Flame retardancy in fabric consisting of cellulosic fiber and modacrylic fiber containing fine-grained MoO3 particles

Flame retardancy of fabrics consisting of modacrylic fiber containing with various dispersed metal compounds and cellulosic fiber has been investigated by means of flame test (ISO15025 procedure A) and limiting oxygen index (LOI). It has been found that excellent flame retardancy is achieved by fine-grained MoO3 particles. The afterflame time in flame test and the LOI value are improved with decreasing particle size of MoO3. The flame retardancy of MoO3 (particle size; 0.1 µm) is comparable to that of Sb2O3. On the other hand, significant improvement in flame retardancy is not observed for other metal compounds although some metal oxides and a hydroxide in the present study are known as flame retardant or smoke suppressing agent in halogen containing polymer in previous studies. In order to clarify the mechanism of the observed flame retardancy by the addition of fine-grained MoO3, we have carried out X-ray fluorescence spectrometry (XRF) measurement of the fabric specimen after the flame test and thermogravimetric analysis (TGA) of various types of samples. These analytical data indicated that MoO3 works as halogen synergist in solid phase and the char of modacrylic fiber formed by addition of MoO3 suppresses decomposition of the cotton blended in the fabric in the range of the ignition temperature.Version of record online: 14 July 2015journal articl

2021年度 立命館大学社会システム研究所公開学術シンポジウム 「パンデミックで見えてくる社会と〈私〉」

articl

A Report on the Workshop of Teaching Portfolio in 2011

大阪府立大学工業高等専門学校は,2009年1月に全国の高等教育機関で初めて学内でティーチング・ポ ートフォリオ作成ワークショップを開催した.その後,2010年1月には第2回,8月には第3回,2011年 1月には第4回となるティーチング・ポートフォリオ作成ワークショップを開催し,現在常勤教員76名中 39名(51%)がティーチング・ポートフォリオを作成している.本稿では,2011年度に開催したワークショ ップの概要について説明した後,ワークショップ参加者の感想を報告する.departmental bulletin pape

核融合科学研究所 核融合工学研究プロジェクト 全体報告書

On the basis of the outstanding progress in high-density and high-temperature plasma experiments in the Large Helical Device (LHD) at National Institute for Fusion Science (NIFS), the conceptual design studies on the LHD-type helical fusion reactor, the FFHR series, have been conducted since 1993. In order to strongly promote this research activity in parallel with the acceleration of the related technological R&D for reactor components, the Fusion Engineering Research Project (FERP) was launched at NIFS in FY2010. The FERP consists of 13 tasks and 44 sub-tasks, each strongly assisted by domestic and international collaborations. The reactor design studies have focused on FFHR-d1, the demo-class reactor having a major radius of 15.6 m, which is four times larger than that of LHD. The similar heliotron magnetic configuration is employed to ensure steady-state operation with 3 GW self-ignited fusion power generation. The design activity has proceeded with the staged program, named “round,” that defines iterative working. The first round is to determine the basic core plasma parameters, the second is to compose all of the three-dimensional designs, the third focuses on construction and maintenance schemes, and the fourth is dedicated to passive safety. Since 2015, a multi-path strategy has been taken to include various options in the design, with FFHR-d1A as the base option. As a remarkable achievement of the reactor design, the Direct Profile Extrapolation (DPE) method is included in the helical systems code, HELIOSCOPE, in order to predict the confinement capability. The radial-build was successfully fixed and the neutronics calculation was carried out for the determined three-dimensional structure. The cost evaluation is also being conducted using these outcomes. The related R&D works in FERP are categorized into five key subjects: (1) large-scale superconducting (SC) magnet, (2) long-life liquid blanket, (3) low-activation structural materials, (4) high heat & particle-flux control, and (5) tritium and safety. Using the remarkable achievements of the related R&D works, the engineering design of FFHR-d1 defines the basic option and challenging option. While the basic option is an extension of the ITER technology, the challenging option includes innovative ideas from the following three purposes: (1) to overcome the difficulties related with the construction and maintenance of three-dimensionally complicated large structures, (2) to enhance the passive safety, and (3) to improve plant efficiency. For the superconducting magnet, the high-temperature superconductor (HTS) using ReBCO tapes is considered as an alternative (challenging) option to the cable-in-conduit conductor using low-temperature superconducting Nb3Sn strands. One of the purposes for selecting the HTS is to facilitate the three-dimensional winding of the helical coils by connecting prefabricated segmented conductors. A mechanical lap joint technique with low joint resistance has been developed and a 3 m-long short-sample conductor has successfully achieved 100 kA- current at a magnetic field of 5 T and temperature of 20 K. Further tests will be carried out in the world-largest 13 T, 700-mm bore superconducting magnet facility. For the tritium breeding blanket, we have chosen, as a challenging option, the liquid blanket with molten salt FLiNaBe from the viewpoint of passive safety. To increase the hydrogen solubility, an innovative idea to include powders of titanium was also proposed. An increase of hydrogen solubility over five orders of magnitude has been confirmed in an experiment, which makes the tritium permeation barrier less necessary for the coating on the walls of cooling pipes. The “Oroshhi-2” testing facility was constructed as a platform for international collaborations, having a twin-loop for testing both molten-salt (FLiNaK) and liquid metal (LiPb) under the perpendicular magnetic field of 3 T, the world’s largest for this purpose. For the structural material of blankets, a dissimilar bonding technique has been developed to join the vanadium alloy, NIFS-HEAT2, and a nickel alloy. For the helical built-in divertor, the diverter tiles could be placed at the backside of the blankets where the incident neutron flux is sufficiently reduced by an order of magnitude. It is thus expected that a copper-alloy could be used for cooling pipes under the bonded tungsten tile, since the maximum neutron fluence is limited to be lower than the allowable limit of ~1 dpa for copper within the operation period. We note that the peak heat flux on the helical divertor is expected to reach or exceed ~20 MW/m² because of the non-uniform strike point distributions, and effective removal of this heat flux is a concern. The maintenance scheme for the full-helical divertor is also a critical issue. To solve these problems, a new concept of liquid divertor has been proposed as a unique idea. Ten units of molten-tin shower jets (falls) are proposed to be installed on the inboard side of the torus to intersect the ergodic layer. It is considered that the vertical flow of tin jets could be stabilized using an internal flow resistance such as wires, chains, and tapes imbedded. In case the liquid divertor actually works, the full-helical divertor would become less necessary, though it should still be situated at the rear. Neutral particles are expected to be efficiently evacuated through the gaps between liquid metal showers. The mission of the NIFS FERP is to establish the scientific and technological basis that demonstrates the engineering feasibility of the helical fusion reactor and to promote the entire fusion engineering research toward the realization of fusion reactors in the mid-21st century. The progress of the NIFS FERP during the second six-year mid-term period in Japan for FY2010-2015 is overviewed in this full report. The numerical targets for the major components, which are the SC magnet, the in-vessel components, and the blanket, were compiled in FY2016,and its summary is also added in this report.research repor

Effect of Magnetic Field Distribution on Recovery Currents of NbTi Superconducting Conductors

ORCID　 0000-0003-1454-8117NbTi conductors for helical coils of the Large Helical Device were developed to satisfy cold-end stability. Their recovery currents were measured with a conductor test facility with 9 T split coils at the National Institute for Fusion Science. The measured recovery currents were higher by 15 to 20 % than that calculated from Maddockʼs equal area theorem with the measured conductor resistance and heat transfer. We have proposed an analytical method to estimate the recovery current in a finite magnetic field using the temperature distribution that is calculated with representative thermal conductivity, resistivity, and heat transfer. In order to check the validity of this method, we carried out simulation with a finite-difference method. The results revealed that the proposed analytical method is applicable with slight underestimation as long as the resistivity can be fitted by a function of temperature only. In addition, the necessary length of the magnetic field higher than 95 % is around three times of the temperature characteristic length of the conductor for measurement of recovery currents with an overestimation of less than 5 %.journal articl

Residual magnetic field induced by superconducting magnets of Large Helical Device

0000-0003-1454-8117The residual magnetic field was measured to investigate its source for the Large Helical Device, the superconducting magnet system of which consists of two Helical Coils (HCs), two Inner Vertical (IV) coils, two Inner Shaping (IS) coils, and two Outer Vertical (OV) coils. NbTi cable-in-conduit conductors were adopted for the IV, IS, and OV coils. Firstly, Hall probes were installed at five periodic positions on the mid-plane of the inner cylinder of the cryostat. Since the residual field was changed by around 0.1 mT at all the positions during the warm-up of the superconducting coils, a major part of the residual field had to be induced by magnetization of the coils. In the next campaign (cool-down, plasma experiment, and warm-up), the Hall probes were moved to the five different vertical positions in order to measure the distribution of the residual field. Calculation of the residual field has been carried out under the assumption that NbTi filaments in each conductor are magnetized in the same direction as the field at the center of the conductor during excitation. From comparison between the measured and calculated values, we conclude that the residual field from the coils that had been excited to high currents should be reduced by around 35%, due to the self-field in strands in the conductor. The best fitted critical current densities of the IV and IS coils are 1.31 and 2.81 × 1010 A/m2, respectively, which are consistent with the field dependence obtained from the magnetization curve of each strand.journal articl

Increased expression of EphA7 correlates with adverse outcome in primary and recurrent glioblastoma multiforme patients-0

G in both tumor cells and endothelial cells (B). MVD in GBM by immunohistochemical staining for vWF, microvessels are represented by brown clusters, which stand out sharply from other tissues. Low tumor vascularity (C) in GBM with low expression of EphA7 as shown A. In contrast, microvessel density was relatively high (D) in GBM with high expression of EphA7 as shown B. Negative staining in normal brain tissue (E). Original magnification, ×400 (A, B, E) and ×200 (C, D).<p><b>Copyright information:</b></p><p>Taken from "Increased expression of EphA7 correlates with adverse outcome in primary and recurrent glioblastoma multiforme patients"</p><p>http://www.biomedcentral.com/1471-2407/8/79</p><p>BMC Cancer 2008;8():79-79.</p><p>Published online 25 Mar 2008</p><p>PMCID:PMC2292196.</p><p></p

Upgrade Plan on NIFS Superconducting Magnet Test Facility

In 1991, the superconducting magnet test facility had been constructed in National Institute for Fusion Science. The facility consists of a helium liquefier/refrigerator with the cooling capacity of 600 W at 4.5 K, large cryostats, a mechanical testing machine, DC power supplies with the maximum current of 75 kA and a distributed control system. So far, the development of the superconducting coils for the Large Helical Device had been conducted and many collaborative works have been carried out. However, the test facility with higher bias filed is needed to develop superconducting coils for fusion reactors and the cooling system, which can supply coolant of various temperature to test samples, is required to apply HTS to large-scale conductors. Therefore, the test facility will be upgraded to promote the development of superconducting coils for fusion reactors. The maximum bias field will be upgraded from 9 T to 15 T to investigate the performance of superconductors under the higher bias field and the existing helium liquefier/refrigerator will be replaced with a variable temperature one to test those under various temperature environment. Consequently, it is expected to progress the development of the superconductors for fusion reactors and the development of large-scale HTS conductors.journal articl

8 ワルファリンと抗血小板薬併用症例の検討(I.一般演題,第269回新潟循環器談話会)

departmental bulletin pape