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Lu, K.; Katsuyama, Jinya; Li, Y.
Journal of Pressure Vessel Technology, 142(2), p.021208_1 - 021208_11, 2020/04
Times Cited Count：6 Percentile：42.63(Engineering, Mechanical)Nuclear Safety Research Center, Sector of Nuclear Safety Research and Emergency Preparedness
JAEA-Review 2018-022, 201 Pages, 2019/01
JAEA-Review-2018-022.pdf:20.61MB
Nuclear Safety Research Center (NSRC), Sector of Nuclear Safety Research and Emergency Preparedness, Japan Atomic Energy Agency (JAEA) is conducting technical support to nuclear safety regulation and safety research based on the Mid-Long Term Target determined by Japanese government. This report summarizes the research structure of NSRC and the cooperative research activities with domestic and international organizations as well as the nuclear safety research activities and results in the period from JFY 2015 to 2017 on the nine research fields in NSRC; (1) severe accident analysis, (2) radiation risk analysis, (3) safety of nuclear fuels in light water reactors (LWRs), (4) thermohydraulic behavior under severe accident in LWRs, (5) materials degradation and structural integrity, (6) safety of nuclear fuel cycle facilities, (7) safety management on criticality, (8) safety of radioactive waste management, and (9) nuclear safeguards.
Lu, K.; Masaki, Koichi; Katsuyama, Jinya; Li, Y.; Uno, Shumpei*
Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 10 Pages, 2018/07
Katsuyama, Jinya; Yamaguchi, Yoshihito; Li, Y.
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 6 Pages, 2018/07
It has become more important to develop methods for evaluating failure behavior of the nuclear components under severe conditions. We are researching on prediction methods of creep deformation and failure behavior of the nuclear components under elevated temperature conditions based on finite element analysis. In this study, as a part of a project called COSSAL, we performed failure analysis of a large scale pipe experiment to validate our prediction methods based on a creep damage evaluation method. We conclude that creep constitutive law that consider material damage can provide the highest accurate analysis.
Katsuyama, Jinya; Osakabe, Kazuya*; Uno, Shumpei; Li, Y.; Yoshimura, Shinobu*
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 9 Pages, 2017/07
A structural integrity assessment methodology based on probabilistic fracture mechanics (PFM) is a rational methodology in evaluating failure frequency of reactor pressure vessels (RPVs) by considering the probabilistic distributions of various influence factors related to the aged degradation. We have developed a PFM analysis code PASCAL to evaluate the failure frequency of RPVs considering the neutron irradiation embrittlement and pressurized thermal shock (PTS) events. We have also developed a guideline on the structural integrity assessment of RPVs based on PFM to improve the applicability of PFM in Japan and to be able to perform the PFM analyses and evaluate through-wall cracking frequency of RPVs. The technical basis for PFM analysis is provided and the latest knowledge is included in the guideline. In this paper, an overview of the guideline and some typical analysis results obtained based on the guideline and Japanese database related to PTS evaluation are presented.
Sakaba, Nariaki; Tachibana, Yukio; Nakagawa, Shigeaki; Hamamoto, Shimpei
Transactions of 18th International Conference on Structural Mechanics in Reactor Technology (SMiRT-18), p.4499 - 4511, 2005/08
Safety demonstration tests using the HTTR are now underway in order to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to research and development for the VHTR, which is one of the Generation IV reactor candidates. The coolant flow reduction test by running down gas circulators, which is one of the safety demonstration tests, is a simulation test of anticipated transients without scram. During the coolant flow reduction test, temperature of the high-temperature helium components and chemistry in the primary circuit are changed rapidly. This paper describes the structural integrity assessments of helium components, e.g. helium pipes, heat exchangers, during the coolant flow reduction test. From the result of this evaluation, it was found that the helium components were kept their structural integrity during temperature and chemistry transient condition in the coolant flow reduction test from the reactor power at 30%. It was also confirmed by this assessment that the coolant flow reduction test will be able to perform with its enough safety margins from the reactor power at 100%.
Takeda, Nobukazu; Nakahira, Masataka
JAERI-Tech 2004-073, 59 Pages, 2005/01
JAERI-Tech-2004-073.pdf:11.36MB
The ITER toamak is composed of major components such as superconducting magnet and vacuum vessel whose operation temperatures are changed from room temperature to 4 K and room temperature to 200
C, respectively. The gravity support of the tokamak is flexible in order to accept the thermal deformation caused by temperature change. This structural feature causes the complex behaviors of the tokamak during seismic events. Therefore, the mechanical characteristics of the flexible support have to be investigated in detail. The present report describes the global plan of the series of vibration tests to estimate the seismic performance of the ITER tokamak. Although it is ideal that the vibration tests are carried out using a full-scale model, scale models are planned due to the limitation of the test facilities. The test results can be estimated by a scaling law. When the scaling law cannot be applied to some performances, the test is performed using a full-scale model. In addition, the other tests such as vacuum vessel and small-scaled models of the support structure are also planned.
Sakaba, Nariaki; Nakagawa, Shigeaki; Furusawa, Takayuki; Tachibana, Yukio
Transactions of the American Nuclear Society, 91, P. 377, 2004/00
Carbon deposition occurred occasionally in the graphite-moderated gas-cooled reactors was evaluated for the reactor pressure vessel, intermediate heat exchanger, etc. using the measured chemical impurity data for the initial condition of the safety demonstration test. By the evaluated result, it is confirmed that the high-temperature components keep their structural integrity during the any temperature transients in safety demonstration tests.
Kogawa, Hiroyuki; Ishikura, Shuichi*; Futakawa, Masatoshi; Kaminaga, Masanori; Hino, Ryutaro
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 7 Pages, 2003/04
The developments of a MW-class spallation neutron source facility are being carried out under the high-intensity proton accelerator project promoted by JAERI and KEK. A mercury target will be used as a neutron source in the facility. The mercury target vessel made of 316LSS will be subjected to pressure wave generated by rapid thermal expansion of mercury due to a pulsed proton beam injection. The pressure wave will make huge stress on the vessel and will deform the vessel, which would cause cavitation in mercury. To estimate the structural integrity of the mercury target vessel, especially beam window, dynamic stress behaviors due to 1MW-pulsed proton beam injection were analyzed by using FEM code. In the analyses, two types of the target vessels with semi-cylindrical and flat type windows were used as analytical models. As the results, it has been understood that the stress generated in the beam window by the pressure wave could be treated as the secondary stress. Also it was confirmed that the flat type window would be more advantageous from the structural viewpoint than the semi-cylindrical type window.
Hanawa, Satoshi; Kojima, Takao; Sumita, Junya; Tachibana, Yukio
JAERI-Tech 2002-024, 46 Pages, 2002/03
JAERI-Tech-2002-024.pdf:3.29MB
no abstracts in English
Shibata, Katsuyuki; Onizawa, Kunio; Li, Y.*; Kato, Daisuke*
Proceedings of 4th International Workshop on the Integrity of Nuclear Components, p.31 - 41, 2002/00
no abstracts in English
Shibata, Katsuyuki; Kato, Daisuke*; Li, Y.*
Nuclear Engineering and Design, 208(1), p.1 - 13, 2001/08
Times Cited Count：21 Percentile：80.29(Nuclear Science & Technology)no abstracts in English
Shibata, Katsuyuki; Onizawa, Kunio; Li, Y.*; Kato, Daisuke*
JAERI-Data/Code 2001-011, 233 Pages, 2001/03
JAERI-Data-Code-2001-011.pdf:7.42MB
no abstracts in English
Ishihara, Masahiro; Hanawa, Satoshi; Iyoku, Tatsuo; Shiozawa, Shusaku
Tanso, 2001(196), p.39 - 48, 2001/02
no abstracts in English
Onizawa, Kunio; Suzuki, Masahide
ISIJ International, 37(8), p.821 - 828, 1997/08
Times Cited Count：3 Percentile：35.47(Metallurgy & Metallurgical Engineering)no abstracts in English
Shibata, Katsuyuki
Nucl. Eng. Des., 174(1), p.79 - 90, 1997/00
Times Cited Count：1 Percentile：14.48(Nuclear Science & Technology)no abstracts in English
Ishihara, Masahiro; Iyoku, Tatsuo; Futakawa, Masatoshi
Zairyo, 42(472), p.15 - 21, 1993/01
no abstracts in English
Inagaki, Yoshiyuki; Iyoku, Tatsuo; *; ; Shiozawa, Shusaku
JAERI-M 90-020, 70 Pages, 1990/02
no abstracts in English
Wan, T.; Kogawa, Hiroyuki; Iwamoto, Hiroki; Takei, Hayanori; Obayashi, Hironari; Sasa, Toshinobu; Naoe, Takashi; Futakawa, Masatoshi
no journal, ,
Japan Atomic Energy Agency (JAEA) has proposed an accelerator-driven system (ADS) for nuclear transmutation, which possesses a Lead-Bismuth Eutectic (LBE) target/coolant system with an LBE enclosure vessel. To solve basic technical issues related to the ADS design, the construction of Transmutation Experimental Facility (TEF) including the ADS Target Test Facility (TEF-T) has been proposed within the framework of Japan Proton Accelerator Research Complex (J-PARC) project. In the TEF-T, the LBE spallation target will be installed, and be bombarded by pulsed proton beams (400 MeV, 25 Hz, 0.5 ms in pulse duration). The target vessel suffers the static stress caused by design pressure and thermal stress due to heat deposition owing to the spallation reaction. To assess the structure integrity of target vessel, the pressure waves generated in LBE, which caused by rapid temperature rising due to the bombardment of pulsed proton beam, should be considered. On the basis of the results of neutronic and thermal-hydraulic analyses, the effects of pressure waves on the structural integrity of target vessel were studied through the dynamic analyses in the present study. The structural integrity of the LBE enclosure vessel for the TEF-T target was assessed.
