Skip to Main Content
Skip Nav Destination
ASTM Selected Technical Papers
Effects of Radiation on Structural Materials
By
JA Sprague
JA Sprague
editor
1
U.S. Naval Research Laboratory
,
Washington, D.C. 20375
;
symposium chairman and editor
.
Search for other works by this author on:
David Kramer
David Kramer
editor
2
Atomics International
USA
Search for other works by this author on:
ISBN-10:
0-8031-0327-1
ISBN:
978-0-8031-0327-6
No. of Pages:
694
Publisher:
ASTM International
Publication date:
1979

Recent experiments on developmental fast-flux test facility (FFTF) cladding (20 percent cold-worked Type 316 stainless steel) have extended the data base to a fast neutron fluence of 8.4 × 1022 neutrons (n)/cm2 (E > 0.1 MeV). The specimens were irradiated in the experimental breeder reactor-II (EBR-II) at temperatures ranging from 371 to 816°C, although peak fluence levels were attained on specimens irradiated near 371 and 649°C only. Tension tests were performed at 232°C, near the irradiation temperature, and, in some cases, above the irradiation temperature. Test specimen strain rates ranged from 4 × 10 -5/s to 4 × 10-2/s.

The data generated on cladding irradiated near 371 °C established that the low-temperature strength and ductility are fluence independent beyond about 5 × 1022 n/cm2 (E > 0.1 MeV). The strength behavior of the irradiated cladding at 538, 593, and 649 °C is essentially the same as exhibited by thermally aged developmental cladding at the same temperatures and times out of the reactor. Up to a fluence of ∼5 × 1022 n/cm2 (E > 0.1 MeV), the 538°C ductility values remain relatively fluence independent after an initial decrease. Higher temperature (593 and 649 °C) ductilities decrease continually with increasing fluence.

Tensile parameter correlations were developed for the prediction of irradiation effects on the tensile properties of 20 percent cold-worked Type 316 stainless steel. These correlations are based on unirradiated tensile property correlations developed using Hart's equation-of-state approach. The basic premise is that the condition of plastic deformation of some materials such as 316 stainless steel can be characterized by a structure parameter (σ*) which describes the material's “hardness.” It is found that irradiation effects can be incorporated into this formulation by parameterizing the changes in σ* with irradiation temperature and fluence. The resulting correlations provide a description of strength and ductility over the temperature range of 371 to 871 °C and strain rates of 10-5 to 10 1/s.

1.
Kawasaki
,
S.
 et al
,
Journal Atomic Energy Society of Japan
, Vol.
14
, No.
6
,
1972
, pp. 283–289.
2.
Fahr
,
D.
,
Bloom
,
E. E.
, and
Stiegler
,
J. O.
in
Proceedings
, British Nuclear Energy Society Conference on Irradiation Creep and Embrittlement,
1973
, p. 167.
3.
Garr
,
K. R.
,
Pard
,
A. G.
, and
Kramer
,
D.
in
Properties of Reactor Structural Alloys After Neutron or Particle Irradiation, ASTM STP 570
,
American Society for Testing and Materials
,
1975
, pp. 143–155.
4.
Fish
,
R. L.
and
Watrous
,
J. D.
,
Irradiation Efects on the Microstructure and Properties of Metal, ASTM STP 611
,
American Society for Testing and Materials
,
1976
, pp. 91–100.
5.
Garr
,
K. R.
and
Pard
,
A. G.
Irradiation Effects on the Microstructure and Properties of Metals. ASTM STP 611
,
American Society for Testing and Materials
,
1976
, pp. 72–90.
6.
Hart
,
E. W.
,
Li
,
C. Y.
,
Yamada
,
H.
, and
Wire
,
G. L.
in
Constitutive Equations in Plasticity
,
Argon
A. S.
, Ed.,
The MIT Press
,
Cambridge, Mass.
,
1975
.
7.
Hart
,
E. W.
,
Acta Metallurgica
 0001-6160, Vol.
18
,
1970
, p. 599.
8.
Wire
,
G. L.
,
Cannon
,
N. S.
, and
Johnson
,
G. D.
, “
Prediction of Transient Mechanical Response of Type 316 Stainless Steel Cladding Using an Equation of State Approach
,” to be published.
9.
Steichen
,
J. M.
, “
High Strain Rate Tensile Properties of 20% C.W. Type 316 Stainless Steel
,” HEDL-TME 74-39,
Hanford Engineering Development Laboratory
, Richland, Wash.,
06
1974
.
10.
Paxton
,
M. M.
, “
Mechanical Properties of Annealed and Cold Worked Type 316 Stainless Steel Fast Reactor Seamless Tubing
,” HEDL-TME 74-11,
Hanford Engineering Development Laboratory
, Richland, Wash.,
02
1974
.
11.
Nuclear Systems Material Handbook
, “
Effect of Strain Rate of 20 C.W. Type 316 SS
,” Vol.
2
, Property Code 2301 (E-1), Rev. 0, 12-22-75.
This content is only available via PDF.
You do not currently have access to this chapter.
Close Modal

or Create an Account

Close Modal
Close Modal