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Journal Articles

Applicability of miniature compact tension specimens for fracture toughness evaluation of highly neutron irradiated reactor pressure vessel steels

Ha, Yoosung; Tobita, Toru; Otsu, Takuyo; Takamizawa, Hisashi; Nishiyama, Yutaka

Journal of Pressure Vessel Technology, 140(5), p.051402_1 - 051402_6, 2018/10

 Times Cited Count:1 Percentile:11.36(Engineering, Mechanical)

JAEA Reports

Confirmation tests for Warm Pre-stress (WPS) effect in reactor pressure vessel steel (Contract research)

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.

Journal Articles

Fracture toughness evaluation of reactor pressure vessel steels by master curve method using miniature compact tension specimens

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:7 Percentile:42.22(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 ($$T_{o}$$) 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 $$T_{o}$$, 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 $$T_{o}$$ using the 0.16T-CT specimens.

Journal Articles

Finite element analysis on the application of Mini-C(T) test specimens for fracture toughness evaluation

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.

Journal Articles

Fracture toughness evaluation of reactor pressure vessel steels by master curve method using Mini-CT specimens

Tobita, Toru; Nishiyama, Yutaka; Otsu, Takuyo; Udagawa, Makoto; Katsuyama, Jinya; Onizawa, Kunio

Proceedings of 2013 ASME Pressure Vessels and Piping Conference (PVP 2013) (DVD-ROM), 8 Pages, 2013/07

Mini-CT (0.16T-CT) specimens up to eight can be taken from broken halves of surveillance Charpy specimens. 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. Reference temperature To of 0.16T-CT specimens was approximately equal to those of 1T-CT and other type of specimens for all materials. We also examined a loading rate effect on TO of Mini-CT specimens for some materials within the specified range in the test method. Higher loading rate gave rise to slightly higher TO. The difference in TO between upper and lower loading rate of the standard was approximately 10$$^{circ}$$C.

Oral presentation

Investigation on surveillance specimen exclusion for heat affected zone materials of reactor pressure vessel steels, 5; Evaluation of the susceptibility to neutron irradiation embrittlement

Tobita, Toru; Katsuyama, Jinya; Fuse, Masaharu; Otsu, Takuyo*; Onizawa, Kunio

no journal, , 

In order to obtain the technical basis to judge the necessity of surveillance specimens from heat-affected zone (HAZ) materials of reactor pressure vessel (RPV) steels, the neutron irradiation program for base metals and simulated HAZ materials was performed at JRR-3. The materials used were fabricated as typical RPV steels varying impurity levels. It was indicated that the fracture toughness of simulated HAZ materials was equivalent to or better than that of base metal before neutron irradiation. Concerning the effects of impurities and metallographic structures on the mechanical properties the neutron irradiation embrittlement susceptibility of simulated HAZ materials was compared with that of base metal. The irradiation embrittlement as well as irradiation hardening of simulated HAZ materials was evaluated to be roughly equivalent to those of base metal. However, it was suggested that there may be a local region in HAZ susceptible more to irradiation embrittlement than base metal.

Oral presentation

Evaluation of fracture toughness of reactor pressure vessel steels by master curve approach, specimen size effect on fracture toughness

Tobita, Toru; Nishiyama, Yutaka; Otsu, Takuyo; Udagawa, Makoto; Onizawa, Kunio

no journal, , 

no abstracts in English

Oral presentation

Development of the fracture toughness test technique using the miniature specimen in the hot laboratory

Otsu, Takuyo; Tobita, Toru; Fuse, Masaharu; Yamaki, Kenichi; Terakado, Hiroshi; Nishiyama, Yutaka

no journal, , 

no abstracts in English

Oral presentation

Fracture toughness evaluation of neutron-irradiated reactor pressure vessel steels using miniature C(T) specimens

Tobita, Toru; Otsu, Takuyo; Nishiyama, Yutaka

no journal, , 

Fracture toughness of the neutron irradiated reactor pressure vessel steel up to a fluence of 1$$times$$10$$^{20}$$ n/cm$$^{2}$$ was performed using the 4 mm-thick miniature C(T) specimen. We evaluate fracture toughness scatter and a relation between the shifts of fracture toughness and Charpy DBTT.

Oral presentation

Evaluation of brittle crack arrest toughness of reactor pressure vessel steels

Tobita, Toru; Otsu, Takuyo*; Takamizawa, Hisashi; Nishiyama, Yutaka

no journal, , 

When the reactor pressure vessel has been subjected to the pressurized thermal shock event, even if an non-ductile crack occurs from the postulated defect at the inner surface of the reactor pressure vessel, the crack may stop before penetrates the vessel wall. In this report, crack-arrest fracture toughness (K$$_{Ia}$$) tests were performed on the three types of reactor pressure vessel steels with different mechanical properties. It was confirmed that the temperature dependency of K$$_{Ia}$$ follows the Master Curve as well as static fracture toughness. In addition, we examined the correlation between the crack-arrest fracture toughness and crack arrest force of instrumentation Charpy test.

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