Glass fiber (GF) insulation is a non-combustible material, light, easy to transport/store, and has excellent thermal insulation performance, so it has been widely used in the piping of nuclear power plants. However, if the GF insulation is exposed to a high-temperature environment for a long period of time, there is a possibility that it may be crushed even with a small impact due to deterioration phenomenon and take the form of small particles. In fact, GF dust was generated in some of the insulation waste generated during the maintenance process. In the previous study, the disposal safety assessment of GF waste was performed under the abnormal condition of the disposal facility to calculate the radiation exposure dose of the public residing/ residents nearby facilities, and then the disposal safety of GF waste was verified by confirming that the exposure dose was less than the limit. However, the revised guidelines for safety assessment require the addition of exposure dose assessment of workers. Therefore, in this study, accident scenarios at disposal facilities were derived and the exposure dose to the workers during the accident was evaluated. The evaluation was carried out in the following order: (1) selection of accident scenario, (2) calculation of exposure dose, (3) comparison of evaluation results with dose limits, and confirmation of satisfaction. The representative accident scenarios with the highest risk among the facility accident were selected as; (a) the fire in the treatment facility, (b) the fire in the storage facility, and (c) fire after a collision of transport vehicles. The internal and external exposure doses of the worker by radioactive plume were calculated at 10m away from the accident point. In evaluation, the dose conversion factors ICRP-72 and FGR12 were used. As a result of the calculation, the exposure dose to workers was derived as about 0.08 mSv, 0.20 mSv, and 0.10 mSv, due to fire accidents (vehicle collision, storage facilities, treatment facilities). These were 0.2%, 0.4%, and 0.2% of the limit, and the radiation risk to workers was evaluated to be very low. The results of this study will be used as basic data to prove the safety of the disposal of GF waste. The sensitivity analysis will be performed by changing the radiation source and emission rate in the future.

Probiotic strain is known to regulate the immune system by colonizing in the intestine and interact with intestinal cell receptors of lymphoid tissue. In this study, safety of Streptococcus thermophilus KCTC14471BP and silk fibroin coating effects was evaluated with respect to mucin binding abilities and immune system modulation. S. thermophilus KCTC14471BP was coated with silk fibroin by adding 1% water-soluble calcium and 0.1% silk fibroin. S. thermophilus KCTC14471BP showed the high activities of leucine arylamidase and β-galactosidase. Regarding the antibiotic resistance tests, S. thermophilus KCTC14471BP was susceptible to ampicillin, vancomycin, gentamicin, kanamycin, streptomycin, erythromycin, clindamycin, tetracycline, and chloramphenicol. S. thermophilus KCTC14471BP coated with silk fibroin showed the improved mucin binding ability from 16.1% to 71.3% and was confirmed to have no cytotoxicity against RAW 264.7 macrophage. S. thermophilus KCTC14471BP coated with silk fibroin showed dose-dependently significant increases in pro-inflammatory cytokines IL-1β, IL-6, and TNF-α. These results suggested that S. thermophilus KCTC14471BP can be expected as a promising probiotic bacteria for applications to food industries such as fermentation or functional foods.

To obtain confidence in the safety of disposal facilities for radioactive waste, it is essential to quantitatively evaluate the performance of the waste disposal facilities by using safety assessment models. Thus, safety assessment models require uncertainty management as a key part of the confidencebuilding process. In application to the numerical modelling, the global sensitivity analysis is widely employed for dealing with parametric and conceptual uncertainties. In particular, the parametric uncertainty can be effectively reduced by minimizing the uncertainty of critical parameters in the safety assessment model. In this paper, the numerical model of each step disposal facility (Silo, Near surface, and Trench type) at Wolsong Low and Immediate Level Waste (LILW) Disposal Center is designed by using a two-dimensional finite element code (COMSOL Multiphysics). In order to determine the critical parameters for non-adsorbed nuclides such as H-3, C-14, Tc-99, we introduced the variance-based sensitivity analysis methodology of the global sensitivity analysis. In the case of Silo type, the density of waste is highly sensitive to the total leakage quantity of all nuclides. Additionally, the initial nuclide concentration of H-3 was identified as another important parameter of H-3. On the other hands, the mass transport coefficient showed a high contribution in C-14 and Tc-99. In other types of disposal facilities, the leaking properties of H-3 are significantly affected by the amount of infiltration water. However, C-14 and Tc-99 were found to be more sensitive to the density of waste.

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.

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. Thus, evaluation on the possible impact of the cellulose degradation would be an important aspect in safety evaluation. In this paper, safety assessments conducted in Sweden and UK are studied, and the factors required to be considered for appropriate safety assessment of cellulose is analyzed. SKB (Sweden) conducted safety assessment of cellulose degradation as a part of long-term safety assessment of SFR. SKB determined that ISA would impact sorption of trivalent and tetravalent radionuclides (Eu, Am, Th, Np, Pa, Pu, U, Tc, Zr and Nb) at concentration higher than 10−4–10−3 M, and impact sorption of divalent radionuclides (Ni, Co, Fe, Be and Pb) at concentration higher than 10−2 M. Then, SKB conservatively set the upper limit of ISA concentration to be 10−4 M and conducted cellulose degradation evaluation on each waste package type, considering the expected disposal environment of SFR. Based on the calculated results, some of the waste packages showed concentration of ISA to be higher than 10−4 M, so SKB conservatively developed waste acceptance criteria to prevent ISA being produced to an extent of affecting the safety of the repository. SKB conducted safety assessment only for the repositories with pH above 12.5 and excluded 1BLA from the safety assessment as the expected pH of 1BLA is around 12, which is insufficient for cellulose to degrade. However, SKB set disposal limit for 1BLA as well, to minimize potential impact in future. Serco (UK) conducted safety assessment of cellulose degradation for the conceptual repository, which is a concrete vault with cementitious backfill. Serco estimated that the pH of repository would maintain around 12.4. Serco conservatively assumed that the pH would be sufficient for cellulose degradation to occur partially, and suggested application of appropriate degradation ratio for safety assessment of cellulose degradation. To conduct appropriate safety assessment of cellulose degradation, an appropriate ISA concentration limit based on radionuclide inventory list, and an appropriate cellulose degradation ratio based on the pH of disposal environment should be determined. As for guidance, below pH 12.5, cellulose degradation is not expected, and between pH 12.5–13, partial cellulose degradation is expected. In future, this study could be used as fundamental data to evaluate safety of the repository.

An objective of a safety assessment for geological disposal is to evaluate the radiological impact by radionuclides release from radioactive wastes. Computational estimation of all radionuclides transport in the disposal system, however, is not neccessary because some radionuclides has negligible effect on radiological doses. For this reason, prioritization of radionuclides list is preceded before the safety assessment. The Korea Atomic Energy Research Institue (KAERI) has assessed the long-term safety of a disposal system for spent nculear fuels. Currently, thirty eight radionuclides and twenty three elements are considered in the safety assessment activity of the KAERI. Nevertheless, a screening process for radionulides selection has not been articulated yet. In this study, we reviewed radionuclides selection process in forign countries to re-establish screening criteria for the KAERI’s radionuclides list. Screeing models of the Swedish Nuclear Fuel and Waste Management Company (SKB), the Deparment of Eenrgy (US DOE), and the Japan Nuclear Cycle Development Istitute (JNC) were compared. We found that each country developed different screening model depending on scenarios of radionuclides release. Nonetheless, there were common properties that determines the importance of radionuclides. These properties for radionuclides include halflife, radiotoxicity (or specific activity), and mobility in underground medium. Based on the review results, we proposed radionuclides selection process to prioritize the importance of radionucldies in the KAERI safety assessment.

The timescale for the post-closure safety assessment of a deep geological repository ranges from ten thousand to a million year. In such a long period of time, the biosphere inevitably undergoes changes. Therefore, the long-term evolution of a biosphere is recognized as an important issue in the post-closure safety assessment of a deep geological repository for spent fuels. In this study, we reviewed the approaches to address the long-term evolution of a biosphere. The major drivers of longterm evolution of a biosphere are the climate change and the resulting landscape development. They can affect the hydrogeological and hydrogeochemical characteristics of a biosphere, and then the radionuclide migration through the biosphere followed by the exposure doses for the critical groups. In addition, human activities and the social developments can affect the climate change resulting in the long-term evolution of a biosphere. To make a biosphere assessment, the long-term evolution scenarios for the biosphere should be formulated considering these climate change, landscape development, and human activities. In addition, features, events, and processes (FEPs) that affect the long-term evolution of a biosphere should be used. According to the Safety Case reports of Finland, the major long-term evolution scenario drivers of a biosphere are local sea-level change due to climate change and land use related to crop type, irrigation procedures, livestock, forest management, construction of a well, and demographics. The climate change causing the local sea-level change can be simulated using various earth system models such as CLIMBER-2, MPI/UW, and UVic and an icesheet model such as SICOPOLIS. The review results of this study and FEPs related to the climate change, the landscape development, and human activities will be used to formulate long-term evolution scenarios for the safety assessment of a deep geological repository for spent fuels.

In South Korea, the master plan for high-level radioactive waste management, announced in 2016, suggested the construction and operation of intermediate storage facilities on a permanent disposal site and specified the adoption of dry storage in consideration of the ease of operation and expansion. As of 2021, the government is again reviewing its overarching policy on the back-end fuel cycles, including intermediate storage and permanent disposal. In the case of dry storage facilities, safety evaluation is being conducted using a combination of deterministic and probabilistic approaches, similar to that of nuclear power plants. The two methods are complementary, of which Probabilistic Safety Assessment (PSA) has the advantage of being able to identify key scenarios affecting safety, but its use in storage facilities has not been highlighted so far. However, depending on the spent fuel management phases such as loading, transportation, and storage, it may be not enough to capture effective and efficient safety evaluation only deterministically, and probabilistic methods may contribute to the evaluation of long-term operation or external events such as an earthquake. There have already been cases where PSA has been performed on a part of the nuclear fuel cycle through previous studies. This paper created the safety assessment model based on open sources such as the released EPRI reports, by targeting arbitrary intermediate storage facilities. The model considered the scenarios for loading, transportation, and storage, with human error respectively. It will be able to be modified and improved to fit domestic and specific intermediate storage facilities in the future.

The decommissioning of a nuclear power plant generates large amounts of radioactive waste, which is of several types. Radioactive concrete powder is classified as low-level waste, which can be disposed of in a landfill. However, its safe disposal in a landfill requires that it be immobilized by solidification using cement. Herein, a safety assessment on the disposal of solidified radioactive concrete powder waste in a conceptual landfill site is performed using RESRAD. Furthermore, sensitivity analyses of certain selected input parameters are conducted to investigate their impact on exposure doses. The exposure doses are estimated, and the relative impact of each pathway on them during the disposal of this waste is assessed. The results of this study can be used to obtain information for designing a landfill site for the safe disposal of low-level radioactive waste generated from the decommissioning of a nuclear power plant.

정부(해양수산부) 및 관계기관에서는 안전관리 취약선박의 해양안전 증진을 위한 많은 정책을 시행하고 있으나, 선박의 안전관리에 대한 위험성평가 부재로 각 제도의 효과성에 다소 아쉬움이 확인되었다. 선박의 위험성평가에 대한 국내 선행연구는 대부분 개별적 해양사고 위험성에 대한 단편적 위험성만을 취급하고 있어 포괄적 해양사고 위험성평가를 위한 새로운 방식이 필요하다. 선박의 안전관 리에 대한 포괄적 위험성평가 적용사례로 항만국통제와 노르웨이 해사국의 점검대상 선정방법이 있다. 본 연구에서는 국내 해양사고 및 등록현황의 높은 비중을 차지하고 있는 연근해어선을 대상으로 포괄적 선박안전관리 위험성평가를 위한 지표를 개발하고 적용하였다. 그 결과, 각 선박별 위험성을 계량적 수치로 확인하였고, 개별적 안전관리 취약요소의 순위를 식별하였다.

In this study, algorithms for analyzing the torsion of buildings under earthquake excitation are developed. The algorithm and formulations to account for the torsional angle are verified by analyzing the seismic acceleration time history data. The method was applied to the reference buildings to examine their operation and usability. The reference application demonstrated that the noise-canceling scheme successfully overcame various obstacles in the field measurements. The developed method is expected to be used as a tool to support a loss assessment system for determining the direction and priority of disaster response in the event of an earthquake.

유방조직 확장기(BTE)는 MR 금기로 분류되고 있으나 이전 연구들에서 제한적인 조건에서 MRI 검사가 가능함을 발표하였다. 본 연구는 미국재료시험 협회에서 권고하는 변위력과 토크뿐 만 아니라 뒤틀림을 측정하여 MR 안전성을 평가하였다. 3.0 T MRI 장치와 350 ㎖, 650 ㎖ BTE를 사용하였다. 변위력은 자기장이 가장 큰 지점에서 측정 기구에 BTE를 매달고 발생된 최대 변위각을 3회씩 측정하였다. 토크는 원형 용기에 생리식염수를 단계적으로 채운 BTE를 놓고 자기장에 의해 발생된 회전력을 45˚ 간격에서 정성적으로 측정하였다. 뒤틀림은 생리식염수를 최대로 채운 뒤 자성 밸브의 뒤틀림 각도를 포켓 경사계를 이용하여 두부, 어깨, 유방, 간, 골반이 중심일 때 유방 위치에서 각각 측정하였다. 측정 결과 변위각 은 350 ㎖의 경우 0 ㏄, 150 ㏄에서 각각 125˚, 54.3˚, 650 ㎖은 0 ㏄에서만 123.3˚로 MR 금기 기준을 초과했다. 토크는 350 ㎖의 경우 모두 +4로 측정되었고, 650 ㎖의 경우 450 ㏄ 생리식염수 미만에서는 모두 +4, 550 ㏄ 이상에서는 +1∼+4로 나타났으며, 일부를 조건을 제외하고 모두 MR 금기 조건을 초과했다. 뒤틀림은 350 ㎖의 경우 5˚∼80˚ 를 나타냈으며. 650 ㎖는 15˚∼90˚를 나타내어 환자에게 통증 유발 가능성을 확인했다. BTE의 변위각, 토크, 뒤틀림을 측정하여 MR 금기로 평가하였으며, 어떠한 조건하에서도 MR 환경에서 환자의 안전을 보장할 수 없다.