ORNL-TM-2647.txt

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4 AN EVALUATION OF THE MOLTEN-SALT REACTOR EXPERIMENT
3 |
HASTELLOY N SURVEILLANCE SPECIMENS - THIRD GROUP
H. E. McCoy, Jr.
- 1
|
3
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NOTICE This document contains information of a preliminary nature

3659
and was prepared primarily for internal use at the Oak Ridge National P! w ¥
Laboratory. It is subject to revision or correction and therefore does

not represent a final report.

PISTRIBUTION OF THIS DOCUMENT 15 UNLIMITED

   
 

 

LEGAL NOTICE

 

This report was prepared as an account of Government sponsored work., Neither the United States,

nor the Commission, nor any person octing on behalf of the Commission:

A, Mokes any warranty or representation, expressed or implied, with respect to the accuracy,
completeness, or usefulness of the information contained in this repart, or that the use of
any information, appoaratus, method, or process disclosed in this report may not infringe
privately awned rights; or

B. Assumes any liabilitias with respact to the use of, or for damages resulting from the use of
any information, apparatus, method, or process disclosed in this report,

As used in the above, “‘person acting on behalf of the Commission’’ includes any employee or

contractor of the Commission, or employee of such contractor, to the extent thot such employee

or contractor of the Commission, or smployee of such coentractor prepares, disseminates, or
provides access to, any information pursuant to his employment or contract with the Commission,
or his employment with such contracter,

 

 

P

v
ORNL~TM-2647

Contract No. W-7405-eng-26

METALS AND CERAMICS DIVISION

AN EVALUATION OF THE MOLTEN-SALT REACTOR EXPERIMENT
HASTELIOY N SURVEILLANCE SPECIMENS — THIRD GROUP

H. E. McCoy, Jr.

———— LEG AL NOHCE—_——T
This report was prepared as zn account of Government sponsored work. Neither the United

States, nor the Commission, nor any perscn acling on bebalf of the Commission:

A, Makes any warranty or representation, expressed or implied, with respect o the accu-
racy, compieteness, or usefulness of the information contained in this report, or that the use
of any information, apparatus, method, or procesa disclosed in this report may not infringe
privately owned rights; or

B. Assumes any liabilities with respect to the use of, or for damagea resulting from the
use of any information, apparatus, method, or process disclosed in this report.

As used in the above, '‘person acting on bebalf of the Commission’ [ncludes any em-
ployee or contractor of the Commlission, or employee of such contractor, to the extent that
such employee or contractor of the Commission, or employee of such contractor prepares,
disseminates, or provides access to, any information pursuant to kls employment or contract
with the Commiesion, or his employment with such contracior.

 

 

 

JANUARY 1970

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

"MENT 1S UNLIMIFED,
DISTRIBUTION OF THIS DOCT ict
iii

g

o
@

o

CONTENTS
Abstract « & ¢« ¢ v @ v e e i e e e e e e e e e e e
Introduction . . . « v ¢« ¢« 4 « 4 ¢ o 6 0 e s e s e
Experimental Details . . & « ¢ v ¢ o v + v v o o o o &
Surveillance Assemblies . . . . . « « + + « « .« .
Materials . . v v v o o & o o o o o 0 0 e a1 a e
Test SpPeCcimens .« + o o« o o o o o s o & o o o o &
Irradiation Conditions . . « « v v o o « o &« &+ o « &
Testing Techniques . . « « + « ¢ o « « o o o o &

Experimental Results o « & o v ¢ v ¢ v o« o v o « o &

Visual and Metallographic Examination . . . . .

Mechanical Property Data — Standard Hastelloy N
Tensile Properties . « o« ¢ o & o o &
Creep-Rupture Properties . . . . . « . « « o« &

Mechanical Property Data — Modified Hastelloy N .
Tensile Properties . . « & « o o o o o « o o &
Creep-Rupture Properties . . « « « « . « +

Metallographic Examination of Mechanical Property
SPECIMEIlS & 4 & & o o o o o o o s s o s s & o «

Discussion of Results . v o ¢ o o o « o o o o » @

Summary and Conclusions . . « « o o o o « « o o o &
Acknowledgments . + ¢ ¢ v 4 4 4 e« 4 @ a2 e e 2 s s
AppendiX .« o & v 4 4 e v e 0 e e e e e e e e e e

g2 0w o w Wwow W

wWowowonNn
o~ 0 2

36
62
66
67
70
AN EVALUATION OF THE MOLTEN-SALT REACTOR EXPERIMENT
HASTELIOY N SURVEILILANCE SPECIMENS — THIRD GROUP

 

 

H. E. McCoy, Jr.

ABSTRACT

We have examined the third group of Hastelloy N
surveillance samples removed from the Molten-Salt
Reactor Experiment. ©Standard Hastelloy N was removed
from the core after exposure to a thermal fluence of
9.4 x 10°° neutrons/cm® over a time period of 15,289 hr
at 650°C and from outside the reactor vessel after
exposure to a thermal fluence of 2.6 x 101° neutrons/cm2
over a time period of 20,789 hr at 650°C. The former
samples were exposed to the fuel salt and the later
samples were exposed to nitrogen plus 2 to 5% 0. The
material seemed dquite compatible with both environments,
Postirradiation tests showed that the fracture strain
was reduced at 25°C and above 500°C, The reduction in
ductility at 25°C is likely due to carbide precipitation
and the reduction above 500°C is due to the presence
of helium from the °B(N,x)”Li transmutation. The
accumulated results in this series of experiments allow
us to follow the changes in fracture strain over the
thermal fluence range 1.3 x 10'% to 9.4 x 10°° neutrons/cm?,

Two heats of modified Hastelloy N were removed
from the core after irradiation to a thermal fluence of
5.3 x 10%<° neu.trons/cm2 over a time period of 9789 hr
at 650°C. The postirradiation properties of these
alloys were better than those of standard Hastelloy N.

 

INTRODUCTION

The Molten-Salt Reactor Experiment (MSRE) is a single region reactor
that is fueled by a molten fluoride salt (65 LiF—29.1 BeF,—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 reactor design

and construction can be found elsewhere.1 We knew that the neutron

 

iR, C. Robinson, MSRE Design and Operations Report, Pt. 1, Descrip-
tion of Reactor Design, ORNL-TM-728 (January 1965).

 

 
environment would produce some changes in the two structural materials —

graphite and Hastelloy N. Although we were very confident of the com-

patibility of these materials with the fluoride salt, we needed to keep -
abreast of the possible development of corrosion problems within the v
reactor itself, TFor these reasons, we developed a surveillance program >

that would allow us to follow the property changes of graphite and
Hastelloy N specimens as the reactor operated.

The reactor went critical on June 1, 1965, and after numerous small
problems were solved, assumed normal operation in May 1966, We have
removed three groups of surveillance samples and the results of tests on

253 were

the Hastelloy N specimens from the first and second groups
reported previously. This report will deal primarily with the results

of tests on the Hastelloy N samples removed with the third group of sur-
veillance samples. The third group included two heats of standard
Hastelloy N that had been used in fabricating the MSRE and two heats

with modified chemistry that had better mechanical properties after irra-
diation and appear attractive for use in future molten-salt reactors.

The respective history of each lot was (1) standard Hastelloy N, exposed
in the MSRE cell to an environment of N, + 2 to 5% 0, for 20,789 hr at
650°C to a thermal fluence of 2.6 x 1017 neutrons/em®, (2) standard
Hastelloy N, exposed in the MSRE core to fluoride salt for 15,289 hr at
650°C to a thermal fluence of 9.4 x 10°° neutrons/cm®, and (3) modified
Hastelloy N, exposed in the MSRE core to fluoride salt for 9789 hr at
650°C to a thermal fluence of 5.3 x 10°° neutrons/ecm*. 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.

 

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

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

 

 

 

 
EXPERIMENTAL DETAILS

Surveillance Assemblijes

The core surveillance assembly was designed by W. H. Cook and
others, and the details have been reported previously.4 The specimens
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 vari-
ous pieces that are joined by pinning and tongue-and-groove Jjoints. 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 Jjoined 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 posi-
tion 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 MSRKE com-
puter so that the temperature matches that of the reactor. Thus, these
specimens are exposed to conditions similar to those in the reactor
except for the static salt and the absence of a neutron flux.

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

These specimens are exposed to the cell environment of Nz + 2 to 5% 0z,

Materials

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

tion of the reactor vessel and heat 5065 was used for forming the top

 

“W. H. Cook, MSR Program Semiann. Progr. Rept. Aug. 31, 1965,
ORNL-3872, p. &7.
 

 

 

 

Table 1. Chemical Analysis of Surveillance Heats
Content, wt %
Element

Heat 5065  Heat 5085 Heat 67-502 Heat 67-504
Cr 7.2 7.3 7.18 6.9%
Fe 3.9 3.5 0.034 .05
Mo 16.5 16.7 12.0 12.4
C 0.065 0.052 0.05 0.07
Si 0.60 0.58 0.015 0.010
Co 0.08 0.15 0.02 0.02
W 0.04 0.07 2.15 0.03
Mn 0.55 0,67 0.14 0,12
v 0.22 0.20 0.06 0.01
P 0.004 0.0043 0.001 0.002
S 0,007 0.004 < 0.002 0.003
Al 0.01 0,02 0.02 0.03
Ti 0,01 < 0.01 0.49 < 0.02
Cu 0.01 0.01 0.04 0.03
0 0.0016 0.0093 0.0002 < 00,0001
N 0,011 0.013 < 0,0001 0.0003
Zr < 0,002 < 0.01 0.01
Hf < 0,01 0.50

Analysis, ppm
B 24,37, 38 1 0.3
20,10

 

and bottom heads., These materials were given a mill anneal of 1 hr at
1177°C and a final anneal of 2 hr at 900°C at ORNL after fabrication.
The chemical compositions of the two modified alloys are given in
Table 1. The modifications in composition were made principally to
improve the alloy's resistance to radiation damage and to bring about
general improvements in the fabricability, weldability, and ductility.5
These alloys were small 100-1b heats made by 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., diam. rod. Two sections of rod were

 

 

°H. E. McCoy and J. R. Weir, Materials Development for Molten-Salt
Breeder Reactors, ORNL-TM-1854 (June 1967). |

 
welded together to make 62-in,-long rods for fabricating the samples.
The rods were annealed for 1 hr at 1177°C in argon and then the reduced

sections were machined,

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, the rods are sawed into small mechanical prop-
erty specimens having a gage section 1/8 in. in diameter 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 machine the 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.

IRRADIATION CONDITIONS

The jrradiation conditions for the three groups of surveillance
specimens that have been removed are summarized in Table 2. The environ-
ment in the core facility is the molten-fluoride fuel salt. The speci-
mens outside the core {designated ''vessel" specimens) are exposed to the

cell environment of Nz + 2 to 5% O».

Testing Techniques

The laboratory creep-rupture tests of unirradiated control specimens
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 femperature accuracy was *0.75%, the guaranteed accuracy of the

Chromel-P—Alumel thermocouples used.

"Qra
 

 

 

 

Table 2. Summary of Exposure Conditions of Surveillance Sam.plesa
Group 1 Group 2 Group 3
Core Core Vessel Core Core Vessel
Standard Modified Standard Standard Modified Standard
Hastelloy N Hastelloy N Hastelloy N Hastelloy N Hastelloy N Hastelloy N
Date inserted 9/8 /65 9/13 /66 8 /24 [65 9/13 /66 6/5/67 8 /24 [65
Date removed 7 /28 66 5/9/67 6/5/67 4/3/68 4 [3/68 5/7/68
Mwhr on MSRE at time 0.0066 gog2 0 8682 36,247 0
of insertion
Mwhr on MSRE at time 8682 36,247 36,247 72,441 72,441 72,441
of removal
Temperature, °C 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
Peak fluence, neutrons/cm2 ) _ )
Thermal (< 0.876 ev) 1.3 x 1040 4.1 x 10%° 1.3 x 10%? 9.4 x 10°° 5.3 x 10%° 2.6 x 1017
Epithermal (> 0.876 ev) 3.8 x 10°° 1.2 x 10%% 2.5 x 101° 2.8 x 10°1 1.6 x 10°% 5.0 x 101°
(> 50 kev) 1.2 x 10°° 3.7 x 10%° 2.1 x 10%° 8.5 x 10°° 4.8 x 10°0° 4.2 x 10%°
(> 1.22 Mev) 3.1 x 10%° 1.0 x 1079 5.5 x 1018 2.3 x 10%° 1.3 x 1020 1.1 x 10°
(> 2.02 Mev) 1.6 x 10%° 0.5 x 1020 3,0 x 1018 1.1 x 10%° 0.7 x 1020 6.0 x 1018
Peak flux, neutrons cm™° sec™  Mw ™t ]
Thermal (< 0.876 ev) 4.1 x 10¥* (b,e) 4.1 x 1012 (b,e) 1.0 x 1011 (v) 4.1 x 10 (p,e) 4.1 x 102 (b,e) 1.0 x 10! (b)
Epithermal (> 0.876 ev) 1.2 x 10%% (e¢) 1.2 x 1013 (e) 1.9 x 101 (b,e) 1.2 x 1012 (e) 1.2 x 1022 (o) 1.9 x 10! (p,e)
(> 50 kev) 3.7 x 10%* (e) 3.7 x 1012 (e) 1.6 x 101t (¢) 3,7 x 1012 (¢) 3.7 x 10+? (e) 1.6 x 1011 (e)
(> 1.22 Mev) 1.0 x 10 (b,e) 1.0 x 10*? (b,c) 4.2 x 101° (v 1.0 x 10*2 (b,e) 1.0 x 102 (b,e) 4.2 x 1019 (1)
(> 2.02 Mev) 0.5 x 10*? {(b,e) 0.5 x 101* (b,e) 2.3 x 101° (v) 0.5 x 10%2 (b,e) 0.5 x 102 (v,e) 2.3 x 1019 (1)
#Information compiled by R. C. Steffy. Revised for full-power operation at 8 Mw.

bExperimentally dete

cCalculated.

rmined.
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 accu-
racy of the strain measurement is difficult to determine, The
extensometer (mechanical and electrical portions) produced measurements
that could be read to about *0.02% strain; however, other factors (tem-
perature changes in the cell, mechanical vibrations, etc.) probably com-
bined 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
exception. In the laboratory, the control system was stabilized at the
desired temperature by use of a recorder with an expanded scale. 1In the
tests in the hot cells, the control point was established by setting the
controller without the aid of the expanded-scale recorder, This error
and the thermocouple accuracy combine to give a temperature uncertainty
of about *1%,

The tensile tests were run on Instron Universal Testing Machines.
The strain measurements were taken from the crosshead travel and gener-
ally are accurate to *2% strain.

The test environment was air in all cases., Metallographic examina-
tion showed that the depth of oxidation was small and we feel that the

environment did not appreciably influence the test results.
EXPERIMENTAL RESULTS

Visual and Metallographic Examination

W, H. Cook was in charge of the disassembly of the core surveillance
fixture. As shown in Fig. 1 the assembly was in excellent mechanical
condition when removed., The graphite and Hastelloy N surfaces were very

clean with markings such as numbers and tool marks clearly visible. The
 

 

 

 

 

 

 

 

 

 

 

 

Fig. 1. Molten-Salt Reactor Experiment Surveillance Speumens
from Run 14 (Stringers RS3, RL2, and RR2).

Hastelloy N was discolored slightly. The Hastelloy N surveillance rods
outside the reactor vessel were oxidized, bfit the oxide was tenacious.
We examined polished cross sections of segments from each of the
surveillance rods., Typical micrographs of the reods of standard
Hastelloy N that were located outside the reactor vessel are shown in
Figs. 2 and 3. The cell environment of N> + 2 to 5% 02 is oxidizing to
the alloy, but there is no evidence of nitriding. There is some inter-
nal oxidation to a depth of 1 to 2 mils and a very thin uniform surface
oxide. The general microstructure is characterized by large MgC-type
carbides that are distributed ,during' the primary working and by finer
MgC-type carbides that formed during ‘the Jong thermal anneal of
20,789 hr at 650°C. | |
 

 

 

 

 

 

 

 

 

 

   

    

  

 

 

 

9
-
s
3
. Fig. 2. Photomicropgraphs of Hastelloy N (Heat :5085) Surveillance
| Specimens Exposed to the Cell Environment of N» + 2 to 5% 02 for
| -~ 20,789 hr at 650°C. 500x. (&) Unetched showing surface oxidation.
~© {(b) Etched (glyceria regia) showing shallow modification of miecrostruc- -
ture due to reaction with cell enviromment. .. -~~~

 
 

10

R-45443

R-45444

)

 

Fig. 3. Photomicrographs of Hastelloy N (Heat 5065) Surveillance
Specimens Exposed to the Cell Environment of N; + 2 to 5% 0, for
20,789 hr at 650°C. 500x. . (a) Unetched showing surface oxidation.
(b) Etched (glyceria regia) showing shallow modification of microstruc-
ture due to reaction with cell environment.
11

Typical photomicrographs of the standard Hastelloy N samples
exposed to the MSRE fuel salt for 15,289 hr at 650°C are shown in
Figs. 4 and 5. There is a thin layer of modified structure less than
0.5 mil thick at the surface. This modified structure was noted pre-
viously.6 As shown in Figs. 6 and 7, a similar product is formed on
the surfaces of the control specimens. We originally suspected that
this product formed only where the Hastelloy N was in intimate contact
with graphite, but a closer examination shows that the modified struc-
ture exists around the complete circumference of the rod and not Just
where it makes tangential contact with graphite. From the practical
standpoint, the modified layer has not changed detectably since we first
observed the samples removed after 4800 hr of exposure and we feel that
it will not influence the mechanical properties of the material.

The amount of fine MgC-type precipitates is not detectably differ-
ent for the material aged while being irradiated (Figs. 4 and 5) and
that aged in the absence of irradiation (Figs. 6 and 7). There does
seem to be more precipitate present in heat 5065 than in heat 5085, an
observation in keeping with the higher carbon content of heat 5065
(Table 1, p. 4). Extracted precipitates were found to be of the MC type
with a lattice parameter of 11.02 A. Some of the material exposed to the
highest thermal fluence of 9.4 X 10<° neu.trons/bm2 was examined in trans-
mission., A typical electron photomicrograph is shown in Fig. 8 where
helium bubbles are clearly visible along several of the grain boundaries.
This sample had not been stressed and the rather large sizes of these
bubbles attests to the initial inhomogeneous distribution of the boron.
However, the observation of bubbles of this size only serves to support
the formation of gas bubbles in metals during irradiation and really is
not informative from a mechanistic standpoint. Deformation takes place
on an atomistic scale (few angstroms) and our ability to see details in
the grain boundaries of such specimens is limited to about 50 A, so we

are not yet able to see details of the size involved in deformation.

 

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

 

 

 

 

 

 

 

 

 

 

 

 

 

R-45437

 

 

 

 

 

 

 

 

R-45438

 

 

 

 

Fig. 4. Photomicrographs of Hastelloy N (Heat 5085) Surveillance
Specimens Exposed to Fuel Salt for 15,289 hr at 650°C. 500x.
(a) Unetched. (b) Etched (glyceria regla) photomlcrogra.phs showing
shallow reaction layer near surface,.
e

oWingf

i

5
aphs

5065) Surveillance
sh

°C
photomicrogr

i
e

=

e
T

e

)

9 hr at 650

3

regisa

ria
face.

of Hastelloy N (Heat
28¢

for 15

e

5
sur

 
 
 

aph
Salt
(g1

grap
near

 

g3gd

5.

tched
shallow reaction laye

ig.

F
cimens Expos
Une

a)

 

 
 
 

 

14

Y-90144

Y-90145

Fig. 6. Photomicrographs of Hastelloy N (Heat 5085) Surveillance
Control Specimens Exposed to Static Barren Fuel Salt for 15,289 hr at
650°C. Note the shallow reaction layer near the surface. (a) Etched,
100X. (b) Etched, 500X. Etchant: glyceria regia. |

 
 

 

 

 

 

 

 

 

 

 

 

¥

 

 

 

 

 

 

 

- Fig. 7. Photomicrggraphs of Hastelloy'N (Heat 5065) Surveillance
- Control Specimens Exposed to Static Barren Fuel Salt for 15,289 hr at
650°C, ~ Note the shallow reaction layer near the surface. (a) Etched.
100x., "(b) As polished, 500>< “(e) Btched. - 500x%. Etchant: glyceria

regia.

 
16

YE-9886

 

-
Fig. 8. Transmission Electron Photomicrograph of Hastelloy N
(Heat 5085) Exposed to the MSRE Core for 15,289 hr at 650°C. The ther-

msl fluence was 9.4 X 102° neutrons /cm2 s enough to transmute most of
the 108 in this material to helium.
17

Typical photomicrographs of the modified Hastelloy N are shown in
Figs. 9 and 10. The grain size in this material is rather large due to
the high preirradiation anneal of 1 hr at 1177°C and the absence of the
carbide stringers. There is a thin layer on the surface, which has the
appearance of being a deposit rather than where material has been removed
by corrosion. X-ray studies indicate the presence of iron. Since these
alloys contain only trace quantities of iron compared with 4 to 5% for the
rest of the material in the MSRE and in the control facility, it is quite
reasonable that iron should be deposited on the surfaces of the modified
alloys. The general microstructure contains a finely dispersed precipi-

tate with larger amounts along the grain boundaries.

Mechanical Property Data — Standard Hastelloy N

Tensile Properties

 

The postirradiation tensile properties of heat 5085 after exposure
to the cell enviromment of Np + 2 to 5% 0p for 20,789 hr are given in
Table A-1 (Appendix) and the fracture strain is plotted as a function of
test temperature in Fig. 11. There are significant reductions in the
fracture strain at 25°C and above 500°C. The fracture strain decreased
with decreasing strain rate at the elevated temperatures. One particu-
larly interesting observation was that the fracture strain at room tem-
perature could be improved by a postirradiation anneal of & hr at 870°C.
This anneal is quite often used as a postweld anneal and is sufficient
to precipitate (or redissolve) carbides and to relieve residual stresses,
but does not cause grain-boundary motion. Thus, the recovery of the
room-temperature ductility by this anneal supports the supposition that
the reduction in ductility at 25°C is due to carbide precipitation.

The results of tensile tests on heat 5065 after exposure to the
MSRE cell environment for 20,789 hr at 650°C are given in Table A-2
(Appendix) and the fracture strain is shown as a function of test tem-
perature in Fig. 12. This heat does not exhibit the reduction in duc-
tility at 25°C, but does show a substantial loss in ductility above
500°C due to irradiation. The effect of strain rate is qualitatively
the same as that shown in Fig. 11 for heat 5085, but the scatter in

experimental results does not allow a quantitative comparison.
 

 

 

 

 

 

 

 

 

 

 

 

R-45441

 

R-45442

 

 

 

 

 

 

 

 

 

(b)

Fig. 9. Photomicrographs of Modified Hastelloy N Contalna.ng 2% W
and 0.5% Ti (Heat 67-502) After Exposure to the MSRE Core for 9789 hr at
650°C and a Thermal Fluence of 5.3 x 10°C neutrons/em?. 500x.

(a) As polished. (b) Etchant: glyceria regia. This structure is also
representative of that of a heat of material containing 0.5% Hf (Heat
67-504) that had a similar exposure.
{
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!
!
I

 

 

 

 

 

 

 

 

-
-

«4

 

 

 

 

 

 

Fig. 10, Photomicrographs of Modified Hastelloy N Containing 2% W

and 0.5% Ti (Heat 67-502) After Exposure to Static Barren Fuel Salt for
9789 hr at 650°C. 500x. (a) Edge and (b) typical matrix. Etchant:
glyceria regia. This structure is also representative of that of a heat

of material containing O.‘S%Hf"_r(heat'-_'6_'7-:503r) that had a similar exposure.

 

 

 
20

ORNL —DWG 69-4464

 

 

60 T
| | I
STRAIN RATE
{(min™Y)
50 .
© PI ANNEAL AT o 0.05
8hr AT 870°C A 0.002

. ,//’,!xr”/”’ | \ ® 0.0005
AN
. N

 

 

 

 

 

 

 

FRACTURE STRAIN (%)

 

 

 

 

 

 

 

 

 

 

 

10 —- . .
\A\b\
. “x\\ik
| A
0 : ‘
0 $00 200 300 400 500 600 700 800 300

TEST TEMPERATURE (°C)
Fig. 11. Postirradiation Tensile Properties of Hastelloy N (Heat

5085) After Irradiation to a Thermal Fluence of 2.6 x 10?7 neutrons/cm®
Over 20,789 hr at 650°C.

ORNL-DWG 69-4462

 

70 : '([

 

 

 

 

 

 

 

 

 

:\5 I
=

=1

24

o

w 0 STRAIN RATE

x {min™") A\ 4

5 o 0.05

= 20 A 0,002 . U\

&

 

 

 

 

® 0.0005 ¢
| &
| :tt Q

 

 

 

 

 

 

 

 

 

 

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

Fig. 12. Postirradiation Tensile Properties of Hastelloy N (Heat
5065) After Irradiation to a Thermal Fluence of 2.6 x 101° neu.trons/cm2
Over 20,789 hr at 650°C.
21

Heats 5085 and 5065 were also removed from the core surveillance
and control facilities after 15,289 hr of exposure. The detailed ten-
sile data on these specimens are given in Tables A-3 through A-6
(Appendix). The variation of the fracture strain with test temperature
is shown in Fig. 13 for heat 5085 after exposure in the core surveil-
lance facility to a thermal fluence of 9.4 x 10°° neutrons/cm2 and in
Fig. 14 for heat 5085 after exposure in the control facility. The irra-
diated samples (Fig. 13) showed reduced fracture strains compared with
the control samples (Fig. 14) over the entire test temperature range
studied. ©Similar plots of the variation of fracture strain with test
temperature are shown for heat 5065 after removal from the core surveil-
lance facility (Fig. 15) and the control facility (Fig. 16). There is a
slight reduction in the fracture strain at 25°C due to irradiation, but
the reduction is not nearly as great as noted for heat 5085 (Fig. 13).
The fracture strain of heat 5065 is reduced above 500°C by irradiation,
and the resulting strains are lower than those for heat 5085 (Fig. 13).
For example, at a test temperature of 650°C and a strain rate of
0.002 min~! heat 5085 (Fig. 13) fractures at a strain of 5% and heat 5065
(Fig. 15) fractures at 3.5%. The unirradiated samples from the control
facility show a greater reduction in fracture strain for heat 5085
(Fig. 14) than for heat 5065 (Fig. 16) at 25°C, but the fracture strains
at 650°C are not appreciably different for the two heats.

Heat 5085 has been carried throughout our surveillance program, so
we now have accumulated enough data to see how the fracture strain is
varying with neutron fluence. The fracture strain is shown as a function
of test temperature in Fig. 17 for heat 5085. Generally, the fracture
strain is decreased with increasing fluence. At low fluence the ductil-
ity reduction at low test temperatures is restricted to 25°C and recovers
at higher test temperatures, but at higher fluences the fracture strain
is reduced over a wider temperature range. We attribute this embrittle-
ment to carbide precipitation and, as shown in Fig. 11, the ductility can
be recovered by postirradiation annealing. The ductility reduction above
500°C is associated with the presence of helium and the increased ten-
dency for intergranular fracture. The fracture strain at a given temper-

ature above 500°C generally decreases with increasing thermal fluence.
ORNL-— DWG 69-4463

 

50 T

H
o

 

 

 

 

op—
_7LH___M

 

 

 

 

 

|
|

 

FRACTURE STRAIN (%)
o

 

 

 

 

 
 

 

 

 

 

 

20 STRAIN RATE ‘ <
‘ (min~ 1) | 'y |
o 0.05 % \\\\j\i\\\
A 0002 ’ & :
10 ' | < ‘
NS
% J , ~—
o | | -
0 100 200 300 400 500 600 700 800 900

TEST TEMPERATURE (°C)

Fig. 13. DPostirradiation Tensile Properties of Hastelloy N (Heat
5085) After Irradiation to a Thermal Fluence of 9.4 X 1029 neutrons /cm®
Over 15,289 hr at 650°C.

QRNL-DWG 69— 4464
60

o 4

~— ‘(\
A /T/I \\

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

S \
= 40
= & X
= \
x \
& \ \
N\
S A \ck
5 STRAIN RATE .
= (min~1) N
w20 © 0.05 N ®
’ A
A 0.002 ;F\H——fl
|
10 - :
o |
0 100 200 300 400 500 600 700 800 900

TEST TEMPERATURE (°C)

Fig. 14. Tensile Properties of Hastelloy N (Heat 5085) After Aging
for 15,289 hr at 650°C in Static Barren Fuel Salt.
23

ORNL-DWG 69-4465

 

 

 

 

 

 

 

 

50
__——-—'_'—-__
40 O | \ N
2
=
g 30 STRAIN RATE
=
5 {min )
i 0 0.05
2 20 a 0.002 \
x
[T

 

10

 

 

 

 

 

 

 

 

 

 

 

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

Fig. 15. Postirradiation Tensile Properties of Hastelloy N (Heat

5065) After Irradiation to a Thermal Fluence of 9.4 x 1020 neutrons /cm?
Over 15,289 hr at 650°C.

ORNL-DOWG 69-4466

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

30 o
O
-~
40 -
\

—_ \
& \
~ STRAIN RATE \ \
&% 20 (min~ Y \
= 0 0.05 \
o & 0.002 T
o \
o
L

10

0

0 100 200 300 400 500 600 700 800 900

TEST TEMPERATURE (°C)

Fig. 16. Tensile Properties of Hastelloy N (Heat 5065) After Aging
for 15,289 hr at 650°C in Static Barren Fuel Salt.
24

ORNL-DWG 68-11774R

 

 

 

 

 

 

 

  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

60
J HEAT 5085 _
50 ‘  a STRAIN RATE = 0.05 min
T
O/I' /’—_“"‘--..“\
A .‘ -7 fi e

£40 ot ‘ ™\ .
~— ; ’; ;W———~ -+ \ Y
% o v \\ \ \ /
o V/ - ' == '—: \ \ //
- ,. . - — Y
D30 b IO . 13 Y% N\
% ‘ 4 A \
5 . S 4800 hr IN STATIC SALT YO\
I Y 19 OSSN A ]
x 20 4 13x10", 11,000 hr 'qu\ AR
L = 26x10'°,20,789 hr AR }\\--\\

v 13x102°, 4800 hr N s Sy

10 = o 94x10%° 15,289 hr ~ o \0\ \‘;\\ <
¢ 1.5x40% 22,533 hr OIS,
| | =
0
0 100 200 300 400 500 600 7OO 800 900 1000

TEST TEMPERATURE (°C)

Fig. 17. Postirradiation Tensile Properties of MSRE Surveillance
Samples.

1

The fracture strains for heat 5085 at a slower strain rate, 0.002 min™-,

are shown in Fig. 18. The behavior as a function of fluence is similar
to that shown in Fig. 17 at a higher strain rate. An important excep-

tion is the observation that as the test temperature is increased, the

fracture strain depends on fluence to a lesser extent.

One very important question that arises is "How much of the reduc-
tion in fracture strain in Fig. 17 can be attributed to thermal aging?"
Several of the tensile results obtained on samples from the control
facility are shown in Fig. 19. The results show generally that the
fracture strain is reduced at all test temperatures by aging at 650°C.
However, the reductions in fracture strain due to aging do not account
for nearly all the reduction noted for irradiated samples. For example,
the fracture strains at test temperatures of 25 and 650°C are 53.0 and
33.6% for as-annealed samples, 38.6 and 19.6% after aging for 15,289 hr
at 650°C (Fig. 19), and 28.5 and 11.0% after irradiation to a fluence
of 9.4 x 102° neutrons/cm? over a period of 15,289 hr at 650°C. We do

not have as much data on the response of heat 5065 to aging, but a
25

ORNL-DWG 69-4467

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

50 !
NEUTRON FLUENCE  TIME AT
{neutrons /cm?< ) 850° C (hr) C\
40 —— & 13x10"% 11,000 UNIRRADIATED —
o 2.6x10° 20,789 el
2 v 13x40% 4800 <
z . o 9.4x10%° 15,289
<{
= o\\x {
w
w N
o
2 20 o
g
T 20 13 AND 2.6 x40™
1.3 x10 O
N |
10 7
9.4 x 102
0 8
0 100 200 300 400 500 600 700 800 900
TEST TEMPERATURE (°C)
Fig. 18. Postirradiation Tensile Properties (Strain Rate of

0.002 min~!) of Hastelloy N (Heat 5085) After Exposure to Various Neutron
Fluences.

ORNL-DWG 694468

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

60 T T G
l ’/-DATA BEING FILLED IN
j | °
50 L <
o |1 g N
/ \ o
- \ P
o 40 00—
& A “ /CK//
z \ -
% 30 \ /,a’
g Y \/I/
> A
'_
Q \
= e
E 20 ——  AGING TIME AT Ye
650°C (hr)
oo
o 4800
10 — A 15,289
0
0 100 200 300 400 500 600 700 800 900
TEST TEMPERATURE (°C)
Fig. 19. Variation of the Tensile Properties (Strain Rate of

0.5 min~!) of Hastelloy N (Heat 5085) with Aging Time in Barren Fuel
Salt at 650°C.
26

comparison of Figs. 14 and 16, pp. 22 and 23, of the fracture strain for
heats 5065 and 5085 after aging for 15,289 hr at 650°C indicates very
similar properties of the two heats at 650°C and higher fracture strain
for heat 5065 at 25°C.

The combined observations on the change in the fracture strain at
a test temperature of 25°C are shown in Fig. 20, The results are plotted
as if aging time were the controlling variable, and the results on the
unirradiated samples (open symbols) indicate that aging time is an impor-
tant factor. However, for the cases where irradiated and unirradiated
samples were tested after similar thermal histories, the irradiated sam-
ples showed the larger reduction in fracture strain. Thus, neutron
fluence as well as time at temperature are important factors in the
reduction of the fracture strain at low temperatures. The question of
whether the role of irradiation is to produce defects that enhance the
growth rates of precipitates or to produce transmutation products that
aid in stabilizing a precipitate nucleus yet remains unanswered. The
data shown for heat 5065 in Fig. 20 indicate that this material is less
susceptible to this type of embrittlement than heat 5085,

Creep-Rupture Properties

 

Some of the samples were tested under creep conditions at 650°C,
The detailed results of these experiments are presented in Tables A-7
through A-9 (Appendix) and some correlations of these results with those

758 will be presented,

obtained from previous surveillance samples
A stress-rupture plot for heat 5085 is shown in Fig. 21. The
results obtained on the unirradiated samples show that the rupture life
is reduced slightly by prolonged aging at 650°C, the magnitude of the
effect decreasing with decreasing stress level, The rupture life is

decreased markedly by irradiation even at thermal fluences as low as

 

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

®H. E. McCoy, An Evaluation of the Molten-Salt Reactor Experiment
Hastelloy N Surveillance Specimen — Second Group, ORNL-TM-2359
(February 1969).
27

ORNL-DWG 69-4469
HEAT IRRADIATED UNIRRADIATED

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

5065 ° o
5085 a a
100
=
g e 5
- THERMAL Q @ o %
» NEUTRON e o e <
W FLUENCE > = = *
o 3 80 I~ (neutrons/cm?) o 0 <, ©
=~ < - o o
o
g Q
@ Lon .
w ot 40 °
= . b
- =
= <
o 8
k20 % .
D O
o
)
o
0o
-10 *
0 2 4 6 8 10 12 14 16 18 20

AGING TIME AT 650°C (100 hr)

Fig. 20. Variation of the Fracture Strain at 25°C with Aging and
Irradiation.

ORNL-DWG 69-—4470

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

70 T\\
\\
60 UNIRRADIATED —]
.\\L ~ \\
A p o
A ~ \’\\ N
My
50 T PRI \\
= 9 ~3 o T e
g SN IS
TR TN RN
8 40 = N §'\\ N
9 ™ ::Nh..‘;\ “‘“:\
0 THERMAL SN /7 “%Eh\ Bou,
¥ 30 =  FLUENCE TIME AT S PR
o (neutrons/cm?) 650°C (hr) flL\\ o [&7
» e UNIRRADIATED 0 “‘\
o0 | ® UNIRRADIATED 4800 I o~
A UNIRRADIATED 15,289 T~ 1,
o 1.3 x 10" 11,000 o TT
226 x 107 20,789 1.3 x 10°° neutrons/cm® (MSRE) AND
101~ o 1.3 x 10%° 4800 3-5 x 10%° neutrons/cm? (ORR)
s LI I [
o Lo frri
167" 10° 10' 10° 10° 10*

RUPTURE TIME (hr)

Fig. 21. Postirradiation Stress Rupture Properties of MSRE Surveil-
lance Specimens (Heat 5085) at 650°C.
28

1.3 x 10%° neutrons/cmz. The rupture life is reduced progressively by
increasing fluence with a large step occurring between fluences of
1.3 and 9.4 x 10°° neutrons /cm®.

The minimum creep rate is shown as a function of stress level in
Fig. 22 for these same samples from heat 5085. The results on unirradi-
ated samples show that the creep rate is increased slightly at high
stfess levels by prolonged aging at 650°C. A few irradiated samples
tested at high stress levels and the samples irradiated to
9.4 x 1020 neutrons/bm2 have a higher minimum creep rate, This is likely
due to the tests being so short and the fracture strains so low that the
steady-state creep period (the period in which the minimum creep rate
usually occurs) was not reached and the rate that we measured was too
high. However, generally at lower stress levels irradiation and aging
seem to have little effect on the minimum creep rate.

The -fracture strains are shown as a function of minimum creep rate
in Fig. 23 for the irradiated samples of heat 5085, The scatterband
shown in this figure was determined from numerous tests on samples of
these same heats of standard Hastelloy N that had been irradiated in the
Oak Ridge Research Reactor (ORR) to thermal fluences of 2 to
5 x 10°9 neutrons/émz. In samples removed from the MSRE the fracture
strain was decreased from about 20% for unirradiated material to about
2.3% by a thermal fluence of 1.3 x 10%° neutrons/cmz. Progressively
higher fluences reduced the fracture strain even further. The limited
number of data points for each lot of material do not give a very strong
indication of the very distinct ductility minimum noted in our ORR irra-
diations. This may be associated with the much larger times involved in
the MSRE irradiations (5000 to 20,000 hr) compared with those in the
ORR (1000 hr).

The effects of aging in the absence of irradiation on the fracture
strain of heat 5085 at 650°C are shown in Fig. 24, After annealing for
2 hr at 900°C the fracture strain is 30 to 40%, almost independent of
strain rate. After aging for 4800 hr at 650°C the fracture strains are
now in the range of 20 to 30%. Aging for 15,289 hr causes a small
further reduction in the fracture strain at high strain rates, but no

significant change in the fracture strain at low strain rates. Thus, the
29

ORNL—DWG 69-44T71

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

70 ‘
l‘/ |
680 // l
/'
A ‘
/ 1 i
A
50 // |
— I']/Il A r
- e 0 P
8 T l |
O 40 ..A.o—;- | d‘
S . A I\ THERMAL
= (L\i-// L FLUENCE TIME AT
« 40 I ¢ {neutrons /cm?) 650°C (hr) !
o 30 e UNIRRADIATED 0 !
i > ,'f/ P o 1.3 x 10" 11,000
w / . 1
20 Ao s 2.6 x 10 20,789
ol L™ o 1.3 x 40%° 4800
// 0 9.4 x 10%° 15,289
10 ® UNIRRADIATED 4800 i
a UNIRRADIATED 15,289
b
o i
1074 1073 1072 101 10° 10

MINIMUM CREEP RATE (%/hr)

Fig., 22, Minimum Creep Rate of Hastelloy N (Heat 5085) Surveillance
Specimens from MSRE at 650°C.

ORNL-DWG 69-4472

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

5

4
;5 RANGE 2-5x102° neuh”ons/c:m2 (ORR)H«..Z/

Vi
23 4
7 19 2 o

@ / _-1.3%40"7 neutrons/cm /é%éé
» / L] ok
) - A
K2 2 6x10"° Yé%ffi 14/24/ 20 1t
= O X V1 L+1.3%10
Q neutrons /cm? /\//A/,4>> » _—7 Aé% A | neutron 2
= N s LT 74BN eutrons/cm
£ S P 4

YT YR

20 2 LTl A
9.4 x10 neutrons/cm - 2
LT T

10-4 10~3 102 10~ 100 101
MINIMUM CREEP RATE (%/hr)

Fig. 23. Variation of Fracture Strain with Strain Rate for
Hastelloy N (Heat 5085) Surveillance Specimens at 650°C.
30

ORNL-DWG 69-4473

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

50 TTTTT T TTT
® AS ANNEALED
4 4800 hr

40 ‘ 0 45,289 hr
§ o . »
— """l-qu-"'-—u__..____- ® —
<
§ 30
= T AT
wn T T~ _.—-——"—""‘—-—-_‘
ul F \m?‘“—-..“ A ‘
a O
— et O
O
g
& |

10 CREEP t TENSILE

TESTS ||| TESTS
. A
102 10~ 10° 10! 102

STRAIN RATE (%/hr)

Fig. 24. Influence of Aging at €50°C on the Fracture Strain of
Hastelloy N (Heat 5085) When Tested at 650°C.

fracture strain in a sample strained at a rate of 0.1%/hr at 650°C is
about 20% for heat 5085 after aging 15,289 hr at 650°C (Fig. 24) and
only 0.5% (Fig. 23) after a thermal fluence of 9.4 X 10°9 neutrons/cm®.

The results of our stress-rupture tests on heat 5065 are summarized
in Fig. 25. Contrary to the behavior of heat 5085 (Fig. 21), thermal
aging seems to have no effect on the creep-rupture properties of
heat 5065. However, the rupture life is reduced appreciably by irradia-
tion and the resulting lifetimes at a given stress are quite comparable
for heats 5065 and 5085,

The minimum creep rate is shown as a function of stress level for
heat 5065 in Fig. 26. The minimum creep rate is increased slightly by
thermal aging at 650°C. The creep rates seem to be high for some of the
irradiated samples, but this is likely associated with the fracture
strain being so low that the minimum creep rate was not established.

The variation of fracture strain with minimum creep rate for the
irradiated samples of heat 5065 is shown in Fig. 27. The ductility of
this heat is about the same as that shown in Fig. 23 for heat 5085,

except for the material irradiated to a thermal fluence of
31

ORNL-DWG 69-4474

 

 

 

 

 

 

 

 

 

 

 

 

 

 

    

 

 

 

 

 

 
  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

70 =\ 1 r M
N |
\\~. :
60 \\\c |
: .\ . .
: n
1N \7'\
50 ~ ‘
~ ™ ™ ;
—_ ? S~ NS0 o m F |
» i { ™ -~ \ \ P
Q
o 40 AVERAGE OR ~ . N | \%}_
o 2-5x1029 neutrons /om® =" T N ™
O ] | \,__\ \\
= \‘\ \\‘
2 | a \ NS ,
W 30 ? = \ r - \\ —— N e ‘
£ LT T SRS\
E ‘ ! : --."'-d..____.“‘ \.-"\ i L
THERMAL FLUENCE  TIME AT Tl \;
20 +— {neutrons/cm?) 650°C (hr) - {- <l i \é\
S| | N
e UNIRRADIATED 0 1~
= UNIRRADIATED 15,289 ; o
t0 }— o 1.3x10'® 11,000 —+- - - R
s 26x10" 20,789 | | | |
o  94x10 4800 | | .i
0 Lol i | Ll
Tol 10° 10! 102 10° 10
RUPTURE TIME (hr)
Fig. 25. Stress Rupture Properties of MSRE Surveillance Specimens
(Heat 5065) at 650°C.
ORNL-DWG 69-4475
70 T T T T T T °
THERMAL A
FLUENCE TIME AT e
60 L (neutrons/cm?) 650°C (hr)
e UNIRRADIATED 0 e
B UNIRRADIATED 15,289 0 p
5o @ 1.3 x 10" 1,000 3
= s 2.6 x 10" 20,789 A L om o
3z 0 9.4 x 10%° 4800 //
O ala
o 40
9 /
w dlfl/ a P
A o < 0
% . | 1
» é
20
) /
%
ol
10
° 2 —4 0 1
107% 1073 10” 10 10 10
MINIMUM CREEP RATE (% /hr)
Fig. 26. Minimum Creep Rate of Hastelloy N (Heat 5065) Surveillance

 

Samples from MSRE at 650°C.
32

ORNL-DWG 69-4476

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

7 ? I T T T T T T
T T T T T T
THERMAL FLUENCE  TIME AT 650°C | s f ;
c ( neutrons/cm?) ( hr) L gk |
a 1.3 x10' 11,000 M S
| s 26x10"® 20,789 |}
: 20 i
5 ———+094x10 4800 i|i =+ , F
32 |1 | Pl ; 20
< i | . AVERAGE 2-5x10
g 1 : | | | neutrons/cm2 (ORR) ) :
= 4 | T L 1l | 1\ e
B f ! | | bl | v’ ,
ol > | Ll VA
P CTHTIZ 27
g & | 4//4,////3 j .
3 7 | 7R
- ] /42,4é2’ | o S |
2 , < i 7777800
/ > | 7
%57, S _ // '
. 711 ]
?,//22/}.//; A 7 | |
1 27%02?@3a/ 7 1]
ez | |
0 O 1 ' ; ! 1 i ! H
10” 1072 10! 10° 10"

 

MINIMUM CREEP RATE (% /hr)

Fig. 27. Variation of Fracture Strain with Strain Rate for
Hastelloy N (Heat 5065) Surveillance Specimens at 650°C.

9.4 x 10°° neutrons/em® for which heat 5065 has much lower fracture
strains.

Our observations on the variation of fracture strain of surveillance
specimens with strain rate are summarized in Fig. 28. Data are included
in this figure for both tensile and creep tests. The scatterband based
on our observations on samples irradiated in the ORR is also shown in
Fig., 28 (ref. 9)., The fracture strain varies widely with fluence at
high strain rates and varies much less at low strain rates. At low
strain rates, the fracture strains for the samples irradiated to
1.3 x 10%° neutrons/cm2 fall above the scatterband and those for samples

irradiated to 9.4 x 10°C neutrons/cm? fall below the scatterband.

 

°H. E. McCoy, "Variation of the Mechanical Properties of Irradiated
Hastelloy N with Strain Rate,” J. Nucl. Mater. 312(1), 67-85 (May 1969).
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

g}
m T T T = o
~ 4 ) + -+ —
P t - + +
il - ] T
N > — N
© N + =
789777 7 AL :
S A MR 5 Y
W - O 0 W @ O - < =
R T R R R R L R UL SRR SRS TGN NR N —
O B, LULAATIALAAEIRARTARASLTRUARERRA AL AU - o
SRR TR AR AR A VR A o 1
O ,7//////////////////// - m _ W_._ o ———]
AT 46% m,w R
//J//////////// AN /7 O 2w
o
—
.  uo //////////////////// L+ 8% . 5
; — 1- IE.V,ZZ T L LT TR TR E LA T IR x * ] &
: — I OIS LEE TR TEERATEREE CR R ST Gy, W w0 o
- _ LR AT G V I vy T
! o R R o W -
“ & o B N N //////f w v
T e 28 . Mt W]
» ~
R S g O efifl%%yfi%y
E o %%///////%
. O
O 8 %74?. %
™M o o Q — =S ERTTEIRERRERRRY = -
o) .G. - o A /rV//// AR -
> & AALARLEA
- M <t N A
N ~ _1 ‘ BRI
" 2 o S\
- = 2 9 SN
< " x @Q N *
J o 0 O ,/ -
= 0 o o !
— o o o Q
B I W > © O O ARy
I S oo ¥ O NN
——— e D) Ny - _ N
Ll - - 6 .u.m
o
0 w
= m o~
[ OO R ———— w — T —_ _ - SRS, _O
e m— — 0w JN « £ - fild»%//%; Y ~
[ — oo E - O T ANNW
— woo k Ll QO O AR
ToHowb= =9 NN
T 2EEQ i WA
e s 1 AR
T / 0
1 1 - L. / . _
*- Q
0 < 0 o = o o» w© M~ W o] <t o o ~ O

(%) NIVYHLS 34NLOVYHS

STRAIN RATE (% /hr)

Fig. 28.
Hastelloy N Surveillance Specimens at 650°C,

Variation of Fracture Strain with Strain Rate for
34

Mechanical Property Data — Modified Hastelloy N

Tensile Properties

 

Two heats of modified Hastelloy N (Table 1, p. 4) were included in
both the core surveillance facility and the control facility.
Heat 67-502 contained a nominal 0.5% Ti and 2% W and heat 67-504 con-
tained a nominal addition of 0.5% Hf. These samples were exposed to
salt 9789 hr at 650°C and those in the MSRE received a thermal neutron
flvence of 5.3 x 10°° neutrons/em®. The detailed results of tensile
tests on these materials are given in Tables A-10 through A-13 (Appendix).

L are compared in

The fracture strains at a strain rate of 0.05 min~
Fig. 29 for the modified alloys. In the unirradiated condition,
heat 67-502 had higher fracture strains than heat 67-504 up to about
600°C where the opposite trend was noted. The shapes of these curves
are different from those shown in Figs. 14 and 16, pp. <22 and 23, for
the standard alloy. The fracture strain of the standard alloy decreases
rapidly with increasing temperature above about 500°C and a ductility
minimum occurs at about 650°C, For the modified alloys, the fracture
strain does not begin to decrease with temperature until about ©600°C
and the strain continues to decrease up to 850°C, the highest tempera-
ture investigated. The fracture strains of the modified alloys after
irradiation, also shown in Fig. 29, are reduced at temperatures above
about 500°C., Both alloys have about the same fracture strains, and a
comparison with Fig. 17, p. 24, shows that the modified alloys irradi-
ated to 5.3 X 10°° neutrons/ecm® have better fracture strains than those
of the standard alloy irradiated to 1.3 x 10%° neutrons/ém?. One test
indicates that the fracture strain of heat 67-504 at 25°C may be reduced
by irradiation. This will have to be checked further.

The fracture strains observed in tensile tests at a strain rate of
0.002 min~! for heat 67-502 and -504 are shown in Fig. 30, The postirra-
diation properties of the two heats are about the same and, although
reduced considerably by irradiation, they are much better than those

shown in Fig. 18, p. 25, for standard Hastelloy N.
35

ORNL—DWG 69— 4477

 

70

 

60

 

 

50 S:\\__t_“/Q///\A %
=y

/ o \
__7/// N,
’A
HEAT IRRADIATED UNIRRADIATED ®
67-502 ® ©
C

N\

 

 

o
o

 

 

 

 

FRACTURE STRAIN (%)

"

 

 

20 &67-504 A A %

M,

*° R

2 100 200 300 400 500 600 70C 8C
TEST TEMPERATURE (°C)

 

 

 

 

 

 

 

 

 

 

 

 

900

Fig. 29. Tensile Properties (Strain Rate of 0.05 min"l) of Two
Compositions of Modified Hastellog N Following Irradiation (Thermal
Fluence of 5,3 x 10°° neutrons/em®) or Annealing (Control Samples) for
9789 hr at 650°C.

ORNL-DWG 69-4478

 

60 i

 

50 -

 

 

b
o

 

w
o

  

 

 

 

 
 
 

HEAT IRRADIATED UNIRRADIATED
67-502 e o

67-504 A N T : : \\

 

 

FRACTURE STRAIN (%)

n
o

 

 

 

 

 

 

 

 

 

 

 

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

‘Fig. 30. Tensile Properties (Strain Rate of 0.002 min™') of Two
Compositions of Modified Hastelloy N Following Irradiation (Thermal
Fluence of 5.3 x 10°° neutrons/cm®) or Annealing (Control Samples) for
9789 hr at 650°C,
36

Creep-Rupture Properties

 

The results of creep-rupture tests on heats 67-502 and -504 are
given in Tables A-14 and A-15 (Appendix). The stress-rupture properties
of heat 67-502 are shown in Fig. 31. Aging for 9789 hr at 650°C
increased the rupture life by about 30% and neutron irradiation to a
fluence of 5.3 x 1070 neu.‘trons/cm2 decreased the rupture life by about
30% compared with the as-annealed material, The minimum creep rate,
shown in Fig., 32, was increased by the aging treatment and the samples
that were aged and irradiated simultaneously had the same minimum creep
rates as the as-annealed material.

The creep-rupture strength of heat 67-504 (Fig. 33) was not altered
appreciably by aging at 650°C, However, irradiation decreased the rup-
ture life of the material by about 50%. The minimum creep rate (Fig. 34)
was increased by the combined irradiation and aging treatment, but was
increased even more by Jjust the aging treatment.

The fracture strains of heats 67-502 and -504 at 650°C are shown
in Fig. 35 as a function of strain rate. In the unirradiated condition,
both heats show a trend of decreasing fracture strain with decreasing
strain rate. Aging improves the fracture strain of both heats and irra-
diation decreases the fracture strain, The irradiated material has a

fracture strain of about 6% under creep conditions.

Metallographic Examination of Mechanical Property Specimens

Several of the tested samples were examined metallographically.
Photomicrographs are shown in Fig. 36 for a sample of heat 5085 that was
tensile tested at 25°C after being exposed to the cell environment for
20,789 hr at 650°C. The fracture under these test conditions would
normally be transgranular, but the fracture shown in Fig. 36 is primarily
intergranular. This change in fracture mode probably accounts for the
fracture strain of this sample being only 33% (Fig. 11, p. 20) compared
with about 55% for unirradiated, unaged material. The photomicrographs
shown in Fig. 37 were made on another sample of heat 5085 from outside
the vessel that was tested at 650°C. The fracture is largely intergran-

ular and there are very few cracks near the fracture,
37

CRNL-DWG 65-4479

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

70 —T O & f
| --\'-:\\ !
N ;
\r.\ ;
60
\QQ\A
\R\E\\\
50 \\ \\.\\
N n
2 *\\‘Q\\
= \Q\‘\
g RS
O "~.\
" o
fl 30
a
|_
W
20 o ANNEALED 4 hr AT 1177°C 1
& ANNEALED, AGED 9,789 hr AT 650°C ‘ !
0 ANNEALED, IRRADIATED AT 650°C TO I
10 L A THERMAL FLUENCE OF
5.3 X 1020 neutrons/cm2
OVER 9789 hr
o Lo e
107" 10° 10" 102 108 10?

RUPTURE TIME {hr)

Fig. 31. Stress-Rupture Properties of Heat 67-502 at 650°C.

ORNL-DWG 69-4480

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

70 T T ’ — #g
I ' l | , 4'/" | ;
1 i P .
, \ /,.v //
60 ! :
| 1 :
i BA/( | i
L (iilVe ‘ |
E o A | |
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o 40 T e ; Tt
e o /’ i !
9 | : / /’y | | .
— ! ‘ / /A l !
® A | .
g % i | T
o | | ; . [
n L o ANNEALED f hr AT 1177°C
20 il & ANNEALED, AGED 9789 hr AT 650°C i
1 o ANNEALED, IRRADIATED AT 650°C TO A
D THERMAL FLUENCE OF 5.3 x1029 neutrons/ecm?
! OVER 9789 hr
0 it = -
R !
| | o
o 11— L1
10™4 1072 1072 T 10° 10'

MINIMUM CREEP RATE (%/hr)

Fig. 32. Variation of Minimum Creep Rate with Stress Ievel for
Heat 67-502 at 650°C.
38

ORNL-DWG 69-44814

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

70 o A
N"“h.._'h
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60 | w
§
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50 <
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o 40 } TS
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8 {E : \\\
g 30 i | \:\‘
& [l !
= o ANNEALED { hr AT {177°C
5 ANNEALED, AGED 9789 hr AT 650°C
20 O ANNEALED, IRRADIATED AT 650°C TO A
THERMAL FLUENCE OF 5.3x1020
neutrons /cm2 OVER 9789 hr
10 | || | |
o ]
Lo ! ,
| L
0 1 1
10~ 1 10° to! 102 103 104

RUPTURE TIME (hr)}

Fig. 33. Stress-Rupture Properties of Heat 67-504 at 650°C.

ORNL-—-DWG 69-4482

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

70 , — —— - , —
. ; . Lo i D P //‘— //f 4
// // |
60 : 4t ] P
o
5O [ e i : '
8 ;
2 ‘
3 a0
2 ;e ' ]
530 e {1
@ | o “
‘ o o ANNEALED fhr AT 4477 °C
20— -~ 17| 771 & ANNEALED, AGED 9789 hr AT 650°C [ ][]
e | o ANNEALED , IRRADIATED AT 650°C TO| |||
1o b L] .. A THERMAL FLUENCE OF 53x10%°
o 1|, neutrons /cm?® OVER 9789 hr
o |
i L | ![J 1

 

1Ll

10

 

C L

0
1074 1072 +072 107!
MINIMUM CREEP RATE {%/hr)

 

L
10°

Fig. 34. Variation of Minimum Creep Rate with Stress Level for
Heat 67-504 at 650°C.
39

ORNL-DWG 69-4483

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

52 T - T T T
T T o I [ ] | 3\ { | 1 ! l I | ; ) ”
HEAT  ANNEALED AGED IRRADIATED | A
67 -502 0 a o N LT e
48 67-504 . A n > ) AT
; i' s » —1 L7
! " A
| ' \’ z‘! _,/ //
44 ‘ l ) /’d“ ' fig? T
T T
| ‘ AT / A
; ‘ 4 " // .
40 e e LA 4 -
‘ . pr / .
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36 E Dt | : -
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10 / _ //
i /
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12 / / T
// f"/ i
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u
. ol ® | ® A7 CREEP TESTS =—|~=TENSILE TESTS
o
4 o
0
1073 1072 107! 1o° ) 102

STRAIN RATE (%/hr)

Fig. 35. Comparison of the Fracture Strains of Two Heats of
Modified Hastelloy N Tested at 650°C in Various Metallurgical Conditions.
 

¥

 

 

  
  

40

R-47891

T R-47892

 

 

 

 

 

 

 

 

 

 

Fig. 36. Photomicrographs of a Hastelloy N (Heat 5085) Sample
Tested at 25°C at a Strain Rate of 0.05 min~?1. Sa.mgle had been irradi=-
ated to a thermal fluence of 2.6 X 10'° neutrons /em* while being exposed
to the cell environment at 650°C for 20,789 hr. 100x. (a) Fracture,
as polished. (b) Fracture, etched. (c) Edge of stressed portion. |
Etchant: aqua regia. ' ' - '
%)

#

41

R-47897

R-47898

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig. 3'7. Photomicrographs of a Hastelloy N (Heat 5085) Sample
Tested at 650°C at a Strain Rate of O. 002 min~}, Sample had been irra-
diated to a thermal fluence of 2.6 X 101? neutrons/em® while being
exposed to the cell environment at 650°C for 20,789 hr, 100x:

(a) Fracture, as polished. (b) Fracture, etched (c) Edge of stressed
portion, etched, Etchant: aqua regia.
 

 

42

A sample of heat 5065 that was removed from outside the veséel'and
tested at 25°C is shown in Fig. 38. The fracture is still partially
intergranular, but the elongated grains indicate that the sample strained
a large amount before failing. As shown in Fig. 12, p. 20, this sample
had a fracture strain of 59.5%. . A sample of heat 5065 was tested at
600°C and typical phdtomicrographs'are shown in Fig. 39. The fracture
strain of this sample was only lf% and the fracture is predominately
intergranular. . |

The thin oxide surface layers do not seem to influence the deforma-
tion in the samples from.outside the vessel., The amount of edge cracking
noted in Figs. 36 through 39 is not unusual for Hastelloy N.

, _Typical_photomicrographs of a sample of heat 5085 aged for 15,289 hr
at 650°C in staticrbarfen fuel salt and tested at 200°C are shown in
Fig. 40. The fracture strain of this sample was quite good (Fig. 14,
p. 22) although the fracture is partiaelly intergranular. A cross section
from the stressed portion of this sample is shown in Fig. 40(b) and shows
that the surface structural modification extends to a depth of about
3 mils. Figure 40(c) gives some further insight into the origin of the
modified structure. The surface indentations are engraved identifica-
tion numbers and would have involved heavy cold working. Note that the
modified structure extends to a greater depth under these impressions.
The febrication sequence was (1) rods fabricated to approximate diameter,
(2) rod segments welded together, (3) annealed for 2 hr at 900°C in
argon, (4) sinterless ground to final diameter, (5) gage portions milled,
(6) identification numbers engraved, and (7) samples put into surveillance
facility. Thus, all surfaces were cold worked; the major diameter by
grinding and the gage diameter by milling., The resulting miéroétructure
likely results from carbide précipitation on the dislocations introduced
by cold working. |

Figure 4lrshowé several photomicrographs of a sample of heat 5085
that was irradiated in the MSRE core to a thermal fluence of
9.4 X 10%° neutrons/em® and tested at 200°C. The fracture is mixed -
transgranular and intergranular. The grain shapes attest to the lower
elongation of the irradiated sample in Fig. 41(c) compared with that of
the unirradiated sample shown in Fig. 40(a). The asdpolished
43

R-47883

  

 

o'

(a) ' v ______________ ke e il o i e ‘

 

 

 

 

 

 

m R-47887

  

(c)

~ Fig. 38. FPhotomicrographs of a Hastelloy N (Heat 5065) Sample
) Tested at 25°C at a Strain Rate of 0.05 min~!, Sample had been irradi-
ated to a thermal fluence of 2.6 X 10'? neutrons/cm® while being exposed
to the cell enviromment at 650°C for 20,789 hr. 100X. (a) Fracture,
o as polished. (b) Fracture, etched. (c) Edge of stressed portion.
W | Etchant: aqua regia. o | |
 

 

 

 

 

 

 

 

 

 

 

exposed to the cell environment at 650°C for 20,789 hr. 2100x.

 

portion, etched. Etchant: aqua regia.

R-47858

 

Fig. 39. Photomicrographs of a Hastelloy N (Heat 5065) Sample
Tested at 600°C at a Strain Rate of O. 002 min~1, a.m;gle had been irra-
diated to a thermal fluence of 2.6 x 101° neutrons/bm while being

(a) Fracture, as polished. (b) Fracture, etched. (e) Edge of stressed
 

 

45

 

   

B Y-92953

 

   

 

, Fig. 40. Photomicrographs of a Hastelloy N (Heat 5085) Sample
 Exposed to Static Barren Fuel Salt for 15,289 hr-at 650°C and Then
. |  Tested at 200°C, (a) Fracture. 100x. (b) Edge of cross section from
' . stressed portion, 500x. (c¢) Edge of cross section from shoulder showing
' impressions made by engraving numbers. 100x. Etchant: glyceria regia.
“ ~- Co T
 

 

 

R-47923

 

 

 

 

 

 

 

 

 

 

 

 

 

 

R-47925

"o

(d) R ,*tz-«.-:..;i'iz . et

 

Fig. 41. Photomicrographs of a Hastelloy N (Heat 5085) Sample
Exposed to Fluoride Salt in the MSRE for 15,289 hr at 650°C_and Then
Tested at 25°C. Thermal Fluence was 9.4 X 102C neutrons/em®. (a) Frac-
ture, as polished. 100x. (b) Fracture, as polished. 500x. (c) Frac-

ture, etched. 100x. (d) Edge of stressed portion, etched. 100x.
Etchant: aqus regia. : »
47

photomicrographs [Fig. 41(a) and (b)] show that the large carbides
actually fracture during testing. Some of these cracks may propagate
and contribute to fracture. (This fracturing occurs in unirradiated and
irradiated samples when tested at relatively low temperatures, e.g.,
Figs. 36, 38, and 40.) There is a very thin layer of the modified struc-
ture near the surface of the tested sample [Fig. 41(d)], but the amount
of edge cracking is not unusually high for a sample that has strained
36% (Fig. 13, p. 22).

The fracture of a sample of heat 5085 that was tested at 650°C
after exposure to a static barren fuel salt for 15,289 hr at 650°C is
shown in Fig. 42. The fracture is mixed transgranular and intergranular
and there is considerable evidence of carbide fracturing. By contrast,
the fracture shown in Fig. 43 is primarily intergranular. The sample in
Fig. 43 was irradiated to a thermal fluence of 2.4 x 10°° neutrons/cm2
and tested at 650°C. At the low fracture strain of only 5% (Fig. 13,
p. 22) there is little evidence of plastic deformation with only a few
fractured carbides and some separations at the carbide-metal interfaces
[Fig. 43(b)].

Samples of heat 5065 that were irradiated to a thermal fluence of
2.4 x 10%° neu.trons/cm2 and the thermal controls both had fracture
strains of over 40% when tested at 25°C (Figs. 15 and 16, p. 23). The
fractures of these samples appear quite similar (Figs. 44 and 45). The
fractures are largely intergranular although the grains have deformed
extensively. ©Some edge cracking occurred on both samples,

When tested at 650°C an unirradiated sample of heat 5065 had a
fracture with transgranular and intergranular components (Fig. 46).
This sample had a fracture strain of about 17% (Fig. 16, p. 23) compared
with 3.5% (Fig. 15, p. 23) for a comparable sample irradiated to a
thermal fluence of 9.4 x 1020 neutrons/cmz. The irradiated sample has
an intergranular fracture with very limited intergranular and carbide
cracking (Fig. 47). The modified layer near the surface has not deformed
differently from the remaining sample. The fracture of a sample modified
with titanium (heat 67-502) that was exposed to barren fuel salt for
9789 hr at 650°C and tested at 25°C is shown in Fig. 48. The grains have
deformed extensively although the fracture is largely intergranular.
 

e i g

Y-92913

 

 

1Y-92914

 

' Fig. 42. Photomicrographs of a Hastelloy N (Heat 5085) Sample
Exposed to Static Barren Fuel Salt for 15,289 hr at 650°C and Then .
Tested at 650°C and a Strain Rate of 0.002 min~!. 100x. (a) Fracture,
as polished. (b) Fracture, etched. Etchant: glyceria regis.
’\

 

;.

)

 

Fig. 43, Photomicrographs of a Hastelloy N (Heat 5085) Sample
Exposed to Fluoride Salt in the MSRE for 15,289 hr at 650°C and Then .
Tested at 650°C and a Strain Rate of 0.002 min~!, Thermal fluence was
9.4 x 100 neutrons/cm®. (a) Fracture, as polished. 100x. (b) Frac-
ture, as polished. 500x. (c) Fracture, etched. 100x. (d) Edge of

stressed portion, etched. 100x. Etchant: aqua regia.” Reduced 14.5%.
 

50

  

Y-92917

(2) |

 

 

 

 

Fig. 44. Photomicrographs of a Hastelloy N (Heat 5065) Sample
Exposed to Static Barren Fuel Salt for 15,289 hr at 650°C and Then Tested
at 25°C end a Strain Rate of 0.05 min~!, 100xX. (a) Fracture, as polished.
(b) Fracture, etched. (c) Edge of stressed portion, etched. Etchant:
glyceria regia. - ) :
 

 

 

 

 

 

 

3

 

 

 

R-47905

Fig. 45. PhotomicrOgraphs of a Hastelloy N (Heat 5065) Sample

Exposed to Fluoride Salt in the MSRE for 15,289 hr at 650°C and Then
Tested: at 25°C and a Strain Rate of 0.05 min "l Thermal fluence was

9.4 x 10°° neutrons/em®. '100x. (a) Fracture, as polished. (b) Fracture,
etched. (c) Edge of stressed portion. Etchant: aqua regia.
 

M e

 

52

 

 

 

 

 

 

 

 

 

 

 

 

 

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Fig. 47. Photomlcrbgraphs of a Hastelloy N (HEat 5065) Sample

‘Exposed to Fluoride Salt in the ‘MSRE for 15,289 hr at 650°C and Then

Tested at 650°C and & Strain Rate of 0.002 min "1, Thermal fluence was
9.4 x 10?0 neutrons/em?. - 100X. - (a) Fracture, as polished. (b) Fracture,
etched, (c) Edge of stressed portion, etched. Etchant: aqua regia.
 

 

 

 

 

Y-92925

 

 

Y-92926

 

 

 

Fig. 48. Photomicrographs of a Hastelloy N (Heat 67-502) Sample
Exposed to Static Barren Fuel Salt for 9789 hr at 650°C and Then Tested
at 25°C. 100X. (a) Fracture, as polished. (b) Fracture, etched.
Etchant: glyceria regia.

0
55

Further photomicrographs of the unstressed portion of this same sample
are shown in Fig. 49. There is a very fine carbide precipitate through-
out this sample and our other studies indicate that these are likely
carbides of the MyC or MC types.lo The grain boundaries also etch quite
heavily indicating the presence of heavy grain-boundary precipitation.
The surface of the sample has a very thin layer that is rich in iron.
This is likely a deposit since this alloy is very low in iron and the
rest of the system is fabricated of material that contains 4 to 5% Fe.

A sample of heat 67-502 that was irradiated and tested at 25°C is
shown in Fig, 50, The fracture strain of this sample was about as high
as that for the unirradiated sample (Fig. 29, p. 35), but the grains are
not nearly as elongated in the irradiated sample (Fig. 50) as in the
unirradiated sample (Fig. 48). There are considerably more grain-
boundary cracks in the irradiated sample than in the unirradiated sample.

The fracture of a sample of heat 67-502 that was exposed to barren
fuel salt for 9789 hr and tested at 650°C is shown in Fig. 51. The
fracture is transgranular, an observation consistent with the extremely
high fracture strain of this sample (Fig. 29, p. 35).

The fracture at 25°C of heat 67-504 (Fig. 52) is quite similar to
that noted for heat 67-502 (Fig. 48). Heat 67-504 is modified with
0.5% Hf and the photomicrographs in Fig. 53 indicate that there may be
more precipitate in this alloy than in heat 67-502 that was modifed with
0.5% Ti and 2% W, However, the apparent amount of precipitation is very
dependent upon the etching procedure, This heat of material is also
very low in iron and iron deposition was noted [Fig. 53(a)l.

The fracture of an irradiated sample of heat 67-504 tested at 25°C
is shown in Fig. 54. The grains have deformed and the fracture was
transgranular. As shown in Fig. 29, p. 35, the fracture strain at 25°C
of heat 67-504 was lower than that of 67-502. However, the fractures
(Figs. 50 and 54) give the opposite impression,

 

'0R. E. Gelbach and S. W. Cook, MSR Program Semiann. Progr. Rept.

Feb. 28, 1969, ORNL-4396, pp. 240-242.

 
 

 

56

Y-92927

 

Y-92928

 

 

 

 

 

 

 

 

 

 

Y-92929

   
 

 

Fig. 49. Photomicrographs of the Unstressed Portion of a Hastelloy N
(Heat 67-502) Sample That Was Exposed to Static Barren Fuel Salt for
9789 hr at 650°C. (a) Typical. 100x. (b) Typical. 500x. (c¢) Edge.
500x. Etchant: glyceria regia. - o

W
 

 

 

57

 

 

 

 

 

' Fig. 50. Photomicrographs of & Hastelloy N (Heat 67-502) Sample
Exposed to Fluoride Salt in the MSRE for 9789 hr at 650°C and Then Tested
‘at 25°C. 100X, (a) Fracture, as polished. (b) Fracture, etched.

(¢) Edge of stressed portion, etched. Etchant: agua regia.
 

58

B Y -92930

 

 

(a)

 

 

Y-92931

 

Fig. 51. Photomicrographs of the Fracture of a Hastelloy N
(Heat 67-502) Sample Exposed to Static Barren Fuel Salt for 9789 hr at
650°C and Then Tested at 650°C and a Strain Rate of 0.002 min~%, 100,
(a) As polished. (b) Etched. Etchant: glyceria regia.

4
 

 

 

 

-
3

 

 

 

jijig. 52;_ Photbmicfbgrgpfiéfof the Fracture of a Hastelloy N

(Heat 67-504) Sample That Was Exposed to Barren Fuel Salt for 9789 hr at

650°C and Tested at 25°C.- 100X, (a) As polished. (b) Etched. Etchant:

glyceria regia. L R |
 

 

 

 

 

 

 

60

Y-92939 | N\

‘N

4

 

 

 

 

 

 

 

 

 

 

e

b B S AT B Pty
3 'Fr?i"«ff . ‘\‘f\ o
Gl .(1;‘}: ! i~

 

 

 

. Fig. 53, Photomicrographs of the Unstressed Portions of a Hastelloy N
(Heat 67-504) Sample Exposed to Barren Fuel Salt for 9789 hr at 650°C.
(a) Edge, as polished. 1000x. (b) Typical microstructure, etched. 2100X.
Etchant: glyceria regia.
 

 

 

 

 

 

 

61

 

R-47929

 

©@F

Fig. 54. Photomlcrographs ‘of a Hastelloy N (Heat 67-504) Sa.mple |
Exposed to Fluoride Salt in the MSRE for 9789 hr at 650°C and Tested at
25°C, 100X, (a) Fracture, as polished. (b) Fracture, etched. (c) Edge
of stressed portion, etched. Etchant: aqua regia. Reduced 10%.
62

Heat 67-504 also failed by a predominately transgranular mode when
tested at 650°C (Fig. 55). This observation_cofreletes quite well with
the high fracture strain of this sample;(Fig.”295 p. 35).

DiScuSSion of Results

The samples of standard Hastelloy N removed from the survelllance
fac111t1es inside the MSRE core and outside the reactor vessel continue
to show excellent compatibility with their respective enviromments of
fuel Salt_and N2 + 2 to 5% 02, We couid not detect the diffusion of fis-
sion products into the samples from the core. The sampies exposed to
the cell environment showed no efiidence of nitriding and only superficial
oxidation. ) o J

We have removed several groups of samples from the MSRE (Table 2, p, 6)
and some valuable observations have been made regarding the changes of
mechanical pfoPerties_with fluenee*and thermal expoeure. The thermal
flux in the center of the core is 40_timesrthat at the vessel wall. Only
the Hastelloy N thimbles for the control rods have been exposed to the
peak flux and have accumulated a'thermal neutron fluence of about
11X ;Ozl‘neutrons/em?. The tensile (Figs. 17 and 18, pp. 24 and'25) and
the creep (Figs. 21, 23, 25, and 27, pp. 27, 29, 31, and 32) properties
of Hastelloy N exposed to such a fluence are very poor; but the thimbles
are stressed only slightly'in compression and the observed property
changes are not catastrophic.

~ The peak vessel thermal fluence is presently (May 1968)
2.6 x 101° neutrons/em?, but the properties have already been altered
markedly, The results that we obtained at higher fluences can be used
to estimate the future properties of the MSEE vessel. Fiéure 56 has been
constructed for heat 5085 and shows how the fractfireusfiréih at different
strain rates varies with thermal fluence at 650°C., The fraction of the
. strain remaining is based on the number shown in parenthesis Wthh is
the value measured in the unirradiated starting material after a 2-hr
anneal at 900°C. The corresponding curve for the burnup of 10p is also
 

63

[ Y-92942

 

 

 

 

Fig. 55. Photomicrographs of the Fracture of a Hastelloy N
(Heat 67-504) Sample Exposed to Static Barren Fuel Salt for 9789 hr at
650°C and Tested at 650°C and a Strain rate of 0,002 min~l, = 100x.

(a) As polished. (b) Etched. Etchant: glyceria regia.

al

 

 

&

 

 
ORNL-DWG 695-4460

 

 

 

 

 

 

 

 

2
102! ® ‘o\\ ‘o
N\
N\
TNy
5

 

 

 

 

\
\
2 \ \\ N

o e b —{ 362,000

 

 

 

 

 

THERMAL FLUENCE (neutrons/cm?2)

 

 

 

 

 

 

 

 

 

 

 

 

20 \ \ —_
10 \ \ \ =
1\ \ L
| - \ 2
\ \ \ ~
. \ \ \ o
\ \ \ =
<
\L\ \ o
S
. . \ - 72,000 32
e
2 | &
é=0.4% /hr é=12% /hr ¥=300%/hr -
Ll
e ¢ l . 36,200 T
1o1s L(30) (30) (33.5) |
0.2 0.4 06 0.8 10

FRACTION REMAINING

Fig. 56. Change in the Fracture Strain of Hastelloy N (Heat 5085)
with Thermal Fluence for Various Strain Rates.
65

shown for comparison (based on a cross section of 1930 barns for MSRE
spectrum’ ),

The strain rate of 300%/hr (0,05 min~!) is that normally used in
tensile tests and the vessel heats presently, after 72,400 Mwhr of oper-
ation, will strain about 20% at 650°C before failure. The fracture
strain decreases with decreasing strain rate to a minimum value of 1.5%
at a strain rate of about O.l%/hr. These property changes are gquite
large when we consider that only 4% of the 1B has been converted to
helium (accumulated thermal fluence =~ 2.6 x 1017 neutrons/cmz). However,
the results from samples exposed to higher fluences in the core indicate
that the property changes are not simply proportional to the quantity of
10B transmuted to helium. Although the intended operating time for the
MSRE has not been established, it likely will not exceed 362,000 Mwhr
(five times the energy already generated). Figure 56 indicates that the
mechanical properties should not worsen drastically from those that pre-
vail after 72,400 Mwhr of operation.

The main uncertainty associated with our estimate of the future
properties of the MSRE vessel is that we have not taken into account the
property changes due to thermal aging. As shown in Figs. 19 and 21,

Pp. 25 and 27, the tensile and creep properties of heat 5085 were changed
just by annealing at 650°C. However, heat 5065 (Fig. 25, p. 31) seemed
to be more resistant to aging. The noted changes in the fracture strain
at lower temperatures could also conceivably become a problem. These
changes also vary from heat to heat and depend on both fluence and aging
time. Thus, the effects of aging cannot be taken into account adequately
to improve on the property estimates made on the basis of Fig. 56.

The postirradiation properties of the two modified alloys (heats
67-502 and -504) were better than those of the standard alloys (compare
Figs 28 and 35, pp. 33 and 39). We cannot explain the property changes
that occurred as a result of aging these alloys in static barren salt,

but our current aging studies should reveal the mechanism.

 

11p, E. Prince, MSR Program Semiann, Progr. Rept. Aug. 31, 1967,
ORNL-4191, p. 58.
66

Metallographic examination showed that the standard Hastelloy N was
very compatible with the fuel salt. The samples removed from outside
the vessel were oxidized but there was no evidence of nitrogen absorption.
The modified alloys were exposed to the fuel salt and an iron-rich layer
was formed on the surface, This transfer occurred because the modified
alloys contain less than 0.1% Fe and the standard Hastelloy N used in
the rest of the system contains 4 to 5% Fe.

SUMMARY AND CONCLUSIONS

We have examined the third group of surveillance samples removed
from the MSRE. The materials involved were (1) standard Hastelloy N
exposed in the MSRE cell to an enviromment of N, + 2 to 5% O, for
20,789 hr at 650°C to a thermal fluence of 2.6 X 109 neutrons/bm2,

(2) standard Hastelloy N exposed in the MSRE core to fluoride salt for
15,289 hr at 650°C to a thermal fluence of 9.4 x 10°C neutrons/em®, and
(3) modified Hastelloy N exposed in the MSRE to fluoride salt for 9789 hr
at 650°C to a thermal fluence of 5,3 X 10°° neutrons/cm®. Postirradia-
tion creep and tensile tests were run to evaluate the influence of the
respective exposures on the mechanical properties. The property change
of most concern is the reduction in the fracture strain. The fracture
strain at 25°C was reduced by an amount related to both the neutron flu-
ence and the exposure time. The fracture strain could be restored by a
postirradiation anneal of & hr at 870°C and we propose that the reduction
in fracture strain at 25°C is due to carbide precipitation. The fracture
strain in tensile tests above 500°C and in creep tests at 650°C was
reduced by irradiation. This embrittlement is associated with the helium
that is produced by the transmutation of 10B in the alloy and cannot be
recovered by postirradiation annealing. We compared the results from
samples irradiated to thermal fluences ranging from 1.3 X 1019 to

9.4 x 1020 neu,trons/cm2 and have an approximate picture of how the frac-
ture strain varies with fluence. This picture is complicated by the fact
that the properties of the unirradiated material change some with aging
at 650°C, and we do not have sufficient data to separate the contribu-

tions of irradiation and aging to the observed embrittlement.
67

The modified alloys had better postirradiation properties than were
measured for the standard Hastelloy N. Although the properties of these
materials were affected appreciably by irradiation, the minimum fracture

strain observed in postirradiation creep tests at 650°C was 6%.

ACKNOWLEDGMENTS

The author is indebted to numerous persons for assistance in this
study.

W. H, Cook and A. Taboada — Design of surveillance assembly and insertion

of specimens.

W, H., Cook and R, C. Steffy — Flux measurements.

J. R. Weir, Jr., R. E. Gehlbach, and C, E. Sessions — Review of manuscript.

E. J. Lawrence and J, L, Griffith — Assembled surveillance and control

specimens in fixture.

P. Haubenreich and MSRE Operation Staff — Exercised extreme care in
inserting and removing the
surveillance specimens,

E. M. King and Hot Cell Operation Staff — Developed techniques for cutting
long rods into individual speci-
mens, determined specimen
straightness, and offered assis-
tance in running creep and
tensile tests.,

B. C. Williams, B, McNabb, and N, O, Pleasant — Ran tensile and creep

tests on surveillance and
control specimens.
E. M, Thomas and J. Feltner — Processed test data.
H. R. Tinch and E, Iee — Metallography on control and surveillance
specimens.
Metals and Ceramics Reports Office — Preparation of manuscript.

Graphic Arts — Preparation of drawings.
APPENDIX
Table A-1. Postirradiation Tensile Properties of Samples of Heat 5085>

 

 

 

Specimen Test Strain Stress, psi Strain, % Reduction True
Number Temperature Rate in Area Strain
(°c) (min™1)  Yield  Ultimate Uniform Total (%) (%)
7976, 25 0.05 46,500 99,100 32.8 32.8 24,51 28.0
7982 25 0.05 46,700 111,900 48,2 48,3 34.19 42,0
7975 200 0.05 38,000 94,600 40,0 41,2 32.11 39.0
797, 400 0.05 3,700 90,800 46,8 47.2 36.46 45.0
7972 500 0.05 33,500 86,700 43.2 4 ,O 36.13 45.0
7970 550 0.05 32,000 76,000 29.5 30.2 30.89 37.0
7968 600 0,05 31,700 68,000 21.0 21.3 24,63 28.0
7966 650 0.05 30,400 60,900 8.8 19,7 14.82 16.0
7964 700 0.05 30,000 54,000 15.7 15.9 8. 64 9.0
7962 760 0.05 30,200 43,800 8.3 g.9 12.30 13.0
7960 850 0.05 33,000 37,500 3.2 3.5 3.98 4.0
7973 400 0.002 35,000 91,500 47,3 47.8 37.20 47.0
7989 500 0.002 37,200 79,000 27.5 28,7 29.55 35,0
7969 550 0.002 33,500 67,400 19.5 20.6 18,94 21.0
7967 600 0,002 31,500 58,000 15.2 15.7 19,07 21.0
7965 650 0.002 31,300 49,900 11.1 11.6 18.61 21.0
7963 700 0.002 30,000 41,900 6.7 7.3 10.02 11.0
7961 760 0.002 28,600 37,000 bty 5.0 8.31 9.0
7981 850 0.002 21,700 21,700 1.0 1.5 4 .91 5.0
7988 500 0.0005 34,300 68,800 20,4 21.1 19.33 21.0
7987 550 0.0005 33,500 64,100 19.0 19.6 18.48 20.0
7986 600 0, 0005 31,800 54,100 13.3 13,7 16,76 18.0
7985 650 0.0005 31,300 46,100 8.5 9.3 6.95 7.0
7984 700 0.0005 30,800 41,100 5.0 5.2 6,45 7.0

 

%1ocated outside the MSRE vessel for 20,789 hr at 650°C. Thermal fluence was
2.6 x 1017 neutrons/cmz.

bGiven a postirradiation anneal of & hr at 870°C,

T
Table A-2. Postirradiation Tensile Properties of Samples of Heat 5065

 

 

 

 

g . Test Strain Stress, psi Strain, % Reduction True
pecimen - .
Number Temperature Ratg in Area Strain
(°C) (min™1)  Yield  Ultimate Uniform Total (%) (%)
7940 25 0.05 49,000 118,800 57.8 59.7 38.44 49,0
7939 200 0.05 40,500 105,200 55.1 57.6 40, 54 52.0
7938 400 0.05 26,600 99,600 55.0 56,3 43,27 57.0
7936 500 0.05 45,400 9%, 800 46,6 47,2 40,83 52.0
7934 550 0.05 36,000 83,800 31.8 32.0 26,83 31.0
7946 600 0.05 33,100 79,800 33.3 33.9 31.33 38.0
7930 650 0.05 30,600 55,400 11.7 11.8 11.69 12,0
7928 700 0.05 31,100 57,200 13.5 13.6 16.83 18.0
7926 760 0.05 29,600 33,700 2.2 3.2 8.17 2.0
7924 850 0,05 32,700 36,800 3.2 3.4 3.99 4.0
7935 500 0.002 36,200 85,200 25.7 26,2 22,89 26,0
7956 500 0.002 38,100 86,900 29.5 30.2 24, 14 28.0
7933 550 0,002 35,600 78,400 26,4 26,7 24,99 29.0
7931 600 0.002 34,600 63,400 14,9 15.1 8.49 9.0
79477 650 0.002 34,100 55,500 12.2 12.5 16.09 18.0
7948 700 0.002 33,500 49,900 8.6 8.7 g8.31 2.0
7925 760 0.002 33,300 40,000 3.5 5.0 12.39 13.0
7945 850 0.002 22,0600 23,100 1.0 2.0 2.86 3.0
7955 400 0.0005 38,200 100,700 52.8 54,2 37.37 47.0
7954 500 0.0005 36,400 80,100 28.3 28.9 25.11 29.0
7953 550 0,0005 37,500 70,200 20.3 20,9 19.59 22.0
7952 600 0.0005 27,800 61,100 16,7 17.0 18.74 21.0
7929 650 0.0005 33,700 52,700 8.8 9.1 16.02 17.0
7950 700 0.0005 32,800 46,300 4.8 6.2 5.69 6.0

 

*Located outside the MSRE vessel for 20,789 hr at 650°C. Thermal fluence was
2.6 x 1017 neutrons /em?,

(48
Table A-3, Tensile Properties of Samples of Heat 5085>

 

 

 

Specimen Test Strain Stress, psi Strain, % Reduction True
Numbe T Temp%rature R?tgl : : : in Area Strain
(°c) (min™")  Yield Ultimate Uniform Total (%) (%)
7888 25 0.05 52,300 95,000 28,7 28.9 19.95 22.0
7889 200 0.05 43,700 92,200 36.1 36.7 30.00 36.0
7890 400 0.05 42,800 84,300 31.9 32.1 31.11 37.0
7892 500 0.05 39,600 81,200 35.0 35.2 31.33 38,0
789 550 0.05 38,900 71,500 22.5 23.6 27,64 32.0
7881 600 0.05 38,100 58,700 11.1 11.8 18.80 21.0
7887 650 0.05 34,400 51,600 10.4 11.0 14.16 15.0
7879 700 0.05 35,900 46,200 5.8 6.3 10,70 11.0
7877 760 0.05 29,700 36,200 4.7 5.4 11.74 12.0
7875 850 0.05 30,400 31,300 1.8 2.2 6. 84 7.0
7891 400 0.002 42,000 86,400 34,6 34,8 30.11 36.0
7893 500 0.002 41,100 68,000 17.0 18.1 20,08 22.0
7895 550 0.002 44,300 64,900 12.4 13.3 16.02 17.0
7880 600 0.002 34,300 46,900 6.4 7.1 16,04 17.0
7886 650 0.002 35,000 42,400 4.5 5.0 13,14 14,0
7878 700 0.002 33,700 36,100 2.1 2.2 7.63 8.0
7876 760 0.002 31,000 31,900 1.8 2.2 4,69 5.0
7874 850 0,002 22,200 22,200 1.2 1.2 1.74 2.0

 

after exposure in the MSRE core for 15,289 hr at 650°C. Thermal fluence was
9.4 % 102° neutrons/em® and the fast fluence > 1.22 Mev was 2.3 X 1049 neutrons/bmz.

£l
 

 

 

 

Table A-4, Tensile Properties of Control Specimens of Heat 50852
Specimen Test Strain Stress, psi Strain, % Reduction True
Number Temperature Rate in Area Strain
( °c) (min™')  Yield  Ultimate Uniform Total (%) (%)
10166 25 0.05 53,900 115,900 38.4 38.6 29.66 35.0
10167 200 0.05 44,500 109,600 46,5 46,8 37,44 47.0
10165 400 0.05 39,600 102,700 49,7 50.2 39.08 50.0
10168 500 0.05 41,200 99,500 47,9 49.1 33,98 42.0
10169 550 0.05 38,600 91,400 39.3 39.5 33.20 40,0
10187 600 0.05 41,500 86,500 24,5 26,0 24,51 28.0
10170 650 0.05 38,000 70,500 18.8 19.6 20.89 23.0
10185 700 0.05 42,300 77,700 22.5 24.0 22.23 29.0
10172 760 0.05 35,800 68,600 24,8 31.3 22.23 29.0
10164 850 0.05 30,200 50,300 12.9 37.6 33.66 51.0
10183 400 0.002 41,700 107,300 48.8 49,3 36,46 45.0
10177 500 0.002 39,200 95,200 32.5 33.3 28.30 33.0
10184 550 0.002 39,600 83,500 25.6 26.6 19.20 21.0
10189 600 0.002 42,700 73,100 15,6 16.1 19,07 21.0
10190 650 0.002 37,700 64,700 17.4 18.0 19. 74 22.0
10173 700 0.002 42,400 68,000 15.8 16.8 19.87 22.0
10182 760 0,002 35,500 53,100 8.6 35.5 32.55 39.0
10174 850 0.002 31,600 31,600 1.3 34.1 35.48 44, O
10175 850 0.002 29,900 29,900 1.5 31.0 32.88 40.0
®pfter exposure to static fluoride salt for 15,289 hr at 650°C.

7L
Table A-5, Tensile Properties of Samples of Heat 5065%

 

 

 

Specimen Test Strain Stress, psi Strain, % Reduction True
Number Temperature Rate in Area Strain
(°c) (min~?)  Yield  Ultimate Uniform Total (%) (%)
7915 25 0.05 51,700 109,300 41,4 41.5 34,14 42.0
7916 200 0.05 62,300 102,400 38,9 39.6 35.87 44,0
7917 400 0.05 41,700 94,600 43.8 46,3 39.02 49,0
7919 500 0.05 44,100 90,600 37.0 38.0 34,63 43,0
7921 550 0.05 43,100 81,400 25.4 25,7 25,83 30.0
7908 600 0.05 38,600 62,100 11.7 11.9 13,28 14,0
7914 650 0.05 39,200 55,600 8.3 8.5 12,53 13.0
7906 700 0.05 36,200 46,800 5.4 5,5 6.75 7.0
7904 760 0.05 33,800 38,100 2.8 2.9 10.99 12.0
7900 850 0.05 34,000 34,000 1.4 1.5 5.18 5.0
7918 400 0.002 46,600 96,900 41,9 43,0 37.57 47.0
7920 500 0.002 44, 300 76.800 16.3 16.6 19.46 22.0
7922 550 0.002 42,400 66,300 12.0 12.4 18.15 20,0
7907 600 0,002 37,200 49,500 5.8 5.9 12.95 14.0
7913 650 0.002 40,400 46,300 3.2 3.4 5.95 6.0
7905 700 0.002 37,400 38,200 1.7 1.8 6.10 6.0
7903 760 0,002 33,200 34,700 1.0 1.0 9.40 9.0
7901 850 0.002 20,500 20,500 1.0 1.0 4,38 4,0

 

qafter exposure in MSRE core for 15,289 hr at 650°C. Thermal fluence was
9.4 x 10°° neutrons/em® and fast fluence > 1.22 Mev was 2.3 X 10°Y neutrons /cm®.

GL
Table A-6. Tensile Properties of Control Specimens of Heat 50652

 

 

 

. Test Strain Stress, psi Strain, % Reduction True
Specimen T £ Rat in A Strai
Number emperature ate in Area rain
(°c) (min~')  Yield  Ultimate Uniform Total (%) (%)
10215 25 0.05 60,900 126,700 46.5 47 .4 39.30 50.0
10214 200 0.05 46.800 114,300 47.2 49,4 37.07 46,0
10210 400 0.05 46,700 110,300 bt b 47 4 39.30 50.0
10194 500 0.05 41,200 103,700 45,7 46,5 36.77 46,0
10213 550 0.05 42,800 98,400 41,2 41,4 35.59 44,0
10217 600 0.05 40,700 848,300 23,2 24.3 20.86 23.0
10211 650 0.05 43,400 78,200 18.4 18.8 16.56 18.0
10191 700 0.05 39,600 71.400 17.0 17.6 19.07 21.0
10195 760 0.05 36,900 69,500 20.3 29.9 21.94 25.0
10196 850 0.05 35,300 46,900 8.6 44, 1 48,82 67.0
10208 400 0.002 50,700 118,500 43,4 46,5 38.34 48,0
10199 500 0.002 46,100 91,800 30.8 31.4 22.48 25.0
10200 550 0.002 46,500 84,100 18.6 19.1 17.75 20.0
10212 600 0.002 41,200 75,400 18.5 19.0 19.20 21.0
10216 650 0.002 4 , 200 73,300 16.0 16.5 16.83 18.0
10202 700 0.002 43,700 69,700 13.8 21.1 18.68 21.0
10198 760 0.002 41,700 50,400 7.5 34.8 36.87 46,0
10192 850 0.002 29,400 29,400 1.3 44,9 44,56 59,0

 

After exposure to static fluoride salt for 15,289 hr at 650°C.

9L
77

Table A-7. Postirradiation Creep-Rupture Properties at 650°C
of Samples of Standard Hastelloy N%

 

 

Test Specimen Stress Rupture Rupture Minimum
Homber fiumber Level Life Strain Creep Rate
(psi) (hr) (%) (%/hr)
Heat 5065
R=717 7941 52,000 3.4 2.3 0.525
R-736 7942 47,000 11.6 1.0 0.0792
R-725 7943 40,000 30.8 0,97 0.0283
R-728 7944 32,400 777 0.72 0.0062
R-780 7949 27,000 499,2 2.0 0.0012
Heat 5085
R=718 777 52,000 1.9 2.4 0.892
R-739 7978 47,000 4ol 1.9 0.400
R-724 7972 40,000 40,2 1.3 0.0209
R-729 7980 32,400 264 .0 1.2 0.0029
R-781 7983 27,000 451.5 1.6 0.0021

 

®ocated outside the MSRE Vessel for 20,789 hr at 650°C.
Thermal fluence was 2.6 X 10'° neutrons/cm®.

Table A-8. Creep-Rupture Properties at 650°C of Samples
of Standard Hastelloy N2

 

 

Te st Specimen Stress Rupture Rupture Minimum
Tber fium;er Tevel Life Strain Creep Rate
(psi) (hr) (%) (%/br)
Heat 5065
R-719 7909 47,000 0.3 0.12 0.20
R-726 7912 32,400 4,7 0.18 0.0214
R-721 7911 27,000 5.1 0.0% 0.0134
R-730 7910 21,500 246,3 0.28 0,.0010
R-783 7923 17,000 1940,2 0.83 0.0002
Heat 5085
R-720 7882 47,000 0
R-722 7884 32,400 12.6 0.52 0.0245
R-723 7885 27,000 50.6 0,81 0.0104
R-731 7883 21,500 217.8 1.1 0.0025
R-782 7873 17,000 2386.8 1.29 0.0005

 

Tafter exposure in the MSRE core for 15,289 hr at 650°C.
Thermal fluence was 9.4 x 10°° neu.trons/cm2 and the fast fluence
>1,22 Mev was 2.3 x 10%° neutrons/em?.
78

Table A-9. Creep-Rupture Properties at 650°C of Control
Specimens of Standard Hastelloy N2

 

 

Test Specimen Stress Rupture Rupture Minimum Reduction
Nonber fiumber level  Life  Strain Creep Rate in Area
(psi)  (nr) (%) (%/nr) (%)
Heat 5065
7240 10205 55,000 14.9 15.5 0.620 16.2
7239 10206 47,000  107.5 23.3 0.145 20.5
7238 10193 40,000 416.6 28.9 0.0426 24,8
7237 10209 32,400 1234.2 27,8 0.0131 32.9
Heat 5085
7236 10178 55,000 10.3 18.2 0.2905 18.7
7235 10188 47,000 65.2 19.6 0.156 18.8
7355 10179 40,000 526.5 20.5 0.025 24 4
7233 10186 38,200 658.3 23.2 0.0230 27.5
7416 10176 32,400 1670.6 23,7 0.0083 29.9

 

®Exposed to static fluoride salt for 15,289 hr at 650°C.
Table A-10. Tensile Properties of Heat 67-502 Surveillance Samples®

 

 

 

Specimen Test Strain Stress, psi Strain, % Reduction True
Number Temperature Rate in Area Strain
(°c) (min~!)  Yield  Ultimate Uniform Total (%) (%)
5329 25 0.05 65,400 119,300 50.3 51.6 34,81 43,0
5330 200 0.05 51,300 104,500 47,0 49.3 42.83 56.0
5331 400 0.05 47,500 93,800 47.9 50.2 45,68 61.0
5333 500 0.05 47,800 109,000 48.0 51.8 42,39 55.0
5335 550 0.05 40,000 104,000 43,5 bte 2 39.02 49,0
5322 600 0.05 41,100 82,800 38,1 40,2 31.46 38.0
5328 650 0.05 44,100 69,600 22.5 24,1 27.13 32.0
5320 700 0.05 36,700 61,600 19.0 20.2 23.35 27.0
5318 760 0.05 33,000 58,000 14.9 16.6 8,67 9.0
5316 850 0.05 31,100 35,300 2.4 3.0 13.29 14.0
5332 400 0.002 52,300 96,800 48 .4 50.0 35.22 43.0
5334 500 0,002 52,600 82,400 29.3 31.4% 26,03 30.0
5336 550 0.002 51,200 66,400 10.5 13.0 22.56 26.0
5321 600 0.002 40,200 62,700 17.1 19.0 19,07 21.0
5327 650 0,002 42,200 62,900 15.6 17.1 15,36 17.0
5319 700 0.002 35,500 55,500 10.2 12.7 13.29 14,0
5317 760 0.002 35,500 42,100 b oo 8.2 9.59 10.0
5315 850 0.002 24,900 25,100 1.1 3.0 9,44 10.0

 

gExgosed to core of MSRE for 9789 hr at 650°C. Thermal fluence was
5.3 x 1079 neutrons /em?.

6L
Table A-11l. Tensile Properties of Heat ©7-504 Surveillance Sam.plesa

 

 

 

Specimen Test Strain Stress, psi Strain, % R?duction Tru?
Number Temperature Rgtgl : ‘ . in Areas Strain
(°C) (min™*)  Yield Ultimate  Uniform  Total (%) (%)
5086 25 0.05 102,300 119,300 26.0 27.0 33.46 41.0
5087 200 0.05 46,400 105,700 43,7 by 4 31.89 38.0
5088 400 0.05 41,000 97,700 440 44,5 37.20 47.0
5020 500 0.05 40,100 93,000 42.8 43,3 37.% 48.0
5092 550 0.05 39,600 112,000 46,0 46,7 36,46 45,0
5079 600 0.05 39,700 101,000 35.3 35.7 33.53 41,0
5085 650 0.05 32,900 70,000 30.0 30.7 31.45 38.0
5077 700 .05 35,000 65,900 23.2 24,2 24, 54 28.0
5075 760 0.05 34,900 57,600 12.8 13.8 18.02 20,0
5073 850 0,05 32,600 40,800 4.8 6,6 9.37 10.0
5089 400 0.002 41,700 99,000 41.8 42,9 36.36 45.0
5091 500 0.002 41,600 91,200 42,2 43,2 28,18 33.0
5093 550 0.002 38,000 78,100 28,2 29.9 21.18 24,0
5078 600 0.002 40,200 69,600 23.7 24.8 17.16 19.0
5084 650 0,002 34,700 63,900 21.9 23.1 16.96 19.0
5076 700 0.002 35,700 56,200 11.9 14.6 13.12 14,0
5074 760 0.002 37,000 42,300 4.3 6,7 11.92 13.0
5072 850 0.002 25,400 26,000 2.1 A 11,50 12.0

 

%Exgosed to core of MSRE for 9789 hr at 650°C. Thermal fluence was
5.3 x 10°° neutrons /em®,

08
 

 

 

Table A-12. Tensile Properties of Control Specimens of Heat 6'7-502a
Specimen Test Strain Stress, psi Strain, % Reduction True
Number Temperature Rate in Area Strain
um (°c) (min™1)  Yield  Ultimate Uniform Total (%) (%)
5381 25 0.05 56,600 113,400 53.7 55.4 50.04 69,0
5390 200 0.05 54,400 111,700 54, 4 56.2 51,17 72.0
5393 400 0.05 48,100 106,100 58.0 60,9 52.51 74,0
5391 500 0.05 42,900 95,600 50.6 52.6 45,31 60.0
5395 550 0.05 45,600 97,600 52.9 54,9 48,16 66,0
5397 600 0.05 44,000 93,500 52.8 55.1 43,09 56.0
5394 650 0.05 51,200 98,200 46,5 48,3 35.48 44,0
5389 700 0.05 41,700 83,600 41,3 43,6 34,19 42,0
5388 760 0.05 39,200 72,800 24,5 39.7 29, 44 35.0
5382 850 0.05 35,200 47,300 8.0 25.0 23.35 27.0
5386 400 0.002 49,300 105,100 54,8 56.9 48,90 67.0
5384 500 0,002 45,000 89,700 51,7 55.7 45,39 ©1.0
5392 550 0.002 38,800 88,100 50.4 52.9 35.48 44,0
5379 600 0. 002 52,400 88,900 29.6 31.2 28.87 34,0
5387 700 0.002 41,700 75,300 18.8 34.1 31.56 38.0
5383 760 0.002 42 , 800 50,800 6.4 27.3 24,75 28.0
5380 850 0,002 29,000 29,400 2.9 18.1 12,07 21.0

 

Safter exposure to static fluoride salt for 9789 hr at 650°C.

8
Table A-13, Tensile Properties of Control Specimens of Heat 67 =504

 

 

 

Speni Test Strain Stress, psi Strain, % Reduction True
pecimen . .
Number Temperature Ratgl ) . in Area Strain
(°C) (min~*) Yield Ultimate Uniform Total (%) (%)
5058 25 0.05 67,500 132,800 50.2 52.3 43,72 57.0
5046 200 0.05 59,700 118,700 43,5 44,0 41.59 54,0
5057 400 0.05 45,500 105,100 46,2 47,2 36,41 45,0
5061 500 0.05 43,100 99,900 50.6 52.1 41,30 53.0
5045 550 0.05 45,300 101,500 47,9 49, 4 36,87 46,0
5055 600 0.05 45,400 97,700 49,1 50.9 36,67 46.0
5065 650 0.05 43,300 94,300 48.8 50.9 37.37 47.0
5067 700 0.05 53,300 9%, 600 30.8 39.5 36,00 45,0
5066 760 0.05 44 , 200 75,400 20.3 43,5 38.44 49,0
5060 850 0.05 34,900 47,800 7.7 38.7 35.48 4,0
5064 400 0.002 53,400 122,800 46,2 47.9 34,19 42,0
5041 500 0.002 61,300 115,500 bty .5 46,1 33,52 41,0
5053 550 0.002 49,200 99,700 47,2 48,5 39.59 50.0
5044 600 0.002 45,800 91,300 41.9 43,7 37.27 47,0
5042 650 0.002 44,900 80,200 30,0 41,2 38,14 48,0
5048 700 0.002 42,100 68,100 16.2 45,8 40, 64 52.0
5047 760 0.002 42,700 50,800 6.0 38.9 41.78 54,0
5056 850 0.002 31,500 32,400 2.9 34.8 34.29 42,0

 

®After exposure to static fluoride salt for 9789 hr at 650°C.

23
83

Table A-14, Creep-Rupture Properties at 650°C of Surveillance
Specimens of Modified Hastelloy N®

 

 

 

. Stress Rupture Rupture Minimum
Ngzszr Sfii;;:in Level Life Strain Creep Rate
(psi) (hr) (%) (%/hr)
Heat 67-502
R-735 729 55,000 26.0 5.8 0.178
R-716 711 47,000 126,4 4,3 0,0248
R-709 704 40,000 655,2 6.4 0,0068
R-735 730 32,400 12747 5.8 0.0019
Heat 67-504
R-733 728 55,000 59,6 7.0 0.0905
R-715 710 477,000 181,7 6.8 0,0272
R-708 703 40,000 467.2 6.9 0.0113
R-732 727 32,400 1643.8 5.9 0.0025
. ®Exposed to the core of the MSRE for 9789 hr at 650°C,
Thermal fluence was 5.3 X 10°° neutrons/cm®,

Table A-15. Creep-Rupture Properties at 650°C of
Control Specimens of Modified Hastelloy 7

 

 

Te gt Specimen Stress Rupture Rupture  Minimum Reduction
Number fium;er Tevel Life Strain  Creep Rate in Ares
(psi)  (hr) (%) (%/br) (%)

Heat 67-502

7232 5401 70,000 3.6 45.6 4,15 39.5

7231 5402 55,000 60,1 38.3 0,308 38.3

7230 5403 47,000 242.2 37.2 0.0775 34,7

7229 5404 40,000 1148.6 31.7 0.0148 29.2
Heat 67-504

7228 5050 70,000 5.5 43,7 3.80 43,9

7227 5043 55,000 108.3 41,7 0.180 39.6

7R26 5049 47,000  329,2 39.4 0.0485 43.3

7225 5062 40,000 1014.6 40.8 0. 0200 48,0

 

%Exposed to static fluoride salt for 9789 hr at 650°C.
1-3.
4=5,

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