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Kasahara, Naoto*; Yamano, Hidemasa; Nakamura, Izumi*; Demachi, Kazuyuki*; Sato, Takuya*; Ichimiya, Masakazu*
Transactions of 26th International Conference on Structural Mechanics in Reactor Technology (SMiRT-26) (Internet), 8 Pages, 2022/07
Utilizing fracture control, we are developing a technology to suppress the expansion of damage caused by an event that exceeds the design assumption. We made a plan to develop three issues; (1) Technology for mitigating failure consequence at extremely high temperatures, (2) Technology for mitigating failure consequence against excessive earthquakes, and (3) Methodology for improving reactor structure resilience.
Yada, Hiroki; Ando, Masanori; Tsukimori, Kazuyuki; Ichimiya, Masakazu*; Anoda, Yoshinari*
Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 9 Pages, 2018/07
Containment vessel (CV) of nuclear power plants is an important structure to prevent radioactive release, however, the safety margin of the CV against pressure are not numerically clarified. The head plate structure is included in CV boundary of fast reactor. In order to develop the evaluation method of the ultimate strength of the head plate structure at beyond the specified limit, pressure failure tests and finite element analysis (FEA) of the head plates subjected to convex side pressure were performed. In the test of the relative thin thickness head plate, non-axisymmetric deformations was observed in post buckling behavior and failure pressure was lower than other cases. To evaluate non-axisymmetric deformations in the test, FEA using 3-D solid model constructed by precise dimensions of the test specimen, moreover, FEA using simplified model with uniform or non-uniform thickness were performed. Through analyses, the feature of the post buckling behavior was discussed.
Tsukimori, Kazuyuki*; Yada, Hiroki; Ando, Masanori; Ichimiya, Masakazu*; Anoda, Yoshinari*
Proceedings of 12th International Conference on Asian Structure Integrity of Nuclear Components (ASINCO-12) (CD-ROM), p.105 - 121, 2018/04
In FBR plants the head plate constitutes a part of the boundary of the containment vessel (CV), therefore, it is an important issue if the function as the boundary is maintained or not in the severe accident. And also it is important to evaluate the leak rate from the penetrated crack of the head plate, in order to estimate the effect of released fission product out of CV. Authors conducted pressure endurance tests of head plate specimens subjected to external pressure, which covered post-buckling behaviors and until crack penetration. In this paper leak rate test results at several pressure levels are introduced and the tendency of leak rate behaviors with relation of the penetrate crack length and the pressure level are discussed. Also, the modeling of head plate thickness distribution for 3-D analyses based on the detailed 3-D measurement data of specimens is discussed, which possibly relates to the 3-D deformation patterns observed in the tests and the length of penetration cracks.
Ando, Masanori; Yada, Hiroki; Tsukimori, Kazuyuki; Ichimiya, Masakazu*; Anoda, Yoshinari*
Journal of Pressure Vessel Technology, 139(6), p.061201_1 - 061201_12, 2017/08
Times Cited Count：1 Percentile：7.39(Engineering, Mechanical)In this study, in order to develop the evaluation method of the ultimate pressure of the bellows structure subject to the internal pressure beyond the specified, the failure test and finite element analysis (FEA) of the bellows structure were performed. The failure modes were demonstrated through the series of tests, and three kind of failure mode were observed. To simulate the buckling and deformation behavior during the test, the implicit and explicit analyses were performed.
Tsukimori, Kazuyuki; Ando, Masanori; Yada, Hiroki; Ichimiya, Masakazu*; Anoda, Yoshinari*; Arakawa, Manabu*
Transactions of 24th International Conference on Structural Mechanics in Reactor Technology (SMiRT-24) (USB Flash Drive), 10 Pages, 2017/08
The analytical treatment of Multi-ply bellows behaviours is difficult compared with that of single-ply bellows, since the uncertainty of friction between plies exists. In this study verification was conducted based on experiments by comparing between two-ply and single-ply bellows test results. Following results were obtained. The spring rate of two-ply bellows is approximately twice of that of single-ply bellows, even if internal pressure is loaded. Typical buckling behaviour of bellows, in-plane squirm, was observed in both cases of two-ply and single-ply bellows. The deformation patterns were similar with each other, but the pressure levels of two-ply bellows were approximately twice of those of single-ply bellows. These means the friction force can be ignored practically. As the conclusion, two-ply bellows analyses can be replaced by the analyses of single-ply bellows model with half pressure load and the effort of numerical analysis can be reduced.
Ando, Masanori; Yada, Hiroki; Tsukimori, Kazuyuki; Ichimiya, Masakazu*; Anoda, Yoshinari*
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 11 Pages, 2017/07
In this study, in order to develop the evaluation method of the pressure toughness of bellows structure under the beyond design base event, the pressure failure tests and finite element analysis (FEA) of the bellows structure subjected to internal pressure were performed. In the test of five convolutions 0.5 mm-thickness bellows specimen with guard pipe, the maximum pressure was larger than those in the tests without guard pipe specimens and ductile failure was observed. On the other hand, in the test of five convolutions 0.5 mm-thickness bellows specimen without guard pipe, local failure was observed. In the test of the six convolutions 1.0 mm-thickness bellows specimen, ductile failure was observed in the both single and double ply bellows. The maximum pressure obtained in all tests were about 10 times larger than the estimated results of limiting design pressure based on in-plain instability by the EJMA standards.
Yada, Hiroki; Ando, Masanori; Tsukimori, Kazuyuki; Ichimiya, Masakazu*; Anoda, Yoshinari*
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 8 Pages, 2017/07
The head plate that composes the boundary between primary and secondary coolant in intermediate heat exchanger of FBR has an important role when the progress of the BDBE is considered. In order to develop the evaluation method of the pressure toughness of the head plate under the BDBE, the pressure failure tests and finite element analysis of two types of head plate subjected to convex side pressure was performed in this study. It can be concluded that a failure mode of a head plate subjected convex side pressure is circumferential through-wall crack caused by straightening following the bending deformation near the rim.
Ando, Masanori; Yada, Hiroki; Tsukimori, Kazuyuki; Ichimiya, Masakazu*; Anoda, Yoshinari*
Proceedings of 2016 ASME Pressure Vessels and Piping Conference (PVP 2016) (Internet), 8 Pages, 2016/07
To clarify the progress of the events under the beyond design basis events, it is necessary to evaluate the pressure toughness of containment vessel adequately. The containment vessel of fast reactor is composed of the various structures, and one of the thinnest boundary structures is bellows structure to absorb the thermal expansion of the coolant piping penetrating the containment vessel. In addition to the containment vessel boundary, evaluating the pressure toughness of reactor coolant and gas boundary is also important because of same reason of that in the containment vessel boundary. In the primary coolant and gas boundary, the cover gas bellows of the intermediate heat exchanger in fast reactor is one of the thinnest structures and has important role when the progress of the BDBE is considered. Therefore, in order to develop the evaluation method of the pressure toughness of bellows structure under the BDBE, the pressure failure tests and finite element analysis of the bellows structure subjected to internal pressure were performed in this study.
Yada, Hiroki; Ando, Masanori; Tsukimori, Kazuyuki; Ichimiya, Masakazu*; Anoda, Yoshinari*
Proceedings of 2016 ASME Pressure Vessels and Piping Conference (PVP 2016) (Internet), 7 Pages, 2016/07
no abstracts in English
Saito, Shinzo; Okamoto, Koji*; Kataoka, Isao*; Sugiyama, Kenichiro*; Muramatsu, Ken*; Ichimiya, Masakazu*; Kondo, Satoru; Yonomoto, Taisuke
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 10 Pages, 2015/05
Konomura, Mamoru; Ichimiya, Masakazu; Mukai, Kazuo
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 9 Pages, 2012/00
Japanese prototype fast breeder reactor, Monju, will soon restart. Monju has three sodium loops with steam generators and a turbin-generator, 280 MWe. Monju is expected to demonstrate the sodium technology and the sodium-water heat exchange technology in Japan. It will be operated at full power operation during around 10 years after restart in order to accumulate operation & maintenance experience and to evaluate its design technology. After the system start-up test (SST), Monju will be operated under full power. In this stage, the main object of Monju operation will be to achieve its initial targets which were fixed when its construction was decided around thirty years ago. The targets are to demonstrate a safe and reliable operation, that is, accumulation of the operation & maintenance experience and evaluation of the design technology, and to establish sodium handling technology. For example, inspection and diagnosis technologies are important for maintenance of a sodium cooled reactor. An out-of-pile sodium test facility will be constructed near Monju in order to make many tests of inspection devices and research many chemical tests. At the same time, the activities for the performance improvement, for example, a new licence, will be prepared in order to utilize Monju as a R&D facility. After the accumulation of operating experience, Monju will be enhanced the performance as a R&D facility in order to demonstrate innovative technologies, for example, irradiation of advanced fuel, longer operation cycle, higher burnup. For this purpose, Monju will be needed to get a new licence and core modification. And Monju on-site non-destructive Post Irradiated Evaluation facility will be expected at this stage. There were many R&D works in Japan with sodium out-of-pile facilities. All the experience were reflected in the design of Monju. Monju will demonstrate a handling of sodium technologies under power plant operation.
Nagura, Fuminori; Ichimiya, Masakazu; Sagayama, Yutaka
Genshiryoku eye, 56(6), p.20 - 23, 2010/06
As for Japan because it is energy security guaranty and the like, Japan continued the research and development of FBR ("Joyo", "Monju") and fuel cycle. In recent years, there are a growing interest in the long-term energy security and terrestrial environmental problem. And FBR cycle has again been observed, by the development of young country and the like. This year, reactor operation was reopened after about 14 years, "Monju" is the only prototype FBR reactor which it works as the Western Worlds, and the result of these research and developments is expected even worldwide. JAEA is carrying on international cooperation of "JAEA-CEA", "JAEA-DOE", GIF, IAEA, for the points of view "The efficient development", "Decrease of development risk", "Establishment of worldwide standard technology", "Monju" establishes to the position as base of international research and development, and JAEA makes the best endeavors to "Monju" carrying out worldwide contribution.
Nagata, Takashi; Ichimiya, Masakazu; Funasaka, Hideyuki; Mizuta, Shunji; Nagura, Fuminori
Nihon Genshiryoku Gakkai-Shi ATOMO
, 51(4), p.234 - 238, 2009/04
no abstracts in English
Nagaoki, Yoshihiro; Kikuchi, Shin; Ichimiya, Masakazu
Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10
"Fast Reactor Cycle Technology Development (FaCT)" project has been conducted since 2006. In this project, design study and research and development (R&D) on innovative technologies for fast reactor (FR) cycle system are implemented in order to present the conceptual designs of commercial and demonstration facilities by 2015 and start operating demonstration fast reactor in 2025. The R&Ds has been stepped forward into the development stage to establish the realization of innovative technologies which bring excellent performance to fast reactor cycle system. The purpose of R&D by 2010 is to decide weather innovative technologies shall be adopted. So promoting R&D of FR, the project governance was organized. Furthermore, several possible R&D have been effectively carried out within the frameworks of international cooperation, such as GNEP, GIF, and INPRO.
Konomura, Mamoru; Ichimiya, Masakazu
Journal of Nuclear Materials, 371(1-3), p.250 - 269, 2007/09
Times Cited Count：18 Percentile：75.3(Materials Science, Multidisciplinary)no abstracts in English
Ichimiya, Masakazu; Mizuno, Tomoyasu; Kotake, Shoji
Nuclear Engineering and Technology, 39(3), p.171 - 186, 2007/06
Critical issues in the development targets for the future fast reactor (FR) cycle system, including odium-cooled FR were to ensure safety assurance, efficient utilization of resources, reduction of environmental burden, assurance of nuclear non-proliferation, and economic competitiveness. A promising design concept of sodium-cooled fast reactor JSFR is proposed aiming at fully satisfaction of the development targets for the next generation nuclear energy system. A roadmap toward JSFR commercializationis described, to be followed up in a new framework of the Fast reactor Cycle Technology development (FaCT) Project launched in 2006.
Minato, Kazuo; Ichimiya, Masakazu; Inoue, Tadashi*
Proceedings of 9th OECD/NEA Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation, p.35 - 43, 2007/00
no abstracts in English
Yamashita, Kiyonobu; Ichimiya, Masakazu; Yamashita, Hidetoshi
Proceedings of 14th Pacific Basin Nuclear Conference (PBNC-14), p.834 - 841, 2004/00
Large-scale Sodium-Cooled Fast Reactor Concept was created. It satisfies development targets such as ensuring safety, economic competitiveness, efficient utilization of resoueces and non-proliferation. Its concept and recent fundamental reseaches will be reported.
Dai-18-Kai Kankoku Genshiryoku Sangyo Kaigi / Kankoku Genshiryoku Gakkai Nenkai, 165 Pages, 2003/00
None
Ichimiya, Masakazu; Mizuno, Tomoyasu; Konomura, Mamoru
Global 2003; International Conference on Atoms for Prosperity: Upda, 0 Pages, 2003/00
An innovative concept of sodium-cooled fast reactor, named JNC Sodium Cooled FR (JSFR) has been created through the Feasibility Study on Commercialized FR Cycle System, aiming at full satisfaction of the development targets. It is evaluated that JSFR possesses the highest potential in terms of technological feasibility to respond the diverse needs of society. JSFR supported by an appropriate fuel cycle is recognized as a promising candidate for the next generation nuclear energy system, such as GenerationIV system.
