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Kageyama, Tomio; Denuma, Akio; Koizumi, Jin*; Odakura, Manabu*; Haginoya, Masahiro*; Isaka, Shinichi*; Kadowaki, Hiroyuki*; Kobayashi, Shingo*; Morimoto, Taisei*; Kato, Yoshiaki*; et al.
JAEA-Technology 2022-033, 130 Pages, 2023/03
JAEA-Technology-2022-033.pdf:9.87MB
Uranium handling facility for development of nuclear fuel manufacturing equipment (Mockup room) was constructed in 1972. The Mockup room has a weak seismic resistance and is deteriorating with age. Also, the original purpose with this facility have been achieved and there are no new development plans using this facility. Therefore, interior equipment installed in this facility had been dismantled and removed since March 2019. After that, the Mockup room was inspected for contamination, and then controlled area in the Mockup room was cancelled on March 29th 2022. A total of 6,549 workers (not including security witnesses) were required for this work. The amount of non-radioactive waste generated by this work was 31,300 kg. The amount of radioactive waste generated by this work was 3,734 kg of combustible waste (103 drums), 4,393 kg of flame resistance waste (61 drums), 37,790 kg of non-combustible waste (124 drums, 19 containers). This report describes about the dismantling and removing the interior equipment in the Mockup room, the amount of waste generated by this work, and procedure for cancellation the controlled area in the facility.
Shinozaki, Masaru; Aita, Takahiro; Iso, Takahito*; Odakura, Manabu*; Haginoya, Masahiro*; Kadowaki, Hiroyuki*; Kobayashi, Shingo*; Inagawa, Takumu*; Morimoto, Taisei*; Iso, Hidetoshi; et al.
JAEA-Technology 2021-043, 100 Pages, 2022/03
JAEA-Technology-2021-043.pdf:7.49MB
It is planned that the MOX (Mixed Oxide) from the decommissioned facilities in Nuclear Fuel Cycle Engineering Laboratories is going to be consolidated and stored stably and safely for a long term in Plutonium Fuel Production Facility of the Plutonium Fuel Development Center of Nuclear Fuel Cycle Engineering Laboratories. For this purpose, it is necessary to pelletize nuclear fuel materials in the facility and store them in the assembly storage (hereinafter referred to as "waste packaging work") to secure storage space in the plutonium material storage. As a countermeasure to reduce the facility risk in this waste packing work, it was decided to construct a new powder weighing and homogenization mixing facility to physically limit the amount (batch size) of nuclear fuel materials handled at the entrance of the process. In order to secure the installation space for the new facility in the powder preparation room (1) (FP-101), the pre-dismantling temporary waste storage facility 3 (FPG-03a, b, c) was dismantled and removed. This facility consists of a granulating and sizing facility, an additive mixing facility, and a receiving and delivering guided facility, which started to be used from January 1993, and was discontinued on February 3, 2012 and became a waste facility. Subsequently, the dismantling and removal of the interior equipment was carried out by pellet fabrication section for glove operation to reduce the amount of hold-up, and before the main dismantling and removal, there was almost no interior equipment except for large machinery. This report describes the dismantling and removal of the glove box and some interior equipment and peripherals of the facility, as well as the Green House setup method, dismantling and removal procedures, and issues specific to powder process equipment (dust, etc.).
Yasuda, Satoshi; Tamura, Kazuhisa; Terasawa, Tomoo; Yano, Masahiro; Nakajima, Hideaki*; Morimoto, Takahiro*; Okazaki, Toshiya*; Agari, Ryushi*; Takahashi, Yasufumi*; Kato, Masaru*; et al.
Journal of Physical Chemistry C, 124(9), p.5300 - 5307, 2020/03
Times Cited Count：14 Percentile：60.14(Chemistry, Physical)Confinement of hydrogen molecules at graphene-substrate interface has presented significant importance from the viewpoints of development of fundamental understanding of two-dimensional material interface and energy storage system. In this study, we investigate H
confinement at a graphene-Au interface by combining selective proton permeability of graphene and the electrochemical hydrogen evolution reaction (electrochemical HER) method. After HER on a graphene/Au electrode in protonic acidic solution, scanning tunneling microscopy finds that H nanobubble structures can be produced between graphene and the Au surface. Strain analysis by Raman spectroscopy also shows that atomic size roughness on the graphene/Au surface originating from the HER-induced strain relaxation of graphene plays significant role in formation of the nucleation site and H storage capacity.
at high temperatures up to 2190 KMorimoto, Kyoichi; Kato, Masato; Ogasawara, Masahiro*
Journal of Nuclear Materials, 443(1-3), p.286 - 290, 2013/11
Times Cited Count：5 Percentile：38.62(Materials Science, Multidisciplinary)In this study, measurement was conducted for the sliced MOX pellets containing 30% of Pu prepared by a conventional powder metallurgy technology. Oxygen-to-metal (O/M) ratios of the samples were adjusted in the range from 1.92 to 2.00. The thermal diffusivities of these samples were measured at temperature up to 2150 K with the laser flash method. Thermal diffusivities of the near-stoichiometric samples obtained in the cooling process were greatly lower than those in the heating process unlike measurement below 1770 K. On the other hand, they were almost identical for the sample of 1.946 in O/M. It was also shown that thermal diffusivity decreased with the temperature but increased with the O/M.
Ono, Shinya*; Inoue, Kei*; Morimoto, Masahiro*; Arae, Sadanori*; Toyoshima, Hiroaki*; Yoshigoe, Akitaka; Teraoka, Yuden; Ogata, Shoichi*; Yasuda, Tetsuji*; Tanaka, Masatoshi*
Surface Science, 606(21-22), p.1685 - 1692, 2012/11
Times Cited Count：8 Percentile：35.38(Chemistry, Physical)Ono, Shinya*; Inoue, Kei*; Morimoto, Masahiro*; Arae, Sadanori*; Toyoshima, Hiroaki*; Yoshigoe, Akitaka; Teraoka, Yuden; Ogata, Shoichi*; Yasuda, Tetsuji*; Tanaka, Masatoshi*
Shingaku Giho, 111(114), p.23 - 27, 2011/07
The initial oxidation on high-index silicon surfaces with (113) and (120) orientations at 820 K has been investigated by real-time X-ray photoemission spectroscopy (Si 2p and O 1s) using 687 eV photons. The time evolutions of the Si
(n=1-4) components in the Si 2p spectrum indicate that the Si
state is suppressed on high-index surfaces compared with Si(001). The O 1s state consists of two components, a low-binding-energy component (LBC) and a high-binding-energy component (HBC). It is suggested that the O atom in strained Si-O-Si contributes to the LBC component. The reaction rates are slower on high-index surfaces compared with that on Si(001).
Morimoto, Kyoichi; Kato, Masato; Ogasawara, Masahiro*
Proceedings of Plutonium Futures; The Science 2010 (CD-ROM), p.339 - 340, 2010/09
Nuclear fuel pellets are stored before loading into a reactor. In some cases, the fuel pellets are left for several years. When uranium-plutonium mixed oxide (MOX) fuel pellets are stored for a long time, lattice defects induced by self-irradiation (
decay) accumulate and these defects affect physical properties of the pellets, i.e. lattice parameter, electrical resistivity and thermal conductivity. The thermal conductivity of fuel pellets is one of the most important properties for fuel design and performance analyses; it is known to decrease due to the defects induced by self-irradiation, but it can be recovered by heating the pellets. In this study, the recovery behavior of thermal conductivity of a MOX fuel pellet stored for long time was investigated as a function of time and temperature, in order to make it easy to analyze the thermal performance of fuel pellets.
solid solutionsMorimoto, Kyoichi; Kato, Masato; Ogasawara, Masahiro*; Kashimura, Motoaki; Abe, Tomoyuki
Journal of Alloys and Compounds, 452(1), p.54 - 60, 2008/03
Times Cited Count：30 Percentile：77.85(Chemistry, Physical)Plutonium and uranium mixed oxide (MOX) fuel with high Pu content have been developed as a fuel of fast reactor (FR). As the storage time of Pu raw material between reprocessing and fabrication increases, americium content of the fabricated MOX fuel increases up to a few percent. In this work, the thermal conductivity of MOX fuel containing Am was investigated as a part of clarifying the effect of Am content on thermal physical properties. The pellets of (Am
Pu
U)O
, (Am
Pu U)O and (Am
Pu U)O were prepared. The oxygen to metal ratio (O/M ratio) of sintered pellet was adjusted to 2.00. The thermal diffusivity measurement was carried out in the range of temperature from 900 K to 1700 K by the laser flash method, and thermal conductivity of these pellets was evaluated. The heat capacity for evaluating thermal conductivity was derived from heat capacity of UO, PuO and AmO by using the Kopp-Neumann rule.
oxideMorimoto, Kyoichi; Kato, Masato; Ogasawara, Masahiro*; Kashimura, Motoaki
Journal of Nuclear Materials, 374(3), p.378 - 385, 2008/03
Times Cited Count：35 Percentile：89.2(Materials Science, Multidisciplinary)The thermal conductivities of (U
PuAm)O solid solutions (x = 0.0 - 0.1) were studied at temperatures from 900 to 1773 K. Thermal conductivities were obtained from the thermal diffusivity measured by laser flash method. The thermal conductivities obtained experimentally up to about 1400K could be expressed by a classical phonon transport model, 
= (A+BT)
, A(x) = 2.89
x + 2.2410
(m K/W) and B(x) = (- 6.70x + 2.48) 10
(m/W). The experimental values of A showed a good agreement with theoretical predictions. The experimental values of B could be fairly expressed by the theoretical prediction in the region x 
0.04, but not deviated from the ones in the region x 
0.04. Although this reason could not be understood well, it is most likely that the uncertainty in the measurement of melting temperature cause this difference.
PuAm)O in high temperature regionKomeno, Akira; Morimoto, Kyoichi; Kato, Masato; Kashimura, Motoaki; Ogasawara, Masahiro*; Sunaoshi, Takeo*
Transactions of the American Nuclear Society, 97(1), p.616 - 617, 2007/11
no abstracts in English
Kato, Masato; Morimoto, Kyoichi; Komeno, Akira; Nakamichi, Shinya; Kashimura, Motoaki; Abe, Tomoyuki; Uno, Hiroki*; Ogasawara, Masahiro*; Tamura, Tetsuya*; Sugata, Hiromasa*; et al.
JAEA-Technology 2006-049, 32 Pages, 2006/10
JAEA-Technology-2006-049.pdf:19.46MBJAEA-Technology-2006-049(errata).pdf:0.32MB
Japan Atomic Energy Agency has developed a fast breeder reactor(FBR), and plutonium and uranium mixed oxide (MOX) having low density and 20-30%Pu content has used as a fuel of the FBR, Monju. In plutonium, Americium has been accumulated during long-term storage, and Am content will be increasing up to 2-3% in the MOX. It is essential to evaluate the influence of Am content on physical properties of MOX on the development of FBR in the future. In this study melting points and thermal conductivities which are important data on the fuel design were measured systematically in wide range of composition, and the effects of Am accumulated were evaluated. The solidus temperatures of MOX were measured as a function of Pu content, oxygen to metal ratio(O/M) and Am content using thermal arrest technique. The sample was sealed in a tungsten capsule in vacuum for measuring solidus temperature. In the measurements of MOX with Pu content of more than 30%, a rhenium inner capsule was used to prevent the reaction between MOX and tungsten. In the results, it was confirmed that the melting points of MOX decrease with as an increase of Pu content and increase slightly with a decrease of O/M ratio. The effect of Am content on the fuel design was negligible small in the range of Am content up to 3%. Thermal conductivities of MOX were evaluated from thermal diffusivity measured by laser flash method and heat capacity calculated by Nuemann- Kopp's law. The thermal conductivity of MOX decreased slightly in the temperature of less than 1173K with increasing Am content. The effect of Am accumulated in long-term storage fuel was evaluated from melting points and thermal conductivities measured in this study. It is concluded that the increase of Am in the fuel barely affect the fuel design in the range of less than 3%Am content.
Kato, Masato; Morimoto, Kyoichi; Kihara, Yoshiyuki; Ogasawara, Masahiro*; Tamura, Tetsuya*; Uno, Hiroki*; Sunaoshi, Takeo*
JNC TN8400 2004-022, 44 Pages, 2005/03
JNC-TN8400-2004-022.pdf:5.43MB
Japan Nuclear Development Institute has developed homogeneous mixed oxide fuel containing minor actinide as a fuel of an advanced fast reactor. Study on the sintering behavior of the fuel was carried out and the heat treatment technique for preparing homogeneous low O/M fuel had been developed. In this report, oxygen potential was measured and phase relation was evaluated, which are needed essentially for developing the new type fuel.Oxygen potential of (Npsub0.02Amsub0.02Pusub0.3Usub0.66)Osub2-X was measured by gas equilibrium method as a function of temperature and O/M ratio. The MOX with MA has slightly higher oxygen potential as compared with that of MOX without MA. And the model of oxygen potential was derived from the measurement results based on lattice defect theory.In samples with low O/M ratio, two fcc phases were observed at room temperature. The temperature of the phase separation was measured and it is observed that the addition of MA have the effect to be decreased the phase separation temperature. In the MOX containing MA and Nd simulated a low decontaminated fuel, the Pu-Am-Nd oxides were precipitated by decreasing O/M ratio in less than 1.96.
; Masaki, Kei; ; Morimoto, Masaaki*; *; Sakurai, Shinji; *; Saido, Masahiro; Inoue, Masahiko*; *; et al.
JAERI-Tech 98-049, 151 Pages, 1998/11
no abstracts in English
Masaki, Kei; ; ; Morimoto, Masaaki*; *; Hosogane, Nobuyuki; Sakurai, Shinji; Saido, Masahiro
Purazuma, Kaku Yugo Gakkai-Shi, 74(9), p.1048 - 1053, 1998/09
no abstracts in English
; Omori, Eiichi; Kato, Yoshiyuki; Suzuki, Hiroshi; Shimoyamada, Tetsuya; Tomiyama, Masahiro; Shimokura, Mitsuharu; Sakuraba, Terumi; Morimoto, Kyoichi; Hagiwara, Masayoshi; et al.
PNC TN8410 98-013, 1028 Pages, 1998/01
PNC-TN8410-98-013.pdf:143.04MB
None
Sakurai, Shinji; Hosogane, Nobuyuki; Masaki, Kei; ; ; *; *; Shimizu, Katsuhiro; Akino, Noboru; Miyo, Yasuhiko; et al.
Fusion Engineering and Design, 39-40, p.371 - 376, 1998/00
Times Cited Count：5 Percentile：44.27(Nuclear Science & Technology)no abstracts in English
Masaki, Kei; ; Morimoto, Masaaki*; ; *; Hosogane, Nobuyuki; Saido, Masahiro
Fusion Technology 1998, p.67 - 70, 1998/00
no abstracts in English
; ; Masaki, Kei; Hosogane, Nobuyuki; Sakurai, Shinji; Morimoto, Masaaki*; Miyo, Yasuhiko; Hiratsuka, Hajime; Akino, Noboru; *; et al.
Proceedings of 17th IEEE/NPSS Symposium Fusion Engineering (SOFE'97), 2, p.365 - 368, 1998/00
no abstracts in English
PuAm)O
(x=1.90
2.00)Morimoto, Kyoichi; Kato, Masato; Komeno, Akira; Kashimura, Motoaki; Abe, Tomoyuki; Ogasawara, Masahiro*; Sunaoshi, Takeo*; Uno, Hiroki*; Tamura, Tetsuya*
no journal, ,
To examine the influences of O/M ratio on the thermal conductivity of MOX fuel, the thermal diffusivities of the MOX fuel with 30% Pu content were measured, and thermal conductivities were evaluated.
Komeno, Akira; Kato, Masato; Morimoto, Kyoichi; Kashimura, Motoaki; Abe, Tomoyuki; Ogasawara, Masahiro*; Sunaoshi, Takeo*; Uno, Hiroki*; Tamura, Tetsuya*
no journal, ,
no abstracts in English
