고준위방사성폐기물 처분 Safety Case 개발을 통한 처분안전성 신뢰도 향상

백민훈,고낙열,정종태,김경수 한국방사성폐기물학회 2016 방사성폐기물학회지 Vol.14 No.4

Many countries have developed a safety case suitable to their own countries in order to improve the confidence of disposal safety in deep geological disposal of high-level radioactive waste as well as to develop a disposal program and obtain its license. This study introduces and summarizes the meaning, necessity, and development process of the safety case for radioactive waste disposal. The disposal safety is also discussed in various aspects of the safety case. In addition, the status of safety case development in the foreign countries is briefly introduced for Switzerland, Japan, the United States of America, Sweden, and Finland. The strategy for the safety case development that is being developed by KAERI is also briefly introduced. Based on the safety case, we analyze the efforts necessary to improve confidence in disposal safety for high-level radioactive waste. Considering domestic situations, we propose and discuss some implementing methods for the improvement of disposal safety, such as construction of a reliable information database, understanding of processes related to safety, reduction of uncertainties in safety assessment, communication with stakeholders, and ensuring justice and transparency. This study will contribute to the understanding of the safety case for deep geological disposal and to improving confidence in disposal safety through the development of the safety case in Korea for the disposal of high-level radioactive waste. 고준위방사성폐기물 심층처분에서 처분안전성의 신뢰도를 향상시킬 수 있는 방안으로 그리고 처분 프로그램 개발 및 인허가를 위해 많은 나라들에서 자국에 적합한 safety case를 개발하고 있다. 본 연구에서는 방사성폐기물 처분을 위한 safety case 의 의의, 필요성, 개발과정들을 정리하고 소개하였다. 그리고 처분안전성을 safety case의 다양한 측면에서 논의하였다. 아울러 스위스, 일본, 미국, 스웨덴, 핀란드 등 해외의 safety case 개발 현황과 현재 KAERI에서 개발 중인 safety case의 개발전략을 간략히 소개하였다. 고준위방사성폐기물 처분안전성의 신뢰도 향상을 위해 safety case 기반 하에서 어떤 노력들이필요한지를 분석하였다. 그리고 국내 상황을 반영하여 신뢰할 수 있는 정보자료의 구축, 안전성 관련 과정들의 이해, 안전성평가의 불확실성 저감, 이해당사자와의 의사소통, 공정성과 투명성 확보 등의 실행 방안을 제안하고 논의하였다. 본 논문에제시된 내용들은 심층처분 safety case를 이해하고, 국내에서 개발하고 있는 고준위방사성폐기물 처분 safety case 개발을 통한 처분안전성 신뢰도 향상에 기여할 수 있을 것으로 기대한다.

Confidence Improvement of Disposal Safety by Development of a Safety Case for High-Level Radioactive Waste Disposal

Min Hoon Baik,Nak-Youl Ko,Jongtae Jeong,Kyung-Su Kim 한국방사성폐기물학회 2016 방사성폐기물학회지 Vol.14 No.4

고준위방사성폐기물 심층처분에서 처분안전성의 신뢰도를 향상시킬 수 있는 방안으로 그리고 처분 프로그램 개발 및 인허가를 위해 많은 나라들에서 자국에 적합한 safety case를 개발하고 있다. 본 연구에서는 방사성폐기물 처분을 위한 safety case 의 의의, 필요성, 개발과정들을 정리하고 소개하였다. 그리고 처분안전성을 safety case의 다양한 측면에서 논의하였다. 아울러 스위스, 일본, 미국, 스웨덴, 핀란드 등 해외의 safety case 개발 현황과 현재 KAERI에서 개발 중인 safety case의 개발전략을 간략히 소개하였다. 고준위방사성폐기물 처분안전성의 신뢰도 향상을 위해 safety case 기반 하에서 어떤 노력들이필요한지를 분석하였다. 그리고 국내 상황을 반영하여 신뢰할 수 있는 정보자료의 구축, 안전성 관련 과정들의 이해, 안전성평가의 불확실성 저감, 이해당사자와의 의사소통, 공정성과 투명성 확보 등의 실행 방안을 제안하고 논의하였다. 본 논문에제시된 내용들은 심층처분 safety case를 이해하고, 국내에서 개발하고 있는 고준위방사성폐기물 처분 safety case 개발을 통한 처분안전성 신뢰도 향상에 기여할 수 있을 것으로 기대한다. Many countries have developed a safety case suitable to their own countries in order to improve the confidence of disposal safety in deep geological disposal of high-level radioactive waste as well as to develop a disposal program and obtain its license. This study introduces and summarizes the meaning, necessity, and development process of the safety case for radioactive waste disposal. The disposal safety is also discussed in various aspects of the safety case. In addition, the status of safety case development in the foreign countries is briefly introduced for Switzerland, Japan, the United States of America, Sweden, and Finland. The strategy for the safety case development that is being developed by KAERI is also briefly introduced. Based on the safety case, we analyze the efforts necessary to improve confidence in disposal safety for high-level radioactive waste. Considering domestic situations, we propose and discuss some implementing methods for the improvement of disposal safety, such as construction of a reliable information database, understanding of processes related to safety, reduction of uncertainties in safety assessment, communication with stakeholders, and ensuring justice and transparency. This study will contribute to the understanding of the safety case for deep geological disposal and to improving confidence in disposal safety through the development of the safety case in Korea for the disposal of high-level radioactive waste.

Safety Function, Performance Targets and Technical Requirements for the EBS System of an Alternative Repository Concept for SNFs in Korea

Jong-Youl Lee,Heui-Joo Choi,Dong-Keun Cho 한국방사성폐기물학회 2023 한국방사성폐기물학회 학술논문요약집 Vol.21 No.2

According to the second high-level radioactive waste management national basic plan announced in December 2021, the reference geological disposal concept for spent nuclear fuels (SNF) in Korea followed the Finnish concept based on KBS-3 type. Also, the basic plan required consideration of the development of the technical alternatives. Accordingly, Korea Atomic Energy Research Institute is conducting analyses of various alternative disposal concepts for spent nuclear fuels and is in the final selection stage of an alternative disposal concept. 10 disposal concepts including reference concept were considered for analysis in terms of disposal efficiency and safety. They were reference concept, mined deep borehole matrix, sub-seabed disposal, deep borehole disposal, multi-level disposal, space disposal, sub-sea bed disposal, long-term storage, deep horizontal borehole disposal, and ice-sheet disposal. Among them, first 4 concepts, mined deep borehole matrix, sub-seabed disposal, deep borehole disposal, multi-level disposal, were selected as candidate alternative disposal concepts by the evaluation of qualitative items. And then, by the evaluation of quantitative and qualitative items with specialists, multi-level disposal concept was being selected as a final alternative disposal concept. Design basis and performance requirements for designing alternative disposal systems were laid in the previous stage. Based on this, the design strategy and main design requirements were derived, and the engineered barrier system of a high-efficiency disposal concept was preliminary designed accordingly. In addition, as an alternative disposal concept, performance targets and related requirements were established to ensure that the high-efficiency repository system and its engineered barrier system components, such as disposal containers, buffer bentonites, and backfill perform the safety functions. Items that qualitatively describe safety functions, performance goals, and related requirements at this stage and items whose quantitative values are changed according to future test results will be determined and updated in the process of finalizing and specifically designing an alternative highefficiency disposal system.

Post-closure safety assessment of near surface disposal facilities for disused sealed radioactive sources

Lee, Seunghee,Kim, Juyoul Elsevier 2017 Nuclear engineering and design Vol.313 No.-

<P><B>Abstract</B></P> <P>Great attention has been recently paid to the post-closure safety assessment of low- and intermediate-level radioactive waste (LILW) disposal facility for disused sealed radioactive sources (DSRSs) around the world. Although the amount of volume of DSRSs generated from industry, medicine and research and education organization was relatively small compared with radioactive wastes from commercial nuclear power plants, some DSRSs can pose a significant hazard to human health due to their high activities and long half-lives, if not appropriately managed and disposed. In this study, post-closure safety assessment was carried out for DSRSs generated from 1991 to 2014 in Korea in order to ensure long-term safety of near surface disposal facilities. Two kinds of disposal options were considered, i.e., engineered vault type disposal facility and rock-cavern type disposal facility. Rock-cavern type disposal facility has been under operation in Gyeongju city, republic of Korea since August 2015 and engineered vault type disposal facility will be constructed until December 2020 in the vicinity of rock-cavern disposal facility. Assessment endpoint was individual dose to the member of critical group, which was modeled by GoldSim, which has been widely used as probabilistic risk analysis software based on Monte Carlo simulation in the area of safety assessment of radioactive waste facilities. In normal groundwater scenario, the maximum exposure dose was extremely low, approximately 1×10<SUP>−7</SUP> mSv/yr, for both disposal options and satisfied the regulatory limit of 0.1mSv/yr. However, in the drinking well scenario, the maximum exposure dose for engineered vault type disposal facility was assessed as 2.022mSv/yr where the value exceeded the regulatory limit of 1mSv/yr. The maximum exposure dose for rock-cavern type disposal facility was calculated to be 0.634mSv/yr, whose results was relatively very close to the regulatory limit considering high uncertainty of long-term environmental conditions. It was demonstrated that DSRSs including the radionuclides of <SUP>14</SUP>C, <SUP>226</SUP>Ra and <SUP>241</SUP>Am significantly contributed to the large portion of exposure dose to the public based on the long-term safety assessment. Therefore, it was recommended that the near surface disposal of DSRSs containing <SUP>14</SUP>C, <SUP>226</SUP>Ra and <SUP>241</SUP>Am should be restricted and managed by long-term interim storage option in order to minimize their potential radiological health effects.</P> <P><B>Highlights</B></P> <P> <UL> <LI> Post-closure safety assessment of near surface disposal facility for DSRS was performed. </LI> <LI> Engineered vault and rock-cavern type were considered for normal and well scenario. </LI> <LI> <SUP>14</SUP>C, <SUP>226</SUP>Ra, <SUP>241</SUP>Am were primary nuclides contributing large portion of exposure dose. </LI> <LI> Near surface disposal of DSRSs containing <SUP>14</SUP>C, <SUP>226</SUP>Ra and <SUP>241</SUP>Am should be restricted. </LI> </UL> </P>

Preliminary Safety Evaluation for Cellulose Disposal at 1st Phase Disposal Facility

Hyun Woo Song,Moonoh Kim,Sang June Park,Suil Bang 한국방사성폐기물학회 2022 한국방사성폐기물학회 학술논문요약집 Vol.20 No.1

Recently, concern regarding disposal of cellulosic material is growing as cellulose is known to produce complexing agent, isosaccharinic acid (ISA), upon degradation. ISA could enhance mobility of some radionuclides, thus increasing the amount of radionuclide released into the environment. Evaluation on the possible impact of the cellulose degradation would be an important aspect in safety evaluation. In this paper, the maximum safe disposal amount cellulose is evaluated considering the disposal environment of silos of 1st phase disposal facility. The key factor governing the impact of cellulose degradation is pH of disposal environment, as cellulose is known to degrade partially at pH above 12.5, and completely at pH above 13. Thus, disposal environment should be analyzed as to determine the extent of degradation. As silos are constructed with large amount of cement, porewater within concrete walls would be of very high pH. However, for high pH porewater to be released into the pores of crushed rock, which is filling up the silos, lower pH groundwater (commonly pH 7) should flow into the silos through the concrete walls. This causes dilution of the high pH concrete porewater, resulting in a lower pH as the silos are filled, reaching to expected pH of 11.8–12.3, which is below cellulose degradation condition. Thus, cellulose degradation is not expected, but to quantitatively evaluate safe disposal amount of cellulose, partial degradation is assumed. Upon literature review, the most conservative ISA concentration, enhancing radionuclide mobility, is determined to be 1.0×10?4 M and to reach this concentration, cellulose mass equivalent to 6wt% of cement of the repository, is required to be degraded. However, this ratio is derived based on complete degradation of cellulose into ISA, so for partial degradation, degradation ratio and yield ratio of ISA should be considered. Commonly, cellulosic material (e.g. cotton, paper, etc.) has degree of polymerization (DP) between 1,000–2,000, and with this DP, degradation ratio is estimated to be about 10%. Furthermore, yield ratio of ISA is known to be 80%. Considering all these aspects, about 1.79×107 kg of cellulose could be disposed, which if converted into number of drums, considering cellulose content of dry active waste, more than 100,000 drums (200 L) could be disposed with negligible impact on safety. Based on the result, negligible impact of cellulose degradation is expected for safety of 1st phase disposal facility. In future, this study could be used as fundamental data for revising waste acceptance criteria.

Improvement Plan for Safety Management System related to Kids Cafe

Myeong-jin Jeong,Myeonggu Lee 국제문화기술진흥원 2020 International Journal of Advanced Culture Technolo Vol.8 No.4

As the number of kids cafes, one of the children's playgrounds, is increasing rapidly, safety accidents in the kids cafe are also increasing rapidly. The facility is also increasing as the need increases, but it is spreading without ensuring safety. In particular, the Ministry of Public Administration and Security for children's play facilities in the kids cafe, and the Ministry of Culture, Sports and Tourism for organic organizations are divided into different departments, so it is not easy for local governments to manage and supervise the actual business, and there are safety blind spots. Kids cafes have changed rapidly according to needs of children and guardians who are users, and there are many problems associated with them. Therefore, we identified problems that may arise due to insufficient safety management systems for kids cafes, investigated the safety management related to kids cafes in advanced countries, and compared and analysed them with domestic systems. As a result of the research, we proposed a safety management reinforcement system, and we hope to contribute to the reduction and prevention of kids café safety accidents.

Geochemical Behavior of Cesium and Cobalt in a Near-Surface and Deep Geological Disposal Environments

Jae-Kwang Lee,Jongtae Jeong,Min-Hoon Baik 한국방사성폐기물학회 2022 한국방사성폐기물학회 학술논문요약집 Vol.20 No.2

Disposal methods of radioactive waste can be mainly divided into near-surface disposal and deep geological disposal. If the radioactive waste is exposed to groundwater for a long time in the disposal environment, no matter how the decommissioning waste generated from the nuclear power plant is disposed of, the radionuclides may be released from the disposal site. Decommissioning waste from nuclear power plant contains radionuclides that are harmful to ecosystem including humans. Radionuclides released from disposal site behave in a complex and sensitive manner affected by geochemical conditions such as soil, geological media and groundwater. Sorption is one of the important mechanisms to retard the migration of radionuclides in a subsurface environment. In this study, geochemical properties of groundwater were collected and analyzed in the disposal environment considering disposal method in order to evaluate the geochemical behavior of radionuclides. The formation of aqueous and precipitated species of cesium and cobalt in a disposal condition were calculated and estimated. The sorption properties were also evaluated and predicted by considering the changes in the geochemical conditions such as pH, redox potential and geological media for the safety assessment.

A Preliminary Study on the Development of Geological Evolution Reconstruction Methodology for Long-term Safety of the Disposal Site

Soolim Jung,Doohee Jeong,Ji-Min Choi 한국방사성폐기물학회 2022 한국방사성폐기물학회 학술논문요약집 Vol.20 No.1

Disposal facilities for radioactive waste shall be sited to provide isolation from the accessible biosphere. The features shall aim to provide this isolation for tens of thousands to a million years after closure. For the safety assessments of repository, the long-term natural evolution and possible events of the site, that can cause disturbances to the facility over the period of interest, should be considered. Geological development processes that the site have been experienced can contribute to understanding and descripting the present-day conditions. Moreover, knowledge of the past is necessary to predict the future evolution of the site. With regard to disposal site, understanding past geological evolution history allows to access the possibility of hazardous events of the site that can cause disturbances to the facility over the period of interest, and to verify the change in the geological environment is within the safe performance range even after the period of interest. In addition, certain parameters that change with the geological evolution can affect the hydrological and geochemical characteristics which are essential to disposal performance. There are various factors in the evolution of the geological environment, but not all are related to disposal safety. The objective of this research is to develop a geological reconstruction method considering factors that should be derived preferentially for the geological characteristics of the disposal site and the evaluation of the long-term safety. As a preliminary study on this, we investigated case studies related to geological reconstruction of overseas disposal research institutes, and reviewed which factors are suitable for the domestic granitoid distribution environment. It is expected that systematic and consistent results will be possible in the future through this methodology.

Preliminary Safety Assessment of Disposal of Glass Fiber Waste

Woo Yong Kim,Moon Oh Kim,Seung Won Hwang,Sang June Park,Bang Suil 한국방사성폐기물학회 2022 한국방사성폐기물학회 학술논문요약집 Vol.20 No.1

In a recent preliminary inspection for disposal, the glass fiber waste (GFW), used as a pipe insulation, was judged as “pending evaluation” because some dust was found in drum opening tests. Therefore, additional inspection is required to ensure that the package corresponds with the acceptance criteria of the particulates. The dust was generated presumably due to GFW being used in a high-temperature environment for a long time, thus being easily degraded and crushed. For this reason, safety issues that may occur in the process of handling, transportation, and disposal are emerging. Therefore, in this study, a preliminary safety assessment of GFW disposal was performed, the exposure dose to the general public was derived, and compared with the dose limit. The evaluation was carried out in the following order: (1) evaluation of GFW radiation source term, (2) selection of accident scenario, (3) calculation of exposure dose, (4) comparison of evaluation results with dose limits, and confirmation of satisfaction. The average radioactivity of the GFW to be disposed of was used as the source term, and the main nuclides were identified as H-3, Fe-55, Co-60, Ni-63, and Pu-241. In general, the types of accidents that can occur at disposal facilities can be classified into falls, fires, collisions during transportation, off-site accidents, and nuclear criticality, and the accident scenarios are selected by analyzing and reviewing the probability of each accident. In this study, the accident analysis and scenarios presented in the safety assessment of the KORAD were reviewed, and the fire in the treatment facility, the fire in the storage facility, and the collision of the transport vehicle were selected as the evaluation scenarios. When an accident occurs, the radioactive material inside the container leaks out and diffuses into the atmosphere. In this evaluation, the internal and external exposure of the general public due to radioactive plume at the site boundary was evaluated and the dose conversion factors from ICRP-72 and FGR 12 were used. Based on the evaluation, general public was exposed to 0.004 mSv, 0.013 mSv, and 0.045 mSv, respectively, due to a fire at a treatment facility, at a storage facility, and in a transport vehicle. Most of the dose is due to internal exposure by Pu-241 nuclide, because the proportion of it in the waste is high, and when inhaled, the internal dose is high by emitting beta rays. It was confirmed that the result of dose was 0.4%, 1.3% and 4.5% of the annual dose limit, sufficiently satisfying the dose limit and safety.

Study on the 3D Site Descriptive Model for the Wolseong LILW Disposal Center’s Site Characteristics Evaluation

( Soo-gin Kim ),( Jae-yeol Cheong ),( Hyun-jin Cho ) 대한지질공학회 2019 대한지질공학회 학술발표회논문집 Vol.2019 No.2

The most important factor of the radioactive waste disposal is to select a suitable geological site and implement a disposal system that meets safety, technical and environmental requirements. The system may slightly depend on the geological conditions and the type of waste in each country, but the ultimate goal is to secure long-term safety. Thus, it is very important to understand the characteristics of the site, before designing to maintain the performance of the disposal system for a long time. The Wolseong Low and Intermediate Level Radioactive waste (hereinafter referred to as the 'LILW') Disposal Center (hereinafter referred to as the 'study area') is a licensed facility for safe management of LILW in KOREA. Isolate LILW to the underground disposal repository consisting of multiple barriers for a long period to prevent leakage into ecosystems and radiation effects on the human body. The study area is the world's first complex LILW disposal facility to be operated within the same site with the gradual development of underground silo, surface and landfill disposal type. This study analyzes geological features and predicts future’s natural phenomena such as groundwater flow by building the 3D site descriptive model (SDM) in the study area in order to enhance the safety of radioactive waste disposal facilities.