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Shibata, Katsuyuki; Onizawa, Kunio; Tanaka, Kazuhisa; Suzuki, Masahide
Proceedings of 2006 ASME Pressure Vessels and Piping Division Conference (PVP 2006)/International Council on Pressure Vessel Technology (ICPVT-11) (CD-ROM), 9 Pages, 2006/07
no abstracts in English
Kawasaki, Nobuchika; Kubayashi, Sumio; Hasebe, Shinichi; Kasahara, Naoto
Proceedings of 2006 ASME Pressure Vessels and Piping Division Conference (PVP 2006)/International Council on Pressure Vessel Technology (ICPVT-11) (CD-ROM), 8 Pages, 2006/07
The coolant leakage by thermal striping phenomenon should be prevented at nuclear power plants and a lot of efforts are made to develop its evaluation methods. The frequency transfer function method can explain temperature and stress response to fluid temperature history using transfer function models; therefore it is considered as an excellent evaluation method. To measure temperature response of structures to fluid and to confirm their frequency characteristics, transient temperature measurement tests were performed by JAEA. In the transient temperature measurement tests, three different frequencies of sinusoidal fluid temperature waves (0.05, 0.2, 0.5Hz) were controlled using frequency controlled thermal fatigue test equipment (SPECTRA) and temperature responses at inner and outer structural surfaces were measured along the test sections. Frequency effects on temperature attenuation during transfer process from fluid to structures were confirmed and the effective heat transfer function in frequency transfer function method was verified by transient temperature measurement test results.
Onizawa, Kunio; Shibata, Katsuyuki; Osakabe, Kazuya; Tanaka, Kazuhisa
Proceedings of 2006 ASME Pressure Vessels and Piping Division Conference (PVP 2006)/International Council on Pressure Vessel Technology (ICPVT-11) (CD-ROM), 8 Pages, 2006/07
Probabilistic fracture mechanics (PFM) analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed at JAEA to evaluate the conditional failure probability of a reactor pressure vessel (RPV) containing a flaw under transient conditions such as pressurized thermal shock (PTS). Some functions of PASCAL have been improved such as Monte Carlo method, probability of crack detection (POD) by inspection and the GUI. Using the improved PASCAL, case studies on some parameters for some typical PTS transients have been performed. When the crack existence probability for the Marshall- and PNNL-type distributions is considered, the failure probability could be the same order of magnitude. The effect of chemical composition of vessel material was studied through the comparison of conditional failure probabilities for some transients. The effect of POD by inspection was also compared and evaluated in terms of inspection time.
Sukekawa, Masayuki*; Isobe, Nobuhiro*; Shibamoto, Hiroshi; Tanaka, Yoshihiko*; Kasahara, Naoto
Proceedings of 2006 ASME Pressure Vessels and Piping Division Conference (PVP 2006)/International Council on Pressure Vessel Technology (ICPVT-11) (CD-ROM), 5 Pages, 2006/00
For expansion of non-creep design area and simplification of design procedures, a rational identification method of creep design area by negligible creep (NC) curves was studied. NC curves of six kinds of stainless and ferrite steels for fast reactors were determined at 1.5Sm (Sm: design stress intensity). These NC curves are based on domestic material data. NC curves provide the relation between temperature and time that does not induce meaningful creep strain under the constant primary stress. As for 316FR steel, which is used for reactor vessel in Japanese fast reactor, non-creep design area is identified with comparing the highest temperature and 425C (constant upper limit for austenite stainless steal) by existing Japanese Guides. However, this temperature limit can be enhanced by NC curve concept when operating (thermal transient) time is long. NC curves under higher primary stress, and the curves under secondary stress were also studied. However, at the present stage, NC curves for stress level 1.5Sm were adopted to identify creep design area. The concept of NC curve was introduced into the interim FDS (fast reactor design standard for commercialized fast reactors in Japan) to simplify the creep design of fast reactor systems. Utilizing these curves, design becomes easier for components which are employed at comparatively lower temperature under normal condition and short holding time at high temperature.
Watanabe, Daigo*; Chuman, Yasuharu*; Otani, Tomomi*; Shibamoto, Hiroshi; Inoue, Kazuhiko*; Kasahara, Naoto
Proceedings of 2006 ASME Pressure Vessels and Piping Division Conference (PVP 2006)/International Council on Pressure Vessel Technology (ICPVT-11) (CD-ROM), 7 Pages, 2006/00
Prevention of thermal ratcheting is an important problem for high temperature components of fast breeder reactors that are subjected to cyclic thermal loads. To clarify ratcheting behaviors, structural model tests were planned. Strain measurement is important for understanding the thermal ratcheting phenomenon, however the conventional measurement by strain gauge is difficult at high temperature. Then, Laser speckle strain measurement system using the dual-beam set-up was developed to apply to high temperature structural model tests. This system was applied to the thermal ratcheting tests, which demonstrated the actual operative conditions of reactor vessels. Through comparison with uniaxial test results obtained by extensometers, the laser speckle method was verified. Measured data of structural model tests were utilized to certify the guidelines of inelastic analysis for design, which provide prediction method of strain in components of fast reactor.
Kasahara, Naoto; Furuhashi, Ichiro*
Proceedings of 2006 ASME Pressure Vessels and Piping Division Conference (PVP 2006)/International Council on Pressure Vessel Technology (ICPVT-11) (CD-ROM), 8 Pages, 2006/00
Dependency of results of structural analysis on constitutive equations obstructs the application of inelastic analysis methods to design. To overcome this difficulty, a realistic inelastic analysis approach assisted by the stress redistribution locus (SRL) has been proposed for the purpose of design application. The basis of this approach is the existence of an unique SRL in each structure. The final goal of this study is to assure the uniqueness of SRL. To clarify control mechanisms of SRLs in structures, elastic-plastic and elastic-creep behaviors of fundamental structures are analyzed.