The MBR process, which uses membrane to separate solids instead of secondary clarifier, has the advantage of maintaining high MLSS (Mixed Liquor Suspended Solids) concentration in bioreactors. In this study, three MBR processes combined with chemical phosphorus removal facility were studied: the phase-isolated MBR process with one recirculation pump, the A2O MBR process with two recirculation pumps, and the MBR process without a oxygen depletion reactor with three recirculation pumps. 116 simulations by EQPS (Effluent Quality Prediction System) were performed, under same design influent characteristics and hydraulic retention time, to see the effects of recirculation ratio changes on effluent water quality. The final effluent limits for total nitrogen and phosphorus were set to as 10 ㎎/L or less and 0.2 ㎎/L or less, respectively. AlCl3 was fed at MBR effluent to make 0.2 ㎎/L of total phosphorus precisely by PI (Proportional-Integral) controller in EQPS. This study showed that the phase isolated MBR process (R1 600%) had the highest nitrogen removal efficiency, with a final effluent T-N concentration of 6.765 mg/L. However, the A2O MBR process (R1 400%, R2 100%) required the lowest AlCl3 flow of 0.742 m3/day, which is approximately 59.1% lower than the average AlCl3 flow of the phase-isolated MBR process. It also produced 4,835 kg/day of sludge, the lowest among the studying MBR processes. It seems that A2O MBR process is the most economic method to treat studying wastewater to meet final effluent nutrient limit targets. However, carbon neutrality must be considered to select the best process to treat studying wastewater and it will be presented later.

고속도로에 적용되고 있는 소음저감대책 현황을 고찰하기 위해서 고속도로 건설 전 수행해야 하는 환경영향평가서와 고속도로 완공 후 소음기준 초과 소음에 대한 소음영향분석보고서를 검토하였다. 환경영향평가서는 시기별로 수립되는 방음대책이 변화하였다. 2013 년 전까지는 방음벽(흡음형, 반사형) 위주의 방음대책이 수립되었다. 그러나 2015년 이후 방음벽 이외의 배수성저소음포장과 소음감쇠 기를 저감방안으로 제시하였다. 그러나 보고서마다 소음저감효과가 서술되어 있거난 서술되어 있지 않았다. 환경영향평가 이후 신규소 음저감대책이 수록된 소음영향분석보고서에서 최근 5년간 보고서 일부를 중심으로 검토하였다. 소음대책 수립을 위해서 20년 후 장래 교통량을 이용하여 방음벽 신설, 방음터널, 소음감쇠기, 배수성저소음포장 등의 소음저감대책을 수립하였으며 방음벽과 배수성저소음 포장을 조합한 대책안이 많이 제시되었다.

본 연구는 축산시설 내 설치된 무창기공형 집열기의 배기 방향 및 유량 변화를 통해 벽체에 전달되는 일사를 차단, 이를 통한 냉방효과를 검증하려는 목적의 기초 연구로서 무창기공형 집열기 시험장치를 제작, 배기 유량 변화에 따른 위치별 온도 변화 및 이를 통한 열성능 평가를 수행하였다. 실험 결과, 무창기공형 집열기의 유량조건별 집열판 표면온도는 최고 27.7℃, 배기온도는 최고 약 10.9℃ 온도 차이를 확인하였다. 무창기공형 집열기의 유량조건별 열교환 유효도는 0.48∼0.62, 효율은 30%∼90%의 분포로 나타났다. 집열판 에너지는 유량이 증가함에 따라 감소, 집열기 내부 에너지는 유량이 증가함에 따라 증가하였다. 이를 통해 농업시설 외벽에 설치된 무창기공형 집열기의 여름철 미운용으로 인한 집열판 및 내부 온도상승과 이로 인한 벽체로의 열전달 등 무창기공형 집열기로 인한 역효과를 방지할 수 있을 뿐만 아니라 집열기 외부로의 강제 배기를 통해 벽체로 직접 투입되는 일사 차단을 통한 냉방효과 또한 구현할 수 있을 것으로 판단된다.

The railroad facilities are intended for long-term operation as the initial acquisition costs necessary for infrastructure construction are high. Therefore, regular maintenance of railroad facilities is essential, and furthermore, system reliability through systematic performance evaluation is required. In this study, the signal control system of railroad electrical equipment was selected as the subject of research and the performance evaluation target facility selection study was conducted using AHP. The results of the study can contribute to the reliability of the signal control system as well as to the reliability of the railroad system, which is a higher system.

PURPOSES : In this study, the luminance of night road markings was measured in a tunnel of length 200 m or less. The purpose of the project is to evaluate the consistency of night road markings. METHODS : In this study, field measurements were conducted to achieve the purpose of this study. Five tunnels with lengths of less than 200 m were selected to measure the luminance value of the road markings. The analysis of the difference in road markings between the inside and outside points of the tunnel and the analysis of alternative tunnels and points were used to assess the consistency of road markings in tunnels. RESULTS : The average luminance of the tunnel’s night road markings was 9.7, and the standard deviation was 3.0. The analysis of variance for the tunnel and point indicated that the p-value was less than 0.05 and was inconsistent. CONCLUSIONS : In conclusion, consistency was assessed by measuring the luminance value of the short tunnel of length 200 m in the Cheongju Sangju Expressway, and it was confirmed that the luminance of the road markings was not consistent with the tunnel and point. Finally, it is necessary to control night lightings on roads outside the tunnel or adjust lighting facilities in the tunnel to enhance the consistency of luminance.

3세대 방사광가속기는 기초 및 응용과학분야의 최첨단연구에 활용하여 우리나 라 기초과학 연구의 활성화에 기여한다는 목적으로 운영되고 있으며, 그 구축 및 운영을 위 해 2016년까지 30년에 가까운 기간 동안 총 7,879억 원 가량이 투입된 거대연구시설이다. 3세 대 방사광가속기 정책과 관련하여 가속기의 필요성이나 효과성에 대해 다양한 평가가 있어 왔지만 대체로 현재 시점의 성공적 결과를 강조하며 향후 정부의 지속적 지원 필요성을 피력 하거나 정책적 개선방안을 포괄적인 차원에서 제시하는 연구들이 주를 이루며, 이 사업이 시 기별로 어떠한 정책적 평가를 받았는지 살펴보고 그 결과가 어디에서 기인하는지 체계적으 로 분석한 연구는 전무하다. 본 연구에서는 3세대 방사광가속기 사업을 시기별로 구분하여 평가하되, 분석기준으로 정책활동의 다섯 가지 기준을 고려하였다. 장기간 진행되는 대형연 구시설 정책을 시기별로 나누어 살펴봄으로써 방사광가속기 및 유사 대형연구시설의 구축과 운영에 있어서 요구되는 정책활동의 원리와 관련된 정책적 시사점을 제시하고자 하였다.

본 연구는 소각시설을 규모별, 유형별로 평가하기 위하여 이행지표를 설정한 후 설문조사를 실시하였다. 설문대상 수는 2012년 기준 전국 184개 중 약 15%인 28개를 선정하였다. 규모별로 대규모 (≥100 t/d), 중규모(50~<100 t/d), 소규모(<50 t/d)로 나누고, 유형별로는 스토커, 열분해, 유동층으로 분류하였다. 이행지표는 기술성, 경제성, 환경성으로 대별한 후 각각 6개, 10개, 30개의 세부항목을 설정 하고 5등급으로 나누어 점수를 부과하였다. 평가결과 소각시설 규모별로는 기술성, 경제성, 환경성 평가 에서 모두 대규모 시설이 소규모 시설 보다 높은 점수를 받았으며, 종합점수에서도 동일한 경향을 나타 내어 소각시설의 규모가 증가할수록 더 높은 평가 점수를 받는 것을 알 수 있었다. 소각시설 유형별로 는 기술성, 경제성, 환경성을 종합 평가한 결과에서 스토커식(65.3점)이 유동상(59.0점) 및 열분해(58.3 점) 보다 우수한 평가결과를 나타내었다. 특히 소각시설 평가시 기술성, 경제성, 환경성 중 경제성에서 뚜렷한 차이를 나타내었다.

유한요소해석 코드인 ANSYS를 이용, 폭 5.6m간이 느타리재배사를 3차원 강뼈대 구조물로 모델링하여 베드기둥 형태 및 파이프 규격에 따른 구조적 안전성을 분석하였으며 허용응력 설계법에 기초해 구조안전 여부를 판단하였다. 전산구조 해석 결과, 베드기둥에 따른 구조적 안전성은 안전적설심의 경우, 직립형 베드기둥 구조가 다른 베드기둥 형태보다 훨씬 높게 나타났으나 안전풍속 측면에서는 형태별 차이가 없는 것으로 나타났다. 서까래 규격과 베드기둥 설치 간격에 따른 구조적 안전성은 본 연구에서 고려한 파이프 규격 범위에 있어서 안전풍속 측면의 경우, 베드기둥 설치간격보다 시설 외부적 요소인 서까래 간격에 더 영향을 받는 것으로 나타났으나 안전적설심 측면에서는 안전 풍속과는 반대로 서까래 간격보다는 시설 내부적 구조물인 베드기둥 간격이 더 중요한 것으로 나타났다. 베드기둥의 좌굴에 대한 안전성은 해석의 모든 경우에서 안전한 것으로 나타났다.

The Republic of Korea (ROK), as a member state of the IAEA, is operating the State’s System of Accounting for and Control (SSAC) and conducting independent national inspections. Furthermore, an evaluation methodology for the material unaccounted for (MUF) is being developed in ROK to enhance capabilities of national inspection. Generally, physical and chemical changes of nuclear material are unavoidable due to the operating system and structure of facilities, an accumulation of material unaccounted for (MUF) has been issued. IAEA developed statistical MUF evaluation method that can be applied to all facilities around the world and it mainly focuses on the diversion detection of nuclear materials in facilities. However, in terms of the national safeguard inspection, an evaluation of accountancy in facilities is additionally needed. Therefore, in this research, a new approach to MUF evaluation is suggested, based on the Guide to the Expression of Uncertainty in Measurement (GUM) that an evaluation of measurement uncertainty factors is straightforward. A hypothetical list of inventory items (LII) which has 6,118 items at the beginning and end of the material balance period, along with 360 inflow and outflow nuclear material items at a virtual fuel fabrication plant was employed for both the conventional IAEA MUF evaluation method and the proposed GUM-based method. To calculate the measurement uncertainty, it was assumed that an electronic balance, gravimetry, and a thermal ionization mass spectrometer were used for a measurement of the mass, concentration, and enrichment of 235U, respectively. Additionally, it was considered that independent and correlated uncertainty factors were defined as random factors and systematic factors for the ease of uncertainty propagation by the GUM. The total MUF uncertainties of IAEA (σMUF) and GUM (uMUF) method were 37.951 and 36.692 kg, respectively, under the aforementioned assumptions. The difference is low, it was demonstrated that the GUM method is applicable to the MUF evaluation. The IAEA method demonstrated its applicability to all nuclear facilities, but its calculated errors exhibited low traceability due to its simplification. In contrast, the calculated uncertainty based on the GUM method exhibited high reliability and traceability, as it allows for individual management of measurement uncertainty based on the facility’s accounting information. Consequently, the application of the GUM approach could offer more benefits than the conventional IAEA method in cases of national safeguard inspections where factor analysis is required for MUF assessment.

Evaluating the effectiveness of the radiation protection measures deployed at the Centralized Radioactive Waste Management Facility in Ghana is pivotal to guaranteeing the safety of personnel, public and the environment, thus the need for this study. RadiagemTM 2000 was used in measuring the dose rate of the facility whilst the personal radiation exposure of the personnel from 2011 to 2022 was measured from the thermoluminescent dosimeter badges using Harshaw 6600 Plus Automated TLD Reader. The decay store containing scrap metals from dismantled disused sealed radioactive sources (DSRS), and low-level wastes measured the highest dose rate of 1.06 ± 0.92 μSv·h−1. The range of the mean annual average personnel dose equivalent is 0.41–2.07 mSv. The annual effective doses are below the ICRP limit of 20 mSv. From the multivariate principal component analysis biplot, all the personal dose equivalent formed a cluster, and the cluster is mostly influenced by the radiological data from the outer wall surface of the facility where no DSRS are stored. The personal dose equivalents are not primarily due to the radiation exposures of staff during operations with DSRS at the facility but can be attributed to environmental radiation, thus the current radiation protection measures at the Facility can be deemed as effective.

Decommissioning of nuclear power plants generates a large amount of radioactive waste in a short period. Moreover, Radioactive waste has various forms including a large volumes of metal, concrete, and solid waste. The disposal of decommissioning waste using 200 L drums is inefficient in terms of economics, work efficiency, and radiation safety. Therefore, The Korea Radioactive Waste Agency is developing large containers for the packaging, transportation, and disposal of decommissioning waste. Assessing disposability considering the characteristics of the radioactive waste and facility, convenience of operation, and safety of workers is necessary. In this study, the exposure dose rate of workers during the disposal of new containers was evaluated using Monte Carlo N-Particle Transport code. Six normal and four abnormal scenarios were derived for the assessment of the dose rate in a near surface disposal facility operation. The results showed that the calculated dose rates in all normal scenarios were lower than the direct exposure dose limitation of workers in the safety analysis report. In abnormal scenarios, the work hours with dose rates below 20 mSv·y−1 were calculated. The results of this study will be useful in establishing the optimal radiation work conditions.

In Korea, Kori Unit 1, a commercial pressurized water reactor (PWR), was permanently shut down in June 2017, and an immediate decommissioning strategy is underway. Therefore, it is essential to understand the characteristics of radioactive waste during the decommissioning process of nuclear power plants (NPP). Because radioactive waste must be handled with care, radioactive waste is treated in a hot cell facility. Hot cell facility handles radioactive waste, and worker safety is essential. In this study, it was dealt with whether or not the radiation safety regulations were satisfied when processing the core beltline metal of the dismantling waste treated at the post irradiation examination facility (PIEF) of the hot cell facility. Core beltline metal used for the pressure vessel in the reactor is carbon steel, and it is continuously irradiated by neutrons during the operation of the NPP. A radiological safety estimation of the behavior of radioactive aerosols during the cutting process within the PIEF was carried out to ensure the safety of the environment and workers. When processing the core beltline metal in PIEF, dominant six nuclides (60Co, 63Ni, 55Fe, 3H, 59Ni, 14C) of aerosol are generated. Accordingly each cutting device, amount of aerosol and value of dose is different. Using a 99.97% efficiency HEPA filter, the emission concentration of the dominant nuclides (60Co, 63Ni, 55Fe, 3H, 59Ni, 14C) in the air source term was satisfied with the emission control standard of Nuclear Safety Commission No. 2016-16. It was confirmed that the radioactivity concentration in the airborne source term inside the PIEF is in equilibrium state, when ventilation is considered. Also, the mass of aerosol and the concentration of airborne source term differed according to the thickness of the saw blade of the cutting tool, and the exposure dose of the worker was different through Monte Carlo N-Particle (MCNP). At that time, 60Co accounted for 95.4% of the exposure dose, showing that 60Co had the highest impact on workers, followed by 55Fe with 2.7%. The worker’s dose limit is satisfied in accordance with Article 2 of the Nuclear Safety Act and the dose limit of radiation-controlled area is found to be satisfied in accordance with Article 3 of the rules on technical standards for radiation safety management at this time.

Material balance evaluation is an important measure to determine whether or not nuclear material is diverted. A prototype code to evaluate material balance has been developed for uranium fuel fabrication facility. However, it is difficult to analyze the code’s functionality and performance because the utilization of real facility data related to material balance evaluation is very limited. It is also restricted to deliberately implement various abnormal situations based on real facility data, such as nuclear diversion condition. In this study, process flow simulator of uranium fuel fabrication facility has been developed to produce various process data required for material balance evaluation. The process flow simulator was developed on the basis of the Simulink-SimEvents framework of the MathWorks. This framework is suitable for batch-based process modeling like uranium fuel fabrication facility. It dynamically simulates the movement of nuclear material according to the time function and provides process data such as nuclear material amount at inputs, outputs, and inventories required for Material Unaccounted For (MUF) and MUF uncertainty calculation. The process flow simulator code provides these data to the material balance evaluation code. And then the material balance evaluation code calculates MUF and MUF uncertainty to evaluate whether or not nuclear material is diverted. The process flow simulator code can simulate the movement of nuclear material for any abnormal situation which is difficult to implement with real process data. This code is expected to contribute to checking and improving the functionality and performance of the prototype code of material balance evaluation by simulating process data for various operation scenarios.

The Republic of Korea is implementing safeguards for domestic nuclear facilities through cooperation with the IAEA. But it is not to evaluate the material balance for the material unaccounted for, MUF in the bulk handling facility. Although the development of a material balance evaluation program is underway, there are no related regulations. The State Regulatory Authority, SRA is performing material balance evaluation, MBE on the facility based on the design information and material balance results of the facility. However, it is not possible to directly derive measurement uncertainty for the facility’s measurement equipment, which is an important variable of MBE. To solve this problem, it is trying to derive a method suitable for the domestic environment by investigating the some measurement uncertainty estimation methods and analyzing characteristics of them. In this study, the traditional measurement uncertainty estimation method, GUM method and GUM-S1 method were studied and the advantages and disadvantages were analyzed. Due to the problems mentioned above, the uncertainty quantification technique currently being used cannot be applied to the evaluation of the domestic material balance. Therefore, we are tying to apply them to the evaluation the domestic material balance through the above three methods or a combination of them appropriately. Through this continuing study, it is expected that it will be possible to present a plan to derive measurement uncertainty optimized for the domestic MBE environment.

The dose was evaluated for the workers transporting the spent resin drums from a spent resin mixture treatment facility. The treatment technology of spent resin mixture waste based on microwave was developed to compensate for the shortcoming of the existing one. The mechanism of the facility for the treatment is divided into separation, desorption, condensation and adsorption process. The treated spent resin that has passed through the microwave reactor flows into the spent resin storage tank. As the treatment time elapses, if spent resin accumulates in the spent resin storage tank, it is moved to the drum of the volume of 200 L. The drum must be moved by the worker, in which case radiation exposure to the drum transport worker occurs. It requires the dose evaluation for drum transport workers in terms of radiation safety. Dose evaluation was performed in consideration of the change in the composition ratio and weight of the spent resin mixture, where the working time for transportation was considered from 10 to 120 minutes in 10-minute increment. In the case of 100 kg of the spent resin mixture, the dose range was derived as 4.62×10−3 – 5.90×10−2 mSv for the 100 kg of spent resin, 4.72×10−3– 5.58×10−2 mSv for the 80 kg of spent resin and 20 kg of zeolite and activated carbon, and 5.38×10−3 – 6.32×10−2 mSv for the 60 kg of spent resin and 40 kg of zeolite and activated carbon. In the case of 150 kg of the spent resin mixture, the dose range was derived as 6.83×10−3 – 8.20×10−2 mSv for the 150 kg of spent resin, 7.13×10−3 – 8.22×10−2 mSv for the 120 kg of spent resin and 30 kg of zeolite and activated carbon, and 8.28×10−3 – 8.86×10−2 mSv for the 90 kg of spent resin and 60 kg of zeolite and activated carbon. The estimated maximum doses for each weight (100 kg and 150 kg of mixture) were confirmed to be 3.16×10−1% and 4.43×10−1% of the annual average dose limit of 20 mSv for radiation workers.

In nuclear power plants and nuclear facilities, radioactive waste containing hazardous substances (Mixed waste) is continuously generated due to research such as radiochemical study and nuclide analysis. In addition, radioactive waste including heavy metals and asbestos is generated during the dismantling process of nuclear power plants. Mixed wastes have both radiation hazards and chemical hazards, and there’s a possibility of synergistic effects generation. However, in most countries except the United States, there are no regulatory standards for the chemical hazards of mixed waste. The regulations applicable to mixed waste in Korea include the Nuclear Safety Act and the Waste Management Act. The Nuclear Safety Act prohibits the acceptance of hazardous radioactive waste in disposal facilities, but there is no definition or characteristic identification procedure for “hazardous.” The Waste Management Act also does not state the regulation for radioactive waste. In the Gyeongju disposal facility in Korea, the leachate in the disposal facility is expected to be a groundwater saturated with concrete and is expected to irradiated by radioactive waste. On the other hands, most of the non-radioactive waste landfill facilities are built on the surface, and the leachate is expected to be rainwater that reacts with the soil. Due to the differences in leaching environments, there’s a potential to overestimate or underestimate the leaching properties of hazardous substances if the standard leaching test is applied. To show for this, a leaching test simulating disposal facility’s environment were applied to sample waste containing heavy metals. The leaching solution was groundwater collected from the area near the Gyeongju disposal facility, which is then saturated with concrete and adjusted to pH 12.5. In addition, gamma-ray irradiation was conducted during the leaching test to observe changes in the leaching behavior of heavy metals in the actual radioactive waste disposal environment. As a result, lead showed significantly increased leaching compared to the standard test method, and cadmium was not detected in all experimental conditions except heavy irradiation. This study suggested that regulations on the hazardous of mixed waste should be settled, which should be established in sufficient consideration of the types and characteristics of substances contained in the waste.

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.