Interim dry cask storage systems comprising AISI 304 or 316 stainless steel canisters have become critical for the storage of spent nuclear fuel from light water reactors in the Republic of Korea. However, the combination of microstructural sensitization, residual tensile stress, and corrosive environments can induce chloride-induced stress corrosion cracking (CISCC) for stainless steel canisters. Suppressing one or more of these three variables can effectively mitigate CISCC initiation or propagation. Surface-modification technologies, such as surface peening and burnishing, focus on relieving residual tensile stress by introducing compressive stress to near-surface regions of materials. Overlay coating methods such as cold spray can serve as a barrier between the environment and the canister, while also inducing compressive stress similar to surface peening. This approach can both mitigate CISCC initiation and facilitate CISCC repair. Surface-painting methods can also be used to isolate materials from external corrosive environments. However, environmental variables, such as relative humidity, composition of surface deposits, and pH can affect the CISCC behavior. Therefore, in addition to research on surface modification and coating technologies, site-specific environmental investigations of various nuclear power plants are required.

Molten salt reactors represent a promising advancement in nuclear technology due to their potential for enhanced safety, higher efficiency, and reduced nuclear waste. However, the development of structural materials that can survive under severe corrosion environments is crucial. In the present work, pure Ni was deposited on the surface of 316H stainless steel using a directed energy deposition (DED) process. This study aimed to fabricate pure Ni alloy layers on an STS316H alloy substrate. It was observed that low laser power during the deposition of pure Ni on the STS316H substrate could induce stacking defects such as surface irregularities and internal voids, which were confirmed through photographic and SEM analyses. Additionally, the diffusion of Fe and Cr elements from the STS316H substrate into the Ni layers was observed to decrease with increasing Ni deposition height. Analysis of the composition of Cr and Fe components within the Ni deposition structures allows for the prediction of properties such as the corrosion resistance of Ni.

원자력 발전소에서 배관 시스템은 냉각수 및 오염수를 운반하고 생성된 증기를 터빈으로 이동시켜 에너지를 생산하는 중요한 설비이다. 국내에 건설된 원자력 발전소의 가동연수가 증가함에 따라 배관 시스템의 물리적, 기계적 성질의 열화현상은 발생할 수 있으며 이를 경년열화로 정의한다. 배관 시스 템의 경년열화는 재료의 피로, 부식(국부감육), 마모 등과 같은 메커니즘을 통해 발생할 수 있으며 재 료의 강도 및 시스템의 성능 저하와 균열을 야기할 수 있다. 지속적이고 안정적인 에너지 생산을 위해 경제성과 정확도를 고려한 원전 배관 시스템의 손상 감시 기술은 필요하다. 따라서 본 연구는 원전 배 관 시스템의 손상 감시 기술을 개발하기 위한 기초적인 연구로써 배관 시스템의 취약요소로 판단되는 elbow의 국부적인 감육에 따른 거동의 변화를 분석하고자 한다.

The decommissioning of nuclear power plants will generate a lot of low and intermediate-level radioactive waste (LILW), and preliminary radioactive evaluation for these wastes should be carried out before decommissioning work. Mainly, Concrete, Carbon Steel, Stainless Steel-304 (SS304) and Inconel are used in many parts of nuclear power plants and considered as main resource of nuclear wastes. Depending on the material location, the number of neutrons irradiated to material varies, which can range from self-disposal waste to LILW. In this paper, activation analysis was performed to compare the radiation dose according to the presence or absence of impurity elements present in SS304. For the calculation, SS304 composition and impurity elements were used as described in the report of NUREG-3474. This report lists 41 impurity elements for SS304 and other materials. Calculation code is used ORIGEN-S module in SCALE 6.1 code. Neutron flux is used as arbitrary value that around 1E+11 level and irradiation time is set as 30 year with 10-year cooling time. In the ORIGEN-S calculation, 1g of SS304 is used for easy calculation of specific activity. The ORIGEN-S calculation results are as follows. All impurity elements contained case calculated 9.32E+07 Bq activity. In the absence of all impurity elements case and most cases shows that total Becquerel value after 10-year cooling time around 9.11E+07 Bq, and Co impurity case had larger result. The calculation was performed again by increasing the amount of impurity substances by 100 times to perform the sensitivity evaluation more reliably. Representatively, Li, N, Co, and Ba impurity elements cases were calculated to have a particularly high Becquerel. Especially Co impurity element case, a total Becquerel of 3.03E+08 was calculated. Accordingly, evaluation of impurities mixed in SS304 must be considered, and in particular, the inclusion rate for Co must be considered.

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.

The microstructure and mechanical properties of the heat-affected zone welded using the GTAW process on SDSS material used in offshore structures were analyzed. The chemical composition of the specimen material was analyzed using XRF. The microstructure of the heat-affected zone where the plate was welded was examined by SEM, and the ferrite volume fraction was assessed using the point count method of ASTM E562. A lot of ferrite was formed in the overheated weldment region, and In the weld cap where the cooling rate was fast, ferrite was not converted back to austenite and the microstructure was not uniform. From the ferrite phase fraction, it was shown that it can be applied to the pitting resistance equivalent numbers through changes in mechanical properties according to welding conditions.

This study aims to evaluate the structural safety of a structural thermal barrier, installed inside the structure of a building and performed the role of a load-bearing element and an insulation simultaneously, contributing to the realization of net-zero buildings. To ensure the reliability of the analysis model, the analysis results derived from LS-DYNA were compared with the experimental results. Based on the results shown through the flexural experiment, the reliability of the thermal cross-section insulation structure model for slabs was validated. In addition, the effect of the UHPC block on the load support performance and its contribution to vertical deflection was verified.

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.

It is reported that 48 pressurized heavy water reactors (PHWRs) are in operation, and 10 PHWRs including Wolsong-1 NPP have been permanently shut down in the world. In the case of PHWRs, which have been permanently ceased, they are managed through the delayed decommissioning method, but there are no cases of dismantling. Therefore, technology development is urgent for the effective decommissioning of PHWRs. Unlike PWRs, PHWRs are separated into coolant system and moderator system. Most of pipes and systems of coolant system are mainly composed of carbon steel, expect of the steam generator tubes which are composed of nickel alloy. On the other hand, the moderator system is composed of stainless steel. In the case of stainless steel, the inner layer of the oxide film is composed of chromium oxide, and the outer layer is composed of iron and nickel oxide in enriched. To remove two oxide layers, it is needs to different decontamination method, the coolant system can perform the system decontamination process through a reduction process, but in the case of the moderator system, the oxidation/reduction process is required because it has a material and oxide film similar to PWRs. In this study, this is evaluated the oxide film removal rate according to the type of stainless steel and temperature in order to remove the oxide film deposited in the moderator system. The experiments were carried out at temperatures of 60, 70, 80 and 90°C, with a concentration of 200 ppm of permanganic acid and nitric acid, and 2,000 ppm of oxalic acid, respectively. The results of the oxide film removal rate test for SUS304 showed 29% at 60°C, 38% at 70 and 80°C, and 41% at 90°C. For SUS403, the oxide film removal rate experiment results showed 62% at 60°C, 85% at 70°C, 94% at 80°C, over 99% at 90°C. The results showed that the removal efficiency of the oxide film increased as the temperature increased. Following the results of experimental, the optimum temperature of oxide removal in composed of the stainless steel material is to be 90°C for decontamination of PHWR.

Laser cutting has many advantages, including high-speed cutting potential, no reaction forces, narrow kerf widths, ease of remote control, and more. This makes it the next generation cutting technology for nuclear decommissioning. For this reason, various groups in countries with nuclear power plants have been working on applying laser cutting to nuclear decommissioning. Our group has also been developing in-air and underwater laser cutting technologies. Previous research has focused on efficiently cutting thicker steels. To accomplish this, a cutting head with a long focusing element with a focal length of 600 mm was utilized. A long focusing head is advantageous for cutting thick objects at high speeds because it can maintain a high power density over a long distance. However, with such a long focused beam, the residual laser power that remains after passing through the target object can cut or damage other unwanted objects located behind the target. Utilizing a short focused element can solve this problem, but if the focal length is too short, the cutting capability will be reduced. In this work, we developed and applied a cutting head that utilizes a focused element with a short focal length of 300 mm. Cutting tests with this head allowed us to cut 10-60 mm thick stainless steel plates at a laser power of 6 kW. We also obtained the maximum cutting speed and kerf width for each thickness while increasing the laser power by 1 kW from 1 to 6 kW. The results obtained in this work are expected to be utilized for safe cutting in future nuclear decommissioning applications.

For Korean nuclear fuel cycle project, it is necessary to design and evaluate the integrity of spent fuel storage. For the design and evaluation of spent fuel storage, it is necessary to evaluate the properties of various materials used in spent fuel storage. The materials previously considered in the design of nuclear power plants were limited to static properties and were listed in design and manufacturing code and standards. However, for the evaluation of the storage containers in scenarios such as transportation and events, dynamic material property evaluations are required. Research on the dynamic properties of materials is generally conducted in the fields of automotive and aerospace, and most of the studies are on metal materials under sheet conditions. Since the structural materials of the storage containers for used nuclear fuel are mostly composed of thick materials, consideration should be given to property evaluation methodology and quantitative comparison. In this study, the mechanical properties of stainless steel material with canister application were evaluated according to the strain rate, and the crack resistance evaluation was also performed. It was confirmed the changes in strength and crack resistance according to the increase in strain rate and observed differences in microstructural hardening behavior.

Long-term safe storage of spent nuclear fuel (SNF) determines sustainability of the current light water reactor (LWR) fleet. In the U.S., SNF is stored in stainless steel canister in dry cask storage system (DCSS) after spending several years in wet pool storage system while there is no DSCC in Republic of Korea. The SNF storage time in DSCC is expected to be multiple decades since no permanent geological repositories are identified in both countries. One limiting factor for extended storage of SNF in DSCC is chloride-induced stress corrosion cracking (CISCC) in the welded regions of the stainless steel canisters. The propensity for the occurrence of CISCC has warranted the development of the mitigation and repair technologies to ensure the safe and long-term storage for both present and new canister although no CISCC failure was reported yet. This study investigates cold spray deposition coatings of 304 L and 316 L stainless steels on prototypical stainless steel canisters such as sensitized flat and C-ring samples. The cold spray technology has been identified as the most promising approach by Extended Storage Collaboration Program (ESCP) driven by Electric Power Research Institute (EPRI). The talk includes microstructural characterization, adhesion strength measurement, residual stress evaluation, and corrosion behavior of the coated materials in boiling MgCl2 solution and electrochemical corrosion tests in NaCl solution. In addition, the capability of repair of cracks on the canister surface using the coating technology will be presented.

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.

A heat exchanger refers to a pressure vessel that indirectly exchanges heat between low-temperature/ high-temperature fluids with a solid wall interposed therebetween, and a shell-and-tube cylindrical heat exchanger is generally applied. The shell-and-tube cylindrical heat exchanger is widely used in ships and there is a problem in that the welding area is narrow and welding defects occur a lot due to high-level welding. In particular, in the case of a ship heat exchanger, if a problem occurs in the welding part during operation, the possibility of a safety accident is high, and repair is not easy. In this study, to solve this problem, the GTAW(Gas Tungsten Arc Welding) method was applied to secure the optimum conditions for pipe welding of STS304 material with a thickness of 5.5mm and to conduct a test. Afterwards, in accordance with the ASME rules, welding performance was verified through cross-sectional observation of welds, mechanical property tests, (tensile strength, bending strength, cryogenic impact strength) and non-destructive testing(PT, RT).

The effect of the laser beam diameter on the microstructure and hardness of 17-4 PH stainless steel manufactured via the directed energy deposition process is investigated. The pore size and area fraction are much lower using a laser beam diameter of 1.0 mm compared with those observed using a laser beam diameter of 1.8 mm. Additionally, using a relatively larger beam diameter results in pores in the form of incomplete melting. Martensite and retained austenite are observed under both conditions. A smaller width of the weld track and overlapping area are observed in the sample fabricated with a 1.0 mm beam diameter. This difference appears to be mainly caused by the energy density based on the variation in the beam diameter. The sample prepared with a beam diameter of 1.0 mm had a higher hardness near the substrate than that prepared with a 1.8 mm beam diameter, which may be influenced by the degree of melt mixing between the 17-4 PH metal powder and carbon steel substrate.