In addition to Korea, various countries such as the United States, the United Kingdom, France, and China are designing small module-type reactors. In particular, a small modular reactor is the power of 300 MWe or less, in which the main equipment constituting the nuclear reactor is integrated into a single container. Depending on the purpose, small modular reactors are being developed to help daily life such as power, heating supply, and seawater desalination, or for power supply such as icebreakers, nuclear submarines, and spacecraft propellants. Small modular reactors are classified according to form. It can be classified into light-water reactors/ pressurized light-water reactors based on technology proven in commercial reactors, and non-lightwater reactors based on fuel and coolant type such as Sodium-cooled Fast Reactor, High temperature gas-cooled reactor, Very high temperature reactor and Moltenn salt reactor. SMRs, which are designed for various purposes, have the biggest difference from commercial nuclear reactors. The size of SMRs is as small as 1/5 of that of the commercial reactors. Several modules may be installed to generate the same power as commercial reactors. Because of the individually operation for each module, load follow is possible. Also, The reactor can be cooled by natural convection because the size is small enough. It is manufactured as a module, the construction period can be reduced. Depending on the characteristics of these SMRs, application for safeguards is considered. There are many things to consider in terms of safeguards. Therefore, it is IAEA inspection or other approaches for SMRs installed and remotely operated in isolated areas, data integrity for remote monitoring equipment to prevent the diversion of nuclear materials, verification method and material accountancy and control for new fuel types and reactors. Since SMR is more compact and technical intensive, safeguards should be considered at the design stage so that safeguards can be efficiently and effectively implemented, which is called the Safeguards by design (SBD) in the IAEA. In this paper, according to the characteristics of SMR, we will analyze the advantages/disadvantages from the point of view of safeguards and explain what should be considered.

The measurement activities to evaluate material balance of nuclear material are usually performed by operator. It is because that the IAEA does not have enough manpower to carry out nuclear measurement accountancy of all nuclear materials in the world. Therefore, the IAEA should consider scenarios which facility operator tries to divert nuclear material for misuse by distorting measurement record. It is required to verify the operator’s measurement data whether it is normal or not. IAEA measures inventory items using their own equipment which is independent of facility operator equipment for verification. Since all inventory lists cannot be verified due to limited resources, the number of items to be verified is determined through statistical method which is called as sample size calculation. They measure for the selected items using their own equipment and compares with operator’s record. The IAEA determines sample size by comprehensively considering targeted diverted nuclear material amount and targeted non-detection probability and performance of measurement equipment. In general, the targeted diverted nuclear material amount is considered significant quantity (plutonium: 8 kg, uranium-235: 75 kg). If the targeted non-detection probability or the performance of the verification equipment is low, the sample size increases, and on the contrary, in the case of high non-detection probability or good performance of verification equipment, even a small sample size is satisfied. It cannot be determined from a single sample size calculation because there are so many sample size combinations for each verification equipment and there are many diversion scenarios to be considered. So, IAEA estimates initial sample size based on statistical method to reduce calculation load. And then they calculate non-detection probability for a combination of initial sample size. Through the iteration calculation, the sample size that satisfies the closest to the target value is derived. The sample size calculation code has been developed to review IAEA’s calculation method. The main difference is that IAEA calculates sample size based on approximate equation, while in this study, sample size is calculated by exact equation. The benchmarking study was performed on reference materials. The data obtained by the code show similar results to the reference materials within an acceptable range. The calculation method developed in this study will be applied to support IAEA and domestic inspection activities in uranium fuel fabrication facility.

목적 : 본 연구는 정상안에서 눈물막 지질층이 시력의 질에 미치는 영향을 알아보고자 하였다. 방법 : 건성안을 포함하여 안과적 질환과 관련된 진단을 받은 적이 없으며 교정시력이 20/20 이상인 30명(60 안)을 대상으로 하였다. 굴절이상, 눈물막 지질층 두께, 대비감도를 검사하였고, 눈물막 지질층은 정량적으로 두께 를 측정하여 3 그룹(15~30 nm, 30~80 nm, ＞80 nm)로 분류하였다. 대비감도는 명소시 상태에서 공간주파수 1.5, 3, 6, 12, 18 cpd(cycles per degree)에 대하여 각각 검사하였다. 결과 : 전체 대상자 중 44안이 80 nm 보다 얇은 1, 2 그룹에 속했고 눈물막 지질층 두께 측정에서 80 nm보다 두꺼운 3 그룹에서만 정상 대비감도를 보였다. 눈물막 지질층 두께가 증가할수록 모든 영역의 공간주파수에서 대 비감도가 향상되는 것을 확인할 수 있었고, 눈물막 지질층 두께가 두꺼워질수록 굴절이상도가 높게 나타났다. 결론 : 교정시력 20/20 이상에서 눈물막 지질층의 불안정이 대비감도 저하에 미치는 영향을 확인하였다. 따라 서 시력검사 과정에서 안정된 눈물막 유지를 위한 검사자의 세심한 조치가 필요하며, 눈물막 지질층과 대비감도 검사 병행은 더 나은 시력의 질을 제공하는데 도움이 될 것으로 사료된다.

The fundamental characteristics of groundwater colloids, such as composition, concentration, size, and stability, were analyzed using granitic groundwater samples taken from the KAERI Underground Research Tunnel (KURT) site by such analytical methods as inductively coupled plasma-mass spectrometry, field emission-transmission electron microscopy, a liquid chromatography-organic carbon detector, and dynamic light scattering technique. The results show that the KURT groundwater colloids are mainly composed of clay minerals, calcite, metal (Fe) oxide, and organic matter. The size and concentration of the groundwater colloids were 10–250 nm and 33–64 μg·L−1, respectively. These values are similar to those from other studies performed in granitic groundwater. The groundwater colloids were found to be moderately stable under the groundwater conditions of the KURT site. Consequently, the groundwater colloids in the fractured granite system of the KURT site can form stable radiocolloids and increase the mobility of radionuclides if they associate with radionuclides released from a radioactive waste repository. The results provide basic data for evaluating the effects of groundwater colloids on radionuclide migration in fractured granite rock, which is necessary for the safety assessment of a high-level radioactive waste repository.

The purpose of using coolant in machining is both to increase a tool life and also to prevent product deformation and thus, stabilize the surface quality by lubricating and cooling the tool and the machining surface. However, a very small amount of cutting mist should be used because chlorine-based extreme pressure additives are used to generate environmental pollutants in the production process and cause occupational diseases of workers. In this study, medical titanium alloy (Ti-6Al-7Nb) was subjected to a processing experiment by selecting factors and levels affecting cutting power in the processing of the Aerosol Dry Lubrication (ADL) method using vegetable oil. The machining shape was a slot to sufficiently reflect the effect of the cutting depth. As for the measurement of cutting force, the trend of cutting characteristics was identified through complete factor analysis. The factors affecting the cutting force of ADL slot processing were identified using the reaction surface analysis method, and the characteristics of the cutting force according to the change in factor level were analyzed. As the cutting speed increased, the cutting force decreased and then increased again. The cutting force continued to increase as the feed speed increased. The increase in the cutting depth increased the cutting force more significantly than the increase in the cutting speed and the feed speed. Through the reaction surface analysis method, the regression equation for predicting cutting force was identified, and the optimal processing conditions were proposed. The cutting force was predicted from the secondary regression equation and compared with the experimental value.

Store-operated Ca2+ entry (SOCE) represents one of the major Ca2+ entry routes in non-excitable cells. It is involved in a variety of fundamental biological processes and the maintenance of Ca2+ homeostasis. The Ca2+ releaseactivated Ca2+ (CRAC) channel consists of stromal interaction molecule and Orai; however, the role and action of Homer proteins as an adaptor protein to SOCE-mediated Ca2+ signaling through the activation of CRAC channels in non-excitable cells still remain unknown. In the present study, we investigated the role of Homer2 in the process of Ca2+ signaling induced by the interaction between CRACs and Homer2 proteins in non-excitable cells. The response to Ca2+ entry by thapsigargin-mediated Ca2+ store depletion remarkably decreased in pancreatic acinar cells of Homer2–/– mice, as compared to wild-type cells. It also showed critical differences in regulated patterns by the specific blockers of SOCE in pancreatic acinar cells of Homer2–/– mice. The response to Ca2+ entry by the depletion in Ca2+ store markedly increased in the cellular overexpression of Orai1 and STIM1 as compared to the overexpression of Homer2 in cells; however, this response was remarkably inhibited by the overexpression of Orai1, STIM1, and Homer2. These results suggest that Homer2 has a critical role in the regulatory action of SOCE activity and the interactions between CRAC channels.

This study is about the production of radiation sources of simulated concrete and soil reference materials to verify the validity of the quality establishment and measurement of the detector (HPGe) of the radioactive soil and concrete waste classification system, which is being developed to quickly and accurately classify nuclear decommissioning waste. Specific activity of gamma nucleus among radioactive wastes is evaluated using gamma spectroscopy. At this time, in order to verify the validity and reliability of measuring equipment, it shall be a standardized substance of the same medium as nuclear decommissioning waste (chemical ingredients, particles, density, etc.) in order to correct the energy and efficiency of gamma nuclide analysis equipment. The CRM used for the detector’s energy correction used a 1 L Marinelli beaker standard correctional radiation source consisting of 10 radioactive isotopes. In order to correct efficiency, in accordance with the production and certification process of the Korea Standards and Research Institute, it has produced artificial simulated radioactive concrete similar to nuclear decommissioning waste (30% for cement, 60% for regulation and 10% for bentonite). The radioactive homogeneity of the simulated concrete reference materials was evaluated using dispersion analysis (ANOVA) in accordance with ISO Guide 35, while 137Cs and 60Co of concrete reference materials were able to obtain homogeneous measurements both in and between bottles. The self-absorption rate of the simulated concrete reference material was determined by the MCNP computer simulation measurement method, and the self-absorption correction coefficients of 137Cs and 60Co were assessed at 0.995 and 0.996, respectively, and the standard value for the radiation of the simulated concrete reference material was calculated on the weighted average of the measurements of 20 samples. The uncertainty about the reference value was calculated by combining measurement uncertainty (Type B evaluation), bottle to bottle standard deviation, and uncertainty within bottle by modifying the formula suggested in ISO Guide 35. The concentration of 137Cs and 60Co of reference materials was divided into high-speed measurement mode and precision measurement mode in consideration of the self-disposal standard. The reference value and uncertainty of expansion among reference materials for high-speed measurement mode were rated at 1,032.7 ± 64.0 Bq·kg−1and 1,083.7 Bq·kg−1, respectively. The standard value and expansion uncertainty for 137Cs and 60Co among reference materials for precision measurement mode were rated at 113.7 ± 10.0 Bq·kg−1 and 122.3 ± 10.3 Bq·kg−1, respectively.

As Kori-1 permanently shut down in Korea, it is expected that a large amount of radioactive waste will be generated during decommissioning of nuclear power plants. Radioactive concrete waste is contaminated up to depth of 100 mm with radionuclides such as 137Cs and 60Co. The radioactive waste should be accurately classified to reduce the cost of disposing of radioactive waste. Therefore, the specific radioactivity of waste must be precisely evaluated by gamma-ray measurements emitted from the radionuclides. In general, the effectiveness of the radioactivity measurement and process is confirmed using certified reference material (CRM) composed of water or agar. However, the decommissioning waste differs from this CRM in apparent density and chemical elements, so the specific radioactivity is underestimated or overestimated. Therefore, reference material composed of the same apparent density and chemical elements as the sample is required to improve the quality of radioactivity measurement. The purpose of this study is to develop a concrete reference material for the nuclear decommissioning waste. The concrete reference material composed of SiO2, CaO, and Al2O3 were manufactured in compliance with ISO Guide 35. 10 bottles were randomly selected for homogeneity test, and 2 samples for analysis were taken from each bottle. The specific radioactivity was measured using an HPGe detector with an efficiency of 30%. The results of the homogeneity test of 137Cs and 60Co satisfied the requirements of ISO Guide 35. Coincidence summing and selfabsorption effects were corrected using the Monte Carlo efficiency transfer code and Monte Carlo NParticle transport code. The reference values of 137Cs and 60Co in the concrete reference material were evaluated in the range of 1,000–1,100 Bq·kg−1 and extended uncertainty was around 7%.