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Journal Articles

Evaluation method of creep-fatigue life for 316FR weldment

Nagae, Yuji; Yamamoto, Kenji*; Otani, Tomomi*

Proceedings of 2015 ASME Pressure Vessels and Piping Conference (PVP 2015) (Internet), 7 Pages, 2015/07

Journal Articles

Development of 2012 edition of JSME code for design and construction of fast reactors, 6; Design margin assessment for the new materials to the rules

Ando, Masanori; Watanabe, Sota*; Kikuchi, Koichi*; Otani, Tomomi*; Sato, Kenichiro*; Tsukimori, Kazuyuki; Asayama, Tai

Proceedings of 2013 ASME Pressure Vessels and Piping Conference (PVP 2013) (DVD-ROM), 11 Pages, 2013/07

New 2012 edition of JSME code for design and construction of fast reactors (FRs code) was published by Japan society of mechanical engineers (JSME). Main topic of the current JSME FRs code 2012 edition is registration of the two new materials, 316FR and Mod.9Cr-1Mo steel. The design margins for the new materials to the rules for the components and piping serviced at elevated temperature described in the JSME FRs code were assessed. To confirm the design margins, a series of the assessment program for the new materials to the conventional design rules was performed using the evaluation of the experimental data and finite element analysis. Through these assessments, the enough design margins for new materials to the rules were confirmed.

Journal Articles

Creep strength evaluation of welded joint made of modified 9Cr-1Mo steel for Japanese Sodium Cooled Fast Reactor (JSFR)

Wakai, Takashi; Nagae, Yuji; Onizawa, Takashi; Obara, Satoshi; Xu, Y.*; Otani, Tomomi*; Date, Shingo*; Asayama, Tai

Proceedings of 2010 ASME Pressure Vessels and Piping Conference (PVP 2010) (CD-ROM), 9 Pages, 2010/07

Journal Articles

Clarification of strain limits considering the ratcheting fatigue strength of 316FR steel

Isobe, Nobuhiro*; Sukekawa, Masayuki*; Nakayama, Yasunari*; Date, Shingo*; Otani, Tomomi*; Takahashi, Yukio*; Kasahara, Naoto; Shibamoto, Hiroshi*; Nagashima, Hideaki*; Inoue, Kazuhiko*

Nuclear Engineering and Design, 238(2), p.347 - 352, 2008/02

 Times Cited Count:21 Percentile:81.14(Nuclear Science & Technology)

The effect of ratcheting on fatigue strength was investigated in order to rationalize the strain limit as a design criterion of commercialized fast reactor systems. Ratcheting fatigue tests were conducted at 550$$^{circ}$$C. Duration of the ratchet straining was set for a certain number of strain cycles taking the loading condition of fast reactors into account, and the number of cycles for strain accumulation was defined as the ratchet-expired cycle. Fatigue lives decrease as the accumulated strain by ratcheting increases. Fatigue life reduction was negligible when the maximum mean stress was less than 25 MPa, corresponding to an accumulated strain of 2.2 percent. Accumulated strain is limited to 2 percent in the present design guidelines and this strain limit is considered effective to avoid reducing fatigue life by ratcheting. Micro-crack growth behaviors were also investigated in these tests in order to discuss the life reduction mechanisms in ratcheting conditions.

Journal Articles

An Experimental validation of the guideline for inelastic design analysis through structural model tests

Watanabe, Daigo*; Chuman, Yasuharu*; Otani, Tomomi*; Shibamoto, Hiroshi*; Inoue, Kazuhiko*; Kasahara, Naoto

Nuclear Engineering and Design, 238(2), p.389 - 398, 2008/02

 Times Cited Count:5 Percentile:37.69(Nuclear Science & Technology)

In this paper, the inelastic analysis procedures for the improved design of future fast breeder reactors were validated through the structural model tests and the evaluation of the experimental results by the inelastic analyses. First, a thermal fatigue test of a 316FR hollow cylinder with two longitudinal weldments was conducted under the condition of combined constant axial load and cyclic movement of axial temperature distribution, which simulated the loading condition near the free surface of coolant sodium in the main vessel of fast breeder reactors (FBRs). Second, the inelastic analyses were carried out in accordance with the recommended procedure by using the measured results of oscillating temperature distribution. Finally, the results of inelastic analyses were compared with the experimental results and it was validated that the recommended practice gave a conservative result for the deformation and a good estimation of strain range for the fatigue life evaluation.

Journal Articles

Prediction of inelastic stress-strain behavior of spherical tube sheet

Igari, Toshihide*; Takao, Nobuyuki*; Otani, Tomomi*; Shibamoto, Hiroshi; Kasahara, Naoto

Nihon Kikai Gakkai 2006-Nendo Nenji Taikai Koen Rombunshu, Vol.1, p.957 - 958, 2006/09

no abstracts in English

Journal Articles

Measurement of thermal ratcheting strain on the structures by the laser speckle method

Watanabe, Daigo*; Chuman, Yasuharu*; Otani, Tomomi*; Shibamoto, Hiroshi; Inoue, Kazuhiko*; Kasahara, Naoto

Proceedings of 2006 ASME Pressure Vessels and Piping Division Conference (PVP 2006)/International Council on Pressure Vessel Technology (ICPVT-11) (CD-ROM), 7 Pages, 2006/00

Prevention of thermal ratcheting is an important problem for high temperature components of fast breeder reactors that are subjected to cyclic thermal loads. To clarify ratcheting behaviors, structural model tests were planned. Strain measurement is important for understanding the thermal ratcheting phenomenon, however the conventional measurement by strain gauge is difficult at high temperature. Then, Laser speckle strain measurement system using the dual-beam set-up was developed to apply to high temperature structural model tests. This system was applied to the thermal ratcheting tests, which demonstrated the actual operative conditions of reactor vessels. Through comparison with uniaxial test results obtained by extensometers, the laser speckle method was verified. Measured data of structural model tests were utilized to certify the guidelines of inelastic analysis for design, which provide prediction method of strain in components of fast reactor.

Oral presentation

R&D issues in structural design standard of fast reactor, 15; Verification of the guideline of inelastic analysis for design by structural model test

Watanabe, Daigo*; Chuman, Yasuharu*; Otani, Tomomi*; Shibamoto, Hiroshi; Inoue, Kazuhiko*; Kasahara, Naoto

no journal, , 

no abstracts in English

Oral presentation

Design study of double wall straight tube stream generator, 4; Development of design method for 3-dimensional perforated curved tube-plate

Shibamoto, Hiroshi; Kasahara, Naoto; Kisohara, Naoyuki; Igari, Toshihide*; Takao, Nobuyuki*; Otani, Tomomi*

no journal, , 

no abstracts in English

Oral presentation

Refurbishment and restart of JMTR, 11; Technology development for attractive irradiation tests, 5; Re-evaluation of thermal scattering law for neutron dosimetry

Imaizumi, Tomomi; Kadotani, Hiroyuki*; Okumura, Keisuke; Katakura, Junichi; Nagao, Yoshiharu

no journal, , 

no abstracts in English

Oral presentation

Development of elevated temperature materials technology toward the demonstration reactor of JSFR, 1; Overview

Asayama, Tai; Nagae, Yuji; Wakai, Takashi; Futagami, Satoshi; Enuma, Yasuhiro*; Otani, Tomomi*

no journal, , 

Mod.9Cr-1Mo steel is going to be used for the coolant system of JSFR which includes piping, intermediate heat exchangers and steam generators so that shortened piping system and compact component design can be realized. Technologies required for this purpose have been developed.

Oral presentation

Development of the design creep rupture equation on Modified 9Cr-1Mo steel for structural material of FBR

Onizawa, Takashi; Asayama, Tai; Otani, Tomomi*

no journal, , 

no abstracts in English

Oral presentation

Design study of small-sized high temperature gas-cooled reactor (MHR-50/100is), 1; Outline of development of small-sized high temperature gas-cooled reactor concept

Shimizu, Katsusuke*; Minatsuki, Isao*; Otani, Tomomi*; Mizokami, Yoritaka*; Oyama, Sunao*; Tsukamoto, Hiroki*; Kunitomi, Kazuhiko; Tachibana, Yukio; Terada, Atsuhiko

no journal, , 

no abstracts in English

Oral presentation

Basic data acquisitions for development of remote decontamination techniques, 1; Project outline and basic data acquisition plan

Yaita, Yumi*; Sakai, Hitoshi*; Endo, Hiroshi*; Otani, Tomomi*; Takahashi, Yoshiaki*; Oikawa, Kageharu*; Murata, Hirotoshi*; Fukushima, Mineo; Kawatsuma, Shinji

no journal, , 

no abstracts in English

14 (Records 1-14 displayed on this page)
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