사범대학 학생은 대학생일 뿐만 아니라 미래의 교사로서 자신의 핵심 소양을 향상시킬 뿐만 아니라 학생의 핵심 소양을 발전시키는 중요한 책 임을 져야 한다. 본 논문의 연구 목적은 사범생의 직업적 인식이 핵심 소양에 미치는 영향을 조사하는 것이다. 연구 방법은 정량적 연구 방법 과 SPSS 26.0 통계 소프트웨어를 사용하여 데이터를 분석하였다. 연구 결과는 사범생의 직업 정체성과 핵심 소양 사이에 유의미한 정적 상관관 계가 있음을 보여주며, 직업적 정체성의 각 차원은 핵심 소양에 예측 효 과가 있는 것을 보여 주었다. 이 연구를 통해 우리는 사범생의 핵심 소 양 형성에 영향을 미치는 요인과 메커니즘을 찾고 사범생의 인재 양성과 기초 교육 발전을 향상시키기 위한 이론적 참고 자료를 제공한다.

The rapid spread of novel infectious diseases, driven by globalization and transportation, has caused significant global damage. This study aimed to deeply explore the meaning of nursing experiences from nurses who performed holistic care for COVID-19 patients. Methods: Nurses from a tertiary hospital's COVID-19 isolation ward from January 1 to March 31, 2021, were included. Data were collected through 1:1 in-depth interviews with 17 nurses who understood the research purpose and agreed to participate. The transcribed interview data were analyzed using thematic analysis. Results: Five themes and 13 sub-themes were derived from 47 codes. The five themes included 'The beginning of new nursing,' 'New nursing while wearing personal protective equipment,' 'End-of-life care for patients,' 'Fruitfulness of nursing for COVID-19 patients,' and 'Demands for better nursing.' Conclusion: Nurses provided holistic and high-complexity nursing care to COVID-19 patients, experienced personal growth, and recognized the importance of colleagues. The unique context of 'isolation' for COVID-19 patients has resulted in profound end-of-life care experiences for nurses, highlighting the demand for support programs to address this issue. Furthermore, it shows the necessity for staffing support tailored to patient care.

In this study, the impact load resulting from collision with the fuel rods of surrogate spent nuclear fuel (SNF) assemblies was measured during a rolling test based on an analysis of the data from surrogate SNF-loaded sea transportation tests. Unfortunately, during the sea transportation tests, excessive rolling motion occurred on the ship during the test, causing the assemblies to slip and collide with the canister. Hence, we designed and conducted a separate test to simulate rolling in sea transportation to determine whether such impact loads can occur under normal conditions of SNF transport, with the test conditions for the fuel assembly to slide within the basket experimentally determined. Rolling tests were conducted while varying the rolling angle and frequency to determine the angles and frequencies at which the assemblies experienced slippage. The test results show that slippage of SNF assemblies can occur at angles of approximately 14° or greater because of rolling motion, which can generate impact loads. However, this result exceeds the conditions under which a vessel can depart for coastal navigation, thus deviating from the normal conditions required for SNF transport. Consequently, it is not necessary to consider such loads when evaluating the integrity of SNFs under normal transportation conditions.

It is crucial to understand the hydro-mechanical behavior of rock mass to assess the performance of natural barriers. As rock fractures serve as both mechanically weak planes and prominent pathways for hydraulic flow, they significantly influence the hydro-mechanical behavior of the rock mass. Hence, understanding the characteristics of rock fractures is necessary to analyze the long-term behavior of natural barriers. In particular, fracture apertures are crucial parameters directly associated with groundwater flow and consequently hold significant importance in determining the hydro-mechanical behavior of natural barriers. Fracture apertures are defined as mechanical and hydraulic apertures, and various studies have been conducted to measure and analyze them. However, direct measurement of mechanical aperture according to changes in normal stress is known to be a challenging task. For this reason, there has been a scarcity of direct comparative findings between mechanical and hydraulic apertures under various normal stress conditions. This study aims to analyze the characteristics of the mechanical and hydraulic apertures according to changes in normal stress based on experimental results. A digital analysis technique using a pressure film image was applied to analyze the mechanical aperture characteristics of the fracture. This technique can be applied by performing a pressure film compression test and a normal stiffness test on a fracture specimen, and has the advantage of being able to derive mechanical apertures under various normal stress conditions. The hydraulic aperture characteristics of the fracture were analyzed based on Cubic law after measuring the flow rate by performing a constant pressure injection test under triaxial compression conditions. By applying various confining pressures, it was possible to examine the hydraulic apertures according to changes in normal stress conditions. Through the experimental results, the relationship between the mechanical and hydraulic apertures of the fracture was summarized under various normal stress conditions. In addition, the experimental results were used to examine the applicability of various empirical equations for mechanical and hydraulic apertures proposed in previous studies. The characteristics of the fracture aperture resulting from this study are significant because they are required in the hydro-mechanical model of natural barriers. Future studies will entail further experiments, with the objective of establishing novel relationships based on the accumulation of experimental data.

There are many types of foam molding methods. The most commonly used methods are the pressure foaming method, in which foam resin is mixed with a foaming agent at high temperature and high pressure, and the normal pressure foaming method, which foams at high temperature without pressure. The polymer resins used for foaming have different viscosities. For foaming under normal pressure, they need to be designed and analyzed for optimal foaming conditions, to obtain resins with low melt-viscosity or a narrow optimal viscosity range. This study investigated how changes in viscosity, molding temperature, and cross-link foaming conditions affected the characteristics of the molded foam, prepared by blending rubber polymer with biodegradable resin. The morphologies of cross sections and the cell structures of the normal pressure foam were investigated by SEM analysis. Properties were also studied according to cross-link/foaming conditions and torque. Also, the correlation between foaming characteristics was studied by analyzing tensile strength and elongation, which are mechanical properties of foaming composites.

본 연구는 우리나라에서 수수-수단그라스 교잡종 (sorghum bicolor L.: SSH)에 대해 극단기상과 정상기상 간 생산량을 비교할 목적으로 수행하였다. SSH 데이터 (n=1,025)는 농촌진흥청의 신품종 적응성 실험보고서(1979 ―2019)로부터 수집하였다. 기상자료는 기상청으로부터 평균기온, 최저기온, 최고기온, 최대 강수량, 누적 강수량, 최대풍속, 평균풍속 및 일조시간을 10일 기준으로 계산하 여 수집하였다. 극단기상과 정상기상 간 구별을 위해 상 자 그림을 이용하여 탐색하였다. 극단기상과 정상기상 간 생산량 차이는 5% 유의수준 하에서 t-검정 및 ANOVA를 통해 확인하였다. 그 결과, 극단기상은 극단적으로 강한 바람을 동반한 봄 가뭄, 극단적으로 높은 강우량을 기록 하는 여름장마와 가을장마가 두드러졌다. 예측 생산량 피 해(kg/ha)는 각각 1,961―6,541, 2,161―4,526 및 508― 5,582로 나타났다. 본 연구는 우리나라의 SSH에 대한 취 약성 및 피해 산정에 도움이 되는 기초자료로서 극단기상 과 정상기상 사이의 생산량 차이를 확인하는 데 의의가 있다.

Currently, the development of evaluation technology for vibration and shock loads transmitted to spent nuclear fuel and structural integrity of spent nuclear fuel under normal conditions of transport is progressing in Korea by the present authors. Road transportation tests using surrogate spent nuclear fuel were performed in September, 2020 using a test model of KORAD-21 transportation cask and sea transportation tests were conducted from September 30 to October 4, 2021. Finally, the shake table tests and rolling test were conducted from October 31 to November 2, 2022. The shake table test was performed with the input load produced conservatively from the data obtained from the road and sea transportation tests. The test input was produced based on the power spectral densities and shock response spectrums from the transportation tests. In addition to the test inputs from the road and sea tests, sine sweep input and half sine input were used to verify the vibration characteristics of assemblies under boundary conditions during normal conditions of transport. Because the input load of the shake table test was produced conservatively, a slightly larger strain than the strain value measured in road and sea transportation tests was measured from the shake table tests. In the case of the sea test, it is considered that the process of enveloping the data in the 20 to 80 Hz range generated by the engine propeller system was performed excessively conservatively. As a result of analyzing the test results for the difference in boundary conditions, it was confirmed that the test conditions of loading the basket generated a relatively large strain compared to the conditions of loading the disk assembly for the same input load. Therefore, it is concluded that a transportation cask having a structure in which a basket and a disk are separated, such as KORAD-21, is more advantageous in terms of vibration shock load characteristics under normal conditions of transport than a transportation cask having an integral internal structure in which a basket and a disk are a single unit. However, this effect will be insignificant because the load itself transmitted to the disk assembly is very small.

Integrity evaluation scheme for Spent Fuel (SF) dry storage has been developed under transportation failure modes. This method especially considered the degradation characteristics of Spent Fuel (SF) during dry storage such as radial and circumferential hydride content, hydride volume fraction, oxide thickness, etc. Hydride and zircaloy cladding are considered as material composite system, using correlation models related to material properties. Critical Strain Energy Density (CSED) is compared with Strain Energy Density (SED), to evaluate cladding integrity. CSED serves as material characteristics, while SED can be considered as boundary condition. To calculate the CSED of cladding in the lateral failure mode, circumferential hydride concentration is used. SED is calculated considering both the bending moment and axial load. On the other hand, in the longitudinal failure case, fuel rod temperature, internal pressure, hoop stress, radial hydride concentration is used to calculate CSED. And pinch force (contact) was considered to evaluate SED. Model validations were conducted by comparing hot cell SF test and existing validated evaluation results. To separately handle normal transportation conditions from hypothetical accident conditions, SED according to stress-strain analysis results was separated into elastic and plastic regions. As a result of applying this scheme for 14×14 SF, failure probability of normal condition was zero, which is the similar result with DOE and same with EPRI. Regarding accident condition, lateral case showed similar result, but longitudinal case showed different but reasonable result, which was due to the different analysis conditions. The proposed methodology which was indigenously developed through this study is named as K-method.

Since the development of endoscopic ultrasound (EUS), the importance in the using EUS in pancreas and biliary tree has gradually increased. In order to properly observe the pancreas and biliary tree in EUS, it is necessary to learn landmark structures at each location, such as the stomach and duodenum, and to learn the positional relationship between them and normal structures. In this article, we will deal with the examination method at each position of the linear and radial EUS and reveal the normal EUS image.

This paper introduces a container loading problem and proposes a theoretical approach that efficiently solves it. The problem is to determine a proper weight of products loaded on a container that is delivered by third party logistics (3PL) providers. When the company pre-loads products into a container, typically one or two days in advance of its delivery date, various truck weights of 3PL providers and unpredictability of the randomness make it difficult for the company to meet the total weight regulation. Such a randomness is mainly due to physical difference of trucks, fuel level, and personalized equipment/belongings, etc. This paper provides a theoretical methodology that uses historical shipping data to deal with the randomness. The problem is formulated as a stochastic optimization where the truck randomness is reflected by a theoretical distribution. The data analytics solution of the problem is derived, which can be easily applied in practice. Experiments using practical data reveal that the suggested approach results in a significant cost reduction, compared to a simple average heuristic method. This study provides new aspects of the container loading problem and the efficient solving approach, which can be widely applied in diverse industries using 3PL providers.

Radioactive source terms are important factor in design, licensing and operation of SMR (Small Modular Reactor). In this study, regulatory requirements and evaluation methodology for normal operation on NuScale SMR, which received standard design certification approval on September 11, 2020 from US NRC, are reviewed. The radioactive waste management system of nuclear power reactor should be designed to limit radionuclide concentration in effluents and keep radioactive effluents at restricted area boundary ALARA according to 10 CFR 20 and 10 CFR 50 Appendix I. Also, in general, the coolant source term to calculate the off-site radiological consequences for normal operation of SMR should be determined by using models and parameters that are consistent with regulatory guide 1.112, NUREG- 0017 and the guidance provided in ANSI/ANS-18.1-1999, and the result should be corrected by reflecting the design characteristics of SMR. The coolant source term of NuScale, unlike the case of large NPPs, cannot rely solely on empirical source term data, because the NuScale source term is based on first principle physics, operational experience from recent industry, and lessons learned from large PWR operation. Fission products in reactor coolant are conservatively calculated using first principle physics in SCALE Code assuming 60 GWD/MTU. The release of fission products from fuel to primary coolant based on industry operational experience is determined as fuel failure fraction of 0.0066% for normal operation source term and 0.066% for design basis source term while coolant source term of large NPP is calculated by using ANSI/ANS-18.1 for normal operation and fuel failure fraction of 1% for design basis source term. Water activation products in reactor coolant are calculated from first principles physics and corrosion activation products are calculated by utilizing current large PWR operating data (ANSI/ANS 18.1- 1999) and adjusted to NuScale plant parameters. Also, because ANSI/ANS 18.1-1999 is not based on first principle physics models for CRUD generation, buildup, transport, plate-out, or solubility, NuScale has incorporated lessons learned by using ERPI’s primary water chemistry and steam generator guidelines to ensure source term is conservative and design of materials used cobalt reduction philosophy to help ensure the coolant source term are conservative. Based on the coolant source term calculated according to the above-described method, the annual releases of radioactive materials in gaseous and liquid effluents from NuScale reactor are evaluated. Currently, Small Modular Reactors such as ARA, SMART 100 are under review for licensing in Korea. This study will be helpful to understand how the reactor coolant system source terms are defined and evaluated for SMR.

In Korea, the NUREG-0017 methodology based on realistic model for reactor coolant concentrations are used to estimate the annual radioactive effluent releases for normal operation of nuclear power plant. The realistic model to estimate the radionuclide concentrations in reactor coolant is formulated as a standard, ANSI/ANS-18.1. This standard has provided a set of the reference radionuclide concentrations and adjustment factors for estimating the radioactivity in the principal fluid systems of target plant. Since ANSI/ANS-18.1 was first published in 1976, it was revised in 1984, 1999, 2016, and most recently in 2020. Therefore, this study analyzed revision history of assessment methodology of radioactive source term of light water reactors, which is ANSI/ANS-18.1. Assessment methodology of radioactive source term given ANSI/ANS-18.1 is by using radionuclide concentrations for reactor coolant and steam generator fluid of the reference plant and adjustment factors, which is modifying radioactive source term according to differences in design parameters between reference plant and target plant. There are three type of reference plant: PWR with u-tube steam generator, PWR with once-through steam generator, and BWR. This study analyzed for PWR with u-tube steam generator. Although the standard was revised, evaluation methodology and formula of adjustment factor have been retained, but some of items have been revised. First revision item is reduction of the number of radionuclides and decrease of radioactive concentration in reactor coolant. In the 1976 version of the standard, there were 71 target radionuclides, but the target nuclides have reduced to 57 in 1984 and 56 after 1999. In the case of radioactive concentration in reactor coolant, as the version of standard was updated, the radioactive concentration of 18 nuclides in 1984, 14 nuclides in 1999, and 25 radionuclides in 2016 was decreased. Most of the radionuclides with decrease radioactivity concentration were fission product, it is resulted from improvement of nuclear fuel performance. Second revision item is change of adjustment factors. After the revision in 2016, the adjustment factors for zinc addition plants using natural or depleted zinc are changed. This study analyzed revision history of evaluation methodology of radioactive source term of light water reactors. Furthermore, result of this study will be contributed to the improvement of understanding of assessment methodology and revision history for the radioactive source term.

This paper intends to present considerations on the question of what is the “load standard” or “design load” for integrity evaluation under normal transportation conditions and what type of design load is good for users. This suggests a direction for subsequent research on producing design loads that transport business companies can utilize without difficulty. Several studies have been conducted to evaluate the integrity of spent nuclear fuel during normal transportation. A representative study recently conducted is the Multi-modal Transportation Test (MMTT) conducted using a commercial spent nuclear fuel cask by US DOE in 2017. In Korea, additional transport tests were planned to acquire sufficient test data under the conditions of road and sea transport considering the Korean situation. As a result, road transport tests were carried out in 2020 and sea transport tests were carried out in 2021. In the road transport test, a driving test that simulates various road conditions and a test that cycled a 4.5 km road eight times were performed. In most cases, the maximum acceleration of less than 1 g occurred, and the maximum strain was less than 48 με. For the sea transport test, the magnitude of both the maximum acceleration and the maximum strain were lower than those in the road transport test. We concluded tentatively that the integrity of spent fuel under normal conditions of transport was satisfactory with a large margin. However, when the storage business is realized and the transport of spent fuel becomes visible, the storage and transport business companies will have to prove the maintenance of the integrity of the spent fuel under normal transport conditions at the request of the regulatory agency. The transport business companies can transport the spent nuclear fuel by using different types of transport casks and different types of trucks and ships from those used in the tests mentioned above. However, it is absurd to have to prove the integrity of spent nuclear fuel by performing expensive tests again. Therefore, in this study, the design load that can be used by transport business companies is to be presented. The design load to be presented should satisfy the following requirements. The design load should be applicable including some differences in the transport cask or transport system, or different design loads should be presented according to the differences. The location where this design load is applied is to be specified (e.g. fuel rod, basket, internal structure). Requirements according to the operating speed of the transport system should be presented together. The type of design load is to be presented (e.g. PSD, SRS, FDS etc.). Other types of standards may be presented. For example, a speed limit for a vehicle carrying spent nuclear fuel may be suggested, or a speed limit for a vehicle passing through a speed bump may be suggested. In order to present such a reliable design load, a multi-axis vibration excitation shaker table test will be carried out. Though this shaker table test, the behavior of the nuclear fuel assembly is closely evaluated by applying the data obtained from the road and sea transport tests previously performed as an input load. In addition, FDS (Fatigue Damage Spectrum) will be produced and applied to experimentally evaluate the durability of fuel assemblies under normal transport conditions.

Currently, the development of evaluation technology for vibration and shock loads transmitted to spent nuclear fuel and structural integrity of spent nuclear fuel under normal conditions of transport is progressing in Korea by the present authors. Road transportation tests using surrogate spent nuclear fuel were performed in September, 2020 using a test model of KORAD-21 transportation cask and sea transportation tests were conducted from September 30 to October 4, 2021. In order to investigate amplification or attenuation characteristics, according to the load transfer path, a number of accelerometers were attached on a ship cargo hold, cradle, cask, canister, disk assembly, basket, and surrogate fuel assemblies and to investigate the durability of spent nuclear fuel rods, strain gages were attached on surrogate fuel assemblies. A ship named “JW STELLA” which has similar deadweight (5,000 ton) of existing spent nuclear fuel transportation ships was used for the sea transportation tests. The ship is propelled by 1,825 hp two main engines with two 4-bladed propellers. There are two major vibration sources in the ship. One is the vibration from waves and the other is the vibration from the engine and propeller system. The sensor locations on the ship were determined considering the vibration sources. The sea transportation test was performed for 5 days, the test data were measured successfully. The ship with the test model was departed from Changwon and sailed to Uljin, sailed west to Yeonggwang and then returned to Changwon. In addition to sailing on a designated test route, circulation test, braking/acceleration test, depth of water test, and rolling test were conducted. As a result of the preliminary data analysis of the sea test, power spectral densities and shock response spectrums were obtained according to the different test conditions. The vibratory loads caused by the wave mainly occurred in the frequency range of 0.1 to 0.3 Hz. The vibratory loads caused by the propeller occurred near the n/rev rotating frequencies, such as 5, 10, 20 Hz etc. However, those frequencies are far from the natural frequencies of local mode of the fuel rods, so it is considered that the vibratory loads from the wave and the propeller do not have a significant influence on the structural integrity of the fuel rods. Among all the test cases, maximum strain occurred at SG31 near the bottom nozzle on the test; the magnitude was 73.62 micro strain. Based on the analyzed road and sea transportation test data, a few input spectra for the shaker table test will be obtained and the shaker table test will be conducted in 2022. It is expected that the detailed vibration characteristics of the assembly which were difficult to identify from the test results can be investigated.

The purpose of this study is to develop the analysis procedures for the evaluation of the structural integrity of the spent fuel in normal condition of transport at sea. Spent nuclear fuel must be transported from the wet storage facility in the nuclear power plant to the intermediate storage facility, and the structural integrity must be maintained in vibration and shock loads during the transportation. In general, the transport of spent nuclear fuel is performed in three kinds of modes: road, rail, and sea. During transport, the spent nuclear fuel is subjected to repeated vibration and shock loads by road surfaces, railroad tracks, and waves of the sea. It should be evaluated whether the structural integrity of the spent fuel is maintained under these load conditions. All nuclear power plants in Korea are located in coastal sites, and the interim storage facility for spent nuclear fuel is highly likely to be decided as a coastal site as well. Therefore, the main mode of the spent nuclear fuel transport is expected to be maritime transport by ships. In this study, the analysis procedure was developed to evaluate the safety of spent fuel at maritime transport by ships, and the procedure for evaluating the integrity of spent fuel under normal conditions of maritime transport were proposed. CFD analysis using SeaFEM was performed for the vibration analysis of the ship by waves, and the structural vibration analysis of the transport system was simulated using the developed in-house codes. The fatigue durability of the cladding was also evaluated using the developed fatigue analysis program and the fatigue analysis used the strain data obtained from the structural analysis. It was concluded that the value of the fatigue damage on the spent fuel cladding during normal conditions of maritime transportation is close to “0” and the structural integrity of the spent fuel is maintained in the same condition.

Because sows are industrially vital for swine production, monitoring for their health or disorder status is important to ensure high reproductive performance. Especially, ambient temperature changes in different season, especially during summer, are directly influenced to the reproductive performance of sows. Although the serum biochemical parameters are widely applied in the veterinary medicine with wide ranges for the physiological process, the values are also influenced by several factors such as age, breed, gender, and stress. In addition, domestic sows in Koreaspecific reference interval (RI) for serum biochemistry has not been established yet. Therefore, the present study was aimed to evaluate seasonal variation of RIs in the serum biochemistry in domestic sows in Korea at different seasons and to establish normal RIs using a RI finding program (Reference Value Advisor). Significant difference (p < 0.05) on the different seasons were identified in several serum biochemical parameters including BUN, CRE, GGT, GLU, ALB, TP, LDH and Na in sows. Therefore, we further established RIs, specific in domestic sows in Korea regardless of season. The established RIs based on the serum biochemical values provide a baseline for interpreting biochemical results in the domestic sows in Korea, regardless of seasonal effect. It may contribute to develop a strategy for better reproductive performance by improving breeding management practice and evaluating health of pig herds, which facilitate to avert the economic loss in summer infertility in sows.