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ORNL-TM- 3063
C o n t r a c t No. W-7405-erg-26
METALS AND CERAMICS DIVISION
AN EVALUATION OF THE MOLTEN-SALT REACTOR EXPERIMENT
HASTELLOY N SURVEILLANCE SPECIMENS - FOURTH GROW
H. E. McCoy, Jr.
MARCH 1971
LDGAL NOTIC
rs, or their employees,
implied, or assumes any
€or the accuracy, cominformation, apparatus,
OAK R I E E NATIONAL LABORATORY
Oak R i d g e , Tennessee
operated by
UNION CARBIDE CORPORATION
for the
U.S. ATOMIC ENERGY COMMISSION
-c
iii
CONTENTS
Page
.
..............................
1
Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . .
1
Experimental Details . . . . . . . . . . . . . . . . . . . . . . . .
2
Surveillance Assemblies . . . . . . . . . . . . . . . . . . . .
2
Materials . . . . . . . . . . . . . . . . . . . . . . . . . . .
3
T e s t Specimens . . . . . . . . . . . . . . . . . . . . . . . . .
5
I r r a d i a t i o n Conditions . . . . . . . . . . . . . . . . . . . . . . .
5
Testing Techniques . . . . . . . . . . . . . . . . . . . . . . .
7
Experimental Results . . . . . . . . . . . . . . . . . . . . . . . .
8
Visual and Metallographic Examination . . . . . . . . . . . . . 8
Mechanical Property Data .
Standard Hastelloy N . . . . . . . . 17
Mechanical Property Data .
Modified Hastelloy N . . . . . . . . 35
Metallographic Examination of Test Samples . . . . . . . . . . . 57
Discussion of Results . . . . . . . . . . . . . . . . . . . . . . .
72
Summary and Conclusions . . . . . . . . . . . . . . . . . . . . . .
87
Acknowledgments . . . . . . . . . . . . . . . . . . . . . . . . . .
87
Abstract
-
6
2%
I
AN EVALUATION OF THE MOLTEN-SALT REACTOR EXPERIMENT
HASTELLOY N SURVEILLANCE SPECIMENS FOURTH GROUP
-
H. E. McCoy, Jr.
ABSTRACT
Two heats of standard Hastelloy N were removed from the
core of t h e MSRE a f t e r 22,533 hr a t 650°C and exposed t o a
thermal fluence of 1.5 x 1O2I neutrons/cm* and a fast fluence
> 50 kev of 1.1X loz1 neutrons/cm2. The mechanical propert i e s have systematically deteriorated with increasing fluence.
However, t h e cha e i n roperties i s due t o t h e helium
produced by t h e 8B(n,CX77Li transmutation and can be reduced
by changes i n chemical composition. Some of these modified
heats have been exposed t o the core of t h e MSRE and show
improved resistance t o irradiation.
The corrosion of t h e Hastelloy N has been l a r g e l y due t o
t h e s e l e c t i v e removal of chromium. The r a t e s of removal a r e
much as predicted from t h e measured diffusion rate of chromium.
Other s u p e r f i c i a l structure modifications have been observed,
b u t ' t h e y l i k e l y result from carbide p r e c i p i t a t i o n along s l i p
bands t h a t were formed during machining.
INTRODUCTION
The Molten-Salt Reactor Experiment (MSRE) i s a single region reac-
t o r t h a t i s fueled by a molten fluoride s a l t (65 LiF-29.1 BeF2-5 ZrF40.9 UF4, mole $), moderated by unclad graphite, and contained by
Hastelloy N (Ni-16 Mo-7 C I Y : Fe-O.05 C, w t $)
.
The details of t h e reac-
t o r design and construction can be found elsewhere.'
neutron environment wou
rials
I
-
- graphite and Ha
We knew t h a t t h e
roduce some changes i n t h e two s t r u c t u r a l mate-
N.
Although we were very confident of
t h e compatibility of thes
e r i a l s with t h e fluoride salt, we needed
t o keep abreast of t h e
e development of corrosion problems within
PO
I
c
lR. C. Robinson, M3RE Design and Operations Report, Pt. 1, Descript i o n of Reactor Design, ORNETM-728 (1965).
z
t h e reactor i t s e l f .
For these reasons, we developed a surveillance
w
program t h a t would allow us t o follow the property changes of graphite
t
and Hastelloy N specimens as t h e reactor operated.
c
The reactor went c r i t i c a l on June 1, 1965.
After many small prob-
lems were solved, normal operation began i n May 1966.
groups of surveillance samples.
groups were
.e;
We removed four
The r e s u l t s of t e s t s on t h e first three
This report deals primarily with t h e r e s u l t s
of t e s t s on samples removed with t h e fourth group.
The fourth group
included two heats of standard Hastelloy N used i n fabricating t h e MSRE
and three heats with modified chemistry t h a t had b e t t e r mechanicalpropert i e s a f t e r i r r a d i a t i o n and appear a t t r a c t i v e f o r use i n f u t u r e molten-
salt reactors. The respective h i s t o r y of each l o t was (1) standard
Hastelloy N, annealed 2 hr a t 900°C and exposed t o t h e USRE core f o r
22,533 hr a t 650°C t o a thermal fluence of 1.5 X
neutrons/cm2,
(2) two heats of modified Hastelloy N, annealed 1hr a t 1177°C and exposed
t o the MSRE core f o r 7244 h r a t 650°C t o a thermal fluence of 5.1
X
neutrons/cm2, and (3) a single heat of modified Hastelloy N, annealed
f o r 1hr a t 1177°C and exposed t o t h e USRE c e l l environment of
N2
*
b
+ 2 to
59 0 2 f o r 17,033 hr a t 650°C t o a thermal fluence of 2.5 X loi9
neutrons/cm2. The r e s u l t s of t e s t s on these materials w i l l be presented
i n d e t a i l , and some comparisons w i l l be made with t h e data f r o m the
groups removed previously.
EXPERIMENTAL DETAIIS
Surveillance Ass emblies
The core surveillance assembly5 was designed by W. H. Cook and
others, and t h e d e t a i l s have been reported previously.
The specimens
2H. E. McCoy, An Evaluation of t h e Molten-Salt Reactor Experiment
Hastelloy N Surveillance Specimen - First Group, ORNL-TM-1997 (1967).
3H. E. McCoy, An Evaluation of t h e Molten-Salt Reactor Experiment
Hastelloy N Surveillance Specimen - Second Group, ORNLTM-2359 (1969).
“H. E. McCoy, An Evaluation of t h e Molten-Salt Reactor Experiment
Hastelloy N Surveillance Specimen Third Group, ORNL-TM-2647 (1970).
5W. H. Cook, MSR Program Semiann. Progr. Rept. Aug. 31, 1965,
ORNG3872, p. 87.
-
t
i.
I,
3
Bd
%
2'
P
"I
a r e arranged i n three stringers.
Each s t r i n g e r i s about 62 i n . long
and consists of two Hastelloy N rods and a graphite section made up of
various pieces t h a t a r e joined by pinning and tongue-and-groove j o i n t s .
The Hastelloy N rod has periodic-reduced sections 1 l/8 i n . long by
1/8 in. i n diameter and can be cut i n t o small t e n s i l e specimens a f t e r it
i s removed from t h e reactor. Three s t r i n g e r s a r e joined together so
t h a t they can be separated i n a hot c e l l and reassembled with one or
more new s t r i n g e r s f o r r e i n s e r t i o n i n t o t h e reactor. The assembled
s t r i n g e r s f i t i n t o a perforated Hastelloy N basket t h a t i s inserted i n t o
an a x i a l p o s i t i o n about 3.6 i n . from t h e core center l i n e .
A control f a c i l i t y i s associated with t h e surveillance program.
It u t i l i z e s a "fuel salt" containing depleted uranium i n a s t a t i c pot
t h a t i s heated e l e c t r i c a l l y .
The temperature i s controlled by t h e MSRE
computer so that t h e temperature matches t h a t of t h e reactor.
Thus,
these specimens a r e exposed t o conditions similar t o those i n t h e react o r except f o r t h e s t a t i c salt and t h e absence of a neutron flux.
There i s another surveillance f a c i l i t y f o r Hastelloy N located
outside t h e core i n a v e r t i c a l p o s i t i o n about 4.5 in. from t h e vessel.
These specimens a r e exposed t o t h e c e l l environment of N2 + 2 t o 5% 0 2 .
Materials
The compositions of t h e two heats of standard Hastelloy N a r e given
i n Table 1. These heats were a i r melted by t h e S t e l l i t e Division of
Union Carbide Corporation.
Heat 5085 was used f o r making t h e c y l i n d r i c a l
portion of t h e reactor v e s s e l and heat 5065 was used f o r forming t h e top
and bottom heads.
These materials were given a m i l l anneal of 1 h r a t
1177°C and a f i n a l anneal of 2 hr a t 900°C a t l O R N L after fabrication.
The chemical compositions of t h e three modified a l l o y s are given
i n Table 1. The modifications i n composition were made p r i n c i p a l l y t o
improve t h e a l l o y ' s resistance t o r a d i a t i o n damage and t o bring about
i
general improvements i n t h e f a b r i c a b i l i t y , weldability, and d u c t i l i t y .
6H. E. McCoy and J. R. Weir, Materials Development f o r Molten-Salt
Breeder Reactors, ORNLTM-l854 (1967).
4
,-
LiJ
Table 1. Chemical Analysis of Surveillance Heats
Heat 5065
Heat 5085
8
Content, w t 4
Heat 7320 Heat 67-551
Heat 67-504
*'
1
Cr
7.3
7.3
7.2
7.0
6.94
Fe
3.9
3.5
< 0.05
0.02
0.05
Mo
16.5
16.7
12.4
12.2
C
0.065
0.052
12.0
0.059
Si
0.60
0.58
0.03
0.02
0.07
0.010
co
0.08
0. I5
0.01
0.03
0.02
W
0.04
6.001
0.03
Mn
0.55
v
0.22
P
0.004
0.07
0.67
0.20
0.0043
S
Al
0.007
0.01
Ti
0.01
< 0.05
0.17
< 0.02
0.002
0.004
0.003
0.02
0.15
< 0.01
0.65
0.028
0.12
< 0.001
0.0006
0.12
0.01
< 0.002
< 0.05
0.003
0.002
0.03
< 0.02
1.1
0.01
0.01
0.02
0.01
0
0.0016
0.0093
0.001
0.0004
< 0.0001
N
0.011
0.013
0.0002
0.0003
0.0003
Hf
B
< 0.1
< 0.1
0.0024
< 0.002
< 0.05
**
0.03
cu
zr
*B
< 0.01
6-
c
0.01
0.50
0.0038
0.00002
0.00003
0.0002
Alloys 67-551 and 67-504 were small, 100-lb heats made by t h e S t e l l i t e
Division of Union Carbide Corporation by vacuum melting.
finished t o 1/2 in. p l a t e by working at 870°C.
They were
We cut s t r i p s 1/2 in. by
1/2 in. from t h e p l a t e s and swaged them t o l/Ct-in.-diam rod.
!bo sec-
t i o n s of rod were welded together t o make 62-in.-long rods f o r fabricating t h e samples.
The rods were annealed f o r 1h r a t 1177°C i n argon
and then t h e reduced sections were machined.
Heat 7320 was a 5000-lb
melt made by t h e Materials Systems Division of Union Carbide Corporation.
,
P
Part of t h e heat was fabricated by t h e vendor t o 5/16-in.-diam rod and
!
was s i n t e r l e s s ground t o obtain t h e needed 1/4 in. stock.
The material
was annealed 1h r a t 1177°C and then t h e reduced sections were machined.
i
,
k
,-
LJ
5
Test Specimens
&.
The surveillance rods inside t h e core a r e 62 i n . long and those
E
r)
'II
outside t h e v e s s e l a r e 84 i n . long.
They both a r e 1/4 i n . i n diameter
with reduced sections 1/8 i n . i n diameter by 1 1/8 i n . long. After
removal from t h e reactor, t h e rods a r e sawed i n t o small mechanical prope r t y specimens having a gage section 1/8 i n . i n diameter by 1 l/8 i n .
long.
The f i r s t rods were machined as segments and then welded together,
but we described previously an improved technique i n which we use a
milling c u t t e r t o machine t h e reduced sections i n t h e rod.3
This tech-
nique i s quicker, cheaper, and requires l e s s handling of t h e r e l a t i v e l y
f r a g i l e rods than t h e previous method of making t h e rods i n t o segments.
The standard Hastelloy N rods were machined and welded together and t h e
modified alloys were prepared by milling.
f.
h
'
.
IRRADLATION CONDITIONS
The i r r a d i a t i o n conditions f o r t h e various groups of surveillance
specimens t h a t have been removed are summarized i n Table 2.
The reac-
t o r operated from June 1965 u n t i l March 1968 w i t h a single change of
f u e l salt i n which t h e r e was a 33% enrichment of 235U- After t h i s
period of operation t h e uranium i n t h e f u e l was stripped by fluorination
and replaced with 233U ( r e f . 7).
This charge of s a l t was used u n t i l t h e
present group of samples was removed.
Referring again t o Table 2, t h e
standard Hastelloy N i n t h e core was exposed t o both salts and t h e modif i e d Hastelloy N i n t h e core was exposed only t o t h e l a t t e r salt.
The
same s a l t has been used i n t h e control f a c i l i t y throughout operation.
The specimens outside the core (designated "vessel" specimens)
were exposed t o t h e c e l l envi
-
nment of
N2
+ 2 t o 5% 02.
7P. N. Haubenreich and J. R. Engle, "Experience with t h e Molteng( 2), 118 (1970).
S a l t Reactor Experiment, " Nucl. Appl. Technol. -
e
1
r
Table 2.
,
Sumnary of Exposure Conditions of Surveillance Samples‘
Group 1
Core
Group 2, Haetelloy N
Vessel
Standard
Core
Standard
Haetelloy
Modified
Date inserted
Date removed‘
Megawatt-hour on EiLsRE
a t time of insertion
Megawatt-hour on EiLsRE
at time of r e m 1
Temperature, ‘C
Time a t temperature, hr
Peak fluence, neutrons/cm2
Thermal (< 0.876 ev)
Epithermal (> 0.876 ev)
(> 50 kev)
(> 1.22 MeV)
(> 2.02 MeV)
Heat
Designations
Group 3, HasteUoy N
Core
Core
Vessel
Standard
Modified
Standard
Group 4, Hastelloy N
Core
Core
Vessel
Standard
Modified
Modified
9/8/65
7/28/66
0.0066
9/13/66
5/9/67
8682
8/24/65
9/13/66
6/5/67
6/5/67
0
4/3/68
8682
4/3/68
36,247
8/24/65
5/7/68
0
9/13/66
6/69
8682
4/ID/68
6/69
72,441
5/7/68
6/69
36,247
8682
36,247
36,247
72,441
72,4441
72,441
92,805
92,805
92,805
650 f 10
482800
550 f 10
5500
650 f 10
650 f 10
u,m
15,289
650 f 10
9789
650 f 10
20,789
650 f 10
22,533
650 f 10
7244
650 f 10
17,033
1.3 x 1019
2.5 X 1019
2.1 x 10’’
5.5 X 10l8
3.0 X 10l8
9.4 x 1020
2.8 X 1021
8.5 x lozo
2.3 X lozo
1.1 X IDzo
5.3 X 10’’
1.6 x loz1
4.8 x lozo
1.3 X 10’’
0.7 X lozo
2.6 X loi9
5.0 x U)l9
4.2 x 1019
1.1 X lo1’
6.0 X ID18
1.5 X loz1
3.7 x loz1
1.1 x loz1
3.1 x 1020
1.5 X lozo
5.1
9.1
1.1
0.8
0.4
5065
5085
5065
5m5
67-502
67-504
5065
5085
5065
5085
7320
67-551
0
1.3 x IDzo 4.1
3.8 x lozo 1.2
1.2 x 1020 3.7
3 . 1 x 1019 1.0
1.6 X lo1’ 0.5
5081
5085
x 1020
x lo2’
x 1020
x 1020
x lozo
21545
2U54
aInformetion ccrmpiled by R. C. Steffy, Reactor Division, OWL, July 1969. Rwieed for full-paver operation a t 8 Mw.
b
lozo
x mZo
x io20
x lozo
x lozo
X
2.5
3.9
3.3
8.6
3.5
X
x
1019
1019
1019
x
x 10l8
x 10l8
67-504
7
Testing Techniques
5
The laboratory creep-rupture t e s t s of unirradiated control specimens were run i n conventional creep machines of t h e dead-load and leverarm types.
The s t r a i n was measured by a d i a l indicator that showed the
t o t a l movement of t h e specimen and p a r t of t h e load t r a i n .
The zero
s t r a i n measurement was taken immediately a f t e r t h e load was applied.
The temperature accuracy was +0.75$,
t h e guaranteed accuracy of t h e
Chromel-P-Alumel thermocouples used.
The p o s t i r r a d i a t i o n creep-rupture tests were run i n lever-arm
machines t h a t were located i n hot c e l l s .
The s t r a i n was measured by an
extensometer with rods attached t o t h e upper and lower specimen grips.
The r e l a t i v e movement of these two rods was measured by a l i n e a r d i f f e r e n t i a l transformer, and t h e transformer s i g n a l was recorded.
The
accuracy of t h e s t r a i n measurement i s d i f f i c u l t t o determine.
The exten-
someter (mechanical and e l e c t r i c a l portions) produced measurements t h a t
could be read t o about +0.02$ s t r a i n ; however, other f a c t o r s (temperat u r e changes i n t h e c e l l , mechanical vibrations, etc. ) probably combine
t o give an o v e r a l l accuracy of &0.1$1s t r a i n .
This i s considerably b e t t e r
than t h e specimen-to-specimen reproducibility t h a t one would expect f o r
r e l a t i v e l y b r i t t l e materials.
The temperature measuring and c o n t r o l
systemwas t h e same as t h a t used i n t h e laboratory with only one exception.
I n t h e laboratory, t h e control system was s t a b i l i z e d a t t h e
desired temperature by use of a recorder w i t h an expanded scale.
t e s t s i n t h e hot c e l l s , t h
c o n t r o l l e r without t h e aid
In t h e
o n t r o l point was established by s e t t i n g t h e
t h e expanded-scale recorder.
This e r r o r
and t h e thermocouple accuracy combine t o give a temperature uncertainty
of about kl$.
The t e n s i l e t e s t s were run on Instron Universal Testing Machines.
The s t r a i n measurements were taken from t h e crosshead t r a v e l and genera l l y a r e accurate t o
*z$ s t r a i n .
The t e s t environment
air i n a l l cases.
Metallographic examina-
t i o n showed t h a t t h e depth of oxidation was small, and we f e e l t h a t the!
environment did not appreciably influence t h e t e s t results.
8
EXPERIMENTAL RESULTS
W
V i s u a l and Metallographic Examination
W. H. Cook was i n charge of t h e disassembly of the core surveillance fixture.
As shown i n Fig. 1, t h e assembly was i n excellent mechan-
i c a l condition when removed.
The Hastelloy N samples were more
discolored than noted previously; however, surface marking such as
numbers were r e a d i l y visible.
The detailed appearance of t h e s t r i n g e r
has been described previously by Cook.8
The Hastelloy N surveillance
rods located outside t h e core were oxidized, but t h e oxide was tenacious.
Metallographic examination of t h e Hastelloy N s t r a p s t h a t held t h e
graphite and metal together revealed intergranular cracks. A t y p i c a l
8W. H. Cook, MSR Program Semiann. Progr. Rept. Aug. 31, 1965,
pp. 87-92.
0-3872,
b
\
c
Fig. 1. Overall V i e w of USRE Surveillance Assembly Removed A f t e r
Run 18. Parts of t h i s assembly had been exposed t o t h e salt f o r
22,533 h r a t 650°C. The center portion i s graphite and t h e long rods
are Hastelloy N.
hd
crack i s s h m i n Fig. 2 and extends t o a depth of about 3 mils.
A
similar s t r a p on t h e modified samples which had been i n t h e reactor f o r
7244 hr had cracks t o a depth of about 1.5 mils.
These s t r a p s are about
0.020 i n . t h i c k and they l i k e l y encountered some deformation while being
removed.
However, t h e cracks were quite uniformly spaced on both sur-
faces of the straps, and t h e i r general appearance attests t o a general
corrosion t h a t rendered t h e grain boundaries extremely b r i t t l e .
Examina-
t i o n of unirradiated control straps f a i l e d t o reveal a similar type of
cracking.
Fig. 2. Typical Microstructure of a Hastelloy N (Heat 5055) After
Exposure t o t h e MSRE: Core f o r 22,533 hr a t 650°C. This materialwas
used f o r s t r a p s f o r t h e surveillance assembly. As polished. 500X.
These observations l e d t o t h e examination of tabs from t h e surveillance stringers.
Small sections were cut from t h e centers of t h e
Hastelloy N surveillance rods.
-
Typical photomicrographs of heat 5065
after exposure t o t h e core f o r 22,533 hr are shown i n Fig. 3 .
The
unetched view i n Fig. 3(a) shows t h e surface layer t h a t led t o the discolored appearance and a single grain boundary t h a t is v i s i b l e .
Much
of t h e surface layer looks metallic, but t h i s i s d i f f i c u l t t o judge on
f
10
Fig. 3. Typical Photomicrographs of Hastelloy N (Heat 5065) Exposed
t o t h e MSRE Core f o r 22,533 h r a t 65OOC. ( a ) As polished. (b) Etchant:
aqua regia. 500~.
11
such a t h i n f i l m .
The etched view i n Fig. 3(b) reveals some carbide
p r e c i p i t a t i o n near t h e surface and grain boundaries t h a t are generally
Heat 5085 was exposed an i d e n t i c a l time and
lined with carbides.
t y p i c a l photomicrographs a r e shown i n Fig. 4.
The unetched view i n
Fig. 4(a) shows a modified grain bouhdary s t r u c t u r e t o a depthiof about
2 mils. Etching [Fig. 4(b)] reveals a grain boundary network of carbides. The g r a i n boundaries near t h e surface seem t o etch d i f f e r e n t l y
f r o m t h e r e s t of t h e sample, but l i t t l e more can be said.
We interpreted these observations as being indicative of some corrosion and performed one further crude experiment t o reveal the depth of
t h i s attack. One t e n s i l e sample had been cut too short f o r testing, and
we bent t h e remaining portion i n a vise. The s k l e was then sectioned
and examined metallographically; the resulting photomicrographs a r e shown
i n Fig. 5. The tension s i d e cracked t o a depth of about 4 mils whereas
the compression s i d e did not crack.
Both sides etched abnormally t o a
depth of about 4 mils.
ZI
Samples of heats 5065 and 5085 t h a t were exposed t o the s t a t i c barr e n s a l t i n t h e control f a c i l i t y were examined.
Figure 6 shows t y p i c a l
A
'i
photomicrographs of heat 5065 a f t e r exposure f o r 22,533 hr.
There i s
some surface roughening, but no structure modification near t h e surface
such as t h a t shown i n Fig. 3 f o r t h e sample *om t h e reactor.
Likewise,
heat 5085 (Fig. 7) showed some surface e f f e c t s t h a t were minor compared
with i t s i r r a d i a t e d counterpart i n Fig. 4.
Thus, t h e r e i s l i t t l e doubt that t h e samples i n t h e core experienced
some modifications, apparently t o a depth of 3 t o 4 mils.
This a l t e r s
up t o about l
2$ of t h e sample cross section and can be expected t o influence t h e mechanicalproperties. We s h a l l dwell further on t h i s very
important subject l a t e r i n t h i s report.
A sample of heat 5085 f r o m t h e core was examined by transmission
electron microscopy.
-
This sample had received s u f f i c i e n t thermal fluence
t o transmute about 97$ of the I 0 B t o helium.
obvious i n Fig. 8.
Another point of concern was t h e formation of voids
i n the laaterial due t o fast neutrons.
c
bles and dislocations were present.
(d
The helium bubbles are
No defects other than helium bub-
u
Q
Fig. 4. Typical Photomicrographs of Hastelloy N (Heat 5085) Exposed
t o t h e MSRE Core for 22,533 hr a t 650°C. (a) As polished. (b) Etchant:
glyceria regia. 5OOx.
f
14
&
V
Fig. 6. Typical Photomicrographs of Heat 5065 After l&osure
S t a t i c Barren Fuel S a l t f o r 22,533 h r a t 650°C. (a) As polished.
(b) Etched. Etchant: glyceria regia. 500X.
to
f
Y,
3
.
1
Fig. 7. Typical Photomicrographs of Heat 5085 Af'ter Exposure t o
S t a t i c Barren Fuel S a l t for 22,533 hr a t 65OOC. (a) As polished.
(b) Etched. Etchant: glyceria regia. 500x.
16
Fig. 8 . Transmission Electron Micrograph of Hastelloy N (Heat 5085)
Exposed t o t h e MSRE Core f o r 22,533 h r a t 650°C. Irradiated t o a thermal
neutron fluence of 1.5 X 10" netrtrons/cm2 and a f a s t neutron fluence of
3.1 X '
'
0
1
neutrons/cm2 (> 1.22 MeV). The white spots a r e helium bubbles
t h a t are located on a twin boundary. 25,OOOX. Reduced 16%.
Mechanical Property Data
- Standard Hastelloy N
Two heats of standard Hastelloy N were exposed t o t h e MSRE core
environment f o r 22,533 h r a t 650°C and received a thermal neutron fluence
of 1.5 X 1021 neutrons/cm2.
Similar control samples were exposed t o
s t a t i c barren f u e l s a l t f o r a corresponding length of time.
The results
of t e n s i l e t e s t s on heat 5085 a r e summa;rized i n Tables 3 and 4 f o r
unirradiated and i r r a d i a t e d samples, respectively.
The f r a c t u r e s t r a i n s
are shown as a function of test temperature i n Fig. 9. With increasing
temperature t h e unirradiated samples exhibit f i r s t an increase i n frac-
ture s t r a i n , then a sharp decrease, and then an increase.
The irra-
diated samples follow t h e same general p a t t e r n except f o r t h e absence
Table 3.
Specimen
Number
Results of Tensile Tests on Control Samples of Heat 5085&
Test
Temperature
("C>
(min-1)
stress^
Ultimate
Yield Tensile
Elongation, $
Uniform
Total
Reduction
i n Area
( 4)
0.05
48,100 108,500
39,300 101,400
35,400 94,900
35,200 98,500
34,000 92,800
35,800 90,100
32,700 88,900
40.5
44.7
48.2
48.6
48.0
37.5
51.6
40.6
45.1
48.9
49.0
48.8
38.2
52.2
32.8
33.4
39.3
36.3
40.0
28.8
37.1
True
Fracture
Strain
($1
40
41
50
45
10409
10408
10407
10406
10405
10416
25
200
400
400
500
500
550
10417
550
0.002
36,500
79,500
29 -4
30.1
24.1
28
10418
10420
10421
10422
10404
1W12
10423
10424
10426
10425
600
600
650
650
650
650
760
760
850
850
0.05
33,400
35,200
32,600
34,300
35,700
30,000
33,900
31,600
33,600
24,100
82,200
69,000
75,200
65,600
64,500
59,700
64,900
52,000
50,600
26,300
37.3
25.7
31.0
23.3
24.9
15.2
25.6
12.2
10.0
4.2
37.9
26.9
32.0
23.9
25.5
22.2
27.5
27.0
33.0
36.7
28.6
19.1
28.9
23.9
19.7
21.2
19.7
31.6
37.1
25.3
34
21
34
27
22
24
22
38
46
29
10410
c
Strain
Rate
0.05
0.05
0.05
0.002
0.05
0.002
0.002
0.05
0.002
0.0002
0.000027
0.05
0.002
0.05
0.002
aAnnealed 2 hr at 900°C; exposed t o a static v e s s e l of barren Ate1 salt f o r
22,533 hr at 650°C.
-51
34
46
Table 4 .
Specimen
Number
8806
8805
88od
8803
8802
8801
8812
8813
88U
8816
8817
8818
8800
8799
8819
8820
8821
8822
Postirradiation Tensile Properties of Hastelloy N (Hea, 5085)'
Test
Temperature
("C)
25
200
400
400
500
500
550
Stress, p s i
Strain
Rate
(min-1)
Yield Tensile
53,900
44,800
0.05
41,600
0.002
41,300
0.05
40,800
39,400
0.002
37,900
0.05
36,200
0.002
0.05
36,700
0.002
39,000
0.05
36,400
0.002
37,200
0.002
35,200
0.000027 32,900
0.05
30,700
0.002
35,100
0.05
32,100
22,200
0.002
89,000
85,600
80,900
80,100
73,200
61,100
62,600
49,100
56,200
48,500
50,800
42,800
37,700
0.05
0.05
550
600
600
650
650
650
650
760
760
850
850
34,500
36,300
35,100
32,700
22,200
Elongation, $
Uniform Total
22.0
26.4
30.2
28.4
24.8
ll.7
14.4
5.8
9.8
4.5
8.8
4.7
2.1
1.2
6.1
1.7
2.3
1.6
in Area
($1
w
True
Fracture
Strain
($1
22.1
27.0
30.5
29.2
25.0
12.4
19.6
26.4
24.3
26.8
23.2
13.9
22
31
28
31
26
U.9
14.3
6.2
10.5
4.9
9.3
5.0
2.4
1.4
6.4
2.0
2.3
1.7
9.1
13.3
6.7
9.1
5.8
2.8
6.1
6.4
2.5
2.1
2.6
15
10
15
U
7
10
6
3
6
7
3
2
3
a
Annealed 2 hr at 900°C prior t o insertion i n reactor. Irradiated t o a thermal
fluence of 1.5 X 1021 neutrons/cm2 over a period of 22,533 hr a t 650°C.
I
7o
-
I
STRAIN
RATE UJIRRAM4TEO
I
I
RRACWTEO
cmin-9
0.05
0.002
0
0
A
A
c
TEST TEMPERATURE FC)
Fig. 9. Fracture S t r a i n s of Hastelloy N (Heat 5085) After Removal
from t h e MSFE and from t h e Control F a c i l i t y . A l l samples annealed 1hr
a t 900°C before i r r a d i a t i o n f o r 22,533 h r a t 650°C t o a thermal fluence
of 1.5 X loz1 neutrons/cmz.
1
,.-.
19
of a d u c t i l i t y increase a t high temperatures.
I
.
However, the levels of
t h e f r a c t u r e s t r a i n s are lower f o r the i r r a d i a t e d material over t h e
e n t i r e temperature range.
For both materials t h e f r a c t u r e s t r a i n
decreases with decreasing s t r a i n r a t e .
Another difference i n behavior
between t h e i r r a d i a t e d and unirradiated samples i s t h a t a t t h e highest
t h e f r a c t u r e s t r a i n begins i t s precipitious drop
s t r a i n rate (0.05 min")
a t a lower temperature f o r t h e i r r a d i a t e d material.
Further c h a r a c t e r i s t i c s of t h e e f f e c t s of i r r a d i a t i o n on t h e tens i l e properties of heat 5085 a r e apparent when t h e r a t i o s of t h e irraThe yield
diated and unirradiated properties a r e compared (Fig. lo).
s t r e s s i s frob 10 t o 20% higher f o r the irradiated material a t t e s t t e m peratures up t o 760°C.
The ultimate t e n s i l e s t r e s s i s about 20% lower
f o r t h e i r r a d i a t e d material and drops even further as t h e t e s t tempera-
ture i s increased above 500°C.
The f r a c t u r e s t r a i n s of t h e i r r a d i a t e d
samples a r e about 50% of those of the unirradiated samples up t o about
5OO0C, above which t h e reduction i s even greater.
-
ORNL- DWG 70 729i
1.4
t.2
Q
5
'
0.4
0.2
0
0
.
bi
100
200
300
.
400
500
600
TEST TEMPERATURE (
'C)
700
800
900
Fig. lo. Comparison of t h e Tensile Properties of Control and Irradiated Hastelloy N (Heat 5085). Samples i r r a d i a t e d t o a fluence of
1.5 X 10" neutrons/cm2 over a period of 22,533 hr a t 650°C. Tested a t
a s t r a i n rate of 0.05 min'l.
20
.
The r e s u l t s of t e n s i l e t e s t s on t h e samples of heat 5065 a r e summarized i n Tables 5 and 6. The f r a c t u r e s t r a i n s of these sanqles a r e
s h a m i n Fig. 11as a function of t e s t temperature.
hd
t
The r e s u l t s a r e
q u i t e similar t o those shown i n Fig. 9 f o r heat 5085. A notable except i o n i s t h e much higher fracture s t r a i n s a t low temperatures of heat 5065
after irradiation. The r a t i o s of t h e i r r a d i a t e d and unirradiated tens i l e properties a r e shown i n Fig. 12. The yield s t r e s s was not a l t e r e d
appreciably by irradiation. The ultimate t e n s i l e s t r e s s was decreased
about 10% by i r r a d i a t i o n a t low t e s t temperatures and up t o 35% a t high
temperatures.
Similarly, the f r a c t u r e s t r a i n was reduced about 10% a t
low temperatures, and t h i s reduction progressed t o about 85% a t high
t e s t temperatures.
The progressive change i n t h e fracture s t r a i n w i t h increasing
fluence i s i l l u s t r a t e d i n Fig. I3 f o r heat 5085.
The f r a c t u r e s t r a i n
has been reduced over the e n t i r e range of t e s t temperatures investigated.
A similar trend was noted a t a slawer s t r a i n r a t e of 0.002 min-l
(Fig.
U),although t h e absolute values of t h e f r a c t u r e s t r a i n s were
lower than noted i n Fig. I3 a t t h e higher s t r a i n r a t e of 0.05 min'l.
These samples have experienced various holding times a t 650°C and some
of t h e property changes can be a t t r i b u t e d t o thermal aging.
The frac-
t u r e s t r a i n s f o r several s e t s of control samples of heat 5085 a r e compared i n Fig. l5. A t low t e s t temperatures and at 760°C and above t h e
f r a c t u r e s t r a i n seems t o show a progressive decrease with increasing
holding time a t 650°C. A t intermediate temperatures t h e behavior i s
more complex.
A t a t e s t temperature of 65OoC, t h e r e s u l t s indicate a
progressive decrease i n f r a c t u r e s t r a i n up t o an exposure time of
15,289 hr and then an increase w i t h f u r t h e r aging time.
Heat 5065 was not included i n one set of surveillance samples, but
there are s u f f i c i e n t data t o follow t h e property changes with fluence
and aging time a t 650°C. The f r a c t u r e s t r a i n i s shown i n Fig. 16 as a
function of temperature f o r various fluences.
A t low t e s t temperatures,
excluding t h e one apparently anomalous point, t h e f r a c t u r e s t r a i n was
*
a c t u a l l y higher f o r i r r a d i a t i o n t o fluences of 1.3 and 2.6 X lo1'
neutrons/cm*.
Higher fluences reduced t h e f r a c t u r e s t r a i n at low t e s t
temperatures, but not t o values as law as noted i n Fig. I3 f o r heat 5085.
t
W
21
Table 5.
Specimen
Number
Results of Tensile Tests on Control S w l e s of Heat 5065'
Test
Temperature
(OC)
10384
10388
10392
10383
10391
10378
10395
10375
10377
10393
10382
10394
10397
1MOO
10385
10398
10390
10379
25
200
400
400
500
500
550
550
600
600
650
650
650
650
760
760
850
850
Strain
Rat e
(min-1)
0.05
0.05
0.05
0.002
0.05
0.002
0.05
0.002
0.05
0.002
0.05
0.002
0.0002
0. oooO27
0.05
0.002
0.05
0.002
Stress, p s i
61,200
44,000
42,600
43,900
42,200
45,800
42,500
43,800
41,800
39,400
41,500
39,200
41,800
42,900
32,900
41,100
35,500
24,800
El0
ation
Tensile
Unifz
126,500
107,900
102,000
106,000
99,400
98,700
98,800
84,600
93,800
76,700
81,900
71,700
71,200
60,800
70,500
42,100
49,500
25,100
48.5
47.5
48.9
47.5
48.1
45.5
48.2
25.5
42.0
25.6
26.4
23.1
18.8
7.5
24.3
5.7
8.7
2.9
4
Tkal
49.8
49.4
50.8
47.7
49.6
46.2
49.9
26.1
43.0
26.126.8
23.6
19.6
19.7
27.3
39.6
38.1
56.5
a
Annealed 2 h r a t 900°C prior t o insertion i n the reactor.
vessel of barren fuel s a l t for 22,533 hr a t 650°C.
Reduction True
in Area Fracture
Strain
( 4)
( 4)
36.77
40.16
45.31
39.30
37.54
34.08
38.M
27.51
34.08
34.29
23.8.:
20.89
17.88
24.57
22.85
45.48
45.31
47.15
46
51
60
50
47
42
48
32
42
42
27
23
20
28
26
61
60
64
Exposed t o a s t a t i c
r
I
-
*
Table 6.
Specimen
Nwnber
8833
8832
8831
8830
8829
8828
8839
8840
8841
8842
8843
8W
8827
8826
8845
8846
8847
8848
L
u
Postirradiation Tensile Properties of Hastelloy N (Heat 5065)&
Test
Temperature
Strain
Rate
(min-i)
25
200
0.05
0.05
400
0.05
400
500
500
550
550
600
600
650
650
650
650
760
760
850
850
0.002
0.05
0.002
0.05
0.002
0.05
0.002
0.05
0.002
0.0002
O.OOO27
0.05
0.002
0.05
0.002
Stress, psi
Yield
56,700
46,000
43,700
43,200
43,600
39,800
41,200
37,800
40,500
Ultimate
Tensile
109,500
99,900
94,300
91,300
82,100
63,200
68,700
57,700
60,100
41,000 50,400
39,700
52,100
40,000 42,800
34,300
36,100
28,400
36,900
37,800
41,400
35,200
35,200
35,400
35,400
20,300 20,300
Reduction
Elongation, $
Uniform Total
42.5
42.8
42.8
44.6
39.0
39.3
35.7
29.1
10.9
15.5
7.1
8.6
4.3
6.0
3.0
2.1
37.2
29.5
n.3
15.7
7.2
9.0
4.5
6.2
3.1
2.2
1.2
2.7
1.3
1.6
1.0
1.1
2.5
1.3
1.5
1.0
Area
3.32
4.44
31.22
27.31
22.98
10.70
16.83
8.93
9.30
7.40
7.37
4.89
3.30
Fracture
True
Strain
3
5
37
32
26
11
18
9
10
8
8
5
3
1.1
1
3.33
2.86
2.22
0.16
3
3
2
0
a
Annealed 2 hr a t 900°C prior t o insertion i n the reactor. Irradiated t o a thermal
fluence of 1.5 X lo2' neutrons/cm2 over a period of 22,533 hr a t 650°C.
22
ORNL- DWG m-7292
60
W
*
50
Y
3
- 40
t
a
a
k30
Ec
2
920
lo
0
0
f00
200
300
400
500
600
700
800
900
TEST TEMPERATURE (OC)
Fig. ll. f i a c t u r e Strains of Hastelloy N (Heat 5065) After Removal
f r o m t h e MSRE and f'romthe Control F a c i l i t y . A l l samples annealed 2 h r
a t 900°C before i r r a d i a t i o n f o r 22,533 hr a t 650°C t o a fluence of
1.5 X loz1 neutrons/cmz.
ORNL-DWG 70-7293
1.2
-
1.0
W
0
s6 0.8
2
0:
5
\
e 0.6
2
z=
0
0
0.4
sa
0.2
n
"
0
(00
200
300
400
500
600
TEST TEMPERATURE PC)
700
800
900
Fig. 12. Comparison of t h e Tensile Properties of Control and Irradiated Hastelloy N (Heat 5065). Samples i r r a d i a t e d t o a fluence of
1.5 X 10" neutrons/cm' over a period of 22,533 hr a t 650°C. Tested a t
a s t r a i n rate of 0.05 m i n - l .
.
23
80
I
I
I
1
I
I
I
I
I
I
400
200
300
70
c
ORNL- OWG 70- 7294
ANNEALED 2 h r AT900°C
1.3 x 40'9 neulrons/cm2, 4400 hr
0 2.6 x
neutrons/cm2, 20,789 hr
v 1.3 x 40'
neutrons/cm2, 4800 hr
0 9.4 x1OZo neutrons/cm2, 45,289 h i
0 4.5 X 40" neutrons/cm2. 22.533 hr
0
A
60
2 50
5a
I-
*
40
W
a
3
Io
a
E 30
20
40
n
.,
0
400
500
600
TEST TEMPERATURE (OC)
700
000
900
Fig. 13. Fracture Strains of Hastelloy N (Heat 5085) After Irrad i a t i o n t o Various Thermal Fluences i n t h e MSRE. Tested a t a s t r a i n
rate of 0.05 rnin-l.
40
LL
w
5 20
3
E
0
0
*
(00
200
ORNL-WG 69-4467R
22,533
300
400
500
600
TEST TEMPERATURE C C )
700
800
9€Jcl
Fig. Ik. Postirradiation Tensile Properties of Hastelloy N
(Heat 5085) After Exposure t o Various Neutron Fluences. Tested a t a
s t r a i n rate of 0.002 min-1.
24
..0
400
2 0 0 3 0 0 4 0 0 5 0 0 6 0 0 7 0 0 8 0 0 9 0 0
TEST TEMPERATURE ('c)
F i g . 15. Variation of t h e Tensile Properties of Hastelloy N
(Heat 5085) with Aging Time i n Barren Fuel S a l t a t 650°C. Tested a t a
strain rate of 0.05 min-l.
ORNL- DWG 70- 7295
70
ll
60
-E
50
z
3 40
t;
e
W
Ea 30
e
20
$0
0
0
(00
200
300
400
500
600
TEST TEMPERATURE CC)
700
800
900
Fig. 16. Variation o f t h e Postirradiation Tensile Properties of
Hastelloy N (Heat 5065) with Thermal Neutron Fluence.
25
A t t e s t temperatures above 550°C t h e f r a c t u r e s t r a i n decreases progres-
s i v e l y with increasing fluence.
A s shown i n Fig. 17 some of these
changes i n f r a c t u r e s t r a i n can be a t t r i b u t e d t o thermal aging a t 650°C.
Except a t t e s t temperatures above 75OoC, t h e f r a c t u r e s t r a i n i s lowest
f o r the material aged 15,289 h r and i s improved a f t e r aging 22,533 hr.
The changes i n properties a t l o w t e s t temperatures a r e l e s s than those
f o r heat 5085 (Fig. 15).
ORNL- DWG 70- 7296
70
60
W
a
30
a
a
lL
20
n v
0
400
200
300
400
500
600
TEST TEMPERATURE CC)
700
000
900
Fig. 17. Effects of Thermal Aging a t 650°C on t h e Tensile Propert i e s of Hastelloy N (Heat 5065) a t a S t r a i n Rate of 0.05 min-l.
A s discussed previously i n t h i s s e r i e s of reports, the changes i n
f r a c t u r e s t r a i n a t low temperatures were not expected.
Although t h e
f r a c t u r e s t r a i n s have not reached values below 2046, we a r e s t i l l i n t e r ested i n i t s progression.
Fig. 18.
Our experience t o date i s summarized i n
If t h e noted e f f e c t s were due s i m p l y t o thermal aging, then
t h e r e s u l t s should c o r r e l a t
correlation does not e x i s t .
i t h t h e time a t 650°C.
Obviously such a
Where data a r e available f o r p a i r s of irra-
diated and unirradiated samples, the i r r a d i a t e d sample has the lower
26
70
ORNL-OWG 70-7297
I
HEAT
0
I
I
I
IRRADIATED
UNIRRADIATED
0
0
4
8
12
16
TIME AT 650T (hr)
1
I
I
20
(Xl03)
Fig. l8. Variation of t h e Fracture S t r a i n a t 25°C with Annealing
T h e and Thermal Fluence.
fracture strain.
embrittlement.
This indicates t h a t i r r a d i a t i o n has a r o l e i n t h e
We previously showed t h a t t h e d u c t i l i t y could be
recovered by an anneal of 8 hr a t 8 7 1 ° C and concluded t h a t t h e changes
m u s t be associated with carbide precipitation.
The higher susceptibil-
i t y of heat 5085 t o t h i s type of embrittlement is not understood, since
heat 5065 a c t u a l l y has a higher carbon concentration (Table 1, p. 4).
The data i n Fig. 18 defy extrapolation, so one cannot conclude whether
t h e room temperature embrittlement i s l i k e l y t o become worse.
In these discussions of t e n s i l e properties we have emphasized t h e
changes i n f r a c t u r e s t r a i n and made l i t t l e mention of strength changes.
Some pertinent data a r e summrized i n Table 7 f o r heat 5085.
The sam-
p l e s were t e s t e d a t 25 and 650°C and include three h i s t o r i e s :
annealed, (2) thermally aged, and (3) irradiated.
(1)as
The changes i n yield
strength a r e l i k e l y significant, but t h e main point i s that they a r e
9H. E. McCoy, An Evaluation of t h e Molten-Salt Reactor Experiment
Hastelloy N Surveillance Specimen- Third Group, ORNLTIG2647 (19701,
p. 17.
x
r-
.
V
27
Table 7. Comparison of t h e Tensile Properties of
Heat 5085 Before and After Irradiationa
Heat Treatment
Ultimate Tensile
Stress, p s i
macture Strain, 4
Yield Stress, p s i
a t 25OC a t 650°C at a o c at 6 5 0 " ~ at 25°C a t 650°C
Annealed 2 hr a t 900°C
51,500
29,600
120,900
75,800
53.1
33.7
Annealed 2 hr at 900°C and
aged 22,533 hr at 650°C
48,100
32,600
108,500
75,200
40.6
32.0
Annealed 2 hr a t 900"C,
irradiated f o r 22,533 h r
a t 650°C t o a thermal
fluence of
1.5 X 1 O 2 l neutrons/cm2
53,900
36,400
89,000
50,800
22.1
9.3
a
S t r a i n r a t e of 0.05 min'l.
small considering t h e long thermal h i s t o r y of 22,533 hr.
The ultimate
t e n s i l e strength i s reduced due t o the lower f r a c t u r e s t r a i n and t h e
resulting f a c t t h a t t h e t e s t i s interrupted before it reaches the high
ultimate strength noted f o r t h e as-anneaJ-ed material.
a r e s h a m i n Table 8 f o r heat 5065.
Similar r e s u l t s
The yield strength i s increased by
aging, but the main e f f e c t s a r e on t h e fracture s t r a i n s and t h e ultimate
t e n s i l e strengths.
Table 8. Comparison of t h e Tensile Properties of
Heat 5065 Before and Af'ter Irradiationa
Heat Treatment
Yield Stress, p s i
Fracture Strain, $
a t 25°C a t 650°C at 25°C at 6 5 0 " ~ a t 25°C a t 650°C
Annealed 2 hr a t 900°C
Annealed 2 h r a t 900°C and 61,200
aged 22,533 hr a t 650°C
Annealed 2 hr a t 900°C,
irradiated f o r 22,533 hr
a t 650°C t o a thermal
fluence of .
1.5 X 1 O 2 I neutrons/cm2
ultimate
Stress, Tensile
psi
56,700
aS t r a i n r a t e of
0.05 min-l.
32,100
41,500
126,400
126,500
81,200
81,900
55.3
49.8
34.3
26.8
39,700
109,500
52,100
42.8
6.2
28
The r e s u l t s of creep-rupture t e s t s on heats 5085 and 5065 from the
fourth group of surveillance samples a r e sumarized i n Tables 9 and 10.
u
*
The r e s u l t s a r e most i n t e r e s t i n g when they a r e campared with those
obtained on previous groups of surveillance samples so t h a t t h e progress i v e changes i n properties can be noted.
c
The stress-rupture properties
of heat 5085 a r e shown i n Fig. 19. The rupture l i v e s of t h e last group
of samples (1.5
X
lo2' thermal neutrons/cm2) a r e not detectably d i f f e r -
ent from those irradiated previously t o a thermal fluence of 9.4 X 1020
I n t h e unirradiated condition, the rupture l i v e s f o r the
samples aged 22,533 hr are s l i g h t l y l e s s than i n t h e as-annealed condineutrons/cm2.
t i o n , but greater than those observed f o r samples aged f o r l e s s e r times.
Table 9.
Test
Number
Specimen
Number
Creep-Rupture Tests on Heat 5085 a t 650°C
Stress
(psi)
Rupture
Ufe
(hr)
Rupture
Strain
( 4)
Reduction
i n Area
Minimum
Creep
23.4
28.5
9.4
( 4)
7978
10406
Unirradiat eda
55 000
22.1
21.0
7897
10415
40,OOO
7896'
7895b
1mU.
10413
2187.1
5.40
5.40
7894b
7893b
10412
104U
32,400
27,000
21,500
2064.9
2.44
17,OOO
2000.0
2.10
1.1
0.288
0.0299
0.0045
0.0026
0.ooll
0
0.0007
!
,
500.1
2500.7
24.5
12.7
Irradiated'
R-969
8811
40,000
0.6
1.1
0.78
R- 958
8810
32,400
31.6
0.54
R-956
8809
27 ,000
R- 957
R- 964
8808
21,500
60.2
362.0
0.57
0.63
0.0067
0.0055
0.0010
8807
17 ,000
426.4
0.69
0.0004
a
Annealed 2 hr a t 900°C, exposed t o s t a t i c barren f u e l s a l t f o r
22,533 hr a t 650°C.
.
bDiscontinued p r i o r t o f a i l u r e .
C
t
Annealed 2 hr a t 900°C, exposed t o MSRE core f o r 22 533 h r a t
65OoC, received thermal fluence of 1.5 X lo2' neutrons/cm
a.
29
Table 10. Creep-Rupture Tests on Heat 5065 a t 650°C
Test
Number
7978
7892
78937,
7890b
788gb
7888
Specimen
Number
104.06
10388
10387
10386
10385
10384
Stress
(psi)
Rupture
Life
Rupture
Strain
( 4)
Reduction
i n Area
( 4)
Unirradiat eda
22.1
21.0
55,000
40,000
651.2
36.3
2144.3
44.7
32,400
8.8
2l87.1
27,000
2373.3
3.6
21,500
2000.0
2.8
17,000
23.4
29.9
37.7
7.4
2.2
0
Minimum
Creep
0.34
0.0331
0.0096
0.0040
0.0013
Irradiat edc
R-966
R-962
R- 954
R-960
R- 963
8838
8836
8836
8835
8834
40,000
32,400
27,000
21,500
17,000
0
22.7
40.5
46.2
1567.3
0.0075
0.0061
0.0051
0.0004
0.25
0.44
0.47
0.86
aAnnealed 2 h r a t 900°C, exposed t o s t a t i c barren f u e l s a l t f o r
22,533 h r a t 650°C.
bDiscontinued p r i o r t o f a i l u r e .
C
Annealed 2 h r a t 9OO"C, exposed t o MSRE core f o r 22 533 h r a t
650"C, received thermal fluence of 1.5 X 1021 neutrons/cm 1
.
The minimum creep r a t e s a r e shown as a function of s t r e s s i n
Fig. 20 f o r heat 5085.
A s noted previously, neither i r r a d i a t i o n nor
aging has a detectable e f f e c t on the creep r a t e . The data deviate from
t h e l i n e shown i n Fig. 20 a t both extremes. A t high s t r e s s e s , t h e irrad i a t e d s a q l e s f a i l a t such l o w s t r a i n s t h a t a minimum creep r a t e i s not
established over long enough eriod f o r measurement.
A t low s t r e s s
l e v e l s t h e data deviate due to t h e semilog p l o t .
The f r a c t u r e s t r a i n i s t h e parameter most a f f e c t e d by i r r a d i a t i o n ,
and a v a r i a t i o n of t h i s parameter with minimum creep r a t e i s shown i n
Fig. 2 1 f o r heat 5085 a t
tion o f t h e fracture str
Tensile and creep t
s t r a i n r a t e s and s t r e s s e s .
.
There has been a continual deterioraincreasing thermal fluence.
e run a t 650'C a t s e v e r a l d i f f e r e n t
e fracture strains fromthese t e s t s are
p l o t t e d together i n Fig. 22 although d i f f e r e n t parameters a r e controlled
30
ORNL-OWG 69-
t
0
-
o 1.3
x 1019
A 2.6 x 1019
o 9.3 x 4OZo
0 9.4 x 1020
0 1.5 x (02'
I 1.1111111
fo-'
400
v
neutrons/cm2 (MSRE) AN<
20,709
4000
(5,289
22.533
3-5 x 40"
neutrons/cm2 (ORR)
'
I 11111111
io'
402
103
.-
' RUPTURE TIME (hi)
Fig. 19. Postirradiation Stress-Rupture Properties of MSRE Surveillance Specimens (Heat 5085) a t 650°C.
ORNL-DWG 69-4474R2
70
60
--I
g
-
50
40
9
v)
30
LT
c
v)
20
40
0
40-3
40-2
io-'
MINIMUM CREEP RATE (%
/ hr)
$00
40'
Fig. 20. Minimum Creep Rate of Hastelloy N (Heat 5085) Surveillance Specimens from the MSRE a t 650°C.
.
_-.
w
31
ORNL-DWG 6 9 - 4 4 7 2 R
io-4
10-3
40-2
lo-‘
MINIMUM CREEP RATE (%/hr)
I00
4 0’
Fig. 21. Variation of Fracture S t r a i n w i t h S t r a i n Rate f o r
Hastelloy N (Heat 5085) Surveillance Specimens a t 650°C.
ORNL-DWG 70-7298
40
I I I111111
UNIRRADIATED
1-
-;
(1’
30 -
1
I
s-”
-
0
20
$
a
e-
-*
e--
IO
W
a
2
8
V
a
a
LL
6
40-2
to‘
lo-’
I02
io3
STRAIN RATE ( % / h t )
Fig. 22. Fracture S t r a i n a t 650°C of Heat 5085 i n the Unirradiated
and Irradiated Conditions. Irradiated t o a thermal fluence of 1.5 X lo2’
neutrons/cm2 over a period of 22,533 h r a t 650°C.
i n t h e two types of tests.
The most s t r i k i n g feature i s . t h e marked
dependence of t h e f r a c t u r e s t r a i n of t h e irradiated samples on t h e s t r a i n
r a t e and t h e r a t h e r weak dependence of the f r a c t u r e s t r a i n s of t h e
unirradiated samples on the s t r a i n r a t e .
”he i r r a d i a t e d samples a t t h i s
I
32
high fluence l e v e l show a general decrease i n f r a c t u r e s t r a i n with
decreasing s t r a i n r a t e , with a possible s l i g h t increase i n f r a c t u r e
.
s t r a i n a t very l o w s t r a i n rates.
The s e n s i t i v i t y of t h e fracture s t r a i n of heat 5085 a t 650°C t o
t h e helium content i s i l l u s t r a t e d i n Fig. 23 f o r t h r e e s t r a i n r a t e s .
This material contains 38 ppm B (Table 1, p. 4) t h a t can y i e l d an equiva l e n t amount of helium when transmuted.
(Several factors contribute t o
t h e relationship that 1ppm of' natural boron by weight leads t o
1.1ppm He on an atomic basis.)
The points shown i n Fig. 23 a l l come
from surveillance samples from t h e lSRE.
The two higher s t r a i n rates
are obtained by t e n s i l e t e s t i n g , and t h e f r a c t u r e s t r a i n decreases with
increasing helium content.
The 0.1% s t r a i n rate i s t h e creep r a t e that
corresponds t o t h e lowest fracture s t r a i n (Fig. 21).
The f r a c t u r e
s t r a i n drops abruptly with the presence of l p p m He and then decreases
gradually with increasing helium content.
The creep properties of heat 5065 are i l l u s t r a t e d i n a s e r i e s of
graphs similar t o t h a t j u s t presented f o r heat 5085.
The stress-rupture
properties a t 650°C a r e shown i n Fig. 24 f o r heat 5065. The rupture
ORNL-OWG 69-7297R2
Fig. 23. Variation of t h e Fracture S t r a i n with Calculated Helium
Content and S t r a i n Rate f o r Hastelloy N (Heat 5085) a t 650°C.
P
33
ORNL-DWG 69-4474R
100
10'
102
to4
RUPTURE TIME (hr)
Fig. 24. Stress-Rupture Properties of MSRE Surveillance Specimens
(Heat 5065) a t 650°C.
times are equivalent f o r t h e samples i r r a d i a t e d t o the two highest
fluences.
The rupture times a r e a l s o quite camparable with those s h a m
i n Fig. 19 f o r heat 5085.
The unirradiated samples t h a t were aged f o r
22,533 h r a t 650°C i n s t a t i c barren f u e l salt have longer rupture l i v e s
than t h e as-received material. The minimum creep r a t e s are shown i n
Fig. 25 and show a lack of s e n s i t i v i t y t o any of t h e variables being
studied. The f r a c t u r e s t r a i n i s shown as a f'unction of s t r a i n r a t e i n
Fig. 26. The f r a c t u r e s t r a i n decreases with increasing fluence up t o
t h e two highest fluence levels, where t h e s t r a i n s are about equivalent.
This heat of material shows a d u c t i l i t y minimum with t h e f r a c t u r e s t r a i n
increasing s l i g h t l y with decreasing creep r a t e except f o r t h e samples
showing t h e lowest fluence.
The t e n s i l e and creep t e s t r e s u l t s f o r heat 5065 have been combined
i n Fig. 27.
These r e s u l t s again show the marked dependence of the frac-
ture s t r a i n on t h e s t r a i n rate f o r t h e i r r a d i a t e d material.
A comparison
with t h e similar p l o t f o r heat 5085 (Fig. 22) reveals some s l i g h t differences i n t h e f r a c t u r e s t r a i n s of t h e two heats, but shows generally
34
70
60
c
50
-nn
8 40
-0
u)
8
a
30
5
20
40
0
40-2
io-'
too
MINIMUM CREEP RATE (%/hr)
F i g . 25. Minimum Creep Rate of Hastelloy N (Heat 5065) Surveillance Samples from the NRE at 650°C.
ORNL-DWG
ERMAL FLUENCE
69-4476R
1
6
5
z
a 4
a
$
I
$3
u
!E
2
4
0
to-'
46'
40to-'
MINIMUM CREEP RATE (%/hr)
ioo
io'
Fgg. 26. Variation of Fracture Strain with Strain Rate for
HasteUoy N (Heat 5065) Surveillance Specimens at 650°C.
E
35
ORNL-DWG 70-7299R
50
40
30
-
e,
5
a
a
f
10
W
8
a
2
V
a
I I I11111
IRRADIATED
I-
a
1
IL
/
TENSILE TESTS
4
. 0 CREEP TESTS
.
___~_
--e
-0-
9.
0
c’
-
I
~~
0-
2-4h----
4’
a/
+
i
/’
I
I I1111
6
/* e**
++*
ro3
P
Fig. 27. Fracture Strains a t 65OOC of Heat 5065 i n t h e Unirradiated and t h e Irradiated Conditions. Irradiated t o a thermal fluence
of 1.5 X 1O2I neutrons/cm’ over a period of 22,533 hr a t 650°C.
Heat 5065
t h a t t h e two heats respond t o i r r a d i a t i o n q u i t e similarly.
contains about 23 ppm B (average of t h e four values i n Table 1, p. 41,
and t h e v a r i a t i o n of t h e f r a c t u r e s t r a i n with helium content i s sham
The behavior i s quite similar t o that noted i n Fig. 23 f o r
i n Fig. 28.
heat 5085.
Thus, t h e lower boron (and hence helium) concentration i n
heat 5065 r e s u l t s i n about t h e same properties as t h e higher boron l e v e l
i n heat 5085.
Mechanical P
Two heats of m o d i
6244 hr a t 650°C and rece
neutrons/cm2 (Table 2, p
titanium
- heat
Data -Modified Hastelloy N
loy N were exposed t o t h e MSRE core f o r
a thermal neutron fluence of 5.1 X lo2’
Both of these alloys were modified with
7320 contained 0.65% and heat 67-551 contained 1.1$.
The t e n s i l e properties of heat 7320 a r e summarized i n Tables 11 and
l2 f o r t h e control and i r r a d i a t e d samples, respectively.
b
The fracture
36
c
0
0.2
0.5
1
2
5
10
20
50
HELIUM CONTENT (ppm)
F i g . 28. Variation of Fracture S t r a i n a t 650°C with Calculated
Helium Content and S t r a i n Rate f o r Heat, 5065.
Table ll. Results of Tensile Tests on Control Samples
of Heat 7320a
Test
Specimen TemperNumber ature
(OC)
9464
9482
9470
9481
9472
9456
9467
9466
9458
9476
9478
9461
9475
9474
9477
9480
9471
9457
25
200
400
400
500
500
550
550
600
600
650
650
650
650
760
760
850
850
Strain
Rate
Stress, psi
Ultimate
(dn-1) Yield
0.05
0.05
0.05
0.002
Tensile
68,800 122,600
60, OOO 1W7,600
56,600 103,300
54, 100 96,000
0.05
99,200
54,500
0.002
53,600
96,700
0.05
51,700 95,500
0.002
49,400
85,300
0.05
52,500 91,900
54,800 80,500
0.002
0.05
49,300
78,600
0.002
47,700
72,800
0.0002
50,800 67,900
0. oo0027 47,800
59, loo
50,100 71,500
0.05
44,200
45, 100
0.002
0.05
45,000
48,200
0.002
27,800
27,500
Elongation, $
Uniform Total
48.2
40.6
46.9
45.2
47.9
45.1
49.2
36.7
41.4
18.2
28.4
14.5
12.3
5.6
12.0
4.2
5.6
1.6
49.7
41.9
49.1
46.2
51.1
46.2
52.0
37.7
45.5
19.7
30.6
15.5
13.7
9.6
13.1
17.2
10.2
10.1
Reduction
in ($)
Area
40.83
43.75
43.09
40.64
44.38
41.21
45.39
32.99
41. 11
23.35
36.67
19.46
U .40
13.88
ll.69
19.07
12.39
ll.97
True
Fracture
Strain
($1
52
58
56
52
59
53
61
40
53
27
46
22
16
15
12
21
13
13
aAnnealed 1hr a t ll77"C and exposed t o a vessel of s t a t i c barren fuel
s a l t f o r 7244 hr a t 650°C.
37
h-,
Table 12.
Postirradiation Tensile Properties of
Hastelloy N (Heat 7320)&
i
9221
9222
9223
9224
9225
9226
9215
92U
9213
9212
9211
9210
9227
9228
9209
9208
9207
9206
25
200
400
400
500
500
550
550
600
600
650
650
650
650
760
760
850
850
0.05
0.05
0.05
0.002
0.05
0.002
0.05
0.002
0.05
0.002
0.05
0.002
0.0002
0.000027
0.05
0.002
0.05
0.002
69,800 116,600
59,000 102,900
54,100 93,500
58,200 94,100
54,900 89,100
55,400 83,300
61,100 88,200
61,000 74,700
60,800 81,700
61,700 69,800
67,400 74,900
57,100 68,500
49,500 58,200
37,500 49,400
56,700 63,700
48,700 48,900
37,900 38,200
24,800 24,800
43.8
40.7
40.5
38.0
39.1
31.1
26.8
9.2
18.4
4.3
8.0
4.3
4.3
2.0
3.8
1.6
1.6
1.0
45.2
35.48
41.7
41.6
38.7
41.1
32.5
27.6
11.1
19.6
5.5
9.5
5.1
4.7
2.9
4.0
3.0
2.2
1.6
35.69
33.09
31.67
36.97
26.61
26.26
21.27
27.74
11.83
7.10
9.08
7.54
3.00
6.I5
6.15
7.25
2.23
&Annealed 1h r a t 1177°C. Irradiated t o a thermal fluence of 5.1
neutrons/cm2 over a period of 7244 hr a t 650°C.
s t r a i n s a r e s h m as a function of t e s t temperature i n Fig. 29.
44
44
40
38
46
31
30
24
32
I3
7
10
8
3
6
6
8
2
lo2'
X
Irra-
d i a t i o n reduces t h e f r a c t u r e s t r a i n over the e n t i r e range of t e s t temperatures, but t h e magnitude of the decrease i s much greater a t elevated
temperatures.
The sharp drop i n fracture s t r a i n with increasing temper-
a t u r e occurs a t a lower temperature f o r t h e i r r a d i a t e d material.
The
r a t i o s of t h e i r r a d i a t e d and unirradiated properties are shown i n
Fig. 30.
The y i e l d s t r e s s i s higher f o r the i r r a d i a t e d material over
the t e s t temperature range of 550 t o 760°C.
The ultimate t e n s i l e strength
i s about t h e same f o r t h e i r r a d i a t e d and unirradiated material but sham
a gradual decline with increasing t e s t temperature.
The f r a c t u r e s t r a i n
decreased due t o irradiation, with the reduction being about 80s a t 850°C.
Some idea of t h e t e n s i l e properties of heat 7320 due t o aging and
i r r a d i a t i o n can be obtained from Table 13. The yield strengths a t 25 and
650°C are increased by aging; at a t e s t temperature of 650°C a f u r t h e r
P
u
increase occurs due t o i r r a d i a t i o n .
The changes i n ultimate t e n s i l e
38
ORNL-DWG 70-7501
60
50
-0
n
H)o
200
300
400
500
600
TEST TEMPERATURE ('C)
700
800
900
Fig. 29. Fracture S t r a i n s of Heat 7320 After Removal from t h e MSRE
and from t h e Control Facility. A l l samples annealed 1h r a t 1177°C
before i r r a d i a t i o n t o a thermal fluence of 5.1 X lo2' neutrons/cm* over
a period of 7% h r at 650°C.
1.4
f.2
0.2
0
0
100
200
300
400
500
600
TEST TEMPERATURE E)
700
800
900
Fig. 30. Comparison of Unirradiated and Irradiated Tensile Pr
t i e s of Heat 7320 After I r r a d i a t i o n t o a Thermal Fluence of 5.1 X 10
neutrons/cm* Over a Period of 7244 hr at 650°C. Tested a t a s t r a i n r a t e
of 0.05 min-l.
TF-
39
Table 13. Comparison of the Tensile Properties of
Heat 7320 Before and A f t e r Irradiationa
Heat Treatment
Yield Stress, p s i
a t 25°C a t 650°C
Annealed 1 hr a t 1177°C
38,500
Annealed 1hr a t ll77"C, 68,800
aged 7244 hr a t 650°C
Annealed 1hr a t 1177"C, 69,800
irradiated f o r 7244 h r
a t 650°C t o a thermal
'Itimate
Stress, p s i
e
f i a c t u r e Strain,
a t 25°C a t 650°C
a t 25°C
a t 650°C
25,600
49,300
107,300
122,600
72,700
78,600
75.4
48.1
53.5
30.6
67,400
116,600
74,900
45.2
9.5
fluence of
5.1 X lo2' neutrons/cm2
a
Tested a t a s t r a i n r a t e of 0.05 min'l.
strength a r e r a t h e r small and l i k e l y within experimental error.
The
f r a c t u r e s t r a i n i s decreased by aging and by irradiation; t h e magnitude
of the decrease i s much l a r g e r at t h e t e s t temperature of 650°C.
The r e s u l t s of t e n s i l e t e s t s on the control and irradiated samples
of heat 67-551 are given i n Tables
i
U- and 15, respectively.
The f r a c t u r e
s t r a i n i s shown as a flxnction of t e s t temperature i n Fig. 31.
The frac-
ture s t r a i n a t low temperatures i s not influenced by irradiation, but
above 400°C t h e f r a c t u r e s t r a i n i s decreased by i r r a d i a t i o n .
However,
t h e f r a c t u r e s t r a i n s a t high temperatures a r e higher than those f o r
e i t h e r heat 7320 (Fig. 29) o r t h e standard a l l o y (Figs. 9 and 11, pp. 18
and 22, respectively).
The r a t i o s of t h e i r r a d i a t e d and unirradiated
t e n s i l e properties a r e shown i n Fig. 32.
The yield stress i s generally
about 25% lower f o r t h e i r r a d i a t e d material.
The u l t h a t e t e n s i l e
strength i s reduced only about lO$ by i r r a d i a t i o n when t e s t e d below 55OoC,
but i s reduced up t o 30%as t h e t e s t temperature i s increased.
The frac-
ture s t r a i n i s unaffected by i r r a d i a t i o n up t o test temperatures of 500°C;
above t h i s temperature the
cture s t r a i n decreases progressively with
increasing t e s t temperatur
Some values of t h e t e n s i l e properties a t
25 and 650°C a r e given i n Table 16.
Aging a t 650°C seems t o increase
t h e yield strength a t both 25 and 650°C; however, t h e strength of t h e
P
i r r a d i a t e d material i s quite close t o t h a t of t h e as-annealed material.
dd
The ultimate t e n s i l e strength i s increased o n l y s l i g h t l y by aging and
Table l-4. Results of Tensile Tests on Control Samples of Heat 6?-55la
Test
Temperature
Specimen
Number
( O C 1
9403
9419
9422
9417
9506
9423
9402
9418
9416
9428
9410
9407
9426
94x7
9424
9421
9425
9420
25
200
400
400
500
500
550
550
600
600
650
650
650
650
760
760
850
850
'Annealed
65OOC.
CI,
t
Strain
Rat e
Stress, p s i
(min-1)
0.05
0.05
0.05
0.002
0.05
0.002
0.05
0.002
0.05
0.002
0.05 ,
0.002
0.0002
0.000027
0.05
0.002
0.05
0.002
Tensile
U1tihate
62,400
54,100
52,300
51,400
48,100
50,400
45,200
47,600
42,800
46,200
41,900
42,700
46,000
39,900
45,600
43,000
41,300
25,500
120,200
108,700
104,300
101,700
89,900
92,100
94,100
80,400
87,800
79,600
84,600
76,300
74,800
60,900
78,000
45,200
5 1,000
26,200
Elongation, $I
Uniform Total
50.3
47.5
48.7
50.4
50.4
42.0
52.0
26.3
46.8
23.1
40.4
27.3
22.3
11.9
21.6
5.5
6.2
2.3
52.3
50.1
50.4
51.5
53.6
43.6
53.7
27.9
49.8
25.3
42.6
31.5
30.2
21.7
28.8
36.2
23.6
22.4
Reduct ion
i n Area
True
Fracture
($1
$1
41.59
45.22
46.66
43.84
46.58
36.67
41.78
31.67
43.09
26.37
37.07
25.42
28.87
23.23
28.87
33.98
26. I&
18.81
Strain
54
60
63
58
63
46
54
38
56
31
46
29
34
26
34
42
30
21
1 hr a t 1177°C p r i o r t o exposure t o a vessel of s t a t i c barren f u e l s a l t f o r 7244 h r a t
Table 15.
Spech e n
Number
9167
9168
9169
9175
9176
9172
9161
9160
9159
9158
9157
9156
9173
9174
9155
9154
9153
9152
Test
Temperature
("C)
25
200
400
400
5 00
500
550
550
600
600
650
650
650
650
760
760
850
850
Postirradiation Tensile Properties of Hastelloy N (Heat 67-551)"
Strain
Rat e
(min-l)
0.05
0.05
0.05
0.002
0.05
0.002
0.05
0.002
0.05
0.002
0.05
0.002
0.0002
0.000027
0.05
0.002
0.05
0.002
Stress, p s i
Ultimate
Tensile
-
49,600
39,800
39,200
40,400
37,000
37,000
34,500
39,500
37,000
41,900
31,300
31,200
35,100
34,800
29,700
29,200
26,700
22,300
107,800
94,700
91,800
88,100
83,400
67,900
80,400
57,100
71,300
61,800
56,600
53,300
57,100
51,300
51,500
38,400
34,700
22,700
Reduction
in Area
Elongation, '$
True
Fracture
Uniform
Total
( 8)
Strain
50.6
50.0
51.0
53.2
50.7
52.6
52.9
36.87
42.15
38.44
39.69
38.34
29.21
34.92
23.35
39.59
22.68
28.76
19.20
L3.74
12.5
8.79
46
55
49
51
48.7
51.3
52.1
25.8
42.8
14.7
31.9
18.0
20.7
16.0
12.3
7.5
13.8
5.7
4.8
2.2
27.4
44.5
17.0
33.9
19.2
22.6
16.9
14.5
12.9
14.8
9.2
6.5
6.2
'Annealed 1 hr a t 1177°C. Irradiated t o a thermal fluence of 5 . 1 X '
0
1
period of 7244 h r a t 650°C.
10.74
7.54
5.24
(8)
48
35
43
27
50
26
34
21
15
13
9
11
8
neutrons/cm2 over a
5
42
Fig. 31. Fracture Strains of Heat 67-551 After Removal from the
MSRE and from t h e Control Facility. All samples annealed 1hr a t 1177°C
before i r r a d i a t i o n t o a thermal fluence of 5 . 1 X lo2' neutrons/cm2 over
7244 h r a t 650°C.
ORNL-OWG
70-7304
1.2
4.0
0.2
0
0
(00
200
300
400
5-90
600
TEST TEMPERATURE CC)
700
800
900
Fig. 32. Comparison of Unirradiated and Irradiated Tensile Propert i e s of Heat 67-551 After I r r a d i a t i o n t o a Therm1 Fluence of 5.1 X lo2'
neutrons/cm2 over a period of 7244 hr at 650°C. Tested a t a s t r a i n r a t e
of 0.05 min-l.
43
dd
Table 16.
Comparison of the Tensile Properties of Heat 67-551
Before and After Irradiationa
4
Heat Treatment
Ultimate Tensile
Yield Stress, p s i
Stress, p s i
Fracture Strain, $
a t 25°C at 650°C at 25°C at 6 5 0 " ~ a t 25°C a t 650°C
Annealed 1hr a t 1177°C
44,600
26,900
113,600
78,900
79.6
57.3
Annealed 1h r a t 1177"C,
aged 7244 hr a t 650°C
62,400
41,900
120,200
84,600
52.3
42.6
Annealed 1h r a t 1177"C,
aged 7244 hr at 650"C,
irradiated t o a thermal
fluence of
5.1 x
neutrons/cm2
49,600
31,300
107,800
56,600
51.0
22.6
&Tested a t a s t r a i n r a t e of 0.05 min-l.
decreased by irradiation.
A t 25°C t h e f r a c t u r e s t r a i n i s reduced from
79.6% t o 52.3% by aging, and i r r a d i a t i o n does not cause any flxrther
change.
A t 650°C t h e f r a c t u r e s t r a i n i s decreased by aging and decreased
f'urther by i r r a d i a t i o n .
The creep-rupture properties of heat 7320 a r e summarized i n
Table 17.
These results are compared i n Figs. 33, 34, and 35 with those
f o r unirradiated samples t h a t were given a p r e t e s t anneal of 1 h r a t
1177°C. The stress-rupture properties i n Fig. 33 show t h a t t h e aging
treatment given t h e controls, 7244 h r a t 75O9C, increased t h e rupture
The i r r a d i a t e d samples f a i l e d a f t e r longer times than did the
unirradiated samples that w e r e s i m p l y annealed 1 hr at 1177"C, but
life.
ruptured i n shorter times than the control samples.
However, the minimum
creep r a t e s shown i n Fig, 34 seem t o f a l l within a common s c a t t e r band
f o r a l l conditions. Thus, neither i r r a d i a t i o n nor aging seems t o have a
detectable e f f e c t on t h e minimum creep r a t e . The f r a c t u r e s t r a i n is
shown as a function of s t r a i n rate i n Fig. 35.
The f r a c t u r e s t r a i n s of
the unirradiated samples vary from about SO$ i n a t e n s i l e test t o about
lO$ i n a long-term creep test. The i r r a d i a t e d samples have lower f'ract u r e s t r a i n s t h a t vary from 9.5$ f o r t h e f a s t e s t t e n s i l e t e s t t o 2 t o 3$
f o r creep tests.
44
Table 17.
Test
Number
Creep-Rupture Tests on Heat 7320 a t 650°C
Specimen
Number
Unirradiated
7276
10329
7277
7279
7278
10252
7280
7013
7425
70l.4
7016
7015
7424
7017
Stress
(psi)
Lify( 4 )
ture
Strain
(hr)
- Annealed
55,000
50,000
47,000
43,000
40,000
35,000
30,000
Minimum
Creep
Reduct ion
i n Area
(7
8
1 h r a t 1177°C p r i o r t o t e s t
4.7
11.5
49.9
103.4
384.6
1602.1
1949.5
28.7
21.2
21.8
18.1
16.3
27.3
10.9
29.6
28.3
31.3
20.5
17.6
30.6
19.9
0,190
32.9
26.3
29.3
0.25
0.045
0.019
0.015
0.0006
0.125
0.0194
0.0069
0.0059
0.0056
0.0017
Unirradiated Controls‘
7885
7991
7886
7887b
7884
9462
9468
9463
9465
9460
55,000
47,000
40,000
40,000
32,400
32.5
224.1
653.7
501.9
2420.2
U.2
15.3
17.9
7.9
6.9
21.4
4.8
Irradiat edc
R-1151
R- 1016
R-951
R-955
R- 967
R-950
‘
9204
9216
9217
9219
9218
9220
63,000
63,000
55,000
47,000
40,000
32,400
1.7
4.0
12.3
95.1
329.9
2083.3
12.2d
2.0
2.7
2.3
3.2
3.1
1.2
0.44
0.20
0.018
0.0074
0.0058
a
Annealed 1 h r a t 1 1 7 7 ° C and exposed t o a vessel of s t a t i c barren
f u e l s a l t f o r 7244 hr a t 650°C.
bDiscontinued p r i o r t o f a i l u r e .
5.1
X
‘Annealed 1h r a t 1177°C. Irradiated t o a thermal fluence of
10” neutrons/cm2 over a period of 7244 h r a t 650°C.
%est loaded so t h a t s t r a i n on loading was included.
t e s t s did not include t h i s s t r a i n .
A l l other
f
45
ds
2
5
5
2
10'
402
5
2
RUPTURE TIME lhrl
Stress-Rupture Properties of Heat 7320 a t 650°C.
Fig. 33.
ORNL-OWG 70-7306
70
c
60
50
-a
.)
0
0
-
40
0
Y)
u)
E
30
II
0
IRRADIATED, S.lX4&
20
to
0
to-'
2
2
5
to-2
2
S
to-'
2
5
COO
MINIMUM CREEP RATE (%/hr)
Fig. 34.
Creep Properties a t 650°C of Heat 7320.
.
46
_.
ORNL- DWG 70-7307
B
I I
IRRADIL
10-3 2
5
10-2 2
5
lo-'
2
5
loo
2
5
(0'
2
5
(02
2
5
103
STRAIN RATE (%/hr)
Fig. 35. Fracture Strains a t 650°C of Heat 7320 i n t h e Unirradiated
and Irradiated Conditions. Irradiated t o a thermal fluence of 5 . 1 x lo2'
neutrons/cm2 over a period of 7244 hr at 650OC.
The r e s u l t s of creep-rupture t e s t s on heat 67-551 with 1.1%
T i are
given i n Table 18.
The stress-rupture properties of t h e c o n t r o l and
irradiated samples a r e compared i n Fig. 36 f o r 650°C.
The i r r a d i a t e d
samples generally f a i l i n s l i g h t l y shorter times than t h e unirradiated
samples.
The minimum creep'rate s e a s t o be unaffected by irradiation,
although t h e r e a r e two data points t h a t seem t o be anomalous (Fig. 37).
The f r a c t u r e s t r a i n s of t h i s a l l o y (Fig. 38) a r e higher i n both t h e
unirradiated and i r r a d i a t e d conditions than those f o r heat 7320. The
f r a c t u r e s t r a i n s of t h e unirradiated samples varied over t h e range of
43 t o 25% and those of t h e i r r a d i a t e d samples varied from 22.6 t o 5.84
over t h e range of s t r a i n r a t e s studied.
One heat of modified Hastelloy N containing 0.49% H f (heat 67-506)
was exposed t o t h e c e l l environment f o r 17,033 h r a t 650°C.
fluence was 2.5 X 1019 neutrons/cm*.
The.therma1
This same heat of material was
included previously i n our surveillance program and was exposed t o a
b
47
bi
Table 18. Results of Creep-Rupture Tests on Heat 67-551 a t 650°C
Test
Number
7872
7871
Specimen
Number
Stress
(psi)
Rupture
Life
Strain
( 8)
(hr)
Reduction
i n Area
( $)
Unirradiated, Annealed 1 h r at 1177°C before t e s t i n g
55,000
U.2
29.5
33.2
6889
6884
47,000
U3.8 33.6
26.0
Minimum
Creep
0.32
0.071
Unirradiated Controlsa
7882
7880
7883
7881
9409
9404
94u.
9408
55,000
47,000
40,000
32,400
44.7
271.2
956.2
2034.9
25.5
29.1
30.4
25.2
28.5
29.6
25.5
29.9
0.190
0.067
0.018
0.0053
b
Irradiated
R- l I 5 O
R- 1036
R-949
R-961
R-948
R- 947
9150
9151
9 162
9165
9164
9163
63,000
63,000
55,000
47,000
40,000
32.400
0.9
0.6
50.5
20.8
2.2
5.8
117.1
8.3
746.9
U85.6
9.1
12.5
2.2
2.6
0.079
0.058
0 0079
0 0053
a
Annealed 1 h r a t 1177°C p r i o r t o exposure t o a vessel of s t a t i c
barren fuel s a l t f o r 7244 h r a t 650°C.
5.1
bAnnealed 1h r a t 1177°C. Irradiated t o a thermal fluence of
lo2' neutrons/cm2 over a period of 7244 h r a t 650°C.
X
C
Test loaded t o include s t r a i n on loading.
include t h i s s t r a i n .
thermal fluence of 5.3
X
lo2'
A l l other t e s t s did not
neutrons/cm* while being a t temperature i n
t h e core f o r 9789 h r ( r e f . 10). I n t h e group of samples presently being
discussed, there were two rods of heat 67-504.
Two heats of material
should have been involved, but p o s t i r r a d i a t i o n chemical analysis revealed
t h a t both rods were made of t h e same material.
This was not discovered,
u n t i l after many of t h e samp es were tested, so we have several tests
under duplicate conditions.
-
loH. E. McCoy, An Evaluation of t h e Molten-Salt Reactqr Experiment
Hastelloy N Surveillance Specimen Third Group, ORNETM-2647 (1970).
-
I
I
/
I
I
I
I
n
N
"2
n
N
-
Q
n
N
"2
n
4%
PI0
ard
0
J
z
6
I
I
I
I
I
I
I
I
I
I
I
'0
N
'h
n
N
W
i
t
9 8
i r
z
!!
5
"f
N
2
N
n
I %
alo
h aO
H
49
ORNL- DWG 70- 7340
60
40
20
0-3 2
5
40-2
2
1
5
.1
lo-'
2
5
1 8
2
STRAIN RATE (%/hr)
TENSILE TESTS
CREEP TESTS
10'
2
5
102
2
Fig. 3 8 . Comparison of t h e Fracture Strains of Unirradiated and
Irradiated Heat 67-551 a t 650°C. Irradiated t o a thermal fluence of
5 . 1 X lo2' neutrons/cm2 over a period of 7244 h r a t 650°C.
The r e s u l t s of the p o s t i r r a d i a t i o n t e n s i l e t e s t s on heat 67-5W
a r e given i n Table 19.
The p o s t i r r a d i a t i o n f r a c t u r e s t r a i n s a r e shown
i n Fig. 39 as a function of temperature and s t r a i n r a t e .
The f r a c t u r e
s t r a i n generally decreases with increasing temperature above about 400°C
and with decreasing s t r a i n r a t e .
those reported previously1'
The r e s u l t s from the present t e s t s and
show some very s t r i k i n g changes i n flracture
s t r a i n with varying h i s t o r i e s (Fig. 40).
The samples were a l l i n i t i a l l y
annealed 1hr a t 1177°C before being given t h e treatment indicated i n
Fig. 40. I n t h e as-annealed
ondition, t h e a l l o y has a fracture s t r a i n
of 70 t o 804 up t o about 6OO0C, where t h e f r a c t u r e s t r a i n drops precipA
itiously.
Aging f o r 9789 hr a t 650°C reduced the f r a c t u r e s t r a i n at
t
lower temperatures and increased it above about 700°C.
However, t h e
-
f r a c t u r e s t r a i n s a r e i n t h e range of 40 t o 50$ over t h e e n t i r e range of
ks
temperatures studied.
I r r a d i a t i o n t o a thermal fluence of 2.5
neutrons/cm2 over a period of 17,033 hr i n N2
+
X
lo1'
2 t o 5% 02 resulted i n
50
Table 19. Postirradiation Tensile Properties of
Hastelloy N (Heat 67-504)&
,
Test
Specimen TemperNuniber
ature
("C)
5 u
51k7
5llO
5162
5109
5u5
5108
5 M
5107
5u3
5106
5l42
5u7
5153
5 m
5154
5096
53.64
5120
5156
5128
5121
5155
5122
5158
5105
5l41
5UO
5104
5098
5123
5159
5l24
5160
5102
5138
5103
5139
25
25
200
200
400
400
400
400
500
500
500
500
550
550
550
550
600
600
600
600
650
650
650
650
650
650
650
650
650
760
760
760
760
760
850
850
850
850
Strain
Rate
Yield
Elongation, $
Uniform Total
(min-1)
0.05
0.05
0.05
0.05
0.05
0.05
0.002
0.002
0.05
0.05
0.002
0.002
0.05
0.05
0.002
0.002
0.05
0.05
0.002
0.002
0.05
0.05
0.05
0.002
0.002
0.0002
O.OOO2
-0.m27
-0.m27
0.05
0.05
0.05
0.002
0.002
0.05
0.05
0.002
0.002
50,300 112,200
50,700 112,800
41,000 96,400
40,900 95,300
36,700 90,800
37,500 90,800
36,700 91,800
41,100 97,900
40,600 92,700
40,900 94,000
41,600 90,500
33,800 78,600
39,700 74,900
34,700 77,800
44,600 62,000
33,200 59,900
31,900 74,500
32,900 70,800
36,400 49,600
30,900 49,500
39,900 55,500
38,800 59,600
30,200 46,200
52,200 54,500
31,700 44,800
32,200 40,800
30,700 38,800
30,200 37,100
30,700 37,200
27,OOO 38,600
33,300 40,100
28,800 38,OOO
33,300 33,300
28,200
34,600
27,300 29,700
27,500
30,900
21,100 21,100
23,700 23,700
45.9
54.6
43.0
45.7
49.8
52.8
52.2
52.7
45.7
49.6
44.5
41.1
34.5
39.7
11.9
20.5
34.7
32.9
8.9
12.6
24.0
U.6
12.8
2.2
u.2
7.4
7.8
4.4
4.8
9.0
6.7
9.6
1.2
4.2
2.9
3.2
1.0
1.0
48.1
56.6
44.8
47.3
52.0
54.8
54.7
53.7
47.2
50.6
46.2
42.6
36.5
43.5
15.1
21.5
36.9
34.0
10.9
15.7
25.8
16.5
15.8
6.2
u.3
9.8
8.8
6.5
6.5
23.7
8.7
13.5
2.7
5.6
4.3
4.5
1.9
2.5
Reduction
i n Area
(4 )
34.63
36.54
38.34
46.66
36.15
34.84
45.71
33.31
32.23
39.97
39.40
36.00
31.45
37.94
16.02
20.38
26.02
35.69
20.38
23.35
22.06
19-15
10.17
3.51
ll.54
13.52
6.65
5.1
5.4
18.94
10.17
13.19
4.63
6.00
2.40
7.10
0.64
0.48
aAnnealed 1 hr a t ll77"C. Irradiated t o a thermal fluence of 0.25
neutrons/cm2 over a period of 17,033 hr at 650°C.
X
1020
Rue
Fracture
Strain
($1
43
45
48
63
45
43
61
41
39
51
50
45
38
48
17
23
30
44
23
27
25
21
11
4
12
15
7
5
6
21
11
u
5
6
2
7
1
0
51
ORNL-OWG 70-734t
60
50
'ii 40
c
z
2
w 30
3
cc
I
c
0
a
(L
LL
o
0.05
A 0.002
a
20
__
0.0002
(0
0
0
400
200
300
400
500
600
TEST TEMPERATURE ("C)
700
800
900
Fig. 39. Fracture Strains of Heat 67-504 After I r r a d i a t i o n t o a
Thermal Fluence of 2.5 X 10'' neutrons/cm2 f o r 17,033 h r a t 650°C.
ORNL-DWG 70-7312
.
"
0
~
0
0
2
0
0
3
0
0
4
a
J
6
~ 0 0 7 0 0 8 0 0 9 0 0
TEST TEMPERATURE (TI
Fig. 40. Comparison of t h e Fracture Strains of Heat 67-504 i n
Various Conditions When Tested at a S t r a i n Rate of 0.05 min'l.
52
f r a c t u r e s t r a i n s of about 50% up t o 50O0C, above which t h e fracture
s t r a i n dropped progressively with increasing temperature.
t o a thermal fluence of 5.3
X
lo2'
Irradiation
t
neutrons/cm2 a t 650°C i n f u e l s a l t
over a period of 9789 h r resulted i n a low f r a c t u r e s t r a i n a t 25°C (which
may be anomalous), f r a c t u r e s t r a i n s of about 45% up t o 500°C, and
decreasing fracture s t r a i n s with increasing t e s t temperature.
The most
s t r i k i n g feature i s t h a t above 550°C t h e f r a c t u r e s t r a i n s a r e lower f o r
material i r r a d i a t e d i n N2 + 2 t o 5% 02 f o r 17,033 h r t o a fluence of
2.5 x loi9 neutrons/cm2 than f o r those i r r a d i a t e d i n f u e l salt f o r
9789 h r t o a fluence of 5.3 X
lo2* neutrons/cm2.
Some values of t h e t e n s i l e properties a t 25 and 650°C are given i n
Table 20. The yield stress a t both t e s t temperatures was increased by
aging and by i r r a d i a t i o n .
The ultimate t e n s i l e s t r e s s a t 25°C was
increased by aging and by irradiation; a t 650°C it was increased by
aging, but decreased.by i r r a d i a t i o n .
We have already discussed the
changes i n the f r a c t u r e s t r a i n .
Table 20.
Comparison o f t h e Tensile Properties of Heat 67-504
After Various Treatmentsa
Heat Treatment
Yield Stress, p s i
a t 25°C a t 650°C
Ultimate
Stress,. p- s i
at 25°C at 650°C
Fracture Strain,
s
a t 25°C
a t 65OOC
73.2
65.9
52.3
50.9
Annealed 1hr a t ll77"C
37,400
25,600
105,000
Annealed a t U77"C and
aged 978gb hr a t 650°C
67,500
43,300
133,000
83,000
94,300
Annealed a t UWOC, i r r a - 102,OOO
diated tob a t h e m 1
fluence of 5.3 x
neutrons/cm2 over
9789 h r ( s a l t
environment)
32,900
ll9,300
70,000
27.0
30.7
Annealed a t ll77"C,
irradiated t o a therm a l fluence of
2.5 X lof9 neutrons/cm2
Over 17,033 hr (N2 + 2
t o 55 92 environment)
38,800
ll2,OOO
59,600
48.1
16.5
50,300
.
BTested a t a s t r a i n r a t e of 0.05 min-1.
bData from: H. E. McCoy, Jr., A n Evaluation of the Molten-Salt Reactor Experiment
1
m
l
.
Hastelloy N Surveillance Specimens Third Group, ORNLTM-2647 (
-
53
The r e s u l t s of creep-rupture t e s t s on heat 67-504 a r e summarized
i n Table 21.
These r e s u l t s and those obtained previously1'
t o prepare Figs. 41, 42, and 43.
were used
The stress-rupture properties i n
Fig. 4 1 show t h a t t h e rupture l i f e was not changed appreciably by aging,
a t l e a s t f o r 9789 h r a t 650°C.
The stress-rupture l i f e was reduced at
l e a s t two orders of magnitude by t h e lower fluence and only a f a c t o r of
2 by t h e higher fluence.
The minimum creep r a t e s i n Fig. 42 show t h a t
t h i s property was not changed a s much as t h e rupture l i f e .
Aging f o r
9789 h r increased t h e minimum creep, but samples i r r a d i a t e d over t h e
same period had a lower creep r a t e than t h e aged samples.
The samples
i r r a d i a t e d t o t h e lower fluence had a creep r a t e an order of magnitude
higher than t h a t of the as-annealed material.
The f r a c t u r e s t r a i n s a r e
'
H
.
E. McCoy, An Evaluation of t h e Molten-Salt Reactor Experiment
Hastelloy N Surveillance Specimen Third Group, ORNETM-264'7 (1970).
-
Table 21.
Test
Number
Creep-Rupture Tests on Heat 67-504 a t 650°C
Specimen
Number
Stress
(psi)
Rupture
Life
Strain
(b)
(4)
Minimum
Creep
I r r a d i a t e d a t 650°C t o a thermal fluence of 2.5 X loi9 neutrons/cm2
R-952
R-953
R-959
R-965
R-971
R- 1017
R- 968
R- 1033
R- 1019
5112
5U8
5113
5u9
51U
5150
5115
5137
5116
55,000
55,000
47,000
47,000
40,000
0.6
1.0
7.6
3.3
u.5
33.8
235.5
268.7
l175.4
6.2
10.9
4.1
3.8
4.2
2.7
9.88
6.6
0.83
0.52
0.42
0.050
0.0071
0.0066
0.0016
32.9
31.2
27.2
28.7
16.3
3.7
0.16
0.029
0.0068
0.0030
8.6
10.6
12.1
Annealed 1h r at 1177°C
7432
7431
6255
6254
6253
6247
6245
4991
4936
4933
0,000
3,000
5,000
7,000
40,000
1.3
17.7
127.9
425.5
876.9
54
mNL-DWG 70-731s
m
U
60
D
50
20
0
5
t
O
0
2
5
d
2
5
r
o
2
2
5
d
2
5
RVPNRE TIME (hr)
Fig. 41.
Stress-Ftupture Properties at 650°C of Heat 67-504.
ORNL-DWG 70-7313
60
50
rn
m
W
E30
rn
20
'h'
0
F i g . 42. Creep-Rupture Properties at 650°C of Heat 67-504.
55
ORNL-OWG 70-73!4R
~
5
io-‘
2
5
io-’
2
5
400
2
5
40’
SOLID P
OPEN POINTS CREEP
o ANNEALED i hr AT 4f77OC
A ANNEALEDANDAGED
9789 hr AT 65OOC
ANNEALED,THERMAL FLUENCE
OF 5.3XtOZ0 neutrons/cm*
0 ANNEALED.THERMAL FLUENCE
OF 2.5xiOt9 neutrons/cm
-
2
5
io2
2
STRAIN RATE (70)
Fig. 4 3 . Comparison of the Fracture Strains of Unirradiated and
Irradiated Heat 67-504 a t 65OOC. [Data f o r samples aged f o r 9789 h r a t
650°C and f o r those i r r a d i a t e d t o 2.5 X lo1’ neutrons/cm2 from:
H. E. McCoy, An Evalwtion of t h e Molten-Salt Reactor Experiment
Hastelloy N Surveillance Specimen - Third Group, ORNLTM-2647 (1970). 1
shown as a f’unction of stra
rate in Fig. 43. Aging increased t h e fracture s t r a i n of t h e unirradiated material. A f t e r i r r a d i a t i o n t h e material
i
1
i r r a d i a t e d t o t h e lower fluence had t h e lowest f r a c t u r e s t r a i n .
Several samples of heat 67-504 were subjected t o tests t h a t were
interrupted.
The r e s u l t s of these t e s t s a r e summarized i n Table 22.
strained continuously a t 650°C a t a s t r a i n r a t e of 0.05 min-l, t h e
If
56
Table 22.
Type of Testa
Results of Interrupted Tensile
Tests on Heat 67-504
Test
Temperature
("C>
A
B
C
D
E
A
F
650
650
650
650
650
66.9
100.4
109.0
104.0
95.4
760
760
35.5
A
- Run uninterrupted
B
- Strained
a
Fracture Strain, $
Unirradiat ed
Irradiated
96.0
16.5
,
25.8
8.7
23.7
a t a s t r a i n rate of 0.05 min-l.
5$, held a t temperature f o r 2 min, cycle
repeated t o failure.
C
- Strained
54, held a t temperature f o r 10 min, cycle
repeated t o f a i l u r e .
D - Strained 55, held at temperature f o r 30 min, cycle
repeated t o f a i l u r e .
-
E
Strained 5$, held a t temperature f o r 60 min, cycle
repeated t o f a i l u r e .
F - Strained 34, held a t temperature f o r 60 min, cycle
repeated t o f a i l u r e .
material had a f r a c t u r e s t r a i n of 66.98 i f unirradiated and 16.5$ i f
irradiated.
If t h e material was s t r a i n e d 5 4 a t 650°C then annealed 1h r
a t 650°C and t h i s p a t t e r n continued u n t i l failure, the unirradiated samp l e failed with 95.44 s t r a i n and t h e i r r a d i a t e d sample with 25.84 s t r a i n .
A sequence of t e s t s was performed on unirradiated samples a t 650°C i n
which t h e annealing time was vaxied from 2 t o 60 min; t h e r e s u l t s show
t h a t t h i s variable had l i t t l e e f f e c t on t h e f r a c t u r e s t r a i n . A t a t e s t
temperature of 760°C t h e fracture s t r a i n was 35.5$ f o r an unirradiated
sample and 8.7% f o r an irradiated material. Samples were a l s o t e s t e d by
s t r a i n i n g 3$, annealing 1h r a t 76OoC, s t r a i n i n g 3$, and repeating t h i s
sequence u n t i l f a i l u r e . The fracture strains were 96.0 and 23.7% f o r t h e
unirradiated and irradiated samples, respectively. These tests have
57
demonstrated t h a t there i s not, f o r each t e s t temperature, a unique
s t r a i n a t which f a i l u r e occurs; rather, t h e fracture s t r a i n depends upon
t h e loading history.
Metallographic Examination of Test Samples
The t e n s i l e samples that were t e s t e d a t 25°C a t a s t r a i n r a t e of
0.05 rnin'l
and a t 650°C a t a s t r a i n r a t e of 0.002 min-l were examined.
An irradiated sample of heat 5085 from the core that was t e s t e d a t 25°C
i s shown i n Fig. 4.4.
The fracture i s primarily intergranular.
There i s
a high frequency of edge cracking t h a t extends t o a depth of about 4 mils.
The microstructure i n the 4-mil region near the surface etches differently.
A t a 650°C t e s t temperature (Fig. 45) the f r a c t u r e i s intergranular with
no evidence of p l a s t i c deformation of t h e individual grains except adjacent t o t h e fracture.
This sample a l s o had t h e modified structure and
edge cracks t h a t extend 10 mils i n t o t h e sample.
The control samples of
heat 5085 t h a t were annealed f o r 22,533 h r a t 650°C i n s t a t i c barren f u e l
s a l t a r e shown i n Figs. 46 and 4'7.
t e s t e d a t 25°C.
The sample sham i n Fig. 46 was
The f r a c t u r e i s predominantly intergranular, but there
are no edge cracks.
There i s a very t h i n layer near the surface where
t h e s t r u c t u r e i s modified s l i g h t l y .
The mLcrostructure i s generally
characterized by large &C-type carbides, many of which cracked during
deformation, and a network of almost continuous carbides along the grain
and twin boundaries.
intergranular.
When t e s t e d a t 650°C (Fig. 47) the f r a c t u r e i s
There is some edge cracking, but not as frequently as
noted i n Fig. 45 for t h e i r r a d i a t e d sample.
Another f a c t o r t h a t should
be considered i n comparing Figs. 45 and 47 i s t h a t t h e i r r a d i a t e d sample
i n Fig. 45 failed after s t r a i n i n g only 5 4 (Table 4, p.
a),whereas
the
unirradiated sample i n Fig. 47 strained 24% (Table 3, p. 17) before fracturing.
The higher s t r a i n should have increased t h e number and depth of
edge cracks.
The observed
t h e exposure t o t h e reactor
o s i t e trend indicates some influence of
ironment on the fracture characteristics
near t h e surface.
Heat 5065 was a l s o exposed t o t h e core f o r 22,533 h r a t 650°C. The
microstructure of a sample t e s t e d at 25°C i s sham i n Fig. 48. The
% X
B'
+
4,
Fig. 45. Photomicrographs of a Hastelloy N (Heat 5085) Sample Tested a t 650°C After Bei
%Exposed
t o t h e Core of t h e MSRE for 22,533 hr a t 650°C and Irradiated t o a Thermal Fluence of 1.5 X 10
neutrons/cm*. (a) Fracture, etched. lOOX. (b) Edge, as polished. loox. ( c ) Edge, as polished. 500X.
( d ) Edge, etched. 1OOX. Etchant: glyceria regia. Reduced 27%.
60
id
Fig. 46. Typical Photomicrographs of a Hastelloy N (Heat 5085)
Sample Tested a t 25°C After Being Exposed t o S t a t i c Barren Fuel S a l t f o r
22,533 h r a t 650°C. ( a ) Fracture, etched. 1OOX. (b) Edge near fiacture,
etched. 1OOX. ( c ) Representative unstressed structure, etched. 500x.
Etchant: glyceria regia. Reduced 24.54.
c
w
61
*
Fig. 47. Typical Photomicrographs of a Hastelloy M (Heat 5085)
Sample Tested a t 650°C ( S t r a i n Rate 0.002 min"')
After Being Exposed t o
S t a t i c Barren Fuel Salt for 22,533 hr a t 650°C. (a) Fracture. lOOX.
(b) Edge near fracture. 1OOX. ( c ) Representative unstressed structure.
500X. Etchant: glyceria regia. Reduced 25$.
29
63
6,
.
f r a c t u r e is mixed transgranular and intergranular. The same s t r u c t u r a l
modification and edge cracking that were noted i n heat 5085 (Fig. 4 4 )
a r e a l s o present i n heat 5065.
shown i n Fig. 49.
A sample of heat 5065 t e s t e d a t 650°C i s
The f r a c t u r e i s intergranular with l i t t l e evidence
of deformation of t h e adjacent grains.
The microstructure a t t h e edge
i s modified t o a depth of a b o u t 4 mils, and cracks extend t o a depth of
about 10 mils.
A control sample t e s t e d a t 25°C i s shown i n Fig. 50.
The f r a c t u r e i s mixed transgranular and intergranular, and the grains
are elongated i n t h e direction of stressing.
There i s no edge cracking.
There a r e copious amounts of carbides, both o f t h e large primary carbides and t h e f i n e r carbides formed during t h e long annealing treatment
a t 650°C. The microstructure of a control sample of heat 5065 t e s t e d a t
650°C (Fig. 51) shows an intergranular f r a c t u r e and some edge cracking.
Again t h e frequency of cracking i s l e s s than t h a t noted f o r t h e irradiated material shown i n Fig. 49.
Heat 7320 was exposed t o t h e MSRE core f o r 7244 h r a t 650°C. Typicalmicrographs of a sample t e s t e d a t 25°C a r e shown i n Fig. 52.
The
f r a c t u r e i s predominantly transgranular and t h e f l a w l i n e s and t h e elongated grains attest t o t h e deformation of t h e matrix.
There are some
edge cracks t o a depth of about 3 mils, but the microstructure i s not
modified a t t h e sample surface.
Microstructures of an i r r a d i a t e d sample
tested a t 650°C a r e shown i n Fig. 53.
The f r a c t u r e is intergranular.
There a r e edge cracks t h a t extend t o a depth of about 10 mils, but t h e
microstructure is not modified near the surface. A control sample t h a t
was t e s t e d a t 25°C i s sham i n Fig. 54.
The f r a c t u r e i s transgranular
with the grains being elongated i n t h e direction of stressing.
no edge cracking.
There i s
The unstressed microstructure gives a h i n t of the
very f i n e matrix p r e c i p i t a t i o n t h a t i s present i n t h i s alloy.
Photo-
micrographs of a control sample t e s t e d a t 650°C are shown i n Fig. 55.
The fracture i s intergranular, and t h e r e a r e no edge cracks such as were
.
noted i n t h e i r r a d i a t e d sample (Fig. 53).
Heat 67-551 was expos
e MSRE core f o r 7244 h r a t 650°C. The
f r a c t u r e a t 25°C (Fig. 56)
ed transgranular and intergranular, and
grains have deformed extensively.
W
The microstructure near the surface
i s not altered, but t h e r e a r e edge cracks t h a t extend t o a depth of about
"
I
Y
*
"
..
..-
...
Exposed
Fig. 49. Photomicrographs of a Hastelloy N (Heat 5065) Sample Tested a t 650°C After Bei
3
1
t o t h e Core o f t h e MSRF: f o r 22,533 h r a t 650°C and Irradiated t o a Thermal Fluence of 1.5 x 10
neutrons/cm2. (a) Fracture, etched. lOOX. (b) Edge, as polished. lOOX. ( c ) Edge, as polished. 5 0 0 ~ .
( d ) Edge, etched. 1OOX. Etchant: glyceria regia. Reduced 26.5%.
P
$
65
i
(sd
Fig. 50. Typical Photomicrographs of a Hastelloy N (Heat 5065)
Sample Tested a t 25°C After Being Exposed t o S t a t i c Barren Fuel S a l t f o r
22 533 hr a t 650°C. (a) Fracture. lOOX. (b) Edge near fracture. lOOX.
( c j Representative unstressed structure. 500X. Etchant: glyceria
regia. Reduced 22%.
66
,
.Fig. 51. Typical Photomicrographs of a Hastelloy N (Heat 5065)
Saniple Tested at 650°C (Strain Rate of 0.002 min-l) After Being Exposed
to Barren Fuel Salt for 22,533 hr at 650°C. (a) Eracture. 1OOx.
(b) Edge near fracture. 1OOX. (c) Representative unstressed structure.
500X. Reduced 21.59.
67
m
hd
Fig. 52. Photo&crographs of a Modified Hastelloy N (Heat 7320)
ing Exposed t o t h e MSRE Core f o r 7244 h r
Sample Tested a t 25°C Aft
a t 650°C and Irradiated t
luence of 5.1 x 1 O 2 O neutrons/cm2.
(a) Fracture, etched. 1OOX. (b) Edge, as polished. 1OOX. ( c ) Edge,
etched. lOOX. Etchant: glyceria regia. Reduced 22%.
68
Fig. 53. Photomicrographs of a Modified Hastelloy N (Heat 7320)
Sample Tested a t 650°C ( S t r a i n Rate, 0.002 min'l) After Being Exposed t o
the MSRE Core for 7244 hr a t 650°C and Irradiated t o a Fluence of
5.1 X lo2' neutrons/cm2. (a) Fracture, etched. 1OOX. (b) Edge, as
polished. 1OOX. ( c ) Edge, etched. 1OOX. Etchant: glyceria regia.
Reduced 224.
69
.
Fig. 54.
Photomicrographs of a Modified Hastelloy N (Heat 7320)
Sample Tested a t 25°C After Being Exposed t o Static Barren Fuel Salt f o r
7244 hr a t 650°C. (a) Fracture.
(b) Edge. 1OOX. ( c ) Typical
unstressed microstructure. 5 0 0 ~ . Etchant: glyceria regia. Reduced 22.5$.
loox.
kd
70
Fig. 55. Photamicrographs of a Modified Hastelloy N (Heat 7320)
Sample Tested a t 650°C (Strain Rate, 0.002 min-l) After Being Exposed t o
S t a t i c Barren F’uel S a l t f o r 7244 h r a t 650°C. (a) Fracture. lOOX.
(b) Edge near fracture. lOOX. (c) Typical unstressed microstructure.
500X. Etchant: gylceria regia. Reduced 23%.
c
bd
71
c
Fig. 56. Photomicrographs of a Modified Hastelloy N (Heat 67-551)
Sample Tested a t 25°C After Being Exposed t o t h e MSRE Core for 7244 hr
a t 650°C and Irradiated t o a Fluence of 5.1 X lo2' neutrons/cm2.
(a) Fracture, etched. lOOX. (b) Edge, as polished. lOOX. ( c ) Edge,
etched. lOOX. Etchant: glyceria regia. Reduced 21%.
72
2 mils.
A t a t e s t temperature of 650°C t h e fracture i s mixed intergran-
ular and transgranular (Fig. 57).
The microstructure near t h e edge i s
not modified, but edge cracks extend t o a depth of about 8 mils. A cont r o l sample t h a t was t e s t e d a t 25°C is shown i n Fig. 58. The f r a c t u r e
b,
.
-
i s l a r g e l y transgranular, and t h e r e are no edge cracks nor s t r u c t u r a l
modifications. The high magnification view shows the f i n e carbide prec i p i t a t e s t h a t form during t h e long thermal anneal. A t 650°C t h e fracture i s intergranular with numerous intergranular cracks scattered
throughout t h e sample (Fig. 59). A few cracks are near t h e surface, but
these a r e t o be expected i n l i g h t of t h e 31.5% s t r a i n (Table U, p. 40)
t h a t occurred i n t h i s sample before failure.
Heat 67-504 was exposed t o t h e MSRE c e l l environment of N2 + 2 t o
5% 02 f o r 17,033 h r a t 650°C and had an oxide film of 1 t o 2 mils:
when
t e s t e d a t 25°C (Fig. 60) t h e fracture was mixed transgranular and i n t e r granular, although m o s t of the deformation occurred within the grains.
There was no edge cracking.
was primarily intergranular.
15 mils.
When tested a t 650°C (Fig. 61) t h e f r a c t u r e
There were edge cracks t o a depth of
Since there were no controls t o compare with these samples, it
i s d i f f i c u l t t o say whether t h e frequency and depth of cracking a r e
g r e a t e r than would be expected.
DISCUSSION OF RESULTS
The mechanical property changes of t h e standard Hastelloy N i n both
t h e irradiated and unirradiated conditions have followed very regular
trends throughout i t s exposure i n t h e MSRE and i n t h e control f a c i l i t y .
Heats 5065 and 5085 have been used throughout t h e surveillance program.
Exposure t o t h e static barren fuel salt up t o 15,289 h r brought about a
gradual reduction of t h e tensile fracture s t r a i n s i n both heats; f'urther
exposure up t o 22,533 h r caused a s l i g h t improvement over t h e s t r a i n s
observed after 15,289 h r ( F i g s . 15 and 17, pp. 24 and 25).
These changes
were small campared with those observed a f t e r i r r a d i a t i o n , and we attribu t e them t o t h e p r e c i p i t a t i o n of grain-boundary carbides. These changes
i n t e n s i l e properties occurred without detectable change i n t h e creep
strength a t 650°C ( F i g s . 19 and 25, pp. 30 and 3 4 ) .
td
73
.
Fig. 57. Photomicrographs of a Modified Hastelloy N (Heat 67-551)
Sample Tested a t 650°C ( S t r a i n Rate, 0.002 min-') After Being Exposed t o
t h e MSRE Core for.7244 h r a t 650°C and Irradiated t o a Fluence of
5.1 X lo2' neutrons/cm2. (a) Fracture, etched. 1OOX. (b) Edge, as
polished. lOOX. (c) Edge, etched. lOOX. Etchant: glyceria regia.
Reduced 20%.
74
Fig. 58. Photomicrographs of a Modified Hastelloy N (Heat 67-551)
Sample Tested a t 25°C After Being Exposed t o S t a t i c Barren Fuel S a l t f o r
7244 h r a t 650°C. (a) Fracture. lOOX. (b) w e . lOOX. ( c ) Typical
unstressed microstructure. 500X. Etchant: glyceria regia. Reduced
20.5%.
75
1
' U
hs of a Modified Hastelloy N (Heat 67-551)
Fig. 59. Photomicrog
Sample Tested a t 650°C ( S t r a i n Rate, 0.002 min'l) After Being Exposed t o
S t a t i c Barren Fuel S a l t f o r 7244 hi. a t 650°C. (a) Fracture. 1OOX.
(b) Edge near f'racture. lOOX. ( c ) Typical unstressed microstructure.
500X. Etchant: glyceria regia. Reduced 20.5$.
m
76
.
4
4
Fig. 61. Photomicrographs of a Modified Hastelloy N (Heat 67-504) Sample Tested a t 650°C ( S t r a i n
Rate, 0.002 rnin-l) Following Exposure t o the MSRE Cell Environment f o r 17,033 hr and Irradiated t o a
Fluence of 2.5 x lo1' neutrons/cm2. (a) Fracture, etched. lOOX. (b) Edge, a s polished. lOOX.
( c ) Edge, etched. lOOX. (d) Edge, as polished. 500X. Etchant: glyceria regia. Reduced 25%.
78
I r r a d i a t i o n of both heats caused a general decrease of t h e f r a c t u r e
s t r a i n w i t h increasing thermal fluence a t test temperatures of 25 t o
b)
6
850°C. The f r a c t u r e s t r a i n s a t low temperatures are lower f o r heat 5065.
We attribute t h e reduction i n f r a c t u r e s t r a i n a t low temperatures t o prec i p i t a t i o n of carbides and demonstrated previously t h a t it can be
We a t t r i b u t e the embrit-
recovered by annealing t o coarsen t h e carbide.12
tlement a t high temperatures t o t h e helium formed by thermal transmutat i o n of l 0 B t o form ‘He and 7 L i .
by annealing.
This embrittlement cannot be recovered
The presence of helium i n the material influences t h e
creep properties a t 650°C by reducing t h e rupture l i f e and t h e s t r a i n a t
fracture; the creep rate i s unaffected (Figs. 19 through 26, pp. 30
through 3 4 ) . The e f f e c t s on t h e rupture l i f e are greater a t higher s t r e s s
levels and decrease t o insignificant below about 10,000 p s i .
mum operating stress13 i n t h e MSRE i s 6000 p s i . )
(The maxi-
The changes i n t h e
f i a c t u r e s t r a i n are of utmost importance and are s l i g h t l y dependent upon
the thermal fluence (Figs. 2 1 and 26).
Figure 23 shows more c l e a r l y the
s e n s i t i v i t y of heat 5085 t o the presence of helium and how t h i s sensitivi t y a t 650°C depends upon the s t r a i n rate.
1.3
X
A thermal fluence of only
lo1’ neutrons/cm2 produced about 1ppm (atomic) He and reduced t h e
f r a c t u r e s t r a i n from 30 t o 2s when the material was crept at a r a t e of
O.ls/hr.
The s t r a i n rate results i n t h e lowest f r a c t u r e s t r a i n s with
slower rates r e s u l t i n g i n s l i g h t l y higher values.
Figure 28, p. 36,
shows that t h i s same general behavior holds f o r heat 5065 although t h e
amounts of helium produced a r e a c t u a l l y lower.
received a thermal fluence of about 4
X
The MSRE vessel has
10” neutrons/cm2, and Figs. 23
and 28 indicate t h a t the fracture s t r a i n s w i l l not drop much more with
continued operation.
fluence of about 2
X
The control rod thimbles have received a thermal
1021 neutrons/cm2.
They should be extremely b r i t t l e ,
but are normally subjected t o only a s l i g h t compressive s t r e s s .
The greater s t r a i n rate s e n s i t i v i t y of the i r r a d i a t e d Hastelloy N
at 650°C i s i l l u s t r a t e d i n Figs. 22 and 27 f o r heats 5085 and 5065,
12H. E. McCoy, An Evaluation of t h e Molten-Salt Reactor Experiment
Hastelloy N Surveillance Specimen - Third Group, ORNL-TM-2647 (1970).
I3R. B. Briggs,
ORNL, privdte cammunication.
.
W
79
19,
respectively.
U
material varies from 32 t o 219, a decrease of 349, and t h e i r r a d i a t e d
For heat 5085 t h e f r a c t u r e s t r a i n of t h e unirradiated
material varies from 9.3 t o 0.5$, a decrease of 959.
i
For heat 5065 t h e
corresponding reductions a r e 359 f o r the unirradiated material and 97%
f o r t h e i r r a d i a t e d material.
The question of primary concern i s whether t h e Hastelloy N has
corroded more r a p i d l y during the previous surveillance period than
previously.
We have at l e a s t t h r e e parameters or phenomena that can be
observed o r measured t h a t l i k e l y a r e dependent upon t h e corrosion r a t e .
The f i r s t measure i s t h e chemical change of t h e s a l t during reactor
operation.
Since the chromium observed i n t h e s a l t can hardly come
fkom anywhere except t h e Hastelloy N, we cannot question t h i s parameter
as a measure of t h e corrosion r a t e .
t h e microstructure.
Second, we can observe changes i n
Figures 2, 3, and 4 , pp. 9, 10, and 12, show typi-
c a l changes, but these observations a r e d i f f i c u l t t o i n t e r p r e t without
additional information.
i
Third, t h e mechanical properties themselves may
be a l t e r e d o r t h e crack patterns t h a t develop may be useful indications
of the e f f e c t s of corrosion on t h e mechanical properties.
I
The property
changes themselves a r e not very useful i n t h i s case because we have only
matched sets of control samples and surveillance samples.
The control
samples a r e exposed t o a static barren f u e l s a l t t h a t does not match the
corrosiveness of t h e MSRE f u e l c i r c u i t .
The surveillance samples a r e
exposed t o a neutron fluence and t o a more corrosive s a l t c i r c u i t .
The
i r r a d i a t i o n e f f e c t s predominate, so the effects of t h e added corrosion
a r e not large enough t o see
observing t h e mechanical property changes.
Thus t h e r e a r e several measures of corrosion i n t h e MSRE, but they a r e
a l l subject t o i n t e r p r e t a t i o n .
Because of t h e importance of t h i s sub-
j e c t , l e t us review t h e p e r t i n e n t f a c t s and observations.
1. Before the last period of operation, t h e salt was removed from
t h e MSRF, and processed
remove t h e uranium and replace it with 233U.
This processing l e f t t h
a l t fairly oxidizing, and it was necessary t o
i a l b y adding beryllium metal. The chromium
a d j u s t t h e oxidation
PO
content of the s a l t was only 40 ppm when placed i n t h e reactor and rose
i
u
over a few weeks t o a l e v e l of about 100 ppm.
This apparent'corrosion
codld be accounted f o r by uniformly removing t h e chromium from t h e
80
Hastelloy N t o a depth of 0.3 m i l .
The t o t a l increase of chromium i n
t h e salt during t h e e n t i r e MSFE operation would require uniform chromium
extraction t o a depth of 0.4 m i l .
t
Hawever, t h e diffusion rate along t h e
g r a i n boundaries can be about lo6 t h e r a t e through t h e bulk grains,14
and it i s more l i k e l y that chromium be removed t o g r e a t e r depths along
t h e g r a i n boundaries.
2.
Electron microprobe examination of the sample i n Fig. 5 , p. 13,
revealed a d e f i n i t e chromium gradient near t h e surface and a surface
chromium concentration of nearly zero.
4x
A diffusion coefficient of
cm2/sec f o r chromium was computed based on t h e shape of t h e
p r o f i l e which i s i n excellent agreement with t h e value of 2 x
cm2/sec
measured by Grimes et a1.I5 (Fig. 62).
3.
Complete loss of chromium from Hastelloy N does not r e s u l t i n
t h e type of microstructural modification t h a t was observed, not does it
cause embrittlement.16 We made laboratory melts containing from 0 t o
9% C r w i t h t h e standard Hastelloy N base composition.
The mount of car-
bide p r e c i p i t a t e increases and t h e g r a i n s i z e decreases as t h e chromium cont e n t increases.
The a l l o y is weaker i n creep a t 650°C without chromium,
but t h e f r a c t u r e s t r a i n s a r e not dependent upon t h e chromium concentration.
4. Similar microprobe scans have been made on control sanrples where
t h e modified s t r u c t u r e i s a l s o present.
The resolution i s much b e t t e r
on these samples where the background r a d i a t i o n i s not present.
The
gradients measured i n a sample that had been i n the c o n t r o l f a c i l i t y f o r
15,289 hr a t 650°C are shown i n F i g . 63.
t h e i r o n i s enriched near the surface.
The chromium i s depleted and
These chemical modifications
extend only a short distance i n t o the material.
These changes would not
be expected t o produce t h e microstructural changes.
14W. R. Upthegrove and M. J. Sinnot, "Grain-Boundary Self-Diffusion
of Nickel," Trans. Am. SOC. Metals 50,
- 1031 (1958).
15W. R. Grimes, G. M. Watson, J. H. DeVan, and R. B. Evans, "RadioTracer Techniques i n t h e Study of Corrosion by Molten Fluorides,"
pp. 559-574 i n Conference on t h e Use of Radioisotopes i n t h e P G s i c a l
Sciences and Industry, September 6 1 7 , 1960, Proceedings, Vol. 111,
International Atomic Energy Agency, Vienna, 1962.
16H. E. McCoy, Influence of Various Alloying Additions on t h e
Strength of Nickel-Base Alloys (report i n preparation).
f
LJ
81
ORNL- OWG 70-4933
hd
d
SMEASURED
(D= 4x
cm2/sec)-
.
CALCULATED( D = 2 x
I
0
2
cm2/sec)
3
4
DISTANCE FROM SURFACE (mils)
Fig. 62. Chromium Gradient i n Hastelloy N Sample Exposed t o t h e
MSRE Core f o r 22,533 hr.
ORNL-OWG 70-7659
12
z
gci
a
I-
W
z
Y 4
8
2
0
0
5
Fig. 6 3 . Concentrat
t o S t a t i c Barren Fuel Sa
650°C.
5.
40
15
20
DISTANCE FROM SURFACE ( p )
25
30
radients i n Hastelloy N (Heat 5085) Exposed
t h e MSRE Control Vessel f o r 15,289 h r a t
The Hastelloy N straps t h a t held t h e surveillance assembly
together f o r 22,533 h r had
ntergranular surface cracks t o a depth of
about 3 mils (ref. 17) (see Fig. 2, p. 9 ) .
Straps t h a t had been i n the
reactor f o r 7244. hr had cracks t o a depth of 1.5 mils.
The s t r a p s are
0.020 i n . t h i c k and should not be stressed during reactor operation, but
.
c-d
17W. H. Cook, MSR Program Semiann. F’rogr. Rept., Aug. 31, 1969,
ORNL4449, pp. 165-168.
82
they were handled considerably during disassembly.
Thus one cannot con-
clude whether t h e cracks were formed during service o r i n handling a f t e r -
wards.
straps
.
The microstructure was not modified near t h e surface of these
6.
-
A surface microstructural modification has been observed spas-
modically during t h i s e n t i r e program.
The photomicrographs of t h e
various l o t s of heat 5085 that were examined a r e presented i n Fig. 64.
The modified microstructure has been present t o some degree i n a l l samp l e s including t h e i r r a d i a t e d and t h e control samples.
It i s d i f f i c u l t
t o say t h a t t h e modification has g r a m any worse.
7.
We have been able t o produce a microstructural modification
q u i t e similar t o t h a t shown i n Fig. 65 by s i n t e r l e s s grinding Hastelloy N
and then annealing it f o r long periods of time i n argon a t 650°C
(Fig. 65).
Thus, t h e layer may w e l l be associated with cold working and
the manner i n which t h i s alters carbide deposition near t h e surface.
8.
I
There i s a d e f i n i t e tendency f o r t h e samples removed from t h e
MSRE t o show progressively more edge cracking during p o s t i r r a d i a t i o n
t e s t i n g as t h e time of exposure increases.
As shown i n Fig. 66 f o r sam-
p l e s s t r e s s e d a t 25"C, t h e frequency of edge cracks'increases w i t h exposure, but t h e maximum depth is constant a t about 4 mils.
The control
samples do not show much edge cracking.
9.
Microprobe scans on i r r a d i a t e d samples f a i l e d t o r e v e a l any
f i s s i o n products i n t h e sample.
However, t h e i n a b i l i t y t o analyze s o l e l y
t h e material i n a g r a i n boundary makes it impossible t o conclude t h a t
no f i s s i o n products are entering t h e metal.
We plan t o dissolve progres-
s i v e layers from t h e sample surfaces f o r chemical analysis t o answer t h i s
quest ion.
The observations on t h e alloys of modified Hastelloy N a r e q u i t e
important.
Heats 7320 (0.5% T i ) and 67-551 (1.1%
T i ) were exposed t o
t h e MSRE core f o r 724.4 h r a t 650°C.
The properties of heat 7320 were
not as good as we have observed f o r other 0.5% Ti-modified a l l o y s irradiated a t 650°C.
However, t h e properties a f t e r i r r a d i a t i o n i n t h e MSRE
are equivalent t o those measured a f t e r i r r a d i a t i o n i n t h e ORR i n a helium
environment.
The properties of heat 67-551 are outstanding.
Both of
f
83
W
NRADIATLD
Y-Ppy?
c
4.m k
8
22.533 k
Fig. 64. Photomicrographs of Unstressed Hastelloy N (Heat 5085)
Control and Irradiated Samples.
Li
Fig. 65. Photomicrograph of Hastelloy N (Heat 5085) R o d That Was
Annealed 1 h r a t 1177OC, Sinterless Ground 5.2 mils, and Annealed f o r
4370 hr a t 650°C i n Argon. 500X. Etchant: glyceria regia.
84
,
Fig. 66. Samples of Hastelloy N (Heat 5085) Stressed a t 25°C. The
control samples were exposed t o s t a t i c barren f u e l s a l t f o r the indicated
time and t h e irradiated samples were exposed t o t h e core of t h e MSFE.
i
these heats seem t o demonstrate some small property changes due t o therm a l aging.
These changes are not large, but m u s t be followed t o longer
times t o make sure that they do not become large.
These two modified heats did not show a s t r u c t u r a l modification
near t h e surface, but did exhibit edge cracking when t e s t e d .
The fre-
quency of edge cracks i s much higher than noted f o r the controls.
One of t h e most conf'using observations was the poor mechanical
properties of heat 67-504 from outside t h e core.
This heat was pre-
viously18 exposed t o t h e core f o r 9789 h r t o a thermal fluence of
5.3 X
lo2* neutrons/cm*.
The present group of samples outside t h e core
ISH. E. McCoy, An Evaluation of t h e Molten-Salt Reactor Experiment
Hastelloy N Surveillance Specimen - Third Group, ORNETM-2647 (1970).
85
was exposed t o N2 + 2 t o 5% 02 f o r 17,033 h r and received a thermal
f
fluence of 2.5 x loi9 neutrons/cm2.
The l a t t e r group had poorer mechan-
i c a l p r o p e r t i e s than t h e group exposed t o t h e higher fluence.
J
The
rupture l i f e a t a given stress l e v e l was l e s s (Fig. 41, p. 54), t h e
minimum creep r a t e higher (Fig. 42, p. 541, and t h e fracture strain lower
(Fig. 43, p. 5 5 ) .
The photomicrographs i n Figs. 60 and 61, pp. 76 and 77,
/
show same surface oxidation but no other unusual features.
Two heats
should have been used, heat 67-502 (2%W + 0.5$ T i ) and heat 67-504
(0.5% H f )
.
0.5$ Hf,
but no tungsten was present.
Postirradiation chemical analyses showed the presence of
Hence, we have concluded t h a t
only heat 67-504 was included in these tests.
Our surveillance program has given us t h e opportunity t o look a t
s e v e r a l alloys of modified canposition.
The creep properties of these
heats a r e compared with those o f standard Hastelloy N i n Fig. 67.
The
curves were drawn through only four or f i v e data points i n each case,
so s l i g h t differences i n slope a r e not s i g n i f i c a n t . The rupture l i v e s
o f t h e modified alloys a r e within a f a c t o r of 2. The rupture l i v e s of
t h e irradiated, modified heats are about equivalent t o those of unirradiated standard Hastelloy N.
The creep rates, shown i n Fig. 67(b), a r e
l w e r f o r t h e modified alloys by as much as a f a c t o r of 3.
and 67-504 have t h e lowest creep rates.
strains are shown i n Fig. 67(c).
Heats 67-502
The p o s t i r r a d i a t i o n f r a c t u r e
Quite a range of f r a c t u r e s t r a i n s
r e s u l t e d with a m i n i m of about 0.5% f o r standard Hastelloy N and a
minimum of about 7%f o r heat 67-551.
Heats 67-502 and 67-504 had frac-
t u r e strains of 6.4%; heat 7320 was not much b e t t e r than standard
.
Hastelloy N with o n l y 2.8%.
None of t h e modified alloys have shown any adverse corrosion behav-
ior i n the salt. Thus, it would appear t h a t we have several alloys t h a t
a r e s u i t a b l e f o r use i n future reactors t h a t operate a t 650°C. However,
as discussed previously, these new alloys a r e very s e n s i t i v e t o i r r a d i a t i o n temperature and the proposed 700°C operating temperature of a
breeder i s too high f o r these a l l o y s . l g
Present work offers encouragement t h a t an a l l o y t h a t is stable a t 700°C can be developed by adding
--
I9H. E. McCoy et a l . , MSR Program Semiann. Progr. Rept. Aug. 31,
1969, ORNE44.49,
p. l84.
0
(DY
rnm
N
I
0
CW
.-
s
YJ
m
I
P
0
I
m
1.
I
3
FRACTURE STRAIN ( %)
8
n
A
0
h)
A
I
0
8
I
ru
-
P
0
STRESS (4000 psi)
0
P
A
8
0
lJl
8
0
P
STRESS (4000psi)
0
UI
d
---
0
m
.-.
%
2
f
is,
87
l a r g e r amounts of titanium o r combined amounts of these elements along
t
with niobium and hafnium.
.
SUMMARY AND CONCLUSIONS
The heats of Hastelloy N used i n fabricating t h e MSRE have s h a m a
systematic deterioration of mechanical properties w i t h increasing neutron
fluence.
The material exposed f o r t h e longest period of time i n t h e
core has reached a thermal fluence of 1.5 x 1021 neutrons/cm2 and a fast
fluence (> 50 kev) of 1.1X
neutrons/cm2.
These values are q u i t e
close t o those anticipated f o r f u t u r e reactors with a 30-year design
life.
The d u c t i l i t y of t h e material was too low, but t h e microstructure
was f r e e of irradiation-induced voids and defects other than helium bub-
bles.
Several heats of t h e modified alloys have been exposed t o the MSRE
and these have b e t t e r p o s t i r r a d i a t i o n properties.
They a l s o seem t o
have good corrosion resistance.
n
The standard Hastelloy N removed from t h e core shows some evidence
of corrosion.
.
The corrosion seems generally t o be due t o t h e s e l e c t i v e
removal of chromium, as predicted by prenuclear tests.
Some observations
t h a t have not been explained adequately are (1)t h e presence of grainboundazy cracks i n t h e s t r a p s t h a t held p a r t s of t h e surveillance assembly together, (2) t h e modified microstructure near the surface, and
(3) t h e formation of intergranular cracks originating from t h e surface
when i r r a d i a t e d materials a r e strained.
One of t h e modified alloys, heat 67-504, was exposed t o the c e l l
environment.
This heat had been previously exposed t o t h e MSRE.
The
fluence was higher i n t h e core, but t h e p o s t i r r a d i a t i o n properties were
superior t o those of t h e material exposed t o t h e c e l l environment.
We
presently have no explanation for t h e observed behavior.
c
The author i s indebte
many people f o r a s s i s t i n g i n t h i s study:
W. H. Cook and A. Taboada f o r design of t h e surveillance assembly and
i n s e r t i o n of t h e specimens; W. H. C o o k and R. C. S t e f f y f o r measurements
88
of flux; J. R. Weir, Jr., R. E. Gehlbach, and C. E. Sessions f o r review
of t h e manuscript; E. J. Lawrence and J. L. Griff'ith f o r assembling t h e
surveillance and control specimens i n the f i x t u r e ; P. Haubenreich and
t h e MSRE Operation S t a f f f o r t h e extreme care w i t h which they inserted and
removed t h e surveillance specimens; E. M. King and t h e Hot C e l l Operation
S t a f f f o r developing techniques f o r cutting long rods i n t o individual
specimens, determining specimen straightness, and assistance i n running
creep and t e n s i l e t e s t s ; B. C. W i l l i a m s , B. McNabb, and H. W. Kline f o r
running tensile and creep tests on surveillance and control specimens;
J. Feltner f o r processing t h e t e s t data; H. R. Tinch and N. M. Atchley
f o r metallography of t h e control and surveillance specimens;
Frances Scarboro of The Metals and Ceramics Division Reports Office f o r
preparing t h e manuscript; and t h e Graphic Arts Department f o r preparing
the d r a m s .
d
A
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