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Dive into the research topics where Yoshiyuki Inagaki is active.

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Featured researches published by Yoshiyuki Inagaki.


Journal of Nuclear Science and Technology | 2007

Numerical Study on Tritium Behavior by Using Isotope Exchange Reactions in Thermochemical Water-Splitting Iodine—Sulfur Process

Hirofumi Ohashi; Nariaki Sakaba; Tetsuo Nishihara; Yoshiyuki Inagaki; Kazuhiko Kunitomi

One potential problem in the hydrogen production system coupled with the high-temperature gascooled reactor (HTGR) is transmission of tritium from the primary coolant to the product hydrogen by permeation through the heat transfer tubes. Tritium accumulation in the process chemicals in the components of a hydrogen plant, a thermochemical water-splitting iodine-sulfur (IS) process, will also be a critical issue in seeking to license the hydrogen plant as a non-nuclear plant in the future. A numerical analysis model for tritium behavior in the IS process was developed by considering the isotope exchange reactions between tritium and the hydrogen-containing process chemicals, i.e., H2O, H2SO4 and HI. The tritium activity concentration in the IS process coupled with the high-temperature engineering test reactor (HTTR), the HTTR-IS system, was preliminarily evaluated in regard to the effects of some indeterminate parameters, i.e., equilibrium constants of the isotope exchange reactions, permeability of tritium through heat transfer tubes, tritium and hydrogen concentrations in the secondary helium coolant, and the leak rate from the secondary coolant loop. The results describing how the tritium activity concentration changes with variations in these parameters and which component has the maximum tritium activity concentration in the IS process are described in this paper.


18th International Conference on Nuclear Engineering: Volume 6 | 2010

Conceptual Structure Design of High Temperature Isolation Valve for High Temperature Gas Cooled Reactor

Shoji Takada; Kenji Abe; Yoshiyuki Inagaki

The high temperature isolation valve (HTIV) is a key component to assure the safety of a high temperature gas cooled reactor (HTGR) connected with a hydrogen production system, that is, protection of radioactive material release from the reactor to the hydrogen production system and combustible gas ingress to the reactor at the accident of fracture of an intermediate heat exchanger and the chemical reactor. The HTIV used in the helium condition over 900 °C, however, has not been made for practical use yet. The conceptual structure design of an angle type HTIV was carried out. A seat made of Hasteloy-XR is welded inside a valve box. Internal thermal insulation is employed around the seat and a liner because high temperature helium gas over 900 °C flows inside the valve. Inner diameter of the top of seat was set 445 mm based on fabrication experiences of valve makers. A draft overall structure was proposed based on the diameter of seat. The numerical analysis was carried out to estimate temperature distribution and stress of metallic components by using a three-dimensional finite element method code. Numerical results showed that the temperature of the seat was simply decreased from the top around 900 °C to the root, and the thermal stress locally increased at the root of the seat which was connected with the valve box. The stress was lowered below the allowable limit 120 MPa by decreasing thickness of the connecting part and increasing the temperature of valve box to around 350 °C. The stress also increased at the top of the seat. Creep analysis was also carried out to estimate a creep-fatigue damage based on the temperature history of the normal operation and the depressurization accident.Copyright


Volume 1: Plant Operations, Maintenance, Engineering, Modifications and Life Cycle; Component Reliability and Materials Issues; Next Generation Systems | 2009

Corrosion Behavior of FBR Structural Materials in High Temperature Supercritical CO2

Tomohiro Furukawa; Yoshiyuki Inagaki; Masanori Aritomi

A key problem in the application of a supercritical carbon dioxide (CO2 ) turbine cycle to a fast breeder reactor is the corrosion of structural materials by supercritical CO2 at high temperature. In this study, corrosion tests on the candidate materials, high-chromium martensitic steel (12Cr-steel) and FBR grade type 316 stainless steel (316FR), were performed for up to approximately 2000h at 400–600°C in supercritical CO2 pressurized at 20MPa. Corrosion due to the high temperature oxidation was measured in both steels. Results showed that the behavior differed greatly. For 12Cr-specimens, weight gain showed parabolic growth as exposure time increased at each temperature. The oxidation coefficient could be estimated by the Arrhenius function. The specimens were covered by two successive layers of oxide, an Fe-Cr-O layer and an Fe-O layer. A partial thin oxide diffusion layer appeared between the base metal and the Fe-Cr-O layer. The corrosion behavior was equivalent to that in supercritical CO2 at 10MPa, and no effects of CO2 pressure were observed in this study. For 316FR specimens, the weight gain was significantly lower than that of 12Cr-specimen, and good resistance against corrosion was observed. No dependency of temperature or immersed time on weight gain was observed under the test conditions, and the value of all tested specimens was within 2g/m2 . Some nodule shape oxide was observed on the surface of the 316FR specimen. Carburizing, known as a factor in the occurrence of breakaway corrosion and/or the degradation of ductility, was observed on the surface of both steels.Copyright


International Journal of Hydrogen Energy | 2007

Application of nuclear energy for environmentally friendly hydrogen generation

Michio Yamawaki; Tetsuo Nishihara; Yoshiyuki Inagaki; Kazuo Minato; Hiroyuki Oigawa; Kaoru Onuki; Ryutaro Hino; Masuro Ogawa


Progress in Nuclear Energy | 2011

Compatibility of FBR structural materials with supercritical carbon dioxide

Tomohiro Furukawa; Yoshiyuki Inagaki; Masanori Aritomi


Journal of Power and Energy Systems | 2010

Corrosion Behavior of FBR Structural Materials in High Temperature Supercritical Carbon Dioxide

Tomohiro Furukawa; Yoshiyuki Inagaki; Masanori Aritomi


Nuclear Engineering and Design | 2006

Development of control technology for HTTR hydrogen production system with mock-up test facility: System controllability test for loss of chemical reaction

Hirofumi Ohashi; Yoshitomo Inaba; Tetsuo Nishihara; Tetsuaki Takeda; Koji Hayashi; Shoji Takada; Yoshiyuki Inagaki


International Journal of Hydrogen Energy | 2011

Lab-scale water-splitting hydrogen production test of modified hybrid sulfur process working at around 550 °C

Toshihide Takai; Shinji Kubo; Toshio Nakagiri; Yoshiyuki Inagaki


Nuclear Engineering and Design | 2014

Components development for sulfuric acid processing in the IS process

Hiroki Noguchi; Shinji Kubo; Jin Iwatsuki; Seiji Kasakara; Nobuyuki Tanaka; Yoshiyuki Imai; Atsuhiko Terada; Hiroaki Takegami; Yu Kamiji; Kaoru Onuki; Yoshiyuki Inagaki


Nuclear Engineering and Design | 2014

Process flow sheet evaluation of a nuclear hydrogen steelmaking plant applying very high temperature reactors for efficient steel production with less CO2 emissions

Seiji Kasahara; Yoshiyuki Inagaki; Masuro Ogawa

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Masuro Ogawa

Japan Atomic Energy Agency

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Seiji Kasahara

Japan Atomic Energy Agency

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Shinji Kubo

Japan Atomic Energy Agency

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Atsuhiko Terada

Japan Atomic Energy Agency

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Jin Iwatsuki

Japan Atomic Energy Agency

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Masanori Aritomi

Tokyo Institute of Technology

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Tetsuo Nishihara

Japan Atomic Energy Agency

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Tomohiro Furukawa

Japan Atomic Energy Agency

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Hirofumi Ohashi

Japan Atomic Energy Agency

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Ikuo Ioka

Japan Atomic Energy Agency

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