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JAEA Reports

Decommissioning of Pre-dismantling Temporary Waste Storage Facility 3 (FPG-03a,b,c) in Plutonium Fuel Production 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.).

Journal Articles

Thermal diffusivity measurement of (U, Pu)O$$_{2-x}$$ at high temperatures up to 2190 K

Morimoto, Kyoichi; Kato, Masato; Ogasawara, Masahiro*

Journal of Nuclear Materials, 443(1-3), p.286 - 290, 2013/11

 Times Cited Count:4 Percentile:35.38(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.

Journal Articles

Time-evolution of thermal oxidation on high-index silicon surfaces; Real-time photoemission spectroscopic study with synchrotron radiation

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:37.41(Chemistry, Physical)

Journal Articles

Characterization of initial oxidation process on high-index silicon surfaces by real-time photoemission spectroscopy

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$$^{n+}$$ (n=1-4) components in the Si 2p spectrum indicate that the Si$$^{2+}$$ 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).

Journal Articles

Thermal recovery evaluation of thermal conductivity in a self-irradiated MOX pellet

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 ($$alpha$$ 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.

Journal Articles

Thermal conductivities of (U,Pu,Am)O$$_{2}$$ solid solutions

Morimoto, Kyoichi; Kato, Masato; Ogasawara, Masahiro*; Kashimura, Motoaki; Abe, Tomoyuki

Journal of Alloys and Compounds, 452(1), p.54 - 60, 2008/03

 Times Cited Count:27 Percentile:77.45(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$$_{0.007}$$ Pu$$_{0.3}$$ U)O$$_{2.00}$$, (Am$$_{0.02}$$ Pu$$_{0.3}$$ U)O$$_{2.00}$$ and (Am$$_{0.03}$$ Pu$$_{0.3}$$ U)O$$_{2.00}$$ 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$$_{2}$$, PuO$$_{2}$$ and AmO$$_{2}$$ by using the Kopp-Neumann rule.

Journal Articles

Thermal conductivities of hypostoichiometric (U, Pu, Am)O$$_{2-x}$$ oxide

Morimoto, Kyoichi; Kato, Masato; Ogasawara, Masahiro*; Kashimura, Motoaki

Journal of Nuclear Materials, 374(3), p.378 - 385, 2008/03

 Times Cited Count:32 Percentile:89.1(Materials Science, Multidisciplinary)

The thermal conductivities of (U$$_{0.68}$$Pu$$_{0.3}$$Am$$_{0.02}$$)O$$_{2-x}$$ 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, $$lambda$$ = (A+BT)$$^{-1}$$, A(x) = 2.89$$times$$x + 2.24$$times$$10$$^{-2}$$ (m K/W) and B(x) = (- 6.70$$times$$x + 2.48) $$times$$ 10$$^{-4}$$ (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.

Journal Articles

Measurement of thermal conductivity of (U$$_{0.68}$$Pu$$_{0.3}$$Am$$_{0.02}$$)O$$_{2-x}$$ in high temperature region

Komeno, 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

JAEA Reports

Evaluation of thermal physical properties for fast reactor fuels; Melting point and thermal conductivities

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:19.46MB
JAEA-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.

JAEA Reports

Development of low decontaminated MOX fuel containing MA IV; Oxygen potential and phase relation

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.

JAEA Reports

Design and installation of W-shaped divertor in JT-60U

; Masaki, Kei; ; Morimoto, Masaaki*; *; Sakurai, Shinji; *; Saido, Masahiro; *; *; et al.

JAERI-Tech 98-049, 151 Pages, 1998/11

JAERI-Tech-98-049.pdf:6.45MB

no abstracts in English

Journal Articles

Inspection of JT-60 W-shaped divertor after the initial operation

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

JAEA Reports

Facility damages by the explosion

; 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

Journal Articles

Development of a compact W-shaped pumped divertor in JT-60U

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:44.99(Nuclear Science & Technology)

no abstracts in English

Journal Articles

The First inspection of JT-60U W-shaped divertor after high power operation

Masaki, Kei; ; Morimoto, Masaaki*; ; *; Hosogane, Nobuyuki; Saido, Masahiro

Fusion Technology 1998, p.67 - 70, 1998/00

no abstracts in English

Journal Articles

Installation of the W-shaped divertor in JT-60U

; ; Masaki, Kei; Hosogane, Nobuyuki; Sakurai, Shinji; Morimoto, Masaaki*; Miyo, Yasuhiko; Hiratsuka, Hajime; Akino, Noboru; *; et al.

Proc. of 17th IEEE/NPSS Symposium Fusion Engineering (SOFE'97), 2, p.365 - 368, 1998/00

no abstracts in English

Oral presentation

Thermal physical properties of MOX Fuels, 3; Measurements of thermal diffusivity for (U$$_{0.68}$$Pu$$_{0.3}$$Am$$_{0.02}$$)O$$_{x}$$ (x=1.90$$sim$$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.

Oral presentation

Thermal physical properties of MOX fuels, 2; Thermal diffusivity measurement of (U, Pu, Am)O$$_{2.00}$$

Komeno, Akira; Kato, Masato; Morimoto, Kyoichi; Kashimura, Motoaki; Abe, Tomoyuki; Ogasawara, Masahiro*; Sunaoshi, Takeo*; Uno, Hiroki*; Tamura, Tetsuya*

no journal, , 

no abstracts in English

Oral presentation

Enthalpy measurement and heat capacity evaluation of (U, Pu, Am)O$$_{2}$$

Komeno, Akira; Morimoto, Kyoichi; Kato, Masato; Kashimura, Motoaki; Abe, Tomoyuki; Ogasawara, Masahiro*; Sunaoshi, Takeo*; Uno, Hiroki*; Tamura, Tetsuya*

no journal, , 

no abstracts in English

Oral presentation

Thermal conductivity of (U$$_{0.68}$$Pu$$_{0.3}$$Am$$_{0.02}$$)O$$_{2-x}$$ at temperatures from 900 to 2200K

Morimoto, Kyoichi; Kato, Masato; Komeno, Akira; Kashimura, Motoaki; Abe, Tomoyuki; Ogasawara, Masahiro*; Sunaoshi, Takeo*; Uno, Hiroki*; Tamura, Tetsuya*

no journal, , 

no abstracts in English

46 (Records 1-20 displayed on this page)