Stress relief cracking in nuclear pressure vessel steels. Final report

Technical Report · Sun Jun 01 04:00:00 EDT 1986

OSTI ID:5772413

McMahon, Jr, C J; Gentner, D H

Specimens from 33 heats of three types of nuclear pressure vessel steel were heat treated to produce a simulated weld HAZ microstructure and tested to determine their resistance to stress relief cracking (SRC). The method employed a special jig in which notched specimens were loaded in 3-point bending to a fixed displacement and then heated to 610/sup 0/C for 6 h in a simulated stress relief treatment. The susceptibility to SRC was measured in terms of the extent of cracking as a function of the notch-opening displacement. The results showed that the forging grade A508-2 was generically much more susceptible to SRC than either the plate grade A533-B or the forging grade A508-3. The four tested heats of the latter steel showed almost no SRC susceptibility within the limitations of these tests. In the other two steels wide heat-to-heat variation in SRC susceptibility was found. 20 refs., 24 figs.

• Research Organization: Pennsylvania Univ., Philadelphia (USA). Dept. of Materials Science and Engineering
• OSTI ID: 5772413
• Report Number(s): EPRI-NP-4627; ON: TI86920335
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
220200* -- Nuclear Reactor Technology-- Components & Accessories
36 MATERIALS SCIENCE
360103 -- Metals & Alloys-- Mechanical Properties
ALLOYS
CONTAINERS
CRACK PROPAGATION
CREEP
CRYSTAL STRUCTURE
IRON ALLOYS
IRON BASE ALLOYS
MATERIALS
MATERIALS TESTING
MECHANICAL PROPERTIES
MICROSTRUCTURE
NUCLEAR FACILITIES
NUCLEAR POWER PLANTS
POWER PLANTS
PRESSURE VESSELS
REACTOR MATERIALS
STEELS
STRESSES
TESTING
THERMAL POWER PLANTS