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Li, Y.; Uno, Shumpei*; Katsuyama, Jinya; Dickson, T.*; Kirk, M.*
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 8 Pages, 2017/07
Lu, K.; Katsuyama, Jinya; Uno, Shumpei; Li, Y.
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 8 Pages, 2017/07
Katsuyama, Jinya; Osakabe, Kazuya*; Uno, Shumpei; Li, Y.; Yoshimura, Shinobu*
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 9 Pages, 2017/07
A structural integrity assessment methodology based on probabilistic fracture mechanics (PFM) is a rational methodology in evaluating failure frequency of reactor pressure vessels (RPVs) by considering the probabilistic distributions of various influence factors related to the aged degradation. We have developed a PFM analysis code PASCAL to evaluate the failure frequency of RPVs considering the neutron irradiation embrittlement and pressurized thermal shock (PTS) events. We have also developed a guideline on the structural integrity assessment of RPVs based on PFM to improve the applicability of PFM in Japan and to be able to perform the PFM analyses and evaluate through-wall cracking frequency of RPVs. The technical basis for PFM analysis is provided and the latest knowledge is included in the guideline. In this paper, an overview of the guideline and some typical analysis results obtained based on the guideline and Japanese database related to PTS evaluation are presented.
Nishida, Akemi; Ota, Yoshimi*; Tsubota, Haruji; Li, Y.
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 10 Pages, 2017/07
Many empirical formulas have been proposed for evaluating local damage to reinforced concrete (RC) structures caused by impacts of rigid projectiles. Most of these formulas have been derived based on impact tests perpendicular to the target structures. Up to now, few impact tests oblique to the target structures have been carried out. This study has been conducted with the purpose of proposing a new formula for evaluating the local damage caused by oblique impacts based on previous experimental and simulation results. In this paper, the results of simulation analyses for evaluating the local damage to a RC panel subjected to normal and oblique impacts by deformable projectile, by using the simulation method that the validity was confirmed through simulation analysis of past impact experimental results. Based on the results of these simulation analyses, the quantitative evaluation of reduction effects in the local damage caused by the difference of the impact angle is investigated.
Lacroix, V.*; Bouydo, A.*; Katsumata, Genshichiro*; Li, Y.; Hasegawa, Kunio
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 7 Pages, 2017/07
Azuma, Kisaburo*; Li, Y.; Hasegawa, Kunio; Shim, D. J.*
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 7 Pages, 2017/07
Azuma, Kisaburo*; Li, Y.; Hasegawa, Kunio
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 6 Pages, 2017/07
Azuma, Kisaburo*; Li, Y.; Hasegawa, Kunio; Xu, S.*
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 9 Pages, 2017/07
Takaya, Shigeru; Fujisaki, Tatsuya*; Tanaka, Masaaki
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 8 Pages, 2017/07
Flow-induced vibration (FIV) of a hot-leg piping is one of main concerns in the design of an advanced loop-type sodium cooled fast reactor. We have been developing numerical analysis models to deal with this issue. In this study, computational fluid dynamics (CFD) simulation of a 1/3 scaled-model of the hot-leg piping was conducted. The results such as velocity profiles and power spectral densities (PSD) of pressure fluctuations were compared with experiment ones. The simulated PSD of pressure fluctuation at the recirculation region agreed well with the experiment. Then, stress induced by FIV was evaluated using pressure fluctuation data calculated by the CFD simulation. The calculated stress generally agrees well the measurement values, which indicates the importance of precise evaluation of the PSD of pressure fluctuation at the recirculation region for evaluation of FIV of the hot-leg piping with a short elbow.
Ando, Masanori; Yada, Hiroki; Tsukimori, Kazuyuki; Ichimiya, Masakazu*; Anoda, Yoshinari*
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 11 Pages, 2017/07
In this study, in order to develop the evaluation method of the pressure toughness of bellows structure under the beyond design base event, the pressure failure tests and finite element analysis (FEA) of the bellows structure subjected to internal pressure were performed. In the test of five convolutions 0.5 mm-thickness bellows specimen with guard pipe, the maximum pressure was larger than those in the tests without guard pipe specimens and ductile failure was observed. On the other hand, in the test of five convolutions 0.5 mm-thickness bellows specimen without guard pipe, local failure was observed. In the test of the six convolutions 1.0 mm-thickness bellows specimen, ductile failure was observed in the both single and double ply bellows. The maximum pressure obtained in all tests were about 10 times larger than the estimated results of limiting design pressure based on in-plain instability by the EJMA standards.
Yada, Hiroki; Ando, Masanori; Tsukimori, Kazuyuki; Ichimiya, Masakazu*; Anoda, Yoshinari*
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 8 Pages, 2017/07
The head plate that composes the boundary between primary and secondary coolant in intermediate heat exchanger of FBR has an important role when the progress of the BDBE is considered. In order to develop the evaluation method of the pressure toughness of the head plate under the BDBE, the pressure failure tests and finite element analysis of two types of head plate subjected to convex side pressure was performed in this study. It can be concluded that a failure mode of a head plate subjected convex side pressure is circumferential through-wall crack caused by straightening following the bending deformation near the rim.
Hasegawa, Kunio; Li, Y.; Katsumata, Genshichiro*; Dulieu, P.*; Lacroix, V.*
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 6 Pages, 2017/07
Net-section stress at the ligament between component free surface and subsurface flaw increases when the ligament distance is short. It can be easily expected that stress intensity factors increase when the subsurface flaw locates near the free surface. To avoid catastrophic failures caused by ligament failure, fitness-for-service (FFS) codes provide flaw-to-surface proximity rules. The proximity rules are used to determine whether the flaws should be treated as subsurface flaws as-is, or transformed to surface flaws. The stress intensity factor for the transformed surface flaw increases furthermore. The increment of the stress intensity factor before and after transformation depends on the location of the subsurface flaw. Although the concept of the proximity rules are the same, the specific criteria for the rules on transforming subsurface flaws to surface flaws differ amongst FFS codes. Particularly, the criteria are different amongst the same organizations of ASME (American Society of Mechanical Engineers). The proximity criteria of the FFS codes in the world were introduced in this paper. In addition, the stress intensity factors based on the different criteria used in the ASME Codes are compared.
Hasegawa, Kunio; Dulieu, P.*; Lacroix, V.*
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 5 Pages, 2017/07
The stress intensity factors of the subsurface flaws are affected by the stress concentrations caused by the notches. The interaction of stress intensity factor increases with increasing stress concentration factor and decreasing the ligament distance between the tips of the subsurface flaws and the notches for a given notch width. Such subsurface flaws shall be transformed to surface flaws at far distance of the notch tips for conservative evaluations. This paper shows the interactions of stress intensity factors of subsurface flaws under stress concentration fields. Based on the interaction, a flaw-to-surface proximity criterion for a circular flaw is proposed under the stress concentration field induced by a notch.
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.
Yamamoto, Tomohiko; Kitamura, Seiji; Iwasaki, Akihisa*; Matsubara, Shinichiro*; Okamura, Shigeki*
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 10 Pages, 2017/07
To design fast reactor (FR) components, seismic response must be evaluated in order to ensure structural integrity. Therefore, a sophisticated analysis method has to be developed to study the seismic response of FR core. The fast reactors are made of several hundred core assemblies in hexagonal arrangement. When a big earthquake occurs, large horizontal displacement and impact force of each core assembly may cause a trouble for control rod insertability and core assembly intensity. Therefore, a seismic analysis method of fast reactor core considering horizontal nonlinear behavior, such as impact, fluid-structure interaction, etc. is needed. Validation of the core assembly vibration analysis code in three dimension (REVIAN-3D) was conducted by a full scale experiment. In this validation, the vertical behavior (raising displacement) and horizontal behavior (Impact force, horizontal response) of the analysis result agreed very well with the experiments.
Ha, Yoosung; Tobita, Toru; Takamizawa, Hisashi; Nishiyama, Yutaka
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 5 Pages, 2017/07
The applicability of miniature-C(T) (Mini-C(T)) specimens to fracture toughness evaluation was investigated for neutron-irradiated reactor pressure vessel (RPV) steel. 
value determined from irradiated Mini-C(T) specimens was in good agreement with that determined from the irradiated pre-cracked Charpy-type (PCCv) specimens. Also, the scatter of the 1T-equivalent fracture toughness values obtained from the irradiated Mini-C(T) specimens was not significantly different from that obtained from the irradiated PCCv. values determined from Mini-C(T) specimens agreed very well with the correlation between Charpy 41J transition temperature and of commercially manufactured RPV steels.
Shimomura, Kenta; Onizawa, Takashi; Kato, Shoichi; Ando, Masanori; Wakai, Takashi
no journal, ,
This paper describes the formulation of material characteristics of austenitic stainless steels at extremely high temperature which meets in some kinds of severe accidents of nuclear power plants. After the severe accident in Fukushima Dai-ichi Nuclear Power Plants, it has been supposed to be very important not only to prevent the occurrence of abnormal conditions, i.e. from the first to the third layer safety, but also to prevent the expansion of the accident conditions, i.e. the fourth layer safety. In order to evaluate the structural integrity under the severe accident condition, material characteristics which can be used in the numerical analyses, such as finite element analysis, were required. However, there were no material characteristics applicable to the structural integrity assessment at extremely high temperature. Therefore, a series of tensile and creep tests was performed for austenitic stainless at extremely high temperature which meets in some kinds of severe accidents of nuclear power plants, namely up to 1000
C. Based on the acquired data from the tests, monotonic stress-strain equation and creep rupture equation applicable to the structural analysis at extremely high temperature, up to 1000C were formulated. As a result, these formulae make it possible to conduct the structural integrity assessment using numerical analysis techniques, such as finite element method.
Takamizawa, Hisashi; Tobita, Toru; Iwata, Keiko; Katsuyama, Jinya; Nishiyama, Yutaka
no journal, ,
no abstracts in English
