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Ando, Masanori; Hirose, Yuichi*; Takano, Masahito*
Journal of Pressure Vessel Technology, 143(6), p.061505_1 - 061505_9, 2021/12
Times Cited Count：0 Percentile：0(Engineering, Mechanical)This study compares and assesses the different fatigue and creep-fatigue life evaluation methods by performing tests of perforated plate made of Mod.9Cr-1Mo steel. Multi-perforated plate was subjected to mechanical cyclic loading at 550
C, and crack initiation and propagation on the surfaces of the holes were observed. A series of finite element analyses were carried out to predict the number of cycles to failure by the several failure life evaluation methods, and these predictions were then compared with the test results. Several types of evaluation methods that use the elastic FEA were applied.
Takano, Kazuya; Sakamoto, Yoshihiko; Morohoshi, Kyoichi*; Okazaki, Hitoshi*; Gima, Hiromichi*; Teramae, Takuma*; Ikarimoto, Iwao*; Botte, F.*; Dirat, J.-F.*; Dechelette, F.*
Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 8 Pages, 2019/05
ASTRID has the objective to integrate innovative options in order to prepare the 4th generation reactors. In ASTRID, large number of tubes are installed above each fuel subassembly to monitor the core. These instrumentations such as thermocouples (TC) and Failed Fuel Detection and Location (FFDL) systems are integrated into Above Core Structure (ACS) with various sizes tubes. In the present study, the routing study for TC tubes and FFDL tubes was performed with 3D modeling and mock-up experiment of the ACS designed for ASTRID with 1500 MW thermal power in order to clarify the integration process and secure the design hypotheses. Although some problems on fabricability were found in the mock-up experiment, the possible solutions were proposed. The present study gives manufacturing feedback to design team and will contribute to increase the knowledge for ACS design and fabricability.
Li, Y.; Katsumata, Genshichiro*; Masaki, Koichi*; Hayashi, Shotaro*; Itabashi, Yu*; Nagai, Masaki*; Suzuki, Masahide*; Kanto, Yasuhiro*
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 10 Pages, 2017/07
In Japan, a PFM analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed by the Japan Atomic Energy Agency to evaluate the through-wall cracking frequencies of Japanese reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock transients. In this study, as a part of the verification activities, a working group was established in Japan, with seven organizations from industry, universities and institutes voluntarily participating as members. The source program of PASCAL was released to the members of the working group. Through one year activities, the applicability of PASCAL for structural integrity assessments of domestic RPVs was confirmed with great confidence. This paper presents the details of the verification activities of the working group including the verification plan, approaches and results.
Ohgama, Kazuya; Ikeda, Kazumi*; Ishikawa, Makoto; Kan, Taro*; Maruyama, Shuhei; Yokoyama, Kenji; Sugino, Kazuteru; Nagaya, Yasunobu; Oki, Shigeo
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04
Li, Y.; Hayashi, Shotaro*; Itabashi, Yu*; Nagai, Masaki*; Kanto, Yasuhiro*; Suzuki, Masahide*; Masaki, Koichi*
JAEA-Review 2017-005, 80 Pages, 2017/03
JAEA-Review-2017-005.pdf:16.85MB
For the improvement of the structural integrity assessment methodology on reactor pressure vessels (RPVs), the probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed and improved in JAEA based on latest knowledge. The PASCAL code evaluates the failure probabilities and frequencies of Japanese RPVs under transient events such as pressurized thermal shock considering neutron irradiation embrittlement. In order to confirm the reliability of the PASCAL as a domestic standard code and to promote the application of PFM on the domestic structural integrity assessments of RPVs, it is important to verify the probabilistic variables, functions and models incorporated in the PASCAL and summarize the verification processes and results as a document. On the basis of these backgrounds, we established a working group, composed of experts on this field besides the developers, on the verification of the PASCAL3 which is a PFM analysis module of PASCAL, and the source program of PASCAL3 was released to the members of working group. Through one year activities, the applicability of PASCAL in structural integrity assessments of domestic RPVs was confirmed with great confidence. This report summarizes the activities of the working group on the verification of PASCAL in FY2015.
Takaya, Shigeru; Kamishima, Yoshio*; Machida, Hideo*; Watanabe, Daigo*; Asayama, Tai
Nuclear Engineering and Design, 305, p.270 - 276, 2016/08
Times Cited Count：3 Percentile：28.28(Nuclear Science & Technology)In our previous study, we proposed a new process for determining the in-service inspection (ISI) requirements using the System Based Code concept. The proposed process consists of two complementary evaluations, one focusing on structural integrity and the other on plant safety. In this study, the ISI requirements for a reactor guard vessel (RGV) and core support structure (CSS) of a prototype sodium-cooled fast breeder reactor in Japan (Monju) were investigated using the proposed process. It was shown that both components had sufficient reliability even assuming unrealistic severe conditions. The failure occurrences of these components were practically eliminated. Hence, it was concluded that no ISI requirements were needed for these components. The proposed process is expected to contribute to the realization of effective and rational ISI by properly taking into account plant-specific features.
Takeda, Takeshi; Onuki, Akira*; Kanamori, Daisuke*; Otsu, Iwao
Science and Technology of Nuclear Installations, 2016, p.7481793_1 - 7481793_15, 2016/00
Times Cited Count：1 Percentile：10.71(Nuclear Science & Technology)Takaya, Shigeru; Watanabe, Daigo*; Yokoi, Shinobu*; Kamishima, Yoshio*; Kurisaka, Kenichi; Asayama, Tai
Journal of Pressure Vessel Technology, 137(5), p.051802_1 - 051802_7, 2015/10
Times Cited Count：2 Percentile：11.54(Engineering, Mechanical)The minimum wall thickness required to prevent seismic buckling of a reactor vessel in a fast reactor is derived using the System Based Code (SBC) concept. One of the key features of SBC concept is margin optimization; to implement this concept, the reliability design method is employed, and the target reliability for seismic buckling of the reactor vessel is derived from nuclear plant safety goals. Input data for reliability evaluation, such as distribution type, mean value, and standard deviation of random variables, are also prepared. Seismic hazard is considered to evaluate uncertainty of seismic load. Minimum wall thickness required to achieve the target reliability is evaluated, and is found to be less than that determined from a conventional deterministic design method. Furthermore, the influence of each random variable on the evaluation is investigated, and it is found that the seismic load has a significant impact.
Takaya, Shigeru; Kamishima, Yoshio*; Machida, Hideo*; Watanabe, Daigo*; Asayama, Tai
Transactions of the 23rd International Conference on Structural Mechanics in Reactor Technology (SMiRT-23) (USB Flash Drive), 10 Pages, 2015/08
In our previous study, a new process for determination of in-service inspection (ISI) requirements was proposed on the basis of the System Based Code concept. The proposed process consists of two complementary evaluations, one focusing on structural integrity and the other on plant safety. In this study, ISI requirements for a reactor guard vessel and a core support structure of the prototype sodium-cooled fast breeder reactor in Japan, Monju, were investigated according to the proposed process. The proposed process is expected to contribute to realize effective and rational ISI by properly taking into account plant-specific features.
Takeda, Takeshi; Onuki, Akira*; Nishi, Hiroaki*
Journal of Energy and Power Engineering, 9(5), p.426 - 442, 2015/05
Takaya, Shigeru; Asayama, Tai; Kamishima, Yoshio*; Machida, Hideo*; Watanabe, Daigo*; Nakai, Satoru; Morishita, Masaki
Journal of Nuclear Engineering and Radiation Science, 1(1), p.011004_1 - 011004_9, 2015/01
A new process for determination of inservice inspection (ISI) requirements was proposed based on the System Based Code concept to realize effective and rational ISI by properly taking into account plant specific features. The proposed process consists of two complementary evaluations, one focusing on structural integrity and the other one on detectability of defects before they would grow to an unacceptable size in light of plant safety. If defect detection was not feasible, structural integrity evaluation would be required under sufficiently conservative hypothesis. The applicability of the proposed process was illustrated through an application to the existing prototype fast breeder reactor, Monju.
Takeda, Takeshi; Onuki, Akira*; Nishi, Hiroaki*
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 13 Pages, 2014/12
Takaya, Shigeru; Watanabe, Daigo*; Yokoi, Shinobu*; Kamishima, Yoshio*; Kurisaka, Kenichi; Asayama, Tai
Proceedings of 2013 ASME Pressure Vessels and Piping Conference (PVP 2013) (DVD-ROM), 6 Pages, 2013/07
In this paper, minimum wall thickness requirement of reactor vessel of fast reactor for seismic buckling is discussed on the basis of the System Based Code (SBC) concept. One of key concepts of SBC is the margin optimization. To implement this concept, reliability design method is employed, and the target reliability for seismic buckling of reactor vessel is derived from nuclear plant safety goals. Input data for reliability evaluation such as distribution type, mean value and standard deviation of random variable are prepared. Seismic hazard is considered to evaluate uncertainty of seismic load. Wall thickness needed to achieve the target reliability is evaluated, and as a result, it is shown that the minimum wall thickness can be reduced from that required by a deterministic design method.
Watanabe, Daigo*; Chuman, Yasuharu*; Asayama, Tai; Takaya, Shigeru; Machida, Hideo*; Kamishima, Yoshio*
Proceedings of 2013 ASME Pressure Vessels and Piping Conference (PVP 2013) (DVD-ROM), 7 Pages, 2013/07
Limit state design was newly developed in order to apply the margin exchange which is one of the innovative concepts of the System Based Code (SBC). It was shown that limit state design method is applicable to plant design instead of current design criteria. In this report, working example of a reactor vessel of a Fast Reactor subject to thermal load is conducted to demonstrate this concept. As the result allowable stress was increased by changing the acceptance criteria from current design criteria to limit state design criteria.
Kato, Atsushi; Hirata, Shingo; Chikazawa, Yoshitaka; Uto, Nariaki; Obata, Hiroyuki*; Kotake, Shoji*; Uzawa, Masayuki*
Proceedings of 2010 International Congress on Advances in Nuclear Power Plants (ICAPP '10) (CD-ROM), p.692 - 699, 2010/06
In the JSFR design, a single rotating plug and an upper inner structure (UIS) with a vertically penetrating slit are proposed, so that the fuel handling machine (FHM) can access any subassembly by horizontal movement of the FHM arm in the slit space. As a result of a full-scale mockup test, excellent performance in normal operation has been shown. In this study, from the viewpoint of achieving reliability of the pantograph type FHM, behavior of the FHM mockup have been investigated under abnormal conditions.
Oba, Ryoji*; Bushnak, A.*; Kojiri, Toshiharu*; Matsuura, Tomonori*; Nagai, Haruyasu
Proceedings of the 3rd International Conference on Water Resources and Arid Environments 2008 and the 1st Arab Water Forum (Internet), 12 Pages, 2008/11
no abstracts in English
Kanagawa, Akihiro*; Iwatsuki, Jin; Tanaka, Nobuyuki; Onuki, Kaoru; Fukui, Hiroshi*; Nishibayashi, Toshiki*
JAEA-Technology 2007-061, 32 Pages, 2007/12
JAEA-Technology-2007-061.pdf:21.03MB
Separation of hydrogen iodide from HIx solution (HI-I
-HO mixture) is one of the technical issues in the development of thermochemical IS process. Application of pervaporation (PV) to the concentration of HIx solution in the IS process pilot test plant was discussed from the viewpoints of process heat mass balance, conceptual design of the apparatus, and the corrosion resistance of the membrane module. Compared with the electro-electrodialysis system, the PV system enables the downsizing of apparatus by using hollow fiber membranes, although it does not improve the thermal efficiency of IS process. Immersion tests of commercially available Nafion hollow fiber membrane module in the HIx solution at 100C indicated the necessity of improving the corrosion resistance of bundle materials.
Ito, Takaya*; Sato, Hiroyuki*; Usui, Yukinori*; Toda, Mikio*
Mitsubishi Juko Giho, 43(4), p.45 - 49, 2006/12
MHI has participated in the FBR development that is a national project from an initial stage as a member of fabricators. Aiming at FBR commercialization until 2050, MHI has actively contributed to "Feasibility Studies on Commercialized Fast Reactor Cycle System" performed mainly by JAEA from 1999. This time, the Sodium-cooled Fast Reactor was chosen at a national level as the main concept. This Sodium-cooled Fast Reactor is proposed by MHI, and for an economy improvement innovation technologies (reduction of the number of the loops, shortening of piping systems, integrated IHX/Pump component and so on) are adopted in this concept. From now on, research and development for FBR commercialization will be accelerated at a national level toward realization of a demonstration reactor until about 2025. MHI also challenges the project as a proposer of this concept.
Toya, Yuichi; Washiya, Tadahiro; Koizumi, Kenji; Morita, Shinichi
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 6 Pages, 2006/07
Japan Atomic Energy Agency (JAEA) has been leading feasibility study on commercialized fast reactor cycle systems in Japan. In this study, we have proposed a new disassembly technology by mechanical disassembly system that consists of a mechanical cutting step and a wrapper tube pulling step. In the mechanical tool system, high durability mechanical cutter cuts the wrapper tube (Slit-Cut (S/C) operation in circle direction), and then the wrapper tube is pulled out and removed from the fuel assembly. Then the fuel pins are cut (Crop-Cut (C/C) operation at entrance nozzle side) and the entrance nozzle is removed. The fuel pins are transported to the shearing machine in next process. The Fundamental tests were carried out with simulated FBR fuel pins and wrapper tube, and cutting performance and wrapper tube pulling performance has been confirmed by engineering scale. As the results, we established the disassembly procedure and the fundamental design of mechanical disassembly system.
U Doppler effect measured in FCA LWR simulating cores; Analysis with JENDL-3.3 library and SRAC system (Joint research)Kawasaki, Kenji*; Ando, Masaki; Okajima, Shigeaki; Fukushima, Masahiro; Nakano, Makoto*; Matsumoto, Hideki*
JAEA-Research 2006-008, 40 Pages, 2006/03
JAEA-Research-2006-008.pdf:3.6MB
Analysis was performed to evaluate prediction accuracy of a neutronics code system for thermal reactor; the SRAC system with the use of the latest nuclear data library JENDL-3.3 for the U Doppler effect measured in the uranium fueled (FCA-XXI-1D2) and MOX fueled (XXII-1 series) cores. The results of the analysis with the diffusion theory showed overestimation by up to +11%. In relatively soft neutron spectra, however, the calculated values agreed with the experimental ones within the experimental errors.
