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.

Radioactive liquid waste generated during the operation of domestic nuclear power plants is treated through a somewhat different liquid radwaste system (LRS) for each plant. Prior to the introduction of standard nuclear power plants, LRS used a concentrated water dry system (CWDS) to evaporate liquid waste and manage it in the form of dry powder. The boron-containing radioactive liquid waste dry powder was solidified using paraffin from 1995 to 2010, and about 3,650 drums (based on 200 L) of paraffin solidified drums are currently stored in nuclear power plants. Paraffin solidification drums do not meet the acceptance criteria for radioactive waste repositories because it is difficult to secure the homogeneity of the solidified body and there are concerns about leaching of radioactive waste due to the low melting point of paraffin. In order to solve this problem and safely permanently dispose of paraffin solidification drums, the characteristics of dry powder paraffin solidification drums containing boron-containing radioactive liquid waste must be analyzed and appropriate treatment technology utilizing the results must be introduced. This study analyzes the physical properties of paraffin, the chemical properties of boron-containing radioactive waste dry powder, and the physicochemical properties of paraffin solidification powder, and proposes an appropriate alternative technology for treating boron-containing radioactive waste dry drum. When disposing of the paraffin solidification drum with boron-containing radioactive liquid waste dry powder, the solidification body must be effectively withdrawn from the drum and the paraffin must be completely separated from the solidification body. When disposing the drum, the solidified material must be effectively extracted from the drum and the paraffin must be completely separated from the solidified material. Afterwards, the paraffin must be self-disposed, and the radioactive waste must be disposed of in accordance with acceptance criteria of repository. We looked at how each characteristic of the paraffin solidification drum with boron-containing radioactive liquid waste dry powder can be utilized in each of the above treatment processes.

The large rectangular and cylindrical concrete drums are stored in nuclear power plant (NPP) for a long time. At the early stage of NPP operation, the treatment technology of boron concentrates and spent resin was not well developed, when compared to current system. Since the waste acceptance criteria (WAC) of the disposal facility was not established, the boron concentrates and spent resins were packaged in 200 L drum. Some of the 200 L drums, which contain relatively high dose rate radioactive waste, were stored in large concrete drum. The concrete drum offers superior shielding effect and allows reduction of radiation exposure to workers. The WAC requires various characteristics: radiological characteristics, physical characteristics, chemical characteristics, etc. The non-destructive method allows the rapid evaluation and estimation of the concrete structure. Also, it is expected that the large concrete exhibits integrity after the measurements. In this paper, the non-destructive method to understand the large rectangular and cylindrical drum is systematically studied. The advantage and disadvantage of the non-destructive methods were compared in this paper. In addition, the optimized methodology to characterize the radioactive waste containing large rectangular and cylindrical drum will be discussed in this paper.

In order to permanently dispose of radioactive waste drums generated from nuclear power plants, disposal suitability must be demonstrated and the nuclides and radioactivity contained in the waste drums, including those in the shielding drums, must be identified. At present, reliable measurements of the nuclide concentration are performed using drum nuclide analysis devices at power plants and disposal facilities during acceptance inspection. The essential functions required to perform nuclide analysis using the non-destructive assay system are the correction for self-attenuation and the dead time correction. Until now, measurements have mainly been performed for drums containing solid waste such as DAW drums using SGS calibration drums with ordinary iron drums. However, for drums containing non-uniform radioactive waste, such as waste filters embedded in cement within shielding drums, a separate calibration drum needs to be produced. In order to produce calibration drums for shielded and embedded waste drums, the design considered the placement of calibration sources, setting of shielding thickness, correction for medium density, and cement mixing ratio. Based on these considerations, three calibration drums were produced. First, a shielding drum with an empty interior was produced. Second, a density correction drum filled with cement was produced to create apparent density on the surface of the shielding drum. Third, a physical model drum was produced containing a mock waste filter and cement filled in the shielding drum.

The spent filters stored in Kori Unit 1 are planned that compressed and disposed for volume reduction. However, shielding reinforcement is required to package high-dose spent filters in a 200 L drum. So, in this study suggests a shielding thickness that can satisfy the surface dose criteria of 10 mSv·h−1 when packaging several compressed spent filters into 200 L drums, and the number of drums required for the compressed spent filter packaging was calculated. In this study, representative gamma-emitting nuclides in spent filter are assumed that Co-60 and Cs-137, and dose reduction due to half-life is not considered, because the date of occurrence and nuclide information of the stored spent filter are not accurate. The shielding material is assumed to be concrete, and the thickness of the shielding is assumed to 18 cm considering the diameter of the spent filter and compression mold. Considering the height of the compressed spent filter and the internal height of the shielding drum, assuming the placement of the compressed spent filter in the drum in the vertical direction only, the maximum number of packaging of the compressed spent filter is 3. When applying a 18 cm thick concrete shield, the maximum dose of the spent filter can packaged in the drum is 125 mSv·h−1, so when packaging 3 spent filters of the same dose, the dose of a spent filter shall not exceed 41 mSv·h−1 and not exceed 62 mSv·h−1when packing 2 spent filters. Therefore, the dose ranges of spent filters that can be packaged in a drum are classified into three groups: 0–41 mSv·h−1, 41–62 mSv·h−1, and 62–125 mSv·h−1based on 41 mSv·h−1, 62 mSv·h−1, and 125 mSv·h−1. When 227 spent filters stored in the filter room are classified according to the above dose group, 207, 3 and 4 spent filters are distributed in each group, and the number of shielding drums required to pack the appropriate number of spent filters in each dose group is 75. Meanwhile, 8 spent filters exceeding 125 mSv·h−1 and 5 spent filters that has without dose information are excluded from compression and packaging until the treatment and disposal method are prepared. In the future, we will segmentation of waste filter dose groups through the consideration of dose reduction and horizontal placement of compressed spent filters, and derive the minimum number of drums required for compressed spent filter packaging.

Currently, KHNP has 24 operating nuclear power plant units with a toal combined capacity of about 23 GWe and two units are under construction. However, permanent stop of Kori unit 1 nuclear power plant was decided in 2017. Accordingly, interest in how to dispose of waste stored inside a permanently stopped nuclear power plant and waste generated as decommissioning process is increasing. KHNP CRI is conducting research on the advancement of plasma torch melting facilities for waste treatment generated during the plant decommissioning and operation period. Plasma torch melting facility is composed of various equipment such as a melting furnace (Melting chamber, Pyrolsis chamber), a torch, an exhaust system facility, a waste supply device, and other equipment. In demonstration test, concrete waste was put in a 200 L drum to check whether it can be pyrolyzed using a plasma torch melting facility. Reproducibility for waste treatment in the form of a 200 L drum and discharge of molten slag could be confirmed, the amount of concrete waste in 200 L Drum that could be treated according to power of plasma torch was confirmed. This demonstration test confirmed the field applicability and stability of plasma torch melting facility, and improved expectations for long-term operation.

With the development of the nuclear industry and the increase in the use of radioactive materials, the generation of radioactive waste is increasing. As the generation of radioactive waste increases, the occurrence of related safety accidents is also increasing, and it is necessary to develop a radioactive waste monitoring technology to prevent such accidents in advance and efficiently manage radioactive waste. In Information and Communication Technology (ICT), various ICT technologies such as Internet of Things (IoT), Augmented Reality (AR), and Virtual Reality (VR) that can help with the safety management of these radioactive wastes are being developed. In this study, a radioactive waste monitoring technology was developed using ICT technology, such as management of the entire cycle history of waste using Quick Response (QR) codes, and development of AR visualization technology for small packages of radioactive waste. In addition, by using IoT technology to collect desired data from sensors and store the results, after the waste drum is loaded in the waste storage, a technology was developed to track and monitor the history and movement of the waste drum from repackaging to transfer to the storage. The data required for monitoring the radioactive waste drum includes location information, whether the drum is open or closed, temperature and humidity, etc. To collect this information, a drum monitoring technology was built with a 2.4 G wireless router, an anchor constituting a virtual zone, a tag to be mounted on the drum container, and a WNT server that collects sensor data. The network tool provided by WirePas was used for network configuration, and the status of gateways and nodes can be monitored by interworking with the WNT server. The configured IoT sensor technology were tested in a waste storage environment. Four anchors were installed and linked to the network to match the virtual zone and the real storage zone, and it was confirmed whether the movement of the tag was recorded on the network while moving the tag including the IoT sensor for analyzing location information. Based on these research results, it can contribute to the safety management of radioactive waste and establishment of Waste Acceptance Criteria (WCP) by and managing the history and monitoring the waste in the entire cycle from repackaging to disposal.

During decommissioning of a nuclear power plant, a large amount of radioactive waste is produced, and it is known to cost more than 300 billion won to dispose the waste. To reduce the disposal cost, it is essential to minimize the number of radioactive waste drums, which can be achieved by detecting and removing hotspot contaminations in the radioactive waste drums. Therefore, a Compton CT system for radioactive waste monitoring is under development, which provides the images of both the internal structure of the drum and the radioactive hotspot(s) in the drum. Based on the acquired information, the activity of hotspots can be estimated. The performance of the system is affected by various geometry factors. Therefore, it is essential to determine optimal configuration by evaluating the effects of the factors on the performance of the system. In the present study, we determined the optimum value of the factors and then predicted the performance of the optimized system by using a simulator based on the Geant4 Monte Carlo simulation. For optimization, the factors were evaluated in terms of structural similarity index measure (SSIM) and measurement time. The considered factors were the activity of the CT source, source to object distance (SOD), object to detector distance (ODD), and projection angle. The simulation result showed that the activities of the CT sources were determined as 23 mCi for 137Cs and 9.6 mCi for 60Co. The optimal SOD and ODD were 180 cm and 40 cm, respectively. The optimal projection angle was evaluated as 4° since it achieves the SSIM of 0.95 faster than other projection angles. With the optimized parameters, the performance of the system was evaluated using the IAEA gamma CT standard phantom containing a hotspot of 137Cs (7.02 μCi). The Compton image was reconstructed using the back-projection algorithm, and the CT image was reconstructed using the filtered back-projection algorithm. The result showed that the location of the hotspot in the Compton image was well identified at the true position. The acquired CT image also well represented the internal structure of the phantom, and the estimated mean linear attenuation coefficient value (μ= 0.0789 cm−1) of the phantom was close to the true value (μ= 0.0752 cm−1). In addition, the hotspot activity estimated by combining the information of the Compton image and CT image was 8.06 μCi. Hence, it was found that the Compton CT system provides essential information for radioactive waste drums.

이 논문은 드럼 세트를 구성하는 악기들 중 스네어 드럼에 사용되는 진동제어장치의 크기와 질량의 변화에 따라 음량과 음색 및 엔벨로프가 어떻게 변화되는지에 대해 연구하였다. 스네어 드럼을 사용하여 연주하는 드럼 연주자 및 음향 엔지니어들에게 진동제어장치의 변화에 따라 도출되는 소리의 특성을 제공하여 표현하고 싶은 스네어 음색을 구사할 수 있는 정보를 제공하는 목적을 가지고 있으며 스네어 드럼의 음색을 구사하는 능력이 향상된다는 것은 음악의 장르와 스타일에 따라 연주자 및 음향 엔지니어의 표현 능력이 향상될 수 있는 긍정적 영향을 끼칠 수 있다. 실험 결과 진동제어장치를 사용하여 면적과 질량의 변화를 준 환경에서 라우드 니스의 변화는 발생하지 않았으며, 주파수 응답의 변화는 미미한 수준이었고, 엔벨로프의 길 이에서는 큰 변화가 발생하였다. 실험과정에서 실제로 듣는 청감상의 스네어 드럼 음색 변화는 매우 크게 나타났으며 특히 질량보다는 면적의 변화에 따른 실험환경에서 큰 음색 변화가 있었다. 결과적으로 진동제어장치는 실질적으로 엔벨로프에만 관여하며, 스네어 드럼은 엔벨 로프의 길이가 짧아 엔벨로프의 앞부분에서 발생하는 음색이 더 중요하다는 일부 인식과 달 리, 엔벨로프의 뒷부분도 인간이 지각하는 음색에 큰 영향을 준다는 연구들이 도출되었다.

The mesh selectivity of a drum shaped pot for finely-striate buccinum (Buccinum striatissimum) was conducted a total of eight times with four different mesh sizes (22, 35, 50 and 60 mm) from May to September, 2019 in the eastern coastal waters of Korea. The size selectivity analysis was estimated by the SELECT method to express logistic selectivity curves. In the results, the catch of finely-striate buccinum was occupied about 90% in the total catch weight. The equation of the master curve of selectivity was estimated to s(R) = exp(-7.778R+9.983)/[1+exp(-7.778R+9.983)]. The relative shell height of 50% selection was 1.284 and the selection range (SR) was 0.282. The optimal mesh size for 50% selection on the minimum maturation size (75 mm, Shell height) was estimated more than 60 mm by the master selectivity curve.

현대의 대중음악은 단순하고 반복적인 음악 스타일이 주를 이루고 있다. 2014년부터 유튜브 영상을 시작으로 활동한 독일의 드러머 아니카 닐즈(Anika Nilles)는 현대의 대중음악 스타일과 다른 홀수 잇단음표를 음악에 활용하여 자신의 음악을 독보적으로 만들어냈다. 가장 기본적이고 대중음악에서도 자주 활용되는 3잇단음표를 대표적으로 활용했으며 또한 음악에서 자주 쓰이지 않고 희소성 있는 5잇단음표, 7잇단음표, 9잇단음표까지 연주에 활용했다. 본 연구의 목적은 닐즈가 활용한 홀수 잇단음표의 활용을 분석하고 그 효과를 도출해 현대음악에서의 발전 가능성을 제시하였고 연구 방법은 5잇단음표, 7잇단음표, 9잇단음표의 3가지 분류로 나누어 각각에서 활용된 기법에 대해 분석하였다. 분석 결과 닐즈는 잇단음표를 활용한 기법의 특징을 두드러지게 표현하고 연주했다는 결론에 이르렀으며 기법의 다양한 적용 방법도 가지고 있었다. 결과적으로 홀수 잇단음표를 활용한 그녀의 연주 기법으로 인해 창의적으로 다양한 콘텐츠의 음악을 만들어 낼 수 있었다는 결론을 도출하였다.

In this study, high thermal efficiency and eco-friendly coffee roaster is developed. The core element of coffee roaster is a roasting drum and most of the coffee roasting drum is a flat cylindrical barrel type. From the traditional shape of cereal puffing machine, we got the idea of the roasting drum. Oak barrel shape is well known high thermal efficiency system in korean heating cook system. Therefore, we designed the coffee roaster containing the oak barrel shaped drum and combustor nozzles which has an ability of roasting 3kg of coffee beans within 10 minutes of roasting time. With our developed coffee roasting machine, the heating speed of roasting process is very fast, and the roasting of coffee beans can be quickly finished. We experimented the roasting time and compared the results with other coffee roasting machines of the same amount of roasting capacity. The results were outstanding in roasting time and the quality of roasting. In spite of smaller fuel consumption, the roasting time was more shorter than flat cylindrical roasting drum.

1980년대 영국은 당시 하우스와 테크노음악이 성행하였다. 대세의 음악을 틀지 않던 몇몇의 디제이들이 차별성을 두기 위해 50~60년대 발매 되었던 재즈음악에 디스코, 하우스, 힙합 리 듬을 믹스 하였고 그 과정에서 애시드 재즈라는 장르가 탄생하게 되었다. 이러한 음악적 배경 으로 애시드 재즈에서의 드럼 연주 스타일은 지속적인 안정감과 긴장감을 표현하기 위해 고스 트 노트와 엑센트를 적극 활용한다. 밴드 자미로콰이의 드러머인 데릭 멕켄지는 엑센트 기법 을 활용하여 다양한 리듬 패턴을 만들어 섹션마다 엑센트 프레이즈의 변화를 주었고 의도에 따라 엑센트 기법을 선택적으로 사용하여 연주에서의 다이나믹을 형성하였다. 본 연구는 밴드 자미로콰이의 드럼 연주기법을 분석하여 대중음악에서 크게 자리 잡는 섹션인 벌스와 코러스 에 대입할 확장 방법을 제시하고 엑센트의 음악 표현 기법은 특정 음표를 강조하여 음악의 중 심 요소인 리듬의 생명력을 부여한다는 결과를 도출한다.