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Katsuyama, Jinya; Katsumata, Genshichiro; Onizawa, Kunio; Watanabe, Tadashi*; Nishiyama, Yutaka
Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 9 Pages, 2014/07
For structural integrity assessment on reactor pressure vessel (RPV) of pressurized water reactor during the pressurized thermal shock (PTS) events, temperature of coolant water and heat transfer coefficient between coolant water and RPV are dominant factors. These values can be determined on the basis of thermal-hydraulics (TH) analysis simulating PTS events. Using these values, structural integrity assessment of RPV is performed by thermal-structural analysis, e.g. loading that affects the crack initiation and propagation is evaluated. In this study, we performed the TH and thermal-structural analyses using three-dimensional model of cold-leg, downcomer and RPV to assess loading conditions during the PTS more accurate. We obtained the loading histories at the reactor core region of RPV where a crack is postulated in the structural integrity assessment. Through the comparison between analysis results and current evaluation method, conservatism of current method will be discussed.
Osakabe, Kazuya*; Masaki, Koichi*; Katsuyama, Jinya; Katsumata, Genshichiro; Onizawa, Kunio
Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 7 Pages, 2014/07
The structural integrity of reactor pressure vessel (RPV) during pressurized thermal shock events is judged to be maintained unless the stress intensity factors at the crack tip is smaller than fracture toughness 
based on deterministic approach in the current Japanese code. Application of a probabilistic fracture mechanics (PFM) analysis method for the structural reliability assessment of RPVs has become attractive recently, because uncertainties of several parameters can be incorporated rationally. According to the PFM analysis method in the U.S., through-wall cracking frequencies (TWCFs) are estimated. In this study, in order to identify the conservatism in the current code, PFM analyses on TWCF have been performed for certain model of RPVs. The result shows that the current assumption in JEAC 4206-2007 is conservative as compared with realistic conditions. Effects of variation of PTS transients on crack initiation frequency and TWCF have been also discussed.
under various selections of temperature conditionsYamamoto, Masato*; Kimura, Akihiko*; Onizawa, Kunio; Yoshimoto, Kentaro*; Ogawa, Takuya*; Mabuchi, Yasuhiro*; Viehrig, H.-W.*; Miura, Naoki*; Soneda, Naoki*
Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 7 Pages, 2014/07
The Master Curve (MC) approach for the fracture toughness reference temperature To is expected to be a powerful tool to ensure the reliability of long-term used RPV steels. In order to get sufficient number of data for the MC approach related to the present surveillance program for RPVs, the use of miniature specimens is important. The test technique for the miniature specimens (Mini-CT) of 4 mm thick had been verified the basic applicability of MC approach by means of Mini-CT for the determination of fracture toughness of typical Japanese RPV steels. A round robin (RR) program was organized to assure the robustness of the technique. As the third step of RR program, blinded tests were carried out. Precise material information was not provided to the participants. From the results obtained, the scatter range in was within the acceptable scatter range specified in the testing standard. The selection of testing temperature seems to give limited effect like that in larger specimens.
Li, Y.; Osakabe, Kazuya*; Katsumata, Genshichiro; Katsuyama, Jinya; Onizawa, Kunio; Yoshimura, Shinobu*
Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 8 Pages, 2014/07
Multiple cracks in the same welded joints have been detected in piping systems of nuclear power plants. Therefore, structural integrity assessments considering multiple cracks and crack initiation in aged piping have been important. Probabilistic fracture mechanics (PFM) is a rational methodology in structural integrity assessment of aged piping in nuclear power plants. Two PFM codes, PASCAL-SP and PRAISE-JNES, have been improved or developed in Japan for the structural integrity assessment considering the age related degradation mechanisms of pipes. In this paper, a benchmark analysis was conducted considering multiple cracks and crack initiation, in order to confirm their reliability and applicability. Based on the numerical investigation in consideration of important influence factors such as crack number, crack location, crack distribution and crack detection probability of in-service inspection, it was concluded that the analysis results of these two codes are in good agreement.
Hasegawa, Kunio*; Li, Y.; Bezensek, B.*; Hoang, P. H.*; Rathbun, H. J.*
Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 10 Pages, 2014/07
Piping components in power plants may experience combined bending and torsion moments during operation. There is a lack of guidance for pipe evaluation for pipes with local wall thinning flaws under the combined bending and torsion moments. ASME BPV Code Section XI Working Group is currently developing fully plastic bending pipe evaluation procedures for pressurized piping components containing local wall thinning subjected to combined torsion and bending moments. Using elastic fully plastic finite element analyses, plastic collapse bending moments under torsions were obtained for 4 to 24 inch diameter pipes with various local wall thinning flaw sizes. The objective of this paper is to introduce an equivalent moment, which combines torsion and bending moments by a vector summation, and to establish the applicable range of wall thinning lengths, angles and depths, where the equivalent moments are equal to pure bending moments.
Okajima, Satoshi; Wakai, Takashi; Ando, Masanori; Inoue, Yasuhiro*; Watanabe, Sota*
Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 8 Pages, 2014/07
Watakabe, Tomoyoshi; Tsukimori, Kazuyuki; Otani, Akihito*; Moriizumi, Makoto; Kaneko, Naoaki*
Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 8 Pages, 2014/07
In recent years, earthquakes over design condition were observed in Japan. Confirming the ultimate strength and design safety margin of mechanical components is important for the seismic integrity. This study focused on piping components, and it was one of the most important mechanical components for protecting boundary of coolant. Failure tests of thick-walled piping components for Light Water Reactors (LWRs) described previously in the literature. According to these tests, the failure mode of thick-walled piping components under seismic cyclic loading was low cycle fatigue. However, failure tests have scarcely been performed on thin-walled piping components pressurized at low levels for Fast Breeder Reactors (FBRs). This paper presents dynamic failure tests of thin-walled piping components in FBRs. Based on the test results, the failure mode, the ultimate strength, and the elastic-plastic behavior are discussed.
Kai, Satoru*; Watakabe, Tomoyoshi; Kaneko, Naoaki*; Tochiki, Kunihiro*; Moriizumi, Makoto; Tsukimori, Kazuyuki
Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 10 Pages, 2014/07
Piping in a nuclear power plant is usually laid across several floors of a single building or adjacent buildings, and is supported at many points. As the piping is excited by a large earthquake through multiple supporting points, seismic response analysis by multiple excitations within the range of plastic deformation of piping material is necessary to obtain the precise seismic response of the piping. This paper reports the validation results of the seismic elastic-plastic time history analysis of piping compared with the results of the shaking test of a 3-dimensional piping model under a plastic deformation range using triple uni-axial shake table.
Asayama, Tai; Wakai, Takashi; Ando, Masanori; Okajima, Satoshi; Nagae, Yuji; Takaya, Shigeru; Onizawa, Takashi; Tsukimori, Kazuyuki; Morishita, Masaki
Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 6 Pages, 2014/07
This paper overviews the ongoing research and development as well as activities for codification of structural codes for the Japan Sodium Cooled Fast Reactor (JSFR). Not only the design and construction code which has been published and updated on a regular basis, codes on welding, fitness-for-service, leak-before-break evaluation as well as the guidelines for structural reliability evaluation are being developed. The basic strategy for the development is to fully take advantage of the favorable technical characteristics associated with sodium-cooled fast reactors; the codes will be developed based on the System Based Code concept. The above mentioned set of codes are planned to be published from the Japan Society of Mechanical Engineers in 2016.
Garcia Rodriguez, D.; Sakakibara, Yasuhide*
Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 9 Pages, 2014/07
The behavior of high temperature (400
C) low-cycle fatigue crack propagation during in-plane bending of an elbow from Monju LMFBR is presented in three stages. 1st, experimental measurements made in a real-size laboratory specimen are presented. This test was carried out under displacement-controlled conditions, with artificial defects introduced in the crown parts, where the maximum stress arises. 2nd, the experimental setup is simulated making use of FEA, in order to obtain the actual stress distribution through the loaded elbow. Finally, based on the FEA data, deterministic fatigue crack propagation based on the J-integral criterion is compared with the experimental data.
Nagae, Yuji; Onizawa, Takashi; Takaya, Shigeru; Yamashita, Takuya
Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 9 Pages, 2014/07
