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Nagata, Hiroshi; Kochiyama, Mami; Chinone, Marina; Sugaya, Naoto; Nishimura, Arashi; Ishikawa, Joji; Sakai, Akihiro; Ide, Hiroshi
JAEA-Data/Code 2024-016, 44 Pages, 2025/03
JAEA-Data-Code-2024-016.pdf:3.54MB
The elemental composition of the structural materials of nuclear reactor facilities is used as one of the important parameters in activation calculations that are evaluated when formulating decommissioning plans. Regarding the elemental composition of aluminum alloys and other materials used as structural materials for test and research reactors, sufficient data is not available regarding elements other than the major elements. For this reason, samples were collected from aluminum alloy, beryllium, hafnium, and other materials that have been used as the main structural materials of JMTR (Japan Materials Testing Reactor), and their elemental compositions were analyzed. This report summarizes the elemental composition data of 78 elements obtained in FY2023.
Yamamoto, Keisuke; Nakagawa, Takuya; Shimojo, Hiroto; Kijima, Jun; Miura, Daiya; Onose, Yoshihiko*; Namba, Koji*; Uchida, Hiroaki*; Sakamoto, Kazuhiko*; Ono, Chika*; et al.
JAEA-Technology 2024-019, 211 Pages, 2025/02
JAEA-Technology-2024-019.pdf:35.35MB
The uranium enrichment facilities at the Nuclear Fuel Cycle Engineering Laboratories of Japan Atomic Energy Agency (JAEA) were constructed sequentially to develop uranium enrichment technology with centrifugal separation method. The developed technologies were transferred to Japan Nuclear Fuel Limited until 2001. And the original purpose has been achieved. Wastewater Treatment Facility, one of the uranium enrichment facilities, was constructed in 1976 to treat radioactive liquid waste generated at the facilities, and it finished the role in 2008. In accordance with the Medium/Long-Term Management Plan of JAEA Facilities, interior equipment installed in this facility had been dismantled and removed since November 2021 to August 2023. This report summarizes the findings obtained through the work related to the contamination inspection methods cancellation the controlled area of Wastewater Treatment Facility from September 2023 to March 2024.
gamma spectrometry intercomparison based on the information obtained through five intercomparisons during the Fukushima mapping projectMikami, Satoshi; Ishikawa, Daisuke*; Matsuda, Hideo*; Hoshide, Yoshifumi*; Okuda, Naotoshi*; Sakamoto, Ryuichi*; Saito, Kimiaki
Journal of Environmental Radioactivity, 210, p.105938_1 - 105938_7, 2019/12
Times Cited Count：4 Percentile：9.63(Environmental Sciences)Five intercomparisons of in situ 
spectrometry by 6-7 participating teams have been conducted between December 2011 and August 2015 at sites in Fukushima prefecture which affected by the fallout of FDNPS accident occurred in March 2011. The evaluated deposition densities agreed within 5-6% in terms of coefficient of variation (CV) for radiocesium (
Cs and 
Cs), by our best achievement, and the ratio of Cs/Cs in deposition density agreed within 1-2% in CV, through five intercomparisons. These results guarantee the accuracy of the measurements of the mapping project. Two different methods for intercomparison were conducted: (1) sequential measurements at an identical point; and (2) simultaneous measurements in a narrow area within 3 m radius. In a comparison between the two methods at a site, no significant difference was observed between the results. The standard protocols for the two different intercomparison methods were proposed based on our experience.
Cu
Al
bulk glassy alloyOnodera, Naoto*; Ishii, Akito*; Ishii, Koji*; Iwase, Akihiro*; Yokoyama, Yoshihiko*; Saito, Yuichi; Ishikawa, Norito; Yabuuchi, Atsushi*; Hori, Fuminobu*
Nuclear Instruments and Methods in Physics Research B, 314, p.122 - 124, 2013/11
Times Cited Count：3 Percentile：25.51(Instruments & Instrumentation)It has been reported that heavy ion irradiation causes softening in some cases of Zr-based bulk metallic glass alloys. However, the fundamental mechanisms of such softening have not been clarified yet. In this study, ZrCuAl bulk glassy alloys were irradiated with heavy ions of 10 MeV I at room temperature. Positron annihilation measurements have performed before and after irradiation to investigate changes in free volume. We discuss the relationship between the energy loss and local open volume change after 10 MeV I irradiation compared with those obtained for 200 MeV Xe and 5 MeV Al. The energy loss analysis in ion irradiation for the positron lifetime has revealed that the decreasing trend of positron lifetime is well expressed as a function of total electronic energy deposition rather than total elastic energy deposition. It means that the positron lifetime change by the irradiation has a relationship with the inelastic collisions with electrons during heavy ion irradiation.
Sakuraba, Naotoshi; Numata, Masami; Komiya, Tomokazu; Ichise, Kenichi; Nishi, Masahiro; Tomita, Takeshi; Usami, Koji; Endo, Shinya; Miyata, Seiichi; Kurosawa, Tatsuya; et al.
JAEA-Technology 2009-071, 34 Pages, 2010/03
JAEA-Technology-2009-071.pdf:21.07MB
As a part of maintenance technology of a large-sized glove box for handling of TRU nuclides, we developed replacement technology for front acrylic panels using the bag-in/bag-out method and applied this technology to replace the deteriorated front acrylic panels at Waste Safety Testing Facility (WASTEF) in Nuclear Science Research Institute of Japan Atomic Energy Agency (JAEA). As a consequence, we could safely replace the front acrylic panels under the condition of continuous negative pressure only with partial decontamination of the glove box. We also demonstrated that the present technology is highly effective in points of safety, workability and cost as compared to the usual replacement technology for front acrylic panels of a glove box, where workers in an air-line suit replace directly the front acrylic panels in a green house.
Sakakibara, Tetsuro; Aoyama, Yoshio; Yamaguchi, Hiromi; Sasaki, Naoto*; Nishikawa, Takeshi*; Murata, Minoru*; Park, J.*; Taniguchi, Shoji*; Fujita, Michiru*; Fukuda, Tomoyuki*; et al.
Proceedings of International Waste Management Symposium 2009 (WM '09) (CD-ROM), 15 Pages, 2009/03
The volume reduction treatment of solid waste system by ultra-high frequency induction furnace (UHFIF) was developed from FY2005 to FY2007. Basic data for melting performance were collected by non-radioactive experiments using the bench scale UHFIF with a crucible capacity of 10 liters. Based on the obtained data, engineering specifications were evaluated for a demonstration scale UHFIF with a crucible capacity of 30 liters. A new demonstration scale UHFIF was constructed and melting experiments of surrogate wastes were carried out by this furnace. It was confirmed that the demonstration scale UHFIF can melt ferrous metal, ceramics and aluminum all together and stabilize aluminum by oxidation to alumina. Density, chemical composition, and surface condition of the solidified substances were analyzed, and homogeneity of the solidified substances was confirmed. Melting behavior in the demonstration scale UHFIF was analyzed by computer simulation and simulation results agreed well with the experimental ones. From the design study for a full scale UHFIF with a crucible capacity of 100 liters, basic specifications were evaluated for the full scale UHFIF. Based on the obtained specification, melting behavior in the full scale UHFIF was analyzed by computer simulation.
Konomura, Mamoru; Ogawa, Takashi; Okano, Yasushi; Yamaguchi, Hiroyuki; Murakami, Tsutomu; Takaki, Naoyuki; Nishiguchi, Youhei; Sugino, Kazuteru; Naganuma, Masayuki; Hishida, Masahiko; et al.
JNC TN9400 2004-035, 2071 Pages, 2004/06
JNC-TN9400-2004-035.pdf:76.42MB
The attractive concepts for Sodium-, lead-bismuth-, helium- and water-cooled FBRs have been created through using typical plant features and employing advanced technologies. Efforts on evaluating technological prospects of feasibility have been paid for these concepts. Also, it was comfirmed if these concepts satisfy design requierments of capability and performance presumed in the feasibilty study on commertialization of Fast Breeder Reactor Systems. As results, it was concluded that the selection of sodium-cooled reactor was most rational for practical use of FBR technologies in 2015.
Yokoyama, Kenji; Hosogai, Hiromi*; Chiba, Go; Kasahara, Naoto; Ishikawa, Makoto
JNC TN9400 2004-022, 162 Pages, 2004/04
JNC-TN9400-2004-022.pdf:13.47MB
In the fast reactor development, numerical simulation using analysis code plays an important role for complementing theory and experiment. In order to efficiently advance the research and development of fast reactors, JNC promotes the development of next generation simulation code (NGSC). In this report, research result by prototyping which carried out for the conceptual design of the NGSC is described. From the viewpoint of the cooperative research with CEA (Commissariat a l'Energie Atomique) in France, a trend survey on several platforms for numerical analysis and an applicability evaluation of CEA's SALOME platform for the NGSC were carried out. As a result of the evaluation, it is confirmed that SALOME had been satisfied the features of efficiency, openness, universality, expansibility and completeness that are required by the NGSC. In addition, it is confirmed that SALOME had the concept of the control layer required by the NGSC and would be one of the important candidates as a platform of the NGSC. In the field of the structure analysis, the prototype of the PARTS.NET code was reexamined from the viewpoint of class structure and input/output specification in order to improve the data processing efficiency and maintainability. In the field of the reactor physics analysis, a development test of a new code with C++ and a reuse test of an existing code written in Fortran was carried out in view of utilizing SALOME for the NGSC.
Yokoyama, Kenji; Hosogai, Hiromi*; Uto, Nariaki; Kasahara, Naoto; Ishikawa, Makoto
JNC TN9400 2003-021, 205 Pages, 2003/04
JNC-TN9400-2003-021.pdf:8.86MB
In the fast reactor development, numerical simulation using analytical codes plays an important role for complementing theory and experiment. It is necessary that the engineering models and analysis methods can be flexibly changed, because the phenomina to be investigated become more complicated due to the diversity of the needs for research. And, there are large problems in combining physical propaties and engineering models in many different fields. Aming to the realization of the next generation code system which can solve those problems, the authors adopted three methods, (1)Multi-language (SoftWIRE.NET, Visual Basic .NET and Fortran) (2)Fortran90 and (3)Python to make a prototype of the next generation code system. As this result, the followings were comfirmed. (1)It is possible to reuse a function of the existing codes written in Fortran as an object of the next generation code system by using visual Basic .NET. (2)The maintenanability of the existing code written by Fortran77 can be improved by using the new features of Fortran90. (3)The toolbox-type code system can be built by using Python.
Yokoyama, Kenji; Uto, Nariaki; kasahara, Naoto; ; Ishikawa, Makoto
Nihon Genshiryoku Gakkai 2003-Nen Aki No Taikai, 2(E64), 343 Pages, 2003/00
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Yokoyama, Kenji; Hosogai, Hiromi*; Uto, Nariaki; Kasahara, Naoto; Nagura, Fuminori; ; ; Ishikawa, Makoto
JNC TN9420 2002-004, 309 Pages, 2002/11
JNC-TN9420-2002-004.pdf:11.4MB
In the fast reactor development, numerical simulation using analytical codes plays an important role for complementing theory and experiment. It is necessary that the engineering models and analysis methods can be flexibly changed, because the phenomina to be investigated become more complicated due to the diversity of the needs for research. And, there are large problems in combining physical propaties and engineering models in many different fields. In this study, the goal is to develop a flexible and general-purposive analysis system, in which the phisical propaties and engineering models are replesented as a programming languare or a diagams that are easily understandable for humans and executable for computers. The authors named this concept the Engineering Modeling Language(EML). This report describes the result of the investigation for latest computer technologies and software development techniques which seem to be usable for a realization of the analysis code system for nuclear engineering as an EML.
Goto, Shunji*; Takeshita, Kunikazu*; Suzuki, Yoshio*; Ohashi, Haruhiko*; Asano, Yoshihiro; Kimura, Hiroaki*; Matsushita, Tomohiro*; Yagi, Naoto*; Isshiki, Maiko*; Yamazaki, H.*; et al.
Nuclear Instruments and Methods in Physics Research A, 467-468(Part1), p.682 - 685, 2001/07
Times Cited Count：146 Percentile：99.12(Instruments & Instrumentation)no abstracts in English
Goto, Shunji*; Takeshita, Kunikazu*; Suzuki, Yoshio*; Ohashi, Haruhiko*; Asano, Yoshihiro; Kimura, Hiroaki*; Matsushita, Tomohiro*; Yagi, Naoto*; Isshiki, M.*; Yamazaki, H.*; et al.
Nuclear Instruments and Methods in Physics Research A, 467-468(Part1), p.682 - 685, 2001/07
no abstracts in English
