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Journal Articles

The Surface composition of asteroid 162173 Ryugu from Hayabusa2 near-infrared spectroscopy

Kitazato, Kohei*; Milliken, R. E.*; Iwata, Takahiro*; Abe, Masanao*; Otake, Makiko*; Matsuura, Shuji*; Arai, Takehiko*; Nakauchi, Yusuke*; Nakamura, Tomoki*; Matsuoka, Moe*; et al.

Science, 364(6437), p.272 - 275, 2019/04

 Times Cited Count:86 Percentile:0.14(Multidisciplinary Sciences)

The near-Earth asteroid 162173 Ryugu, the target of Hayabusa2 sample return mission, is believed to be a primitive carbonaceous object. The Near Infrared Spectrometer (NIRS3) on Hayabusa2 acquired reflectance spectra of Ryugu's surface to provide direct measurements of the surface composition and geological context for the returned samples. A weak, narrow absorption feature centered at 2.72 micron was detected across the entire observed surface, indicating that hydroxyl (OH)-bearing minerals are ubiquitous there. The intensity of the OH feature and low albedo are similar to thermally- and/or shock-metamorphosed carbonaceous chondrite meteorites. There are few variations in the OH-band position, consistent with Ryugu being a compositionally homogeneous rubble-pile object generated from impact fragments of an undifferentiated aqueously altered parent body.

Journal Articles

Development of new type passive autocatalytic recombiner, 1; Experimental study on degradation of catalyst

Kamiji, Yu; Matsumura, Daiju; Taniguchi, Masashi*; Nishihata, Yasuo; Tanaka, Hirohisa*; Hirata, Shingo*; Hara, Mikiya; Hino, Ryutaro

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 4 Pages, 2015/05

In a severe accident at a nuclear power plant, a large amount of hydrogen can be released to primary containment vessel or reactor building. Passive autocatalytic recombiner (PAR) is one of the most effective systems for hydrogen mitigation and safety accident management. The new type PAR is under developing to improve conventional PARs, especially its size and weight. In this study, the influence of steam coexistence for the automotive catalyst activity was experimentally examined. These results show that the steam slightly affects the reaction start up and catalyst activity.

Journal Articles

Development of new type passive autocatalytic recombiner, 1; Characterization of monolithic catalyst

Kamiji, Yu; Taniguchi, Masashi*; Nishihata, Yasuo; Nagaishi, Ryuji; Tanaka, Hirohisa*; Hirata, Shingo*; Hara, Mikiya; Hino, Ryutaro

Proceedings of 2nd International Conference on Maintenance Science and Technology (ICMST-Kobe 2014), p.87 - 88, 2014/11

For hydrogen mitigation, a new type passive autocatalytic recombiner is under development. In this study, the activation energy of hydrogen-oxygen recombination reaction was examined to clarify the basic characteristics of the catalyst. In addition, the degradation of the catalyst by $$gamma$$-ray irradiation simulating the environmental condition in nuclear power plants was also examined. As a result, the activation energy was experimentally estimated at 5.75 kJ/mol. Besides, no significant differences were observed in the compositional distribution from the EPMA results between the non-irradiated and the irradiated catalyst. However, the irradiated catalyst showed much more activity because of larger specific surface area of the catalyst and surface area of the precious metals. It showed that $$gamma$$-ray irradiation up to 1.0 MGy can increase activity of the catalyst.

Journal Articles

Development of design evaluation tools for the JSFR fuel transfer pot

Chikazawa, Yoshitaka; Hirata, Shingo; Obata, Hiroyuki*

Nuclear Engineering and Design, 273, p.1 - 9, 2014/07

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

JSFR is going to adopt an advanced fuel handling system. As for ex-vessel spent fuel handling, a transfer pot with two fuel subassembly positions has been developed so as to shorten refueling period increasing plant availability. The pot is required to provide sufficient cooling capability in case of transportation malfunction. In this study, a three dimensional analysis model for heat transfer evaluation of the JSFR fuel transfer pot has been developed. The heat transfer models inside and outside the pot have been validated by reference experiments. Using the developed three-dimensional model, the JSFR fuel transfer pot has been analyzed. For a simpler design tool, a two dimensional analysis model has been developed. Comparison of the three and two dimensional analyses shows that two dimensional analyses could estimate pot cooling performance conservatively.

Journal Articles

Heat transfer experiments on fuel subassembly transfer pot for JSFR

Chikazawa, Yoshitaka; Kato, Atsushi; Hirata, Shingo*; Obata, Hiroyuki*

Journal of Nuclear Science and Technology, 51(6), p.798 - 808, 2014/06

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

JSFR is going to adopt an advanced fuel handling system. As for ex-vessel spent fuel handling, a transfer pot with two fuel subassembly positions has been developed so as to shorten refueling period increasing plant availability. The pot is required to provide sufficient cooling capability in case of transportation malfunction. To evaluate cooling capacity of the transfer pot, a mockup pot has been fabricated and heat transfer experiments have been conducted on the mockup pot.

Journal Articles

Under-sodium endurance experiment of selector valve in failed-fuel detection and location system of JSFR

Aizawa, Kosuke; Fujita, Kaoru; Hirata, Shingo*; Kasahara, Naoto*

Nuclear Technology, 183(1), p.1 - 12, 2013/07

In the design of Japan Sodium-cooled Fast Reactor (JSFR), a selector-valve mechanism is adopted for its failed-fuel detection and location (FFDL) system. Since JSFR has only two FFDL units for about 600 fuel subassemblies, one FFDL unit must handle much larger number of subassemblies than in previous designs. In addition, during long plant life of 60 years, the wear length of the selector-valve will become longer than those of past reactors. Therefore, the endurance of the selector-valve becomes important. To demonstrate the manufacturability and endurance of the selector-valve, a full size mock-up was manufactured, and an endurance experiment of the mock-up model under high-temperature sodium were conducted. The cross-section observation, hardness measurement, and chemical assay results after the endurance experiment showed that the coating layer on the sliding surface still remains. Thus, the endurance of the JSFR selector-valve was demonstrated.

Journal Articles

Thermal analysis on shipping cask for JSFR fresh fuel

Kato, Atsushi; Chikazawa, Yoshitaka; Uto, Nariaki; Hirata, Shingo; Obata, Hiroyuki*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 12 Pages, 2012/00

A basic feasibility of the helium gas cask has been evaluated by thermal analyses. There have been conducted two analyses: whole cask and detail inside subassembly analyses. The detail inside subassembly analysis has shown that the temperature distribution is mainly governed by thermal conductivity and natural convection of coolant helium hardly contributes heat removal. In the case of a cask with five subassemblies with 2.2 kW decay heat per each, the maximum cladding temperature is evaluated to be 361 $$^{circ}$$C satisfying cladding temperature limit of 395 $$^{circ}$$C. Those results have shown the basic feasibility of the helium gas fresh fuel shipping cask.

Journal Articles

Development of transfer pot for JSFR ex-vessel fuel handling

Hirata, Shingo; Chikazawa, Yoshitaka; Kato, Atsushi; Uto, Nariaki; Obata, Hiroyuki*; Kotake, Shoji*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 10 Pages, 2012/00

In Fast Reactor Cycle Technology Development (FaCT) project, Japan Sodium-cooled Fast Reactor (JSFR) is going to adopt an advanced fuel handling system. As for ex-vessel spent fuel handling, a pot which contains two fuel subassemblies simultaneously and is applicable size to compact reactor vessel, has been developing so as to shorten a refueling period leading to an improvement of plant availability. The pot is required to provide sufficient cooling capability even in case of transportation malfunction during transportation of spent fuel subassemblies with high decay heat. In the present study, experimental and analytical studies to evaluate the cooling capacity of the pot are summarized.

Journal Articles

Development of the JSFR fuel handling system and mockup experiments of fuel handling machine in abnormal conditions

Kato, Atsushi; Hirata, Shingo; Chikazawa, Yoshitaka; Uto, Nariaki; Obata, Hiroyuki*; Kotake, Shoji*; Uzawa, Masayuki*

Proceedings of 2010 International Congress on Advances in Nuclear Power Plants (ICAPP '10) (CD-ROM), p.692 - 699, 2010/06

In the JSFR design, a single rotating plug and an upper inner structure (UIS) with a vertically penetrating slit are proposed, so that the fuel handling machine (FHM) can access any subassembly by horizontal movement of the FHM arm in the slit space. As a result of a full-scale mockup test, excellent performance in normal operation has been shown. In this study, from the viewpoint of achieving reliability of the pantograph type FHM, behavior of the FHM mockup have been investigated under abnormal conditions.

Journal Articles

Endurance sodium experiment of selector-valve for failed fuel detection and location system in sodium-cooled large reactor

Aizawa, Kosuke; Fujita, Kaoru; Hirata, Shingo; Kasahara, Naoto

Proceedings of 2010 International Congress on Advances in Nuclear Power Plants (ICAPP '10) (CD-ROM), p.645 - 652, 2010/06

A conceptual design study of an advanced large-sized (1500MWe class) sodium-cooled fast reactor (named JSFR) has progressed in the FaCT project in Japan. JSFR adopts a selector-valve mechanism for the failed fuel detection and location (FFDL) system. The drive shaft rotates and moves vertically in order to select the channel. And, the drive shaft is in contact with the selector-valve drum by spring load. Thus, a mechanical wear could occur between the drive shaft and the drum of the selector-valve FFDL system. There is concern about manufacturing capability and endurance of the JSFR selector-valve. To demonstrate manufacturing capability and endurance of the JSFR selector-valve, a mock-up was manufactured and an endurance experiment under high temperature sodium has been conducted.

Journal Articles

Conceptual design for Japan sodium-cooled fast reactor, 3; Development of advanced fuel handling system for JSFR

Kato, Atsushi; Hirata, Shingo; Chikazawa, Yoshitaka; Uto, Nariaki; Obata, Hiroyuki; Kotake, Shoji

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9281_1 - 9281_6, 2009/05

One of the most important challenges to commercialize a Fast Reactor is to increase economic competitiveness. For that purpose, Japan Sodium cooled Fast Reactor (hereafter JSFR) aims to simplify the plant system and reduce the raw and processed material by adopting innovative technologies. In the JSFR design, a single rotating plug and a reactor upper inner structure (hereafter UIS) with a vertically penetrating slit are proposed, so that the fuel handling machine (hereafter FHM) can access any subassembly by horizontal movement of the FHM arm in the slit space. The feature of this FHM enables no need for the UIS removal when the rotational plug moves round above the core, which can achieve a compact reactor vessel to enhance the economic competitiveness. We fabricated the full scale FHM test equipment to perform comprehensive tests in the air for demonstrating the feasibility of the key characteristics of this FHM concept.

Journal Articles

Conceptual design of the blanket tritium recovery system for the prototype fusion reactor

Kakuta, Toshiya*; Hirata, Shingo*; Mori, Seiji*; Konishi, Satoshi; Kawamura, Yoshinori; Nishi, Masataka; Ohara, Yoshihiro

Fusion Science and Technology, 41(3), p.1069 - 1073, 2002/05

Research-and-development of the supercritical water-cooled prototype fusion reactor which has cost competitiveness has been performed in Japan Atomic Energy Research Institute (JAERI). It is necessary to establish immediately the design concept of the blanket tritium recovery system which collects tritium continuously and safely from the supercritical water-cooled blanket because fuel self-sufficiency is inevitable in the prototype reactor. The candidate systems are; 1) batch-processing cryogenic molecular sheave bed recovery system with cryogenic temperature operation, 2) continuous processing Pd membrane penetration recovery system with high vacuum operation. In the present study, however, the third candidate system, the hydrogen pump system with protonic conductors, was investigated. As a result of the study, it was made clear that the system with minimized energy consumption and minimized accidental tritium release could be realized by using the hydrogen pump for the blanket tritium recovery system of the prototype fusion reactor.

JAEA Reports

Accident identification in tritium processing systems of international thermonuclear experimental reactor in engineering design activity

Enoeda, Mikio; D.F.Holland*; Matsuda, Yuji; Ohira, Shigeru; Okuno, Kenji; *; Hirata, Shingo*

JAERI-Tech 95-050, 90 Pages, 1995/11

JAERI-Tech-95-050.pdf:2.98MB

no abstracts in English

Journal Articles

Experimental and analytical study on membrane detritiation process

Hirata, Shingo*; Kakuta, Toshiya*; Ito, H.*; Suzuki, T.*; Hayashi, Takumi; Ishida, Toshikatsu*; Matsuda, Yuji; Okuno, Kenji

Fusion Technology, 28(3), p.1521 - 1526, 1995/10

no abstracts in English

JAEA Reports

Characteristics of pebble packing into in-pile mockup on fusion blanket

Nakamichi, Masaru; Kawamura, Hiroshi; Sagawa, Hisashi; Ishida, Toshikatsu*; Hirata, Shingo*; *

JAERI-M 93-060, 30 Pages, 1993/03

JAERI-M-93-060.pdf:1.18MB

no abstracts in English

Journal Articles

Packing characteristics of solid breeder for fusion reactor blanket

Nakamichi, Masaru; Kawamura, Hiroshi; Sagawa, Hisashi; Ishida, Toshikatsu*; Hirata, Shingo*; *

FAPIG, 0(132), p.22 - 28, 1992/11

no abstracts in English

JAEA Reports

Test apparatus for ITER blanket pebble packing behavior

Enoeda, Mikio; Yoshida, Hiroshi; Hirata, Shingo*; *

JAERI-M 92-104, 29 Pages, 1992/07

JAERI-M-92-104.pdf:0.77MB

no abstracts in English

JAEA Reports

Test apparatus for ITER blanket cooling water distributor design

Yoshida, Hiroshi; Enoeda, Mikio; Hirata, Shingo*; Ito, H.*

JAERI-M 92-070, 45 Pages, 1992/05

JAERI-M-92-070.pdf:1.34MB

no abstracts in English

JAEA Reports

Japanese contributions to blanket design for ITER

Kuroda, Toshimasa*; Yoshida, Hiroshi; Takatsu, Hideyuki; Seki, Yasushi; Noda, Kenji; Watanabe, H.; Koizumi, Koichi; Nishio, Satoshi; *; *; et al.

JAERI-M 91-133, 191 Pages, 1991/08

JAERI-M-91-133.pdf:5.79MB

no abstracts in English

JAEA Reports

Japanese contributions to ITER testing program of solid breeder blankets for DEMO

Kuroda, Toshimasa*; Yoshida, Hiroshi; Takatsu, Hideyuki; *; Mori, Seiji*; *; *; Hirata, Shingo*; Miura, H.*

JAERI-M 91-063, 72 Pages, 1991/04

JAERI-M-91-063.pdf:1.55MB

no abstracts in English

30 (Records 1-20 displayed on this page)