Concrete is the primary building material for nuclear facilities, making it one of the most common forms of radioactive waste generated when decommissioning a nuclear facility. Of the total waste generated at the Connecticut Yankee and Maine Yankee nuclear power plants in the United States, concrete waste accounts for 83.5% of the total for Connecticut Yankee and 52% for Maine Yankee. In order to dispose of the low- to medium-level radioactive concrete waste generated during the decommissioning of nuclear power plants, it is necessary to analyze the radioactivity concentration of gamma nuclides such as Co-58, Co-60, Cs-137, and Ce-144. Gamma-ray spectroscopy is commonly used method to measure the radioactivity concentration of gamma nuclides in the radioactive waste; however, due to the nature of gamma detectors, gamma rays from sequentially decaying nuclides such as Co-60 or Y-88 are subject to True Coincidence Summing (TCS). TCS reduces the Full Energy Peak Efficiency (FEPE) of specific gamma ray and it can cause underestimation of radioactivity concentration. Therefor the TCS effect must be compensated for in order to accurately assess the radioactivity of the sample. In addition, samples with high density and large volume will experience a certain level of self-shielding effect of gamma rays, so this must also be compensated for. The Radioactive Waste Chemical Analysis Center at the Korea Atomic Energy Research Institute performs nuclide analysis for the final disposal of low- and intermediate-level concrete waste. Since a large number of samples must be analyzed within the facility, the analytical method must simultaneously satisfy accuracy and speed. In this study, we report on the results of evaluating the accuracy of the radioactivity concentration correction by applying an efficiency transfer method that appears to satisfy these requirements to concrete standard reference material.

Plasma melting technology uses electrical arc phenomena such as lightning to create hightemperature sparks of about 1,600 degrees or more to meet waste disposal requirements through treatment and reduction without distinguishing radioactive waste generated during nuclear power plant operation and dismantling according to physical characteristics. Decommissioning radioactive waste scabbed concrete occurs when polishing and cutting the contaminated structure surface to a certain depth. In this study, Scabbed concrete containing paint was confirmed for volume reduction and disposal safety using plasma treatment technology, and it is planned to be verified through continuous empirical tests.

Decommissioning waste is generated with various types and large quantities within a short period. Concrete, a significant building material for nuclear facilities, is one of the largest decommissioning wastes, which is mixed with aggregate, sand, and cement with water by the relevant mixing ratio. Recently, the proposed treatment method for volume reduction of radioactive concrete waste was proven up to scale-up testing using unit equipment, which involved sequentially thermomechanical and chemical treatment. According to studies, the aggregate as non-radioactive material is separated from cement components with contaminated radionuclides as less than clearance criteria, so the volume of radioactive concrete waste is decreased effectively. However, some supplementation points were presented to commercialize the process. Hence, the process requires efficiency as possible to minimize the interface parts, either by integration or rearranging the equipment. In this study, feasibility testing was performed using integrated heating and grinding equipment, to supplement the possible issue of generated powder and dust during the process. Previously, heat treatment and grinding devices were configured separately for pilot-scale testing. But some problems such as leakage and pipe blockage occurred during the transportation of generated fine powder, which caused difficulties in maintaining the equipment. For that reason, we studied to reduce the interface between the equipment by integrating and rearranging the equipment. To evaluate the thermal grinding performance, the fraction of coarse and concrete fines based on 1mm particle size was measured, and the amount of residual cement in each part was analyzed by wet analysis using 4M hydrochloric acid. The result was compared with previous studies and the thermomechanical equipment could be selected to enhance the process. Therefore, it is expected that the equipment for commercialization could be optimized and composed the process compactly by this study.

During the operation of nuclear power plant (NPP), the concentrates and spent resin are generated. They show relatively high radioactivity compared to other radioactive waste, such as dry active waste, charcoals, and concrete wastes. The waste acceptance criteria (WAC) of disposal facility defines the structure and property of treated waste. The concentrates and spent resin should be solidified or packaged in high integrity container (HIC) to satisfy the WAC in Korea. The Kori NPP has stored history waste. The large concrete package with solidified concentrates and spent resin. The WAC requires identification of 18 properties for the radioactive waste. Since some of the properties are not clearly identified, the large concrete packages could not satisfy the WAC in this moment. The generation of the large concrete package (rectangular type and cylindrical type), pretreatment of the package, treatment of inner drum, process development for clearance waste, etc. will be discussed in this paper. In addition, the conceptual design of whole treatment process will be discussed.

In the Kori power plant radioactive waste storage, the concentrated waste and spent resin drums generated in the past are repacked and stored in large concrete drums. Four 200 L drums of solidified concentrated waste are packed in the square concrete. One 200 L drum of spent resin is packed inside the round concrete. In order to build a foundation for disposal of large concrete drums that generated in the past, it is necessary to develop a large concrete drum handling device and disposal suitability evaluation technology. In order to build handling equipment and establishment of disposal base, such as weight and volume, of square and round concrete containers must be identified. In addition, waste information, such as the production record of the built in drum and the type of contents, is required. Therefore, this study plans to comprehensively review the characteristics of the waste by investigating the structure of square and round concrete containers and the records of internal drum production.

In the decommissioning site of Korean Research Reactor 1&2 (KRR-1&2), according to Low and Intermediate-level Radioactive Waste Disposal Acceptance Criteria of the Korea Radioactive Waste Agency (WAC-SIL-2022-1), characteristics of radioactive waste was conducted on approximately 550 drums of concrete and soil waste for a year starting from 2021. Among them, 50 drums of concrete waste transported and disposed to Gyeongju LILW disposal facility at the end of 2022. For the remaining approximately 500 drums of concrete and soil waste stored on-site, they were reclassified into two categories: permanent disposal grade and clearance grade. This classification was based on calculating the sum of fractions (SOF) per drum for each radionuclides. The plan is to dispose of around 200 drums in the permanent disposal grade and about 300 drums in the clearance grade by the end of 2023. Since concrete and soil decommissioning wastes are generated in large quantities over a short period with similar origins, they were grouped within five drums as suggested by the acceptance criteria. Mixed samples were collected from each group and used for radionuclide analysis. When utilizing mixed samples, three distinct samples are collected and analyzed for each group. The maximum value among these three radionuclide analysis results is then uniformly applied as the radionuclide concentration value for all drums within that group. Radioactive nuclides contained in similar types of radioactive waste with similar origins can be expected to have some statistical distribution. However, There has been no verification as to whether the maximum value among the three mixed samples exists within the statistical distribution or if it deviates from this distribution to represent a different value. In this study, we confirmed characteristics of radionuclide concentration distribution by examining and comparing radionuclide concentration distributions for radioactive wastes drum grouped for nuclear characteristic among 50 concrete wastes drum disposed in year 2022 and 500 concretes & soils drum scheduled for disposal (clearance or permanent disposal) in year 2023. In particular, when comparing tritium to other nuclides, it was observed that the standard deviation for the distribution of maximum values was approximately 318 times larger.

Most of the radioactive wastes generated during the nuclear fuel processing activities conducted by KEPCO Nuclear Fuel Co., Ltd. are classified as the categories of intermediate and low-level radioactive waste. These radioactive waste materials are intended for permanent disposal at a designated disposal site, adhering strictly to the waste acceptance criteria. To facilitate the safe transportation of radioactive waste to the disposal site, it is necessary to ensure that the waste drums maintain a level of criticality that complies with the waste acceptance criteria. This necessitates the maintenance of subcritical conditions, under immersion or optimal neutron moderation conditions. This paper presents a criticality safety assessment of concrete radioactive waste under the most conservative conditions of immersion and moderation conditions for waste drums. Specifically, In order to send radioactive waste, which is the subject of criticality analysis, to a disposal facility, pre-processing operations must be performed to ensure compliance with waste accepatance criteria. To meet the physical characteristics required by the accepance criteria, particles below 0.2 mm should not be included. Thus, a 0.3 mm sieve is used to separate particles lager than 0.3 mm, and only those particles are placed in drums. The drums should be filled to achieve a filling ratio of at least 85%. A criticality analysis was conducted using the KENO-VI of SCALE. The Criticality Safety Analysis Results of varying the filling ratio of concrete drums from 85% to 100% presented in an effective multiplication factor of 0.22484. Additionally, the effective multiplication factor presented to be 0.25384 under the optimal moderation conditions. This demonstrates full compliance with the USL and criticality technology standards set as 0.95.

At domestic nuclear power plant, concrete containers are stored to store waste generated before waste acceptance criteria (WAC) was established. Concrete container store concentrated waste liquid and waste resin. In order to disposal radioactive waste to a disposal site, it is necessary to conduct a characteristic evaluation inside the waste to check whether it satisfies the WAC. Two types of concrete containers are stored: round and square. The round type is filled with one 200-liter drum, and the square type is filled with four 200-liter drums. In the case of a round shape, the top lid is fastened with bolts, so it is possible to collect samples after opening the top lid without the need for additional equipment. However, in the case of a square shape, there is no top lid, and concrete is poured to cure the lid, so the separate equipment for characteristic evaluation is required. It is necessary to install a workstation for sample collection on the top of the concrete container, equipment for coring the top of the concrete container, and a device to prevent concrete dust scattering. Currently, the design of equipment for evaluating the characteristics of concrete containers has been completed, and equipment optimization through mock-up test will be performed in the future.

Concentrated effluent and spent ion exchange resins (IERs) from nuclear power plants (NPPs) were generated prior to the establishment of a disposal facility site and waste acceptance criteria have been temporarily stored at the NPPs because their suitability for disposal has not been confirmed. In particular, at the Kori Unit 1, which was the first to start the commercial operation in South Korea, the initially generated concentrated effluent and IERs are repackaged in large size of concrete containers and stored without provided regulation standard. The concentrated effluent is package as cementitious form in 200 L drums and repackaged in concrete containers, case of the IERs were solidified or dehydrated and repackaged in round concrete container. In this study, we review and propose a disposal plan for concentrated effluent and IERs repackaging drums that have not been confirmed to be suitable for disposal from the first operating nuclear power plant, Kori Unit 1, 2. First, the concentrated effluent was stored in four 200 L drums respectively, and then, it was again stored in concrete container and which was poured on top using grouted concrete. Therefore, the process was required by cutting concrete container for extracting the internal drums at first. Internal radioactive waste should be crushed to the suitable waste criteria and solidified, finally disposal in to the polymer concrete high integrity container (PC-HIC). IER was repackaged and disposal in square type of 200 L concrete drums respectively covered the cap. So, extracting the internal drums should be extracted after removing the cap of external concrete container. Cement solidification drums can be crushed and re-solidified or disposed in the PC-HIC. Stored IER after dehydrated can be disposal in PC-HIC. In conclusion, the container was used as a package that repackaging the concentrated effluent and IER was separated into two different types of waste depending on the level of contamination of radioactivity, the polluted area is disposed of as radioactivity contamination or the unspoiled area will be treated as self-disposal waste.

In Natural Analogue Study, Concrete is one of the important engineering barrier components in the Multi-thin wall concept of radioactive waste disposal and plays the most important role in disposal sites. The concrete barrier at the disposal site loses its role as a barrier due to various deterioration phenomena such as settlement, earthquake, and ground movement, causing the disposed waste to leak into the natural ecosystem. Some of the key factor is deterioration triggered by sulfate attack. Concrete deterioration induced by sulfate is commonly manifested in an extensive scale when a concrete structure makes contact with soil or water, aggravating its performance. In this study, an accelerated concrete deterioration evaluation experiment was performed using a total of three experimental methods to evaluate the reaction between concrete and water. The first experiment was a deterioration evaluation using Demi. Water, the second was a deterioration evaluation using KURT groundwater after extraction, and the last experiment was a concrete deterioration evaluation using KURT groundwater and sodium sulfate. For all of these experiments, accelerated concrete deterioration experiments were conducted after immersion for a total of 365 days, and specimens were taken out at 30-day intervals and characterization analysis such as SEM and EDS was performed. Experimental analyzes have shown that various chemical species are generated or destroyed over time. In the future, we plan to continue to conduct a total of three concrete deterioration evaluation experiments above, and additionally evaluate the chemical reaction between bentonite and concrete.

Concrete structures of spent nuclear fuel interim storage facility should maintain their ability to shield and structural integrity during normal, off-normal and accident conditions. The concrete structures may deteriorate if the interim storage facility operates for more than several decades. Even if deterioration occurs, the concrete structures must maintain their own functions such as radiation shielding protection and structural integrity. Therefore, it is necessary to establish an analysis methodology that can evaluate whether the deteriorated concrete structure maintains its integrity under not only normal or off-normal condition but also accident condition. In this study, dynamic material testing was conducted on concrete cores extracted from HANARO exterior wall during seismic reinforcement construction. HANARO was constructed at the Korea Atomic Energy Research Institute in 1995, following strict nuclear quality assurance standards. In order to conduct the dynamic material testing of the extracted concrete cores, self-disposal had to be performed because the concrete cores were extracted and stored in a radiation controlled area. A self-disposal application was prepared and submitted based on the radionuclide analysis results, and it was finally approved in April 2023. Then, a test was performed by processing a specimen for dynamic property testing using a self-disposed concrete core. The concrete cores were processed to create specimens for dynamic material testing and the dynamic material testing was performed to obtain stress-strain diagrams according to the strain rate.

In the case of dry storage facilities, slipping of the cask or tip-over are dangerous phenomena. For this reason, in dry storage facilities, measures against slipping and tip-over or related safety evaluations are important. Accidental conditions that can cause cask slippage and tip-over in dry storage facilities include natural phenomena such as floods, tornadoes, tsunamis, typhoons, earthquakes, and artificial phenomena such as airplane crashes. However, among natural phenomena, earthquakes are the most important natural phenomenon that causes tip-over. Also, many people had the stereotype that Korea is an earthquake-safe zone before 2016. However, earthquakes become a major disaster in Korea due to the 2016 Gyeongju earthquake and the 2017 Pohang earthquake, followed by the Goesan earthquake in October 2022. In this paper, seismic analysis was performed based on dry storage facilities including multiple casks. Design variables for the construction of an analysis model for dry storage facilities were investigated, and seismic analysis was performed. To evaluate tip-over accident during earthquake, seismic load was used from 0.2 g PGA to 0.8 g PGA and these earthquakes were followed Design Response Spectrum (DRS) in RG 1.60. The friction coefficient of concrete pad was used from 0.2 to 1.0. As a result of the analysis, tip-over accident could not find in the analysis from 0.2 g to 0.6 g. However, tip-over was appeared at friction coefficients of 0.8 and 1.0 at 0.8 g PGA. Tip-over angular velocity of cask was derived by seismic analysis and was compared with formula and tip-over analysis results. As a result, a generalized dry storage facility analysis model was proposed, and dry storage facility safety evaluation was conducted through seismic analysis. Also, tip-over angular velocity was derived using seismic analysis for tip-over analysis.

Due to the aging of a building, 38.8% (about 2.82 million buildings) of the total buildings are old for more than 30 years after completion and are located in a blind spot for an inspection, except for buildings subject to regular legal inspection (about 3%). Such existing buildings require users to self-inspect themselves and make efforts to take preemptive risks. The scope of this study was defined as the general public's visual self-inspection of buildings and was limited to structural members that affect the structural stability of old buildings. This study categorized possible damage to reinforced concrete to check the structural safety of buildings and proposed a checklist to prevent the damage. A damage assessment methodology was presented during the inspection, and a self-inspection scenario was tested through a chatbot connection. It is believed that it can increase the accessibility and convenience of non-experts and induce equalized results when performing inspections, according to the chatbot guide.

In this study, the structural performance of the specimen fabricated through 3D printing was evaluated through monotonic loading experiments analysis to apply to 3D printed structures. The compression and flexural experiments were carried out, and the experimental results were compared to the finite element model results. The loading directions of specimens were investigated to consider the capacity of specimens with different curing periods, such as 7 and 28 days. As a result, the strength tended to increase slightly depending on the stacking direction. Also, between the 3D-printed panel composite and the non-reinforced panel, the bending performance depended on the presence or absence of composite reinforcement.

본 논문에서는 유한요소해석 프로그램을 통해 파괴 거동 유형별 철근콘크리트 기둥 및 폭발 하중을 모델링하였으며, 실제 실험과 의 동적 응답을 비교하여 모델의 적합성을 입증하였다. 개발한 모델을 이용하여 폭발 하중에 대한 부재의 동적 응답을 확인하기 위해 폭발 하중 시나리오를 설정하였으며 해당 시나리오별 폭발 하중에 대한 시간에 따른 변위 및 응력 결과를 도출하였다. 동적 응답을 통 해 폭발 하중에 대한 기둥의 성능평가(Ductility, Residual)를 수행하였으며 이를 비교 및 분석하였다.

PURPOSES : The purpose of this study was to improve the performance of concrete pavements by decreasing measurement deviations using an Internet of Things (IoT)-based air content measurement device. METHODS : We calculated the properties of concrete which varied according to the air content. For a low measurement deviation, the concrete pavement performed according to the design standard. To confirm the difference in the performance of the concrete pavement for various air contents, we verified the change in the relative dynamic modulus according to the number of freeze–thaw cycles for each value of the air content. In addition, we analyzed the number of durability cracks according to the freeze–thaw cycles in the field. RESULTS : We confirmed that IoT-based measurement equipment improved the performance of pavements without changing their mixing designs or specifications. We confirmed that the performance of concrete pavements changed even with variations in air content within the range of quality standards. Using IoT-based air content management, we confirmed the reduction in concrete pavement durability cracks without changing the mixing design. CONCLUSIONS : We confirmed that IoT-based air-content management improved pavement performance. The feasibility of extending this concept to manage other concrete properties such as the chloride content should be acknowledged. Future research will require laboratory tests to understand the variation in concrete properties with varying air contents and to consider diverse load conditions.

PURPOSES : This study aims to understand the effect of defective dowel bar installation on jointed concrete pavement (JCP), which can cause joint freezing, spalling, cracking, and faulting and finally shorten the lifespan of the pavement. METHODS : A comprehensive field survey was undertaken at an expressway construction site in South Korea to assess dowel bar installation conditions. In addition, finite element (FE) analysis was used to simulate JCP behavior with both vertical and horizontal dowel misalignments. Different temperature conditions, including a change of -55 °C and gradient of -0.1 °C/mm, were integrated into the FE model to examine horizontal slab contraction and simultaneous slab curling. RESULTS : The analysis revealed pronounced slab behaviors under specific temperature changes, particularly when combined with dowel misalignments. The simultaneous effects of horizontal contraction and slab curling owing to temperature changes and gradients became more evident in the presence of dowel misalignments. CONCLUSIONS : The results confirmed that dowel bar misalignment considerably affected the behavior of the JCP, thereby emphasizing the importance of the proper installation of dowel bars.

PURPOSES : Previously, the expansion state of the concrete pavement in which AAR occurred could not be determined. Because the current situation has not been evaluated, it has been difficult to prepare an appropriate response. In this study, a method for calculating the expansion amount of concrete pavement using the stiffness damage test (SDT) is proposed. METHODS : The SDT method was examined through a literature review. For the laboratory tests, specimens that generated AAR were produced based on the mix design (2018) of the Korea Expressway Corporation. SDT was used to calculate various mechanical properties, and their correlation with the expansion amount was reviewed. RESULTS : Using the SDT, various mechanical properties(elastic modulus, hysteresis area, plastic deformation, plastic deformation index, stiffness damage index, and nonlinear index) were calculated based on the expansion rate of the AAR. The elastic modulus was evaluated as the best predictor of the expansion rate. Thus, if the elastic modulus is calculated using SDT, a prediction equation can be used to calculate the amount of AAR expansion. This equation will need to be supplemented by further research. CONCLUSIONS : SDT was used to confirm that the expansion state due to the AAR of the concrete pavement could be indirectly evaluated. Among the mechanical properties related to SDT, the elastic modulus was found to be the most suitable for predicting the amount of expansion.

PURPOSES : This study provides fundamental information on the temperature variations in tunnel structures during severe fire events. A fire event in a tunnel can drastically increase the internal temperature, which can significantly affect its structural safety. METHODS : Numerical simulations that consider various fire conditions are more efficient than experimental tests. The fire dynamic simulator (FDS) software, based on computational fluid dynamics (CFD) and developed by the National Institute of Standards and Technology, was used for the simulations. The variables included single and multiple accidents involving heavy goods vehicles carrying 27,000 liters of diesel fuel. Additionally, the concrete material characteristics of heat conductivity and specific heat were included in the analysis. The temperatures of concrete were investigated at various locations, surfaces, and inside the concrete at different depths. The obtained temperatures were verified to determine whether they reached the limits provided by the Fire Resistance Design for Road Tunnel (MOLIT 2021). RESULTS : For a fire caused by 27,000 liters of diesel, the fire intensity, expressed as the heat release rate, was approximately 160 MW. The increase in the carrying capacity of the fire source did not significantly affect the fire intensity; however, it affected the duration of the fire. The maximum temperature of concrete surface in the tunnel was approximately 1400 ℃ at some distance away in a longitudinal direction from the location of fire (not directly above). The temperature inside the concrete was successfully analyzed using FDS. The temperature inside the concrete decreased as the conductivity decreased and the specific heat increased. According to the Fire Resistance Design for Road Tunnel (MOLIT 2021), the internal temperatures should be within 380 ℃ and 250 ℃ for concrete and reinforcing steel, respectively. The temperatures were found to be approximately 380 ℃ and 100 ℃ in mist cases at depths of 5 cm and 10 cm, respectively, inside the concrete. CONCLUSIONS : The fire simulation studies indicated that the location of the maximum temperature was not directly above the fire, possibly because of fire-frame movements. During the final stage of the fire, the location of the highest temperature was immediately above the fire. During the fire in a tunnel with 27,000 liters of diesel, the maximum fire intensity was approximately 160 MW. The capacity of the fire source did not significantly affect the fire intensity, but affected the duration. Provided the concrete cover about 6 cm and 10 cm, both concrete and reinforcing steel can meet the required temperature limits of the Fire Resistance Design for Road Tunnel (MOLIT 2021). However, the results from this study are based on a few assumptions. Therefore, further studies should be conducted to include more specific numerical simulations and experimental tests that consider other variables, including tunnel shapes, fire sources, and locations.

전 세계 대부분의 국가들은 탄소배출량을 줄이기 위한 노력을 지속하고 있다. RC구조물의 탄소배출량을 줄이기 위해 수십 년간 건설 분야의 많은 연구자들이 철근콘크리트 구조물에서 철근을 FRP보강근으로 대체하기 위한 연구를 수행하여 왔다. 북미지역을 비롯한 일부 지역에서는 해양구조물이나 도로 등에 CFRP보강근을 사용한 바 있다. 그러나 건축물에는 철근을 CFRP보강근으로 대체한 사례와 적용을 위한 연구는 거의 진행되지 않았다. 따라서 본 연구에서는 예제건물을 선정하여 철근콘 크리트구조로 설계한 후, 철근을 CFRP보강근으로 대체하여 설계함으로써, 철근콘크리트건물에서 철근을 CFRP보강근으로 대체 하였을 때 철근량을 비교하였다. 그 결과 슬래브의 철근비가 0.005미만으로서 CFRP의 사용량을 줄일 수 있을 뿐만 아니라 안전 율 측에서도 우수한 것으로 나타났다.