탄소섬유 강화 플라스틱 (Carbon fiber reinforced plastics, CFRP)은 고함량의 탄소섬유 (Carbon fiber, CF)와 고분자로 이루어진 복합재료로서, 뛰어난 기계적 성능으로 항공우주, 자동차, 토목 등 다 양한 산업 분야에서 사용되고 있다. 하지만 사용량 증가에 따른 폐기물의 환경문제와 추출한 재활용 탄소섬유 (Recycled carbon fiber, rCF)의 적용 가능 분야의 한계로 인해 재활용이 제한적인 실정이 다. 본 연구에서는 rCF와 CF 혼입 시멘트계 전자파 복합재를 제작하여 그 성능을 비교 분석하기 위 한 실험을 수행하였다. 구성재료는 시멘트, 잔골재, 고성능 감수제를 사용하였으며, 비교 분석을 위해 CF와 rCF를 각각 6 mm, 12 mm 길이를 0.1, 0.3, 0.5, 1.0 wt.% 함량으로 사용하였다. 전자파 복합 재의 흡수 성능 향상을 위해 각각 다른 함량의 다층 구조를 형성하였으며, 전자파 투과를 낮은 함량에 서 높은 함량 방향이 되도록 측정을 진행하였다. 전자파 차폐성능은 재령 28일 이후 네트워크 분석기 를 사용하여 자유 공간에서 측정하였으며, C-band (4~8 GHz)와 X-band (8~12 GHz) 주파수 영역 에서의 반사율과 투과율을 각각 측정하였다.

Recent advancements in electronic devices and wireless communication technologies, particularly the rise of 5G, have raised concerns about the escalating electromagnetic pollution and its potential adverse impacts on human health and electronics. As a result, the demand for effective electromagnetic interference (EMI) shielding materials has grown significantly. Traditional materials face limitations in providing optimal solutions owing to inadequacy and low performance due to small thickness. MXene-based composite materials have emerged as promising candidates in this context owing to their exceptional electrical properties, high conductivity, and superior EMI shielding efficiency across a broad frequency range. This review examines the recent developments and advantages of MXene-based composite materials in EMI shielding applications, emphasizing their potential to address the challenges posed by electromagnetic pollution and to foster advancements in modern electronics systems and vital technologies.

This study focuses on the development of coatings designed for storage containers used in the management of radioactive waste. The primary objective is to enhance the shielding performance of these containers against either gamma or neutron radiation. Shielding against these types of radiation is essential to ensure the safety of personnel and the environment. In this study, tungsten and boron cabide coating specimens were manufactured using the HVOF (High-Velocity Oxy Fuel) technuqe. These coatings act as an additional layer of protection for the storage containers, effectively absorbing and attenuating gamma and neutron radiation. The fabricated tungsten and boron carbide coating specimens were evaluated using two different testing methods. The first experiment evaluates the effectiveness of a radiation shielding coating on cold-rolled steel surfaces, achieved by applying a mixture of WC (Tungsten Carbide) powders. WC-based coating specimens, featuring different ratios, were prepared and preliminarily assessed for their radiation shielding capabilities. In the gamma-ray shielding test, Cs-137 was utilized as the radiation source. The coating thickness remained constant at 250 μm. Based on the test results, the attenuation ratio and shielding rate for each coated specimen were calculated. It was observed that the gammaray shielding rate exhibited relatively higher shielding performance as the WC content increased. This observation aligns with our findings from the gamma-ray shielding test and underscores the potential benefits of increasing the tungsten content in the coating. In the second experiment, a neutron shielding material was created by applying a 100 μm-thick layer of B4C (Boron Carbide) onto 316SS. The thermal neutron (AmBe) shielding test results demonstrated an approximate shielding rate of 27%. The thermal neutron shielding rate was confirmed to exceed 99.9% in the 1.5 cm thick SiC+B4C bulk plate. This indicates a significant reduction in required volume. This study establishes that these coatings enhance the gamma-ray and neutron shielding effectiveness of storage containers designed for managing radioactive waste. In the future, we plan to conduct a comparative evaluation of the radiation shielding properties to optimize the coating conditions and ensure optimal shielding effectiveness.

In the nuclear environment, sensors ensure safety, monitoring, and operational efficiency under various operating conditions. These sensors come in various forms, each tailored to specific purposes, including nuclear safety and security, waste treatment and storage, gas leak detection, temperature and humidity monitoring, and corrosion detection. Ensuring the longevity of sensors without the need for frequent replacements is a vital goal for researchers in this field. This paper explores materials that can act as shields to protect sensors from harsh environmental conditions (high radiation and temperatures) to enhance their lifetime. The types of material that had been explored were divided into categories: metal and non-metal. Fourteen types of metal and seven different plastic materials were studied and focused on their characteristics and current applications. Considering properties like melting point, intensity, and conductivity, plastic materials are chosen to be examined as sensor shielding material. A preliminary experiment was conducted to verify signal characteristics changes by shielding material. Metal material and plastic material each were placed in the middle of the granite and the target sensor. The result showed that when metal is between the granite and the sensor, the density and impedance are higher in granite than in the metal. This leads to signal attenuation and a shift in resonance frequency, while plastic does not. Therefore, PPS (Polyphenylene sulfide) and PAI (Polyamide-imide) have lower density and impedance than granite while also possessing heat, moisture, and radiation resistance for effective shielding.

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.

After the Fukushima disaster, overseas nuclear power plants have established conditions for issuing a red alert in the event of fuel damage within the spent fuel pool and they have already implemented conditions for issuing a blue alert when fuel is exposed above the water surface. In South Korean nuclear power plants, a real-time monitoring system is in place to oversee the exposure of spent fuel to the surface within the spent fuel pool. To achieve this, a water level indicator gauge is installed within the spent fuel pool, allowing for continuous real-time monitoring. This paper conducted a comparative assessment of radiation levels from water level monitoring system in two units’ spent fuel pools based on the low water levels (1 feet from the storage rack), utilizing the radiation analysis code (MCNP).

To ensure radiological safety margin in the transport and storage of spent nuclear fuel, it is crucial to perform source term and shielding analyses in advance from the perspective of conservation. When performing source term analysis on UO2 fuel, which is mostly used in commercial nuclear power plants, uranium and oxygen are basically considered to be the initial materials of the new fuel. However, the presence of impurities in the fuel and structural materials of the fuel assembly may influence the source term and shielding analyses. The impurities could be radioactive materials or the stable materials that are activated by irradiation during reactor power operation. As measuring the impurity concentration levels in the fuel and structural materials can be challenging, publicly available information on impurity concentration levels is used as a reference in this evaluation. To assess the effect of impurities, the results of the source term and shielding analyses were compared depending on whether the assumed impurity concentration is considered. For the shielding analysis, generic cask design data developed by KEPCO-E&C was utilized.

RUCAS (Recycling-Underlying Computational Dose Assessment System), a dose assessment program based on the RESRAD-RECYCLE framework, is designed to evaluate dose for recycling scenarios of radioactive waste in metals and concrete. To confirm the validity of the recycling scenarios provided by RUCAS, comparative evaluations will be conducted with RESRAD-RECYCLE for metal radioactive waste recycling scenarios and with MicroShield® for concrete radioactive waste recycling scenarios. In the evaluation of metal recycling scenarios without shielding, RUCAS showed similar results when compared to both MicroShield® and RESRAD-RECYCLE. This validates the function of dose assessments using RUCAS for metal recycling scenarios. However, when shielding was present, RUCAS produced results that were comparable to MicroShield®, but differed from those of RESRAD-RECYCLE. The underestimation of dose values up to 1.66E+08 times difference by RESRAD-RECYCLE could potentially decrease reliability and safety in evaluated doses, further emphasizing the importance of RUCAS. Because validation is also necessary for the expanded calculation capabilities resulting from methodological changes of RUCAS (i.e., various radiation source geometries), based on prior validations, it was determined that additional validations are required for different radiation source materials and shielding conditions. In case where the radiation source and shielding materials were identical, RUCAS and MicroShield® produced similar results according to both the Kalos et al. (1974) and Lin and Jiang (1996) methodologies. This demonstrates that the that differences in methodology are inconsequential when considering the same source and shielding materials. However, when the atomic number of the radiation source materials was larger than that of shielding material (HZ-LZ condition), RUCAS obtained results similar to MicroShield® only for the Kalos et al. (1974) methodology. While Lin and Jiang (1996) methodology yield higher results than MicroShield®. Lastly, in case where the atomic number of the radiation source material was smaller than that of the shielding material (LZ-HZ condition,) both methodologies yielded results comparable to MicroShield®. In conclusion, the validity of RUCAS’s shielding calculations has been verified, confirming improvements in dose assessment compared to RESRAD-RECYCLE. Additionally, we observed that shielding effectiveness calculations differ depending on the methodology of build-up effect. If the validity of these methodologies is confirmed, it is expected that selecting the most advantageous methodology for each condition will enable more rational dose assessments. Consequently, in future research, we plan to evaluate the validity of Lin and Jiang (1996) methodology using particle transport codes based on the Monte Carlo method, such as MCNP and Geant 4, rather than MicroShield®.

A radiation shielding resin with thermal stability and high radiation shielding effect has been developed for the neutron shielding resin filled in the shielding shell of dry storage/transport cask for spent nuclear fuel. Among the most commercially available neutron shielding resins, epoxy and aluminum hydroxide boron carbide are used. But in case of the resin, hydrogen content enhances the neutron shielding effect through optimization of aluminum hydroxide, zinc borate, boron carbide, and flame retardant. We developed a radiation shielding material that can increase the boron content and have thermal stability. Flame retardancy was evaluated for thermal stability, and neutron shielding evaluation was conducted in a research reactor to prove the shielding effect. As a result of the UL94 vertical burning test, a grade of V-0 was received. Therefore, it was confirmed that it had flame retardancy. According to an experiment to measure the shielding rate of the resin against neutron rays using NRF (Neutron Radiography Facility), a shielding rate of 91.54% was confirmed for the existing resin composition and a shielding rate of 96.30% for the developed resin composition. A 40 M SANS (40 M Small Angle Neutron Scattering Instrument) neutron shielding rate test was performed. Assuming aging conditions (6 hours, 180 degrees), the shielding rate was analyzed after heating. As a result of the experiment, the developed products with 99.8740% and 99.9644% showed the same or higher performance.

For deep geological repository of the spent nuclear fuel, the fuel assemblies loaded in the storage cask are transferred to the disposal cask and the operation is performed in the fuel handling hot cell at the fuel re-packaging facility. As the fuel handling hot cell shielding is accomplished by the concrete wall and the viewing glass window, the required shielding thickness was evaluated for both materials. The ordinary concrete is applied to hot cell wall and two kinds of glasses, i.e., single layer of lead glass and double layer of lead glass and borosilicate glass, are considered for the viewing glass window. A bare spent PWR fuel assembly exposed to the environment in the hot cell was considered as the neutron and gamma radiation sources. The neutron and gamma transport calculations were performed using the MAVRIC program of the SCALE code system for the dose rate evaluation. The dose limit of 10 μSv/h is applied as the target dose to establish the required shielding thickness. The concrete wall of 94 cm thickness reduces the total dose rate to 6.9 μSv/h, which is the sum of neutron dose and gamma dose. Penetrating the concrete wall, both of the neutron dose and the gamma dose decrease constantly with shield thickness and the gamma dose is always dominant through whole penetrating distance. Single layer lead glass of 74 cm thickness reduces total dose rate to 6.2 μSv/h. Applying double layer shield glass combined of lead glass and borosilicate glass, the total dose rate reduces to 3.6 μSv/h at same shield thickness of 74 cm. Through the shield glass, gamma dose decreases rapidly and neutron dose decreases slowly compared with those for concrete wall. In result, neuron dose becomes dominant on the window glass shielding. The more efficient dose reduction of double layer glass is achieved by the borosilicate glass’s superior neutron shielding power. Thus, the use of double layer glass of lead glass and borosilicate glass is recommended for the viewing glass of the fuel handling hot cell. Finally, it is concluded that about 1 m thick concrete wall and 75 cm thick viewing glass window are sufficient for the radiation shielding of the hot cell at the spent fuel repackaging facility.

Radiation dose rates for spent fuel storage casks and storage facilities of them are typically calculated using Monte Carlo calculation codes. In particular, Monte Carlo computer code has the advantage of being able to analyze radiation transport very similar to the actual situation and accurately simulate complex structures. However, to evaluate the radiation dose rate for models such as ISFSI (Independent Spent Fuel Storage Installation) with a lot of spent fuel storage casks using Monte Carlo computational techniques has a disadvantage that it takes considerable computational time. This is because the radiation dose rate from the cask located at the outermost part of the storage facility to hundreds of meters must be calculated. In addition, if a building is considered in addition to many storage casks, more analysis time is required. Therefore, it is necessary to improve the efficiency of the computational techniques in order to evaluate the radiation dose rate for the ISFSI using Monte Carlo computational codes. The radiation dose rate evaluation of storage facilities using evaluation techniques for improving calculation efficiency is performed in the following steps. (1) simplified change in detailed analysis model for single storage cask, (2) create source term for the outermost side and top surface of the storage cask, (3) full modeling for storage facilities using casks with surface sources, (4) evaluation of radiation dose rate by distance corresponding to the dose rate limit. Using this calculation method, the dose rate according to the distance was evaluated by assuming that the concrete storage cask (KORAD21C) and the horizontal storage module (NUHOMS-HSM) were stored in the storage facility. As a result of calculation, the distance to boundary of the radiation control area and restricted area of the storage facility is respectively 75 m / 530 m (KORAD21C case), and 20 m / 350 m (NUHOMS-HSM case).

Environmental regulations of the International Maritime Organization (IMO) are getting stricter, and the demand for replacing the fuel of ships with eco-friendly fuels instead of heavy oil in the shipbuilding and marine industries is increasing. Among eco-friendly fuels, LNG (liquefied natural gas) is currently the most popular fuel. This is because it is an alternative that can avoid the IMO's environmental regulations by replacing fuel. In PART 1, as a basic study of laser welding of high manganese steel materials, a fiber laser bead-on-plate experiment was conducted using nitrogen protective gas, and the effect of each factor on the penetration shape was analyzed through cross-sectional observation. In PART II, argon and helium shielding gases, not the nitrogen shielding gas used in PART I, were tested under the same experimental conditions and the effect of the shielding gas on penetration during laser welding was conducted.

Environmental regulations of the IMO (International Maritime Organization) are becoming more and more conservative. In order to respond to IMO, the demand for replacing the fuel of ships with eco-friendly fuels instead of conventional heavy oil is increasing in the shipbuilding and offshore industries. Among eco-friendly fuels, LNG (Liquefied Natural Gas) is currently the most popular fuel. LNG is characteristically liquefied at -163 degrees, and at this time, its volume is reduced to 1/600, so it is transported in a cryogenic liquefied state for transport efficiency. A tank for storing this should have sufficient mechanical/thermal performance at cryogenic temperatures, and among them, high manganese steel is known as a material with high price competitiveness and satisfying these performance. However, high manganese steel has a limitation in that the mechanical performance of the filler metal is lower than that of the base metal called ‘under matching’. In this study, to overcome this limitation, a basic study was conducted to apply the fiber laser welding method without filler metal to high manganese steel. To obtain efficient welding conditions, in this study, bead-on-plate welding was performed by changing the fiber laser welding speed and output using helium shielding gas, and the effect of each factor on the penetration shape was analyzed through cross-sectional observation.

The Co-60 is a radioactive material widely used in domestic and foreign medical, industrial, health and research fields. Currently, world market for the Co-60 is about 80 MCi/yr and is expected to grow to 150 MCi/yr by 2025. For the Co-60, Nordion of Canada occupies about 80% of the world market. In the case of Korea, a small amount of sources with relatively low radioactivity intensity are produced using research reactors, but most of the Co-60 is entirely dependent on imports. Accordingly, although the technical feasibility of the Co-60 production technology using the PHWR was evaluated, it was evaluated as a negative result on the additional construction of a hot cell, core management, safety analysis and economic feasibility. Canada, the main producer of the Co-60, is also conducting research on the Co-60 production technology using PWR with GE-Hitachi and Westinghouse as the number of PHWR is expected to decrease. In Korea, it is necessary to preoccupy the Co-60 production technology and auxiliary technology using the PWR by utilizing excellent technology, and active research is being conducted to secure unique nuclear power technology that does not depend on foreign countries. Therefore, in this study, the thickness and weight of the radioactive shielding required for handling (transport) of Co-60 produced using the PWR were calculated.

The spent filters stored in Kori Unit 1 are planned that compressed and disposed for volume reduction. However, shielding reinforcement is required to package high-dose spent filters in a 200 L drum. So, in this study suggests a shielding thickness that can satisfy the surface dose criteria of 10 mSv·h−1 when packaging several compressed spent filters into 200 L drums, and the number of drums required for the compressed spent filter packaging was calculated. In this study, representative gamma-emitting nuclides in spent filter are assumed that Co-60 and Cs-137, and dose reduction due to half-life is not considered, because the date of occurrence and nuclide information of the stored spent filter are not accurate. The shielding material is assumed to be concrete, and the thickness of the shielding is assumed to 18 cm considering the diameter of the spent filter and compression mold. Considering the height of the compressed spent filter and the internal height of the shielding drum, assuming the placement of the compressed spent filter in the drum in the vertical direction only, the maximum number of packaging of the compressed spent filter is 3. When applying a 18 cm thick concrete shield, the maximum dose of the spent filter can packaged in the drum is 125 mSv·h−1, so when packaging 3 spent filters of the same dose, the dose of a spent filter shall not exceed 41 mSv·h−1 and not exceed 62 mSv·h−1when packing 2 spent filters. Therefore, the dose ranges of spent filters that can be packaged in a drum are classified into three groups: 0–41 mSv·h−1, 41–62 mSv·h−1, and 62–125 mSv·h−1based on 41 mSv·h−1, 62 mSv·h−1, and 125 mSv·h−1. When 227 spent filters stored in the filter room are classified according to the above dose group, 207, 3 and 4 spent filters are distributed in each group, and the number of shielding drums required to pack the appropriate number of spent filters in each dose group is 75. Meanwhile, 8 spent filters exceeding 125 mSv·h−1 and 5 spent filters that has without dose information are excluded from compression and packaging until the treatment and disposal method are prepared. In the future, we will segmentation of waste filter dose groups through the consideration of dose reduction and horizontal placement of compressed spent filters, and derive the minimum number of drums required for compressed spent filter packaging.

As the decommissioning of nuclear power plants increases, there is an increasing interest in the amounts of radioactive waste. Especially, the radiation dose limit for packaging of radioactive wastes shall not exceed 2 mSv·h−1 and 0.1 mSv·h−1 on contact and at 2 m, respectively in South Korea. The DEMplus provides various environmental geometry and all properties such as materials, absorptions, and reflections and the estimation of the radiation dose rates is based on the radiation interactions of the designed 3D geometry model. With the consideration of the radiation dose rate by using DEMplus and its strategy of packaging plan, the radiation shielding was optimized and estimated in this paper. The modular shielded containers (MSC) with shielding inserted were used for radioactive wastes that require shielded packaging. In order to verify the accuracy of the estimated radiation dose rate by using DEMplus, the estimated results were compared with those obtained using MicroShield. The trends of the estimated radiation dose rates using DEMplus and the estimation of MicroShield were similar to each other. The results of this study demonstrated the feasibility of using DEMplus as a means of estimating the radiation dose limit in packaging plan of the radioactive waste.

Preparation of advanced functional materials from agricultural waste by eco-friendly processing route is inevitable for sustainable development. This work demonstrates the development of carbon/silica (C/SiO2) and carbon/silicon carbide (C/ SiC) composite foam monoliths of low thermal conductivity, high EMI shielding performance and reasonable compressive strength from rice husk. The C/SiO2 and C/SiC composite foams are obtained by carbonization and subsequent carbothermal reduction, respectively, of rice husk–sucrose composites consolidated by filter-pressing rice husk powder dispersed in sucrose solutions of various concentrations (300–600 g L− 1). The amorphous nature of silica in C/SiO2 and the presence of β-SiC in C/SiC are evidenced from XRD and TEM analysis. The compressive strength and thermal conductivity are depending on the foam density which is tailored by sucrose solution concentration. The compressive strength in the ranges of 0.32–1.67 and 0.19–1.19 MPa are observed for C/SiO2 and C/SiC foams, respectively, with density in the ranges of 0.26–0.37 and 0.18–0.29 g cm− 3. The C/SiO2 and C/SiC exhibited thermal conductivity in the ranges of 0.150–0.205 W m− 1 K− 1 and 0.165–0.431 W m− 1 K− 1, respectively. The C/SiO2 and C/SiC composite foams show absorption dominated EMI shielding effectiveness in the ranges of 18–38.5 dB and 20–43.7 dB, respectively. The inherent pore channels and corrugated surface structure in rice husk, electrically conducting carbon and dielectric SiO2 and SiC contribute to the total EMI shielding.