지하도로는 주변 도심개발 등으로 수평적 도로용량 확장이 어려운 구간을 대상으로 입체적 활용을 통해 교통혼잡을 해소하기 위한 수단으로써 주목받고 있다. 그러나 지하도로는 공간 특성상 지상에 비해 다양한 불확실성이 수반된다는 면에서 위험요인이 크고 직접 건설비용이 큰 사업으로 인식되기도 하며, 최근 관련 안전·방재 및 설계 요소 관련 법령과 지침이 지속적으로 제·개정됨에 따라 사업 비 변동요인이 증가하고 있다. 따라서 향후 증가할 지하도로 민간투자사업 추진시에 발생하는 다양한 위험요인에 대한 관리방안 마련 이 필요한 시점이다. 이에 본 연구는 지하고속도로 민간투자사업의 특수성을 반영한 기본계획 단계의 기술부문 평가지표에 대한 배점 개선안을 제시하는 것을 목적으로 한다. 이를 위해 현재 제3자 제안 공고상 기술 부문의 평가 배점을 검토하고, 신규 평가항목이 포함 된 후보 평가지표를 제시하였다. 또한 분석적 계층화 기법(AHP) 기반의 전문가 설문조사를 수행하여 평가지표의 우선순위를 도출해 최종적인 지하 고속도로 평가지표의 배점을 도출하였다. 평가항목별 배점 결과 건설계획 분야에는 ‘공사비, 보상비 등 사업비 선정(10 점)’, ‘장대터널계획 종합(25점)’, ‘시·종점부 도로접속 및 교통운영(18점)’ 평가항목이 추가되었다. 사업관리 및 운영계획 분야에는 ‘안 정적 운영을 위한 계획수립(76점)’, ‘운영계획의 창의성(24점)’, ‘재해 및 재난관리계획(21점)’ 등의 순으로 배점이 수정되었으며, 특히 시설 및 장비확보 계획, 유지보수 계획, 교통 및 영업관리 계획 등이 고려되는 ‘안정적 운영을 위한 계획수립’ 평가항목의 배점이 크게 증가하였다. 본 연구에서 제시한 배점 개선안을 통해 기본설계 단계에서부터 지하 고속도로 민간투자사업의 위험요인과 특수성을 고 려하여 사업을 추진할 수 있을 것으로 판단된다.

After the permanent shut down of Kori Unit 1, various decommissioning activities will be implemented, including decontamination, segmentation, waste management, and site restoration. During the decommissioning period, waste management is among the most important activities to ensure that the process proceeds smoothly and within the expected timeframe. Furthermore, the radioactive waste generated during the operation should be sent to a disposal facility to complete the decommissioning project. Square and cylindrical concrete re-package drums were generated during the 1980s and 1990s. The square, containing boron concentrates, and cylindrical, containing spent resin, concrete re-package drums have been stored in a radioactive waste storage building. Homogeneous radioactive waste, including boron concentrates, spent resin, and sludge, should be solidified or packaged in high-integrity containers (HICs). This study investigates the sequential segmentation process for the separation of contaminated and non-contaminated regions, the re-packaging process of segmented or crushed cement-solidified boron concentrate, and re-packaging in HICs. The conceptual design evaluates the re-packaging plan for the segmented and crushed cement-solidified waste using HICs, which is acceptable in a disposal facility, and the quantity of generated HICs from the treatment process.

Korea Atomic Energy Research Institute’s Post Irradiated Examination Facility safely stores spent nuclear fuel using a wet storage method to conduct research. Here, in order to remove the radioactivity released into the water, the stored water is passed through an ion exchange resin tower, and the radionuclides are exchanged with the bead-shaped ion exchange resin filled inside to lower the radioactivity concentration. At this time, because the stored water passes in one direction, clogging of the ion exchange resin occurs. If this phenomenon continues, the flow rate of the water treatment process decreases and operation efficiency decreases, so a backwashing process is necessary to re-mix the ion exchange resin and secure the flow rate again. In this study, the flow rate reduction trend according to the lifespan of the ion exchange resin and the flow rate recovery according to the backwash process operation amount were analyzed. The flow rate reduction trend of the ion exchange process was analyzed immediately after the backwashing process was started. In addition, the amount of flow recovery according to the backwash process operation amount was evaluated by the amount of waste generated during the backwash process and the number of days of operation until the backwash process was needed again. As a result, the flow rate of the ion exchange process decreased rapidly right after the backwash process until the position of the ion exchange resins was stabilized, and then stabilized. After that, it gradually decreased and reached the point where the backwash process was necessary. However, the decline trend was analyzed to be the same regardless of the lifespan of the ion exchange resin. In addition, the amount of waste generated during the operation of the backwash process was increased in the order of 400 L, 600 L, 1,100 L, 1,400 L, 3,500 L, and 4,200 L to increase the amount of operation of the backwash process. As a result, the number of days of ion exchange resin operation was 285 days, 338 days, and 342 days, was analyzed as 422 days, 322 days, and 720 days. Based on this study, it was confirmed that the flow rate reduction trend is the same regardless of the lifespan of the ion exchange resin, and as the backwash process operation increases, the number of days the ion exchange process can be operated increases, but there is a turning point where the waste treatment cost exceeds the number of days of operation.

Molten salt reactors have several advantages over conventional light water reactors. These include producing less nuclear waste, operating at higher power efficiency and inherent safety due to the low operating pressure. NaCl-MgCl2 eutectic salt is one of the candidates for the molten salt reactor coolant. However, because the salt is very hygroscopic, structural material corrosion occurs resulting in the high cost to maintain. To mitigate corrosion there have been many studies for the dehydration of the salt, especially focusing on the magnesium chloride. The reason is that the moisture adsorbed to the magnesium chloride undergoes hydrolysis over 200 degrees Celsius and decomposes to MgOHCl while the moisture associated with the NaCl is easily liberated during the heating procedure without chemical reaction. As the operating temperature of the molten salt is between 500 and 700 degrees Celsius, the MgOHCl is believed as the main cause for the structural corrosion. In this research, thermal dehydration of the salt with elemental Mg, for the NaCl-MgCl2 eutectic, was studied based on the previous dehydration methods and considering scalable and easy to handle. The MgOHCl was removed both through the thermal decomposition and the reduction by Mg metal. After the removal of MgOHCl, based on the difference between the freezing points and the density, the salt cooled down very slowly to ensure the separation between the purified salt and the disposals such as MgO and remaining Mg metals. The efficiency of the dehydration method was determined by the concentration of the MgOHCl. The concentration was determined by cyclic voltammetry and the result was compared with undehydrated salt and salt dehydrated thermally without the addition of Mg metal. To qualify and quantify the MgOHCl content through the cyclic voltammetry, it was necessary to observe the signal by adding MgOHCl to each sample. Based on the thermogravimetric analysis result of MgCl2· 6H2O, MgOHCl powder was formed through heating the MgCl2·6H2O.

Radioactive waste generated during nuclear power plant decommissioning is classified as radioactive waste before the concentration is identified, but more than 90% of the amount generated is at a level that can be by clearance. However, due to a problem in the analysis procedure, the analysis is not carried out at the place of on-site but is transported to an external institution to identify concentration, which implies a problem of human error because 100% manual. As a way to solve this problem, research is underway to develop a mobile radioactive waste nuclide analysis facility. The mobile radionuclide analysis facility consists of a preparation room, a sample storage room, a measurement room, a pretreatment room, and a waste storage room, and is connected to an external ventilation facility. In addition, since the automation module is built-in from the sample pre-threatening step to the separation step, safety can be improved and rapid analysis can be performed by being located in the decommissioning site. As an initial study for the introduction of a mobile nuclide analysis facility, Visiplan, a preliminary external exposure evaluation code, was used to derive the analysis workload by a single process and evaluate the exposure dose of workers. Based on this, as a follow-up study, the amount of analysis work according to the continuous process and the exposure dose of workers were evaluated. As a result of the evaluation, the Regulatory dose limit was satisfied, and in future studies, internal and external exposure doses were evaluated in consideration of the route of movement, and it is intended to be used as basic data in the field introduction process.

In case of damaged spent fuels, it would require additional treatment for their transportation and storage to capture the radioactive fission products in a defined space. The canning container for the damaged spent fuels is one way to seal the radioactive fission products inside the container. In the Post Irradiation Examination Facility (PIEF) of KAERI, the Quiver container has been introduced for canning damaged spent fuels from Westinghouse Sweden. The main container body has been manufactured for particle-tightness of spent fuel. In addition, drying equipment is being prepared for gas-tightness of spent fuel. The drying equipment can remove water and fill the inert gas inside the container. Before drying inside the container, we evaluated the volatile fission products inventory because volatile fission products could be released during the drying process. Despite assuming highly conservative hypotheses for the inventory remaining in damaged fuel rods, the amount that could be released during the drying process was less and dose rate levels around the evacuation piping system were low.

In this paper, the mechanical properties of glass fiber reinforced plastic (GFRP) rebar, which has been applied as an concrete reinforcement, and produced carbon fiber reinforced plastic (CFRP) grid were compared to develop a concrete reinforcement material with excellent mechanical properties. In addition, the mechanical properties of CFRP prepared with each molding process were evaluated. Three molding processes were evaluated: prepreg oven bagging, reaction injection molding (RIM), and pultrusion. The tensile strength of the CFRP grid prepared through pultrusion was 2.85GPa, the elastic modulus was 169.81GPa, and the strain was 1.68%, which was 2.85 times better in tensile strength, and 2.83 times better in elastic modulus compared mechanical properties of GFRP rebar. The strain was confirmed to be equivalent to GFRP rebar.

본 연구에서는 비가열 떡 제조업체 3곳을 대상으로 원 재료, 제조설비 및 계절별 제조공정에 대한 미생물 오염 도를 분석하였고, 여름철 불림 시간에 따른 미생물 오염 도 및 불림 수 온도 조절을 통한 미생물 저감 효과를 확 인하고자 하였다. 3업체의 원재료 일반세균수는 2.69-5.08 log CFU/g 범위로 검출되었으나 제조공정 중 불림공정에 서 미생물 오염도가 급격히 증가함을 확인하였다. 계절에 따른 제조공정별 미생물 오염도 분석결과, 여름철 불림공 정에서 일반세균 및 대장균군이 7.01 및 3.96 log CFU/g 로 다른 계절에 비해 유의적으로 높게 나타났고, 이후 공 정에서도 높은 오염도를 유지하여 냉동공정에서 일반세균 이 6.24 log CFU/g로 법적인 기준을 초과하여 검출되었다. 여름철 불림 초기 수온은 19.1oC에서 불림 12시간 후 26.8oC까지 상승하였고, 불림시간에(3, 6, 9, 12 h) 따른 제조공정별 미생물 오염도 분석결과, 불림시간이 길어질수 록 미생물 오염도가 유의적으로 높게 나타났고, 불림 9시 간 이후부터 냉동공정의 일반세균 수치가 냉동식품의 법 적인 기준을 초과하여 검출되었다. 여름철 불림 수의 온 도 상승을 억제하기 위하여 얼음팩을 이용하여 불림 수온 을 조절한 결과, 불림 12시간까지 20.1oC로 유지되어 조 절 전보다 약 7oC가량 낮게 나타났다. 이에 따른 제조공 정별 미생물 오염도 분석결과, 3업체의 불림 12시간 이후 냉동공정 일반세균 평균값이 4.42 log CFU/g로 조절 전보 다 1.77 log CFU/g 감소한 것으로 확인되었다. 이상의 결 과로 볼 때, 안전한 비가열 냉동떡 생산을 위해서는 업체 에 맞는 불림시간 및 불림 수 온도조절 등의 선행요건 관 리기준 설정이 필요하며, 이러한 선행요건 관리기준의 적 용으로 비가열 냉동떡 제조 HACCP system의 안전성이 확립될 수 있을 것으로 판단되었다.

Radioactive contamination of soil on the site of a nuclear facility has a characteristic that radioactive nuclides are adsorbed into the pores between soil particles, making it quite difficult to decontaminate. For this reason, research on the development of various decontamination processes is being actively conducted. In this study, among various decontamination studies, a soil decontamination process using supercritical carbon dioxide was presented. The decontamination process uses supercritical carbon dioxide as the main solvent, which has a higher penetration power than other materials. Therefore, the process consists of the process of desorbing and extracting the target radionuclides between particles of soil. However, since nuclides exist as ions in the soil, polar chelating ligand material was introduced as an additive to nonpolar supercritical carbon dioxide for smooth chemical reactions in the soil. Thereafter, from the viewpoint of improving process continuity and efficiency, an alcohol material was introduced as an auxiliary solvent for liquefaction of chelating ligand in a solid state. Through prior research on the selection of a solvent for liquefaction of chelating ligand, ethanol and 2-propanol were finally selected based on whether the chelating ligand was dissolved. However, if the auxiliary solvent in which the chelating ligand is dissolved is to be combined with radionuclides in the soil, it must first be well dissolved in supercritical carbon dioxide, the main solvent. Therefore, in this study, the solubility of ethanol and 2-propanol in supercritical carbon dioxide was measured and the suitability was evaluated. The temperature conditions were carried out at 40°C, the same as the previously designed decontamination process, and the measurement was conducted by adjusting the pressure and volume through a syringe pump and a variable volume device. In addition, solubility was measured based on the observation of the ‘cloud point’ in which the image becomes cloudy and then bright. As a result of the experiment, several solubility points were measured at a pressure of 150 bar or less. If the flow rate ratio of supercritical carbon dioxide and auxiliary solvent derived from the results is applied to the soil decontamination process, it is expected that the process efficiency will increase in the future.

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.

본 연구에서는 하수 재이용을 위한 역삼투막 공정에서 전처리 정밀여과막(MF) 손상에 대한 누출되는 다양한 수 질변화로써 막 손상 검지 방안을 제시하였다. 이를 위하여 역삼투막 유입수질 적합성 평가지표인 SDI (silt density index)를 3에서 5의 범위 내에서 막 손상 시 검지 감도를 정량화하기 위하여 전처리 분리막이 1에서 3가닥 파단에 따라 SDI는 1.92에 서 6.11까지 증가한 결과를 확인할 수 있었다. 일반적으로 3을 기준으로 역삼투막 유입수질로 설정하였을 때 분리막이 3가닥 까지 파단이 되어야만 막 손상 검지가 가능하다는 것을 의미하며 역삼투막의 오염은 잠재적으로 가속화되어 효율을 저하시 킬 수 있다. 또한 이때 누출되는 입자성과 유기물질에 대하여 0.45 μm 이상의 크기만 걸러주는 입자계수는 입도분포별 막 파 단 개수에 따라 일정한 패턴을 확인할 수 없었으며, TOC 농도는 약 2배의 변화패턴으로써 SDI와의 상관관계로써 TOC가 막 손상 수질지표로써 신뢰성이 높은 것으로 확인되었다. 수질분석결과와 더불어 USEPA에서 제시하는 막 손상 검지 방법 중 압력손실시험과 이를 기반으로 LRVDIT 모델의 적합성 평가를 한 결과 막 손상 또는 역삼투막 공정으로 유입되는 막오염물질 을 신속하게 확인할 수 있는 SDI 및 TOC를 포함한 LRVDIT 모니터링과 UCL 설정을 병행해야 한다.

The off-site dose calculation is regularly carried out at the nuclear power plants in order to evaluate off-site dose from gaseous and liquid effluent during normal operation. In 2009, the off-site calculation program (K-DOSE60) was developed in accordance with ICRP-60 by KHNP. This software needs meteorological data, gaseous and liquid effluent data, and various other input parameters to evaluate off-site dose. As a result, it takes a certain amount of time for the user to enter accurate input data and verify calculated results, and it is difficult to intuitively determine them because of providing textbased calculated results. Therefore, in this study, the improvement of the calculation program was considered so that a more reliable and effective evaluation could be performed when calculating the off-site dose. The main improvements of the off-site dose calculation program (ODCP) are as follows. First, it is developed as the network-based program to link with meteorological data, and gaseous and liquid effluent data to remove input errors and simplify data transfer. Second, through validation process of input data, input errors are eliminated. Third, the input data and calculated results are visually provided so that the user can easily determine the evaluation results. Fourth, database of input and calculated results is constructed to facilitate evaluation result history management.

In this study, the removal efficiency of PFCs(perfluorinated compounds) in the GAC(granule activated carbon) process based on the superheated steam automatic regeneration system was investigated in laboratory scale and pilot-scale reactor. Among PFCs, PFHxS(perfluorohexyl sulfonate) was most effectively removed. The removal efficiency of PFCs was found to be closely related to the EBCT, and the removal efficiencies of PFOA(perfluorooctanoic acid), PFOS(perfluorooctyl sulfonate), and PFHxS were 43.7, 75, and 100%, respectively, under the condition of EBCT of 6 min. Afterward, PFOA, PFOS, and PFHxS exhibited the earlier breakthrough time in the order. After that, GAC was regenerated, and the removal efficiency of the PFCs before and after regeneration was compared. As a result, it was shown that the PFCs removal efficiency in the regenerated GAC process were higher, and that of PFOA was improved to 75%. The findings of this study indicate the feasibility of the superheated steam automatic regeneration system for the stable removal of the PFCs, and it was verified that this technology can be applied stably enough even in field conditions.

소변검사 전 냉/해동 반복과 해동 과정에 따라 대표적인 임상 화학검사 측정값의 변화를 확인 함으로써 소변검사의 안정성과 품질 개선방안을 모색하고자 하였다. 조사 대상자는 모두 건강한 남성 10명이었으며 이들의 소변 검체를 이용하여 냉/해동 안정성(freeze and thaw stability) 실험을 진행하였다. Micro-albumin과 Amylase의 경우 시간이 경과 됨에 따라 37℃에서 는 통계적 유의성은 없었으나, 42℃와 60℃에서는 시간의 경과에 따른 결과가 통계적으로 유의한 변동 이 있었고, BUN, Creatinine, Uric acid와 Glucose에서는 통계적으로 유의한 변동이 있었다. Long term의 안정성 결과, 7일이 지난 후에는 Glucose의 변이는 증가하였고, 60℃에서는 Amylase가 감 소하는 양상을 보였다. Glucose와 Amylase의 경우 시간의 경과에 따른 결과가 통계적으로 유의한 변동 이 있었다. 신뢰성 있는 검사결과를 얻기 위해서는 소변 시료의 채취, 보관 및 저장 등을 비롯한 요검사 의 정확한 표준화가 필요하며 생체 물질별 안정성 확보를 위한 조건들의 체계적 연구가 필요하다.