본 연구의 목적은 디지털 트랜스포메이션의 도입이 가속화 된 경영 환경 속에서 조직 내 인적자원개발(human resource development: HRD) 담당자들이 경험하는 업무 변화에 대해 알아보고, 그에 따른 필요역량 및 업무 수행 시 겪는 어려움을 탐색하는데 있다. 이러한 목적을 달성하기 위해 HRD 업무를 담당하는 현직 종사자 10명을 대상으로 하여 일대일 심층 인터뷰를 실시하였다. 그 결과 디지털 트랜스포메이션시대에 HRD 담당자의 업무는 NCS 기반의 능력단위를 기준으로 하여 전통적인 업무와 비교했을 때 강화, 감소 또는 새롭게 추가된 업무가 있는 것으로 나타났다. 또한 필요 역량은 지식, 기술, 태도, 능력이라는 범주로 나누어 도출하였는데, 테크놀로지 활용 기술 및 변화에 적극적이고 긍정적으로 대응하려는 태도 및 능력이 중요한 것으로 나타났다. 전체적으로 디지털 트렌드 및 스킬에 대한 지속적인 학습으로 인해 업무량은 증가한 것으로 나타났다. 이상의 결과를 바탕으로 본 연구에서는 조직원들의 변화를 이끌어가야 하는 HRD 담당자들에게 실천적 시사점을 제시하고, 연구의 제한점 및 향후 연구 방향에 대하여 제언하였다.

본 연구는 2022년 2월 발발한 러시아-우크라이나 전쟁과 분쟁에 대응 하는 국제기구의 역할을 살펴보는 것을 목적으로 한다. 구체적으로 본 연구는 분쟁과 국제기구 간 관계에 대한 기존 논의를 확인하고, 우크라 이나 전쟁이 촉발한 국제기구의 최근 활동에 대해 살펴본다. 특히 UN 총회는 안전보장 이사회의 상임이사국 거부권 행사에 대해 유엔 총회의 견제 역할을 강조하는 4월 26일 결의안을 통과시켰고, 이는 제한적이지 만 기존의 유엔 개혁에 대한 진일보한 활동으로 평가된다. 우크라이나 전쟁 직후 안전보장이사회의 상임이사국 중 하나인 러시아의 지속적인 반대로 유엔 차원에서의 강력한 우크라이나 지원 및 러시아 규탄이 나타 나지 않았다. 이에 대한 대응으로 유엔 총회는 안전보장이사회의 상임이 사국 거부권 행사 행위를 총회에서 설명해야하는 결의안을 채택하였다. 이를 통해 본 연구는 국제기구가 분쟁발발을 억지하거나 분쟁을 종식시 키는 적극적 역할을 수행하기 어렵지만 국제분쟁이 국제기구의 기존 개 혁논의를 촉진시켜 분쟁에 대응하는 국제기구의 역할을 다각화하는 요인 이라는 점을 제시한다.

클라우제비츠는 『전쟁론』에서 전쟁은 개별 상황마다 그 본질을 약간씩 변화시키기 때문에 카멜레온과도 같다고 보았다. 클라우제비츠의 이러한 주장은 모든 것이 불확실한 전쟁 상황에서 다양한 요소들이 상호작용을 하여 승패에 영향을 미칠 수 있음을 언급한 것이다. 이 글에서는 불확실성 의 요인이 되는 우연과 마찰, 정보, 위험, 육체적 고통을 토대로 이를 해 소하기 위한 5가지 방안을 도출하여 제1연평해전과 연평도 포격전 사례를 분석하였다. 분석 결과 첫째, 실전적 교육훈련, 둘째, 과감한 권한 위임, 셋째, 부대 운용의 융통성, 넷째, 간단없는 작전지속 지원, 다섯째, 피아 전술 전기 숙달 등을 통해 불확실성을 해소하고 전승을 달성할 수 있었음 을 확인할 수 있었다. 이것이 한국군에 주는 함의로는 첫째, 창의적인 사 고력과 탄력적인 부대지휘 능력을 배양하기 위한 교육훈련의 필요성, 둘 째, 전술교리 적용에 융통성 없이 집착하는 문화 지양, 셋째, 군인으로서 ‘인생의 첫 전투’를 어떻게 수행할 것인가에 대한 준비가 필요하다. 비록 군사과학 기술이 발전하여 전장 상황을 가시화할 수 있게 되더라도 전쟁 의 근본적인 본질인 불확실성은 전쟁 수행에 커다란 영향을 미치는 변수 라는 것을 망각해서는 안 될 것이다.

As part of the third ATM Challenge, we performed a series of atmospheric dispersion simulations for routine releases of Xe-133 from ordinarily operating nuclear facilities such as Medical Isotope Production Facilities (MIPFs), Nuclear Power Plants (NPPs), and Research Reactors (RRs) in the Northern Hemisphere using our ATM, Lagrangian Atmospheric Dose Assessment System (LADAS), with Numerical Weather Prediction (NWP) data produced by the Korea Meteorological Administration (KMA). The simulation time period is 6 months, from June to November in 2014, and we used the stack emission data except for CNL (Canada) and IRE (Belgium) in accordance with the scenario of the third ATM Challenge 2019. In addition, the simulations were done individually for all MIPFs, NPPs, and RRs. We utilized 3-hourly KMA’s Unified Model Global Data Assimilation and Prediction System (UM-GDAPS) data with 0.35°×0.23° horizontal resolution as input meteorological fields and extracted hourly time series results for Xe-133 activity concentrations with few different resolutions such as 0.5°×0.5°, 0.35°×0.23°, and 0.1°×0.1° at several IMS stations in the Northern Hemisphere which were in normal operation in 2014. Considering previously reported values of daily Xe-133 release amounts for CNL and IRE, measured signals at some IMS stations (such as CAX17, DEX33, SEX63, and USX75) were well reproduced from the simulation results.

Kori unit 1 was permanently shut down in 2007 and is currently awaiting approval for decommissioning and dismantling (D&D). The wastes generated during decommissioning is estimated to be approximately 14,500 of 200 L drums. In this study, the treatment process of decommissioning wastes will be reviewed through the case of the US Zion nuclear power station (ZNPS). Zion unit 1 and 2 received an operating license in 1973 and were permanently shut down and the spent nuclear fuel was transferred to the pool in 1998. The decommissioning was carried out according to the following five steps; (1) safe storage (SAFSTOR) dormancy, (2) preparation for decommissioning, (3) establishment of independent spent fuel storage installation (ISFSI) and transfer of the spent fuel and greater than class C radioactive materials, (4) decommissioning operations and (5) site restoration. The total volume of waste generated during decommissioning was expected to be approximately 1.7×105 m3. This is far above the Kori unit 1 waste estimation because ZNPS has a history of accidents and includes soil waste. Wastes were treated differently according to their properties and locations.

Trojan Nuclear Power Plant (NPP), a four-loop PWR designed by Westinghouse and owned by Portland General Electric (PGE), reached its initial threshold in 1975 and was operational until November 1992. PGE received a Possession Only License from the NRC in May 1993. In 1995, limited decommissioning activities began at the Trojan, including the completion of a large components removal project to remove and dispose of four steam generators and pressurizers from the containment building. In April 1996, the NRC approved a plan to dismantling the Trojan NPP and began more aggressive component removal activities. At the end of 1998, part of the radioactive drainage system began to be removed, and embedded piping decontamination and survey activities began. Trojan NPP has more than 8,840 m of contaminated pipelines throughout the power block. Most of Trojan NPP’s contaminated embedded piping can generally be divided into four categories drainage piping, ventilation ducts, buried process piping, and other items. For the Trojan NPP, the complete removal of contaminated and embedded piping without damaging the building would have significantly increased costs due to the structural considerations of the building and the depth of the embedded pipe. Therefore, Trojan NPP has chosen to conduct the Embedded Pipe Remediation Project (EPRP) to clean and in situ survey of most of the embedded piping to meet the Final Site Survey (FSS) acceptance criteria, with much success. This study provides a discussion of EPRP activities in the Trojan NPP, including classification and characterization of affected piping, modeling of proposed contamination acceptance criteria, and evaluation of various decontamination and survey techniques. It describes the decontamination tools, techniques, and survey equipment and the condition of work and cost estimate costs used in these projects. To identify embedded piping and drains at the Trojan NPP, based on frequent site surveys, plan sketches showing an overview of system flow paths and connections and database were developed to identify drain inputs and headers. This approach effort has been a successful method of remediation and site survey activities. The developed database was a valuable asset to the EPRP and a Work Breakdown Structure (WBS) code was assigned to each drains and headers, allowing the embedded piping to be integrated into the decommissioning cost estimation software (Decon. Expert) and schedule, which aided in decommissioning cost estimation. Also, regular database updates made it easy to check the status of the decommissioning project data. The waste system drain at Trojan NPP was heavily contaminated. The goal of the remediation effort is to completely remove all removable contamination and to reduce the fixed contamination below the decided contamination acceptance criteria. Accordingly, Hydrolysis, Media blast, Chemical decontamination and Pipe removal were considered as remediation option. Trojan NPP’s drainage pipe decontamination option did not cause a significant corrosion layer inside the pipe and media blast was chosen as the main method for stainless steel pipe. In particular, the decommissioning owner decontaminates most of the embedded piping in-situ to meet the FSS acceptance criteria for economic feasibility in Trojan NPP. The remaining pipe was filled with grout to prevent leaching and spreading of contamination inside the pipe. In-situ decontamination and survey of most of these contaminated pipes are considered the most cost-effective option.

A significant amount of piping is embedded in nuclear power plants (NPPs). In decommissioning these materials must be removed and cleaned. It can then be evaluated for radioactivity content below the release level. MARSSIM presents Derived Concentration Guideline Levels (DCGLs) that meet release guidelines. Calculating DCGL requires scenarios for the placement of embedded pipe and its long-term potential location or use. Some NPPs choose to keep the embedded pipes in the building. Because others will dismantle the building and dispose of the piping in-situ, determining the disposal option for embedded piping requires the use of measurement techniques with the sensitivity and accuracy necessary to measure the level of radioactive contamination of embedded piping and meet DCGL guidelines. The main measuring detectors used in NPPs are gas counters that are remotely controlled as they move along the inside of the pipe. The Geiger-Mueller (GM) detector is a detector commonly used in the nuclear field. Typically, this GM detector used 3-detectors that cover the entire perimeter of the pipe and are positioned at 120-degrees to each other. This is called a pipe crawler. It is very insensitive to gamma and X-ray, only measures beta-emitter and does not provide nuclide identification. The second method is a method using a high-resolution gamma-ray detector. Although not yet commercialized in many places, embedded piping is a scanning method. The technique only detects gamma-emitting nuclides, but some nuclides can be identified. Gamma-ray scanning identifies the average concentration per pipe length by the detector collimator. It is considerably longer than a pipe crawler. In addition, several techniques, including direct measurement of dose rate and radiochemical analysis after scraping sampling, are used and they must be used complementary to each other to determine the source term. Expensive sampling and radiochemical analysis can be reduced if these detectors are used to measure the radioactivity profile and to perform waste classification using scaling factor. In the actual Trojan NPP, a pipe crawler detector was used to survey the activity profile in a 26 foot of an embedded pipe. These results indicate that the geometric averaging of the factors and a dispersion values for each nuclide are constant within the accuracy factors. However, in order to accurately use the scaling factor in waste classification, it must have sample representativeness. Whether the sample through smear or scraping is representative of the radionuclide mixture in the pipe. Since the concentration varies according to the thickness of the deposit and depending on the location of the junction or bend, a lot of data are needed to confirm the reliability of the nuclide mixture. In this study, the reliability of the scaling factor, sampling representativeness and concentration measurement accuracy problems for waste classification in decommissioning NPP were evaluated and various techniques for measuring radioactive contamination on the inner surface of embedded pipes were surveyed and described. In addition, the advantages and limitations of detectors used to measure radioactivity concentrations in embedded piping are described. If this is used, it is expected that it will be helpful in determining the source term of the pipe embedded in the NPPs.

As the management procedure for self-disposal wastes stored in the radiation controlled area within the Korea Atomic Energy Research Institute (KAERI) have been established, and the types and quantities of wastes are increasing. In order to carry out the disposal of wastes with various generation histories, we expanded the processing range from surface contaminated waste, which was already in progress, to volumetric contaminated waste. In this paper, a case study of self-disposal of volumetric contaminated radioactive waste for which final disposal has been completed is described. In order to carry out of self-disposal of volumetric contaminated waste, it is important to collect representative samples and prove their representativenss. Based on this, a treatment plan was established after reviewing the history of the waste to be disposed of, and the treatment work was carried out as follows; waste collection, classification by size and shape, radiation (activity) measurement, sampling of representative samples, radioactivity concentration analysis, notification to regulatory bodies and question-and-answer, final disposal. The waste is judged have no potential for contamination because the polywood used to set the flat floor between the steel frame and floorboards in the experimental greenhouse didn’t come into contact with radioactive material. However, due to the conservative approach to the presence or absence of contamination, the treatment plan was established assuming volumetric contaminated waste. The type of waste is single wood, and the major contaminating radionuclides are Sr-85 and Cs-137. After the waste was collected and sorted by size and shape, it was weighed and a representative sampling amount and sampling method were set up. A direct method of surface contamination was performed on the entire area, and the representative sample was divided into three groups of homogenized population samples using the trisection method, with three points (upper/middle/below) were sampled at a 200:1 ratio, and radioactivity concentration analysis was conducted. After confirming that the concentration was below the allowable concentration for selfdisposal, the final disposal was completed after receiving approval after reporting to the regulatory body. As a result of radioactivity concentration analysis of representative samples, the maximum radioactivity concentration for each nuclide was Sr-85: < MDC (0.00178), Cs-137 : 0.00183 Bq/g (Sr-85 : 1 Bq/g, Cs-137 : 0.1 Bq/g), which meets the nuclide allowable concentration standard. It was confirmed that the total maximum fraction of 0.02 Bq/g satisfies the criteria (In the case of mixed nuclides, the sum of the fraction is less than 1). This paper introduces the establishment and implementation of self-disposal procedures based on the experience of self-disposal of radioactive waste with volumetric contaminants, and is going to utilize it as a basic material for self-disposal of radioactive waste with volumetric contaminants that will continue in the future and contribute to the reduction of radioactive wastes.

Kori-1, the nuclear power plants in South Korea, first started operation in April 1978 and was suspended permanently in 2017. The saturation rate time of spent nuclear fuel generated by major nuclear power plants operating in Korea are getting closer. If we fail to dispose spent nuclear fuel, which is equivalent to high-level radioactive waste, the nuclear power plants will have to be shutdown. High-level radioactive waste is permanently disposed through a deep geological disposal system because it contains long-term half-life nuclides and emits high energy. To select the deep geological disposal site and construct the disposal facilities, it is necessary to establish appropriate regulatory policies accordingly. The status of database construction in OECD-NEA, NRC, SITEX, and IAEA, which provides safety regulations for deep geological disposal system, stipulates each requirement for dismantling nuclear power plants. However, details such as specific figures are not specified, and guidelines for the disposal of high-level radioactive wastes are not clearly distinguished. In Korea, the CYPRUS program, an integrated database system, has been developed to support comprehensive performance evaluation for high-level waste disposal. However, due to several difficult situations, maintenance and upgrades have not been performed, so the research results exist only in the form of raw data and the new research results have not been reflected. Other than that, there is no preemptive basis for regulating the deep geological disposal system. With real-time database, we can develop a regulatory system for the domestic deep disposal system by systematically analyzing the regulatory condition and regulatory case data of international organizations and foreign leading countries. The database system processed and stored primary data collected from nuclear safety reports and other related data. In addition, we used relational database and designed table to maximize time and space efficiency. It is provided in the form of a web service so that multiple users can easily find the data they want at the same time. Based on these technologies, this study established a database system by analyzing the legal systems, regulatory standards, and cases of major foreign leading countries such as Sweden, Finland, the United States, and Japan. This database aims to organize data for each safety case component and further prepare a safety regulatory framework for each stage of development of disposal facilities suitable for the domestic environment.

In Korea, research on the development of safety case, including the safety assessment of disposal facility for the spent nuclear fuel, is being conducted for long-term management planning. The safety assessment procedure on disposal facility for the spent nuclear fuel heavily involves creating scenarios in which radioactive materials from the repository reach the human biosphere by combining Features, Events and Processes (FEP) that describe processes or events occurring around the disposal area. Meanwhile, the general guidelines provided by the IAEA or top-tier regulatory requirements addressed by each country do not mention detailed methods of ‘how to develop scenarios by combining individual FEPs’. For this reason, the overall frameworks of developing scenarios are almost similar, but their details are quite different depending on situation. Therefore, in order to follow up and clearly analyze the methods of how to develop scenarios, it is necessary to understand and compare case studies performed by each institution. In the previous companion paper entitled ‘Research Status and Trends’, the characteristics and advantages/disadvantages of representative scenario development methods were described. In this paper, which is a next series of the companion papers, we investigate and review with a focus on details of scenario development methods officially documented. In particular, we summarize some cases for the most commonly utilized methods, which are categorized as the ‘systematic method’, and this method is addressed by Process Influence Diagram (PID) and Rock Engineering System (RES). The lessons-learned and insight of these approaches can be used to develop the scenarios for enhanced Korean disposal facility for the spent nuclear fuel in the future.

본 연구는 디지털 미디어 환경변화 속에서 평생교육 강사들의 배움경험을 알아보기 위한 질적 사례연구이다. 디지털 미디어는 미래의 정보화 사회에 중요한 수단으로 계속 발전되어 왔지만, 코로나19로 인해 그 속도가 빨라졌다. 이러한 변화는 교육 현장도 예외가 없었다. 그 가운데 가장 큰 변화는 디지털 미디어 기반 온라인 수업이라고 할 수 있다. 특히, 공교육 현장과 달리 평생교육 현장에 있는 강사들에게는 체계적 지원 없이 교육 환경변화에 따른 준비와 책임이 발생했다. 이러한 책임 앞에 놓여있는 평생교육 강사들이 디지털 미디어 환경변화에 따라 새로이 경험하게 되는 배움 과정은 학습자의 배움과 교수설계 및 평생교육 기관의 경쟁력에도 영향을 미치는 중요한 요소라고 할 수 있다. 따라서 본 연구는 디지털 미디어 환경변화라는 현상의 맥락 안에서 평생교육 강사들의 배움경험을 알아보기 위해 특정 기준을 정하고 그 기준을 중심으로 연구 대상자를 선정했다. 연구 대상자로 선정된 평생교육 강사들은 디지털 미디어 환경변화에 적응하기 위한 내적, 외적 노력을 기울인 강사들이다. 이들을 대상으로 심층면담을 진행하고 이를 분석했다. 그 결과 첫째, 사례자들은 교육환경의 변화가 두렵기도 했지만, 변화를 장애라고 생각하지 않았다. 그들은 각자 자신에게 맞는 배움 방식을 선택하고 실천하며 자기 주도적으로 환경변화에 적응하려는 모습을 보였다. 둘째, 사례자들이 디지털 미디어 환경변화에 적응을 위해 발휘한 배움의 문제해결력은 사례자들 자신을 새롭게 발견하고 성장할 수 있게 만들어 준 새로운 배움으로 이어졌다. 셋째, 사례자들은 배움을 통해 교육적 환경변화에 적응을 위해 받은 도움이 다른 누군가에 돌려줘야 할 배움의 동반성장임을 깨닫게 되었다.