Updated on 2022/05/19

写真a

 
FURUYA, Masahiro
 
Affiliation
Faculty of Science and Engineering, Graduate School of Advanced Science and Engineering
Job title
Professor(without tenure)

Research Institute

  • 2020
    -
    2022

    理工学術院総合研究所   兼任研究員

Degree

  • Delft University of Technology   Ph.D.

Research Experience

  • 2019.04
    -
    Now

    Waseda University   Cooperative Major in Nuclear Energy, Graduate School of Advanced Science and Engineering   Professor

  • 2014.07
    -
    Now

    Central Research Institute of Electric Power Industry   Nuclear Technology Research Laboratory   Deputy Associate Vice President

  • 2016.04
    -
    2020.03

    Tokyo Institute of Technology   Major in Nuclear Engineering, Department of Transdisciplinary Science and Engineering Graduate School of Environment and Society   Visiting Professor

  • 2015.11
    -
    2016.03

    東京工業大学大学院   原子核工学専攻   連携教授

  • 1998.07
    -
    2014.06

    (財)電力中央研究所   原子力技術研究所   主任研究員

  • 1993.04
    -
    1998.06

    Central Research Institute of Electric Power Industry   Komae Research Laboratory   Research Scientist

  • 1992.07
    -
    1992.11

    Argonne National Laboratory   Reactor Engineering Division   Research Scientist

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Professional Memberships

  •  
     
     

    ATOMIC ENERGY SOCIETY OF JAPAN

  •  
     
     

    THE JAPANESE SOCIETY FOR MULTIPHASE FLOW

  •  
     
     

    THE HEAT TRANSFER SOCIETY OF JAPAN

  •  
     
     

    THE JAPAN SOCIETY OF MECHANICAL ENGINEERS

 

Research Areas

  • Analytical chemistry

  • Energy chemistry

  • Inorganic materials and properties

  • Composite materials and interfaces

  • Fluid engineering

  • Thermal engineering

  • Marine engineering

  • Nuclear engineering

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Research Interests

  • 放射線誘起表面活性

  • 放射線

  • 海洋工学

  • 放射線,X線,粒子線

  • 船舶・海洋構造物

  • 材料加工・処理

  • メカニズム

  • 放射能

  • 防食技術

  • 海洋保全

  • 酸化金属皮膜

  • 熱流動

  • ステンレス

  • 原子炉防食

  • 放射線・X線・粒子線

  • 金属材料

  • 応力腐食割れ

  • 電気化学

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Papers

  • Measurement of forced convection subcooled boiling flow through a vertical annular channel with high-speed video cameras and image reconstruction

    Atsushi Ui, Masahiro Furuya, Takahiro Arai, Kenetsu Shirakawa

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   59 ( 2 ) 148 - 162  2022.02

     View Summary

    In order to develop models regarding subcooled boiling flow and its heat transfer, we conducted subcooled boiling experiments to investigate subcooled bubble formation and measured development process of the bubbles using a test loop with a vertical annulus flow path under atmospheric conditions. Two-phase flow parameters regarding subcooled boiling, such as instantaneous bubble velocity, turbulent velocity components, void fraction, onset of nucleate boiling (ONB) and onset of significant void (OSV), were obtained with two high-speed video cameras in high resolution in temporally and spatially. The identical bubbles were identified from the two video images taken by the high-speed video cameras, and the trajectories were reconstructed in three-dimensional. The instantaneous bubble velocities and turbulence velocity components regarding bubble transport were quantified with the trajectory data. It was observed that the bubbles were rising along the surface of the heater rod moved largely in the circumferential direction, and it was found from the result that velocity component for circumferential direction was enough large compared with that for radial direction.

    DOI

  • Controlling Microdroplet Inner Rotation by Parallel Carrier Flow of Sesame and Silicone Oils

    Hibiki Yoshimura, Daiki Tanaka, Masahiro Furuya, Tetsushi Sekiguchi, Shuichi Shoji

    Micromachines   13 ( 1 )  2021.12

     View Summary

    We developed a method for passively controlling microdroplet rotation, including interior rotation, using a parallel flow comprising silicone and sesame oils. This device has a simple 2D structure with a straight channel and T-junctions fabricated from polydimethylsiloxane. A microdroplet that forms upstream moves into the sesame oil. Then, the largest flow velocity at the interface of the two oil layers applies a rotational force to the microdroplet. A microdroplet in the lower oil rotates clockwise while that in the upper oil rotates anti-clockwise. The rotational direction was controlled by a simple combination of sesame and silicone oils. Droplet interior flow was visualized by tracking microbeads inside the microdroplets. This study will contribute to the efficient creation of chiral molecules for pharmaceutical and materials development by controlling rotational direction and speed.

    DOI

  • Validation of droplet-generation performance of a newly developed microfluidic device with a three-dimensional structure

    Yoshito Nozaki, Dong Hyun Yoon, Masahiro Furuya, Hiroyuki Fujita, Tetsushi Sekiguchi, Shuichi Shoji

    Sensors and Actuators A: Physical   331   112917 - 112917  2021.11

     View Summary

    We fabricated a microfluidic device with a three-dimensional (3D) structure and verified its droplet-generation performance for the stable production of droplets of around 10 μm in size. We compared the performance of the 3D device with that of conventional simple T-junction and cross-junction structures. The continuous phase sheared the dispersed phase into droplets from eight directions in the 3D device, compared with only one direction in the T-junction device and two in the cross-junction device. Droplets were produced efficiently over a wide range of fluid properties and flow conditions with the 3D device, unlike with the two conventional planar devices. Fluidic experiments were conducted using mineral oil with a surfactant as the continuous phase, deionized (DI) water as the dispersed phase, and DI water with glycerin to change the viscosity of the dispersed phase. The minimum droplet length was 47.2 μm in the T-junction device, 39.0 μm in the cross-junction device, and 22.4 μm in the 3D device when using a water and glycerin mixture with a viscosity of 9.0 mPa·s. Compared with the conventional devices, smaller droplets were produced using our 3D device, indicating that it has excellent droplet-generation performance.

    DOI

  • Measurement of forced convection subcooled boiling flow and rod surface temperature distribution

    Atsushi Ui, Masahiro Furuya, Takahiro Arai, Kenetsu Shirakawa

    Nuclear Engineering and Design   381  2021.09

     View Summary

    In order to obtain high-resolution data for modelling of boiling two-phase flow and its validation, we designed and constructed a test loop with a vertical annulus flow path and conducted subcooled boiling experiments to investigate subcooled bubble incipience and its development process under atmospheric condition. Three kinds of the state-of-the art measurement techniques were applied to quantify key parameters such as radial and vertical distributions of void fraction, bubble velocity, interfacial area concentration (IAC), Sauter mean diameters, high-resolution temperature distribution on rod surface, bubble transport behavior, and turbulent velocity components as well as onset of nucleate boiling (ONB), and onset of significant void (OSV).

    DOI

  • Study on improvement for the prediction accuracy of natural circulation flow rate by investigating void fraction correlation

    Shunsuke Yoshimura, Takuma Yamaguchi, Keisuke Ino, Masahiro Furuya, Shinichi Morooka

    Nuclear Engineering and Design   380  2021.08

     View Summary

    Natural circulation is a key technology for developing the molten core cooling system without an external power source from the lessons of the severe accident at Fukushima-Daiichi Nuclear Power Station. This study is devoted to quantify the void fraction which is an important parameter for the driving force of natural circulation flow, and to evaluate the effect of the void fraction correlation on the prediction accuracy of the natural circulation flow rate. Test was conducted at atmospheric pressure and room temperature, using the upward air–water two phase flow. Vertical tubes with an inner diameter of 36 and 25 mm were used as the test section. The void fraction was measured by three different methods: quick-closing valve method, pressure drop method, and conductive void-probe method. The following conclusions are obtained from this study: (1) The data of the natural circulation flow rate, void fraction and pressure drop for the upward air–water two phase flow at atmospheric pressure and room temperature were obtained to develop and verify the new model. (2) By improving the void correlation, it was found that the prediction accuracy of the natural circulation flow rate could be improved by about 10% to 5%, that is, the prediction error can be halved in the range of this study. (3) The natural circulation flow rate for 25 mm test section was saturated with increasing the air flow rate at higher air flow condition. The model cannot predict this tendency. From the point of design of the actual molten core cooling system, the model improvements in this region are necessary in the future.

    DOI

  • Efficient Generation of Microdroplets Using Tail Breakup Induced with Multi-Branch Channels

    Daiki Tanaka, Satsuki Kajiya, Seito Shijo, Dong Hyun Yoon, Masahiro Furuya, Yoshito Nozaki, Hiroyuki Fujita, Tetsushi Sekiguchi, Shuichi Shoji

    Molecules   26 ( 12 ) 3707 - 3707  2021.06

     View Summary

    In recent years, research on the application of microdroplets in the fields of biotechnology and chemistry has made remarkable progress, but the technology for the stable generation of single-micrometer-scale microdroplets has not yet been established. In this paper, we developed an efficient and stable single-micrometer-scale droplet generation device based on the fragmentation of droplet tails, called "tail thread mode", that appears under moderate flow conditions. This method can efficiently encapsulate microbeads that mimic cells and chemical products in passively generated single-micrometer-scale microdroplets. The device has a simple 2D structure; a T-junction is used for droplet generation; and in the downstream, multi-branch channels are designed for droplet deformation into the tail. Several 1-2 mu m droplets were successfully produced by the tail's fragmentation; this continuous splitting was induced by the branch channels. We examined a wide range of experimental conditions and found the optimal flow rate condition can be reduced to one-tenth compared to the conventional tip-streaming method. A mold was fabricated by simple soft lithography, and a polydimethylsiloxane (PDMS) device was fabricated using the mold. Based on the 15 patterns of experimental conditions and the results, the key factors for the generation of microdroplets in this device were examined. In the most efficient condition, 61.1% of the total droplets generated were smaller than 2 mu m.

    DOI

  • Evaluation of structural effect of BWR spacers on droplet flow dynamics

    Riichiro Okawa, Masahiro Furuya, Takahiro Arai, Tsugumasa Iiyama

    Nuclear Engineering and Design   377  2021.06

     View Summary

    We have established an experimental system to visualize a droplet flow in a simulated BWR fuel sub-channel optically and measure the diameter and velocity of droplet after passing through a spacer. For representative spacers of ferrule and grid type, an effect of them on downstream droplets was evaluated with the experimental system. When a ferrule type spacer was simulated and implemented in both the center and side sub-channel, the vertical velocity of droplets got faster especially in the range of small diameter compared to the case of no spacer. When a grid type spacer was simulated and implemented in the center sub-channel especially, a large dispersion of vertical velocity of droplets occurred especially in the range of small diameter compared to the case of no spacer. By a computational fluid dynamics analysis for gas phase flow to drive the droplets in the sub-channel, it was confirmed qualitatively that the characteristics of droplet behavior observed in this experiment were dependent on the structure and geometry of spacer and sub-channel. Furthermore, it was revealed that a relation between a droplet diameter and velocity can be organized with a non-dimensional function derived from a momentum equation of particle in driving fluid and its drag coefficient has linear correlation with a gas Froude number.

    DOI

  • Precipitation profile and dryout concentration of sea-water pool-boiling in 5 × 5 full-height BWR bundle

    Masahiro Furuya, Riichiro Okawa, Takahiro Arai, Kenetsu Shirakawa

    Nuclear Engineering and Design   375   111083 - 111083  2021.04

     View Summary

    Sea water shall be injected into water-cooled nuclear reactors during severe accidents, which are located along coastal side to flood the nuclear fuel, which is heated by residual heat. Precipitation growth to narrower flow path area is a key to gain the confidence of accident mitigation procedure to cool down the reactor core during accidental conditions. A pool boiling experiment was conducted with a simulated 5 × 5 full-height BWR fuel-rod bundle with condensed (two and half times higher concentration) sea water. The temperature on the center rod surface in the top spacer rose rapidly, since the flow area inside the top spacer was filled with the precipitated salt. Dryout below the top spacer escalated temperatures of the heater surface. On the other hand, the heater above the top spacer was cooled stably by pool boiling. An example calculation estimates that the dryout due to salt precipitation may occur 19 h after sea water injection for an ABWR, which had operated at 3.926 GWt for 13 months on the basis of critical dryout concentration of 50 wt%.

    DOI

  • Development of Subchannel Void Sensor for Wide Pressure and Temperature Ranges and Its Application to Boiling Flow Dynamics in a Heated Rod Bundle

    Takahiro Arai, Masahiro Furuya, Kenetsu Shirakawa

    Nuclear Technology    2021

     View Summary

    A subchannel void sensor (SCVS) acquires the two-phase flow in a rod bundle as the time-series data of cross-sectional distributions. Herein, the temperature and pressure ranges of an SCVS were extended to include the rated conditions of boiling water reactors. The improved SCVSs were installed in a 5 × 5 heated rod bundle at eight height levels. In a boiling experiment using the rod bundle, the three-dimensional distributions of the boiling two-phase flow were measured over a wide pressure range (up to 7.2 MPa). The new experimental data were compared with existing experimental data and the results of a subchannel analysis. Experimental results were consistent with those of a high-energy X-ray computed tomography study of a heated rod bundle with the same geometry and under the same heat and flow conditions as those used in our study. The subchannel analysis code reproduced the experimental results fairly well, and the obtained database is applicable for validating and improving thermal-hydraulic analysis codes.

    DOI

  • Kinetic energy evaluation for the steam explosion in a shallow pool with a spreading melt layer at the bottom

    Kiyofumi Moriyama, Masahiro Furuya

    Nuclear Engineering and Design   360   110521 - 110521  2020.04

     View Summary

    Steam explosion experiments with a melt layer spreading at the bottom of a shallow water pool, namely the PULiMS-E6 and SES-S1 by KTH, Sweden, were simulated by the steam explosion simulation code, JASMINE. The observed impulses in the experiments were successfully reproduced by simulations with assumed premixing conditions. With those simulation results, the adequacy of the kinetic energy evaluation method used for the experiments were examined by comparison of the kinetic energy directly obtained in the simulation, Ek, and the one evaluated based on the impulse and the water mass limited to the center area above the premixing zone, Ekic. It showed that the impulse based kinetic energy evaluation gives about five times overestimation. The impact of the water pool geometry on the validity of the impulse based kinetic energy evaluation method was further examined by a parametric study with variations of the pool geometry in the simulations of PULiMS-E6 and SES-S1 as well as high pressure bubble expansion simulations. The results for the relation of Ekic/Ek and the geometric factors were consistent between the cases for the experiments and the bubble expansion. The results showed that: (1) for the shallow water pool regime, Ekic/Ek shows a trend of convergence to 4–5, (2) for deep water pool regime, the impulse based kinetic energy evaluation with the whole water mass, Eki, rather than Ekic, gives a good estimation. A set of empirical formulas was obtained for Ekic/Ek.

    DOI

  • Visualization of three-dimensional boiling two-phase flow in a fuel rod bundle

    新井崇洋, 古谷正裕

    ΑΤΟΜΟΣ   62 ( 10 ) 560 - 564  2020

    DOI CiNii J-GLOBAL

  • Evaluation of the Removal Properties of Iodine and Organic Iodide by a AgNO3 Solution

    金井大造, 古谷正裕, 西村聡

    日本原子力学会和文論文誌   19 ( 1 ) 16 - 23  2020

     View Summary

    <p> When excess pressure and temperature are added to a containment vessel of a nuclear power plant during a severe accident, damage of the containment vessel and the release of radioactive materials into the environment are expected. Filtered containment venting systems (FCVS) installed in exhaust systems reduce the release of radioactive materials by the use of multistage filters. In actual FCVS, several types of alkaline solution are used in the scrubbing stage to capture the radioactive iodine (I2). Therefore, it is difficult to remove CH3I by the alkaline solution, and molecular sieves such as silver zeolites installed downstream of the scrubbing stage are used to capture the organic iodide (CH3I). Silver nitrate aqueous solution (AgNO3 aq.) is highly reactive against iodine; hence, by using AgNO3 aq., the removal of organic iodide in the scrubbing stage can be expected. We have conducted basic evaluations of the removal properties of iodine and organic iodide by AgNO3 aq. using small-scale and pool-scrubbing (inner diameter: 300 mm) test sections at ambient temperature and pressure. The small-scale test results show that AgNO3 aq. has the same I2 removal performance as sodium hydroxide aqueous solution (NaOH aq.). Moreover, the pool-scrubbing test results show that a CH3I decontamination factor (DF) of over 50 can be expected under the conditions of a AgNO3 concentration ≥ 10 wt% and submergence ≥ 1.14 m.</p>

    DOI CiNii J-GLOBAL

  • High-energy x-ray ct measurement of void fraction distribution around part length rods in a rod bundle at high pressure and temperatures

    Takahiro Arai, Atsushi Ui, Masahiro Furuya, Riichiro Okawa, Tsugumasa Iiyama, Shota Ueda, Kenetsu Shirakawa, Kazuaki Kito

    International Conference on Nuclear Engineering, Proceedings, ICONE   3  2020

     View Summary

    A boiling experiment was conducted to acquire a threedimensional void fraction distribution in a rod bundle with part length rods. The test section was a 5 × 5 heated rod bundle that partially simulated a BWR rod bundle, and three PLRs were arranged together in the corner. The heated length of the fulllength rod was 3.71 m, which is equal in length to an actual fuel rod in BWRs, whereas the heated length of the PLR was 1.85 m. The rod bundle exhibited an axially and radially uniform heat flux. Further, the local void fraction was quantified by normalizing the intensities in the CT images of the boiling twophase flow with those obtained under the liquid-phase and gasphase conditions. The cross-sectional void fraction distribution was acquired at six height levels. The experimental results exhibited the void distribution around PLRs with respect to a wide range of flow conditions, inlet temperatures, inlet flow rates, bundle thermal powers, and system pressures.

  • Effect of channel geometries on two-phase mixture level swell and its fluctuation amplitude

    Takahiro Arai, Masahiro Furuya, Kenetsu Shirakawa

    MECHANICAL ENGINEERING JOURNAL   7 ( 3 )  2020

     View Summary

    Gas-liquid two-phase flow in a stagnant pool is an important phenomenon in designing and operating industrial facilities. When gas is mixed or boiling occurs in stagnant water, the actual water level appears higher than the original water level. The actual water level is called a two-phase mixture level and largely depends on the flow channel geometries, dimensions, and flow conditions. This study focuses on the influence of channel geometries, circular pipes and rod bundles, on the two-phase mixture level and its fluctuation behavior. An air-water experiment using circular pipes with inner diameters of 50 and 224 mm and 5 x 5 and 10 x 10 rod bundles was conducted, and the two-phase mixture level swell was visually observed. As the inlet gas flow rate increased, the two-phase mixture level basically increased regardless of the channel geometry. The fluctuation amplitude was remarkably increased by formulating the slug bubbles covering the entire diameter in the small pipe with a diameter of up to 50 mm. In the rod bundles and large pipe with a diameter of 224 mm, no slug bubble was sustained, and the two-phase water level and its fluctuation amplitude were relatively small compared with those of the small pipe.

    DOI

  • Comparison of the Structure and Phase Changes of Carbon-Coated SiO and Li-Doped Carbon-Coated SiO During Repeated Charge-Discharge Cycling

    Takakazu Hirose, Kohta Takahashi, Takumi Matsuno, Yusuke Osawa, Masahiro Furuya, Reiko Sakai, Chinami Matsui, Hiroyuki Koide

    JOURNAL OF THE ELECTROCHEMICAL SOCIETY   167 ( 12 )  2020.01

     View Summary

    Carbon-coated SiO (SiO-C), which is a high-capacity anode material, experiences a significant capacity drop in the initial charge-discharge cycles. In contrast, Li-doped SiO-C (Li-SiO-C), which has been recently developed, exhibits a significantly smaller capacity drop. To explain this difference, we performed a detailed investigation of the structures and phase changes associated with the charge-discharge cycling of these materials by comparing their Si structures and electronic states obtained from solid-state magic-angle spinning nuclear magnetic resonance and Si K-edge X-ray absorption fine structure measurements. The results show that, in the case of SiO-C, the Li(4)SiO(4)generated during charge is partially decomposed during discharge in the initial charge-discharge cycles. These generation and decomposition behaviors are most intense during the first 20 cycles. We believe that this phenomenon is the cause of the increased irreversible capacity observed in the initial cycles of SiO-C. In addition, we confirmed that Li2SiO3, a component of Li-SiO-C, is relatively stable electrochemically, although some of it gradually converts into Li(4)SiO(4)during charge-discharge cycling. The presence of Li(2)SiO(3)at the outset implies that less Li(4)SiO(4)is generated during charging compared to SiO-C, which we believe explains the lack of a significant capacity drop in the initial cycles of Li-SiO-C.

    DOI

  • TRACE code demonstration of thermal stratification in BWR suppression pool

    Riichiro Okawa, Masahiro Furuya

    Nuclear Engineering and Design   355   110357 - 110357  2019.12  [Refereed]

     View Summary

    An analytical model was developed to describe thermal stratification in a primary containment vessel (PCV) and transient thermal-hydraulics coupled with a reactor pressure vessel (RPV) using TRACE code version 5.0 patch level 4. Geometries of a dry well (D/W) and a suppression chamber (S/C) were represented by a nodalization of TRACE code to simulate multi-dimensional flow in the PCV. An additive loss coefficient (so called ‘K-factor’) was focused as a sensitivity parameter to limit flow rate in a pool. For the first step, a validation analysis was conducted against a steam discharge experiment of S/C. The TRACE result was in good agreement with the measurement and showed a thermally-stratified temperature distribution in the S/C pool. For the second step, an analysis to simulate the accident at Fukushima Daiichi Unit 3 power plant (1F3) was conducted. It was proved to be able to explain the pressure increase in the PCV at the beginning of accident by demonstrating thermal stratification in the S/C pool. Sensitivity study revealed an optimal K-factor value for a macroscopic viscous drag in a liquid phase fluid to demonstrate thermal stratification in a pool.

    DOI

  • Density and viscosity of liquid ZrO2 measured by aerodynamic levitation technique

    Toshiki Kondo, Hiroaki Muta, Ken Kurosaki, Florian Kargl, Akifumi Yamaji, Masahiro Furuya, Yuji Ohishi

    Heliyon   5 ( 7 ) e02049  2019.07  [Refereed]

     View Summary

    Liquid ZrO2 is one of the most important materials involved in severe accident analysis of a light-water reactor. Despite its importance, the physical properties of liquid ZrO2 are scarcely reported. In particular, there are no experimental reports on the viscosity of liquid ZrO2. This is mainly due to the technical difficulties involved in the measurement of thermo-physical properties of liquid ZrO2, which has an extremely high melting point. To address this problem, an aerodynamic levitation technique was used in this study. The density of liquid ZrO2 was calculated from its mass and volume, estimated based on the recorded image of the sample. The viscosity was measured by a droplet oscillation technique. The density and viscosity of liquid ZrO2 at temperatures ranging from 2753 K to 3273 K, and 3170 K–3471 K, respectively, were successfully evaluated. The density of liquid ZrO2 was found to be 4.7 g/cm 3 at its melting point of 2988 K and decreased linearly with increasing temperature, and the viscosity of liquid ZrO2 was 13 mPa at its melting point.

    DOI

  • Three dimensional void distribution measurement of salt-water pool-boiling in 5 × 5 bundle geometry with X-ray CT system

    Masahiro Furuya, Hiroki Takiguchi, Riichiro Okawa, Takahiro Arai

    Annals of Nuclear Energy   129   207 - 213  2019.07  [Refereed]

     View Summary

    Boiling of sea water may occur in a pressure vessel of light water (nuclear) reactors to flood the nuclear fuel as an accident management procedure. Another salt water is borated water, which will be injected into the reactor core as a neutron absorber to avoid recriticality. Boiling behavior of such salt water including these mixtures is a key to gain the confidence of accident strategy to cool down the reactor core during accidental conditions. Pool boiling experiments were conducted with a simulated 5 × 5 fuel-rod bundle with three different fluids: water, condensed (two and half times higher concentration) sea water and its mixture solution of sea water and borated water. Three-dimensional void-fraction distributions in the rod bundles were quantified by the high-energy X-ray CT system with a linear accelerator. There are no significant differences in void fraction distributions between condensed sea water and mixture solution. The void-fraction has a peak at the center on horizontal plane for all the fluids. The two salt-waters shift boiling incipience toward downstream and decrease void swell level so that vertical void-fraction profiles of the salt waters are steeper than that of water. This is because created bubbles in the salt waters were smaller than those in water. The spacer has a mixing effect to increase void fraction at upstream of the spacer and homogenize the void fraction on the horizontal plane.

    DOI

  • Research of realistic FP behavior evaluation under severe accident-Accident analyses of unit 3 in the Fukushima-Daiichi nuclear power station with MAAP-

    西義久, 阿部数馬, 神田憲一, 中村康一, 西村聡, 宇井淳, 古谷正裕

    日本機械学会年次大会講演論文集(CD-ROM)   2019   S08108  2019

    DOI CiNii J-GLOBAL

  • Performance evaluation of venturi scrubber used in filtered containment venting system

    金井大造, 古谷正裕, 西村聡

    日本機械学会年次大会講演論文集(CD-ROM)   2019   S08107  2019

     View Summary

    <p>Filtered containment venting systems (FCVS) installed in the exhaust systems reduce the release of the radioactive materials by the multistage filters. Venturi-scrubber, wet scrubbing, static mixer and metal-fiber filter remove aerosol. This study has focused on the venturi-scrubber which remove aerosol by collision of droplets with particles and conducted evaluation of aerosol removal performance and hydro-dynamic behavior. Performance of venturi scrubber depend on its geometries such as throat cross sectional area, spread angle of expanding section. This study used several type of venturi scrubbers, the test results show that the aerosol DF depend on aerosol diameter and the venturi-scrubber used in this study shows DF ≥ 100 for aerosol (diameter: 1.0 μm). Moreover, this study evaluated the aerosol removal performance using CFD calculation and previous model (Ueoka and Kawakami model). By adjusting the constant included in the collision coefficient (ε), the venturi scrubber performances predicted from the CFD calculation are in good agreement with the experimental results.</p>

    DOI CiNii J-GLOBAL

  • Suppression Effect of Steam Explosion with Carbon Dioxide Dissolved

    古谷正裕, 新井崇洋, 飯山継正, 大川理一郎

    日本機械学会年次大会講演論文集(CD-ROM)   2019   S08119  2019

    DOI CiNii J-GLOBAL

  • Data assimilation with subchannel analysis code CTF on NUPEC BWR BFBT test matrix

    Atsushi Ui, Yoshiro Kudo, Masahiro Furuya

    18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019     5052 - 5063  2019

     View Summary

    In order to gain the reliability of subchannel analysis on three-fluid two-phase flow in nuclear fuel assemblies, implemented models are expected to describe the detailed three-dimensional two-phase flow in fuel assemblies. Especially, crossflow model limits the prediction performance of subchannel analysis codes, it is important to develop a model that can analyze the phenomenon appropriately. In this study, CTF results are validated against the NUPEC BWR Full-Size Fine-mesh Bundle Test (BFBT). In BFBT, void fraction distribution across 8×8 rod bundles was measured to confirm the effects of radial/axial power distribution and unheated rods. Moreover, uncertainties of void fraction were quantified. In order to evaluate the prediction performance of the CTF code for the BFBT, bundle-averaged void fraction difference and the residual void fraction difference for each subchannel were defined. Subchannels were classified into several groups considering the grid spacer pressure loss coefficients set for each channel and the characteristics of the subchannel considering location, such as corner subchannel, adjacent corner subchannel, etc. Sensitivity parameters affecting void fraction and/or cross flow were selected with the Kriging method, and the response surface model represented by these sensitivity parameters was created. the simulation-driven MCDA method using the alternative model was applied for optimizing sensitivity parameters by data assimilation, and a set of parameters to accurately calculate the bundle-averaged void fraction difference was identified with the Metropolis method. CTF analysis with the parameter set identified by the data assimilation was conducted, and it was confirmed that the average value of the bundle-averaged void fraction improved with the parameter set by the data assimilation so that the predicted value would match the experimental value.

  • X-ray radiography for two-phase mixture level fluctuation during boil-off in rod bundle for wide pressure range

    T. Arai, M. Furuya, H. Takiguchi, Y. Nishi, K. Shirakawa

    18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019     5701 - 5708  2019

     View Summary

    In the case of a loss in a coolant accident when the water level of the reactor core falls below the top of active fuel (TAF), the actual water level, namely the two-phase mixture level, is an important factor to predict the coolability of the core. The two-phase mixture level depends on the liquid inventory and temporal and spatial variations of the void fraction in the core, and can be greatly affected by the system pressure. A boil-off experiment in which the water level decreased by evaporation without any water supply, was conducted using a 5×5 heated rod bundle over a wide pressure range from 0.1 to 7.2 MPa. The heated length of the rod was 3.71 m, which is equal in length to an actual fuel rod in boiling water reactors (BWRs). The 5×5 rod bundles had an axially and radially uniform power profile, and eight pairs of fine thermocouples were embedded in the heated region to acquire the axial distribution of the rod surface temperature. The temporal fluctuation of the two-phase mixture level was visualized by a linear-accelerated driven X-ray real-time radiography at 400 frames per second. The experimental results exhibit the influence of the two-phase mixture level variation on the rise of rod-temperature under boil-off conditions.

  • Analysis of hemispherical vessel ablation failure involving natural convection by MPS method with corrective matrix

    Nozomu Takahashi, Guangtao Duan, Masahiro Furuya, Akifumi Yamaji

    International Journal of Advanced Nuclear Reactor Design and Technology   1   19 - 29  2019

    DOI

  • Precipitation profile and dryout concentration of sea-water pool-boiling in 5 × 5 bundle geometry

    Masahiro Furuya, Riichiro Okawa, Takahiro Arai, Hiroki Takiguchi, Kenetsu Shirakawa

    Nuclear Engineering and Design   341   38 - 45  2019.01  [Refereed]

     View Summary

    As an accident management procedure of light water (nuclear) reactors which are situated along sea shore, sea water will be injected into the reactor pressure vessel to flood the nuclear fuel which is heated by residual heat. Another salt water is borated water, which will be injected into the reactor core as a neutron absorber to avoid recriticality. Precipitation behavior of such salt water including these mixtures is a key to gain the confidence of accident strategy to cool down the reactor core during accidental conditions. Pool boiling experiments were conducted with a simulated 5 × 5 fuel-rod bundle with condensed (two and half times denser) sea water and a mixture solution of sea water and borated water. Three-dimensional salt-precipitation distributions in the rod bundles were quantified with X-ray CT system. For both solutions, salt precipitated downstream and close to the top of active fuel (TAF) height where the void fraction is the highest. The condensed sea water yields wider precipitation region in height direction than mixture solution does. Mixture solution may give localized precipitates at the same height, which is just below TAF and uniformly spread on the horizontal plane. For both solutions, dryout concentration is larger as collapsed solution level is higher. This is because that lower collapsed solution level gives longer boiling-length and higher void-fraction, which results in larger amount of salt precipitations. The proposed salt concentration is useful to evaluate dryout concentration, which is the almost constant salt concentration for heat flux levels within the experimental ranges.

    DOI

  • TRANSIENT BOILING AND CROSS FLOW IN 5x5 ROD BUNDLE WITH RAPID HEATING

    Hiroki Takiguchi, Masahiro Furuya, Takahiro Arai, Kenetsu Shirakawa

    PROCEEDINGS OF THE 26TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, 2018, VOL 6A    2018  [Refereed]

     View Summary

    Rapid thermal elevation in nuclear reactor is an important factor for nuclear safety. It is indispensable to develop a three-dimensional nuclear thermal transient analysis code and confirm its validity in order to accurately evaluate the effectiveness of the running nuclear safety measures when heating power of reactor core rapidly rises. However, the heat transfer characteristics such as reactivity feedback characteristics due to moderator density and the technical knowledge explaining the uncertainty are insufficient. In particular, the cross propagation behavior of vapor bubble (void) in cross section of fuel assembly is not grasped.This study evaluates the cross propagation void behavior in a simulated fuel assembly at time of rapid heat generation with a thermal hydraulic test loop including a 5 x5 rod bundle having the heat generation profile in the flow cross sectional direction. In this paper, the branching heat output condition of transient cross propagation was investigated from visualization of high speed video camera and void fraction measurement by wire mesh sensor with the inlet flow rate 0.3m/s and the inlet coolant temperature 40 degrees C, which are based on the transient safety analysis condition. In addition, we applied the particle imaging velocimetry (PIV) technique to measure liquid-phase velocity profile of the coolant in the transient cross flow and experimentally clarified the relationship with the cross flow.

  • Three-dimensional velocity vector determination algorithm for individual bubble identified with Wire-Mesh Sensors

    Masahiro Furuya, Taizo Kanai, Takahiro Arai, Hiroki Takiguchi, Horst-Michael Prasser, Uwe Hampel, Eckhard Schleicher

    Nuclear Engineering and Design   336   74 - 79  2018  [Refereed]

     View Summary

    The bubble pairing scheme was devised to quantify three-dimensional velocity of each bubble. We used two sets of Wire-Mesh Sensors to identify locations of each bubble according to bubble identification algorithm, which was developed by HZDR. The devised scheme was applied to the vertical upward air-water flow at 0.64 m/s for both air and water superficial velocities in a large diameter pipe (i.d. 224 mm). The bubble pairing scheme visualized the developing process of two-phase flow: large bubbles coalesced with each other to move toward the center, while the rest of bubbles broke up into smaller bubbles and decelerated.

    DOI

  • Investigation on influence of crust formation on VULCANO VE-U7 corium spreading with MPS method

    Yusan Yasumura, Akifumi Yamaji, Masahiro Furuya, Yuji Ohishi, Guangtao Duan

    Annals of Nuclear Energy   107   119 - 127  2017  [Refereed]

     View Summary

    In a severe accident of a light water reactor, the corium spreading behavior on a containment floor is important as it may threaten the containment vessel integrity. The Moving Particle Semi-implicit (MPS) method is one of the Lagrangian particle methods for simulation of incompressible flow. In this study, the MPS method is further developed to simulate corium spreading involving not only flow, but also heat transfer, phase change and thermo-physical property change of corium. A new crust formation model was developed, in which, immobilization of crust was modeled by stopping the particle movement when its solid fraction is above the threshold and is in contact with the substrate or any other immobilized particles. The VULCANO VE-U7 corium spreading experiment was analyzed by the developed MPS spreading analysis code to investigate influences of different particle sizes, the corium viscosity changes, and the “immobilization solid fraction” of the crust formation model on the spreading and its termination. Viscosity change of the corium was influential to the overall progression of the spreading leading edge, whereas termination of the spreading was primarily determined by the immobilization of the leading edge (i.e., crust formation). The progression of the leading edge and termination of the spreading were well predicted, but the simulation overestimated the substrate temperature. Further investigations may be necessary for the future study to see if thermal resistance at the corium-substrate boundary has significant influence on the overall spreading behavior and its termination.

    DOI

  • 9th International Conference on Multiphase Flow (ICMF2016)

    Journal of the Atomic Energy Society of Japan   58 ( 10 ) 621 - 621  2016

    DOI CiNii

  • Multi-dimensional void fraction measurement of transient boiling two-phase flow in a heated rod bundle

    Takahiro Arai, Masahiro Furuya, Taizo Kanai, Kenetsu Shirakawa, Yoshihisa Nishi

    MECHANICAL ENGINEERING JOURNAL   2 ( 5 )  2015  [Refereed]

     View Summary

    A subchannel void sensor (SCVS) was developed to measure the cross-sectional distribution of void fraction in a 5x5 heated rod bundle with o.d. 10 mm and heated length 2000 mm, and applied to a boiling two-phase flow experiment under the atmospheric pressure condition assuming at an accident or in a spent fuel pool in a boiling water reactor (BWR). The SCVS comprises 6-wire by 6-wire and 5-rod by 5-rod electrodes. The wire electrodes of 0.2 mm in diameter are arranged in lattice patterns between the rod bundle, while the electric conductance value in a region near one wire and another corresponds to local void fraction in the central-subchannel region. The local void fractions at 32 points (= 6x6-4) can be obtained as a cross-sectional distribution. The local void fractions near the rod surface at 100 points (= 4x25) can be also estimated by the conductance value in a region between one wire and one rod. The devised sensors are installed at five height levels along the axis to acquire two-phase flow behavior. A pair of SCVS is mounted at each level and placed 30 mm apart to estimate the one-dimensional phasic velocity distribution based on the cross-correlation analysis of both layers. The temporal resolution of void fraction measurement is 1600 frames (cross-sections) per second. The axial and radial power profile of the heated rod bundle are uniform, and eight pairs of sheath thermocouples are embedded on the heated rod to monitor its surface temperature distribution. The boiling two-phase flow experiment, which simulated a boil-off process, was conducted with the devised SCVS and experimental data was acquired under various inlet flow velocity, rod bundle power and inlet subcooling conditions. The experimental results were presented by the axial and cross-sectional distributions of void fraction, phasic velocity and bubble-chord length.

    DOI

  • Concurrent upward liquid slug dynamics on both surfaces of annular channel acquired with liquid film sensor

    Arai, T., Furuya, M., Kanai, T., Shirakawa, K.

    Experimental Thermal and Fluid Science   60   337 - 345  2015  [Refereed]

     View Summary

    The interfacial behavior of upward liquid film flow is an important phenomenon to evaluate interfacial transfer accompanying the entrainment and deposition of droplets. This research focuses on a vertical annular channel, and an air-water liquid film flow experiment was conducted under atmospheric pressure conditions. The diameters of inner and outer pipes in the annular channel were 12 and 18. mm respectively. The experiment featured multi-point electrode sensors installed in both the inner and outer pipe surface at the same height, and the ability to measure the liquid film distribution on both surfaces in the annular channel simultaneously. As for the sensor structure, 10. ×. 32 measuring points were arranged in a lattice pattern on the sensor surface and the spatial resolution was 2. ×. 2. mm, hence the liquid film thickness distribution could be measured rapidly, at over 1250 slices per second. Since the sensor was manufactured by a flexible multilayer substrate, it was applicable to a cylindrical channel surface. In the experiment, water was supplied from the inner pipe surface and uniformly distributed in the circumferential direction, whereupon liquid film distributions were measured 300. mm downstream from the water supply position. The time series data of the liquid film distribution demonstrated circumferential distributions of liquid film thickness and interfacial wave velocity. When the superficial gas velocity was smaller than 20. m/s, a liquid film formed on both inner and outside pipe surfaces, regardless of the superficial liquid velocity. With increasing superficial gas velocity, the film thickness of the outer pipe surface became thinner than that of the inner pipe surface. Measurement of the liquid film thickness on both surfaces of the annular channel also showed that a liquid slug with wavelength of several millimeters passed concurrently through both surfaces in the annular channel.

    DOI

  • An evaluation model to predict steam concentration in a BWR reactor building

    Kondo, M., Yoneda, K., Furuya, M., Nishi, Y.

    Journal of Nuclear Science and Technology   52 ( 11 ) 1369 - 1382  2015  [Refereed]

     View Summary

    When there is no power for cooling the spent fuel pool and conditioning the air in a boiling water reactor (BWR) reactor building, water vapor is generated from the pool and it affects the atmosphere in the building. To consider the impact of the steam in preparing emergency operation procedures, the building atmosphere under various conditions is to be evaluated with reasonably low computational cost. A lumped parameter model to predict the transient behavior of the building atmosphere was developed, in which the evaporation from the spent fuel pool and the condensation to the wall were taken into consideration. A transient behavior of temperature and vapor concentration in a BWR operating floor was predicted with the model. The results and the prediction speed were compared to those of a three-dimensional computational fluid dynamic calculation, and it was confirmed that the model could obtain almost the same results about 280,000 times faster. Parameter studies are conducted with the model, and dominant parameters to the evaporation and the condensation were clarified.

    DOI

  • Residual stress distribution in oxide films formed on Zircaloy-2

    Sawabe, T., Sonoda, T., Furuya, M., Kitajima, S., Takano, H.

    Journal of Nuclear Materials   466   658 - 665  2015  [Refereed]

     View Summary

    In order to evaluate residual the stress distribution in oxides formed on zirconium alloys, synchrotron X-ray diffraction (XRD) was performed on the oxides formed on Zircaloy-2 after autoclave treatment at a temperature of 360° C in pure water. The use of a micro-beam XRD and a micro-sized cross-sectional sample achieved the detailed local characterization of the oxides. The oxide microstructure was observed by TEM following the micro-beam XRD measurements. The residual compressive stress increased in the vicinity of the oxide/metal interface of the pre-transition oxide. Highly oriented columnar grains of a monoclinic phase were observed in that region. Furthermore, at the interface of the post-first transition oxide, there was only a small increase in the residual compressive stress and the columnar grains had a more random orientation. The volume fraction of the tetragonal phase increased with the residual compressive stress. The results are discussed in terms of the formation and transition of the protective oxide.

    DOI

  • 3D simulation of eutectic interaction of Pb-Sn system using Moving Particle Semi-implicit (MPS) method

    Andi Mustari, A.P., Oka, Y., Furuya, M., Takeo, W., Chen, R.

    Annals of Nuclear Energy   81   26 - 33  2015  [Refereed]

     View Summary

    The eutectic reaction phenomenon was analyzed based on the Moving Particle Semi-implicit (MPS) method. The improved MPS code was applied to three-dimensional Pb-Sn system experiments conducted at Central Research Institute of Electric Power Industry (CRIEPI). The experiments dealt with solid-solid contact materials at 225 and 205 °C. The mass diffusion process was modeled based on Fick's second law. The criterion of eutectic reaction was modeled based on binary phase diagram. The calculated penetration rates, i.e. height reduction rates, were compared with the experimental measurements. The results obtained by the MPS simulations exhibit good agreement with the experiments carried out at 205 and 225 °C. MPS simulation predicted that Sn particle will liquefy earlier than Pb particle.

    DOI

  • Experiments and volume-of-fluid (VOF) simulations of a three-fluid dam-break

    M. Furuya, Y. Oka, M. Satoh, S. Lo, T. Arai

    WIT Transactions on Engineering Sciences   83   363 - 371  2014.07  [Refereed]

     View Summary

    Three-fluid dam-break experiments were conducted to observe the mixing and stratification processes of three immiscible fluids (two-liquids and one gas) by gravity. Two liquids (silicone oil and salt water) were separated with a vertical wall and filled in a rectangle container. Fluid motions are visualized by four sets of video cameras which are synchronized with each other. Parametric study reviles the effects of two key fluid properties: kinematic viscosity (or molecular weight) for silicone oil and density (or concentration) for sodium chloride aqueous water. After the withdrawal of a vertical partition plate, two liquids intersected earlier at the center, while the fluids stuck on the walls. The kinematic viscosity and density difference affect the three-dimensional mixing and stratification processes significantly. These visual databases are suitable for a code validation on the interfacial phenomena. Computational multi-phase fluid dynamics analysis was conducted with Star CCM+ version 8.04. The numerical results agree with experimental results accordingly for all those key parameters. The physics behind the contact angle and interfacial tensions are thoroughly investigated parametrically.

    DOI

  • Comparison of 12-and 16-Core Prostate Biopsy in Japanese Patients with Serum Prostate-Specific Antigen Level of 4.0-20.0 ng/mL

    Yasuhide Miyoshi, Masahiro Furuya, Jun-ichi Teranishi, Kazumi Noguchi, Hiroji Uemura, Yumiko Yokomizo, Shinpei Sugiura, Yoshinobu Kubota

    UROLOGY JOURNAL   11 ( 3 ) 1609 - 1614  2014.05  [Refereed]

     View Summary

    Purpose: In the present study, we compared 12- with 16-core biopsy in patients with prostate-specific antigen (PSA) levels of 4.0-20.0 ng/mL.Materials and Methods: Between 2003 and 2010, 332 patients whose serum PSA level was between 4.0 and 20.0 ng/mL underwent initial transrectal ultrasound (TRUS)-guided needle biopsy. Of those patients, 195 underwent 12-core biopsy and 137 underwent 16-core biopsy.Results: In the 12-core prostate biopsy group, 66(33.8%) patients were found to have prostate cancer. On the other hand, in the 16-core prostate biopsy group of 137 patients, 61(44.5%) were found to have prostate cancer. Among all patients, the prostate cancer detection rate was slightly higher in the 16-core biopsy group than in the 12-core biopsy group. Moreover, in patients with prostate volume > 30 mL or PSA density (PSAD) <0.2, the prostate cancer detection rate was significantly higher in the 16-core biopsy group than in the 12-core biopsy group. There was no significant difference in pathological tumor grade, indolent cancer probability, or biopsy complication rate between the two groups.Conclusion: In order to detect prostate cancer, 16-core prostate biopsy is safe and feasible for Japanese patients with serum PSA level of 4.0-20.0 ng/mL.

    PubMed

  • DEVELOPMENT OF A MULTI-DIMENSIONAL MEASUREMENT SENSOR OF VOID FRACTION AND PHASIC VELOCITY FOR BOILING TWO-PHASE FLOW IN A 5x5 HEATED ROD BUNDLE

    Takahiro Arai, Masahiro Furuya, Taizo Kanai, Kenetsu Shirakawa, Yoshihisa Nishi

    PROCEEDINGS OF THE 22ND INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING - 2014, VOL 2A    2014  [Refereed]

     View Summary

    A subchannel void sensor (SCVS) was developed to measure the cross-sectional distribution of a void fraction in a 5x5 heated rod bundle with o.d. 10 mm and heated length 2000 mm, and applied in a boiling two-phase flow experiment under the atmospheric conditions assumed in an accident and spent fuel pool. The SCVS comprises 6-wire by 6-wire and 5-rod by 5-rod electrodes. Wire electrodes 0.2 mm in diameter are arranged in latticed patterns between the rod bundle, while a conductance value in a region near one wire and another gives a local void fraction in the central-subchannel region. 32 points (= 6x6-4) of the local void fraction can be obtained as a cross-sectional distribution. In addition, a local void fraction near the rod surface can be estimated by a conductance value in a region near one wire and one rod using the simulated fuel rods as rod electrodes, which allows 100 additional points (=4x25) of the local void fraction to be acquired. The devised sensors are installed at five height levels to acquire two-phase flow dynamics in an axial direction. A pair of SCVS is mounted at each level and placed 30 mm apart to estimate the one-dimensional phasic velocity distribution based on the cross-correlation analysis of both layers. The time resolution of void measurement exceeds 800 frames (cross-sections) per second. The heated rod bundle has an axially and radially uniform power profile, and eight pairs of sheath thermocouples are embedded on the heated rod to monitor its surface temperature distribution. The boiling two-phase flow experiment, which simulated a boil-off process, was conducted with the devised SCVS and experimental data was acquired under various experimental conditions, such as inlet-flow velocity, rod-bundle power and inlet subcooling. The experimental results exhibited axial and radial distribution of two-phase flow structures, i.e. void-fraction and phasic-velocity distributions quantitatively.

  • Effect of two-phase flow structure in decontamination factor of filtered containment venting system

    Taizo Kanai, Yoshihisa Nishi, Masahiro Furuya, Kenetsu Shirakawa, Takahiro Arai, Satoshi Nishimura, Nobuyuki Tanaka, Masaaki Satake

    International Conference on Nuclear Engineering, Proceedings, ICONE   1  2014  [Refereed]

     View Summary

    In order to gain the best use of filtered containment venting systems (FCVSs), the decomtamination factor of FCVSs is to be investigated as a function of system parameter including steam flow rate, pressure, temperature, water level, and operating time. A full-height test facilities were designed and constructed in Central Research Institute of Electric Power Industry (CRIEPI), Japan to evaluate the decontamination factor (DF) in FCVSs. The target types are the orifice and the venturi FCVSs. The height and the internal diameter of the cylindrical test vessel is 8 m and 0.5 m. Bubbly flows were visualized through the view window up to 0.8 MPa and 170 °C. Steam bubbles in 0.2 wt% sodium thiosulfate and 0.5 wt% sodium hydroxide were found to be much smaller than those in water. The DF were evaluated for the aerosol, elemental iodine and organic iodine. The installed aerosol optical spectrometer measures the number density and the diameter of aerosols. The concentrations of elemental iodine were quantified with an inductively-coupled plasma with mass spectrometry (ICP-MS). The concentration of organic iodine was quantified with a gas chromatography with mass spectrometry (GC-MS). In order to investigate two-phase flow dynamics in the vessel, separate effect tests were conducted with air-water test facility. The height of cylindrical test vessel is 8 m. Visual observation was conducted for two internal diameter levels: 0.05 and 0.5 m. High speed video frames were recorded through the transparent (acrylic) vessel wall. Wire-Mesh Sensors (WMS) were installed to acquire a cross-sectional void fraction to compare with DF in the facility. On the basis of the obtained database, we develop the FCVSs performance evaluation technique and propose an optimal FCVSs operation method for a further safety improvements of the nuclear power plant.

    DOI

  • Experimental and numerical study of stratification and solidification/ melting behaviors

    Li, G., Oka, Y., Furuya, M.

    Nuclear Engineering and Design   272   109 - 117  2014  [Refereed]

     View Summary

    Given the severe accident of a light water reactor (LWR), stratification and solidification/melting are important phenomena in melt corium behavior within the reactor lower head, influencing the decay heat distribution and ablation of penetration tube and vessel wall. Numerical calculation is a necessary and effective approach for mechanistic study of local melt corium behavior. In this study, the improved moving particle semi-implicit (MPS) method was applied for investigating the stratification and solidification/melting phenomena. The implicit viscous term calculation technique and stability improvement technique were adopted to enable MPS to simulate the stratification process of materials with high viscosity in phase transition stage. The solid-liquid phase transition model was also coupled with MPS method. The validation experiment was carried out with low-melting-point metal tin and NeoSK-SALT. The layer configurations and temperature profiles obtained from MPS calculation showed good agreement with the experimental results. Meanwhile, the calculation results indicated that the material freezing behavior could affect the layer formation, and the layer configurations also significantly influenced the temperature profiles and heat flux distributions. The present results demonstrated that MPS method has the capacity to understand the local melt behavior in detail that is relevant to stratification and phase transition. © 2014 Elsevier B.V.

    DOI

  • Early construction and operation of the highly contaminated water treatment system in Fukushima Daiichi Nuclear Power Station (IV) - Assessment of hydrogen behavior in stored Cs adsorption vessel

    Kondo, M., Arai, T., Nishi, Y., Furuya, M., Kanai, T., Morita, R., Uchiyama, Y., Satake, M., Shirakawa, K., Nauchi, Y., Koyama, T., Ishikawa, K., Suzuki, S.

    Journal of Nuclear Science and Technology   51 ( 7-8 ) 916 - 929  2014  [Refereed]

     View Summary

    Hydrogen diffusion behavior in a cesium adsorption vessel is assessed. The vessel is used to remove radioactive substance from contaminated water, which is proceeded from Fukushima accident. Experiment and numerical calculation are conducted to clarify the characteristics of natural circulation in the vessel. The natural circulation arising from the temperature difference between inside and outside the vessel is confirmed. We develop an evaluation model to predict the natural circulation and its prediction agrees well with the results obtained by the experiment and the calculation. Using the model, we predict steady and transient behavior of hydrogen concentration. Results indicate that hydrogen concentration is kept lower than the flammability limit when the short vent pipe is open. © 2014 Atomic Energy Society of Japan. All rights reserved.

    DOI

  • Development of a multi-dimensional measurement sensor of void fraction and phasic velocity for boiling two-phase flow in a 5×5 heated rod bundle

    Takahiro Arai, Masahiro Furuya, Taizo Kanai, Kenetsu Shirakawa, Yoshihisa Nishi

    International Conference on Nuclear Engineering, Proceedings, ICONE   2A  2014  [Refereed]

     View Summary

    A subchannel void sensor (SCVS) was developed to measure the cross-sectional distribution of a void fraction in a 5×5 heated rod bundle with o.d. 10 mm and heated length 2000 mm, and applied in a boiling two-phase flow experiment under the atmospheric conditions assumed in an accident and spent fuel pool. The SCVS comprises 6-wire by 6-wire and 5-rod by 5-rod electrodes. Wire electrodes 0.2 mm in diameter are arranged in latticed patterns between the rod bundle, while a conductance value in a region near one wire and another gives a local void fraction in the central-subchannel region. 32 points (= 6×6-4) of the local void fraction can be obtained as a crosssectional distribution. In addition, a local void fraction near the rod surface can be estimated by a conductance value in a region near one wire and one rod using the simulated fuel rods as rod electrodes, which allows 100 additional points (=4×25) of the local void fraction to be acquired. The devised sensors are installed at five height levels to acquire two-phase flow dynamics in an axial direction. A pair of SCVS is mounted at each level and placed 30 mm apart to estimate the onedimensional phasic velocity distribution based on the crosscorrelation analysis of both layers. The time resolution of void measurement exceeds 800 frames (cross-sections) per second. The heated rod bundle has an axially and radially uniform power profile, and eight pairs of sheath thermocouples are embedded on the heated rod to monitor its surface temperature distribution. The boiling two-phase flow experiment, which simulated a boil-off process, was conducted with the devised SCVS and experimental data was acquired under various experimental conditions, such as inlet-flow velocity, rod-bundle power and inlet subcooling. The experimental results exhibited axial and radial distribution of two-phase flow structures, i.e. void-fraction and phasic-velocity distributions quantitatively.

    DOI

  • Detailed observation of seawater precipitation with 5x5 mockup fuel bundle based on x-ray computated tomography technique

    Retsu Kojo, Akitoshi Hotta, Masahiro Furuya, Miyuki Akiba, Harutaka Hoshi

    International Conference on Nuclear Engineering, Proceedings, ICONE   6  2014  [Refereed]

     View Summary

    In order to conform to the new regulatory standard in Japan, seawater is regarded as the alternative water source both for BWRs (Boiling water reactors) and PWRs (Pressurized water reactors). For preventing further accident evolutions occurred in Fukushima Daiichi nuclear power plants, seawater was injected into the reactors for more than one week. With long-term seawater injection, sea salt compositions are condensed and many of them will precipitate when saturated concentrations are exceeded. Unlike corrosion issues, impacts of sea salt precipitation on the heat removal has not been studied widely in the past because it has not been regarded as the alternative water source before Fukushima Daiichi nuclear power plants accident. The existent knowledge base of boric acid precipitation under LOCA conditions was studied. Based on the existent study on impacts of boric acid precipitation under LOCA conditions, the experimental project of seawater consisting of four experiments was proposed. The existent database of physical properties, such as viscosity, of seawater is rather poor under severe accident conditions. In addition, it is likely that boric acid will be injected with water or seawater to prevent re-criticality. It is known that physical properties of boric acid vary widely under high temperature conditions. Measurement of viscosity of the seawater-boric acid mixture was conducted in high temperature using a rotational viscometer. Under conditions equivalent to the estimated bulk coolant conditions under a long-term cooling phase of severe accidents. In this range, no obvious change of viscosity is expected. Then a detailed structure of seawater precipitates was observed using a mockup fuel bundle with 5x5 in the square lattice and 500 mm in the length. Images of precipitates were taken using the X-ray CT. The water level, the concentration of sea salt and the heat flux are employed as experimental parameters. The heat flux, bubble stirring in downstream of spacers and heat loss by a non-heated channel box were identified as influential factors to local and overall precipitates in fuel bundles.

    DOI

  • Development of electrocatalyst to reduce carbon dioxide

    Furuya, M.

    Nihon Kikai Gakkai Ronbunshu, B Hen/Transactions of the Japan Society of Mechanical Engineers, Part B   79 ( 799 ) 286 - 290  2013  [Refereed]

     View Summary

    The carbon-doped copper oxide is devised as an electrocatalyst to reduce carbon dioxide under ambient pressure and temperature. The electrode was immersed into a KHCO3 aqueous solution with CO2 bubbling. The electric potential was maintained at -1.64 V vs SHE. The electrode prepared at 900 oC gives the maximum production rate of ethylene (25 %), ethanol (6.9 %) and 1-propanol (3.6 %). The production rate of methane, from which is harmful to separate ethylene, was suppressed to one fifteenth of that of ethylene. In contrast to a thermally-oxidized copper-oxide layer, the doped-carbon and a high ratio of Cu2O to CuO in the devised electrocatalyst may result in the higher productivity and selectivity. ©2013 The Japan Society of Mechanical Engineers.

    DOI

  • Experiments and MPS analysis of stratification behavior of two immiscible fluids

    Li, G., Oka, Y., Furuya, M., Kondo, M.

    Nuclear Engineering and Design   265   210 - 221  2013  [Refereed]

     View Summary

    Stratification behavior is of great significance in the late in-vessel stage of core melt severe accident of a nuclear reactor. Conventional numerical methods have difficulties in analyzing stratification process accompanying with free surface without depending on empirical correlations. The Moving Particle Semi-implicit (MPS) method, which calculates free surface and multiphase flow without empirical equations, is applicable for analyzing the stratification behavior of fluids. In the present study, the original MPS method was improved to simulate the stratification behavior of two immiscible fluids. The improved MPS method was validated through simulating classical dam break problem. Then, the stratification processes of two fluid columns and injected fluid were investigated through experiments and simulations, using silicone oil and salt water as the simulant materials. The effects of fluid viscosity and density difference on stratification behavior were also sensitively investigated by simulations. Typical fluid configurations at various parametric and geometrical conditions were observed and well predicted by improved MPS method. © 2013 Elsevier B.V. All rights reserved.

    DOI

  • Three-dimensional phasic velocity determination methods with wire-mesh sensor

    Kanai, T., Furuya, M., Arai, T., Shirakawa, K., Nishi, Y.

    International Journal of Multiphase Flow   46   75 - 86  2012  [Refereed]

     View Summary

    A gas-liquid two-phase flow in a large diameter pipe exhibits a three-dimensional flow structure. The wire-mesh sensor (WMS) can acquire a quasi-three-dimensional void fraction distribution. Furthermore, the WMS can acquire a phasic-velocity distribution on the basis of the time lag of void signals between both sets of WMS. Previously, the acquired phasic velocity was one-dimensional distributions.The authors propose a method to estimate the three-dimensional phasic-velocity distribution from the same WMS data. A three dimensional velocity vector was determined on the basis of cross-correlation analysis. The flow direction is determined by the WMS measuring-point combination, whereby the cross-correlation coefficient between both sets of WMS measuring points reveals the peak. In addition, the flow structure can be extracted by size on the basis of a wavelet analysis.The proposed method was applied for two sets of 64. ×. 64 mesh sensors in an air-water flow in a vertical pipe with inner diameter of 224. mm. The proposed method can successfully visualize a swirl flow structure where large and small bubbles tend to move respectively in inward and outward directions in turn. © 2012 Elsevier Ltd.

    DOI

  • Development of a subchannel void sensor and two-phase flow measurement in 10×10 rod bundle

    Arai, T., Furuya, M., Kanai, T., Shirakawa, K.

    International Journal of Multiphase Flow   47   183 - 192  2012  [Refereed]

     View Summary

    An accurate and detailed experimental database is crucial for modeling the multidimensional two-phase flow and for validating the numerical calculation results. In particular, a two-phase flow in the rod bundle flow channel is so complicated that it is difficult to measure a multidimensional flow structure. Based on the available reference, a point-measurement sensor for acquiring void fractions and bubble velocity distributions do not infer interactions of the subchannel flow dynamics, such as a cross flow and flow distribution, etc. In order to acquire multidimensional two-phase flow in a 10 × 10 rod bundle with an o.d. of 10. mm and length of 3110. mm, a new sensor consisting of 11 × 11 wire and 10 × 10 rod electrodes was developed. The electrical potential in the proximity region between the two wires creates a void fraction in the central subchannel, like a so-called wire-mesh sensor. A unique feature of the devised sensor is that the void fraction near the rod surface can be estimated from the electrical potential in the proximity region between one wire and one rod, meaning the additional 400 points of void fraction and phasic velocity in the 10 × 10 rod bundle can be acquired. The devised sensor demonstrates multidimensional flow structures, i.e. void fraction, phasic velocity, sauter mean diameter and interfacial area concentration distributions. Acquired data exhibit complexity of two-phase flow dynamics in a rod bundle flow channel, such as coalescence and the breakup of bubbles in transient phasic velocity distributions. © 2012 Elsevier Ltd.

    DOI

  • Validation of trace code for flashing-induced density wave oscillations in SIRIUS-N facility, which simulates ESBWR

    Masahiro Furuya, Yoshihisa Nishi, Nobuyuki Ueda

    International Conference on Nuclear Engineering, Proceedings, ICONE   3 ( 1 ) 713 - 719  2012  [Refereed]

     View Summary

    The TRACE code was validated against the flashing-induced density wave oscillation in the SIRIUS-N facility at low pressure (from 0.1 to 0.5 MPa) as a part of the international CAMP-Program of USNRC. The SIRIUS-N facility is a scaled copy of natural circulation BWR (ESBWR). Stability map of TRACE agrees with that of SIRIUS-N facility at low subcooling region, though instability observed in the lower heat flux and higher subcooling region from the stability limit of experiment. The TRACE code demonstrates the flashing-induced density wave oscillation characteristics: The oscillation period correlates well with the transit time of single-phase liquid in the chimney regardless of the system pressure, inlet subcooling, and heat flux. Unlike Type-I and II density wave oscillations, the inlet or exit throttling does not affect stability boundary and oscillation amplitude of flashing-induced density wave oscillations significantly. Increasing pressure decreases oscillation amplitude. The comprehensive validation confirms that the TRACE code can demonstrate thermal-hydraulic stability of natural circulation BWRs. Copyright © 2012 by ASME.

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  • Microstructure of oxide layers formed on zirconium alloy by air oxidation, uniform corrosion and fresh-green surface modification

    Sawabe, T., Sonoda, T., Furuya, M., Kitajima, S., Kinoshita, M., Tokiwai, M.

    Journal of Nuclear Materials   419 ( 1-3 ) 310 - 319  2011  [Refereed]

     View Summary

    Cladding materials with superior corrosion resistance and anti-hydrogen pickup have been developed for high burnup nuclear fuel. We have suggested a surface modification of the cladding materials for this purpose and invented a new surface modification method "Fresh-Green". The Fresh-Green treatment oxidizes and carbonizes a material surface in the same process. Zircaloy-2 with the Fresh-Green treatment showed the improvement of corrosion resistance in autoclave tests. In order to investigate the effect of surface modifications on the corrosion resistance, a synchrotron radiation experiment and a TEM observation were performed on different oxide layers formed on Zircaloy-2. The oxide layers were formed by air-oxidation, an autoclave test and the Fresh-Green treatment. Crystal structures of all the samples were transformed as Zr > Zr3O > tetragonal ZrO2 > monoclinic ZrO2 from the matrix to the surface. Columnar grains of monoclinic zirconia were arranged unidirectionally in the Fresh-Green oxide layer treated at a low temperature. Diffusing capacity for oxygen influenced the crystal structure of the oxide layers. © 2011 Elsevier B.V. All rights reserved.

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  • Effect of salt additives on film boiling heat transfer and mechanism of quenching temperature rise

    Arai, T., Furuya, M.

    Heat Transfer - Asian Research   40 ( 2 ) 101 - 113  2011  [Refereed]

     View Summary

    A high-temperature stainless-steel sphere was immersed into various salt solutions to investigate the film boiling behavior at vapor film collapse. The film boiling behavior around the sphere was observed with a digital video camera. Both surface temperature of the sphere and solid-liquid contact behavior were measured. Results of the experiment showed that salt additives enhanced condensation heat transfer, and the observed vapor film was thinner. Furthermore, the frequency of direct contact between the sphere surface and coolant increased. The quenching temperature increased with increased salt concentration, and was highly correlated with ion molar concentration, which represents the density of ions regardless of the type of salt. © 2010 Wiley Periodicals, Inc.

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  • Safety Evaluation Based on Uncertainty and Scaling Analyses with Statistical Approach by CRIEPI

    FURUYA Masahiro, NISHI Yoshihisa

    Journal of the Atomic Energy Society of Japan   52 ( 2 ) 86 - 90  2010.02  [Refereed]

    DOI CiNii

  • Innovative ultra rapid cooling and atomizing process utilizing vapor explosion and production of new functional powders

    Masahiro Furuya, Takahiro Arai

    Proceedings of the World Powder Metallurgy Congress and Exhibition, World PM 2010   1  2010  [Refereed]

     View Summary

    We propose CANOPUS method, which is innovative rapid cooling and liquid atomization process utilizing a sustainable small-scale spontaneous vapor explosion with a vapor explosion promoter as a quenchant. In this manner, one can utilize vapor explosion securely and efficiently to produce the fine amorphous powder. The cooling rate of the CANOPUS method is up to 1.5·108 K/s, which is 280 times higher than that of the conventional water atomizing and is thousands times higher than that of the widely-used gas atomizing. CANOPUS method was successfully applied to a highly viscous (18 Pa s) Mora coal gasification slug, resulting in 30 m powders, which can be used for one of the raw materials of cement. CANOPUS process produces the following highly functional powders: (1) homogeneous materials without segregation, (2) amorphous powders, (3) powders with control grain boundary size, and (4) fine-scale powders. The CANOPUS method allows us to produce new functional materials with excellent tenacious, corrosion resistant, and soft magnetic properties for more practical use, which was difficult to amorphize and control grain boundary size by the other cooling methods so far.

  • Study on corrosion control in reactors using Radiation Induced Surface Activation (RISA) - Mechanism behind stainless steel durability due to RISA against crevice corrosion

    Shogo Mabuchi, Tatsuya Hazuku, Shin Ichi Motoda, Tomoji Takamasa, Susumu Uematsu, Masahiro Furuya

    International Conference on Nuclear Engineering, Proceedings, ICONE   2   651 - 656  2010  [Refereed]

     View Summary

    This study examines a corrosion control technique for corrosion-resistant materials or of stainless steel in piping for nuclear reactors. This employs an effect of Radiation Induced Surface Activation (RISA). The experimental results revealed: (1) The mechanism behind the corrosion control proposed by the previous report was confirmed to be appropriate. This via tests that measured the amount of dissolved oxygen and iron ions, in the solution. (2) The corrosion control technique was confirmed to be useful for stainless steel with any kind of metal oxide film coating on the surface. (3) It was also shown to be useful even in actual seawater, due to biological effects, which is a far more severe environment for corrosion control than simple salt water. The corrosion control technique for corrosion-resistant material using RISA in seawater has therefore been shown to offer a significant potential for practical applications in naval architecture and marine structures. Copyright © 2010 by ASME.

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  • Visual observation of fine-scale mixing morphology during vapor explosion and droplet entrapping processes

    Masahiro Furuya, Takahiro Arai

    2010 14th International Heat Transfer Conference, IHTC 14   1   289 - 296  2010  [Refereed]

     View Summary

    The successive stages of vapor explosion were video-framed with an exposure time of 500 ns. In order to attain good repeatability and visibility, a smooth round water droplet was impinged onto a molten alloy surface. This configuration suppresses pre-mixing process prior to triggering of vapor explosion. The cluster of bubble generated by spontaneous bubble-nucleation covered the whole contact area at 0.1 ms after the impingement. Prominent fine mixing between two liquids were found to start at 0.6 ms that resulting in vapor explosion. Droplet entrapping phenomenon frequently occurred on an oxide layer, since coherent mixing was prevented due to unevenly formed oxide layer. © 2010 by ASME.

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  • Transient response of BWR core flow during simulated generator load rejection event

    Masahiro Furuya, Takashi Hara, Shinya Mizokami

    International Conference on Nuclear Engineering, Proceedings, ICONE   3   625 - 631  2009  [Refereed]

     View Summary

    Integral Effects Test (IET) was conducted to investigate the effects of flow redistribution during the generator load rejection event by using the SIRIUS-F facility, which simulates boiling two-phase flow in a BWR core. Owing to the automatic controllers of a recirculation pump inverter and fine-control valves in the facility, the time series of signals of heat flux and mass flux were observed to agree well with those of target rapid flow-decrease events in the previous experimental series. This paper addresses the simulated generator load rejection event, during which the flow and power gradually decrease and the flow takes a turn toward recovery. As a result of the two-parallel channel experiment, mass flux of a hot channel is lower than that of the other during the initial stage. When the void fraction becomes smaller, mass flux of the hot channel is observed to become higher. This phenomenon can be accurately demonstrated with the TRAC-BF1 code as well. The code does, therefore, predict the boiling two-phase flow in a BWR core even at such flow-decrease event. During the event, differential pressure along each channel between the upper and lower plena decreases by several tens of kPa. The relative perturbations of the differential pressure between both channels, however, remain less than 0.4 %, which is a significantly small amount. In conclusion, the differential pressures between the upper and lower plena of two-parallel channels are, therefore, identical to each other regardless of the power. Copyright ©2009 by ASME.

    DOI

  • Study of Mass transfer effect on Flow Accelerated Corrosion

    Yoneda, K., Inada, F., Morita, R., Furuya, M.

    Nihon Kikai Gakkai Ronbunshu, B Hen/Transactions of the Japan Society of Mechanical Engineers, Part B   75 ( 751 ) 427 - 428  2009  [Refereed]

     View Summary

    Flow Accelerated Corrosion (FAC) requires considerable attention in plant piping management, for its potential of catastrophic pipe rupture of main piping systems. In view of fluid dynamics, the most essential factor to be considered is mass transfer at the inner surface of the pipe. Mass transfer coefficients are determined by fluid properties and piping geometry, however, no universal correlation exists, which is adaptable to various types of piping elements with strong turbulence. In this study, the modeling of mass transfer coefficient was progressed based on Chilton-Colburn analogy and utilizing "effective friction velocity" from the hydraulics in the viscous sub-layer along the wall. FAC experiments with PWR condensate water condition and CFD for the flow were conducted with a contracted rectangular duct. By considering the turbulent velocity of the viscous layer into the mass transfer coefficient, the correlation with the FAC thinning rate improved, effectively.

    DOI

  • Effect of salt additives on film boiling heat transfer and mechanism of quenching temperature rise

    Arai, T., Furuya, M.

    Nihon Kikai Gakkai Ronbunshu, B Hen/Transactions of the Japan Society of Mechanical Engineers, Part B   75 ( 758 ) 1932 - 1938  2009  [Refereed]

     View Summary

    A high-temperature stainless-steel sphere was immersed into various salt solutions to test film boiling behavior at vapor film collapse. The film boiling behavior around the sphere was observed with a digital-video camera. Because salt additives enhanced condensation heat transfer, the observed vapor film was thinner. Surface temperature of the sphere was measured. Salt additives increased the quenching (vapor film collapse) temperature, because frequency of direct contact between sphere surface and coolant increased. Quenching temperature rises with increased salt concentration. The quenching temperature is well correlated with ion molar concentration, which is a number density of ions, regardless of the type of hydrated salts.

    DOI

  • Effect of nanofluid on the film boiling behavior at vapor film collapse

    Takahiro Arai, Masahiro Furuya

    International Conference on Nuclear Engineering, Proceedings, ICONE   3   633 - 638  2009  [Refereed]

     View Summary

    A high-temperature stainless-steel sphere was immersed into Al 2O3 nanofluid to investigate film boiling heat transfer and collapse of vapor film. Surface temperature is referred to the measured value of thermocouples embedded into and welded onto a surface of the sphere. A direct contact between the immersed sphere and Al2O3 nanofluids is quantified by the acquired electric conductivity. The Al 2O3 nanofluid concentration is varied from 0.024 to 1.3 vol%. A film boiling heat transfer rate of Al2O3 nanofluid is almost the same or slightly lower than that of water. A quenching temperature rises slightly with increased the Al2O3 nanofluid concentrations. In both water and Al2O3 nanofluid, the direct contact signals between the sphere and coolant were not detected before vapor film collapse. Copyright ©2009 by ASME.

    DOI

  • Effect of hydrated salt additives on film boiling behavior at vapor film collapse

    Arai, T., Furuya, M.

    Journal of Engineering for Gas Turbines and Power   131 ( 1 )  2009  [Refereed]

     View Summary

    A high-temperature stainless steel sphere was immersed into various salt solutions to investigate film boiling behavior at vapor film collapse. The film boiling behavior around the sphere was observed with a high-speed digital-video camera. Because the salt additives enhance the condensation heat transfer, the observed vapor film was thinner. The surface temperature of the sphere was measured. Salt additives increased the quenching (vapor film collapse) temperature because the frequency of direct contact between the sphere surface and the coolant increased. Quenching temperature increases with increased salt concentration. The quenching temperature, however, approaches a constant value when the salt concentration is close to its saturation concentration. The quenching temperature is well correlated with ion molar concentration, which is a number density of ions, regardless of the type of hydrated salts. © 2009 by ASME.

    DOI

  • Effect of Ni/Fe ratio and Ni concentration on crud deposition behavior on heated zircaloy-4 surface in simulated pwr primary water

    Hirotaka Kawamura, Masahiro Furuya

    14th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors 2009   2   1124 - 1135  2009  [Refereed]

     View Summary

    Crud deposition on fuel cladding surface would become significant issue in Japanese PWRs, because the large amounts of crud deposition has led to an increase of the dose rate in the primary coolant system and become a root cause of axial offset anomalies (AOA). In order to clarify the contribution factors of the deposition, the effects of heat flux, Ni/Fe ratio and nickel concentration in the test solution on the crud deposition were investigated in a simulated Japanese PWR fuel cycle chemistry at 325°C under sub-cooled boiling and non-irradiated condition. From the test results, it was revealed that the crud layer composed of NiFe2O4 and NiO was formed on the fuel cladding. The amounts of deposited crud layer increased with increase of heat flux and Ni/Fe ratio in the test solution. Ni contents in the crud layer increased with increase of Ni concentration in the test solution.

  • Crevice corrosion control for stainless steel using radiation-induced surface activation

    Taichi Kato, Tatsuya Hazuku, Shin Ichi Motoda, Tomoji Takamasa, Mamoru Hishida, Takanori Kumata, Hiroaki Abe, Masahiro Furuya

    International Congress on Advances in Nuclear Power Plants 2009, ICAPP 2009   3   2088 - 2094  2009  [Refereed]

     View Summary

    When a semiconductor film is irradiated by γ-rays, excited electrons are transferred to a base metal in contact with the film, resulting in cathodic-anodic reactions and surface activation of the metal oxide film. The authors first produced radiation-induced surface activation (RISA) in 2000 and have used it in the development of a new corrosion protection method. This report describes a corrosion mitigation technique based on RISA to prevent crevice corrosion in stainless steel, using low-intensity radiation. Experimental results show that an electrode potential of -100 mV vs. Ag/AgCl was produced and maintained on TiO2-coated SUS304 stainless steel specimens immersed in artificial seawater and in close contact with a small, sealed 60Co source (external irradiation) or activated by neutron irradiation to become self-exciting, with no corrosion observed for more than 7 days. In contrast, the potential of a specimen without a radiation source decreased to less than -280 mV vs. Ag/AgCl and crevice corrosion occurred beneath the O-ring within a few days. The corrosion control mechanism was explored by measurement of dissolved oxygen and iron ions in the solution.

  • Interfacial phenomena of radiation-induced and photo-induced

    Yoshio Honjo, Masahiro Furuya, Tomoji Takamasa, Koji Okamoto

    International Conference on Nuclear Engineering, Proceedings, ICONE   3   353 - 360  2008  [Refereed]

     View Summary

    When a metal oxide is irradiated by gamma rays, the irradiated surface becomes hydrophilic. This surface phenomenon is called as radiation induced surface activation (RISA). In order to investigate radiation-induced and photo- induced hydrophilicity, the contact angles of water droplets on a titanium dioxide surface were measured in terms of irradiation intensity and time for gamma rays of cobalt-60 and for ultraviolet rays. Reciprocals of the contact angles increased in proportion to irradiation time before the contact angles reached their super-hydrophilicity state. The reciprocals of contact angles correlate well with integrated intensity by a straight line, regardless of the irradiation intensity and time. Radiation induced and photo-induced hydrophilicity phenomena are identical to each other in this regard. In addition, an effect of ambient gas was investigated. In pure argon gas, the contact angle remains the same against the irradiation time. This clearly shows that a certain humidity in ambicnt gas is required to take thc placc of RISA hydrophilicity. A single crystal titanium dioxide (100) surface was analyzed by X-ray photoelectron spectrometry (XPS). After irradiation with gamma rays, a peak was found in the 0 Is spectrum, indicating the adsorption of dissociative water to a surface 5-fold coordinate titanium site, and the formation of a surface hydroxyl group. We conclude that the RISA hydrophilicity is caused by chemisorption of the hydroxyl group on the surface. © 2008 by ASME.

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  • Effect of surface property of molten metal pools on triggering of vapor explosions in water droplet impingement

    Furuya, M., Arai, T.

    International Journal of Heat and Mass Transfer   51 ( 17-18 ) 4439 - 4446  2008  [Refereed]

     View Summary

    Small-scale experiments have been conducted to investigate the triggering mechanism of vapor explosions. In order to attain good repeatability and visibility, a smooth round water droplet was impinged onto a molten alloy surface. This configuration suppresses premixing events prior to triggering. Six molten metal and alloys were used as the pool liquid. The lower limit of the contact temperature in the vapor explosion region closely agrees with the spontaneous bubble nucleation temperature of water. The upper limit of the initial molten alloy temperature decreases when an oxide layer forms on the surface causing an increase of the emissivity of thermal radiation that has a stabilizing effect on the vapor film. When an oxide layer was formed on the surface, a water droplet was occasionally entrapped into a molten alloy dome, since the oxide layer prevents the droplet from evaporating coherently. The vapor explosion region obtained for the mirror surface is a conservative estimate, since that for the oxide surface fell into the internal region of the vapor explosion for the mirror surface. © 2008 Elsevier Ltd. All rights reserved.

    DOI

  • Effect of hydrated salt additives on film boiling behavior at vapor film collapse

    Takahiro Arai, Masahiro Furuya

    International Conference on Nuclear Engineering, Proceedings, ICONE   3   323 - 332  2008  [Refereed]

     View Summary

    A high-temperature stainless-steel sphere was immersed into various salt solutions to test film boiling behavior at vapor film collapse. The film boiling behavior around the sphere was observed with a high-speed digital-video camera. Because salt additives enhanced condensation heat transfer, the observed vapor film was thinner. Surface temperature of the sphere was measured. Salt additives increased the quenching (vapor film collapse) temperature, because frequency of direct contact between sphere surface and coolant increased. Quenching temperature rises with increased salt concentration. The quenching temperature, however, approaches a constant value when the slat concentration is close to its saturation concentration. The quenching temperature is well correlated with ion molar concentration, which is a number density of ions, regardless of the type of hydrated salts. © 2008 by ASME.

    DOI

  • Collaborative filtering based on a weighted maximum margin matrix factorization

    Furuya M, Oba S, Ishii S

    Proceedings of the 13th International Symposium on Artificial Life and Robotics, AROB 13th'08     543 - 546  2008  [Refereed]

  • Visualization of filament from molten copper droplet before vapor explosion in highly-subcooled water pool

    Takahiro Arai, Masahiro Furuya

    Proceedings - 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH-12    2007  [Refereed]

     View Summary

    Experiments were conducted in which a molten copper droplet was released into a pool of water. Spontaneous vapor explosion did not occur when the water temperature was 50°C or higher. Spontaneous vapor explosions, however, occurred at a rate of 70% when water temperature was 20°C A series of high-speed video images explores the triggering process of spontaneous vapor explosions: (1) when a molten copper droplet is released into pool of water, a vapor film forms and separates the copper droplet and the surrounding water that is a boiling film, (2) a filament of molten copper grows from the surface and deforms the vapor film, (3) since the filament of molten copper has a small heat capacity and is rapidly quenched, the vapor film condenses along its surface, and the molten copper droplet around the filament directly contacts water, and finally (4) triggering of spontaneous vapor explosions occurs from the filament to the whole molten copper droplet. When filament growth was observed, it triggered a spontaneous vapor explosion in almost all cases. When filament growth was not observed, spontaneous vapor explosions were not observed and vapor films therefore stably formed around the molten copper droplets. The authors thus conclude that filaments from the molten copper trigger spontaneous vapor explosions in highly subcooled water.

  • Regional stability estimation of 1/3 MOX BWR-5 core with SIRIUS-F facility

    Masahiro Furuya, Takanori Fukahori, Shinya Mizokami, Jun Yokoya

    Proceedings - 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH-12    2007  [Refereed]

     View Summary

    The SIRIUS-F facility was designed and constructed for highly accurate simulation of channel instability, core-wide instability and regional instability of ABWR. A real-time simulation was performed by the modal-point kinetics of reactor neutronics and fuel-rod thermal conduction on the basis of a measured void fraction in a reactor core section of the facility. A noise analysis method was performed to calculate decay ratios and resonance frequencies from dominant poles of transfer function on the basis of the AR method using time series measurement data of a core inlet flow of the facility. By utilizing this method, one can estimate stability in any specific operating point online without assuming excess conservative conditions. Channel and regional stability experiments were conducted for a wide range of operating conditions including maximum power points along the minimum pump speed line and the natural circulation line of the BWR-5 core with one third of MOX fuels installed. The decay ratios and resonance frequencies are in good agreement with the analyses results calculated by design analysis code, ODYSY The SIRIUS-F experimental results demonstrated stability characteristics such as a stabilizing effect of the power, and revealed a sufficiently large stability margin even under hypothetical conditions of power level.

  • Effects of mechanical constraint and heat capacity on vapor explosions

    Masahiro Furuya, Takahiro Arai

    Proceedings of the 5th International Conference on Nanochannels, Microchannels and Minichannels, ICNMM2007     469 - 476  2007  [Refereed]

     View Summary

    Small-scale experiments have been conducted to investigate the triggering mechanism of vapor explosions. In order to attain good repeatability and visibility, a smooth round water droplet was impinged onto a molten alloy surface. This configuration suppresses premixing events prior to triggering. The effect of the water droplet curvature was found to be negligibly small when the droplet diameter is larger than 4.5 mm. Vapor explosion conditions were identical for the molten tin pool depths ranging from 0.5 to 40 mm. The experimental results and the heat conduction analysis suggest that the length scale required for atomizing and fine mixing in the triggering event of the vapor explosion are sufficiently smaller than the molten tin pool depth of 0.5 mm. Six different kinds of materials were used as the pool liquid. The lower limit of the contact temperature in the vapor explosion region closely agrees with the spontaneous nucleation temperature of water. The upper limit of the initial molten alloy temperature decreases when an oxide layer forms on the surface causing an increase of the emissivity of thermal radiation that has a stabilizing effect on the vapor film. When an oxide layer formed on the surface, a water droplet was occasionally entrapped into a molten alloy dome, since the oxide layer prevents the droplet from evaporating coherently. The vapor explosion region obtained for the mirror surface is a conservative estimate, since that for the oxide surface fell into the internal region of mirror surface. Copyright © 2007 by ASME.

    DOI

  • Development of BWR regional stability experimental facility SIRIUS-F, which simulates thermal hydraulics-neutronics coupling, and stability evaluation of ABWRs

    Furuya, M., Fukahori, T., Mizokami, S.

    Nuclear Technology   158 ( 2 ) 191 - 207  2007  [Refereed]

     View Summary

    To investigate the stability of a boiling water reactor (BWR), the SIRIUS-F facility was designed and built for highly accurate simulation of thermal-hydraulic (channel) instabilities and coupled thermal hydraulics-neutronics instabilities of the BWR. By using two sets of measured void-fraction distributions in a reactor core section of the SIRIUS-F facility, a real-time void-reactivity feedback simulation was performed on the basis of the modal point kinetics of reactor neutronics and fuel rod thermal conduction. A noise analysis method was performed to calculate decay ratios and resonance frequencies from dominant poles of transfer function based on the AR method using time-series measurement data of a core inlet flow of the facility. Channel and regional stability experiments were conducted for a wide range of operating conditions, including maximum power points along the minimum pump speed line and the natural circulation line of advanced BWR plants. The experimentally obtained decay ratios and resonance frequencies are in good agreement with those calculated by the linear stability analysis code ODYSY. The SIRIUS-F experimental results demonstrated stability characteristics as a function of power and revealed a sufficiently large stability margin even under hypothetical power level conditions.

    DOI

  • Corrosion product deposition behavior on heated zircaloy-4 surface in simulated PWR primary water

    Hirotaka Kawamura, Masahiro Furuya

    Canadian Nuclear Society - 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems 2007   1   369 - 384  2007  [Refereed]

     View Summary

    Japanese PWR utilities desire to employ long-term fuel cycle and high bum-up operations. Corrosion product (crud) deposition on fuel cladding surface would become significant issue in Japanese PWRs, because the large amounts of crud deposition has led to an increase of the dose rate in the primary coolant system and become a root cause of axial offset anomalies (AOA) for other utilities. In order to clarify the contribution factors of the deposition, the effects of steaming rate, pHi-, and boron contents in the test solution on the crud deposition were investigated under sub-cooled boiling and non-irradiated condition. From the test results, it was revealed that the spinel oxide layer was formed on the fuel cladding surface in a simulated Japanese PWR fuel cycle chemistry at 325°C. The oxide layer was easily formed on the heated surface of the fuel cladding, and the oxide thickness was increased with increase of steaming rate and decrease of pHT.

  • Study of improvement of quenching process on high-temperature surface by radiation induced surface activation

    Takamasa, T., Hazuku, T., Fukuhara, Y., Hirose, Y., Abe, K., Uematsu, J., Okamoto, K., Furuya, M., Mishima, K., Hibiki, T.

    KURRI Progress Report     179  2006  [Refereed]

  • Atmospheric Corrosion Control on the basis of Radiation Induced Surface Activation

    FURUYA Masahiro, TAKAMASA Tomoji, OKAMOTO Koji, SAEGUSA Toshiari

    Transactions of the Atomic Energy Society of Japan   4 ( 1 ) 84 - 86  2005.03  [Refereed]

    DOI CiNii

  • Regional stability estimation of natural circulation BWRs using SIRIUS-N facility

    Furuya, M., Inada, F., Van Der Hagen, T.H.J.J.

    Journal of Nuclear Science and Technology   42 ( 4 ) 341 - 350  2005  [Refereed]

     View Summary

    The SIRIUS-N facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of the BWR. A real-time simulation was performed in the digital controller for heat conduction in a fuel-rod and modal point kinetics of reactor neutronics using measured void fractions in reactor core sections of the thermal-hydraulic loop. In order to estimate decay ratios, an auto-regressive method has been successfully applied for time series data of the core inlet flow rate. Experiments were conducted with the SIRIUS-N facility for the rated operating condition of 3.13GWt natural circulation BWR. Channel and regional stability decay ratios were determined to be 0.38 and 0.54, respectively, which indicates sufficient margin for the instabilities. Experiments were extended to evaluate the stability sensitivity of design parameters such as the power profile, void reactivity coefficients, core inlet sub-cooling, and the fuel rod time constant. © 2005 Taylor & Francis Group, Ltd.

    DOI

  • Radiation induced surface activation on Leidenfrost and quenching phenomena

    Takamasa, T., Hazuku, T., Okamoto, K., Mishima, K., Furuya, M.

    Experimental Thermal and Fluid Science   29 ( 3 ) 267 - 274  2005  [Refereed]

     View Summary

    Improving the limit of boiling heat transfer or critical heat flux requires that the cooling liquid can contact the heating surface, or a high wettability, highly hydrophilic heating surface, even if a vapor bubble layer is generated on the surface. In our previous study, contact angle, an indicator of macroscopic wettability, of a water droplet on metal oxide at room temperature was measured by image processing of the images obtained by a CCD video camera. The results showed that the surface wettability on metal oxide pieces of titanium, zircaloy No. 4, SUS-304, and copper was improved significantly by the radiation induced surface activation (RISA) phenomenon. To delineate the effect of RISA on heat transferring phenomena, the Leidenfrost condition and quenching of metal oxides irradiated by y-rays were investigated in this study. In the Leidenfrost experiment, when the temperature of the heating surface reached the wetting limit temperature, water-solid contact vanished because a stable vapor film existed between the droplet and the metal surface; i.e., a Leidenfrost condition obtained. The wetting limit temperature increased with integrated irradiation dose. After irradiation, the wet length and the duration of contact increased, and the contact angle decreased. In the quenching test, high surface wettability, or a highly hydrophilic condition, of a simulated fuel rod made of SUS was achieved, and the quenching velocities were increased up to 20-30% after 300 kGy 60Co γ-ray irradiation. © 2004 Elsevier Inc. All rights reserved.

    DOI

  • Flashing-induced density wave oscillations in a natural circulation BWR - Mechanism of instability and stability map

    Furuya, M., Inada, F., Van Der Hagen, T.H.J.J.

    Nuclear Engineering and Design   235 ( 15 ) 1557 - 1569  2005  [Refereed]

     View Summary

    Experiments were conducted to investigate two-phase flow instabilities due to flashing in a boiling natural circulation loop with a chimney at low pressure. The SIRIUS-N facility was designed to have non-dimensional values nearly equal to those of typical natural circulation boiling water reactor (BWR). The observed instability is suggested to be flashing-induced density wave oscillations, since the oscillation period correlated well with the passing time of single-phase liquid in the chimney section regardless of system pressure, heat flux, and inlet subcooling. Stability maps were obtained in reference to the inlet subcooling and the heat flux at the system pressures of 0.1, 0.2, 0.35, and 0.5 MPa. The flow became stable below a certain heat flux regardless of the channel inlet subcooling. The stable region enlarged with increasing system pressure. Thus, the stability margin becomes larger in a startup process of a reactor by pressurizing the reactor sufficiently before heating according to the stability map. © 2005 Elsevier B.V. All rights reserved.

    DOI

  • Development of SIRIUS-N facility with simulated void-reactivity feedback to investigate regional and core-wide stability of natural circulation BWRs

    Furuya, M., Inada, F., Van Der Hagen, T.H.J.J.

    Nuclear Engineering and Design   235 ( 15 ) 1635 - 1649  2005  [Refereed]

     View Summary

    The SIRIUS-N facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of a natural circulation BWR. A real-time simulation was performed in the digital controller for modal point kinetics of reactor neutronics and fuel-rod conduction on the basis of measured void fractions in reactor core sections of the thermal-hydraulic loop. Stability experiments were conducted for a wide range of thermal-hydraulic conditions, power distributions, and fuel rod time constants, including the nominal operating conditions of a typical natural circulation BWR. The results show that there is a sufficiently wide stability margin under nominal operating conditions, even when void-reactivity feedback is taken into account. The stability experiments were extended to include a hypothetical parameter range (double-void reactivity coefficient and inlet core subcooling increased by a factor of 3.6) in order to identify instability phenomena. The regional instability was clearly demonstrated with the SIRIUS-N facility, when the fuel rod time constant matches the oscillation period of density wave oscillations. © 2004 Elsevier B.V. All rights reserved.

    DOI

  • Development of BWR Regional Stability Experimental Facility SIRIUS-F, which Simulates:Thermalhydraulics-Neutronics Coupling in Reactor Core, and Stability Evaluation of ABWR

    FURUYA Masahiro, FUKAHORI Takanori, MIZOKAMI Shinya

    Transactions of the Atomic Energy Society of Japan   4 ( 2 ) 93 - 105  2005  [Refereed]

     View Summary

    The SIRIUS-F facility was designed and constructed for highly accurate simulation of channel, core-wide and regional instabilities of an ABWR. A real-time simulation is performed for the modal-point kinetics of reactor neutronics and fuel-rod conduction on the basis of a measured void fraction in a reactor core section of the facility.<BR>A noise analysis method was performed to calculate decay ratios from dominant poles of transfer function on the basis of the AR method by applying time series of a core inlet flow rate. By utilizing this method, one can estimate stability at any specific operating point online without assuming excess conservative conditions.<BR>Channel and regional stability experiments were conducted for a wide range of operating conditions including maximum power points along the minimum pump speed line and the natural circulation line of the ABWR. The decay ratios and the resonance frequencies are in good agreement with those from the design analysis code, ODYSY. The SIRIUS-F experimental results demonstrated stability characteristics such as a stabilizing effect of the power, and reviled a sufficiently large stability margin even under hypothetical conditions of power enlargement.

    DOI CiNii

  • Characteristics of type-I density wave oscillations in a natural circulation BWR at relatively high pressure

    Furuya, M., Inada, F., Van Der Hagen, T.H.J.J.

    Journal of Nuclear Science and Technology   42 ( 2 ) 191 - 200  2005  [Refereed]

     View Summary

    Experiments were conducted to investigate two-phase flow instabilities in a boiling natural circulation loop with a chimney at high pressure. The SIRIUS-N facility was designed to have non-dimensional values which are nearly equal to those of a typical natural circulation BWR. The observed oscillations are found to be density wave oscillations, since the void fractions in the chimney inlet and exit are out of phase. They belong to the Type-I category, since they occur at low flow qualities, according to the Fukuda—Kobori's classification. Moreover, the oscillation period correlates well with the passing time of bubbles in the chimney section regardless of the system pressure, the heat flux, and the inlet subcooling. Two distinct phenomena are found in relation between the oscillation period and liquid passing time in the chimney, indicating that the driving mechanisms of the instabilities are different between low and high pressures. Stability maps were obtained in reference to the inlet subcooling and the heat flux at the system pressures of 1, 2, 4, and 7.2 MPa. The flow became stable below a certain heat flux regardless of the channel inlet subcooling. The stable region enlarges with increasing system pressure. Thus, the stability margin becomes larger in a startup process of a reactor by pressurizing the reactor sufficiently before withdrawing the control rods. The obtained stability map demonstrates that the nominal operating condition of the ESBWR has a significant stability margin to the unstable region. © 2005 Taylor & Francis Group, Ltd.

    DOI

  • Study on improvement of leidenfrost and quenching properties by radiation induced surface activation

    Takamasa, T., Hazuku, T., Fukuhara, Y., Tamura, N., Takano, M., Hayashi, T., Okamoto, K., Imai, Y., Furuya, M., Mishima, K., Hibiki, T.

    KURRI Progress Report     34  2004  [Refereed]

  • Mechanism and Evaluation of Flashing-Induced Density Wave Oscillations in Natural Circulation BWR

    FURUYA Masahiro, INADA Fumio

    Transactions of the Atomic Energy Society of Japan   3 ( 2 ) 141 - 150  2004  [Refereed]

     View Summary

    Experiments were conducted to investigete two-phase flow stability of a natural circulation BWR due to flashing at low pressure. The facility used in the experiment was designed to have non-dimensional values which are nearly equal to those of typical natural circulation BWR. The observed instability is suggested to be the flashing induced density wave oscillations, since the oscillation period was nearly one and a half to two times the passing time in the chimney section, and correlated well with a single line regardless of system pressure, heat flux, and inlet subcooling. Stability maps were obtained in reference to the core inlet subcooling and the heat flux at the system pressures of 0.1, 0.2, 0.35, and 0.5MPa. The flow became stable below a certain heat flux regardless of the channel inlet subcooling. The stable region enlarged with increasing system pressure. According to the stability map, the stability margin becomes larger in a startup process of a reactor by pressurizing the reactor sufficiently before withdrawing control rods.

    DOI CiNii

  • Electrical-chemical reaction of radiation induced surface activation phenomenon

    Takamasa, T., Hazuku, T., Fukuhara, Y., Tamura, N., Takano, M., Hayashi, T., Okamoto, K., Imai, Y., Furuya, M., Mishima, K., Hibiki, T.

    KURRI Progress Report     124  2003  [Refereed]

  • Effects of polymer, surfactant, and salt additives to a coolant on the mitigation and the severity of vapor explosions

    Furuya, M., Kinoshita, I.

    Experimental Thermal and Fluid Science   26 ( 2-4 ) 213 - 219  2002  [Refereed]

     View Summary

    Effects of mitigation and suppression of vapor explosion have been investigated by adding a surfactant, polymer, and neutral salt into the water droplet impinging onto a molten alloy pool surface. This configuration was selected to attain good reproducibility and visibility, because premixing events prior to the triggering restrained. In addition, an adequate amount of a surfactant can be adsorbed at the vapor/liquid interface compared to that in the case of conventional configurations such as a molten alloy droplet injecting into a water pool. Dilute anionic and nonionic surfactant aqueoues solutions did not affect the triggering conditions or the pressure pulse generated by vapor explosion, even if the surfactant was added up to a density 25 times higher than the critical micelle concentration. Polymeric additives, which increase the viscosity of a fluid, have little suppression effect on vapor explosion. Spontaneous vapor explosion was, however, suppressed by a 200 wppm polyethylene glycol (molecular weight of 4 x 106) solution, since deposition of the solute due to cloudy-point phenomenon may stabilize the vapor film and prevent the solution from mixing finely. In order to exert this effect, molecular weight should be heavier so that a cloudy-point temperature is below the boiling point at the tested system pressure. Additionally, this threshold concentration became denser as the impingement velocity increased. Thus, a denser concentration and heavier molecular weight should be used to suppress vapor explosion when the vapor film may be subjected to large inertia and/or external force. When a neutral salt was added, the initial molten alloy temperature range where vapor explosion occurred shifted to a higher temperature and became wider. Vapor explosion was observed on the molten zinc surface, which does not explode spontaneously for water. © 2002 Elsevier Science Inc. All rights reserved.

    DOI

  • Effect of liquid density differences on boiling two-phase flow stability

    Furuya, M., Manera, A., Van Bragt, D.D.B., Van der Hagen, T.H.J.J., De Kruijf, W.J.M.

    Journal of Nuclear Science and Technology   39 ( 10 ) 1094 - 1098  2002  [Refereed]

     View Summary

    In order to investigate the effect of considering liquid density dependence on local fluid temperature in the thermal-hydraulic stability, a linear stability analysis is performed for a boiling natural circulation loop with an adiabatic riser. Type-I and Type-II instabilities were to investigate according to Fukuda-Kobori's classification. Type-I instability is dominant when the flow quality is low, while Type-II instability is relevant at high flow quality. Type-II instability is well known as the typical density wave oscillation. Neglecting liquid density differences yields estimates of Type-II instability margins that are too small, due to both a change in system-dynamics features and in the operational point. On the other hand, neglecting liquid density differences yields estimates of Type-I stability margins that are too large, especially due to a change in the operational point. Neglecting density differences is thus non-conservative in this case. Therefore, it is highly recommended to include liquid density dependence on the fluid subcooling in the stability analysis if a flow loop with an adiabatic riser is operated under the condition of low flow quality. © 2002 Taylor and Francis Group, LTD.

    DOI

  • A Study on Thermo-Hydraulic Instability of Boiling Natural Circulation Loop with a Chimney. 4th Report. An Analytical Consideration of the Stability and Thermo-Hydraulic Characteristics in the Chimney in High Pressure.

    INADA Fumio, FURUYA Masahiro, YASUO Akira

    TRANSACTIONS OF THE JAPAN SOCIETY OF MECHANICAL ENGINEERS Series B   68 ( 666 ) 511 - 518  2002  [Refereed]

     View Summary

    Thermo-hydraulic instabilities of a boiling natural circulation loop with a chimney under high pressure were investigated using linear stability analysis. Drift-flux model was used for two-phase flow model. The instability regions as well as the thermo-hydraulic characteristics in the chimney such as wavy feature were examined, which were compared with the characteristics in low pressure. Instability could occur when exit quality was relatively low, which was the same manner as the characteristics in low pressure. In high-pressure, void was generated near channel exit, and void wave propagated in the chimney. In low pressure, steam was generated only near the chimney exit due to gravity induced flashing, and single-phase enthalpy wave, that is, temperature wave propagated in single-phase flow region. Though flow could be very stable in the high pressure and high power condition, the decay ratio of higher mode could be larger than that of lower mode.

    DOI CiNii

  • A Linear Stability Analysis of a Vapor Film in Terms of the Triggering of Vapor Explosions.

    FURUYA Masahiro, MATSUMURA Kunihito, KINOSHITA Izumi

    TRANSACTIONS OF THE JAPAN SOCIETY OF MECHANICAL ENGINEERS Series B   68 ( 675 ) 3176 - 3182  2002  [Refereed]

     View Summary

    A Detailed analytical model to explain the vapor film collapse was developed to evaluate the occurrence conditions of self-triggering vapor explosions. The following conclusions were drawn based on linear stability analysis using the thermo-dynamic property of water, by linearizing and perturbing basic equations (Rayleigh-Lamb-Plesset's bubble momentum equation, the mass conservation equation, the state equation for ideal gas, and the Clausius-Clapeyron equation). The vapor film stabilizes with the reduction of the hot-liquid diameter, decreasing the condensation heat transfer coefficient, and increasing the thermal radiation coefficient. The cold-liquid viscosity and surface tension have a stabilizing effect, though this effect is negligibly small where the hot-liquid diameter is over 1 mm. The analysis predicts the vapor explosion occurrence limits obtained experimentally by other researchers to within approximately 10 K.

    DOI CiNii

  • Development of Core-Wide and Regional Stability Test Facility, SIRIUS, that Simulates Void Reactivity Feedback, and Stability Evaluation

    FURUYA Masahiro, KUBO Yuichiro, INADA Fumio, YASUO Akira

    JOURNAL OF THE ATOMIC ENERGY SOCIETY OF JAPAN   43 ( 10 ) 1027 - 1038  2001.10  [Refereed]

     View Summary

    The SIRIUS facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of the BWR. A real-time simulation was performed in the digital controller for modal point kinetics of reactor neutronics and fuel-rod conduction on the basis of measured void fractions in reactor core sections of the thermal-hydraulic loop, which simulates thermal-hydraulics of a natural circulation BWR. A solid-state, series-regulated power supply, that plays a role of simulation output, was designed to attain fast response speed without loss of accuracy.
    A noise analysis method was performed to calculate decay ratios from dominant poles of a transfer function by applying AR method to time series of the core inlet flow rate. Experiments were conducted with the SIRIUS facility for the nominal operating condition of 3.13GWt natural circulation BWR. Channel and regional stability decay ratios were determined to be 0.38 and 0.54, respectively, which indicates sufficient margin for the instabilities, Experiments were extended to evaluate the stability sensitivity of the design parameters such as the power profile on the basis of three-dimensional steady-state analysis, the void reactivity coefficients, the core inlet subcooling, and the thermal conductance of the fuel rod.

    CiNii

  • Inlet throttling effect on the boiling two-phase flow stability in a natural circulation loop with a chimney

    Furuya, M., Inada, F., Yasuo, A.

    Heat and Mass Transfer/Waerme- und Stoffuebertragung   37 ( 2-3 ) 111 - 115  2001  [Refereed]

     View Summary

    Experiments have been conducted to investigate an effect of inlet restriction on the thermal-hydraulic stability. A Test facility used in this study was designed and constructed to have non-dimensional values that are nearly equal to those of natural circulation BWR. Experimental results showed that driving force of the natural circulation at the stability boundary was described as a function of heat flux and inlet subcooling independent of inlet restriction. In order to extend experimental database regarding thermal-hydraulic stability to different inlet restriction, numerical analysis was carried out based on the homogeneous flow model. Stability maps in reference to the core inlet subcooling and heat flux were presented for various inlet restrictions using the above-mentioned function. Instability region during the inlet subcooling shifted to the higher inlet subcooling with increasing inlet restriction and became larger with increasing heat flux.

    DOI

  • Experimental Study on Convective Heat Transfer with Thin Porous Bodies.

    NISHI Yoshihisa, KINOSHITA Izumi, FURUYA Masahiro

    TRANSACTIONS OF THE JAPAN SOCIETY OF MECHANICAL ENGINEERS Series B   67 ( 661 ) 2274 - 2280  2001  [Refereed]

     View Summary

    Experimental studies are made on the convective heat transfer of three types of thin porous bodies. Heat transfer performances, flow patterns and temperature profiles near the porous bodies are compared with each other. The heat transfer performance of porous bodies with the largest pore diameter is large. It became clear that the high heat transfer performance depends on an excellent heat transportation ability inside the pore and near the surface of the porous bodies.

    DOI CiNii

  • Thermo-hydraulic instability of boiling natural circulation loop induced by flashing (analytical consideration)

    Inada, F., Furuya, M., Yasuo, A.

    Nuclear Engineering and Design   200 ( 1 ) 187 - 199  2000  [Refereed]

     View Summary

    An analytical study is presented on the thermo-hydraulic stability of a boiling natural circulation loop with a chimney at low pressure start-up. The effect of flashing induced by the pressure drop in the channel and the chimney due to gravity head on the instability is considered. A method to analyze linear stability is developed, in which a drift-flux model is used. The analytical result of a stability map agrees very well with the experimental one obtained in a previous report. Instability does not occur when the heater power is too low to generate voids in the chimney and only natural circulation of single phase can be induced. Instability tends to occur when boiling occurs only near the chimney exit due to flashing. This instability phenomenon has some similarities with density wave oscillation, such as the phase difference of temperature between the boiling region and non-boiling region, and the oscillation period which is near to the time required for fluid to pass through the chimney. However, there are also some differences from density wave oscillation, such as the boiling region is very short, and pressure fluctuation can affect void fraction fluctuation.

    DOI

  • Compact reversed shear tokamak reactor with a superheated steam cycle

    Okano, K., Asaoka, Y., Yoshida, T., Furuya, M., Tomabechi, K., Ogawa, Y., Sekimura, N., Hiwatari, R., Yamamoto, T., Ishikawa, T., Fukai, Y., Hatayama, A., Inoue, N., Kohyama, A., Shinya, K., Murakami, Y., Senda, I., Yamazaki, S., Mori, S., Adachi, J., Takemoto, M.

    Nuclear Fusion   40 ( SPEC. ISS. 3 ) 635 - 646  2000  [Refereed]

     View Summary

    The compact reversed shear tokamak CREST is a cost competitive reactor concept based on a reversed shear high β plasma and water cooled ferritic steel components. The moderate aspect ratio A = 3.4 and the elongation κ = 2.0 of CREST are very similar to the case of the ITER advanced mode plasma. Presentation of such a concept based on the ITER project should be worth while for formulating a fusion development strategy. The achievement of a competitive cost of electricity (COE) is the first priority for electric power industries. High β and high thermal efficiency are the most effective parameters for achieving a competitive COE. In order to achieve a high efficiency power plant, a superheated steam cycle has been adopted which permits a high thermal efficiency (η = 41%). Current profile control and high speed plasma rotation by neutral beam current drive stabilize the ideal MHD activity up to the Troyon coefficient βN = 5.5. A cost assessment has shown that CREST could generate about 1.16 GW(e) electric power at a competitive cost.

    DOI

  • Vapor explosion in the droplet impingement system

    M Furuya, Kinoshita, I, Y Nishi

    HEAT TRANSFER 1998, VOL 5     471 - 476  1998  [Refereed]

     View Summary

    In order to investigate the triggering mechanism of vapor explosion, small-scale experiments have been conducted to observe and examine explosive fragmentation characteristics due to thermal interaction between an impinging droplet and a molten alloy pool. The process that initiates vapor explosion was clearly observed in high-speed videotaping frames: a hemispherical water droplet fragmented from the bottom surface where spontaneous bubble-nucleation took place immediately after making direct contact with a molten alloy pool surface, and small particles of the alloy also fragmented and dispersed due to the high pressure that resulted from vapor explosion. Conditions in which vapor explosion occurred were investigated for various initial temperatures of the droplet and molten alloy peel. It is found from wave height for deformed molten alloy surface that explosion intensity diminishes with decreasing droplet subcooling.

  • Thermal-Hydraulic Instability of Boiling Natural Circulation Loop with a Chimney (3rd Report, Instability at High System Pressure)

    FURUYA Masahiro, INADA Fumio, YASUO Akira

    Transactions of the Japan Society of Mechanical Engineers. Series B.   63 ( 612 ) 2757 - 2763  1997.08  [Refereed]

     View Summary

    Experiments have been conducted to investigate thermal-hydraulic instabilities at system pressure ranging from 1 to 7.2 MPa in a boiling natural circulation loop with a chimney. Stability maps in reference to the system pressure, the channel inlet subcooling, and heat flux are presented. Two different types of instabilities were observed in the facility at relatively low and high system pressures. Both of the instability mechanisms were clarified by investigation of the transient flow pattern and the response of the driving force of the circulation to momentum energy.

    CiNii

  • Visualization and Image-Processing of Multiphase Flow by Neutron Radiography

    TAKENAKA Nobuyuki, ASANO Hitoshi, FUJII Terushige, NISHI Y, FURUYA M, KINOSHITA I, MATSUBAYASHI Masahito

    Neutron Radiography (5)   pp. 383-394.  1997  [Refereed]

  • Thermohydraulic instability of boiling natural circulation loop with a chimney. (Part I. Linear stability analysis using homogeneous two-phase flow model and experiment on thermohydraulic instability induced by flashing)

    Inada, F., Furuya, M., Yasuo, A.

    Heat Transfer - Japanese Research   24 ( 7 ) 563 - 576  1995  [Refereed]

     View Summary

    Instability of a boiling natural circulation loop with a chimney at low pressure and low heater power was investigated by linear stability analysis and experiment. A homogeneous and thermodynamic equilibrium model for two-phase flow was used. The effect of flashing induced by pressure drop in the heated channels and the chimmey was considered. The effects of coupling between two boiling channels were investigated. It was found that in-phase-mode instability was apt to occur when channel inlet subcooling was large and boiling began in the chimney. In-phase-mode instability easily occurred when channel length was shortened and chimney length was increased. Out-of-phase-mode instability was apt to occur when chimney length was decreased and boiling began in the channel.

  • Thermohydraulic Instability of Boiling Natural Circulation Loop with a Chimeny. 1st Report, Linear Stability Analysis Using Homogenous Two-Phase Flow Model and Experiment on Thermohydraulic Instability Induced by Flashing.

    Inada Fumio, Furuya Masahiro, Yasuo Akira

    TRANSACTIONS OF THE JAPAN SOCIETY OF MECHANICAL ENGINEERS Series B   61 ( 591 ) 4067 - 4073  1995  [Refereed]

     View Summary

    Instability of a boiling natural circulation loop with a chimney at low pressure and low heater power was investigated by linear stability analysis and experiment. A homogeneous and thermodynamic equilibrium model for two-phase flow was used. The effect of flashing induced by pressure drop in the heated channels and the chimney was considered. The effects of coupling between two boiling channels were investigated. It was found that in-phase-mode instability was apt to occur when channel inlet subcooling was large and boiling began in the chimney. In-phase-mode instability easily occurred when channel length became short and the chimney became long. Out-of-phase-mode instability was apt to occur when chimney length became small and boiling began in the channel. It was suggested that in-phase-mode instability was density weve oscillation induced by flashing in the chimney and out-of-phase-mode instability was density wave oscillation induced by boiling in the channels. The analytical results agreed qualitatively with experimental results.

    DOI CiNii

  • A study on Thermo-Hydraulic Instability of Boiling Natural Circulation Loop with a Chimeny. 2nd Report, Experimental Approach to Clarify the Flow Instability in Detail.

    Furuya Masahiro, Inada Fumio, Yasuo Akira

    TRANSACTIONS OF THE JAPAN SOCIETY OF MECHANICAL ENGINEERS Series B   61 ( 591 ) 4074 - 4080  1995  [Refereed]

     View Summary

    Experiments are are conducted to investigate two-phase flow instabilities in a boiling natural circulation loop with a chimney induced by flashing in the chimney at lower pressure. The type of instability that occurred in the experiments is suggested to be density wave oscillations induced by flashing in the chimney. The differences from other instabilities such as geysering, flow pattern transition instability, and natural circulation oscillations are discussed on the basis of the dynamic characteristics, the oscillation period, and the transient flow resume.

    DOI CiNii

  • Direct-contact heat-transfer characteristics between a melted alloy and water

    Kinoshita, I., Nishi, Y., Furuya, M.

    Heat Transfer - Japanese Research   24 ( 4 ) 397 - 407  1995  [Refereed]

     View Summary

    As a candidate for an innovative system generator for fast-breeder reactors, a heat exchanger with direct-contact heat transfer between a melted alloy and water was proposed. In this study, the effect of pressure on the heat-transfer characteristics and the required degree of superheating of the melted alloy above water-saturation temperatures are evaluated during the direct-contact heat-transfer experiment by injecting water into Wood's alloy. As a result of the experiment, the product of the degree of Wood's alloy superheating above the water-saturation temperature and the depth of the feed-water injection point is constant for each pressure. This constant increases as the pressure rises.

  • Direct contact heat transfer characteristics between melting alloy and water

    Kinoshita, Izumi, Nishi, Yoshihisa, Furuya, Masahiro

    Nippon Kikai Gakkai Ronbunshu, B Hen/Transactions of the Japan Society of Mechanical Engineers, Part B   61 ( 588 ) 3038 - 3043  1995  [Refereed]

  • A study on thermohydraulic instability of a boiling natural circulation loop with a chimney. (Part II. Experimental approach to clarify the flow instability in detail)

    Furuya, M., Inada, F., Yasuo, A.

    Heat Transfer - Japanese Research   24 ( 7 ) 577 - 588  1995  [Refereed]

     View Summary

    Experiments were conducted to investigate two-phase instabilities in a boiling natural circulation loop induced by flashing in a chimney at lower pressure. The type of instability that occurred in the experiments is suggested to be density wave oscillations induced by flashing in the chimney. The differences from other instabilities such as geysering, flow pattern transition instability, and natural circulation oscillations are discussed on the basis of the dynamic characteristics, the oscillation period, and the transient flow resume.

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    J-GLOBAL

  • Estimation of the In-Depth Debris Status of Fukushima Unit-2 and Unit-3 with Multi-Physics Modeling (2) Points of discussion on debris conditions in Units 2 and 3

    佐藤一憲, 山路哲史, 古谷正裕, 大石佑治, LI Xin, 間所寛, 深井尋史

    日本原子力学会春の年会予稿集(CD-ROM)   2020  2020

    J-GLOBAL

  • Visualization of pool boiling in a heated rod bundle using X-ray radiography and two-phase mixture level fluctuation characteristics

    新井崇洋, 古谷正裕, 白川健悦

    日本伝熱シンポジウム講演論文集(CD-ROM)   57th  2020

    J-GLOBAL

  • Measurement of three-dimensional void fraction distribution in heated rod bundle with unheated rod

    植田翔多, 新井崇洋, 宇井淳, 古谷正裕, 大川理一郎, 白川健悦

    日本伝熱シンポジウム講演論文集(CD-ROM)   57th  2020

    J-GLOBAL

  • Steam Explosion Retardant by Mixing Sodium Bicarbonate and Citric Acid

    古谷正裕, 新井崇洋

    日本伝熱シンポジウム講演論文集(CD-ROM)   57th  2020

    J-GLOBAL

  • Evaluation of effect of spacer and swirl vanes on droplet deposition in bundle flow channel by visualization

    大川理一郎, 古谷正裕, 新井崇洋, 飯山継正

    日本伝熱シンポジウム講演論文集(CD-ROM)   57th  2020

    J-GLOBAL

  • Study on the boiling mechanism under the high pressure region (2) Bubble growth process on visualized boiling images with semantic segmentation algorism

    古谷正裕, 小野綾子, 吉田啓之

    日本原子力学会秋の大会予稿集(CD-ROM)   2020  2020

    J-GLOBAL

  • Research and development for understanding two-phase flow behavior inside a fuel bundle (8) Void fraction distribution around non-heated rods in a rod bundle

    新井崇洋, 宇井淳, 古谷正裕, 大川理一郎, 飯山継正, 植田翔多, 白川健悦

    日本原子力学会秋の大会予稿集(CD-ROM)   2020  2020

    J-GLOBAL

  • Evaluation of swirl vanes on droplet flow passing through BWR spacer

    大川理一郎, 古谷正裕, 新井崇洋, 飯山継正

    日本原子力学会秋の大会予稿集(CD-ROM)   2020  2020

    J-GLOBAL

  • Estimation of the In-Depth Debris Status of Fukushima Unit-2 and Unit-3 with Multi-Physics Modeling (3) Overview of the Project and Progress of the First Year

    山路哲史, 古谷正裕, 大石佑治, 佐藤一憲, 深井尋史, LI Xin, 間所寛

    日本原子力学会秋の大会予稿集(CD-ROM)   2020  2020

    J-GLOBAL

  • Effect of center-positioned unheated rod in 5x5 rod bundle on void behavior during saturated pool boiling

    植田翔多, 新井崇洋, 宇井淳, 古谷正裕, 大川理一郎, 白川健悦

    日本原子力学会秋の大会予稿集(CD-ROM)   2020  2020

    J-GLOBAL

  • Estimation of the In-Depth Debris Status of Fukushima Unit-2 and Unit-3 with Multi-Physics Modeling (4) Consideration on possible RPV-boundary failure modes of Unit 2

    佐藤一憲, 山路哲史, 古谷正裕, 大石佑治, LI Xin, 間所寛, 深井尋史

    日本原子力学会秋の大会予稿集(CD-ROM)   2020  2020

    J-GLOBAL

  • Void-fraction measurement with high spatial resolution in a 5×5 rod bundle by linear-accelerator-driven X-ray computed tomography over a wide pressure range

    Takahiro Arai, Masahiro Furuya, Hiroki Takiguchi, Yoshihisa Nishi, Kenetsu Shirakawa

    Flow Measurement and Instrumentation   69  2019.10

     View Summary

    Void fraction (i.e., the volume fraction occupied by gas) is a key parameter for determining the coolability and neutron-moderating performance of a water-cooled nuclear reactor. To develop computational multi-fluid dynamics models for determining the void-fraction distribution, experimental data of comparable quality are required. We have developed a high-energy X-ray computed tomography (CT) system to acquire three-dimensional void-fraction distributions. The CT system comprises a linear-accelerator-driven high-energy X-ray source and a linear detector array. We quantified a boiling two-phase flow in a 5 × 5 heated rod bundle at high pressure, simulating a fuel-rod bundle in a boiling water reactor (BWR). Because the axial travel of the CT system is 4 m and includes the entire BWR fuel-rod bundle, we optimized the CT imaging conditions and reconstruction method for rod-bundle visualization to reduce uncertainties due to density fluctuations in the boiling flow and imaging artifacts. We conducted a boiling experiment at a low flow rate and low thermal power and acquired three-dimensional distributions of the void fraction over a wide pressure range of 0.1–7.2 MPa. The experiment provided three-dimensional void-fraction distributions with high spatial resolution, especially in subchannel regions surrounded by rods, and the results are suitable for validating three-dimensional thermal-hydraulic analysis codes.

    DOI

  • Two-phase mixture level swell during pool scrubbing in terms of flow geometry

    Takahiro Arai, Masahiro Furuya, Hiroki Takiguchi, Kenetsu Shirakawa

    International Conference on Nuclear Engineering, Proceedings, ICONE   2019-May  2019.05

     View Summary

    When boiling occurs in a pool or gas is injected into a stagnant water pool, the actual liquid level, i.e., the two-phase mixture level, becomes higher than the original liquid level, i.e., the collapsed water level, and depends on the gas held in the liquid pool. The two-phase mixture level is an important indicator of the core cooling of a nuclear reactor and of the long-term operation of the filtered containment venting system under accident conditions. To clarify the relationship between the two-phase mixture and collapsed water levels, we conducted an air-water two-phase flow experiment in which we used different channel geometries, i.e., circular pipes and rod bundles, to inject air at atmospheric pressure into stagnant water. Then, we used visual observation to obtain the swell and fluctuation amplitude of the two-phase mixture level. The results indicate that with increases in the superficial gas velocity, the flow channel geometry significantly affects the swell of the two-phase mixture level, and that the dominant bubble scale in the rod bundle was influenced by both the hydraulic diameter and the channel-box scale. We compared the height and fluctuation amplitude of the two-phase mixture level with existing and newly acquired experimental data and models.

  • Ablation analysis with MPS for proposing ex-vessel corium spreading management in light water reactors

    Masafumi Katta, Guangtao Duan, Akifumi Yamaji, Masahiro Furuya

    International Conference on Nuclear Engineering, Proceedings, ICONE   2019-May  2019.05

     View Summary

    In a postulated sever accident of a light water reactor (LWR), molten core debris (corium) may breach the reactor pressure vessel and be released to the ex-vessel containment floor. A core catcher manages ex-vessel corium cooling by uniformly spreading the corium into a large space, but it requires a dedicated plant design. In contrast, corium shields have been back-fitted to some boiling water reactors to prevent excessive amount of corium to flow into sump pits, where effective corium cooling may be difficult. However, corium shields can only block the corium flow and cannot contribute to uniform spreading of the ex-vessel corium. This study proposes a preliminary concept of “Debris Spreading Floor”, which can be applied to any types of reactor plants including back-fitting to the existing plants. More specifically, the existing containment floor is overlaid with sacrificial material and refractory material is placed around sump pits. It is intended to allow the original function of sump pits to collect leaking water under normal, abnormal transient and design basis accident conditions. However, under postulated severe accident condition, spreading of ex-vessel corium is promoted by ablating itself with the hot corium and guiding corium spreading away from sump pits. To develop the concept, mechanistic analysis of corium spreading, which can consider influence of substrate ablation, is needed. The Moving Particle Semi-implicit (MPS) method is a Lagrangian particle method and thus suitable for mechanistic simulation of free-surface spreading flow involving solid / liquid phase change and interactions. In this research, effect of the proposed concept is firstly presented with the MPS simulations. Preliminary simulations in 2D show that, amount of corium flowing into sump pits is reduced by the concept. Secondly, validity of the MPS simulations is quantitatively discussed by simulating the experiments with simulant. The experiment was carried out by Central Research Institute of Electric Power Industry (CRIEPI) by pouring liquid Pb-Bi onto a Pb-Bi block so that the inflow liquid spreads on the block surface while it also ablates the block. Sensitivity analyses have been carried out with different initial conditions, calculation resolutions, subscale models and parameters of the MPS simulations to identify the key models and parameters for quantitative prediction of the melt / substrate interactions.

  • Large-break LOCA analysis with modified boiling heat-transfer model in TRACE code

    Riichiro Okawa, Masahiro Furuya

    Nuclear Engineering and Design   346   97 - 111  2019.05

     View Summary

    Numerical analyses were conducted to replicate several tests for simulating a double-ended cold-leg large break loss-of-coolant accident (LBLOCA) in the Loss-of-Fluid Test (LOFT) using the TRACE (version 5/patch level 4) code. Analytical results by the original TRACE code were so conservative that especially a first peak of cladding temperature was estimated higher than the experimental data at the blowdown phase and subsequent temperature drop corresponding to the temporal quench was not seen. We were interested in minimum film boiling temperature (T min ) as a heat transfer model factor estimating the quench at the moment, investigated correlation equations for T min in previous studies and especially focused on ones given as a function of coolant mass flow because the complicated flow transient and decompression in the core region at the blowdown phase was interpreted as having an influence on the cladding temperature behavior. There are several correlations meeting the above condition but it was revealed that they are insufficient to apply for high pressure especially. Therefore, a new term including an effect of mass flow flux and time derivative of pressure was defined and added with a proportional coefficient hypothetically to the current correlation in the TRACE code for modification. The LOFT analyses were conducted again using the modified TRACE code, and it was shown by applying roughly the same proportional coefficient to all the cases of LOFT analyses that estimation of the cladding temperature behavior was improved more precisely at the blowdown phase. Also, the transition during the phase was explained phenomenologically with the wall heat transfer mode and boiling curve.

    DOI

  • Three-Dimensional Flow Dynamics of High-Temperature Viscous Fluid and Ablation of Pedestal Floor

    古谷正裕, 古谷正裕, 山路哲史, 大石佑治

    混相流シンポジウム講演論文集(Web)   2019  2019

    J-GLOBAL

  • 近赤外分光法による五ほう酸ナトリウムの濃度拡散挙動の観察

    飯山継正, 古谷正裕, 新井崇洋

    日本原子力学会秋の大会予稿集(CD-ROM)   2019  2019

    J-GLOBAL

  • データ同化手法に基づくサブチャンネル解析コードのモデルパラメータの適正化

    宇井淳, 工藤義朗, 工藤義朗, 古谷正裕

    日本原子力学会秋の大会予稿集(CD-ROM)   2019  2019

    J-GLOBAL

  • BWRスペーサの構造が液滴流動に及ぼす影響評価

    大川理一郎, 古谷正裕, 新井崇洋, 飯山継正

    日本原子力学会秋の大会予稿集(CD-ROM)   2019  2019

    J-GLOBAL

  • 炭酸ガス溶解による水蒸気爆発抑制法の開発

    古谷正裕, 新井崇洋, 飯山継正, 大川理一郎, 宇井淳

    日本原子力学会秋の大会予稿集(CD-ROM)   2019  2019

    J-GLOBAL

  • RIA時の沸騰遷移に関する研究(高温待機時RIA)(6)高圧条件下における急速発熱時の3×3バンドル過渡限界熱流束

    新井崇洋, 古谷正裕, 白川健悦, 宇井淳, 大川理一郎, 佐合優一, 原田健一

    日本原子力学会秋の大会予稿集(CD-ROM)   2019  2019

    J-GLOBAL

  • RIA時の沸騰遷移に関する研究(高温待機時RIA)(5)研究経過及び概要

    原田健一, 佐合優一, 古谷正裕, 土田嗣美, 宮地孝政

    日本原子力学会秋の大会予稿集(CD-ROM)   2019  2019

    J-GLOBAL

  • フィルタベント性能評価のための基盤技術の開発と活用

    金井大造, 古谷正裕, 西義久, 西村聡

    日本保全学会学術講演会要旨集   16th  2019

    J-GLOBAL

  • BWR燃料サブチャンネルのスペーサ下流域における液滴付着効果の可視観察

    大川理一郎, 古谷正裕, 新井崇洋, 飯山継正

    日本伝熱シンポジウム講演論文集(CD-ROM)   56th   ROMBUNNO.F123  2019

    J-GLOBAL

  • マイクロバブルを用いた蒸気爆発抑制対策

    古谷正裕, 新井崇洋

    日本伝熱シンポジウム講演論文集(CD-ROM)   56th   ROMBUNNO.B325  2019

    J-GLOBAL

  • Observation of Concentration Diffusion Behavior of Hydrazine Hydrate

    滝口広樹, 古谷正裕, 新井崇洋

    混相流   33 ( 1 ) 71 - 76  2019

     View Summary

    <p>Highly accurate measurement of thermal properties, reaction rates and mass transfers is necessary to grasp thermal flow characteristics of nuclear reactor plant because these properties are unsteadily sensitive to concentration, temperature, pressure and scale. On the real scale of the plant, however, there are problems such as the complexity of three-dimensional flow and the effect of convection. In this study, a novel technology that measures mass transfer and reaction rate of nuclear application material with high accuracy and high speed by applying microchannel which makes one-dimensional laminar flow, has developed. In this paper, a micro flow control system that can adjust flow with a minimum volume of 1 μL/min. by a pressurized pump, was fabricated. In order to observe concentration diffusion of the flow field in real time, a near infrared spectroscopic visible observation device was developed and combined with the flow system. The mutual diffusion coefficient was measured for hydrazine hydrate, which is added for the purpose of preventing corrosion of secondary piping in nuclear power plant. It revealed that the absorption depends on the concentration near 1555 nm for Fourier transform near-infrared spectrophotometer. The diffusion coefficients of 5, 10 and 20 wt% for hydrazine hydrate were quantified with the luminance profiles in the absorption peak wavelength.</p>

    DOI CiNii J-GLOBAL

  • Multi‐physicsモデリングによるEx‐Vessel溶融物挙動理解の深化(6)全体概要とMPS法によるspreading解析の高度化(3)

    山路哲史, 古谷正裕, 大石佑治, JUBAIDAH, DUAN Guangtao

    日本原子力学会春の年会予稿集(CD-ROM)   2019   ROMBUNNO.2I18  2019

    J-GLOBAL

  • 硝酸銀水溶液の有機ヨウ素除去特性に関する基礎検討

    金井大造, 古谷正裕, 西村聡

    日本原子力学会春の年会予稿集(CD-ROM)   2019   ROMBUNNO.3I03  2019

    J-GLOBAL

  • 流れに平行な円柱のL/D比が抗力係数に及ぼす影響

    飯山継正, 古谷正裕, 白川健悦, 前田義明, 川芳昭, 小池訓弘, 島田太郎

    日本原子力学会春の年会予稿集(CD-ROM)   2019   ROMBUNNO.1I14  2019

    J-GLOBAL

  • BWR燃料サブチャンネルにおけるスペーサを通過する液滴流動の可視観察

    大川理一郎, 古谷正裕, 新井崇洋, 滝口広樹, 飯山継正

    日本原子力学会春の年会予稿集(CD-ROM)   2019   ROMBUNNO.1I10  2019

    J-GLOBAL

  • 高圧・低流量域におけるバンドル内沸騰二相流のサブチャンネル解析

    新井崇洋, 古谷正裕, 宇井淳, 大川理一郎

    日本原子力学会春の年会予稿集(CD-ROM)   2019   ROMBUNNO.2I20  2019

    J-GLOBAL

  • Transient boiling flow in 5 × 5 rod bundle under non-uniform rapid heating

    Hiroki Takiguchi, Masahiro Furuya, Takahiro Arai, Kenetsu Shirakawa

    Nuclear Engineering and Design   340   447 - 456  2018.12

     View Summary

    Rapid thermal elevation in boiling water reactor (BWR) is an important factor for nuclear safety and there is a need to develop an analysis code for the transient phenomenon and its validation process. To evaluate the thermal property of transient boiling and its uncertainty, corroborative experimental information is crucial. In particular, the lateral propagation behavior of a vapor bubble (void) in the cross-sectional direction of fuel assembly has yet to be determined. This study evaluates the void propagation behavior in a 5 × 5 rod bundle with cross-sectional heat distribution that causes only the 3 × 3 rod bundle to generate heat; assuming rapid heating under atmospheric pressure. In this paper, using the maximum heat output applied to the nine heated rods as a parameter, from the visualization of the void behavior and the measurement of the local void fraction, the heat output conditions under circumstances where lateral propagation of voids occurs and where voids are only localized in the heated region are summarized. We quantified the time difference initially detected and the time-averaged void fraction according to the lateral propagation level.

    DOI

  • 鉛直二重管の強制対流サブクール沸騰の計測と気泡挙動の再構成

    宇井淳, 古谷正裕, 滝口広樹, 白川健悦, 新井崇洋

    日本伝熱シンポジウム講演論文集(CD-ROM)   55th   ROMBUNNO.G223  2018

    J-GLOBAL

  • 大気圧から7MPaまでの5×5発熱管群内プール沸騰における二相水位変動

    新井崇洋, 古谷正裕, 滝口広樹, 西義久, 白川健悦

    日本伝熱シンポジウム講演論文集(CD-ROM)   55th   ROMBUNNO.G212  2018

    J-GLOBAL

  • Boiling concentration of seawater and heat-transfer deterioration with salt precipitation in 5X5 bundle geometry

    古谷正裕

    日本機械学会動力・エネルギー技術シンポジウム講演論文集(CD-ROM)   23rd ( 0 ) A212  2018

     View Summary

    In order to investigate heat transfer deterioration due to salt precipitation, a pool boiling experiment was conducted with a full-height 5×5 rod-bundle of BWR simulated fuel in sea water. The temperature on the center rod surface at the top spacer rose rapidly, since the flow area inside the top spacer was filled with the precipitated salt. Dryout occurred below the spacer, which results in the temperature escalation of the heater surface. On the other hand, the heater above the top spacer was cooled continuously by pool boiling.

    DOI CiNii J-GLOBAL

  • Study on Eutectic Melting Behavior of Control Rod Materials in Core Disruptive Accidents of Sodium-Cooled Fast Reactors: Project Overview and Progress by JFY2017

    山野秀将, 高井俊秀, 古川智弘, 斉藤淳一, 菊地晋, 江村優軌, 東英生, 福山博之, 西剛史, 太田弘道, LIU Xiaoxing, 守田幸路, 中村勤也, 太田宏一, 古谷正裕

    日本機械学会動力・エネルギー技術シンポジウム講演論文集(CD-ROM)   23rd ( 0 ) A224 - 427  2018

     View Summary

    It is necessary to simulate a eutectic melting reaction and relocation behavior of boron carbide (B4C) as a control rod material and stainless steel (SS) during a core disruptive accident in an advanced sodium-cooled fast reactor designed in Japan. On that account, a new project has been started to conduct eutectic melting experiments, thermophysical property measurement of the eutectic melt, and physical model development for the eutectic melting reaction. The eutectic experiments involve the visualization experiments, eutectic reaction rate experiments and material analyses. The thermophysical properties are measured in the range of liquid and solid states. The physical model is developed for a severe accident computer code based on the measured data of the eutectic reaction rate and the physical properties. This paper describes the project overview and progress by JFY2017.

    DOI CiNii J-GLOBAL

  • 蒸気爆発抑制材の高温溶融物に対する抑制効果

    古谷正裕, 新井崇洋

    日本伝熱シンポジウム講演論文集(CD-ROM)   55th   ROMBUNNO.G232  2018

    J-GLOBAL

  • 急速発熱5×5バンドル内における過渡沸騰と横断流動

    滝口広樹, 新井崇洋, 古谷正裕, 白川健悦

    日本伝熱シンポジウム講演論文集(CD-ROM)   55th   ROMBUNNO.G211  2018

    J-GLOBAL

  • Observation of Mutual Diffusion Behavior of Hydrazine Hydrate by Near-infrared Spectroscopic Technique

    滝口広樹, 古谷正裕, 新井崇洋

    混相流シンポジウム講演論文集(Web)   2018   WEB ONLY  2018

    J-GLOBAL

  • 高温溶融ニッケルと錫に対する蒸気爆発抑制材の有効性

    古谷正裕, 新井崇洋

    日本原子力学会春の年会予稿集(CD-ROM)   2018   ROMBUNNO.3J04  2018

    J-GLOBAL

  • 沸騰二相流におけるボイド率および流体温度の同時計測手法の開発

    滝口広樹, 古谷正裕, 新井崇洋

    日本原子力学会春の年会予稿集(CD-ROM)   2018   ROMBUNNO.2B21  2018

    J-GLOBAL

  • TRACEコードを用いた福島第一原子力発電所3号機事故における格納容器の熱水力挙動の解析

    大川理一郎, 古谷正裕

    日本原子力学会春の年会予稿集(CD-ROM)   2018   ROMBUNNO.3K09  2018

    J-GLOBAL

  • Multi‐physicsモデリングによるEx‐Vessel溶融物挙動理解の深化(2)全体概要とMPS法によるSpreading解析の高度化

    山路哲史, 古谷正裕, 大石佑治, DUAN Guangtao

    日本原子力学会春の年会予稿集(CD-ROM)   2018   ROMBUNNO.3J08  2018

    J-GLOBAL

  • Multi‐physicsモデリングによるEx‐Vessel溶融物挙動理解の深化(4)全体概要とMPS法によるSpreading解析の高度化(2)

    山路哲史, 古谷正裕, 大石佑治, DUAN Guangtao

    日本原子力学会秋の大会予稿集(CD-ROM)   2018   ROMBUNNO.1H09  2018

    J-GLOBAL

  • 5×5発熱バンドル内沸騰二相流の流体温度とボイド率の同時計測

    滝口広樹, 古谷正裕, 新井崇洋

    日本原子力学会秋の大会予稿集(CD-ROM)   2018   ROMBUNNO.2I16  2018

    J-GLOBAL

  • Multi‐physicsモデリングによるEx‐Vessel溶融物挙動理解の深化(5)ペデスタル床複雑構造に拡がる溶融物の三次元流動

    古谷正裕, 山路哲史, 大石佑治

    日本原子力学会秋の大会予稿集(CD-ROM)   2018   ROMBUNNO.1H10  2018

    J-GLOBAL

  • TRACEコードを用いたBWRプラントにおける水素拡散挙動の解析

    大川理一郎, 古谷正裕

    日本原子力学会秋の大会予稿集(CD-ROM)   2018   ROMBUNNO.2I09  2018

    J-GLOBAL

  • 浅い水プール底における蒸気爆発のシミュレーション

    森山清史, 古谷正裕

    日本原子力学会秋の大会予稿集(CD-ROM)   2018   ROMBUNNO.2I21  2018

    J-GLOBAL

  • 燃料集合体内冷却水の気液二相流の挙動解明に向けた研究開発(6)大気圧二重管体系における強制対流サブクール沸騰場の二相流計測

    宇井淳, 滝口広樹, 古谷正裕, 新井崇洋, 白川健悦, 木藤和明

    日本原子力学会秋の大会予稿集(CD-ROM)   2018   ROMBUNNO.2I15  2018

    J-GLOBAL

  • 燃料露出過程の炉内流動評価(16)大気圧から7MPaまでの低流量条件における発熱管群体系のボイド率相関式

    新井崇洋, 古谷正裕, 滝口広樹, 西義久, 白川健悦

    日本原子力学会秋の大会予稿集(CD-ROM)   2018   ROMBUNNO.2I17  2018

    J-GLOBAL

  • Visual Observation of Three-Dimensional Flow Dynamics in Complex Structure and CMFD simulation

    FURUYA Masahiro, YAMAJI Akifumi, OHISHI Yuji

    The Proceedings of Mechanical Engineering Congress, Japan   2018 ( 0 ) S0530306  2018

     View Summary

    <p>A pedestal-floor structure of a nuclear-reactor containment-vessel was additive manufactured with two sump-pits and a doorway at a reduced scale of 1:100 and 1:50. A silicone-oil jet impinged and spread on the pedestal floor. Such three-dimensional flow was visualized from three cameras. A volume of fluid (VOF) simulation was successfully reproduced this three-dimensional spreading behavior, silicone-oil flow-thickness evolution and drainage weights.</p>

    DOI CiNii J-GLOBAL

  • Flashing-induced density wave oscillations in a boiling natural circulation system

    Masahiro Furuya

    Advances in Multiphase Flow and Heat Transfer   3   280 - 299  2018

     View Summary

    This chapter addresses characteristics of flashing-induced density wave oscillations on the basis of the experimental results in a boiling natural circulation system with an adiabatic chimney. Flashing is caused by the sudden increase of vapor generation due to the reduction in hydrostatic head, since saturation enthalpy changes with pressure. Flashing-induced density wave oscillations may, therefore, occur at low pressure. The oscillation period correlates well with the passing time of bubbles in the chimney section regardless of the system pressure, the heat flux, and the inlet subcooling. According to the stability map, the flow became stable below a certain heat flux regardless of the channel inlet subcooling. The stable region enlarged with increasing system pressure. Therefore, the stability margin becomes larger by pressurizing the loop sufficiently before heating.

    DOI

  • Severe accident analysis with spatial discretized model by MAAP - (2) parametric study on Fukushima-daiichi unit-3-

    Yoshihisa Nishi, Kenichi Kanda, Kazuma Abe, Satoshi Nishimura, Koichi Nakamura, Masahiro Furuya, Atsushi Ui

    International Conference on Nuclear Engineering, Proceedings, ICONE   6B  2018

     View Summary

    Accident analyses of unit 3 in the Fukushima-Daiichi nuclear power station were performed with MAAP (Modular Accident Analysis Program) 5.03. In the analysis, the operations of RCIC, HPCI, SRV were simulated and assuming the operation of alternative water injection to reproduce the measured pressure and temperature. As a result of parametric evaluation, analysis results consistent with measured values are obtained. In this paper, the results of the sensitivity analysis are reported. Simultaneously, attempts were made to analyze the transient and deposition amount of fission product (FP) in the reactor building (R/B) and a model dividing each room of R/B in detail was created to confirm the FP behavior. Based on these analyses, the deposition amount of the primary containment vessel (PCV) (drywell (D/W) and wetwell (W/W)) and R/B could be estimated using detailed model of the R/B dividing into nodes in the MAAP simulation.

    DOI

  • Transient boiling and cross flow in 5×5 ROD bundle with rapid heating

    Hiroki Takiguchi, Masahiro Furuya, Takahiro Arai, Kenetsu Shirakawa

    International Conference on Nuclear Engineering, Proceedings, ICONE   6A  2018

     View Summary

    Rapid thermal elevation in nuclear reactor is an important factor for nuclear safety. It is indispensable to develop a three-dimensional nuclear thermal transient analysis code and confirm its validity in order to accurately evaluate the effectiveness of the running nuclear safety measures when heating power of reactor core rapidly rises. However, the heat transfer characteristics such as reactivity feedback characteristics due to moderator density and the technical knowledge explaining the uncertainty are insufficient. In particular, the cross propagation behavior of vapor bubble (void) in cross section of fuel assembly is not grasped. This study evaluates the cross propagation void behavior in a simulated fuel assembly at time of rapid heat generation with a thermal hydraulic test loop including a 5×5 rod bundle having the heat generation profile in the flow cross sectional direction. In this paper, the branching heat output condition of transient cross propagation was investigated from visualization of high speed video camera and void fraction measurement by wire mesh sensor with the inlet flow rate 0.3m/s and the inlet coolant temperature 40oC, which are based on the transient safety analysis condition. In addition, we applied the particle imaging velocimetry (PIV) technique to measure liquid-phase velocity profile of the coolant in the transient cross flow and experimentally clarified the relationship with the cross flow.

    DOI

  • Severe accident analysis with spatial discretized model by MAAP - (1) parametric study on Fukushima-daiichi Unit-2-

    Kenichi Kanda, Yoshihisa Nishi, Kazuma Abe, Satoshi Nishimura, Koichi Nakamura, Masahiro Furuya, Atsushi Ui

    International Conference on Nuclear Engineering, Proceedings, ICONE   6B  2018

     View Summary

    Accident analyses of the Fukushima-Daiichi unit-2 nuclear power plant were performed with MAAP (Modular Accident Analysis Program) version 5.03. We assumed RCIC, SRV operation and alternative water injection in order to reproduce the measured pressure and temperature values in RPV and PCV. From parametric studies, it was found that the analysis results were in good agreement with the measured data. In this paper, the results of the parametric studies are reported. Furthermore, spatial discretization of compartments (such as rooms in the reactor building, etc.) into small parts successfully demonstrated the transient distribution and deposition of fission products (FPs) across the rooms. Such special discretization is particularly important for the forensic investigation of severe accidents and the deposited amount in the R/B might be estimated by using this detailed model.

    DOI

  • Temperature Profile Measurement in Simulated Fuel Assembly Structure with Wire-Mesh Technology

    Hiroki Takiguchi, Masahiro Furuya, Takahiro Arai

    Science and Technology of Nuclear Installations   2018  2018

     View Summary

    When light water reactor (LWR) is subject to a cold shutdown, it needs to be cooled with pure water or seawater to prevent the core melting. To precisely evaluate the cooling characteristics in the fuel assembly, a measurement method capable of installing to the fuel assembly structure and determining the temperature distribution with high temporal resolution, high spatial resolution, and in multidimension is required. Furthermore, it is more practical if applicable to a pressure range up to the rated pressure 16 MPa of a pressurized water reactor (PWR). In this study, we applied the principle of the wire-mesh sensor technology used in the void fraction measurement to the temperature measurement and developed a simulated fuel assembly (bundle) test loop with installing the temperature profile sensors. To investigate the measurement performance in the bundle test section, it was confirmed that a predetermined temperature calibration line with respect to time-average impedance was calculated and became a function of temperature. To evaluate the followability of measurement in a transient temperature change process, we fabricated a 16 × 16 wire-mesh sensor device and measured the hot-water jet-mixing process into the cold-water pool in real time and calculated the temperature profile from the temperature calibration line obtained in advance from each measurement point. In addition, the sensors applied to three-dimensional temperature distribution measurement of a complex flow field in the bundle structure. The axial and cross-sectional profiles of temperature were quantified in the forced flow field with nonboiling when the 5×5 bundle was heated by energization.

    DOI

  • Prescription for use of vapor explosion retardant into salt water

    Masahiro Furuya, Takahiro Arai

    International Heat Transfer Conference   2018-August   2439 - 2446  2018

     View Summary

    Vapor explosion has been causing disasters in many industries such as metalwork and paper industries. One of the countermeasures is retardant additives into water to stabilize the vapor film which separates two liquids. A spontaneous vapor explosion of a molten tin jet at 700 oC was suppressed with only 0.03 wt% polyethylene glycol aqueous solution for molecular weight of 4×106 g/mol. This is because the solute deposited near the vapor-liquid interface due to the cloud-point phenomenon, that stabilizes vapor film and prevents the solution from mixing finely. Salts are known additives to act as vapor-explosion promoter. Increasing salt concentration requires denser PEG solution to suppress vapor explosion: e.g. 0.03 wt% PEG for water, while 0.07 wt% PEG for sea water and 3 wt% sodium chloride aqueous solution. These salt solutions were selected for practical relevance in industrial disasters. A solid sphere quenching experiment indicates that this threshold concentration of PEG can be determined by the quenching temperature of the solid sphere: the contact temperature of the solid sphere with solution must be sufficiently low (e.g. spontaneous-bubble nucleation temperature of the solution) to suppress the vapor explosion.

    DOI

  • Topics on Boiling: From Fundamentals to Applications

    Tomohiko Yamaguchi, Gyoko Nagayama, Takaharu Tsuruta, Yuyan Jiang, Shigeo Maruyama, Kunito Okuyama, Yasushi Saito, Koichi Suzuki, Manabu Tange, Ichiro Ueno, Tomohiro Osawa, Yasusuke Hattori, Takahito Saiki, Jun Ando, Kazuna Horiuchi, Yusuke Koiwa, Hitoshi Asano, Kazuhisa Yuki, Yasuyuki Takata, Yoshiyuki Abe, Raffaele Savino, Serizawa Yoshihiro, Hidetoshi Ohkubo, Yutaka Abe, Shinpei Saitho, Masahiro Furuya, Satoru Momoki, Chieko Kondou, Shigeru Koyama, Mamoru Ozawa

    Boiling: Research and Advances     443 - 777  2017.06

     View Summary

    This chapter deals with the various topics on boiling with regard to aspects of the fundamentals and applications to introduce the development of each author's research in recent decades. The first four sections investigate the physics of boiling as phase change phenomena, including thermodynamic phase equilibrium state (Section 6.1), molecular dynamics of phase change (Section 6.2), computational analysis of boiling in micro-nano scale (Section 6.3), and transient boiling under rapid heating (Section 6.4). Section 6.5 deals with two-phase distribution measurement using neuron radiography. The following three sections then examine a specific boiling regime during highly subcooled boiling, called microbubble emission boiling (MEB). Each section treats the overall characteristics of MEB (Section 6.6), the occurrence conditions of MEB (Section 6.7), and vapor collapses in subcooled liquid related to MEB (Section 6.8). The next four sections are devoted to heat transfer augmentation with various techniques: thermal spray coating (Section 6.9), porous media (Section 6.10), patterned wettability refinement (Section 6.11), and self-rewetting fluid (Section 6.12). The last seven sections describe topics on applications of boiling. Sections 6.13 and 6.14 introduce boiling research in steel industries. Sections 6.15 and 6.16 explore vapor explosion. Boiling of refrigerant is discussed with heat pump systems in Section 6.17 and with automobile air conditioners in Section 6.18. Boiling related to emergency cooling core systems is considered in Section 6.19.

    DOI

  • Numerical analysis of the melt behavior in a fuel support piece of the BWR by MPS

    Ronghua Chen, Lie Chen, Kailun Guo, Akifumi Yamaji, Masahiro Furuya, Wenxi Tian, G. H. Su, Suizheng Qiu

    Annals of Nuclear Energy   102   422 - 439  2017.04

     View Summary

    The fuel support piece in a boiling water reactor (BWR) is used to brace fuel assemblies. The channel within the fuel support piece is determined to be a potential corium relocation path from the core region to the lower head during the severe accident of BWR. In the present study, the improved ∗∗∗Moving Particle Semi-implicit (MPS) method was adopted to simulate the flow and solidification behavior of the melt in a fuel support piece. The MPS method was first validated against the Pb-Bi plate ablation test that was performed by CRIEPI. The predicted ablation mass of the plate agreed well with the experimental results. Then the flowing and freezing behaviors of molten stainless steel (SS) and zircaloy in the fuel support piece were simulated by MPS method with a three dimensional particle configuration, respectively. In this study, the flow and solidification behavior of SS was simulated first. After all the SS passed through the channel, the flowing behavior of Zr in the fuel support piece was simulated. The simulation results indicated that the crust layer formed on the inner surface of the fuel support piece during the melt discharging process. The fuel support piece was plugged by the solidified zircaloy particles in the lower initial temperature case. The fuel support piece kept intact in all the calculation that were performed under the assumed order of melt injection. The present results could help to reveal the progression of a BWR severe accident.

    DOI

  • Development of Integrated Analysis Tool for Filtered-Containment Venting System

    金井大造, 古谷正裕

    日本機械学会動力・エネルギー技術シンポジウム講演論文集(CD-ROM)   22nd ( 0 ) A113  2017

     View Summary

    <p>A filtered containment venting system (FCVS) reduces containment pressure and amount of the radioactive release to the environment during a severe accident of nuclear power plant. CRIEPI has developed a FCVS analysis tool which evaluate removal performance of aerosol, iodine and organic iodine in each process of FCVS as decontamination factor as a function of the FCVS system parameters. The paper addresses the integrated analysis tool of FCVS simulator based on the experimental database.</p>

    DOI CiNii J-GLOBAL

  • 先進的レベル2PRA評価手法の開発(4)フィルタベント総合性能評価ツールの開発

    金井大造, 古谷正裕, 中村康一, 遠藤寛, 西義久

    日本原子力学会秋の大会予稿集(CD-ROM)   2017   ROMBUNNO.1B16  2017

    J-GLOBAL

  • 燃料集合体内冷却水の気液二相流の挙動解明に向けた研究開発(1)全体研究計画

    木藤和明, 上遠野健一, 宇井淳, 古谷正裕

    日本原子力学会秋の大会予稿集(CD-ROM)   2017   ROMBUNNO.1E01  2017

    J-GLOBAL

  • 軽水炉のシビアアクシデント下の海水・ホウ酸注入時の影響に関する試験(7)5×5バンドル流路内での海水沸騰濃縮と析出塩による閉塞

    古谷正裕

    日本原子力学会秋の大会予稿集(CD-ROM)   2017   ROMBUNNO.2F10  2017

    J-GLOBAL

  • RIA時の沸騰遷移に関する研究(低温時RIA)(2)大気圧下における急速発熱時の3×3バンドル過渡限界熱流束

    新井崇洋, 古谷正裕, 白川健悦, 渡辺瞬, 滝口広樹, 佐合優一, 原田健一

    日本原子力学会秋の大会予稿集(CD-ROM)   2017   ROMBUNNO.1C16  2017

    J-GLOBAL

  • Multi‐physicsモデリングによるEx‐Vessel溶融物挙動理解の深化(1)全体計画

    山路哲史, 古谷正裕, 大石佑治, DUAN Guangtao

    日本原子力学会秋の大会予稿集(CD-ROM)   2017   ROMBUNNO.3F05  2017

    J-GLOBAL

  • TRACEコードを用いたLOCA模擬試験解析における沸騰伝熱モデルの改良効果

    大川理一郎, 古谷正裕

    日本原子力学会秋の大会予稿集(CD-ROM)   2017   ROMBUNNO.3E13  2017

    J-GLOBAL

  • 燃料集合体内冷却水の気液二相流の挙動解明に向けた研究開発(3)大気圧サブクール沸騰試験

    宇井淳, 古谷正裕, 滝口広樹, 白川健悦, 新井崇洋

    日本原子力学会秋の大会予稿集(CD-ROM)   2017   ROMBUNNO.1E03  2017

    J-GLOBAL

  • ナトリウム冷却高速炉の炉心損傷事故時の制御棒材の共晶溶融挙動に関する研究(1)プロジェクト全体概要

    山野秀将, 高井俊秀, 古川智弘, 江村優軌, 倉田正輝, 東英生, 福山博之, 西剛史, 太田弘道, LIU Xiaoxing, 守田幸路, 古谷正裕

    日本原子力学会秋の大会予稿集(CD-ROM)   2017   ROMBUNNO.2P01  2017

    J-GLOBAL

  • RIA時の沸騰遷移に関する研究(低温時RIA)(1)全体計画

    佐合優一, 原田健一, 古谷正裕, 金子浩久, 宮地孝政

    日本原子力学会秋の大会予稿集(CD-ROM)   2017   ROMBUNNO.1C15  2017

    J-GLOBAL

  • Dose evaluation in the BWR reactor building with MAAP-DOSE

    Kazuma Abe, Kenichi Kanda, Satoshi Nishimura, Masahiro Furuya, Yoshihisa Nishi

    2017 International Congress on Advances in Nuclear Power Plants, ICAPP 2017 - A New Paradigm in Nuclear Power Safety, Proceedings    2017

     View Summary

    The dose assessment of the main control room has been required in order to show its residence under the present Japanese new regulatory standards. Furthermore, to assess the in-plant dose such as reactor building where a lot of operations are performed is considered to contribute to the evaluation of accessibility in the training scenario for future voluntary safety improvements. The severe accident analysis code MAAP (Modular Accident Analysis Program) has "DOSE" module that dose calculation and accident progress analysis can be implemented simultaneously. It is expected to contribute in the evaluation of more detailed accident scenario because temporal trend of the dose corresponding to FP (Fission Product) amount can be obtained for each accident sequence. This study intends to obtain the knowledge to handle the DOSE module of MAAP by performing a dose calculation of the reactor building of a BWR plant. The FP migration behavior and the dose rate in the reactor building and the main control room were evaluated against two accident sequences, the only SBO and the superimposition of SBO and LOCA, by using BWR4 Mark-I containment model which is distributed as a sample of MAAP. The dose calculation result reflecting the accident progress was obtained from the result such as the LOCA and SBO sequence showed relatively high doses because of the early pressure vessel damage. The interphase between the dose and the FP mass was good for each accident sequence, especially for the noble gas. The SBO sequence showed the overall higher dose and the FP amount flowing into the reactor building was also larger. Noble gases, CsOH and CsI were confirmed as FP nuclides with a high degree of influence on dose, and the contribution of noble gas with large release amount was particularly higher. Since the noble gas is released to the external environment along with the time, the contribution of CsOH and CsI became higher in the latter stage of the scenario. In addition, as one of the factors that have an impact on the dose rate in the building during a severe accident, drywell spray operation is considered. The effect of spray on dose can be considered as below, the capture of FP, cooling of the debris, and the shielding effect of the water. The effect of greatly reducing the dose in the SBO scenario was confirmed. In addition, it was confirmed that the cooling debris by spray had a large influence on the dose.

  • Development of temperature profile sensor at high temporal and spatial resolution

    Hiroki Takiguchi, Masahiro Furuya, Takahiro Arai

    2017 International Congress on Advances in Nuclear Power Plants, ICAPP 2017 - A New Paradigm in Nuclear Power Safety, Proceedings    2017

     View Summary

    In order to quantify thermo-physical flow field for the industrial applications such as nuclear and chemical reactors, high temporal and spatial measurements for temperature, pressure, phase velocity, viscosity and so on are required to validate computational fluid dynamics (CFD) and subchannel analyses. The paper proposes a novel temperature profile sensor, which can acquire temperature distribution in water at high temporal (a millisecond) and spatial (millimeter) resolutions. The devised sensor acquires electric conductance between transmitter and receiver wires, which is a function of temperature. The sensor comprise wire mesh structure for multipoint and simultaneous temperature measurement in water, which indicated that three-dimensional temperature distribution can be detected in flexible resolutions. For the demonstration of the principle, temperature profile in water was estimated according to predetermined temperature calibration line against time-averaged impedance. The 16×16 grid sensor visualized fast and multi-dimensional mixing process of a hot water jet into a cold water pool.

  • Surface modification technology of fuel cladding, fresh green to mitigate corrosion and hydrogen-pickup in high-temperature steam environment

    Masahiro Furuya

    2017 International Congress on Advances in Nuclear Power Plants, ICAPP 2017 - A New Paradigm in Nuclear Power Safety, Proceedings    2017

     View Summary

    Fuel claddings would be exposed to a hightemperature steam-flow after a boiled-up sequence of a severe accident of light water reactors as well as spent fuel pools. Zirconium alloys are common cladding material for light water reactors. All the zirconium alloys generate additional heat by the exothermic reaction of oxidation in high-temperature steam. Subsequent temperature rise accelerates the rate of oxidation. The countermeasure to suppress oxidation of zirconium alloy is key to mitigate severe accidents. We invent a surface modification technology of zirconium alloy, Fresh Green, to oxidize and to carbonize a zirconium-alloy surface in the same process. XRD and XPS analyses indicated that a modified layer on zirconium surface is carbon-doped zirconium dioxide, in which some of oxygen atoms in monoclinic zirconium dioxide are replaced by carbon. Experiments are conducted for three representative conditions using an autoclave at relatively high temperature: uniform corrosion at 400 degrees Celsius for 336 hours, nodular corrosion at 500 degrees Celsius for 24 hours, supercritical water corrosion (at 400 degrees Celsius, 24 hours). The Fresh Green surface modification reduces both the oxidation rate and hydrogen-pickup less than a half of that with untreated base material. This is because the Fresh Green layer is closely packed and adhered intimately to the base material. Hydrides were found in the base material without treatment, though they were scarcely observed in the Fresh Green treated specimen. Since the operating temperature of the Fresh Green process is lower than the final annealing temperature in a manufacturing process, the process does not affect the material property and crystal structure. The Fresh Green process, which can be operated at slightly higher pressure than ambient, can be introduced in the conventional manufacturing process without complicity.

  • Spray cooling performance during spent fuel pool accident with MAAP code

    Kenichi Kanda, Satoshi Nishimura, Masaaki Satake, Kazuma Abe, Masahiro Furuya, Yoshihisa Nishi

    2017 International Congress on Advances in Nuclear Power Plants, ICAPP 2017 - A New Paradigm in Nuclear Power Safety, Proceedings    2017

     View Summary

    The effectiveness evaluation of the safety measures under severe accidents is demanded; not only for nuclear reactors but also spent fuel pools (SFPs) after the Fukushima-Daiichi nuclear power plant accident which hit the Tokyo Electric Power Company on March 11, 2011. This paper addresses models of the spent fuel pool with accident analysis code, MAAP and a sensitivity analysis of the spray cooling parameters following a loss of coolant accident (LOCA) to evaluate the fuel coolability. In this study, spray water is injected from the upper space of SFP onto fuel racks using an SFP model attached to a MAAP code. The instantaneous LOCA in SFP was hypothetically assumed, which meant the spent fuels were directly exposed to the atmosphere from the onset of the event. Furthermore, the sensitivity of the spray cooling parameters was analyzed to investigate the influence on how the instantaneous LOCA sequence in the SFP will progress. The spray cooling model of the MAAP code includes a range of input parameters such as the spray-flow rate, spray water temperature, and so on. For example, when the spray-flow rate suffices to cool the fuel, the maximum temperature of the fuel cladding decreases after the onset of spray cooling with a particular time delay. Furthermore, increasing the spray-flow rate helps reduce any time delay and accelerate cooling. Conversely, when the spray-flow rate is insufficient for cooling, the fuel cladding temperature continues to rise and ends up damaging the fuel. The spray-flow rate significantly impacts on the cooling of the fuel accordingly. The MAAP code demonstrates a time trace of the maximum cladding temperature in terms of spray cooling parameters. In addition, certain spray cooling parameters may impact on the behavior of fission products (FP). For example, particles drifting in the atmosphere in aerosol form can be easily adsorbed by spray droplets when the spray-flow rate increases, which further eliminates FP. We confirmed the impact on FP removal of the spray.

  • Applicability of steam-explosion retardant for molten nickel and tin at high-temperature

    Masahiro Furuya, Takahiro Arai

    17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017   2017-September  2017

     View Summary

    Steam explosion has been a potential threat to containment boundary during severe accident in light water reactors. Polyethylene glycol (PEG) aqueous solution is an effective triggering retardant for both spontaneous and external-triggered steam explosions. In order to investigate steam explosion controllability at high temperature and for high meting-point metal, experiments are extended for molten nickel and tin up to 1800oC. Steam explosion was suppressed for both molten tin and nickel in a 0.1 wt% PEG (molecular weight of 4×106 g/mol) aqueous solution. The stability of PEG was investigated with gel permeation chromatography, ion chromatography, and high-performance liquid chromatography. The results indicated that PEG may be precipitated in the liquid side without producing byproducts and suppress steam explosions even for the molten jet at 1800 oC.

  • Accident progression analyses of a BWR with filtered containment venting system during long term severe accident conditions

    Atsushi Ui, Masahiro Furuya, Taizo Kanai

    17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017   2017-September  2017

     View Summary

    It is expected that filtered containment venting system (FCVS) would play an important role to reduce the amount of fission products (FPs) released to the environment during severe accident conditions. Experimental studies for FCVS have been conducted in CRIEPI, and decontamination factors (DFs) regarding internal impaction at nozzles, Venturi’s scrubbers, metal fibrous filters, and Zeolite filters have been measured and formulated, respectively. The DF models were implemented into the MELCOR input data assuming a typical advanced boiling water reactor (ABWR). Several accident progression analyses, such as large break loss-of-coolant accident (LOCA), long term station blackout (LTSBO), and loss-of-all power supply, were conducted for the BWR with a FCVS that consists of scrubbers, metal and Zeolite filters connected to its suppression pool and dry well. The result indicated that as the water level of the FCVS was decreased, efficiency of scrubber was decreased, and the total DF was governed by metal fibrous filter. After most of the water in FCVS was evaporated, FCVS temperature increased, and FPs once captured by the FCVS and deposited on wall inside of the system were transported the environment. Time-depended release ratios of Cs, CsI, Te, Sr to the environment were compared under condition with FCVS and that without FCVS. The results obtained effectiveness of FCVS for each FP. The results also gave some insights for implementation of FCVS and severe accident management for BWRs.

  • Temperature-insensitive solution for accurate conductance measurement

    Hiroki Takiguchi, Masahiro Furuya, Takahiro Arai

    International Conference on Fluid Flow, Heat and Mass Transfer    2017

     View Summary

    In order to design boilers, evaporators and condensers, a phase distribution is one of the most important factor to be determined. Although electrodes which are immersed into a coolant media give useful information such as void fraction and phasic velocity, one may suffer temperature variation which affect electric conductance of the coolant. Potential gradient and electromagnetic wave which are generated in a test liquid due to energization, give critical measuring accuracy in the form of electric noise. If the test liquid has temperature dependency of electric conductivity, time-displacement of liquid temperature during the transient boiling phenomenon gives additional uncertainty to the measured value. Figure 1 illustrates experimental apparatus, which consists of a pre-heater, test section, separator, condenser and circulation pump. A heater pipe is inserted in the test section and heated by direct current. Wire-Mesh Sensors (WMSs) were inserted near by the heated section. The experiment was conducted in non-boiling (no bubbles in a solution) condition to quantify temperature dependence of each solute. The test fluids are dilute aqueous solutions, whose solutes are vitriolic solutes (K2SO4 and Na2SO4), nitric solutes (KNO3 and NaNO3) and fluorescence (Rhodamine-B and Uranine) which is used in PIV, LDV. The experimental procedures are tabulated as follows. 1. Calibration using ion-exchanged water - rest potential of water was measured at intervals of 10° from 30° to 90°. The value became larger in high temperature condition. - 2. Setup of solution concentration (Table 1) - Gain value of detector was set 1, and the concentration of each solute was calibrated that the electric potential of its solution is equal to one of water in the case of 90°. - 3. Measurement of temperature dependency of electric potential (dV/dT, V: Voltage, T: Temperature) and SN ratio 4. Measurement of electric conductivity, pH and dissolved oxygen level (not mention in this extended abstract) Results and discussion were given as follows. Time-series data of electric potentials of water and Rhodamine-B solution were indicated in Figure2. As shown in these trends, these all data were normalized with time-averaged one of 90°. It was confirmed that amplification method of electric potential with solute concentration predominates that one method with detector gain in order to improve SN ration of the signal. Temperature sensitivity index (dV/dT) and SN ratio were illustrated in Figure 3. It shows the normalized dV/dT data when one of water is 100. It was clarified that the solutes except Uranine have the temperature dependency-reduction effect to 70 % of one water. Furthermore, SN ratio of all solutions were increased about 3 times than one of water. Finally, thermos-physical properties of test liquid solutions were illustrated in Figure 4 and 5, respectively. These graphs were normalized with the values at 30° each other. Fig. 4 indicates that addition of solute gives reduction effect of temperature change of electric conductivity. Especially in the case of Rhodamine-B, as shown in Fig. 5, its temperature dependence was indicated equal tendency with one of water because this fluorescence is not ionized. As the measurement of void fraction, reducing temperature dependency (dV/dT) without changing thermos-physical properties of water is important in order to avoid unnecessarily mixing the sensitive properties of electric conductivity. In conclusion, it was identified that Rhodamine-B has the following effects, reduction of temperature dependency of electric potential, improvement of SN ratio without changing water properties (pH, DO).

    DOI

  • Development of an aerosol decontamination factor evaluation method using an aerosol spectrometer

    Taizo Kanai, Masahiro Furuya, Takahiro Arai, Yoshihisa Nishi

    Nuclear Engineering and Design   303   58 - 67  2016.07

     View Summary

    During a severe nuclear power plant accident, the release of fission products into containment and an increase in containment pressure are assumed to be possible. When the containment is damaged by excess pressure or temperature, radioactive materials are released. Pressure suppression pools, containment spray systems and a filtered containment venting system (FCVS) reduce containment pressure and reduce the radioactive release into the environment. These devices remove radioactive materials via various mechanisms. Pressure suppression pools remove radioactive materials by pool scrubbing. Spray systems remove radioactive materials by droplet-aerosol interaction. FCVS, which is installed in the exhaust system, comprises multi-scrubbers (venturi-scrubber, pool scrubbing, static mixer, metal-fiber filter and molecular sieve). For the particulate radioactive materials, its size affects the removal performance and a number of studies have been performed on the removal effect of radioactive materials. This study has developed a new means of evaluating aerosol removal efficiency. The aerosol number density of each effective diameter (light scattering equivalent diameter) is measured using an optical method, while the decontamination factor (DF) of each effective diameter is evaluated by the inlet outlet number density ratio. While the applicable scope is limited to several conditions (geometry of test section: inner diameter 500 mm × height 8.0 m, nozzle shape and air-water ambient pressure conditions), this study has developed a numerical model which defines aerosol DF as a function of aerosol diameter (d) and submergences (x).

    DOI

  • 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics

    Journal of the Atomic Energy Society of Japan   58 ( 2 ) 127 - 127  2016

    DOI CiNii

  • スクラビング時の蒸気凝縮効果と液位変動

    金井大造, 古谷正裕, 新井崇洋, 西義久

    日本伝熱シンポジウム講演論文集(CD-ROM)   53rd   ROMBUNNO.J214  2016

    J-GLOBAL

  • 高時間・空間分解能温度分布センサーの開発

    滝口広樹, 新井崇洋, 古谷正裕

    日本伝熱シンポジウム講演論文集(CD-ROM)   53rd   ROMBUNNO.K134  2016

    J-GLOBAL

  • ワイヤメッシュセンサを用いた気泡挙動計測のための手法開発

    金井大造, 古谷正裕, 新井崇洋, 白川健悦

    混相流シンポジウム講演論文集(CD-ROM)   2016   ROMBUNNO.C115  2016

    J-GLOBAL

  • 停滞水条件下における垂直円管及びバンドル内の二相水位変動

    新井崇洋, 古谷正裕, 金井大造, 滝口広樹, 白川健悦

    混相流シンポジウム講演論文集(CD-ROM)   2016   ROMBUNNO.C114  2016

    J-GLOBAL

  • ワイヤメッシュセンサを用いたボイド率計測における蛍光緩衝剤の添加効果

    滝口広樹, 古谷正裕, 新井崇洋, 渡辺瞬

    混相流シンポジウム講演論文集(CD-ROM)   2016   ROMBUNNO.E213  2016

    J-GLOBAL

  • 垂直矩形管内二相流における液相及び気相速度分布計測手法の開発

    滝口広樹, 古谷正裕, 新井崇洋, 金井大造

    日本原子力学会春の年会予稿集(CD-ROM)   2016   ROMBUNNO.3C11  2016

    J-GLOBAL

  • MAAP5.03による福島第一原子力発電所建屋内FP分布の試計算(2)3号機計算結果

    阿部数馬, 神田憲一, 西義久, 西村聡, 中村康一, 古谷正裕, 宇井淳

    日本原子力学会春の年会予稿集(CD-ROM)   2016   ROMBUNNO.1D15  2016

    J-GLOBAL

  • 燃料露出過程の炉内流動評価(11)事故時低流量下での発熱バンドル流路内の沸騰二相流発達に及ぼす圧力影響

    新井崇洋, 古谷正裕, 白川健悦, 西義久

    日本原子力学会春の年会予稿集(CD-ROM)   2016   ROMBUNNO.2C21  2016

    J-GLOBAL

  • 軽水炉のシビアアクシデント下の海水・ホウ酸注入時の影響に関する試験(4)5×5バンドル流路で沸騰濃縮された海水とホウ酸水との混合液の塩析出挙動

    古谷正裕

    日本原子力学会春の年会予稿集(CD-ROM)   2016   ROMBUNNO.1C12  2016

    J-GLOBAL

  • MAAP5.03による福島第一原子力発電所建屋内FP分布の試計算(1)2号計算結果

    神田憲一, 阿部数馬, 西義久, 西村聡, 中村康一, 古谷正裕, 宇井淳

    日本原子力学会春の年会予稿集(CD-ROM)   2016   ROMBUNNO.1D14  2016

    J-GLOBAL

  • フィルタベントシステムの運用高度化(6)エアロゾル除染性能の数理モデル化

    金井大造, 古谷正裕, 新井崇洋, 西義久

    日本原子力学会春の年会予稿集(CD-ROM)   2016   ROMBUNNO.1D19  2016

    J-GLOBAL

  • MAAPによる使用済燃料プール事故解析に関する評価(1)冷却水喪失時(LOCA)時のスプレイ冷却特性評価

    神田憲一, 西村聡, 佐竹正哲, 阿部数馬, 古谷正裕, 西義久

    日本原子力学会秋の大会予稿集(CD-ROM)   2016   ROMBUNNO.2K08  2016

    J-GLOBAL

  • 軽水炉のシビアアクシデント下の海水・ホウ酸注入時の影響に関する試験(6)5×5バンドル流路内での塩水プール沸騰時のボイド率分布

    古谷正裕, 滝口広樹, 新井崇洋, 白川健悦

    日本原子力学会秋の大会予稿集(CD-ROM)   2016   ROMBUNNO.1K12  2016

    J-GLOBAL

  • 燃料露出過程の炉内流動評価(13)高エネルギX線CTを用いた高圧・低流量条件下での5×5発熱バンドル内ボイド率分布計測

    新井崇洋, 古谷正裕, 白川健悦, 西義久

    日本原子力学会秋の大会予稿集(CD-ROM)   2016   ROMBUNNO.1K14  2016

    J-GLOBAL

  • MAAP‐DOSEを用いたBWR原子炉建屋内線量評価

    阿部数馬, 神田憲一, 西村聡, 古谷正裕, 西義久

    日本原子力学会秋の大会予稿集(CD-ROM)   2016   ROMBUNNO.3K06  2016

    J-GLOBAL

  • フィルタベントシステムの運用高度化(7)液性とヨウ素除染性能の関係

    金井大造, 古谷正裕, 新井崇洋, 西義久

    日本原子力学会秋の大会予稿集(CD-ROM)   2016   ROMBUNNO.2E07  2016

    J-GLOBAL

  • Iodien decontamination performance in filtered venting systems

    KANAI Taizo, FURUYA Masahiro, ARAI Takahiro, NISHI Yoshihisa

    The Proceedings of the National Symposium on Power and Energy Systems   21st ( 0 ) B125  2016

     View Summary

    <p>During a severe nuclear power plant accident, increasing of containment pressure are assumed to be possible. When the containment is damaged by excess pressure or temperature, radioactive materials are released from the containment. A filtered containment venting system (FCVS) reduces containment pressure and the radioactive release into the environment. The FCVS removes radioactive materials via various process. A pool scrubbing is an important process and flow dynamics (inertial collision, submergence, bubble diameter) and chemical conditions (pH, solubility) affects the radioactive materials removal performance. For an iodine (I2), chemical conditions are important factor in this process. This study has evaluated the relationships between the iodine decontamination performance and the chemical conditions in the pool scrubbing process.</p>

    DOI CiNii J-GLOBAL

  • Additive manufacturing of ceramic porous media with heating and temperature-measurement devices

    FURUYA Masahiro

    The Proceedings of Mechanical Engineering Congress, Japan   2016 ( 0 ) S0440106  2016

     View Summary

    <p>In order to investigate boiling heat transfer of a porous body with a crack, alumina porous body was additivemanufactured on the basis of powder bed fusion method. Alumina was selected for its heat resistant property and similarity of a crust of molten core. The manufactured porous body with 30% porosity has a slit-shape crack. Additive manufacturing became possible with the optimized operating parameters, powder size distribution, and the specific sintering assistant. A metal foil was attached on the back side of porous body and heated by applied direct electric current. The temperature profile of the metal foil was estimated on the basis of acquired electric-resistance distribution. As a results of pool boiling experiment in a salt-water pool, the boiling heat-transfer performance was acquired as a function of heat-transfer levels.</p>

    DOI CiNii J-GLOBAL

  • Aerosol decontamination performance in filtered venting systems

    KANAI Taizo, FURUYA Masahiro, ARAI Takahiro, NISHI Yoshihisa

    The Proceedings of Mechanical Engineering Congress, Japan   2016 ( 0 ) S0820102  2016

     View Summary

    <p>During a severe nuclear power plant accident, increasing of containment pressure are assumed to be possible. When the containment is damaged by excess pressure or temperature, radioactive materials are released from the containment. A filtered containment venting system (FCVS) reduces containment pressure and the radioactive release into the environment. The FCVS removes radioactive materials via various process. In the pool scrubbing process, flow dynamics (inertial impact, submergence, bubble diameter) and chemical conditions (pH, solubility) affect the removal performance of the radioactive materials. This study has forcued on the aerosol removal effect by an inertial impact.</p>

    DOI CiNii J-GLOBAL

  • Time-resolved measurement of subchannel void distribution for boiling two-phase flow in 5x5 rod bundle at high-pressure and high-temperature

    Takahiro Arai, Masahiro Furuya, Kenetsu Shirakawa, Yoshihisa Nishi

    International Topical Meeting on Advances in Thermal Hydraulics 2016, ATH 2016     193 - 200  2016

     View Summary

    A subchannel void sensor (SCVS) was improved to acquire time-resolve boiling two-phase flow in a 5-by-5 heated rod bundle at high-pressure and high-temperature. The SCVS consists of 6-by-6 wire electrodes and 5-by-5 rod electrodes. The SCVS can acquire a time series data of local void fraction at 32 points of central subchannel region in addition to 100 points of rod surface regions between rods. The devised sensors are installed in the heated rod bundle at eight height levels to acquire two-phase flow dynamics and its development. The axial and radial power profile of the heated rod bundle are uniform, and eight pairs of sheath thermocouples are embedded on the heated rod to monitor its surface temperature distribution. The paper addresses the axial and cross-sectional distributions of void fraction at various pressure and rod bundle power conditions, which simulates an accidental condition in reactor core and in a spent fuel pool of a boiling water reactor (BWR). The experimental data are suitable to validate the computational multi-fluid dynamics (CMFD) and subchannel codes.

  • ステンレス鋼と炭化ホウ素との共晶溶融自然対流の観察と三次元ラマン分光分析

    古谷正裕, 師岡愼一

    日本伝熱シンポジウム講演論文集(CD-ROM)   52nd   ROMBUNNO.F223  2015

    J-GLOBAL

  • 炭化ホウ素とステンレス鋼との共晶溶融移動観察と三次元ラマン分光測定

    古谷正裕

    日本原子力学会春の年会予稿集(CD-ROM)   2015   ROMBUNNO.H07  2015

    J-GLOBAL

  • フィルタベントシステムの運用高度化(5)ヨウ素除染性能

    金井大造, 古谷正裕, 新井崇洋, 西義久, 白川健悦, 田中伸幸

    日本原子力学会春の年会予稿集(CD-ROM)   2015   ROMBUNNO.H52  2015

    J-GLOBAL

  • 燃料露出過程の炉内流動評価(7)大気圧下でのバンドル内沸騰二相流動に及ぼす径方向出力分布の影響

    新井崇洋, 古谷正裕, 金井大造, 白川健悦, 西義久

    日本原子力学会春の年会予稿集(CD-ROM)   2015   ROMBUNNO.H22  2015

    J-GLOBAL

  • 燃料露出過程の炉内流動評価(9)高圧条件下での5×5発熱バンドル流路内沸騰二相流のボイド率分布計測

    新井崇洋, 古谷正裕, 白川健悦, 西義久

    日本原子力学会秋の大会予稿集(CD-ROM)   2015   ROMBUNNO.C57  2015

    J-GLOBAL

  • 水蒸気爆発抑制材の開発と使用法

    古谷正裕, 新井崇洋

    日本機械学会年次大会講演論文集(CD-ROM)   2015   ROMBUNNO.S1740106  2015

    J-GLOBAL

  • 高圧条件下での発熱バンドル内沸騰二相流におけるボイド率分布計測手法の開発

    新井崇洋, 古谷正裕, 白川健悦, 西義久

    日本機械学会年次大会講演論文集(CD-ROM)   2015   ROMBUNNO.G0600404  2015

    J-GLOBAL

  • S1740106 Development and Prescription of Steam Explosion Retardant

    FURUYA Masahiro, ARAI Takahiro

    The Proceedings of Mechanical Engineering Congress, Japan   2015 ( 0 ) _S1740106 - -_S1740106-  2015

     View Summary

    Polyethylene glycol (PEG) was found to be a reliable retardant of steam explosions. In order to investigate the effects of molecular weight of PEG, concentration and salt additives on the controllability of steam explosions, experiments were conducted for tin drops immersed in a solution pool. Steam explosion was suppressed with a 0.03 wt% PEG solution for molecular weight of 4 million. This is because the cloudy-point phenomenon stabilizes vapor film and prevents the solution from mixing finely by the precipitated solute near the steam-water interface. The molecular weight must be selected in reference to the cloudy-point temperature to be lower than saturation temperature by a certain degrees at the target pressure. Steam explosion may occur in a PEG solution by adding 1 wt% of sodium chloride, because such salts act as steam explosion promoter and reduce the cloudy-point temperature significantly.

    DOI CiNii

  • Spatial Scaling Effect of Interfacial Tension and Contact Angle on Interfacial Motion in Three-Fluid Dam Break

    FURUYA Masahiro, OKA Yoshiaki, SATO Makoto, LO Simon, ARAI Takahiro

    JAPANESE JOURNAL OF MULTIPHASE FLOW   28/29 ( 5/1 ) 579 - 582  2015

     View Summary

    VOF numerical simulations were conducted for a three-fluid dam break problem to investigate the scaling effect on mixing and stratification processes of three immiscible fluids by gravity. Two liquids (silicone oil and salt water) were initially separated with a vertical partition plate and filled in a rectangle container. Computational fluid dynamics analysis was conducted with Star CCM+ version 9.02. The contact angle affects the interfacial motion significantly in a relatively small-scale domain, while the interfacial tension affects it in a relatively large-scale domain.

    DOI CiNii J-GLOBAL

  • G0600404 Development of void fraction distribution measurement for boiling two-phase flow in heated rod bundle under high pressure condition

    ARAI Takahiro, FURUYA Masahiro, SHIRAKAWA Kenetsu, NISHI Yoshihisa

    The Proceedings of Mechanical Engineering Congress, Japan   2015 ( 0 ) _G0600404 - -_G0600404-  2015

     View Summary

    We devised the Subchannel void sensor (SCVS), which consists of 6-by-6 wire electrodes and 5-by-5 rod electrodes to acquire boiling two-phase flow dynamics in a 5-by-5 rod bundle at high-pressure and high-temperature conditions. SCVS can acquire a time series data of local void fraction at 32 points of central subchannel region and at 100 points of near rod surface regions. The experimental results exhibit the axial and cross-sectional distributions of void fraction, which are suitable to validate the CFD and subchannel codes.

    DOI CiNii

  • ICONE23-1789 ABLATION AND MELTING RELOCATION OF HEMISPHERE VESSEL DUE TO NATURAL CONVECTION

    Furuya Masahiro, Oka Yoshiaki

    The Proceedings of the International Conference on Nuclear Engineering (ICONE)   2015 ( 0 ) _ICONE23 - 1-_ICONE23-1  2015

     View Summary

    In order to investigate ablation failure mode of lower head with molten core during a severe accident of light water reactors, ablation and melting relocation experiments were conducted with a hemisphere vessel with a drain hole. Three different silicone oil were used to investigate an effect of fluid viscosity to simulate a molten core. The hemisphere vessel is molded with lead bismuth eutectic alloy. The vessel wall thinning and melt relocation occurred just below the silicone oil level by ablation due to natural convection. For the lowerviscosity silicone oil, it results in breaking all around the vessel wall. On the contrary for the higher-viscosity silicone oil, the drain hole were ablated as well which enhance drainage flow. The time series of ablated molten wall and silicone oil weights drained from the hole were quantified separately on the basis of measured volume and weight of drained fluids for the code validation.

    DOI CiNii

  • ICONE23-1653 DECONTAMINATION PROCESS OF ELEMENTAL IODINE WITH FILTERED CONTAINMENT VENTING SYSTEM

    Kanai Taizo, Furuya Masahiro, Arai Takahiro, Tanaka Nobuyuki, Nishi Yoshihisa, Shirakawa Kenetsu, Nishimura Satoshi, Satake Masaaki

    The Proceedings of the International Conference on Nuclear Engineering (ICONE)   2015 ( 0 ) _ICONE23 - 1-_ICONE23-1  2015

     View Summary

    In the event of a nuclear power plant accident pressure within the containment can increase. A Filtered Containment Venting System (FCVS) allows for over-pressure release through multi-scrubbers (Venturi-scrubber, bubbling-scrubber, metal fiber filter and molecular sieve) and reduces the radioactive release. However, FCVS performance changes depending on operational conditions, e.g. steam flow rate, pressure, operating time and so on. The Central Research Institute of Electric Power Industry (CRIEPI)full-height FCVS test facility is constructed to measure FCVS performance under several conditions and can evaluate the decontamination factor (DF) of three major targets (aerosol, elemental iodine (I_2) and organic iodine (CH_3I)). This project is intended to acquire a systematic database of FCVS performance and optimize the FCVS operation procedure. The CRIEPI test vessel is about 8 m high, with an internal diameter of 0.5 m. FCVS performance tests were conducted under the following conditions: maximum pressure and temperature of 0.8 MPa and 170℃, inlet gas flow of steam (〜1600 kg/h) and air (〜300 kg/h) and containing aerosol/ iodine/ organic iodine. If fission product iodine gas is released into the environment during a severe accident, it will have a major impact on public health. This paper addresses the iodine decontamination performance by the bubbling effect. Iodine is effectively soluble in an alkaline solution. Accordingly, 0.5 wt% sodium hydroxide (NaOH) or a mixture of 0.2 wt% sodium thiosulfate (Na_2S_2O_3) and 0.5 wt% NaOH is used as an iodine filter (absorber) and during the experiment, an alkaline solution with composition equivalent to the actual equipment is used. The concentration of elemental iodine is quantified with an Inductively-Coupled Plasma with Mass Spectrometry (ICP-MS), while iodine DF is defined by the concentration ratio at the inlet and outlet. Iodine DF shows low dependence on flow dynamics, but dependence on solution property. Where iodine concentration is low, DF is high (between 10^4 and 10^5) and vice versa when the iodine concentration (saturate) increases.

    DOI CiNii

  • ICONE23-1551 COOLABILITY HEIGHT OF 5×5 HEATED ROD BUNDLE IN REFERENCE TO COLLAPSED LEVEL AND BOILING TWO-PHASE FLOW DYNAMICS

    Arai Takahiro, Furuya Masahiro, Kanai Taizo, Shirakawa Kenetsu, Nishi Yoshihisa

    The Proceedings of the International Conference on Nuclear Engineering (ICONE)   2015 ( 0 ) _ICONE23 - 1-_ICONE23-1  2015

     View Summary

    In accidents when the water level of the reactor core descends below the top of the active fuel, the cooling limit height is a key factor in determining the accident mitigation procedure and the boiling two-phase flow in a fuel-rod bundle exhibits multi-dimensional and complex flow structures during such boil-off process. A rod bundle boil-off experiment was conducted to determine the three-dimensional void-fraction distribution and axial profile of the rod-surface temperature during the boil-off process under atmospheric pressure conditions. The 5×5 rod bundle, featuring a heated length of 2 m, had an axially and radially uniform power profile, with eight pairs of sheath thermocouples embedded on the heated rod to monitor its surface temperature distribution. The void-fraction distribution was acquired with five pairs of SubChannel Void Sensor (SCVS) as time-series data. The experimental results showed the relationship between an effective cooling level and boiling two-phase flow dynamics in the rod bundle.

    DOI CiNii

  • Ablation and melting relocation of hemisphere vessel due to natural convection

    Masahiro Furuya, Yoshiaki Oka

    International Conference on Nuclear Engineering, Proceedings, ICONE   2015-January  2015

     View Summary

    In order to investigate ablation failure mode of lower head with molten core during a severe accident of light water reactors, ablation and melting relocation experiments were conducted with a hemisphere vessel with a drain hole. Three different silicone oil were used to investigate an effect of fluid viscosity to simulate a molten core. The hemisphere vessel is molded with lead bismuth eutectic alloy. The vessel wall thinning and melt relocation occurred just below the silicone oil level by ablation due to natural convection. For the lower-viscosity silicone oil, it results in breaking all around the vessel wall. On the contrary for the higher-viscosity silicone oil, the drain hole were ablated as well which enhance drainage flow. The time series of ablated molten wall and silicone oil weights drained from the hole were quantified separately on the basis of measured volume and weight of drained fluids for the code validation.

  • Coolability height of 5×5 heated rod bundle in reference to collapsed level and boiling two-phase flow dynamics

    Takahiro Arai, Masahiro Furuya, Taizo Kanai, Kenetsu Shirakawa, Yoshihisa Nishi

    International Conference on Nuclear Engineering, Proceedings, ICONE   2015-January  2015

     View Summary

    In accidents when the water level of the reactor core descends below the top of the active fuel, the cooling limit height is a key factor in determining the accident mitigation procedure and the boiling two-phase flow in a fuel-rod bundle exhibits multi-dimensional and complex flow structures during such boil-off process. A rod bundle boil-off experiment was conducted to determine the three-dimensional void-fraction distribution and axial profile of the rod-surface temperature during the boil-off process under atmospheric pressure conditions. The 5×5 rod bundle, featuring a heated length of 2 m, had an axially and radially uniform power profile, with eight pairs of sheath thermocouples embedded on the heated rod to monitor its surface temperature distribution. The void-fraction distribution was acquired with five pairs of SubChannel Void Sensor (SCVS) as time-series data. The experimental results showed the relationship between an effective cooling level and boiling two-phase flow dynamics in the rod bundle.

  • Decontamination process of elemental iodine with Filtered Containment Venting System

    Taizo Kanai, Masahiro Furuya, Takahiro Arai, Nobuyuki Tanaka, Yoshihisa Nishi, Kenetsu Shirakawa, Satoshi Nishimura, Masaaki Satake

    International Conference on Nuclear Engineering, Proceedings, ICONE   2015-January  2015

     View Summary

    In the event of a nuclear power plant accident pressure within the containment can increase. A Filtered Containment Venting System (FCVS) allows for over-pressure release through multi-scrubbers (Venturi-scrubber, bubbling-scrubber, metal fiber filter and molecular sieve) and reduces the radioactive release. However, FCVS performance changes depending on operational conditions, e.g. steam flow rate, pressure, operating time and so on. The Central Research Institute of Electric Power Industry (CRIEPI) full-height FCVS test facility is constructed to measure FCVS performance under several conditions and can evaluate the decontamination factor (DF) of three major targets (aerosol, elemental iodine (I2) and organic iodine (CH3I)). This project is intended to acquire a systematic database of FCVS performance and optimize the FCVS operation procedure. The CRIEPI test vessel is about 8 m high, with an internal diameter of 0.5 m. FCVS performance tests were conducted under the following conditions: maximum pressure and temperature of 0.8 MPa and 170°C, inlet gas flow of steam (∼1600 kg/h) and air (∼300 kg/h) and containing aerosol/iodine/organic iodine. If fission product iodine gas is released into the environment during a severe accident, it will have a major impact on public health. This paper addresses the iodine decontamination performance by the bubbling effect. Iodine is effectively soluble in an alkaline solution. Accordingly, 0.5 wt% sodium hydroxide (NaOH) or a mixture of 0.2 wt% sodium thiosulfate (Na2S2O3) and 0.5 wt% NaOH is used as an iodine filter (absorber) and during the experiment, an alkaline solution with composition equivalent to the actual equipment is used. The concentration of elemental iodine is quantified with an Inductively-Coupled Plasma with Mass Spectrometry (ICP-MS), while iodine DF is defined by the concentration ratio at the inlet and outlet. Iodine DF shows low dependence on flow dynamics, but dependence on solution property. Where iodine concentration is low, DF is high (between 104 and 105) and vice versa when the iodine concentration (saturate) increases.

  • Boiled-up level and boiling two-phase flow dynamics in 5 × 5 heated rod bundle during boil-off process under atmospheric pressure conditions

    Takahiro Arai, Masahiro Furuya, Taizo Kanai, Kenetsu Shirakawa, Yoshihisa Nishi

    International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015   9   7312 - 7322  2015

     View Summary

    In the case of an accident and when a water level of a reactor core falls below the top level of active fuel, the cooling limit height becomes a key factor for determining the accident mitigation procedure. To predict the cooling limit height, it is important to clarify a two-phase mixture level in a rod bundle during the boil-off process. The two-phase mixture level depends on the collapsed level and void fraction distribution. During the boil-off process, a boiling two-phase flow in the rod bundle exhibits multidimensional and complex flow structures. The paper addresses a three-dimensional void fraction distribution and a two-phase mixture level in 5 × 5 heated rod bundles during the boil-off process, under atmospheric pressure conditions. The heated rod length is 3.7 m, which is the same as the fuel rod in boiling water-reactor (BWR). The 5x5 rod bundles have an axially and radially uniform power profile, and eight pairs of sheath thermocouples are embedded in the heated rod to monitor their axial surface temperature profiles. The diameter of the heated rod is 10 mm, and the rod pitch is 13 mm. The void fraction distributions were acquired with eight pairs of subchannel void sensors (SCVS) as time series data. The two-phase mixture level was evaluated by side-viewing images acquired with two high-speed digital video cameras. The experimental result exhibits a relationship of the boiling two-phase flow dynamics to the two-phase mixture level, and the void fraction during the boil-off process.

  • Suppression measures and effective triggering retardant of steam explosions

    Masahiro Furuya, Takahiro Arai

    International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015   8   6579 - 6589  2015

     View Summary

    Steam explosion has been a potential threat during severe accident in light water reactors. We had reported that Polyethylene glycol (PEG) is a reliable retardant of steam explosions. Experiments were conducted to investigate the effects of concentration, molecular weight and salt additives on the controllability of steam explosions. Steam explosion was suppressed with a 0.03 wt% PEG solution for molecular weight of 4 million. This is because the cloudy-point phenomenon stabilizes vapor film and prevents the solution from mixing finely by the precipitated solute near the steam-water interface. The stabilizing effect of vapor film was confirmed in a solid stainless-steel sphere quenching experiment as well. The molecular weight must be selected in reference to the cloudy-point temperature to be lower than saturation temperature by a certain degrees at the target pressure. At atmospheric pressure, a molecular weight of 4 million is demonstrated to suppress steam explosions. The effective concentration became denser when large share stress and/or external force act on the vapor film. Steam explosion may occur in a PEG solution by adding lwt% of sodium chloride, because such salts act as steam explosion promoter.

  • Analysis of metal vessel wall ablation experiment with high temperature liquid by MPS method

    Daisuke Masumura, Yoshiaki Oka, Akifumi Yamaji, Masahiro Furuya

    International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015   9   7401 - 7413  2015

     View Summary

    In a severe accident of a light water reactor, ablation of the reactor pressure vessel (RPV) lower head by corium is a key phenomenon, which affects progression of the accident. The Moving Particle Semi- implicit (MPS) method is one of particle methods that calculate behavior of incompressible fluid by semi-implicit method. In preceding studies, MPS models have been developed to analyze phenomena such as heat conduction, phase change, natural convection, thermal stratification, and radiation heat transfer. These phenomena are expected to play key roles in the lower head ablation. This paper aims to investigate whether the MPS method is capable of analyzing the lower head ablation phenomenon, which involves complex interactions of the above mentioned phenomena. The small-scale experiment carried out at Central Research Institute of Electric Power Industry (CRIEPI) using Pb-Bi vessel and silicone oil was analyzed. The heat transfer model was modified for evaluation of heat transfer between the vessel and the oil. The results were compared both qualitatively and quantitatively with the experiment. The former is the comparison of the simulation and experiment regarding phenomena that the liquid ablates the metal vessel and discharges through the vessel wall, which showed good agreement. The latter are comparisons of the calculated liquid temperature, ablation start time and discharge start time with respect to the corresponding measurements. The analyses have shown that the MPS method is capable of analyzing ablation phenomenon qualitatively, but needs further development for quantitative prediction, including investigations on influence of the particle size used in the simulation.

  • Radial and axial development of boiling Two-Phase flowin A 5 × 5 Heated rod bundle under atmospheric pressure condition

    Takahiro Arai, Masahiro Furuya, Taizo Kanai, Kenetsu Shirakawa, Yoshihisa Nishi

    Multiphase Science and Technology   27 ( 2-4 ) 203 - 213  2015

     View Summary

    The boiling two-phase flow in the fuel rod bundle of a boiling water reactor exhibits multidimensional and transient flow dynamics. Coolability of the fuel is the key to ensuring safety, particularly in the event of any accident that involves the coolant injection malfunctioning and the core becoming uncovered. This paper addresses boiling two-phase flow dynamics in a 5 £ 5 heated rod bundle under atmospheric pressure. The diameter of the heated rod is 10 mm, the rod pitch is 13 mm, and the heated length is 3.71 m. The radial power profile is the key experimental parameter: (i) uniform, (ii) center peak, (iii) side peak, and (iv) corner peak. The cross-sectional void-fraction distribution of a total of 132 points was acquired at more than 800 frames (cross sections) per second with a SubChannel Void Sensor (SCVS). The axial void-fraction distribution was also acquired with eight pairs of SCVS, which explain the radial and axial development of boiling two-phase flow in the rod bundle.

    DOI

  • MAAP5.01及びMELCOR2.1を用いた軽水炉代表プラントの過酷事故解析 : BWR-5/Mark-Ⅱ改良型プラントの全交流電源喪失解析の比較

    西村 聡, 日渡 良爾, 古谷 正裕

    電力中央研究所報告. L   ( 13006 ) 巻頭1 - 3,1-31  2014.06

    CiNii

  • Development of Measurement Method of Void Fraction Distribution for Boiling Water Flow in Heated Rod Bundle

    新井 崇洋, 古谷 正裕, 金井 大造

    混相流   27 ( 5 ) 647 - 654  2014.03

    CiNii

  • Ultra Rapid Cooling and Atomizing Method Utilizing Self-Sustained Vapor Explosions

    古谷 正裕

    化学工学   78 ( 3 ) 180 - 183  2014.03

    CiNii

  • 直接通電発熱バンドル内における二相流計測手法の開発

    渡辺瞬, 新井崇洋, 西義久, 古谷正裕, 白川健悦, 金井大造

    日本伝熱シンポジウム講演論文集(CD-ROM)   51st   ROMBUNNO.A313  2014

    J-GLOBAL

  • 部分露出状態にある発熱管群の沸騰伝熱

    新井崇洋, 古谷正裕, 金井大造, 白川健悦, 西義久

    日本伝熱シンポジウム講演論文集(CD-ROM)   51st   ROMBUNNO.A222  2014

    J-GLOBAL

  • 半球容器内自然対流によるアブレーション・溶融移動

    古谷正裕, 岡芳明

    日本伝熱シンポジウム講演論文集(CD-ROM)   51st   ROMBUNNO.I334  2014

    J-GLOBAL

  • フィルタベント内部流動とエアロゾル除染能力の相関

    金井大造, 古谷正裕, 新井崇洋, 白川健悦, 西義久

    混相流シンポジウム講演論文集(CD-ROM)   2014   ROMBUNNO.E332  2014

    J-GLOBAL

  • 三流体ダム崩壊の界面移動における接触角および界面張力の寸法効果

    古谷正裕, 岡芳明, 佐藤誠, LO Simon, 新井崇洋

    混相流シンポジウム講演論文集(CD-ROM)   2014   ROMBUNNO.A223  2014

    J-GLOBAL

  • 粉と化学工学 持続的蒸気爆発による超急冷微粒子製造法

    古谷正裕

    化学工学   78 ( 3 ) 180 - 183  2014

    J-GLOBAL

  • MPS法による溶融物挙動解析(5)半球容器のアブレーション・溶融移動試験

    古谷正裕, 岡芳明

    日本原子力学会春の年会予稿集(CD-ROM)   2014   ROMBUNNO.M26  2014

    J-GLOBAL

  • フィルタベントシステムの運用高度化(3)気液二相流動構造とエアロゾル除染性能

    金井大造, 古谷正裕, 新井崇洋, 西義久, 白川健悦, 田中伸幸, 西村聡, 佐竹正哲

    日本原子力学会春の年会予稿集(CD-ROM)   2014   ROMBUNNO.L10  2014

    J-GLOBAL

  • 燃料露出過程の炉内流動評価(3)大気圧下でのバンドル内水位低下に伴うコラプスト水位と冷却可能水位の相関

    新井崇洋, 古谷正裕, 金井大造, 白川健悦, 西義久

    日本原子力学会春の年会予稿集(CD-ROM)   2014   ROMBUNNO.M49  2014

    J-GLOBAL

  • MAAP5.01及びMELCOR2.1を用いた軽水炉代表プラントの過酷事故解析―BWR‐5/Mark‐II改良型プラントの全交流電源喪失解析の比較―

    西村聡, 日渡良爾, 古谷正裕, 西義久

    電力中央研究所原子力技術研究所研究報告   ( L13006 ) 37P  2014

    J-GLOBAL

  • MAAP及びMELCORを用いた軽水炉代表プラントの過酷事故解析―BWR‐5/Mark‐II改良型プラントの全交流電源喪失解析の比較―

    西村聡, 日渡良爾, 古谷正裕, 西義久

    日本原子力学会秋の大会予稿集(CD-ROM)   2014   ROMBUNNO.J42  2014

    J-GLOBAL

  • フィルタベントシステムの運用高度化(4)高温高圧蒸気流中のエアロゾル除染能力評価

    金井大造, 古谷正裕, 新井崇洋, 西義久, 白川健悦, 田中伸幸, 西村聡, 佐竹正哲

    日本原子力学会秋の大会予稿集(CD-ROM)   2014   ROMBUNNO.J20  2014

    J-GLOBAL

  • 直接通電加熱バンドルにおける沸騰二相流のボイド及び温度分布計測法

    渡辺瞬, 新井崇洋, 古谷正裕, 西義久, 白川健悦, 金井大造

    日本原子力学会秋の大会予稿集(CD-ROM)   2014   ROMBUNNO.K17  2014

    J-GLOBAL

  • 燃料露出過程の炉内流動評価(4)大気圧下でのバンドル内沸騰二相流動に及ぼす有効発熱長の影響

    新井崇洋, 古谷正裕, 金井大造, 白川健悦, 西義久

    日本原子力学会秋の大会予稿集(CD-ROM)   2014   ROMBUNNO.J29  2014

    J-GLOBAL

  • 軽水炉のシビアアクシデント下の海水・ホウ酸注入時の影響に関する試験(3)5×5バンドル流路で沸騰濃縮された海水の塩析出挙動

    古谷正裕, 小城烈, 秋葉美幸, 星陽崇, 堀田亮年

    日本原子力学会秋の大会予稿集(CD-ROM)   2014   ROMBUNNO.J34  2014

    J-GLOBAL

  • Development of Measurement Method of Void Fraction Distribution for Boiling Water Flow in Heated Rod Bundle

    ARAI Takahiro, FURUYA Masahiro, KANAI Taizo, SHIRAKAWA Kenetsu, NISHI Yoshihisa

    JAPANESE JOURNAL OF MULTIPHASE FLOW   27/28 ( 5/1 ) 647 - 654  2014

     View Summary

    SubChannel Void Sensor (SCVS) consisting of electrodes with 6-wire by 6-wire and 5-rod by 5-rod electrodes has been developed to acquire boiling two-phase flow dynamics in 5×5 rod bundle geometry. The SCVS can acquire a local void fraction in 32 points (=6×6-4) of central subchannel regions and 100 points (=4×25) of near rod surface regions. The temporal resolution is up to 5000 frames (cross sections) per second. The devised sensors are installed every 500mm in the heated rod bundle and exhibits the quasi three-dimensional boiling two-phase flow structures, i.e. void fraction distribution.

    DOI CiNii J-GLOBAL

  • Development of Electrocatalyst to Reduce Carbon Dioxide

    古谷正裕

    日本機械学会論文集 B編(Web)   79 ( 799 )  2013

    J-GLOBAL

  • 発熱バンドル流路内沸騰二相流におけるボイド率及び相速度の多次元高速計測手法の開発

    新井崇洋, 古谷正裕, 金井大造, 白川健悦

    混相流シンポジウム講演論文集(CD-ROM)   2013   ROMBUNNO.E114  2013

    J-GLOBAL

  • 燃料棒溶融数値解析手法の開発((1)溶融実験)

    古谷正裕, 永武拓, 高瀬和之, 吉田啓之, 永瀬文久

    混相流シンポジウム講演論文集(CD-ROM)   2013   ROMBUNNO.E121  2013

    J-GLOBAL

  • 燃料棒溶融数値解析手法の開発((2)溶融実験の数値解析結果)

    永武拓, 高瀬和之, 古谷正裕, 吉田啓之, 永瀬文久

    混相流シンポジウム講演論文集(CD-ROM)   2013   ROMBUNNO.E122  2013

    J-GLOBAL

  • 二流体の重力層分離過程(1)可視化実験

    古谷正裕, 岡芳明, 近藤雅裕, LI Gen

    日本機械学会年次大会講演論文集(CD-ROM)   2013   ROMBUNNO.S083022  2013

    J-GLOBAL

  • MAAP5コードによるBWRプラントのシビアアクシデント解析

    西村聡, 日渡良爾, 古谷正裕, 西義久

    日本機械学会年次大会講演論文集(CD-ROM)   2013   ROMBUNNO.S083021  2013

    J-GLOBAL

  • S083022 Stratification Behavior of Two Fluids by Gravity : (1) Experimental Observation

    Furuya Masahiro, OKA Yoshiaki, KONDO Masahiro, LI Gen

    The Proceedings of Mechanical Engineering Congress, Japan   2013 ( 0 ) _S083022 - 1-_S083022-5  2013

     View Summary

    Two experimental campaigns were conducted to observe mixing and stratification processes of two immiscible fluids by gravity. Parametric study reviles the effects of two key fluid properties: kinematic viscosity (or molecular weight) for silicone oil and density (or concentration) for sodium chloride aqueous water. After withdrawal of a vertical partition plate, two liquids intersects earlier at the center, while the fluids stick on the walls. Another experiments were conducted with the salt water injecting on to the silicone oil pool. The kinematic viscosity and density difference affect three-dimensional mixing and stratification processes significantly. These visual databases are suitable for a code validation on the interfacial phenomena. Key Words : Gravity-driven stratification, Injection, Mixture, Phase separation, Liquid-liquid, Flow visualization

    DOI CiNii

  • S083021 Severe Accident Analysis of a BWR Plant with MAAP5 Code

    NISHIMURA Satoshi, HIWATARI Ryoji, FURUYA Masahiro, NISHI Yoshihisa

    The Proceedings of Mechanical Engineering Congress, Japan   2013 ( 0 ) _S083021 - 1-_S083021-4  2013

     View Summary

    Severe accident analysis was performed for a station blackout (TBU) sequence in a typical BWR-5 plant with modified Mark-II type containment by MAAP5.01 code. Core heatup, degradation and relocation processes, reactor vessel failure and resultant impact on the containment including hydrogen production were investigated. Sensitivity analysis focusing on direct contact heating (DCH), which affect on the consequence of TBU sequence, was also performed. If extensive DCH does not occur in the TBU sequence, failure of the containment vessel can be postponed. On the other hand, concrete ablation at a floor and a side wall inside the pedestal due to molten core - concrete interaction (MCCI) significantly increases, because a large amount of debris with high temperature stays inside the pedestal.

    DOI CiNii

  • S083023 Stratification Behavior of Two Fluids by Gravity : (2) Analysis with MPS Method

    Li Gen, Oka Yoshiaki, Furuya Masahiro, Kondo Masahiro

    The Proceedings of Mechanical Engineering Congress, Japan   2013 ( 0 ) _S083023 - 1-_S083023-5  2013

     View Summary

    The Moving Particle Semi-implicit (MPS) method is improved to analyze the stratification behavior of highly viscous fluids. The stratifying processes of silicone oil and salt water are simulated in 2D and 3D respectively, and the time history of fluid shapes is compared with experimental results. Characteristic fluid shapes for high and low silicone oil viscosity conditions are well predicted and the results show good agreement with experiments.

    DOI CiNii

  • Development of Electrocatalyst to Reduce Carbon Dioxide

    FURUYA Masahiro

    TRANSACTIONS OF THE JAPAN SOCIETY OF MECHANICAL ENGINEERS Series B   79 ( 799 ) 286 - 290  2013

     View Summary

    The carbon-doped copper oxide is devised as an electrocatalyst to reduce carbon dioxide under ambient pressure and temperature. The electrode was immersed into a KHCO3 aqueous solution with CO2 bubbling. The electric potential was maintained at -1.64 V vs SHE. The electrode prepared at 900&deg;C gives the maximum production rate of ethylene (25 %), ethanol (6.9 %) and 1-propanol (3.6 %). The production rate of methane, from which is harmful to separate ethylene, was suppressed to one fifteenth of that of ethylene. In contrast to a thermally-oxidized copper-oxide layer, the doped-carbon and a high ratio of Cu2O to CuO in the devised electrocatalyst may result in the higher productivity and selectivity.

    DOI CiNii J-GLOBAL

  • Three-dimensional measurement of the two-phase flow development in a large pipe using bubble-tracking method

    KANAI TAIZO, FURUYA MASAHIRO, ARAI TAKAHIRO, SHIRAKAWA KENETSU

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2013 ( 0 ) 294 - 294  2013

     View Summary

    内径224mm大口径円筒管内二相流の軸方向発達過程を二枚セットのワイヤメッシュセンサにより計測した。得られた三次元気泡信号に対して、ワイヤメッシュセンサ間で類似の気泡を追跡する気泡追跡法を開発し、この手法により気泡径分布・三次元速度分布を評価した。

    DOI CiNii J-GLOBAL

  • Development of Analytical Method for Behavior of Fuel Melting by Particle Method:(2)Melt Relocation Experiments

    FURUYA MASAHIRO, NAGATAKE TAKU, YOSHIDA HIROYUKI, TAKASE KAZUYUKI, NAGASE FUMIHISA

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2013 ( 0 ) 252 - 252  2013

     View Summary

    燃料の溶融移動挙動を把握するとともに解析の検証データを得るため、円柱状金属試験片を融点以上の雰囲気に設置し、溶融変形をビデオ観察した。変形の時系列変化を金属物性値と対比して考察した。

    DOI CiNii J-GLOBAL

  • Development of Analytical Method for Behavior of Fuel Melting by Particle Method:(1)Outline of the Research Plan

    Nagatake Taku, Takase Kazuyuki, Furuya Masahiro, Yoshida Hiroyuki, Nagase Fumihisa

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2013 ( 0 ) 251 - 251  2013

     View Summary

    福島第一原子力発電所では、全電源喪失により燃料棒の冷却が出来なくなり炉心溶融に至った。溶融した燃料は圧力容器下部にデブリとして堆積していると考えられており、従って、廃炉措置を行うにあたり、溶融燃料取り出しや再臨界評価のためのデブリ分布等の情報が必要である。本研究では、福島第一原子力発電所廃炉措置等に資するため、燃料溶融及び溶融燃料の落下挙動に着目し、炉心構成要素である燃料棒・チャンネルボックス・制御棒の溶融過程を解析可能である粒子法をもとにした数値解析手法の開発を目的とする。本報では、本解析手法開発の研究計画の概要について報告する。

    DOI CiNii J-GLOBAL

  • Experiments on Heat Removal Influenced by Injection of Seawater and Boric Acid under Severe Accident Conditions in LWRs:(1) Concept and Design of Experiment

    HOTTA AKITOSHI, HOSHI HARUTAKA, FURUYA MASAHIRO

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2013 ( 0 ) 271 - 271  2013

     View Summary

    原子炉施設におけるシビアアクシデント発生時のAM&lt;策として、水源が枯渇した後に原子炉及び使用済燃料プールに海水注入され、これがホウ酸と混合する状況において、炉心及びデブリベッドの冷却性に及ぼす影響を把握するための試験を計画している。ここでは、その構想をまとめる。

    DOI CiNii J-GLOBAL

  • Concurrent upward liquid film dynamics on both surfaces of annular channel with liquid film sensor

    ARAI TAKAHIRO, FURUYA MASAHIRO, KANAI TAIZO, SHIRAKAWA KENETSU

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2013 ( 0 ) 293 - 293  2013

     View Summary

    曲率を有する平面上の液膜分布を高速計測可能な液膜センサを開発し、環状流路を構成する直径12mm内管および直径18mm外管それぞれの表面に液膜センサを配置することで、液膜挙動を両面で同時に計測した。水-空気体系において、気液混合部から下流側300mmの位置に形成される液膜挙動を10&amp;times;32計測点の液膜分布として5000断面/秒で取得し、気液混合条件および入口みかけ流速の影響を液膜厚さ,波速などによって評価した。

    DOI CiNii J-GLOBAL

  • 燃料露出過程の炉内流動評価(1)全体計画

    西義久, 新井崇洋, 古谷正裕, 白川健悦, 金井大造, 西村聡, 佐竹正哲, 茶木雅夫

    日本原子力学会秋の大会予稿集(CD-ROM)   2013   ROMBUNNO.K11  2013

    J-GLOBAL

  • 燃料露出過程の炉内流動評価(2)サブチャンネルボイドセンサを用いた5×5バンドル流路内沸騰二相流のボイド率分布及び相速度分布計測

    新井崇洋, 古谷正裕, 金井大造, 白川健悦, 西義久

    日本原子力学会秋の大会予稿集(CD-ROM)   2013   ROMBUNNO.K12  2013

    J-GLOBAL

  • 使用済燃料プールの事故時冷却特性評価―MAAPコードを用いた冷却材喪失事故解析―

    西村聡, 日渡良爾, 古谷正裕, 西義久

    日本原子力学会秋の大会予稿集(CD-ROM)   2013   ROMBUNNO.K44  2013

    J-GLOBAL

  • フィルタベントシステムの運用高度化(1)全体計画

    古谷正裕, 金井大造, 田中伸幸, 西村聡, 新井崇洋, 佐竹正哲, 白川健悦, 西義久

    日本原子力学会秋の大会予稿集(CD-ROM)   2013   ROMBUNNO.K26  2013

    J-GLOBAL

  • フィルタベントシステムの運用高度化(2)小口径管を用いた流動と除染係数の相関

    金井大造, 古谷正裕, 田中伸幸, 西村聡, 新井崇洋, 佐竹正哲, 白川健悦, 西義久

    日本原子力学会秋の大会予稿集(CD-ROM)   2013   ROMBUNNO.K27  2013

    J-GLOBAL

  • 軽水炉のシビアアクシデント下の海水・ホウ酸注入時の影響に関する試験(2)海水及びホウ酸水混合溶液の析出物

    小城烈, 堀田亮年, 星陽崇, 川部隆平, 古谷正裕

    日本原子力学会秋の大会予稿集(CD-ROM)   2013   ROMBUNNO.K23  2013

    J-GLOBAL

  • Evaluation of cooling characteristics for spent fuel pool accidents : Analysis of loss-of-cooling and loss-of-coolant accident in SFP with MAAP code

    西村 聡, 日渡 良爾, 古谷 正裕

    電力中央研究所報告. L   ( L12007 ) 巻頭1 - 3,1-31  2013

    CiNii J-GLOBAL

  • Bubble Identification and Three-dimensional Phasic Velocity Measurement of Two Phase Flow Dynamics in a Large Vertical Pipe

    KANAI Taizo, FURUYA Mashahiro, ARAI Takahiro, SHIRAKAWA Kennetsu, NISHI Yoshihisa

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2012 ( 0 ) 307 - 307  2012

     View Summary

    ワイヤメッシュセンサで大口径管内二相流を計測し、得られた信号から個々の気泡信号を識別し、それぞれの気泡の移動速度を求めた。

    DOI CiNii

  • Three-Dimensional Gas-Phasic Velocity Measurement of Two-Phase Flow in Terms of Bubble Size in a Large-Diameter Pipe

    金井 大造, 古谷 正裕, 新井 崇洋, 白川 健悦, 西 義久

    JAPANESE JOURNAL OF MULTIPHASE FLOW   25 ( 5 ) 443 - 450  2012

     View Summary

    A wire-mesh sensor (WMS) can acquire a void fraction distribution at a high temporal and spatial resolution and also estimate the velocity of a vertical rising flow by investigating the signal time-delay of the two WMSs. The authors propose to extend this time series analysis to estimate the multi-dimensional velocity profile via cross-correlation analysis between a point of upstream WMS and multiple points downstream. Moreover, bubbles behave in various ways according to size, which is used to classify them into certain groups.

    CiNii

  • 最適評価コードTRACEによるフラッシング誘起密度波振動の予測性

    古谷正裕, 西義久, 植田伸幸

    日本機械学会年次大会講演論文集(CD-ROM)   2012   ROMBUNNO.S083034  2012

    J-GLOBAL

  • F202 Development of Electrocatalyst to Reduce Carbon Dioxide

    FURUYA Masahiro

    The Proceedings of the National Symposium on Power and Energy Systems   2012 ( 0 ) 421 - 422  2012

     View Summary

    The carbon-doped copper oxide is devised as an electrocatalyst to reduce carbon dioxide under ambient pressure and temperature. The electrode was immersed into a KHCO_3 aqueous solution with CO_2 bubbling. The electric potential was maintained at -1.64 V vs SHE. The electrode prepared at 900 ℃ gives the maximum production rate of ethylene (25 %), ethanol (6.9 %) and 1-propanol (3.6 %). The production rate of methane, from which is harmful to separate ethylene, was suppressed to one fifteenth of that of ethylene. In contrast to a thermally-oxidized copper-oxide layer, the doped-carbon and a high ratio of Cu_2O to CuO in the devised electrocatalyst may result in the higher productivity and selectivity.

    DOI CiNii

  • S083034 Predictability of Best-Estimate Code, TRACE for Flashing-Induced Density Wave Oscillations

    FURUYA Masahiro, Nishi Yoshihisa, Ueda Nobuyuki

    The Proceedings of Mechanical Engineering Congress, Japan   2012 ( 0 ) _S083034 - 1-_S083034-5  2012

     View Summary

    The best-estimate code TRACE is validated with the stability database of SIRIUS-N Facility at low pressure. The SIRIUS-N facility simulates thermal-hydraulics of the economic simplified BWR (ESBWR) Numerical results exhibits flashing-induced density wave oscillation characteristics, since the oscillation period correlates well with single-phase liquid transit time through the chimney region regardless of the system pressure, inlet subcooling and heat flux Unlike Type-I and II density wave oscillations, the inlet or exit throttling does not affect stability boundary and oscillation amplitude of flashing-induced density wave oscillations significantly. Stability maps in reference to the subcooling and heat flux obtained from the TRACE code agrees with those of the experimental data at low subcooling region, though instability observed in the lower heat flux and higher subcooling region from the stability limit of the experiment.

    DOI CiNii

  • 管群流路内におけるサブチャンネル間相互作用

    新井崇洋, 古谷正裕, 金井大造, 白川健悦

    日本伝熱シンポジウム講演論文集(CD-ROM)   49th   ROMBUNNO.C211  2012

    J-GLOBAL

  • 気泡追跡法による個別気泡速度計測

    金井大造, 古谷正裕, 新井崇洋, 白川健悦, 西義久

    日本伝熱シンポジウム講演論文集(CD-ROM)   49th   ROMBUNNO.C124  2012

    J-GLOBAL

  • Mechanical Property of Surface Modification Layer 'Fresh Green' on the Zircaloy Which Suppresses Corrosion and Hydrogen-Pickup

    FURUYA MASAHIRO, KITAJIMA SHOICHI, SONODA TAKESHI, SAWABE TAKASHI, TOKIWAI MORIYASU, KINOSHITA MOTOYASU

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2012 ( 0 ) 467 - 467  2012

     View Summary

    ジルコニウム合金の表面改質技術フレッシュグリーン(FG)は緻密で密着性が高い保護皮膜であることから、耐食性と耐水素吸収性が向上する。そのメカニズムを検討するため、三点曲げ試験および球形ナノインデンテーションを実施し、FG皮膜に対して機械的特性(亀裂発生応力、弾性率、降伏せん断応力など)を測定し、大気酸化皮膜と比較して考察した。

    DOI CiNii J-GLOBAL

  • Measurement of Two-phase Flow in Fuel rod Bundle using Subchannel Void Sensor and Applicability of Drift Flux Models

    Arai Takahiro, Furuya Masahiro, Kanai Taizo, Shirakawa Kenetsu

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2012 ( 0 ) 309 - 309  2012

     View Summary

    10×10バンドル流路内気液二相流を断面分布として高速計測可能なサブチャンネルボイドセンサを開発し、空気-水二相流動実験へ適用した。さまざまな流動条件に対して得られた実験結果に基づきボイド率分布,相速度分布,気泡コード長分布などによって二相流遷移過程を評価するとともに、既存のドリフトフラックスモデルの予測精度を検証した。

    DOI CiNii J-GLOBAL

  • 大口径円筒管内二相流における気泡弁別と三次元気泡速度測

    金井大造, 古谷正裕, 新井崇洋, 白川健悦

    日本原子力学会春の年会予稿集(CD-ROM)   2012   ROMBUNNO.G04  2012

    DOI J-GLOBAL

  • OpenFOAMを用いた気液二相上昇流の解析

    近藤雅裕, 新井崇宏, 金井大造, 古谷正裕, 西義久

    計算工学講演会論文集(CD-ROM)   17   ROMBUNNO.G-5-4  2012

    J-GLOBAL

  • 二酸化炭素の電気還元触媒の開発

    古谷正裕

    日本機械学会動力・エネルギー技術シンポジウム講演論文集   17th   421 - 422  2012

    J-GLOBAL

  • Flow Structure of Vertical Downward Two-Phase Stream

    Kanai Taizo, Furuya Mashahiro, Arai Takahiro, Shirakawa Kenetsu, Nishi Yoshihisa

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2012 ( 0 ) 235 - 235  2012

     View Summary

    下降流では、モデル開発に必要な実験データおよび流動メカニズムに関する知見が上昇流に比べ不足している。本研究では、ワイヤメッシュセンサ多次元二相流計測手法を水空気系小口径円筒管内下降流及び上昇流計測に適用した。得られた平均ボイド率、平均気泡速度の比較から下降流と上昇流との流動特性の違いを評価することで、下降流に関する既存ドリフトフラックスモデルの適用性を検証した。

    DOI CiNii J-GLOBAL

  • Corrosion and Hydrogen-Pickup Resistance of Zircaloy Cladding with Surface Modification Layer of Platinum and Carbon Co-doped Zirconium Oxide

    FURUYA MASAHIRO, KITAJIMA SHOICHI, SONODA TAKESHI, SAWABE TAKASHI, TOKIWAI MORIYASU, KINOSHITA MOTOYASU, TANAKA NOBUYUKI

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2012 ( 0 ) 426 - 426  2012

     View Summary

    ジルコニウム合金の表面を炭化と酸化を同時に進行させる表面改質技術フレッシュグリーン(FG)は密着性が高く、皮膜が緻密で保護性が高いことから、耐食性と耐水素吸収性が向上する。さらに白金を共担持することにより、耐食性のみならず耐水素吸収性を著しく高めることが360℃高温水中腐食試験で判明した。白金・炭素共担持FG皮膜はMott-Schottky分析によりドナー密度が高く、自然浸漬電位が貴であることが水素吸収を抑制したと考えられる。

    DOI CiNii J-GLOBAL

  • Local stress measurement of oxide film on Zircaloy-2 using synchrotron micro X-ray beam

    Sawabe Takashi, Sonoda Takkeshi, Kitajima Shoichi, Furuya Masahiro

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2012 ( 0 ) 380 - 380  2012

     View Summary

    燃料被覆管のブレイクアウェイ現象は、被覆管の腐食を加速させる。ブレイクア ウェイの周期は、合金組成や水化学環境により異なり、被覆管表面に形成される。酸化膜の内部応力の違いによると推定される。酸化膜は数&amp;mu;mの厚さのため、詳細 な応力解析には高い分解能が必要である。そこで、 放射光によるマイクロビームX線を利用した被覆管酸化膜の局所X線回折と応力解析手法を検討した。&lt;br&gt; X線回折に寄与する体積を最小化するため、ビーム幅を0.2&amp;mu;mとして、高温水中腐食により形成した酸化膜を深さ方向に透過法を用いてX線回折測定した。単斜晶酸化ジルコニウムの格子面間隔は、基材との界面より表面側へ1&amp;mu;mの付近から顕著に狭く、圧縮応力は膜内部の方が界面付近よりも大きいことを示唆している。これは、正方晶酸化ジルコニウムのマルテンサイト変態による体積膨張が圧縮応力と して作用したためと考えられる。

    DOI CiNii J-GLOBAL

  • Measurement of Two-phase Flow in Pool Rod-Bundle System using Subchannel Void Sensor

    ARAI TAKAHIRO, FURUYA MASAHIRO, KANAI TAIZO, SHIRAKAWA KENETSU

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2012 ( 0 ) 232 - 232  2012

     View Summary

    10&amp;times;10バンドル流路内気液二相流を断面分布として高速計測可能なサブチャンネルボイドセンサを開発し、入口液相みかけ流速が0であるプール体系の空気-水二相流動実験へ適用した。実験結果からボイド率分布,相速度分布,気泡コード長分布の軸方向発達過程を評価することにより、既存モデルの予測精度を検証した。

    DOI CiNii J-GLOBAL

  • Fundamental Study of Corrosion Control in Marine and Offshore Structures Using Radiation Induced Surface Activation (RISA)-3rd Report: Temperature Dependence on RISA Corrosion Control Effect

    下村祐介, 馬渕祥吾, 波津久達也, 元田慎一, 賞雅寛而, 植松進, 古谷正裕

    マリンエンジニアリング   47 ( 6 ) 896 - 901  2012

     View Summary

    The temperature dependence on radiation induced surface activation (RISA) corrosion control effect for stainless steel was examined in artificial and natural seawater in this paper. The results obtained are as follows: 1) The RISA effect was evident up to 30 °C, but was hardly recognizable over 45 °C. 2) The RISA effect was observed regardless of the seasons, and the anodic current decreased with increase in the temperature. 3) It was confirmed that corrosion control mechanism proposed in the previous paper was appropriate. The increase of dissolved oxygen was measured in natural seawater by the appearance of RISA effect.

    DOI CiNii J-GLOBAL

  • Fundamental Study of Corrosion Control in Marine and Offshore Structures Using Radiation Induced Surface Activation (RISA) -3rd Report: Temperature Dependence on RISA Corrosion Control Effect

    Shimomura Yusuke, Mabuchi Shogo, Hazuku Tatsuya, Motoda Shin-ichi, Takamasa Tomoji, Uematsu Susumu, Furuya Masahiro

    Marine Engineering   47 ( 6 ) 896 - 901  2012

     View Summary

    The temperature dependence on radiation induced surface activation (RISA) corrosion control effect for stainless steel was examined in artificial and natural seawater in this paper. The results obtained are as follows: 1) The RISA effect was evident up to 30 °C, but was hardly recognizable over 45 °C. 2) The RISA effect was observed regardless of the seasons, and the anodic current decreased with increase in the temperature. 3) It was confirmed that corrosion control mechanism proposed in the previous paper was appropriate. The increase of dissolved oxygen was measured in natural seawater by the appearance of RISA effect.

    DOI CiNii

  • Three-Dimensional Gas-Phasic Velocity Measurement of Two-Phase Flow in Terms of Bubble Size in a Large-Diameter Pipe

    KANAI Taizo, FURUYA Masahiro, ARAI Takahiro, SHIRAKAWA Kenetsu, NISHI Yoshihisa

    JAPANESE JOURNAL OF MULTIPHASE FLOW   25 ( 5 ) 443 - 450  2012

     View Summary

    A wire-mesh sensor (WMS) can acquire a void fraction distribution at a high temporal and spatial resolution and also estimate the velocity of a vertical rising flow by investigating the signal time-delay of the two WMSs. The authors propose to extend this time series analysis to estimate the multi-dimensional velocity profile via cross-correlation analysis between a point of upstream WMS and multiple points downstream. Moreover, bubbles behave in various ways according to size, which is used to classify them into certain groups.

    DOI CiNii J-GLOBAL

  • Analysis of crystal strain in oxide layers formed on Zircaloy-2 by application of μ-XRD method with micro-sized specimen

    澤部 孝史, 園田 健, 古谷 正裕

    電力中央研究所報告. L   ( L11002 ) 巻頭1 - 3,1-20  2012

    CiNii J-GLOBAL

  • Applicability of Best-Estimate Code TRACE to Thermal-Hydraulic Stability : Flashing-Induced Density Wave Oscillations in Natural Circulation BWRs

    古谷 正裕, 西 義久, 植田 伸幸

    電力中央研究所報告. L   ( L11006 ) 1 - 13,巻頭1-3  2012

    CiNii J-GLOBAL

  • Local Measurement and Characteristics of Two-phase Flow in Rod Bundle with Subchannel Void Sensor

    新井 崇洋, 古谷 正裕, 白川 健悦

    電力中央研究所報告. L   ( L11011 ) 1 - 13,巻頭1-3  2012

    CiNii J-GLOBAL

  • Development of Three-Dimensional Individual Bubble-Velocity Measurement Method by Bubble Tracking

    金井 大造, 古谷 正裕, 新井 崇洋

    電力中央研究所報告. L   ( L11014 ) 1 - 17,巻頭1-3  2012

    CiNii J-GLOBAL

  • Three-dimensional measurement and flow development process of the two-phase flow in a large-diameter pipe

    KANAI Taizo, FURUYA Mashahiro, ARAI Takahiro, SHIRAKAWA Kennetsu

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2011 ( 0 ) 255 - 255  2011

     View Summary

    ワイヤを正方格子状に配したワイヤメッシュセンサ(Wire-Mesh Sensor: WMS)は、ワイヤが交差する近接点の導電率から気泡の有無の二次元分布を高速・高精度で計測である。さらに、WMSは近接配置した2組の計測信号の時間差により気泡速度も計測可能である。本研究では、大気圧水‐空気条件において大口径円管内二相流動試験を行い、WMSを二相流動計測に適用し、ボイド率および気泡径の二次元分布計測、気泡速度の三次元計測を行った。気泡速度の三次元計測は上流側WMSの1計測点に対し下流側WMSのある計測点領域(Nx×Ny)間で相互相関分析することにより行った。さらに、気泡信号を気泡径ごとに弁別、抽出し、気泡の大きさごとの三次元速度計測も行うことで、大口径円筒管における二相流動の発達過程を気泡サイズ毎に評価した。

    DOI CiNii

  • Development of Measurement Method of Multi-dimensional Gas Velocity Distributions and Application to Swirl Flow Dynamics

    KANAI Taizo, FURUYA Mashahiro, ARAI Takahiro, SHIRAKAWA Kennetsu

    Proceedings of National Heat Transfer Symposium   2011 ( 0 ) 307 - 307  2011

     View Summary

    ワイヤメッシュセンサ(WMS)はコンダクタンス法に基づき断面ボイド率分布を計測でき、近接した2層目のWMSとのボイド信号の相互相関により気泡速度を推定出来る。本研究では多次元速度分布を推定するため、大口径円管内を上昇する二相流動を対象に、上流側WMSの1計測点と、下流側の21×21計測点との間で相互相関分析を行い、相互相関係数が最大になる組み合わせで速度ベクトルが定めた。さらに、WMS信号のWavelet解析を行い、気泡信号を周波数(気泡の大きさ)ごとに抽出し、気泡の大きさ毎の多次元速度を求めた。本手法を旋回流に対して適用した結果、気泡の旋回成分が計測され、大気泡は円筒管中心へ移動することを定量的に評価した。

    DOI CiNii

  • Visualization technique of gas-liquid two-phase flow structure in rod bundle

    Arai Takahiro, Furuya Masahiro, Kanai Taizo, Shirakawa Kenetsu

    Proceedings of National Heat Transfer Symposium   2011 ( 0 ) 306 - 306  2011

     View Summary

    管群流路内の気液二相流動を多次元的に高い時間分解能で計測可能なセンサを開発した。開発したセンサは、10×10本の管群流路において各管のギャップに11×11本のワイヤ電極ならびに100本のロッド電極により構成され、各電極間のコンダクタンス計測結果に基づきボイド率を推定する。隣接する管に囲まれた流路であるサブチャンネルの中央に対応するワイヤ同士の近接点121箇所(11×11)、および管ギャップに対応する管とワイヤの近接点400箇所(100×4)の二相流計測が可能である。水空気二相流動試験により、管群流路内の軸方向ならびに径方向ボイド混合特性をボイド率分布、相速度分布などによって定量的に評価した。

    DOI CiNii

  • ICONE19-44020 Multi-dimensional Two-Phase Flow Measurements in a Large-Diameter Pipe Using Wire-Mesh Sensor

    KANAI Taizo, FURUYA Masahiro, ARAI Takahiro, SHIRAKAWA Kenetsu, NISHI Yoshihisa, UEDA Nobuyuki

    The Proceedings of the International Conference on Nuclear Engineering (ICONE)   2011 ( 0 ) _ICONE1944 - _ICONE1944  2011

     View Summary

    The authors developed a method of measurement to determine the multi-dimensionality of two phase flow. A wire-mesh sensor (WMS) can acquire a void fraction distribution at a high temporal and spatial resolution and also estimate the velocity of a vertical rising flow by investigating the signal time-delay of the upstream WMS relative to downstream. Previously, one-dimensional velocity was estimated by using the same point of each WMS at a temporal resolution of 1.0-5.0s. The authors propose to extend this time series analysis to estimate the multi-dimensional velocity profile via cross-correlation analysis between a point of upstream WMS and multiple points downstream. Bubbles behave in various ways according to size, which is used to classify them into certain groups via wavelet analysis before cross-correlation analysis. This method was verified by air-water straight and swirl flows within a large-diameter vertical pipe. A high-speed camera is used to set the parameter of cross-correlation analysis. The results revealed that for the rising straight and swirl flows, large scale bubbles tend to move to the center, while the small bubble is pushed to the outside or sucked into the space where the large bubbles existed. Moreover, it is found that this method can estimate the rotational component of velocity of the swirl flow as well as measuring the multi-dimensional velocity vector at high temporal resolutions of 0.2s.

    DOI CiNii

  • 管群流路内における気液二相構造の可視化手法

    新井崇洋, 古谷正裕, 金井大造, 白川健悦

    日本伝熱シンポジウム講演論文集(CD-ROM)   48th   ROMBUNNO.E321  2011

    J-GLOBAL

  • Validation of TRACE Code With Type-I Density Wave Oscillations in Natural Circulation Loop

    FURUYA MASAHIRO, NISHI YOSHIHISA, UEDA NOBUYUKI

    Proceedings of National Heat Transfer Symposium   48th ( 0 ) 193 - 193  2011

     View Summary

    自然循環ループにおける密度波振動には、水頭損失が支配的なI型と、摩擦損失が支配的なII型に分類される。本研究では、自然循環沸騰水型原子炉の一つであるESBWRを摸擬したSIRIUS-N設備で観測したI型密度波振動を対象に、システム動特性解析コードTRACE ver5を用いた数値解析を行った。その結果、解析では実験で得られた静特性および安定境界を妥当な精度で予測した。また不安定現象の特徴である振動周期と通過時間の相関を精度良く予測できることから、TRACEコードが自然循環ループの安定性評価に使用することの有用性を確認した。

    DOI CiNii J-GLOBAL

  • 大口径円筒管内二相流の三次元相速度分布計測手法の開発

    金井大造, 古谷正裕, 新井崇洋, 白川健悦

    日本伝熱シンポジウム講演論文集(CD-ROM)   48th   ROMBUNNO.E322  2011

    J-GLOBAL

  • Measurements of Multidimensional Phasic-Velocity Distribution in a Large-Diameter Pipe Using Wire-Mesh Sensor

    TAIZO KANAI, MASAHIRO FURUYA, TAKAHIRO ARAI, KENETSU SHIRAKAWA, YOSHIHISA NISHI, NOBUYUKI UEDA

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2011 ( 0 ) 250 - 250  2011

     View Summary

    大口径円筒管内の二相流動は多次元性を持つため、多次元二相流計測手法は気泡流やチャーン流のモデル化において重要な計測手法である。ワイヤメッシュセンサ(WMS)はボイド率分布を高速かつ高空間分解能で計測可能である。これまで二段のWMSにより気泡のWMS通過時間差から主流方向一次元相速度計測も行われている。本研究では、大口径円筒管(内径:224mm)内二相流動をWMSにより計測した。上流側WMSのある点の信号と、下流側WMS複数点の信号との間で相互相関分析を行い、最も高い相関係数を示す方向と時間差から多次元速度ベクトルを求めた。また、小気泡は液相流動による影響が大きく、大気泡は多次元せん断流となる。多次元相速度ベクトルは気泡径により大きく異なるため、Wavelet解析を用いて気泡径毎に信号を抽出し、気泡径毎に相速度ベクトルを評価した。

    DOI CiNii J-GLOBAL

  • Electrochemical Investigation of Surface Modification Layer 'Fresh Green' on the Zircaloy Which Suppresses Corrosion and Hydrogen-Pickup

    FURUYA MASAHIRO, KITAJIMA SHOICHI, SONODA TAKESHI, SAWABE TAKASHI, TOKIWAI MORIYASU, KINOSHITA MOTOYASU

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2011 ( 0 ) 491 - 491  2011

     View Summary

    ジルコニウム合金の表面改質技術フレッシュグリーン(FG)は緻密で密着性が高い保護皮膜であることから、耐食性と耐水素吸収性が向上する。そのメカニズムを検討するため、FG皮膜の電気化学特性を試験した。基材にジルカロイ2を用い、対照試験として、FG皮膜の他に大気中で酸化した(OD)皮膜と360℃高温水中で酸化させた(AC)皮膜を用いた。EISの結果、ACとOD皮膜は利得も位相遅れもほぼ同じであるが、FG皮膜はコンダクタンスが高い二層構造から構成されていることが判明した。Mott-Schottky分析によりFG皮膜はACとOD皮膜よりドナー密度が高い。さらに自然浸漬電位が高い。これらは酸化皮膜内にカーボンがドープされていることが原因と推察される。これらの電気化学特性がFG処理被覆管材の耐食性を向上させていると考えられる。

    DOI CiNii J-GLOBAL

  • Development of Multidimensional Two-phase Flow Measurement Sensor in Fuel Bundle

    Arai Takahiro, Furuya Masahiro, Kanai Taizo, Kondo Masahiro, Shirakawa Kenetsu

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2011 ( 0 ) 248 - 248  2011

     View Summary

    10×10バンドル燃料集合体を模擬した流路内の二相流動の三次元分布を高い時間分解能で計測可能なセンサを開発した。開発したセンサは、模擬燃料棒ギャップに配置した11×11本のワイヤ電極ならびに模擬燃料棒である100本のロッド電極により構成され、各電極間のコンダクタンス計測結果に基づきボイド率を推定する。サブチャンネル中央に対応するワイヤ同士の近接点121箇所(11×11)、および燃料棒ギャップに対応する模擬燃料棒とワイヤの近接点400箇所(100×4)の二相流計測(ボイド率、気泡コード長、相速度)が可能である。水空気二相流動試験により、燃料集合体内の鉛直上昇二相流動を定量的に計測できることを検証した。

    DOI CiNii J-GLOBAL

  • 気泡追跡法による大口径管内の気泡動力学と相互干渉

    古谷正裕, 金井大造, 新井崇洋, 白川健悦, HAMPEL Uwe, SCHLEICHER Eckard

    日本混相流学会年会講演会講演論文集   2011   214 - 215  2011

    J-GLOBAL

  • バンドル流路内気液二相流発達過程の三次元計測

    新井崇洋, 古谷正裕, 金井大造, 白川健悦

    日本混相流学会年会講演会講演論文集   2011   210 - 211  2011

    J-GLOBAL

  • 大口径円筒管内二相流の三次元相速度分布計測

    金井大造, 古谷正裕, 新井崇洋, 白川健悦

    日本混相流学会年会講演会講演論文集   2011   212 - 213  2011

    J-GLOBAL

  • Raman Spectrometry of Surface Modification Layer 'Fresh Green' on the Zircaloy Which Suppresses Corrosion and Hydrogen-Pickup

    FURUYA MASAHIRO, KITAJIMA SHOICHI, SONODA TAKESHI, SAWABE TAKASHI, TOKIWAI MORIYASU, KINOSHITA MOTOYASU

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2011 ( 0 ) 462 - 462  2011

     View Summary

    ジルコニウム合金の表面改質技術フレッシュグリーン(FG)は緻密で密着性が高い保護皮膜であることから、耐食性と耐水素吸収性が向上する。そのメカニズムを検討するため、FG皮膜のラマン分光分析を行った。基材にジルカロイ2を用い、対照試験として、FG皮膜の他に大気中で酸化した(OD)皮膜と360℃高温水中で酸化させた(AC)皮膜を用いた。ラマン分光分析の結果、フレッシュグリーン皮膜では、ドープされる炭素の一部はsp2軌道を有する非晶質炭素の形態であると推定される。フレッシュグリーン皮膜表面から約0.5umの深さに於いて、Gバンドの強度が平均値の3倍以上になるピークが観測されている。フレッシュグリーン皮膜では、表面では単斜晶の割合が大きく、0.6umより深い位置から、基材との界面近傍までにおいて正方晶の割合が相対的に高くなる領域が存在している。この境界はGバンドのピークと符合している。これらの電気化学特性がFG処理被覆管材の耐食性を向上させていると考えられる。

    DOI CiNii J-GLOBAL

  • Development of subchannel void sensor acquiring three-dimensional two-phase flow in rod bundle

    ARAI TAKAHIRO, FURUYA MASAHIRO, KANAI TAIZO, SHIRAKAWA KENETSU

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2011 ( 0 ) 257 - 257  2011

     View Summary

    バンドル流路内気液二相流を高速で三次元計測可能なボイドセンサを開発した。開発したセンサは、10×10バンドル流路において隣接するロッドに囲まれたチャンネルの中央サブチャンネル121箇所(11×11)、およびロッド表面サブチャンネル400箇所(10×10×4)を複数の流路断面に対して同時計測可能である。空気-水二相流動実験により取得したボイド率分布,相速度分布,気泡コード長分布に基づき既存モデルと比較することで、バンドル内気液二相流の発達特性を評価した。

    DOI CiNii J-GLOBAL

  • 大口径円筒管内二相流の三次元計測と流動発達過程

    金井大造, 古谷正裕, 新井崇洋, 白川健悦

    日本原子力学会秋の大会予稿集(CD-ROM)   2011   ROMBUNNO.P22  2011

    J-GLOBAL

  • Influence and applicability of wire-mesh sensor to acquire two phase flow dynamics

    金井 大造, 古谷 正裕, 新井 崇洋

    電力中央研究所報告 L 研究報告   ( L10003 ) 1 - 12,巻頭1〜3  2011

    CiNii J-GLOBAL

  • Development of multidimensional two-phase flow measurement sensor in rod bundle

    新井 崇洋, 古谷 正裕, 白川 健悦

    電力中央研究所報告 L 研究報告   ( L10002 ) 1 - 13,巻頭1〜3  2011

    CiNii J-GLOBAL

  • Applicability of best-estimate analysis TRACE in terms of natural circulation BWR stability

    古谷 正裕, 植田 伸幸, 西 義久

    電力中央研究所報告 L 研究報告   ( L10005 ) 1 - 13,巻頭1〜3  2011

    CiNii J-GLOBAL

  • Development of three-dimensional phasic-velocity distribution measurement in a large-diameter pipe

    金井 大造, 古谷 正裕, 新井 崇洋

    電力中央研究所報告 L 研究報告   ( L10006 ) 1 - 17  2011

    CiNii J-GLOBAL

  • Development of Two-phase Flow Measurement Sensors and Gas-Liquid Two-phase Flow Dynamics

    FURUYA Masahiro, ARAI Takahiro, KANAI Taizo, SHIRAKAWA Kenetsu

    Journal of the Visualization Society of Japan   31 ( 122 ) 92 - 97  2011

     View Summary

    &nbsp;&nbsp;We have devised the methods to measure a liquid-film thickness distribution and three-dimensional velocity in a large diameter pipe or complex rod bundle geometry. <br>(1) A pair of an excitation electrode and a working electrode is implemented onto a three-layer printed circuit with a 1.8 mm lengthwise and crosswise apart from each other. This printed circuit can bend to form a cylindrical shape with 12 mm in diameter. The cylindrical printed circuit demonstrates liquid film perturbations in an annuls flow path. <br>(2) We have proposed the methods to estimate three-dimensional velocity profile in terms of length scale of a two-phase flow mixture on the basis of acquired data with commonly used wire-mesh sensors. We visualize a swirl flow dynamics such that a water flow moves outside and in turn gas shifts to inside of a pipe. <br>(3) We visualized 121 points of center subchannel in a 10 &times; 10 rod bundle geometry by inserting a thin wire into each rod gap. In addition, we visualized 400 points of rod gap in the combination of rods as an excitation electrode and thin wires as a measurement electrode. The devised 'subchannel void sensors' were installed in eight layers acquiring 521 points / layer at every 1 ms.

    DOI CiNii J-GLOBAL

  • Boiling Heat Transfer

    FURUYA Masahiro

      49 ( 209 ) 15 - 16  2010.10

    CiNii

  • チタンの表面改質による水処理技術の開発

    田中 伸幸, 古谷 正裕, 常磐井 守泰, 本多 孝

    化学工学 = Chemical engineering   74 ( 4 ) 171 - 173  2010.04

    CiNii

  • Safety Evaluation Based on Uncertainty and Scaling Analyses with Statistical Approach by CRIEPI

    FURUYA Masahiro, NISHI Yoshihisa

    Ατομοσ   52 ( 2 ) 86 - 90  2010.02

    CiNii

  • Corrosion Control and Anti-Hydrogen-Pickup of Zircaloy Cladding by Surface Modification Technology, 'Fresh Green':(1) Corrosion Control of Zr-Nb Alloy and Zircaloy-2

    FURUYA MASAHIRO, KITAJIMA SHOICHI, SONODA TAKESHI, SAWABE TAKASHI, TOKIWAI MORIYASU, KINOSHITA MOTOYASU

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2010 ( 0 ) 594 - 594  2010

     View Summary

    ジルカロイ-2の表面を薄いカーボンドープ酸化ジルコニウム層に表面改質するフレッシュグリーン処理を施した。ラマン分光分析を用いて分析を行った結果、フレッシュグリーン皮膜では、GバンドとDバンドに特徴的な共鳴ピークが現れ、カーボンはZrC結合以外にもsp2軌道を有するグラファイト構造でドープされていることが判明した。

    DOI CiNii

  • Corrosion Control and Anti-Hydrogen-Pickup of Zircaloy Cladding by Surface Modification Technology, 'Fresh Green':(2) Microstructure of Modified Layers

    SAWABE TAKASHI, SONODA TAKESHI, FURUYA MASAHIRO, KITAJIMA SHOICHI, KINOSHITA MOTOYASU, TOKIWAI MORIYASU

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2010 ( 0 ) 595 - 595  2010

     View Summary

    ジルカロイ-2の表面を薄いカーボンドープ酸化ジルコニウム層に表面改質するフレッシュグリーン処理を施した。SPring-8放射光施設においてマイクロX線を用いてフレッシュグリーン皮膜の結晶構造を解析した結果、皮膜の酸化ジルコニウムの結晶構造は表面から金属との界面にかけて単斜晶であり、界面近傍では正方晶であることが明らかとなった。

    DOI CiNii

  • Development of High-Speed and Multidimensional Measurement Sensor for Liquid Film Flow Dynamics:Optimization of Measurement Parameters with Electrochemical Impedance Spectrometry

    FURUYA MASAHIRO, ARAI TAKAHIRO, KANAI TAIZO

    Proceedings of National Heat Transfer Symposium   47th ( 0 ) 96 - 96  2010

     View Summary

    著者らは液膜流動を高い時間・空間分解能で計測する高密度多点電極法を開発した。センサーはn行m列の電極対で構成され、それぞれn行の電圧印加とm列の電流走査でn+m本の電極リードにまとめることで高密度化を達成している。液膜厚さの計測原理は電極間のコンダクタンス計測であるため、電圧印加時に形成される電気二重層の特性を知ることが重要である。本研究では、電気化学インピーダンス法により等価回路を求め、電気二重層の静電容量や溶液抵抗などを推定した。センサーの過渡応答により計測速度の制限値などを見出した。

    DOI CiNii J-GLOBAL

  • Evaporation and Quenching Characteristics of Nanofluid and Salt Solusion

    ARAI Takahiro, FURUYA Masahiro

    Proceedings of National Heat Transfer Symposium   47th ( 0 ) 91 - 91  2010

     View Summary

    高温ステンレス球をナノ流体および塩水に浸漬するクエンチ実験を行った。塩水では水溶液濃度の増大に伴いクエンチ温度が上昇したが、ナノ流体ではその傾向は顕著に見られなかった。この違いを考察するために、常温常圧で気液界面からの蒸発速度を観測する実験を行った。その結果、塩水では濃度の増大に伴い蒸発速度が減少したが、ナノ流体では水とほぼ変わらない値であった。よって、ナノ流体と塩水とのクエンチ温度の変化は蒸発速度と相関があることが明らかになった。

    DOI CiNii J-GLOBAL

  • Applicability of Wire-Mesh Sensor to Acquire Bubbly Flow Dynamics

    Taizo KANAI, Masahiro FURUYA, Takahiro ARAI

    Proceedings of National Heat Transfer Symposium   47th ( 0 ) 95 - 95  2010

     View Summary

    ワイヤメッシュセンサでは断面のボイド率分布や流速分布を連続的に計測できるが、侵襲法であるため二相流構造に影響を及ぼす。本研究では内径50mmの円管を用い、気泡上昇の可視画像と対比することにより、様々な気泡径および気泡形状に対してワイヤメッシュセンサが界面変形に与える影響を明らかにした。また、気泡上昇速度の計測精度を気泡径および径方向測定位置の関数として表し、二相流計測手法としてのワイヤメッシュセンサの適用性を把握した。

    DOI CiNii J-GLOBAL

  • 炭素ドープ酸化チタンを用いた人工関節用摺動部材の開発

    藤原邦彦, 植月啓太, 石坂春彦, 土居憲次, 藏本孝一, 田中伸幸, 古谷正裕, 常盤井守泰

    日本人工関節学会プログラム・抄録集   40th   337  2010

    J-GLOBAL

  • 我が国の最先端原子力研究開発 第17回 統計的安全評価手法に関する電力中央研究所の取組み

    古谷正裕, 西義久

    ΑΤΟΜΟΣ   52 ( 2 ) 86 - 90  2010

    DOI J-GLOBAL

  • Development of high-speed and multidimentional measurement method for liquid film behavior

    Arai Takahiro, Furuya Masahiro

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2010 ( 0 ) 257 - 257  2010

     View Summary

    液膜流動を高い時間・空間分解能で計測する手法を開発し、液膜厚さの過渡変化が計測可能であることを示した。本手法は、曲率を有する構造体として燃料被覆管を模擬した体系への適用も可能である。

    DOI CiNii J-GLOBAL

  • 持続可能社会への化学工学の貢献 チタンの表面改質による水処理技術の開発

    田中伸幸, 古谷正裕, 常磐井守泰, 本多孝

    化学工学   74 ( 4 ) 171 - 173  2010

    J-GLOBAL

  • 高密度多点電極法による液膜厚さ計測技術の開発

    新井崇洋, 古谷正裕, 金井大造

    電力中央研究所原子力技術研究所研究報告   ( L09008 ) 1 - 11,巻頭1〜3  2010

    CiNii J-GLOBAL

  • フレッシュグリーン表面改質技術によるジルカロイ被覆管の耐食性・耐水素吸収向上(1)Zr‐Nb合金およびジルコニウムの耐食性

    古谷正裕, 北島庄一, 園田健, 澤部孝史, 常磐井守泰, 木下幹康

    日本原子力学会秋の大会予稿集(CD-ROM)   2010   ROMBUNNO.G13  2010

    J-GLOBAL

  • フレッシュグリーン表面改質技術によるジルカロイ被覆管の耐食性・耐水素吸収特性向上(2)皮膜部の結晶構造解析

    澤部孝史, 園田健, 古谷正裕, 北島庄一, 木下幹康, 常磐井守泰

    日本原子力学会秋の大会予稿集(CD-ROM)   2010   ROMBUNNO.G14  2010

    J-GLOBAL

  • Vertical Upward Bubbly Flow Mechanism in a Large Vertical Pipe:(1) Measurement of Bubbly Flow Dynamics by Wire-Mesh Sensor

    TAIZO KANAI, MASAHIRO FURUYA, TAKAHIRO ARAI, MASAHIRO KONDO, KENETSU SHIRAKAWA

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2010 ( 0 ) 343 - 343  2010

     View Summary

    ワイヤメッシュセンサ(WMS)は断面ボイド率分布や界面移動速度分布を高速に計測できる。本研究では、内径224mmの配管内鉛直上昇流におけるWMSの計測精度を高速度カメラ等との比較により検証した。

    DOI CiNii J-GLOBAL

  • Measurement of liquid film distribution on simulated fuel rod surface

    ARAI Takahiro, FURUYA Masahiro, KANAI Taizou, SHIRAKAWA Kenetsu, KONDO Masahiro

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2010 ( 0 ) 349 - 349  2010

     View Summary

    燃料棒を模擬した円柱形状表面の液膜流動を2次元分布として高い時間分解能で計測可能な高密度多点電極センサを開発した。開発したセンサを用いて水空気による液膜流動試験を実施することにより、模擬燃料棒表面における垂直上昇液膜流動を計測できることを示した。

    DOI CiNii J-GLOBAL

  • Vertical Upward Bubbly Flow Mechanism in a Large Vertical Pipe:(2) Validation of Multi-Dimensional Air-Water Two-Phase Flow CFD Code

    KONDO Masahiro, FURUYA Masahiro, KANAI Taizou, ARAI Takahiro, SHIRAKAWA Kenetsu

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2010 ( 0 ) 344 - 344  2010

     View Summary

    内径224mmの円筒管内垂直上昇気液二相流を対象に数値流体解析(CFD)を行った。CFDには汎用流体解析コードを用いた。実験で得られたボイド率や気泡上昇速度の径方向分布を解析結果と比較して精度を検証した。また、二相流相関式のパラメータの感度を検討した。

    DOI CiNii J-GLOBAL

  • 先進オリエントサイクル研究構想の進展を―使用済み燃料からレアメタル・レアアースの分離・回収 被覆管材料などへの利用

    常磐井守泰, 古谷正裕

    原子力eye   56 ( 10 ) 29 - 31  2010

    J-GLOBAL

  • Friction Property for Carbon-doped Titanium Oxide Film on a Titanium Material Surface

    Horiuch Tatsuya, Miyaji Yoshiaki, Kurisu Hiroki, Yamamoto Setsuo, Furuya Masahiro, Goto Minoru

    Abstract of annual meeting of the Surface Science of Japan   30th ( 0 ) 396 - 396  2010

     View Summary

    炭素ドープと酸化処理を同時に施す新たな表面改質処理(フレッシュグリーン処理)を施したチタンのガス放出特性と真空中の摩擦特性について調べた結果を報告する.

    DOI CiNii J-GLOBAL

  • Surface Modification Technology, Fresh Green, to mitigate Corrosion of Titanium and Zirconium

    田中伸幸, 古谷正裕, 常磐井守泰

    チタン   58 ( 2 ) 128 - 131  2010

    CiNii J-GLOBAL

  • Effect of Nanoparticle Suspension on Vapor Film Collapse

    新井崇洋, 古谷正裕

    日本伝熱シンポジウム講演論文集(CD-ROM)   46th ( 0 ) 1 - 1  2009

     View Summary

    直径30mmの高温固体ステンレス球を、ナノ粒子を懸濁させた水、いわゆるナノ流体中に浸漬させ、膜沸騰蒸気膜の崩壊挙動を観測した。固体球表面に埋め込んだ熱電対による固体球表面温度を計測および膜沸騰蒸気膜挙動の可視観測を実施した。各沸騰様相における固体球と水の接触を検出するために、ナノ流体と固体球に電極を挿入し、固体球とナノ流体の間を流れる電流を計測した。ナノ粒子の種類および濃度を変化させることによってナノ粒子懸濁の影響を評価した。

    DOI CiNii J-GLOBAL

  • Transient Response of BWR Core Flow during Simulated Power and Flow Decrease Events

    古谷正裕, 原貴, 溝上伸也

    日本伝熱シンポジウム講演論文集(CD-ROM)   46th ( 0 ) 180 - 180  2009

     View Summary

    沸騰水型原子炉(BWR)の炉内流動を模擬した試験設備SIRIUS-Fを用いて、出力および流量が減少する過渡事象を模擬する実験を行った。実験の結果、出力の異なる並行2チャンネルの流量配分は、初期は出力が高いチャンネルの方が質量流束が低い。しかしながらボイド率がある程度低減すると、質量流束が逆転する。BWR過渡事象解析コードTRAC-BF1では、この質量流束逆転現象を精度良く再現した。このような出力・流量急減事象においても、同コードはBWR炉内の沸騰二相流を精度良く予測できることを確認した。

    DOI CiNii J-GLOBAL

  • 放射線誘起表面活性に関する研究

    賞雅寛而, 波津久達也, 福原豊, 嘉村明彦, 西大路隆司, 熊田崇徳, 班目春樹, 岡本孝司, 阿部弘亨, 間淵幸雄, 仲川勉, 吉廻智江, 早川和寿, 古谷正裕, 白鳥義行

    UTNL-R(東京大学大学院工学系研究科原子力専攻)   ( 0476 ) 23 - 24  2009

    J-GLOBAL

  • チタンの表面改質技術「フレッシュグリーン」による水浄化技術の開発

    田中伸幸, 古谷正裕, 常磐井守泰, 堀江正明, 本多孝

    化学工学会年会研究発表講演要旨集(CD-ROM)   74th ( 0 ) 586 - 586  2009

    DOI CiNii J-GLOBAL

  • Vapor film collapse caused by filament growth of molten copper and triggering of vapor explosion

    新井崇洋, 古谷正裕

    日本機械学会関東支部総会講演会講演論文集   15th   115 - 116  2009

    J-GLOBAL

  • Visual Observation of Fine-scale Mixing Morphology during Vapor Explosion and Droplet Entrapping Processes

    古谷正裕, 新井崇洋

    日本機械学会関東支部総会講演会講演論文集   15th   117 - 118  2009

    J-GLOBAL

  • Study of Mass Transfer Effect on Flow Accelerated Corrosion

    米田公俊, 稲田文夫, 森田良, 古谷正裕

    日本機械学会論文集 B編   75 ( 751 ) 427 - 428  2009

     View Summary

    Flow Accelerated Corrosion (FAC) requires considerable attention in plant piping management, for its potential of catastrophic pipe rupture of main piping systems. In view of fluid dynamics, the most essential factor to be considered is mass transfer at the inner surface of the pipe. Mass transfer coefficients are determined by fluid properties and piping geometry, however, no universal correlation exists, which is adaptable to various types of piping elements with strong turbulence. In this study, the modeling of mass transfer coefficient was progressed based on Chilton-Colburn analogy and utilizing "effective friction velocity" from the hydraulics in the viscous sub-layer along the wall. FAC experiments with PWR condensate water condition and CFD for the flow were conducted with a contracted rectangular duct. By considering the turbulent velocity of the viscous layer into the mass transfer coefficient, the correlation with the FAC thinning rate improved, effectively.

    DOI CiNii J-GLOBAL

  • 膜沸騰蒸気膜のクエンチ特性に及ぼす塩水およびナノ流体の影響

    新井崇洋, 古谷正裕

    日本機械学会動力・エネルギー技術シンポジウム講演論文集   14th   447 - 448  2009

    J-GLOBAL

  • チタンの表面改質「フレッシュグリーン」による水処理技術の開発

    田中伸幸, 古谷正裕, 常磐井守泰, 堀江正明, 本多孝

    化学工学会秋季大会研究発表講演要旨集(CD-ROM)   41st ( 0 ) 168 - 168  2009

    DOI CiNii J-GLOBAL

  • Effect of Thermophisical Properties of Coolant on Vapor Film Collapse Based on Film Boiling Stability Model

    Arai Takahiro, Furuya Masahiro

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2009 ( 0 ) 312 - 312  2009

     View Summary

    既存の高温固体球浸漬実験から得られた知見を基に、膜沸騰蒸気膜の線形安定性解析の定数を定め、熱伝達係数や熱物性値が蒸気膜崩壊条件に及ぼす影響を考察した。

    DOI CiNii J-GLOBAL

  • Corrosion Control and Anti-Hydrogen-Pickup of Zircaloy-2 Cladding by Surface Modification Technology, 'Fresh Green'

    FURUYA MASAHIRO, TOKIWAI MORIYASU, KITAJIMA SHOICHI, ITO KUNIO, ETOH YOSHINORI, AOMI MASAKI, SAKAMOTO KAN

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2009 ( 0 ) 486 - 486  2009

     View Summary

    ジルカロイ-2の表面を薄いカーボンドープ酸化ジルコニウム層に表面改質するフレッシュグリーン処理を施した。360℃飽和水中に153日間浸漬した結果、腐食増量および水素吸収量が低減する結果を得た。

    DOI CiNii J-GLOBAL

  • 高性能光触媒「フレッシュグリーン」による水浄化技術の開発

    田中伸幸, 古谷正裕, 常磐井守泰, 本多孝

    環境技術学会研究発表大会及び特別講演会予稿集   9th   63 - 64  2009

    J-GLOBAL

  • Development of MIG Brazing Method of Titanium to Stainless Steel

    古谷正裕, 常磐井守泰, 田中伸幸, 堀江正明

    日本機械学会年次大会講演論文集   2009 ( Vol.6 ) 321 - 322  2009

    J-GLOBAL

  • Effect of Salt Additives on Film Boiling Heat Transfer and Mechanism of Quenching Temperature Rise

    新井崇洋, 古谷正裕

    日本機械学会論文集 B編   75 ( 758 ) 1932 - 1938  2009

     View Summary

    A high-temperature stainless-steel sphere was immersed into various salt solutions to test film boiling behavior at vapor film collapse. The film boiling behavior around the sphere was observed with a digital-video camera. Because salt additives enhanced condensation heat transfer, the observed vapor film was thinner. Surface temperature of the sphere was measured. Salt additives increased the quenching (vapor film collapse) temperature, because frequency of direct contact between sphere surface and coolant increased. Quenching temperature rises with increased salt concentration. The quenching temperature is well correlated with ion molar concentration, which is a number density of ions, regardless of the type of hydrated salts.

    DOI CiNii J-GLOBAL

  • Mechanism of Hydrophilicity by Radiation-Induced Surface Activation

    HONJO Yoshio, FURUYA Masahiro, TAKAMASA Tomoji, OKAMOTO Koji

    Journal of Power and Energy Systems   3 ( 1 ) 216 - 227  2009

     View Summary

    When a metal oxide is irradiated by gamma rays, the irradiated surface becomes hydrophilic. This surface phenomenon is called as radiation-induced surface activation (RISA) hydrophilicity. In order to investigate gamma ray-induced and photoinduced hydrophilicity, the contact angles of water droplets on a titanium dioxide surface were measured in terms of irradiation intensity and time for gamma rays of cobalt-60 and for ultraviolet rays. Reciprocals of the contact angles increased in proportion to the irradiation time before the contact angles reached its super-hydrophilic state. The irradiation time dependency is equal to each other qualitatively. In addition, an effect of ambient gas was investigated. In pure argon gas, the contact angle remains the same against the irradiation time. This clearly indicates that certain humidity is required in ambient gas to take place of RISA hydrophilicity. A single crystal titanium dioxide (100) surface was analyzed by X-ray photoelectron spectrometry (XPS). After irradiation with gamma rays, a peak was found in the O1s spectrum, which indicates the adsorption of dissociative water to a surface 5-fold coordinate titanium site, and the formation of a surface hydroxyl group. We conclude that the RISA hydrophilicity is caused by chemisorption of the hydroxyl group on the surface.

    DOI CiNii

  • 10511 Vapor film collapse caused by filament growth of molten copper and triggering of vapor explosion

    ARAI Takahiro, FURUYA Masahiro

    The Proceedings of Conference of Kanto Branch   2009 ( 15 ) 115 - 116  2009

     View Summary

    Experiments were conducted in which a molten copper droplet released into water pool. Spontaneous vapor explosion did not occur when water temperature was over 50℃. Spontaneous vapor explosion did, however, occur when water temperature was lower than 50℃. A high-speed video frames explored the stages of vapor explosions: (1) a vapor film was formed and separates the copper droplet and surrounding water, (2) a filament of molten copper grew from the surface and deformed the vapor film, (3) the vapor film collapsed along the filament surface, and finally (4) triggering of vapor explosions occurred from the filament to the whole molten copper droplet. When the filament growth was observed, it triggered the vapor explosion in almost all .cases. When.. not, . vapor explosion was not observed and the vapor film was, therefore, stably formed around the molten copper droplet. We concluded that the filament form the molten copper triggered vapor explosion in a highly subcooled water.

    DOI CiNii

  • 10512 Visual Observation of Fine-scale Mixing Morphology during Vapor Explosion and Droplet Entrapping Processes

    FURUYA Masahiro, ARAI Takahiro

    The Proceedings of Conference of Kanto Branch   2009 ( 15 ) 117 - 118  2009

     View Summary

    The fine-scale mixing morphology of vapor explosion and droplet entrapping was visualized by high-speed digital video cameras at an interface between a water droplet and a molten tin pool. The cluster of bubble generated by spontaneous bubble-nucleation covered the whole contact area at 0.1ms after the impingement. Prominent fine mixing between two liquids were found to start at 0.6ms that resulting in vapor explosion. Droplet entrapping phenomenon frequently occurred on an oxide layer, since coherent mixing was prevented due to unevenly formed oxide layer.

    DOI CiNii

  • E210 Effect of salt solutions and nanofluids on quenching characteristics of vapor film

    ARAI Takahiro, FURUYA Masahiro

    The Proceedings of the National Symposium on Power and Energy Systems   2009 ( 14 ) 447 - 448  2009

     View Summary

    A high-temperature stainless-steel sphere was immersed into an A12O3 nanofluid to investigate film boiling heat transfer and collapse of vapor film. Surface temperature is referred to the measured value of thermocouples embedded into and welded onto bottom surface of the sphere. The Al_2O_3 nanofluid concentration is varied from 0 to 15 wt%. Comparing between Al_2O_3 nanofluid and salt solution, The film boiling heat transfer coefficient and quenching temperature for Al_2O_3 nanofluid remain the same within the experimental Al_2O_3 nanofluid concentration. In contrast to the nanofluid, those values for salt solution significantly increased with increasing concentration of the salt solution.

    DOI CiNii

  • 放射線誘起表面活性(RISA)を用いた船舶・海洋構造物の耐食材防食技術に関する基礎研究II : RISAによるすきま腐食抑制メカニズム(所外発表論文等概要)

    植松 潤一, 波津久 達也, 元田 慎一, 賞雅 寛而, 植松 進, 古谷 正裕

    海上技術安全研究所報告   8 ( 4 ) 437 - 437  2009

    CiNii

  • Feasibility of Long-Life and Corrosion-Resistant Canister with MIG Brazing of Titanium to Stainless Steel

    古谷正裕, 常磐井守泰, 田中伸幸, 堀江正明

    チタン   57 ( 2 ) 89 - 96  2009

    CiNii J-GLOBAL

  • Corrosion control and anti-hydrogen-absorption of fuel cladding by surface modification technology, fresh green

    古谷 正裕, 常磐井 守泰, 田中 伸幸

    電力中央研究所報告 L 研究報告   ( L08014 ) 1 - 14,巻頭1〜3  2009

    CiNii J-GLOBAL

  • J0402-2-1 Development of MIG Brazing Method of Titanium to Stainless Steel

    FURUYA Masahiro, TOKIWAI Moriyasu, TANAKA Nobuyuki, Horie Masaaki

    The proceedings of the JSME annual meeting   2009 ( 6 ) 321 - 322  2009

     View Summary

    In order to store nuclear spent fuels for a long term, we propose the concept of stainless steel canister with titanium cladding. A parametric study on MIG brazing for titanium and stainless steel revealed that Cu-1Mn-3Si alloy (MG960) is promising MIG brazing material. The optimized brazing conditions showed adequate structural strength such as JIS G 0601 shear strength, tensile shear stress and peel strength.

    DOI CiNii

  • Effect of Oxide Film Type on Crevice Corrosion Protection Caused by

    K Akihiko, Hazuku Tatsuya, Motoda Shinichi, Takamasa Tomoji, Uematsu Susumu, Furuya Masahiro

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2008 ( 0 ) 643 - 643  2008

     View Summary

    SUS304鋼上にAlO3溶射被膜やプラズマ溶射被膜を成膜した試験片にRI放射線源によりγ線を照射しながら、この試験片の腐食溶液中での母材の溶出量および電気化学的変化を計測した。これらの条件で、放射線誘起表面活性によるSUS304鋼の隙間腐食抑制効果を確認した。

    DOI CiNii

  • Effects of Salt Precipitation and Particle Suspension on Vapor Film Collapse

    新井崇洋, 古谷正裕

    日本伝熱シンポジウム講演論文集(CD-ROM)   45th ( 0 ) 27 - 27  2008

     View Summary

    塩の析出および粒子懸濁が膜沸騰蒸気膜の崩壊挙動に及ぼす影響を明らかにするため、水溶液中に直径30mmの高温固体ステンレス球を浸漬させる蒸気膜崩壊実験を実施した。固体球表面温度の計測ならびに膜沸騰蒸気膜挙動の可視観測を実施した。各沸騰様相における固液接触による固体球と水の接触を検出するために、水溶液と固体球に電極を挿入し、固体球と水溶液の間を流れる電流を計測した。塩の析出の影響については、あらかじめ塩を析出させた固体球を用いた蒸気膜崩壊実験により評価し、粒子懸濁の影響については、懸濁水溶液を用いた蒸気膜崩壊実験により評価した。

    DOI CiNii J-GLOBAL

  • 放射線誘起表面活性に関する研究

    賞雅寛而, 波津久達也, 福原豊, 菊池貴好, 嘉村明彦, 班目春樹, 岡本孝司, 阿部弘亨, 友澤秀征, 間淵幸雄, 仲川勉, 吉廻智江, 早川和寿, 古谷正裕, 白鳥義行

    UTNL-R(東京大学大学院工学系研究科原子力専攻)   ( 0473 ) 19 - 20  2008

    J-GLOBAL

  • Behavior of Vapor Film Collapse and Hydrated Salt Precipitation in Hydrated Salt Solution

    Arai Takahiro, Furuya Masahiro

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2008 ( 0 ) 332 - 332  2008

     View Summary

    800℃に加熱した高温固体ステンレス球を80℃の水和塩水溶液に浸漬する実験を実施し、固体球周囲に形成される膜沸騰蒸気膜挙動および固体球表面への水和塩析出挙動を可視観測した。塩が析出しやすい水溶液および塩が析出しにくい水溶液に同様の実験を実施し、膜沸騰蒸気膜の崩壊挙動と水和塩析出挙動の相関を明らかにした。

    DOI CiNii J-GLOBAL

  • Transient Boiling Two-Phase Flow with a sudden decrease of flow rate

    FURUYA MASAHIRO, HARA TAKASHI, MIZOKAMI SHINYA, ONO HIROSHI

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2008 ( 0 ) 331 - 331  2008

     View Summary

    SIRIUS-F設備を用いて、BWRの流量急減事象を模擬した実験を行った。流量および出力が急減する場合においても、出力の異なる2並行流路(模擬炉心チャンネル)の出入口間の差圧は同じ値を示した。実験で得られた流量や差圧等の沸騰二相流挙動は、過渡時においても数値解析と良い一致を見せた。

    DOI CiNii J-GLOBAL

  • Anti Corrosion by Radiation Induced Surface Activation in Porous Films

    Honjo Yoshio, Furuya Masahiro, Okamoto Koji, Takamasa Tomoji, Yasunaga Tatsuya, Fujisawa kyosuke

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2008 ( 0 ) 483 - 483  2008

     View Summary

    多孔質皮膜としてジルコニア熔射皮膜を鉄母材に成膜し、0.1%NaCl水溶液中でコバルト60からのγ線を照射して防食性能を調べた。溶出した鉄の量からγ線照射による防食効果が認められた。

    DOI CiNii J-GLOBAL

  • A Consideration on Pipe-Wall Thinning Mechanisms from an Aspect of Fluid-Mechanics

    稲田文夫, 米田公俊, 森田良, 藤原和俊, 古谷正裕

    材料と環境   57 ( 5 ) 218 - 223  2008

     View Summary

    The contribution of the fluid mechanics to the piping wall thinning phenomena was investigated. It was shown that the fluid force to the wall was quite different between flow accelerated corrosion (FAC) and erosion. The turbulent mass transfer, which is one of the primary factors of FAC, was analogous to the turbulent heat transfer. The model that the molecular transport in the viscous sublayer nearby soon of wall was predominant was practicable. In addition, the mass transport was predicted using commercial codes of computational fluid dynamics. Some prediction results of the mass transfer in orifice and the elbow using above techniques were explained.

    DOI CiNii J-GLOBAL

  • Study of Mass Transfer Effect on Flow Accelerated Corrosion

    米田公俊, 稲田文夫, 森田良, 藤原和俊, 古谷正裕

    日本機械学会動力・エネルギー技術シンポジウム講演論文集   13th   21 - 22  2008

    J-GLOBAL

  • 流量急減事象におけるBWR炉内流動応答

    古谷正裕, 原貴, 溝上伸也, 本谷朗

    日本混相流学会年会講演会講演論文集   2008   102 - 103  2008

    J-GLOBAL

  • Corrosion control by Radiation-Induced Surface Activation

    HONJO Yoshio, FURUYA Masahiro, TAKAMASA Tomoji, OKAMOTO Koji, FUJIKAWA Seigo

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2008 ( 0 ) 642 - 642  2008

     View Summary

    鉄母材上にジルコニア溶射皮膜を成膜した試験片にγ線を照射しながら、この試験片の腐食溶液中での母材の溶出量および電気化学的変化を計測した。このとき、放射線誘起表面活性(RISA)によると考えられる効果が見いだされた。この結果からRISA効果の発現メカニズムについて新たな知見を見いだした。

    DOI CiNii J-GLOBAL

  • Correlation of Vapor Film Collapse Temperature in Terms of Hydrated Salt Solutions and Their Concentrations

    Arai Takahiro, Furuya Masahiro

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2008 ( 0 ) 360 - 360  2008

     View Summary

    水溶液中に直径30mmの高温固体ステンレス球を浸漬させる蒸気膜崩壊実験を実施した。固体球表面温度の計測ならびに膜沸騰蒸気膜挙動の可視観測を実施した結果、蒸気膜崩壊温度に及ぼす水溶液種類および濃度の影響を、水溶液中のイオン数密度を表すイオンモル濃度によって統一的に整理できることを示した。

    DOI CiNii J-GLOBAL

  • Improvement of Hydrogen Absorption of Zircaloy-2 Cladding by Surface Modification Technology, 'Fresh Green'

    TOKIWAI MORIYASU, FURUYA MASAHIRO, TANAKA NOBUYUKI, HORIE MASAAKI

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2008 ( 0 ) 580 - 580  2008

     View Summary

    ジルカロイ-2の表面を薄いカーボンドープ酸化ジルコニウム層で覆う表面改質処理(フレッシュグリーン処理)を施した。高温蒸気中における水素吸収の改善効果を調べた結果、水素吸収量が大幅に低下するとの結果を得た。

    DOI CiNii J-GLOBAL

  • 酸化被膜の種類が放射線誘起表面活性(RISA)におよぼすすきま腐食抑制効果の影響

    嘉村明彦, 波津久達也, 元田慎一, 賞雅寛而, 植松進, 古谷正裕

    日本原子力学会秋の大会予稿集(CD-ROM)   2008   K11  2008

    J-GLOBAL

  • Improvement of Corrosion Resistance of Titanium by Surface Modification Technology, 'Fresh Green'

    TANAKA NOBUYUKI, TOKIWAI MORIYASU, FURUYA MASAHIRO, HORIE MASAAKI

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2008 ( 0 ) 573 - 573  2008

     View Summary

    チタンの表面を薄いカーボンドープ酸化チタン層で覆う表面改質処理(フレッシュグリーン処理)を施した。耐食性の改善効果を調べた結果、耐食性が問題とされていた1%沸騰塩酸や30%沸騰ギ酸などに対して、全面腐食耐性が大きく改善された。

    DOI CiNii J-GLOBAL

  • Development of Surface Modification Techonology, 'Fresh Green' for Titanium, Zirconium and Hafnium

    FURUYA MASAHIRO, TANAKA NOBUYUKI, TOKIWAI MORIYASU, HORIE MASAAKI

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2008 ( 0 ) 572 - 572  2008

     View Summary

    4A族であるチタン、ジルコニウム、およびハフニウムの表面を改質処理する技術(フレッシュグリーンと命名)を開発した。この処理によって、表面は薄い酸化物皮膜で覆われるが、その皮膜には炭素が含まれる。このカーボンドープ酸化物皮膜は緻密で密着力が強く、高い硬度、耐食性や耐摩耗性が発現する。

    DOI CiNii J-GLOBAL

  • Fundamental Study of Corrosion Control in Marine and Offshore Structures Using Radiation Induced Surface Activation (RISA)-2nd Report: Mechanism behind Stainless Steel Durability Due to RISA Against Crevice Control

    植松潤一, 波津久達也, 元田慎一, 賞雅寛而, 植松進, 古谷正裕

    マリンエンジニアリング   43 ( 5 ) 803 - 808  2008

     View Summary

    This study examines a corrosion control technique for corrosion-resistant materials or of stainless steel. This employs an effect of Radiation Induced Surface Activation (RISA) . The experimental results revealed: (1) The mechanism behind the corrosion control proposed by the previous report was confirmed to be appropriate. This via tests that measured the amount of dissolved oxygen and iron ions, in the solution. (2) The corrosion control technique was confirmed to be useful for stainless steel with any kind of metal oxide film coating on the surface. (3) It was also shown to be useful even in actual seawater, due to biological effects, which is a far more severe environment for corrosion control than simple salt water. The corrosion control technique for corrosion-resistant material using RISA in seawater has therefore been shown to offer a significant potential for practical applications in naval architecture and marine structures.

    DOI CiNii J-GLOBAL

  • Stability Estimation of ABWR on the Basis of Noise Analysis

    FURUYA Masahiro, FUKAHORI Takanori, MIZOKAMI Shinya, YOKOYA Jun

    Journal of Power and Energy Systems   2 ( 1 ) 421 - 434  2008

     View Summary

    In order to investigate the stability of a nuclear reactor core with an oxide mixture of uranium and plutonium (MOX) fuel installed, channel stability and regional stability tests were conducted with the SIRIUS-F facility. The SIRIUS-F facility was designed and constructed to provide a highly accurate simulation of thermal-hydraulic (channel) instabilities and coupled thermalhydraulics-neutronics instabilities of the Advanced Boiling Water Reactors (ABWRs). A real-time simulation was performed by modal point kinetics of reactor neutronics and fuel-rod thermal conduction on the basis of a measured void fraction in a reactor core section of the facility.<br>A time series analysis was performed to calculate decay ratio and resonance frequency from a dominant pole of a transfer function by applying auto regressive (AR) methods to the time-series of the core inlet flow rate. Experiments were conducted with the SIRIUS-F facility, which simulates ABWR with MOX fuel installed. The variations in the decay ratio and resonance frequency among the five common AR methods are within 0.03 and 0.01 Hz, respectively. In this system, the appropriate decay ratio and resonance frequency can be estimated on the basis of the Yule-Walker method with the model order of 30.

    DOI CiNii

  • Availability of vapor explosion promoters based on quantification of vapor film collapse enhancement

    新井 崇洋, 古谷 正裕

    電力中央研究所報告 L 研究報告   ( L07008 ) 1 - 13,巻頭1〜3  2008

    CiNii J-GLOBAL

  • Transient response of BWR core flow during simulated rapid flow decrease events

    古谷 正裕, 原 貴, 溝上 伸也

    電力中央研究所報告 L 研究報告   ( L07006 ) 1 - 16,巻頭1〜3  2008

    CiNii J-GLOBAL

  • Feasibility of long-life and corrosion-resistant canister with titanium cladding

    古谷 正裕, 常磐井 守泰, 三枝 利有

    電力中央研究所報告 L 研究報告   ( L07013 ) 1 - 22,巻頭1〜3  2008

    CiNii J-GLOBAL

  • Stability Estimation of BWR-5 Plant with SIRIUS-F Facility

    FURUYA Masahiro, FUKAHORI Takanori, MIZOKAMI Shinya, YOKOYA Jun

    Progress in Multiphase Flow Research   3 ( 2008 ) 111 - 124  2008

     View Summary

    Channel stability, core-wide stability, and regional stability experiments were conducted for a wide range of operating conditions of a BWR-5 with UOX (uranium dioxide) type 9&times;9A fuels installed, including maximum power points along the minimum pump speed line and the natural circulation line. The SIRIUS-F facility, used in this study, was designed and constructed for highly accurate simulation of void-reactivity feedback of BWRs. The decay ratios and the resonance frequencies are in good agreement with those from the design analysis code, ODYSY. The SIRIUS-F experimental results demonstrated stability characteristics such as a stabilizing effect of the power, and revealed a sufficiently large stability margin even under hypothetical conditions of power enlargement.

    DOI CiNii J-GLOBAL

  • A110 Study of Mass Transfer Effect on Flow Accelerated Corrosion

    YONEDA Kimitoshi, INADA Fumio, MORITA Ryo, FUJIWARA Kazutoshi, FURUYA Masahiro

    The Proceedings of the National Symposium on Power and Energy Systems   2008 ( 13 ) 21 - 22  2008

     View Summary

    Flow Accelerated Corrosion (FAC) requires considerable attention in plant piping management, for its potential of catastrophic pipe rupture of main piping systems. In view of fluid dynamics, the most essential factor to be considered is mass transfer at the inner surface of the pipe. Mass transfer coefficients are determined by fluid properties and piping geometry, however, no universal correlation exists, which is adaptable to various types of piping elements with strong turbulence. In this study, the modeling of mass transfer coefficient was progressed based on Chilton-Colburn analogy and utilizing "Effective Friction velocity" from the hydraulics in the viscous sub-layer along the wall. FAC experiments with PWR condensate water condition and CFD for the flow were conducted with a contracted rectangular duct. By considering the turbulent velocity of the viscous layer into the mass transfer coefficient, the correlation with the FAC thinning rate improved, effectively.

    DOI CiNii

  • ICONE15-10382 STABILITY ESTIMATION OF ABWR ON THE BASIS OF NOISE ANALYSIS

    Furuya Masahiro, Fukahori Takanori, Mizokami Shinya, Yokoya Jun

    The Proceedings of the International Conference on Nuclear Engineering (ICONE)   2007 ( 0 ) _ICONE1510 - _ICONE1510  2007

     View Summary

    In order to investigate the stability of a nuclear reactor core with an oxide mixture of uranium and plutonium (MOX) fuel installed, channel and regional stability tests were conducted with the SIRIUS-F facility. The SIRIUSF facility was designed and constructed to provide a highly accurate simulation of thermal-hydraulic (channel) instabilities and coupled thermalhydraulics-neutronics instabilities of the Advanced Boiling Water Reactors (ABWRs). A realtime simulation was performed by modal point kinetics of reactor neutronics and fuel-rod thermal conduction on the basis of a measured void fraction in a reactor core section of the facility. A time series analysis was performed to calculate decay ratio and resonance frequency from a dominant pole of a transfer function by applying AR methods to the time-series of the core inlet flow rate. Experiments were conducted with the SIRIUS-F facility, which simulates ABWR with MOX fuel installed. The variations in the decay ratio and resonance frequency among the five common AR methods are within 0.03 and 0.01 Hz, respectively. In this system, the appropriate decay ratio and resonance frequency can be estimated on the basis of the Yule-Walker method with the model order of 30.

    DOI CiNii

  • Stability Estimation of ABWR Core with MOX Fuels by Neutronic-Thermalhydraulic Coupling Stability Facility, SIRIUS-F

    古谷正裕, 深堀貴憲, 溝上伸也, 横谷淳

    日本伝熱シンポジウム講演論文集(CD-ROM)   44th   A331  2007

    J-GLOBAL

  • Effects of Sort and Concentration of Hydrated Salts on Quenching Temperature

    新井崇洋, 古谷正裕

    日本伝熱シンポジウム講演論文集(CD-ROM)   44th   A242  2007

    J-GLOBAL

  • Effect of vapor explosion promotor on particle size distribution

    ARAI Takahiro, FURUYA Masahiro

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2007 ( 0 ) 264 - 264  2007

     View Summary

    1500℃の溶融錫を内径2mmのノズルから常温の冷却材プール中に連続滴下させる実験を、冷却材として水ならびに5、10、20wt%塩化カルシウム水溶液を用いて実施した。その結果、水の場合には安定した膜沸騰形成のためにほとんど蒸気爆発することなく固化に至ったが、塩化カルシウム水溶液の場合には水溶液濃度の上昇とともに蒸気爆発頻度が増大した。得られた固化物の粒径分布を比較すると、水の場合には9割以上が直径1mm以上であったのに対して、20wt%塩化カルシウム水溶液の場合には8割以上が直径1mm以下となり、上限付対数正規分布とよい一致がみられた。

    DOI CiNii J-GLOBAL

  • 自然循環BWR炉の熱的性能,安全性に関する試験

    古谷正裕

    日本原子力学会春の年会要旨集(CD-ROM)   2007   TD09  2007

    J-GLOBAL

  • Electrochemical response of metal-oxide by gamma ray irradiation

    Honjo Yoshio, Furuya Masahiro, Takamasa Tomoji, Okamoto Koji, Hiroishi Daisuke, Tomozawa Hidemasa, Fujisawa Kyosuke

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2007 ( 0 ) 90 - 90  2007

     View Summary

    ステンレスの表面に酸化チタンの薄い皮膜を成膜し、電気化学的に酸化チタンの自然浸漬電位を測定した。この皮膜にγ線もしくは紫外線を照射した場合の電位応答を調べた。酸化チタンの皮膜はスパッタ成膜であり、厚さは 100nmから1000nmである。この厚さの範囲では、短波長紫外線(UV-C)照射による電位変化および応答の速さは皮膜厚さに依らない傾向を示した。これは半導体としての酸化チタン皮膜の応答として説明できるものである。一方、γ線を照射した場合には、自然浸漬電位の値が皮膜厚さに対して相関しておらず、振る舞いは単純でないが、皮膜がステンレス上に成膜された複合材料であることを考慮に入れて定性的な説明を与える。

    DOI CiNii J-GLOBAL

  • SIRIUS‐F設備を用いたBWR‐5炉心の安定性評価

    古谷正裕, 深堀貴憲, 溝上伸也, 横谷淳

    日本混相流学会年会講演会講演論文集   2007   294 - 295  2007

    J-GLOBAL

  • Fundamental Study of Corrosion Control in Marine and Offshore Structures Using Radiation Induced Surface Activation (RISA)-Improvement of SUS304 Stainless Steel Durability Against Crevice Corrosion in Seawater

    元田慎一, 植松潤一, 波津久達也, 賞雅寛而, 植松進, 古谷正裕

    マリンエンジニアリング   42 ( 4 ) 675 - 681  2007

     View Summary

    A corrosion mitigation technique based on radiation induced surface activation (RISA) from the gamma ray irradiation on a metal surface is reported in this paper. This study aimed to develop a RISA method to prevent crevice corrosion in SUS304 stainless steel using low-intensity radioactive material. Experiment showed that an electrode potential of -100 mV vs. Ag/AgCl was produced and maintained on TiO2-coated SUS304 stainless steel specimens immersed in artificial seawater and in close contact with a small, sealed60Co source or activated by spontaneous neutron irradiation, with no corrosion observed for more than 7 days. On the contrary, the potential of the specimen without a radiation source decreased less than-280 mV vs. Ag/AgCl and crevice corrosion occurred beneath the O-ring within few days. The RISA effect of low-intensity radioactive material has the potential to prevent crevice corrosion of SUS304 stainless steel in actual seawater.

    DOI CiNii J-GLOBAL

  • Development of Innovative Nuclear Technology Based on the Radiation Induced Surface Activation:(1) Summay of the Project Results

    Takamasa Tomoji, Hazuku Tatsuya, Okamoto Koji, Mishima Kaichiro, Morooka Shinichi, Furuya Masahiro

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2007 ( 0 ) 280 - 280  2007

     View Summary

    放射線誘起表面活性(Radiation Induced Surface Activation, RISA)は、今世紀初頭に我が国で初めて確認された金属酸化皮膜への放射線照射によって生じる表面活性効果であり、その効果により放射線環境下の伝熱・防食特性向上をはかることができる。本稿では、昨年度末まで4年計画で行われた経産省革新的実用原子力技術開発費補助事業「放射線誘起表面活性効果による高性能原子炉に関する技術開発」のRISAによる伝熱特性向上及びメカニズム解明などの研究結果を概説する。

    DOI CiNii J-GLOBAL

  • Feasibility of Long-Life and Corrosion-Resistant Canister with Titanium Cladding

    FURUYA MASAHIRO, TOKIWAI MORIYASU, HONJO YOSHIO, HORIE MASAAKI, Saegusa Toshiari

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2007 ( 0 ) 347 - 347  2007

     View Summary

    使用済燃料をコンクリートキャスク等で貯蔵する場合の長期耐久性に優れたキャニスターを提案する。通常のステンレス製キャニスターをチタン板で被覆する構造である。これによりステンレス製キャニスターの腐食や応力腐食割れの懸念が無くなる。従来、チタンとステンレス鋼の溶接においては、脆い金属間化合物が析出し、強度が不足することが知られてきたが、銅合金を用いることによってチタンとステンレス鋼との溶接を行い、接合部は十分な強度があることが判明した。コスト評価を実施した結果、本概念ではチタン使用量が少なく、低コストで製造できる見通しを得た。なお、本キャニスターの概念は、放射性廃棄物を処分する場合などで用いる容器の長期耐久性確保にも有効である。

    DOI CiNii J-GLOBAL

  • Visualization of vapor film during film boiling in hydrated salt solutions and evaluation of condensation heat transfer

    ARAI Takahiro, FURUYA Masahiro

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2007 ( 0 ) 289 - 289  2007

     View Summary

    高温固体ステンレス球を冷却材プールに浸漬させ、高温固体球周囲に形成される膜沸騰蒸気膜を可視観測した。膜沸騰時の可視観測画像から蒸気膜厚さおよび凝縮熱伝達率を算出した結果、水和塩の添加によって蒸気膜厚さが減少し、凝縮熱伝達が向上することが判明した。

    DOI CiNii J-GLOBAL

  • 水アトマイズ銅粉の比表面積の増加方法

    水谷民穂, 小田英治, 藤原弘, 飴山恵, 古谷正裕, 新井崇洋

    粉体粉末冶金協会講演概要集   2007   221  2007

    J-GLOBAL

  • Investigation on promoting method of vapor explosion to quench and atomize high melting point metals

    新井 崇洋, 古谷 正裕

    電力中央研究所報告 L 研究報告   ( L06013 ) 1 - 13,巻頭1〜5  2007

    CiNii J-GLOBAL

  • Effect of Salt Additives to Water on the Severity of Vapor Explosions and on the Collapse of Vapor Film

    新井崇洋, 古谷正裕

    Thermal Science and Engineering   15 ( 3 ) 91 - 100  2007

     View Summary

    We proposed ultra rapid solidification and atomization technique, CANOPUS (Cooling and Atomizing based on NOble Process Utilizing Steam explosion), using small-scale vapor explosions to make an amorphous metal. The CANOPUS method is suitable for rapid cooling and atomization process, which utilizing sustainable small-scale vapor explosions. In order to apply the CANOPUS method to a high melting point metal, it is necessary to make a small-scale vapor explosion occur at a high temperature of the molten metal. Small-scale experiment is conducted to develop the vapor explosion promotor in which spontaneous vapor explosion can occur at a high temperature of a molten metal. Spontaneous vapor explosion do not occur when water at 80&deg;C is used as a coolant. However, spontaneous vapor explosion occurs when water at 80&deg;C with salt additives is used as a coolant. Specifically, lithium chloride solution generates spontaneous vapor explosions at the highest temperature of the molten tin in the experiment. In order to clarify the triggering mechanism of the spontaneous vapor explosion when the promotor is used as a coolant, a high-temperature solid stainless steel sphere is immersed into a coolant. The interfacial temperature of the stainless steel sphere is measured, and the behavior of a vapor film around the stainless steel sphere is observed with a digital video camera. As a result, salt additives resulted in an increase of quench temperature in all salt solutions. The quenching curves of each coolant indicate that the salt additives improve the film boiling heat transfer. The improvement of the film boiling heat transfer causes an unstable formation of the vapor film and a rise of the quench temperature. It is clarified that the salt additives to water promotes a vapor film collapse. Comparing two experiments, the quench temperature of each solution is in close agreement with the upper limit of the molten tin temperature that causes spontaneous vapor explosion. This result suggests that the vapor film collapse triggers spontaneous vapor explosion.

    DOI CiNii J-GLOBAL

  • Stability estimation of BWR-5 plant with SIRIUS-F facility

    古谷 正裕, 深堀 貴憲, 溝上 伸也

    電力中央研究所報告 L 研究報告   ( L06004 ) 巻頭1 - 3,1〜23  2007

    CiNii J-GLOBAL

  • Improvement of the Wettability of Metal Oxide Surface by Radiation Induced Surface Activation

    SAYANO AKIO, KANO FUMIHISA, SAITO NORIHISA, ABE HIROAKI, OKAMOTO KOJI, TAKAMASA TOMOJI, MIYANO MASAMI, FURUYA MASAHIRO, YOSHIKAWA MASAHITO

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2006 ( 0 ) 558 - 558  2006

     View Summary

    酸化金属表面に放射線(γ線)を照射することにより皮膜表面が親水化する現象が見出されている。本報告では、照射による親水化面の安定性に関する結果を中心に報告する。

    DOI CiNii

  • Surface Science of Radiation-Induced and Photo-Induced Hydrophilicity

    古谷正裕, 本城義夫, 阿部弘亨, 岡本孝司, 賞雅寛而, 鹿野文寿, 宮野征巳

    日本伝熱シンポジウム講演論文集(CD-ROM)   43rd   A211  2006

    J-GLOBAL

  • Effect of Salt Additives to Water on the Severity of Vapor Explosions and on the Collapse of Vapor Film

    新井崇洋, 古谷正裕

    日本伝熱シンポジウム講演論文集(CD-ROM)   43rd   J114  2006

    J-GLOBAL

  • Corrosion Control Based on Radiation Induced Surface Activation

    古谷正裕, 賞雅寛而, 岡本孝司, 安永龍哉, 植松進

    マリンエンジニアリング   41 ( 2 ) 278 - 284  2006

    DOI CiNii J-GLOBAL

  • Vapor Film Collapse Structure at High Temperature and Triggering of Vapor Explosion

    ARAI Takahiro, FURUYA Masahiro

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2006 ( 0 ) 235 - 235  2006

     View Summary

    1400度の銅液滴を水中に滴下する実験を行い、蒸気爆発が発生する場合には、溶融銅のフィラメントが観測され、その後のトリガリングの基点になることを明らかにした。

    DOI CiNii J-GLOBAL

  • Anticorrosion Effect Based on Radiation Induced Surface Activation of Atmospheric Oxidation Film

    FURUYA MASAHIRO, HONJO YOSHIO, TAKAMASA TOMOJI, OKAMOTO KOJI, HIROISHI DAISUKE, NAKAGAWA MASAKI

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2006 ( 0 ) 410 - 410  2006

     View Summary

    ステンレス鋼表面に高温大気中で酸化皮膜を設け、γ線照射を行い、自然浸漬電位を測定した。その結果、電位は100mV以上卑化し、大気中酸化皮膜でも放射性誘起表面活性による腐食緩和効果が発現することが判明した。

    DOI CiNii J-GLOBAL

  • 高融点金属を急冷・微粒化させる蒸気爆発促進材の開発

    新井崇洋, 古谷正裕

    電力中央研究所原子力技術研究所研究報告   ( L05013 ) 1 - 13,巻頭1〜5  2006

    CiNii J-GLOBAL

  • SIRIUS‐F設備を用いた1/3 MOX燃料装荷BWR‐5炉心の安定性評価

    古谷正裕, 深堀貴憲, 溝上伸也, 横谷淳

    電力中央研究所原子力技術研究所研究報告   ( L05020 ) 1 - 22,巻頭1〜5  2006

    CiNii J-GLOBAL

  • Promoting Mechanism of Vapor Explosion by salt additives

    ARAI Takahiro, FURUYA Masahiro

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2006 ( 0 ) 319 - 319  2006

     View Summary

    錫を用いた小規模自発的蒸気爆発実験ならびに高温固体球を用いた蒸気膜生成・崩壊実験を行った。その結果、塩の添加によってクエンチ温度が上昇し、蒸気膜が崩壊しやすくなることで、より高温で蒸気爆発が発生するようになることを示した。この蒸気爆発促進効果は、塩添加によって膜沸騰熱伝達が向上し、生成される蒸気膜が薄くなったことに起因することを明らかにした。

    DOI CiNii J-GLOBAL

  • Stability Evaluation of BWR-5 Plant with 1/3 MOX Fuel Installed Using Regional Stability Experimental Facility SIRIUS-F

    FURUYA MASAHIRO, FUKAHORI TAKANORI, MIZOKAMI SHINYA, YOKOYA JUN

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2006 ( 0 ) 299 - 299  2006

     View Summary

    BWR-5に9×9A型のMOX燃料を全体の1/3装荷した炉心を想定し、領域安定性試験設備SIRIUS-Fを用いて最低ポンプ速度最大出力点など広い範囲で試験を行い、運転範囲において、十分な安定性余裕があることを確認した。また許認可解析コードODYSYの減幅比(DR)および共振周波数は試験結果と精度良く一致した。

    DOI CiNii J-GLOBAL

  • 放射線誘起表面活性による酸化金属表面のぬれ性向上(2)

    佐谷野顕生, 鹿野文寿, 斎藤宣久, 阿部弘亨, 岡本孝司, 賞雅寛而, 古谷正裕, 宮野征巳, 吉川正人

    日本原子力学会秋の大会予稿集(CD-ROM)   2006   G41  2006

    J-GLOBAL

  • Radiation induced hydrophilicity of metal-oxide and surface adsorption substance

    Honjo Yoshio, Furuya Masahiro, Abe Hiroaki, Okamoto Koji, Takamasa Tomoji, Kano Fumihisa, Ono Shoichi, Miyano Masami

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2006 ( 0 ) 548 - 548  2006

     View Summary

    金属酸化物にガンマ線照射をすると放射線誘起表面活性(RISA)により超親水性が発現すると考えられる。ガンマ線照射した酸化チタン単結晶(100)面を表面分析し、超親水性と表面状態の関係を示す。親水性を発現した酸化チタン単結晶表面ではブリッジ酸素が減少していることがXPSからわかった。酸化チタンの光触媒としての詳細な研究から、欠損したブリッジ酸素には水が化学吸着して表面水酸基を形成すると考えられており、表面水酸基が超親水化に重要である。一方、昇温脱離分析を行ったところ、表面からの水脱離温度が高い値を示した。このことから、結合エネルギーの高い表面吸着水の存在が超親水性発現に重要であることがわかった。

    DOI CiNii J-GLOBAL

  • Triggering of Vapor Explosion with Copper in Highly Subcooled Water

    新井崇洋, 古谷正裕

    日本機械学会年次大会講演論文集   2006 ( Vol.3 ) 151 - 152  2006

    J-GLOBAL

  • Stability Estimation on the Basis of Noise Analysis

    古谷正裕, 深堀貴憲, 溝上伸也, 横谷淳

    日本機械学会年次大会講演論文集   2006 ( Vol.3 ) 203 - 204  2006

    J-GLOBAL

  • 4604 Triggering of Vapor Explosion with Copper in Highly Subcooled Water

    ARAI Takahiro, Furuya Masahiro

    The proceedings of the JSME annual meeting   2006 ( 3 ) 151 - 152  2006

     View Summary

    Experiments were conducted in which a molten copper droplet released into water pool. Spontaneous vapor explosion did not occur when water temperature was 50℃. Spontaneous vapor explosion did, however, occur at a rate of 70%, when water temperature was 20℃. A high-speed video frames explored the stages of vapor explosions: (1) a vapor film was formed and separates the copper droplet and surrounding water, (2) a filament of molten copper grew from the surface and deformed the vapor film, (3) the vapor film collapsed along the filament surface, and finally (4) triggering of vapor explosions occurred from the filament to the whole molten copper droplet. When the filament growth was observed, it triggered the vapor explosion in almost all cases. When not, vapor explosion was not observed and the vapor film was, therefore, stably formed around the molten copper droplet. We concluded that the filament form the molten copper triggered vapor explosion in a highly subcooled water.

    DOI CiNii

  • Corrosion Control Based on Radiation Induced Surface Activation

    FURUYA Masahiro, TAKAMASA Tomoji, OKAMOTO Koji, YASUNAGA David T., UEMATSU Susumu

    Marine Engineering   41 ( 2 ) 278 - 284  2006

     View Summary

    When a semiconductor film is irradiated by gamma rays, excited electrons are transferred to abase metal in contact with the film, resulting in a drop of corrosion potential. The authors propose a corrosion mitigation method based on radiation induced surface activation (RISA) phenomena by supplying gamma rays from outside the material, or based on a self-excited methodology activating the film and/or the base metal. The corrosion potential of ZrO2 coated SUS304L was shifted down to the range between -90 mV and -300 mV vs. SSE by gamma-ray irradiation. The corrosion potential was further shifted down to -600 mV when a CoCr intermediate layer was inserted between the ZrO2 spray coating film and the SUS304L base metal. Iron specimens with a spray coating film of TiO2, ZrO2, and Al2O3 were immersed in a 3 wt% sodium chloride aqueous solution. Pitting and general corrosion were observed on both the specimens kept in a darkroom and illuminated with ultraviolet rays. Pitting and general corrosion were, however, suppressed on all three specimens irradiated with gamma rays.

    DOI CiNii

  • Surface Modification Technology “Fresh Green” Provides New Applications for Titanium

    常磐井守泰, 古谷正裕, 田中伸幸

    チタン   54 ( 1 ) 54 - 56  2006

    CiNii J-GLOBAL

  • Evaluations of Super-Hydrophilic Titanium Oxide Compound Fabricated by Plasma Thermal Spray Coating

    樋口貞雄, 宮島清富, 古谷正裕, 小林陽, 河村浩孝, 田辺一夫

    電気学会論文誌 A   126 ( 8 ) 857 - 862  2006

     View Summary

    The present paper concerns the evaluations of the long period super-hydrophilicity of titanium oxide films fabricated by plasma thermal spraying on the aluminum substrate. The fabricated titanium oxide and tungsten doped titanium oxide kept super-hydrophilic surface over 315 days without UV ray illumination. Especially, tungsten doped titanium oxide kept super-hydrophilic surface including the area without dropping of water for 315 days.

    DOI CiNii J-GLOBAL

  • 4631 Stability Estimation on the Basis of Noise Analysis

    FURUYA Masahiro, FUKAHORI Takanori, MIZOKAMI Shinya, YOKOYA Jun

    The proceedings of the JSME annual meeting   2006 ( 3 ) 203 - 204  2006

     View Summary

    A time series analysis was performed to calculate decay ratio and resonance frequency from a dominant pole of a transfer function by applying AR methods to time-series of the core inlet flow rate. Experiments were conducted with the SIRIUS-F facility, which simulates ABWR with MOX fuel installed. The variations in the decay ratio and resonance frequency among five common AR methods are within 0.03 and 0.01Hz respectively. In this system, the appropriate decay ratio and resonance frequency can be estimated on the basis of Yule-Walker Method with the model order of 30.

    DOI CiNii

  • Characteristics of type-1 density wave oscillations in a natural circulation BWR at relatively high pressure

    M Furuya, F Inada, THJJ Van Der Hagen

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   42 ( 2 ) 191 - 200  2005.02

     View Summary

    Experiments were conducted to investigate two-phase flow instabilities in a boiling natural circulation loop with a chimney at high pressure. The SIRIUS-N facility was designed to have non-dimensional values which are nearly equal to those of a typical natural circulation BWR. The observed oscillations are found to be density wave oscillations, since the void fractions in the chimney inlet and exit are out of phase. They belong to the Type-1 category, since they occur at low flow qualities, according to the Fukuda-Kobori's classification. Moreover, the oscillation period correlates well with the passing time of bubbles in the chimney section regardless of the system pressure, the heat flux, and the inlet subcooling. Two distinct phenomena are found in relation between the oscillation period and liquid passing time in the chimney, indicating that the driving mechanisms of the instabilities are different between low and high pressures. Stability maps were obtained in reference to the inlet subcooling and the heat flux at the system pressures of 1, 2, 4, and 7.2 MPa. The flow became stable below a certain heat flux regardless of the channel inlet subcooling. The stable region enlarges with increasing system pressure. Thus, the stability margin becomes larger in a startup process of a reactor by pressurizing the reactor sufficiently before withdrawing the control rods. The obtained stability map demonstrates that the nominal operating condition of the ESBWR has a significant stability margin to the unstable region.

    DOI CiNii

  • Mechanical application of radiation induced surface activation ( RISA ).

    賞雅寛而, 三島嘉郎, 岡本孝司, 古谷正裕

    機械の研究   57 ( 3 ) 343 - 350  2005

    J-GLOBAL

  • Corrosion Control by Radiation Induced Surface Activation Using a Radio Isotope Source

    Nakamura Daisuke, Motoda Shinichi, Hazuku Tatsuya, Takamasa Tomoji, Furuya Masahiro, Okamoto Koji, Uematsu Susumu

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   43rd ( 0 ) 436 - 436  2005

     View Summary

    酸化金属を溶射したSUS304 試験片をRI 放射線源と接触させることで、酸化被膜に微弱放射線が供給され、放射線誘起表面活性による腐食緩和効果が発現した。

    DOI CiNii J-GLOBAL

  • Atmospheric Corrosion Control on the basis of Radiation Induced Surface Activation

    FURUYA Masahiro, TAKAMASA Tomoji, OKAMOTO Koji, SAEGUSA Toshiari

    Transactions of the Atomic Energy Society of Japan   4 ( 1 ) 84 - 86  2005

    DOI CiNii J-GLOBAL

  • Development of BWR Regional Stability Experimental Facility SIRIUS-F, which Simulates Thermalhydraulics-Neutronics Coupling in Reactor Core, and Stability Evaluation of ABWR

    古谷正裕, 深堀貴憲, 溝上伸也

    日本原子力学会和文論文誌   4 ( 2 ) 93 - 105  2005

     View Summary

    The SIRIUS-F facility was designed and constructed for highly accurate simulation of channel, core-wide and regional instabilities of an ABWR. A real-time simulation is performed for the modal-point kinetics of reactor neutronics and fuel-rod conduction on the basis of a measured void fraction in a reactor core section of the facility. A noise analysis method was performed to calculate decay ratios from dominant poles of transfer function on the basis of the AR method by applying time series of a core inlet flow rate. By utilizing this method, one can estimate stability at any specific operating point online without assuming excess conservative conditions. Channel and regional stability experiments were conducted for a wide range of operating conditions including maximum power points along the minimum pump speed line and the natural circulation line of the ABWR. The decay ratios and the resonance frequencies are in good agreement with those from the design analysis code, ODYSY. The SIRIUS-F experimental results demonstrated stability characteristics such as a stabilizing effect of the power, and reviled a sufficiently large stability margin even under hypothetical conditions of power enlargement.

    DOI CiNii J-GLOBAL

  • SIRIUS‐F設備を用いた全MOX燃料装荷ABWR炉心の領域安定性評価

    古谷正裕, 稲田文夫, 深堀貴憲, 溝上伸也, 横谷淳

    電力中央研究所原子力技術研究所研究報告   ( L04011 ) 1 - 22,巻頭1〜5  2005

    CiNii J-GLOBAL

  • Evaluations of super-hydrophilic titanium oxide compound applied for audible noise reduction

    樋口貞雄, 宮島清富, 小林陽, 古谷正裕, 河村浩孝, 田辺一夫

    電力中央研究所材料科学研究所研究報告   ( Q04018 ) 21P  2005

    J-GLOBAL

  • Stability Evaluation of ABWR with MOX Fuel Installed Using Regional Stability Experimental Facility SIRIUS-F

    FURUYA MASAHIRO, FUKAHORI TAKANORI, MIZOKAMI SHINYA, YOKOYA JUN

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2005 ( 0 ) 237 - 237  2005

     View Summary

    ABWRに9×9A型のMOX燃料を全数装荷した炉心を想定し、領域安定性試験設備SIRIUS-Fを用いて最低ポンプ速度最大出力点など広い範囲で試験を行った。その結果、同炉は運転範囲において、十分な安定性余裕があることを確認した。また許認可解析コードODYSY の減幅比(DR)および共振周波数は試験結果と精度良く一致した。

    DOI CiNii J-GLOBAL

  • Improvement of the Wettability of Metal Oxide Surface by Radiation Induced Surce Activation

    Sayano Akio, Kano Fumihisa, Saito Norihisa, Okamoto Koji, Abe Hiroaki, Imai Yasuyuki, Takamasa Tomoji, Furuya Masahiro, Yoshikawa Masahito

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2005 ( 0 ) 396 - 396  2005

     View Summary

    大気中においてコバルト60によるガンマ線照射をして表面を超親水化させたジルカロイ_-_4試験体に対し、各種処理の影響を調べた。照射前の接触角が約77度であるのに対し、線量率10 kGy/h で34時間照射を行った場合、接触角は10度前後まで減少し、高い親水性を示すようになる。この親水化した試験体を大気中にて24時間保持した場合には接触角の変化は認められない。さらにこの試験体を常温真空中で1時間保持したもの、および大気中100℃で1時間保持したものでは接触角が60度前後まで上昇し、ほぼ照射前のレベルに近いところまで接触角が回復していた。一方、同じ試験体を常温純水中に保持後マグネットスターラにより1時間攪拌して流水中における影響を調べたところ、接触角は15度前後を示し、高い親水性を維持していた。

    DOI CiNii J-GLOBAL

  • Regional stability estimation of natural circulation BWRs using SIRIUS-N facility

    Furuya, M., Inada, F., Van Der Hagen, T.H.J.J.

    Journal of Nuclear Science and Technology   42 ( 4 ) 341 - 350  2005

     View Summary

    The SIRIUS-N facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of the BWR. A real-time simulation was performed in the digital controller for heat conduction in a fuel-rod and modal point kinetics of reactor neutronics using measured void fractions in reactor core sections of the thermal-hydraulic loop. In order to estimate decay ratios, an auto-regressive method has been successfully applied for time series data of the core inlet flow rate. Experiments were conducted with the SIRIUS-N facility for the rated operating condition of 3.13 GWt natural circulation BWR. Channel and regional stability decay ratios were determined to be 0.38 and 0.54, respectively, which indicates sufficient margin for the instabilities. Experiments were extended to evaluate the stability sensitivity of design parameters such as the power profile, void reactivity coefficients, core inlet subcooling, and the fuel rod time constant.

    DOI CiNii

  • Development of Visible Light Photocatalyst with Superior Durability and High Catalytic Activity, ‘Fresh Green’

    古谷正裕

    電力中央研究所狛江研究所報告   ( T03067 ) 1 - 9,巻頭1〜3  2005

    CiNii J-GLOBAL

  • Boiling Characteristics of Salt Additive Droplet on a Hot Surface

    MATSUMURA Kunihito, KAMINAGA Fumito, FURUYA Masahiro

      12 ( 4 ) 37 - 38  2004.07

    CiNii

  • Innovative Nuclear Technologies Based on Radiation Induced Surface Activation(RISA) (2)

    YASUNAGA TATSUYA, SHIMOJO JUN, FURUYA MASAHIRO, UEMATSU SUSUMU, SEKIMURA NAOTO, TAKAMASA TOMOJI

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2004 ( 0 ) 286 - 286  2004

     View Summary

    放射線誘起表面活性(RISA)効果による防食機能を有する材料として、母材にSUS316L、中間層としてCo-CrあるいはNi-Crを用い、酸化金属皮膜ZrO2, TiO2を溶射した材料をとりあげて評価した。この材料が原子炉環境に適用できる耐久性を有するかを見極めるため、硬度評価、磨耗評価、断面評価、剥離評価等の機械的特性を実施した。また、非放射線照射下での高温高圧水環境下での耐久性も評価した。

    DOI CiNii

  • Innovative Nuclear Technologies Based on Radiation Induced Surface Activation (RISA) (1)

    FUJISAWA KYOSUKE, TAKAMASA TOMOJI, MOROOKA SHINICHI, HISHIDA MAMORU, FURUYA MASAHIRO, UEMATSU SUSUMU, NAKAMURA HIDEO, MISHIMA KAICHIRO, SEKIMURA NAOTO, OKAMOTO KOJI

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2004 ( 0 ) 285 - 285  2004

     View Summary

    放射線誘起表面活性(Radiation Induced Surface Activation :RISA)効果を利用して、原子炉内部の構造材や被覆管等に同皮膜を施工することにより、(1)構造物の腐食電位を卑化させて耐食性を向上させる、(2)原子炉放射線環境下における炉心伝熱特性(限界熱流速、事故時の再冠水速度)を向上させる、ことにより経済性・安全性に優れる高性能原子炉に関する技術を開発している。ここでは、その概要について報告する。

    DOI CiNii

  • ディーゼル排ガスに含まれるナノ粒子中化学成分のオンライン測定技術の開発

    田中伸幸, 津崎昌東, 古谷正裕, 川添浩平, 出口祥啓

    大気環境学会年会講演要旨集   45th   475  2004

    J-GLOBAL

  • Innovative Nuclear Technologies Based on Radiation Induced Surface Activation (RISA) (3)

    ONO SHOICHI, MIYANO MASAMI, HISHIDA MAMORU, YASUNAGA TATSUYA, FURUYA MASAHIRO, TAKAMASA TOMOJI

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   42nd ( 0 ) 287 - 287  2004

     View Summary

    軽水炉プラントの高経年化にともない、炉内構造材の応力腐食割れ(SCC)対策が重要課題となっている。SCC環境因子の定量的指標としてはステンレス鋼の電気化学的腐食電位(ECP)が用いられ、-230mV以下であればSCCの発生やき裂進展が抑制できるとされている。最近、室温の水溶液中において、ジルコニア(ZrO2)被覆ステンレス鋼にγ線照射することにより放射性誘起表面活性(RISA)が発現し、ECPは大きく低下することが明らかにされ、水素注入や貴金属注入に替わる新しい防食技術として期待されている。今回、BWR冷却水を模擬した高温水環境においても、RISAによるECP低下効果が顕著に発現することが明らかになったので報告する。

    DOI CiNii J-GLOBAL

  • 放射線誘起表面活性効果による高性能原子炉に関する技術開発(1)

    藤沢匡介, 師岡慎一, 菱田護, 古谷正裕, 植松進, 中村秀夫, 三島嘉一郎, 関村直人, 賞雅寛而

    日本原子力学会春の年会要旨集   42nd   287  2004

    J-GLOBAL

  • 放射線誘起表面活性による防食技術と水素製造

    古谷正裕

    日本原子力学会春の年会要旨集   42nd   SO2  2004

    J-GLOBAL

  • 放射線誘起表面活性効果による高性能原子炉に関する技術開発(2)

    安永龍哉, 下条純, 古谷正裕, 植松進, 関村直人, 賞雅寛而

    日本原子力学会春の年会要旨集   42nd   288  2004

    J-GLOBAL

  • 放射線誘起表面活性とは何か?

    賞雅寛而, 波津久達也, 班目春樹, 岡本孝司, 三島嘉一郎, 古谷正裕, 友沢秀征

    日本原子力学会春の年会要旨集   42nd   SO1  2004

    J-GLOBAL

  • Study on the evaporation characteristics of salt additive droplet

    松村邦仁, 神永文人, 古谷正裕

    日本伝熱シンポジウム講演論文集   41st ( Vol.1 ) 7 - 8  2004

    J-GLOBAL

  • Radiation-Induced and Photo-Induced Hydrophilicity

    古谷正裕, 松村哲夫, 賞雅寛而, 岡本孝司, 本城義夫, 宮野征巳

    日本伝熱シンポジウム講演論文集   41st ( Vol.3 ) 645 - 646  2004

    J-GLOBAL

  • 炉内流動と核反応度との結合を模擬したBWR領域安定性試験設備SIRIUS‐Fの開発とABWRの安定性評価

    古谷正裕, 稲田文夫, 深堀貴憲, 太田武

    電力中央研究所原子力技術研究所研究報告   ( L04001 ) 1 - 17,巻頭1〜5  2004

    CiNii J-GLOBAL

  • Mechanism and Evaluation of Flashing-Induced Density Wave Oscillations in Natural Circulation BWR

    FURUYA Masahiro, INADA Fumio

    Transactions of the Atomic Energy Society of Japan   3 ( 2 ) 141 - 150  2004

     View Summary

    Experiments were conducted to investigete two-phase flow stability of a natural circulation BWR due to flashing at low pressure. The facility used in the experiment was designed to have non-dimensional values which are nearly equal to those of typical natural circulation BWR. The observed instability is suggested to be the flashing induced density wave oscillations, since the oscillation period was nearly one and a half to two times the passing time in the chimney section, and correlated well with a single line regardless of system pressure, heat flux, and inlet subcooling. Stability maps were obtained in reference to the core inlet subcooling and the heat flux at the system pressures of 0.1, 0.2, 0.35, and 0.5 MPa. The flow became stable below a certain heat flux regardless of the channel inlet subcooling. The stable region enlarged with increasing system pressure. According to the stability map, the stability margin becomes larger in a startup process of a reactor by pressurizing the reactor sufficiently before withdrawing control rods.

    DOI CiNii J-GLOBAL

  • Development of BWR Regional Stability Experimental Facility SIRIUS-F, which Simulates Thermalhydraulics-Neutronics Coupling in Reactor Core, and Stability Evaluation of ABWR

    FURUYA MASAHIRO, INADA FUMIO, FUKAHORI TAKANORI, OHTA TAKESHI

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2004 ( 0 ) 248 - 248  2004

     View Summary

    ABWRを模擬した領域安定性試験設備SIRIUS-Fを開発した。最低ポンプ速度最大出力点など広い範囲で試験を行った結果、許認可解析コードODYSY のDRおよび共振周波数はSIRIUS-Fの試験結果と精度良く一致した。

    DOI CiNii J-GLOBAL

  • 溶射法により作製した二酸化チタン膜の光吸収特性および親水性

    樋口貞雄, 宮島清富, 水津竜夫, 古谷正裕, 田辺一夫

    応用物理学会学術講演会講演予稿集   65th ( 2 ) 531  2004

    J-GLOBAL

  • 放射線誘起表面活性による沸騰改善と可視化

    岡本孝司, 今井康之, 古谷正裕, 賞雅寛而

    日本機械学会年次大会講演論文集   2004 ( Vol.8 ) 153 - 154  2004

    J-GLOBAL

  • Effect of Surface Wettability Caused by Radiation Induced Surface Activation on Leidenfrost Condition

    TAKAMASA Tomoji, HAZUKU Tatsuya, MISHIMA Kaichiro, OKAMOTO Koji, IMAI Yasuyuki, FURUYA Masahiro

    Thermal Sci. Engng.   12 ( 1 ) 35 - 36  2004.01

    CiNii

  • Development of Corrosion Control Method on the basic of Radiation Induced Surface Activation(Research Institute at Outside Organizations)

    古谷 正裕, 植松 進, 賞雅 寛而, 岡本 孝司, 広石 大介, 河村 浩孝

    Papers of National Maritime Research Institute   3 ( 5 ) 595 - 595  2004

    CiNii

  • 耐久性と触媒活性を向上させた可視光応答型光触媒

    古谷 正裕

    日本機械学會誌   107 ( 1028 ) 560 - 560  2004

    DOI CiNii

  • Boiling Enhancement and Visualization on Radiation Induced Surface Activation

    OKAMOTO Koji, IMAI Yasuyuki, FURUYA Masahiro, TAKAMASA Tomoji

    The Reference Collection of Annual Meeting   2004 ( 8 ) 153 - 154  2004

    DOI CiNii

  • Development of Visible Light Photocatalyst with Superior Durability and High Catalytic Activity, ‘Fresh Green’

    古谷正裕

    日本機械学会年次大会講演論文集   2004 ( Vol.6 ) 311 - 312  2004

     View Summary

    A new process was developed to form high performance photocatalyst film of titanium dioxide by oxidizing titanium metals in the burning flame of acetylene gases. The oxide film by this process has much superior resistance to scratch, wear and chemical reactions to a conventional titanium dioxide spray coating film. Further, this film shows superior performances in the visible-light photo-catalysis; ten times or more of photo-current density and wide absorption wave range up to 490nm comparing with the conventional spray coating film up to 410nm.

    DOI CiNii J-GLOBAL

  • Corrosion Control on the Basis of Radiation Induced Surface Activation

    FURUYA Masahiro

    放射線化学   78 ( 78 ) 35 - 38  2004

    CiNii J-GLOBAL

  • Development of Production Method of Fibers and Powders from Highly-Viscous Coal Gasification Slag

    古谷正裕, 市川和芳, 山本武志

    電力中央研究所狛江研究所報告   ( T03066 ) 1 - 11,巻頭1〜3  2004

    CiNii J-GLOBAL

  • Occurrence Conditions of Sustainable Minute Bubble Emission Boiling for High Heat Flux Cooling

    FURUYA Masahiro

    Transactions of the Atomic Energy Society of Japan   2 ( 2 ) 115 - 120  2003.06

     View Summary

    In order to investigate the conditions causing minute bubble emission boiling, critical heat flux experiments were conducted in the pool boiling system with different thermal capacities of heat transfer surfaces and method of heating. Stable minute bubble emission boiling was observed for a 10 mm-thick copper cylinder heated by thermal radiation. The critical heat flux obtained was 6.0 MW/m2. When the heat flux exceeded above approximately 3 MW/m2, a large vapor bubble formed on the heat transfer surface, then was condensed immediately and dispersed into minute bubbles. A 4 mm-thick silicon carbide heat transfer surface burnouted at 1.9 MW/m2. When the heat transfer surface has a small thermal capacity such as the metal foil, commonly used in CHF experiments, a temporary loss of heat removal due to large bubble formation will cause a rapid temperature increase and will result in burnout. Minute bubble emission boiling could occur stably when the time constant determined by the heating method and thermal capacity exceeds the time required for a large bubble to condense in the subcooled fluid. © 2003, Atomic Energy Society of Japan. All rights reserved.

    DOI CiNii

  • Radiation Induced Surface Activation

    TAKAMASA Tomoji, OKAMOTO Koji, MISHIMA Kaichiro, FURUYA Masahiro

    Journal of the Atomic Energy Society of Japan / Atomic Energy Society of Japan   45 ( 2 ) 112 - 117  2003

     View Summary

    &lt;p&gt; プラズマ照射された酸化金属被膜にkGy/hのオーダーの強&lt;i&gt;γ&lt;/i&gt;線照射を行うことにより, 表面が活性化され伝熱特性の改善 (濡れ性および限界熱流束の向上) が生じる。この放射線による表面活性は放射線誘起表面活性 (Radiation Induced Surface Activation : RISA) と呼ばれ, またRISAによる沸騰熱伝達改善は放射線誘起沸騰改善 (Radiation Induced Boiling Enhancement) と呼ばれている。本稿では, 伝熱特性の向上を中心に, RISAの原理をはじめ, 腐食低減および放射線計測などの原子炉関連の応用研究の現状を解説する。&lt;/p&gt;

    CiNii

  • 水素エネルギー社会と原子力 原子力を利用した水素製造技術の全容 [5] 放射線誘起表面活性を用いた高効率水素製造技術

    古谷正裕

    原子力eye   49 ( 1 ) 30 - 33  2003

    J-GLOBAL

  • 放射線誘起表面活性を利用した自励防食法の開発

    古谷正裕, 植松進, 賞雅寛而, 波津久達也, 岡本孝司, 広石大介, 河村浩孝, 白鳥義行

    日本原子力学会春の年会要旨集   41st   745  2003

    J-GLOBAL

  • Study on the quenching characteristics of solid sphere(Effect of salt additive to a coolant)

    松村邦仁, 古谷正裕, 神永文人

    日本伝熱シンポジウム講演論文集   40th ( Vol.1 ) 139 - 140  2003

    J-GLOBAL

  • Amorphization of Practical Materials by Innovative Rapid Cooling and Atomization Process, CANOPUS

    古谷正裕, 藤江政武

    日本伝熱シンポジウム講演論文集   40th ( Vol.3 ) 803 - 804  2003

    J-GLOBAL

  • Research on surface activity induced by radiation.

    今井康之, 班目春樹, 岡本孝司, 賞雅寛而, 古谷正裕

    共用設備管理部門年報   ( UTRCN-K-34 ) 59 - 61  2003

    J-GLOBAL

  • Occurrence Conditions of Sustainable Minute Bubble Emission Boiling for High Heat Flux Cooling

    FURUYA Masahiro

    Transactions of the Atomic Energy Society of Japan   2 ( 2 ) 115 - 120  2003

     View Summary

    In order to investigate the conditions causing minute bubble emission boiling, critical heat flux experiments were conducted in the pool boiling system with different thermal capacities of heat transfer surfaces and method of heating.<BR>Stable minute bubble emission boiling was observed for a 10 mm-thick copper cylinder heated by thermal radiation. The critical heat flux obtained was 6.0MW/mm2. When the heat flux exceeded above approximately 3MW/mm2, a large vapor bubble formed on the heat transfer surface, then was condensed immediately and dispersed into minute bubbles.<BR>A 4 mm-thick silicon carbide heat transfer surface burnouted at 1.9MW/mm2. When the heat transfer surface has a small thermal capacity such as the metal foil, commonly used in CHF experiments, a temporary loss of heat removal due to large bubble formation will cause a rapid temperature increase and will result in burnout.<BR>Minute bubble emission boiling could occur stably when the time constant determined by the heating method and thermal capacity exceeds the time required for a large bubble to condense in the subcooled fluid.

    DOI CiNii J-GLOBAL

  • Development of Corrosion Control Method on the Basis of Radiation Induced Surface Activation:Effect of Intermediate Layer

    FURUYA MASAHIRO, FUJISAWA KYOUSUKE, SHIMOJO JUN, SHIRATORI YOSHIYUKI, KAWAMURA HIROTAKA, MATSUMURA TETSUO, TAKAMASA TOMOJI, OKAMOTO KOJI, HIROISHI DAISUKE, SEKIMURA NAOTO, UEMATSU SUSUMU, YASUNAGA TATSUYA

    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan   2003 ( 0 ) 383 - 383  2003

     View Summary

    放射線誘起表面活性を発現する酸化ジルコニウムと母材との間にCoCr中間層を成膜した結果、電気的コンダクタンスが良好になり自然浸漬電位がさらに卑化し、腐食緩和効果がより高まることが判明した。

    DOI CiNii J-GLOBAL

  • ICONE11-36523 SURFACE WETTABILITY AND LEIDENFROST CONDITION ON OXIDIZED METALS CAUSED BY RADIATION INDUCED SURFACE ACTIVATION

    Koga Tatsuya, Hazuku Tatsuya, Takamasa Tomoji, Okamoto Koji, Imai Yasuyuki, Mishima Kaichiro, Furuya Masahiro

    The Proceedings of the International Conference on Nuclear Engineering (ICONE)   2003 ( 0 ) 438 - 438  2003

    DOI CiNii

  • TED-AJ03-362 RADIATION INDUCED SURFACE ACTIVATION : WETTABILITY CAUSED BY SELF INDUCED ACTIVATION :

    HAZUKU Tatsuya, TAKAMASA Tomoji, IMAI Yasuyuki, OKAMOTO Koji, FURUYA Masahiro

    Proceedings of the ... ASME/JSME Thermal Engineering Joint Conference   2003 ( 6 ) 292 - 292  2003

     View Summary

    Improving the limit of boiling heat transfer or critical heat flux requires that the cooling liquid can contact the heating surface, or a high-wettability, highly hydrophilic heating surface, even if a vapor bubble layer is generated on the surface. We investigated surface wettability using metal oxides irradiated by γ-rays under normal room conditions. Contact angle, an indicator of macroscopic wettability, was measured by image-processing of the images obtained by a CCD video camera. The results showed that the surface wettability on oxidized metal pieces of titanium, zircaloy No.4,SUS-304,and copper was improved significantly by the Radiation Induced Surface Activation (RISA) phenomenon. Highly hydrophilic conditions of the test pieces were achieved after 500-kGy irradiation by ^<60>Co γ-rays. To check the RISA with radioactive metal under non-γ-ray environment use, surface wettability was investigated using metal oxides irradiated by radiation rays emitted from radioactive in the metals in room temperature conditions. The contact angle decreased linearly with integrated irradiation of neutron rays and kept in low contact angle for longer than 400 hours. These results revealed that the self-induced RISA existed with radioactive metals.[figure]

    CiNii

  • ICONE11-36164 EFFECTS OF CHIMNEY ON THERMO-HYDRAULIC AND CORE INSTABILITY OF BOILING NATURAL CIRCULATION LOOP

    Inada Fumio, Furuya Masahiro, Yasuo Akira

    The Proceedings of the International Conference on Nuclear Engineering (ICONE)   2003 ( 0 ) 251 - 251  2003

    DOI CiNii

  • Radiation Induced Surface Activation

    TAKAMASA Tomoji, OKAMOTO Koji, MISHIMA Kaichiro, FURUYA Masahiro

    Journal of the Atomic Energy Society of Japan / Atomic Energy Society of Japan   45 ( 2 ) 112 - 117  2003

    DOI CiNii J-GLOBAL

  • Development of Corrosion Control Method on the Basis of Radiation Induced Surface Activation.

    古谷正裕, 河村浩孝, 松村哲夫, 賞雅寛而, 岡本孝司, 広石大介, 植松進, 安永龍哉, 藤沢匡介

    電力中央研究所狛江研究所報告   ( T02038 ) 1 - 10,巻頭1〜5  2003

    CiNii J-GLOBAL

  • 蒸気爆発を活用した超急冷・微粒化手法の開発と非晶質金属の製造

    古谷 正裕

    日本機械学會誌   106 ( 1013 ) 298 - 298  2003

    DOI CiNii

  • (11)チムニを有する沸騰自然循環ループの不安定流動に関する研究 : 第4報,高圧時の安定性とチムニ内熱水力挙動の解析的検討(論文)(日本機械学会賞〔2002年度(平成14年度)審査経過報告〕)

    稲田 文夫, 古谷 正裕, 安尾 明

    日本機械学會誌   106 ( 1014 ) 339 - 339  2003

    DOI CiNii

  • Investigation of Production Method of Fibers and Powders from Highly Viscous Slag

    古谷正裕, 市川和芳, 山本武志, 川芳昭

    日本機械学会熱工学コンファレンス講演論文集   2003 ( 0 ) 77 - 78  2003

     View Summary

    A gas atomization method is successfully applied for fiber and powder production from a highly viscous slag. The slag suction pressure becomes higher, when a flow guide was inserted at the exit of the gas nozzle. When the molten coal gasification slag temperature was 1500℃, a smooth continuous fiber measuring approximately 9μm in diameter was produced, which could be used as a heat insulating media and a buffer material. Powders having a diameters of approximately 9μm were produced at 1600℃, which could be used as raw material for cement.

    DOI CiNii J-GLOBAL

  • Effect of liquid density differences on boiling two-phase flow stability

    M Furuya, A Manera, DDB van Bragt, THJJ van der Hagen, WJM de Kruijf

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   39 ( 10 ) 1094 - 1098  2002.10

     View Summary

    In order to investigate the effect of considering liquid density dependence on local fluid temperature in the thermal-hydraulic stability, a linear stability analysis is performed for a boiling natural circulation loop with an adiabatic riser. Type-I and Type-II instabilities were to investigate according to Fukuda-Kobori's classification. Type-I instability is dominant when the flow quality is low, while Type-II instability is relevant at high flow quality. Type-II instability is well known as the typical density wave oscillation. Neglecting liquid density differences yields estimates of Type-II instability margins that are too small, due to both a change in system-dynamics features and in the operational point. On the other hand, neglecting liquid density differences yields estimates of Type-I stability margins that are too large, especially due to a change in the operational point. Neglecting density differences is thus non-conservative in this case. Therefore, it is highly recommended to include liquid density dependence on the fluid subcooling in the stability analysis if a flow loop with an adiabatic riser is operated under the condition of low flow quality.

    DOI CiNii

  • Heat Transfer in Two-Phase Flow : Fundamentals-Interfacial Phenomena Session

    FURUYA Masahiro

    伝熱 : journal of the Heat Transfer Society of Japan   41 ( 170 ) 6 - 7  2002.09

    CiNii

  • THE 12TH INTERNATIONAL HEAT TRANSFER CONFERENCE

    FURUYA Masahiro, NAKABEPPU Osamu, ITAYA Yoshinori, AOKI Kazuo, TADA Yukio, ISHIGURO Hiroshi, NAKAYAMA Akira, OSAKABE Masahiro, HIRASAWA Shigeki, SHIRAKASHI Ryo, YOKOBORI Seiichi, OHTA Haruhiko, SATO Kimitoshi, TOMIMURA Toshio, INADA Shigeaki, UTAKA Yoshio, MAEKAWA Toru, NAKABE Kazuyoshi, MUNAKATA Tetsuo, MATSUBARA Koji, SHIBAHARA Masahiko

    Journal of the Heat Transfar Society of Japan   41 ( 170 ) 6 - 35  2002

    DOI CiNii

  • A Study on Thermo-Hydraulic Instability of Boiling Natural Circulation Loop with a Chimney. 4th Report. An Analytical Consideration of the Stability and Thermo-Hydraulic Characteristics in the Chimney in High Pressure.

    稲田文夫, 古谷正裕, 安尾明

    日本機械学会論文集 B編   68 ( 666 ) 511 - 518  2002

     View Summary

    Thermo-hydraulic instabilities of a boiling natural circulation loop with a chimney under high pressure were investigated using linear stability analysis. Drift-flux model was used for two-phase flow model. The instability regions as well as the thermo-hydraulic characteristics in the chimney such as wavy feature were examined, which were compared with the characteristics in low pressure. Instability could occur when exit quality was relatively low, which was the same manner as the characteristics in low pressure. In high-pressure, void was generated near channel exit, and void wave propagated in the chimney. In low pressure, steam was generated only near the chimney exit due to gravity induced flashing, and single-phase enthalpy wave, that is, temperature wave propagated in single-phase flow region. Though flow could be very stable in the high pressure and high power condition, the decay ratio of higher mode could be larger than that of lower mode.

    DOI CiNii J-GLOBAL

  • 放射線機能触媒を用いた高効率ハイブリッド水素製造法の開発

    古谷正裕, 藤嶋昭, 広石大介, 岡本孝司, 賞雅寛而, 魚谷正樹

    日本原子力学会春の年会要旨集   40th   SO10  2002

    J-GLOBAL

  • 電磁波照射によるナノ磁性微粒子の発熱を利用した腫よう細胞ターゲティング療法の開発

    木谷貴雄, 古谷正裕, 長尾洋昌, 佐藤裕子, 内山明彦

    日本ME学会大会論文集   41st   198  2002

    J-GLOBAL

  • 放射線誘起表面活性を用いた防食

    古谷正裕, 賞雅寛而, 岡本孝司

    マリンエンジニアリング学術講演会講演論文集   67th   125 - 128  2002

    J-GLOBAL

  • Spontaneous Nucleation and Fine-Scale Mixing Structure in the Vapor Explosion.

    古谷正裕, 木下泉

    日本伝熱シンポジウム講演論文集   39th ( Vol.1 ) 3 - 4  2002

    J-GLOBAL

  • Radiation Induced Boiling Enhancement. 5th Report, Leidenfrost Temperature and Quenching Condition.

    今井康之, 深町典博, 賞雅寛而, 岡本孝司, 三島嘉一郎, 古谷正裕

    日本伝熱シンポジウム講演論文集   39th ( Vol.2 ) 461 - 462  2002

    J-GLOBAL

  • Development of Corrosion Control Method on the Basis of Radiation Induced Surface Activation

    FURUYA Masahiro, UEMATSU Susumu, TAKAMASA Tomoji, OKAMOTO Koji, HIROISHI Daisuke, KAWAMURA Hirotaka

    Transactions of the Atomic Energy Society of Japan   1 ( 2 ) 242 - 243  2002

    DOI CiNii J-GLOBAL

  • 超高粘性スラグを対象とした球形化・微粒化手法の検討

    古谷正裕, 市川和芳, 西村聡

    日本混相流学会年会講演会講演論文集   2002   125 - 126  2002

    J-GLOBAL

  • ジルカロイ‐2被覆管の耐食性に及ぼす熱流束,ボイド率および温度の影響

    河村浩孝, 神戸弘巳, 古谷正裕, 平野秀朗

    日本原子力学会秋の大会予稿集   2002   584  2002

    J-GLOBAL

  • 放射線誘起表面活性を利用した防食法の開発

    古谷正裕, 植松進, 賞雅寛而, 岡本孝司, 広石大介, 河村浩孝, 白鳥義行

    日本原子力学会秋の大会予稿集   2002 ( 2 ) 586 - 243  2002

     View Summary

    BWR構造材の応力腐食割れ低減方策として,冷却材に水素を注入したり,構造材に貴金属を担持させることにより,腐食電位を応力腐食割れ発生のしきい値より卑化させる手法が一部試みられている。本研究ではこれらの代替として,放射線誘起表面活性効果(Radiation Induced Surface Activation)を利用して腐食電位を低下させる手法を提案する。すなわち,酸化金属皮膜にγ線などの放射線を照射することにより,軌道電子が伝導帯に励起され,同時にホールができることによりアノード電流が流れるという,非消耗型の腐食緩和手法である。

    CiNii J-GLOBAL

  • A Linear Stability Analysis of a Vapor Film in Terms of the Triggering of Vapor Explosions.

    古谷正裕, 松村邦仁, 木下泉

    日本機械学会論文集 B編   68 ( 675 ) 3176 - 3182  2002

     View Summary

    A Detailed analytical model to explain the vapor film collapse was developed to evaluate the occurrence conditions of self-triggering vapor explosions. The following conclusions were drawn based on linear stability analysis using the thermo-dynamic property of water, by linearizing and perturbing basic equations (Rayleigh-Lamb-Plesset's bubble momentum equation, the mass conservation equation, the state equation for ideal gas, and the Clausius-Clapeyron equation). The vapor film stabilizes with the reduction of the hot-liquid diameter, decreasing the condensation heat transfer coefficient, and increasing the thermal radiation coefficient. The cold-liquid viscosity and surface tension have a stabilizing effect, though this effect is negligibly small where the hot-liquid diameter is over 1 mm. The analysis predicts the vapor explosion occurrence limits obtained experimentally by other researchers to within approximately 10 K.

    DOI CiNii J-GLOBAL

  • A linear stability analysis of a vapor film in terms of the triggering of vapor explosions

    Furuya, M., Matsumura, K., Kinoshita, I.

    Journal of Nuclear Science and Technology   39 ( 10 ) 1026 - 1032  2002

     View Summary

    A detailed analytical model to explain the vapor film collapse was developed to evaluate the occurrence conditions of self-triggering vapor explosions. The following conclusions were drawn based on linear stability analysis using the thermo-physical property of water, by linearizing and perturbing basic equations (Rayleigh-Lamb-Plesset's bubble momentum equation, the mass conservation equation, the state equation for ideal gas, and the Clausius-Clapeyron equation). The vapor film stabilizes with the reduction of the hot-liquid diameter, decrease of the condensation heat transfer coefficient, or increase of the thermal radiation coefficient. The cold-liquid viscosity and surface tension have a stabilizing effect, though this effect is negligibly small when the hot-liquid diameter is over 1 mm. The analysis predicts the vapor explosion occurrence limits obtained experimentally by other researchers within approximately 10 K. A simple correlation for the stability boundary is proposed by simplifying the above detailed model: the difference in cold-liquid temperature at the stability boundary between these models is less than 1 K when the condensation heat transfer coefficient is over 104W/m2·K and the hot-liquid temperature is lower than 2,000°C. © 2002 Taylor and Francis Group, LTD.

    DOI CiNii

  • Effect of Zinc Injection on BWR Fuel Cladding Corrosion. (Part 1). Study on an Accelerated Corrosion Condition to Evaluate Corrosion Resistance of Zircaloy-2 Fuel Claddiing.

    河村浩孝, 神戸弘巳, 古谷正裕

    電力中央研究所狛江研究所報告   ( T01040 ) 1 - 14,巻頭1〜4  2002

    CiNii J-GLOBAL

  • Visualization of Liquid-Liquid Interfacial Phenomena and Occurrence Conditions of Vapor Explosion.

    古谷正裕, 木下泉

    日本機械学会年次大会講演論文集   2002 ( Vol.4 ) 257 - 258  2002

     View Summary

    The triggering process of vapor explosion was visualized by high-speed digital video cameras at an interface between a water droplet and a molten tin pool. A cluster of bubble generated by spontaneous bubble-nucleation covered the whole contact area at 0.10ms after the impingement. Prominent fine mixing between two liquids were found to start at 0.55ms that resulting in vapor explosion. A water micro jet toward the molten tin surface might induce the fine mixing after the bubbles had expanded excessively and had contacted the subcooled water.

    DOI CiNii J-GLOBAL

  • Effect of Dissolved Non-condensible Gas on Vapor Explosion.

    古谷正裕, 木下泉, 藤江政武

    日本機械学会熱工学部門講演会講演論文集   2002 ( 0 ) 335 - 336  2002

     View Summary

    In order to investigate the triggering mechanism of vapor explosion, experiments were carried out in the system of a carbonate water droplet impinging onto a molten lead-bismuth pool. Carbon dioxide dissolved into water had a suppression effect on vapor explosion. The non-condensible gas reduced condensation rate, which resulted in restraining a water micro jet from impinging on to the molten lead-bismuth surface to induce the fine mixing after the bubbles had expanded excessively and had contacted the subcooled water.

    DOI CiNii J-GLOBAL

  • Development of Innovative Rapid Cooling and Liquid Atomization Process, CANOPUS, and Application to Atomization of High-Viscous Fluld

    FURUYA Masahiro, ICHIKAWA Kazuyoshi, NISHIMURA Satoshi

    微粒化シンポジウム講演論文集 = Symposium (ILASS-Japan) on Atomization   10   17 - 22  2001.12

    CiNii

  • Development of Rapid Cooling and Fragmentation Process Making the Best Use of Vapor Explosion Phenomenon.

    古谷正裕

    日本伝熱シンポジウム講演論文集   38th ( Vol.3 ) 831 - 832  2001

    J-GLOBAL

  • Effects of Mechanical Constraint and Thermal Capacitance on the Vapor Explosions.

    古谷正裕, 木下泉, 西村聡

    日本機械学会関東支部総会講演会講演論文集   7th   245 - 246  2001

    J-GLOBAL

  • Evaluation of the Flow at the Contraction of a Heat Exchanger(Part 2) - Effect of Thermal-Hydraulic Factors on Scale Deposition at the Contraction.

    米田公俊, 安尾明, 稲田文夫, 古谷正裕

    電力中央研究所狛江研究所報告   ( T00034 ) 19P  2001

    J-GLOBAL

  • 矩形流路の縮流部における減圧沸騰現象

    米田公俊, 安尾明, 稲田文夫, 古谷正裕

    日本混相流学会年会講演会講演論文集   2001   95 - 96  2001

    J-GLOBAL

  • BWR炉心および領域安定性試験設備SIRIUSの開発

    古谷正裕, 稲田文夫, 安尾明

    日本混相流学会年会講演会講演論文集   2001   7 - 10  2001

    J-GLOBAL

  • 溶融銅とナトリウムの熱的相互作用に関する実験的研究 (III) 破砕に及ぼすウェーバー数の影響

    西村聡, 古谷正裕, 木下泉, 杉山憲一郎, 岡田亮兵

    日本原子力学会秋の大会予稿集   2001   576  2001

    J-GLOBAL

  • BWR安定性試験設備SIRIUSによる自然循環BWRの炉心および領域安定性評価

    古谷正裕, 稲田文夫, 安尾明

    日本原子力学会秋の大会予稿集   2001   513  2001

    J-GLOBAL

  • Flashing at the Contraction of a Heat Exchanger.

    米田公俊, 安尾明, 稲田文夫, 古谷正裕

    日本機械学会年次大会講演論文集   2001 ( Vol.5 ) 275 - 276  2001

    J-GLOBAL

  • Core-Wide and Regional Stability Evaluation of Natural Circulation BWR on the Basis of Noise Analysis.

    古谷正裕, 稲田文夫, 安尾明

    日本機械学会年次大会講演論文集   2001 ( Vol.5 ) 289 - 290  2001

    J-GLOBAL

  • Experimental Study on Convective Heat Transfer with Thin Porous Bodies.

    西義久, 木下泉, 古谷正裕

    日本機械学会論文集 B編   67 ( 661 ) 2274 - 2280  2001

     View Summary

    Experimental studies are made on the convective heat transfer of three types of thin porous bodies. Heat transfer performances, flow patterns and temperature profiles near the porous bodies are compared with each other. The heat transfer performance of porous bodies with the largest pore diameter is large. It became clear that the high heat transfer performance depends on an excellent heat transportation ability inside the pore and near the surface of the porous bodies.

    DOI CiNii J-GLOBAL

  • Development of Core-Wide and Regional Stability Test Facility, SIRIUS, that Simulates Void Reactivity Feedback, and Stability Evaluation.

    古谷正裕, 久保雄一郎, 稲田文夫, 安尾明

    日本原子力学会誌   43 ( 10 ) 1027 - 1038  2001

     View Summary

    The SIRIUS facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of the BWR. A real-time simulation was performed in the digital controller for modal point kinetics of reactor neutronics and fuel-rod conduction on the basis of measured void fractions in reactor core sections of the thermal-hydraulic loop, which simulates thermal-hydraulics of a natural circulation BWR. A solid-state, series-regulated power supply, that plays a role of simulation output, was designed to attain fast response speed without loss of accuracy.<BR>A noise analysis method was performed to calculate decay ratios from dominant poles of a transfer function by applying AR method to time series of the core inlet flow rate. Experiments were conducted with the SIRIUS facility for the nominal operating condition of 3.13GWt natural circulation BWR. Channel and regional stability decay ratios were determined to be 0.38 and 0.54, respectively, which indicates sufficient margin for the instabilities. Experiments were extended to evaluate the stability sensitivity of the design parameters such as the power profile on the basis of three-dimensional steady-state analysis, the void reactivity coefficients, the core inlet subcooling, and the thermal conductance of the fuel rod.

    DOI CiNii J-GLOBAL

  • Development of Rapid Cooling Process Utilizing Vapor Explosion Phenomenon and Application to Amorphizaion of Practical Materials.

    古谷正裕

    日本機械学会熱工学部門講演会講演論文集   2001   423 - 424  2001

    J-GLOBAL

  • 革新的な超急冷・液体微粒化手法CANOPUSの開発と高粘性流体の微粒化

    古谷正裕, 市川和芳, 西村聡

    微粒化シンポジウム講演論文集   10th   17 - 22  2001

    J-GLOBAL

  • 829 Effects of Mechanical Constraint and Thermal Capacitance on the Vapor Explosions

    FURUYA Masahiro, KINOSHITA Izumi, NISHIMURA Satoshi

    The Proceedings of Conference of Kanto Branch   2001 ( 7 ) 245 - 246  2001

     View Summary

    Experiments werec conducted to investigate the triggering phenomenon in vapor explosion for various molten alloy pool depths and water droplet diameters in a droplet impingement system. The effect of the water droplet curvature Was found to be negligibly small when the droplet diameter is larger than 4.5mm. Vapor explosion conditions were identical for the depths ranging from 0.5 to 40mm. The length scale required in fragmentation and fine-scale mixing in the triggering event was considered to be less than 0.5mm.

    DOI CiNii

  • Experimental Investigation of Natural Circulation BWR Core-Wide and Regional Stability on the basis of Time Series Analysis.

    古谷正裕, 稲田文夫, 安尾明

    電力中央研究所狛江研究所報告   ( T01001 ) 1 - 12,巻頭1〜4  2001

    CiNii J-GLOBAL

  • K-2209 Flashing at the Contraction of a Heat Exchanger

    YONEDA Kimitoshi, YASUO Akira, INADA Fumio, FURUYA Masahiro

    The proceedings of the JSME annual meeting   1 ( 1 ) 275 - 276  2001

     View Summary

    In heat exchangers of power plants, tube supports may be exposed to scale deposition. As the explanation for this phenomenon, flashing is said to be one of the factor. In this study, flashing at the leading edge of contraction in a rectangular flow path were observed. Pressure drop and steam void characteristics at the flashing occurrence point were measured. The phenomenon were also investigated by photographs. Between two types of contraction, with same contraction ratio and different hydraulic diameter, considerable difference of boundary temperature of flashing occurrence were seen. Most of the bubbles detected at the leading edge of contraction were very small while those at the upstream included bubbles of many sizes.

    DOI CiNii

  • K-2216 Core-Wide and Regional Stability Evaluation of Natural Circulation BWR on the Basis of Noise Analysis

    FURUYA Masahiro, INADA Fumio, YASUO Akira

    The proceedings of the JSME annual meeting   1 ( 1 ) 289 - 290  2001

     View Summary

    A Noise analysis method was performed to calculate decay ratios from dominant poles of a transfer function by applying AR method to time series of the core inlet flow rate. Experiments were conducted with the SIRIUS facility, which simulates a representative natural circulation BWR. Channel and regional stability decay ratios at the nominal operating condition were determined to be 0.38 and 0.54, respectively, which indicates sufficient margin for the instabilities. Experiments were extended to investigate the effects of the design parameters on stability.

    DOI CiNii

  • B222 Development of Rapid Cooling Process Utilizing Vapor Explosion Phenomenon and Application to Amorphizaion of Practical Materials

    FURUYA Masahiro

    Proceedings of thermal engineering conference   2001 ( 0 ) 423 - 424  2001

     View Summary

    Sustainable small-scale spontaneous vapor explosion was proposed to utilize for the innovative rapid cooling and the liquid atomizing processes, namely the CANOPUS method. On the basis of CANOPUS method, a highly viscous coal gasification slag was atomized into 30μm powders for use of cement raw materials. The cooling rate of Al_<89>-Si_<11> particle was up to 1.5×l0^8K/s, which is more than 280 times higher than the conventional cooling method.

    DOI CiNii

  • ボイド反応度フィードバックを模擬したBWR安定性試験設備SISIUSの開発

    古谷 正裕, 稲田 文夫, 安尾 明

    電力中央研究所報告 研究報告 T   ( 1 ) 1 - 19,巻頭1〜4  2000.09

    CiNii

  • Effects of Additives on Triggering of Vapor Explosion.

    古谷正裕, 坂本陽, 木下泉, 西村聡

    日本伝熱シンポジウム講演論文集   37th ( Vol.2 ) 457 - 458  2000

    J-GLOBAL

  • 蒸気爆発トリガリング事象における界面活性剤添加効果

    古谷正裕, 坂本陽, 木下泉, 西村聡

    日本原子力学会春の年会要旨集   38th   485  2000

    J-GLOBAL

  • 気泡微細化沸騰の発生条件に関する基礎的検討

    古谷正裕, 安尾明

    日本混相流学会年会講演会講演論文集   1st   251 - 252  2000

    J-GLOBAL

  • Development of SIRIUS facility that simulates void reactivity feedback and thermal-hydraulics of BWRs.

    古谷正裕, 稲田文夫, 安尾明

    電力中央研究所狛江研究所報告   ( T00001 ) 24P  2000

    J-GLOBAL

  • (7)蒸気爆発におけるトリガリング事象の研究

    古谷 正裕

    日本機械学會誌   103 ( 978 ) 296 - 296  2000

    DOI CiNii

  • Interfacial Phenomena in the Triggering Process of Vapor Explosions.

    古谷正裕, 木下泉, 西村聡

    日本機械学会年次大会講演論文集   2000 ( Vol.1 ) 807 - 808  2000

     View Summary

    The triggering process of vapor explosion was visualized by high-speed digital video cameras at an interface between a water droplet and a molten alloy pool. A ring of 3mm diameter was seen at 0.01ms after the impingement and is considered to be a cluster of bubbles generated by spontaneous bubble-nucleation. The cluster of bubble covered the whole contact area at 0.05ms. Prominent fine mixing between two liquids were found to start at 0.70ms that resulting in vapor explosion.

    DOI CiNii J-GLOBAL

  • Suppression Effect of Viscous and Surfactant Additives on Vapor Explosions.

    古谷正裕, 木下泉

    電力中央研究所狛江研究所報告   ( T99091 ) 1 - 4,1〜12  2000

    CiNii J-GLOBAL

  • Experimental Evaluation on Regional and Core-Wide Stability of Natural Circulation BWRs.

    古谷正裕, 稲田文夫, 安尾明

    電力中央研究所狛江研究所報告   ( T00002 ) 1 - 14,巻頭1〜4  2000

    CiNii J-GLOBAL

  • A Study on Thermo-Hydraulic Instability of Boiling Natural Circulation Loop with a Chimney. A Consideration of the Effect of Axial Higher Mode.

    稲田文夫, 古谷正裕, 安尾明

    日本機械学会動力・エネルギー技術シンポジウム講演論文集   7th ( 7 ) 269 - 274  2000

     View Summary

    Thermo-hydraulic instabilities of a boiling natural circulation loop with a chimney under low and high pressure were investigated using linear stability analysis. The effect of nuclear coupling was also considered. Both in low- and high-pressure conditions, instability could occur when exit quality was relatively low. In low-pressure condition, flashing only near the exit of the chimney could induce instability, and enthalpy wave of single-phase flow was generated in the chimney. In high-pressure condition, void was generated near channel exit, and void wave propagated in the chimney. In the high pressure and high power condition, though flow could be very stable, the decay ratio of higher mode could be larger than that of lower mode. The sensitivity on decay ratio to the thermal power, inlet subcooling, void reactivity feedback coefficient and so on could be very low when there was a long chimney.

    DOI CiNii J-GLOBAL

  • An Effect of Thermo-Physical Property of Hot Liquid on the Vapor Explosion.

    古谷正裕, 木下泉, 西義久, 西村聡

    日本伝熱シンポジウム講演論文集   36th ( Vol.3 ) 635 - 636  1999

    J-GLOBAL

  • Study on Mechanical Interaction between Molten Alloy and Water.

    西村聡, 植田伸幸, 西義久, 古谷正裕, 木下泉

    電力中央研究所狛江研究所報告   ( T98009 ) 34P  1999

    J-GLOBAL

  • Development of Vapor Collapse Model for Vapor Explosions.

    古谷正裕, 木下泉, 松村邦仁

    電力中央研究所狛江研究所報告   ( T98070 ) 1 - 4,1〜13  1999

    CiNii J-GLOBAL

  • An analytical estimation of thermo-hydraulic stability of natural circulation BWR with nuclear coupling.

    稲田文夫, 古谷正裕, 安尾明

    電力中央研究所狛江研究所報告   ( T98076 ) 1 - 4,1〜10  1999

    CiNii J-GLOBAL

  • Experimental Study on Forced Convection Boiling Heat Transfer on Molten Alloy.

    西村聡, 植田伸幸, 西義久, 古谷正裕, 木下泉, 山口剛司

    日本伝熱シンポジウム講演論文集   35th ( Vol.1 ) 191 - 192  1998

    J-GLOBAL

  • lnvestigation of Triggering Phenomenon in the Vapor Explosion.

    古谷正裕, 木下泉

    電力中央研究所狛江研究所報告   ( T97044 ) 20P  1998

    J-GLOBAL

  • A Study on Thermo-Hydraulic Instability of a Boiling Natural Circulation Loop with Chimney. An Analytical Consideration about Instability of High Pressure Condition.

    稲田文夫, 古谷正裕, 安尾明

    日本機械学会全国大会講演論文集   76th ( Vol.3 ) 457 - 458  1998

    J-GLOBAL

  • Surface Property Effect on the Vapor Explosions.

    古谷正裕, 木下泉, 西義久, 西村聡

    日本機械学会全国大会講演論文集   76th ( Vol.3 ) 419 - 420  1998

    J-GLOBAL

  • Visualization of Liquid Metal Flow by Neutron Radiography

    TAKENAKA Nobuyuki, ASANO Hitoshi, FUJII Terushige, NISHI Yoshihisa, FURUYA Masahiro, KINOSHITA Izui, MATSUBAYASHI Masahito

    Journal of the Visualization Society of Japan   17 ( 1 ) 59 - 60  1997

     View Summary

    Liquid metal flows were visualizaed by using neutron radiography. Single-phase lead-bismuth eutectic flow was visualized by a tracer method. High speed visualization of water evaporation behaviors in lead-bismuth-tin alloy was carried out. Thermal neutron radiography system at JRR-3M in JAERI was used. It was shown that neutron radiography was applicable to visualize single-and two-phase flow of heavy metal.

    CiNii

  • Visualization of Direct Contact Heat Transfer Between Water and Molten Alloy by Neutron Radiography. (1st Report).

    西義久, 古谷正裕, 木下泉, 竹中信幸, 松林政仁

    日本伝熱シンポジウム講演論文集   34th ( Vol.2 ) 523 - 524  1997

    J-GLOBAL

  • Thermal-Hydraulic Instability of a Boiling Natural Circulation Loop with a Chimney. 3rd Report: Effect of Inlet Throttling.

    古谷正裕, 稲田文夫, 安尾明

    日本伝熱シンポジウム講演論文集   34th ( Vol.1 ) 205 - 206  1997

    J-GLOBAL

  • Effect of the Inlet Throttling on the Thermal-Hydraulic Instability of the Natural Circulation BWR.

    古谷正裕, 稲田文夫, 米田公俊

    電力中央研究所狛江研究所報告   ( T96024 ) 15P  1997

    J-GLOBAL

  • Feasibility Study on Applicability of Direct Contact Heat Transfer SGs for FBRs.

    木下泉, 西義久, 古谷正裕

    電力中央研究所狛江研究所総合報告   ( T49 ) 39P  1997

    J-GLOBAL

  • Visualization of Steam Bubbles with Evaporation in Molten Alloy.

    西義久, 古谷正裕, 木下泉, 竹中信幸, 松林政仁

    電力中央研究所狛江研究所報告   ( T96058 ) 15P  1997

    J-GLOBAL

  • Thermal Interaction between Molten Alloy Pool and Impingement Droplet.

    古谷正裕, 木下泉, 西義久

    日本原子力学会秋の大会予稿集   1997   485  1997

    J-GLOBAL

  • Visualization of Liquid Metal Flow by Neutron Radiography.

    竹中信幸, 浅野等, 藤井照重, 西義久, 古谷正裕, 木下泉, 松林政仁

    可視化情報学会誌   17 ( Suppl 1 ) 59 - 60  1997

     View Summary

    Liquid metal flows were visualizaed by using neutron radiography. Single-phase lead-bismuth eutectic flow was visualized by a tracer method. High speed visualization of water evaporation behaviors in lead-bismuth-tin alloy was carried out. Thermal neutron radiography system at JRR-3M in JAERI was used. It was shown that neutron radiography was applicable to visualize single-and two-phase flow of heavy metal.

    DOI CiNii J-GLOBAL

  • Thermal-Hydraulic Instability of Boiling Natural Circulation Loop with a Chimney. 3rd Report. Instability at High System Pressure.

    古谷正裕, 稲田文夫, 安尾明

    日本機械学会論文集 B編   63 ( 612 ) 2757 - 2763  1997

     View Summary

    Experiments have been conducted to investigate thermal-hydraulic instabilities at system pressure ranging from 1 to 7.2 MPa in a boiling natural circulation loop with a chimney. Stability maps in reference to the system pressure, the channel inlet subcooling, and heat flux are presented. Two different types of instabilities were observed in the facility at relatively low and high system pressures. Both of the instability mechanisms were clarified by investigation of the transient flow pattern and the response of the driving force of the circulation to momentum energy.

    DOI CiNii J-GLOBAL

  • Convective Heat Transfer Characteristics of Thin Porous Plates. Effect of Pore Size under Natural Convection.

    西義久, 木下泉, 古谷正裕

    日本伝熱シンポジウム講演論文集   33rd ( Vol 2 ) 391 - 392  1996

    J-GLOBAL

  • Thermal-Hydraulic Instability of a Boiling Natural Circulation Loop with a Chimney. 2nd Report. Instability under the Higher System Pressure.

    古谷正裕, 稲田文夫, 安尾明

    日本伝熱シンポジウム講演論文集   33rd ( Vol 3 ) 821 - 822  1996

    J-GLOBAL

  • Thermal-Hydraulic Instability of the Natural Circulation BWR. 7th Report. Analytical Estimation of the Instability at the Higher System Pressure.

    古谷正裕, 稲田文夫, 安尾明

    電力中央研究所狛江研究所報告   ( T95066 ) 15P  1996

    J-GLOBAL

  • Visulization of Direct Contact Heat Transfer between Water and Molten Alloy.

    西義久, 古谷正裕, 木下泉, 竹中信幸, 松林政人

    電力中央研究所狛江研究所報告   ( T95061 ) 20P  1996

    J-GLOBAL

  • Thermal-Hydraulic Instability of the Natural Circulation BWR. 6th Report: Occurrence Condition and Mechanism of the Instability at the Higher System Pressure.

    古谷正裕, 稲田文夫, 安尾明

    電力中央研究所狛江研究所報告   ( T95065 ) 20P  1996

    J-GLOBAL

  • Investigations of the Occurrence Conditions and the Mechanism of Minute Bubble Emission Boiling.

    古谷正裕

    電力中央研究所狛江研究所報告   ( T95095 ) 29P  1996

    J-GLOBAL

  • Study on Enhancement Heat Transfer of Reactor Vessel Auxiliary Cooling System of Fast Breeder Reactor.

    西義久, 木下泉, 植田伸幸, 古谷正裕

    電力中央研究所狛江研究所総合報告   ( T45 ) 69P  1996

    J-GLOBAL

  • Development of Steam Generators for FBR with Direct Contact Heat Transfer between Melting Alloy and Water.

    木下泉, 西義久, 古谷正裕

    日本機械学会動力・エネルギー技術シンポジウム講演論文集   5th   124 - 129  1996

    J-GLOBAL

  • Critical Heat Flux and Convective Heat Transfer with a Two-Dimensional Liquid Jet Impinging on Flat and Concave Surfaces.

    Furuya Masashiro, Inoue Akira, Tanno Ryuji

    TRANSACTIONS OF THE JAPAN SOCIETY OF MECHANICAL ENGINEERS Series B   61 ( 591 ) 4094 - 4100  1995

     View Summary

    The diverter surface of the ITER Fusion Engineering Reactor is exposed to strong radiation locally up to 20 MW/m&lt;SUP&gt;2&lt;/SUP&gt;. We have proposed a diverter cooling system which consists of concave surfaces cooled by two-dimensional liquid jets. Experiments are conducted to investigate local heat transfer coefficeint and critical heat flux on flat and concave surfaces under various cooling conditions. Based on photographic study, a critical heat flux model was derived by modifying a Haramura&#039;s model to take account of the subcooling effect. The proposed correlation based on this model was in good agreement with the experimental data.

    CiNii

  • Thermohydraulic Instability of Boiling Natural Circulation Loop with a Chimeny. 1st Report, Linear Stability Analysis Using Homogenous Two-Phase Flow Model and Experiment on Thermohydraulic Instability Induced by Flashing.

    Inada Fumio, Furuya Masahiro, Yasuo Akira

    TRANSACTIONS OF THE JAPAN SOCIETY OF MECHANICAL ENGINEERS Series B   61 ( 591 ) 4067 - 4073  1995

     View Summary

    Instability of a boiling natural circulation loop with a chimney at low pressure and low heater power was investigated by linear stability analysis and experiment. A homogeneous and thermodynamic equilibrium model for two-phase flow was used. The effect of flashing induced by pressure drop in the heated channels and the chimney was considered. The effects of coupling between two boiling channels were investigated. It was found that in-phase-mode instability was apt to occur when channel inlet subcooling was large and boiling began in the chimney. In-phase-mode instability easily occurred when channel length became short and the chimney became long. Out-of-phase-mode instability was apt to occur when chimney length became small and boiling began in the channel. It was suggested that in-phase-mode instability was density weve oscillation induced by flashing in the chimney and out-of-phase-mode instability was density wave oscillation induced by boiling in the channels. The analytical results agreed qualitatively with experimental results.

    CiNii

  • A study on Thermo-Hydraulic Instability of Boiling Natural Circulation Loop with a Chimeny. 2nd Report, Experimental Approach to Clarify the Flow Instability in Detail.

    Furuya Masahiro, Inada Fumio, Yasuo Akira

    TRANSACTIONS OF THE JAPAN SOCIETY OF MECHANICAL ENGINEERS Series B   61 ( 591 ) 4074 - 4080  1995

     View Summary

    Experiments are are conducted to investigate two-phase flow instabilities in a boiling natural circulation loop with a chimney induced by flashing in the chimney at lower pressure. The type of instability that occurred in the experiments is suggested to be density wave oscillations induced by flashing in the chimney. The differences from other instabilities such as geysering, flow pattern transition instability, and natural circulation oscillations are discussed on the basis of the dynamic characteristics, the oscillation period, and the transient flow resume.

    CiNii

  • Thermal Interaction and Evaporation Characteristics of a Droplet Impinging on a Melt Surface.

    古谷正裕, 木下泉, 西義久

    日本伝熱シンポジウム講演論文集   32nd ( Vol 3 ) 791 - 792  1995

    J-GLOBAL

  • Convective Heat Transfer Characteristics and Flow Resistance of Thin Porous Plates.

    西義久, 木下泉, 古谷正裕

    日本伝熱シンポジウム講演論文集   32nd ( Vol 3 ) 645 - 646  1995

    J-GLOBAL

  • Heat Transfer Characteristics of a Direct Contact Heat Exchanger with Low Melting Point Alloy and Water. Part 3: Effect of Pressure on Stem Temperature.

    木下泉, 西義久, 古谷正裕

    日本伝熱シンポジウム講演論文集   32nd ( Vol 3 ) 793 - 794  1995

    J-GLOBAL

  • Evaporation Characterics of a Water Droplet Impinging on a Melt Surface (2nd Report: Splashing Phenomena).

    古谷正裕, 木下泉, 西義久

    電力中央研究所狛江研究所報告   ( T94044 ) 28P  1995

    J-GLOBAL

  • Thermo-Hydraulic Instability of the Natural Circulation BWR (5th Report). An Analytical Method to Estimate Thermo-Hydraulic Instability at Start-up.

    稲田文夫, 古谷正裕, 安尾明

    電力中央研究所狛江研究所報告   ( T94061 ) 30P  1995

    J-GLOBAL

  • Innovative Steam Generator for FBRs with Direct Contact Heat Transfer. Evaluation of Size, Safety and Material Compatibility.

    木下泉, 西義久, 古谷正裕

    電力中央研究所狛江研究所報告   ( T95006 ) 21P  1995

    J-GLOBAL

  • Development of a Direct Contact Heat Transfer Steam Generator for FBRs.

    木下泉, 西義久, 古谷正裕

    日本機械学会全国大会講演論文集   73rd ( Vol 3 ) 91 - 92  1995

    J-GLOBAL

  • Study on Enhancement of Convective Heat Transfer with Thin Porous Plates. Effect of Pore Diameter.

    西義久, 木下泉, 古谷正裕

    日本機械学会熱工学部門講演会講演論文集   1995   46 - 48  1995

    J-GLOBAL

  • Splashing Phenomena in the System of a Droplet Impinging on the Melt Pool.

    古谷正裕, 木下泉, 西義久

    日本機械学会熱工学部門講演会講演論文集   1995   135 - 137  1995

    J-GLOBAL

  • Direct Contact Heat Transfer Characteristics between Melting Alloy and Water.

    木下泉, 西義久, 古谷正裕

    日本機械学会論文集 B編   61 ( 588 ) 3038 - 3043  1995

     View Summary

    As a candidate for an innovative steam generator for fast breeder reactors, a heat exchanger with direct contact heat transfer between melting alloy and water was proposed. The evaluation of heat transfer characteristics of this heat exchanger is one of the research subjects for the design and development of the steam generator. In this study, the effect of the pressure on heat transfer characteristics and the required degree of superheating of melting alloy above water saturation temperature are evaluated during the direct contact heat transfer experiment by injecting water into Wood's alloy. In the experiment, the pressure, the temperature of the Wood's alloy, the flow rate of feed water, and the depth of the feed water injection point are varied as parameters. As a result of the experiment, the product of the degree of Wood's alloy superheating above water saturation temperature and the depth of the feed water injection point is constant for each pressure. This constant increases as the pressure rises.

    DOI CiNii J-GLOBAL

  • Thermohydraulic Instability of Boiling Natural Circulation Loop with a Chimeny. 1st Report, Linear Stability Analysis Using Homogenous Two-Phase Flow Model and Experiment on Thermohydraulic Instability Induced by Flashing.

    稲田文夫, 古谷正裕, 安尾明

    日本機械学会論文集 B編   61 ( 591 ) 4067 - 4073  1995

     View Summary

    Instability of a boiling natural circulation loop with a chimney at low pressure and low heater power was investigated by linear stability analysis and experiment. A homogeneous and thermodynamic equilibrium model for two-phase flow was used. The effect of flashing induced by pressure drop in the heated channels and the chimney was considered. The effects of coupling between two boiling channels were investigated. It was found that in-phase-mode instability was apt to occur when channel inlet subcooling was large and boiling began in the chimney. In-phase-mode instability easily occurred when channel length became short and the chimney became long. Out-of-phase-mode instability was apt to occur when chimney length became small and boiling began in the channel. It was suggested that in-phase-mode instability was density weve oscillation induced by flashing in the chimney and out-of-phase-mode instability was density wave oscillation induced by boiling in the channels. The analytical results agreed qualitatively with experimental results.

    DOI CiNii J-GLOBAL

  • A study on Thermo-Hydraulic Instability of Boiling Natural Circulation Loop with a Chimeny. 2nd Report, Experimental Approach to Clarify the Flow Instability in Detail.

    古谷正裕, 稲田文夫, 安尾明

    日本機械学会論文集 B編   61 ( 591 ) 4074 - 4080  1995

     View Summary

    Experiments are are conducted to investigate two-phase flow instabilities in a boiling natural circulation loop with a chimney induced by flashing in the chimney at lower pressure. The type of instability that occurred in the experiments is suggested to be density wave oscillations induced by flashing in the chimney. The differences from other instabilities such as geysering, flow pattern transition instability, and natural circulation oscillations are discussed on the basis of the dynamic characteristics, the oscillation period, and the transient flow resume.

    DOI CiNii J-GLOBAL

  • Critical Heat Flux and Convective Heat Transfer with a Two-Dimensional Liquid Jet Impinging on Flat and Concave Surfaces.

    古谷正裕, 井上晃, 丹野隆治

    日本機械学会論文集 B編   61 ( 591 ) 4094 - 4100  1995

     View Summary

    The diverter surface of the ITER Fusion Engineering Reactor is exposed to strong radiation locally up to 20 MW/m2. We have proposed a diverter cooling system which consists of concave surfaces cooled by two-dimensional liquid jets. Experiments are conducted to investigate local heat transfer coefficeint and critical heat flux on flat and concave surfaces under various cooling conditions. Based on photographic study, a critical heat flux model was derived by modifying a Haramura's model to take account of the subcooling effect. The proposed correlation based on this model was in good agreement with the experimental data.

    DOI CiNii J-GLOBAL

  • Heat Transfer Characteristics of a Direct Contact Heat Exchanger with Low Melting Point Alloy and Water. Part 2: Effect of pressure.

    木下泉, 西義久, 古谷正裕

    日本伝熱シンポジウム講演論文集   31st ( Pt 3 ) 1030 - 1032  1994

    J-GLOBAL

  • Leidenfrost Phenomena on Melt Surfaces.

    古谷正裕, 木下泉, 西義久

    日本伝熱シンポジウム講演論文集   31st ( Pt 2 ) 445 - 447  1994

    J-GLOBAL

  • Density Wave Oscillation of Boiling Natural Circulation Loop with a Chimney at Low Pressure. In the case of 2 boiling channels.

    稲田文夫, 古谷正裕, 大川富雄

    日本機械学会通常総会講演会講演論文集   71st ( Pt 3 ) 791 - 793  1994

    J-GLOBAL

  • Effect of Pressure on Direct Contact Heat Transfer Characteristics between Liquid Metal and Water.

    木下泉, 古谷正裕

    電力中央研究所狛江研究所報告   ( T93041 ) 24P  1994

    J-GLOBAL

  • Thermo-hydraulic instability of natural circulation BWR. (3rd Report). The effect of coupling between two channels on the stability at low pressure start-up.

    稲田文夫, 古谷正裕, 大川富雄

    電力中央研究所狛江研究所報告   ( T93037 ) 24P  1994

    J-GLOBAL

  • Evaporation of Water Droplet on Heated Low Melting Point Alloy.

    木下泉, 古谷正裕

    電力中央研究所狛江研究所報告   ( T93036 ) 26P  1994

    J-GLOBAL

  • Thermo-Hydraulic Instability of the Natural Circulation BWR. 4th Report: Experimental Approach to Clarify the Flow Instability Phenomena and the Conditions in Detail.

    古谷正裕, 稲田文夫

    電力中央研究所狛江研究所報告   ( T93061 ) 42P  1994

    J-GLOBAL

  • Direct Contact Heat Transfer Characteristics Between Low Melting Point Alloy and Water.

    木下泉, 西義久, 古谷正裕

    日本機械学会熱工学部門講演会講演論文集   1994   201 - 203  1994

    J-GLOBAL

  • Density Wave Oscillations of Boiling Natural Circulation Loop with a Chimney under Lower System Pressure. Experimental Approach to Clarify the Flow Instability in Detail.

    古谷正裕, 稲田文夫

    日本機械学会動力・エネルギー技術シンポジウム講演論文集   4th   435 - 440  1994

    J-GLOBAL

  • Critical Heat Flux of a Concave Surface Cooied by a Two-Dimensional Inpinging Liquid Jet.

    古谷正裕, 丹野隆治, 井上晃

    日本伝熱シンポジウム講演論文集   30th ( Pt 2 ) 433 - 435  1993

    J-GLOBAL

  • Critical Heat Flux in a Forced Convective Boiling with a Two-Dimensional Liquid Jet.

    古谷正裕, 丹野隆治, 井上晃

    日本機械学会全国大会講演論文集   71th ( Pt D ) 468 - 470  1993

    J-GLOBAL

  • UNSTEADY 3-DIMENSIONAL BEHAVIOR OF NATURAL-CONVECTION IN HORIZONTAL ANNULUS .4. APPLICATION OF LES TO TURBULENT NATURAL-CONVECTION

    N TANIGUCHI, Y USUI, M FURUYA, Y MIKI, K FUKUDA, S HASEGAWA

    JOURNAL OF THE ATOMIC ENERGY SOCIETY OF JAPAN   34 ( 10 ) 996 - 1004  1992.10

     View Summary

    The large eddy simulation (LES) method is applied to turbulent natural convection in a horizontal concentric annulus. Numerical results are obtained for Rayleigh numbers based on the gap width from 2.51 x 10(6) to 1.18 x 10(9). These results are compared with other researchers' experimental data and the validity of LES is investigated. It is found at LES is very useful even if at high Ra, at which the DNS is not applicable. However, in the cases where temperature differences between cylinders are large, effect of dependency of physical properties on temperature is not negligible, which causes the poor coincidence betwenn the experiments and the numerical results.

    CiNii

  • Meeting Report of CCITT and CCIR (1991.6~1991.11).

    塩見京太, 鷲田力, 元永康則, 中尾康二, 古谷信雄, 木原賢一, 和田正裕, 大村好則, 村上仁己

    国際通信の研究   ( 147 )  1992

    J-GLOBAL

  • Unsteady Three-Dimensional Behavior Natural Convection in Horizontal Annulus. (IV). Application of LES to Turbulent Natural Convection.

    谷口昇, 碓井志典, 古谷正裕, 三木康臣, 福田研二, 長谷川修

    日本原子力学会誌   34 ( 10 ) 996 - 1004  1992

     View Summary

    The large eddy simulation (LES) method is applied to turbulent natural convection in a horizontal concentric annulus. Numerical results are obtained for Rayleigh numbers based on the gap width from 2.51&times;106 to 1.18&times;109. These results are compared with other researchers' experimental data and the validity of LES is investigated. It is found that LES is very useful even if at high Ra, at which the DNS is not applicable. However, in the cases where temperature differences between cylinders are large, effect of dependency of physical properties on temperature is not negligible, which causes the poor coincidence betwenn the experiments and the numerical results.

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  • Direct numerical simulation and large eddy simulation of natural convection in cylindrical annuli.

    谷口昇, 福田研二, 三木康臣, 確井志典, 古谷正裕, 長谷川修

    日本機械学会計算力学講演会講演論文集   3rd   161 - 162  1990

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    特許第4662122号

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    Patent

    J-GLOBAL

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    特許第4656496号

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    Patent

    J-GLOBAL

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    特許第4623510号

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

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    特許第4623502号

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

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    特許第4623503号

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

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    特許第4604153号

    賞雅 寛而, 波津久 達也, 関村 直人, 岡本 孝司, 阿部 弘亨, 三島 嘉一郎, 中村 秀夫, 柴本 泰照, 植松 進, 古谷 正裕, 小野 昇一, 師岡 慎一, 秋葉 美幸, 鹿野 文寿, 安永 龍哉, 藤沢 匡介, 千草 剛, 下条 純

    Patent

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    特許第4597713号

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

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    特許第4587302号

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    Patent

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    特許第4578274号

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

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    特許第4555704号

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

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    Patent

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    特許第4541929号

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    Patent

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    特許第4541928号

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

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    特許第4534144号

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

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    J-GLOBAL

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    田中 伸幸, 古谷 正裕, 常磐井 守泰, 堀江 正明

    Patent

    J-GLOBAL

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    特許第4502325号

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

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    特許第4480014号

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    Patent

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    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

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    Patent

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    特許第4450320号

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    Patent

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    特許第4437200号

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

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    特許第4428706号

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

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    特許第4428705号

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

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    Patent

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    特許第4399851号

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

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    特許第4392840号

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

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    特許第4368315号

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    特許第4360469号

    古谷 正裕, 市川 和芳, 山本 武志

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    特許第4233027号

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    特許第4201962号

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    特許第4194378号

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    特許第4066320号

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    特許第4059772号

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    Patent

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    古谷 正裕

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    特許第3980050号

    古谷 正裕

    Patent

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    特許第3948738号

    古谷 正裕

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    特許第3948739号

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    Patent

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    Patent

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    Patent

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    Patent

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    Patent

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    J-GLOBAL

  • 多機能性皮膜形成用コーティング組成物

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 冷風生成装置

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 体着具

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 鉄鋼製品及びその製造方法

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 防汚建築物

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 油処理設備

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 電力供給機器

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 遊興具

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 船舶

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 熱交換器

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 鏡面を有する多機能材

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 飛行体

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 耐熱部材

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • キッチン製品および炭素ドープ酸化チタン表層を備えた食器洗浄機

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 耐放射線部材及びそれを用いた原子力発電システム

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • ロケット部品

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • トイレクリーニングシステム

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 鉄道車両

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 建築用資材

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 有機物分解システム

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 医家向け抗菌製品およびその取り扱い方法

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 灌漑装置及び灌漑用部材、並びに、灌漑システム

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 空気清浄装置乃至空気清浄システム

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 脱臭装置乃至脱臭システム

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 厨房システム

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 産業用機械

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 物品保管庫

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • カッター

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 作業用機械

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 防腐装置

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 耐環境性機器

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 宗教用品

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 飲食料品用部材及び食器洗浄装置

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 浄化装置

    古谷 正裕, 常磐井 守泰, 高橋 毅, 小林 博和, 田中 伸幸, 三上 己紀, 黒田 昌宏

    Patent

    J-GLOBAL

  • 沸騰水型原子炉用燃料集合体および沸騰水型原子炉用チャンネルボックス

    賞雅 寛而, 波津久 達也, 関村 直人, 岡本 孝司, 阿部 弘亨, 三島 嘉一郎, 中村 秀夫, 柴本 泰照, 植松 進, 古谷 正裕, 宮野 征巳, 師岡 慎一, 秋葉 美幸, 鹿野 文寿, 安永 龍哉, 藤沢 匡介, 千草 剛, 下条 純

    Patent

    J-GLOBAL

  • 多機能材の製造方法

    古谷 正裕

    Patent

    J-GLOBAL

  • 多機能材

    古谷 正裕

    Patent

    J-GLOBAL

  • 防食性に優れた機能性被覆の形成法

    賞雅 寛而, 波津久 達也, 関村 直人, 岡本 孝司, 阿部 弘亨, 三島 嘉一郎, 中村 秀夫, 柴本 泰照, 植松 進, 古谷 正裕, 小野 昇一, 師岡 慎一, 秋葉 美幸, 鹿野 文寿, 安永 龍哉, 藤沢 匡介, 千草 剛, 下条 純

    Patent

    J-GLOBAL

  • 微粒子の製造方法及び製造装置

    古谷 正裕

    Patent

    J-GLOBAL

  • 高粘性スラグの繊維化及び微粒化装置

    古谷 正裕, 市川 和芳, 山本 武志

    Patent

    J-GLOBAL

  • 超親水性薄膜及びその形成方法

    樋口 貞雄, 宮島 清富, 古谷 正裕, 田辺 一夫

    Patent

    J-GLOBAL

  • 超親水性部材の製造方法

    樋口 貞雄, 宮島 清富, 古谷 正裕, 田辺 一夫

    Patent

    J-GLOBAL

  • 炭素ドープ酸化チタン層を有する多機能材

    古谷 正裕

    Patent

    J-GLOBAL

  • 炭素ドープ酸化チタン層を有する基体の製造方法

    古谷 正裕

    Patent

    J-GLOBAL

  • 耐食性に優れた金属構造体、前記金属構造体を製造するための材料および前記金属構造体の製法

    安永 龍哉, 藤沢 匡介, 下条 純, 古谷 正裕

    Patent

    J-GLOBAL

  • 防食用部材および該防食用部材を取り付けた金属構造体

    下条 純, 藤沢 匡介, 安永 龍哉, 古谷 正裕

    Patent

    J-GLOBAL

  • 癌治療装置および方法、並びに体液処理装置と方法

    古谷 正裕, 佐藤 裕子, 長尾 洋昌, 内山 明彦

    Patent

    J-GLOBAL

  • 放射線検出器および放射線検出方法

    古谷 正裕, 賞雅 寛而, 岡本 孝司, 友澤 秀征

    Patent

    J-GLOBAL

  • 蒸気発生器の伝熱管支持構造体

    米田 公俊, 古谷 正裕

    Patent

    J-GLOBAL

  • 光触媒装置および光触媒の照射方法

    古谷 正裕

    Patent

    J-GLOBAL

  • 放射線触媒装置並びに水素製造方法及びこれを利用した水素製造装置

    古谷 正裕, 藤嶋 昭

    Patent

    J-GLOBAL

  • 微粒子の製造方法及び製造装置、並びに微粒子

    特許第3461345号

    古谷 正裕

    Patent

    J-GLOBAL

  • アモルファス金属の製造方法、アモルファス金属微粒子の製造方法及び製造装置、並びにアモルファス金属微粒子

    特許第3461344号

    古谷 正裕

    Patent

    J-GLOBAL

  • 構造物の清浄化方法並びに防食方法、およびこれらを利用する構造物

    賞雅 寛而, 岡本 孝司, 古谷 正裕

    Patent

    J-GLOBAL

  • 微細化沸騰を利用した冷却方法

    古谷 正裕

    Patent

    J-GLOBAL

  • 微粒子の製造方法及び製造装置

    古谷 正裕

    Patent

    J-GLOBAL

  • アモルファス金属の製造方法及び製造装置

    古谷 正裕

    Patent

    J-GLOBAL

  • 直接接触伝熱型蒸気発生器

    西 義久, 木下 泉, 古谷 正裕

    Patent

    J-GLOBAL

▼display all

Research Projects

  • 伝熱面微細構の三次元多孔質造形と表面改質技術による冷却限界の飛躍的向上

    日本学術振興会  科学研究費助成事業 基盤研究(C)

    Project Year :

    2021.04
    -
    2024.03
     

    古谷 正裕

  • Construction of multiple micro interfaces using 3D microfluidic devices and application to functional chemical reactions

    Japan Society for the Promotion of Science  Grants-in-Aid for Scientific Research Grant-in-Aid for Scientific Research (A)

    Project Year :

    2020.04
    -
    2023.03
     

  • Fundamental Study on Development of Corrosion Control Technique for Anti-corrosion Materials in Marine and Offshore Structures using Radiation Induced Surface Activation

    Japan Society for the Promotion of Science  Grants-in-Aid for Scientific Research Grant-in-Aid for Scientific Research (B)

    Project Year :

    2007
    -
    2008
     

    TAKAMASA Tomoji, HAZUKU Tatsuya, ISHIMARU Takashi, MOYODA Shin-ichi, FURUYA Masahiro

  • Fundamental Study on Corrosion Control using Radiation Induced Surface Activation for Marine Use

    Japan Society for the Promotion of Science  Grants-in-Aid for Scientific Research Grant-in-Aid for Scientific Research (B)

    Project Year :

    2005
    -
    2006
     

    TAKAMASA Tomoji, ISHIMARU Takashi, MOTODA Shinichi, HAZUKU Tatsuya, FURUYA Masahiro

     View Summary

    A corrosion mitigation technique based on radiation induced surface activation (RISA) from the gamma ray irradiation on a metal surface is reported in this paper. This study aimed to develop a RISA method to prevent crevice corrosion in SUS304 stainless steel using low-intensity radioactive material. Experiment showed that an electrode potential of -100 mV vs. Ag/AgCl was produced and maintained on TiO_2-coated SUS304 stainless steel specimens immersed in artificial seawater and in close contact with a small, sealed ^<60>Co source or activated by spontaneous neutron irradiation, with no corrosion observed for more than 7 days. On the contrary, the potential of the specimen without a radiation source decreased less than -280 mV vs. Ag/AgCl and crevice corrosion occurred beneath the O-ring within few days. The RISA effect of low-intensity radioactive material has the potential to prevent crevice corrosion of SUS304 stainless steel in actual seawater.

  • Corrosion Control for Internal Structure of Nuclear Reactor Based on Radiation Induced Surface Activation Phenomenon

    Japan Society for the Promotion of Science  Grants-in-Aid for Scientific Research Grant-in-Aid for Scientific Research (B)

    Project Year :

    2003
    -
    2004
     

    TOMOJI Takamasa, MOTODA Shinichi, HAZUKU Tatsuya, UEMATSU Susumu, OKAMOTO Koji, FURUYA Masahiro

     View Summary

    When a semiconductor film is irradiated by γ-ray, excited electrons are transferred to a base metal in contact with the film, resulting in a drop of corrosion potential. This study proposes a corrosion mitigation method based on radiation induced surface activation(RISA) phenomena by supplying γ-rays from outside the material, or based on a self-excited methodology activating the film and/or the base metal. The results of the study revealed that electrons in the oxide film were activated by γ-ray irradiation and transferred from the oxide to the adjacent base metal. This made the corrosion potential of the metal less noble through a process in which the radiation induced a surface activation phenomenon, even in Zirconium oxide (ZrO_2) films which had large band gap energies. The mitigating method for the corrosion of metals was developed by utilizing this effect. The method involved external γ-rays irradiation introduced by activation of oxide films and/or base metals corrosion potential of ZrO_2 coated SUS304L was shifted down to the range between 90 mV and 300 mV vs. SSE by γ-ray irradiation. The corrosion potential was further shifted down to 600 mV when a CoCr intermediate layer was inserted between the ZrO_2 spray coating film and the SUS304L base metal. Iron specimens with a spray coating film of TiO_2,ZrO_2, and Al_2O_3 were immersed in a 3 wt% sodium chloride aqueous solution. Pitting and general corrosion were suppressed on all three specimens irradiated with γ-rays.
    These results clearly show that the corrosion potential of stainless steel could be less noble up to a level of -600 mV vs. SSE with this methodology and these processes would have a remarkable effect in mitigating the corrosion of stainless steels used in internal structures of nuclear reactors.

  • 放射線誘起表面活性特定研究に関する企画調査

    日本学術振興会  科学研究費助成事業 基盤研究(C)

    Project Year :

    2003
     
     
     

    賞雅 寛而, 古谷 正裕, 友澤 秀征, 岡本 孝司, 伊達 広行, 師岡 慎一

     View Summary

    本企画調査では、放射線、金属材料、熱流動、電気化学などが複合した全く新しい現象である放射線誘起表面活性の研究の方向性を確認するその一歩として、各専門研究者を集約した共同研究体制により、1)原子炉内放射線誘起熱伝達技術調査・分析、2)放射線誘起熱伝達メカニズム調査分析、3)放射線誘材料の調査・分析、4)宇宙環境における放射線誘起表面活性の調査・分析、5)放射線誘起腐食防止効果の技術調査・分析、及び6)放射線誘起電気特性調査・分析が行われ、文献1-12に示されるように、非常に大きな成果を得た。
    その成果をもとに、本企画調査の主要な目的であった科学研究費補助金平成16年度発足『特定領域研究』申請「放射線誘起表面活性による機能創出技術の構築」がなされている。この特定領域研究の目的は、放射線・電磁波・材料・電気化学・腐食・伝熱流動・計測などの研究分野の研究者を有機的に結合する研究領域を設定することにより、放射線誘起活性現象のメカニズム解明及び最適化を行い、放射線誘起表面活性による新たな利用分野の創成特に非放射線環境における放射線規制値未満の微弱放射能による自励放射線誘起表面活性によって生成される機能、創出技術を構築することである。この特定領域研究の実施により、(1)発展段階の観点から見て成長期の初期段階にあり、学術研究における先導的又は基盤的意義を有する研究領域である放射線誘起表面活性の利用研究の急速な展開、及び(2)放射線誘起表面活性現象は多くの研究分野が関係しているために、この領域の研究の発展は互いの研究分野に新しい情報を提供し、各学術研究分野の水準向上・強化につながり、かつそれらの発展に大きく寄与すると、それぞれ考えられる。また放射線誘起表面活性技術をエネルギー環境技術として確立することにより、我が国が技術立国として今後も存在し、世界に貢献できる重要な手段が確立されると期待される。

  • 放射線誘起表面活性現象を利用した原子炉内構造物防食特性改善

    日本学術振興会  科学研究費助成事業 萌芽研究

    Project Year :

    2002
     
     
     

    賞雅 寛而, 古谷 正裕, 岡本 孝司, 波津久 達也

     View Summary

    研究代表者らは、酸化ジルコニウムなどの広いバンドギャップを有する被膜にγ線を照射することにより、励起電子が被膜に接する金属に移行して電位を卑化させる放射線誘起表面活性現象を見いだした。本研究の目的は、この現象を利用して、金属の腐食を緩和する方法を実験的に確認することである。
    実験に用いられた試験片形状は全て幅20mm、長さ50mm、厚さ1mmであり、母材は純度99.99%の鉄、SUS304L、およびSUS316Lである。これに酸化被膜として酸化チタン、酸化ジルコニウム、または酸化アルミニウムを表面に溶射により成膜した。表面のみを試験対象とするため、裏面および端部にはエポキシ系樹脂を塗布した。3wt%塩化ナトリウム水溶液中に17時間浸漬した酸化アルミニウムを約200μm溶射した99.99%鉄試験片は、暗室保管の場合、一部孔食が見られ、ほぼ全面に腐食が進行した。一方、γ線を照射した場合には、このような腐食様相はほとんど見受けられなかった。溶液浸漬時間を40h,64hとした実験を行った結果、暗室の場合には腐食が更に進行したが、γ線照射の場合は腐食の進行が遅いことが判明した。これらの結果から、放射線照射により顕著な防食効果を発揮することが示された。線量率を増大させることにより、一層高い腐食緩和効果が生じ、また酸化チタン以外の酸化ジルコニウムおよび酸化アルミニウムについても、γ線照射により高い防食特性が得られることがわかった。
    このように炉内構造物表面の酸化金属被膜(SUS、ジルカロイなど)がγ線と反応して、放射線誘起表面活性を生じ、腐食の進行を低減できることが確認された。このような技術のさらなる開発は特に、原子炉高経年化使用、近年社会的問題になった原子炉構造物のひび割れ防止、また原子炉安全上大きな課題となっている応力腐食割れの抑制を目的に開発を急ぐ必要がある。

  • A RESEARCH ON RAPID EVAPORATION OF THE MICRO BUBBLE AND ITS INJECTION-COLLAPSE BEHAVIOR

    Japan Society for the Promotion of Science  Grants-in-Aid for Scientific Research Grant-in-Aid for Scientific Research (C)

    Project Year :

    1999
    -
    2000
     

    INADA Shigeaki, FURUYA Masahiro

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    The purpose of this study is to clarify the miniaturization phenomenon in which large number of minute droplets intensely scatters to the atmosphere, when the droplet collided on the heating surface. Especially, the generation behavior of the minute bubble(micro bubble), which arose, when the liquid-solid contact occurred by a collapse of the thin vapor film formed between the heating surface and the droplet, was caught for the elucidation of the minituarization mechanism. In quartz heating surface used until now, the nucleate boiling behavior continued to the high-temperature range, and the dispersion phenomena of the intense and minute droplets could not be remarkably recognized. Then, sapphire disk of which thermal conductivity is good, ITO(Indium Tin Oxide)film plane which thinly spattered conductive film to the glass plane and the plane which thinly spattered chromium coating on the quartz plane were used as a heating surface. Boiling behavior on each plane was caught by high-speed camera from each back surface, and it was compared with the quartz plane. Evaporation lifetime property of the nucleate boiling condition(boiling of the condition in which heating surface and liquid film sufficiently got wet)on the ITO film plane and the chromium coating plane is similar to the data in the quartz plane. The lifetime property on the sapphire closs to the lifetime for metal surface. The part of divided droplet would contact with heating surface through the thin vapor film, when the temperature of the chromium coating plane exceeds 450°C.At this time, it could be confirmed that the dispersion of the explosive minute liquid particle was generated. It was proven that the result of inducing this explosive dispersion was a micro bubble.

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