Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Shimodaira, Masaki; Tobita, Toru; Takamizawa, Hisashi; Katsuyama, Jinya; Hanawa, Satoshi
Proceedings of ASME 2020 Pressure Vessels and Piping Conference (PVP 2020) (Internet), 7 Pages, 2020/08
In JEAC 4206 which prescribes the methodology for assessing the structural integrity of reactor pressure vessels (RPVs), an under-clad crack (UCC) at the inner surface of RPV is postulated, and it is required that the fracture toughness of RPV steels is higher than stress intensity factor for at the crack tip during the pressurized thermal shock event. In the present study, to investigate the effect of cladding on the fracture toughness, we performed three-point bending fracture toughness tests and finite element analyses (FEAs) for an RPV steel containing an UCC or a surface crack, and the constraint effect for UCC was also discussed. As the result, we found that the fracture toughness for UCC was considerably higher than that for surface crack. On the other hand, the FEAs showed that the cladding decreased the constraint effect for UCC.
Chimi, Yasuhiro; Iwata, Keiko; Tobita, Toru; Otsu, Takuyo; Takamizawa, Hisashi; Yoshimoto, Kentaro*; Murakami, Takeshi*; Hanawa, Satoshi; Nishiyama, Yutaka
JAEA-Research 2017-018, 122 Pages, 2018/03
JAEA-Research-2017-018.pdf:44.03MB
Warm pre-stress (WPS) effect is a phenomenon that after applying a load at a high temperature fracture does not occur in unloading during cooling, and then the fracture toughness in reloading at a lower temperature increases effectively. Engineering evaluation models to predict an apparent fracture toughness in reloading are established using experimental data with linear elasticity. However, there is a lack of data on the WPS effect for the effects of specimen size and surface crack in elastic-plastic regime. In this study, fracture toughness tests were performed after applying load-temperature histories which simulate pressurized thermal shock transients to confirm the WPS effect. The experimental results of an apparent fracture toughness tend to be lower than the predictive results using the engineering evaluation models in the case of a high degree of plastic deformation in preloading. Considering the plastic component of preloading can refine the engineering evaluation models.
Tobita, Toru; Nishiyama, Yutaka; Otsu, Takuyo; Udagawa, Makoto; Katsuyama, Jinya; Onizawa, Kunio
Journal of Pressure Vessel Technology, 137(5), p.051405_1 - 051405_8, 2015/10
Times Cited Count：14 Percentile：54.45(Engineering, Mechanical)We conducted a series of fracture toughness tests based on the Master curve method for several specimen size and shapes, such as 0.16T-CT, pre-cracked Charpy type, 0.4T-CT and 1T-CT specimens, in commercially manufactured 5 kinds of A533B class1 steels with different impurity contents and fracture toughness levels. The reference temperature (
) values determined from the 0.16T-CT specimens were overall in good agreement with those determined from the 1T-CT specimens. The scatter of the 1T-equivalent fracture toughness values obtained from the 0.16T-CT specimens was equivalent to that obtained from the other larger specimens. The higher loading rate gave rise to a slightly higher , and this dependency was almost the same for the larger specimens. We suggested an optimum test temperature on the basis of the Charpy transition temperature for determining using the 0.16T-CT specimens.
Takamizawa, Hisashi; Tobita, Toru; Otsu, Takuyo; Katsuyama, Jinya; Nishiyama, Yutaka; Onizawa, Kunio
Proceedings of 2015 ASME Pressure Vessels and Piping Conference (PVP 2015) (Internet), 7 Pages, 2015/07
Fracture toughness evaluation by the Master Curve method using miniature compact tension (Mini-C(T)) specimens taken from the broken halves of surveillance Charpy specimens has been proposed. We performed finite element analysis (FEA) to examine the difference in constraint effect of the crack tip for the different size C(T) and precracked Charpy v-notch specimens. The constraint effect for Mini-C(T) specimens in terms of the T-stress and Q-parameter was similar to the larger C(T) specimens. To optimize the fatigue precracking conditions for the Mini-C(T) specimen fabrication, plastic zone distribution analysis was performed. We confirmed the fatigue precrack length and the availability of the mitigated crack shape for Mini-C(T). We also obtained the fracture toughness data for different sizes specimen. It was shown that To obtained from the Mini-C(T) specimens is in reasonably good agreement with that from others. We compared the fracture toughness data with T41J based fracture toughness curves proposed in recent study. All of the data were well enveloped by the proposed lower bound curve.
Hoshino, Takanori; Kitagaki, Toru; Yano, Kimihiko; Okamura, Nobuo; Ohara, Hiroshi*; Fukasawa, Tetsuo*; Koizumi, Kenji
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
Wakai, Eiichi; Otsuka, Hideo*; Matsukawa, Shingo; Furuya, Kazuyuki*; Tanigawa, Hiroyasu; Oka, Keiichiro*; Onuki, Somei*; Yamamoto, Toshio*; Takada, Fumiki; Jitsukawa, Shiro
Fusion Engineering and Design, 81(8-14), p.1077 - 1084, 2006/02
Times Cited Count：11 Percentile：60.27(Nuclear Science & Technology)no abstracts in English
Onizawa, Kunio; Suzuki, Masahide
JSME International Journal, Series A, 47(3), p.479 - 485, 2004/07
In the structural integrity assessment of reactor pressure vessel, fracture toughness values are estimated by assuming that the radiation effect on fracture toughness is equivalent to that on Charpy properties. Therefore, it is necessary to establish the correlation between both properties especially on irradiation embrittlement. In this paper, we present the fracture toughness data obtained by applying the master curve approach that was adopted recently in the ASTM test method. Materials used in this study are five ASTM A533B class 1 steels and one weld metal. Neutron irradiation for Charpy-size specimens as well as standard Charpy-v specimens was carried out at the Japan Materials Testing Reactor. The shifts of the reference temperature on fracture toughness due to neutron irradiation are evaluated. Correlation between the fracture toughness reference temperature and Charpy transition temperature is established. Based on the correlation, the optimum test temperature for fracture toughness testing and the method to determine a lower bound fracture toughness curve are discussed.
Onizawa, Kunio; Shibata, Katsuyuki; Kato, Daisuke*; Li, Y.*
JSME International Journal, Series A, 47(3), p.486 - 493, 2004/07
The probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed in JAERI. This code can evaluate the conditional probabilities of crack initiation and fracture of a reactor pressure vessel (RPV) under transient conditions such as pressurized thermal shock (PTS). Based on the temperature and stress distributions in the vessel wall for four PTS sequences in a typical 3-loop PWR, parametric PFM analyses are performed using PASCAL on the variables such as pre-service inspection model, crack geometry, fracture toughness curve and irradiation embrittlement prediction equation. The results showed that the good perfomance inspection model had a significant effect on the fracture probability and reduced it by more than 3 orders of magnitude. The fracture probability calculated by the fracture toughness estimation method in Japan was about 2 orders of magnitude lower than that by the USA method. It was found that the treatment of a semi-elliptical crack in PASCAL reduced the conservatism in a conventional method that it is transformed into an infinite length crack.
Onizawa, Kunio; Shibata, Katsuyuki; Suzuki, Masahide; Kato, Daisuke*; Li, Y.*
RPV Integrity and Fracture Mechanics (PVP-Vol.481), p.11 - 17, 2004/07
At JAERI, the prpbabilistic fracture mechanics analysis code PASCAL has been developed. Using the PASCAL, the treatment methods of an embedded crack and the fracture toughness evaluation methods were studied on the probability of crack initiation and fracture of a RPV. For calculating the stress intensity factor (SIF) of an embedded crack, the ASME and CRIEPI procedures were introduced into PASCAL. Under a severe pressurized thermal shock condition, the crack growth analysis models with different SIF calculation points and crack growth directions are compared. The results showed that the crack tip at the inner side was most important to the fracture probability. The conditional fracture probability for an embedded crack when the same crack size distribution is assumed is approximately one order of magnitude lower than that of a surface crack. Fracture toughness curves based on Weibull distribution were incorporated into PASCAL. The comparison between these new curves and the current curves showed little effect on the conditional fracture probability of a RPV.
Onizawa, Kunio; Shibata, Katsuyuki; Kato, Daisuke*; Li, Y.*
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 8 Pages, 2003/04
Probabilistic Fracture Mechanics (PFM) has been used in the fields of reliability analysis for important structural components. At JAERI, the PFM analysis code PASCAL has been developed. This code evaluates the conditional probabilities of crack initiation and fracture of a reactor pressure vessel (RPV) under transient conditions such as pressurized thermal shock (PTS). Four cases of PTS transients were selected based on the severity for a typical 3-loop PWR. Based on thermal stress analyses, PFM analyses were performed by using PASCAL code focusing on some important variables on the RPV fracture probability. The results showed that non-destructive examination methods had a significant effect on the fracture probability by more than three orders of magnitude. The comparisons of the results using fracture toughness estimation methods between in Japan and USA, and crack geometries between a semi-elliptical surface crack and an infinite surface crack are also made.
Jitsukawa, Shiro; Tamura, Manabu*; Van der Schaaf, B.*; Klueh, R. L.*; Alamo, A.*; Petersen, C.*; Schirra, M.*; Spaetig, P.*; Odette, G. R.*; Tavassoli, A. A.*; et al.
Journal of Nuclear Materials, 307-311(Part1), p.179 - 186, 2002/12
Times Cited Count：160 Percentile：99.28(Materials Science, Multidisciplinary)Reduced activation ferritic/martensitic steel is the primary candidate structural material for ITER Test Blanket Modules and DEMOnstration fusion reactor because of its excellent dimensional stability under irradiation and lower residual activity as compared with the Ni bearing steels such as the austenitic stainless steels. In this paper, microstructural features, tensile, fracture toughness, creep and fatigue properties of a reduced activation martensitic steel F82H (8Cr-2W-0.04Ta-0.1C) are reported before and after irradiation, in addition to the design concept used for development of this alloy. A large number of collaborative test results including those generated under the IEA working group implementing agreements are collected and are used to evaluate the feasibility of use of F82H steel as one of the reference alloys. The effect of metallurgical variables on the irradiation hardening is reviewed and compared with the results obtained from irradiation experiments.
Ishiyama, Shintaro; Eto, Motokuni
Nihon Genshiryoku Gakkai-Shi, 43(11), p.1159 - 1166, 2001/11
Times Cited Count：2 Percentile：19.66(Nuclear Science & Technology)no abstracts in English
Onizawa, Kunio; Suzuki, Masahide
Proceedings of Asian Pacific Conference on Fracture and Strength '01(APCFS '01) and International Conference on Advanced Technology in Experimental Mechanics '01 (ATEM '01), p.140 - 145, 2001/00
In the structural integrity assessment of reactor pressure vessel, fracture toughness values are estimated by assuming that the radiation effect on fracture toughness is equivalent to that on Charpy properties. Therefore, it is necessary to establish the correlation between both properties especially on irradiation embrittlement. In this paper, we present the fracture toughness data obtained by applying the master curve approach that was adopted recently in the ASTM test method. Materials used in this study are five ASTM A533B class 1 steels and one weld metal. Neutron irradiation for Charpy-size specimens as well as standard Charpy-v specimens was carried out at the Japan Materials Testing Reactor. The shifts of the reference temperature on fracture toughness due to neutron irradiation are evaluated. Correlation between the fracture toughness reference temperature and Charpy transition temperature is established. Based on the correlation, the optimum test temperature for fracture toughness testing and the method to determine a lower bound fracture toughness curve are discussed.
Onizawa, Kunio; Suzuki, Masahide
Proceedings of the 8th Japanese-German Joint Seminar on Structural Integrity and NDE in Power Engineering, p.62 - 69, 2001/00
To assure the structural integrity of reactor pressure vessel (RPV) throughout its operational life, fracture toughness of the steel after neutron irradiation must be determined. In this report the investigation on the master curve approach using Charpy-size specimens is presented for the precise evaluation of fracture toughness on irradiation embrittlement. Using some Japanese A533B-1 steels, fracture toughness tests in the transition range were performed varying specimen thickness. Charpy-size specimens were also irradiated at Japan Materials Testing Reactor. Applying the master curve method and JEAC method as well, the specimen size effect, temperature dependence and the lower bound were evaluated. The shifts of reference temperature of fracture toughness and Charpy transition temperature due to neutron irradiation were also compared and found to be almost equivalent.
Onizawa, Kunio; Suzuki, Masahide
ISIJ International, 37(8), p.821 - 828, 1997/08
Times Cited Count：3 Percentile：35.47(Metallurgy & Metallurgical Engineering)no abstracts in English
Ishiyama, Shintaro; Eto, Motokuni
Zairyo, 41(463), p.475 - 481, 1992/04
no abstracts in English
Nakajima, Hideo; Yoshida, Kiyoshi; Tsuji, Hiroshi; R.L.Tobler*; I.S.Hwang*; M.M.Morra*; R.G.Ballinger*
Advances in Cryogenic Engineering, Materials, Vol.38, p.207 - 215, 1992/00
no abstracts in English
Nishiyama, Yutaka; Fukaya, Kiyoshi; Suzuki, Masahide; Kodaira, Tsuneo; Oku, Tatsuo*
Effects of Radiation on Materials; 15th International Symposium (ASTM STP 1125), p.1287 - 1303, 1992/00
no abstracts in English
Eto, Motokuni; Ishiyama, Shintaro; Nishiyama, Yutaka; Oku, Tatsuo*; T.D.Burchell*
Int. Symp. on Carbon New Processing and New Applications; Extended Abstracts,Vol. 1, p.170 - 173, 1990/11
no abstracts in English
