아스팔트는 점성이 높은 도로포장 재료로서 골재와 채움재를 결합시키는 역할을 하는 중요한 재료이다. 아스팔트 콘크 리트는 생산 및 시공 과정에서 혼합물이 장비에 부착되는 현상이 발생하여 생산 효율 및 성능 저하를 유발하기도 한다. 이와 같은 문제를 해결하기 위하여 석유류와 식물성 기름을 부착방지제(Asphalt Release Agent)로 사용하고 있다. 하지 만 석유류는 단시간에 아스팔트를 용해시켜 혼합물의 성능 저하를 유발할 수 있고, 발생하는 유증기로 인해 작업자를 위 험에 노출시킬 수 있다. 식물성 기름은 석유류에 비해 아스팔트에 대한 용해력은 다소 작지만 석유류와 동일한 문제를 야기하며 제품 단가로 인해 높은 비용을 발생시킨다. 이에 따라 아스팔트 혼합물 생산 및 시공현장에서 저독성이며 친환 경적인 부착방지제의 중요도가 상승하고 있다.

The Korea Atomic Energy Research Institute (KAERI) has facilities that are operated for the purpose of treating radioactive wastes and storing drums before sending them to a disposal site. Domestic regulations related to nuclear facility require radiological dose assessment resulting from release of gaseous radioactive effluent of nuclear facilities. In this study, ICRP-60-based dose conversion factors were applied to evaluate the radiation dose to residents in the event of operation and accident for the radioactive waste management facilities in KAERI. The radioactive gaseous effluent generated from each facility diffuse outside the exclusion area boundary (EAB), causing radiation exposure to residents. To evaluate the external exposure dose, the exposure pathways of cloudshine and radioactive contaminated soil were analyzed. The internal exposure dose was estimated by considering the exposure from respiration and ingestion of agricultural and livestock products. The maximum individual exposure dose was evaluated to be 1.71% compared to the dose limit. The assumed situation used for accidental scenarios are as follows; A fire inside the facility and falling of radioactive waste drum. It was a fire accident that caused the maximum exposure dose to individual and population living within an 80 km radius of the site. At the outer boundary of the low population zone (LPZ), the maximum effective dose and thyroid equivalent dose were estimated as 8.92 E-06% and 5.29 E-06%, respectively, compared to the dose limit. As a result of evaluating the radiological exposure dose from gaseous emissions, the radioactive waste treatment facilities and its supplementary facilities meet the regulations related to nuclear facility, and are operated safely in terms of radiological environmental impact assessment.

While many countries consider direct disposal of the spent nuclear fuels, they need to consider long-term disposal scenarios with severe accidents such as the contact between underwater and the spent nuclear fuel due to large defect of the canister. Radionuclides releases rapidly with contacting water or slowly with dissolution of UO2 matrix. The former is known as the ‘Instant Release’, and the latter is ‘Congruential Release’. Even though the instant release fractions (IRF) are much smaller than the congruential ones, IRF has to be treated carefully due to the fact that the instant releases lead to much larger value of the exposure dose rates than the congruential ones which proceed very slowly. It is known that the exposure dose rates by the instant releases are ~25 times larger than the one by the congruent release. The radionuclides from UO2 matrix migrate to the grain boundary, make bubbles, and make tunnels, which leads to instant releases of some radionuclides. The radionuclides in the gap between UO2 pellet and cladding can be also instantly released. In addition, the radionuclides in the crud are instantly released. But in this paper, nuclides from the crud are not regarded, due to the lack of the leaching data. Meanwhile, there’re some nuclides that released from the construction materials like the cladding, the Rod Cluster Control Assembly (RCCA), or the other metal parts. In this work, IRF values for major IRF nuclides such as Cs, I, Cl, Se for the reference PWR spent fuels of South Korea were evaluated based on the rationale from literatures’ review. In particular, these evaluations were done as the function of fission gas release (FGR), average discharge burnup, and fuel dimensions. In addition, the values of IRF for the other nuclides were also suggested based on the other institutes.

Given the situation in the Republic of Korea that all nuclear power plants are located at the seaside, the interim storage facility is also likely to be located at seaside and the maritime transportation of Spent Nuclear Fuel is considered inevitable. The Republic of Korea does not have an independently developed maritime transportation risk assessment code, and no research has been conducted to evaluate the release rate of radionuclides from a submerged transportation cask in the sea. Therefore, there is a need to develop a technology that can assess the impact of immersion accidents and establish a regulatory framework for maritime transportation accidents. The release rate of radionuclides should be calculated from the flow rate through a flow path in the breached containment boundary. According to the cask design criteria, it is anticipated that even under severe accident conditions, the flow path size will be very small. Previous studies have evaluated fluid flow passing through micro-scale channel by integrating internal and external flows within and around a transport cask. As part of the evaluation, a comprehensive “Full-Field Model” incorporating external flow fields and a localized “Local-Field Model” with micro-scale flow paths were constructed. Sub-modeling techniques were employed to couple the flow field calculated by the two models. The aforementioned approach is utilized to conduct the evaluation of fluid flow passing through micro-scale flow paths. This study aims to evaluate fluid flow passing through micro-scale flow paths using the aforementioned CFD (Computational Fluid Dynamics) method and aims to code the findings. The Gaussian Process Regression technique, a machine learning model, is utilized for developing a mathematical metamodel. The selected input parameters for coding are organized and their respective impacts are analyzed. The range of these selected parameters is tailored to suit domestic environments, and computational experiments are planned through Design of Experiments. The flow path size is included as an input parameter in the coded model. In cases where the flow path size becomes extremely small, making it impractical to use CFD techniques for calculations, Poiseuille’s law is employed to calculate the release rate. In this study, a model is developed to evaluate the release rate of radionuclides using CFD and mathematical equations covering the whole possible range of flow path size in a lost cask in the deep sea. The model will be used in the development of a maritime transportation risk assessment code suitable for the situation and environment in Korea.

모바일 게임 시장 규모가 시간이 지날수록 커지고 있다. 그만큼 한국뿐만 아니라 전 세계적으로 모바일 게임 을 개발하고 출시하고 있는 개발사들이 늘어나고 있는데, 한해 출시되는 수백 개 이상의 모바일 게임들 사이 에서 성공적으로 유저들을 유치하고 수익을 창출하기 위해서는 단순히 모바일 게임을 완성도 있고 재미있게 개발하는 것으로 끝나는 것이 아닌 출시 이전부터 출시 이후 운영 까지 다양한 개발 외적인 전략을 활용하여 경쟁력을 확보해야 한다. 본 연구에서는 모바일 게임의 완성도를 결정하는 게임 내적인 요인들을 제외하고 출 시, 마케팅, 운영 등 모바일 게임 개발 외적인 요인들 중 모바일 게임의 성공에 도움이 될 수 있는 키워드들 을 추출하고 이에 대한 전략들을 정의함으로써 향후 모바일 게임 개발사들이 게임을 개발하는데 있어 참고할 수 있도록 하였다.

In case of Korea, unlike overseas nuclear power plants, adjacent units are located in permanently stopped nuclear power plants. Radioactive substances from airborne and liquid effluents are released into the environment from the NPP, and the radioactivity of the released substances must be reported to the regulatory authorities. Radioactive effluents are released into the environment not only in operation but also after permanent shutdown. Due to domestic conditions in which multiple units exist on the same site, it is necessary to consider radioactive effluents generated after permanent shutdown of NPPs. In particular, liquid effluent may have an increased tritium concentration due to draining the spent fuel pool. This paper summarizes the annual liquid emissions of PWR power plants that have been permanently shut down. The data was obtained from the Nuclear Regulatory Commission’s (NRC) annual radioactive effluent release report, which provides information on the annual emissions power plants into the environment. The liquid emissions of each plant were organized into an annual table, providing an overview of the amount of liquid released by each plant. This study aims to raise awareness about the potential environmental impact of permanently shut down nuclear power plants and the need for proper management of their liquid emissions. The findings of this study can used by operator, policymakers, and other stakeholders to make informed decisions regarding the decommissioning and management of nuclear power plants.

In underground repository environments, various types of engineered barriers are installed to hinder the mobility of radionuclides. Cement admixtures, especially used to improve workability for concrete, are composed of fairly high organic molecules and have a dispersing effect through bonding with the C-S-H of the concrete. Previous studies have shown that complex-forming organics like EDTA, NTA, and ISA have a significant effect on the mobility of radionuclides, but the studies on the behavior and stability of combined complexes in hydrated cement are lacking. So, we selected a commonly used polycarboxylic-ester (PCE) type cement admixture and stable Co as a surrogate of Co-60 to perform desorption experiments from hydrated cement containing the admixture. Radioactive Co is known to be a common contaminant in nuclear fission and medical facilities and considered to exist as a relatively stable phase in repositories. In addition, the evaluation of cobalt can be a standard of safety issue for other radionuclides with the presence of cement admixture in repository. In this study, cement samples were prepared at water/cement ratio of 0.55 and cured for 28 days at 23-25°C and at least 80% of humidity with varying cement admixtures of 0.0, 0.1, and 2.0wt%. To evaluate the stability of cobalt in the weathered cement, a 0.001 M HCl solution was used to simulate cement weathering conditions on a hot plate at 60°C for 1 day using a solid/liquid ratio of 1:100. Degree of weathering was confirmed using XRD analysis. The adsorption experiments were performed by adding 0.0042 mmol of cobalt (CoCl2, Sigma-Aldrich, anhydrous ≥ 98.0%) to the weathered cement for 3 days using a platform shaker at 200 rpm, and the supernatant was separated using a syringe filter (<0.20 um) before ICP-MS analysis to determine the amount of Co adsorption. Cobalt desorption was tested for the Co-adsorbed cement using 0.019 mmol of calcium (Ca(NO3)2·4H2O, Sigma-Aldrich, 99%) for 3 hours to 14 days. The results showed that adsorbed cobalt with and without cement admixture was stably bound to cement, and did not increase any noticeable Co release by 2.0wt% PCE admixture. However, additional experiments using varying contents of PCE and other admixtures should be conducted to provide a standard for assessing the safety of cement admixtures in repositories.

Bentonite is a promising buffer material for high-level radioactive waste (HLW) disposal due to the high nuclides sorption capacity and swelling property. However, bentonite has the potential to generate colloid particles, with small particle sizes less than 1,000 nm when in contact with groundwater. The bentonite colloids easily form pseudo-colloid with the released nuclides and migrate through the water-conducting rock to the biosphere. Therefore, understanding the generation and migration of bentonite colloids is crucial in assessing the safety of the HLW repository. In this study, an artificial fracture system was prepared to investigate colloid release from compacted bentonite. A 250 mm diameter acrylic artificial fracture system was used, with 30 mm of compacted calcium bentonite installed. Artificial groundwater flow was injected into the system at a flow rate of 250 μL/h, and every 6 mL of leachate was collected by a fraction collector. A film-type pressure sensor was equipped to monitor the swelling pressure, and the swelling was observed using a digital microscope. The results indicate that the compacted bentonite formed a mineral ring originating from the swelling of the bentonite, and the end of the ring generated colloid particles due to chemical erosion. Although the release rate of colloids increased with increasing flow rate, the colloid ratio depended on the low ionic strength of the injected artificial groundwater. This work contributes to the understanding of the chemical erosion and colloid release mechanism of compacted bentonite.

Two sets of analyses for the cases of groundwater release to well and sea ecosystems were conducted for the environmental impact assessment of high-level radioactive waste disposal facilities. After obtaining the respective BDCF (Biosphere Dose Conversion Factor) results for the scenarios of well-farming and marine water fishing using different biosphere assessment conceptual models implemented in ECOLEGO, they were compared each other. The purposes of these analyses are to identify reference generic biosphere conceptual models and to get insight on model uncertainty. In this study, the endpoint used for the comparison of the ECOLEGO biosphere models was the socalled Biosphere Dose Conversion Factor (BDCF), which is defined as the maximum value of the total dose to the exposed group, in Sv/yr, resulting from a continuous unit release of 1 Bq/yr during the whole simulation time either to the well compartment (BDCF_Well) or to the marine water compartment (BDCF_Sea). The radionuclides considered in the comparison were Cs-137, I-129, Nb-94, Ni-59, Ni- 63, Sr-90 and Tc-99. The conceptual models used in the biosphere assessment of the releases to a well are based on models that have been used by the DOE (simple-soil model) and SKB (complex-soil model) in safety assessments of radioactive waste repositories, respectively. Difference between two conceptual models used in the assessment of the releases to a sea is the number of compartments representing the sea; i.e., one model represents the sea with one compartment for the water and one for the sediment (singlecompartment model), whereas the alternative model uses two compartments for the water and the sediments: one for the inner coast and one for the outer coast (double-compartment model). The results of the BDCF_Well to a farmer obtained with the DOE and SKB models are shown to be very close to each other. Despite the differences in conceptual models and parameters, the results are within a maximum difference of a factor of 4. The results from the SKB model were higher for all radionuclides. The values of the BDCF_Sea obtained with the single- and double-compartment models are shown to be larger differences with a maximum order of 2. For all studied radionuclides, the double-compartment model produces higher BDCFs than does the single-compartment model. The differences would be due to activity concentrations in both water and sediments. Since the hydrodynamic behavior assumed for flow in the sea could significantly influence the dilution volumes and hence the concentrations, it is found that site-specific investigations are necessary to establish an appropriate marine biosphere conceptual model.

일본 정부는 2021년 4월 13일 일방적으로 후쿠시마 원전 오염수를 해양에 방 류하기로 결정했다. 일본의 원전 오염수 해양 방류는 후쿠시마 주변 해역 및 연안의 해양환경에 미치는 방사능 오염뿐만 아니라 우리나라 연안과 태평양 전 역으로 방사능이 유입되어 피해가 발생할 수 있으며, 해양 생태계 변화 등의 심각한 문제도 발생할 것이다. 일본의 원전 오염수 방류 결정에 대한 다양한 대응 방안이 검토되고 있으며, 유엔해양법협약을 중심으로 국제해양법재판소 제소 및 심각한 피해의 가능성과 상황의 긴급성이 있는 경우 잠정조치 신청도 할 수 있다. 이 논문에서는 일본의 원전 오염수 방류 결정에 대한 문제점을 살펴보고, 해 양환경 오염으로 인한 국제환경분쟁의 의의와 주요 특성에 관한 내용을 검토하 고, 소송에 대비하여 유엔해양법협약 관련 규정 및 판례들을 검토하여 시사점 을 도출하였다. 일본의 후쿠시마 원전 오염수 방류가 임박한 시점에서 문제 해 결을 위하여 최우선으로 유엔해양법협약을 근거하여 ITLOS 제소 및 잠정조치 신청을 하는 사법적 해결방안을 모색하고 동시에 한일 양국 간 또는 국제사회 의 공동 협력체계를 통한 외교적(비사법적) 해결방안 등 실효적인 해결방안을 제시하고자 한다.

The anaerobic digestion process produces methane while stabilizing sludge. As of 2020, 62 anaerobic digesters in public wastewater treatment plants are operational in Korea. Many researchers have studied to improve digester performance. Thermal hydrolysis technology is one of the pre-treatment methods for treating sludge. Reduced retention time and enhanced biogas production are the main advantages of sludge disintegration at relatively high temperatures and pressures. But nutrients like nitrogen and phosphorus are released from the pre-treated sludge. Phosphorus is a non-renewable resource that is essential to food production. Wastewater receives 20% of the total phosphate discharge, while 90% of the influent phosphorus load is in sludge. For efficient phosphorus recovery, it is essential to comprehend the phosphorus release characteristics during wastewater treatment, including anaerobic digestion. Biological or chemical processes can achieve phosphorus removal to comply with the effluent discharge limits regulations. The three primary sources of phosphorus in sludge are aluminum-bound phosphorus (Al-P), polyphosphate in phosphorus-accumulating organisms (PAOs), and iron-bound phosphorus (Fe-P). Anaerobic digestion is the typical method for recovering carbon and phosphorus. However, previous research has demonstrated that most phosphorus in anaerobic digestion occurs as a solid phase coupled with heavy metals. Therefore, the poor mass transfer rate results in a slow phosphorus release. Due to the recent growth in interest and significance of phosphorus recovery, many researchers have studied to improve the quantity of phosphorus released into the liquid phase through chelation addition, process operation optimization, and disintegration using sludge pre-treatment. The study aims to investigate characteristics of the phosphorus release associated with the thermal hydrolysis breakdown of sludge and propose a method for recovering phosphorus in a wastewater treatment plant. When solubilizing sludge using thermal hydrolysis pre-treatment, organic phosphates, inorganic phosphates, and polyphosphates are converted into ortho-phosphate. Therefore, applying thermal hydrolysis, anaerobic digestion, and phosphorus recovery processes (struvite formation or microbial electrolysis cells) can recover carbon and phosphorus.

Concrete is used as the main engineering barrier in low and intermediate level radioactive waste disposal facilities. As the time passed, the radionuclides stored in repository may contact with groundwater and leak into the ecosystem through the rock media. In this process, the radionuclides can react with calcite via sorption or coprecipitation, because calcite is the major mineral of concrete. Under the various background conditions in repository, frequent dissolution-precipitation reactions can happen. Dissolution of Sr-coprecipitated calcite may be different from that of SrCO3(s) which can mislead the safety performance of radioactive Sr and the estimate of Sr mobility based on the solubility of SrCO3(s). Strontium is not only one of the fission products but also emits beta rays with a long half-life almost 29 years. The strontium may be released or retarded by the dissolution-precipitation reactions in repository. In this study, the dissolution of Sr-coprecipitated with respect to calcite was tested in various environment conditions. The Sr-coprecipitated calcite, (Sr,Ca)CO3(s) was synthesized by coprecipitation method in alkaline condition. The 250 mL of 0.1 M of CaCl2 solution was mixed with 250 mL of 1.14 mM SrCl2·6H2O solution. Then, independently prepared 500 mL of 0.1 M Na2CO3 solution was mixed with the mixed solution of CaCl2 and SrCl2. The precipitates could be made and they were aged for 3 days at room temperature. Then, the supernatant was separated by the centrifugation and the solid at the bottom was dried in an oven at temperature 80°C. After that, the Srcoprecipitated calcite powder was washed using the DI water several times and dried again before use. Characterization of solid powder was conducted by XRD and SEM, and the ICP-MS and ICP-AES were used to analyze the concentrations of Ca and Sr. The batch dissolution experiment was conducted with a solid-to-solution ratio of 10 g/L groundwater in polyethylene tubes. The oxidative groundwater was synthesized by simulating the chemical composition of KAERI Underground Research Tunnel (KURT) DB-3 groundwater. Different temperatures and pHs were prepared and tested for the release of Sr and Ca from the coprecipitated (Sr,Ca)CO3(s) to compare the results with the release of Sr and Ca from SrCO3(s) and CaCO3(s), respectively. Such as, these results will be used to provide better understanding of Sr release and mobility in various repository environments.

Background: The muscles associated with tension-type headache (TTH) are connected to the scalp fascia, which is firmly attached to the superciliary ridge of the frontal bone. However, there is a paucity of data on treatments for TTH that directly target the scalp fasciae. Objectives: We aimed to validate a new manual therapy to treat TTH by applying myofascial relaxation techniques to the scalp and to examine the changes in quality of life and headache characteristics after treatment. Design: pretest-posttest control group design study. Methods: Total 24 participants with TTH (10 males, 14 females) participated in the study. Manual therapy group (MT) received standard manual therapy and scalp myofascial release technique group (SMT) received scalp myofascial release technique twice per week for 4 weeks. Brief Pain Inventory (BPI) and Headache Impact Test (HIT) for quality of life, headache frequency, duration, and Visual Analog Scale (VAS) for intensity were assessed at the pre- and post-treatment. Results: There was a significant improvement within groups in headache frequency (MT P<.05, SMT P<.001), duration (MT P<.05, SMT P<.01), VAS (MT P<.01, SMT P<.001), HIT (MT P<.001, SMT P<.001), and BPI (MT P<.001, SMT P<.001) between the pre- and post-treatment. Group differences were significant for headache duration (P=.027), but the others were not. MT and SMT reduced the impact of headaches on daily life. Also Headache frequency, duration, intensity, and BPI were improved. Conclusion: MT and SMT reduced the impact of headaches on daily life. Also Headache frequency, duration, intensity, and BPI were improved. It has been suggested that both MT and SMT can be used as a non-invasive treatment to treat the TTH, and to improve the quality of life.

Several countries, including Korea, are considering the direct disposal of spent nuclear fuels. The radiological safety assessment results published after a geological repository closure indicate that the instant release is the main radiation source rather than the congruent release. Three Safety Case reports recently published were reviewed and the IRF values of seven long-lived radionuclides, including relevant experimental results, were compared. According to the literature review, the IRF values of both the CANDU and low burnup PWR spent fuel have been experimentally measured and used reasonably. In particular, the IRF values of volatile long-lived nuclides, such as 129I and 135Cs, were estimated from the FGR value. Because experimental leaching data regarding high burnup spent nuclear fuels are extremely scarce, a mathematical modelling approach proposed by Johnson and McGinnes was successfully applied to the domestic high burnup PWR spent nuclear fuel to derive the IRF values of iodine and cesium. The best estimate of the IRF was 5.5% at a discharge burnup of 55 GWd tHM−1.

Attention has been paid to the source term released after Chernobyl and Three Mile Island (TMI), which were the representative accidents of nuclear power plants, and has been studied several times in order to predict and evaluate radiation source term, which can be released in the event of a virtual accident. In particular, the impact of the accident was assessed on the basis of Deterministic Safety Analysis (DSA) and after the WASH-1400, the technology of the Probabilistic Safety Assessment (PSA) was introduced, supplementing safety by taking into account the existence of uncertainty. After the Fukushima accident, a SOARCA report was published to evaluate the specific classification of each type of accident, the realistic progress of the accident, and the leakage of radioactive materials. In this paper, the evaluation methodology and results of the source term of severe accident before and after the Fukushima accident were compared, and the evaluation methods applied to domestic nuclear power plants were compared. Prior to the Fukushima accident, the behavior of the accident and source term were evaluated for Loss of Coolant Accident (LOCA), which led to design based accidents, Total Loss of Feed Water (TLOFW) followed by Station Blackout (SBO) the results were compared to Chernobyl and TMI based on the resulting data to evaluate safety and reliability. After the Fukushima accident, the Interfacing System Loss of Coolant Accident (ISLOCA) and the Steam Generator Tube Rupture (SGTR), which is classified as containment’s bypass accident, were included for predictive assessment. This is due to the analysis that the risk of cancer and early mortality are affected. MACST facilities and strategies were added to domestic nuclear power plants, and accidents with a high core damage frequency were mainly interpreted. In addition, source term was evaluated with the addition of a Basement Melt-Through (BMT) accident that had not previously been considered as a focus. As a result of the comparison of source term evaluation, accidents can be caused by a number of unidentified problems, and Korea’s experience on Level 2 and 3 has not been accumulated, making it difficult to predict the results of source term evaluation or lack of reliability.

The Kori Unit 1 and Wolsong Units 1, commercial reactors in South Korea, were permanently shut down due to the expiration of their design lifetime. Therefore, nuclear power plants that have been permanently shut down must be dismantled, and the site must be finally released after removing the remaining radionuclides. Domestic regulatory standards for site remediation should not exceed 0.1 mSv per year based on effective dose. In addition, it is necessary to calculate the preliminary Derived Concentration Guideline Levels (DCGL) to prove that the conditions are met. Therefore, in this study, the input factor considering the geological characteristics of the site of Kori Unit 1 was investigated, and the preliminary Derived Concentration Guideline Levels were calculated and compared with the results of previous studies. As a result of comparative analysis, 60Co, 134Cs, and 137Cs, which are gamma-ray emitting radionuclides, had similar values to DCGL of previous studies A and B. However, 63Ni, a beta-rayemitting nuclide, was 5.94×104 Bq·g−1 in this study and 8.47×101 Bq·g−1in previous study B, resulting in a difference of about 700 times. In addition, in the case of 90Sr, this study and previous study A were derived similarly, but this study was 5.34×101 Bq·g−1 and previous study B was 1.18×10−1 Bq·g−1, resulting in a difference of about 450 times. This difference is judged to be because, unlike this study using only the industrial worker scenario, in the case of previous study B, the resident farmer scenario was mixed and used, which considers the internal exposure caused by ingestion of food produced in the contaminated area. In this study, it was confirmed that DCGL according to the change of geological factors of the site did not have a significant effect on gamma-ray-emitting nuclides. However, it was confirmed that considering the intake of food affects the DCGL of beta-ray-emitting nuclides. Therefore, there is a need to conduct future studies applying intake input factors that meet domestic conditions.