국내 건축물에서는 노후한 철근콘크리트 구조물의 안전성이 중요한 문제로 대두되고 있다. 구조물 부분이나 전체의 무너짐으로 인해 경제적 손실을 초래할 수 있으며, 이는 주로 구성 재료의 내구성 결 함으로 인해 발생한다. 여러 노후화 인자 중 동결융해와 부식은 주요한 열화 요인으로 작용한다. 동결 지역의 구조물은 동결융해가 위험 요소로 작용할 수 있으며, 해양 구조물은 해수에 존재하는 염소이온 에 의해 부식될 수 있다. 이러한 문제를 해결하기 위해서는 복합 열화 작용과 철근콘크리트 부재의 성 능 저하 관계를 이해하는 것이 필요하다. 본 연구는 동결융해와 부식의 복합적 피해가 RC 보의 거동 에 미치는 영향을 실험적으로 조사하였다. 7개의 RC 보를 제작하여 각각 다른 수준의 열화 조건을 부여한 후 휨 시험을 실시하였다.

Fundamental aspects of creating passivation layers for corrosion resistance in nuclear engineering applications, specifically the ability to form complete layers versus porous ones, are being explored in this study. Utilizing a laser ablation technique, 1,064 nm fire at 10 Hz with 60 pulses per shot and 0.5 mm between impact points, aluminum samples are treated in an attempt to create a fully formed passivation layer that will be tested in a LiCl-KCl eutectic salt. By placing these samples into an electrochemical environment mimicking a pyroprocessing system, corrosion rates, resistances and material characteristics are tested for one week and then compared between treated and untreated samples. In initial testing, linear sweep voltammetry indicates corrosion current density for the untreated sample at −0.038 mA·cm−2 and treated samples at −0.024 mA·cm−2 and −0.016 mA·cm−2, respectively. This correlates to a control sample corrosion rate of −0.205 mm·yr−1 and treated rates of −0.130 mm·yr−1 and −0.086 mm·yr−1 for samples 1 and 2. In addition, electrochemical impedance spectroscopy circuits show application of a longer-lasting porous passivation layer on the treated metal, compared to the naturally forming layer. However, the current technique fails to create a uniform protection layer across the sample.

In this study, we investigated the suppression of the corrosion of cast iron in a copper–cast iron double-layered canister under local corrosion of the copper layer. The cold spray coating technique was used to insert metals with lower galvanic activity than that of copper, such as silver, nickel, and titanium, between the copper and cast iron layers. Electrochemically accelerated corrosion tests were performed on the galvanic specimens in KURT groundwater at a voltage of 1.0 V for a week. The results revealed that copper corrosion was evident in all galvanic specimens of Cu–Ag, Cu–Ni, and Cu–Ti. By contrast, the copper was barely corroded in the Cu–Fe specimens. Therefore, it was concluded that if an inactive galvanic metal is applied to the areas where local corrosion is concerned, such as welding parts, the disposal canister can overcome local or non-uniform corrosion of the copper canister for long periods.

Heavy water primary system decontamination technology is essential to reduce worker exposure and improve safety during maintenance and decommissioning of nuclear facilities. Advanced decontamination technology development aims to secure controlled decontamination technologies that can reduce the cost of radiation exposure and dramatically reduce the amount of secondary waste generated when decontaminating large equipment and large-area facilities. We conducted a study to identify candidate corrosion inhibitors through the literature and analyze the degree of corrosion of carbon steel samples. Countries with advanced nuclear technology have developed chemical decontamination technology for the entire nuclear power generation system and applied it to the dismantling and maintenance of nuclear power plants. In the decontamination process, the corrosion oxide film must be removed. If the base metal is corroded by the decontaminant in this process, additional secondary waste is generated and treatment costs increase. Therefore, it is necessary to develop a corrosion inhibitor that inhibits the corrosion of the carbon steel base metal in the decontamination process to generate a secondary waste liquid that is favorable for waste reduction and treatment. In this presentation, a study was conducted to analyze the extent of corrosion on a carbon steel base material and identify candidate materials for corrosion inhibition testing. Samples were analyzed using optical microscopy and EPMA analysis to determine the thickness of the corroded oxide film. EPMA analysis also allowed us to map the elemental distribution of the carbon steel corrosion layer, which we plan to quantify in the future. The candidate materials for organic-based corrosion inhibitor were also selected based on their inhibition mechanism; having high electronegative elements for coordinate covalent bonding at metal surface and hydrophobic nonpolar group for preventing access of corrosive substances.The selection of candidate materials for corrosion inhibition testing was based on the mechanism of the corrosion inhibitor. Organic-based corrosion inhibitors are adsorbed by donor-acceptor interactions between metal surfaces and highly electronegative elements. Corrosion can also be inhibited by arranging hydrophobic nonpolar groups on metal surfaces in the solution direction to prevent access of corrosive substances.

After the major radioactivation structures (RPV, Core, SG, etc.) due to neutron irradiation from the nuclear fuel in the reactor are permanently shut down, numerous nuclides that emit alpha-rays, beta-rays, gamma-rays, etc. exist within the radioactive structures. In this study, nuclides were selected to evaluate the source term for worker exposure management (external exposure) at the time of decommissioning. The selection of nuclides was derived by sequentially considering the four steps. In the first stage, the classification of isotopes of major nuclides generated from the radiation of fission products, neutron-radiated products, coolant-induced corrosion products, and other impurities was considered as a step to select evaluation nuclides in major primary system structures. As a second step, in order to select the major radionuclides to be considered at the time of decommissioning, it is necessary to select the nuclides considering their half-life. Considering this, nuclides that were less than 5 years after permanent suspension were excluded. As a third step, since the purpose of reducing worker exposure during decommissioning is significant, nuclides that emit gamma rays when decaying were selected. As a final step, it is a material made by radiation from the fuel rod of the reactor and is often a fission product found in the event of a Severe accident at a nuclear power plant, and is excluded from the nuclide for evaluation at the time of decommissioning is excluded. The final selected Co-60 is a nuclide that emits high-energy gamma rays and was classified as a major nuclide that affects the reduction of radiation exposure to decommissioning workers. In the future, based on the nuclide selection results derived from this study, we plan to study the evaluation of worker radiation exposure from crud to decommissioning workers by deriving evaluation results of crud and radioactive source terms within the reactor core.

The primary heat transport system consists mainly of the in-core fuel channels connected to the steam generators by a system of feeder pipes and headers. The feeders and headers are made of carbon steel. Feeders run vertically upwards from the fuel channels across the face of the reactor and horizontally over the refueling machine to the headers. Structural materials of the primary systems of nuclear power plants (NPPs) are exposed to high temperature and pressure conditions, so that the materials employed in these plants have to take into accounts a useful design life of at least 30 years. The corrosion products, mainly iron oxides, are generated from the carbon steel corrosion which is the main constituent of the feeder pipes and headers of this circuit. Typical film thickness on CANDU-PHWR surface is 75μm or 30mg/cm2. Deposits on PHWR tends to be much thicker than PWR due to use of carbon steel and also for the source of corrosion products available on the carbon steel surface. Degradation of carbon steel for the feeder pipes transferring the primary system coolant by flow-assisted corrosion in high temperature has been reported in CANDU reactors including Point Lapreau, Gentully-2, Darlington and Bruce NPPs. The formation of Fe3O4 film on a carbon steel surface reduces the dissolution rate of steel substantially. The protectiveness of the Fe3O4 film over the carbon steel is affected by the environmental factors and the operational parameters of the feeder pipes, including the velocity, wall shear stress, solution pH, temperature, concentration of dissolved iron, quality of solution, etc. For effective chemical decontamination of these thick oxides containing radionuclides such as Co-60, it is necessary to understand the corrosion behaviors of feeder pipes and the characteristics of oxide formed on it. In this work, we investigated the growth of oxide films that develop on type SA-107 Gr. B carbon steel in high temperature water and steam environment by scanning electron microscopy (SEM) and glow discharge optical emission spectrometry (GD-OES) for the quantification and the solidstate speciation of metal oxide films. This study was especially focused to set the experimental tests conditions how to increase the oxide thickness up to 50 m by changing the oxidation conditions, such as solution chemistry and thermo-hydraulic conditions both temperature and pressure and so on.

It is crucial to be sure about the safety of nuclear facilities for human resources who are in danger of radioactive emission, also diminishing the volume of the wastes that are buried under the ground. Chemical decontamination of nuclear facilities can provide these demands at the same time by dissolving the oxide layer, which radionuclides such as 60Co and 58Co have been penetrated, of parts that are utilized in nuclear plants. Although there are many commercial methods to approaching its aim and they perform a high decontamination factor, they have some issues such as applying organic acids which have the ability to chelate with radionuclides that can be washed by underground water, have large quantities of radioactive waste and damage to the surface by severe intergranular attack. A new method has been introduced by KAERI’s scientist which is named the HyBRID Process, in this process the main solution is the acidic form of Hydrazine. In this process, like other acid-washing processes, there is a chance of corrosion on the metal surface which is not desired. The metal surface is able to be protected during dissolving process by adding some organic and inorganic corrosion inhibitors such as PP2 and PP3. There is a very new research topic about ionic liquids (ILs) as corrosion inhibitors which illustrates a vast potential for this application due to their tunable nature and the variety of options for cationic and anionic parts. The key factors for ILs corrosion inhibitors such as the hardness properties are summarized. In this study, we review to the fundamentals and development of corrosion inhibitors for chemical decontamination and give an prospect with emphasis on the challenges to be overcome.

Copper, mainly used as a material for outer canister, generates various corrosion products under aerobic and anaerobic conditions in the operational and/or post-closure phases of the deep geological repository. These products could affect performance of engineering barrier system (EBS) through interaction with surrounding bentonite that makes up the buffer and backfill materials. Accordingly, in this study, we suggested research items to be conducted to minimize degradation of EBS due to copper corrosion products, based on the phenomenological review results for copper corrosion mechanisms and interaction between resultant product and bentonite in the deep geological disposal environment. During the post-closure phase, condition in the disposal facility changes form aerobic to anaerobic over time, and thereby, causes and products of copper corrosion vary. Under aerobic condition, copper corrosion is mainly induced by oxygen (O2) in the repository, chloride (Cl-) and carbonate (CO3 2-) ions from groundwater flowing into the facility, resulting in corrosion products such as cuprite (Cu2O), tenorite (CuO), atacamite (CuCl2·3Cu(OH)2) and malachite (Cu2CO3(OH)2). And, copper corrosion under anaerobic condition is primarily due to hydrogen sulfide (H2S) and sulfate (SO4 2-) in groundwater flowing into the facility, leading to formation of chalcocite (Cu2S) and covellite (CuS) as corrosion products. Depending on environment of the disposal facility, copper corrosion products are dissolved and ionized to Cu2+ in groundwater, and subsequently adsorbed on the nearby smectite. Then, it causes a cation exchange reaction with exchangeable cations in the interlayer of smectite. As a result of reviewing the previous experiments, it was confirmed that Cu2+-exchanged bentonite has a slightly reduced basal spacing and swelling capacity. From the results as above, there is a possibility that performance of EBS may be degraded due to copper corrosion products. To minimize its effect of degradation in the domestic facility, items to be further studied are as follows: (a) Method for reducing copper corrosion such as selection of appropriate material and structure for the canister, and (b) How to control dissolution of copper canister product into groundwater through predicting type and ionization process. The results of this study could be directly used to developing design concept of EBS for the domestic disposal facility and to establishing roadmap of future R&D programs.

Currently, Korea is considering a disposal system based on Sweden’s KBS-3 model to dispose of high-level waste. The disposal system uses a multi-barrier concept to protect high-level waste with canister, buffer, backfill, and natural rock. In Korea, copper and iron are being considered for external and internal canisters, and bentonite is being considered as a buffer material. This is a similar choice to many overseas disposal systems. However, unlike the rolling, extrusion, and forging manufacturing methods being considered overseas for manufacturing external canister, domestic research is currently underway on manufacturing external copper canister using cold spray coating. The canister manufacturing method may vary depending on unit cost and manufacturing convenience. However, the properties of metal vary slightly depending on the manufacturing method of the metal. In this case, the characteristics of the canister may vary slightly depending on the canister manufacturing method, and eventually the corrosion resistance may also vary slightly. In order to understand how the copper canister manufacturing method affects corrosion resistance, corrosion rates were calculated and compared through electrochemical corrosion experiments at domestic groundwater ion concentration.

In this study, a third metal layer with a higher corrosion potential than copper was introduced between the copper and cast iron layer to strengthen the corrosion resistance of the copper layer which is considered as a corrosion resistant barrier in the disposal container for spent nuclear fuel. Three types of corrosion-resistant metals, silver, nickel, and titanium, were selected as the intermediate insertion layer, and the galvanic specimens of two bonded metals were exposed to KURT (KAERI Underground Research Tunnel) groundwater and a high voltage of 1.0 V was applied to corrode the specimens at electrochemically accelerated condition. Corrosion of copper part was confirmed in Cu-Ti, Cu-Ni, and Cu-Ag galvanic specimens, but copper part was not corroded in Cu-Fe galvanic specimen. If the corrosion-resistant intermediate layer proposed in this study works properly, the local corrosion problem of copper disposal canister is expected to be some degree solved, which can apply to a welding part or a stress concentrated part.

For the sake of future generations, the management of radioactive waste is essential. The disposal of spent nuclear fuel (SNF) is considered an urgent challenge to ensure human safety by storing it until its radioactivity drops to a negligible level. Evaluating the safety of disposal facilities is crucial to guarantee their durability for more than 100,000 years, a period sufficient for SNF radioactivity to become ignored. Past studies have proposed various parameters for forecasting the safety of SNF disposal. Among these, radiochemistry and electrochemistry play pivotal roles in predicting the corrosion-related chemical reactions occurring within the SNF and the structural materials of disposal facilities. Our study considers an extreme scenario where the SNF canister becomes compromised, allowing underground water to infiltrate and contact the SNF. We aim to improve the corrosion mechanism and mass-balance equation compared with what Shoesmith et al. proved under the same circumstances. To enhance the comprehensibility of the chemical reactions occurring within the breached SNF canister, we have organized these reactions into eight categories: mass diffusion, alpha radiolysis, adsorption, hydrate formation, solidification, decomposition, ionization, and oxidation. After categorization, we define how each species interacts with others and calculate the rate of change in species’ concentrations resulting from these reactions. By summing up the concentration change rates of each species due to these reactions, we redefine the mass-balance equations for each species. These newly categorized equations, which have not been explained in detail previously, offer a detailed description of corrosion reactions. This comprehensive understanding allows us to evaluate the safety implications of a compromised SNF canister and the associated disposal facilities by numerically solving the mass-balance equations.

One of the options for spent fuel dry storage systems is to store them in canisters using metal or concrete casks close to shore. The interaction between the austenitic stainless steel and the chloride atmosphere generated from the sea creates detrimental conditions leading to chloride induced stress corrosion cracking (CISCC) in the canister. The corrosion integrity of the canister in the concrete cask is very important because the canister is sealed and used for a long period of time. A canister made of austenitic stainless steel has several welding lines on the wall and lid, which are generated during the welding process and have high residual tensile stress. The interaction between the austenitic stainless steel and the chloride atmosphere generated from the sea creates detrimental conditions leading to chloride induced stress corrosion cracking (CISCC) in the canister. The corrosion integrity of the canister in the concrete cask is very important because the canister is sealed and used for a long period of time. In order to evaluate such soundness, an accelerated test capable of simulating the CISCC crack propagation phenomenon of the canister weld is required. In this study, a test device for performing the CISCC simulation test was constructed using the DCPD device. The direct current potential drop (DCPD) technique is a widely accepted method of monitoring crack initiation and growth in controlled laboratory tests. Total 10 types of test specimens with varying welds, base metal, salinity and stress were selected and a sealed chamber with DCPD test apparatus were designed and constructed to evaluate them. The chamber for CISCC simulation was manufactured as a sealed with a solution containing 10% MgCl2. A 1/2 CT specimen with precracked pre-cracks was loaded into the prepared container, and gauze was attached from the bottom for smooth delivery to the specimen to facilitate penetration of chloride. After the test, the measured DCPD data were correlated with Electron Back scattered Diffraction (EBSD) data.

Pyroprocessing is a crucial method for recovering nuclear fuel materials, particularly uranium and transuranic elements (TRU), through electrochemical reactions in a LiCl or LiCl-KCl molten salt system, which is highly stable medium at elevated temperatures. In the electrochemical reduction stage, actinide metal oxides are effectively transformed into their metallic forms and retained at the cathode within a molten LiCl-Li2O environment at 650°C. Simultaneously, oxygen ions (O2-) are generated at the cathode and then transported through the molten salt to be discharged at the anode, where they combine to form oxygen gas (O2) on the anode’s surface. One notable challenge in this electrochemical process is the generation of various byproducts during the anode oxide reduction step, including oxygen, chlorine, carbon dioxide, and carbon monoxide. Consequently, significant amounts of corrosion products tend to accumulate on the upper region of the anode’s immersion area over time. This report introduces a novel solution to mitigate corrosion-related challenges within the specified temperature range. We propose a selective oxidation treatment for the NiCrAl-based 214 Haynes alloy, involving exposure to 1,100°C in a reducing atmosphere. The objective is to stimulate the growth of protective α-Al2O3 scales on the alloy’s surface. The resulting oxide scales have undergone thorough characterization using SEM, EDS, and XRD techniques. The pre-grown alumina scale has demonstrated commendable adherence and thermal stability, even when subjected to a chlorine-oxygen mixed atmosphere at the specified temperature.

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.

Molten salt reactor (MSR) uses fluoride or chloride based molten salt as a coolant of the system, and fuel materials are dissolved in the molten salt, therefore it can be act as both coolant and nuclear fuel. A few issues have arisen from early-stage research and development program of MSR from Oak Ridge National Laboratory, including corrosion of structural materials and fission product management. For investigating the effect of additives on corrosion of structural materials, Mg(OH)2 and MgCl2*6H2O are added into the NaCl-MgCl2 eutectic salt. Prepared chloride salt is injected into the autoclave in the glove box, as well as corrosion coupons for candidate structural materials for molten chloride salt reactor, SS316, Alloy 600, and C-276 are also prepared. The temperature is set as 700°C. After 500 h corrosion experiment, the samples are taken out from the autoclave, and they are analyzed with scanning electron microscopy (SEM) and energy-dispersive X-ray spectroscopy (EDS). SS316 samples show weight loss with all salt conditions, while Alloy 600 and C-276 show weight gain after the corrosion experiment.

Safe management of spent nuclear fuel (SNF) is a key issue to determine sustainability of current light water reactor (LWR) fleet. However, none of the countries are actually conducting permanent disposal of SNFs yet. Instead, most countries are pursuing interim storage of spent nuclear fuels in dry cask storage system (DCSS). These dry casks are usually made of stainlesssteels for resistibility against cracking and corrosion, which can be occurred over a long-term storage period. Nevertheless, some corrosion called Chloride-Induced Stress Corrosion Cracking (CISCC) can arise in certain conditions, exacerbating the lifetime of dry casks. CISCC can occur if the three conditions are satisfied simultaneously: (i) residual tensile stress, (ii) material sensitization, and (iii) chloride-rich environment. A residual tensile stress is developed by the two processes. One is the bending process of stainless-steel plates into a cylindrical shape, and the other is the welding process, which can incur solidification-induced stress. These stresses provide a driving force of pit-to-crack transition. Around the fusion weld areas, chromium is precipitated at the grain boundary as a carbide form while it depletes chromium around it, leading to material susceptible to pitting corrosion. It is called sensitization. Finally, coastal regions, where nuclear power plants usually operate, tend to have a higher relative humidity and more chloride concentration compared to inland areas. This high humidity and chloride ion concentration initiate pitting corrosion on the surface of stainless-steels. To prevent initiation of CISCC, at least one of the three conditions should be removed. For this, several surface engineering techniques are under investigation. One of the most promising approaches is surface peening method, which is the process that impacts the surface of materials with media (e.g., small pins, balls, laser pulse). By this impact, plastic deformation on the surface occurs with compressive stress that counteracts with pre-existing residual tensile stress, so this approach can prevent pit-to-crack transition of stainless-steels. Also, cold spray deposition can prevent CISCC. Cold spray deposition is a method of spraying fine metal powder to a substrate by accelerating them to supersonic velocity with propellant gas. As a result, a thin coating composed of the feedstock powders can protect the substrate from outer corrosive environments. In addition, the impact of the feedstock powder on the substrate during the process provides compressive stress, similar to the peening method.

In this study, the physical properties and fracture characteristics according to the tensile load are evaluated on the materials of the polymeric filler and carbon fiber-based composite sleeve technique. The polymeric filler and the composite sleeve technique are applied to areas where the pipe body thickness is reduced due to corrosion in large-diameter water pipes. First, the tensile strength of the polymeric filler was 161.48~240.43 kgf/cm2, and the tensile strength of the polyurea polymeric filler was relatively higher than that of the epoxy. However, the tensile strength of the polymeric filler is relatively very low compared to ductile cast iron pipes(4,300 kgf/cm2<) or steel pipes(4,100 kgf/cm2). Second, the tensile strength of glass fiber, which is mainly used in composite sleeves, is 3,887.0 kgf/cm2, and that of carbon fiber is up to 5,922.5 kgf/cm2. The tensile strengths of glass and carbon fiber are higher than ductile cast iron pipe or steel pipe. Third, when reinforcing the hemispherical simulated corrosion shape of the ductile cast iron pipe and the steel pipe with a polymeric filler, there was an effect of increasing the ultimate tensile load by 1.04 to 1.06 times, but the ultimate load was 37.7 to 53.7% compared to the ductile cast iron or steel specimen without corrosion damage. It was found that the effect on the reinforcement of the corrosion damaged part was insignificant. Fourth, the composite sleeve using carbon fiber showed an ultimate load of 1.10(0.61T, 1,821.0 kgf) and 1.02(0.60T, 2,290.7 kgf) times higher than the ductile cast iron pipe(1,657.83 kgf) and steel pipe(2,236.8 kgf), respectively. When using a composite sleeve such as fiber, the corrosion damage part of large-diameter water pipes can be reinforced with same level as the original pipe, and the supply stability can be secured through accident prevention.

일반적인 불투과형 사방댐은 제체 대부분이 콘크리트로 구성되어 분리 및 해체가 어려워 조속한 보강이 난해하다는 단점이 있다. 이러한 단점을 보완하기 위한 투과형 강재 스크린 사방댐이 제작되어 사용 중에 있지만, 강재가 부식되는 등 여러 문제를 보이는 실정이다. 이에 따라, 선행 연구에서는 강재 스크린 구조물을 내부식성이 뛰어난 GFRP로 대체하여 성능 및 안전 성을 검증하였지만, 동수압만 고려되는 등 다양한 조건이 고려되지 않은 것으로 확인되었다. 그렇기에 본 연구에서는 선행 연구 에서 수행된 GFRP 스크린 구조물뿐만 아니라, CFRP도 고려하였으며, 보다 다양한 표면 유속 별 하중 조건 및 유목, 토석류 등 과 같은 다양한 조건에서의 성능 비교 및 안전성 평가를 수행하였다. 해석 결과, GFRP와 CFRP는 강재 대비 제체에 작용하는 응력이 29.79∼91.73%가량 감소된 성능을 보이며, 이 외에도 충분한 안전성과 경제성을 겸비함을 확인하였다. 결론적으로 GFRP 및 CFPR 스크린 구조물은 강재 투과형 사방댐을 대체하여 사용하기에 충분하다고 판단되지만, 이는 수치 해석을 통한 결과이 므로 향후 실제 실험이 진행될 필요가 있을 것으로 판단된다.

This study aimed to address the limitations of traditional plasma nitriding methods by implementing a short-term plasma oxy-nitriding treatment on the surface of AISI 420 martensitic stainless steel. This treatment involved the sequential formation of nitride and oxide layers, to enhance surface hardness and corrosion resistance, respectively. The process resulted in the formation of a 20 μm-thick nitride layer and a 3 μm-thick oxide layer on the steel surface. Initially, the hardness increased by 2.2 times after nitriding, followed by a subsequent decrease of approximately 31 % after oxidation. While the nitriding process reduced corrosion resistance, the subsequent oxidation process led to the formation of a passive oxide film, effectively resolving this issue. The pitting corrosion of the oxide passive film started at 82.6 mVssc, providing better corrosion resistance characteristics than the nitride layer. Consequently, the trade-off between surface hardness and corrosion resistance in plasma oxy-nitrided AISI 420 martensitic stainless steel is anticipated to be recognized as an innovative and comprehensive surface treatment process for biomedical components.

Corrosion-related challenges remain a significant research topic in developing next-generation Molten Salt Reactors (MSRs). To gain a deeper understanding of preventing corrosion in MSRs, previous studies have attempted to improve the corrosion resistance of structural alloys by coating surfaces such as alumina coating. To conduct a corrosion test of coating alloys fully immersed in molten salt, it’s important to ensure that the coating application process is carefully carried out. Ideally, coating all sides of the alloy is necessary to avoid gaps like corners of the alloy, while only applying a one-sided coating alloy can lead to galvanic corrosion with the base metals. Using the droplet shape of eutectic salt applied to only one side of the coating alloy would avoid these problems in conventional corrosion immersion tests, as corrosion would occur solely on the coating surface. Although the droplet method for corrosion tests cannot fully replicate corrosion in the MSRs environment, it offers a valuable tool for comparing and evaluating the corrosion resistance of different coating surfaces of alloys. However, the surface area is important due to the effect of diffusion in the corrosion of alloy in molten salt environments, but it is difficult to unify in the case of droplet tests. Therefore, understanding the droplet-alloy properties and corrosion mechanism is needed to accurately predict and analyze these test systems’ behavior highlighting unity for corrosion tests of different coating surfaces of alloys. To analyze the molten salt droplet behavior on various samples, pelletized eutectic NaCl-MgCl2 was prepared as salt and W-, Mo-coating, and base SS316 as samples. At room temperature, the same mass of pelletized eutectic NaCl-MgCl2 was placed on different samples under an argon atmosphere and heated to a eutectic point of 500°C in a furnace. After every hour, the molten droplets were hardened by rapid cooling at room temperature outside the furnace. The mass loss of salts and the contact area of the samples were measured by mass balance and SEM. The shape, surface area to volume ratio, and evaporation of the droplets of NaCl-MgCl2 per each coating sample and hour were analyzed to identify the optimal mass to equalize the contact coating surface of alloys with salts. Furthermore, We also analyzed whether their results reached saturation of corrosion products through ICP-MS. This will be significant research for the uniformity of the liquid-drop shape corrosion test of the coating sample in molten eutectic salts.