Enzymatic flow injection method for rapid determination of choline in urine with electrochemiluminescence detection

In order to determine trace choline in human urine, a flow injection analysis (FIA) system has been developed by coupling an enzyme reactor with an electrochemiluminescence (ECL) detector. The enzyme reactor is prepared by covalently immobilizing choline oxidase (ChOx) onto the aminopropyl-controlled pore glass beads, which are then carefully packed into a micro column. The enzyme reactor catalyzes the production of hydrogen peroxide that is in direct proportion to the concentration of choline. In this study, the enzymatically produced hydrogen peroxide was detected by an ECL detector positioned at the down stream of an enzyme react:or based on the luminol/H2O2 ECL system. Under the optimized condition, the enzymatic FIA/ECL provided high sensitivity for the determination of choline with the detection limit as low as 0.05 mu M (absolute detection limit was at sub pmol level). The method was successfully applied in the determination of choline in the samples of human urine, and the analytical results were in good agreement with those obtained by using the microbore HPLC with an immobilized enzyme reactor-electrochemical detection system.ArticleBioelectrochemistry. 79(1):147-151 (2010)journal articl

New evaluation method for the curing degree of rubber and its nanocomposites using ATR-FTIR spectroscopy

A nondestructive method of evaluating the curing degree (crosslinking density) of cured rubbers and their nanocomposites based on attenuated total reflection Fourier-transform infrared (ATR-FTIR) spectroscopy was proposed and applied to fluorine-based rubber (FKM), in which triallyl-isocyanurate (TAIC) was employed as a curing agent. ATR-FTIR spectroscopy demonstrated that the CO band in TAIC at 1699 cm−1 decreased in intensity and broadened as the curing reaction progressed. A calibration model relating the crosslinking density in the FKM with the full width at half maximum (FWHM) of the band of the CO bond was developed. The model showed good applicability to both FKM and FKM nanocomposites with various nanofillers, including single- and multi-walled carbon nanotubes (CNTs), carbon black and silica particles. Interestingly, when the fillers, especially CNTs, were added to the rubber, the FWHM was more sensitive to the change in the crosslinking density than the change in the Young's modulus

Full Report on the NIFS Fusion Engineering Research Project for the Mid-Term of FY2010-2015

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