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Ohshima, Hiroyuki; Morishita, Masaki*; Aizawa, Kosuke; Ando, Masanori; Ashida, Takashi; Chikazawa, Yoshitaka; Doda, Norihiro; Enuma, Yasuhiro; Ezure, Toshiki; Fukano, Yoshitaka; et al.
Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07
This book is a collection of the past experience of design, construction, and operation of two reactors, the latest knowledge and technology for SFR designs, and the future prospects of SFR development in Japan. It is intended to provide the perspective and the relevant knowledge to enable readers to become more familiar with SFR technology.
F from 
O
?Tang, T. L.*; Uesaka, Tomohiro*; Kawase, Shoichiro; Beaumel, D.*; Dozono, Masanori*; Fujii, Toshihiko*; Fukuda, Naoki*; Fukunaga, Taku*; Galindo-Uribarri, A.*; Hwang, S. H.*; et al.
Physical Review Letters, 124(21), p.212502_1 - 212502_6, 2020/05
Times Cited Count：14 Percentile：73.46(Physics, Multidisciplinary)The structure of a neutron-rich F nucleus is investigated by a quasifree (
) knockout reaction. The sum of spectroscopic factors of 
orbital is found to be 1.0 
0.3. The result shows that the O core of F nucleus significantly differs from a free O nucleus, and the core consists of 
35% O
, and 65% excited O. The result shows that the O core of F nucleus significantly differs from a free O nucleus. The result may infer that the addition of the 
proton considerably changes the neutron structure in F from that in O, which could be a possible mechanism responsible for the oxygen dripline anomaly.
Otani, Akihito*; Kai, Satoru*; Kaneko, Naoaki*; Watakabe, Tomoyoshi; Ando, Masanori; Tsukimori, Kazuyuki*
Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 10 Pages, 2018/07
This paper demonstrates an application result of the JSME Seismic Code Case to an actual complex piping system. The secondary coolant piping system of Japanese Fast Breeder Reactor, Monju, was selected as a representative of the complex piping systems. The elastic-plastic time history analysis for the piping system was performed and the piping system has been evaluated according to the JSME Seismic Code Case. The evaluation by the Code Case provides a reasonable result in terms of the piping fatigue evaluation that governs seismic integrity of piping systems.
Wakai, Takashi; Kobayashi, Sumio; Kato, Shoichi; Ando, Masanori; Takasho, Hideki*
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 7 Pages, 2017/07
This paper describes a thermal fatigue test on a structural model with a dissimilar welded joint. In the present design of JSFR, there may be dissimilar welded joints between ferritic and austenitic steels especially in IHX and SG. Creep-fatigue is one of the most important failure modes in JSFR components. However, the creep-fatigue damage evaluation method has not been established for dissimilar welded joint. To investigate the evaluation method, structural test will be needed for verification. Therefore, a thermal fatigue test on a thick-wall cylinder with a circumferential dissimilar welded joint between Mod.9Cr-1Mo steel and 304SS was performed. Since the coefficients of thermal expansion of these steels were significantly different, buttering layer of Ni base alloy was installed between them. After the completion of the test, deep cracks were observed at the HAZ in 304SS, as well as at HAZ in Mod.9Cr-1Mo steel. There were many tiny surface cracks in BM of 304SS. According to the fatigue damage evaluation based on the finite element analysis results, the largest fatigue damage was calculated at HAZ in 304SS. Large fatigue damage was also estimated at BM of 304SS. Fatigue cracks were observed at HAZ and BM of 304SS in the test, so that analytical results are in a good agreement with the observations. However, though relatively small fatigue damage was estimated at HAZ in Mod.9Cr-1Mo steel, deep fatigue cracks were observed in the test. To identify the cause of such a discrepancy between the test and calculations, we performed a series of finite element analyses. Some metallurgical investigations were also performed.
Okajima, Satoshi; Wakai, Takashi; Ando, Masanori; Inoue, Yasuhiro*; Watanabe, Sota*
Journal of Pressure Vessel Technology, 138(5), p.051204_1 - 051204_6, 2016/10
Times Cited Count：2 Percentile：13.41(Engineering, Mechanical)Kogure, Toshihiro*; Yamaguchi, Noriko*; Segawa, Hiroyo*; Mukai, Hiroki*; Motai, Satoko*; Akiyama, Kotone*; Mitome, Masanori*; Hara, Toru*; Yaita, Tsuyoshi
Microscopy, 65(5), p.451 - 459, 2016/10
Times Cited Count：54 Percentile：96.98(Microscopy)Shimomura, Kenta; Kato, Shoichi; Wakai, Takashi; Ando, Masanori; Hirose, Yuichi*; Sato, Kenichiro*
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 9 Pages, 2015/05
This paper describes experimental and analytical works to confirm that the design standard for SFR components sufficiently covers possible failure mechanisms. Creep-fatigue damage evaluation method in JSME design standard for SFR components has been constructed based on experiments and/or numerical analyses of conventional austenitic stainless steels, such as 304SS. Since the material characteristics of Mod.9Cr-1Mo steel are substantially different from those of austenitic stainless steels, it is required to verify the applicability of the design standards to the SFR components made of Mod.9Cr-1Mo steel. A series of uni-axial creep-fatigue tests were conducted using double-ended notch bar specimens made of Mod.9Cr-1Mo steel under displacement controlled condition with 30 minute holding. The curvature radii of the specimens were 1.6mm, 11.2mm and 40.0mm. The specimen having 1.6mm notch and 11.2mm notch failed from outer surface but the specimen having 40.0mm notch showed obvious internal crack nucleation. In addition, though total duration time of the creep-fatigue test was only 2,000 hours, a lot of creep voids and inter granular crack growth were observed. To clarify the cause of such peculiar failure, some additional experiments were performed, as well as some numerical analyses. We could point out that such a peculiar failure aspect might result from corresponding stress distribution in the cross section. As a result of a series of investigations, possible causes of such peculiar failure could be narrowed down. A future investigation plan was proposed to clarify the most significant cause.
Okajima, Satoshi; Wakai, Takashi; Ando, Masanori; Inoue, Yasuhiro*; Watanabe, Sota*
Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 8 Pages, 2014/07
Asayama, Tai; Wakai, Takashi; Ando, Masanori; Okajima, Satoshi; Nagae, Yuji; Takaya, Shigeru; Onizawa, Takashi; Tsukimori, Kazuyuki; Morishita, Masaki
Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 6 Pages, 2014/07
This paper overviews the ongoing research and development as well as activities for codification of structural codes for the Japan Sodium Cooled Fast Reactor (JSFR). Not only the design and construction code which has been published and updated on a regular basis, codes on welding, fitness-for-service, leak-before-break evaluation as well as the guidelines for structural reliability evaluation are being developed. The basic strategy for the development is to fully take advantage of the favorable technical characteristics associated with sodium-cooled fast reactors; the codes will be developed based on the System Based Code concept. The above mentioned set of codes are planned to be published from the Japan Society of Mechanical Engineers in 2016.
Wakai, Takashi; Inoue, Osamu*; Ando, Masanori; Kobayashi, Sumio
Transactions of the 22nd International Conference on Structural Mechanics in Reactor Technology (SMiRT-22) (CD-ROM), 9 Pages, 2013/08
Watahiki, Shunsuke; Hanakawa, Hiroki; Imaizumi, Tomomi; Nagata, Hiroshi; Ide, Hiroshi; Komukai, Bunsaku; Kimura, Nobuaki; Miyauchi, Masaru; Ito, Masayasu; Nishikata, Kaori; et al.
JAEA-Technology 2013-021, 43 Pages, 2013/07
JAEA-Technology-2013-021.pdf:5.12MB
The number of research reactors in the world is decreasing because of their aging. On the other hand, the necessity of research reactor, which is used for human resources development, progress of the science and technology, industrial use and safety research is increasing for the countries which are planning to introduce the nuclear power plants. From above background, the Neutron Irradiation and Testing Reactor Center began to discuss a basic concept of Multipurpose Compact Research Reactor (MCRR) for education and training, etc., on 2010 to 2012. This activity is also expected to contribute to design tool improvement and human resource development in the center. In 2011, design study of reactor core, irradiation facilities with high versatility and practicality, and hot laboratory equipment for the production of Mo-99 was carried out. As the result of design study of reactor core, subcriticality and operation time of the reactor in consideration of an irradiation capsule, and about the transient response of the reactor to the reactivity disturbance during automatic control operation, it was possible to do automatic operation of MCRR, was confirmed. As the result of design study of irradiation facilities, it was confirmed that the implementation of an efficient mass production radioisotope Mo-99 can be expected. As the result of design study with hot laboratory facilities, Mo-99 production, RI export devised considered cell and facilities for exporting the specimens quickly was designed.
Hanawa, Satoshi; Nakamura, Takehiko; Uchida, Shunsuke; Kus, P.*; Vsolak, R.*; Kysela, J.*; Sakai, Masanori*
Nuclear Technology, 183(1), p.136 - 148, 2013/07
Times Cited Count：2 Percentile：18.55(Nuclear Science & Technology)no abstracts in English
Hirose, Yuichi*; Ando, Masanori; Onizawa, Takashi; Wakai, Takashi; Sato, Kenichiro*
Proceedings of 2013 International Congress on Advances in Nuclear Power Plants (ICAPP 2013) (USB Flash Drive), 6 Pages, 2013/04
The purpose of this study is to develop assessment of structural integrity for JSFR's primary system made from 316FR steel and Mod.9Cr-1Mo steel in severe accidents that sodium temperature exceeds the design basis temperature as 650 
C. It is important of sodium boundary to prevent damages in high-temperature environment. From this standpoint, the way of stress calculation, evaluation formula including limiting value, safety factor and cumulative damages are considered. This paper provides example to apply these assessment for JSFR under development in Japan.
Shinohara, Masanori; Motegi, Toshihiro; Saito, Kenji; Haga, Hiroyuki; Sasaki, Shinji; Katsuyama, Kozo; Takada, Kiyoshi*; Higashimura, Keisuke*; Fujii, Junichi*; Ukai, Takayuki*; et al.
JAEA-Technology 2012-032, 29 Pages, 2012/11
JAEA-Technology-2012-032.pdf:6.57MB
An event, in which one of WRMs were disabled to detect the neutron flux in the reactor core, occurred during the period of reactor shut down of HTTR in March, 2010. The actual life time of WRM was unexpectedly shorter than the past developed life time. Investigation of the cause of the outage of WRM toward the recovery of the life time up to the past developed life is one of the issues to develop the technology basis of HTGR. Then, two experimental investigations were carried out to reveal the cause of the malfunction by specifying the damaged part causing the event in the WRM. One is an experiment using a mock-up sample test which strength degradation on assembly accuracy and heat cycle to specify the damaged part in the WRM. The other is a destructive test in FMF to specify the damaged part in the WRM. This report summarized the results of the destructive test and the experimental investigation using the mock-up to reveal the cause of malfunction of WRM.
Watahiki, Masatoshi; Akai, Masanori; Nakai, Koji; Iemura, Keisuke; Yoshino, Masanori*; Hirano, Hiroshi*; Kitamura, Akihiro; Suzuki, Kazunori
Nihon Genshiryoku Gakkai Wabun Rombunshi, 11(1), p.101 - 109, 2012/02
Gloveboxes used for plutonium fuel development and fabrication are eventually dismantled for replacement or decommissioning. Since equipment interior and the inner surface of gloveboxes are contaminated in radioactive materials, glovebox dismantling work is performed by workers wearing an air fed suit with mechanical tools in a plastic enclosure system to control the spread of contamination. Various improvements of enclosure system are implemented including modification of the rooms to decontaminate and undress the air fed suit and introduction of inflammable filter and safety film near the size reduction workspace against fire. We describe the countermeasures deployed in the enclosure system against potential hazards and how these devices work in the real dismantling activities.
Imaizumi, Tomomi; Miyauchi, Masaru; Ito, Masayasu; Watahiki, Shunsuke; Nagata, Hiroshi; Hanakawa, Hiroki; Naka, Michihiro; Kawamata, Kazuo; Yamaura, Takayuki; Ide, Hiroshi; et al.
JAEA-Technology 2011-031, 123 Pages, 2012/01
JAEA-Technology-2011-031.pdf:16.08MB
The number of research reactors in the world is decreasing because of their aging. However, the planning to introduce the nuclear power plants is increasing in Asian countries. In these Asian countries, the key issue is the human resource development for operation and management of nuclear power plants after constructed them, and also the necessity of research reactor, which is used for lifetime extension of LWRs, progress of the science and technology, expansion of industry use, human resources training and so on, is increasing. From above backgrounds, the Neutron Irradiation and Testing Reactor Center began to discuss basic concept of a multipurpose low-power research reactor for education and training, etc. This design study is expected to contribute not only to design tool improvement and human resources development in the Neutron Irradiation and Testing Reactor Center but also to maintain and upgrade the technology on research reactors in nuclear power-related companies. This report treats the activities of the working group from July 2010 to June 2011 on the multipurpose low-power research reactor in the Neutron Irradiation and Testing Reactor Center and nuclear power-related companies.
Sakanaka, Shogo*; Akemoto, Mitsuo*; Aoto, Tomohiro*; Arakawa, Dai*; Asaoka, Seiji*; Enomoto, Atsushi*; Fukuda, Shigeki*; Furukawa, Kazuro*; Furuya, Takaaki*; Haga, Kaiichi*; et al.
Proceedings of 1st International Particle Accelerator Conference (IPAC '10) (Internet), p.2338 - 2340, 2010/05
Future synchrotron light source using a 5-GeV energy recovery linac (ERL) is under proposal by our Japanese collaboration team, and we are conducting R&D efforts for that. We are developing high-brightness DC photocathode guns, two types of cryomodules for both injector and main superconducting (SC) linacs, and 1.3 GHz high CW-power RF sources. We are also constructing the Compact ERL (cERL) for demonstrating the recirculation of low-emittance, high-current beams using above-mentioned critical technologies.
Sakai, Kenji; Oi, Motoki; Kai, Tetsuya; Watanabe, Akihiko; Nakatani, Takeshi; Higemoto, Wataru; Shimomura, Koichiro*; Kinoshita, Hidetaka; Kaminaga, Masanori
JAEA-Technology 2009-042, 44 Pages, 2009/08
JAEA-Technology-2009-042.pdf:35.33MB
A general control system for the Materials and Life Science Experimental Facility (MLF-GCS) at J-PARC has an advanced and independent system for control of the mercury target, including a large amount of mercury, three moderators with supercritical hydrogen, and cooling systems with radioactive water. Although the MLF-GCS is an independent system, it works closely with the accelerator and other facility control systems within J-PARC. The MLF have succeeded in the first proton beam injection and neutron beam generation in May 2008, and succeeded the muon beams generation in September 2008. The design and construction of the MLF-GCS has finished before the first proton beam injection. It has been operated stably and efficiently in the off- and on- beam commissioning. This paper reports on the design, construction and operation of the MLF-GCS.
Koshizuka, Seiichi*; Liu, J.*; Morita, Koji*; Arima, Tatsumi*; Zhang, S.*; Tobita, Yoshiharu; Yamano, Hidemasa; Ito, Takahiro*; Naito, Masanori*; Shirakawa, Noriyuki*; et al.
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), 1 Pages, 2009/05
A computer code, named COMPASS, is developed for multi-physics analysis of core disruptive accidents of sodium-cooled fast reactors (SFRs). A meshless method, called MPS method, is employed since complex thermal-hydraulics and structural problems with various phase change processes have to be analyzed. Verification for separeted basic processes and validation for practical phenomena are carried out. COMPASS is also expected to investigate molten fuel discharge to avoid re-criticality in large size SFR cores. Both MOX and metal fuels are considered. Eutectic reactions between the metal fuel and the cladding material are investigated by phase diagram calculation, classical and first-principles molecular dynamics. Basic studies relevant to the numerical methods support the code development of COMPASS. Parallel processing is implemented by OpenMP to treat large-scale problems. A visualization tool is also prepared by using AVS.
Oi, Motoki; Kai, Tetsuya; Kinoshita, Hidetaka; Sakai, Kenji; Kaminaga, Masanori; Futakawa, Masatoshi
Nuclear Instruments and Methods in Physics Research A, 600(1), p.120 - 122, 2009/02
Times Cited Count：1 Percentile：12.35(Instruments & Instrumentation)The Materials and Life Science Experimental Facility (MLF) of J-PARC has the general control system (MLF-GCS) that controls all subsystems of the MLFAccording to classifying into each function, the MLF-GCS consists of three layers of a PLC (programmable logic controller) link layer, server layer and external network layer. The PLC link layer is an inner layer and core part of the MLF-GCS. The server layer acquires various data from the inner and outer layer. The server systems also protect the core part of the MLF-GCS from network troubles of external LANs by mediating between the inner and outer layer. The server systems play an important role for realizing advanced and independent control in the MLF. A modeling and construction of the server systems have been almost finished, and an improvement and optimization of them are now in progress. This paper gives an overview of the server systems for the MLF-GCS and reports on their development status.
