ORNL-TM-3063.txt

 

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Contract No. W-7405-eng-26
METALS AND CERAMICS DIVISION
AN EVALUATION OF THE MOLTEN-SALT REACTOR EXPERIMENT
HASTELLOY N SURVEILLANCE SPECIMENS — FOURTH GROUP
H. E. McCoy, Jr.
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MARCH 1971
_ _LEGAL NOTICE——— i
This " report - was prepared a3 an account. of work |
sponsored by the United States Government, Neither |
1 the United States nor the United States Atomic Energy
-] Commission, nor any of their employees, nor any -of
{ | their contractors, subcontractors, or their employees, | :
.| makes any warranty, express or implied, or assumes any
- | legal liability or responsibility for the accuracy, com-
| pleteness or usefulness of any information, apparatus,
-product or ‘process disclosed, or represents that its use
wouid not infriflge priva;flely ow@ed rights. ok
| | | OAK RIDGE NATIONAL LABORATORY
* _ , , ~ Oak Ridge, Tennessee
L o - operated by
o I .~ UNION CARBIDE CORPORATION
P _ for the
; ‘ j ' | U.S. ATOMIC ENERGY COMMISSION

; n ” | | ~ DISTRIBUTION OF THIS DOCUMENT IS UN
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- CONTENTS

Abstract . . . . . . . . . .. 0 o0 e L
Introduection . . . . . . .
Experimental Details .
Surveillance Assemblies e e e e e e e e
Materials . . . . . . . . .
Test Specimens . . . . . . . . . .
Irradiation Conditions . . . . . . . . . . .
Testing Techniques .
Experimental Results . . . . . . . . « « « &
Visual and Metallographic Examination
Mechanical Property Data — Standard Hastelloy N
Mechanical Property Data — Modified Hastelloy N
Metallographic Examination of Test Samples .
Discussion of Results . . . . . . . . . . « . . .
Sumary and Conclusions 7

Acknowledgments . . . . . . .

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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 the MSRE after 22,533 hr at 650°C and exposed to a
thermal fluence of 1.5 X 1021 neutrons/cm? and a fast fluence
> 50 kev of 1.1 X 10%! neutrons/cm®. The mechanical proper-
ties have systematically deteriorated with increasing fluence.
However, the change in ropertles is due to the helium
produced by the 1 B(n,x)”1Li transmutation and can be reduced
- by changes in chemical composition. Some of these modified
heats have been exposed to the core of the MSRE and show
improved resistance to irradiation.

The corrosion of the Hastelloy N has been largely due to
the selective removal of chromium. The rates of removal are
much as predicted from the measured diffusion rate of chromium.
Other superficial structure modifications have been observed,
but ‘they likely result from carbide precipitation along slip
bands that were formed during machining.

 

INTRODUCTION

The Molten~Salt Reactor Experiment (MSRE) is a single region reac-
tor that is fueled by a molten fluoride salt (65 LiF—29.1 BeFo—5 ZrF,—
0.9 UF;, mole %), moderated by unclad graphite, and contained by

- Hastelloy N (Ni~16 Mo-7 Cr—4 Fe-0.05 C, wt %). The details of the reac-

tor design and constructidniban_be'fbund elsewhere.® We knew that the

neutron environment would produce some changes in the two structural mate-

- rials - graphlte and Hastelloy N. Although we were very confldent of
~ the compatiblllty of these materlals with the fluoride salt,'we needed

to keep abreast'ofrthe_p9581ble;development of-corros1on.problems within

 

| 1R c. Roblnson, MSRE De31gn and Operations Report ' Pt. 1, Descrlp-

tion of Reactor Design, ORNL-TM-728 (1965).
 

 

the reactor itself. For these reasons, we developed a surveillance

(

program that would allow us to follow the property changes of graphite

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and Hastelloy N specimens as the reactor operated.

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The reactor went critical on June 1, 1965. After many small prob-
lems were solved, normal operation began in May 1966. We removed four
groups of surveillance samples. The results of tests on the first three
groups were reported.2_4 This report deals primarily with the results
of tests on samples removed with the fourth group. The fourth group
included two heats of standard Hastelloy N used in fabricating the MSRE
and three heats with modified chemistry that had better mechanical proper-
ties after irradiation and appear attractive for use in future molten-
salt reactors. The respective history of each lot was (1) standard
Hastelloy N, annealed 2 hr at 900°C and exposed to the MSRE core for
22,533 hr at 650°C to a thermal fluence of 1.5 X 1022 neutrons/cm?,

(2) two heats of modified Hastelloy N, annealed 1 hr at 1177°C and exposed
to the MSRE core for 7244 hr at 650°C to & thermal fluence of 5.1 x 1022
neutrons/cm?, and (3) a single heat of modified Hastelloy N, annealed 3
for 1 hr at 1177°C and exposed to the MSRE cell enviromment of N, + 2 to e
5% 0, for 17,033 hr at 650°C to a thermal fluence of 2.5 X 101°
neutrons/cm?. The results of tests on these materials will be presented
in detail, and some comparisons will be made with the data from the

groups removed previously.
EXPERIMENTAL DETATLS

Surveillance Assemblies

The core surveillance assembly’ was designed by W. H. Cook and

others, and the details have been reported previously. The specimens

 

2H. E. McCoy, An Evaluation of the Molten-Salt Reactor Experiment
Hastelloy N Surveillance Specimen — First Group, ORNL-TM-1997 (1967).

3. E. McCoy, An Evaluation of the Molten-Salt Reactor Experiment
Hastelloy N Surveillance Specimen — Second Group, ORNL-TM-2359 (1969). .

“H. E. McCoy, An Evaluation of the Molten-Salt Reactor Experiment
Hastelloy N Surveillance Specimen — Third Group, ORNL-TM-2647 (1970).

°W. H. Cook, MSR Program Semiann. Progr. Rept. Aug. 31, 1965, o/
ORNL-3872, p. 87. | - S .

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are arranged in three stringers. Each stringer is about 62 in. long
and consists of two Hastelloy N rods and a graphite section made up of
various pieces that are joined by pinning and tongue-and-groove joints.
The Hastelloy N rod has periodic-reduced sections 1 1/8 in. long by
1/8 in. in diameter and can be cut into small tensile specimens after it
is removed from the reactor. Three stringers are joined together so
that they can be separated in a hot cell and reassembled with one or
more new stringers for reinsertion into the reactor. The assembled
stringers fit into a perforated Hastelloy N basket that is inserted into
an axial position about 3.6 in. from the core center line.

A control facility is associated with the surveillance program.
It utilizes a "fuel salt" containing depleted uranium in a static pot
that is heated electrically. The temperature is controlled by the MSRE
computer so that the temperature matches that of the reactor. Thus,
these specimens are ekposed to conditions similar to those in the reac-
tor except for the static salt and the absence of a neutron flux.

There is another surveillance facility for-Hastelloy N located
outside the core in a vertical position about 4.5 in. from the vessel.

These specimens are exposed to the cell enviromment of N> + 2 to 5% O,.

Materials

The compositions of the two heats of standard Hastelloy N are given
in Table 1. These heats were air melted by the Stellite Division of
Union Carbide Corporation. Heat 5085 was used for making the cylindrical

portion of the reactor VéSSél'and heat 5065 was used for forming the top

~and bottom heads. These materialS'wefe given a mill anneal of 1 hr at

1177°C and a final anneal of 2 hr at 900°C atxORNL after fabrication.

The chemical compositions of the three modified alloys are given

 in Table 1. The modifications in composition were made principally to

imprové the alloy's,resiStanéejtdnradiation demage and to bring about

-general improvements in the fabricability, weldability, and ductility.®

 

®H. E. MeCoy and J. R. Weir, Materials Development for Molten-Salt
Breeder Reactors, ORNL-TM-1854 (1967).

 

 
 

4

Table 1. Chemical Analysis of Surveillance Heats

 

Element

Content, wt %

 

 

(»

Heat 5065 Heat 5085 Heat 7320 Heat 67-551 Heat 67-504
Cr 7.3 7.3 7.2 7.0 6.9
Fe 3.9 3.5 < 0.05 0.02 0.05
Mo 16.5 16.7 1 12.0 12.2 12.4
C 0.065 0.052 0.059 0.028 0.07
si 0.60 0.58 0.03 0.02 0.010
Co 0.08 0.15 0.01 0.03 0.02
W 0.04 0.07 < 0.05 0.001 0.03
Mn 0.55 0.67 .  0.17 - 0.12 0.12
v 0.22 . 0.20 <0.02 < 0.001 0.01
P 0.004 0.0043 0.002 0.0006 0.002
S 0.007 0.004 0.003 < 0.002 0.003
Al 0.01 0.02 - 0.15 < 0.05 0.03
Ti 0.01 < 0.01 0.65 1.1 < 0.02
Cu 0.01 0.01 0.02 0.01 0.03
0 0.0016 0.0093 0.001 0.0004 < 0.0001
N 0.011 0.013 0.0002 - 0.0003 0.0003
Zr <O0.1 < 0.002 < 0.05 < 0.01 0.01
Hf < 0.1 0.50
B 0.0038 0.00002 0.0002 0.00003

0.0024

 

Alloys 67-551 and 67-504 were small, 100-1b heats made by the Stellite
Division of Union Carbide Corporation by vacuum melting. They were
finished to 1/2 in. plate by working at 870°C. We cut strips 1/2 in. by
1/2 in. from the plates and swaged them to 1/4-in.-dism rod. Two sec-
tions of rod were welded together to make 62-in.-long rods for fabri-
cating the samples. The rods were annealed for 1 hr at 1177°C in argon
and then the reduced sections were machined. Heat 7320 was a 5000-1b
melt made by the Materials Systems Division of Union Carbide Corporation.
' Part of the heat was fabricated by the vendor to 5/16-in.-diam rod and .
was sinterless ground to obtain the needed 1/4 in. stock.k The material

was annealed 1 hr at 1177°C and then the reduced sections were machined.
d.). of »

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Test Specimens

The surveillance rods inside the core are 62 in. long and those
outside the vessel are 84 in. long. They both are 1/4 in. in diameter
with reduced sections 1/8 in. in diameter by 1 1/8 in. long. After
removal from the reactor, fihé rods are sawed into small mechanical prop-
erty specimens having a gage section 1/8 in. in diemeter by 1 1/8 in.
long.

The first rods were machined as segments and then welded together,
but we described previously an improved technique in which we use a
milling cutter to machineAthe_reduced sections in the rod.? This tech-
nique is Quicker, cheaper, and requires less handling of the relatively
fragile rods than the previous method of making the rods into segments.
The standard Hastelloy N rods were machined and welded together and the

modified alloys were prepared by milling.

IRRADTATION CONDITIONS

The jrradiation conditions for the various groups of surveillance
specimens that have been removed are sumarized in Table 2. The reac-
tor operated from June 1965 until March 1968 with a single change of
fuel salt in which there was a 33% enrichment of 235U. After this
period of operation the uranium in the fuel was'stripped'by Tluorination
and replaced with 233y (ref,m7). This charge of salt was used until the

present group of samples was removed. Referring again to Table 2, the

standard Hastelloy N in the core was exposed to both salts and the modi-

fied Hastelloy N in the core was exposed only to the 1atter salt. The

same salt has been used_ln_thg control facility throughout operation.

The specimens outside the core (designated "vessel" specimens)

 were'exposed to the cell envifonment of N» + 2 to 5% 0s.

 

7P. N. Haubenrelch and J R. Engle, "Experlence w1th the Molten-

Salt Reactor Experlment " Nucl Appl. Technol. 8(2) 118 (1970)

 

 
 

 

 

 

 

Table 2. Summary of Exposure Conditions of Surveillance Samplesa’
‘Growp 1 4
Care Group 2, Hastelloy N Group 3, Hastelloy N Group 4, Hastelloy N
Standard Core .Vessel Core Core Vessel Core Core Vessel
Haptelloy § Modified Standard Standard Modified Standard Standard Modified Modified
Date inserted 9/8/65 9/13/66 8/24 /65 9/13/66 6/5/67 8/24 /65 9/13/66 4/10/68 5/7/68
Date removed’ 7/28/66 5/9/67 6/5/67 4[3/68 4/3/68 5/7/68 6/69 6/69 6/69
Megawatt-hour on MSRE 0.0066 8682 0 8682 36,247 0 8682 72,441 36,247
at time of insertion
Megawatt-hour on MSRE 8682 36,247 36,247 72,441 72,441 72,441 92,805 92,805 92,805
at time of removal -
Temperature, °C 650 + 10 650 + 10 650 + 10 650 + 10 650 + 10 650 + 10 650 + 10 650 + 10 650 + 10
Time at temperature, hr 4800 5500 11,000 15,289 9789 20,789 22,533 7244 17,033
Peak fluence, neutrons/cm?
Thermal (< 0.876 ev) 1.3 x 1020 4,1 x 1020 1.3 x 101° 9.4 x 1020 5,3 x 1020 2.6 x 1019 1.5 x 10%} 5,1 x 102° 2,5 x 1019
Fpithermal (> 0.876 ev) 3.8 x 1020 1,2 x 102} 2,5 x 10!® 2.8 x 10! 1,6 x 10! 5.0 x 10! 3.7 x 10! 9,1 x 10?° 3,9 x 101?
(> 50 kev) 1.2 x 1020 3,7 x 10°° 2,1 x 1017 8.5 x 1020 4.8 x 1020 4,2 x 10'? 1,1 x 10*! 1.1 x 102°© 3.3 x 101°
(>1.22 Mev) 3.1 x10'% 1.0x 10?0 5,5 x 1018 2.3 x10°° 1,3x10%0 1,1 x10'° 3.,1x10%° 0.8x10%° 8,6 x 1018
(> 2.02 Mev) 1.6 x 101° 0.5 x 1029 3,0 x 101% 1,1 x 1020 0.7 x10%°® 6,0 x 10'% 1.5 x 102° 0.4 x 102° 3,5 x 108
Heat 5081 21545 5065 5065 67-502 5065 5065 7320 67-504
Designations 5085 21554 5085 5085 67-504 5085 5085 67-551

 

®Information compiled by R. C. Steffy, Resctor Division, ORNL, July 1969.

¢ o

Revised for fullepower operation at & Mw.

 
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Testing Techniques

The laboratory creep-rupture tests of unirradiated control speci-
mens were run in conventional creep machines of the dead-load and lever-
arm types. The strain was measured by a dial indicator that showed the
total movement of the specimen and part of the load train. The zero
strain measurement was taken immediately after the load was applied.

The temperature accuracy was +0.75%, the guaranteed accuracy of the
Chromel-P—-Alumel thermocouples used.

The postirradiation creep-rupture tests were run in lever-arm
machines that were located in hot cells. The strain was measured by an
eXtensometer‘with rods attached to the upper and lower specimen grips.
The relative movement of these two rods was measured by a linear differ-
ential transformer, and the transformer signal was recorded. The
accuracy of the strain measurement is difficult to determine. The exten-
someter (mechanical and electrical portions) produced measurements that
could be read to about *0.02% strainj however, other factors (tempera-
ture changes in the cell, mechanical vibrations, etc.) probably combine
to give an overall accuracy of +0.1% strain. This is considerably better
than the specimen-to-specimen reproducibility that one would expect for
relatively brittle materials. The temperature measuring and control
system was the same as that used in the laboratory with only one excep-
tion. In the laboratory, the ccntrol system was stabilized et the
desired temperature by use of a recorder with an expanded scale. In the
tests in the hot cells, the control point was established by settlng the

controller Wlthout the ald of the expanded-scale recorder. This error

_'and the thermocouple accuracy combine to give a temperature uncertainty

of about *1%.-
The ten31le tests were run on Instron Universal Testlng Machines.

The strain measurements were’ taken from the crosshead travel and gener-

. _ally are accurate to *2% straln.

The test environment was air in all cases. Metallegrepnic examina-

tion showed that the depth of oxldat;on was small, and we feel that the

environment did not sppreciasbly influence the test results.

 
 

EXPERIMENTAL RESULTS

Visual and Metallographic Examination J _ e

W. H. Cook was in charge of the disassembly of the core surveil-
lance fixture. As shown in Fig. 1, the assembly was in excellent mechan-
-ical condition when removed. The Hastelloy N samples were more
discolored than noted previously; however, surface marking such as
numbers were readily visible. The detailed appearance of the stringer
has been described previously by Cook.® The Hastelloy N surveillance
rods located outside the core were oxidized, but the oxide was tenacious.

Metallographic examination of the Hastelloy N straps that held the
graphite and metal together revealed intergranular cracks. A typical

 

8W. H. Cook, MSR Program Semiann. Progr. Rept. Aug. 31, 1965,
ORNL-3872, pp. 87-92.

- R-4B618

     

 

Fig. 1. Overall View of MSRE Surveillance Assembly Removed After
Run 18. Parts of this assembly had been exposed to the salt for , .
22,533 hr at 650°C. The center portion is graphite and the long rods gfi;
are Hastelloy N. -
 

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“s_Haste1loy N'survelllance rods. Typlcal.photomlcrographs of heat 5065
. after exposure to the core for 22,533 hr are shown in Fig. 3. The

crack is shown in Fig. 2 and extends to a depth of about 3 mils. A
similar strap on the modified samples which had been in the reactor for
7244 hr had cracks to a defith of about 1.5 mils. These straps are about
0.020 in. thick and they likely encountered some deformation while being
removed. However, the cracks were quite uniformly spaced on both sur-
faces of the straps, and their genefal gppearance attests to a general
corrosion that rendered the grain boundaries extremely brittle. Examina-
‘tion of unirradiated control straps failed to reveal a similar type of

cracking.

R-48571

 

 

 

 

 

 

Fig. 2. Typlcal Mlcrostructure of a Hastelloy N (Heat 5055) After
Exposure to the MSRE Core for 22,533 hr at 650°C. This material was
used for straps for the surveillsnée assembly. As polished. 500X.

These observatlons led to the examlnatlon of tabs from the surveil-

lance stringers. Small sections were cut from the centers of the

unetched view in Flg. 3(a) shows the surface layer that led to the dis-
colored appearance and a s1ngle graln boundary that is visible. Much
of the surface layer looks metallic, but this is difficult to judge on

 

 
 

 

 

 

 

Fig. 3.

 

10

R-48860

Typical Photomicrographs of Hastelloy N (Heat 5065) Exposed

to the MSRE Core for 22,533 hr- a.t 650°c, (a) As polished. (b) Etchant:

aqua regia.

500x.
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11

such a thin film. The etched view in Fig. 3(b) reveals some carbide
prec1p1tat10n near the surface and grain boundaries that are generally
lined with carbides. Heat 5085 was exposed an identical time and
typical photomicrographs are shown in Fig. 4. The unetched view in
Fig. 4(a) shows a modified grain boundary structure to a depthiof about
2 mils. Etching [Fig. 4(b)] reveals a grain boundary network of car-
bides. The grain boundaries near the surface seem to etch differently
from the rest of the sample, but little more can be said.

We interpreted these observétions as being indicative of some cor-

rosion and performed one further crude experiment to reveal the depth of

this attack. One tensile sample had been cut too short for testing, and

we bent the remaining portion in a vise. The sample was then sectioned
and examined metallographically; the resulting photomicrographs are shown
in Fig. 5. The tension side cracked to a depth of about 4 mils whereas
the compression side did not crack. Both-sides etched abnormally to a
depth of about 4 mils. | |

' Samples of heats 5065 and 5085 that were exposed to the static bar-
ren salt in the control fac111ty'were examined. Figure 6 shows'typical
photomlcrographs of heat 5065 after exposure for 22,533 hr. There is
some surface roughening, but no structure modification near the surface

such as that shown in Fig. 3 for the sample from the reactor. Likewise,

" heat 5085 (Fig. 7) showed some surface effects that were minor compared

with its irradiated counterpart in Fig. 4.
Thus, there is little doubt that the samples in the core experienced
some modifications, apparently to a depth of 3 to 4 mils. This alters

up to about 12% of the sample cross section and can be expected to influ-
ence the mechanical properties;“ We shall dwell fUrther on thls very

1mportant subaect later in this report._ _
A sample of heat 5085 fram the core was examined by transm1sszon |

electron mlcroscopy ThlS sample had received suff1c1ent thermal fluence

o transmute about 97% of the 108 4o helium. The helium bubbles are

obvious in Fig. 8. Another point of concern‘was the formation of voids
in the material due to fast neutrons. No defects other than helium bub-

bles and dislocations were present.
 

 

 

 

 

 

 

 

Fig. 4.
to the MSRE Core for 22,533 hr at 650° (a) As polished.- (b) Etchant°
glycerla regia. 500x

 

 

 

 

 

 

 

 

 

 

 

 

Typical Photomlcrographs of Hastelloy N (Heat 5085) Exposed
( a) b «) " "’) oy (

8 R-50280

LIRS a4

     

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

       

 

 

Fig. 5. Typical Photomlicrographs of a Hastelloy N (Heat 5085) Sample Exposed to the MSRE Core for
22,533 hr at 650°C. The sample was bent in a vise. (a) As polished, tension side. (b) As polished,

tension side. (¢) Etched, tension side. (d) Etched, compression side. Etchant: aqua regia. 500X.
Reduced 27%.
 

 

 

 

 

 

 

 

Fig. 6.
Static Barren Fuel Salt for 22,533 hr at 650°C. (a) As polished.

(b) Etched.

14

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Typical Photomicrographs of Heat 5065 After Exposure to

Etchant: glycerie regia. 500X.

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‘Fig. 7. Typical Photomicrographs of Heat 5085 After Exposure to
~Static Barren Fuel Salt for 22,533 hr at 650°C. (a) As polished.
(b) Etched. Etchant: glyceria regia. 500X.

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17

MEchanical-Pfoperty Data — Standard Hastelloy N

Two heats of standard Hastelloy N were exposed to the MSRE core
environment for 22,533 hr at 650°C and received a thermal neutron fluence
of 1.5 x 10?1 neutrons/em®. Similar control samples were exposed to
static barren fuel salt for a corresponding length of time. The results
of tensile tests on heat 5085 are summarized in Tables 3 and 4 for |
unirradiated and irradiated samples, respectively. The fracture strains
are shdwn as a function of test temperature in Fig. 9. With increasing
temperature the unirradiated samples exhibit first an increase in frac-

ture strain, then a sharp decrease, and then an increase. The irra-

diated samples followrthe same general pattern except for the absence

Table 3. Results of Tensile Tests on Control Samples of Heat 5085>

 

Strain ___§E£E§§;_B§i_ Elongation, % Reduction True

 

Test

Sfii;;gin z:gg:r- (gngl) Yield 312;??2? Uniform Total in(2§ea Fg::;gie
(°c) (%)
10410 25 - 0.05 ' 48,100 108,500  40.5  40.6 32.8 40
10409 200  0.05 39,300 101,400  44.7  45.1 33.4 41
10408 400 0.0 35,400 94,900  48.2  48.9 39.3 50
10407 400  0.002 35,200 98,500  48.6  49.0 36.3 45
10406 500  0.05 34,000 92,800 - 48.0  48.8  40.0 51
10405 500  0.002 35,800 90,100  37.5  38.2 28.8 34
10416 550  0.05 32,700 88,900 51.6 52.2 37.1 46
10417 550 0.002 36,500 79,500 29.4 30.1 24.1 28
10418 - 600  0.05 33,400 82,200  37.3 7.9  28.6 34
10420 600  0.002 35,200 69,000  25.7  26.9 19.1 21
10421 650  0.05 32,600 75,200  31.0  32.0 28.9 3%
10422 650  0.002 34,300 65,600  23.3  23.9 23.9 27
10404 650  0.0002 35,700 64,500 . 24.9 25.5 . " 19.7 - 22
10412 650  0.000027 30,000 59,700  15.2  22.2  2L.2 24
10423 760 0.05 33,900 64,900  25.6  27.5  19.7 - 22
10424 760 ~ 0.002 31,600 52,000  12.2  27.0 31.6 38
10426 < 850  0.05 33,600 50,600  10.0  33.0  37.1 - 46
10425 850  0.002 24,100 . 26,300 4.2 3607 25.3 29

 

.+ Banmealed 2 hr -at 900°C;-expééed'fo a static vessel of'baireh'fuél galt for
22,533 hr at 650°C. , | |

 
 

18

Table 4. Postirradiation Tensile Properties of Hastelloy N (Heat 5085)%

 

 

Test i True
Stress, psi :
Specimen Temper- Strain __—L-E_—Ultimat Elongation 4 Reduction Fracture
Namber  ature  oO0%,) Yield goqy®  Uniform Total  1RQY€®  gtrain
(°c) 7 | (%)
8806 25 0.05 53,900 89,000 22.0 22.1 19.6 22
8805 200 0.05 44,800 85,600 26.4 27.0 26.4 31
8804 400 0.05 41,600 80,900 30.2 30.5 24.3 28
8803 400 0.002 41,300 80,100 28.4 29.2 26.8 31
8802 500 0.05 40,800 73,200 24.8 25.0 23.2 26
8801 500 0.002 39,400 61,100 1n.7 12.4 13.9 15
8812 550 0.05 37,900 62,600 144 14.9 4.3 15
8813 550 0.002 36,200 49,100 5.8 6.2 - 2.1 10
8814 600 0.05 36,700 56,200 9.8 10.5 13.3 14
8816 600 0.002 39,000 48,500 4.5 4.9 6.7 7
8817 650 0.05 36,400 50,800 8.8 9.3 9.1 10
8818 650 0.002 37,200 42,800 4.7 5.0 5.8 6
8800 650 0.002 35,200 37,700 2.1 2.4 2.8 3
8799 650 0.000027 32,900 34,500 . 1.2 1.4 6.1 6
8819 760 0.05 30,700 36,300 6.1 6.4 6.4 7
8820 760 0.002 35,100 - 35,100 1.7 2.0 2.5 3
8821 850 0.05 32,100 32,700 2.3 2.3 2.1 2
8822 850 0.002 22,200 22,200 1.6 1.7 2.6 3

 

®annealed 2 hr at 900°C prior to insertion in reactor. Irradiated to a thermal
fluence of 1.5 X 1021 neutrons/cm? over a period of 22,533 hr at 650°C.

ORNL- DWG 70-7290

 

 

80
t | { 1
STRAN
RATE UNRRADIATED RRADIATED
70 bt
0.05 o .
0.002 a a
0.0002 a .
60 I 0.000027 - -

 

 

 

TR
NN

FRACTURE STRAN (%)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

p

30 "E-ix ] L

//”/ \\0\ \w-__

. e
20

\
0 ‘\\\ -
'--..-,_.-
A ha
=TT

° 0 100 200 300 400 500 600 700 800 800

TEST TEMPERATURE (°C)

- Fig. 9. Fracture Strains of Hastelloy N (Heat 5085) After Removel
from the MSRE and from the Control Facility. All samples annealed 1 hr
at 900°C before irradiation for 22,533 hr at 650°C to a thermal fluence
of 1.5 X 10?2 neutrons/cm?. |

i»
«}

+)

)

" a strain rate of C.05 min

19

of a ductility increase at high temperatures. However, the levels of
the fracture strains are lower for the irradiated material over the
entire temperature range. For both materials the fracture strain
decreases with decreasing strain rate. Another difference in behavior
between the irradiasted and unirradiated samples is that at the highest
strain rate (0.05 min~!) the fracture strain begins its precipitious drop
at a lower temperature for the irradiated material.

Further characteristics of the effects of irradiation on the ten-
sile properties of heat 5085 are apparent when the ratios of the irra-
diated and unirradisted properties are compared (Fig. 10). The yield
stress is from 10 to 20% higher for the irradiated material at test tem-
peratures up to 760°C. The ultimate tensile stress is about 20% lower
for the irradiated material and drops even further as the test tempera-
ture is increased above 500°C. The fracture strains of the irradiated
samples are about 50% of those of the unirradiated samples up to about

500°C, above which the reduction is even greater.

ORNL- DWG 70 - 720¢

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1.4
1.2 O
e
_ © <&
a N
E 1.0 ~ o
9 o S~
Q .
% 0.8 Ky &
2 0
b : a
< I : F—l |
[=] : a
g 0.6 — !_._\ \._____
E Coo
0.4 —
O YIELD STRESS - o | ~~g
0.2 = A ULTIMATE STRESS " — ' ~
0 FRACTURE STRAN = | 1. . ‘ \\

 

O 100 200 300 ' 400 500 600 700 _ 80O 900
: © TEST TEMPERATURE (°C) ,

.Fig; 10. Comparison 6fr£helTehsile'Properties of Control and Irra-
diated Hastelloy N (Heat 5085). Samples irradiated to a fluence of

1.5 x 1021 neutrons/cm? over a period of 22,533 hr at 650°C. Tested at
e 7

 

 
 

20

The results of tensile tests on the samples of heat 5065 are sum-
marized in Tables 5 and 6. The fracture strains of these samples are
shown in Fig. 11 as a function of test temperature. The results are
quite similar to those shown in Fig. 9 for heat 5085. A notable excep-
tion is the much higher fracture strains at low temperatures of heat 5065
after irradiation. The ratios of the irradiated and unirradiated ten-
sile properties are shown in Fig. 12. The yield stress was not altered
appreciably by irradiation. The ultimate tensile stress was decreased
about 10% by irradiation at low test temperatures and up to 35% at high
temperatures. Similerly, the fracture strain was reduced about 10% at
low temperatures, and this reduction progressed to about 85% at high
test temperatures.

The progressive change in the fracture strain'with increasing
fluence is illustrated in Fig. 13 for heat 5085. The fracture strain
has been reduced over the entire range of test temperatures investigated.
A similar trend was noted at a slower strain rate of 0.002 min~1
(Fig. 14), although the absolute values of the fracture strains were
lower than noted in Fig. 13 at the higher strain rate of 0.05 min~!.
These samples have experienced various holding times at 650°C and some
of the property changes can be attributed to thermal aging. The frac-
ture strains for several sets of control semples of heat 5085 are com-
pared in Fig. 15. At low test temperatures and at 760°C and above the
fracture strain seems to show a progressive decrease with increasing
holding time at 650°C. At intermediate temperstures the behavior is
more complex. At a test temperature of 650°C, the results indicate a
progressive decrease in fracture strain up to an exposure time of
A15,289 hr and then an increase with further aging time. |

Heat 5065 was not included in one set of surveillance samples, but
there are sufficient daté to follow the property changes with fluence
and aging time at 650°C. The fracture strain is shown in Fig. 16 as a
function of temperature for various fluences. At low test temperatures,
excluding the one'apparently anomalous point, the fracture strain was |
actually higher for irradiation to fluences of 1.3 and 2.6 X 101°
neutrons /cm?. Higher fluences reduced the fracture strain at low test
temperatures, but not to values as low &s noted in Fig. 13 for heat 5085.

e

m
3

a)

"

1

ay

21

Table 5. Results of'Tensile Tests on Control Samples of Heat 506Sa

 

 

Test . ' : True
. —oLress, pst
Specimen Temper- S;:::n StressUltiiate Elongation, % Rzfi?:::Zn Fracture
Number ature . “oqy Yield Uniform Total Strain
u (min~1) Tensile (%)

(°c) (%)
10384 25 0.05 61,200 126,500 48.5 49.8 36.77 46
10388 200  0.05 42,000 107,900 475  49.4  40.16 51
10392 400  0.05 42,600 102,000 48.9  50.8  45.31 60
10383 400  0.002 43,900 106,000 47.5 4707 39.30 50
10391 500  0.05 42,200 99,400 48.1  49.6  37.54 47
10378 500 0.002 45,800 98,700 45.5 46.2 34.08 42
10395 550 0.05 - 42,500 . 98,800 48.2 49.9. 38.14 48
10375 550 0.002 43,800 84,600 25.5 26.1 27.51 32
10377 600  0.05 41,800 93,800 42.0  43.0  34.08 42
10393 600 0.002 39,400 76,700 25.6 26.1 34.29 42
10382 650  0.05 41,500 81,900 26.4  26.8  23.84 27
10394 650 0.002 39,200 71,700 23.1 23.6 20.89 23
10397 650 0.0002 41,800 71,200 18.8 19.6 17.88 20
10400 650 0.000027 42,900 60,800 7.5 - 19.7 24.57 28
10385 760 0.05 32,900 70,500 24.3 27.3 22.85 26
10398 760 0.002 41,100 42,100 5.7 39.6 45,48 61
10390 850  0.05 35,500 49,500 8.7  38.1  45.31 60
10379 850 0.002 24,800 25,100 2.9 56.5 47.15 64

 

Bpnnealed 2 hr at 900°C prior to insertion in the reactor. Exposed to a static
vessel of barren fuel salt for 22,533 hr at 650°C.

Table 6. Postirradiation Tensile Properties of Hastelloy N (Heat 5065)%

 

 

Test . . ' . True

Specimen = Temper- St;::n : Stressuizizat Elongation, % R:guggzzn Fracture

Number atgre (min=1) Yield Tensilee Uniform Total (%) Strain
(°c) (%)
8833 25 0.05 56,700 109,500 42.5 42.8 3.32 3
8832 200 0.05 46,000 99,900 42.8 4.6 4.44 5
8831 400 0.05 43,700 - 94,300 | 39.0 39.3 31.22 37
8830 400 0.002 43,200 91,300 35.7 37.2 27.31 32
8829 500 0.05 43,600 82,100 29.1 29.5 22.98 26
8828 500 0.002 39,800 63,200 10.9 11.3 10.70 11
8839 550 0.05 41,200 68,700 15.5 15.7 16.83 18
8840 - 550 0.002 - 37,800 - 57,700 7.1 7.2 8.93 9
8841 600 0.05 - 40,500 60,100 8.6 9.0 9.30 10
8842 600 0.002 - 41,000 - 50,400 4.3 4.5 7.40 8
8843 650 - 0.05 39,700 52,100 - 6.0 6.2 ~ 7.37 8
8844 650 0.002. 40,000 42,800 3.0 3.1 4.89 5
8827 650 0.0002 34,300 36,100 2.1 2.2 3.30 3
. 8826 - . 650 0.000027 28,400 -36,900 1.1 1.2 1.1 1
- 8845 760 0.05 37,800 41,400 2.5 2.7 3.33 3
- 8846 - 760 0.002 - 35,200 . 35,200 1.3 1.3 2.86 3
8847 - 850 0.05 35,400 . 35,400 1.5 1.6 2.22 2
8848 g50 - 0.002 - 20,300 20,300 1.0 1.0  0.16 0

 

®annealed 2 hr at 900°C prior:to-insertion in the reactor. Irradiated to a thermal
fluence of 1.5 X 1020 neutrons/em? over a period of 22,533 hr at 650°C.

 
 

22

ORNL-~ DWG 70-7292

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

60 :
o T T N\ /
; ~ | fi /
= 40 \ K A 2
8 AY,
&
= .
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w 30 / -
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L RATE UNIRRADIATED IRRADIATED \ ~=>2
(mifl_i) -~ @
1 0.05 o ® 'Y |
01 0002 - a a AN
0.0002 o s ~L T
0.000027 ° - —p
o | _ | ! -
o 100 200 300 400 500 600 700 800 900

TEST TEMPERATURE {°C)

Fig. 11. Fracture Strains of Hastelloy N (Heat 5065) After Removal
from the MSRE and from the Control Facility. All samples annealed 2 hr
at 900°C before irradiation for 22,533 hr at 650°C to a fluence of
1.5 X 10?1 neutrons/cm?.

ORNL—DWG TO- 7293

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1.2
0
o
1.0 /—"" ~— o
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o« a ULTIMATE STRESS
o FRACTURE STRAIN o
0.2 \\ =
‘-—______
o
0
o 100 200 300 400 500 600 700 800 500

TEST TEMPERATURE (°C)

Fig. 12. Comparison of the Tensile Properties of Control and Irra-
diated Hastelloy N (Heat 5065). Samples irradiated to a fluence of
1.5 X 102! neutrons/cm® over a period of 22,533 hr at 650°C. Tested at
a strain rate of 0.05 min~1.

b1

(n
.

n

0

23

ORNL-—-OWG 70~-7294

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

80
© ANNEALED 2 hr AT 900°C
a 1,3 x 10" neutrons /em2, 1100 hr
0 2.6 % 10'? neutrons/cm2, 20,789 hr
70 20 2
v.4.3 x10%Y neuirons/cm<, 4800 hr
094 x1029 neutrons/cmé<, 15, 289 hr
< {5x 102 neufrons/cma. 22,533 hr
60 o
o 7 ] 0
& 50 . L ! a /
z ] )
b
g / v °
40
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=2 v
- :/ Q
5 .
& 30 N
[ et
20 \ v ”\ T \\\
10 3 %\#—-
0
0 100 200 300 400 500 600 700 800 900

Fig. 13. Fracture Strains of Hastelloy N (Heat 5085) After Irra-~
diation to Various Thermal Fluences in the MSRE.

rate of 0.05 min™?!.

TEST TEMPERATURE (°C)

Tested at & strain

ORNL-DWG €9-4467R

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

* I |
NEUTRON FLUENCE  TIME AT
{neutrons /em®) 650° C {hr}
a0 |— 8 13x10% 11,000 UNIRRADIATED /o,/
o 26x0° 20,789
2 v 13x4020 4800
=z o 94x10%° 15289 . A\ -
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£ 20 5
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0
0 100 200 300 -

 

'n-:sf TEMPERATURE (°C)

Fig. 14 Postlrradiation Ten31le Propertles of Hastelloy N
(Heat 5085) After Exposure to. Various. Neutron Fluences.
 strain rate of 0.002 min~l.

Tested at a

 
 

ORNL-DWG 69-4468R

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

60 l o
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50 1, T /
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T 20— AGING TIME AT Ny
650°C (hr)
oo
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0 — A 15,289
e 22,533
Lo
0 100 200 300 400 500 600 700 800

TEST TEMPERATURE (°C)

Fig. 15. Variation of the Tensile Properties of Hastelloy N
(Heat 5085) with Aging Time in Barren Fuel Salt at 650°C.

strain rate of 0.05 min™1l.

Tested at a

ORNL-DWG 70—7295

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

70
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4 1.3 x 10'? neutrons /cm? ‘\ ‘]'
o 2.6 x40' neutrons/em?
ol o zc"neu rons/ mz (5% 102! \ ‘b\ o
9.5 x 10" neutrons /cm X\ \ N .
a
¢ 1.5u10""-m:ulrons/.:fn2 \&\ \
1&-—_5
o | —
0 100 200 300 400 5C0 600 700 800

TEST TEMPERATURE (°C)

900

Fig. 16. Variation of the Postirradiation Tensile Propertles of
Hastelloy N (Heat 5065) with Thermal Neutron Fluence.

i
4

1)

)

25

At test temperatures above 550°C the fracture strain decreases progres-
sively with increasing fluence. As shown in Fig. 17 some of these
changes in fracture strain can be attributed to thermal aging at 650°C.
Except at test temperatures above 750°C, the fracture strain is lowest
for the material aged 15,289 hr and is improved after aging 22,533 hr.
The changes in properties at low test temperatures are less than those
for heat 5085 (Fig. 15).

ORNL—OWG 70-7296

 

70

 

60 ' /[

 

g

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FRACTURE STRAIN (%)
o H
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AGING TIME "/
{hr}
B0
{0 +— o 15,289
a 22,533
o | _
0 100 200 300 400 500 600 700 800 900

- TEST TEMPERATURE (°C)

Fig. 17. Effects of Thermal Aging at 650°C on the Tensile Proper-

 ties of Hastelloy N (Heat 5065) at a Strain Rate of O. 05 min~1.

As d1scussed prev1ous1y in thls series of reports, the changes in
fracture strain at low temperatures were not expected. Although the
fracture stralns have not reached values below 20%, we are still 1nter-
ested in its progress1on Our experlence to date is summarized in

F1g._18. If the noted effects were due simply to thermal aging, then

" the results should correlate with the time at 650°C. Obviously such a

_ correlation does not exist. ‘Where data are available for_pairs of irra-

diated and unirradiated samples, the irradiated sample has the lower

 

 
 

26

ORNL-DWG 70-7297

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

70 T T I
HEAT IRRADIATED UNIRRADIATED
5065 ¢ o
60 - 5085 = ’ o .
® 50 o
o o
®
)
Pl .
40 - a
=
& . .
- 1 1 |
w ! 1 1
w 30 ¥ ! s 1
&« ' : 1
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L 29 S o 8 g' b
Q e Q 2 %
= * = ® o
" " < ° o
10
0
0 q 8 t2 16 20  (xt0%)

TIME AT 650°C (hr)

Fig. 18. Variation of the Fracture Strain at 25°C with Annealing
Time and Thermal Fluence.

fracture strain.

embrittlement.

must be associated with carbide precipitation.

This indicates that irradistion has a role in the
We previously showed that the ductility could be
recovered by an anneal of 8 hr at 871°C and concluded that the changes

9

The higher susceptibil-

ity of heat 5085 to this type of embrittlement is not understood, since
heat 5065 actually has a higher carbon concentration (Table 1, p. 4).
The data in Fig. 18 defy extrapolation, so one cannot conclude whether

the room temperature embrittlement is likely to become worse.

‘In these discussions of tensile properties we have emphasized the

changes in fracture strain and made little mention of strength changes.

Some pertinent data are summarized in Table 7 for heat 5085. The sam-
ples were tested at 25 and 650°C and include thrée histbries: (1) as
annealed, (2) thermally aged, and (3) irradiated. The changes in yield
strength are likely significant, but the main point is that they are

 

°H. E. McCoy, An Evaluation of the Molten-Salt Reactor Experiment
Hastelloy N Surveillance Specimen — Third Group, ORNI~TM-2647 (1970),

p. 7.

"
0

3)

27

Table 7. Comparison of the Tensile Properties of
Heat 5085 Before and After Irradiation®

 

Ultimate Tensile
Heat Treatment Yield Stress, psi Stress, psi Fracture Strain, %
| at 25°C at 650°C at 25°C at 650°C at 25°C at 650°C

 

Annealed 2 hr at 900°C 51,500 29,600 120,900 75,800 53.1 33.7
Annealed 2 hr at 900°C and 48,100 32,600 108,500 75,200 40.6 32.0
aged 22,533 hr at 650°C « : |
Annealed 2 hr at 900°C, 53,900 36,400 89,000 50,800 22.1 9.3

irradiated for 22,533 hr
at 650°C to a thermal
fluence of

1.5 X 10?1 neutrons/cm?

 

8Strain rate of 0.05 min~1.

small considering the long thermal history of 22,533 hr. The ultimate
tensile strength is reduced due to the lower fracture strain and the
resulting fact that the tést is interrupted before it reaches the high

" ultimate strength.noted for the as-annealed material. Similar results

are shown in Table 8 for heat 5065. The yield strength is increased by
aging,‘but the main effects are on the fracture strains and the ultimate

tensile strengths.

Table 8. Comparison of the Tensile Properties of
Heat 5065 Before and After Irradiation®

 

Ultimate Tensile Fracture Strain, %

Stress, psi ~ =0°
#t 25°C at 650°¢ °¢ °C at 650°C

Yield Stress, psi
at 25°C at 650°C

 

Heat Treatment

 

 

 

Amealed 2 hr at 900°C 56,700 32,100 126,400 81,200  55.3 34.3

Annealed 2 hr at 900°C and 61,200 41,500 126,500 81,900  49.8 26.8
aged 22,533 hr at 650°C o | -

_ Annealed 2 hr at 900°C, 56,700 39,700 109,500 52,100  42.8 6.2

irradiated for 22,533 hr
at 650°C to a thermal
fluence of . _

1,5 X 102! neutrons/cm?

 

aStrain rate of O.OSVmin'1.~;‘

 
 

28

The results of creep-rupture tests on heats 5085 and 5065 from the &a)"
fourth group of surveillance samples are summarized in Tables 9 and 10. -

iy

The results are most interesting when they are compared with those

obtained on previous groups of surveillance samples so that the progres- .
sive changes in properties can be noted. The stress-rupture properties
of heat 5085 are shown in Fig. 19.
of samples (1.5 X 102! thermal neutrons/cm?) are not detectably differ-

ent from those irradiasted previously to a thermal fluence of 9.4 X 102°

The rupture lives of the last group

neutrons/cm?. In the unirradiated condition, the rupture lives for the
samples aged 22,533 hr are slightly less than in the as-annealed condi-

tion, but greater than those observed for samples aged for lesser times.

 

 

 

 

Table 9. Creep-Rupture Tests on Heat 5085 at 650°C
Test Specimen Stress Rupture . Rupture Reduction Mgnimum
: . . . reep
Number Number (psi) %;fi? S?;sln 1n(25ea Rate
(%/hr)
Unirradiateda .
7978 1406 55,000 22.1  21.0 23.4 0.288 .
7897 10415 40,000 500.1  24.5 28.5 0.0299
7896° 10414 32,400  2500.7  12.7 9.4 0.0045
7895° 10413 27,000  2187.1  5.40 5.40  0.0026
7894° 10412 21,500  2064.9  2.10 2.44  0.0011
7893° 10411 17,000  2000.0 1.1 0 0.0007
Irradiated®
R-969 8811 40,000 0.6 = 1.1 0.78
R-958 8810 32,400 31.6  0.54 0.0067
R-956 8809 27,000 60.2  0.57 '0.0055
R-957 8808 21,500 362.0  0.63 0.0010
R-964 8807 17,000 426.4  0.69 0.0004

 

®Annealed 2 hr at 900°C, exposed to static barren fuel salt for

22,533 hr at 650°C.

bDiscontinn.ed prior to failure.

CAnnealed 2 hr at 900°C, exposed to MSRE core for 22
650°C, received thermal fluence of 1.5 X 10?! neutrons/cm®.

3

533 hr at
'fl

29

Table 10. Creep-Rupture Tests on Heat 5065 at 650°C

 

 

: . Minimum
Test Specimen Stress Rggture _ Ruptu?e R?ductlon Creep
. ife Strain in Area
Number Number (pSl ) (hr) ( %) ( %) Rate
| (%/hr)
Unirra.dia.teda
7978 10406 55,000 22.1 21.0 23.4 0.34
7892 10388 40,000 = 651.2 36.3 29.9 - 0.0331
'7891b 10387 32,400 = 2144.3 4 7 37.7 0.0096
7890b 10386 27,000 2187.1 8.8 7.4 0.0040
7889b 10385 21,500 2373.3 3.6 2.2 0.0013
7888 10384 17,000 2000.0 2.8 0
| Irradiatedc
R-966 8838 40,000 0
R-962 8836 - 32,400 - 22.7 0.25 0.0075
R-954 8836 27,000 - 40.5 0.44% 0.0061
R-960 8835 21,500 46.2 0.47 0.0051
R-963 8834 17,000  1567.3 0.86 0.0004%

 

@pnnealed 2 hr at 900°C, exposed to static barren fuel salt for
22,533 hr at 650°C. |

bDiscontinued prior to failure.

CAnnealed 2 hr at 900°C, exposed to MSRE core for 225533 hr at
650°C, received thermal fluence of 1.5 X 1021 neutrons/cm?.

The minimum creep rates are shown as & function of stress in
Fig. 20 for heat 5085. As noted previously, neither irradiation nor

aging has a detectable effect on the creep rate. The data deviate from

- the line shown in Fig. 20 atfboth extremes. At high stresses, the irra-

diated samples fail at such low Strains that a minimum creep rate is not

 established over‘long‘éfiough'pefiod for measurement. At low stress

f_leVelS-the data deviate due to the semilog plot.

The fracture strain is the parameter_most affected by irradiation,

" and a variation of this parameter with minimum creep rate is shown in
- Fig.'Zl for heat 5085 at_650°c;;_There has been a continual deteriora-

tion of the fracture strain with increasing thermal fluence.

' Tensile and creep tests were run at 650°C at several different

. strain rates and stresses. The fracture strains from these tests are

plotted together in Fig. 22 although different parameters are controlled
 

30

ORNL~DWG 69-4470R

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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© 30 -  FLUENCE TIME AT ] Sl 3
e {neutrons/cm?) 650°C (hr) \é;c-._._ o| [&
« e UNIRRADIATED O / T~
20 |~ ® UNIRRADIATED 4800 / P T
a UNIRRADIATED 15,289 o] \
¢ UNIRRADIATED 22,533 L
ol ° 13 x w0 11,000 3 x10%° neutrons/ecm? (MSRE) AND
a26 x 10" 20,789 3-5 x 10%% neutrons/cm? (ORR)
o 1.3 x 10%° 4800
oL 094 x 102 15,289
< 1.5 x 103 22,533
L L [t iEL
1o 10° . o' 102 103 104
RUPTURE TIME (hr)
Fig. 19. Postirradiation Stress-Rupture Properties of MSRE Surveil-

 

 

 

lance Specimens (Heat 5085) at 650°C.

70 /
A
60 A
A
7
o8 ¢ 4
£
50 // p
. wt a A ;
2 ‘XB&T
o 40 IR £ o H
S &Il -AA THERMAL
= L‘l// FLUENCE TIME AT
o 4o ly| o1V 0 (neutrons/cm?) 650°C (hr)
fi 30 AT e UNIRRADIATED © il
e ‘ A o 4.3 x to% 11,000
20 Ly Ao a 2.6 x 4% 20789 ||
&4 o/f/ o 1,3 x 10%° © 4800
P4l ¢ 9.4 x 102 . 15,289
10 o 1.5 x 102! 22,533 1}
= UNIRRADIATED 4800
a UNIRRADIATED 45,289
o ¢ UNIRRADIATED 22,533
1074 1073 4072 -t 100 10!
MINIMUM CREEP RATE (% / hr)
Fig. 20. Minimum Creep Rate of Hastelloy N (Heat 5085) Surveil-

ORNL—-DWG 69-4471R2 -

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

lance Specimens from the MSRE at 650°C.
”»

4

a

FRACTURE STRAIN (%)

Fig. 21.

N w H

FRACTURE STRAIN (%)

i

0
104

 

31

ORNL-DWG 69-4472R

RANGE 2-5x1029 neutrons/cm? (ORR)

LT T

1.3x10%° neutrons/cm2

2.6x10'° 1.3x10%°

neutrons neutrons/cm?2

9.4x1020 1.5x102!
neutrons /cm? - neutrons /em?2
10~3 102 10! 100 10!

MINIMUM CREEP RATE (%/hr)

Variation of ‘Fracture Strain with Strain Rate for

Hastelloy N (Heat 5085) Surveillance Specimens at 650°C.

ORNL-DWG 70-7298

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

40
IR
UNIRRADIATED Lk
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30 =
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STRAIN RATE (%/hr)

F':i.g. 22. Fracture Strain at 650°C of Heat 5085 in the Unirradiated

~and Irradiated Conditions. TIrradiated to a thermal fluence of 1.5 X 1021

neutrons/cm® over a period of 22,533 hr at 650°C.

in the two types of tests. The most striking feature is.the marked

dependence of the fracture strain of the irradiated samples on the strain

rate and the rather weak dependerice of the fracture strains of the

unirradiated samples on the strain rate.

The irradiated samples at this

 
 

 

32

high fluence level show a general decrease in fracture strain»with
decreasing strain rate, with a possible slight increase in fracture
strain at very low strain rates.

The sensitivity of the fracture strain of heat 5085 at 650°C to
the helium content is illustrated in Fig. 23 for three strain rates.
This material contains 38 ppm B (Table 1, p. 4) that can yield an equiv-
alent amount of helium when transmuted. (Several factors contribute to
the relationship that 1 ppm of natural boron by weight leads to
1.1 ppm He on an atomic basis.) The points shown in Fig. 23 all come
from surveillance samples from the MSRE. The two higher strain rates
are obtained by tensile testing, and the fracture strain decreases with
increasing helium content. The 0.1% strain rate is the creep rate that
corresponds to the lowest fracture strain (Fig. 21). The fracture
strain drops abruptly with the presence of 1 ppm He and then decreases
gradually with increasing helium content.

The creep properties of heat 5065 are illustrated in a series of
graphs similar to that just presented for heat 5085. The stress-rupture
properties at 650°C are shown in Fig. 24 for heat 5065. The rupture

ORNL-DWG €9-7297R2

o
THERMAL FLUENCE \g” ‘:\o‘ &
(neutrons/cm?) 3 q® » 0"

36

32

28

FRACTURE STRAIN (%)
- - N N
o n » o D

F

 

0o 02 0.5 { 2 5 10 20 50
~ HELIUM CONTENT (ppm}

Fig. 23. Variation of the Fracture Strain with Calculated Helium
Content and Strain Rate for Hastelloy N (Heat 5085) at 650°C.
n

3}

3

33

ORNL-DWG 68-4474R

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

80 : : T T7TTT — T TTT717
THERMAL FLUENCE TIME AT
(neutrons/m?) 650°C (hr)
70 . - ¢ UNIRRADIATED o |
\\ u UNIRRADIATED 15,289
i "\ ¢ UNIRRADIATED 22,533
\\\ o |3x10 11,000
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. . '\\‘ 094 MO21 15,289
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g i . . . \-.::,:.. ‘ N
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g:_l 30 "--.___-- Q L ‘
k= 9 -""-L.“_ \\:-..\ \
9 ---\-o.. ™
20 S q
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107! 10° T} 102 10° 104

RUPTURE TIME (hr)

Fig. 24. Stress-Rupture Propertles of MSRE Surveillance Specimens
(Heat 5065) at 650°C.

times are equivalent for the samples irradiated to the two highest

fluences. The rupture times are also quite comparable with those shown
in Fig. 19 for heat 5085. The unirradiated samples that were aged for
22,533 hr at 650°C in static barren fuel salt have longer rupture lives
than the as-receivedrmaterial. The minimum creep rates are shown in
Fig. 25 and show a lack of sensitivity to any of the variables being
studied. The fracture strain is shown as a function of strain rate in

Fig. 26. The fracture straln decreases with increasing fluence up to

the two highest fluence levels, where the strains are about equivalent.

This heat of materlalrshcws a-ductlllty minimum with the fracture strain

~ increasing slightly with’ decreasing creep rate except for the samples

”'ShOWIHg the lowest fluence.

The tensile and creep test results for heat 5065 have been combined

in Fig. 27. These results agaln ‘show the marked dependence of the frac-

ture straln on the straln rate for the irradiated materlal. A comparison
with the similar plot for heat 5085 (Fig. 22) reveals some slight dif-

ferences in the fracture strains of the two heats, but shows generally

 
 

ORNL-DWG 69-4475R

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

] T 1 reinrn T. T T V00178 A .
THERMAL FLUENCE  TIME AT ,/
(neutrons/em?) €50°C (hr}) /’
60 | ® UNIRRADIATED )
® UNIRRADIATED 15,289 Z
¢ UNIRRADIATED 22,533 <ol [
e 1.3x10" 14,000 h
50 — a 26x10" 20,789 v
= 0 9.4 x 10%° 15,280 || LA "
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8 40 180
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20 o ¥ TRl
¢ | o 4 /
’d
10
0
104 03 3072 10~! 10° 10'

MINIMUM CREEP RATE (% /hr)

Fig. 25. Minimum Creep Rate of Hastelloy N (Heat 5065) Surveil-
lance Samples from the MSRE at 650°C.

ORNL-DWG 69-4476R

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

7
RIEBRERLEREE
THERMAL FLUENCE TIME AT 650°¢C
6 { neutrons/cm?2) (hr)
a 1,3 x10% 11,000
a 26x%x10% 20,789
5 o 9.4x1022‘: 15,289
3 v 1.5%10 22,533
e AVERAGE 2-5x10%°
z neutrons/cm® {ORR)
4 \’
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1074 03 1072 10! 10° 10!

MINIMUM CREEP RATE (% /hr)

Fig. 26. Variation of Fracture Strain with Strain Rate for
Hastelloy N (Heat 5065) Surveillance Specimens at 650°C.
»n

FRACTURE STRAIN (%)

35

ORNL-DWG 70-7299R

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

50
L TTHIT l -
| UNIRRADIATED| | |
40
30
. /)bf‘
o —TT11]
20 . v
10 |
8 P TTHN
IRRADIATED ' P
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e TENSILE TESTS e | ‘,.—/
4 © CREEP TESTS L pe
| I~ |
J-—-"""‘
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o Y
10 1073 102 10! 10° 10’ 102 10°

STRAIN RATE (% /hr)

Fig. 27. Fracture Strains at 650°C of Heat 5065 in the Unirra-
diated and the Irradiated Conditions. Irradiated to a thermal fluence
of 1.5 X 10?1 neutrons/cm?® over a period of 22,533 hr at 650°C.

that the two heats respond to irradiation quite similarly. Heat 5065
contains about 23 ppm B (average of the four values in Table 1, p. 4),
and the variation of the ffacture strain with helium content is shown

in Fig. 28. The behavior is quite similar to that noted in Fig. 23 for
heat 5085. Thus, the lower borori (and hence helium) :concéntration in
heat 5065 results in .a‘béfitthe same properties as the higher boron level
in heat 5085. IR -

Mechanical Property Data — Modified Hastelloy N

Two heats of modifiéd. .H—astell_oy N were exposéd to the MSRE core for
6244 hr at 650°C and 'recie’iifed”_a,i;hérmal neutron fluence of 5.1 X 1020 |
neutrons/cm? (Table 2 , P 6) Both of these alloys were modified with
titanium — heat 7320 cdzitia}i-néd' 0.65% and heat 67-_-551 contained 1.1%.

_ The tensile pr0pertieS of ihea'f 7320 are summarized in Tables 1l and

12 for the control and irradiated samples, respectively. The fracture

 
 

Helium Content and Strain Rate for Heat 5065.

36

ORNL-DWG TO-7300

THERMAL FLUENCE (neutrons/cm?)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

50

40 1L3%10'° 2.6 x10'° 9.4x10%° 1.5 x10%
——AN
' ! 11 IR
DEFORMATION RATE
N~ %/ he)
2
< o 300
S \\\\\\\\\. .12
2 . : \\‘ & 04
z N N
£ <
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8 : -
\ ] o
\\hn-. \.--.h-
0 L ey
o 0.2 0.5 1 2 5 10 20
' HELIUM CONTENT (ppm)
Fig. 28. Variation of Fracture Strain at 650°C with Calculated

 

 

Table 11. Results of Tensile Tests on Control Samples
of Heat 7320%

Test . : : True
Specimen Temper- Straln *‘§E£E§%izgiigg Elongation, % Rzgfiizzzn Fracture
Number a?‘j-'g()’-‘ (min 1) Yield . -:7e  Uniform Total (%) Si(:;!)a.in
6ls 25 0.05 68,800 122,600 48.2  49.7  40.83 52
9482 200 0.05 60,000 107,600 40.6  41.9  43.75 58
9470 400  0.05 56,600 103,300 46.9 49.1 43,09 56
9481 400  0.002 54,100 96,000 45.2  46.2  40.64 52
9472 500 0.05 54,500 99,200 47.9 51.1 44 .38 59
9456 500 0.002 53,600 96,700 45,1 46.2 41.21 53
9467 550 0.05 51,700 95,500 49.2 52.0 45.39 61
9466 550 0.002 49,400 85,300 36.7 37.7 32.99 40
9458 600 0.05 52,500 91,900 41.4  45.5 4111 53
9476 600 0.002 54,800 80,500 18.2 19.7 23.35 27
9478 650 0.05 49,300 78,600 - 28.4 30.6 36.67 46
9461 650 0.002 47,700 72,800 14.5 15.5 19.46 22
9475 650 0.0002 50,800 67,900 12.3 13.7 14.40 16
9474 650 0.000027 47,800 59,100 5.6 9.6 13.88 15
9477 760 0.05 50,100 71,500 12.0 13.1 11.69 12
9480 760 0.002 44,200 45,100 4.2 17.2 19.07 21
9471 850 0.05 45,000 48,200 5.6 10.2 12.39 13
9457 850 0.002 27,500 27,800 1.6 10.1 11.97 13

 

®Annealed 1 hr at 1177°C and exposed to a vessel of static barren fuel
salt for 7244 hr at 650°C
»

i

»

"

»

37

Table 12. Postifradiation Tensile Properties of
Hastelloy N (Heat 7320)%

 

 

Test . . True
DLress, psl

Specimen Temper- ngiln St?essUltfi % Elongation, % R:duiilon Fracture

Number ature Ate. | Yield 1@& © tal n rrea Strain
°¢) (min=%) Tensile Uniform Tota (%) (%)
9221 25 0.05 69,800 116,600 43.8 45.2 35.48 b
9222 200 0.05 59,000 102,900 40.7 41.7 35.69 by
9223 400 0.05 54,100 93,500 40.5 41.6 33.09 40
9224 400 0.002 58,200 94,100 38.0 38.7 31.67 38
9225 500 0.05 54,900 89,100 39.1 41.1 36.97 46
9226 - 500 0.002 55,400 83,300 31.1 32.5 26.61 31
9215 550 0.05 61,100 88,200 26.8 27.6 26.26 30
9214 550  0.002 61,000 74,700 9.2 1.1 21.27 24
9213 600 0.05 60,800 81,700 18.4 19.6 27 .74 32
9212 600" 0,002 61,700 69,800 4.3 5.5 11.83 13
9211 650 0.05 67,400 74,900 8.0 9.5 7.10 7
9210 650 0.002 57,100 68,500 4.3 5.1 9.08 10
9227 650 0.0002 49,500 58,200 4.3 4.7 7.54% 8
9228 650 0.000027 37,500 49,400 2.0 2.9 3.00 3
9209 760 0.05 56,700 63,700 3.8 4.0 6.15 6
9208 760 0.002 48,700 48,900 1.6 3.0 - 6.15 6
9207 850 0.05 37,900 38,200 1.6 2.2 7.25 8
9206 850 0.002 24,800 24,800 1.0 1.6 2.23 2

 

@annealed 1 hr at 1177°C. Irradiated to a thermal fluence of 5.1 X 102°
neutrons/ecm? over a period of 7244 hr at 650°C.

strains are shown as a function of test temperature in Fig. 29. Irra-
diation reduces the fracture strain over the entire range of test tem~
peratures, but the magnitude of the decrease is much greater at elevated
temperatures. The sharp drop in fracture strain with increasing temper-
ature occurs at a lower temperature for the irradiated material. The
ratios of the irradiated“éfid'unifradiated.pr0perties are shown in

Fig. 30. The yield stresSjis higher'for the irradisted material over

the test temperature rangeof'550.to 760°C. The ultimate-tensile strength
is about the same for the irradiated and unirradiated material but shows
& gradual decline with increasing test temperature. The fracture strain
 decreased due to irradiation,rfiith the reduction being about 80% at 850°C.
'~ Some idea of the tensile properties of heat 7320 due to aging and
irradiation can be obtained from Table 13. The yield strengths at 25 and
" 650°C are increased by aging;fat a test temperature of 650°C a further

increase occurs due to irradiation. The changes in ultimate tensile

 
 

 

38

ORNL—-DWG 707304

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

60
50 O e k\
-
— — ol —
1r———~—-—-—.___~0 \
= 40 —
& \
s N \
2 ~
& \ )
& \
& \
-
2 | \
& 50 l—— STRAIN RATE UNIRRADIATED IRRADIATED \
' (min™") \ t\ A
005 ° . ' \
0002 a . \o-
. Qo002 o u \4 M P—r
fo 0000027 o - — e -
~ P D
- _:g.__,_ .
o
0 100 200 300 400 500 600 700 800 900

TEST TEMPERATURE (°C)

Fig. 29. Fracture Strains of Heat 7320 After Removal from the MSRE
and from the Control Facility. All samples annealed 1 hr at 1177°C

before irradiation to a thermal fluence of 5.1 X 10°° neutrons/cm® over
a period of 7244 hr at 650°C.

ORNL—- DOWG 70-7302

 

2 * / /\\

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

2 10}-e ¢ e
: A :i 4 A
E o ‘———__-\“ A = 4
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=
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o
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=
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3
o © YIELD STRESS \
S 04 & ULTIMATE STRESS \
« 0 FRACTURE STRAIN \
a a
\-I-h.
0.2
o

 

 

 

 

] 100 200 300 400 500 600 700 800 900
: TEST TEMPERATURE (°C)

Fig. 30. Comparison of Unirradiated and Irradiated Tensile Proger-
ties of Heat 7320 After Irradiation to a Thermal Fluence of 5.1 X 10<0
neutrons/cm® Over & Period of 7244 hr at 650°C. Tested at a strain rate
of 0.05 min~1.
 

»

»

39

Table 13. Comparison of the Tensile Properties of
Heat 7320 Before and After Irradiation®

 

. ' Ultimate Tensile .
Yield Stress, psi St 1 Fracture Strain, %
Heat Treatment 2t 25°C at 650°C ——enoo8s P82  SET555C at 650°C

 

| at 25°C at 650°C
Amnealed 1 hr at 1177°C 38,500 25,600 107,300 72,700  75.4 53.5

Annealed 1 hr at 1177°C, 68,800 49,300 122,600 78,600  48.1 30.6
aged 7244 hr at 650°C .
Annealed 1 hr at 1177°C, 69,800 67,400 116,600 74,900  45.2 9.5

irradiated for 7244 hr
at 650°C to a thermal
fluence of '
5.1 X 1029 neutrons/cm?

 

Srested at & strain rete of 0.05 min~l.

strength are rather small and likely within experimental error. The

fracture strain is decreased by aging and by irradiation; the magnitude

 

of the decrease is much larger at the test temperature of 650°C.

The results of tensile tests on the control and irradiated samples o
of heat 67-551 are given in Tables 14 and 15, respectively. The fracture |
strain is shown as a function of test temperature in Fig. 31. The frac-
ture strain at low temperatures is not influenced by irradiation, but
above 400°C the fracture strain is decreased by irradiation. However,
the fracture strains at high temperatures are higher than those for
either heat 7320 (Fig. 29) or the standard alloy (Figs. 9 and 11, pp. 18
and 22, respectively). The ratios of the irradiated and unirradiated
tensilé properties are Shdwn in'Fig. 32. The yield stress is generally
about 25% lower for the irradiated material. The ultimate tensile
strength is reduced only abOut_lO% by irradiation when tested below 550°¢C,

‘but is reduced up to 30% as the test temperature is increased. The frac-

ture strain is unaffected byrirradiation up to test temperatures of 500°C;

~above this temperature:théifractfire strain decreases progressively with

- increasihg test_temperature;f Some values of the tensile properties at

25 and 650°C are given_in_Tqblerlé, Aging at 650°C seems to increase
the yield_fitrength at both 25 and 650°C; however, the strength of the
irradiated material is quite close to that of the as-annealed material.

The ultimate tensile strength is increased only slightly by aging and
 

Table 14. Results of Tensile Tests on Control Samples of Heat 67-551%

 

 

 

 

rai Stress, psi o True
Specimen Temg:izture S;aten ,fiit£m3te Elongation, % Rifi?fi:ian Fracture

Number (°c) (min=1) Yield  pepgile ~ Uniform Total (%) S?;!;in
9403 25 0.05 62,400 120,200 50.3 52.3 41.59 - 54
9419 200 0.05 54,100 108,700 47.5 50.1 45,22 60
9422 400 0.05 52,300 104,300 48.7 50.4 46.66 63
9417 400 0.002 51,400 101,700 50.4 51.5 43 .84 58
9506 500 0.05 48,100 89,900 50.4 53.6 46.58 63
9423 500 0.002 50,400 92,100 42.0 43.6 36.67 46
9402 550 0.05 45,200 94,100 52.0 - 53.7 41.78 54
9418 550 0.002 47,600 80,400 26.3 27.9 31.67 38
9416 600 0.05 42,800 87,800 46.8 49.8 43.09 56
9428 600 0.002 46,200 79,600 23.1 25.3 26.37 31
9410 650 0.05 41,900 84,600 40.4 42.6 37.07 46
9407 650 0.002 - 42,700 76,300 27.3 - 31.5 25.42 29
9426 650 0.0002 46,000 74,800 22.3 30.2 28.87 34
9415 650 0.000027 39,900 60,900 11.9 21.7 23.23 26
9424 760 0.05 45,600 78,000 21.6  28.8 28.87 34
9421 760 0.002 43,000 45,200 5.5 36.2 33.98 42
9425 850 0.05 41,300 51,000 6.2 23.6 26.14 30
9420 850 0.002 25,500 26,200 2.3 22.4 18.81 21

 

gAnnealed 1l hr

650°C.

-

at 1177°C prior to exposure to a vessel of static barren fuel salt for 7244 hr at

0%

 
Table 15. Postirradiation Tensile Properties of Hastelloy N (Heat 67-551)%

 

 

 

 

. Test Strain Reduction True
Specimen . Temperature Rate .Stresgl P?i Elongation, % in Area Fracture
Number (°c) (min"l)  viela iUl Tniforn  Total = (g) S%;?ln
9167 25 0.05 49,600 107,800 50.6 51.0  36.87 46
9168 200 0.05 39,800 94,700 50.0 53.2 42.15 55
9169 - 400 - 0.05 . 39,200 91,800 48.7 50.7 38.44 49
9175 400 0.002 40,400 88,100 51.3 52.6  39.69 51
9176 . - 500 - 0.05 - 37,000 83,400 52.1 52.9 38.34 48
9172 500 1 0.002 37,000 67,900 25.8 27.4  29.21 35
9161 550 0.05 34,500 80,400 42.8 445 34.92 43
2160 550 0.002 - 39,500 57,100 14.7 17.0 23.35 27
9159 600 0.05 37,000 71,300 31.9 33.9 39.59 50
9158 600 0.002 41,900 61,800 18.0 19.2 22.68 26
9157 - 650 0.05 31,300 56,600 20.7 22.6 28.76 34
9156 650 0.002 31,200 53,300 16.0 16.9 19.20 21
9173 650 0.0002 35,100 57,100 12.3 14.5 13,74 15
9174 650 0.000027 34,800 51,300 7.5 12.9 12.5 13
9155 760 0.05 20,700 51,500 13.8 14.8 8.79 9
9154 760 0.002 20,200 38,400 5.7 9.2 10.74 11
9153 - 850 0.05 26,700 34,700 4.8 6.5 7 .54 8
9152 850 0.002 22,300 22,700 2.2 6.2 5.2 5

 

®annealed 1 hr at 1177°C.
period of 7244 hr at 650°C.

Irradiated to a thermal fluence of 5.1 X 1020 neutrons/cm? over a

v
 

42

. - ORNL-DWG TO-T303R
60

 

 

 

 

 

F-9
(=}
r
/
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-
-
nd'
/
. //
I.‘
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FRACTURE STRAIN (%)
3

 

 

 

 

 

 

 

 

 

\ .
20 I sTRAIN : N
RATE  UNIRRADIATED IRRADIATED Mad—a. N
y ~ ~
(min~') s\ ‘.\
10 - 0.002 A A R
0.0002 o . e,
0.000027 > -
0 100 200 300 400 500 600 700 800 900

TEST TEMPERATURE (°C)

Fig. 31. Fracture Strains of Heat 67-551 After Removal from the
MSRE and from the Control Facility. All samples annealed 1 hr at 1177°C
before irradiation to a thermal fluence of 5.1 X 102° neutrons/cm® over
7244 hr at 650°C.

ORNL—DWG 70-7304

 

 

 

1.0

 

 

O.B o . \

 

 

 

0.6

~ © YIELD STRESS NG
04

a ULTIMATE STRESS

* @ FRACTURE STRAIN \\

 

 

 

 

 

RATIO {IRRADIATED/UNIRRADIATED)

0.2

 

 

 

 

 

 

 

 

 

 

 

o

 

0 100 200 300 400 500 600 700 800 900
TEST TEMPERATURE (°C)

Fig. 32. Comparison of Unirradisted and Irradiated Tensile Proper=-
ties of Heat 67-551 After Irradiation to a Thermal Fluence of 5.1 X 10°°
neutrons/cm® over a period of 7244 hr at 650°C. Tested at & strain rate
of 0.05 min~1.
43

Table 16. Comparison of the Tensile Properties of Heat 67-551
Before and After Irradiation®

 

Ultimate Tensile

Yield Stress, psi Stress, psi Fracture Strain, %
Heat Treatment _o__z_l’_;_ —_— s °
at 25°C at 650°C at 25°C at 650°C at 25°C at 650°C

 

 

Annealed 1 hr at 1177°C = 44,600 26,900 113,600 78,900 79.6 57.3
Annealed 1 hr at 1177°C, 62,400 41,900 120,200 84,600  52.3 42.6
aged 7244 hr at 650°C

Annealed 1 hr at 1177°C, 49,600 31,300 107,800 56,600 51.0 R2.6

aged 7244 hr at 650°C,
irradiated to & thermal
fluence of

5.1 X 1020 neutrons/cm®

 

Srested at & strain rate of 0.05 min~!

decreased by irradiation. At 25°C the fracture strain is reduced from
79.6% to 52.3% by aging, and irradiation does not cause any further
change. At 650°C the fracture strain is decreased by aging and decreased
further by irradiation.

The creep-rupture properties of heat 7320 are summarized in
Table 17. These results are compared in Figs. 33, 34, and 35 with those
for unirradiated samples that were given a pretest anneal of 1 hr at
1177°C. The stress-rupture properties in Fig. 33 show that the aging
treatment given the controls, 7244 hr at 750°C, increased the rupture
life. The irradiated samples failed after longer times than did the
unirradiated sa.mples that were sn.mply annealed 1 hr at 1177°C, but
'ruptured in shorter times than the control semples However, the minimum
- ereep rates shown in Fig. 34 seem'to fall within & common scatter ‘band
~ for all conditions._ Thus, neither 1rrad1at10n nor aging seems to have a
detectable effect on the minimum creep rate The fracture strain is
shown as a function of strain rate in Fig. 35, The fracture stfains of
the unlrradlated samples vary from about 30% in a tensile test to about
10% in a long-term creep test. The irradiated samples have lower frac-
ture strains that vary from 9 5% for the fastest tensile test to 2 to 3%

for creep tests.

 

 
 

 

 

 

Table 17. Creep-Rupture Tests on Heat 7320 at 650°C
: . Minimum
: . Rupture Reduction
Test Specimen Str?ss Iife Strein in Ares Creep
Number Number (psi) (br) (%) (%) Rate
(%/nr)
Unirradiated — Annealed 1 hr at 1177°C prior to test
7013 7276 55,000 4.7 28.7 29.6 0.190
7425 10329 50,000 11.5 21.2 28.3 0.125
7014 7277 47,000 49.9 21.8 31.3 0.019%
7016 7279 43,000 103.4 18.1 20.5 0.0069
7015 7278 40,000  384.6 16.3 17.6 0.0059
Th24 10252 35,000 1602.1 27.3 30.6 0.0056
7017 7280 30,000 1949.5 10.9 19.9 0.0017
Unirradiated Controls®
7885 9462 55,000 32.5 1.2 32.9 0.25
7991 9468 47,000 224.1 15.3 26.3 0.045
7886 9463 40,000 653.7 17.9 29.3 0.019
7887 9465 40,000 501.9 7.9 _1.4 0.015
7884 9460 32,400 2420.2 6.9 4.8 0.0006
Irradiated®
R-1151 9204 63,000 1.7 12.24 1.2
R-1016 9216 63,000 4.0 2.0 0.44
R-951 9217 55,000 12.3 2.7 0.20
R-955 9219 47,000 95.1 2.3 0.018
R-967 9218 40,000 329.9 3.2 0.0074
R-950 9220 32,400 2083.3 3.1 0.0058

 

Annealed 1 hr at 1177°C and exposed to & vessel of static barren

fuel salt for 7244 hr at 650°C.

bDiscontinued prior to failure.

Cannealed 1 hr at 1177°C.

Irradiated to & thermsl fluence of
5.1 x 10?9 neutrons/cm? over a period of 7244 hr at 650°C

dTest loaded so that strain on loadlng was included.
tests did not include this strain.

All other

¥
»

STRESS (1000 psi)

45

ORNL=—DWG 70— 7305

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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20 _ & UNIRRADIATED, ANNEALED 1hr HT7°C
® IRRADIATED, 5.1 * 1029 neutrons/em?
10
0
W 2 5 . 2 5 102 2 5 w0 2 s 40

RUPTURE TIME (hr}

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig. 33. Stress-Rupture Properties of Heat 7320 at 650°C.
70 ORNL-DWG TO-7306
\
0 Ll
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20 ' - & UNIRRADIATED, ANNEALED fhr AT #177°C | | 1]
10
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~_ MINIMUM CREEP RATE (%/hr) L
Fig. 34. Creep Properties at 650°C of Heat 7320.
 

 

46

ORNL- DWG 70~ 7307

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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STRAIN RATE (% /hr)

Fig. 35. Fracture Strains at 650°C of Heat 7320 in the Unirradiated
and Irradiated Conditions. Irradiated to a thermal fluence of 5.1 X 10%°
neutrons/cm® over a period of 7244 hr at 650°C.

The results of creep-rupture tests on heat 67-551 with 1.1% Ti are
given in Table 18. The stress-rupture properties of the control and
irradiated samples are compared in Fig. 36 for 650°C. The irradiated
samples generally fail in slightly shorter times than the unirradiated
samples. The minimum creep rate seems to be unaffected by irradiation,
although there are two data points that seem to be anomalous (Fig. 37).
The fracture strains of this alloy (Fig. 38) are higher in both the
unirradiated and irradiated conditions then those for heat 7320. The
fracture strains of the unirradiated samples varied over the range of
43 to 25% and those of the irradiated samples varied from 22.6 to 5.8%
over the range of strain rates studied.

One heat of modified Hastelloy N containing 0.49% Hf (heat 67-504)
was exposed to the cell environment for 17,033 hr at 650°C. The .thermal -
fluence was 2.5 X 10'° neutrons/em?. This same heat of material was

included previously in our surveillance program and was exposed to a
47

Table 18. Results of Creep-Rupture Tests on Heat 67-551 at 650°C

 

 

 

 

. Minimum
Test Specimen Stress —T_BEEEEEE“-T— R?ductlon Creep
7o . Life Strain in Area
Number Number (psi) (br) (%) (%) Rate
(%/nr)
Unirradiated, Annesled 1 hr at 1177°C before testing
7872 6889 55,000  14.2 29.5 33.2 0.32
7871 6884 47,000 3.8 33.6 26.0 0.071
Unirradiated Controls™
7882 9409 55,000 44.7  25.5 28.5 0.190
7880 9404 4'7 ,000 271.2 29.1 29.6 0.067
7883 9411 40,000 956.2 30.4 25.5 0.018
7881 9408 32,400 2034.9 25.2 29.9 0.0053
Irradiated’
R-1150 9150 63,000 0.9 20.8° 2.2
R-1036 9151 63,000 0.6 2.2 2.6
R-949 2162 55,000 50.5 5.8 0.079
R-~961 9165 47,000 117.1 8.3 0.058
R-948 9164 40, 000 "746.9 9.1 0.0079
R=-947 2163 32, 400 1485.6 12.5 0.0053

 

®pmnealed 1 hr at 1177°C prior to exposure to a vessel of static
barren fuel salt for 7244 hr at 650°C.

Pannesled 1 hr at 1177°C. Irradiated to a thermal fluence of
5.1 x 10?° neutrons/cm® over a period of 7244 hr at 650°C.

®Test loaded to include straln on loading. All other tests did not
include this strain.

thermal fluence of 5.3 X 10%° neutrons/cm while being at temperature in
the core for 9789 hr (ref. lO) In the group of samples presently being
discussed, there were twoquds of heat 67-504. Two heats of material
should have been_involved,fbut'postirradiation'chemical analysis revealed
that both rods were made ofrthe same material. This was not discovered,
until after many of the samples were tested, 80 we have several tests

under dupllcate conditions.

 

10H. E. MecCoy, An Evaluation of the Molten-Salt Reactor Experiment
Hastelloy N Surveillance Specimen — Third Group, ORNI-TM-2647 (1970).

 
ORNL-DWG T0-7308

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

ettt

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

w

Irra-

10
60 T
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N
a b\\
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50
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= N
e 40
8 T
£ \\\
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w 30
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o & UNIRRADIATED, ANNEALED
{hr AT HTT7°C
20
" 10
o
5 10° 5 w0 2 5 w00 2 0° = 5
RUPTURE TIME {hr)

Fig. 36. Stress-Rupture Properties at 650°C of Heat 67-551.
diated to a thermal fluence of 5.1 X 102?° neutrons/cm® over a period of
7244 hr.

0 ORNL~DWG TO— 7309
| st
60 / = 1
- o s
50 "/
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_ "
% v /./
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© UNIRRADIATED CONTROLS
20 © IRRADIATED -
A UNIRRADIATED, ANNEALED thr AT HTT*C
0
0 L
w3 s w? 2 5 0 2 5 W 2 5
INIMUM CREEP ATE (%/hr} ‘
Fig. 37. Creep-Rupture Properties at 650°C of Heat 67-551.

Irra-

diated to a thermal fluence of 5.1 X 10?° neutrons/cm? over a period of

7244 hr at 650°C.
 

»

49

ORNL- OWG 70— 7310

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

60
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UNIRRADIATED ¢
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STRAIN RATE (% /tr)

Fig. 38. Comparison of the Fracture Strains of Unirradiated and
Irradiated Heat 67-551 at 650°C. Irradiated to a thermal fluence of
5.1 X 10?° neutrons/cm?® over a period of 7244 hr at 650°C.

The results of the postirradiation tensile tests . on heat 67-504
are given in Table 19. Thé'postirradiation fracture strains are shown
in Fig. 39 as & function ofltemperature and strain rate.  The fracture
strain generally decreases with increasing temperature above about 400°C
and with decreasing strain réte;'rThe results from the present tests and
those reported préviously10 show_éqmé'Very striking-éhahges'in fracture

strain with varying historiesr(Fig.’40). The samples were all initially

~annealed 1 hr at 1I77°C.béfore.being:given the treatment indicated in

Fig.VAO. In the as-annealedzébnditiOn,ithe alloy has a fracture strain

of 70 to 80% up to about'600°C,_where the fracture strain drops precip-

‘jtiously. Aging for 9789 hr at 650°C reduced the fracture strain at
lower temperatures and increased it above about 700°C. However, the

- fracture strains are in'the;range of 40 to 50% over the entire range of

temperatures studied. Irradiation to a thermal fluence of 2.5 X 101°
neutrons/cm?* over a period of 17,033 hr in N» + 2 to 5% O, resulted in

 
50

Table 19. Postirradiation Tensile Properties of
' Hastelloy N (Heat 67-504)%

 

 

 

Test Stress, psi A True
Strain Elongation, % Reduction

Specimen Temper- Ultimate Fracture
Number  ature (E:zf 1) Tield Tensile Uniform  Total in(:;l)'ea. Strain

(°c) (%)
5111 25 0.05 50,300 112,200 45.9 48.1 34.63 43
5147 25 0.05 50,700 112,800 54.6 56.6 36.54 45
5110 200 0.05 41,000 96,400 43.0 44.8 38.34 48
5162 200 0.05 40,900 95,300 45.7 47.3 46.66 63
5109 400 0.05 36,700 90,800 49.8 52.0 36.15 45
5145 400 0.05 37,500 90,800 52.8 54.8 34.84 43
5108 400 0.002 36,700 91,800 52.2 54.7 45.71 61
5144 400 0.002 41,100 97,900 52.7 53.7 33.31 41
5107 500 0.05 40,600 92,700 45.7 47.2 32.23 39
5143 500 0.05 40,900 94,000 49.6 50.6 39.97 51
5106 500 0.002 41,600 90,500 44 .5 46.2 39.40 50
5142 500 0.002 33,800 178,600 41.1 42.6 36.00 45
5117 550 0.05 39,700 74,900 34.5 36.5 31.45 38
5153 550 0.05 34,700 77,800 39.7 43.5 37.94 48
5118 550 0.002 44,600 62,000 11.9 15.1 16.02 17
5154 550 0.002 33,200 59,900 20.5 21.5 20.38 23
5096 600 0.05 31,900 74,500 34.7 36.9 26.02 30
5164 600 0.05 32,900 70,800 32.9 34.0 35.69 44,
5120 600 0.002 36,400 49,600 8.9 10.9 20.38 23
5156 600 0.002 30,900 49,500 12.6 15.7 23.35 27
5128 650 0.05 39,900 55,500 - 24.0 25.8 22.06 25
5121 650 0.05 38,800 59,600 14.6 16.5 19.15 21
5155 650 0.05 30,200 46,200 12.8 15.8 10.17 11
5122 650 0.002 52,200 54,500 2.2 6.2 3.51 4
5158 650 0.002 31,700 44,800 13.2 - 14.3 11.54 2
5105 650 0.0002 32,200 40,800 7.4 9.8 13.52 15
5141 650 0.0002 30,700 38,800 - 7.8 8.8 6.65 7
5140 650 -0.000027 30,200 37,100 v b 6.5 5.1 5
5104 650 -0.000027 30,700 37,200 4.8 6.5 5.4 6
5098 760 0.05 27,000 38,600 9.0 23.7 18.94 21
5123 760 0.05 33,300 40,100 6.7 8.7 10.17 11
5159 760 0.05 28,800 38,000 9.6 13.5 13.19 14
5124 760 0.002 33,300 33,300 1.2 2.7 4.63 5
5160 760 0.002 28,200 34,600 4.2 5.6 6.00 6
5102 850 0.05 27,300 29,700 2.9 4.3 2.40 2
5138 850 0.05 27,500 30,900 3.2 4.5 7.10 7
5103 850 0.002 21,100 - 21,100 1.0 1.9 0.64 1
5139 850 0.002 23,700 23,700 1.0 2.5 0.48 0

 

%pnnealed 1 hr at 1177°C. Irradiated to & thermsl fluence of 0.25 x 1020
neutrons/cm® over a period of 17,033 hr at 650°C.
 

0

"

»

60

50

40

30

FRACTURE STRAIN (%}

20

10

o

Fig. 39. Fracture Strains of Heat 67-504 After Irradiation to a
Thermal Fluence of 2.5 X 101° neutrons/em® for 17,033 hr at 650°C.

Fig. 40. Comparison of the Fracture Strains of Heat 67-504 in

51

ORNL-DWG TO-T731t

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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STRAIN RATE
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TEST TEMPERATURE (°C)

ORNL=-DWG 70-73(2

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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20 IRRADIATED TO 5.3 x402% neutrons/em® X
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200 . 300 - 400 - 50O 600 700 800 900

_TEST TEMPERATURE (°C)

900

Various Conditions When Tested at a Strain Rate of 0.05 min~!.

 
 

52

fracture strains of about 50% up to 500°C, above which the fracture
strain dropped progressively with increasing temperature. Irradiation
to a thermal fluence of 5.3 X 102C neutrons/cm® at 650°C in fuel salt
over a period of 9789 hr resulted in a low fracture strain at 25°C (which
may be anomalous), fracture strains of about 45% up to 500°C, and
decreasing fracture strains with increasing test temperature. The most
striking feature is that above 550°C the fracture strains are lower for
material irradiated in N, + 2 to 5% Oy for 17,033 hr to a fluence of
2.5 X 101° neutrons/em? than for those irradiated in fuel salt for
9789 hr to a fluence of 5.3 X 10%C neutrons/cm®.
Some values of the tensile properties at 25 and 650°C are given in
Table 20. The yield stress at both test temperatures was increased by
aging and by irradiation. The ultimate tefisile stress at 25°C was
increased by aging and by irradiation; at 650°C it was increased by
aging,ybut decreased. by irradiation. We have already discussed the

changes in the fracture strain.

Table 20. Comparison of the Tensile Properties of Heat 67-504
After Various Treatments®

 

Ultimate Tensile
Stress, psi

et 25°C at 650°C

Fracture Strain, %
at 25°C at 650°C

Yield Stress, psi
at 25°C at 650°C

Heat Treatment

 

 

Annealed 1 hr st 1177°C 37,400 25,600 105,000 83,000 73.2 65.9

Annealed at 1177°C and 67,500 43,300 133,000 94,300 52.3 50.9
aged 9789 hr at 650°C

Annealed at 1177°C, irra- 102,000 32,900 119,300 70,000 27.0 30.7

diated toP a thermal
fluence of 5.3 x 102°
neutrons/cm? over
9789 hr (salt
enviromment)

Annealed at 1177°C, 50,300 38,800 112,000 = 59,600 48.1 16.5
irradiated to a ther-

mal fluence of

2.5 % 101? neutrons/cm?

over 17,033 hr (N, + 2

to 5% 0, enviromment)

 

®Tested at a strain rate of 0.05 min-l.

bData. from: H. E. McCoy, Jr., An Evaluation of the Molten-Salt Reactor Experiment

Hastelloy N Surveillance Specimens — Third Group, ORNL-TM-2647 (1970).
L}

53

The results of Creep-rupture tests on heat 67-504 are summarized
in Table 21. These results and those obtained previously'! were used
to prepare Figs. 41, 42, and 43. The stress-rupture propertiés in
Fig. 41 show that the rupture'lifé was not changed appreciably'by aging,
at least for 9789 hr at 650°C. The stress-rupture life was reduced at
least two orders of megnitude by the lower fluence and only a factor of
2 by the higher fluence. The mi ndmum creep rates in Fig. 42 show that
this property was not changed as much as the rupture life. Aging for
9789 hr increased the'minimfim creep, but samples irradiated over the
same period had a lower creep rate than the aged samples. The samples
irradiated to the lower fluénce had a creep rate an order of magnitude

higher than that of the as-annealed material. The fracture strains are

 

11H. E. McCoy, An Evaluation of the Molten-Salt Reactor Experiment
Hastelloy N Surveillance Specimen — Third Group, ORNL-TM-2647 (1970).

Table 21. Creep-Rupture Tests on Heat 67-504 at 650°C

 

 

 

 

 

_ : Minimum
Test Specimen Stress LifeRupturgtrain Creep
Number Number (psi) () (%) Rate
| (%/hr)
Irradiated at 650°C to a thermal fluence of 2.5 X 1019 neutrons/ecm?
R-952 ' 5112 : 55,000 0.6 - 8.6 9.88
R-953 5148 55,000 1.0 10.6 6.6
R-959 5113 . 47,000 7.6 12.1 0.83
R-965 5149 47,000 - 3.3 6.2 0.52
R-971 5114 40,000 1.5 10.9 0.42
R-1017 5150 - . 40,000 33.8 4.1 0.050
R-968 5115 - - 32,400 235.5 3.8 0.0071
R-1033 5137 . 32,400 - 268.7 4.2 0.0066
R-1019 - 5116 - 27,000 1175.4 2.7 0.0016
S . Annealed 1 hr at 1177°C
7432 6247 70,000 13 329 3.9
- 7431 6245 63,000 - 17.7 31.2  O0.16
6255 491 55,000 127.9  27.2  0.029
- 6254 . 4936 ¢ 47,000 0 425.5 28.7 0.0068

 

6253 4933 -~ 40,000  876.9  16.3 0.0030

 
 

54

ORNL-DWG 707315

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

70 o 1
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60 S
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™
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o Ll © ANNEALED 1 wr AT n77%c
O ANNEALED AND AGED 9789 hr AT 650°C
® THERMAL FLUENCE OF 5.3 x 1020 neutrons fom?
& THERMAL FLUENCE OF 2.5 x10'? neutrons /cm2
o LLIIII | EEEEL P 1t
5 w2 s « 2 s 102 2 5 W 2 5

RUPTURE TIME {(hr)

Fig. 41. Stresé-Rupture Properties at 650°C of Heat 67-504.

ORNL~DWG 70-7313

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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© ANNEALED 1 br AT H77°C
10 O ANNEALED AND AGED 9789 hr AT 650°C )
® THERMAL FLUENCE OF 5.3 x 1020 neutrons/cm?
: ® THERMAL FLUENCE OF 2.5 x 10" neutrons /em?
o R
0? 2 s w2 2. s ' 2 s 10 2 5 10

MINIMUM CREEP RATE (% /hr)

Fig. 42. Creep-Rupture Properties at 650°C of Heat 67-504.

n
 

Wy

55

ORNL-DWG T70-7314R

 

50 1711
UNIRRADIATED

 

 

40

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

i SOLID POINTS - TENSILE TESTS
e OPEN POINTS - CREEP TESTS
30 = - : © ANNEALED 4 hr AT #77°C
1" g & ANNEALED AND AGED
9789 hr AT 650°C
a ANNEALED,THERMAL FLUENCE
OF 5.3%10° neutrons/em? 11
¢ ANNEALED, THERMAL FLUENCE
T OF 2.5x10'® neutrons/cm?2

L 1

 

 

i
1
1°
| ]

 

20

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

{0

IRRADIATED

FRACTURE STRAIN (%}

 

2 5 0% =2 5 10 2 5 0% 2 5 10
STRAIN RATE (%)

10° 2 5 0% 2 5 107

Fig. 43. Compa.:r'lson of" the Fra.cture Strains of Unirradiated and
Irradiated Heat 67-504 at 650°C. [Data for samples aged for 9789 hr at
650°C and for those irradiated to 2.5 x 10'° neutrons/em® from:

H. E. McCoy, An Evaluation of the Molten-Salt Reactor Experiment
Hastelloy N Surveillance Specimen'—‘Thlrd Group, ORNL-TM-2647 (1970).]

Vsshown as a function of - straln rate in Fig. 43. Aging increased the frac-

ture strain of the unlrradlated materlal After 1rrad1at10n the materlal

71rrad1ated to the lower fluence had the lowest fracture straln

‘Several samples-of heat 67-504 were subjected to tests that were
interrupted. The results of'theSe tests are summarized in Table 22. If

strained continuously at 650°C at a strain rate of 0.05 min~!, the

 
56

Table 22. Results of Interrupted Tensile
Tests on Heat 67-504

 

 

 

 

a Test Fracture Strain, %
Type of Test Tem%sg?ture Unirradiated Irradiated
A 650 66.9 16.5
B 650 100.4 |
C 650 109.0
D 650 104.0 .
E 650 95.4 25.8
A 760 35.5 8.7
F 760 9.0 23.7
a 1

A — Run uninterrupted at a strain rate of 0.05 min™-.

B — Strained 5%, held at temperature for 2 min, cycle
repeated to failure.

C — Strained 5%, held at temperature for 10 min, cycle
repeated to failure.

D — Strained 5%, held at temperature for 30 min, cycle
repeated to failure.

E — Strained 5%, held at temperature for 60 min, cycle
repeated to failure.

F — Strained 3%, held at temperature for 60 min, cycle
repeated to failure.

material had a fracture strain of 66.9% if unirradiated and 16.5% if
irradiated. If the material was strained 5% at 650°C then annealed 1 hr
at 650°C and this pattern continued until failure, the unirradiated sam-
ple failed with 95.4% strain and the irradiated sample with 25.8% strain.
A sequence of tests was performed on unirradiated samples at 650°C in
which the annealing time was varied from 2 to 60 min; the results show
that this variable had little effect on the fracture strain. At a test
temperature of 760°C the fracture strain was 35.5% for an unirradiated
sample and 8.7% for an irradiated material. Samples were also tested by'
straining 3%, annealing 1 hr at 760°C, straining 3%, and repeating this
sequence until failure. The fracture strains were 96.0 and 23.7% for the

unirradiated and irradiated samples, respectively. These tests have
 

 

57

demonstrated that there is not, for each test temperature, & unique
strain at which failure occurs; rather, the fracture strain depends upon

the loading history.

Metallographic Examination of Test Samples

The tensile samples that were tested at 25°C at a strain rate of
0.05 min~! and at 650°C at a strain rate of 0.002 min~! were examined.
An irradiated sample of heat 5085 from the core that was tested at 25°C
is shown in Fig. 44. The fracture is primarily intergranular. There is
a high frequency of edge cracking that extends to a depth of about 4 mils.
The microstructure in the 4-mil region near the surface etches differently.
At a 650°C test temperature (Fig. 45)'the fracture is intergranular with
no evidence of plastic deformation of the individual grains except adja-
cent to the fracture. This sample aiSo had the modified structure and
edge cracks that extend 10 mils into the sample. The control samples of
heat 5085 that were annealed for 22,533 hr at 650°C in static barren fuel
salt are shown in Figs. 46 and 47. The sample shown in Fig. 46 wasr
tested at 25°C. The fracture is predominantly intergranular, but there
are no edge cracks. There is a very thin layer near the surface where
‘the structure is modified slightly. The microstructure is generally
characterized by large M¢C-type carbides, many of which cracked during
deformation, and a network of almost continuous carbides along the grain
and twin boundaries. When tested at 650°C (Fig. 47) the fracture is
intergranular. There is some edge cracking, but not as frequently as
noted in Fig. 45 for the irradiated sample. Another factor that should
be considered in comparlng Flgs 45 and 47 is that the irradiated sample
in Fig. 45 failed after stralnlng only 5% (Table 4, p. 18), whereas the
" unirradiated sample in Fig. 47 strained 24% (Table 3, p. 17) ‘before frac-
turing. The higher straln should have increased the number and depth of
edge cracks. The observed opp051te trend indicates some influence of
the exposure to the reactor env1ronment on the fracture characteristlcs
- near the surface. _

'Heat 5065 was also exposed to the core for 22 533 hr at 650 C The
microstructure of a sample tested at 25°C is shown in Fig. 48. The

 
     

R-50900 [ Y NN R -50902

 

 

  

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(c)

o
o L o :
59! : .\fi-v. ot 4 .

Fig. 44. Photomicrographs of a Hastelloy N (Heat 5085) Sample Tested at 25°C After Being Exposed
to the MSRE Core for 22,533 hr at 650°C and Irradiated to a Thermal Fluence of 1.5 X 102! neutrons/cm?
(a) Fracture, etched. 100x. (b) Edge, as polished. 100x. (c) Edge, as polished. 500x. (d) Edge,
etched. 100X. Etchant: glyceria regia. Reduced 27%. ‘

*»

('w I & t ’ » . r “ (

8¢

 
 

 

   

 

 

 

 

 

 

-

 

Fig. 45. Photomicrographs of a Hastelloy N (Heat 5085) Sample Tested at 650°C After Be1n§ Exposed

to the Core of the MSRE for 22,533 hr at 650°C and Irradiated to a Thermal Fluence of 1.5 X 1041
neutrons/em?. (a) Fracture, etched. 100X. (b) Edge, as polished. 100x. (c) Edge, as polished.
(d) Edge, etched. 100X. Etchant: glyceria regia. Reduced 27%.

500x.

6S

 

 

 
 

Y-98217

A Y-98218

TR

 

Fig. 46. Typical Photomicrographs of a Hastelloy N (Heat 5085)
Sample Tested at 25°C After Being Exposed to Static Barren Fuel Salt for
22,533 hr at 650°C. (a) Fracture, etched. 100X. (b) Edge near fracture,
etched. 100X. (c) Representative unstressed structure, etched. 500x.
Etchant: glyceria regia. Reduced 24.5%.
 

 

 

 

 

 

 

 

Fig. 47. Typical'Phé#Qmicrdgraphs.of'a Hastelloy N (Héafi 5085)'
Sample Tested at 650°C (Strain Rate 0.002 min"1) After Being Exposed to

Static Barren Fuel Salt for 22,533 hr at 650°C. (a) Fracture. . 100x.

(b) Edge near fracture. 100X. (c) Representative unstressed structure.
500X. Etchant: glyceria regia. Reduced 25%. . B

 
 

c9

 

Fig. 48. Photomicrographs of a Hastelloy N (Heat 5065) Sample Tested at 25°C After Being Exposed
to the MSRE Core for 22,533 hr at 650°C and Irradiated to a Thermal Fluence of 1.5 X 10?! neutrons/cm?.
(a) Fracture, etched. 100X. (b) Edge, as polished. 100x. (c) Edge, as polished. 500x. (d) Edge,
etched. 100X. Etchant: glyceria regia. Reduced 26%.

 
 

63

fracture is mixed transgranular and intergranular. The same structural
modification and edge'cracking that were noted in heat 5085 (Fig. 44)
are also present in heat 5065. A sample of heat 5065 tested at 650°C is
shown in Fig. 49. The fracture is intergranular'with little evidence
of deformation of the adjacent grains. The microstructure at the edge
is modified to a depth of about 4 mils, and cracks extend to a depth of
about 10 mils. A control sample tested at 25°C is shown in Fig. 50.

The fracture is mixed transgranular and intergranular, and the grains
are elongated in the direction of stressing. There is no edge cracking.
There are copious amounts of carbides, both of the large primary car-
bides and the finer carbides formed during the long annealing treatment
at 650°C. The microstructure of a control sample of heat 5065 tested at
650°C (Fig. 51) shows an intergranular fracture and some edge cracking.
Again the frequency of cracking is less than that noted for the irra-
diated material shown in Fig. 49.

Heat 7320 was exposed to the MSRE core for 7244 hr at 650°C. Typi-
cal micrographs of a sample tested at 25°C are shown in Fig. 52. The
fracture is predominantly transgranular and the flow lines and the elon-
gated grains attest to therdeformation of the matrix. There are some
edge cracks to a depth of about 3 mils, but the microstructure is not
modified at the sample surface. Microstructures of an irradiated sample
tested at 650°C are shown in'Fig. 53. The fracture is intergranular.
There are edge cracks that extend to a depth of about 10 mils, but the
microstructure is not modified near the surface. A control sample that
was tested at 25°C is shOthifi'Fig 54, The fracture is transgranular
with the grains being elongated in the direction of streSS1ng There is
no edge cracking. The unstressed microstructure gives a hint of the
'very fine matrix precipltatlon thatris ‘present in this alloy. -Photo-
“micrographs of a control sample tested at 650°C are shown-in'Fig. 55‘
The fracture is 1ntergranular, and there are no edge cracks such as were
.noted in the irradiated sample (Fig 53).

Heat 67-551 was exposed 1n the MSRE core for 7244 hr at 650°C. The
fracture at 25°C (Fig. 56) is mixed transgranular and 1ntergranular and
grains have deformed extensxvely The microstructure near the surface
is not altered, but there are edge cracks that extend to a depth of about

 
3 R-50883

B R-50885

R-50387

 

‘Fig. 49. Photomicrographs of & Hastelloy N (Heat 5065) Sample Tested at 650° After Being Exposed
to the Core of the MSRE for 22,533 hr at 650°C and Irradiated to a Thermal Fluence of 1.5 X 104!
neutrons/em?. (a) Fracture, etched. 100X. (b) Edge, as polished. 100x. (c) Edge, as polished. 500x.
(d) Edge, etched. 100x. Etchant: glyceria regia. Reduced 26.5%. |
 

&1

 

 

 

212

o

 

 

 

 

"

 

 

o

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

   

 

»

Fig. 50. Typical Photomicrographs of a Hastelloy N (Heat 5065)
Sample Tested at 25°C After Being Exposed to Static Barren Fuel Salt for
. 22.533 hr at 650°C. (a) Fracture. 100X. (b) Edge near fracture. 100x.
L (cj Representative unstressed structure. 500X. Etchant: glyceria

Q regia. Reduced 22%.

 

 

 
Y-98214

   

 

b s e

   

—=v-9521s
R P s

 

 

 

 

 

Fig. 51. Typical Photomicrographs of a Hastelloy N (Heat 5065)
Sample 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) Fracture. 100x.

(b) Edge near fracture. 100X. (c¢) Representative unstressed structure.
500X. Reduced 21.5%. | . -

 

 
&

[T

67

R-50961

  
  

 

R-50963

 

 

 

| R-50965

 

 

 

(c) -

 

Fig. 52. Photomicrographs of & Modified Hastelloy N (Heat 7320)

Sample Tested at 25°C After ‘Being Exposed to the MSRE Core for 7244 hr

at 650°C and Irradiated to & Fluence of 5.1 X 10%° neutrons/cm®.
(a) Fracture, etched. 100x.- (b) Edge, as polished. 100X.. (c) Edge,
etched. 100X. Etchant: glyceria regia. Reduced 22%.

 
R-50916

    

 

 

 

 

Fig. 53. Photomicrographs of a Modified Hastelloy N (Heat 7320)
Sample Tested at 650°C (Strain Rate, 0.002 min~!) After Being Exposed to
the MSRE Core for 7244 hr at 650°C and Irradiated to a Fluence of
5.1 X 10%° neutrons/cm*. (a) Fracture, etched. 100X. (b) Edge, as

polished. 100x. (e) Edge, etched. 100X. Etchant: glyceria regia.
Reduced 22%. | o |
 

 

69

Y-98205

 

  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

#

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

: ' Fig. 54. Photomicrographs of a Modified Hastelloy N (Heat 7320)
Sa.mple Tested at 25°C After iBe:mg Exposed to Static Barren Fuel Salt for

. 7244 hr at 650°C. (a) Fracture. 100x. (b) Edge. 100x. {c) Typical

: unstressed microstructure. 500X. Etchant: glyceria regia. Reduced 22.5%.

 
 

 

 

 

N Y-98208

 

 

 

   

(c) E";;::;u: & .

Fig. 55. Photomlcrographs of a Modified Hastelloy N (Heat 7320)
Sample Tested at 650°C (Strain Rate, 0.002 min~1) After Being Exposed to
Static Barren Fuel Salt for 7244 hr at 650°C. (a) Fracture. 100x.

(b) Edge near fracture. 100X. (c) Typical wstressed microstructure.
500%. Etchant: gyleeria regia. Reduced 23%.
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

R-50955

    
   

()

R-50957

- © R o e -
. Fig. 56. Photomicrographs of a Modified Hastelldy,N_(HEat 67-551)

Sample Tested at 25°C After Being Exposed to the MSRE Core for 7244 hr

at 650°C and Irradiated to a Fluence of 5.1 X 10%° neutrons/em?®.
(a) Fracture, etched. 100X. (b) Edge, as polished. 100X. (c) Edge,
etched. 100X. Etchant: glyceria regia. Reduced 21%.

 
 

 

72

2 mils. At a test temperature of 650°C the fracture is mixed intergran-
ular and transgranular (Fig. 57). The microstructure near the edge is
not modified, but edge cracks extend to a depth of about.8-mils. A con-
trol sample that was tested at 25°C is shown in Fig. 58. The fracture
is largely transgranular, and there are no edge cracks nor structural
modifications. The high magnificétion view shows the fine carbide pre-
cipitates that form during the long thermal anneal. At 650°C the frac-
ture is intergranular with numerous intergranular cracks scattered
throughout the sample (Fig. 59). A few cracks are near the surface, but
these are to be expected in light of the 31.5% strain (Table 14, p. 40)
that occurred in this sample before failure.

Heat 67-504 was exposed to the MSRE cell environment of N, + 2 to
5% 0, for 17,033 hr at 650°C and had an oxide film of 1 to 2 mils: when
tested at 25°C (Fig. 60) the fracture was mixed transgranular and inter-
granular, although most of the deformation occurred within the grains.
There was no edge cracking. When tested at 650°C (Fig. 61) the fracture
was primarily intergranular. There were edge cracks to a depth of
15 mils. Since there were no controls to compare with these samples, it
is difficult to say whether the frequency and depth of cracking are
greater than would be expected.

DISCUSSION OF RESULTS

The mechanical property changes of the standard HastellOy N in both
the irradiated'and unirradiated conditions have fdllowed very. regular
trends throughout its exposure in the MSRE and in the control facility.
Heats 5065 and 5085 have been used throughout the surveillance program.
Exposure to the static barren fuel salt up to 15,289 hr brought about a
gradual reduction of the tensile fracture strains in both heats; further
exposure up to 22,533 hr caused a slight improvement ovef the strains
observed after 15,289 hr (Figs. 15 and 17, pp. 24 and 25). These changes
were small compared with those observed after irradiation, afid we attrib-
ute them to the precipitation of grain-boundary carbides. These changes
in tensile properties occurred without detectable change in the creep
strength at 650°C (Figs. 19 and 25, pp. 30 and 34).
 

o

73

R-50924

 

 

 

 

 

R-50928

   

©)
Fig. 57. Photomicrographs of a Modified Hastelloy N (Heat 67-551)
Sample Tested at 650°C (Strain Rate, 0.002 min~!) After Being Exposed to

the MSRE Core for 7244 hr at 650°C and Irradiated to a Fluence of

5.1 X 10?° neutrons/em?. (a) Fracture, etched. 100x. (b) Edge, as

‘polished. 100X. (c) Edge, etched. 100X. Etchant: glyceria regia.

Reduced 20%.

 
i
i
1
1

 

 

 

 

i Y-98199

     

 

 

 

 

Fig. 58. Photomicrographs of a Modified Hastelloy N (Heat 67~551)

Semple Tested at 25°C After Being Exposed to Static Barren Fuel Salt for

7244 hr at 650°C. (a) Fracture. 100X. (b) Edge. 100x. (e) Typical
unst;essed microstructure. 500X. Etchant: glyceria regia. Reduced
20.5 . 7 ) . . . . .

o
75

: Q _ : 9202

 

 

 

 

 

 

 

 

 

 

 

 

Y-98203

   
   

b

()

Y-98204

 

(c)

. - . Fig. 59. Photomicrographs of a Modified Hastelloy N (Heat 67-551)
Sample Tested at 650°C (Strain Rate, 0.002 min-!) After Being Exposed to
Static Barren Fuel Salt for 7244 hr at 650°C. (a) Fracture. 100X.

‘ (b) Edge near fracture. 100x. (c) Typical unstressed microstructure.

u 500%X. Etchant: glyceria regia. Reduced -20.5%.

 

 
     

R-50470

  

R-50472

  

PO

  

e

(b)

R~50473

   

)

Fig. 60. Photomicrographs of a Modified Hastelloy N (Heat 67-504) Sample Tested at 25°C Following
‘Exposure to the MSRE Cell Environment for 17 »033 hr and Irradiation to a Fluence of 2.5 x 101°
neutrons/cm®. (a) Fracture, etched. 100X. (b) Edge, as polished. 100X. (c) Edge, etched. 100X.
(d4) Edge, as polished. 500X. Etchant: glyceria regia. Reduced 27%. -

(‘ ¥ " ‘ h ¥ o - C

94
c - - C

R-50875 ' ) ' R-50877

 

 

 

 

(D

    

Fig. 61. Photomicrographs of a Modified Hastelloy N (Heat 67-504) Sample Tested at 650°C (Strain
Rate, 0.002 min~!) Following Exposure to the MSRE Cell Environment for 17,033 hr and Irradiated to a
Fluence of 2.5 X 101° neutrons/cm®. (a) Fracture, etched. 100x. (b) Edge, as polished. 100x.

(c) Edge, etched. 100x. (&) Edge, as polished. 500X. Etchant: glyceria regia. Reduced 25%.

Ll

 
 

78

Irradiation of both heats caused a general decrease of the fracture
strain with increasing thermal fluence at test temperatures of 25 to
850°C. The fracture strains at lowltemperatures are lower for heat 5065.
We attribute the reduction in fracture strain at low temperatures to pre-
cipitation of carbides and demonstrated previously that it can be'
recovered by annealing to coarsen the carbide.l? We attribute the embrit-
tlement at high temperatures to the helium formed by thermal transmute-
tion of 1B to form “He and "Li. This embrittlement cannot be recovered
by annealing. The presence of helium in the material influences the
creep properties at 650°C by reducing the rupture life and the straih at
fracture; the creep rate is unaffected (Flgs 19 through 26, pp. 30
through 34). The effects on the rupture life are greater at hlgher stress
levels and decrease to 1n51gn1flcant below about 10,000 psi. (The maxi-
mum operating stress!® in the MSRE is 6000 psi.) The changes in the
fracture strain are of utmost importance and are slightly dependent fipon
the thermal fluence (Figs. 21 and 26). Figure 23 shows more clearly the
sensitivity of heat 5085 to the presefice of helium and how this sensitiv-
ity at 650°C depends upon the strain rate. A thermal fluénge_of only
1.3 X 10'° neutrons/cm® produced sbout 1 ppm (atomic) He and reduced the
fracture strain from 30 to 2% when the material was crept at a rate of
O.l%/hr. The strain rate results in the lowest fracture strains with
slower rates resulting in slightly higher values. Figure 28, pl 36,
shows that this same general behavior holds for heat 5065 although the
amounts of helium produced are actually lower. The MSRE vessel has
received a thermal fluence of about 4 X 10'? neutrons/em®, and Figs. 23
and 28 indicate that the fracture strains will not drop much more with
continued operation. The control rod thimbles have received a thermal
fluence of about 2 X 10°! neutrons/cm?. They should be extremely brittle,
but are normally subjected to only a slight compressive stress.

The greater strain rate senSitivity of the irradiated Hastelloy N
at 650°C is illustrated in Figs. 22 and 27 for heats 5085 and 5065,

 

124, BE. McCoy, An Evaluation of the Molten-Salt Reactor Experiment
Hastelloy N Surveillance Specimen — Third Group, ORNL~TM-2647 (1970)

13R B. Briggs, ORNL, prlvate communication.
79

- respectively. For heat 5085 the fracture strain of the unirradiated
material varies from 32 to 21%, a decrease of 34%, and the irradiated
‘material varies from 9.3 to 0.5%, a decrease of 95%. For heat 5065 the
corresponding reductions are-35%'for the unirradiated material and 97%
for the irradiated material.

The question of primary concern is whether the Hastelloy N has
corroded more rapidly during the previous surveillance period than
previously. We have at least three parameters or phenomena that can be
observed or measured that likely are dependent upon the corrosion rate.
The first measure is the chemical change of the salt during reactor
operation. Since the chromium observed in the salt can hardly come
from anywhere except the Hastelloy N, we cannot question this parameter
as a measure of the corrosion rate. Second, we can observe changes in
the microstructure. Figures 2, 3, and 4, pp. 9, 10, and 12, show typi-
cal changes, but these observations are difficult to interpret without
additional information. Third, the mechanical properties themselves may
be altered or the crack patternS'that develop may be useful indications
of the effects of corrosion on the mechanical properties. The property
changes themselves are not very useful in this case because we have only
matched sets of control samples and surveillance samples. The control
samples are exposed to a static barren fuel salt that does not match the
corrosiveness of the MSRE fuel circuit. The surveillance samples are
exposed to & neutron fiueficeiand to a more corrosive salt circuit. Thé
irradiation effects predominate, s0 the effects of the added corrosion
are not large enough to seé;by Obéérving the mechanical property changes.
Thus there are several measures of corrosion in the MSRE, but they are
‘all subject to interprefatign.'Because-of the importance of this sub-
Ject, let us review thé.pérfiinént'facts and observations.

1. Before the last,periodwof.dperatibn, the salt was removed from
the MSRE and processed to remove the uranium and replace it with 233U,
 This processing left the salt fairly oxidizing, and it was necessary to
adjust theboxidationpotentiai'by adding beryllium“metal._ Thé chromium
contént:ofithe salt\was'bnlj 40 ppm when placed in the reactor and rose
over a few weeks to a level Qf'abofit 100 ppm. This apparénfhcorrosion

codld be accounted for by uniformly removing the chromium from the

 
 

 

80

Hastelloy N to a depth of 0.3 mil. The total increase of chromium in
the salt during the entire MSRE operation would require uniform chromium
extraction to a depth of 0.4 mil. However, the diffusion rate along the
grain boundaries can be about 10° the rate through the bulk grains,l*
and it is more likely that chromium be removed to greater depths along
the grain boundaries.

2. Electron microprobe examination of the sample in Fig. 5, p. 13,
revealed a definite chromium gradient near the surface and a surface
chromium concentration of nearly zero. A diffusion coefficient of
4 x 10"14 cm?/sec for chromium was computed based on the shape of the
profile which is in excellent agreement with the value of 2 X 10714 em?/sec
measured by Grimes et al.l’ (Fig. 62).

3. Complete loss of chromium from Hastelloy N does not result in
the type of microstructural modification that was observed, not does it
cause embrittlement.1© VWe made laboratory melts containing from O to
9% Cr with the standard Hastelloy N base composition. The. amount of car-
bide precipitate increases and the grain size decreases as the chromium con-
tent increases. The alloy is weaker in creep at 650°C without chromium,
but the fracture strains are not dependent upon the chromium concentration.

4. Similar microprobe scans have been made on control samples where
the modified structure is also present. The resolution is much better
on these samples where the background radiation is not present. The
gradients measured in a sample that had been in the control facility for
15,289 hr at 650°C are shown in Fig. 63. The chromium is.depleted and
the iron is enriched near the surface. These chemical modifications
extend only a short diétance into the material. These changes wotld not

be expected to produce the microstructural changes.

 

14W. R. Upthegrove and M. J._Sinnot, "Grain-Boundary Self-Diffusion
of Nickel," Trans. Am. Soc. Metals 50, 1031 (1958).

1°W. R. Grimes, G. M. Watson, J. H. DeVan, and R. B. Evans, "Radio-
Tracer Techniques in the Study of Corrosion by Molten Fluorides,"
pp. 559-574 in Conference on the Use of Radioisotopes in the Physical
Sciences and Industry, September 6-17, 1960, Proceedings, Vol. III,
‘International Atomic Energy Agency, Vienna, 1962.

 

16H. E. McCoy, Influence of Various Alloying Additions on the
Strength of Nickel-Base Alloys (report in preparation).

 
ORNL-DWG 70-4933

 

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CHRCMIUM CONCENTRATION (%)

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DISTANCE FROM SURFACE (mils)

- Fig. 62. Chromium Gradient in Hastelloy N Sample Exposed to the
MSRE Core for 22,533 hr. . N o

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SURFACE LAYER : MATRIX
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.
CONCENTRATION (wt %)
o
\'

 

| ¢ g g —2 —‘—.Cr_

' d
»
\.
o
4

 

 

 

 

 

 

 

 

 

 

 

o 5 40 15 20 25 30
DISTANGE FROM SURFAGE (u)

Fig. 63. Concentration Gradients in Hastelloy N (Heat 5085) Exposed
to Static Barren Fuel Salt in the MSRE Control Vessel for 15, 289 hr at
650°C

,5. The Hastelloy N straps that held the survellla.nce assembly
together for 22, 533 hr ha.d 1ntergranular surface cracks to a depth of
about 3 mils (ref. 17) (see Fig.-2, p- 9). Straps that had been in the

| reactor for 7244 hr ha.d cracks to a depth of 1. 5 mlls. | The straps are
© 0.020 in. thlck and should not be stressed dur1ng reactor 0perat10n, but

 

| o | 17W. H. Cook, MSR Program Semiann; “‘Progr.‘Re'pt., Aug. 31, 1969,
g ORNI~4449, pp. 165-168.

 
 

82

they were handled considerably during disassembly. Thus one cannot con-
clude whether the cracks were formed during service or in handling after-
wards. The microstructure was not modified near the surface of thése
stréps, |

6. A surface microstructural modification has been observed spas-
modically during this entire pregram. The photomicrographs of the
various lots of heat 5085 that were examined are presented in Fig. 64.
The modified microstructure has been present to some degree in all sam-
ples including the irradiated and the control samples. It is difficult
to say that the modification has grown any worse.

7. We have been able to produce a microstructural modification
quite similar to that shown in Fig. 65 by sinterless grinding Hastelloy N
and then annealing it for long periods of time in argon at 650°C
(Fig. 65). Thus, the layer may well be associated with cold working and
the manner in which this alters carbide deposition near the surface. |

8. There is a definite tendency for the samples removed from the
MSRE to show progressively more edge cracking during postirradiation
testing as the time of exposure increases. As shown in Fig. 66 for sam-
ples stressed at 25°C, the frequency of edge cracks 'increases with expo-
sure, but the maximum depth is constant at about 4 mils. The control
samples do not show much edge cracking.

9. Microprobe scans on irradiated samples failed to reveal any
fission products in the sample. However, the inability to analyze solely

the material in a grain boundary makes it impossible to conclude that
| no fission products are entering the metal. We plan to dissolve progres-
sive layers from the sample surfaces for chemical analysis to answer this
question. !

The observations on the alloys of modified Hastelloy N are quite
important. Heats 7320 (0.5% Ti) and 67-551 (1.1% Ti) were exposed to
the MSRE core for 7244 hr at 650°C. The properties of heat 7320 were
not as good as we have observed for other 0.5% Ti-modified alloys irra-
diated at 650°C. However, the properties after irradiation in the MSRE
are equivalent to those measured after irradiation in the ORR in & helium

environment. The properties of heat 67-551 are outstanding. Both of
 

   

 

1.3 x 10® nevtrong/em®

 
    

A¥ ‘"Q‘x‘:; Sy
o A

    

 

 

 

 

 

 

 

1.5x lw nevtrons/cm’

Tig. 64. Photomicrographs of Unstressed Hastelloy N (Heat 5085)
Control and Irradiated Samples.

R Y -09693

-y

 

 

 

 

 

 

 

 

Fig. 65. Photomicrograph of Hastelloy N (Heat 5085) Rod That Was
Amnealed 1 hr at 1177°C, Sinterless Ground 5.2 mils, and Annealed for
4370 hr at 650°C in Argon. 500X. Etchant: glyceria regia.
 

84

 

 

 

 

 

 

 

 

 

 

Fig. 66. Samples of Hastelloy N (Heat 5085) Stressed at 25°C. The"
control samples were exposed to static barren fuel salt for the indicated
time and the irradiated samples were exposed to the core of the MSRE.

these heats seem to demonstrate some small property'changes due to ther-
mal aging. These changes are not large, but must be followed to longer
times to make sure that they do not become large.

These two modified heats did not show a structural modification
near the surface, but did exhibit edge cracking when tested. The fre-
guency of edge cracks is much higher than noted for the controls.

One of the most confusing observations was the poor mechanical
properties of heat 67-504 from outside the core. This heat was pre-
viouslyl® exposed to the core for 9789 hr to a thermal fluence of

5.3 X 102%° neutrons/cm®. The present group of samples outside the core

 

18y, E. McCoy, An Evaluation of the Molten-Salt Reactor Experiment
Hastelloy N Surveillance Specimen — Third Group, ORNI-TM-2647 (1970).

 

 
85

was exposed to N, + 2 to 5% 0, for 17,033 hr and received a thermal
fluence of 2.5 x 101° neutrons/cmz. The latter group had poorer mechan-
ical properties than the group exposed to the higher fluence. The
rupture life at a given stress level was less (Fig. 41, p. 54), the
minimum creep rate higher (Fig. 42, p. 54), and the fracture strain lower
(Fig. 43, p. 55). The photomlcrographs in Figs. 60 and 61, pp. 76 and 77,
show some surface oxidation but no othe;‘unusual features. Two heats
should have been used, heat 67-502 (2% W + 0.5% Ti) and heat 67-504

(0.5% Hf). Postirradiation chemical analyses showed the presence of
0.5% Hf, but no tungsten was present. Hence, we have concluded that
only heat 67-504 was included in these tests.

Our surveillance program has given us the opportunity to look at
several alloys of modified composition. The creep properties of these
heats are compared with those of standard Hastelloy N in Fig; 67. The
curves were drawn through only four or five data points in each case,'
SO siight differences in s;ppg_are not significant. The rupture lives
of .the modified‘alloys'are within a factor of 2. The rupture lives of
the irradiated, modified heats are about equivalent to those of uwnirra-
diated standard Hastelldy N. The creep rates, shown in Fig. 67(b), are
lofier for fhe mbdified alloys by as much as a factor of 3. Heats 67-502
and 67-504 have the lowest creep rates. The postirradiation fracture
strains are shown in Fig. 67(c). Quite a range of fracture strains
resulted with a minimum of about 0.5% for standard Hastelloy N and a
minimm of about 7% for heat 67-551. Heats 67-502 and 67-504 had frac-
ture strains of 6.4%; heat 7320 was not much better than standard
Hastelloy N with only 2.8%.

ane of the modified alloys have shown any adverse corrosion behav-
ior in the salt. Thus, it would appear that we have several alloys that
are suitable for use in future reactors that operate at 650°C. However,
as discussed previously, these new alloys are very sensitive to irradia-
tion temperature and the proposed 700°C operating temperature of a
breeder is too high for these alloys.l® Present work offers encourage-
ment that an alloy that is stable at 700°C can be developed by adding

 

194. E. MbCoy'gz_gl., MSR Program Semiann. Progr. Rept. Aug. 31,
1969, ORNL-4449, p. 184.

 
 

 

STRESS (1000 psi)

70
60
50
40
30
20

10

ORNL-DWG 70-3980

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

TTT T T 7 FTTITd Tt 1T 1T i
N STANDARD HASTELLOY N UNIRRADIATED ’;
N ;
NG J]‘N ‘i
. » ‘
67-502, 5'3*1029‘"Q‘ M
\\\ ORNL-DWG 70-7316
. \\ 70
P,
\\\ %102 l Vs
N 7320, 5.1 A
T~ 0 11 4 Kr
o \ 7
\-.,...,,,._ ) 4 N 273- :?gzb L /flr/(/lfs'r-ssizo
- -4 . ~ 1
T S~ \\\\\ o T / 5.4X 10
"‘\. ) A - b *' ”/"
20 <1 [ TT™ N 71
.3x10%? STANDARD HASTELLOY N | | \ . A
| o~ 67-504, - A L/
o ™ 5.3x402 '8 67-502, ” # y ,/
9.4x10°CSTANDARD HASTELLOY N I o 40 5.3 X 1020~ ; //’ v
8 >
Q / ’/’u
~ V4 R
& A //fl /.4
W Z A 7 STANDARD HASTELLOY N
= "7 RV UNIRRADIATED, IRRADIATED
4 < |
y /‘// | -
A &
(™ —
10° 10! 102 10° 10° 20 ” // 7320, 20 z
RUPTURE TIME (hr} yd 4 d L ' | g
7 ‘ &
2 w
10 |— g
G
| <
i E
0 ?‘
2 5 32 2 5 w02 2 s ot 2 5 0 2
MINIMUM CREEP RATE (%/hr)
|
Fig. 67. Postirradiation Creep Properties at 650°C of Several Modified Alloys That Were Included in the
and heat numbers are shown by ea.c%n 1

ine (see Table 1, p. 4, for the chemical compofitien).

 

86

ORNL-DWG 70-3979

6 67-504, 5.3%10%°
AND
67-502, 5.3x10%°

7320, 5.1x10%°

1.3x40%%, STANDARD HASTELLOY N

9.4%10%°, STANDARD HASTELLOY N

0

04 1073 1072 10!
MINIMUM CREEP RATE (%/ hr)

Surveillance Program. The thermal fluence

 

 

9
 

87

larger smounts of titanium or combined amounts of these elements along

with niobium and hafnium.

SUMMARY AND CONCLUSIONS

The heats of Hastelloy N used in fabricating the MSRE have shown &
systematic deterioration of mechanical properties with increasing neutron
fluence. The material expoeed for the longest period of time in the
core has reached a thermal-fluence of 1.5 x 10?1 neutrons/cm? and a fast
fluence (> 50 kev) of 1.1 X 102! neutrons/cm®. These values are quite
close to those anticipated fer future reactors with a 30-year design |
life. The ductility of the material was too low, but the microstructure
was free of irradiation-induced voids and defects other than helium bub-
bles. Several heats of the modified alloys have been exposed to the MSRE
and these have bhetter postirradiation properties. They also seem to
have good corrosion resistance.

The standard Hastelloy N removed from the core shows some evidence
of corrosion. The corrosion seems generally to be due to the selective
removal of chromium, as predieted by prenuclear tests.- Some observations
that have not been explained adequately are (1) the presence of grain-
boundary cracks in the straps that held parts of the surveillance assem-
bly together, (2) the modified microstructure near the surface, and
(3) the formation of intergranular'cracks originating from the surface
when irradiated materials are strained.

' One of the modified alloys, heat 67-504, was ‘exposed to the cell
enviromment. This heat had been previousLy exposed to the MSRE. The
fiuence was. higher in the core, but the postirradiation properties were
_superlor to those of the material exposed to the cell env1ronment We

presently have no explanatlon for the observed behevmor.

ACKNOWLEDGMENTS

The author- is 1ndebted to many people for 3851st1ng in this study

- W. H. Cook and A. Taboada for design of the surveillance assembly and

insertion of the specimens; W. H. Cook and R. C. Steffy for measurements

 
 

 

88

of flux; J. R. Weir, Jr., R. E. Gehlbach, and C. E. Sessions for review
of the manuscript; E. J. Lawrence and J. L. Griffith for assembling the
surveillance and control specimens in the fixture; P. Haubenreich and

the MSRE Operation Staff for the extreme care with which they inserted and
removed the surveillance specimens; E. M. King and the Hot Cell Operation
Staff for developing techniques for cutting long rods into individual

- ‘specimens, determining specimen straightness, and assistance in running
creep and tensile tests; B. C. Willjams, B. McNabb, and H. W. Kline for
running tensile and creep tests on surveillance and control specimens;

J. Feltner for processing the test data; H. R. Tinch and N.:M. Atchley
for metallography of the control and surveillance specimens;

Frances Scarboro of The Metals and Ceramics Division Reports Office for
-preparing the menuscript; and the Graphic Arts Department for preparing
the drawings. | |

1y

*
89.

ORNL-TM-3063
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55. Nancy Cole e 103. J. W. Koger

56. C. W. Collins .~ ~104. H. W. Kohn

57. E. L. Compere R 105. R. B. Korsmeyer

58. W. H. Cook S 106. A. I. Krakoviak ‘

59. J. W. Cooke T - 107. T. S. Kress

60. L. T. Corbin . - = 108. J. A. Lane

61. J. L. Crowley o B 109. R. B. Lindauer

62. F. L. Culler - - ... . 110.. E. L. Long, Jr.

63. D. R. Cuneo o 111. A. L. Lotts
~64. J. E. Cunningham - - 112. M. I. Iundin

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66. J. H. DeVan 114. R. E. MacPherson
 

 

90

115. D. L. Manning 155. J. L. Scott
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117. R. W. McClung 157. J. H. Shaffer
118-123. H. E. McCoy , 158. W. H. Sides
124. D. L. McElroy 159. G. M. Slaughter
125. C. K. McGlothlan 160. A. N. Smith
126. C. J. McHargue 161. F. J. Smith
127. H. A. Mclain , 162. G. P. Smith
128. B. McNabb 163. 0. L. Smith
129. L. E. McNeese ' 164. P. G. Smith
130. J. R. McWherter 165. T. Sp:.ewak
131. A. S. Meyer 166. R. C. Steffy
132. R. L. Moore 167. R. A. Strehlow
133. D. M. Moulton 168. R. W. Swindeman
13%4. T. R. Mueller | 169. J. R. Tallackson
135. H. H. Nichol 170. R. E. Thoma
136. J. P. Nichols 171. D. B. Trauger
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201.
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