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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.
Nishimura, Akihiko
Dai-62-Kai Koha Senshingu Gijutsu Kenkyukai Koen Rombunshu, p.79 - 86, 2018/12
Thermal Energy Storage (TES) is important to stabilize increasing large amount of fluctuating renewable energy. For safety operation of TES, remote sensing by Fiber Bragg Grating (FBG) sensors is expected. FGB sensors were fabricated using precisely focused picosecond laser pulses. For the best use of heat resistant characteristic, we demonstrated to embed the FBG sensors in metal mold using colloidal silver adhesive. The FBG sensors were tested using a sodium circulation loop in JAEA Tsuruga site. Sodium was circulated with temperature of 500 degree. During emergency cooling, sudden shrinking of the loop was recorded. The application of FBG sensors to advanced remote monitoring for next generation TES plant is proposed.
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
Masaki, Koichi; Miyamoto, Yuhei*; Osakabe, Kazuya*; Uno, Shumpei*; Katsuyama, Jinya; Li, Y.
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 7 Pages, 2017/07
A probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed by Japan Atomic Energy Agency (JAEA). PASCAL can evaluate failure frequencies of Japanese reactor pressure vessels (RPVs) during pressurized thermal shock (PTS) events based on domestic structural integrity assessment models and data of influence factors. In order to improve the engineering applicability of PFM to Japanese RPVs, we have performed verification of the PASCAL. In general, PFM code consists of many functions such as fracture mechanics evaluation functions, probabilistic evaluation functions including random variables sampling modules and probabilistic evaluation models, and so on. The verification of PFM code is basically difficult because it is impossible to confirm such functions through the comparison with experiments. When a PFM code is applied for evaluating failure frequencies of RPVs, verification methodology of the code should be clarified and it is important that verification results including the region and process of the verification of the code are indicated. In this paper, our activities of verification for PASCAL are presented. We firstly represent the overview and methodology of verification of PFM code, and then, some verification examples are provided. Through the verification activities, the applicability of PASCAL in structural integrity assessments for Japanese RPVs was confirmed with great confidence.
Yamano, Hidemasa; Xu, Y.*; Sago, Hiromi*; Hirota, Kazuo*; Baba, Takeo*
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1029 - 1038, 2016/04
This study conducted the flow-induced vibration evaluation of the primary hot-leg piping in the demonstration reactor design of advanced loop-type sodium-cooled fast reactor in order to confirm the integrity of the piping. Following the description of the primary hot-leg piping design and a design guideline of the flow-induced vibration evaluation, this paper describes mainly the flow-induced vibration evaluation and thereby the integrity assessment. In the fatigue evaluation for the flow-induced vibration, the pipe stresses considering the stress concentration factor and so on, at representative locations were less than the design fatigue limit. Therefore, this evaluation confirmed the integrity of the primary hot-leg piping in the demonstration reactor.
Suzuki, Ryosuke*; Matsubara, Masaaki*; Yanagihara, Seiji*; Morijiri, Mitsugu*; Omori, Atsushi*; Wakai, Takashi
Procedia Materials Science, 12, p.24 - 29, 2016/00
Times Cited Count：2 Percentile：74.36(Engineering, Mechanical)In this study, the plastic collapse strength of asymmetry multiple circumferential notched stainless steel pipes subjected to combined axial tension and bending is investigated experimentally and is compared with the theoretical plastic collapse strength. In addition, the potential is discussed for the simplification of structural integrity evaluation of multiple cracked piping. The integrity of the asymmetry multiple circumferential notched stainless steel pipes subjected to combined axial tension and bending can be evaluated conservatively using the theoretical plastic collapse strength for the pipe with multiple notches calculated based on the elastic-perfectly plastic model.
-ray irradiated conditionYamamoto, Masahiro; Sato, Tomonori; Komatsu, Atsushi; Nakano, Junichi; Ueno, Fumiyoshi
Proceedings of European Corrosion Congress 2015 (EUROCORR 2015) (USB Flash Drive), 7 Pages, 2015/09
In Fukushima-Daiichi Nuclear Power Station, decommissioning procedures are continuing and it will take more than 30 years. As some structures are made of carbon steel, degradation by corrosion is large problem for structural reliability. To clarify an irradiation effect for corrosion of carbon steel, corrosion test was con-ducted in 
Co -ray irradiated condition. Corrosion test results showed that corrosion rates of -ray irradiated condition increased with -ray dose rates. The oxidant concentrations were also increased with -ray dose rate. From these results, a new estimation method for corrosion rate of carbon steel in water with -ray irradiated condition using radiolysis calculation is introduced and discussed.
Nakano, Takafumi; Sato, Fuminori; Shirozu, Hidetomo; Nakanishi, Ryuji; Fukuda, Kazuhito; Tachibana, Ikuya
Nihon Genshiryoku Gakkai-Shi ATOMO
, 57(1), p.14 - 20, 2015/01
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.
Saito, Shigeru; Tsujimoto, Kazufumi; Kikuchi, Kenji; Kurata, Yuji; Sasa, Toshinobu; Umeno, Makoto*; Nishihara, Kenji; Mizumoto, Motoharu; Ouchi, Nobuo; Takei, Hayanori; et al.
Nuclear Instruments and Methods in Physics Research A, 562(2), p.646 - 649, 2006/06
Times Cited Count：25 Percentile：83.96(Instruments & Instrumentation)JAERI is conducting R&D on the Accelerator Driven System (ADS) to transmute minor actinides (MAs) contained in the high-level radioactive waste under the OMEGA (Options Making Extra Gains from Actinides and fission products) program. The present study discusses the design of the ADS plant and various R&D on the ADS. The reference design of ADS plant in JAERI is the 800 MWth, Pb-Bi eutectic (LBE) cooled, tank-type subcritical reactor loaded with (MA+Pu) nitride fuel. LBE is selected as a spallation target material. In our results of the optimization study on the neutronics of the ADS, we have adopted the maximum multiplication factor (k
) of 0.97. From the results of the thermal-hydraulic analysis around the LBE spallation target, partition wall and flow control nozzle are required to keep the structural integrity around the core and the beam window. Feasibility of beam window was also discussed for transient conditions of proton beam.
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.
Hidaka, Akihide; Suzuki, Masahide
JAERI-Conf 2003-014, 178 Pages, 2003/09
JAERI-Conf-2003-014.pdf:19.17MB
The Workshop on Reactor Safety Research focusing on the integrity of aged components was held at the Tokai Research Establishment on March 17, 2003. The purpose of the Workshop was to obtain useful information to proceed with the reactor safety research in future and to resolve the issues on the integrity evaluation of aged components through the discussions followed by the presentations on the results of the research at JAERI on all the research subjects assigned to JAERI in the Five-Year Program of Safety Research for Nuclear Installations established by the Nuclear Safety Commission, and on those of the studies at JAERI on the integrity of core shrouds of BWR plants. Thirty-eight people from outside JAERI including the press such as Nihon Television Network Corporation and Shin-Ibaraki Shinbun and fifty-seven people from JAERI attended the Workshop. This proceeding compiles all the viewgraphs presented in the workshop, the opinions of participants for forum and the answers, and summary of questionnaire on workshop.
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
Onizawa, Kunio; Suzuki, Masahide
Effects of Radiation on Materials: 20th International Symposium (ASTM STP 1405), p.79 - 96, 2001/07
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
