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ORNL P
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MICROCOPY RESOLUTION TEST CHART
NATIONAL QURI:AU OF STANDAROS - 1963
ORNUP 3
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Civil 3: PATIO
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NOV 1 5 1985
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ma
hata k

imati na
.. BASUD FOR ANNOUNCEMENT
I MICITAR SCIECE ABSTRACTS
bride tara limpen Day the
1..
IRRADIATION EFFECTS IN STAINLESS STHELS AT HIGH TEMPERATURES
meona duket the
i
W. R. Martin and J. R. Weir, Jr.
Metals and Ceramics Division
Oak Ridge National Laboratory
Oak Ridge, Tennessee .
n
Contract No. ,W-7405-eng-26
pod mene ihan itu membuat membeli meminta maana
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IRRADIATION EFFECTS IN STAINLESS STBELS AT HIGH TEMPERATURES
-- --
W. R. Martin and J. R. Weir, Jr.
-
ABSTRACT
Milli!T
... L'Eri!:
. ieft ;9.17 13:14
•
The effect of irradiation on the elevated-temperature mechanical .
properties of structural materials is described using type 304 stainless
steel as an example. The general result is that the grain boundary
fracture process, but not the deformation process, is affected. All i
data suggest the primary cause to be helium generated from (na)
reactions. Several metallurgical techniques for improving the
ductilities of irradiated alloys are, suggested, and experimental data
are given for type 304 stainless steel for which the degree of improve- ,
ment is demonstrated.
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IRRADIATION EFFECTS IN STAINLESS STEELS AT HIGH TEMPERATURES*
W. R. Martin and J. R. Weir, Jr.
INTRODUCTION
..11
"17**.8!!
· The effect of irradiation on the stress-strain behavior of
structural austenitic stainless steels+ tested above 500°C is different
from that observed for the alloys tested below 500°C, as illustrated in
mig. 1. Irradiating an alloy below 500°C produces a structure with a
higher yield stress, a lower strain-hardening coefficient (and hence a
lower uniform strain), and lower ductility in terms of total elongation
than the unirradiated alloy when tested below 500°C. In many instances,
no change in true fructure strain or stress is observed for stainless
steels irradiated to less than 1 x 102 neutrons/cm.
Above 500°C the effect of irradiation on these materials is dif- .
ferent; first, there are reductions in ductility in terms of uniform and
fracture strains as well as total elongation, and the magnitude of
these reductions is highly strain-rate sensitive. Generally, changes
in yield stress and in the stress-strain relationship are not observed.
Since the ductility is reduced, the ultimate tensile strengths and
fracture strengths are also reduced, and the magnitude of this reduction
is very sensitive to the work hardening capabilities of the alloy. The
effects are schematically illustrated in Fig. 2. They are first noted
at an homologous temperature of about 0.5, which represents the approxi-
mate temperature at which self diffusion becomes important. and high-
temperature creep becomes a consideration for reactor design. ...
1.14
1.
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P.
Lt.
#Research sponsored by the U.S. Atomic Energy Commission under
contract with the Union Carbide. Corporation.
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ELEVATED-TEMPERATURE EMBRIT 'LEMENT BY HELIUM
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There are two very significant differences between the irradiation
effects below and above 500°C. First, the radiation damage affecting
the low-temperature properties can be eliminated by postirradiation
heat treatment. Much of the damage is removed in situ if the irradiatio
temperature is above approximately 350°C. However, postirradiation
annealing of iron- and nickel-base alloys at temperatures between 1050
and 1300°C does not remove the damage effecting the low ductility at
elevated temperatures. The ti:ermal stability of the defect causing ..
the low ductility has been illustrated previously. A second signifi.
cant difference is the dependence upon the neutron energy or the energy:
spectrum of the rieutron flux. The low-temperature damage is due to the
production of vacancies and interstitials by fast neutrons. The thermal
neutrons contribute only a small fraction of the damage. However, the . ;
magnitude of the elevated-temperature embrittlement correlates best with
thermal neutrons4 for normal iron- and nickel-base alloys. Table 1
Asli :illustrates this for two stainless steels.
:
1?14:HT
MA!:;IN
20
·0051:1:13
.!,1817:01
Table 1. Dose Dependence of High-Temperature
Ductility in Irradiated Stainless Steels
Experi-
ment
Dora, neutrons/cm2
Thermal Fission
Neutrons Neutrons
Elonga-
tion
(%)
o
48
x 1018
1.9 x 1016
2.2 x
1.5 x 20
1. Do at
요요요
​хохххо
x x x
x 2017
o
1.6 x 1019
1.6 x 1020
7 x 1020
3 x 1019
3 x 1019
5 x 2020
20-25-Nb stainless steel tested at
750 °C by Roberts and Harries at Harwell.
bType 304 stainless steel tested at
840°C by Martin at ORNL.
.
imn.
Of all the thermal neutron reactions possible, the (n,a) reaction
with 20 B appears to be the one capable of producing a defect structure
with thermal stability previously noted. This reaction also produces a
13thiwn isotope, but the work of Higgins and Roberts' suggests that the
helium produced is more embrittling than the lithium. It was not until
recently6,7 that the (n,a) reactions with other elements and fast
neutrons were considered important (see Table 2). However, we emphasize
that helium (alpha particles) can be produced from the reaction of fast
neutrons with Fe, Ni, N, and other elements and that the quantities
a
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produced will be significant for normal engineering allovs i
a dose level above 1 x 1021 neutrons/cm?. We believe then that the
embrittlement at elevated temperatures is due to the production of helium
gas. The total amount of helium in the alloy can be calculated for a
given alloy composition and neutron dose. A correlation of helium con-
tent with short-time tensile ductilities is given in Fig. 3. This curve
illustrates that as the deformation temperature is increased the amount
of helium needed to initiate the embrittling process decreases. Curves
such as the one shown in Fig. 3 are typical for most structural . ..
materials. The LOB from which the majority of helium is generated in
these commercial heats is segregated in grain boundaries. Thus the
concentration of helium at the grain boundary can be several orders of
magnitude greater than the average value calculated and presented in .
Fig. 3. The actual amount is dependent upon grain size, preirradiation
heat treatment, etc.
1.1.
:
IN
Table 2. Effect of Boron Content on High-Temperature
Irradiation Damage to Type 304 Stainless Steel
*
i
"..!!!
Irradiation
Exposure
Elongation
Boron
Content . at at
(ppm)
704
842°C
Helium
Concentration, ppm
1on, PD
Fast
(n,a) (n,a)
108
Total
6
At high levels of
irradiation exposure
(approx 3 x 1020
neutrons/cm)
0.015
0.023
0.11
10.13
3.9
0.11
3.9
0.015
0.023
, 0.091
0.110
3.3
0.00004
0.00150
0.3 0.315
0.3 . 0.323
0.3 0.391
0.410
0.3: 3.6
0.00006 0.0001
0.00006 0.0015
13
.:39
At low levels of
irradiation exposure
(approx 5 x 1016
neutrons/cm) :
---
---
-
-
Enhancement of the Intergranular Fracture Process
The nature of the irradiation embrittlement is such that the
nucleation and rate of propagation of wedge-type grain boundary cracks
· are affected. This is illustrated in Fig. 4, which shows the variation
in number and length of grain boundary cracks.at fracture. For example,
the unirradiated material strained to 14% would not have any grain
.. boundary cracks visible at this magnification. The mechanism by which
helium affects the process of intergranular fracture is not known,
although some have been proposed.
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The irradiation effects noted thus far for stainless steels are
applicable for a number of structural materials. The variation in
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ductilities of irradiated alloys at elevated temperatures is given in ."
Table 3, illustrating that irradiation embrittlement. at elevated tempera- .:
tures is a general phenomenon and not one specific to one given class of .
alloys. Also, the effects described for short-time tensile tests are in :
general applicable to high-temperature creep.
..
Table 3. Postirradiation Ductility of Materials
Creep-Rupture
Ductility
at
700 °C 800°C
Tensile Ductility
at at
i 700°C 800 °C
Material
at
Reference
0.7–10.0
6-20
25.
5-ll.
1
0.2-0.5
4-20
1-11
9
10,111
2,11
2
178
12
.
110!17
NOIN
Inconel 600
Hastelloy N
Hastelloy X
Rene' 41
Nimonic PE16
A 286
304 stairless
stee'í
20 Cr-25 Ni-Nb
316 stainless
; steel
1-3
2
3
..
11-30
3-40
1-2
1-3
-09:20
1.9
4-30
15-20
:
0.4-10.0
14-22
13,12
12,14
0-10
Dose = 1020 to 1021 neutrons/cm2.
METHODS FOR IMPROVING THE DUCTILITY
Irradiation studies at the Oak Ridge National Laboratory with type
304 stainless steel indicate that the magnitude of irradiation embrittle-
ment can be reduced by several metallurgical techniques. . These methods
are believed to be successful because the quantity of helium per unit
length of the grain boundary is reduced and, in most cases, the stress :
necessary to nucleate and propagate a grain boundary crack is increased
A detailed discussion of the mechanism can be found elsewhere. 15,16
The influence of grain size on the ductility of irradiated stainless
steel is illustrated in Table 4. Grain boundary precipitates 26 are
important also. The effect of increased carbon content is shown in
Table 5. Note that at 900°C where all the carbon is in solid solution
for both the low and high-carbon steels the ductilities as
effect of carbon precipitate is illustrated for the higher carbon steel :
by the data in Table 6. These data show that aging at elevated temper&-.
tures to induce precipitation and spheroidization of carbides also
improves the postirradiation ductility. We believe that the spheroidim
:zation of grain boundary carbides may retard the propagation of grain
boundary cracks in much the same way, as proposed by Weaver. 17,18
TE
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le 4. Influence of Grain Size on
Irradiation Embrittlement
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ASTM
Grain
Size
Total Elongation, %
at 704°C
at 842°C
Unirradiated Irradiated Unirradiated Irradiated
30
23
36
.
g-10
39
58
និង
47
16
+
1
VAI
To improve the ductility of irradiated materials, one must devise ways
to reduce the helium concentration at the grain boundaries. It is only
the helium bubbles at the grain boundary that are believed to be
deleterious; that is, helium generated within the grains would not be
as harmful if these atoms were unable to move to the grain boundaries.
For many zeactor applications, the preponderance of helium generated is
18,
due to the transmutation of LB. Boron normally segregates to the
grain boundaries in the solid state and therefore a large quantity of **
helium is generated at these boundaries. If one could form a stable
boron compound, insoluble either in the melt or at a very high tempera-
ture efter solidification, it would be possible to get a homogeneous
distribution of this compound. Therefore, the helium generated would
stay at the precipitate-matrix interface, and the quantity at the grain
boundaries would be greatly reduced. The precipitate-matrix interface
would also serve as a depository for helium generated from other elements
and fast neutrons. Thus in principle this system should result in
material with a lower susceptibility to elevated-temperature embrittle-
ment in thermal .and fast neutron environments.
This approach for improving the ductility of irradiated alloys is
currently being investigated using the. 18-8 stainless steel. Among the
most stable borides in this stainless steel are those of titanium. We
have now accumulated data from two different irradiations, and typical
data are given in Table 7. It is observed that additions of approxi-
mately 0.3% Ti in type 304 stainless steel are beneficial. These recent
experiments illustrate that much can be done to improve the ductilities
of irradiated materials.
ANI TYRI!...:
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Table 5. Elevated-Temperature Ductility of Unirradiated and Irradiated.
AISI Types 304L' and 304 Stainless Steels.
True Uniform Strain', %
Deformation Irradiation
Total Elongation, %
Carbon
Temperature Temperature Content Unirradi Irradi Unirradi- Irradi-
(°C)
(wt %) • ated
ated
ated
ated
(°C)
650
700
700
800
22.5
20.9
23.6
22.2
15.1
21.8
12.1
20.5
10.5
42.2
36.8
39.8
41.7
800
11.4
.
.
i
704
16.9
28.5
17.6
29.9
11.3
12.4
13.1
23.0
13.3
20.3
6.4
704
(IF APPLICAULES
CLASSIFICATION
900
900
700
700
800
800
*** 900
900
700
700
800
800
900
900
11.2
17.1
11.2
14.1
23.7
17.7
17.8
18.0
14.0
17.8
*17.4
10.2
10:44
10.0
9.3
0.02
0.06
0.02
0.06
0.02
0.06
0.02
0.06
0.02
0.06
0.02
0.06
0.02
0.06
0.02
0.06
0.02
0.06
0.02
0.06
0.02
0.06
0.02
0.06
6.1
37.4
38.9.
41.5
36.2
36.6
31.4
28.0
33.0
40.7
30.0--
20.0
:;: 6.50.
fa 17lilir!
110!! 3.! 112;;'11
775
14.6
7.1
9.6
5.0
7.6
9.1
.-15.7
7.2
12.6
2.9
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31.7
4.0
nimi ori
842
700
700
14.1
9.8
7.0
10.4
10.4
10.2
10.8
6.7
10.6
2.0
800
38.4
31.9
36.8
39.0
24.5
25.6
5.8
800
900
900
8.8
2.6
2.2
3.0
Alloys irradiated to a neutron dose of 3.5 x 1020 neutrons/com? (E > 1 Mev) and
4.5 x 1020 neutrons/cm2 (thermal).
Specimens strained at a rate of 0.2% per min.:
.
.
:12
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Table 6. The Effect of Aging on the Postirradiation
Ductility of Type 304 Stainless Steel
Preirradiation
Heat Treatwent
Deformation
Temperature
(°C)
. Elongation
842
1 hr at 1036°C)
704
1 hr at 1036"C; 100 hr . 704
at 800 °C
1 hr at 1036 °C
1 hr at 1036°C; 100 hr 842
at 800°C
Abose = 1021 neutrons/cm2 (thermal).
Table 9. ^ Effect of Titanium Content on the
Postirradiation Ductility of Types 304
and 304L Stainless Steel
..
Element
(wt %)
Ductility
Irradiated. Irradiated
at 50°C at 700 °C
Titanium
Carbon
29
0.3
1.2
0.02
0.02
0.02
0.06
0.06
0.06
0.3
1.2
Tensile tests conducted at 842°C.
Dose = 1 x 1020 neutrons/cm?
Dose = 5 x 1020 neutrons/cm2..
.
SUMMARY
.
In summary, irradiation damage at elevated temperatures is an
"effect on the process of intergranular fracture, and thus the degree of .
embrittlement is highly sensitive to strain rate and temperature. The
*** embrittlement 18 apparently due to helium, generated from thermal (n,Q)
and fast (na) reactions. We have illustrated several techniques for
improving the ductility in irradiated stainless steel alloys and suggest
that/similar techniques should be applicable to all structural alloys.
:
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REM
.
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.:
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REFERENCES
1. 'W. R. Martin and J. R. Weir, Jr., Am. Soc. Testing Meter. Spec.
Tech. Publ. 380, 251 (1965).
2. F. C. Robertshaw, J. Moteff, F. D. Kingsburg, and M. A. Pugacz, .
Am. Soc. Testing Mater. Spec. Toch. Publ. 341, 372 (1963).
- 3. W. R. Martin and J. R. Weir, Jr., Nature 202(4936), 997 (1964).
4. A. C. Roberts and D. R. Harries, Nature 200(4908), 772 (1963).
5. P.R.B. Higgins and A. C. Roberts, Nature 206(4990), 1249 (1965).
6. S. H. Bush, J. Moteff, and J. R. Weir, Jr., "Radiation Damage in
Fast Reactor Components, " paper presented at the ANS Topical Meeting
on Fast Reactor Technology, Detroit, Michigan, April 26-28, 1965.
To be published in Proceedings.
7. W. R. Martin and J. R. Weir, Jr., "Irradiation Embrittlement of
Low-Boron Type 304 Stainless Steel." to be published in Nature.
R. S. Barnes, Some. Mechanisms of Radiation-Induced Mechanical
Property Changes, AERE Report R-4655 (July 1964).
9. N. E. Hinkle, Am. Soc. Testing Mater. Spec. Tech. Publ. 341, 345
(1963).
10. W. R. Martin and J. R. Weir, Jr., Nuclear Applications 1, 160 :
(April 1965).
11. Private communication from J. Moteff.
12. P.C.L. Pfeil and D. R. Harries, Am. Soc. Testing Mater. Spec. Tech.
Publ. 380, 202 (1965).
13. J. T. Venard and J. R. Weir, Jr., Am. Soc. Testing Mater. Spec.
Tech Publ. 380, 202 (1965).
14. W. R. Martin, unpublished data.
15. W. R. Martin and J. R. Weir, Jr., Influence of Grain Size on the
Irradiation Embrittlement of Stainless Steel at Elevated
Temperatures, ORNL-TM-1043 (March 1965).
16. W. R. Martin and J. R. Weir, Jr., Trans. Am. Nucl. Soc. 8(1), 2
(1965).
¢. W. Weaver, J. Inst. Metals 88, 296 (1959-60).
C. W. Weaver, Acta Met. 8, 343 (1960). | : ENOTYPING..

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FIGURE CAPTIONS
• ORNL - AEC - OFFICIAL
ORNL-DWG
64-1326R
Fig. 1. Effect of Irradiation on the Stre88-Strain
Curves.
ORNL-DWG
'64-1325R
· Fig. 2. Influence of Irradiation at Temperatures
Above 1/2 Im on the Postirradiation Tensile Properties of
Stainless Steel..
ORNL-DWG
65-4619
Fig. 3. Ductility of Irradiated Type 304 Stainless
Steel.
Photo
64248
Fig. 4. Comparison of Intergranular Cracking for.
(a) Unirradiated and (b) Irradiated Type 304 Stainless
Steel Strained at 0.002 in./min at 704°C.
........
..
. CESTO?L!
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07
1
ORAL-DWG 64-1326R

-
1
i...
+ + +
..
..
IRRADIATED
UNIRRADIATED
'.
STRESS
MATERIAL DEFORMED AND IRRADIATED
AT LOW TEMPERATURE (7« 127m)
::::
:

MATERIAL DEFORMED AND IRRADIATED AT
ELEVATED TEMPERATURE (T> 1/2 TMI
UNIRRADIATED
IRRADIATED
STRESS
STRAIN
:
..
......
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ORNL-DWG 64-4325R2

- YIELD AND ULTIMATE
-ENGINEERING STRESS
IRRADIATED
UNIRRADIATED /
TENSILE PROPERTIES
RATIO O
STRAIN RATES
&q><2> <3
TRUE TENSILE-
1 STRESS
STRUE FRACTURE
| STRESS
UNIFORM ELONGATION
AND TRUE
FRACTURE STRAIN
0 0.2 0.4 0.6 0.8 1.0
DEFORMATION TEMPERATURE
ABSOLUTE MELTING POINT OF ALLOY
Influence of Irradiation at Temperatures Above 12Tm on the
Post Irradiation Tensile Properties of Stainless Steel.

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ORNL-DWG 65-4619
NOTE: TENSILE TESTS AT A STRAIN RATE OF 0.002/
I min. ALL MATERIAL GIVEN PREIRRADIATION
LIMITS OF DUCTILITY FOR ANNEAL AT 1038 °C FOR 1 hr ASTM GRAIN
UNIRRADIATED STEEL
SIZE~ NO. 5
704 °C
REPRESENTS ADJUSTED DATA
OF HARRIES FOR 20-25
STAINLESS STEEL AT 750 °C.
NO. 8 GRAIN SIZE AT A STRAIN
RATE OF~ 0.0002/min
TOTAL ELONGATION (%)
CONC.~'4x10-14
AT 842 °C
CONC.~1.5x10-9 to
AT 704 °C 1
10 LTENSILE TESTED
TYPE OF 304 SS —
842 °C 700 °C
A 0.02 CARBON HEAT; 18144BC
A 0.06 CARBON HEAT; 33733 AND 48950
0.00004 0.0004 0.004 . 0.04 0.1
CALCULATED ATOM FRACTION HELIUM (ppm)
Ductility of Irradiated Type 304 Stainless Steel.
10
-
400

PHOTO 64248
lilj .

14
12
16
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9
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3
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2
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9
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(a) UNIRRADIATED MATERIAL FRACTURED AT 0.50 TRUE STRAIN

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..
...
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maman.mod
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(6) IRRADIATED MATERIAL (7x 1020 nvt ) FRACTURED AT 0.14 TRUE STRAIN
Comparison of Intergranular Cracking for Unirradiated (o) and
Irradiated (6) Type 304 Stainless Steel Strained at 0.002 in./min
at 704°C.
END

DATE FILMED
12/13 /65