This study presents a methodology to determine the radionuclides of concern that are expected to be found during the final status survey of Kori Unit 1 decommissioning. The methodology involved reflecting the evaluation results of ORIGEN based on reference documents such as NUREG/CR-3474, NUREG/CR-4289, NUREG/CR-0130, WINCO-1191, and representative fuel loading. A list of potential radionuclides of concern was provided by reflecting the list of radionuclides of concern included in the Kori Unit 1 decommissioning plan. To select the radionuclides of concern, we analyzed the approach of US decommissioning plants based on the recommendations of NUREG-1757 Vol.2 Rev.1 and excluded certain radionuclides from the list. The final list of 23 radionuclides of concern was derived by excluding radionuclides that have a short half-life, low specific activity, analytically difficult to measure, inert gases, or naturally occurring radionuclides. This methodology can be applied to other nuclear power plants, such as the Wolsong Nuclear Power Plant, by reflecting the unique characteristics of the reactor.

Wolsong unit 1, the first PHWR (Pressurized Heavy Water Reactor) in Korea, was permanent shut down in 2019. In Korea, according to the Nuclear Safety Act, the FDP (Final Decommissioning Plan) must be submitted within 5 years of permanent shutdown. According to NSSC Notice, the types, volumes, and radioactivity of solid radioactive wastes should be included in FDP chapter 9, Radioactive Waste Management, Therefore, in this study, activation assessment and waste classification of the End shield, which is a major activation component, were conducted. MCNP and ORIGEN-S computer codes were used for the activation assessment of the End shield. Radioactive waste levels were classified according to the cooling period of 0 to 20 years in consideration of the actual start of decommissioning. The End shield consists of Lattice tube, Shielding ball, Sleeve insert, Calandria tube shielding sleeve, and Embedment Ring. Among the components composed for each fuel channel, the neutron flux was calculated for the components whose level was not predicted by preliminary activation assessment, by dividing them into three channel regions: central channel, inter channel, and outer channel. In the case of the shielding ball, the neutron flux was calculated in the area up to 10 cm close to the core and other parts to check the decrease in neutron flux with the distance from the core. The neutron flux calculations showed that the highest neutron flux was calculated at the Sleeve insert, the component closest to the fuel channel. It was found that the neutron flux decreased by about 1/10 to 1/20 as the distance from the core increased by 20 cm. The outer channel was found to have about 30% of the neutron flux of the center channel. It was found that no change in radioactive waste level due to decay occurred during the 0 to 20 years cooling period. In this study, activation assessment and waste classification of End Shield in Wolsong unit 1 was conducted. The results of this study can be used as a basis for the preparation of the FDP for the Wolsong unit 1.

Wolsong Unit 1 is about 679 MW Pressurized Heavy Water Reactor (PHWR). Canada AECL was responsible for Nuclear Steam Supply System (NSSS) design and supply. Wolsong Unit 1 was operated from 1983 to 2019. Currently, Wolsong Unit 1 is under safety management after permanent shutdown. Wolsong unit 1, a heavy water reactor, has the following characteristics. • Unlike Light Water Reactor, vertical reactors, Heavy Water Reactor is installed horizontally. • The internal structure of the reactor is more complex than that of a light water reactor (380 pressure tubes in reactor as called Calandria) • The Calandria Vault, a large concrete structure filled with light water, is located outside of Calandria In the case of the decommissioning plan of PHWR in Canada, they have adopted a deferred decommissioning strategy that decommissioning begins after permanent shutdown and long-term safety management (30 to 40 years). So, Decommissioning of PHWR in Canada is expected to start in the 2050s. Nuclear Safety Act stipulates that if a commercial nuclear power plant is permanently suspended, the utility must submit a Final Decommissioning Plan (FDP) within 5 years. So, KHNP, the utility, is developing the FDP for Wolsong Unit 1 and have a plan to submit it to the government by the end of 2024. And then licensing review is expected to take at least two years. The key milestone for decommissioning project has a plan to start decommissioning in 2027 and complete it by 2034, but this is flexible depending of the government’s approval for decommissioning and the completion of prerequisites such as spent fuel transfer, etc. KHNP has prepared a strategy and system consisting of three areas such as R&D, Engineering and licensing document development to prepare the final decommissioning plan for Wolsong Unit 1. The promotion system for the preparation of the FDP for Wolsong Unit 1 is consisted of Engineering (HAS Characterization, Process/Work Package/Cost Estimation, Dismantling Safety Evaluation, Radiological Environmental Report, Radioactive Waste Treatment and Facility Construction), R&D (COG cooperation, KHNP R&D Results), Kori unit 1 lessons learned, etc. KHNP have the plan that the FDP Draft development by the end of 2023, reflecting engineering services results, R&D results, COG technical cooperation results and lessoned learned on Kori Unit 1. After collecting opinions from residents through a public hearing, the FDP will be submitted to the government by the end of 2024. It is expected that there will be many difficulties in the development process as it is the world’s first FDP development for the commercial Pressurized Heavy Water Reactors.

Kori Unit 1 is about 600MW Pressurized Light Water Reactor as WH type. KHNP got an approval for construction and operation of Kori unit 1 on May 31, 1972 and started commercial operation from Apr. 29, 1987. And then, it was decided to permanently suspend it on Jun. 18, 2017 after 40 years of commercial operation. The Nuclear Safety Act stipulates that if a commercial nuclear power plant is permanently suspended, the utility must submit a Final Decommissioning Plan (FDP) within 5 years. So, KHNP, the utility, developed a FDP for Kori Unit 1 and submitted it to the government in May 2021. In South Korea, the FDP is to be prepared in accordance with the relevant notices and consists of 11 major chapters such as (1) Decommissioning Plan Overview, (2) Project management, (3) Status of Site and Environmental, (4) Decommissioning Strategies and Method, (5) Ease of Decom. Design characteristic, (6) Safety Analysis, (7) Radiation Protection, (8) Decontamination and Dismantling, (9) Radioactive Waste Management, (10) Environmental Impact Analysis, (11) Fire Protection and (12, 13) Etc., References and Glossary. KHNP has prepared a strategy and system consisting of three areas such as R&D, Engineering and licensing document development to prepare the final decommissioning plan for Kori Unit 1. The promotion system for the preparation of the FDP for Kori Unit 1 is composed of Engineering (HAS Characterization, Dismantling Safety Evaluation, Radiological Environmental Report, Radioactive Waste Treatment and Facility Construction), R&D(KHNP R&D Results such as Process/Work Package /Cost Estimation, Safety Analysis, Contamination and Exposure, Guide for Detailed Characteristic, Site Safety Analysis, RV & RVI Dismantling Process, etc.), Overseas case lessons learned(Taiwan unit 1 NPP FDP and Spain Zorita NPP FDP analysis) and Development of Licensing Document. KHNP completed the initial completion of the Final Decommissioning Plan for Kori Unit 1 in March 2020 and carried out collecting residents’ opinions through public hearings. And then, KHNP supplemented the results of the residents’ opinions and applied for license to the Nuclear Safety and Security Commission in May, 2021. Now, KHNP are responding to the FDP licensing review.

Korea Atomic Energy Research Institute is developing a radionuclide management processes as a conditioning technology to reduce the burden of spent fuel disposal. The radionuclide management process refers to a process managing radionuclides with similar properties by introducing various technology options that can separate and recover radionuclides from spent fuels. In particular, it is a process aimed at increasing disposal efficiency by managing high-heat, high-mobility, and high-toxic radionuclides that can greatly affect the performance of the disposal system. Since the radionuclide management process seeks to consider various technology options for each unit process, it may have several process flows rather than have a single process flow. Describing the various process flows as a single flow network model is called the superstructure model. In this study, we intend to develop a superstructure model for the radionuclide management process and use it as a model to select the optimal process flow. To find the optimal process flow, an objective function must be defined, and at the fuel cycle system level multiple objectives such as effectiveness (disposal area), safety (explosure dose), and economics (cost) can be considered. Before performing the system-level optimization, it is necessary to select candidates of process flow in consideration of waste properties and process efficiency at the process level. In this study, a sensitivity analysis is conducted to analyze changes in waste properties such as decay heat and radioactivity when the separation ratio varies due to the performance change for each unit process of the radionuclide management process. Through this analysis, it is possible to derive a performance range that can have waste properties suitable for following waste treatment, especially waste form manufacturing. It is also possible to analyze the effect of waste properties that vary according to the performance change on waste storage and management approaches.

A phosphorylation (phosphate precipitation) technology of metal chlorides is considering as a proper treatment method for recovering the fission products in a spent molten salt. In KAERI’s previous precipitation tests, the powder of lithium phosphate (Li3PO4) as a precipitation agent reacted with metal chlorides in a simulated LiCl-KCl molten salt. The reaction of metal chlorides containing actinides such as uranium and rare earths with lithium phosphate in a molten salt was known as solidliquid reaction. In order to increase the precipitation reaction rate the powder of lithium phosphate dispersed by stirring thoroughly in a molten salt. As one of the recovery methods of the metal phosphates precipitated on the bottom of the molten salt vessel cutting method at the lower part of the salt ingot is considered. On the other hand, a vacuum distillation method of all the molten salt containing the metal phosphates precipitates was proposed as another recovering method. In recent study, a new method for collecting the phosphorylation reaction products into a small recovering vessel was investigated resulting in some test data by using the lithium phosphate ingot in a molten salt containing uranium and three rare earth elements (Nd, Ce, and La). The phosphorylation experiments using lithium phosphate ingots carried out to collect the metal phosphate precipitates and the test result of this new method was feasible. However, the reaction rate of test using lithium phosphate ingot is very slower than that of test using lithium phosphate powder. In this presentation, the precipitation reactor design used for phosphorylation reaction shows that the amount of molten salt transferred to the distillation unit will reduce by collecting all of the metal phosphates that will be generated using lithium phosphate powder into a small recovering vessel.

The current storage capacity of the spent nuclear fuel at the Kori unit 2 has reached approximately 94% saturation, excluding emergency core capacity. As South Korea has not yet constructed any interim storage facilities to store spent nuclear fuel, one of possible options is to install high density storage racks in spent fuel pool at the reactor site to increase its capacity. The high density storage rack is a technology developed to allow the storage to have more spent nuclear fuel than originally planned, while still ensuring safety. It achieves this by reducing the storage gap between the spent nuclear fuel. The purpose of this study is to investigate an appropriate storage capacity for spent fuel pool in the Kori unit 2 and the factors to be considered during the high density storage rack installation. Given that the Kori unit 2 is planning continued operation and Korea Hydro & Nuclear Power (KHNP) is preparing to install a temporary storage facility for spent nuclear fuel at the Kori nuclear site, it is reasonable to consider the installation of high density storage racks in the spent fuel pool as a storage solution for these issues. When evaluating the capacity of the spent fuel pool, the amount of spent nuclear fuel generated by other reactors in Kori nuclear site and the amount of spent nuclear fuel generated by continued operation of the Kori unit 2 should be taken into account. This study aims to consider these factors and evaluate the capacity of the spent fuel pool. Furthermore, when installing high density storage rack for the spent nuclear fuel, it should be noted that the existing storage racks at the Kori unit 2 are welded to the liner plate, which may require additional cutting work. Therefore, it is necessary to review the suitable method for the cutting work. Additionally, assuming that divers need to access the area near the storage racks or cutting & welding devices require radiation protection in the area, it is essential to analyze the expected radiation level with computational code and propose appropriate measures to limit work time or establish a work zone. Thus, this study evaluates appropriate capacity of spent fuel pool and work methods for the installation of high density storage rack in the spent fuel pool at the Kori unit 2. It is expected that this paper contributes to install high density storage racks in SFP safely.

Background: While efforts have been made to differentiate fall risk in older adults using wearable devices and clinical methodologies, technologies are still infancy. We applied a decision tree (DT) algorithm using inertial measurement unit (IMU) sensor data and clinical measurements to generate high performance classification models of fall risk of older adults. Objects: This study aims to develop a classification model of fall risk using IMU data and clinical measurements in older adults. Methods: Twenty-six older adults were assessed and categorized into high and low fall risk groups. IMU sensor data were obtained while walking from each group, and features were extracted to be used for a DT algorithm with the Gini index (DT1) and the Entropy index (DT2), which generated classification models to differentiate high and low fall risk groups. Model’s performance was compared and presented with accuracy, sensitivity, and specificity. Results: Accuracy, sensitivity and specificity were 77.8%, 80.0%, and 66.7%, respectively, for DT1; and 72.2%, 91.7%, and 33.3%, respectively, for DT2. Conclusion: Our results suggest that the fall risk classification using IMU sensor data obtained during gait has potentials to be developed for practical use. Different machine learning techniques involving larger data set should be warranted for future research and development.

Reactor pressure vessels and steam generators generated in the process of dismantling nuclear power plants or replaced steam have various shape and occupy a considerable amount of the disposal site when disposed of in original shape. For the development of domestic technologies related to the disposal of large wastes, it is necessary to secure technologies for reducing large radioactive metal wastes, including technologies such as decontamination, cutting, melting, and residual radioactivity evaluation. Cases of disposal of steam generators in Europe and the United States were investigated. Except for u-tubes, steam generators are less contaminated or easily decontaminated, so it is possible to reduce the volume of waste subject to final disposal by exempting a significant amount through appropriate treatment. Korea Hydro & Nuclear Power Co. is currently temporarily storing 24 steam generators at 41.6 billion won. This paper presents a method to exempt more parts of the steam generator and reduce the volume of waste by properly combining mechanical cutting thermal cutting and melting to dispose of the steam generator. Currently the decontamination and dismantling industries of nuclear facilities are gradually expanding around the world. Therefore, it is necessary to localize the treatment technology for metal waste generated during maintenance and dismantling. The result of this study can be used to establish waste reduction and disposal method for dismantling steam generators.

3D modeling is a technology for representing real objects in a virtual 3D space or modeling and reproducing the physical environment. 2D drawings for viewing the existing building structure have limitations that make it difficult to understand the structure. By implementing this 3D modeling, specific visualization became possible. 3D technology is being applied in a wide range of preevaluation work required for nuclear decommissioning. In Slovakia, 3D modeling was applied to determine the optimal cutting strategy for the primary circuit before dismantling the VVER type Bohunice V1. In Japan, the Decommissioning Engineering Support System (DEXUS) program has been developed that incorporates VRDose, a decommissioning engineering support system based on 3D CAD models. Through this, the cutting length of the work object and the quantity of containers for packaging waste are calculated, exposure dose calculations in various dismantling scenarios, and cost estimation are performed. Korea also used 3D technology to evaluate the decommissioning waste volume for Kori Unit 1 and to evaluate the optimal scenario of the decommissioning process procedure for the research reactor Unit 1. 3D technology is currently being used in various pre-decommissioning evaluations for VVER, PWR, and research reactors. Overseas, a program that matches various decommissioning pre-evaluation tasks with cost estimation has also been developed. However, most 3D technologies are mainly used as a support system for dose evaluation and amount of decommissioning waste calculation. In this study, 3D modeling was performed on the PHWR structure, and physical and radiological information about the structure was provided. The information on the structure can present the unit cost for the work object by confirming the standard of the applied unit cost factor (UCF). The UCF presents the unit cost for repeated decommissioning operations. The decommissioning cost of the work object can be obtained by multiplying the UCF by the number of repetitions of the work. If the results of this study are combined with the process evaluation and waste quantity estimation performed in previous studies, it is judged that it will be helpful in developing an integrated NPP decommissioning program that requires preliminary evaluation of various tasks. In addition, it is judged that a clear cost estimation of the object to be evaluated will be possible by matching the 3D work object with the UCF.

In operating or permanently shut down nuclear power plants which were built between 1970s and 1990s, asbestos was widely used for ceiling materials, wall materials, and gaskets. Furthermore, it was mainly treated as a heat-resistant material like insulation. In Kori Unit 1, radioactive asbestos was replaced or removed through maintenance and repair in the containment building during the operation period of about 40 years, but radioactive asbestos still remains that need to be partially dismantled. Generally, it is more difficult to handle because it belongs to two different waste categories, radioactive waste and hazardous waste. In addition, the risk increases further due to radioactivity with the asbestos hazards itself. Therefore, it is very important to accurately determine the amount of radioactive asbestos waste and to evaluate the treatment method and disposal reduction rate before the decommissioning is started. According to the Korean Waste Management Act, three methods are recommended for the asbestos (hazardous waste) treatment: landfill, solidification, and high-temperature melting. Landfill is commonly used in Korea and the United States while high-temperature melting and solidification are additionally recommended only in Korea. Considering the situation in Korea, landfill is not appropriate due to the limitations of landfill capacity and potential risks (hazards still remain). Therefore, the other two methods can be considered sufficiently in terms of safety, detoxification, and reduction rate. This paper evaluates the amount of radioactive asbestos waste at Kori Unit 1 based on the actual asbestos building material data (as of February 2022) of the Asbestos Management Comprehensive Information Network. Vitrification is considered as a sufficient alternative for treating radioactive asbestos waste. And, it is checked whether the vitrified waste through the high-temperature melting method, plasma torch, meets the requirements such as detoxification, compressive strength and leachability for storage and disposal stability. It is expected to be useful to prepare a radioactive mixed waste management standard and to reduce the disposal cost through the reduction of final waste.

It is necessary to prepare for cutting and storing waste materials in the reactor vessel internals (RVI) for successful decommissioning of the nuclear power plant (NPP). Since RVI contain massive components and relatively highly activated, their decommissioning process should be conducted carefully in terms of radiological and industrial safety. To achieve efficient decommissioning waste management, this study presents radiation level of RVI and cutting optimization was performed for intermediate level waste. As a result of the radiation evaluation, a part of the core side and the upper part of RVI were evaluated as intermediate-level waste, and other components were evaluated as very low-level or lowlevel waste. For intermediate-level waste cutting, the minimum cutting method that can be put into a container was reviewed in consideration of the size, thickness, and cutting method of the interior product. The final segmentation parts are expected to be loaded into two storage containers.

In the design of a spent-fuel (SF) storage, the consideration of burnup credit brings the benefits in safety and economic views. According to it, various SF burnup measurement systems have been developed to estimate high fidelity burnup credit, such as FORK and SMOPY. Recently, there are a few attempts to localize the SF burnup measurement system in South Korea. For the localization of SF burnup measurement systems, it is very important to build the isotope inventory data base (DB) of various kinds of SFs. In this study, we performed DeCART2D/MASTER core follow calculations and McCARD single fuel assembly (FA) burnup analyses for Hanbit unit 3 and confirmed the characteristic of the isotope inventory over burnup. Firstly, the core follow calculations for Cycles 1~7 were performed using DeCART2D/MASTER code system. The core follow calculation is very realistic and practical because it considers the design conditions from its nuclear design report (NDR). Secondly, the Monte Carlo burnup analyses for single FAs were conducted by the McCARD Monte Carlo (MC) transport code. The McCARD code can utilize continuous energy cross section library and treat complex geometric information for particle transport simulation. Accordingly, the McCARD code can provide accurate solutions for burnup analyses without approximations, but it needs huge computing resources and time burden to perform whole-core follow calculations. Therefore, we will confirm the effectiveness of the single McCARD FA burnup analyses by comparing the DeCART2D/MASTER core follow results with the McCARD solution. From the results, the use of single FA burnup analyses for the establishment of the DBs will be justified. Various FAs, that have different 235U enrichments and loading pattern of fuel rods and burnable absorbers, were considered for the burnup analyses. In addition, the results of the sensitivity analyses for power density, initial enrichment, and cooling time will be presented.

Various optical devices are utilized to fire artillery in Korean Army. To transport and store the optical devices, some cases are made up for a set. The covers of cases are made of Glass Fiber Reinforced Plastic(GFRP) and resin. Because of materials of case, it has a disadvantage in productivity for manufacturing covers and assembling components. In this study, ABS resin is presented for alternative and verify possibility through FEM analysis. Comparing with GFRP, max stress-ratio of ABS is decreased 19.6% for bottom, 8.2% for side and 25.4% for edge. Also each strain is changed 17.3%, 180.3% and 17.7%. According to the research results, ABS resin is considered possible for alternative. And the number of components is decreased for around 73.8%.

저격수는 전장에서 적의 사기와 전투의지에 지대한 영향을 미치며 4차 산업혁명으로 첨단무기가 발달한 현대 전장에서도 적에게 치명적인 피해 를 입히게 된다. 따라서 본 연구의 목적은 4차 산업혁명 시대에 맞추어 한국 육군의 특수작전부대가 안전화작전 단계 시 북한의 도시지역에서 저격전 수행 능력 향상을 위한 발전방향을 제시하는데 있다. 연구 방법 은 저격전의 개념을 정립하고 저격전의 영향요인(전장환경, 위협인식, 대 응의지)에 따른 저격전 전법(작전수행개념, 조직편성, 무기체계)을 중심으 로 분석하였고, 미군의 이라크 도시지역작전 저격전 사례분석을 통해 교 훈을 도출하여 한국군에 적용하는 사례분석 방법을 적용하였다. 연구 결 과 저격전의 영향요인에 따라 저격전 전법이 발전함을 확인하였고, 안정 화작전 단계에서 북한지역 도시지역작전 시 특수작전부대 저격전의 영향 요인에 따른 창의적인 저격전 전법을 제시했다. 따라서 한국 육군 특수 작전부대는 단순한 저격수 운용이 아닌 합동작전개념의 저격전을 발전시 키기 위해 저격전의 영향요인에 능동적으로 대응하여 저격전 전법을 창 의적으로 발전시켜야 할 것이다.

The restraints are used in the intensive care unit as a way to restrict the movement of the whole body or part of the body for the safety of the patient, such as maintaining medical equipment, preventing falls and patient self tube removal through regulating the patient's behavior. However, the physical restraints are used for the physical safety of the patient, but that is cause a variety of physical and psychological complications. Thus, nurses in intensive care units who apply restraint on face an ethical dilemma in the confrontation between the nurse's duty to improve the patient's health and protect the patient's safety and the value of the patient's autonomy and dignity. We have two examples are proposed to identify the ethical dilemma situation faced by nurses, to induce an active attitude towards elimination by establishing exactly restraint - removal criteria. In addition, it is proposed to seek a balance of individual values through active communication between the patient and the medical staff regarding the application of the restraint.

With the enforcement of the “Act on decisions on life-sustaining treatment for patients at the end of life” in February 2018, discussion on advanced care planning (ACP) has increased. However, as decisions on life-sustaining treatments are still made in the intensive care unit, deaths related to the suspension of life-sustaining treatment account for a large proportion of deaths in the intensive care unit. The nurses encounter challenges in supporting the patient's dignified death; they experience an ethical dilemma in the ambiguity due to a lack of guidance on legal responsibilities regarding decisions on life-sustaining treatment. In order for the nurses to perform as a supporter providing care to the patients and as a advocate during the process of decision-making on life-sustaining treatment, there should be a systemic change to ensure the nurses' participation. In addition, an open and continuous discussion should be proposed to cultivate nurses’ ethical sensitivity and moral courage. This paper reports two ethical examples related to the decisions on life-sustaining treatment occurred in intensive care units of a tertiary hospital.

일반적으로 과거 건설공사는 그동안 축적된 기술기준과 경험 그리고 성과를 바탕으로 한 설계와 시공단계에서부터 준공단계까지의 사업관리를 통해 그 품질 향상을 도모해왔다. 그러나 최근 제4차 산업혁명 등 환경변화의 바람은 건설 산업에 있어 기술적 측면은 물론 관리적 측면에 대해서도 많은 영향을 미치고 있다. 이러한 상황에서 건설산업의 각 분야에서 요구되는 생산성 향상, 공정거래 구현 등에 대한 사회적 기대는 건설산업에 대한 변화의 모티브로 작용하고 있다. 정부도 역시 이와 같은 시대적 변화에 대응코자 관계법령 등 제반 규정을 정비 하는데 지속적인 노력을 기울이고 있는 상황에서, 본 연구는 건설사업관리와 관련한 법률규정을 기준으로 품질제고, 비용절감, 공기단축 등을 효율적으로 관리하는 방안에 대해 검토하였다. 여러가지 관리방안 중에 본 논문은 발주자와 관련업계 등에게 가장 민감하게 작용하고 있고 상호간 이해의 충돌이 빈번한 공공부문측면에서 공동주택 건설사업관리 용역대가산정에 대한 제반문제 등을 살펴보고 이에 대한 개선방안을 도출하고자 연구를 진행하였다.