Development of Mini-Compact Tension Test Method for Determining Fracture Toughness Master Curves for Reactor Pressure Vessel Steels

The Master Curve Approach for assessing fracture toughness of reactor pressure vessel (RPV) steels has been accepted throughout the world. The Master Curve Approach using fracture toughness data obtained from RPV steels in Japan has been investigated in order to incorporate this approach into the Japanese Electric Association (JEA) Code 4206, “Method of Verification Tests of the Fracture Toughness for Nuclear Power Plant Components”. This paper presents the applicability of the Master Curve Approach for Japanese RPV steels.

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JIC Fracture Toughness of Nuclear Reactor Pressure Vessel Steels in Transition and Upper Shelf Regions

Tetsu-to-Hagane ◽

10.2355/tetsutohagane1955.68.8_1032 ◽

1982 ◽

Vol 68 (8) ◽

pp. 1032-1039 ◽

Cited By ~ 1

Author(s):

Tsuneo KODAIRA ◽

Nobuya NAKAJIMA ◽

Masakatsu MATSUMOTO ◽

Kiyoshi FUKAYA

Keyword(s):

Fracture Toughness ◽

Pressure Vessel ◽

Nuclear Reactor ◽

Reactor Pressure Vessel ◽

Pressure Vessel Steels ◽

Reactor Pressure

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DEVELOPMENT OF TEST PROCEDURES. FRACTURE TOUGHNESS OF REACTOR PRESSURE VESSEL STEELS. Final Report.

10.2172/4788745 ◽

1969 ◽

Author(s):

P.B. Crosley ◽

E.J. Ripling

Keyword(s):

Fracture Toughness ◽

Pressure Vessel ◽

Reactor Pressure Vessel ◽

Final Report ◽

Test Procedures ◽

Pressure Vessel Steels ◽

Reactor Pressure

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Crack Arrest Test Results of Unirradiated and Irradiated German RPV Steels

Volume 6A: Materials and Fabrication ◽

10.1115/pvp2013-97954 ◽

2013 ◽

Author(s):

Florian Obermeier ◽

Julia Barthelmes ◽

Elisabeth Keim ◽

Hieronymus Hein ◽

Hilmar Schnabel ◽

...

Keyword(s):

Fracture Toughness ◽

Lower Bound ◽

Pressure Vessel ◽

Charpy Impact ◽

Crack Arrest ◽

Reactor Pressure Vessel ◽

Test Method ◽

Fracture Toughness Test ◽

Test Results ◽

Reactor Pressure

In the CARISMA[1] and CARINA[2] projects comprehensive tensile, Charpy-impact and fracture toughness tests were performed for unirradiated and irradiated original reactor pressure vessel (RPV) steel specimens from German pressurized water reactors (PWR) up to neutron fluences in the range of 60 operational years and beyond. In addition, crack arrest fracture toughness tests were performed to demonstrate the crack arrest behavior of the materials. To determine the crack arrest properties of ferritic steels, the designated test method according to ASTM E1221 [3] was used. However, in particular for irradiated reactor pressure vessel materials with higher irradiation embrittlement, the prescribed standard test specimen does not always provide adequate test results. During starter notch preparation annealing effects occurred in the heat affected zone (HAZ) of the brittle weld of the starter notch causing crack arrest in the HAZ after unstable crack initiation. Therefore a modified test method to perform crack arrest tests with so called duplex specimens was investigated. In this paper this modified method and the test results of five base and four weld metals with a fluence up to 4,69E+19 cm−2 (E >1 MeV) are discussed. The available test results show that the duplex specimen is an appropriate alternative to the standard compact crack arrest (CCA) specimen. The measured KIa fracture toughness data are enveloped by the “lower bound” of the ASME KIa-curve indexed with RTNDTj or TKIa but not all data are enveloped by indexing the “lower bound” curve with RTT0 like described in the ASME Code Case N-629 [4]. Furthermore correlations of the crack arrest test results with Charpy-impact and fracture toughness test results will be shown.

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