Abstract
The Ramberg-Osgood strain-hardening exponents and coefficients are characterized for an unirradiated ASTM A302-B steel over a wide range of temperatures from −129 to 260°C. The strain-hardening exponent increases only slightly with temperature over this range, while the coefficient decreases with increasing temperature. Tensile specimens irradiated to 0.002, 0.029, and 0.046 dpa exhibited significant increases in the strain-hardening exponent with increasing neutron irradiation level.
Issue Section:
Research Papers
1.
Bloom, J. M., and Malik, S. N., 1982, “A Procedure
for the Assessment of Integrity of Nuclear Pressure Vessels and Piping
Containing Defects,” EPRI NP-2431, Electric Power Research
Institute.
2.
Chakravarty
J.
K.
Wadekar
S.
L.
Sinha
T.
K.
Asundi
M.
K.
1983
,
“Dynamic Strain-Ageing of A203D Nuclear Structural
Steel
,” Journal of Nuclear Materials
, Vol.
119
, pp.
51
–58
.3.
James
L.
A.
1994
,
“The Effect of Temperature and Cyclic Frequency Upon Fatigue
Crack Growth Behavior of Several Steels in an Elevated Temperature Aqueous
Environment
,” ASME JOURNAL OF PRESSURE VESSEL
TECHNOLOGY
, Vol. 116
, May, pp.
122
–127
.4.
Jung, C. W., and Murty, K. L., 1988, “Effect of
Temperature and Strain Rate on Upper Shelf Fracture Behavior of A533B Class 1
Pressure Vessel Steel,” ASTM STP 969, pp. 392–401.
5.
Kang
S.
S.
Kim
I.
S.
1992
,
“Dynamic Strain-Aging Effect of Fracture Toughness of Vessel
Steels
,” Nuclear Technology
, Vol.
97
, No. 3
, pp.
336
–343
.6.
Kirk
M.
T.
Dodds
R.
H.
1993
,
“J and CTOD Estimation Equations for Shallow Cracks in Single
Edge Notch Specimens
,” Journal of Testing and
Evaluation
, Vol. 21
, No. 4
, pp.
228
–238
.7.
Kumar, V., German, M. D., and Shih, C. F., 1981, “An
Engineering Approach for Elastic-Plastic Fracture Analysis,” EPRI NP-1931,
Electric Power Research Institute.
8.
Little, E. A., and Hudson, J. A., 1986, “Role of
Metallurgical Parameters in the Dynamic Strain Aging of A533B Nuclear Steel,”
Proceedings, Second International Symposium on Environmental
Degradation of Materials in Nuclear Power Systems—Water Reactors,
ANS, pp. 640–644.
9.
Marshall, C. W., Landow, M. P., and Wilkowski, G. M.,
1990, “Effect of Dynamic Strain Aging on Fracture Resistance of Carbon Steels
Operating at Light-Water Reactor Temperatures,” ASTM STP 1074, pp.
339–360.
10.
Miglin, M. T., VanDerSluys, W. A., and Futato, R. J.,
1985, “Effects of Strain Aging in the Unloading Compliance J Test,” ASTM STP
856, pp. 150–165.
11.
Pelli, R., To¨rro¨nen, K., Salonen, S., and Rahka,
K., 1978, “Strain Ageing of Nuclear Pressure Vessel Steel A533B and A508CL2,”
Time and Load Dependent Degradation of Pressure Boundary
Materials, IWG-RRPC-79-2, International Atomic Energy Agency, pp.
182–189.
12.
Ramberg, W., and Osgood, W. R., 1943, “Description of
Stress Strain Curves by Three Parameters,” Technical Note 902, National Advisory
Committee on Aeronautics.
13.
Seman, D. J., Kallenberg, G. P., and Towner, R. J.,
1971, “Fracture Toughness of Low-Strength Steels,” WAPD-TM-895 (available from
the National Technical Information Service).
14.
Shure
K.
1976
, “Updated Measure of Radiation Damage
Exposure
,” Nuclear Technology
, Vol.
31
, No. 3
, pp.
375
–384
.15.
Stofanak, R. J., Poskie, T. J., Li, Y. Y., and Wire,
G. L., 1993, “Irradiation Damage Behavior of Low Alloy Steels Wrought and Weld
Metals,” Proceedings, Sixth International Symposium on Environmental
Degradation of Materials in Nuclear Power Systems—Water Reactors,
TMS, pp. 757–764.
16.
Wilkening
W.
W.
deLorenzi
H.
G.
Barishpolsky
M.
1984
, “Elastic-Plastic Analyses of Surface Flaws
in a Reactor Vessel
,” ASME JOURNAL OF PRESSURE
VESSEL TECHNOLOGY
, Vol. 106
, Aug., pp.
247
–254
.17.
Yoon
K.
K.
Bloom
J.
M.
Pavinich
W.
A.
Slager
H.
W.
1985
,
“Application of the Failure Assessment Diagram to the
Evaluation of Pressure-Temperature Limits for a Pressurized Water
Reactor
,” ASME JOURNAL OF PRESSURE VESSEL
TECHNOLOGY
, Vol. 107
, May, pp.
192
–196
.
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