demand for LNG Carrier and LNG fuel ships are increasing due to global carbon neutrality declaration and ship emissions regulation of IMO, domestic shipyards pay technology fees(about 5~10% of ship price per vessel) to GTT company in France for making LNG cargo hold. Localization of LNG cargo hold is needed to reduce technology fees and engage technological competitiveness, it is important to secure the critical technology like automation process development of insulation system process. Especially, the automation rate of membrane-type insulation system is very low due to interference caused by corrugation and difficulty in securing optimal variable welding condition. In this study, to solve this problem, automatic welding is performed using developed automatic welding equipment on STS304L steel which is used in flat and corner area of membrane-type LNG cargo hold's lap joint. After welding, Cross-sectional observations and Tensile strength tests were conducted to evaluate reliability of equipment and welding condition. As a result of the test, it was confirmed that the strength of the welded zone exceeded that of base material, and secured the optimal welding condition to apply automatic welding.

In this study, the effect of crown heating on the cultivation environment, budding, flowering and yields of strawberry was analyzed. In December, January, and February, when the outside temperature was low, the average strawberry crown temperature at daytime in the test zone was 1.3°C higher than that in the control zone, and the average strawberry crown temperature at nighttime in the test zone was 2.7°C higher than that in the control zone. The average bed temperature at daytime in test zone was 1.7°C higher than that in the control zone, and the average bed temperature at nighttime in test zone was 2.4°C higher than that in the control zone. As a result of performing correlation analysis and regression analysis on strawberry crown temperature and budding period, the correlation coefficient was -0.86, which tended to be shorter as the crown temperature was higher, and the determination coefficient was 0.74. The total yields of strawberry during test period were 392.6 g/plant for test greenhouse and 346.0 g/plant for control greenhouse respectively. As for the quality of strawberries, the ratio of 2L (very large) grades and L (large) grades was 62.4% in the test greenhouse and 58.5% in the control greenhouse, indicating that the proportion of high quality strawberries was higher in the test greenhouse.

As aerial application increasing, with social concerning in pesticide drift rises, so this study attempts to establish a test bench that can repeatedly and continuously evaluate this. To this end, this study first analyze ISO 22866 and ASABE S561.1 among the international standard test methods related to pesticide fugitive evaluation. A test bench was established at the Naju practice field of Chonnam National University in accordance with international standards, and field tests were carried out (ISO 22866, ASABE S561.1) to verify effectiveness. A test bench that established in this study and a pesticide drift recovery protocol by aerial application can improve the experimental environment where field experiments were complex and it was difficult to achieve the same conditions. In addition, it will be possible to construct a database of pesticide drift that takes into account various factors that affect pesticide drift substances, which is expected to improve the reliability of the data, as well as quantitative evaluation of pesticide drift in the air.

Transition metal chalcogenides are promising cathode materials for next-generation battery systems, particularly sodium-ion batteries. Ni3Co6S8-pitch-derived carbon composite microspheres with a yolk-shell structure (Ni3Co6S8@C-YS) were synthesized through a three-step process: spray pyrolysis, pitch coating, and post-heat treatment process. Ni3Co6S8@C-YS exhibited an impressive reversible capacity of 525.2 mA h g-1 at a current density of 0.5 A g-1 over 50 cycles when employed as an anode material for sodium-ion batteries. However, Ni3Co6S8 yolk shell nanopowder (Ni3Co6S8-YS) without pitch-derived carbon demonstrated a continuous decrease in capacity during charging and discharging. The superior sodium-ion storage properties of Ni3Co6S8@C-YS were attributed to the pitchderived carbon, which effectively adjusted the size and distribution of nanocrystals. The carbon-coated yolk-shell microspheres proposed here hold potential for various metal chalcogenide compounds and can be applied to various fields, including the energy storage field.

The effects of exogenous sodium nitroprusside (SNP, nitric oxide donor) on the growth, yield, photosynthetic characteristics, and antioxidant enzyme activity of kimchi cabbage (Brassica rapa L. subsp. pekinensis (Lour.) Hanelt) was studied under the low temperature conditions. Kimchi cabbages were treated with SNP of three concentrations (7.5, 15, 30 mg·L-1) for three times at four-day intervals and exposed to low temperature (16/7°C) stress for seven days. SNP treatment induced increases of net photosynthetic rate (Pn), stomatal conductance (Gs), intracellular CO2 concentration (Ci) and transpiration rate (Tr) under the stress condition with the highest level after the third treatment. The contents of malondialdehyde (MDA) and H2O2 were significantly lower in the treatment of SNP compared to the non-treated control. The activity of ascorbate peroxidase (APX), catalase (CAT), peroxidase (POD) and superoxide dismutase (SOD), increased in treated plants by up to 38, 187, 24 and 175%, respectively compared to the non-treated control. SNP-treated and untreated plants had similar growth characteristics. Compared to the control group, SNP-treatment increased fresh weight and leaf area by 5%. Overall, our findings suggest that the application of sodium nitroprusside to the leaves contributes to reducing physiological damage and enhancing the activities of antioxidant enzymes, thereby improving low temperature stress tolerance in kimchi cabbage.

PURPOSES : The purpose of this study was to evaluate the common performance of asphalt pavements, determine the timing of preventive maintenance, and determine the optimal timing of application of the preventive maintenance methods by analyzing PMS data. METHODS : Using PMS data on asphalt pavement performance on highways, we derived the major damage factors and evaluated them according to the public period and traffic level. Among the factors evaluated, we determined those that could be improved by preventive maintenance, calculated the amount of change annually, and derived the timing of the application of the preventive maintenance method through correlation analysis. RESULTS : Among highway PMS data factors, crack variation was found to affect preventive maintenance, which increased rapidly after five years of performance. Traffic analysis showed that changes increased rapidly in the fifth, sixth, and seventh years when AADT exceeded 20,000, exceeded 10,000, and was under 10,000, respectively. Analysis of the amount of crack variation according to the pavement type showed that crack variation increased rapidly in the overlay section compared to the general AP section. CONCLUSIONS : Crack variation is the performance factor that was expected to be effective in preventive maintenance, and the PMS data showed that the initial application time of the preventive maintenance method varied by one year, depending on the traffic volume.

본 연구는 바이오매스 전소 화력발전소에서 배출되는 저회에 유기 물질을 혼합하여 펠렛화한 후 배추 재배 미치는 영향을 평가하기 위해 수행되었 다. 배추 재배는 노지(UC, upland cultivation)와 시설재배지(GC, greenhouse cultivation)에서 진행되었으며, 처리조건은 대조구(Cont, control)와 저회(BA, bottom ash), 유기질 비료(OF, organic fertilizer), 저회 펠렛(BP, pelleted bottom ash) 처리구로 나누었고, 배추 이식 7일 전에 저회, 유기질 비료 및 저회 펠렛을 각각 1 ton 10a-1 수준으로 시용하였다. 배추의 생체중은 재배지역에 상관없이 OF와 BP 처리구에서 가장 높았으며, Cont 처리구에 비해 각각 15.7-41.6 및 9.7-51.5% 범위로 생체중이 증가하였다. BP 처리구에서 배추 수확 후 토양의 화학적 특성은 전반적으로 OM, TN, Avail. P2O5 등의 함량이 개선되는 효과를 보여주었으며, 이는 OF 처리구와 유사하였다. 노지와 시설재배지에서 배출된 N2O 배출량은 BP 처리가 OF 처리에 비해 29.7 및 11.0%로 각각 감소되는 결과를 보였다. 따라서, 저회 펠렛의 시용은 관행처리구와 비교해 볼 때, 배추 생육, 토양 특성 및 N2O 배출적인 측면에서 효과가 있는 것으로 판단된다.