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Ohshima, Hiroyuki; Asayama, Tai; Furukawa, Tomohiro; Tanaka, Masaaki; Uchibori, Akihiro; Takata, Takashi; Seki, Akiyuki; Enuma, Yasuhiro
Journal of Nuclear Engineering and Radiation Science, 9(2), p.025001_1 - 025001_12, 2023/04
This paper describes the outline and development plan for ARKADIA to transform advanced nuclear reactor design to meet expectations of a safe, economic, and sustainable carbon-free energy source. ARKADIA will realize Artificial Intelligence (AI)-aided integrated numerical analysis to offer the best possible solutions for the design and operation of a nuclear plant, including optimization of safety equipment. State-of-the-art numerical simulation technologies and a knowledge base that stores data and insights from past nuclear reactor development projects and R&D are integrated with AI. In the first phase of development, ARKADIA-Design and ARKADIA-Safety will be constructed individually, with the first target of sodium-cooled reactor. In a subsequent phase, everything will be integrated into a single entity applicable not only to advanced rectors with a variety of concepts, coolants, configurations, and output levels but also to existing light-water reactors.
Yada, Hiroki; Takaya, Shigeru; Enuma, Yasuhiro
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 7 Pages, 2020/08
Amano, Katsunori; Enuma, Yasuhiro; Futagami, Satoshi; Inoue, Tomoyuki*; Watanabe, Sota*
Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 7 Pages, 2016/06
In the framework of GIF, SDC and SDG for the generation IV SFRs have been developed in the circumstance of worldwide deployment of SFRs. JAEA and MFBR have been investigating design study of an advanced loop-type SFR to satisfy SDC in the feasibility study of SDG for SFR. In this study, the ability of the pump/IHX in the advanced loop-type SFR for the safety measures has been evaluated. In addition to the safety measures, maintainability and reparability are taken into account in the advanced loop-type SFR design study. The pump/IHX has been modified to satisfy these requirements. This paper describes the modifications for the ability to withstand a severe earthquake, the reliability of the guard vessel in the primary coolant leak, and the reliability of expansion joints in a sodium-water reaction. The evaluations of thermal transient, structural vibration with pump rotation and wear-out of IHX tubes, that has been adversely effected by the modifications, were described as well.
Wakai, Takashi; Machida, Hideo*; Yoshida, Shinji*; Yanagihara, Seiji*; Suzuki, Ryosuke*; Matsubara, Masaaki*; Enuma, Yasuhiro
Engineering Failure Analysis, 56, p.484 - 500, 2015/10
Times Cited Count：1 Percentile：13.02(Engineering, Mechanical)Enuma, Yasuhiro; Kawasaki, Nobuchika; Orita, Junichi*; Eto, Masao*; Miyagawa, Takayuki*
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
In the frame work of generation IV international forum, safety design criteria and safety design guideline for the generation IV sodium-cooled fast reactors have been developing. JAEA, JAPC, MFBR have been investigating design study for JSFR to satisfy SDC. In addition to the safety measures, maintainability, reparability and manufacturability are taken into account in the JSFR design study. This paper describes the design of main components. Enlargement of the access route for the inspection devices and addition of the access routes were carried out for the reactor structure. The pump-integrated IHX was modified for the primary heat exchanger, which was installed for the decay heat removal in the IHX at the upper plenum, to be removable for improved repair and maintenance. For the steam generator, protective wall tube type design is under investigation as an option with less R&D risks.
Ando, Masanori; Hirose, Yuichi*; Karato, Takanori*; Watanabe, Sota*; Inoue, Osamu*; Kawasaki, Nobuchika; Enuma, Yasuhiro*
Journal of Pressure Vessel Technology, 136(4), p.041406_1 - 041406_10, 2014/08
Times Cited Count：8 Percentile：43.22(Engineering, Mechanical)To compare and assess the creep-fatigue life evaluation methods for stress concentration point, a series of creep-fatigue test was performed with notched specimens made of Mod.9Cr-1Mo steel. Mechanical creep-fatigue tests and thermal creep-fatigue test were performed. A series of Finite Element Analysis was also carried out to predict the number of cycles to failure by the several creep-fatigue life evaluation methods. Then these predictions were compared with the test results. Several types of evaluation methods such are stress redistribution locus (SRL) method, simple elastic follow-up method and the methods described in the JSME FRs code were applied. Through the comparisons, it was appeared that SRL method gave rational conservative prediction of the creep-fatigue life for all conditions tested in this study. The JSME FRs code gave an evaluation over 70 times conservative lives comparing with the test results.
Ando, Masanori; Hasebe, Shinichi; Kobayashi, Sumio; Kasahara, Naoto*; Toyoshi, Akira*; Omae, Takahiro*; Enuma, Yasuhiro*
Nuclear Engineering and Design, 275, p.408 - 421, 2014/08
Times Cited Count：4 Percentile：33.4(Nuclear Science & Technology)To clarify the failure mode of a semispherical tubesheet structure originally designed for SG in the JSFR, a cyclic thermal loading test was performed using a tubesheet model test structure. The tubesheet model made of Mod.9Cr-1Mo steel was subjected to 1,873 cycles of severe thermal transient loads using a large-scale sodium loop, in which sodium heated to 600
C and 250C was flowed repeatedly with periods for each transient of 2 and 1 h, respectively. After the test, the test model was inspected by PT. Then, observation using a SEM and hardness testing were performed. A thermal-hydraulic analysis was also performed to validate the measured temperature history during the thermal transient. Through these examinations and evaluation with thermal-hydraulic analysis, the manner of failure in the tubesheet under cyclic thermal loading is discussed.
Ando, Masanori; Hasebe, Shinichi; Kobayashi, Sumio; Kasahara, Naoto*; Toyoshi, Akira*; Omae, Takahiro*; Enuma, Yasuhiro*
Nuclear Engineering and Design, 275, p.422 - 432, 2014/08
Times Cited Count：11 Percentile：67.73(Nuclear Science & Technology)In this study, the strength of a tubesheet test model simulating a semispherical tubesheet structure subjected to cyclic thermal transients was evaluated using the finite element analysis (FEA). A test model made of Mod.9Cr-1Mo steel was subjected to 1,873 cycles of severe thermal transient loading using a large-scale sodium loop, in which elevated-temperature sodium at 600C and 250C was flowed repeatedly and kept at the final temperature for 2 and 1 h, respectively. Heat transfer analysis and stress analysis were performed using the sodium temperature data measured during the test. Then, the elastic and inelastic stress analysis results were used to investigate the failure mechanism by creep-fatigue damage and evaluate the failure strength. The evaluation based on the results of inelastic analysis estimated the number of cycles to failure within a factor of 3.
Ando, Masanori; Hasebe, Shinichi; Kobayashi, Sumio; Kasahara, Naoto*; Toyoshi, Akira*; Omae, Takahiro*; Enuma, Yasuhiro*
Nihon Kikai Gakkai M&M 2013 Zairyo Rikigaku Kanfarensu Koen Rombunshu (CD-ROM), p.OS1510_1 - OS1510_3, 2013/10
To validate the failure mode and assess creep-fatigue damage evaluation, a thick cylinder test model made of Mod.9Cr-1Mo steel was subjected to 1,873 cycles of accelerated thermal transient loading using a large-scale sodium loop through which liquid sodium at 600C and 250C flowed repeatedly, with the period of each transient being 2 h and 1 h, respectively. After completion of the test, liquid penetrant testing, a surface observation and hardness testing were performed to characterize failure mode. Based on the finite element analysis, creep-fatigue life was evaluated by applying the JSME FRs code. The failure cycles evaluated by rules described in the JSME FRs code was shown to have a safety margin of greater than 300 times for this system.
Wakai, Takashi; Machida, Hideo*; Arakawa, Manabu*; Yoshida, Shinji*; Enuma, Yasuhiro
Journal of Pressure Vessel Technology, 135(1), p.011401_1 - 011401_9, 2013/02
Times Cited Count：2 Percentile：14.61(Engineering, Mechanical)This was carried out to establish crack opening displacement (COD) evaluation methods used in Leak-Before-Break (LBB) assessment of sodium pipes of the Japan Sodium cooled Fast Reactor (JSFR). A COD assessment method applicable to thin walled large diameter pipe made of modified 9Cr-1Mo steel was proposed in this study. In this method, COD was calculated by classifying the components of COD; elastic, local plastic and fully plastic. In addition, the verification of this method was performed by comparing with the results of a series of four-point bending tests at elevated temperature using thin wall modified 9Cr-1Mo steel pipe containing a circumferential through wall crack. The rational leak rate evaluation method in LBB assessment was proposed.
Ando, Masanori; Hasebe, Shinichi; Kobayashi, Sumio; Kasahara, Naoto*; Toyoshi, Akira*; Omae, Takahiro*; Enuma, Yasuhiro*
Nuclear Engineering and Design, 255, p.296 - 309, 2013/02
Times Cited Count：19 Percentile：83.5(Nuclear Science & Technology)To verify the failure mode and assess creep-fatigue damage, a thick cylinder test model made of Mod.9Cr-1Mo steel was subjected to 1,873 cycles of accelerated thermal transient loading using a large-scale sodium loop through which liquid sodium at 600 C and 250 C flowed repeatedly, with the period of each transient being 2 h and 1 h, respectively. After completion of the test, the test model was inspected using liquid penetrant testing. Observations using a scanning electron microscope and hardness testing were then performed to characterize creep-fatigue damage in the structural model subjected to cyclic thermal transient loading in a sodium environment. Finite element analysis were performed to evaluate the relationship between creep-fatigue damage and the observed crack conditions.
Ando, Masanori; Hirose, Yuichi*; Date, Shingo*; Watanabe, Sota*; Enuma, Yasuhiro*; Kawasaki, Nobuchika
Journal of Pressure Vessel Technology, 134(6), p.061403_1 - 061403_12, 2012/12
Times Cited Count：5 Percentile：29.84(Engineering, Mechanical)To verify the methods of estimating strain range for discontinues structures, low cycle fatigue tests were carried out with notched specimens. All the specimens were made of Mod.9Cr-1Mo. Displacement control fatigue tests and thermal fatigue tests were performed by ordinary uni-axial push pull test machine and equipment generating the thermal gradient in the notched plate by induction heating. Elastic and inelastic finite element analysis were also performed to estimate strain range for predicting fatigue life. Then these predictions were compared with the test results. Several methods such as stress reduction locus (SRL) method, simple elastic follow-up (SEF) method, Neuber's law and the procedures employed by elevated temperature design codes were applied. Through these comparisons, the applicability and conservativeness of these strain range estimation methods are discussed.
Ando, Masanori; Isobe, Nobuhiro*; Kikuchi, Koichi*; Enuma, Yasuhiro*
Nuclear Engineering and Design, 247, p.66 - 75, 2012/06
Times Cited Count：7 Percentile：50.57(Nuclear Science & Technology)The effect of ratcheting deformation on fatigue and creep-fatigue life in Mod.9Cr-1Mo steel was investigated. Uniaxial fatigue and creep-fatigue testing with superimposed strain were performed to evaluate the effect of ratcheting deformation on the failure cycle. In the fatigue tests with superimposed strain at 550C, slight reductions of failure lives were observed. All of the numbers of cycles to failure in the fatigue tests with superimposed strain were within a factor of 1.5 of that of the fatigue test without superimposed strain at 550C. The apparent relationship between failure cycles and testing parameters was not observed. It was assumed that suppression of mean stress generation by cyclic softening reduces the effect of ratcheting strain. In the creep-fatigue tests with superimposed strain, test results indicated that the accumulated stain was negligible.
Chikazawa, Yoshitaka; Enuma, Yasuhiro; Kisohara, Naoyuki; Yamano, Hidemasa; Kubo, Shigenobu; Hayafune, Hiroki; Sagawa, Hiroshi*; Okamura, Shigeki*; Shimakawa, Yoshio*
Proceedings of 2012 International Congress on Advances in Nuclear Power Plants (ICAPP '12) (CD-ROM), p.677 - 686, 2012/06
Evaluation of Earthquake and Tsunami on JSFR has been analyzed. For seismic design, safety components are confirmed to maintain their functions even against recent strong Earthquakes. As for Tsunami, some parts of reactor building might be submerged including component cooling water system whose final heat sink is sea water. However, in the JSFR design, safety grade components are independent from component cooling water system (CCWS). The JSFR emergency power supply adopts a gas turbine system with air cooling, since JSFR does not basically require quick start-up of the emergency power supply thanks to the natural convection DHRS. Even in case of long station blackout, the DHRS could be activated by emergency batteries or manually and be operated continuously by natural convection.
Ando, Masanori; Isobe, Nobuhiro*; Date, Shingo*; Kikuchi, Koichi*; Enuma, Yasuhiro*
Zairyo, 61(4), p.377 - 384, 2012/04
The effect of ratcheting deformation and pre-strain on fatigue and creep-fatigue life in Mod.9Cr-1Mo steel was investigated. Uni-axial fatigue and creep-fatigue tests with pre-strain and progressive strain were performed to evaluate the effect of pre-strain and ratcheting strain on the failure cycle. In the fatigue tests with pre-strain, failure lives were not declined. In the fatigue tests with progressive strain, slight reductions of failure lives were observed, however, they were within a factor of 1.5 of the failure life in normal fatigue test. In the creep-fatigue tests with progressive strain, the same conclusion of the fatigue tests was obtained. In both kinds of tests, maximum mean stresses during the tests were insignificant and/or generated in early cycle in the tests, and this character is considered as a reason of that the effect of ratcheting deformation on the fatigue and creep-fatigue lives are insignificant.
Kurome, Kazuya*; Kawamura, Masaya*; Enuma, Yasuhiro*; Tsujita, Yoshihiro*; Sato, Mitsuru*; Futagami, Satoshi; Hayafune, Hiroki
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 11 Pages, 2012/00
Ando, Masanori; Hirose, Yuichi*; Karato, Takanori*; Watanabe, Sota*; Inoue, Osamu*; Kawasaki, Nobuchika; Enuma, Yasuhiro*
Proceedings of 2011 ASME Pressure Vessels and Piping Conference (PVP 2011) (CD-ROM), 11 Pages, 2011/07
Ando, Masanori; Tezuka, Taiji*; Nakamura, Toshio*; Okawa, Tomohiro*; Enuma, Yasuhiro*; Kawasaki, Nobuchika; Tsukimori, Kazuyuki
Proceedings of 2011 ASME Pressure Vessels and Piping Conference (PVP 2011) (CD-ROM), 9 Pages, 2011/07
Ando, Masanori; Tezuka, Taiji*; Nakamura, Toshio*; Okawa, Tomohiro*; Enuma, Yasuhiro*; Kawasaki, Nobuchika; Tsukimori, Kazuyuki
Proceedings of 2011 ASME Pressure Vessels and Piping Conference (PVP 2011) (CD-ROM), 8 Pages, 2011/07
Nagae, Yuji; Enuma, Yasuhiro*; Oshikiri, Masato*; Date, Shingo*; Toyoshi, Akira*; Aizawa, Taiki*; Koyama, Yoichi*; Yanagisawa, Yusuke*
Tainetsu Kinzoku Zairyo Dai-123-Iinkai Kenkyu Hokoku, 52(2), p.161 - 169, 2011/07
no abstracts in English
