J-R curve characterization of irradiated low-shelf nuclear vessel steels
Conference
·
· Am. Soc. Test. Mater., Spec. Tech. Publ.; (United States)
OSTI ID:6410156
The J-R curve behavior of irradiated nuclear pressure vessel steels is characterized in the ductile upper-shelf regime in order to provide a materials basis with which to assess the margin of safety against fracture for water reactor vessels exhibiting a low upper-shelf C /SUB v/ energy. With the single specimen compliance technique, the R-curve is shown to follow a power-law behavior for small crack extension, and this phenomenon has led to a proposed new indexing procedure for J /SUB Ic/. In addition, a specimen size dependence of the R-curve has been suggested by the results of similar compact tension specimens up to 100 mm thick and which have sufficiently deep side grooves to produce a straight crack-front extension. R-curve data are presented in terms of a J versus T instability diagram which couples material and structural parameters, thereby permitting an analysis to be made of the margin against failure in terms of J. Also, a correlation has been suggested between the R-curve parameters and C /SUB v/ shelf energy; this could enhance the structural significance of C /SUB v/ reactor surveillance data.
- Research Organization:
- Materials Eng. Assoc. Inc., Lanham, MD
- OSTI ID:
- 6410156
- Report Number(s):
- CONF-8110313-
- Conference Information:
- Journal Name: Am. Soc. Test. Mater., Spec. Tech. Publ.; (United States) Journal Volume: STP-803
- Country of Publication:
- United States
- Language:
- English
Similar Records
J-R curve characterization of irradiated low upper shelf welds
Extrapolation of the J-R curve for predicting reactor vessel integrity
Extrapolation of the J-R curve for predicting reactor vessel integrity
Technical Report
·
Sat Mar 31 23:00:00 EST 1984
·
OSTI ID:6899508
Extrapolation of the J-R curve for predicting reactor vessel integrity
Technical Report
·
Tue Dec 31 23:00:00 EST 1991
·
OSTI ID:5764532
Extrapolation of the J-R curve for predicting reactor vessel integrity
Technical Report
·
Tue Dec 31 19:00:00 EST 1991
·
OSTI ID:10121770
Related Subjects
22 GENERAL STUDIES OF NUCLEAR REACTORS
220200* -- Nuclear Reactor Technology-- Components & Accessories
36 MATERIALS SCIENCE
360103 -- Metals & Alloys-- Mechanical Properties
360106 -- Metals & Alloys-- Radiation Effects
BENCHMARKS
CONTAINERS
CRACK PROPAGATION
DIAGRAMS
FRACTURE MECHANICS
FRACTURE PROPERTIES
INSPECTION
IRRADIATION
LIMITING VALUES
MATERIALS TESTING
MECHANICAL PROPERTIES
MECHANICAL TESTS
MECHANICS
PHYSICAL RADIATION EFFECTS
PRESSURE VESSELS
RADIATION EFFECTS
TESTING
220200* -- Nuclear Reactor Technology-- Components & Accessories
36 MATERIALS SCIENCE
360103 -- Metals & Alloys-- Mechanical Properties
360106 -- Metals & Alloys-- Radiation Effects
BENCHMARKS
CONTAINERS
CRACK PROPAGATION
DIAGRAMS
FRACTURE MECHANICS
FRACTURE PROPERTIES
INSPECTION
IRRADIATION
LIMITING VALUES
MATERIALS TESTING
MECHANICAL PROPERTIES
MECHANICAL TESTS
MECHANICS
PHYSICAL RADIATION EFFECTS
PRESSURE VESSELS
RADIATION EFFECTS
TESTING