Study on underclad cracking in nuclear reactor vessel steels

Journal Article · Fri Feb 01 00:00:00 EST 1985 · J. Pressure Vessel Technol.; (United States)

DOI:https://doi.org/10.1115/1.3264400· OSTI ID:5903426

Horiya, T; Takeda, T; Yamato, K

Susceptibility to underclad cracking in nuclear reactor vessel steels, such as SA533 Grade B Class 1 and SA508 Class 2, was studied in detail. A convenient simulation test method using simulated HAZ specimens of small size has been developed for quantitative evaluation of susceptibility to underclad cracks. The method can predict precisely the cracking behavior in weldments of steels with relative low crack susceptibility. The effect of chemical compositions on susceptibility to the cracking was examined systematically using the developed simulation test method and the following index was obtained from the test results: U = 20(V) + 7(C) + 4(Mo) + (Cr) + (Cu) - 0.5(Mn) + 1.5 log(X) X = Al . . . Al/2N less than or equal to 1 X = 2N . . . Al/2N > 1 It was confirmed that the new index (U) is useful for the prediction of crack susceptibility of the nuclear vessel steels; i.e., no crack initiation is detected in weldments in the roller bend test for steels having U value below 0.90.

• Research Organization: Nippon Steel Corporation, Kawasaki, Kanagawa
• OSTI ID: 5903426
• Journal Information: J. Pressure Vessel Technol.; (United States), Vol. 107:1
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
FRACTURES
REACTOR SAFETY
PRESSURE VESSELS
FRACTURE PROPERTIES
REACTOR VESSELS
CRACK PROPAGATION
DEFECTS
DESTRUCTIVE TESTING
FORECASTING
HEAT AFFECTED ZONE
MATERIALS TESTING
POWER REACTORS
STEEL-ASTM-A508
STEEL-ASTM-A533
WELDED JOINTS
ALLOYS
CONTAINERS
FAILURES
IRON ALLOYS
IRON BASE ALLOYS
JOINTS
MECHANICAL PROPERTIES
REACTORS
SAFETY
STEELS
TESTING
ZONES
220900* - Nuclear Reactor Technology- Reactor Safety