A facile and efficient method was developed to prepare highly stretchable and conductive graphene conductors with wrinkled structures by the mechanical stretching and shrinking of elastomeric substrates, in which graphene inks were printed on a prestretched elastomeric substrate. Stretchable and exfoliated graphene inks were prepared by mixing graphite and Ecoflex in a shear-assisted fluid dynamics reactor. The resultant graphene conductor exhibited excellent stretchability at 150% strain and high electrical conductivity of 64 ± 1.2 S m− 1. The resistance of the conductor did not change in bent, twisted, and stretched states. The resistance did not change during 10,000 cycles of stretching/releasing, with a maximum strain of 150%. Based on the graphene conductor, a stretchable conductometric sensor with a two-electrode configuration was fabricated to measure impedance changes at different concentrations of electrolyte ions. This sensor exhibited a good and linear sensitivity curve (298.61 Ω mM− 1, R2 = 0.999) in bent and stretched states.

This study was performed to assess the cosmic-ray effect caused by altitude in the aerial gammaray measurement. For the gamma-ray measurement experiment by altitude, the aerial survey system composed of four 4×4×16 inches large volume NaI (Tl) detectors was used. The aerial survey system was installed in a rotor-craft to stably keep its flight altitude and position. In addition, in order to avoid to time-dependent shielding effects with the amount of fuel, a rotor-craft of which the fuel tank is not located beneath the cabin floor was selected. In this study, the ROI (Region Of Interest) was set to the 3~6 MeV range to assess the cosmic-ray contribution to the gamma-ray spectrum that could ignore the contribution of the dominant natural radionuclides. The gamma-ray spectra measured inside and outside of the rotor-craft on the ground were compared to evaluate the shielding effects of the aircraft body. As a result, the count rate of the 40K photo peak was decreased by about 10% when measuring the inside compared to the outside. On the other hand, the total count rate of the 3~6 MeV region was decreased by about 0.7% under the same condition. Therefore, the aircraft body effect was insignificant in 3~6 MeV region considering the relative uncertainty of 0.04~0.78% (1σ). In addition, the count rate in the 3~6 MeV range according to altitude was evaluated to assess the cosmic-ray effect. In order to evaluate the change in the ROI count rate according to the altitude, the gamma-ray spectrum was measured in the range of 300~2,000 m above the sea to avoid the effect of terrestrial radiation. As a result, the relationship between altitude and count rate in the 3~6 MeV range showed a high correlation with the R2 value of 0.99, when the approximate equation was derived in the form of a quadratic polynomial. Also, the count rate of 3~6 MeV at 50~500 m above the ground was estimated using the correlation equation, and this value was compared with the measured count rate. As a result of comparing the average value of estimated count rate and measured count rate, the relative difference is less than 2%. Considering the relative uncertainty of 0.78~4.11% (1σ), it was possible to evaluate the count rate of the 3~6 MeV region relatively accurately. The results of this study could be used for further study on background dose corrections in aerial survey.

An induction melting facility includes several work health and safety risks. To manage the work health and safety risks, care must be taken to identify reasonably foreseeable hazards that could give rise to risks to health and safety, to eliminate risks to health and safety so far as is reasonably practicable. If it is not reasonably practicable to eliminate risks to health and safety, attention have to be given to minimize those risks so far as is reasonably practicable by implementing risk control measures according to the hierarchy of control in regulation, to ensure the control measure is, and is maintained so that it remains, effective, and to review and as necessary revise control measures implemented to maintain, so far as is reasonably practicable, a work environment that is without risks to health or safety. The way to manage the risks associated with induction melting works is to identify hazards and find out what could cause harm from melting works, to assess risks if necessary – understand the nature of the harm that could be caused by the hazard, how serious the harm could be and the likelihood of it happening, to control risks – implement the most effective control measures that are reasonably practicable in the circumstances, and to review control measures to ensure they are working as planned.

For safe management of spent nuclear fuels, they should be delivered to repository or waste disposal site. As the amount of spent nuclear fuel transportation is expected to increase in the future due to the provision of an intermediate storage facility, the necessity to secure transportation cask is emerging. In order to secure the spent nuclear fuel transportation cask, it is necessary to analyze the regulatory processes for domestic and foreign spent nuclear fuel transportation cask. In this study, the regulatory processes for domestic and foreign spent nuclear fuel transportation cask was analyzed. In this study, the IAEA, US, and Korea spent nuclear fuel transportation cask regulatory processes were analyzed. The domestic and foreign spent nuclear fuel transportation cask regulatory processes consist of design phase, manufacturing phase, and operation phase. In the design stage, the transport requirements are designed in accordance with the safety requirements of international organizations and countries. The application to be submitted when applying for approval should include a safety analysis report, evidence proving compliance with safety requirements et al. In the manufacturing stage, it is a stage to check whether the safety requirements are satisfied before the first use after manufacturing the transportation cask. Inspections include welding inspection, leakage inspection, shielding inspection, and thermal inspection. In the operation stage, it is a stage of periodically performing inspections for continuous maintenance of the package when the transportation cask is used. The inspection items to be performed are similar to the manufacturing stage and typically include performance inspection of components and leakage inspection. In this study, domestic and foreign spent nuclear fuel transportation cask regulatory processes were analyzed. It was found that the domestic and foreign spent nuclear fuel transportation cask regulatory processes consist of the design phase, the manufacturing phase, and the operation phase. The results of this study can be used as basic data for policy decision-making for the spent nuclear fuel cask.

본 연구에서 Periphyllus acerihabitans를 국내에서 최초로 보고한다. 이 종의 분포지역, 기주식물, 무시성충의 형태학적 정보와 분류키를 제공 하였다.

The Kori Unit 1 and Wolsong Units 1, commercial reactors in South Korea, were permanently shut down due to the expiration of their design lifetime. Therefore, nuclear power plants that have been permanently shut down must be dismantled, and the site must be finally released after removing the remaining radionuclides. Domestic regulatory standards for site remediation should not exceed 0.1 mSv per year based on effective dose. In addition, it is necessary to calculate the preliminary Derived Concentration Guideline Levels (DCGL) to prove that the conditions are met. Therefore, in this study, the input factor considering the geological characteristics of the site of Kori Unit 1 was investigated, and the preliminary Derived Concentration Guideline Levels were calculated and compared with the results of previous studies. As a result of comparative analysis, 60Co, 134Cs, and 137Cs, which are gamma-ray emitting radionuclides, had similar values to DCGL of previous studies A and B. However, 63Ni, a beta-rayemitting nuclide, was 5.94×104 Bq·g−1 in this study and 8.47×101 Bq·g−1in previous study B, resulting in a difference of about 700 times. In addition, in the case of 90Sr, this study and previous study A were derived similarly, but this study was 5.34×101 Bq·g−1 and previous study B was 1.18×10−1 Bq·g−1, resulting in a difference of about 450 times. This difference is judged to be because, unlike this study using only the industrial worker scenario, in the case of previous study B, the resident farmer scenario was mixed and used, which considers the internal exposure caused by ingestion of food produced in the contaminated area. In this study, it was confirmed that DCGL according to the change of geological factors of the site did not have a significant effect on gamma-ray-emitting nuclides. However, it was confirmed that considering the intake of food affects the DCGL of beta-ray-emitting nuclides. Therefore, there is a need to conduct future studies applying intake input factors that meet domestic conditions.