The ductility of the system based on the capacity of each structural member constituting the seismic force-resisting system is a significant factor determining the structure’s seismic performance. This study aims to provide a procedure to supplement the current seismic design criteria to secure the system’s ductility and improve the seismic performance of the steel ordinary moment frames. For the study, a nonlinear analysis was performed on the 9- and 15-story model buildings, and the formation of collapse mechanisms and damage distribution for dynamic loads were analyzed. As a result of analyzing the nonlinear response and damage distribution of the steel ordinary moment frame, local collapse due to the concentration of structural damage was observed in the case where the influence of the higher mode was dominant. In this study, a procedure to improve the seismic performance and avoid inferior dynamic response was proposed by limiting the strength ratio of the column. The proposed procedure effectively improved the seismic performance of steel ordinary moment frames by reducing the probability of local collapse.

압축센스(Compressed SENSE) 기법은 검사 시간을 획기적으로 단축할 수 있으나, 시간 단축을 위한 기법적용 시 가속계수를 증가시키면 인공물의 발생이 영상에서 증가하는 문제점이 있다. 이에 인공물이 발생하지 않으면서 검사 시간을 최대한 단축할 수 있는 최적의 압축센스 가속계수를 제시하고자 하였다. 연구 방법은 인공물이 발생하지 않는 가속계수 1.0을 기준으로 0.5 간격씩 5.0까지 무릎관절 자기공명영상의 팬텀 실험과 임상실험 영상을 획득한 후, 방사선사 10명이 5점 척도로 영상을 평가하여 유의한 차이가 있는지 판단하였다. 연구 결과 T1 강조영상과 T2 강조 영상 모두 팬텀 실험은 가속계수 2.0 이하로 하였을 때 임상실험은 3.0 이하로 하였을 때 기준이 되는 1.0 영상과 차이가 없었다. 결론적으로 무릎관절 자기공명영상 검사 시 인공물이 발생하지 않으며 검사 시간을 최대로 단축할 수 있는 최적의 압축센스 가속계수는 팬텀 실험의 경우 2.0, 임상실험의 경우 3.0이 적정하리라 판단된다.

The public safety management guidelines were introduced in 2019 and continue to be dedicated to advancing comprehensive measures for public safety management, with a primary focus on prioritizing the safety and well-being of the public within governmental institutions. To achieve this goal, our previous study developed a establishment procedure of risk assessment-based safety inspection system for public institutions that order construction projects and applied it to highway construction projects to evaluate its effectiveness. To enhance the practicality of the establishment procedure, it is essential to collect and analyze feedback from stakeholders regarding its performance and suitability. This study conducted a survey involving 200 participants who had experience with the establishment procedure, and performed statistical analyses to evaluate its performance and applicability. The survey results indicated that the participants reported a high level of satisfaction (scoring 4 and above on a 5-point Likert scale) in several areas: specialization of safety inspection items for different types of work (with a satisfaction rate of 65%), the evaluation process for safety ratings (64.5%), and their willingness to recommend the procedure to other institutions (75.5%). In the factor analysis with Varimax rotation, two factors emerged: (1) a specialization factor related to safety inspection items, and (2) a grading factor associated with safety evaluation results. Regression analyses of these factors unveiled significant positive relationships with improvements in safety and health performance, including the prevention of fatal accidents, heightened safety responsibility, and raised safety inspection standards. The establishment procedure of safety inspection system developed in our previous study can play a crucial role in reducing accidents resulting in fatalities and injuries at construction sites, ultimately contributing to a safer working environment for all involved parties.

The purpose of this report is to provide a summary of the Phase 1 Final Status Survey (FSS) Final Report results and overall conclusions which conduct that the Zion Nuclear Power Station (ZNPS) facility and site meets the 25 mrem(0.25 mSv)per year release criterion as established in Nuclear Regulatory Commission Regulation (NRC) 10 CFR 20.1402 “Radiological Criteria for Unrestricted Use”. The FSS results provided assessment and summarize that any residual radioactivity results in a Total Effective Dose Equivalent (TEDE) to an Average Member of the Critical Group (AMCG) that does not exceed 25 mrem per year, and the residual radioactivity has been reduced to levels that are as low as reasonably achievable (ALARA). The release criterion is translated into site-specific Derived Concentration Guideline Levels (DCGLs) for assessment and summary. ZionSolutions, a decommissioning service provider, estimates that a total of four (4) FSS Final Reports be generated and submitted to the NRC during the decommissioning project. ZionSolutions established the Characterization/License Termination (C/LT) Group, within the Radiation Protection division, with sufficient management and technical resources to fulfill project objectives. The C/LT Group is responsible for the safe completion of all surveys related to characterization and final site closure. Approved site procedures and detailed Technical Support Documents (TSD) direct the FSS process to ensure consistent implementation and adherence to applicable requirements. The development and planning phase was initiated in 1999 by the “Zion Station Historical Site Assessment” (HSA) and the initiation of the characterization process for FSS. Develop the information necessary to support FSS design, including the development of Data Quality Objectives (DQOs) and survey instrument performance standards. DQOs are qualitative and quantitative statements derived from the DQOs process that clarify technical and quality objectives. The next step, FSS design utilizes the combination of traditional scanning surveys, systematic sampling protocols and investigative/judgmental methodologies to evaluate survey units relative to the applicable release criteria for open land sample plans. To aid in the development of an initial suite of potential radionuclides of concern for the decommissioning of ZNPS, the analytical results of representative characterization samples collected at the site were reviewed. At this FSS design step, the Radionuclides of Concern (ROC) are determined. As Co-60 and Cs-137 account for 99.5% of the analysis results of concrete core sampling data form ZNPS’s Containment Building and Auxiliary Building, they are determined and used as the basic ROC in the survey design. Additionally, site information is described and Historical Site Assessment (HSA) is performed. Data collected for the initial HSA will be used to establish the initial regional survey unit and corresponding MARSSIM classification. Next, an assessment of the collected data is performed using the DQO process, and a survey methodology is established by selecting a sampling method and measuring instrumentation. These result judgments provide guidance for C/LT Engineer to interpret findings using the Data Quality Assessment (DQA) process, which analysis Recorded data, Missing values, Deviation from established procedure, and Analysis flags. In conclusion, FSS is the process used to demonstrate that the ZNPS facility and site comply the radiological criteria for unrestricted use specified in 10 CFR.20. The purpose of FSS Sample Plan is to describe the methods to be used in planning, designing, conducting, and evaluating the FSS.

The KINAC resident inspectors are responsible for conducting on-site regulatory and intergovernmental support tasks related to safeguards, physical protection, and cybersecurity in each NSSC regional office. In nuclear material accounting and control, resident inspectors primarily perform tasks such as national inspections and technical support for IAEA inspections. However, with the increasing cases of non-compliance with the advance notification procedure by operators, there is a growing need for improvement in the role of resident inspectors in on-site regulation. In response to this situation, the safeguards division in KINAC has analyzed and improved the on-site check procedures of resident inspectors at LWR facilities. The existing procedure outlines the process where resident inspectors receive the advance notification documents submitted by operators and utilize them as a reference for conducting weekly checks during the overhaul period when IAEA surveillance cameras are installed. Additionally, according to the attached forms specified in the procedure, resident inspectors are required to submit the check results report to the director of the safeguards division in KINAC every week and to the NSSC every month. The inspection items include checking the execution and changes of advance notification, verifying unnotified matters, discussing other issues, assessing the integrity of things such as the operational status of IAEA surveillance equipment, and so on. On April 13-14, 2023, the Safeguards division organized a two-day resident inspector’s work-sharing workshop to discuss improvements in the on-site check procedures of resident inspectors at LWR facilities. During the workshop, a comparison and analysis were conducted between the existing procedures and actual on-site activities. Unnecessary tasks such as advance notification document reception and monthly reporting were eliminated, and the focus was shifted towards emphasizing essential tasks. The opinions of resident inspectors were taken into account to derive directions for improvement. The existing procedure was applicable only during Overhaul periods for resident inspectors. It has been improved by removing this limitation, allowing its use during routine times. Furthermore, the procedure has been enhanced by clarifying its purpose, scope, users, and definitions of terms and specifying responsibilities and authorities. Unnecessary terminology has been eliminated. Remarkably, the definition of advance notification has been detailed, and the reporting of check results has been simplified through weekly task reporting. The Safeguards division in KINAC has strived to enhance the efficiency and simplification of on-site regulatory activities for resident inspectors at LWR facilities by improving their on-site check procedures. These improvement activities are expected to aid resident inspectors in effectively performing a wide range of tasks, including safeguards, physical protection, cybersecurity, and government support. In the future, it will be possible to continue refining the on-site check procedures by sharing the results of using the procedure in meetings and gathering various opinions from resident inspectors.

This study investigated the impact of 26 EFL college students’ familiarity levels with the visual information of TOEIC listening test items on test scores and test performance procedure. Data collected measured students’ degree of familiarity with the visuals via a pre-test and their listening test scores, and responses to the post-test questionnaire were analyzed. To analyze the data, one-way ANOVA was conducted to locate the interdependency between the students’ familiarity levels with visuals and their test scores. Results found that there was no statistically significant differences in test scores regardless of the students’ levels of familiarity with the visuals. Additionally, the correlations between the familiarity with the visuals and the students’ test scores were low. However, the students experienced difficulties with unfamiliar visuals while taking the test. These findings indicate that, although test-takers’ familiarity with visuals does not predict their listening test scores, there is a need to tailor the visuals when developing listening test-items.

According to the “Law on protection and response measures for nuclear facilities and radiation”, Nuclear Power Plant (NPP) licensees should conduct periodic exercises based on hypothetical cyberattack scenarios, and there is a need to select significant and probable ones in a systematic manner. Since cyber-attacks are carried out intentionally, it is difficult to statistically specify the sequences, and it is not easy to systematically establish exercise scenarios because existing engineering safety facilities can be forcibly disabled. To deal with the above situation, this paper suggests a procedure using the Probabilistic Safety Assessment (PSA) model to develop a cybersecurity exercise scenario. The process for creating cyber security exercise scenarios consists of (i) selecting cyber-attack-causing initiating events, (ii) identifying digital systems, (iii) assigning cyber-attack vectors to a digital system, (iv) determining and adding type for operator’s response, (v) modifying a baseline PSA model, and (vi) extracting top-ranked minimal cut sets, and (vii) selecting a representative scenario. This procedure is described in detail through a case study, an expected cyber-attack scenario General Transient-Anticipated Transient Without Scram (GTRN-ATWS). It refers to an accident scenario for ATWS induced by GTRN. Since ATWS is targeted for cyber training in some NPPs, and GTRN is one of the most common accidents occurring in NPPs, GTRN-ATWS was chosen as an example. As for the cyber-attack vector, portable media and mobile devices were selected as examples based on expert judgment. In this paper, only brief examples of GTRN-ATWS events have been presented, but future studies will be conducted on an analysis of all initiating events in which cyber-attacks can occur.

이 연구는 국가 위기관리체계 중 통합방위사태와 경계태세의 발령기준 을 개선하기 위해 분석하였다. 우리 정부와 군은 평시 북한의 침투 및 도발과 같은 상황에 대응하기 위해 경계태세 및 통합방위사태와 같은 국 가 위기관리체계를 지속 발전시켜 왔다. 그러나 북한의 도발 상황을 고 려하여 통합방위사태와 경계태세가 시기적절하게 선포 및 발령되어야 함 에도 불구하고 그렇지 못한 사례가 여럿 있다. 따라서 통합방위사태 선 포절차와 경계태세의 발령조건을 반드시 개선할 필요성이 있다. 이를 위 해 기존 선행연구 사례를 통해 공통적인 연구의 경향들을 분석하였고, 위와 같은 문제가 발생하는 원인을 관련 법령 분석을 통해 연구문제를 도출하였다. 또 과거 위기대응사례를 분석하여 연구문제에 대한 결론을 내림과 동시에 발전 방향을 제시하였다. 이를 통해 본 논제가 주는 함의 는 첫째, 북한의 태도가 급변하여 예측할 수 없는 군사적 도발을 하더라 도 유연하게 대응하고 신속하게 대처할 수 있는 위기관리태세를 갖추어 야 한다는 것이며 둘째, 이러한 적시적인 위기관리를 통해 정부는 우리 국민들의 안전을 수호하는 역할을 수행해야 한다는 점을 시사하고 있다.

Chronic fibroinflammatory pancreatitis causes irreparable damage to the pancreatic parenchyma. This frequently results in food restrictions, painkiller addiction, and serious quality of life impairment in children. We observed a 13-year-old girl who had previously been hospitalized multiple times and had undergone acute pancreatitis arrived with excruciating abdominal pain and recurrent hematemesis. A chronic intra-pseudocyst bleed and an ampulla of Vater hemorrhage were discovered during an upper gastrointestinal endoscopy. A 3×2×1 cm pancreatic head pseudocyst was discovered on the magnetic resonance cholangiopancreatography, however the computed tomography scan revealed a pancreatic head pseudocyst, pancreatic duct stones, and substantial peripancreatic inflammation. The multidisciplinary team determined that Frey’s surgery was the most efficient method to minimize her suffering. No complications occurred during the healing phase following surgery, and two years later, neither recurrence bleeding nor abdominal pain appeared. In summary, Frey’s method is a secure and efficient intervention when applied by a multidisciplinary team.

Kr-85 has a half-life of 10.7 years and it stays in the atmosphere for a long time. However it does not accumulate as an noble gas but only emits beta particles. Therefore its contribution to environmental radiation dose is lower than any other radionuclides. Kr-85 is one of the main fission products produced by nuclear fission reaction and artificial radionuclide that does not exist in nature. For these reasons, monitoring Kr-85 from the atmosphere is meaningful so that the nuclear-related facilities are recommended to control and regulate environmental emissions. Post Irradiation Examination Facility (PIEF) which located in KAERI is a facility that conducts various material and chemical experiments using the irradiated nuclear fuels. Therefore, various radionuclides can present in gaseous effluent including Kr-85. To prevent the environmental hazards and guarantee the radiation safety of the public, nuclear facilities are recommended to be equipped with stack radiation/radioactivity monitoring system, so that the Kr-85 concentration in gaseous effluent is controlled within the regulatory criteria. Particularly, the Kr-85 concentration of gaseous effluent is commonly monitored by the stack monitoring system connected to the process ventilation system from the hot cell. The monitoring system supply the information such as beta count rate, dose rate and flow rate, etc. Due to the concentration of Kr-85 in gaseous effluent is subject to regulatory guide lines, a systemized procedure for calculating Kr-85 concentration of the stack exhaust is necessary. Furthermore, the emission should be monitored whether it satisfies the regulatory standard or does not. This paper performed discussion on the process of calculating the concentration of Kr-85 in the gaseous effluent of PIEF stack from the monitoring system (NGM209, MGP), and the amount of Kr- 85 over the last 2 years emissions was calculated. In addition to calculating effluent rate of radioactive Kr-85, the Minimum Detectable Concentration (MDC) and Decision Threshold (SD) were calculated. As a result, the calculated Kr-85 concentration was below the SD during the entire period. It is considered that there are no environmental emissions of Kr-85.

In 2017, Kori unit 1 nuclear power plant was permanently shut down at the end of its life. Currently, Historical Site Assessment (HSA) for MARSSIM characteristics evaluation is being conducted according to the NUREG-1575 procedure, this is conducted through comprehensive details such as radiological characteristics preliminary investigation and on-site interview. Thus, the decommissioning of nuclear power plant must consider safety and economic feasibility of structures and sites. For this purpose, the establishment of optimal work plan is required which simulations in various fields. This study aims to establish procedure that can form a basis for a rational decommissioning plan using the virtual nuclear power plant model. The mapping procedure for 3D platform implementation consisted of three steps. First, scan the inside and outside of the nuclear power plant for decommissioning structure analysis, 3D modeling is performed based on the data. After that, a platform is designed to directly measure the radiation dose rate and mapped the derived to the program. Finally, mapping the radiation dose rate for each point in 3D using the radiation dose rate calculation factor according to the time change the measured value created on the 3D mapping platform. When the mapping is completed, it is possible to manage the exposure dose of workers according to the ALARA principle through the charge of radiation dose rate over time because of visualization of the color difference to the radiation dose rate at each point. For addition, the exposure dose evaluation considering the movement route and economic feasibility can be considered using developed program. As the interest in safety accidents for workers increases, the importance of minimum radiation dose and optimal work plan for workers is becoming increasingly important. Through this mapping procedure, it will be possible to contribute to the establishment of reasonable process for dismantling nuclear power plant in the future.

Despite the increasing interest in Deep Borehole Disposal (DBD) for its capability of minimizing disposal area, detailed research about DBD operation system design should be conducted before the DBD can be implemented. Recently, DBD operation system applying wireline emplacement (WE) technique is under study due to its high flexibility and capability of minimizing surface equipment. In this study, a conceptual WE system, and operation procdure is introduced. The conceptual WE system consists of 3 main stations, which from the top are hoisting station (HS), canister connection station (CCS) and basement (BS). In HS, WE is controlled and monitored. The WE is controlled using wireline drum winch and sheaves, and load on wireline is measured using a load cell. HS also has a pressure control system (PCS), which monitors internal pressure of the system, and a lubricator, which act as housing for joint device, allowing the joint device to be easily inserted into the borehole. The joint device is used to connect the disposal canister to wireline for emplacement/retrieval. In CCS, a rail transporter brings a transport cask containing disposal canisters, then the transport cask is connected to the hoisting system and a PCS in the BS. The main component located at canister station are a sliding shielding door (SSD), and a slip. The SSD is used to prevent canister from falling into borehole during the connecting operation and prevent radiation from BS to affect the workers. The slip is located beneath the SSD and is used to hold the disposal canister before it is lowered into the borehole. In BS, PCS is installed to prevent overflow and blowout of borehole fluid. The PCS consists of wireline pressure valve, christmas tree and BOP, which all are a type of pressure valve to seal the borehole and release pressure inside the borehole. The WE procedure starts with transporting transport cask to CCS. The transport cask is connected to lubricator, and PCS. Joint device is lowered down to be connected with disposal canisters, then pulled up to check the load on the wireline. After the check-up, SSD is opened, and disposal canister is lowered into the borehole. When desired depth is reached, joint device is disconnected and retrieved for next emplacement. In this study, the conceptual deep borehole disposal system design implementing WE technique is introduced. Based on this study, further detailed design could be derived in future, and feasibility could be tested.

양식장 환경관리를 위하여 여러 나라에서 양식장 환경관리시스템을 구축하고 있으며, 어장환경평가가 그중 하나이다. 2013년에 개정된 국내 어장관리법에는 어류가두리양식장 면허를 연장할 때 어장환경평가를 받도록 명시되어 있다. 어장환경평가는 과학적 분석을 통한 평가를 바탕으로 양식장 환경 개선 조치를 실행하여, 지속 가능한 어업생산 조성, 어장 생산력 증가, 어업인의 소득 증가를 목적으로 한다. 어장환경평가의 평가지표는 저서다모류 군집을 기초로 한 생물지수인 저서동물지수(BHI)와 총유기탄소량(TOC)이며, 두 항목을 점 수화하여 단일등급으로 산출하여 평가에 사용한다. 본 연구는 어장환경평가지표인 BHI와 TOC의 선정 과정과 산출된 평가 등급의 생태학 적 의미를 설명한다.

In Korea, research on the introduction of dry storage facility is being conducted as an alternative to saturation of temporary storage facilities for spent nuclear fuel. The introduction of dry storage facilities requires a radiological impact assessment on the workers of the facility, and for this, an appropriate exposure scenario must be derived through work procedure analysis. In this study, the procedure for storing spent nuclear fuel in dry storage facilities was analyzed based on the case of evaluating the radiological impact of workers in dry storage facilities abroad. We investigated cases of radiological impact assessment on workers at on-site dry storage facilities by PNNL, Dominion, and P. F. Weck. PNNL and Dominion analyzed the storage work procedure of the VSC (Vertical Storage Cask) method using CASTOR V/21, TN-32, respectively, and conducted a radiological impact assessment. P. F. Weck analyzed the storage work procedure of various spent nuclear fuel casks for VSC and HSM (Horizontal Storage Module), conducted a radiological impact assessment. As a result of comparing the procedure for storing spent nuclear fuel by case, it was found that the storage procedure was determined by the storage method and the cask type. In the case of VSC method, canister-type casks and basket-type casks are used, and the storage procedure are partially different according to each. Canister-type cask requires repackaging from transfer overpack to storage overpack, but basket-type cask doesn’t require that procedure. In the case of the HSM method, only the canister type cask was found to be used. However, the storage procedure was different depending on the type of HSM system. Depending on the type of HSM system, the necessity of cask for on-site transport was different. In this study, we investigated and analyzed the work procedure according to the storage method of dry storage facilities abroad. It was found that the dry storage procedure of spent nuclear fuel different according to the storage method and type of cask. The results of this study can be used as basic when deriving the exposure scenario for spent nuclear fuel dry storage workers suitable for the domestic situation.

The use of dampers is being considered a means to improve the seismic performance of buildings. It may take considerable time and effort to find an optimal design solution since repeated three-dimensional nonlinear time history analyses are required. Therefore, a preliminary design procedure for seismic retrofit using hysteretic dampers was proposed in this study. In the proposed procedure, the amount of retrofit (required number of dampers) is estimated from the capacity curve of the building before retrofit and allowable story drift of the building. In combining the capacity curves of the building and the dampers, the deformation demand for the dampers can be easily checked against their deformation capacity. The equations to transform the device displacement to roof displacement for the combination of capacity curves are developed. The proposed procedure was applied to the seismic retrofit design of sample buildings. The study found that the estimated capacity curve was very close to the actual capacity curve obtained from the pushover analysis, which can determine an appropriate configuration to meet the required seismic performance.

Based on the nonlinear static analysis and the approximate seismic evaluation method adopted in “Guidelines for seismic performance evaluation for existing buildings, two methods to calculate strength demand for retrofitting individual structural walls in unreinforced masonry buildings are proposed.” The displacement coefficient method to determine displacement demand from nonlinear static analysis results is used for the inverse calculation of overall strength demand required to reduce the displacement demand to a target value meeting the performance objective of the unreinforced masonry building to retrofit. A preliminary seismic evaluation method to screen out vulnerable buildings, of which detailed evaluation is necessary, is utilized to calculate overall strength demand without structural analysis based on the difference between the seismic demand and capacity. A system modification factor is introduced to the preliminary seismic evaluation method to reduce the strength demand considering inelastic deformation. The overall strength demand is distributed to the structural walls to retrofit based on the wall stiffness, including the remaining walls or otherwise. Four detached residential houses are modeled and analyzed using the nonlinear static and preliminary evaluation procedures to examine the proposed method.

In this study, factors such as improvement of a fishing process and safety, reduction of the labor force and headcount and development of the automation technology for offshore (eel and crab) pot fishing vessels were analyzed. A questionnaire survey was conducted to analyze and select the key factors using independent/paired sample t-test and correlation analysis, and a living lab was operated with ship owners, skippers and experts to discuss practical needs of the site. From the result of questionnaire survey and field requirements, it was possible to understand the level of awareness of ship safety, general safety equipment, fishing work process and fishing safety equipment from the point of view of the field. In addition, there were differences in the measurement results of each items because the working environment and experience were different according to the position of the ship owner and the skipper. The results of the questionnaire survey and various perceptions of field stakeholders were reflected when analyzing the fishing system and fishing process to choose the development equipment applicable to the field. From the analysis results, the selected development equipment based on the fishing equipment and process currently in operation are pot washing device, catch separation and fish hold injection device, length limit regulations and bait ejection device after use, automatic main line winding device, bait crusher, automatic (crab) pot hauling separator and so on.