Correlation between cleavage fracture toughness and charpy impact properties in the transition temperature range of reactor pressure vessel steels

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.

Radionuclides release from re-irradiated fuel under high temperature and pressure conditions; -ray measurements of VEGA-5 test

Hidaka, Akihide; Kudo, Tamotsu; Nakamura, Takehiko; Kanazawa, Toru; Kiuchi, Toshio; Uetsuka, Hiroshi

JAERI-Tech 2003-009, 30 Pages, 2003/03

JAERI-Tech-2003-009.pdf:1.73MB

The VEGA (Verification Experiments of radionuclides Gas/Aerosol release) program is being performed at JAERI to clarify mechanisms of radionuclides release from irradiated fuel during severe accidents and to improve source term predictability. The fifth VEGA-5 test was conducted in January 2002 to confirm the reproducibility of decrease in cesium release under elevated pressure that was observed in the VEGA-2 test and to investigate the release behavior of short-life radionuclides. The PWR fuel of 47GWd/tU after 8.2 years of cooling was re-irradiated at Nuclear Safety Research Reactor (NSRR) for 8 hours before the heat-up test. After that, the two pellets of 10.9g without cladding were heated up to about 2,900K at 1.0MPa under the inert He condition. The experiment reconfirmed the decrease in cesium release under elevated pressure. The release data on short-life radionuclides such as Ru-103 and Ba-140 that has never been observed in the previous VEGA tests without re-irradiation was obtained using the gamma ray measurement.

Conceptual design study of accelerator-driven systems for nuclear waste transmutation

Takizuka, Takakazu; ; Sasa, Toshinobu

Proc. of 2nd Int. Conf. on Accelerator-Driven Transmutation Technologies and Applications, 1, p.179 - 185, 1996/06

no abstracts in English

Data report for ROSA-IV/LSTF 5% hot leg break experiment Run SB-HL-01

Kukita, Yutaka; ; ; Anoda, Yoshinari; Kumamaru, Hiroshige; Suzuki, Mitsuhiro; ; Yamamoto, Nobuo; Nakamura, Hideo; Yonomoto, Taisuke; et al.

JAERI-M 89-225, 117 Pages, 1990/01

JAERI-M-89-225.pdf:3.53MB

no abstracts in English

Recirculation pump suction line 75 and 25% split break LOCA tests of ROSA-III; Runs 929 and 930 with HPCS failure

Nakamura, Hideo; Tasaka, Kanji; Suzuki, Mitsuhiro; Anoda, Yoshinari; Kumamaru, Hiroshige; Yonomoto, Taisuke; Murata, Hideo

JAERI-M 89-131, 260 Pages, 1989/09

JAERI-M-89-131.pdf:6.11MB

no abstracts in English

BWR LOCA Lintegral Test Simulating a 100% Main Steam Line Break Outside Reactor Containment Vessel in ROSA-III Program, Rum 955; Analogy of Steam Line Break and Recirculation Line Small Break

; ; ; ; ; Tasaka, Kanji

JAERI-M 87-044, 151 Pages, 1987/03

JAERI-M-87-044.pdf:3.76MB

no abstracts in English

BWR 200% recirculation pump suction line break LOCA tests,Runs 942 and 943 at ROSA-III without HPCS; Effects of initial fluid conditions on LOCA

; Tasaka, Kanji; ; ; ; ; ; Koizumi, Yasuo

JAERI-M 86-038, 275 Pages, 1986/03

JAERI-M-86-038.pdf:6.4MB

no abstracts in English

BWR Main Steam Line Break LOCA Tests Runs 951,954 and 956 at ROSA-III; Break Area Effects with HPCS Failure-

; Tasaka, Kanji; ; ; ; ; ; ; Koizumi, Yasuo

JAERI-M 85-202, 295 Pages, 1985/12

JAERI-M-85-202.pdf:7.72MB

no abstracts in English

Recirculation pump discharge line break test at ROSA-III for a boiling water reactor

; Tasaka, Kanji; ; ; ;

Nuclear Technology, 70, p.189 - 203, 1985/00

https://doi.org/10.13182/NT85-A33643

no abstracts in English

A Main steam line break experiment at ROSA-III-Run 952 (Standard run with full ECCS)

; ; ; Tasaka, Kanji; ; ; ; ;

JAERI-M 84-229, 153 Pages, 1984/12

JAERI-M-84-229.pdf:4.01MB

no abstracts in English