The applicability of miniature compact tension (Mini-C(T)) specimens to fracture toughness evaluation of neutron-irradiated reactor pressure vessel (RPV) steels was investigated. Three types of RPV steels neutron-irradiated to a high-fluence region were prepared and manufactured as Mini-C(T) specimens according to Japan Electric Association Code (JEAC) 4216-2015. Through careful selection of the test temperature by considering previously obtained mechanical properties data, valid fracture toughness, and reference temperature (To) was obtained with a relatively small number of specimens. Comparing the fracture toughness and To values determined using other larger specimens with those determined using the Mini-C(T) specimens, To values of both unirradiated and irradiated Mini-C(T) specimens were found to be the acceptable margin of error. The scatter of 1T-equivalent fracture toughness values of both unirradiated and irradiated materials obtained using Mini-C(T) specimens did not differ significantly from the values obtained using larger specimens. The correlation between the Charpy 41 J transition temperature (T41J) and the To values agreed very well with that of the data in the literature, regardless of specimen size and fracture toughness of the materials before irradiation. Based on these findings, it was concluded that Mini-C(T) specimens can be applied to fracture toughness evaluation of neutron-irradiated materials without significant specimen size dependence.
Skip Nav Destination
Article navigation
October 2018
Research-Article
Applicability of Miniature Compact Tension Specimens for Fracture Toughness Evaluation of Highly Neutron Irradiated Reactor Pressure Vessel Steels
Yoosung Ha,
Yoosung Ha
Nuclear Safety Research Center,
Japan Atomic Energy Agency,
Shirakata 2-4, Tokai-mura,
Naka-gun 319-1195, Ibaraki-ken, Japan
e-mail: ha.yoosung@jaea.go.jp
Japan Atomic Energy Agency,
Shirakata 2-4, Tokai-mura,
Naka-gun 319-1195, Ibaraki-ken, Japan
e-mail: ha.yoosung@jaea.go.jp
Search for other works by this author on:
Tohru Tobita,
Tohru Tobita
Nuclear Safety Research Center,
Japan Atomic Energy Agency,
Shirakata 2-4, Tokai-mura,
Naka-gun 319-1195, Ibaraki-ken, Japan
e-mail: tobita.tohru@jaea.go.jp
Japan Atomic Energy Agency,
Shirakata 2-4, Tokai-mura,
Naka-gun 319-1195, Ibaraki-ken, Japan
e-mail: tobita.tohru@jaea.go.jp
Search for other works by this author on:
Takuyo Ohtsu,
Takuyo Ohtsu
Nuclear Safety Research Center,
Japan Atomic Energy Agency,
Shirakata 2-4, Tokai-mura,
Naka-gun 319-1195, Ibaraki-ken, Japan
e-mail: ohtsu.takuyo@jaea.go.jp
Japan Atomic Energy Agency,
Shirakata 2-4, Tokai-mura,
Naka-gun 319-1195, Ibaraki-ken, Japan
e-mail: ohtsu.takuyo@jaea.go.jp
Search for other works by this author on:
Hisashi Takamizawa,
Hisashi Takamizawa
Nuclear Safety Research Center,
Japan Atomic Energy Agency,
Shirakata 2-4, Tokai-mura,
Naka-gun 319-1195, Ibaraki-ken, Japan
e-mail: takamizawa.hisashi@jaea.go.jp
Japan Atomic Energy Agency,
Shirakata 2-4, Tokai-mura,
Naka-gun 319-1195, Ibaraki-ken, Japan
e-mail: takamizawa.hisashi@jaea.go.jp
Search for other works by this author on:
Yutaka Nishiyama
Yutaka Nishiyama
Nuclear Safety Research Center,
Japan Atomic Energy Agency,
Shirakata 2-4, Tokai-mura,
Naka-gun 319-1195, Ibaraki-ken, Japan
e-mail: nishiyama.yutaka93@jaea.go.jp
Japan Atomic Energy Agency,
Shirakata 2-4, Tokai-mura,
Naka-gun 319-1195, Ibaraki-ken, Japan
e-mail: nishiyama.yutaka93@jaea.go.jp
Search for other works by this author on:
Yoosung Ha
Nuclear Safety Research Center,
Japan Atomic Energy Agency,
Shirakata 2-4, Tokai-mura,
Naka-gun 319-1195, Ibaraki-ken, Japan
e-mail: ha.yoosung@jaea.go.jp
Japan Atomic Energy Agency,
Shirakata 2-4, Tokai-mura,
Naka-gun 319-1195, Ibaraki-ken, Japan
e-mail: ha.yoosung@jaea.go.jp
Tohru Tobita
Nuclear Safety Research Center,
Japan Atomic Energy Agency,
Shirakata 2-4, Tokai-mura,
Naka-gun 319-1195, Ibaraki-ken, Japan
e-mail: tobita.tohru@jaea.go.jp
Japan Atomic Energy Agency,
Shirakata 2-4, Tokai-mura,
Naka-gun 319-1195, Ibaraki-ken, Japan
e-mail: tobita.tohru@jaea.go.jp
Takuyo Ohtsu
Nuclear Safety Research Center,
Japan Atomic Energy Agency,
Shirakata 2-4, Tokai-mura,
Naka-gun 319-1195, Ibaraki-ken, Japan
e-mail: ohtsu.takuyo@jaea.go.jp
Japan Atomic Energy Agency,
Shirakata 2-4, Tokai-mura,
Naka-gun 319-1195, Ibaraki-ken, Japan
e-mail: ohtsu.takuyo@jaea.go.jp
Hisashi Takamizawa
Nuclear Safety Research Center,
Japan Atomic Energy Agency,
Shirakata 2-4, Tokai-mura,
Naka-gun 319-1195, Ibaraki-ken, Japan
e-mail: takamizawa.hisashi@jaea.go.jp
Japan Atomic Energy Agency,
Shirakata 2-4, Tokai-mura,
Naka-gun 319-1195, Ibaraki-ken, Japan
e-mail: takamizawa.hisashi@jaea.go.jp
Yutaka Nishiyama
Nuclear Safety Research Center,
Japan Atomic Energy Agency,
Shirakata 2-4, Tokai-mura,
Naka-gun 319-1195, Ibaraki-ken, Japan
e-mail: nishiyama.yutaka93@jaea.go.jp
Japan Atomic Energy Agency,
Shirakata 2-4, Tokai-mura,
Naka-gun 319-1195, Ibaraki-ken, Japan
e-mail: nishiyama.yutaka93@jaea.go.jp
Contributed by the Pressure Vessel and Piping Division of ASME for publication in the JOURNAL OF PRESSURE VESSEL TECHNOLOGY. Manuscript received February 8, 2018; final manuscript received June 14, 2018; published online August 2, 2018. Assoc. Editor: Yun-Jae Kim.
J. Pressure Vessel Technol. Oct 2018, 140(5): 051402 (6 pages)
Published Online: August 2, 2018
Article history
Received:
February 8, 2018
Revised:
June 14, 2018
Citation
Ha, Y., Tobita, T., Ohtsu, T., Takamizawa, H., and Nishiyama, Y. (August 2, 2018). "Applicability of Miniature Compact Tension Specimens for Fracture Toughness Evaluation of Highly Neutron Irradiated Reactor Pressure Vessel Steels." ASME. J. Pressure Vessel Technol. October 2018; 140(5): 051402. https://doi.org/10.1115/1.4040642
Download citation file:
Get Email Alerts
Cited By
Constraint Effect and Crack-Tip Plastic Zone of the Pipe With Internal Inclined Surface Cracks Under External Pressure
J. Pressure Vessel Technol (April 2025)
Research on the Accelerated Life Test Method for Gaskets and the Verification of the Accuracy of the Life Prediction
J. Pressure Vessel Technol (June 2025)
Optimization of High-Vapor Pressure Condensate Pipeline Commissioning Schemes in Large Uplift Environments
J. Pressure Vessel Technol (June 2025)
Technical Basis for Revising the Fatigue Crack Growth Rates for Ferritic Steels in the ASME Code Section XI
J. Pressure Vessel Technol (June 2025)
Related Articles
Fracture Safety Analysis Concepts for Nuclear Pressure Vessels, Considering the Effects of Irradiation
J. Basic Eng (June,1971)
Applicability of the ASME Exemption Curve for Chinese Pressure Vessel Steel Q345R
J. Pressure Vessel Technol (December,2015)
An Investigation of the Structural Integrity of a Reactor Pressure Vessel Using Three-Dimensional Computational Fluid Dynamics and Finite Element Method Based Probabilistic Pressurized Thermal Shock Analysis for Optimizing Maintenance Strategy
J. Pressure Vessel Technol (October,2018)
Fracture Toughness Evaluation of Reactor Pressure Vessel Steels by Master Curve Method Using Miniature Compact Tension Specimens
J. Pressure Vessel Technol (October,2015)
Related Proceedings Papers
Related Chapters
Investigation of Some Problems In Developing Standards for Precracked Charpy Slow Bend Tests
Developments in Fracture Mechanics Test Methods Standardization
Lessons Learned: NRC Experience
Continuing and Changing Priorities of the ASME Boiler & Pressure Vessel Codes and Standards
Czech and Slovakian Codes
Companion Guide to the ASME Boiler and Pressure Vessel Code, Volume 3, Third Edition