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Aoshima, Atsushi; Suzuki, Yoshiharu; Namekawa, Takashi
Nihon Genshiryoku Gakkai-Shi ATOMO
, 55(12), p.733 - 736, 2013/12
no abstracts in English
Kitamura, Akihiro; Namekawa, Takashi; Hiramatsu, Kosuke*; Sankai, Yoshiyuki*
Nuclear Technology, 184(3), p.310 - 319, 2013/12
Times Cited Count：7 Percentile：49.28(Nuclear Science & Technology)A remote control system to operate manipulator arm by robot suit HAL, hybrid assisted limb, is examined for in-cell equipment maintenance application. In this integrated system operator wear exoskeletal structured HAL and movement and intention of operator are transferred through computer system to the slave type manipulator arm. The system provides position control scheme and bioelectrical signals control scheme. In order to evaluate the effectiveness and usefulness of the system, we implemented remote handling experiments using mock up equipment and compared the performances of remote operation conducted by the abovementioned two schemes with that by 3D mouse control scheme. Among these three control schemes, bioelectrical signals control scheme had outperformed the others. Under the bioelectrical signals control scheme, the system achieved expected results in executing direct contact tasks of equipment maintenance with small operation time and small variation.
Kitamura, Akihiro; Nakai, Koji; Namekawa, Takashi; Watahiki, Masatoshi
Nuclear Engineering and Design, 241(7), p.2614 - 2623, 2011/07
Times Cited Count：5 Percentile：38.65(Nuclear Science & Technology)We developed a remote control system to display recreated three dimensional information of workspace from measured data obtained by laser rangefinder and to operate manipulator arm remotely. In order to evaluate the effectiveness and usefulness of developed system, we implemented remote handling experiments using mock up equipment and compared the performances of remote operation conducted by the present control system with that by the usual camera based control system. Impressions of operator on the performance of each system are collected and NASA TLX tests are conducted. It was observed that the present system reduced workload stresses and reinforced visual information during remote operation.
Suzuki, Mitsutoshi; Namekawa, Takashi; Asano, Takashi; Niita, Koji*
Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10
The advanced MOX fabrication process in FaCT project has been studies to investigate an overall characteristic of nuclear material accounting of Pu in the proposed confinement box. Flow field induced by a forced convection inside the box is numerically simulated to evaluate the MOX particle behavior and a radiation field due to the spontaneous and induced neutrons emitted from Cm and Pu is calculated using PHITS code. The possibility of remote-monitoring techniques using non-destructive assay to apply to a future safegurads measure is invetigated.
Funasaka, Hideyuki; Nakamura, Hirofumi; Namekawa, Takashi
Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10
On the FaCT project in Japan, six development issues for the advanced aqueous reprocessing system are identified. Crystallization is the one of the major technologies to recover a large amount of uranium from the dissolver solution. The residual solution is loaded to the extraction process for U, Pu and Np co-recovery. Raffinate solution from the extraction process contains FP and MA, where MA is extracted by chromatography. These processes are expected to be more efficient in costs, wastes management, and the plutonium proliferation. As for the fuel fabrication technology, six development issues are identified. The source powder preparation technology of the conversion and granulation process is the most essential for the simplified pelletizing method technology. Remote operation and remote maintenance technique is essential to handle the low decontaminated TRU fuel in a hot cell. Current development status and plan of this integrated system until around 2015 is reported.
Ito, Masanori; Funasaka, Hideyuki; Namekawa, Takashi
Proceedings of European Nuclear Conference 2007 (ENC 2007) (CD-ROM), 7 Pages, 2007/09
The FaCT project in Japan is implemented purposing to decide the adoption of innovative technologies by 2010 and to judge the prospect of the applicability of innovative technologies to the commercialized fast reactor cycle system by 2015. Innovative technologies to be developed are identified as six development issues for advanced aqueous reprocessing system and six ones for simplified pelletizing method fuel fabrication system. As for reprocessing technology, uranium crystallization and extraction chromatography are important. Wide range of development work from chemical fundamental study to engineering-scale equipment operation are planned and practiced. As for fuel fabrication, conversion and granulation are important to get the source MOX powder with good flowability. Development of modular equipments with grater remote maintenance performance and repairing system in a hot cell is essential to put the low decontaminated TRU fuel fabrication into practical use.
Kawaguchi, Koichi; Namekawa, Takashi
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.290 - 295, 2007/09
The JAEA has developed advanced FBR cycle system since 1999 as the Feasibility Study (FS). Several combination of fuel and reactor type, reprocessing method and fuel fabrication method were studied. As the result of FS, the combination of oxide fuel, sodium cooling reactor, advanced aqueous reprocessing system and simplified pelletizing fuel fabrication system is chosen as the most promissing fuel cycle system. In the Fast Reactor Cycle technology (FaCT), six development issues for simplified pelletising technology were selected. TRU fuel handling technology, which is heat removal from nuclear fuel material, is one of these issues. Accumulation of decay heat of MA which is contained in TRU fuel cause oxidation of fuel powder, fuel pellet and cladding tube. Authors designed concept of powder hoppper, O/M adjusting furnace and fuel assembling equipment with heat removal function, and evaluated temperature distribution using thermal hydraulics analysis technique. As a result, it is shown that it is possible to cool fuel materials with specific heat generation up to 20 W/kgHM enough, though more detailed study is required for comprehensive equipments.
Namekawa, Takashi; Kawaguchi, Koichi; Koike, Kazuhiro; Haraguchi, Shingo; Ishii, Satoru
Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10
New concepts of future fabrication system for low decontaminated transuranium (TRU) fuel was designed. It is essential to establish a remote fabrication system in the hot cell, because materials with high levels of radiation and high generation of heat are handled in the system. The configuration of the present systems including aspects related to economy was clarified through the conceptual design study. Technical feasibility of each concepts is follwiong. For oxide fuels, the simplified pelletizing method has a high technical feasibility for the process, and its expected practical use is possible at early stage, because this method is based on wealth results of a conventional method. The sphere-pack method has the advantage of lesser dispersion of the fine powder due to the use of solution and granule in the process. However this system will shoulder additional cost for the liquid waste treatment process due to need dispose of a large bulk of process liquid waste. The vipack method has the possibility of economical improvement if simplification of the process can be achieved. However this system has some problems for quality assurance etc. For the casting method of metal fuel, high economical efficiency is generally expected of small-scale facilities, although verification of fabrication of the TRU alloy slug is required. For nitride fuel, technology developments for N-15 enrichment and recycling, and nitride conversion process etc. In particular, for coated particle fuel fabrication, crucial technology developments are required on coating and assembly process.
Osaka, Masahiko; Namekawa, Takashi; Kurosaki, Ken*; Yamanaka, Shinsuke*
Journal of Nuclear Materials, 344(1-3), p.230 - 234, 2005/09
Times Cited Count：13 Percentile：65.42(Materials Science, Multidisciplinary)The oxygen potential isotherms of (U,Pu,Am)O were represented by a chemical thermodynamic model proposed by Lindemer et al. It was assumed in the present model that (U,Pu,Am)O consisted of the chemical species PuO
, Pu
O, AmO, Am
O and UO in a pseudo-quaternary system by treating the reduction rates of Pu and Am as identical; furthermore an interaction between AmO and UO was introduced. The agreement between analytical and experimental isotherms was good, but the analytical values slightly overestimated the experimental values especially in the case of lower Am content. Adding an interaction between AmO and PuO to the model resulted in a better representation.
,Pu
,Am
)O solid solutionsOsaka, Masahiko; Sato, Isamu; Namekawa, Takashi; Kurosaki, Ken*; Yamanaka, Shinsuke*
Journal of Alloys and Compounds, 397(1-2), p.110 - 114, 2005/07
Times Cited Count：26 Percentile：79.02(Chemistry, Physical)Oxygen potentials of (U,Pu)O containing AmO, (UPuAm)O, were measured as functions of the oxygen to metal (O/M) ratios and temperatures by hermogravimetric analysis (TGA) and discussed from thermodynamic viewpoints. The (UPuAm)O solid solution was prepared by a traditional powder metallurgical route. Oxygen partial pressure (pO) was adjusted in the region from 10
to 10
MPa by using CO/H and HO /H equilibria. Changes of microgram order in the specimen weight were continuously monitored with a TGA apparatus and the dependences of oxygen potentials on the O/M ratios were obtained at 1123K, 1273K and 1423K. The oxygen potentials of (UPuAm)O were significantly higher than those in (U
Pu
)O. The partial molar entropy and enthalpy of oxygen have been calculated and compared with those of Am O, (Am
U) O, and (UPu)O. Finally, the oxygen potential isotherms of (U,Pu,Am)O were examined in terms of defect structure.
Kawaguchi, Koichi; Namekawa, Takashi; Suzuki, Yoshihiro; Haraguchi, Shingo*
Proceedings of 13th International Conference on Nuclear Engineering (ICONE-13) (CD-ROM), 7 Pages, 2005/05
The conceptual design study for advanced FBR fuel fabrication system has been performed for the purpose that the feature of small-scale fabrication system in the transition stage from LWR to FBR fuel cycle. On the small-scale of 50 ton heavy metal per year fabrication system, dry type fabrication systems have superior cost performance than the wet type, although waste amount is larger.
; Measurement and ModelingOsaka, Masahiko; Sato, Isamu; Namekawa, Takashi; Tanaka, Kenya; Ishida, Takashi*
JNC TN9400 2004-076, 34 Pages, 2005/03
JNC-TN9400-2004-076.pdf:0.75MB
Oxygen potentials of a nonstoichiometric oxide, (U
Pu
Am
)O
, were measured as a function of oxygen to metal (O/M) ratio at 1123 K, 1273 K and 1423 K by thermogravimetric analysis (TGA) with CO
/H and HO/H gas equilibria. The oxygen potentials of (UPuAm)O were significantly higher than those of MOX containing no Am, although the content of Am is only 4.5%. The defect structure was examined from slopes of the oxygen partial pressure versus deviation from stoichiometry. A high slope value was observed in (UPuAm)O, which is attributed not to defects or defect clusters, but to possible formation of an ordered intermediate phase. This fact was supported by the shape of oxygen partial molar enthalpy and entropy. The oxygen potential isotherms of (U,Pu,Am)O were represented by a chemical thermodynamic model proposed by Lindemer et al. It was assumed in the present model that (U,Pu,Am)O consisted of the five chemical species in a pseudo-quaternary system by treating the reduction rates of Pu and Am as identical; furthermore an interaction between and was introduced. The agreement between analytical and experimental isotherms was good.
Ohno, Shuji; Sato, Isamu; Nakagiri, Toshio; Hirosawa, Takashi; Miyahara, Shinya; Namekawa, Takashi
JAERI-Review-2004-021, p.199 - 208, 2004/10
Out-of-pile experiments on the release of fission products (FPs) under transient heating conditions were carried out for mixed oxide fuels. The fuels used in the experiments, the plutonium content of which was 30wt%, were irradiated up to 65 GWd/t in the experimental fast reactor JOYO. The experiments consisted of two runs, FP-1 and FP-2. In FP-1, the fuel sample was first heated to 2,000
C and then up to 3,000C. The holding time was 30 min at each temperature. In FP-2, the terminal temperatures were 1,500C and 2,500C, and the holding time was 30 min in the same manner. The release of Cs, a volatile FP, was detected as soon as the fuel sample was heated up. The release rate, after peaking in several minutes, decreased gradually via a diffusion process in the fuel matrix. The "gross" diffusion coefficient agreed well with the data reported in other experiments using LWR fuels. The release fractions were identical in both experiments, namely Cs 100%, Sb~100%, Ru 10%, Ce 0% and Eu 0%.
by an in-cell remote processYoshimochi, Hiroshi; Nemoto, Masanao; Mondo, Kenji; Koyama, Shinichi; Namekawa, Takashi
Journal of Nuclear Science and Technology, 41(8), p.850 - 856, 2004/08
Focusing on the cover layer materials (as the Radon Barrier Materials), which could have the effect to restrain the radon from scattering into the air and the effect of the radiation shielding, we produced the radon barrier materials with crude bentonite on an experimental basis, using the rotary type comprehensive unit for grinding and mixing, through which we carried out the evaluation of the characteristics thereof.
Sato, Koji; Koma, Yoshikazu; Inoue, Akira; Yonezawa, Shigeaki; Takata, Takeshi; Nakabayashi, Hiroki; Namekawa, Takashi; Kawaguchi, Koichi
JNC TN9400 2004-036, 1051 Pages, 2004/06
JNC-TN9400-2004-036.pdf:90.95MB
The plant concept concerning the fuel cycle systems (combination of the reprocessing and the fuel fabrication )has been constructed to reduce their total cost by the introduction of various innovative techniques and to apply their utmost superior efficiency from such standpoints of a decrease in the environmental burden, better resource utilization and proliferation resistance improvement by the low decontamination transuranium element (TRU) recycle. For economical efficiency, less than 0.8 yen/kWh which is the demand value (total of the reprocessing expense and the fuel fabrication expense) of the fuel cycle expense satisfied each combination case at 200 tHM/y scale provisionally set for large-scale facilities. On the other hand, the combination case with a low breeder reactor core has satisfied the demand value, with improvement of the average burnup by the radial direction blanket fuel deletion contributing to the decrease of the fuel cycle expense at 50 tHM/y scale provisionally set for small-scale facilities.
Yamanaka, Shinsuke*; Uno, Masayoshi*; Kurosaki, Ken*; Osaka, Masahiko; Sato, Isamu; Namekawa, Takashi; Kato, Masato; Kihara, Yoshiyuki
JNC TY9400 2004-001, 111 Pages, 2004/04
JNC-TY9400-2004-001.pdf:2.81MB
As the fundamental study for MOX containing MA fuel, property measurement and its evaluation has been performed by JNC and Osaka university, respectively. Each results were compared, and the properties of MOX containing MA has been evaluated totally.
N
Inert MatrixYano, Toyohiko*; Osaka, Masahiko; Namekawa, Takashi
JNC TY9400 2004-002, 84 Pages, 2004/03
None
Koyama, Shinichi; Osaka, Masahiko; Namekawa, Takashi; Ito, Masahiko
Proceedings of 7th International Conference on Nuclear Criticality Safety (ICNC 2003), 0 Pages, 2003/10
Japan Nuclear Cycle Development Institute (JNC) is now totally promoting the development of commercialized fast reactors to realize stable supply of energy in future. One of the important items is to develop hign-performance fuel. For that purpose,it is e
Koyama, Shinichi; Osaka, Masahiko; Sekine, Takashi; Morozumi, Katsufumi; Namekawa, Takushi;
Journal of Nuclear Science and Technology, 40(2), p.998 - 1013, 2003/02
Times Cited Count：23 Percentile：80.7(Nuclear Science & Technology)None
Koyama, Shinichi; Namekawa, Takashi
JNC TN9400 2002-060, 28 Pages, 2002/12
JNC-TN9400-2002-060.pdf:0.91MB
Solubility test was performed to evaluate the fundamental solubility characteristics of neptunium and insoluble residue after dissolution of MOX fuels irradiated in LWRs. (1)Neptunium was observed in insoluble residue of unirradiated MOX fuel as well as dissolved solution. (2)
Np content in dissolution liquid of unirradiated fuels ranged from 0.06 to 0.13% and that of irradiated fuels ranged from 0.03 to 0.1%. (3)Np content in MOX fuels decreased exponentially with increasing burn-up. (4)The solubility of neptunium in unirradiated fuels ranged from 84.1 to 98.1% and that in irradiated fuels ranged from 95.9 to 99.9%. (5)The solubility of neptunium in unirradiated and irradiated fuels slightly decreased both with initial plutonium content.
