ORNL-TM-1544.txt

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ORNL- TM- 1544

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i SOLUTIONS TO THE PROBLEMS OF HIGH-TEMPERATURE
P IRRADIATION EMBRITTLEMENT
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i W.R. Martin J.R. Weir
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ORNL-TM-1544

Contract No. W-7405-eng-26

METALS AND CERAMICS DIVISION

SOLUTIONS TCO THE PROBLEMS OF HIGH-TEMPERATURE
IRRADIATION EMBRITTLEMENT

W. R. Martin J. R. Weir

Paper presented at the Sixty-Ninth Annual Meeting of the
American Soclety for Testing and Materials, Atlantic City, N. J.,
June 27=July 1, 1966.

JUNE 1966

OAK RIDGE NATTONATL LABORATORY
Oak Ridge, Tennessee
operated by
UNION CARBIDE CORPORATION
for the
U.S. ATOMIC ENERGY COMMLISSION

QAK RIDGE NATIONAL LABORATORY LIBRARIES

3 4456 0549970 b
Lé

Abstract

The effect of irradiation on the high-temperature mechanical prop-
erties of structural materials is described using type 304 stainless steel
as an example. The general effect is one in which the grain-boundary
fracture process, but not the deformation process, 1s affected. The data
suggest the primary cause to be helium generated from (n,a) reactions. -
oeveral metallurgical techniques for improving the ductilities of irradi-
ated alloys are suggested and experimental data on type 304 stainless

steel are given for which the degree of improvement is demonstrated.

Introduction

Irradiation embrittlement of iron- and nickel-base alloys at tempera-
tures above 500°C is a problem of immense importance to the success of
nuclear reactors. The integrity of structural components and fuel element
cladding can depend upon properties that are affected or related to the
alloy ductility. Because of the paucity of material data at conditions ’
appropriate to the high-temperature reactor environment, a large effort
has been centered in recent years in the area of irradiation damage.

High-temperature embrittlement of iron- and nickel-base alloys is
characteristically different than that observed at temperatures below
500°C. Within the lower temperature range, the damage produced by the
displacement of atoms hardens the lattice, lowers the work hardening rate,
but does not significantly affect the fracture stress. Material may
exhibit low ductility in terms of uniform strains and total elongations

without a large change in the true fracture strain.

 

i B S R g 6 A B E <D e i nh L d et A L G e -
 

 

In contrast, the embrittlement at high temperatures is not assoclated
with large changes in the bulk strength of the alloy. 1In tensile {ests,
the stress necessary to produce a given strain is unchanged; but the
irradiated alloy fractures at true strains smaller than that of the unir-
radiated alloy. In creep tests, the strain-time relationship in the
irradiated and unirradiated materials is approximately equivalent. However,
because of the embrittlement, the stress-rupture properties of the alloy
are reduced. The embrittlement at high temperature is one related primarily
to the fracture process whereas the low ductility observed at lower tem-
peratures is a result of a change in the deformation (stress-strain)
behavior. The change from low- to high-temperature behavior is observed
at approximately 600°C for stainless steels.

Efforts to find commercial alloys given standard heat treatments that
are not affected by irradiation have not been successful. The sclution
to this problem may lie in the investigation of the many metallurgical
variables available while concurrently investigating the mechanisms of
irradiation damage. From these research programs, a general hypothesis
for the mechanism of embrittlement has been derived. Although a complete
understanding of the embrittlement is not available, one is able to propose
possible solutions to the problem and conduct research within these areas

in order to determine their effectiveness.

Nature of the Embrittlement

Our interpretation of the high-temperature embrittlement has as its
basis our work and the recent literature. (1-12) The most important

observation 1s that the high-temperature embrittlement is associated with
intergranular fracture. Materials tested under conditions that tend to
produce Intergranular fracture in the absence of irradiation damage are
generally highly susceptible to high-temperature irradiation embrittlement.
The general behavior of the irradiated alloys is that they fracture at
strains at which grain-boundary cracks in unirradiated alloys are either
absent or only slightly perceptible. It appears that once a grain-
boundary crack is nucleated, it is easier to propagate that crack than
to initiate another. At fracture one normally finds fewer cracks in the
irradiated alloy than in the unirradiated material. There are exceptions,
particularly in the alloys in which intergranular cracks propagate readily
in the absence of irradiation damage. With these alloys, the number of
cracks near the fracture in both irradiated and unirradiated material is
small. The Irradiation effect then appears to produce easier nucleation
and propagation of grain-boundary cracks.

Since the early work of Hinkle, (1) Robertshaw et al, (2) and
Hughes and Coley, (3) it now appears that helium generated from (n,«)
reactions is responsible for the embrittlement. Harries (4) demonstrated
that the magnitude of embrittlement was a function of 1%B content and
related to the thermal neutron fluence for alloys in which the total boron
content was constant. The lOB(n,a) reaction with thermal neutrons occurs
at a much greater rate than that of the other elements normally present
in structural alloys. Subsequently other investigators (5) have con-
firmed these findings. Since the 1OB(n,O!) reaction produces ‘Li as well
as 4He, it was not clear whether the lithium or helium was producing the
damage. Higgins and Roberts, (6) in their work on bombarding steels

with «-particles and lithium ions, showed that the helium was the principal

 

 
 

 

 

cause of embrittlement. The solubility of the inert gases in metals is
probably quite low. Therefore once helium bubbles are formed in a

material the gas remains essentially in place. Postirradiation heat
treatments do not remove the irradiation damage in either nickel- or
iron-base alloys. (7) The magnitude of embrittlement may be altered
somewhat by postirradiation heat treatments, but these are effects probably
related to changes in metallurgical structure that affect the process of
intergranular fracture.

Although there is a relationship between thermal neutrons, boron
concentration, and embrittlement, the neutrons in fast reactors (without
a significant neutron flux in the thermal energy range) can produce
sufficient helium to cause embrittlement. In a fast neutron spectrum,

3 to 4 Mev neutrons reacting with all of the elements will produce about
one-tenth the helium generated from a stainless steel containing 5 ppm B
irradiated to an equivalent thermal neutron fluence. Therefore, irradia-
tion embrittlement due to helium will be a problem of some magnitude

in fast reactors as well as in thermal reactors.

Helium-induced embrittlement is not peculiar to the stainless steels
and nickel-base alloys. In 1959, Rich (8) proposed that gas bubbles
served as crack nuclei in irradiated beryllium. All of the work in study-
ing irradiated beryllium, (9) fission-gas bubble (lO,ll) behavior in
fuels, and cyclotron-induced helium bubbles (12) in aluminum and copper
serves as a basis for predicting the behavior of helium in iron- and

nickel-base alloys.
In 1965, Barnes (13) presented a critical stress model for helium-
induced embrittlement in which the helium bubbles serve as crack nuclei.
Vacancy flow into the bubbles allows them to grow and connect, thus form-

ing a large grain-boundary crack.

Improvement in Ductility of Irradiated Alloys

Metallurgical solutions to the problem of high-temperature embrittle-
ment appear to be related to one of two primary theses. The first is to
reduce the tendency for grain-boundary fracture in the unirradiated and
irradiated materials. The second assumes that the embrittlement is related
to helium and hence one attempts to reduce the quantity of helium that
resides in the boundary during deformation.

We (14,15) cited some time ago several approaches which could be
taken. These were
1. decreasing the grain size,

2. appropriately distributing the grain-boundary precipitate,

3. lowering the concentration of helium produced in the alloy, and

4. distributing the precipitate within the matrix that traps helium
bubbles and prevents their ultimate concentration within the grain
boundaries.

small-grain size and spherodized grain-boundary carbides should
decrease the tendency for intergranular fracture by increasing the stress
necessary to nucleate a grain-boundary crack according to the following

expression. (16,17)

o = (L2pyy/mL)t/2 (1)

 

 
 

 

 

where
on = stress necessary to nucleate a crack,
= shear modulus,
Yq = free energy of surface,
I =

length of the sliding interface.

The rate of propagation of the crack will be related to the maximum stress
at the tip of the crack. This stress, O ax’ is given in the expression

developed by Zener (18) and Tnglis: (19)

o = [(2t/L)Y/2 + (L/2t)/210_ (2)

where

O
a

i

applied shear stress,

I

radius of curvature at the tip of the sliding interface, and
L = length of sliding interface.
This expression may be simplified for the conditions when L >> t;

now

Oy = Ga(L/Zt)l/2 . (3)
Since the length of the boundary is proportional to the grain size, 1t
follows that grain size can greatly influence intergranular fracture.

When the grain size is decreased, a higher stress 1s required to nucleate
the wedge-type fracture and also the rate at which cracks propagate should
be reduced. Grain-boundary carbides could also act in the same manner.
The length of the sliding interface would be the interparticle spacing

if the strength of the carbide-matrix interface is large compared to the
cohesive strength of the boundary. Weaver (20,21) and Garofalo (22) have
shown that the heat treatment of nickel- and iron-base alloys to produce
coherent grain-boundary precipitate can improve the ductility. Garofalo (22)
cnumerated the following conditions for grain-boundary precipitate to be
beneficial:
1. high cohesion between particle and matrix,
2. interparticle spacing of 1 to 2 p to allow grain-boundary migration,
and
3. rounded particles that have high shear strength.
Examples of the beneficial effect of grain size on tensile properties
are given in Table 1 and the creep data at 650°C in Table 2. A 100-hr

heat treatment at 800°C following a l-hr sclution anneal at 1036°C was

Table 1 -- Grain Size Dependence of Postirradiation Short-Time

Tensile Strength and Ductility of Type 304 Stainless Steel

 

 

 

 

Deformation Yield Strength (psi) Ductility (%)
Temperature True Uniform  Total Elongation
ASTM 9 ASTM 5
(°c) ASTM 9 ASTM 5 ASTM 2 ASTM 5
500 23.5 x 10° 18.8 x 10°® 23.6 24,7 32.2 31.8
600 19.3 13.1 25.5 24,1 34,7 32.0
700 16.7 12.4 19.5 19.1 36.3 25.2
800 15.6 11.6 15.2 14.9 29.6 16.2

900 8.2 8.7 10.1 6.2 24,7 10.2

 
 

Table 2 -- Grain Size Dependence of Postirradiation Stress

Rupture of Type 304 Stainless Steel at 650°C

 

Strength in Terms Ductility (%)

 

 

 

 

Stress of Time to Rupture
(psi) (hr) Elongation at Rupture
ASTM © ASTM 5 ASTM 9 ASTM 5
30 x 10° 11.3 1.5 4ti .0 11.4
25 79.0 5.5 29.5 9.0
20 191.0 109.5 17.2 3.8
15 514.4 194.4 9.2 3.5

 

given type 304 stainless steel in order to produce the desired grain-
boundary carbide distribution. These microstructures are shown in Fig. 1.
The effect of aging is illustrated in Tables 3 and 4 for tensile and creep
conditions, respectively. In terms of ductility, the fine-grain size 1is
superior to the coarse grain in the aged and unaged conditions. However,
because the coarse-grain material creeps at a rate lower than the fine-
grain material, the time to rupture for the aged coarse-grain material
offers the best properties for those conditions investigated to date.
Another approach to improve the ductility is to assume that the radiation-
induced grain-boundary embrittlement is due to the generation of helium
from the (n,a) reactions. Helium bubbles at the grain boundary would be
expected to be deleterious. However, helium generated within the grains
would not be harmful. The primary way that these helium atoms could be

swept into the boundary would be by a dislocation mechanism as illustrated

v el o e et R el TR TR g T R AT T T TR S R T A e S e S IR T T
¢ Y R-29127
; 7. o " - ;
,’ . d
: .. B \‘,» "‘. o’
{ ? % .
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e ,3~M'. e t" p—
<
- R
S m—
-
4 1
% ot
$ -

  

\_....
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Y

 

(b)

Fig. 1 -- Microstructures of Irradiated Stainless Steel Having Been
Creep Tested at 630°C. 20,000 psi stress. (a) Preirradiation heat
treatment of 1 hr at 1036°C, fractured at 3.8% strain. (b) Preirradiation
heat treatment of 1 hr at 1036°C followed by a 100-hr aging at 800°C,

fractured at 14.3% strain. 1000x.
 

11

Table 3 -- Effect of Preirradiation Aging on the Short-Time Tensile

Properties of Type 304 Stainless Steel

 

Ductility (%)
Deformation Strain

 

 

 

Temperature Rate Yield Strength True Total
(°c) (min-1) (psi) Uniform Elongation
Unaged Aged Unaged Aged Unaged Aged
704 20 12.7 x 10° 12.2 x 10° 24.3 25.8 34.7 37.2
0.2 13.1 15.6 15.6 18.1 20.5 30.5
842 20 .11.8 10.6 10.4 12.8 13.7 17.5
0.2 10.9 10.2 5.1 9.4 7.6 13.9

 

Table 4 ~- Effect of Preirradiation Aging on the
Postirradiation Stress Rupture of Type 304

Stainless Steel at 650°C

 

Strength in Terms Ductility (%)

 

 

 

 

Stress of Time to Rupture
Elongation at Rupture
(psi) (nr)
Unaged Aged Unaged Aged
30 x 107 1.5 14.8 11.4 24,2
25 5.5 50.8 9.0 25.1
20 109.5 664.1 3.8 14.3

15 194.4 3638.0 3.5 7.8

 
12

by Barnes (13) for copper. Therefore, to improve the ductility of irradi-
ated materials, one must devise ways to reduce the helium concentration
at the grain boundaries.

For many reactor applications, the preponderance of helium generated
is due to the transmutation of 19B. Boron, a horophillic element, normally
segregates to the grain boundaries in the solid state and therefore a
large quantity of helium is generated near these boundaries. If one could
form a stable boron compound, insoluble either in the melt or at a very
high temperature after solidification, it would be possible to get a
homogeneous distributicn of this compound. Therefore, helium generated
would tend to stay at the precipitate-matrix interface and hence the
quantity at the grain boundaries would be greatly reduced. These precipi-
tates having an incoherent interface, would also serve as traps for helium
generated from other elements and fast neutrons. Thus in principie, this
system should result in material with a lower susceptibility to high-
temperature embrittlement in thermal and fast-neutron envircnments.

We have chosen to first investigate the iron-base systems, In particu-
lar 18-8 stainless steel. Among the most stable borides are those of
titanium. We have now accumulated data from two different irradiations,
and typical data are given in Fig. 2 for a steel containing 0.02 wt % C.
Data for the 0.06 wt % C alloy are given in Table 5. It is clear that
small additions of titanium greatly improve the ductility of type 304
stainless steels. Titanium additions at the level required to meet the
chemical specificatlons for type 321 stainless steel are above the range

for which one observes the maximum ductility.

 
 

 

 

13

ORNL-DWG 66— 239

 

 

 

 

 

 

 

STRAIN RATE OF 2%/ min
70 A 1 *
UNIRRAm/.E\TED
60 IRRADIATED 4X102° neutrons/cm?

(THERMAL) |

\ 1.5X 1019 neutrons/cm? (FAST)

\

o) |
ol \ A
~_

 

 

 

 

 

DUCTILITY, TOTAL ELONGATION (%)

 

 

 

 

 

 

 

 

 

— ¢
30
20

¢ )\C’\fi
10 |

REGULAR 304 S5

) .

0 0.2 0.4 0.6 08 1.0 1.2 14

PERCENT Ti

Fig. 2 -- Ductility at 842°C of Irradiated Austenitic Stainless Steel

as a Function of Titanium Content.
14

Table 5 -- Influence of Titanium on the High-~
Temperature Irradiation Embrittlement of

18-8 Stainless Steels Having 0.06 wt % C

 

 

 

Titanium Total Elongation (%)a,b at
(wt %) 650°C 700°C 842°C
0.0 31.0 31.5 20.0
0.2 40.0 38.0 45.1
0.3 34.7 28,5 35.2
0.4 31.5 26.5 25.0
0.5 28.8 21.5 19.0
0.6 24,1 19.5 17.9
0.8 22.9 20.5 19.1
1.0 30.6 23.5 23.5
1.2 29.8 23.0 28.5

 

aMeasured in 1-in.-gage length for

tests at a strain rate of 0.2%/min.

bSpecimenS irradiated to a fluence

level of 1 x 10?0O neutrons/cm? (thermal) and

1.5 x 10'? neutrons/cm?® (E > 1 Mev).

 
 

 

15

Titanium additions in the range up to 0.2 wt % greatly reduce the
magnitude of irradiation embrittlement in type 304 stainless steel. The
lower ductilities of the higher titanium-bearing alloys are not understood.
Although the alloys containing titanium have a grain size smaller than
the unstabilized grade, we believe the effect of titanium to be as
hypothesized earlier. The helium bubbles in the as-irradiated stainless
steels are not always of a size resolvable in the electron microscope.

A 1-hr postirradiation anneal at 1200°C will produce bubbles in the grain
boundaries of the reguiar 304 stainless steel but not in the 0.2 wt % Ti-
bearing steel. These photomicrographs are compared in Fig. 3. On the
other hand, one can find evidence suggesting bubble attachment to intra-
granular precipitate in the titanium-bearing steels. Figure 4 illustrates
the possible bubble attachment. It is possible that these void areas may
be a result of sample preparation for electron microscopy examination.

We believe the concept of intragranular precipitate serving as sinks
for bubbles to be valid. We have in our own studies experienced difficulty
in getting the correct precipitate distribution and size. Additionally
producing the proper distribution of boron in these precipitates may prove
too difficult in many alloy systems. Thus the improvement of alloys for
use in thermal reactors may prove to be more difficult than the use of
this concept for fast reactor irradiations. In the latter case the pre-
cipitate need not contain the boron. The precipitate may also be formed
in situ, such as the titanium precipitate, or an inert oxide could be
added during fabrication. This latter approach may be desirable if the
thermal stability of precipitates during long exposures at high

temperatures is small in the base alloy system.
16

Standard Type 304 Stainless Steel

 

ORNL Type 304 Stainless Steel Modified with Titanium "

   
 

Grain boundary

-~

Fig. 3 -- Comparison of Grain Boundaries in Irradiated Stainless Steels

&

After Postirradiation Annealing Treatments.

 
17

- YE-9171

 

Fig. 4 -- Electron Transmission Micrograph of 0.2 wt % Ti-Bearing

Stainless Steel after Postirradiation Anneal of 1 hr at 1200°C. 50,000x.

 
18

The final proposal for reducing the embrittlement concerns the con-
centration of helium produced. In fast reactor irradiations, there appears
little hope, as there is no single (n,d) reaction that produces the bulk
of the helium as is the case in thermal reactors with the 1%B(n,x)
reaction. Nitrogen could be a major contributor if the concentrations
become larger than those presently in our commercial alloys. In the
thermal reactors, one can reduce the boron content to concentrations of
10'8, but this appears impractical for commercial application. There is .
another lower limit for boron because even in thermal reactors, the fast
(n,a) reaction produces helium at a rate comparable with the thermal
neutrons and boron at a concentration of about 0.2 ppm for neutron fluences
less than 10°° neutrons/cm?.

The boron level in type 304 stainless steel given an electron-beam
remelt treatment was reduced from 2.9 to 0.015 ppm. Data at a neutron
fluence of 4.5 x 1020 neutrons/cm? have been published earlier (23) for
these alloys and others. We have investigated the embrittlement of these
steels at lower neutron fluences in order to evaluate the relative con-
tribution of helium generated from fast and thermal (n,a) reactions. These
data are given in Table 6 and Fig. 5. The atom fraction of helium plotted
in Fig. 5 is the total helium content. It is apparent that with the lower
boron levels, fast (n,a) reactions are of significant importance. A
correlation of this type 1s surprising since the boron, and hence the
helium produced therefrom, is believed to be segregated to grain boundaries
whereas the helium from fast (n,a) reactions will be produced throughout
the material. If this were true, one would not expect a correlation from

a simple addition of helium produced by both reactions. We believe the -

 

 
 

Table 6 -- Ductility of Stainless Steel as a Function of Boron Concentration and Irradiation Fluence
At 8.6 x 1017 neutrons/cn? At 7.8 x 1018 neutrons/cm? At 2.7 x 10'? neutrons/cm®
(thermal) and (thermal) and (thermal) and

Natural Boron o o . 1016 neutrons/cn® (E > 1 Mev) 7.8 x 1017 neutrons/em? (E > 1 Mev) 2.7 x 109 neutrons/cm® (E > 1 Mev)

 

 

 

 

 

Concentration
Helium Content Helium Content Helium Content
(ppm) Elongation Elongation Elongation
(atom fraction) (atom fraction) (atom fraction)
(%) (%) (%)
Thermal Total Thermal Total Thermal Total
0.015 4. 6 x 10711 1.4 x 10710 32 6 x 10719 1.4 x 10°° 27 2 x107? 5 x 10°?
0.110 43 4 x 10710 5 x 10710 31 3 x 10-2 4 x 1079 26 1 x 10"8 1 x 1078
0.150 ... 5 x 10710 6 x 10710 31 5 x 1077 6 x 107° 24 2 x 1078 2 x 1078

3.900 23 2 x 10"8 2 x 1078 18 2 x 1077 2 x 1077 18 6 x 1077 6 x 1077

 

6T
BORON {ppb)
A 15
C 110
g A 150
® 33500
>
=
_
E TESTED [N VACUU
=]
0O
STRAIN RATE OF 0.2% PER min
10”10 1072 10~8 1077 10—®
LOG OF ATCM FRACTION OF HELIUM
Fig. 5 -- Irradiation Embrittlement of Boron-Stainless Steels

 

20

ORNL-DWG 66-405R

700°C ag a Function of Total Helium Content.

e st o e b EBAe b b MR b Es e B 3 e e e R

 

at

 
21

most plausible explanation is that most of the helium generated from the
threshold reactions is swept into the grain boundaries during deformation;
thus the majority of helium within the samples is at grain boundaries and
one can then get a reasonable correlation by simple addition of the helium
from both sources. This means that the actual concentration at grain
boundaries for a given ductility is many orders of magnitude greater than
that given in Fig. 5. A reasonable figure based on segregation may be a
factor of 1000. Although replication techniques using Faxfilm are prone
to exhibit artifacts we attempted fractographic evaluation of these low
boron bearing alloys in order to evaluate the bubble density along the
fractured boundary. Typical photographs are given in Figs. 6 and 7. We
observed protrusions on the replica that cast shadows, and this i1s what we
would expect for cavities on the metal fracture surface. The density of
these protrusions appears to be related to the total helium content in the
sample and not to neutron fluence or boron content. Assuming, therefore,
that these are cavities along the grain boundary and not artifacts, it

is not clear as to their role in the fracture process.

Summary

The irradiation of stainless steels at temperatures in the range of
600°C and above results in an embrittlement of the alloy that is signi-
ficantly different than the neutron displacement damage which is of
principal importance at temperatures below 600°C. This embrittlement at
high temperatures does not necessarily affect the strength of an alloy,

in terms of the stress necessary to produce a given value of strain. The

 
22

Photo J-349

 

Photo J-352

 

(b)

Fig. 6. Fractographs of Irradiated Stainless Steels Containing

0.015 ppm B after Fracture at 842°C. (a) Fracture at 18% strain with

calculated helium atom fraction of 5 x 10711, (b) Fracture at 12% strain

with calculated helium atom fraction of 4 x 1010, 6, 500x.

 
23

Photo J-318

 

 

Fig. 7. Fractographs of Irradiated Stainless Steels Containing
3.9 ppm B after Fracture at 842°C. (a) Fracture at ll% strain with a
calculated helium atom fraction of 2 x 10-?. (b) Fracture at 10% strain

with a calculated helium atom fraction of 2 x 1078, 6, 500x.

 
24

embrittlement can be severe, and ductilities less than 1% have been
observed for creep conditions.

Irradiation affects the ability of the alloy to resist intergranular
fracture and most experiments point to the principal cause as one related
to the production of helium from two sources. The first one is the
10B(n,o) reaction with neutrons having thermal energies and the second
source 1s from reactions between the major alloy constituents and fast
neutrons having energies in the 3-Mev range.

Solutions to the problem of embrittlement are then related to
(1) alterations of the unirradiated alloy that affect the process of inter-
granular fracture and (2) modifications of the alloy that reduce the
amount of helium that is located in the grain boundaries of the irradiated
alloy.

Metallurgical variables, such as grain size, annealed vs cold-work
structures, are variables that greatly affect the dvctilities of alloys
in the temperature range for which the alloys fracture intergranularly.
After irradiatlion a fine-grain size alloy can be an order of magnitude
more ductile than the coarse grain. Grain sizes in the range of ASTM &
to 11 are preferred.

To reduce the amount of helium at the grain boundary, one must produce
the helium at sites other than the grain boundary and prevent the movement
of helium to the grain boundary. A reduction in boron content can reduce
the amount of helium produced, but is not a complete solution because of
the helium generated from nickel, iron, chromium, nitrogen, and other
elements. Thus, a more suitable approach would appear to be a desegregation

of boron and the production of helium sinks within the matrix of the grains

 
 

 

25

which will greatly reduce the quantity of helium moving to the grain
boundary. Titanium additions to stainless steel are believed to form
complex metal borides dispersed homogeneously within the matrix and the
precipitate-matrix interfaces serve as a depository for helium. Other
approaches for improving high-temperature ductility are available; one

of these is proper aging of existing grades of stainless steels.

Acknowledgments

The authors thank their colleagues, E. E. Bloom, J. W. Woods,
J. 0. Stiegler, T. E. Nolan and R. E. McDonald for their assistance in
our irradiation damage program. We also thank V. Bullington, D. Gates,
H. Kline, and K. W. Boling for their kind attention to details in

conducting the irradiation and deformation experiments.
26

References

1. N. E. Hinkle, "Effect of Neutron Bombardment on Stress-Rupture

Properties of Some Structural Alloys," Radiation Effects on Metals and

 

Neutron Dosimetry, ASTM STP 341, Am. Soc. Testing Mater., 1963, p. 344.

 

2. F. C. Robertshaw, J. Moteff, F. D. Kingsbury, and M. A. Pugacz,
"Neutron Trradiation Effects in A286, Hastelloy X and René 41 Alloys,"
Radiation Effects on Metals and Neutron Dosimetry, ASTM STP 341, Am. Soc.
Testing Mater., 1963, p. 372,

3. N. A. Hughes and J. Coley, "The Effects of Neutron Irradiation
at Elevated Temperatures on the Tensile Properties of Some Austenitic

Stainless Steels,” Journal of Nuclear Materials, Vol. 10, 1963, p. 60.

 

4. P.C.L. Pfeil and D. R. Harries, "Effect of Trradiation in

Austenitic Steels and Other High-Temperature Alloys," Flow and Fracture

 

of Metals and Alloys in Nuclear Environments, ASTM STP 380, Am. Soc.
Testing Mater., 1965, p. 202.

5. W. R. Martin and J. R. Weir, "Irradiation Effects in Stainless
Steels at High Temperatures,”" Proceedings of Sodium Components Development
Program, USAEC CONF-650620, U.S. Atomic Energy Commission, 1966, pp. 36—46.

6. P.R.B. Higgins and A. C. Roberts, "Reduction in Ductility of
Austenitic Stainless Steel after Irradiation,” Nature, Vol. 206, No. 4990,
1965, pp. 12491250,

7. W. R. Martin and J. R. Weir, "Effect of Post-irradiation Heat
Treatment on the Elevated Temperature Embrittlement of Irradiated Stainless

Steel," Nature, Vol. 202, No. 4936, June 1964, p. 997.

 

 
 

 

27

8. J. B. Rich, G. P. Walters, and R. S. Barnes, "The Mechanical

Properties of Some Highly Irradiated Beryllium," Journal of Nuclear

 

Materials, Vol. 4, 1961, p. 287.
9. J. R. Weir, "The Effect of High-Temperature Reactor Irradiation

on Some Physical Properties of Beryllium," International Conference on

 

the Metallurgy of Beryllium, London, 1961, Chapman and Hall, London,

 

1963, pp. 395—409.
10. D. Kramer, W. V. Johnston, and C. G. Rhodes, "Reduction of
Fission-Product Swelling in Uranium Alloys by Means of Finely Dispersed

Phases," Journal of the Institute of Metals, Vol. 93, No. 5, 1964—65,

 

Pp. 145152,
11. W. H. Chatwin and E. D. Hyam, "Grain Boundary Holes in Irradiated
Uranium,"” TRG Report 325(W), United Kingdom Atomic Energy Authority, 1966.
12. C. E. Ells, "Effect of Temperature During Irradiation on the

Behavior of Helium in Aluminum,"” Journal of Nuclear Materials, Vol. 5,

 

1962, pp. 147—149,

13. R. 8. Barnes and R. S. Nelson, "Theories of Swelling and Gas
Retention in Reactor Materials,” AERE-R 4952, Atomic Energy Research
Establishment, Harwell, June 1965,

14. W. R. Martin and J. R. Weir, "Effect of Elevated-Temperature

Irradiation on Hastelloy N," Nuclear Applications, Vol. 1, No. 2, 1965,
J

 

pp. 160~167,
15. W. R. Martin and J. R. Weir, "Influence of Grain Size on the
Irradiation Embrittlement of Stainless Steel at Elevated Temperatures,”

Journal of Nuclear Materials, Vol. 18, No. 2, 1966, pp. 108-118.

 
 

28

16, D. McLean, Grain Boundaries in Metals, Clarendon Press,
Oxford, 1957, p. 336.

17. A. N. Stroh, "The Formation of Cracks as a Result of Plastic
Flow," Proceedings of the Royal Society (London) Series A: Vol. 223,
1954, p. 404.

18. C. Zener, Elasticity and Anelasticity of Metals, University of
Chicago Press, Chicago, 1948, p. 158.

19, C. E. Inglis, Institute of Naval Architects (London) Transactions,
Vol. 55, 1944, p. 1Z6.

20. C. W. Weaver, "The Influence of Annealing Twins on Intergranular
Creep Cracking," Journal of the Institute of Metals, Vol. 88, 1959-60,

p. 296.
21. C. W. Weaver, "Application of Stroh's theory to intercrystalline

creep cracking," Acta Metallurgia, Vol. 8§, 1960, p. 343.

 

22. TF. Garofalo, Fundamentals of Creep and Creep-Rupture in Metals,
Macmillan Company, New York, 1965.

23. W. R. Martin, J. R. Weir, R. E. McDonald, and J. C. Franklin, “
"Irradiation Embrittlement of Low-Boron Type 304 Stainless Steel,” Nature,

Vol. 208, No. 5005, 1965, pp. 7374.

 
615,
16.
17.
18.
19.
20.
21.
22.
23.
24,
25,

71.
2.
73.
7475,
76.
77,
78.
79-80.
81.
82.
83.
84.
85-87,
88.
89.
20.
91-105.

 

ot
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