This study analyzes the impact of occupational health risk assessments on the safety and health levels and the safety behaviors of workers in manufacturing workplaces. An online survey was conducted among 3,172 companies, yielding 637 responses. The statistical analysis on the collected responses revealed three key findings. First, the safety and health levels (safety importance, safety comprehension, safety awareness) positively influence the outcomes of occupational health risk assessments(safety practice, safety management, safety improvement) and safety behaviors (activity change, safety check). Second, the effectiveness of occupational health risk assessments has a positive impact on safety behaviors. Lastly, the effectiveness of occupational health risk assessments partially mediate the relationship between safety and health levels and safety behaviors. These findings are expected to contribute to the promotion of risk assessments in the field of industrial health and to enhancing safety performance by improving workplace safety, health levels, and safety behavior.

The purpose of this study is to analyze effect of Army Risk Assessment System(ARAS) which is used to prevent safety accident in ROK army. Based on prior research, we select 4 indicators which are related to accident prevention effect and analyze the differences before and after ARAS operation for each indicators by using Paired-Samples T-Test. Also, we analyze the correlation between degree of ARAS operation and status of safety accidents of 112 ROK Army units. We conduct an evaluation of each function within the system using IPA method. The results of this study are as follows; All 4 indicators are improved compared to before ARAS operation, and the differences are statistically significant. Also, there is negative correlation between the degree of ARAS operation and the occurrence of safety accidents. So, the operation of ARAS has a positive effect on preventing safety accidents. Finally among the 15 functions of ARAS, 4 functions require improvement. The findings of this study have implications for proposing necessity of computerized system in enforcing Risk Assessment. Also, whether or not operating ARAS is important, but it is also important to operate it well. Lastly, We propose improvement plans for each function to operate it well.

STCW 협약 A-VI에 의거, 승선 필수 증서를 발급받기 위해 소화 과정 최저 훈련 이수를 거치게 되는 공간이 수소화 훈련장 이다. 선박 화재의 상황과 유사한 장소에서 발생한 화재 진압 훈련을 위한 실습 장소의 특성상, 재실자의 안전을 보장할 수 있는 운영 을 위해 안전성을 수치화하여 평가하고 기준을 수립하는 연구의 필요성이 대두되었다. 화재 안전 평가를 위해 FDS를 기반으로 한 Pyrosim을 활용하여 제연설비 유무에 따른 Case를 설정, vector의 분석을 통한 연기 유동 및 열기에 대한 평가를 수행하였다. 피난 안전 평가는 Pathfinder를 통해 허용 피난시간, 총 피난시간을 수치화된 결과로 해석, 비교하여 안전성을 분석하였다. 각 Case에 대한 안전성 을 평가함으로 제연설비별 기능의 적정성을 수치, 시각적으로 나타내었으며, 현재 운영 상태는 안전성이 양호한 것으로 평가하였다. 집 진설비가 정지한 비상상황은 각 피난시간과 111.2초의 여유시간으로 나타내어 수소화 훈련장의 비상상황에 대한 피난시간의 기준으로 활용할 것을 제시하였다.

최근 원자력 지진 PSA(Probabilistic Safety Assessment)를 토대로 산업시설물의 지진 PSA를 수행하는 연구가 진행되었다. 해당 연 구는 원자력 발전소와 산업시설물의 차이를 파악하고, 최종적으로 운영정지를 목표로 하는 고장수목(Fault Tree)를 구축한 후 시각적 확률도구인 베이지안 네트워크(Bayesian Network, BN)으로 변환하였다. 본 연구는 선행연구를 기반으로 지진으로 유발된 구조손상 으로 인해 발생 가능한 화재･폭발에 대해 PSA를 수행하고자 하였다. 이를 위해 화재･폭발을 사건수목(Event Tree)으로 표현하고, BN 으로 변환하였다. 변환된 BN은 화재･폭발 모듈로서 선행연구에서 제시된 고장수목 기반 BN과 연계되어 최종적으로 지진 유발 화재･ 폭발 PSA를 수행할 수 있는 BN 기반 방법론이 개발되었다. 개발된 BN을 검증하기위해 수치예제로서 가상의 가스플랜트 Plot Plan을 생성하였고, 가스플랜트의 설비 종류가 구체적으로 반영된 대규모 BN을 구축하였다. 해당 BN을 이용하여 지진 규모에 따른 전체시 스템의 운영정지 확률 및 하위시스템들의 고장확률 산정과 더불어 역으로 전체시스템이 운영 정지되었을 때 하위시스템들의 영향도 분석과 화재･폭발 가능성을 산정하여 다양한 의사결정을 수행할 수 있음을 제시함으로써 그 우수성을 확인하였다.

교통사고는 인적요인, 도로 기하구조, 교통류, 환경적 요인 등 복합적인 요인에 의해 발생하고 속도는 교통사고와 밀접한 연관성이 있다. 또한, 교통사고는 교통 혼잡도와 관련이 있으며 사고와 실시간 교통상황 간의 상관관계를 통해 사고 발생 개연성을 추정하고 도 로 안전성 분석이 필요하다. 모바일 센서와 통신 기술의 급속한 발전으로 스마트폰 보급률이 증가하였으며 내장된 센서를 기반으로 생성된 차량 주행 데이터 수집이 가능하다. 기술의 발달로 데이터 수집이 쉬워졌음에도 불구하고, 스마트폰을 기반으로 수집된 위험 운전 이벤트를 활용한 도로 위험도 평가에 대한 연구는 부족한 실정이다. 본 연구는 스마트폰 센서 기반의 위험 운전 이벤트 데이터 중 하나인 급감속 위험 운전 이벤트 데이터를 도로 위험도 평가 기법에 활용하는 것을 목적으로 한다. 급감속 위험 운전 이벤트 데이 터는 주행 차량이 3초간 속도를 40km/h 이상 감소하는 위험 이벤트가 발생할 때 시간과 위치를 기록한 자료를 의미한다. 본 연구의 범위는 대한민국 내 인구와 교통량이 많은 지역인 수도권을 대상으로 서울, 경기, 인천을 연결하는 고리 형태의 도로인 수도권제1순환 선을 대상으로 하였다. 먼저, 개별 차량 데이터는 좌표 기반의 내비게이션 데이터로 집계하여 VDS 링크 데이터와 매칭하였다. 다음으 로는 개별 차량의 위험 운전 이벤트 데이터와 차량 검지기의 교통 매개변수를 결합한 새로운 지표를 개발하였다. 또한, 시·공간적 교 통류의 특성을 반영하여 다양한 도로 위험도 평가 방법에 활용하고자 하였다. 마지막으로 위험 운전 지표와 이력 자료를 기반으로 통 계적으로 유의한 안전성능함수를 개발하였으며, 다양한 시간 단위의 집계 수준을 활용하여 도로 구간별 최적의 모형을 제안하였다. 본 연구는 스마트폰 센서를 기반으로 식별한 개별 차량의 위험과 교통류 차원의 위험을 결합하여 새로운 위험 지표를 개발하고 도로 위 험도 평가에 활용한다는 것에 의의가 있다. 결과물은 향후 스마트폰 센서 기반 개별 차량 위험 운전 이벤트 데이터와 교통 조건을 통 합하는 도로 위험도 평가의 기초자료로써 활용될 것으로 기대된다.

With South Korea increasingly focusing on nuclear energy, the management of spent nuclear fuel has attracted considerable attention in South Korea. This study established a novel procedure for selecting safety-relevant radionuclides for long-term safety assessments of a deep geological repository in South Korea. Statistical evaluations were performed to identify the design basis reference spent nuclear fuels and evaluate the source term for up to one million years. Safety-relevant radionuclides were determined based on the half-life criteria, the projected activities for the design basis reference spent nuclear fuel, and the annual limit of ingestion set by the Nuclear Safety and Security Commission Notification No. 2019-10 without considering their chemical and hydrogeological properties. The proposed process was used to select 56 radionuclides, comprising 27 fission and activation products and 29 actinide nuclides. This study explains first the determination of the design basis reference spent nuclear fuels, followed by a comprehensive discussion on the selection criteria and methodology for safety-relevant radionuclides.

Activated carbon (AC) is used for filtering organic and radioactive particles, in liquid and ventilation systems, respectively. Spent ACs (SACs) are stored till decaying to clearance level before disposal, but some SACs are found to contain C-14, a radioactive isotopes 5,730 years halflife, at a concentration greater than clearance level concentration, 1 Bq/g. However, without waste acceptance criteria (WAC) regarding SACs, SACs are not delivered for disposal at current situation. Therefore, this paper aims to perform a preliminary disposal safety examination to provide fundamental data to establish WAC regarding SACs SACs are inorganic ash composed mostly of carbon (~88%) with few other elements (S, H, O, etc.). Some of these SACs produced from NPPs are found to contain C-14 at concentration up to very-low level waste (VLLW) criteria, and few up to low-level waste (LLW) criteria. As SACs are in form of bead or pellets, dispersion may become a concern, thus requiring conditioning to be indispersible, and considering VLL soils can be disposed by packaging into soft-bags, VLL SACs can also be disposed in the same way, provided SACs are dried to meet free water requirement. But, further analysis is required to evaluate radioactive inventory before disposal. Disposability of SACs is examined based on domestic WAC’s requirement on physical and chemical characteristics. Firstly, particulate regulation would be satisfied, as commonly used ACs in filters are in size greater than 0.3 mm, which is greater than regulated particle size of 0.2 mm and below. Secondly, chelating content regulation would be satisfied, as SACs do not contain chelating chemicals. Also, cellulose, which is known to produce chelating agent (ISA), would be degraded and removed as ACs are produced by pyrolysis at 1,000°C, while thermal degradation of cellulose occurs around 350~600°C. Thirdly, ignitability regulation would be satisfied because as per 40 CFR 261.21, ignitable material is defined with ignition point below 60°C, but SACs has ignition point above 350°C. Lastly, gas generation regulation would be satisfied, as SACs being inorganic, they would be targeted for biological degradation, which is one of the main mechanism of gas generation. Therefore, SACs would be suitable to be disposed at domestic repositories, provided they are securely packaged. Further analysis would be required before disposal to determine detailed radioactive inventories and chemical contents, which also would be used to produce fundamental data to establish WAC.

Among nuclear power plants in the Republic of Korea, Kori Unit 1 and Wolsong Unit 1 have been permanently shut down, and Kori Unit 1 is preparing to be decommissioned. According to the decommissioning plan (DP) of Kori Unit 1, a radioactive waste processing complex will be built on the Kori site to reduce radioactive waste generated during decommissioning actively, and various types of decommissioning waste are expected to be treated in the complex. It is judged that matters related to the safety assessment of the complex are not included in the DP since the equipment and treatment processes have not been determined. IAEA GSR Part 5 states that radioactive waste processing complex shall be operated according to national regulations and the conditions imposed by the regulatory body. However, it has been confirmed that separate regulatory requirements for the complex have not yet been established in Korea. It is expected that the Regulation on Technical Standards for Nuclear Facilities, etc. will be applied mutatis mutandis. Liquid and gaseous radioactive materials can be expected to be released into the sea or atmosphere during the operation of the complex. Accordingly, it should be proved that standards such as discharge limits of radioactive effluents are met. Although the assessment of radioactive effluent discharged from nuclear power plants to the environment is systematically conducted, it has been confirmed that the safety assessment framework for radioactive effluents discharged from the complex has not yet been established. Currently, the SAFRAN Tool is based on SADRWMS (Safety Assessment Driving Radioactive Waste Management Solutions), an IAEA safety assessment methodology for pre-disposal management, which uses Pathway Dose Factors (PDFs) derived from generic environmental models. Therefore, in order to conduct a more detailed safety assessment of the complex on a specific site, site characteristic data should be reflected. Although safety assessment using the SAFRAN Tool was conducted at the Thailand Institute of Nuclear Technology (TINT) facility, detailed data were not provided, and PDFs reflecting site characteristic data were not applied. Also, no other studies that considered many types of waste and provided detailed data on the safety assessment were not confirmed. Therefore, this study developed K-CRAFT (Kyung Hee – Comprehensive RAdioactive waste treatment Facility safety assessment Tool), this tool that can derive PDFs by reflecting site characteristic data based on the SADRWMS methodology and conducted preliminary safety assessment for the complex which will be built in Kori site by this tool.

Over the years, in the field of safety assessment of geological disposal system, system-level models have been widely employed, primarily due to considerations of computational efficiency and convenience. However, system-level models have their limitations when it comes to phenomenologically simulating the complex processes occurring within disposal systems, particularly when attempting to account for the coupled processes in the near-field. Therefore, this study investigates a machine learning-based methodology for incorporating phenomenological insights into system-level safety assessment models without compromising computational efficiency. The machine learning application targeted the calculation of waste degradation rates and the estimation of radionuclide flux around the deposition holes. To develop machine learning models for both degradation rates and radionuclide flux, key influencing factors or input parameters need to be identified. Subsequently, process models capable of computing degradation rates and radionuclide flux will be established. To facilitate the generation of machine learning data encompassing a wide range of input parameter combinations, Latin-hypercube sampling will be applied. Based on the predefined scenarios and input parameters, the machine learning models will generate time-series data for the degradation rates and radionuclide flux. The time-series data can subsequently be applied to the system-level safety assessment model as a time table format. The methodology presented in this study is expected to contribute to the enhancement of system-level safety assessment models when applied.

The effect of various physicochemical processes, such as seawater intrusion, on the performance of the engineered barrier should be closely analyzed to precisely assess the safety of high-level radioactive waste repository. In order to evaluate the impact of such processes on the performance of the engineered barrier, a thermal-hydrological-chemical model was developed by using COMSOL Multiphysics and PHREEQC. The coupling of two software was achieved through the application of a sequential non-iterative approach. Model verification was executed through a comparative analysis between the outcomes derived from the developed model and those obtained in prior investigations. Two data were in a good agreement, demonstrating the model is capable of simulating aqueous speciation, adsorption, precipitation, and dissolution. Using the developed model, the geochemical evolution of bentonite buffer under a general condition was simulated as a base case. The model domain consists of 0.5 m of bentonite and 49.5 m of granite. The uraninite (UO2) was assigned at the canister-bentonite interface as the potential source of uranium. Assuming the lifetime of canister as 1,000 years, the porewater mixing without uranium leakage was simulated for 1,000 years. After then, the uranium leakage through the dissolution of uraninite was initiated and simulated for additional 1,000 years. In the base case model, where the porewater mixing between the bentonite and granite was the only considered process, the gypsum tended to dissolve throughout the bentonite, while it precipitated in the vicinity of bentonite-granite boundary. However, the precipitation and dissolution of gypsum only showed a limited effect on the performance of the bentonite. Due to the low solubility of uraninite in the reduced environment, only infinitesimal amounts of uranium dissolved and transported through the bentonite. Additional cases considering various environmental processes, such as seawater or cement porewater intrusion, will be further investigated.

Since the first operation of the Gori No. 1 nuclear power plant in Korea was started to operate in 1978, currently 24 nuclear power plants have been being operated, out of which 21 plants are PWR types and the rest are CANDU types. About 30% of total electricity consumed in Korea is from all these nuclear power plants. The accumulated spent nuclear fuels (SFs) generated from each site are temporarily being stored as wet or dry storage type at each plant site. These SFs with their high radiotoxicity, heat generating, and long-lived radioactivity are actually the only type of high-level radioactive waste (HLW) in Korea, which urgently requires to be disposed of in deep geological repository. Studies on disposal of HLW in various kind of geological repositories have been carried out in such countries as Sweden, Finland, United States, and etc. with their own methodologies and management policies in consideration of their situations. In Korea long-term R&D research program for safe management of SF has also been conducted during last couple of decades since around 1997, during which several various alternative type of disposal concepts for disposal of SNFs in deep geological formations have been investigated and developed. The first concept developed was KAERI Reference Disposal System (KRS) which is actually very much similar to Swedish KBS-3, a famous concept of direct disposal of SF in stable crystalline rock at a depth of around 500 m which has been regarded as one of the most plausible method worldwide. The world first Finnish repository which is expected to begin to operate sooner or later will be also this type. Since the characteristics of SF discharged from domestic nuclear reactors have been changed and improved, and burnup has sometimes increased, a more advanced deep geological repository system has been needed, KRS-HB (KRS with High Burnup SF) has been developed and in consideration of the dimensions of SNFs and the cooling period at the time point of the disposal time, KRS+, a rather improved disposal concept has also been subsequently developed which is especially focused on the efficient disposal area. Recently research has concentrated on rather advanced disposal technology focused on a safer and more economical repository system in recent view of the rapidly growing amount of accumulated SF. Especially in Korea the rock mass and the footprint area for the repository extremely limited for disposal site. Some preliminary studies to achieve rather higher efficiency repository concept for disposal of SF recently have already been emphasized. Among many possible ones for consideration of design for high-efficiency repository system, a double-layered system has been focused which is expected to maximize disposal capacity within the minimum footprint disposal area. Based on such disposal strategy a rather newly designed performance assessment methodology might be required to show long-term safety of the repository. Through the study some prerequisites for such methodological development has been being roughly checked and investigated, which covers FEP identification and pathway and scenario analyses as well as preliminary conceptual modeling for the nuclide release and transport in nearfield, far-field, and even biosphere in and around the conceptual repository system. Through the study such scenarios and models has been implemented to development of a safety assessment by utilizing GoldSim development tool for a rough quantitative comparison with existing disposal options and simple illustration purpose as well as for showing how to develop and implementation of the model to GoldSim templet.