This paper explores the potential application of carbon nanotubes (CNTs) in the construction industry, as CNTs can effectively serve as nano-fillers, bridging the voids and holes in cement structures. However, the limited dispersibility of CNTs in water necessitates the use of dispersing agents for achieving uniform dispersion. In this study, two kinds of cement superplasticizers, polycarboxylate ether (PCE) and sulfonated naphthalene formaldehyde (SNF) were employed as dispersing agents to improve the interfacial affinity between CNTs and cement, and to enhance the strength of the cement nanocomposites. Contact angle experiments revealed that the utilization of PCE and SNF effectively addressed the interface issues between CNTs and cement. As a result, the cement nanocomposite with a CNT to PCE ratio of 1:2 exhibited an approximately 6.6% increase in compressive strength (73.05 MPa), while the CNT:SNF 1:2 cement composite showed a 4.7% increase (71.72 MPa) compared to plain cement (68.52 MPa). In addition, the rate of crack generation in cement nanocomposites with CNTs and dispersing agents was found to be slower than that of plain cement. The resulting cement nanocomposites, characterized by enhanced strength and durability, can be utilized as safer materials in the construction industry.

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.

산오이풀(Sanguisorba hakusanensis)은 한국의 자생식물 이며 정원소재로써 가치가 있지만, 생육 및 생리적 특성 및 정 원 적응 여부에 대하여 알려진 정보가 많지 않아 이용에 어려 움을 겪고 있다. 본 연구에서는 자생식물인 산오이풀의 관수 주기 및 NaCl 농도에 따른 생장, Fv/Fm, NPQ, 성분 변화, 무기성분 변화를 조사하여 내건 및 내염성 보유 여부, 생육 한 계 범위, 스트레스 환경에서 생육을 유지하기 위한 반응을 파 악하고자 했다. 실험 결과 NaCl 무처리구의 관수주기에 따른 성분 분석에서 엽록소 함량의 감소를 제외하고 유의한 차이가 나타나지 않았으나 이는 토양수분함량이 건조 스트레스를 유 발할 정도로 감소하지 않았기 때문으로 보인다. 염 처리에서 는 2주 이후 급격한 스트레스 반응이 나타났고 3주차부터 고 사하기 시작하여 6주차에 모든 개체가 최종 고사했다. 이러한 결과는 2주까지 염 스트레스에 의해 유발되는 2가지 스트레 스 중 초기에 나타나는 삼투 스트레스에는 저항하였으나 이후 나타나는 NPQ의 감소 등 이온 스트레스에 의해 유발된 광합 성 기구 붕괴로 인해 정상적인 생육을 유지할 수 없었기 때문 에 나타난 것으로 보인다. 그러나 무기이온 분석은 이온 스트 레스에 저항하기 위한 메커니즘의 존재 가능성을 시사하였다. 상대적으로 염 농도가 낮을 때에는 세포내 Ca2+ 및 K+ 수준이 높았는데, 이는 Ca2+ 수준이 높아짐에 따라 Na+를 세포 밖으 로 방출시키는 단백질, Na+를 K+와 함께 수송하는 단백질이 기능하여 Na+축적을 지연시키는 반응이 있었음을 시사한다. 그러나 NaCl을 고농도로 처리했을 때는 이러한 반응이 관찰 되지 않았다. 따라서 산오이풀은 염 스트레스에 의해 야기되 는 삼투 스트레스에 강한 저항성을 가지고 있고 이온 독성을 줄이기 위한 메커니즘으로 Na+ 세포내 축적을 지연시키는 것으로 보이지만, 심한 염 스트레스를 받았을 때 나타나는 급격 한 반응에서 이러한 메커니즘이 기능하지 못하고 이온독성에 매우 취약한 것으로 여겨진다. 본 연구를 통해 자생식물인 산 오이풀의 활용을 늘리는 데 기초적인 자료를 제공할 수 있을 것으로 생각된다.

The thermal treatment of radioactive waste attracts great attention. The thermal treatment offers lots of advantages, such as significant volume reduction, hazard reduction, increase of disposal safety, etc. There are various thermal technologies to waste. The developed technologies are calcination, incineration, melting, molten salt oxidation, plasma, pyrolysis, synroc, vitrification, etc. The off-gas treatment system is widely applied in the technologies to increase the safety and operation efficiency. The thermal treatment generates various by-product and pollutants during the process. The dust or fly ash are generated as a particulate from almost every radioactive waste. The treatment of PVC related components generates hydrogen chloride, which usually brings corrosion of facility. The treatment of rubber and spent resin generates sulfur oxide, SOx. The treatment of nitrile rubber generates nitrogen oxide, NOx. The incomplete combustion of radioactive waste usually generates carbon oxide, COx. The process temperature also affects the generation of off gas, such as NOx and/or COx. Various off gas treatment components are organized for the proper treatment of the previously mentioned materials. In this study systematical review on off gas treatment will be reported. Also, worldwide experiences and developed facility will be reported.

The primary purpose of high temperature process of radioactive waste is to satisfy the waste acceptance criteria and volume reduction. The WAC offers the guideline of waste form fabrication process. The WAC is defined as quantitative or qualitative criteria specified by the regulatory body, or specified by and operator and approved by the regulatory body, for radioactive waste to be accepted by the operator of a repository for disposal, or by the operator of a storage facility for storage. The main objective of WAC is to protect staff and general public and environment by the containment of radioactive material, limit external radiation level, and prevent criticality. The WAC also offers systematic management of radioactive waste by standardization of waste management operations, facilitation waste tracking, ensure safe and effective operation of operating facilities, etc. Since the high temperature process for radioactive waste is considered in many countries, lots of codes and standards are considered. In many WACs, compressive strength, thermal cycle stability, radiation exposure stability, free liquid, and leachability are evaluation to understand the effect of solidified form to the disposal facility. In this paper, systematical review on waste form will be discussed. In addition, brief result of characterization of waste form will be compared.

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.