이 연구에서는 CFRP 케이블의 압착형 정착장치의 강종에 따른 피로성능을 평가하였다. 실험결과, CFRP 케이블 압착형 정착시스템은 정착장치의 강종에 상관없이 도로교설계기준의 포스트텐션용 정착 구에 대한 피로시험 조건을 만족시켰다. 그러나 피로시험이후 인장성능 실험에서 SM45C 강종의 정착 장치는 최대내력 도잘전에 케이블과 정착장치 사이에서 미끄러짐이 발생하였으며, 듀플렉스 강종의 정 착장치는 최대 내력 도달과 동시에 CFRP 케이블이 파단되었다.

강섬유 보강 콘크리트(Steel Fiber Reinforced Concrete, SFRC)는 콘크리트의 취성적인 거동을 연성 거동으로 보완해주는 재료로서 사용되고 있다. SFRC는 과거 비 구조체에서 섬유 보강 콘크리트는 균 열을 제어하기 위한 목적으로만 사용되었지만 현재 구조적인 재료로서의 연구가 진행되고 있다. 콘크 리트의 압축강도는 강도계산에 사용되지만 재료의 인성(Toughness)을 평가하기 위해서는 응력-변형률 곡선이 필요하다. 본 연구에서는 섬유의 혼입량을 체적비 0.5%로 두어 나선철근의 유무 및 철근의 피 치를 변수로 두어 압축실험을 통해 인성비(Toughness Ratio)를 측정하였으며, 선행연구에 제안된 예측 식을 통해 실험 값과 해석 값을 비교하였다. 사용한 강섬유는 직경 0.75mm, 길이 60mm 및 인장강도 1,100MPa인 후크형 모양의 섬유를 사용하였으며, SFRC의 압축 실험은 200tonf 용량의 UTM을 사용 하여 응력-변형률 곡선을 확인하기 위해 0.5mm/min의 속도로 변위제어 하였다. Plain 시편의 압축강 도는 19.6MPa로 나타났으며, 52mm간격의 나선철근을 넣은 경우 33.3MPa, 섬유 0.5% 혼입한 경우 29MPa로 각각 70.1% 및 48.0% 높은 압축강도가 나타났다. 52mm 간격의 나선철근을 넣은 시편의 인성비는 0.34로 측정되었으며, 0.5%의 섬유가 혼입된 52mm 및 36mm간격으로 보강된 시편의 경우 각 0.32 및 0.49로 -6.5% 및 44.3%의 증감이 나타났다. 따라서 섬유의 보강으로 인한 SFRC의 인성 거동을 확인하였다. 예측식의 경우 52mm 나선철근으로 보강한 시편의 경우 압축강도 -0.06 및 변형 률 -7.37%의 오차율이 나타났으며, 36mm의 경우 각 11.51% 및 -36.5%의 오차율이 나타났다. 따라 서 예측식을 통해 나선철근으로 보강된 SFRC 강도예측을 확인하였다.

The operating parameters considered in this study include evaporating and condensing temperature, degree of subcooling and degree of superheating. in R744-R717 cascade refrigeration system and R744 two-stage compression refrigeration system with the range of low temperature -50℃∼-30℃. The coefficient of performance(COP) of R744-R717 cascade refrigeration system is about 16∼20% greater than that of R744 two-stage compression refrigeration system in the range of evaporation temperature of -50℃∼-30℃. R744 two-stage compression refrigeration system is unstable because its coefficient of performance changes significantly depending on the evaporating temperature and total compression work, and compression efficiency decreases. In this case, not efficient for long-term use. Whereas R744-R717 R744-R717 is a cascade refrigeration system using eco-friendly refrigerants. And this system is a high-efficiency refrigeration system that performs well even under various operating conditions. This is why it can be configured by selecting a refrigerant suitable for the high temperature side and the low temperature side.

The bentonite buffer material is a crucial component for disposing of high-level radioactive waste (HLW). Several additives have been proposed to enhance the performance of bentonite buffer materials. In this study, unconfined compression tests were conducted on bentonite mixtures as well as pure bentonite buffer material. Joomunjin and silica sands were added at a 30% ratio, and graphite was added at 3% along with bentonite. The unconfined compression strength (UCS) and elastic modulus of pure bentonite were found to be 20% to 50% higher than those of bentonite mixtures under similar dry density and water content conditions. This decrease in strength can be attributed to the reduced cross-sectional area available for bearing the applied load in the bentonitemixture. Furthermore, the 3% graphite-bentonite mixture exhibited a 10% to 30% higher UCS and elastic modulus compared to the 30% sand-bentonite mixtures. However, since the strength properties of additive-bentonite mixtures are lower than those of pure bentonite, it is essential to evaluate thermohydraulic-mechanical functional criteria when considering the use of bentonite mixtures as buffer materials.

Nuclear fuel assemblies are exposed to high temperature and high pressure environments underwater for long periods of time in a reactor, leading to deterioration of the assembly structure. These assembly consists of fuel rods, grids, a top nozzle, a bottom nozzle and guide tubes. In particular, the integrity of the guide tube made of Zircaloy-4 is a very important part in handling the assembly. In the Post Irradiation Examination Facility (PIEF), there are 14×14 Westinghouse STD assemblies that have lost their handleability due to the top nozzle being removed for damaged fuel rod test. To handle these assemblies, it is reasonable to use cut guide tubes whenever possible. Therefore, it is necessary to determine the irradiation embrittlement state of the guide tube before designing or manufacturing parts that can connect the top nozzle and the guide tubes. Therefore, in this paper, the location for installing the top nozzle-guide tube connection parts was selected in the height range of 3,460 to 3,713 mm, and guide tube specimens were made within that range. Offset strain was derived from the load-displacement curve obtained through compression testing to confirm whether the ductility of guide tubes was maintained. As a result, there was no significant difference in strength and ductility of the guide tube within the above length range. In addition, it was confirmed that the ductility was maintained enough to install the top nozzle-guide tube connection parts. Therefore, it is judged that there will be no problem even if the top nozzle-guide tube connection parts are installed in the guide tube to restore the handleability of the assemblies.

Radioactive waste (hereinafter referred to as mixed waste) containing hazardous substances (heavy metals, organic and inorganic waste liquids, asbestos, etc.) has been continuously generated from domestic nuclear power plants, nuclear facilities, and other industrial facilities, and heavy metals were released during the dismantlement of Kori Unit 1 and Wolseong Unit 1. Lead, cadmium, mercury, arsenic), asbestos, decontamination waste liquid (organic/inorganic waste liquid), etc. may be generated. Although hazardous waste related to the nuclear industry continues to be generated, only the regulation direction for hazardous substances is presented in the provisions related to hazardous substances in the delivery regulations for low and intermediate-level radioactive waste and the acceptance criteria for low and intermediate-level radioactive waste disposal facilities. In particular, because there is no clear definition of “hazardousness” and specific standards such as concentration and characteristics for classification of hazardous substances, as well as hazard removal procedures when the hazardousness of radioactive waste is confirmed, no hazardous substances have been delivered in Korea to date and many mixed wastes are stored at each generation facility or at the NPP. As a plan to improve delivery standards related to mixed waste is being prepared recently, it is believed that if the acceptance standards are revised accordingly, it will be possible to confirm the suitability for disposal of drums produced after the establishment of the acceptance standards in 2015. However, it is believed that securing disposal suitability for waste that was packed in 200L drums and compressed under super high pressure in the absence of specific technical standards and regulatory guidelines for the disposal of radioactive waste containing hazardous substances would still remain a difficult problem. In this report overseas acceptance standards related to hazardous waste were reviewed and a plan to secure the disposal suitability of 200 L drums compressed with of super high pressure was proposed.

The spent filters used to purify radioactive materials and remove impurities from primary systems at nuclear power plants (NPPs) have been stored for long periods in filter storage rooms at NPPs due to concerns about the unproven safety of the treatment method, absence of disposal facilities, and risk of high radiation exposure. In the storage room at Kori Unit 1, there are approximately 227 spent filters of 9 different types. The radiation dose rates of filters range from 0.01 to 500 mSv/hr. Recently, a comprehensive plan has been established for the treatment and disposal of radioactive waste that has not yet been treated to facilitate decommissioning of NPPs. As a follow-up measure, compression and packaging optimization processes are being developed to treat the spent filters. KHNP plans to dispose of the spent filters after compressing, packaging, and immobilizing them. However, the spent filters are currently stored without being sorted by type or radiation intensity. If the removal and packing of the filters are done randomly without a plan for the order of withdrawal and subsequent processes, issues may arise such as a decrease in drum loading efficiency and exceeding the dose limit of the package. In this study, the number of drums needed to pack the spent filters was calculated, considering the filter size, weight, quantity, dose rate, shielding thickness of drum, and loadable quantity in a shielding drum (SD). Then, the spent filters that can be loaded on each drum were classified into one group. In addition, the withdrawal order for each group was set so that the filter withdrawal, compression, and packaging processes could be performed efficiently. The spent filter groups are as follows: (1) compression/12 cm SD (17 groups), (2) compression/16 cm SD (6 groups), (3) non-compression/ intermediate storage container (17 groups, additional radiation attenuation required due to high dose rate), and (4) unclassified (5 groups, determined after measurement due to lack of filter information). The withdrawal order of the groups was determined based on several factors, including visual identification of the filter, ease of distribution after withdrawal, work convenience, and safety. Due to the decay of radioactivity over time, the current dose rate of the spent filters is expected to be much lower than at the time of waste generation. Therefore, in the future, sample filters will be taken from the storage room to measure their radioactivity and radiation dose rate. Based on these measurements, a database of radiological characteristics for the 227 filters will be created and used to revise the filter grouping.

When damaged nuclear fuel is stripped and re-fabricated into stabilized pellets, it is necessary to analyze the characteristics of the stabilized pellets, such as density, leaching behavior, and compressive strength, for final disposal. In this study, simulated nuclear fuel with UO2 and burn-up of 35 GWd/tU and 55 GWd/tU was used to measure the compressive strength of the stabilization pellet. In order to change the density of the sintered pellet, a sintered pellet was prepared by heat treatment at 1,550°C and 1,700°C for 6 hours in a reducing atmosphere of 4% H2/Ar. In the case of UO2, the density was 10.4 g/cm3 (94.5% of T.D.) and 10.6 g/cm3 (96.6% of T.D.) depending on the sintering temperature (1,550°C, 1,700°C). In the case of simulated fuel with a burn-up of 35 GWd/tU, the density was 8.8 g/cm3 (80.9% of T.D.) and 10.2 g/cm3 (93.6% of T.D.) depending on the sintering temperature (1,550°C, 1,700°C). In the case of simulated fuel with a burn-up of 55 GWd/tU, the density was 8.3 g/cm3 (77.0% of T.D.) and 10.0 g/cm3 (92.3% of T.D.) depending on the sintering temperature (1,550°C, 1,700°C). It was found that the compressive strength of simulated nuclear fuel decreased with increasing burn-up and increased with increasing density. In the case of UO2, the compressive strengths were 717.8 MPa and 897.4 MPa when the densities were 10.4 g/cm3 and 10.6 g.cm3, respectively. In the case of simulated nuclear fuel with a burn-up of 35 GWd/tU, the compressive strengths were 472.1 MPa and 732.3 MPa when the densities were 8.8 g/cm3 and 10.2 g/cm3. In the case of simulated nuclear fuel with a burn-up of 55 GWd/tU, the compressive strengths were 301.4 MPa and 515.5 MPa when the densities were 8.3 g/cm3 and 10.0 g/cm3, respectively.

목적 : 온열 안구 마사지기 사용 시 실시간으로 눈꺼풀 온도 및 눈물막 파괴 시간을 측정하여 차이를 평가하고 자 하였다. 방법 : 연구 대상은 성인 30명 (23.7±2.60세)을 대상으로 하였고 OSDI설문지 점수에 따라 건성안 그룹과 정 상안 그룹으로 나누었다(건성안 15명, 정상안 15명). 온열 안구 마사지기는 OA-MA011(OA, Seoul, Korea), OA-MA 030(OA, Seoul, Korea), DP-EM50(Caremedi, Hanam, Korea)를 사용하였고, A400 열화상카메라 (FLIR, Wilsonville, Clackamas County USA) 사용하여 착용 전과 후 1, 3, 5, 10, 15분 측정하고, 탈거 후 5, 10, 15분 시간별로 측정 후 평균값을 기록하였다. 그리고 Kekatography 5 m(OCULUS, Wetzlar, Germany)를 사용하여 마사지기 착용하기 전과 착용 15분 뒤에 눈물막 파괴 시간을 측정해서 비교 분석하였다. 결과 : 15분 착용 시 눈꺼풀 온도는 건성안 그룹은 39.77±0.69℃, 정상안 그룹은 39.57±0.77℃ 측정되었 으며, 눈꺼풀 온도 측정 시 건성안, 정상안 그룹 모두 5분 이상 착용 시 통계적으로 유의성이 있었다(p=0.033). 눈물막 파괴시간은 착용하기 전보다는 향상이 있었지만 정상범위에 못 미치는 것으로 확인하였고 건성안에서는 통 계적으로 유의성이 있었지만(p=0.041) 정상안 그룹에서는 통계적으로 유의성이 없었다(p=0.084). 결론 : 안구건조증으로 인한 개선 방법으로 안구마사지기를 사용했을 때 10분 이상 착용 시 눈꺼풀 높은 온도 에서 유지가 되었고 눈물막 파괴 시간은 15분 이상 착용 및 온도 40℃ 유지 시 눈물막 파괴 시간에 영향을 줄 수 있었다. 따라서 온열 기능의 마사지기를 사용 시 정확한 사용 방법과 눈꺼풀 온도가 유지되며 착용 시간은 충분한 상태에서 진행해야 할 것으로 사료 된다.

In this study, the process of compressing/packaging the spent filters of Kori Unit 1, which was conceptually presented in the previous study, is advanced so that disposal suitability for each step can be secure efficiently. In particular, the differences between the previous study and this study are that the disposable filters are screened using an In-Situ Object Counting System (ISOCS), and the method of collecting representative samples for development of scaling factor is specified. The process of compressing/packaging the spent filters consists of 7 stages as follows. 1) Collecting: The spent filters temporarily stored in the filter room are collected by dose and type remotely using a robot system to minimize the radiation exposure of workers according to a pre-established packaging plan. 2) Screening: The gamma activity concentration of the spent filters received by the robot system is measured by ISOCS. The spent filters below the low-level waste concentration limit and the surface dose are transferred into the compression system, while the others are returned in the filter room again. 3) Sampling: The external perforator drilling/cutting the filter was developed for sampling required for the new scaling factors. Since the sampling is collected remotely, the risk of exposure to workers can be reduced. The newly developed scaling factor will be used to verify the disposal suitability of the packages. 4) Compression: According to the pre-established plan, the spent filter collected by dose and type, is supplied to the compression system considering the dose and radionuclide inventory. Whether to additionally store the compressed filter in the drum is determined by checking the accumulated dose. 5) Immobilization: Immobilization with a safety material is necessary when inhomogeneous wastes, like spent filters, have the total radionuclide concentration with a half-life of more than 20 years is 74,000 Bq/g or more and for filling rate or non-dispersible treatment of particulates. 6) Packaging and Analysis: Waste information is labelled onto the package after the measurements of surface dose rate and surface contamination. Finally, using the drum assay system, the gamma radionuclide concentration is measured to identify at least 95% of the total radioactivity concentration of the package. 7) Temporary Storage and Delivery: The packages are moved to temporary storage in the plant prior to disposal. After establishing the plan for delivery and applying for a takeover request to KORAD, if the acceptance inspection is passed, the packages are transported to the disposal facility.

Diesel engine has the advantages of strong power, low fuel consumption and good durability, so it has been widely used in transportation, automobile, ship and other fields. However, the nitrogen oxides(NOx) and particulate matter(PM) emitted by diesel engines have become one of the main causes of air pollution. Especially during idling, the engine temperature is low, and there are more residual exhaust gases in the combustion chamber, resulting in the formation of more harmful emissions. In this study, performance of a single cylinder, four-stroke, direct injection (DI) diesel engine fueled with diesel–biodiesel mixtures has been experimentally investigated. The findings show that a remarkable improvement in PM–NOx trade-off can be achieved by burning diesel-bioethanol blend fuels.

Background: Low back pain (LBP) is a representative disease, and LBP is characterized by muscle dysfunction that provides stability to the lumbar spine. This causes physical functional problems such as decreased posture control ability by reducing the muscular endurance and balance of the lumbar spine. Pelvic compression using instruments, which has been used during recent stabilization exercises, focuses on the anterior superior iliac spine of the pelvis and puts pressure on the sacroiliac joint during exercise, making the pelvis more symmetrical and stable. Currently, research has been actively conducted on the use of pelvic compression belts and non-elastic pelvic belts; however, few studies have conducted research on the application effect of pelvic compression using instruments. Objects: This study aimed to investigate whether there is a difference in trunk muscular endurance and dynamic and static balance ability levels by applying pelvic stabilization through a pelvic compression device between the LBP group and the non-LBP group. Methods: Thirty-nine subjects currently enrolled in Daejeon University were divided into 20 subjects with LBP group and 19 subjects without LBP (NLBP group), and the groups were compared with and without pelvic compression. The trunk muscular endurance test was performed with 4 movements, the dynamic balance test was performed using a Y-balance test, and the static balance test was performed using a Wii balance board. Results: There was a significant difference the LBP group and the NLBP group after pelvic compression was applied to all tests (p < 0.05). In the static and dynamic balance ability test after pelvic compression was applied, there was a significant difference in the LBP group than in the NLBP group (p < 0.05). Conclusion: These results show that pelvic compression using instruments has a positive effect on both those with and without LBP and that it has a greater impact on balance ability when applied to those with LBP.

During normal and off-normal conditions, the concrete structures of dry storage system for spent nuclear fuel must maintain structural integrity. A stress-strain curve is the most important key factor for structural integrity evaluation. The ASTM C39 specifies the concrete specimen geometry for the static compression test. However, there is no standard specimen size for intermediate stain rate, and it is not easy to maintain consistency among all test results because the failure tendency is different from each other. In order to account for the strain rate effects on concrete, the dynamic increase factor (DIF) is conventionally addressed by dividing dynamic strength by static strength. However, the DIF value considers only the strength of concrete and does not describe the overall behavior of concrete, such as a stress-strain relation. The objective of this study is to propose proper specimen geometry for the concrete dynamic compression test by several parametric study. The static compression simulation results with the specimen specified in ASTM C39 showed the constant strain distribution in a cylindrical specimen. However, as the strain rate increases, the strain state in specimen showed a nonuniform with the same geometry of ASTM C39. The non-uniform strain state in the specimen deteriorates the consistency and accuracy of the compression test. Therefore, we presented the specimen shape and size to form a uniform strain state through radial direction by drilling a hole in the axial direction. We analyzed two specimens using ABAQUS with the concrete damaged plasticity model, one with a hole at the center and the other without the hole. As a result, the strain distribution became more uniform than the specimen without the hole. Based on the results, we proposed the specimen shape and size for the intermediate strain rate compression test.

본 논문에서는 CAD 시스템에서 사용하는 NURBS 기저함수를 사용하는 아이소-지오메트릭 해석(Isogeometric analysis) 방법과 기 하학적으로 엄밀한 빔 모델링(geometrically exact beam model)을 활용하여 회전과 병진 운동이 결합된 새로운 형태의 메타물질 (metamaterial)에 대한 해석을 진행하였다. 이차원 셀 구조는 자유형상변환(Free-form deformation) 법과 적절한 내삽법(Interpolation) 을 통해 원통 위에 입혀졌다. 원통의 치수와 셀 개수가 비틀림 각도에 미치는 영향이 매개변수 연구(parametric study)를 통해 확인되었 다. 비틀림과 병진 운동이 결합된 구조의 메커니즘에 대해 수치 예제를 통해 알아보았다.

Background: Compression of the sacroiliac joint (SIJ) enables lumbo-pelvic muscles to increase pelvic stability and reduce pain. Many previous studies assessed the effects of pelvic compression belts on the stabilization of the lumbar and SIJs. However, there are a lack of studies to date on the effects of pelvic compression taping (PCT). Objectives: To investigate the effects of PCT on the muscle activity of lumbopelvic muscles during prone hip extension (PHE). Design: Pre-post test design. Methods: In total, 19 healthy male and female participants performed PHE with and without PCT. A surface electromyography system was used to record the muscle activity of multifidus, erector spinae, internal oblique abdominal, and gluteus maximus muscles, and the activities of these muscles before and after taping were compared. Results: During PHE, the activities of the multifidus and erector spinae muscles (P<.05) significantly increased after taping, while that of the internal oblique abdominal muscle significantly decreased after taping (P<.05). However, the activity of the gluteus maximums muscle did not significantly increase or decrease after taping (P>.05). Conclusion: PCT for the SIJ improves the stability of the lumbo-pelvic region, and the use of PCT with functional training would lead to positive clinical effects.