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Yamaguchi, Yoshihito; Mano, Akihiro; Li, Y.
Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 10 Pages, 2024/03
The steam generator (SG) is an important component of a pressurized water reactor. In addition, local wall-thinning has been reported in SG tubes. The burst differential pressure, considering both the internal and external pressures from the primary and secondary coolant systems, should be predicted for the failure probability evaluation or structural integrity assessment of SG tubes. In this study, based on the results of burst tests performed in Japan and the United States, we improved the existing burst pressure estimation method for SG tubes with wall-thinning. In addition, as an example of the utilization of the improved burst pressure estimation method, the conditional failure probabilities for SG tubes with local wall-thinning, which is necessary for probabilistic risk assessment and risk-informed decision making, are calculated considering the dimensions of the wall-thinning.
Yamaguchi, Yoshihito; Mano, Akihiro; Katsuyama, Jinya; Masaki, Koichi*; Miyamoto, Yuhei*; Li, Y.
JAEA-Data/Code 2020-021, 176 Pages, 2021/02
JAEA-Data-Code-2020-021.pdf:5.26MB
In Japan Atomic Energy Agency, as a part of researches on the structural integrity assessment and seismic safety assessment of aged components in nuclear power plants, a probabilistic fracture mechanics (PFM) analysis code PASCAL-SP (PFM Analysis of Structural Components in Aging LWR - Stress Corrosion Cracking at Welded Joints of Piping) has been developed to evaluate failure probability of piping. The initial version was released in 2010, and after that, the evaluation targets have been expanded and analysis functions have been improved based on the state-of-the art technology. Now, it is released as Ver. 2.0. In the latest version, primary water stress corrosion cracking in the environment of Pressurized Water Reactor, nickel based alloy stress corrosion cracking in the environment of Boiling Water Reactor, and thermal embrittlement can be taken into account as target age-related degradation. Also, many analysis functions have been improved such as incorporations of the latest stress intensity factor solutions and uncertainty evaluation model of weld residual stress. Moreover, seismic fragility evaluation function has been developed by introducing evaluation methods including crack growth analysis model considering excessive cyclic loading due to large earthquake. Furthermore, confidence level evaluation function has been incorporated by considering the epistemic and aleatory uncertainties related to influence parameters in the probabilistic evaluation. This report provides the user's manual and analysis methodology of PASCAL-SP Ver. 2.0.
Katsuyama, Jinya; Uno, Shumpei*; Watanabe, Tadashi*; Li, Y.
Frontiers of Mechanical Engineering, 13(4), p.563 - 570, 2018/12
Times Cited Count：2 Percentile：12.21(Engineering, Mechanical)For the structural integrity assessments on reactor pressure vessels (RPVs) under pressurized thermal shock (PTS) events, thermal hydraulic (TH) behavior of coolant water is one of the most important influence factors. Configuration of plant equipment and their dimensions, and operator action have large influences on TH behavior. In this study, to investigate the influence of operator action on TH behavior during a PTS event, we developed an analysis model for a typical Japanese plant, and performed TH and structural analyses. Two different operator action times were assumed based on the Japanese and US' rules. From the analysis results, it was clarified that differences in operator action times have a significant effect on TH behavior and loading conditions, that is, following the Japanese rule may lead to lower stresses compared to that when following the US rule because earlier operator action caused lower pressure in the RPV.
Nishimura, Akihiko; Takenaka, Yusuke*; Furusawa, Akinori; Torimoto, Kazuhiro; Ueda, Masashi; Fukuda, Naoaki*; Hirao, Kazuyuki*
E-Journal of Advanced Maintenance (Internet), 9(2), p.52 - 59, 2017/08
no abstracts in English
Nishimura, Akihiko; Terada, Takaya; Takenaka, Yusuke*; Furuyama, Takehiro*; Shimomura, Takuya
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 6 Pages, 2014/07
Since 2007, JAEA has been developing laser based technologies of structural health monitoring. The FBG sensor made by femtosecond laser processing will be the best candidate. To make the best use of the heat resistant characteristic, the FBG sensor was embedded in metal mold by laser cladding. A groove was processed to the surface of a SUS metal plate. We used a QCW laser to weld a filler wire on the plate. A series of weld beads perfectly formed a sealing clad on the groove. Though the FBG sensor was buried tightly, no degradation on the reflection spectrum was detected after the processing. The FBG sensor could detect the vibration of the plate caused by impact shocks and audio vibration. The reflection peak of the FBG sensor under laser cladding condition was shifted to be 6 nm. We demonstrated that the corresponded temperature derive from the reflection peak shift reached 600 degrees in heat shock experiments. The installation procedure of a FBG sensor using a portable laser cladding machine was described.
Tsuji, Nobumasa*; Shibata, Taiju; Sumita, Junya; Ishihara, Masahiro; Iyoku, Tatsuo
FAPIG, (169), p.13 - 17, 2005/03
no abstracts in English
Kurihara, Ryoichi
JAERI-Tech 2004-052, 39 Pages, 2004/07
The problems in the thermal structural design of the plasma facing component such as the blanket first wall and the divertor plate which receives very high heat flux were examined in the design of the fusion power reactors. Compact high fusion power reactor must give high heat flux and high-speed neutron flux from the plasma to the first wall and the divertor plate. In this environmental situation, the micro cracks should be generated in material of the first wall. Structural integrity of the first wall would be very low during the operation of the reactor, if those micro-cracks grow in a crack having significant size by the fatigue or the creep. The crack penetration in the first wall can be a factor which threatens the safety of the fusion power reactor. This paper summarizes the problems on the structural integrity in the first wall made of the SiC/SiC composite material or the ferritic steel.
Onizawa, Kunio; Tsutsumi, Hideaki*; Suzuki, Masahide; Shibata, Katsuyuki; Ueno, Fumiyoshi; Kaji, Yoshiyuki; Tsukada, Takashi; Nakajima, Hajime*
JAERI-Tech 2003-073, 125 Pages, 2003/08
JAERI-Tech-2003-073.pdf:11.62MB
Concerning the cracks due to stress corrosion cracking (SCC) observed on the core shrouds of BWRs, a study was conducted on structural integrity evaluation based on crack growth analysis. The cracks investigated were those observed on the regions of lower ring and support ring of the core shroud at Kashiwazaki-Kariwa Nuclear Power Station (NPS) Unit-3, and that on the middle shell region of the core shroud at Fukushima Daiichi NPS Unit-4 of Tokyo Electric Power Company. It was confirmed through data analysis of past SCC growth rate experiments applicable to the condition of the ring regions that the SCC growth rate prescribed in the JSME rule was conservative. The analysis on the core shroud rigidity with a crack indicated that the rigidity reduction was small enough not to affect the dynamic seismic response for the regions studied. Through the comparison of the required area in a cracked section or the allowable crack length, and crack growth analysis results, it was confirmed that the integrity of the core shrouds would be maintained even 4 effective full power years later.
Kogawa, Hiroyuki; Ishikura, Shuichi*; Haga, Katsuhiro; Kinoshita, Hidetaka; Kaminaga, Masanori; Hino, Ryutaro
Proceedings of ICANS-XVI, Volume 3, p.1295 - 1304, 2003/07
no abstracts in English
Li, Y.*; Kato, Daisuke*; Shibata, Katsuyuki; Onizawa, Kunio
Nihon Kikai Gakkai Rombunshu, A, 69(678), p.239 - 245, 2003/02
no abstracts in English
Kaji, Yoshiyuki; Gu, W.*; Ishihara, Masahiro; Arai, Taketoshi; Nakamura, Hitoshi*
Nuclear Engineering and Design, 206(1), p.1 - 12, 2001/05
Times Cited Count：9 Percentile：56.08(Nuclear Science & Technology)no abstracts in English
Shibata, Katsuyuki; Kato, Daisuke*; Li, Y.*
Nihon Genshiryoku Gakkai-Shi, 43(4), p.387 - 396, 2001/04
Times Cited Count：2 Percentile：19.66(Nuclear Science & Technology)no abstracts in English
Shibata, Katsuyuki; Onizawa, Kunio; Kato, Daisuke*; Li, Y.*
Nihon Kikai Gakkai 2001-Nendo Nenji Taikai Koen Rombunshu, p.389 - 390, 2001/00
no abstracts in English
Kato, Daisuke*; Li, Y.*; Shibata, Katsuyuki; Onizawa, Kunio
Nihon Kikai Gakkai 2001-Nendo Nenji Taikai Koen Rombunshu, p.391 - 392, 2001/00
no abstracts in English
Onizawa, Kunio; Suzuki, Masahide
Nihon Kikai Gakkai Heisei-12-Nendo Zairyo Rikigaku Bumon Koenkai Koen Rombunshu, p.431 - 432, 2000/10
no abstracts in English
Onizawa, Kunio; Tobita, Toru; Suzuki, Masahide
Effects of Radiation on Materials (ASTM STP 1366), p.204 - 219, 2000/00
no abstracts in English
Kaji, Yoshiyuki; Gu, W.*; Ishihara, Masahiro; Arai, Taketoshi; Nakamura, Hitoshi*
Transactions of 15th Int. Conf. on Structural Mechanics in Reactor Technol. (SMiRT-15), 2, p.133 - 139, 1999/00
no abstracts in English
; 1, Inspection results at dismantlementFujisaki, Katsuo; Inagaki, Yoshiyuki; ; ; ; ; Sekita, Kenji; Morisaki, Norihiro; *; Iwatsuki, Jin*; et al.
JAERI-Tech 97-053, 57 Pages, 1997/10
no abstracts in English
