ORNL-TM-3064.txt

 

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-

OAK RIDGE NATIONAL LABORATORY

operated by
UNION CARBIDE CORPORATION
for the
U.S. ATOMIC ENERGY COMMISSION

ORNL- TM- 3064

 

MASTER

- INFLUENCE OF TITANIUM, ZIRCONIUM, AND HAFNIUM ADDITIONS ON THE
RESISTANCE OF MODIFIED HASTELLOY N TO IRRADIATION DAMAGE
= AT HIGH TEMPERATURE - PHASE |

: H. E. McCoy, Jr.

THIS DOCUMEN
TCo |
UNCLASsiFIEp TMED AS

DIVIS]
By SION OF CLASSIFICATION
DATE 155

s

y NOTICE This document contains information of a preliminary nature P82;}
3

ond was prepared primarily for internal use at the Oak Ridge National
Laboratory. It is subject to revision or correction and therefore does

not represent a final report.

PISTRIBUTION OF THIS DOCUMENT 1S UNLIMITER
 

 

This report was prepared as an account of work sponsored by the United
States Government, Neither the United States nor the United States Atomic
Energy Commission, nor any of their employees, nor any of their contractors,
subcontractors, or their employees, makes any warranty, express or implied, or
assumes any legal liability or responsibility for the accuracy, completeness or
usefulness of any information, apparatus, product or process disclosed, or
represents that its use would not infringe privately owned rights.

 

 
ar

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ORNL~-TM-3064

Contract No. W-7405-eng-26

METALS AND CERAMICS DIVISION

INFLUENCE OF TITANTUM, ZIRCONIUM, AND HAFNIUM ADDITIONS ON THE
RESISTANCE OF MODIFIED HASTELLOY N TO IRRADIATION DAMAGE

AT HIGH TEMPERATURE — PHASE I

H. E. McCoy, Jr.

- JANUARY 1971

 

 

LEGAL NOTICE
This report was prepared as an account of work
sponsored by the United States Government, Neither
the United States nor-the United States Atomic Energy
Commission, nor any .of their employees, nor any of
their contractors, subcontractors, or their employees,
makes any warranty, express or implied, or assumes any

; legal liability or responsibility for the accuracy, com- |-
| pleteness or usefulness of any information, apparatus,

product or process disclosed, or represents that its use
would not in_fringe privately _owned rights, :

 

OAK RIDGE NATIONAL LABORATORY

~ Oak Ridge, Tennessee
operated by

UNION CARBIDE CORPORATION

for the

U.S. ATOMIC ENERGY COMMISSION

DISTRIBUTION OF THIS DOCUMENT IS UNLIMITER

 
 

 

 

 
 
 
 

 

L&

 

 

 
 

o) l‘

iii

CONTENTS

Abstract .
Introduction . . . . . « . . « .+ . .
Experimental Details .
Test Materials . . . . . .
Irradiation Conditions .
Testing Procedure
Experimental Results .
Alloys Containing Titanium .
Alloys Containing Zirconium
Alloys Containing Hafnium
Alloys Containing No Additions .
Discussion of Results
Summary . . .
Acknowledgments . . . . . . . . . .

Page

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50
100
124
135
143
143
 

 

 

 

 

 

 
 

)

INFLUENCE OF TITANIUM, ZIRCONIUM, AND HAFNIUM ADDITIONS ON THE
RESISTANCE OF MODIFIED HASTELLOY N TO IRRADIATION DAMAGE
AT HIGH TEMPERATURE — PHASE 1

H. E. McCoy, Jr.

ABSTRACT

The influence of small additions of Ti, Zr, and Hf on
the mechanical properties of a modified Hastelloy N with the
nominal composition Ni—12% Mo—7% Cr—0.2% Mn—0.05% C is
described in this report. It deals specifically with test
results from numerous, small, laboratory melts and several
100-1b melts from commercial vendors. Additions of Ti, Zr,
and Hf had beneficial effects on the properties of the alloy
both unirradiated and after irradiation. Irradiation temper-
ature had a marked effect upon the properties of all alloys
investigated. Generally, good properties were observed when
the irradiation temperature was 650°C or less and poor when
the temperature was 700°C or higher. We attributed this
large effect of irradiation temperature to coarsening of the
carbide structure at the higher temperature.

 

INTRODUCTION

Previous studies!™ showed Hastelloy N (Ni-16% Mo—7% Cr—4% Fe-0.05% C)
susceptible to a type of damage produced by irradiation at high tempera-
tures that results in reduced stress-rupture éroperties and fracture duc-
tility. One approach to solv1ng thls problem is that of making slight
changes in the chemistry of the alloy We initially modified the base
comp031t10n to Ni—12% Mo—7% Cr—0.2% Mn—0.05% C to obtain_an“alldy that

 

- 14, E. McCoy and J. R. Welr, "Stress—Rupture ProPertles of Irra-

diated and Unirradiated Hastelloy N Tubes," Nucl. Aggl. 4(2) 96104

(1968).
°H. E. MtCoy, "Varlation of Mechanical Propertles ‘of Trradiated

Hastelloy N with Strain Rate,"” J. Nucl. Mater. 31(1), 67-85 (1969).

34. E. McCoy, An Evaluation of the Molten-Salt Reactor Experiment
Hastelloy N Surveillance Specimens — First Group, ORNI~TM-1997 (1967).

 

 

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

 
 

 

was free of the stringers of the MgC type of carbide (where M is the
metallic component) characteristically found in standard Hastelloy N.
We then made a number of laboratory melts with various additions of Ti,
Zr, and Hf, since they form very strong, stable diborides® and should
tie up the boron as compounds rather than allow it to segregate to the
grain boundaries. The more random distribution of the boron should
reduce the radiation damage by reducing the concentration at grain
boundaries of helium produced by the 1°B(n,x) reaction.

The results of tensile and cfeep tests on these experimental alloys
show that the resistance to irradiation damage is improved by the addi-
tion of Ti, 2r, or Hf. Results of tests on the first small (100-1b) heats
of commercial alloys that contained these same alloy additions also indi-
cate improved properties.

EXPERIMENTAL DETATILS
Test Materials

Our alloys were nonconsumably arc melted in an argon atmosphere
from melting stock of commercial purity. A starting charge of about
2 1b was consolidated and melted several times. The charge was then
placed in another arc-melting furnace with & hearth for drop casting.

The alloy was again melted and drop cast into a l-in.-diam X 6-in.-long
ingot. The ingot was swaged to 1/4-in.-diam rod by the following sched-
ule: from 1 in. to 3/4 in. at 1177°C, from 3/4 in. to 7/16 in. at 871°C,
and from 7/16 in. to 1/4 in. at ambient temperature.

The commercial melts were melted by vacuum induction and cast into
six 4-in.-diem ingots that weighed abéut 16 1b each. They were initially
forged at 1177°C; the final working was done at room temperature. The
final product was small 1/2-in.-thick plates with 40% cold work.

The chemical analysis of each alloy used in this study is given in
Table 1.

 

°B. Post, "Refractory Binary Borides," p. 340 in Boron, Metallo-
Boron Compounds and Boranes, ed. by R. M. Adams, Interscience, New York,
1964.

L
Weight
of Melf;
(1v)

fppmf
B

Other

Ti

Composition

(wt %)

Chemical Compositions of Alloys
Si

Fe

‘Table 1.

Cr

 

 

Mo

)

 

 

_Alloy'
Number

PN AN AN NANANNNNDANNNNANANNDONANNN N

8 83 8388
— —~ ~ ~ ~
NN NN ~t =1 N
OCOO0OO0OO0C0O00O0 ~-NOOOON o o ™
OO0 000000 COO0CO0OO0O0O0O0 A NO OO
o O VO W O
O ™ _I_O_I_BL.O:J
oo OCOO0CO00O0
ol nou o f
© o HHH YN H A
mom mmmm T m
0
Y Mo M M — O wnun O NN AdHO
QOJOOOOOOO. 01257B9%3000000000
00000000 000001000000000000
VVVVVVVYV V Vv V
i M
HeAO HnNO o0 Oddrd A0 0O A4 A4+ 0O
0133568 90400000000000000001
0OOOOOO101000000000000000000
V VVVVYV VVVVYV Vv

NI NN A 0N 1N 1A 100 N o o
mWnumwnununununuq,q,mumwnumWnunumw1¢1+qu,1*1_1_1_1*9~nu
OO0 BC0O0dODOOOd00D000000OD
0000000000000 000000000000OO0
VVVVVVVVVYV VVVVYV vV V V
—~ O O~ oy H N ~ 0\ N O D~ n
DRSO 8RYNa 08NSy DO MMM NN

OCOOHO0000C000H00 Abo0ooooAAO
CO000000O00O00O0O000 00000000000

VvV V

O00O00OO0O0OO0OOAMHMHOAMHMNM OOV HHW 0
‘ 9~9~9%9~9~9~9~9#9~9~9~9~9~1_9~9~9~9~1*1*9~1*mmw*wu wu1*
O000O0O0O0O0O0O0O0O0O0OO0O0

.....

 
 

 

neutrons cm

The rods were fabricated into the test specimens shown in Fig. 1.
Our work with this specimen has shown that the rest results are reproduc-
ible and agree quite well with those obtained for more massive specimens.
However, there is sufficient stress concentration to cause some of the
more brittle specimens to break in the radius at the end of the gage

length.

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Fig. 1. Specimen Used for Tests of Mechanical Properties.

Irradiation Conditions

The results presehted here were obtained from specimens irradiated
in several experiments. These experiments were carried out in three
reactors: the Engineering Test Reactor (ETR) at Idaho Falls, Idaho,

and the Oak Ridge Research Reactor (ORR) and Molten Salt Reactor Experi-
ment (MSRE) at Oak Ridge, Tennessee. A core facility was used in the
ETR where the thermal and fast (> 1 Mev) fluxes were each 3.2 X 1014

2 sec”™? and the fluence was 6 X 1020 neutrons/em?. The ETR
experiments were uninstrumented, and the design temperatures were either
below 150°C or 600 * 100°C. Melt wires included in these experiments
indicated that the operating temperatures were within the desired range.
The experiments in the ORR were performed in a poolside facility where
the peak fluxes were 6 X 1013 neutrons ecm™® sec™! (thermal) and

5 x 1012 peutrons cm™2 sec”! (> 2.9 Mev). The experiments were run for
4!

.

)

either one or two cycles so that the thermal fluence was 2 to 5 x 1020

neutrons/cm? and the fast fluence was 2 to 4 x 10'° neutrons/em®. These

'experimehts were instrumEntéd; and the temperatures were controlled at

650 to 871°C. The‘experiments in the MSRE were run in the center of the
core, where the thermal flux was 4.1 X 1012 neutrons cm™? sec™! and the
fast (> 1.22 Mev) flux was 1.0 X 10’2 neutrons cm™? sec™!. The samples
were exposed to a noncorrosive fluoride salt environment for 4800 hr at
645 + 10°C and received arthermAl fluence of 1.3 X 102° neutrons/cm® and
a fast fluence of 3.1 X lOlg'neutrons/cm?; - Since we could not see any
consistent effects of thermal-neutron'fluence over the small range
enéountered in these experiments, we present our date in this paper as

having been obtained at a common fluence.

Testing Procedure

The creep-rupture tests after irradiation were run in lever-arm
creep machines in the hot'céllszatrthe Oak Ridge National LaboratoryQ
The strain was méasured 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 differential transformer, and the transformer
signal was recorded. The aécuracy of the strain measurements is about
+0.1%, considering fhe effects‘of'aMbiént temperature variations in the

hot cell, mechanical vibrations, etc. This accuracy is considerably

‘better than the specimen-to-specimen reproducibility that one would
. expegt.for.this_allqy.;'The systems for measuring and controlling temper-

 ature combine to give a temperature uncertainty of about *1%.

‘The tensile testS‘weféfrun.Qn Instron Universal Testing Machines.

' The strain measurements were taken from the crosshead travel.

. EXPERIMENTAL RESULTS
 A11oystohtaining Titanium

‘The alloys with additions of titanium fell into three groups. The
first group, alloys 100 through 107, contained from less than 0.0l to
 

 

 

 

6

1.04% Ti. These alloys were of the nominal base composition of Ni—12% Mo—
7% Cr—0.2% Mn—0.05% C. Typical microstructures of these alloys after a
1-hr anneal at 1177°C are shown in Fig. 2. The alloys were all free of
intermetallic precipitates and contained only a small amount of carbide
precipitates. Generally, the amount of carbide increased-and the

grain size decreased with increasing titanium concentration. The second
group, alloys 75 and 76, contained 1% Ti and'were made to explore the
influence of varying carbon content. Typical microstructures of these
alloys are shown in Fig. 3. The third material, a 100-1b commercial
alloy from Special Metals Corporation, designated heat 21545, contained
0.49% Ti. Typical microstructures of this alloy are shown in Fig. 4.
This alloy had a nominal composition quite close to that of alloy 104,
and the microstructures of the two alloys after a l-hr anneal at 1177°C
were quite similar except that the commercisl alloy contained patches of
precipitate. Close examination showed these regions to be quite high in
molybdenum. Some of the precipitates were carbides of the MC type and
others were almost pure molybdenum. The grain size after annealing for
100 hr at 871°C was quite fine, and copious quantities of carbide pfe-
cipitates were present. '

Alloys 100 through 107 were tested under & variety of conditioms.
Tensile tests were run at 650°C on samples that had been annealed 1 hr
at 1177°C and aged 1000 hr at 650°C to duplicate the thermal history of
irradiated semples. The results of these tests are summarized in
Table 2. The variation of the tensile and yield strengths with titaniuwm
content is shown in Fig. 5. The yield strength increased with increasing
titanium. The tensile strength varied in a poorly defined manner: at
a strain rate of 0.05 min~!, it increased systematically with increasing

titenium, and at a rate of 0.002 min~! it reached a maximm at 0.55% Ti.
The variation of the fracture strain with titanium content is shown in

' Fig. 6. At a strain rate of 0.05 min~!, the addition of 0.11% Ti greatly
increased the fracture strain; further additions decreased the fracture
strain. At a strain rate of 0.002 min~!, the fracture strain improved
greatly with the addition of 0.11% Ti, and further additions caused a

gradual improvement.

¥

n
\_,’

 

 

 

 

 

-Alloy 104 §

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig. 2. Photomierographs of Alloys That Contain Various Amounts of
Titanium. The basic composition is Ni—12% Mo-7% Cr—0.2% Mn~0.05% C.
gi?" Etchant: glyceria regia. 100x. Reduced 38%.

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Fig. 3. Photomicerographs of Alloys (a) 75 and (b) 76. The nominal
chemical compositions are Ni—-12% Mo—7% Cr-0.2% Mn—1% Ti-0.05% C and
Ni—-12% Mo—7% Cr—0.2% Mn—1.0% Ti-0.10% C, respectively. Etchant:
glyceria regia. 100x. =

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‘Photomicrographs of Heat 21545, which Containe

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nnealed 100 hr at 871°C, 100

(a) Annealed 1 hr at 1177
Sample (b) etched with hydroch

49% Ti.

d O.
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ith glyceria regia.

itate after anneal of 1 hr at 1177
Reduced 28%.

Samples (a), (c), and (d) ‘etched w

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' . . a
Table 2. Tensile Properties of Several Unirradiated
- Pitanium-Modified Alloys at 650°C

 

 

, P - Reductio;
Alloy Specimen Strain Stress, psi Elongation, % z #:rlzn
- Rete = yie1g Uitimate G m Totel 1B Are
Number  Number (min=1) Tensile
x 103 x 103
100 1906 0.05 7.1 77 .4 43,6 44,2 34.2
100 1910 0.002 30 73.7 27.8 29.7 25.6
101 1933 0.05 27.9 82.6 62.1 64.8 4 2
101 1926 0.002 27.4 75.2 33.0 37.3 29.2
102 1943 0.05 . 31.1 87.1 51.2 53.2 42.7
102 1952 0.002 30.9 78.4 31.4 37.5 - 32.8
103 1962 0.05 31.5 87.7 51.4 54.0 47.3
103 1968 0.002 28.9 69.4 24.5 37.5 33.8
104 1987 0.05 31.1 87.5 51.0 53.0 42.4
104 1984 0.002 35.4 87.9 28.2 36.7 - 28.8
105 2007 0.05 34.2 93.6 51.0 52.8 45.0
105 2004 0.002 34.3 84.1 31.9 40.3 32.4_
106 2023 0.05 34.8 100.7 45,4 47.3 40.3
106 2026 0.002 33.9 79.8 26.0 37.8 35.1
107 2039 0.05 36.4 99.2 43.8 46.1 36.4
107 2041 0.002 33.7 75.1 22.3 38.5 32.5

 

®Annealed 1 hr at 1177°C and aged 1000 hr at 650°C.

ORNL-DWG £9-14398R
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° RATE = min~!

STRESS (ps)
3

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* - TITANIUM CONTENT (wt %)

Fig. 5. Influence of Titanium Content on the Yield and Ultimate
Tensile Strengths at 650°C after Annealing 1 hr at 1177°C and Aging
1000 hr at 650°C.

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TITANIUM CONTENT (wt %}

Fig. 6. Influence of Titanium Content on the Tensile Fracture Strain
at 650°C of Ni—12% Mo—7% Cr-0.2% Mn—0.05% C Alloys.

Alloys 100 through 107 were irradiated at 650°C to a thermal fluence
of 2.5 X 1020 peutrons/em?. The results of tensile tests at 650°C after
irradiation are summarized in Table 3. The fracture strains measured at
650°C are plotted as a function of titanium content in Fig. 7. The data

exhibit considerable variation, but show no systematic variation with

- titanium concentration.'_The*results of these tests are analyzed further

in Fig. 8, where the ratios offfiheivarious tensile proyerties'of the irra-

'7 :diéted_and unirradiated*&libYS aréECOmpared. The yield stress is increased

by irradiation, and the tensile stress and fracture strain are decreased.

However, the ratios are éfily{modérately dependent updh the titanium level.
Another set of samples of alloys 100 through 107 was irradiated at
50°C to a thermal fluence of 9 x 10'° neutrons/cm®. The results of ten-

sile tests of these sam§les ét:650 and 760°C are given in_Tabl¢_4. Com-

‘parison of these;results WithfltbOSé in Table 3 for the samples irradiated

to a higher fluence at a higher temperature reveals that the fracture

strains are higher for the samples irradiated at 50°C. The fracture
 

 

12

Table 3. Tensile Properties of Several Titanium-Modified
Alloys Tested at 650°C after Irradiastion®

 

 

' 3 Reducti
Alloy Specimen Strain _ Stress, psi Elongation, % ?.nuAr on
Number Number ,o%€ . Yield Ultimate Gyporm Total - ( )ea
(min~1) Tensile : ®
x 10° x 103
100 1908 0.05 31.5 66.3 26.3 26.6 23.3
100 1912 0.002 26.3  48.8 134 1.1 13.1
101 1924 0.05 35.5 68.9 23.2 - 23.7 19.1
101 1925 0.002 36.9 57.2 10.7 10.9 12.5
102 1949 0.05 35.1 74.8 33.7 35.1 28.8
- 102 1951 0.002 37.7 67.8 18.5 19.2 14.9
103 1978 0.002 35.3 65.9 21.9 22.8 18.8
104 2006 0.002 39.0 65.2 14.5 15.4 15.6
105 1989 0.002 35.5 69.7 21.9 22.7 31.6
106 2021 0.002 39.2  65.1  12.0  12.6 Yo b
107 2051 0.002 39.3 71.6 15.9 16.3 13.0

 

®pnnealed 1 hr at 1177°C before irradiation at 650°C to a thermal
fluence of 2.5 X 102° neutrons/em?.

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IRRADIATION CONDITIONS
TEMPERATURE = 650°C
28 TIME 24000 hr .
THERMAL FLUENCE =
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TITANIUM CONTENT {(wt %)

Fig. 7. Influence of Titanium Content on the Tensile Fracture
Strain at 650°C of Ni-12% Mo~7% Cr-0.05% C Alloys after Irradiation.
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TITANIUM CONTENT (wt %}

Fig. 8. Ratios of the Unirradiated and Irradiated Tensile Proper-
ties of Alloys Modified with Titanium. Tested at 650°C at a Strain Rate
of 0.002 min~2,

Table 4. Tensile Properties® of Several Titanium-Modified
Alloys Tested at 650 and 760°C after Irradiationb:

 

 

 

 

:Alléy Sfiecimefi s ——-§E£§§§fii%%%553- Elongation, i R:g“i:i:n
Number NUmber i Yield Tensile Uniform Total‘ . (%)
x 10° X 103
_ - Test Temperature 650°

100 1907 26,4 - 60.3 21.3 21.9 24.0
101 . 191 27.6 - . 62:.0. . 216 22.1 18.3
102 1944 31.3 71.4 22.8 23.4 19.4
1103 . 1975 30,7 © . 4.9 23.8. 24.4 - 27.2
104 1992 :29.8 71.8 25.1 25.7 ©  33.2
105 .. 2003 - 344 - 761 23.5 - 23.8 - 22.8
106 2025 33.4. ... .80.8 26.0 . 26,4 34.5
107 2043 39.6 _-_89 3 o R3.4 23.8 211

o Test Temperature 760°C _ ) , . -

100 1905 .. 274 .. 48.8 84 0 9.5 9.3
101 1940 0 L 27.60 0 51.0 8.1 -.9.2 *.5.6
102 - 1948 . 30,6 53,9 8.6 8.9 9.4
103 - 1eTT 26,8 © 53,8 - 9.3 10.3 71004 .
104 1995 .. . 30.0 54.4 8.9 .10.0 8.1
-105 2016 - .- 31.8 - - 56,7 8.8 11.2 . 8.6
106 2024 30.6- - 57.0 9.4 12.8 11.6
107 C 20567 . 31,5 . 58.9 9.3 13.2 “12.1

 

%strain rate = 0.002 min~l.

bAnnealed 1 hr at 1177°C before irradiatlon at 50°C to a thermal
fluence of 9 X 101° neutrons/cm®.’
 

 

14

strains measured at 650°C after irradiation ranged from 22 to 26% but
did not vary systematically with titanium content. However, at a test
temperature of 760°C, the fracture strain did show a slight dependence
on titanium level; it varied from 9.5% with no titanium present to 13.2%
with 1.04% Ti present. |

Several unirradiated samples of alloys 100 through 107 were creep
tested at 650°C. The results of these tests are summarized in Table 5
and plotted in Figs. 9 through 11. The results of stress-rupture tests
in Fig. 9 show a marked improvement in rupture life at a given stress
level with increasing titanium content. The results are somewhat
scattered, and there are some questions about the exact location of the
line for each heat. The measurements of minimum creep rate in Fig. 10
show that titanium decreased the creep rate. All of these tests were
run without an extensometer, and the precision of the measurements was
not very good. However, the data indicate that the creep rate did not
vary appreciably with titanium content from 0.1l to 1.04% Ti (ailoys 101
and 107). The data also indicate that titanium concentration may have
no influence on the creep rate at low stress levels. The fracture strains
are showm in Fig. 11 as a function of the minimum creep rate. Titanium
additions definitely improve the fracture strain. The fracture strains
of alloys 106 and 107, which contained 0.81 and 1.04% Ti, respectively,
were almost independent of creep rate; the other alloys exhibited
decreasing fracture strain with decreasing creep rate.

Several of these alloys were heat treated to simulate irradiated
samples by anneals of 1 hr at 1177°C and 1000 hr at 650°C. The results
of tests on these samples are summarized in Table 6. A comparison of
these results with those in Figs. 9 through 11 for samples that were
annealed 1 hr at 1177°C indicates how these alloys respond to aging. The
rupture life, minimum creep rate, and fracture strain were all increased
by this aging treatment.

Alloys 100 through 107 were irradiated at 650°C and creep tested
at the same temperature after irradiation. The results of these tests
are given in Table 7. All of these samples were tested at a stress level
of 32,400 psi. The rupture lives and fracture strains are shown as func-

tions of titanium content in Fig. 12. The rupture life was increased
J
)

15

Table 5. Creep-Rupture Properties of U‘nirra.dia.teda Titanium-
Modified Alloys at 650°C

 

 

. Rupture Minimum Reduction
Alloy Specimen Test Stress I3 Creep Elongation .
. if'e in Area
Number Number Number (psi) Rate (%)
ST Ry (%)
X 10%
100 1915 5725 - 55 3.4 0.43 22.5 20.8
100 1917 6099 47 20.8 0.55 14.5 12.5
100 1909 5554 40 47.4 0.088 12.5 9.5
100 - 1914 6108 32.4 185.1 0.0077 6.3 6.2
100 1921 6172 27 g27.8 0.0022 4.9 4.0
101 1930 5753 70 - 0.6 2.9 34.5 28.3
101 1937 5692 55 23.4 0.064 21.0 21.3
101 - 1929 6203 47 52.1 0.050 16.0 14.0
101 1934 6203 40 - 0 292.1  0.0060 14.1 13.2
101 1932 6104 32.4 792.8 0.0037 10.5 12.8
102 1958 5751 70 3.3 0.56 28.5 22.8
102 1950 5726 55 32.4 0.047 20.0 15.4
102 1961 6100 47 162.4 0.016 15.8 16.9
102 1956 5669 40 506.5 0.0076 13.7 14.7
103 1979 5747 - 70 3.2 0.45 28.3 22.8
103 1976 5687 55 69.9 0.029 18.5 13.9
103 1966 6087 47 157.4 0.010 14.0 11.6
103 1980 5663 40 . 541.5 0.0082 15.6 18.4
104 1998 6101 70 3.5 0.88 32.3 20.8
104 1991 5566 55 41.6 0.060 18.8 14.0
104 1982 5959 36,5 2718.9 0.0033 13.9 21.4
104 - 1990 5638 - 40 1001.2  0.0066 ; 19.9
105 2018 = 5754 70 7.7 0.25 23.4 18.5
105 2001 5698 - 55 . 94.0  0.019 . 15.9 13.7
105 2017 6089 - 47 - 179.7 0.015 15.5 19.6
105 2000 5586 40 - 476.6 0.0078 16.6 - 13.8
- 106 . 2031. 5750 .70 . 11.6  0.20 - 25.0 19.9
106 2029 5691 .55 . 95.4  0.042 - 21.3. 20.0
106- . 2022 5658 51 158.0 0.038 - 21.9 17.6
106 2032 5988 - 47 372.9 0.015 22.1 18.5
106 - 2034 6107 - 40 832.4 0.014 21.4 21.2
107 2054 5760 70 5.5 1.22 22.5 19.1
107 2046 5644 55 322,67 0.028 20.3 20.3
107 2055 6106 = 47  375.6 0.022 23.1 23.1
107 - 2040 5633 40 14°90.5 0.010 26.5 27.3

 

®pnnealed 1 hr at 1177°C béfore testing.
 

ORNL-DWG 69- 14394 7

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

(x10%) TR TTTT
»
ST
60 LIS 102
M N [T 103
”'t'.\-\.~ \a B b | $) m\// 104
50 100 ™M 101\\ n a 105
"yl { N &
K \"‘l ‘{\ \
7 NN ISONUTNL
2 a0 S TTRETNG
o ALLOY  TITANIUM nl NI ¥
i NO. CONTENT (%) N 7{.,.\\ N
= 30~ 0100 <0.08 -t ~
a 101 0.1 bl
o 102 0.39
20 - « 103 0.3
° 104 0.55
e 105 0.66
01~ 4 108 0.81
= ¢ 107 1.04
ol L L1l
10° 10! 102 o 10

RUPTURE TIME (hr)

Fig. 9. Stress-~Rupture Properties at 650°C of Several Alloys Con-
taining Titanium. All samples annealed 1 hr at 1177°C before testing.

ORNL-DWG 69 -14393

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

(x 103) it Q-‘ 9-
N
8§ \§>
60 Rl
NN i
N
TR °
5 SN
YRR
9 \§ L8 /9-"""
& 40 N i =l
o §§s§> 1T ALoy  TiTaNium
a D |~ NO.  CONTENT (%)
£ .0 @ |3+ ©100 <001 |
» P 2401 0.11
- 0102 0.39
20 " 103 0.31 _
¢ 104 0.55
° 105 0.66
0 4106 0.8 §
107 . 104
o | W
107> 102 T | 10°
MINIMUM CREEP RATE (%/hr)
Fig. 10. Creep Properties at 650°C of Several Alloys Containing .

 

Titanium. All samples annealed 1 hr at 1177°C before testing.
17

ORNL-DWG 69-14392

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

35
o////
30 _ _ 10130250230442;,'
| P
4 // |
25 4 106,107
’; ‘ /‘--—"—’-—----—-
e a ¢ ______..-74" o +
Z A [ L
= 20 A 100
E l//"1 //
tad - 7 8
(5 Jr i 3
- 11t TITANIUM
2 417 i ALLOY NO. CONTENT (%)
o A L~ o 100 <0.0f
10 > A 10 0.11
1 o 102 0.39
|5 T] 103 0.3!
5 = o 104 0.55
* 105 - 0.66
A 106 0.81 .
¢ 107 1.04
O L L Ll 1 1 1 1 1 11§
1073 1072 10~ 10° : 10

MINIMUM CREEP RATE (%/hr)

-~ Fig. 11. Fracture Strains at 650°C of Several Alloys Containing
Titanium. All samples annealed at 1177°C before testing-.

Table 6. Creep Properties of Several Ageda'
Titanium-Modified Alloys at 650°C

 

 

Minimum .
Alloy Specimen ' Test  Stress _Rgg;ure ‘Creep Elorgation R?ductlon
. e in Area
Number Number Number (psi) (hr) Rate (%) (%)
o o - (%/nr)
x 103
100 1922 5723b 55 5.9 0.61 23.0 19.2
100 1919 5630 17.8 1007 0.0007 1.0 0
101 1936 5724, 55 48.6 0.12 24.5 21.9
101 1935 5783 32.4 1250.7 0.021 19.2 12.5
102 1960 57015' 55 88.0 0.140 37.5 29.7
102 1959 5780 32.4 1659.1 0.0051 10.3 7.2
103 - 1973 5729, 55 54.0 0.21 36.4 25.6
103 1972 57577  32.4 1006.9 0.0046 7 6.2 3.3
104 1999 -,57305, 55 101.2  0.095 36.0 26.9
104 19%  5758° 32.4 1006.9 0.0054 5.8 3.3
105 2013 57285' 55 66.6 0.119 38.0 32.4
105 2011 5785°  32.4 1704.5 0.0033 5.9 4.5
106 2035 . 573L 55 1075 0.14 45,0 35.0
106 2027 5779 . 32.4 '1655.8  0.0039 7.5 4.8
107 2052 5654, 42.7 902.2  0.027 42.2 35.0
107 2047 . 5784° 524 1704.5 0.004 6.0 4.8

 

®Sampies ammealed 1 hr at 1177°C and aged 1000 hr at. 650°C.

Test discontinued before failure.
 

Table 7. Creep Properties of Irradiated® Alloys That Contain Various Amounts of Titanium

 

 

: Minimum True
Alloy Specimen Test Titanium Rupture Creep Elongation Reduction Fracture
Content Life in Area

Number Number Number Rate (%) Strain
100 1916 R-146 < 0.01 105.4 0.0099 1.1 5.70 5.90
101 1928 R=-147 0.11 169.9 0,0084 1.6 1.44 1.50
103 1971 R-150 0.31 570.1 0.0077 - 5.2 2.57 2.60
102 1954 R-148  0.39  1327.3 0.0034 - 5.8 4.63 4.70
104 2005 R=-149 0.55 1446.3 0.0048 8.7 ' 5.86 6.00
105 1983 R-151 0.66 1540.3 0.0046 9.2 \ b b
107 2042 R-153 1.04 2928.8 0.0026 12.8 | b b

 

®Irradiated at 650°C to a thermal fluence of 2.5 X 102° neutrons/cm?. Tested at 650°C

32,400 psi.

bDiameter not measured after test.

and
Lt

19

ORNL-DWG 66-410878R

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1800
1600 f— = —+—51
1400 *100

58

1200

£

4 4000 ANNEALED 1 hr AT 4177°C

L 1 E PRIOR TO IRRADIATION

- : , _

X oo #125 - ~———1- IRRADIATION TEMPERATURE =650°C-

E UNIRRADIATED, THERMAL FLUENCE=

3 STD HASTELLOY N 2.5x10 neutrons/cm®. .

@ 600 o527 | TEST TEMPERATURE=650°C

- STRESS=32,350 psi
400
200 1o
o1 {

 

 

 

 

 

 

 

 

 

 

 

 

0 0Ot 02 03 ©C4 05 06 O7 OB 09 10
TITANIUM CONTENT (wt %)

Fig. 12. Influence of Titanium on the Creep Propertles of
Ni-12% Mo—7% Cr—0.05% C after Irradiation. Tested at 650°C at a stress
level of 32,400 psi.

by a factor of 25 and therfracture'strein by a factor of 10 by the addi-
tion of 1.04% Ti. These results are compared with those of the unirra-
diated control samples in Table 8. This comparison is fraught with the
uncertainties footnoted-in'Teble 8,'but it indicates that irradiated
samples containing O.55%-Ti or more have rupture lives that are about
0.7 times the unirradiated value, minimum creep rates that are about
equal to that of the unirradiated material, and fracture strains that
are about 0.7 times the unirradiated value. _

Alloys 100 through 107 were also 1rrad1ated at 30°C to a thermal

 fluence of 9 x 101° neutrons/cm and creep-rupture tested ‘The results

of these tests are given 1n.Teble“9,. These samples were teeted at dif-

ferent stress levels than these'used in thefexperiment.described pre-

- viously (Table 7). However, a_rough comparison'showsfthat the different
irradiation temperetures;(650ivéi50°C)"and the different;flfiEhces (25 vs

9 x 10%° neutrons/cma) had'iiffile'effeCt on the test results. Both
experlments show clearly the 1ncreas1ng rupture llfe and fracture strain
with 1ncre331ng titanlum level. -

The combined results of the tens11e and creep tests allow a deter-

mination of the varlatlon of the fracture strain with strain rate at 650°C.
Table 8. Comparison of the Creep Properties at 32,400 psi and 650°C of
Unirradiated and Irradisated TitaniumsMbdified.Alloys

 

 

 

 

 

Unirradiated Jrradiated ( .)
‘ Ratio (Irradiated:Unirradiated
Titanium Minimum Minimum
Content Rupture Creep Fracture Rupture Creep Fracture Rupture Minimum Fracture
Life Strain Life Strain . Creep
(%) Rate Rate Life Strain
(hr) (%/hr) (%) (hr) (4/hr) (%) Rate
< 0.01 185.1% 0.0077* 6.3% 105.4  0.0099 1.1 0.57 1.3 0.17
0.11 1250.7 ~ 0.021  19.2 169.9  0.008% 1.6 0.14 0.40 0.08
0.31  1900°  0.0046 12.5%°  570.4  0.0077 5.2 0.30 1.7 0.42
0.39  1900%° 0.0051 12.5%7P 1327.3  0.0034 5.8 0.70 0.62 0.47
0.55 1900%°  0.0054 12.5%®  1446.3 0.048 8.7 0.76 0.89 0.70
0.66 1900%°  0.0033 12.5%:P 1540.3  0.0046 9.2 0.81 1.4 0.74
1.04 3000%®  o.0:40 21%/P 2928.8 - 0.0026  12.8 0.75 0.65 0.61

 

a'Values from unaged samples.
bExtrapolated.

" ‘ -

oc

 
®h

| Table 9. Creep Properties of Irradiateda Alloys That Contain Various Amounts of Titanium

 

 

- : e | Titanium Test , Rupture Minimm e
Mber  Mmber  Nmber  Cnient  Temerstuwe QIS i e S
o o . x 10° o o
L100  aom R-257 < 0.01 650 40 68.9 0.019 179
101 1927 Re264 0.11 650 40 226.1 0.0037 . 1.55
103 1974 R-286  0.31 650 - 40 114.0 0.013  3.07
102 1957 R-282 0.39 650 40 255.4 0.010 5.04
104 1981 R-287 - 0.55 650 47 114.8 0.025  5.25
105 2008 R-290  0.66 650 47 200.9 0.0078 - 3.89
106 2030 R-288 0.8l 650 47 318.9 0.0072 8.25
107 2044 R-289 1.04 650 47 374.0 0.0073  7.55

 

®pnnesled 1 hr at 1177°C before

irradiation at 50°C to a thermal fluence of 9 x 101° neutrons/ém?.

TC
 

 

22

The date for samples irradiated at 650°C to a fluence of 2.5 x 103
neutrons/cm?® were used to construct Fig. 13. The strain rate is a con-
trolled parameter in a tensile test and was used direectly in this figure.
The stress is controlled in a creep test, and the resulting minimum
creep rate was used in Fig. 13. The lines on this graph simply Jjoin the
date points, and their slopes have little significance. The most impor-
tant observation is that the addition of titanium causes a greater
improvement in the fracture strain at low strain rates than at high
strain rates.

ORNL-DWG 69-143N

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

35
I TTHNT iE!HHH lllH
ALLOY NO. ANIUM CONTENT (%)
¢ 100 <0.014 /
T av02 0.39 _ 14
+ 104 0.55 /
0 107 1.04
25 / 7
8 1
z / 1
g 20
o L&
- I 1
n 14 //
& —+491HIIE 2
§ 15 L1 y——— ,::._‘.,—
g o =117 st bt - ’ur
- ™ L.
0 et -
—— 4 -
_U‘ = ’.—‘ - ) L~
5 abn T o 17
LUHT |creer TENSILE
_-T TESTS TESTS
o Lol 1T
10-3 102 10~ 00 10! 102

STRAIN RATE (%/hr)

Fig. 13. Variation of Fracture Strain with Strain Rate at 650°C
for Several TltanlumFBearlng Allogs Irradiated at 650°C to a Thermal
Fluence of 2.5 X 1029 neutrons/cm

The results presented thus far for alloys 100 through 107 were
obtained at a test temperature of 650°C for samples that were irradiated
at 650°C or lower. The results in Table 10 illustrate the effects of
varying the temperatures at which specimens were irradiated and tested.
One group of samples was irradiated and tested at 760°C. These results
show & general improvement in properties with increasing-titanium level.
Thé stress-rupture results for this series of tests are compared in
Fig. 14 with those for tests on samples irradiated and tested at 650°C.
Several samples were irradiated at 50°C and tested at 760°C. .These were

~tested under different conditions, but the trend of increasing rupture
Table 10.

 Creep Properties of Alloys That Contain Various Amounts of Titanium

After Irradiation® at Various Temperatures

 

 

 

 

 

D -_ el - o Minimum
Alloy -~ Specimen . Test Titanium Lemperature, C Stress Ru?ture Creep Elongation
Number Number - Number  Content Irra- Test (psi) Life Rate (%)
eE RS (wt %)  diation P (nr)
- (%/nr)
x 103
: S 2.5 X 10°° neutrons/cm® (thermal) |
100 ©.3791° | R-455 < 0.01 657 760 12.5 1.0 0.10 0.21
101~ 3801 - R-454 = 0.11 760 760 12.5 0.8 0.10 0.20
102 .. 3814 . R-453 . 0.39 760 760 12.5 18.1 0.0252 0.71
103 - 3821 ' R-452 . 0.31 760 760 12.5 12.4 0.0333 0.66
104 . 3608 - R-451 . 0.55 760 760 12.5  340.8 0.0086 5.72
104 - 3616 : R-763 . : 760 650 25.5  203.3 0.0042 1.36
105 © 3831 R=450 0.66 77 760 12.5  271.6 0.0066 3.05
106 - 3841 R-458 0.81 774 760 12.5  159.0 0.0196 5.40
107 3855 R-784 1.04° 760 650 30 49 4 0.0238 1.58
107 3851 R-449 S 760 760 12.5 1000, 0.0039 5,19
| - | - 15 242 0.0185 4.88
17.5 38 0.0638 3.92
| _ | 9 x 1019 neutrons/ecm? (thermal)
100 1913 R-254 < 0.01 50 760 15 15.1 0.037 0.83
100 1918 R-308° < 0.01 50 760 15 50.8 0.020 1.67
104 1986 - R-255.  0.55 50 760 15 358.8 0.016 11.9
104 1985 ' R=309° 0.55 50 760 15 316.3 0.011 11.3
107 2045 -~ R-261_ 1.04 50 760 15 914.2 0.015 30.2
107 2035 R-310° = 1.04 50 760 15 734.5 0.016 25.1
S 20 4.8 0.19 2.7

 

* ®Annealed 1 hr at 1177°C before irradiation. -
Pannealed 1000 hr at 650°C after irradiation..

Cload incressed at the intervals indicated.

£C
 

 

24

ORNL-DWG 68-B668R2

w04

57 *
AND & / IRRADIATED AND TESTED AT 760

TESTED AT 650°C '7_5TRESS'42,500 psi :
STRESS=322350 psi

RUPTURE LIFE (hr)

ANNEALED 4hr AT H77*C
BEFORE 'RRADIATION

5 TO 2-3%10%%neutrons/cm?
NUMBERS INDICATE FRACTURE
STRAINS.

 

/‘
0 0.2 04 06 C8 1.0 1.2 4
TITANIUM CONTENT (wt %)

100

Fig. 14. Effect of Titanium Content on Variation of Creep Proper-
ties after Irradiation.

life and elongation is the same as for samples irradiated at 760°C. The
only property that seems appreciably different is the fracture strain:
higher values were obtained for samples irradiated at 50°C. The most
dramatic effect of irradiation temperature is demonstrated by the com-
bination of irradiation at 760°C and testing at 650°C (compare the
results in Tables 7 and 10). The general effect is that the higher irra-
diation temperature results in a shorter rupture life, a higher minimum
creep rate, and a lower fracture strain. For example, alloy 107 was
irradiated at 650°C and tested at 32,400 psi and 650°C. The rupture life
was 2928.8 hr, the minimum creep rate was 0.0026%/hr, and the fracture
strain was 12.8%. The same alloy irradiated at 760°C and tested at
30,000 psi and 650°C had a rupture life of only 49.4 hr, a minimum creep
rate of 0.024%/hr, and a fracture strain of 1.6%.
¥

25

Alloys 75 and 76 were similar in compositien to alloy 107, and only

- limited testing has been carried out. The results of tensile tests are

summarized in Table 11. The results for the unirradiated samples agree
well with those for alloy 107 (Table 2)." Alloy 75 contains 0.062% C,
and alloy 76 contains 0.117% C; therefore, the trends of higher strength

~and lower fracture strain for alloy 76 are reasonable. Irradiation

increased strength and decreased fracture strain (Table 11). The fracture
strains agree reasonably well with those for alloy 107 irradiated at 650°C
(Table 3) and 50°C (Table 4). The samples annealed for 100 hr at 871°C
had a very small grain size, and the fracture strains were lower for

these samples The hlgher carbon content of alloy 76 resulted in slightly

lower fracture stralns in ten51le tests

- Table 11. Tensile Properties of Irradlated and Unirradiated Alloys 75

(0. 99% Ti, 0.062% c) and 76 (1. 01% Tl, 0. 117% C) at 650°C

 

Anneal StreS§, psi Reduc-

 

 

. Strain Elongation, %
Alloy Specimen Before Ultimate ———o——=—22 7 +{ion in
Number Number Irra- (iifif Yield o oiqe (Uniform Total ppon
diation . (%)
| X 10° x 103
Unirradiated
75 1197 121 . 0.05 31 89.3 45.0  45.8 36.3
75 1183 121 0.002 31.4 86.4 - 32.2 1 33.2 25.9
76 1300 121 0.05 39.3 100.3 34.6 34.8 26.9
76 1184 121 0.002  40.5 92.6  25.1 28.7  24.2
Irra.diatedb : '
75 1010 121 0.05 89.5 99.4  16.0 17.2  20.5
75 1011 121 - 0.002 - 90 30.4 5.1 6.1 11.7
75 1118 16  0.05 616 78 8.9 8.9  13.2
75 1119 16  0.002  60.8 66.5 6.4 6.4  10.2
76 /1012 121 0.05 - - 89.5 102.3 - 13.0 13.2 19.1
76 1013 121 0.002 91.8 9.3 5.0 5.3 117
76 1121 16 0.05 ~  62.6 76.8 7.1 7.1 8.6
I 76 1122 0 16 0 0.002 0 6l.4 0 65.7 - 5.2 5.2 - 8.6

 

Anneallng de51gnations .1215 annealed 1 hr at 1177°C.
"16 = annealed 100 hr at 871°C.

bIrra.dlated at 150 c to a thermal fluence of 5.8 x 1020 neutrons/cm .

non-.
 

 

26

Several samples of alloys 75 and 76 were creep tested at 650°C; the
results of these tests are summarized in Table 12. The proPefties of the
unirradiated alloy are qnité comparable with those for alloy 107 (1.04% Ti)
that are given in Table 5. BSame samples were irradiated at 760°C and
tested at 650°C; the test results are given in Table 12. Both of these
alloys had extremely good properties under these conditions. - A comparison .
with the results for heat 107 (1.04% Ti) in Table 10 reveals the superi-
ority of alloys 75 and 76. We have no explanation for this observation.

—

Table 12. Creep Properties of Irradiated and Unirradiated Alloys 75
(0.99% Ti, 0.062% C) and 76 (1.01% Ti, 0.117% C) at 650°C

 

Minimum

 

Alloy Specimen Test  Stress Rupture Creep Elongation Reduction
. Life in Area
Number Number Number (psi) (br) Rate (%) (%)
(%/br)
x 103
Unirradisted”
75 8293 5276 55 128.5 0.0325 21.9 - 18.4
75 8341 5458 40 1919.9 0.0075 33.3 - 26.3
76 8292 5275 55 92.1 0.0880 15.6 22.7
| Irradisted™’ P
75 10266 R-888 47 248.1 0.0442 13.1
75 10265 R-855 35 2669.2 0.0061 23.6
76 10269 R-889 47 276.0 0.0540 19.6
76 10268 R-852 35 2688.4 0.0059 22.0

 

%annealed 1 hr at 1177°C.
Prrradiated at 760°C to & thermal fluence of 3 X 1020 neutrons/cm?.

The first commercial heat of a titanium-modified alloy was a 100-1b
vacuum-melted alloy from Special Metals Corporation, designated 21545.
This alloy was tested as received (40% cold work), after annealing 100 hr
at 871°C (fine grain size, Fig. 4), and after annealing 1 hr at 1177°C
(coarse grain size, Fig. 4). The results of tensile tests on these
samples are summarized in Table 13. The fracture strains at a strain
rate of 0.05 min~! are shown in Fig. 15 as a function of test temperature.

The fracture strain of the material as received was about 10% at 25°C and
”"

27

Table 13. Tensile Properties of Unirradiated Alloy 21545 (0.49% Ti)

 

 

 

a s -Test Stress si Lo Reduc-
Specimen gzgzii S;zizn Temper- i Uitgmate -E%QEEEEEEEL—% tion in
Number : , .\ ature ield Tensile Uniform Total  p.,..q
Test (min~1) =/,
_ (°¢) (%)
| ,. x 102  x 10°
3432 121 0.05. 25 39.4 109.4 68.9 72.5 63.7
3451 121 0.05 200 31.8 99.2 70.9 74.9 58.1
3447 121 0.05 - 427 1 29.4 95.6 73.1 78.8 58.6
3435 121 0.05 500 28.1 9l1.1 78.9 81.9 57.5
2531 - 121 0.05 550 29.1 93.4 72.5 - 75.2 53.3
2521 121 0.002 550 - 30.5 9.1 65.5 67.9 43.9
3436 121 0.05 600 26 .4 85.9 70.9 75.3 52.0
3444, 121 0.002 600 27 74 .8 41.8 43.0 37.5
3446 121 1 0.05 650 24.9 76 41.6 43.5  38.2
2530 121 0.002 650 22.1 67 .4 21.2 22.0 30.7
3425 121 - 0.05 760 25.1 67.6 31.0 33.2 21.8
3429 121 0.002 760 24.7 50.8 12.0 19.2 20.4
2523 121 0.002 760 23.1 45 .4 10.1 25.2 26.5
3448 121 0.05 800 23.7 59 20.5 27.7 22.6
3445 121 0.002 800 - - 23.2 41.3 8.5 18.6 22.7
1565 000 0.05 25 159.7 . 170.4 8.5 11.9 17.5
2519 000 0.05 25 164.2 171.6 6.3 10.7 37.8
1566 000 0.05 427 131.5 147.1 11.4 15.2 28.8
1544 000 0.05 650  119.4  131.6 = 8.0 11.8  20.9
1545 000 0.002 650 113.4 116.8 3.4 8.9 13.2
1567 - 000 0.05° 760 92.7 93.4 3.0 24.9 41.9
1549 16 0.05 .25 . 54.2 122.9 48.7 52.0 56.6
1550 i6 0.05 427 37 102.2 49.8  54.0 42.3
- 2544 16 0.05 500 43.3 103.6 42.1 48,1 42.2
- 2543 16 0.05 550 - 42.5 99.8 - 43.9 45.9 35.0
2520 16 0.05 600 40.4 g9 30.0 31.5  27.3
1765 16 0.05 650 37.9 87.6 39.9 46.1 37.2
1766 - 16 -0.002 . 650 .. 42,2 - 73 19.4 - 48.3  53.2
2540 - 16 - 0.05 -~ 700  40.9 . 70.7- 21.2 = 41.8 42.3
- 2542 16 . -0.002° 700 - 41.4 59.6  .12.3 46.0 28.3
. 2546 16 - 0,05 760 - 416 56 . 11.7 . 53.5 - 58.3
2454 16 0.05 760 - 38 50.4° 13.0 .  56.2 T b
2547 16 0.002 760 37 - 37.3 1.5 53.3 68.1
2522 - 16 - 0.05 ¢ 800 .  36.9 44 .9 11.0 - 57.8 - 70.2
2541 16 0.002 800 30 30 1.1 7.7 29.8
2455 16 0.05 871 31.8  32.1 1.3 = 57.4 78.1
2456 - - 16 - 0.002. 871  17.6- 7.6 1.0 - - 47.7 41.6
24,57 16 - 0.05 982 17.9  17.9 0.8 59.8 = 64.6
2536 46 0.002 600 - 26.4 ' 76.9 45,8 47.6 0 32.6
2535 46 0.002 650 = 26.5 . 74.6 355 .38.5  31.0
2537 - 46 0.002 700 26.5 56.2 15.6 34,3  31.0

 

aAnnealing designations: 121 = annealed 1 hr at 1177°C; OOO = as
received (40% cold work); 16 = annealed 100 hr at 871°C; 46 = annealed 1 hr
at 1177°C plus 1000 hr at 650°C.
 

 

 

28

- ORNL-DWG 69-14390

 

80 T 5

 

o 1 hr AT H77°C

 

 

60

 

100 hr AT B74°C

- NP
30 _“ \\\

]

o

 

 

 

 

FRACTURE STRAIN {%)
H
o
1.

 

AS RECEIVED e |

 

jop

 

10

 

 

 

 

 

 

 

 

 

 

 

 

] 00 200 300 400 500 600 700 800 900 1000
TEST TEMPERATURE {°C)

Fig. 15. Fracture Strains of Alloy 21545 (0.49% Ti) in Tensile
Tests at a Strain Rate of 0.05 min~1.

increased gradually with increasing temperature. The material annealed
for 100 hr at 871°C had a fracture strain of 52% at 25°C with a minimum
ductility of 32% at 600°C. After an anneal of 1 hr at 1177°C, the frac-
ture strain was 75% at temperatures between 25 and 600°C and then dropped
precipitously with increasing temperature. Three samples were annealed
for 1 hr at 1177°C and aged for 1000 hr at 650°C. This anneal did not
change the strength properties appreciably, but did increase the fracture
strain. Alloy 104 (0.55% Ti) was given the same anneal before testing,
and & comparison between the results for this alloy in Table 2 and those
for heat 21545 (0.49% Ti) in Table 13 shows that the commercial alloy is
slightly weaker and that both alloys have comparable fracture strains.
The results of tensile tests on irradiated samples of heat 21545
are given in Table 1l4. The results for samples irradiated at 650°C or
below were used in constructing Fig. 16 for a strain rate of‘b.OOZ min=1.
The samples that were irradiated as received and those irradiated after
annealing 100 hr at 871°C had comparable fracture strains. The samples
~annealed for 1 hr at 1177°C were more ductile. A comparison with thev'
Table 14, Tensile Pr0perties of Alloy 21545 (0.49% Ti) after Irradiation™

 

 

 

 

 

- b | s , ‘ ‘
. Anneal " oR Strain ‘ Stress, psi . Reduction
Specimen o Temperature, °C ‘ 2k Elongation, % :
Numbe? B;zzie‘ | Irra@iation Test (Sizfl) Yield gig:?ize Uniform  Total 1n(2§ea
| | x 103 x 103
1554 000 . 600 550 0.002 111.6 134.9 11.5 15.1 14.7
1553 000 o 00 600 0.002 110.1 127.9 11.2 13.4 11.7
1551 000 . - 650 650 0.05 107.0 124.9 12.5 5.1 16.2
1552 000 . 650 - 650 0.002 107.7 113.3 5.1 9.5 11.7
2524 12y . 590 650 - 0.05 29.0 64 .4 36.0 37.0 32.0
2525 121 . . 590 650 0.002 28.9 58.8 21.6 21.9 17.1
2548 121 . 600 650 0.002  30.0 61.3 21.1 21.4 18.3
2550~ 121 . 600 760  0.002 29.2 51.7 10.3 13.6 19.7
2953 121 760 650 0.002 22.4 22.4 1.0 1.4 3.2
5954 - . 12y . .760 . = 760 0.002 - 15.8 15.8 0.7 0.7 0.81
1560 - 16 . 600 . 550 0.002 42.9 85.3 12.1 12.4 20.5
‘1557 16 600 . 600 0.002 39.3 69.1 13.3 13.4 8.6
1558 . 16 600 650 0.05 38.2 71.5 14.7 15.1 16.2
1559 - 016 600 - - 650 0.002 - 38.8 62.4% 9.7 - 9.7 13.2
37.8 37.8 1.4 4.1 7.1

561 . 16 600 760  0.002

 

'airradiatedjto;a‘thermal fluence of,2:to 5 x 1020 neutrons/cm®.

annealed 1 hr at 1177°C,
as received (40% cold work),
annealed 100 hr at 871°C.

.bAnnealihgldesignétionég 121
I A ~ 000
16

I i

6¢
 

 

30

ORNL-DWG 69-14389

o AS RECEIVED
a 100 hr AT 871°C
g { hr AT 1177°C

N\
N

40

o, &, 0 UNIRRADIATED
e, a, & IRRADIATED

 

 

 

 

 

30

 

FRACTURE STRAIN (%)

 

 

 

 

 

a

 

 

 

 

 

 

 

 

550 600 650 700 750 800 850
TEST TEMPERATURE (°C)

Fig. 16. Fracture Strains of Alloy 21545 (0.49% Ti) in Tensile
Tests at a Strain Rate of 0.002 min~1.

results for alloy 104 (0.55% Ti) in Tables 3 and 4 shows that the com-
mercial alloy was more ductile and slightly weaker. Two samples in
Table 14 were irradiated at 760°C. The samples had much lower fracture
strains than those irradiated at 650°C or below, indicating again & very
strong effect of irradiation temperature.

Samples of heat 21545 were subjected to creep tests in several
annealed conditions; the results of these tests are presented in Table 15 i
for unirradiated samples and in Table 16 for irradiated samples. The
stress-rupture properties at 650°C in the unirradiated condition are
.shown in Fig. 17. The as-received material had the longest rupture life
at a given stress level and the lowest fracture strain. Annealing at
871°C produced a material with fine grain size that failed in much
shorter times with fracture strains of about 50%. Annealing for 1 hr at
1177°C gave longer rupture lives with intermediate fracture strains of
about 30%. Aging this material at 650 or 760°C before testing had only

minor effects on the rupture life or the fracture strain. However, close
o

31

Creep Properties of Unirradiated Alloy 21545 (0.49% Ti)

 

 

3422

 

 

Table 15.
Anneal® " Rupture Minimm Reduction
Specimen Test Before Stress Iife Creep Elongation Ar
Number Number (psi) ~ Rate (%) n srea
Test (hr) (4/nr) ! (%)
x 103
Test Temperature 650°C -
1564 5535 000 70 174.1  0.0093 14.1 20.5
1546 5578 000 55 1317.8  0.0009 18.8 31.2
3441 6294 121 70 0 41.1 32.6
3404 6295 121 62 17.7 0.288 27.2 26.7
5250 6417 121 60 26.8 0.062 27.0 23.5
3406 6296 121 55 39.9  0.148 22.6 17.0
5251 6418 121 52 - 170.9 0.015 24.8 27.2
3416 6297 121 47 197.6 0.037 . 26.3 27.7
5244 6419 121 44 466.1 0.011 26.3 22.6
3418 6298 121 40 564.2 0.015 31.3 29.7
5246 6420 121 37 1478.6  0.0035 34.6 31.7
3413 6299 121 35 1720.9  0.0063 30.1 30.0
5247 6421 121 30 4383.2 0.0013 22.2 26.7
1822 5564 816 40 248.1 0.0688 34.4 20.5
1771 5409 16 55 20.1 1.37 52.5 47.3
1773 5580 16 - 40 242.6  0.120 - 59.4 56.6
1768 5449 16 .30 1171.5 0.017 56.7 51.5
2539 6167 46 40 - 893.7 0.0129 38.7 32.1
2538 6161 L6 ' 32.4 2077.6 0.0041 30.8 24,7
2528 7716 148 40 708.9 0.0318 33.3
5534 7715 148 55 26.7 0.625 34.4 34.2
3400 7718 156 40 527.4  0.0334 36.1 41.4
2533 7717b 156 55 33.7 0.570 32.1 31.8
5508 6384b 103 70 1.9 . 39.2 30.6
5509 6385b 103 63 5.57';4.53 46,1 33.1
5510 6386b 103 55 5.7 2.50 7.4 34.3
5511 6387b 103 47 50.3 0.50 43.9 65.9
5512 fi6388b 103 40 177.2  0.165 55.3 45,1
5513 6428 103 32,4 438.8 - 0.050. 40.1 50.3
' o ' Test Temperature 760°C
3424 7041 121 30 28.7 0.519 -  37.8 32.6
3431 7042 121 25 71.1 0.213 39.4 35.0
6358 121 20 . 280.0 0.060 - 36.2 35.1

 

 ®Annealing designations: 121 = annealed 1 hr at 1177°C; 000 = as
received (40% cold work); 816 = annealed 8 hr at 871°C; 16 = annealed

100 hr at 871°C; 46 = annealed 1 hr at 1177°C plus 1000 hr at 650°C;

148 =-annealed 1 hr at 1177°C plus 1000 hr at 760°C; 156 = annealed 1 hr

at 1177°C, 510 hr at 550°C, 1500 hr at 760°C, and 1 hr at 1177°C;

103 = annealed 100 hr at 871°C, 5600 hr in fluoride salt at 650°C.

belten Salt Reactor Experiment surveillance controls.
 

 

32

Table 16. Creep Properties of Alloy 21545 (0.49% Ti) after Irradiation

 

 

 

 

a .
Specimen Test Anneal Irradiation Stress - Rupture Creep - Elongation Reduction
Number Number Before Temperature (psi) Life Rate (4) in Area
Irradiation (°c) P (hr) (%/r) (%)
x 103
Test Temperature 650°C

5235 R-676 121 650 60 12.9 0.445 7.0 28.6
2558 R-415 121 600 63 8.6 0.237 3.1 19.6
2554 R-328 121 600 55 24.2 0.1 4.0 23.2
2549 R-325 121 600 47 264.8 0.022 10.4 18.3
5237 R-653 121 650 50 65.0 0.023 16.8 23.8
5242 R-657 121 650 50 178.0  0.011 8.0 19.5
2527 R-258 121 590 47 87.4 0.071 10.0

3402 R-909 121 654 47 33.5 0.258 13.2 31,1
5236 R-644 121 650 bh 285.0  0.0040 18.9 26.3
2553 R-249 121 600 40 780.6 0.0091 16.5

2526 R-232 121 590 40 453.8 0.010  11.3

2529 R-348 121 590 40 261.2  0.017 8.2 17.3
5238 R-654 121 650 40 574.6  0.016 13.4 22.8
3426 R-771 121 760 7 1.6 0.019 0.33 5.4
3414 R-913, 121 760 2 6.1 0.056 0.42 11.6
5240 R-841 121 650 27 355.5 0.0081 4.3 9.1
5239 R-817 121 650 27 109.4  0.0070 1.3 5.2
3409 121 704 7 1531.8 8.77

3397 121 760 21.5 223.4

1767 R-352 16 650 40 4.9  0.373 2.0 4.8
1563 R-125 16 600 40 72.6  0.077 6.0

1769 R-353 16 650 35 43.4  0.039 3.4 6.4
1562 R-127 16 600 32.4 356.9  0.019 7.3

1770 R-354 16 650 30 111.7 0.019 3.3 4.0
1774 R-416 16 650 27 430.8  0.0053 3.9 4.8
1772 R-421 16 650 25 1056.5 0.0056 6.5 ©10.7
2448 R-346 16 150 55 3.5 1.59 6.1 10.6
2449 R-378 16 150 47 16.1  0.40 7.8 11.1
2450 R-347 16 150 40 66.4 0.13 9.9 12.9
2451 R-362 16 150 35 193.9  0.037 16.2 9.9
2452 R-419 16 150 30 596.0  0.0075 9.9 11.4
2453 R-420 16 150 27 721.8  0.0059 6.9 10.9
9100 R-317° 15 650 47 2.8  0.46 1.6
9099 R-314° 16 650 40 13.1  0.085 2.7
9102 R-319° 16 650 32.4 51.1 0.013 3.4
9103 R-323° 16 650 27 124.1  0,0089 3.7
3430 R-829 113 760 27 7.8 0.024 0.32 13.5
3433 R-906 113 760 21.5 108.4  0.0052 1.3 6.3

Test Temperature 760°C

2556 R-39% 121 650 25 351.0 0.,0058 . 13.1 7.3
2555 R-329 121 600 20 127.4  0.0328 13.3 12.9
5234 R-652 121 650 20 200.2  0.037 13.8 13.3

a 121 760 25 0.50 0.50 0.66

d 121 760 20 1.5 0.42 1.0

a 121 760 1 94.0 0.03% 4.7

a 22 760 12 101.0 0.040 5.2

a 22 760 11 296.0  0.020 7.0

 

aA::meaa.l:i.ng designations: 121 = annealed 1 hr at 1177°C; 16 = amnealed 100 hr at 871°C;
113 = annealed 1 hr at 1177°C plus 100 hr at 650°C; 22 = annealed 1 hr at 1177°C plus 100 hr
at 871°C.

bJin:meeal.led 5000 hr at 760°C after irradiation.

CMolten Salt Reactor Experiment surveillance specimens exposed to MSRE core for 5600 hr
at 650°C.

9. G. Harman, personal communication, 1970.

ik
- C

L 1]

33

ORNL-DWG €9-14388

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

(“03) - !K l a{{4}
60 (271% 57 I
| 34 '
\:}\\ l a{9
53\ 23| N
50 \\\ .\ 0?5 :
N
.'i%\ 26
2 34 j;‘:'” 5
% TR qugs
\
o \\ 30 '3|
‘1.3 ~\__ p >0
x 30 T
E o ANNEALED thr AT 4§77°C SINT TN
a AS RECEIVED
o ANNEALED 10Ohr AT B71°C
20 ® ANNEALED ihr AT 1177°C,
1000 hr AT 650°C
o ANNEALED {hr AT {{77°C,
10 000 hr AT 760°C
1o° g s0! 102 10° 104

RUPTURE TIME (hr)

Fig. 17. Creep-Rupture Properties of Unirradiated Alloy 21545
(0.49% Ti) at 650°C. The number by each datum point designates the
fracture strain. _

examination of the data in Table 15 reveals that the aging treatment of
1000 hr at 760°C increased the minimum creep rate.

The results of creep-rupture tests at 650°C after irradiation are
shown in Figs. 18 through 20 for material annealed 1 hr at 1177°C. The
irradiation temperatures are indicated by each datum point. At irradia-
tion temperatures of 650°C or bélow, the rupture life was reduced only a
small emount, and the minimum creep rate was unchanged. At an irradia-
tion temperature of 760°C and test temperature of 650°C, the rupture
life decreased at least one order of magnitude, and the minimum creep
rate increased one order of,fiagnifiude.' The corresponding fracture
strains shown in Fig. 20 are 3% and higher for the samples irradiated
at 650°C or below and about 0.5% for samples irradiated at 760°C. Tests
R-841 and R-817 (Table 16) demonstrate the ability of annealing at 760°C

- .after irradiation. to degrade;the-properties_at-650°C;2 This observation

supports the premise]that?afStrubtural chahgé'during annealing at 760°C

is responsible for thegpobr_prdpeities'after irradiation. A comparison

 of the results for heat 21545Win Figs. 18 through 20_With those in Fig. 14
for alloy 104 (0.55% Ti) shows good qualitative agreement. '

The results of postirradiation creep tests on material having a fine

grain size (annealed at 871°C) are also given in Table 16. The fracture
 

 

ORNL-DWG 69-14387

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

{x103)
500 -
60 650 ¢—e
N N\JUNIRRADIATED
600.\T\
. o
50 ! 650 oL ><e 650
654 ¢| 5904N 6009

- 65|0 5090
a 40 590 ee 1¢.600
e I 650
2 I
i IRRADIATED N
= 30 TIT ~ N
» 7608 |l|* 760 | 704

20 * 760

o UNIRRADIATED
10 s IRRADIATED
o 0 1 2 3 4
10 10 10 10 10

RUPTURE TIME (hr)

Fig. 18. Stress-Rupture Properties of Alloy 21545 (0.49% Ti) at
650°C. All materisl annealed 1 hr at 1177°C. Number by each datum point
designates the irradiation temperature in degrees Celsius. :

ORNL-DWG 69-44386

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

(x40%) /{/ )
9
600e A
60 p 2 s 650 1
6007
650 ,JT °
50 650 e——e4 411+
1111600e 70| | |$590 | 654
201 60 50
Zao IR0 250
2 b
0 d
£ 30 o
- P :
» 1 7604 e 760
20
° UNIRRADIATED
¢ |RRADIATED
10
0 = =3 -2 y 0
10 10 10 {0 {0

MINIMUM CREEP RATE (%/hr)

Fig. 19. Creep Properties at 650°C of Alloy 21545 (0.49% Ti). All
material annealed 1 hr at 1177°C. Number by each datum point designates
“irradiation temperature in degrees Celsius.
35

ORNL~-DWG 69-14385

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

20
65'0‘\ |I
NEDC  leg50
- 16 J '
8 N ]
% . \-(;5? * 654
i
E 2 5904 N |
1 17| |ees0
5 g 650 N\
350 &2
o 5901 N *650
T N L~
4 ' 600 b
\Io-/’
760 | |760 f°|°
o Io(]
1073 402 10! 10° 10!

STRAIN RATE {%/hr)

 Fig. 20. Fracture Strain of Alloy 21545 (0.49% Ti) at 650°C after
Irradiation. All material annealed 1 hr at 1177°C. Number by each datum
point designates the irradiation temperature in degrees Celsius.

strains fdr'these.samples were generally lower than those for the mate-
rial amealed 1 hr at 1177°C. The stress-rupture properties of this
material are shown in Fig. 21. In addition to the rupture life being
low for the fine-grained, unirradiated material, the rupture life was
reduced at least 50% by irradiation.

Some of the fine-grained material of heat 21545 was included in the
surveillance program for the MSRE. The detailed results of these tests

ORNL-DWG 69-14384

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

(x10%) —
171 LN
_ o UNIRRADIATED
60 1 e IRRADIATED
|§O | \\
50 oo
: 15077 1 N
| e N
3 40 620 !500-600\‘-:\\
- - {
& 650+ | 30 oo,
il .
650 1. - §~
£ a0 . } 1504 i
@ 6500 | p150
. o
: 650
20
10
0
ok ol - 102 10® 0%

RUPTURE TIME (hr)

Fig. 21. Stfess-RuptfiiéTfroperties of Alloy 21545 (0;49%7Ti) at
650°C. Annealed 100 hr at 871°C to produce a fine grain size. Number by
each datum point indicates the irradiation temperature in degrees Celsius.
 

36

were reported.previously6 end will be discussed only briefly in this
report. These samples were exposed to molten fluoride salt for 5600 hr
at 650°C while receiving a thermal-neutron fluence of 4.1 X 1020 |
neutrons/cm?. The stress-rupture properties of these sampleé are shown
in Fig. 22. The unirradiated control samples were exposed to barren
fuel salt for an equivalent length of time and show a slight reduction
in rupture life (compare Figs. 17 and 22). Irradiation decreased the
rupture life about an order of magnitude, and the fracture strains
ranged from 1.6 to 3.7%. Since the metal was quite compatible with the
salt, we attribute the changes to the long thermal anneal and the neutron
éxposure. ‘ |

Several creep tests were run at 760°C on both the unirradiated and
irradiated materials. The results of these tests are given in Tables 15
and 16, and the stress-rupture properties are shown graphically in
Fig. 23. The rupture life at 760°C was not reduced by irradiation at
650°C or below, but was reduced drastically by irradiation at 760°C. The
- data in Table 16 also show that the higher irradiation temperature
increased the minimum creep rate and decreased the fracture strain.

The microstructures of several samples were examined after testing.
Typical photomicrographs of alloys 100, 102 (0.39% Ti), 104 (0.55% Ti),
and 107 (1.04% Ti) after extended aging at 650°C are shown in Fig. 24.
The carbide precipitation along grain boundaries and within the matrix
was light in alloy 100, heavier in alloy 102, and still heavier for
alloys 104 and 107. The fracture of alloy 100 after creep testing at
40,000 psi and 650°C is shown in Fig. 25. Numerous intergranular cracks
were present, and the fracture was intergranuler. The short time of
about 50 hr at temperature was not sufficient to cause cafbide precipita-
tion. The fracture of alloy 104 (0.55% Ti) after testing at 40,000 psi
and 650°C is shown in Fig. 26. Details of importance are the carbide
stringers, general carbide precipitation, intergranular cracking, and a
predominantly intergranular fracture. This sample was at température
for 1000 hr, and copious amounts of carbide precipitate formed. Photo-
micrographs of the fracture of alloy 107 (1.04% Ti) are shown in Fig. 27.

 

6H. E. McCoy, Jr., An Evaluation of the Molten-Salt Reactor Experi-
ment Hastelloy N Surveillance Specimens — Second Group, ORNL-TM 2359 (1969).
37

ORNL-DWG 69-14383

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

3 o
{X10%} (39)-\ ] ”
\\ ‘(‘46)
N
60 . \\
(47}
50 {1.6) \\
~e \,(:M
e N
o I
5 \\~h (2.7} h
& a0 o \25’
wn
i \
W N, (34) N 40)
= N
" 30 '\ )
\\\(2.7) h
20
o UNIRRADIATED
e IRRADIATED
10
0 _
100 10! 102 103

RUPTURE TIME (br)

. Fig. 22. Stress-Rupturé Properties of Alloy 21545 (0.49% Ti) at
650°C. All material amnealed 100 hr at 871°C and exposed to fluoride
salt for 5600 hr at 650°C. Number by each datum point indicates fracture

strain.

ORNI-DWG 69-14382

 

(x10°)
_ 20
-g :
w
L)
x 10
k- _
w 8 _ o b :
[ ° UNIRRADIATED
5 * IRRADIATED
a | |
107! 109 10! 102 10°

RUPTURE TIME (hr)

. Fig. 23. Stress-Rupture'Properties of Alloy 21545 (0.49% Ti) at
.760°C. All material annealed 1 hr at 1177°C. Number by each datum
point indicates the irradiation temperature in degrees Celsius.
             

 

 

   

 

 

 

 

 

 

“r Ya79314

L))
o

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

i

Alloy 107

 

 

 

 

 

 

 

 

 

 

Fig. 24. Microstructures of Several Titanium-Modified Alloys Annealed 1 hr at 1177°C, Aged
1000 hr at 650°C, and Stressed 1000 hr at 650°C at 32,400 psi. Etchant: glyceria regia, 500X.
Reduced 22.5%.

 
C

39

Y=97291

 

'(a)

s Y=97292

———

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

%‘f : : . L

- Fig.. 25. Miecrostructures _of-'A]_.lOy 100 after Annealing 1 hr at 1177°C
and Testing at 40,000 psi and 650°C. - (a) Fracture, 100x; (b) typical,
500x. Etchant, glyceria regia. B T o . :
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

     

 

(b) typical, 500x.

1 Y=97294

 

Photomicrographs of Alloy 104 (0.55% Ti) after Annealing
1 hr at 1177°C and Stressing at 40,000 psi at 650°C.
Etchant, glyceria regia.

(a) Fracture, 100x;

 

N

 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig. 27. Photomicrographs of Alloy 107 (1.04% Ti) ‘after Annealing
‘1 hr at 1177°C and Testing at 650°C and 40,000 psi. (a) Fracture, 100X;
(b) typical, 500X. Etchant, glyceria regia. S -
 

 

42

The sample was stressed at 40,000 psi at 650°C and fractured after
1490 hr. The grain size was quite small, and there were large quantities
~of precipitate pfesent. The fracture was predominantly intergranular.

| :Ailoy 21545 (0.49% Ti) was examined metallographically after being
stressed at 40,000 psi at 650°C. The fracture was predominantly inter-
granular (Fig. 28). The sample was at temperature 564 hr, and some very
fine précipitation occurred in the matrix and at grain boundaries. The
stringers of precipitate were present in the annealed material before
testing (Fig. 4).

Alloy 100 (< 0.01% Ti) was examined after irradieation and creep
testing at 650°C. Typical microstructures are shown in Fig. 29. This
sample failed after only 1.1% strain (Table 7). There were a few inter-

-granular cracks, but the deformation seemed restricted largely to the

zone near the fracture. Some fine carbides were formed at the grain
boundaries and in the matrix. The microstructure of alloy 104 (0.55% Ti)
is shown in Fig. 30 after irradiation and creep testing at 650°C. This
alloy failed after 10.0% strain; the fracture was largely intergranular.
Alloy 107 (1.04% Ti) failed after 11.1% strain. The microstructure,

shown in Fig. 31, consists of an intergranular fracture with some carbide
precipitation. ‘

Alloy'2l545 (0.49% Ti) was examined after exposure to the MSRE core
and control facility. The material was annealed at 871°C before exposure
to give a fine grain size (Fig. 4). The samples were then exposed to
fluoride salt at €50°C for 5600 hr. Some of the samples were exposed to
barren fuel salt for controls, and others were irradiated in the MSRE
core. Microstructures of an irradiated sample that was tested at 25°C
are shown in Fig. 32. The fracture was transgranular. There was some
edge cracking due partially to slight changes in composition near the
surface. The comparative control sample is shown in Fig. 33. The frac-

| ture was transgranular, and there was less edge cracking. Another irra-
diated sample was tested at 650°C; typical photomicrographs are shown
in Fig. 34. The fracture was intergranular, and there were a few inter-
granular cracks near the fracture. There were some edge cracks. The
control sample shown in Fig.i35 was more ductile, with 40.8% strain com-
pared with only 6.2% for the irradiated sample shown in Fig. 34. The

higher strain led to extensive intergranular cracking.
 

)

(D)

43

g e e e g e e e e R T Y7207
i ! o " ,‘."...7 S o f” . . ' ',"W . -
| " s ey 1 - b o

   
   
  

Y=-97298

' Fig. 28. _Photomicrogf&bhs'of'Alloy.21545 (0.49% Ti) after Annealing

1 hr at 1177°C and Testing at 650°C and 40,000 psi. - (a) Fracture, 100x;

(b) typical, 500X. Etchant: glyceria regia.
 

 

| / R-34188

 

 

 

Fig. 29. Photomicrographs of Alloy 100 (< 0.01% Ti) after Annealing
1 hr at 1177°C, Irradiation for 1000 hr at 650°C to a Thermal Fluence of
2.5 x 102° neutrons/cmz, and Testing at a Stress of 32,400 psi at 650°C.
Fractured in 105.%4 hr with 1.1% strain. (a) Fracture, 100X; (b) typical,
500x. Etchant: aqua regia.
&)

45

 

 

 

 

 

 

 

 

 

 

 

(®)

 

 

Fig. 30. Photomlcrographs of Alloy 104 (0.55% Ti) Annealed 1 hr at
1v7°c, Irradla.ted for 1000 hr &t 650°C to a Thermal Fluence of
2.5 x 1020 neutrons/cm?®, and Stressed at 32,400 psi at 650°C. Fractured
in 1446.3 hr with 10.0% straln. (a) Fracture, 100x; (b) typlca.l 500)(.
Etchant: aqua regia. ‘ '
 

 

 

 

 

 

46

e

 

 

 

 

 

 

 

 

 

 

 

 

 

) bt

Fig. 31. Photomicrographs of Alloy 107 (1.04% Ti) Annealed 1 hr at
1177°C, Irradiated for 1000 hr at 650°C to a Thermsl Fluence of
2.5 X 1020 neutrons/cm?, and Stressed at 32,400 psi at 650°C. Fractured
in 2497.2 hr with 11.1% strain. (a) Fracture (flattened end caused by
handling), 100x; (b) fracture (blunted tip caused by handling), 500x..

My

e
 

 

(2)

Fg}-{- 4

 

 

 

 

 

 

Fig. 32. Photomicrographs of a Titanium-Modified Hastelloy N Surveillance Sample (Heat 21545)
Tested at 25°C at a Strain Rate of 0.05 min~!. Exposed in the MSRE core for 5500 hr at 650°C to a

thermal fluence of 4.1 X 1029 neutrons/em?.

(a) Fracture, 100x; (b) edge of sample about 1/2 in. from

fracture, 100x; (c) edge of sample showing edge cracking, 500xX. Etchant: aqua regia.- Reduced 8%.

L7
 

 

 

 

 

48

Y=-83570

 

 

A w"i’ RS ' - St TN S T e

Fig. 33. Photomicrograph of the Fracture of a Titanium-Modified
Hastelloy N Surveillance Sample (Heat 21545) Tested at 25°C at a Strain
Rate of 0.05 min™*. Exposed to a static fluoride salt for 5500 hr at
650°C before testing. Note the shear fracture and the absence of edge
cracking. 100X. Etchant: glyceria regia. '

 

i
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

6%

 

 

 

 

 

 

 

 

 

 

 
  

 

(c) = R : : / v e
- Fig. 34. Photomicrographs of a Titanium-Modified Hastelloy N Surveillance Sample (Heat 21545)
Tested at 650°C at a Strain Rate of 0.002 min~'. Exposed in the MSRE core for 5500 hr at 650°C to a.
thermal fluence of 4.1 X 10?° neutrons/cm?. Failed after 6.2% strain. (a) Fracture, 100x; (b) edge

of sample about 1/2 in. from fracture, 100x; (c¢) edge of sample showing edge cracking, 500x. Oxide
formed during the tensile test. Etchant: aqua regia. Reduced 10%.
 

 

 

50

 

Fig. 35. Photomicrograph Showing a Portion of the Fracture of a
Titanium-Modified Hastelloy N Surveillance Sample (Heat 21545) Tested at
650°C at a Strain Rate of 0.002 min~l. Failed after 40.8% strain.

Exposed to static fluoride salt for 5500 hr at 650°C before testing. 100x.
Etchant: glyceria regia. |

Alloys Containing Zirconium

Three groups of alloys were studied that contained various amounts
of zirconium. A series of 2-1b laboratory melts containing from less
than 0.03 to 1.18% Zr was made from virgin melting stock. These alloys
are designated 100 and 108 through 112; their chemical compositions are
given in Table 1. Typical microstructures of these alloys after a l-hr
anneal at 1177°C are shown in Fig. 36. The addition of the first 0.1% Zr
(alloy 108) markedly decreased the grain size. Further additions up to
the maximum 1.18% Zr caused only slight additional refinement in grain
size. The second group consisted of two alloys that contained 1% Zr
with additions of 0.1 and 1.0% Re. The third series of alloys consisted
‘of two 100-1b heats from Special Metals Corporation that contained 0.05
and 0.35% Zr, designated heats 21555 and 21554, respectively. The alloy
   

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig. 36. Typical Photomicrographs of Alloys 100 and 108 Through 112. Annealed 1 hr at 1177°C.
(a) Alloy 100 (< 0.03% Zr), (b) alloy 108 (0.10% Zr), (e) alloy 109 (0.28% zr), (d) alloy 110 (0.53% zr),
(e) alloy 111 (0.75% Zr), and (f) alloy 112 (1.18% Zr). 100X. Etchant: glyceria regia. Reduced 42.5%.

TG
 

 

52

with the higher zirconium content from Sfiecial Metals Corporation was
initially forged at 1177°C and cracked profusely. A second alloy was
melted and forged at 1066°C, and a good product was obtained. This
experience gave the first indication that zirconium caused the alloy to
be hot-short and that it would likely have poor weldability. Typical
photbmicrographs of these two alloys are shown in Figs. 37 and 38. As
received, the material contained 40% cold work; annealing at 871°C
produced a banded structure with a fine grain size and numerous carbides.
The banding is likely due to the working pattern, but there was also
some variation in the density of the carbides.

Alloys 100 and 108 through 112 were tested to determine their ten-
sile properties. Samples of each alloy were annealed 1 hr at 1177°C and
“aged for 1000 hr at 650°C to match the thermal history of samples that
were irradiated. The results of tensile tests on the unirradiated
samples are given in Table 17. The variation of‘the strength parameters
at 650°C with zirconium content is shown in Fig. 39. The yield strength
was not affected appreciably until the zirconium level exceeded 0.3%,
‘but it increased markedly abofie this level. The ultimate tensile strength
increased about 10% with the addition of 0.1% Zr, but very little further
strengthening occurred as the zirconium content increased. In the
absence of aging, the strength was not affected by zirconium content.

The fracture strains are shown in Fig. 40 as a function of zirconium con-
tent. A very distincet maximum for aged material occurred at a concentra-
tion of about 0.3% Zr. The fracture strain was almost doubled by the
addition of even 0.1% Zr. The fracture strain at the lower strain rate
was lower, & trend that is usually observed. In unaged material, a
maximum occurred in the fracture strain for zirconium concentrations of
0.5 to 0.7%.

The same experimentfil alloys were irradiated at 650°C to a thermal
fluence of 2.5 X 102° neutrons/cm®. The results of tensile tests on
these samples after irradiation are summarized in Table 18, and the
variations of the fracture strains with zirconium content are shown in
Fig. 41. The postirradiation results also show & ductility maximum at
about 0.3% Zr. The results for the irradiated and unirradiated samples
are compared in Fig. 42, where the ratios of the various tensile properties
53

 

 

 

Fig. 37. Typical Photomicrograph of Alloy 21555 (0.05% Zr).
Annealed 1 hr at 1177°C. 100x. Etchant: glyceria regia.

Y-72251

 

~ Fig. 38. Typical Photdmicrograph'of Alloy 21554 (O.BS%IZr).
Annealed 1 hr at 1177°C. 100X. Etchant: glyceria regia.

“f
 

 

 

 

54

Table 17. Tensile Properties of Several Unirradiated
Zirconium-Modified Alloys at 650°C

 

. Reduc-
Strain Stress, psi L s Heat
Elongation tion in
2 -
Ultimate —-—-5-————L—fi Treat

Ares
Tensile Uniform Total (%) ment

Alloy Specimen e
Number Number (min'l) Yield

 

x 102 x 103
108 3870

0.05 25.7 68.3 28.6 29.8 23.7 a
109 /3880 0.05 29.4 178.6 33.7 34.7 26.8 a
110 3890 0.05 25.9 75.6 42,8 4.8 30.6 a
111 3900 0.05 26.7 75.5 42.3 44,6 39.9 a
112 3909 0.05 26.5 69.6 32.4 34.4 33.3 a
108 2084 0.05 28.9 86.4 - 46.1 48.9 42.6 b
108 2078 0.002 29.2 86.2 47.3 50.0 47.2 b
109 2061 0.05 29.1 85.9 54.8 57.7 50.9 b
109 2062 0.002 29.4  84.7 52.5 56.2 47.3 b
110 2105 0.05 43.8 85.8 49.0 56.8 53.1 b
1130 - 2114 0.002 44,6 87 48.9 51.9 48.7 b
111 2132 0.05 50.5 92.2 45.6 50.3 53.2 b
111 2134 0.002 52.9 80.9 35.6 b 4 47.5 b
112 2146 0.05 51.2 89.1 26.9 28.8 28.1 b
1312 2148 0.002 55.2 89.3 22.6 23.3 25.8 b

 

®Annealed 1 hr at 1177°C.
Prnnealed 1 hr at 1177°C plus 1000 hr at 650°C.

fa
55

ORNL—-DWG 69-14381R2

 

(x10%)

90 T —~
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70 —— _ 4
60
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% 80 //g:—-—f ®
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40 YiELD//

 

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20
e STRAIN RATE =0.05 min~%, AGED
0 — o STRAIN RATE =0.002 min=!, AGED
: A STRAIN RATE =0.05 min—', UNAGED
0 | | | 1 |
o 0.2 0.4 08 0.8 1.0 1.2 1.4

ZIRCONIUM CONTENT {wt %)

Fig. 39. 'Influence of Zirconium Content on Tensile Properties
after Annealing 1 hr at 1177°C and Aging 1000 hr at 650°C.

Table 18.. Tens1le PrOperties of Several Zirconium-Modified Alloys at
' 650°C after Irradiation®

 

 

 

 

Alloy Specimen Strain _ Stress, 1.331 Elongation, % Re.aductlon

Number Number (overy  yiela UitiMte G Total ()
(min“l) 1ie Tensile niiorn - (%)
| N X 103 x 10% | |

108 2075 0.05 - 33.9 . 72.6 28.0 29.0 - 18.7
108 2087 0.002  35.3 - 67.4 22.1 23,5 18.9
109 2074 0.002  37.2 65.2 24.5 - 29.8 28.5
110 - 2109 0.05  4l.2 63.4 16.5 18.5 17.9
110 . 2116 0.002 = 43,4 58.1 13.4 17.1 13.6

111 2141 0.002  49.8 60.8 8.5 . 10.0 12.1
112 2147 0.05 53,9 . 74.6 - 14.9° 17.2 - 13.4
112 2152 0.002 - 56.5 - 68.7 7.1 7.7 11.9

-

 

Annea.led l hr at 1177° ¢ before 1rrad1at10n at 650°C to a thermal
fluence of 2.5 X 1020 neutrons/cm?.
 

 

‘ : ORNL-DWG 69-14380R3 ORNL-DWG 69~-14379R
60 : 32

N S

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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® STRAIN RATE =0,05 min—!, AGED 4 © STRAIN RATE = 0.002 min
o5 © STRAIN RATE =0,002 min~', AGED
a - STRAIN RATE =0.05 min~!, UNAGED \ \
0
t ‘ 0 0.2 0.4 0.6 0.8 1.0 1.2
20 ZIRCONIUM CONTENT (w! %)
0 02 04 06 08 1.0 1.2 1.4
ZIRCONIUM CONTENT (wt%) Fig. 41. Influence of Zirconium Content on

] Fracture Strain at 650°C after Irradiation. Annealed
Fig. 40. Influence of Zirconium Content on 1 hr at 1177°C before irradiation.

Fracture Strain. The unaged samples were annealed
1 hr at 1177°C and the sged samples were annesaled
1l hr at 1177°C and aged 1000 hr at 650°C before
testing.

9¢

 
-

57

) ORNL- DWG 63-14378R
1.4 : T I T

_ © YIELD STRESS

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

4=a A ULTIMATE STRESS
-
1.2 5 \\ o TOTAL ELONGATION
[ \|
[ \
! \,
1.0 N - S O
¥ g N
é 5
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w E A a a A
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0.4 . N :
\w?“‘Fs~__ °
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ZIRCONIUM CONTENT (wt %)

Fig. 42. Effects of Zirconium on the Unirradiated and Irradiated
Tensile Properties at 650°C and a Strain Rate of 0.002 min~l.

are also shown. The yield strength was increased slightly by irradiation
in alloys that contained 0.10 and'df28% Zr (alloys 108 and 109) and
decreased slightly in the other alloys. The ultimate tensile strength
was decreased about 30% in.éaéhfalloy. The fracture strain was reduced
about 50% by irradiation.in'alloysrthat contained 0.3% Zr or less and
reduced about 70% for alloysrfhat contained more zirconium.

- Samples of alldys 100 and 108-through 112 were also irradiated at

. 43°C to a thermal fluencelofj9”x lO19 neutrons/cm?. These samples were

tested at 650 and 760°C;-the résu1ts are given in Table 19. The frac-

ture strains are shown in Fig. 43 as a function of zirconium concentra- -

tion. The peak in fracture strain at 650°C with zirconium concentration

still occurs, but less sharply and at a different concentration than

fobservéd previously (Fig;;Al); ~The fracture strains are,also higher in

' the experiment at lower temperature. We do not know whether to attribute

these differences to the variation in irradiatibn temperature or in

fluence. However, the circumstantial evidence indicates that the irra-

diation temperature is the most important factor. Recall that the yield
strengths of the aged control samples varied from 28,000 to 55,000 psi
 

58

Table 19. TenSile:Properties of Several Zirconium-
Modified Alloys after Irradiation®

 

 

 

 

 

 

Stress, psi - s | Reduction
gfiifiir Sfi:;'::f-n Yielg  Uivimate Unflggiatm;étg in Area
_ Tensile | (%)
x 103 x 10
Test Temperature 650°C |
108 2086 31.5 62.9 18.9 19.2 17.5
109 2064 34.7 88.8 37.8 38.9 24.0
110 2113 33.1 79.3 . 38.6 40.5 34.9
111 2139 33.6 80.8 33.3 37.4 33.8
112 2156 33.8 80 29.9 31.5 21.8
Test Temperature 760°C
108 2088 - 27.4 52.3 10.0 12.3 9.8
109 2058 29.8 58.3 10.2 25.6 28.3
110 2111 27.7 - 56.7 11.6 25.6 27.3
111 2128 32.2 56.2 - 12.4 30.7 34.2
112 2153 29 - 61.8 15.5 18.0 17.5

 

of 9

®Annealed 1 hr at 1177°C, irradiated at 50°C to a thermal fluence
X 10'? neutrons/cm® and tested at a strain rate of 0.002 min~?!.

ORNL-DWG 69-14377R

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

45

40 J"K

/ g \\;
TN \

_30 / = AN
e\:
= A /
225 / /
.—
w
ul 0
o
E 20 o /
N
& _ N\

5

0650° C
4 760° C
STRAN RATE = 0.002 min~"
10 14
5
o
o 0.2 04 06 08 1.0 1.2 14

ZIRCONIUM CONTENT {wt %)

Fig. 43. Fracture Strains of Several Zirconium~Bearing Alloys after

Irradiation at 43°C to a Thermal Fluence of 9 X 10!° neutrons/cm?.
59

and varied systematically with zirconium content (Fig. 39). Irradiation
at 650°C did not change these values markedly (Table 18). The yield
strengths of the alloys irradiated at 43°C are all about 33,000 psi and
do not vary detectably with zirconium content {Table 19). Thus, the
thermal treatment seems-ne¢eésary to bring about the changes character-
istic of.alloys with various concentrations of zirconium.

Some of the samples irradiated at 4350 were tested at 760°C. The.
fracture strain varied markedly with zirconium content (Fig. 43). How-
ever, the peak was shifted”to a higher zirconium level than was noted at
a test temperature of 650°C.

Alloys 100 and 108 through 112 were also creep-rupture tested. The
results of several tests on material annealed for 1 hr at 1177°C are
shown in Table 20 and in Figs. 44 through 46. The stress-rupture proper-
ties, shown in Fig. 44, illustrate the potent effect of zirconium in
increasing the rupture life at a given level of stress. The rupture life
of alloy 109 (0.28% Zr) seems anomalously high, but the other alloys
followed the trend of inoreasifig rupture life with increasing zirconium
content. The creep-rupture strength, shown in Fig. 45, was also
affected by the zirconium concentration. Again, alloy 109 was the only
exception to the trend of increa51ng strength with increasing zirconium
content. erconlum.lmproved_the fracture strain (Fig. 46). The addition
of only 0.1% Zr (alloy 108) improved the fracture strain, compared to
that of alloy 100, several hundred percent at low strain rates. Higher

’concentratlons of zirconlum improved the fracture strailn at high strain

'rates, but brought about little added 1mprovement at low strain rates.

- Some samples were aged for lOOO hr at 650°C to correspond to the

thermal treatment of irradiated samples The results of creep-rupture
. tests on these samples are glven in Table 21. A comparlson'W1th the
| proPerties of the unaged s&mples in Table 20 shows that this aging treat-
~ ment did not affect the pr0pert1es ‘appreciably. |

Alloys 100 and 108 through 112 were irradiated to a thermal fluence

of 2.5 X 1020 neutrons/cm ,at:650°c_and-sub3ected to creep-rupture tests

- at 650°C after irradiation5 'The results of these tests are given in

Table 22. These Samples were stressed at 32,400 psi for about 675 hr
 

 

60

Table 20. Creep Properties at 650°C of Several Alloys That Contain
Various Concentrations of Zirconium?

 

 

: Rupture Minioum Reduction
Alloy Specimen Test  Stress Iife Creep Elongation in Area
Number Number Number (psi) (1r) Rate (%)
(%/nr)
X 103
100 1915 5725 55 3.4 0.43 22.5 20.8
100 1917 6099 47 20.8 0.55 14.5 12.5
100 1909 5554 40 47 .4 0.088 12.5 9.5
100 1914 6108 32.4 135.1 0.0077 6.3 6.2
100 1921 6172 27 827.8 0.0022 4.9 4.0
108 2077 5689 55 4.1 0.16 17.5 14.8
108 2082 5991 47 178.7 0.025 18.8 15.9
108 2091 5721 40 348.5 0.018 25.0 21.4
108 2094 6162 32.4 2137.3 0.0059 31.3 28.4
109 2068 57178 70 30.7 0.128 32.4 28.3
109 2066 5712 55 178.1 0.015 32.0 31.0
109 2057 5740 51 358.5 0.013 31.0 36.0
109 2072 5987 47 851.2 0.0075 34.1 34.8
109 2065 5909 40 1474.3 0.0042 46.9 35.3
110. 2115 5746 70 12.5 0.138 39.3 26.9
110 2104 5673 55 200.7 0.028 35.9 31.1
110 2117 6096 47 283.0 0.030 38.2 38.6
110 2112 5665 40 709.0 0.0080 18.8 25.6
111 2130 5771 70 26.9 0.174 35.0 26.9
111 2135 5688 55 192.8 0.027 30.9 29.7
111 2126 6091 47 589.4 0.014 36.6 33.7
112 2159 5752 70 39.2 0.181 38.2 "31.6
112 2151 5908 62 169.5 0.038 34.3 27.2
112 2145 5666 55 537.4 0.011 30.5 26.9
112 2163 5641 51 673.1 0.010 37.5 28.8
112 2158 6166 47 1254.8 0.0056 32.0 29.9

 

®pnnealed 1 hr at 1177°C before testing.
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ALLOY Zr 100
NO. CONTENT (%)
20 ° 100 <0.03
e {08 0.10
2 {109 0.28
A {10 0.53
10— 1o t 075
e 42 1.48
o | | ' | | Hl
100 _ 10! 102 103

RUPTURE TIME (hr)

Fig. 44. Stress-Rupture Properties at 650°C of Several Alloys That
Contain Various Concentrations of Zirconium. All material was annealed
1 hr at 1177°C before testing. '

ORNL-DWG €9-14375

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* MINIMUM CREEP RATE (%/hr)

: Fig. 45. Influénce of Zirconium on Creep Rates at 650°C. All
- alloys annealed 1 hr at 1177°C before testing.
 

 

Fig. 46.

62

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10~

10°

Fracture Strains During Creep Tests at 650°C for Alloys

That Contain Various Amounts of Zirconium. All samples annealed 1 hr at
1177°C before testing.

Table 21. Stress Properties at 650°C of Several Ageda

Alloys That Contain Zirconium

 

fiupture

Minimum

 

Alloy Specimen Test Stress Lif Creep Elongation Reduction
. e in Area
Number Number Number (psi) (hr) Rate (%) (%)
(%/nr)
x 103
100 1922 5’723b 55 5.9 0.61 23.0 19.2
100 1919 5630 17.8 1007 0.0007 1.0 0
108 2090 5680b 55 280.9 0.042 40.6 38.8
108 2085 5911 32.4 1000 0.0096 15.0 9.8
109 2071 5'727b 55 152.7 0.068 38.4 44 7
109 2063 6092 32.4 1000 0.0044 3.5
110 2123 5682b 55 197.7 0.098 37.5 35.0
110 2120 6086 32.4 1002.2 0.0012
111 2144 5674b 55 305.5 0.052 29.7 28.3
111 2137 6144 32.4 1002.2 0.0011 3.1 0.03
112 2154 563‘7b 55 530.1 0.030 34.4 29.6
112 2149 5629 32.4 1000.6 0.0016 3.1 0.01

 

®annealed 1 hr at 1177°C and aged 1000 hr at 650°C before testing.

Discontinued before failure.

e
9 b/ . o ' .

Table;22. Creep PropertieSfa$'650°C of Irradiated® Alloys that Contain Various Amounts of Zirconium -

 

Stress,b psi

 

 

 

Alloy Specimen = Test Zég;zz;?n —— ;2,400 ___ %0, 000 :
Number lNumber . Number (wt %) - Creep filongation‘ Creep Rugture Elongation R?duction Frazgfire
o | . Rate (%) Rate ?;f? (%) 1n(%§ea ‘Strain
EERTE ~ (%/nr) | (%/br) (%)
100 1916  R-146 < 0.03  0.0099 1.1
108 2076  R-163  0.10  0.009% 8.2 0.147  53.1 7.3 19.0 21.1
109 2070  R-171  0.28  0.0034 2.3 0.025  301.3  10.8 17.2 18.9
110 2121 R-172 0.53 0.0013 = 0.97 = 0.009% 429.5 5.9 1.1 1.7
111 2140 B195  0.75  0.0037 3.3 0.035  127.1 5.7 13.9 15.0

112 2155  R-199 1.1 0.0033 . 5.1 0.080  46.5 3.9 6.8 7.0

 

%11 samples annealed l:hr at 1177°C before irradiation at 650°C to a thermal fluence of 2.5 x 1029
neutrons/cm?.

bAll tests‘excépt‘R-146‘were initially stressed at 32,400 psi for about €75 hr and then stressed at
40,000 psi. Test R-146 failed while loaded at 32,400 psi after 105.4 hr.

£9
 

 

64

and then stressed at 40,000 psi until failure. (This increase in stress
level was necessary to cause failure in a reasonsble time.) The rupture
life at the higher stress is shown in Fig. 47 as a function of zirconium
content. The strain during the first 675 hr at 32,400 psi is given in

parentheses, and the strain at the higher stress level is shown by each

ORNL-DWG 67-3529R2

 

 

 

 

 

 

 

600 TEST CONDITIONS | l
TEMPERATURE = 650°C
675 br AT o = 32,400 psi
500 INDICATED TIME AT 40,000 psi
IRRADIATION CONDITIONS
5.9(0.97)f TEMPERATURE = 650°C
"\ | THERMAL FLUENCE=
400 7 \ 2.5x102° fleuirons/crn2 )
300 —410.8 (2.3) \

RUPTURE LIFE (hr)

200 / | \
1/ N

N

100
/O/?.l'a {8.2) \

— ]
l . 3.9 (5.1}

 

 

 

 

 

 

 

 

 

 

o 0.2 0.4 0.6 0.8 1.0 1.2
ZIRCONIUM CONTENT (wt%)

Fig. 47. Influence of Zirconium Content on the Creep-Rupture
Properties at 650°C after Irradiation. All samples annealed at 1177°C
before irradiation.

roint. The rupture life increased rapidly with zirconium content up to
a maximum for an alloy containing 0.53% Zr. The alloys containing less
zirconium had better fracture strains under these test conditions; how-
ever, all alloys containing 0.1% Zr or more had excellent fracture strains
after irradiation. The minimum creep rates for these same samples are
given in Fig. 48. The creep rates at both stress levels exhibited a
definite minimum for the alloy containing 0.53% Zr. The creep rates for
the unirradiated control samples are also shown in Fig. 48 at the
32,400 psi stress level. The minimum creep rate was not affected detect-
ably by irradiation except for the alloys contéining 0.75 and 1.18% Zr.

A secofid irradiation experiment exposed samples to a thermal-neutron

fluence of 9 X 101° neutrons/cm?® at 50°C. These results are summarized

e
o

65

ORNL—DWG 69-14373R

IRRADIATED

40,000 psi -

MINIMUM CREEP RATE (%/hr)

_ IRRADIATED

32,400 psi

J
UNIRRADIATED, AGED

 

0 0.2 04 06 08 10 12 . 44
ZIRCONIUM CONTENT (wt %)

Fig. 48. Influence of Zirconium Content on the Creep Rates at 650°C.
All samples annealed 1 hr at 1177°C before irradiation at 650°C.

in Table 23. Because of the different stress levels involved, the
results are difficult to compare with those presented previously. How-

ever, the following points can be established by cross-plotting results:

Table 23. Creep Properties at 650°C of Irradiated® Alloys
That Contain Various Amounts of Zirconium

 

 

e s Minimum
Alloy Specimen Test Zéifi:fi:fn Stress Rgif::re Creep Elongation
Number Number  Number (vt %) (psi) (1r) Rate %
| T (%/nr)
| x 10?

100 1911 R-257 < 0.03 40 68.9 0.019 - 1.79
108 2080  R-283 0.10 40 314.9 0.014 8.03

108 o 55 5.1 0.9 5.05

109 - 2059 R-285  0.28 47b 158.6 0.010 2.42

109 . _ ... oo 557 43,0 0.116 5.26
- 110 2110 R-292 0.53 47 453.3 0.0063 7.52

111 2136 R-293  0.75 47 714.6  0.0076 15.5

112 2161 R-307 -1.18 47 905.7 0.0028 - 5.32

112 | o 55b 74.5  0.059 5.24

 

®annealed 1 hr at 1177°C before irradiation at 50°C to a thermal
fluence of 9 X 10!'° neutrons/cm?.

Proad increased during test.
 

 

66

(1) the rupture life increased and the creep rate decreased continuously

with increasing zirconium (Table 23); (2) the rupture lives of these

samples were greater than those of samples irradiated at 650°C (Table 22);

(3) the minimum creep rates were lower than those observed before irra-
diation (Fig. 45); and (4) the fracture strains were all quite good.

Tests were run at 760°C on several samples; the results are shown
in Table 24. For samples irradiated and tested at 760°C, the minimum
creep rate decreased with increasing zirconium (Fig. 49). The fracture
strain seemed to go through a maximum in alloys that contained 0.1 to
0.28% Zr. Test samples of ailoy.108 (0.1% Zr) indicated that the irra-
diation temperature had very little effect when the test temperature was
760°C. A test sample of this same composition also indicated that
annealing at 650°C after irradiation at 43°C had very little effect on
the properties. A single sample of alloy 112 (1.18% Zr) was irradiated
at 760°C and tested at 650°C. The rupture life at 30,000 psi was only
444.9 hr compared with more than 675 hr for a sample stressed at
32,400 psi after irradiation at 650°C (Table 22). However, the fracture
strain was quite good for both samples. |

Two other alloys were made that contained a nominal 1% Zr. These
two alloys, designated 146 and 147, also contained 0.1 and 1% Re, respec-
tively, an addition that was made in an attempt to improve the weldabil-
ity. These alloys were creep tested in the irradiated and unirradiated
conditions; the results are given in Tables 25 and 26, and the stress-
rupture properties are shown graphically in Fig. 50. The stress-rupture
properties of the unirradiated samples are about the same for both heats
and quite similar to those shown in Fig. 44 for alloy 108 (1.18% Zr).
Thus, rhenium does not appear to be a very influential addition; the
properties are basically those of an alloy containing 1% Zr. The single
test of an unirradiated sample of each heat that had been aged for
1000 hr at 760°C indicates an instability in this alloy. This anneal

decreased the rupture life and increased the minimum creep rate. Irra-

- diation at 650°C reduced the rupture life and fracture strain when

samples were tested at 650°C (Fig. 50). Irradiation at a higher temper-
ature further reduced rupture life, but the fracture strain remained

reasonably high.

i»

iy
¥ . 3 v ' ’ .

Table 24. Creep Properties of Irradiated® Alloys That Contain Various Amounts of Zirconium

 

 

 

' Minimum
Alloy Specimen Test Zérconium Temperature, °C Neutron Stresg Hupture Creep Elongation
ontent —= Fluence Life
Number — Number . Number (wt %) Trradistion Test (neut Ten?) (psi) (br) Rate (%
: _ | neutrons/cm (%/nr)
- s | x 1020 x 103

100 © 3791 - R-455 < 0.03 657 760 2.5 12.5 1.0 0.10 0.21
108 2089 ‘37284.0‘ 0.10 43 760 0.9 15 468.1 0.037 36.1
108 2093  R-291° = 0.10 43 760 0.9 15 380.9 0.049 28.3
108 3861  R-448 0.10 750 760 2.5 12.5 665.2 0.026 35.3
109 3871 R-456%  0.28 750 760 2.5 12.5 1009.6 0.009%  14.4
. 15 310.4  0.062 23.8

o | 17. 7.9  0.37 3.9

110 13881 R-447°  0.53 760 760 2.5 12.5 1007.4 0.008%  11.1
L 15 262.4 0.014 3.8

| 7.5  167.6 0.038 16.6

111 3891 R-459°  0.75 740 760 2.5 12.5 1296.3 0.0047  10.4
17.5  308.6 0.043 19.2

112 3901 R-457°  1.18 740 760 2.5 12.5 1008.8 0.0013 2.5

| | | 15 310.4 0.013 4.2

| | 17.5 218.4 0.041 12.4
112 3910 . R-8l4 1.18 760 650 2.6 30 4449 0.017 10.1

 

®Annealed. 1 hr at 1172°C before irradiation.
Ppnnealed 1000 hr at 650°C after irradiation.

CStress increased during test as indicated.

L9
 

€8

ORNL-DWG 69 -14372R

MINIMUM CREEP RATE (%/hr)

 

 

o 0.2 0.4 0.6 0.8 1.0 1.2
ZIRCONIUM CONTENT (wt%)

Fig. 49. Variation of the Minimum Creep Rate at 760°C and 12,500 psi

with Zirconium Concentration. Samples annealed 1 hr at 1177°C and irra-
diated at 760°C to a thermal fluence of 2.5 X 10?° neutrons/cm®.

ORNL-DWG 69-14374

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

3 T
(x107) \:\\i
\\
N
60
6504\ 4650 \QQ
50 \\ \\\\
. < A ® [ ) Al I
721 | | |696 """\\|\ 6507 ‘%
3 N
- ~~
240 7604 N 20606 \\\ +A-o->—2081_'r% 30%
= N
o e N | 1111
w0 f2e ‘I\\ N\ 270 5% STRAIN
30 760, |60 | | |||
T4 T0 8% STRAIN
20
o 146
A447
OPEN SYMBOLS-UNIRRADIATED
10 CLOSED SYMBOLS-1RRADIATED
0
100 10! 102 103 104

RUPTURE LIFE (hr)

Fig. 50. Stress-Rupture Properties at 650°C of Alloys 146 and 147.
All samples annealed 1 hr at 1177°C before testing. Numbers by each
point designate the irradiation temperature. |

in
Table 25. Creep Properties of Irradiated and Unirradiated Alloy 146

 

 

 

a, T : - Thermal Minimum .
Specimen Test 22?§i2 Temperature, °C Neutron Stress ‘Rggg:re Creep Elongation R?i“fi;:;?
Number = Number _ ' .. .. = Irradiation Test Fluence (psi) (hr) Rate (%) + (%)
i - ‘ - (neutrons/cm?) ‘ (%/nr)
o . X 1020 x 103

3515 6647 - 121 650 70 36.1 0.113 29.6 24.7

3516 @ 6648 121 650 63 90.1 0.037 24.9 22.2

3517 6649 121 | 650 55 311.3  0.012 20.1 20.6
3504 6136 121 650 47 865.5 0.0049 22.3 17.2

3503 6135 121 650 40 1003.2° 0.0023 5.4 2.4

3511 7769 148 650 47 110.5 0.16 20.9 15.8
3518 . 6650 121 760 25 352.2 0,016 32.2 32.8

3519 6651 121 760 20 984.1 0.0039 29.1 51.9

3509 R-678 121 650 650 2.6 55 38.4 0.043 2.8

3510 R-656 121 650 650 2.6 47 177.4  0.021 4.9

3508 R-648 121 650 650 2.6 47 42.7 0.046 2.3

3522 . R=-915 121 696 650 2.6 47 7.9 0.20 1.8

3523 R-916 121 721 650 2.6 &7 2.7 1.0 3.0
3526 R-928 121 696 650 2.6 40 153.5 0.024 4.9

3512 R-801 121 760 650 2.6 40 59,0 0.071 4.8

3528  R-935 - 121 721 650 2.6 35 226.3 0.022 7.2

3513 R-830 . 121 760 650 2.6 30 306.3 0.016 5.4

3506 R-552 = 121 . 760 760 2.4 15 380.1 0.0211 10.6

 

aAnnea.ling designations: 121 = annealed 1 hr at 1177°C,

148 = annealed 1 hr at 1177°C plus 1000 hr at 760°C.

bTest“discontinued before failure.

69
 

 

 

 

Table 26. Creep Properties of Irradiated and Unirradiated Alloy 147
a Thermal Minimum
Specimen Test gg?g:i Temperature, °C Neutron ?trefis Rgi;:re Creep Elo?g?tion R:guiii:n
Number  Number Irradietion Test Fluence psi Rate
Irradiation (neutrons,/cn?) (hr) (%/nr) (%)
x 1020 x 103

4031 6652 121 650 70 51.8° 0.069 26.2 20.0

4032 6653 121 650 63 106.2 0.038 23.3 16.6

4033 6654 121 650 55 461.1 0,10 22.1 17.8

4045 6116 121 650 \ 47 1289.5b 0.0054 25.3 21l.5

4044 6117 121 650 40 1147.8° 0.0014 4.2 2.6
4038 7770 148 650 47 53.8 0.19 12.7 14.5

4034 - 6655 121 760 25 453.3 0.018 55.4 37.7

4035 6656 121 760 20 1240.1  0.0039 29.2 34.1

4028 R-679 121 650 650 2.6 55 65.2 0.0089 1.4

4030  R-687 121 650 650 2.6 47 508.1 0.00e8 4.8

4027 R-647 121 650 650 2.6 47 164.2 0.0095 3.2

4037 R-917 121 727 650 2.6 47 5.0 0.29 3.5

4051 R-802 121 760 650 2.6 40 136.4 0.027 4.3

4041 R-934 121 727 650 2.6 35 274.1 0.017 5.8

4052 R-832 121 760 650 2.6 30 804.4 0.0078 8.2

4049 R-553 121 685 760 2.4 15 1082.1 0.011 15.6

 

a'J.f!mnea.liz"lg designations: 121

148

—
—1
—

annealed 1 hr at 1177°C,

annealed 1 hr at 1177°C plus 1000 hr at 760°C.
bTest discontinued before failure. '

0L

 
 

71 -

The two 100~1b commercial melts, heat 21555 (0.05% Zr) and 21554
(0.35% Zr) were evaluated in both tensile and creep tests. The tensile
results are shown in Tables 27 and 28. The tests on heat 21555 after
irradiation showed that the coarse-grained material had superior ductil-
ity and that the fine-grained material had higher yield strength. The
tensile results for heat 21554 (0.35% Zr) are summarized in Table 28,
and the fracture strain for the fine-grained material is shown in Fig. 51.
The behavior is similar to that observed for standard Hastelloy N, but
at 760°C the ductility of the zirconium-modified alloy was much lower.
The strong dependence of fracture strain on strain rate at 760°C was
also quite unusual where the fracture strain was higher at the lower
strain rate. After irradiation, the fracture strain of'the fine-grained
material was generally reduced, and the reduction was larger at higher
test temperatures (Fig. 52). The ratios of the various properties in
the irradiated and unirradiated alloys are shown in Fig. 53. The yield
and ultimate tensile strengths were not affected appreciably by irradia-
tion. The elongation remained high with increasing temperature and
dropped at test temperatures above 600°C. The high ratio at 760°C is
due to the very low fracture strain of the unirradiated sample. The
reduction in area follows the normal pattern for standard Hastelloy N.

Further analysis of the results in Table 28 shows the effects of

various annealing treatments. Several samples were annealed for 100 hr

-at 871°C followed by 1000 hr at 650°C. The strengths were lower and the

fracture strains higher than those of the sampleé held at 650°C for a
much longer time. TWd-sfimfiléé fiere'annealed for 1 hr at 1177°C and then
for 1000 hr at '760°C. - Thié'tréétment gave good strength and.fiuctility
characteristics. Sdme‘sampieéiwere irradiated that had heat treatments
higher for samples annealedifor 1 hr at 1177°C than for those annealed
at 871°C. e - - |
~ Irradiated and-unirfadiatedfsamples.of heats 21555 (0.05% Zr) and
21554 (0.35% Zr) WEre:téétéd ifi!¢féep. The creep results at 650°C for
heat 21555 are given in Tables 29 and 30, and the1stress-rupture proper-
ties are shown in Fig.;54.i ih'the‘unirradiated'conditions, the optimum

rupture life was obtained after a l-hr anneal at 1177°C. Further aging
 

 

' Table 27. Tensile Properties of Irradiated and Unirradiated Alloy 21555 (0.05% Zr)

 

 

 

 

 

00 hr at 871°C.

Anneal® e Strai St i Reducti
Specimen Before Temperature, °C Neutron rain ress, bS Elongation eductlon
Number Irra- Irradistion Test Fluence Ratfl Vielg Uitimate Goreorm Total 0 Area
diation | (neutrons/em?) (min~1) Tensile - (%)
x 1020 x 10° x 103
2488 36 600 0.002 24 .4 4.1 50.2 52.5 37.6
2493 36 650 0.05 24 .4 74.2 55.0 59.0 40.1
2494 - 36 650 0.002 | 25.2 66.8 33.0 42.6 35.2
2496 36 700 0.002 22.8 54.6 20.2 40.8 41.3
2478 148 25 0.05 43.9 1l16.1 54.3 56.9 b 2
- 2479 148 | 650 0.002 27.3 66.9 30.7 33.7 31.5
2480 121 540 650 2.5 0.05 26.6 63.6 43.3 45.0 33.8
2481 121 | 540 650 2.5 0.002 27.1 60 27.5 28.3 20.5
2464 16 540 650 2.5 0.05 36.8 60.2 11.3 11.9 12.6
2465 16 540 650 2.5 0.002 37.7 48.2 5.8 6.0 7.9
a'Anneal:i.ng designations: 36 = 100 hr at 871°C plus 1000 hr at 650°C,
148 = 1 hr at 1177°C plus 1000 hr at 760°C,
121 = 1 hr at 1177°C,
16 = 100

cL

 
73

Table 28. Tensile Properties of Irradiated and Unirradiated
Alloy 21554 (0.35% Zr)

 

 

 

Annea.la Thermal- Stress, psi
Specimen Before Temperature, °C Neutron S;r:in Ultimate M RgducA:ion
Number Irra- Irradlation Test Fluence ( minfl) Yield Tensile Uniform fTotal n(%)ea
diation (neutrons/cm?)
x 102%° x 10?7 x 103
5543 104 25 0.05 67.7 128.7 42.3 47.0 56.33
5544, 104 200 0.05 60 115.9 39.9 42.9 59.80
5545 104 400 0.05 54.7 109.2 40.7 b 4 46,82
5553 104 400 0.002 54.2 111.8 45.8 48.3 53.59
5546 104 450 0.05 53 107.2 41.3 44.3 54..56
5554 104 450 0.002 - 55.9 110.5 42.7 48.1 48.26
5547 104 500 0.05 51.4 106.9 42.5 44 .8 49.79
5555 104 500 0.002 58.6 111.4 39.9 45.3 43.69
5548 104 550 0.05 58.7 106.9 38.%7 43.2 42.73
5556 104 550 0.002 59.8 101.3 26.4 28.0 24.70
5549 104 600 0.05  53.4 102 37.0 30.3 31.44
5557 104 600 0.002 56.8 91 22.7 25.0 24.18
5550 104 650 0.05 46.4 87.5 27.6 30.4 31.88
5558 104 650 - 0,002 55.8 75.5 13.3 46.6 63.15
5551 104 760 0.05 43,7 59 4,0 8.1 64.27
5559 104 760 0.002 37.9 37.9 1.6 47 .4 69.45
5552 104 850 0.05 36.9 37.7 7.5 56.5 69.70
5560 104 850 0.002 22 22 100.0C 46.9 60.94
10094 121 25 -0.05 37.5 102.2 72.6 76.3 62.6
10095 121 200 0.05 33.2 100.1 64,0 65.3 46.3
10096 121 400 0.05 29.8 96.8 71.8 77.8 6l.8
10097 121 500 0.05 25.6 84 .4 79.9 83.7 50.7
10098 121 550 0.05 26.9 8.8 75.8 78.3 54.5
10099 121 600 0.05 27.4% 89.1 74.3 78.0 49.6
10103 121 600 0.002 26.2 81 55.7 57.1 47.2
10100 121 650 0.05 25.7 g81.2 65.8 68.5 44,1
10104 121 : 650 0.002 25.9 72.2 45.5 46.5 32.6
10101 121 700 0.05 25.8 72.5 47.0 48.7 38.9
10102 121 760 0.05 24.7 66 38.0 40.2 23.6
10105 121 760 0.002 24 45.5 11.5 33.5 29.9
2431 36 600 0.002 39.8 81 29.2 46.8 48.5
2436 36 650 C.05 26.5 71.3 51,6 60.6 42.5
2534 36 650 0.05 26 78.3 50.8 54 .4 35.1
2437 36 650 0.002 27.2 69.1 43,5 49.6 31.4
2438 36 650 0.002 25.5 65.5 41.0 45.7 39.2
2439 36 700 0.002 25.2 56.7 18.7 34.4 31.2
10092 148 25 0.05 47.3 114.5 53.5 55.1 41.7
10093 148 650 ¢.002 30.1 68.9 37.3 47.3 42.9
2110 16 650 25 4.1 0.05 64.2 122.1 44,1 £7.5 39.08
9111 16 650 200 4.1 0.05 61 116.1 43,7 46.0 51.90
9112 16 650 400 - 0.05 52.7 109.1 42.5 46.2 34,96
- 9131 16 650 400 o Al 0.002 64.6 - 112.6 39.5 45.2 36.26
9114 16 €50 450 . - 4. - 0,05 56 1no.2 39.8. 42.3 33.24
9132 16 650 - 450 4.1 0.002 57.4 107.9 41.8 45,7 33.94
9115 16 650 500 4.1 0.05 56.9 109.9 40.4 43.4 33.48
- 9113 16 T 650 500 ~dyal 0.002 50.5 119.7 43.4 45.6 25.54
9l11é6 16 . 650 .. - 550 4. 0.05 60.2 10'7.77 4.5 47.5 36.52
9121 16 ~650 - 550 4.1 0,002 43.6 91.7 22.5. 23.0 21.15
9117 16 650 600 b1 0.05 52.7 94.1 31.0 32.0 28.09
9122 16 650 600 4.1 0.002 50.3  80.3 15,7  16.8  18.42
9118 16 ) 650 650 - 4.1 0.05 ‘ - ' 25.54
9123 16 650 - 650 4.1 0.002  48.1 65.6 10.9. 11.2 12.39
9119 16 650 - - 76D 4.1 0.05 42.4 55.4 6.6 © 6.8 10.89
9124 16 650 760 4il 0.002 37.6  37.9 17 - 5.4 - 4.7
9120 16 650 - - 850 4,1 0.05 3.8 . '38.6 3.1 6.4 - 6.27
9125 16 - 650 850 R - 0.002 3.8 12,1 5.1 10.4 - 13.88
2405 Lo 121 L 570 650 2.5 0.05 28.6 63.1 43.0 -  48.8 31.9 .
2406 121 575 - 650 - 2,5 - 0.002 30.5 56.6 28.4 32.6 34.0
2395 16 550 650 2.5 0.05 36.3 60.2 12.0 - 12.4 14.3
2396 16 ‘550 650 2.5 0.002 38.7 50.9 6.3 6.5 8.8
a’Anneav.l:l.ug designations: 16 = 100 hr at 871°C,
' 36 = 100 hr at 871°C plus 1000 hr at 650°C,
148 = 1 hr at 1177°C plus 1000 hr at 760°C,
121 = 1 hr at 1177°C,
104 = 100 hr at 871°C plus 5600 hr at 650°C.
 

 

T4

ORNL-DWG 68-7761

 

60

 

 

 

 

 

 

 

 

 

L
50 = !
_ bl ~ 1 T
g 1 - \__\ i
40 a
3 {
2 i
-
2 | LINL
G STRAIN RATE \ i
d (min_') \j \\ "
:": 20 » 005 A {
e o 0.002 _ \\ ]’
10 \j

 

 

 

 

 

 

 

 

 

 

 

 

0
O 100 200 300 400 500 600 700 800 900

TEST TEMPERATURE (°C})

Fig. 51. Variation of Ductility with Temperature for Control Speci-
mens of Zirconium-Modified Hastelloy N (Heat 21554). Annealed 100 hr
at 871°C and aged 5600 hr at 650°C in barren salt.

ORNL-DWG 68-7762

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

60
50
S ®
& T e \
N *
- 40 \
& \ \
-
<
o \ A
g 30 \
o \
ul °\
a 20 5\
K STRAIN RATE \
- (min"') N\ \:
10 * 005 ‘\\7_ 7
o 0.002 N>
N

 

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

Fig. 52. Variation of Ductility with Temperature for Molten Salt
Reactor Experiment Surveillance Specimens of Zirconium-Modified
Hastelloy N (Heat 21554). Thermal fluence was 4.1 X 1020 neutrons/em?.
Samples annealed 100 hr at 871°C before irradiastion over a period of
5600 hr at 650°C.

N

- ,
Table 29. Creep Properties of Unirradiated Alloy 21555 (0.05% Zr)

 

 

. Anneal® Test Rupture Minimumn . . Reduction
Sgfi;;fiin' | N$;;Zr - Before Temperature ?;::?S Life E:E:P Elo?%?tlon in Area
mber et (°¢) (br) (%/0r) (%)
o | , | x 103
2500 . .6312 = . 121 650 70 3.5 1.49 37.3 36.1
2506 6313 - 121 - 650 62 17.0 0.39 32.5 35.3
2506 . 6314 - 121 650 55 - 62.5 0.12 28.1 26.1
2507 6315 . 121 . . 650 47 231.3 0.026 39.7 33.2
2501 6316 - 121 650 40 451.5 0.0085 34.7 32.0
2509 . 6317 121 650 35 1115.6 0.0065 31.8 32.8
2498 5863 - 37 650 . 40 511.2 0.0153 . 41.5 37.7
2497 6163 37 ' 650 32.4 1673.8 0.0050 34.1 23.1
2461 7669 148 650 - 62 3.4 3.59 31.4 38.3
2459 7492 148 650 47 77.1 0.24 33.8 37.8
2460 - 7668 148 650 40 272.9 0.062 29.8 37.5
2492 5859 36 | 650 47 77.9 0.34 60.0 56.3
2490 6164 36 650 32.4 1028.9 0.028 65.9 51.0
2505 17282 121 760 30 4.4 0.64 38.4 39.1
2499 6357 121 760 20 183.8 0.087 34.6 27.1
0.023 26.8 26.9

2512 . 7281 121 760 15 527.3

 

 aAnnealing designations: 121 = annealed 1 hr at 1177°C,

37 = annealed 1 hr at 1177°C plus 1000 hr at 650°C,
148 = annealed 1 hr at 1177°C plus 1000 hr at 760°C,
36 = annealed 100 hr at 871°C plus 1000 hr at 650°C.

Gl
 

 

76

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b -

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

 

o
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Fig. 53. Comparison of the Tensile Properties of Irradiated and
Unirradiated Zirconium-Modified Hastelloy N (Heat 21554). Thermal fluence
was 4.1 X 102° neutrons/cm®. Annealed 100 hr at 871°C and aged or irra-
diated over a period of 5600 hr at 650°C. Tested at a strain rate of
0.05 min~1. -

Table 30. Creep Properties of Irradiated” Alloy 21555 (0.05% Zr)

 

 

 

 

Annea.lb Rupture Minimum
Specimen Test Before Stress }E fe Creep Elongation
Number Number Irra- (psi) (hr) Rate (%)
diation (%/hr)
x 103 :
Test Temperature 650°C
2483 R-381 121 47 106.4 0.032 6.7
2482 R-259 121 40 410.9 0.0089 10.0
2467 R-349 16 47 O
2466 R-268 16 40 20.0 0.11 2.6
2469 R-350 16 35 45.7 0.032 3.0
Test Temperature 760°C , |
2485 R-227 121 15 110.1- 0.036 7.2
2484 R—355_ 121 12.5 349.5 0.0095 7.7
2472 R=-237 16 15 34.2 0.0096 5.0
2471 R=-351 16 12 123.1 0.027 6.

 

& Irradiated st 540°C to a thermal neutron fluence of 2.5 X 1020
neutrons/cm<.

annealed 1 hr at 1177°C,

bAnnea.l:i_ng designations: 121 _
- annealed 100 hr at 871°C.

16

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o ANNEALED 1hr AT H77°C, 1000 hr AT 760°C
o ANNEALED 100 hr AT 871°C,1000 hr AT 650°C
0 OPEN SYMBOLS — UNIRRADIATED
CLOSED SYMBOLS — IRRADIATED
L L 1T
10° 10! 102 103 10*

RUPTURE TIME (hr)

Fig. 54. Stress-Rupture Properties at 650°C of Alloy 21555 (0.05% Zr).

for 1000 hr at 650°C produced no detectable effect, but aging for 1000 hr
at 760°C decreased the time to failure. Annealing at 871°C to produce a
fine grain size resulted in shorter rupture lives. Irradiated samples
had been annealed at 11771andfl871?0 before irradiation. As shown in

Fig. 54, irradiation reduced the rupture life of the larger grained mate-

rial by a factor of about 2 and that of the finer grained material by a
factor of about 10. '

The minimum creep rate of heat 21555, shown in Fig. 55, was not

-affecfed appreciably by'irradiation. ‘The lowest minimum;cfeeprraté was
obtained by annealing at 1177°C; aging at 650°C had no effect; aging at
760°C increased the. creep rate, and anneallng at 87l°C to glve a small

grain size produced a high creep rate.

The results of creep tests on heat 21554 (O. 35% Zr) are given in
Tables 31 and 32. The stress-rupture properties at 650°C are shown in

F;g. 56. The same'general t;ends with regard to heat treatment were

'observed for this heat and for the heat with lower zirconium content

(heat 21555). The optimum rupture life occurred after a l-hr anneal at
 

 

 

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o ANNEALED t hr AT H77°C,1000 hr AT 760°C
¢ ANNEALED 100 hr AT 871°C,1000 hr AT 650°C
10 OPEN SYMBOLS - UNIRRADIATED
CLOSED SYMBOLS - IRRADIATED

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100

1073 02 107!
MINIMUM CREEP RATE (% /hr)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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Fig. 55. Creep Rates at 650°C of Alloy 21555 (0.05% Zr).

e

I
"9

Table 31. Creep Properties of Unirradiated Alloy 21554 (0.35% zr)

 

Anneal® Rupture " Reduction

 

 

 

Specimen Test Stress . Creep Elongation -
Number  Number B;foze (psi) ?;i? Rate (%) 1n(%§ea
- = (%/tr) -~

'x 103 |
: Test Temperature 650°C .
2442 6306 121 70 - 7.3 1.38 49.4 43.8
2443 6307 - 121 62 - 32.0 0.20 38.1 43,1
2444 - 6308 121 55 9%4.4 0.062 31.6 26.7
2445 - 6309 121 - 47 395.2 0.018 26.8 27.7
2446 6310 121 40 803.8 0.0071 12.7 28.0
2441 5913 37 40 627.7 1.38 20.5 12.8
10086 7667 148 62 5.5 2.45 54.9 41.8
10084 - 7491 148 47 157.5 0.14 46.0 44,0
10085 . 7666 148 40 597.4 0.0313 36.5 33.4
2435 5862 36 47 128.3 0.20 63.0 50.6
2491 5858 36 40 226.3 0.13 - 60.8 50.9
2434 5857 36 40 324.2 0.056 35.5 51.9
2433 6168 36 32.4 1390 9 0.018 55.1 424
5535 6399 104 70 3.3 8.38 60.6 48.8
5536 6398 104 63 8.3 3.20 50.0 50.1
5537 .~ 6397 104 55 28.2 1.01 55.8 45.0
5538 6389 104 - 47 58.1 0.37 38.3 58.3
5539 6390 104 40 228.7 0.12 69.5 56.6
5540 - 6425 104 47 84.1 0.36. 75.1 .56.0
5541 = 6426 104 40 244.0 0.11 63.1 50.0
5542 | 6427 - 104 32.4 709.8 0.032 56.1 50.8
Test Temperature 760°C

2421 . 7286 121 30 35.6 0.33 30.3 25.9
2415 6356 121 20 354.1 0.047 32.9 : 30.7
2427 . 7285 121 15 1073.0  0.0123 28.7 31.3
2404 6056 0 16 - 20  85.0 0.35 0

61.7 45,

 

%Annealing designatione: 121 = annealed 1 hr at 1177°C,

- 37 = annealed 1 hr at 1177 C plus 1000 hr
EOTIRRR at 650°C,
148 = annealed 1hrat 1177 C. plus 1000 hr
. at 760°C,
. 36 = annealed. lOO hr at 871°C plus 1000 hr
S at es0°c,
104 = annealed 100 hr at 871°C plus 5600 hr
at 650°C,

16-=,annealed 100 hr at 871°C.
 

 

Table 32. Creep Properties of Irradiated Alloy 21554 (0.35% Zr)

 

 

 

 

8 Thermsal- Minimum
Specimen = Test gggg;i %Zi;gi:tigg Neutron Stress Rgggzre Creep Elongation
Number Number . - etion (°C) Fluence (psi) (br) Rate
(neutrons/cm?) . (%/hr)
x 1020 X 107
Test Temperature 650°C
2408 R-382 121 590 2.5 47 140.2 0.024 4.75
2424 R-479 121 650 2.5 47 111.3 0.034 5.05
2407 R=-260 121 590 2.5 40 336.3 0.0036 2.50
2420 R-480 121 650 2.5 40 279.0 0.0032 12.6
2398 R-265 16 550 2.5 47 3.9 0.35 1.51
2397 R-269 16 550 2.5 40 76.7 0.051 4.74
2402 R-356 16 550 2.5 35 163.7  0.0068 3.43
9127 R-316 61 650 4.1 47 11.1 0.34 4 .57
9126 R-311 61 650 4.1 40 19.3 0.14 3.99
9128 R-318 6l 650 4.1 32.4 65.4  0.013 2.36
9129 R-322 61 650 4.1 27 204.9 0.0088 3.45
- Test Temperature 760°C

2409 R-364 121 590 2.5 20 71.5 0.014 1.95
2410 R-239 121 590 2.5 15 264.1 0.009%4 6.52
2412 R-470 121 760 2.5 15 9.1 0.092 1.24
2403 R-174 16 760 2.3 20 5.5 0.36 2.95
2401 R-236 16 566 2.5 15 13.3 0.10 2.62
2400 R-357 16 . 560 2.5 12.5 222.8 0.0092 5.62

 

®Annealing designations: 121 = annealed 1 hr at 1177°C; 16 = annealed 100 hr at 871°C;
61 = annealed 200 hr at 871°C. -

 

08
81

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20 A ANNEALED 1 hr AT 1177°C, 1000 hr AT 650°C

0 ANNEALED { hr AT 1477°C, 1000 hr AT 760°C

¢ ANNEALED 100 hr AT 871°C, 1000 hr AT 650°C-
o ANNEALED 100 hr AT 871°C, 5600 hr AT 650°C
10 OPEN SYMBOLS -UNIRRADIATED

CLOSED SYMBOLS-IRRADIATED

UL Ul L LT

o
10° 10 102 10° 104
: RUPTURE TIME (hr) :

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig. 56. Stress-Rupture Properties at 650°C of Alloy 21554 (0.35% Zr).

1177°C. Irradiation reduced the rupture life of the material annealed
at 1177°C by a factor of,zlto 3 and that of the fine-grained material by
a factor of about 10. The variation of the minimum creep rate with heat
treatment and irradiation is shown in Fig. 57. The lowest minimum creep
rate was observed for the material annealed at 1177°C. No effect of
irradiation on the minimum cfeep rate was evident.

The fracture strains are shown as a function -of minimum creep rate

in Fig. 58 for material annealed at 1177°C and tested in the unirradiated

condition at 650°C. The results for heat 21554 (0.35% Zr) show a clear

trend of decreasing fracture strain with decreasing creep rate. The

- _results for heat 21555 (O 05% Zr) show a simila.r trend, but the data

seem to :f‘all on two separa.te l:.ne segments. All of the :E'ra.cture strains
are quite high and fall la.rgely in the range of 20 to 40%. The results
in Tables 29 and 31 show tha.t the sa.mples of both alloys that. were |

annealed at 871°C to give a fine gra:Ln size had fracture strains of 40

to 60%.
 

 

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20 & ANNEALED {hr AT 1177°C, 1000 hr AT €50°C

O ANNEALED fhr AT 4177°C, 4000 hr AT 760°C

¢ ANNEALED 100 hr AT 871°C, {000 hr AT 650°C
< ANNEALED 100 hr AT B71°C, 5600 hr AT 650°C

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

10 OPEN SYMBOLS —~UNIRRADIATED
CLOSED SYMBOLS~IRRADIATED
DT TR
103 102 1o-! 100 10t

MINIMUM CREEP RATE, %/hr

Fig. 57. Creep Properties at 650°C of Alloy 21554 (0.35% 2Zr).

ORNL-DWG 6€9-14366

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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MINIMUM CREEP RATE (%/hr)

Fig. 58. Fracture Strains at 650°C of Alloys 21554 (0.35% Zr) and
21555 (0.05% Zr). All samples amnealed 1 hr at 1177°C before testing.
83

The fracture strains of irradiated samples of heats 21554 and 21555
are shown in Fig. 59 as a function of strain rate. The results generally
follow a trend of increasing fracture strain with decreasing creep rate,

although one point definitely deviates from this trend.

ORNL-DWG 69- 14365

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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MINIMUM CREEP RATE (% /br)
Fig. 59. Fracture Strains at 650°C of Alloys 21554 (0.35% Zr) and

21555 (0.05% Zr) after Irradiation. All samples annealed 1 hr at 1177°C
before irradiation. . A

The results for a test temperature of 760°C are fragmentary

~ (Tables 29 through 32).,.The~$tress-rupture'cheracteristics are shown

in Figs. 60 and 61 for heats 21555 (0.05% Zr) and 21554 (0.05% Zr).
Rupture life was reduced by irradiation, but the samples annealed at

"1177 C before irradiation had 1onger rupture lives than those annealed

at 871°C. The fracture stralns were reduced by irradiation, and the
results for heat 21554 (Table 32) 1ndlcate that the reductlon was greater

-~ as the 1rrad1at10n temperature increased.

Several tested samples,uere examlned metallographically. Typical
photomicrographs of'several'alloys that contained zirconium are shown
in Fig. 62. These samples were annealed 1 hr at 1177°C, aged 1000 hr
 

 

 

 

 

84

ORNL-DWG 69-14364

STRESS (psi)

o ANNEALED thr AT $177°C

¢ ANNEALED 100 hr AT 871°C

OPEN SYMBOLS-UNIRRADIATED

CLOSED SYMBOLS-IRRADIATED AT 540°C

 

10
RUPTURE TIME (hr)

Fig. 60. Stress-Rupture Properties of
Alloy 21555 (0.05% Zr) at 760°C.

ORNL-DWG €9-14363
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20
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o ANNEALED 1he AT 1477°C
o . © ANNEALED 100 hr AT 874°C
OPEN SYMBOLS-UNIRRADIATED
CLOSED SYMBOLS IRRADIATED AT
INDICATED TEMPERATURES
4

100 10! : 102 : o3 2
RUPTURE TIME (hr)

Fig. 61. Stress-Rupture Properties of
Alloy 21554 (0.35% Zr) at 760°C.

e
 

 

Fig. 62. Photomicrographs of Zirconium-Modified Alloys Annealed for 1 hr at 1177°C, Aged for
1000 hr at 650°C, and Stressed at 32,400 psi for 1000 hr. 500x. (a) Alloy 108 (0.10% Zr), (b) alloy 109
(0.28% Zr), (e) alloy 110 (0.53% 2Zr), and (d) alloy 111 (0.75% Zr). Etchant: glyceria regia.
Reduced 26%. |

¢8
 

 

86

at 650°C, and stressed for 1000 hr at 650°C. Alloy 108 (0.10% Zr) had

a very fine carbide precipitate typical of that observed in alloy 100
(Fig. 24), which did not contain zirconium. In alloy 109 (0.28% Zr) the
precipitate was coarser and located predominantly along the grain bound-
aries. Alloys 110 (0.53% Zr) and 111 (0.75% Zr) contained a very fine
precipitate that could not be resolved at a magnificatién of 500%.
Alloy 112 (1.18% Zr) had an even higher concentration of precipitate
(Fig. 63). Figure 63(c) shows that the precipitate was acicular. The
grain boundaries had a denuded zone, and there was a denuded region asso-
ciated with the oxide at the surface. The precipitate was extracted and
found by x-ray diffraction to have a lattice parameter of 4.68 A, equal
to that for zirconium carbide. We also strained some samples of

alloy 112 at 25°C after they had been aged at 650°C to produce the
precipitate. The fracture strains were in excess of 50%, thus indicating
that this microstructure was not brittle.

Some of the aged samples were tensile tested at 650°C (Table 17).
The fracture of alloy 112 after testing at 650°C at a strain rate of
0.002 min~! is shown in Fig. 64. The fracture strain was 23.3%, and the
fracture was transgranular.

The fracture of alloy 110 (0.53% Zr), which was tested in creep at
650°C, is shown in Fig. 65. This alloy was at temperature 709 hr; and
small quantities of precipitate are present in the microstructure. This
precipitate cannot be resolved at low magnification and appears as dark
spots. There are numerous intergranular cracks, and the fracture is
intergranular. Typical photomicrographs of alloy 112 (1.18% Zr) after
creep testing at 650°C are shown in Fig. 66. This sample was at temper-
ature 1255 hr, and large quantities of precipitate are present in the
microstructure. There are numerous intergranular cracks; and the frac-
ture has both intergranular and transgranular components. _

Typical fractures of alloy 21555 (0.05% Zr) after testing at 650°C
and 32,400 psi are shown in Fig..67. When this alloy was annealed
100 hr at 650°C before testing, the grain size was quite small. This
sample was strained 65.9% before it fractured (Table 29), predominantly
transgranularly [Fig. 67(a)]. Annealing for 1 hr at 1177°C and testing

in creep resulted in the microstructure shown in Fig. 67(b). The
 

87

L I S R R SRR Y-T9307

 

    

 

 

(c)
| | Fig. 63. Photomicrographs of Zirconium-Modified Alloy 112 (1.18% 2r)
: | Amnealed for 1 hr at 1177°C, Aged for 1000 hr at 650°C, and Stressed at
32,400 psi for 1000 hr. (a) Typical, 500x; (b) edge of sample, 500x;
and (c) typical, 1000X. Etchant: glyceria regia. Reduced 23%.

 
 

 

 

 

 

Fig. 64. Photomic¢rographs of Alloy 112 (1.18% Zr) Annealed for 1 hr
at 1177°C, Aged for 1000 hr at 650°C, and Tensile Tested at 650°C at a
Strain Rate of 0.002 min“!. (a) Fracture, 100x; (b) typical, 500x.

.
 

  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig. 65. Photomicrographs of Alloy 110 (0.53% zr) Annealed 1 hr at
1177°C and Tested at 650°C and 40,000 psi. (a) Fracture, 100x; (b) typi-
cal, 500X. Etchant: glyceria regia. ' ' '

 
 

 

 

 

mflflg\
g2

A R

 

% Zr) Annealed 1 hr at 1177°C and Tested at 650

C and

glyceria regia.

o

Etchant

and (c¢) typical, 500X.

100X,

s

Photomicrographs of Alloy 112 (1.18
a) Fracture, 100x; (b) typical

Fig. 66.
47,000 psi. (
Reduced 20%.
S8 Y -89841

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig. 67. Photom:.crographs of Alloy 21555 (O. 05% 7r) Creep Tested at 32,400 psi and 650°C.
(a) PFracture of sample ammealed for 100 hr at 871°C and 1000 hr at 650°C before test; (b) fracture of
sample annealed 1 hr at 1177°C before test; and (c¢) fracture of sample annealed 1 hr at 1177°C and
1000 hr at 650°C. . 100x. Etcha.nt ly'cer:.a. regia. Reduced 21%.

 

T6
 

 

92

fracture was still largely intergranular, and the fracture strain was §Ei;
31.8%. Annealing at 1177°C and aging for 1000 hr at 650°C resulted in- .
,a'fracture strain of 34.1% (Table 29). The microstructure contained
only a few small precipitates, and the fracture was still largely trans- ®
granular. Alloy 21554 (0.35% Zr) was subjected to & similer annealing
sequence. After annealing for 100 hr at 871°C and 1000 hr at 650°C and
testing at 650°C the fracture shown in Fig. 68(a) was obtained. This
alloy originally had stringers present (Fig. 38), and numerous cracks
formed ardund these stringers during testing. The fracture was trans-
granular, and the fracture strain was 55.1% (Table 31). Annealihg for
1 hr at 1177°C produced a coarser grain size. The fracture shown in
fig. 68(b) was mixed intergranular and transgranular, and the fracture
strain was about 20%. The approximately 1000 hr at tgmperature during
the creep test was sufficient to produce some precipitation. Another
sample was annealed 1 hr at 1177°C and aged for 1000 hr beforerfiesting.
This sample failed with about 20% strain. The fracture was mixed trans-
granular and intergranular [Fig. 68(a)]. Copious quantities of very fine
precipitate formed. .
Alloy 108 (0.10% Zr) was also examined metallographically after
jrradiation and creep testing. As shown in Fig. 69, a very fine precipi-
tate was formed that was quite similar to that shown in Fig. 62 for an
unirradiated sample. Alloy 110 (0.53% Zr) had more precipitate after
jrradiation than the control sample (compare Figs. 70 and 62). As shown
in Fig. 70 there was a denuded zone along the boundaries and some banding -
of the precipitate, likely due to inhomogeneous distribution of zirconium
in the alloy. The fracture was largely intergranular. Alloy 112 |
(1.18% Zr) had profuse quantities of the precipitate; the microstructure,
however, was quite similar to that for the unirradiated sample (compare
Figs. 71 and 63). The sample could not be etched heavily enough to show
the grain boundaries without obliterating the precipitate. The fraéture
was mixed intergranular and transgranular. - ,- -
Typical photomicrographs of alloy 21554 (0.35% Zr) are shown in . -
Figs. 72 through 75. The sample shown in Fig. 72 was irradiated and
tested at 25°C and failed transgranularly after straining 47.5%;,.Ihé» o
comparable control sample is shown in Fig. 73. This sample'failed _- ‘ kfi;
 

 

 

Y439467

 

 

 

 

 

 

 

 

 

 

Fig. 68. Photomicrographs of Alloy 21554 (0.35% Zr) Creep Tested at 650°C. (a) Fracture of
sample annealed 100 hr at 871°C and 1000 hr at 650°C before test; (b) fracture of a sample annealed
1 hr at 1177°C before test; and (¢) fracture of a sample annealed 1 hr at 1177°C and 1000 hr at 650°C
before test. 100X. Etchant: glyceria regia. Reduced 21%.

 
 

 

 

 

 

9%

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

' R-35432

T st

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig. 69. Photomicrographs of Alloy 108 (0.10% Zr) Annealed 1 hr
at 1177°C, Irradiated at 650°C for 1000 hr to a Thermal-Neutron Fluence
of 2.5 X 10%° neutrons/em?, and Creep Tested at 650°C. Stressed at

32,400 psi for 675 hr (8.2% strain) and 40,000 psi for 53 hr (8.3% strain).
(a5 Fracture, blunted end caused in handling, 100X; (b) typical, 500x.
Etchant: glyceria regis.

i
C - ' C

R-35438

  
   

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(a) &3 s (©) T i

‘Fig. 70. Photomicrographs of Alloy 110 (0.53% Zr) Annealed 1 hr at 1177°C, Irradiated at 650°C
for 1000 hr to 'a Thermal-Neutron Fluence of 2.5 X 1020 neutrons/cm?, and Creep Tested at 650°C.
Stressed at 32,400 psi for 675 hr (0.97% Strain) and 40,000 psi for 430 hr (5.9% Strain). (a) Fracture,
100x; (b) .edge of gage length, 100X; and (c) typical, 1000x. Etchant: glyceria regia. Reduced 24%.
 

     

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig. 71. Photomicrographs of Alloy 112 (1.18% Zr) Annealed 1 hr at 1177°C, Irradiated at 650°C
for 1000 hr to a Thermal-Neutron Fluence of 2.5 X 1020 neutrons/ecm?, and Creep Tested at 650°C.
Stressed at 32,400 psi for 677 hr (5.8%'Strain) and 40,000 psi for 47 hr (3.9% Strain). (a) Fracture,
100x; (b) fracture, 500x; and (c) typical, 1000x. Etchant: glyceria regia. Reduced 24%.

.

 

 

96

 
 

0y R- 41484

 

 

 

 

 

L6

 

 

 

Fig 72. Photomlcrographs of a erconlumrMbdlfled Hastelloy N Surveillance Semple (Heat 21554 )
Tested at 25°C at a Strain Rate of 0.05 min~l. Exposed 1n the Molten Salt Reactor Experiment core for
5500 hr ‘at 650°C to a thermal-neutron fluence of 4.1 X 102° neutrons/cm?®. (a) Fracture, 100x; (b) edge
of sample about 1/2 in. from fracture, lOOX, and (e) edge of sample showing edge cracking, 500x.
Etchant: aqua regia. Reduced 24%.
 

98

 

 

Fig. 73. Photomicrograph of the Fracture of a Zirconium-Modified
Hastelloy N Sample (Heat 21554) Tested at 25°C at a Strain Rate of
0.05 min~1. Exposed to a static fluoride salt for 5500 hr at 650°C
before testing. Note the shear fracture and the absence of edge cracking.
; 100xX. Etchant: glyceria regia. '

 

 

 

 

"

ik
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

(b)

‘Fig} 74}‘ Photomicrographs df'a-ZirconiumEMbdified Hastelloy N Surveillance Sample (Heat 21554)
Tested at 650°C at a Strain Rate of 0.002 min‘l. Exposed in the Molten Salt Reactor Experiment core
for 5500 hr at 650°C to a thermal fluence of 4.1 X 1020 neutrons/cm®. (a) Fracture, 100x; (b) edge of

sample about 1/2 in. from fracture, 100X; and (c) edge of sample showing edge cracking. Oxide formed
during the tensile test. 500%. Etchant: aqua regia. Reduced 22%.

 

66
 

 

 

 

 

 

 

Fig. 75. Photomicrograph Showing a Portion of the Fracture of a
Zirconium-Modified Hastelloy N Sample (Heat 21554) Tested at 650°C at a
Strain Rate of 0.002 min~1. Exposed to static fluoride salt for 5500 hr
at 650°C before testing. 100X. Etchant: glyceria regia.

transgranularly after straining 47.0%. An irradiated sample tested at
650°C is shown in Fig. 74; it fractured intergranularly after 11.2%
strain. The comparable control sample, shown in Fig. 75, strained 46.6%
before failing. There are numerous intergranular separations, and the

fracture is largely intergranular.

Alloys Containing Hafnium

Two groups of alloys were used to study the effects of hafnium
content bn the mechaniéal.proPerties. The first group was composed of
five laboratory melts that contained from 0.06 to 0.43% Hf (designated
alloys 152 through 156), and the second group consisted of two 100-1b
commercial melts that contained 0.08 and 0.50% Hf (designated
alloys 67-503 and 67-504, respectively). The chemical compositions of
these alloys are given in Table 1, and typical photomicrographs are
shown in Figs. 76 through 80. Alloy 153 (Fig. 76) contained only 0.06% Hf
. and had the largest grain size of the group. Alloys-l54 (0.10% Hf) and

ik
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

10T

 

 

 

 

 

Fig. 76. Ph_otomicrographs of Alloys Modified with Hafnium Annealed 1 hi at 1177°C. (a) Alloy 153
(0.06% Hf); (b) alloy 154 (0.10% Hf) and (c) alloy 155 (0.13% Hf). 100x. Etchant: glyceria regis.
Reduced 18%. -
 

 

 

 

 

 

 

16% Hf.
glyceria

 

 

 

 

 

 

 

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Etchant

 

 

 

 

 

 

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102
ographs of Alloy 152 Modif
(a) 100x and (b) 500x.

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Fig. 78. Photomicrographs of Alloy 156 Modified with 0.43% Hf.
Annealed 1 hr at 1177°C. (a) 100x and (b) 500x. Etchant: glyceria
regia. o T - _ |
 

 

 

 

 

 

 

 

 

 

i3

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(®) £

104

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig. 79. Typical Photomicrographs of Heat 67-503 (0.08% HT)

Annealed 1 hr at 1177°C.

Etchant:

glyceria regia.

Longitudinal section. (a) 100X and (b) 500x.

£,

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9

b

105

Y-84279

 

Y-84280

()}

. Fig. 80. Typical_Phdtbmiéiographs of Heat 67-504 (0.50% Hf) Annealed

‘1 hr at 1177°C. Longitudinal section. (a) 100X and (b) 500x. Etchant:

glyceria regia.
 

 

106

155 (0.13% Hf) had much smaller grain sizes. Alloy 152 contained
0.16% Hf, and the grain size was refined by copious quantities of carbide
precipitates at grain boundaries. When the hafnium content was increased
to 0.43%, as in alloy 156, the pfecipitation became more general with
precipitation at grain boundaries and in the matrix. The two commercial
alloys had larger grain sizes than their counterparts in the series of
laboratory melts (Figs. 79 and 80). ’

The results of creep-rupture tests at 650°C on unirradiated and
irradiated samples of alloys 152 through 156 are given in Tables 33 and

34, respectively. The stress-rupture properties for the unirradiated

Table 33. Stress Properties at 650°C of Several
Hafnium-Modified Alloys@

 

 

Minimum ' .
Alloy Specimen Test Stress Rgggzre‘ Creep Elongation R?gugizzn
Number Number Number (psi) - Rate (%) 1
_ (B} (g/nr) (%)
x 103

152 2979 7506 35 1906.3 0.0015 15.0 20.5
152 2964 6537 70 3.% 1.23 23.4 22.3
152 2967 6538 55 21.1 0.146 4.1 15.6
152 2965 6539 47 67.4 0.051 12.2 13.3
152 2966 6540 40 234 .4 0.018 11.7 11.4
153 6554 7507 35 216.5 0.0050 9.3 11.3
153 6453 6658 70 0.3 17.69 41.8 29.3
153 2864 6535 55 6.6 0.194 20.4 17 .4
153 2863 6536 47 76.7 0.023 16.2 18.2
153 6545 6687 40 114.2 0.013 11.9 11.8
153 6544 6686 55 . 7.4 0.223 22.6 18.0
154 6559 6771 70 0.2 3.5 ' 32.0 26.5
154 6556 6689 40 86.0 0.023 7.8 5.5
154 6567 7509 55 5.0 0.188 19.7 24.8
154 6566 7508 47 41.0 0.0352 13.1 18.7
155 8495 7554 70 0 48 .4 37.3
155 2873 6531 55 33.6 0.112 14.0 17.9
155 2874 6532 47 90.2 0.043 13.2 14.0
156 6642 6692 70 1.9 1.25 22 .% 22.0
156 6643 6693 55 12.0 0.152 14.7 11.9
156 2878 6694 &7 257. 0.0214 18.5 15.6
156 6650 7510 35 1210.5 0.011 22.2 20.3

 

8211 materials annealed 1 hr at 1177°C before testing.

"
4

107

Table 34. Creep Properties at 650°C of Irradiated® Alloys
Containing Various Amounts of Hafnium

 

 

Minimum :
Alloy Specimen Test gafnium Stress Tpture Creep Elongation
- ontent . Life
Number Number  Number (wt %) (psi) (hr) Rate (%)
(%/hr)
x 103
100 1916 R-146 32.4 105.4 0.0099 1.1
152 2961 R-373 0.16 47 10.7 0.240 3.2
152 2963 R-208 0.16 40 87.3 0.044 4,28
152 2962 R-371 0.16 32.4 459.5 0.0068 5.14
153 2860 R-212 0.06 - 40 35.6 0.053 2.38
153 2861 R=-222 0.06 35 202.2 0.0065 3.07
153 2862 R-383 0.06 32.4 224..8 0.010 3.32
154 2865 R-374 0.10 47 7.1 0.253 2.25
154 2867 R-188 0.10 40 41.5 0.0548 2.70
154 2866 R-372 0.10 32.4 138.3 0.0115 3.09
155 2872 R-200 0.13 40 63.1 0.0385 2.78
155 2871 R-221 0.13 35 221.9 0.0080 1.93
155 2870 R-238 0.13 35 150.9 0.0084 2.85
156 2877 R-251 0.43 47 80.8 0.101 10.45
156 2875 R-190 0.43 40 395.1 0.0277 15.75
156 2876 R-213 0.43 35 1707.2 0. 0064 14.20

 

%A11 materials annealed 1 hr at 1177°C before irradiation at 650°C
to a thermal fluence of 2,3 X 102C neutrons/cm?.

alloys are shown in Fig. 81.__Therresu1ts for alloy 100, which contained
no detectable hafnium, are used for comparative purposes. The general
observation is that the rupture life increased with increasing hafnium
content up to about 0.16%. 'The streSs-rupture properties of the irra-

‘diated alloys shown in Fig 82 were even more dependent upon the hafnium

content than those of the unirradlated material. In this condltlon,

o alloy 156 (0.43% Hf) had far superlor stress-rupture proPertles.

The minimum creep rates of alloys 100 and 152 through 156 are
shown in Fig. 83. The alleysrthat.contained hanfium had lower creep
rates, but the magnitudes showed no clear dependence upon the concentra-

" tion between 0.06 and 0.43%'Hf;_ Irradiation caused a slight increase

in the creep rate of mostrailpys; alloy 156 (0.43% Hf) seemed unaffected.
_The fracture strains ef;the.alloys in the unirradiated condition are

shown in Fig. 84. Generally, the alloys followed the trend of decreasing
 

108

ORNL-DWG 69-14362

 

 

(x10%)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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£ NO. CONTENT (%) - T. _l H

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b o 153 0.06 N

v 154 0.10 ™ 100

o 155 0.13

20 o 156 0.43

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| 0 . | -
107! . 10° 10" 102 103 10

RUPTURE TIME (hr)

Fig. 81. Stress-Rupture Properties at 650°C. of Several Alloys That

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Contain Hafnium. All samples annealed 1 hr at 1177°C before testing.
ORNL-DWG 69-14361
(x10%)
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& ALLOY IHf F P ig::”'\ Il
a 30— "ho CONTENT () [ 100 '521 155
:.E_, o 100
| & 152 0.46
20 o 153 0.06
v 154 0.10
o 155 0.13
10— o 156 0.43
LTI .
10° 10? 10° 10*
RUPTURE LIFE (hr)

Fig. 82. Creep-Rupture Properties at 650°C of Irradiated Alloys
That Contain Various Concentrations of Hafnium. All samples annealed
1 hr at 1177°C before irradiation at 650°C.

 

 
 

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109

ORNL-DWG 69-14360

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

(x40%). ] i
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T a0 NVL
= ST 11
8 4 §§%Q§§;w ALLOY NO.  Hf CONTENT (%)
€ o e o 400 L
« N a {52 0.16
o 153 0.06
154 0.10
20 : 155 0.13 111
o 156 _ 0.43
10 ' l - CLOSED SYMBOLS -IRRADIATED  {H1H
OPEN SYMBOLS -UNIRRADIATED
. L T LT L
1073 . 1072 107! 100 10'

MINIMUM CREEP RATE (%/hr)

Fig. 83. Creep Rates at 650°C of Alloys That Contain Various Con-
centrations of Hafnium. All samples annealed 1 hr at 1177°C.

ORNL-DWG 69-14359

 

40
ALLOY NO.  Hf CONTENT (%)
100
35 152
153
154
30 155
~ 156
&~
= 25
3
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. @ 20
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03 10 1ot 10° 40"

'MINIMUM CREEP RATE (% /hr)

' Fig. 84. TFracture Strains at 650°C of Alloys That Contain Various
"Amounts of Hafnium. All samples annealed 1 hr at 1177°C before testing.
110

fracture strain with decreasing creep rate. Alloy 156 (0.43% Hf) had ./
slightly superior ductility, particularly at the lower strain rates. »
The fracture strains after irradiation are shown in Fig. 85. Alloy 156

(0.43% Hf) had high fracture strains, and alloy 152 (0.16% Hf) had frac- 2

ture strains somewhat- improved over those of the other alloys.

ORNL-DWG 69-14358

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

: | HHIIIH |
| 14 1 ALLOY : 1
CONTENT (%)
©° {OO
’ a 152 0.6
12 o 153 0.06
- v 154 0.10
[ L ¢ 155 0.3
w 10 - 156 0.43
a
§
2 .
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= A
A
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1073 102 wo! 10° 10’

MINIMUM CREEP RATE (%/hr)

Fig. 85. Fracture Strains at 650°C of Irradiated Alloys That Con-
tain Various Amounts of Hafnium. All samples annealed 1 hr at 1177°C
before irradiation at 650°C.

The commercial alloys, 67-503 (0.08% Hf) and 67-504 (0.50% Hf) were
subjected to various tests. The results of tensile tests on alloy 67-503
are given in Table 35. The results on irradiated tests showed clearly
that the anneal of 1 hr at 1177°C before test gave SUperior fracture
strains at 650°C.  More detailed testing was carried out on heat 67-504
(0.50% Hf). This alloy was included in the surveillance program for the
MSRE. Samples were exposed to a molten fluoride fuel salt in the core
for 9800 hr at 650°C and received a thermal-neutron fluence of 5.3 X 1020
heutrons/cm?. Control samples were exposed to static barren salt for

9800 hr at 650°C. Some test results on these samples are presented in .

e
Table 35, Tensile Properties of Irradiated and Unirradiated Alloy 67-503 (0.08% Hf)

 

 

 

 

 

Anneal® ' o Thermal~ ~ Stress, psi -
Specimen Before Temperature, ¢ Neutron ngiin Yield ‘Ultimate _Eigfgifiggi_fi R§§u§:i:n
Number = Irra- = Irra- Pegt , Fluence (min=1) Tensile  Uniform Total (2)
~diation diation - ‘(neutrons/cm?) -
| o : » o x10%0 o x 102 x10° o
6195 816 . . - 650 - - 0,05 48 . 9.4 - 28.1 33.7  27.1
6196 8l . . . 650 - o 0.002  50.6 74.5 14.6 33.8 31.5
6197 = 816 o 760 0.002 37.1 37.8 1.5 24.9  22.7
6193 121 650 0.002 24.3 55.8 21.7 22.9 17.4
6194 121 - 760 - 0.002 31.2 45.3 6.3 2.5 12.9
4999 123 | 760 0.002 27.7 41.2 - 5.5 6.7 4.7
6229 816 650 650 2.0 0.002 33.% 38.9 3.7 4. 6.7
6234 816 650. 650 2.0 0.05 35.1 46,2 6.0 7.0 1.1
4941 816 760 760 2.4 0.002 8.6 8.7 1.7 1.9 6.9
4974 121 760 760 2.4 0.002 14.3 14.7 1.6 1.7 3.2
4949 121 - 650 - 650 2.4 0.002 26.9 40.2 8.0 8.6 10.0
4966 123 760 760 2.4 0.002 7.8 7.8 0.7 0.8 2.1
%Annealing'designationsf 816 =-annealed 8 hr at 871°C; 121 = annealed 1 hr at 1177°C; 123 = annealed

TIT
 

 

112

Table 36; details of these tests were discussed previously.” The frac-
ture strains at a strain rate of 0.05 min~™! are shown in Fig. 86 as a
function of test temperature. The material in the unirradiated condition
was characterized by a high fracture strain of about 50% that decreased
gradually with increasing temperature above 650°C. Irradietion caused
a general decrease in the fracture strain; the decrease was largest at
test temperatures of 25°C and above 600°C. The decrease at 25°C was
likely associated with carbide precipitation and has been observed for
many other modified and standard heats.” The decrease at high tempera-
tures is due to helium produced during irradiatidn. Similar results for
a lower strain rate of 0.002 min~! are shown in Fig. 87. The fracture
strains were lower than those observed at the higher strain rate. Addi-
tional results given in Table 36 show that the higher anneal of 1 hr at
1177°C resulted in better fracture strains after irrediation than the
anneal of 8 hr at 871°C. The samples irradiated in the MSRE for 9800 hr
had slightly higher fracture strains than those irradiated in the ORR
for about 1100 hr. This may have resulted from the different thermal
histories or from the drastically different flux spectra.

 

7H. E. McCoy, An Evaluation of the Molten-Salt Reactor Experiment
Hastelloy N Surveillance Specimens — Third Group, ORNL-TM-2647 (1970).

 

ORNL-DWG 69-14357

 

 

 

 

 

60
o §
50 UNIRRADIATED —3r—
———_L—‘ . \
/ " IRRADIATED "\"\ \
40 / \
O

 

8
*N
/

FRACTURE STRAIN{")

 

3
d

 

 

 

 

 

 

 

 

 

 

 

 

o 100 200 300 400 $00 _ 600 700 g0 900
TESY TEMPERATURE {°C}

Fig. 86. Variation ot Fracture Strains of Alloy 67-504 with Test Témr.
perature at a Strain Rate of 0.05 min~!. Annealed 1 hr at 1177°C. Irra-
diated at 650°C.

",

"
¥

Table 36. Tensile Properties of Irradiated and Unirradiated
Alloy 67-504 (0.50% Hf)

 

a -
Anneal Temperature, °C Thermal: Strain . Stress, psi Elongation, % Reduction

 

Specimen  Before Neutron
NMumber  Trra- Irra- Test Fluence (R?‘tfl ) Yield %fifi%fie Uniform Total 1B (fi €8
diation diation (neutrons/cm?) ‘™
x 1070 x 10° x 103
5058 151 . 25 0.05 67.5 133 50.2 52.3 43.7
5046 151 200 0.05 59,7 119 43.5 44,0 41.6
5057 151 400 0.05 45.5 105 46.2 &47.2 36.4
5064 151 %00 0.002 53.% 123 46.2 47.9 34.2
5061 151 500 0.05 43.1 99.9 .50.6 52.1 41.3
5041 151 500 0.002 61.3 116 44 .5 46.1 33.5
5045 151 550 0.05 45.3 102 47.9 49.4 36.9
5053 151 550 0.002 49.2 99.7 47.2 48.5 39.6
5055 151 600 0.05 45.4 97.7 49.1 50.9 36.7
5044 151 600 0.002 45.8 91.3 41,9 43.7 37.3
5065 151 : 650 0.05 43.3 9%.3 48.8 50,9 37.4
5042 151 650 0.002 44,9 80.2 30.2 41.2 38.1
5067 151 700 0.05 53.3 9% .6 30.8 39.5 36.0
5048 151 700 0.002 42.1 68.1 16.2 45.8 40.6
5066 151 760 0.05 &4 .2 75.4 20.3 43.5 38.4
5047 151 760 0.002 42.7  50.8 6.0  38.9 418
5060 151 850 0.05 34.9 47.8 7.7 38.7 35.5
5056 151 850 0.002 3.5 32.4 2.9 34.8 34.3
6255 121 25 0.05 37.4 104.7 66.6 73.2 73.7
6256 121 200 0.05 33.1 104.2 68.7 71.3 52.2
6257 121 400 0.05 27.1 88 77.1 80.1 60.6
6258 121 500 0.05 30 96.6 67.3 73.9 50.0
6259 121 550 0.05 27.8 91.2 79.0 g82.3 62.5
6260 121 600 0.05 25.5° 87.4 73.6¢ 79.6 56.3
6265 121 600 ¢.002 4.4 71.4 45,7 47.1 34.7
¢ 6262 121 650 0.05 25.6 83 63.9 66,9 45,1
6266 121 650 0.002 23.9 63.6 33.1 34,2 26.9
6263 121 700 0.05 25.7 76.2 46.7 49.3 32.4
6264 121 760 .05 24 65.8 33.5 35.5 25.7
6267 121 760 7 0.002 35.8 42 6.3 11.3 16.4
5086 121 650 25 5.3 0.05 - 102 119 26.0 27.0 33.5
5087 121 650 200 5.3 0.05 46.4 106 43.9 LY A 31.9
5088 121 650 400 5.3 0.05 41 97.6 44,0 4.5 37.2
5089 121 659 400 5.3 0.002 41.8 99 41.8 42.9 36.4
5090 121 650 500 5.3 0.05 40.2 93 42.8 43.3 37.9
5091 121 650 500 5.3 0.002 41.6 9l1.2 42.2 43.2 28.2
V5092 . 0 121 650 © 550 5.3 0.05 39.6 92 46.0 46,7 36.5
5083 - 121 650 550 5.3 7 -0.002 38 78.2 28.2 "29.9 21.2
5079 121 650 600 5.3 0.05 39.7 82.6 35.3 35.7 33.5
5078 ) 121 650 600 5.3 . 0.002 40.2 69.6 23.7 24.8 17.2
5085 121 650 650 - 5.3 - 0.05 32.9 70 30.0 30.7 31.5
5084 2 650 | - 650 5.3 0.002 34.7 63.9 21.9  23.1 17.0
5077 121 650 © 700 5.3 0.05 35 65.9 23.2 24.2 24,5
s076 121 650 700 5.3 0.002  35.8  56.2 1.9 = 1:4.6  13.1
5075 121 650 760 5.3 0.05 34.9 57.6 12.8 13.8 18.0
5074 121 - 650 760 5.3 0.002 37 42.3 4.3 6.7  11.9
5073 121 650 850 5.3 0.05 32.6 40.8 4.8 6.6 9.4
- 5072 12 650 . .850 5.3 - 0,002 25.4 26 v2.1 bl 11.5
- 6238 816 650 - 650 2.0 0.002 34.8 50.4 - 9.2 9.9  13.2
6239 816 650 650 2.0 0.05 34.5 54.5 -10.3 12,0 19.4
T A945 © 816 760 760 2.0 - 0.002 18.5 18.6 S 24 2.8 4.9
4927 121 650 - 650 2.4 . 0.002 30.4 50.4 12.8 16.8 7.0
4986 - 121 760 CoTed 2.4 0.002 13.5 12.5 S o 1.1 6.0
4970 123 760 . 760 2.4 0,002 9.3 9.3 1.2 1.3 5.1

 

, , a'lflmnea].:lng designations: 151 = annesled 1 hr at 1177°C plus 9800‘hrrat 650°C; 121 = amnnealed 1 hr at
. 1177°C; 816 = annealed 8 hr at 871°C; 123 = annealed 1 hr at 1260°C.
 

 

ORNL-DWG 69-14356

 

50 r
UN!RI?ADIATED e o l

IRRADIATED -1,-\ ¢ S "
\

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1\\ |
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FRACTURE STRAIN (%)
8 8
-
\

 

 

 

 

 

 

 

 

 

 

 

 

0 100 200 300 400 500 600 700 800 S00
"~ TEST TEMPERATURE {°C)

Fig. 87. Variation of Fracture Strain of Alloy 67-504 with Test
Temperature at a Strain Rate of 0.002 min~!. Annealed 1 hr at 1177°C.
Irradiated at 650°C.

The results of creep-rupture tests on heat 67-503 are given in
Table 37. The stress-rupture properties are shown graphically in
Fig. 88, and the creep characteristics are shown in Fig. 89. Irradia-
tion reduced the rupture life more at 760°C than at 650°C. Irradiation
at 650°C had no effect on the minimum creep rate, but irradiation at
760°C increased the minimum creep rate (Fig. 89).

Results of creep-rupture tests for heat 67-504 (0.50% Hf) are sum-
marized in Table 38. Data are included from the MSRE surveillance
program. The stress-rupture characteristics are shown in Fig. 90. This
heat had a rupture life eight to ten times higher than that of heat 67-503
(0.08% Hf). The samples aged in fluoride salt for 9800 hr had slightly
shorter rupture lives than those annealed 1 hr at 1177°C before test.
Irradiation at 650°C reduced the rupture life, but less for the samples
irrediated in the MSRE than for those irradiated in the ORR. Irradia-
tion at 760°C in the ORR drastically decreased the rupture life. The
minimm creep rates for these samples are shown graphically in Fig. 91.
The creep Strength of heat 67-504 is about ten times that of heat 67-503.
The minimum creep rate was unaffected by aging at 650°C for 9800 hr in
static salt or by short-term (1000-hr) irradiation at 650°C. Irradia-
tion at 650°C in the MSRE increased the creep rate, and irradiation at

760°C drastically increased the creep rate.
~ Table 37;-fCréep Pr0pefties of Irradiated® and Unirradiated Alloy 67-503 (0.08% HF)

 

b

 

 

— ' ' ' Minimum .
Specimen  Test Anneal _Temperature, °C_ Stress Xupture " Fracture  Reduction
Number  Number Ir?:fi;;:ion Irradiation Test (psi) %;fi? Rate Stg;;n 1n(23ea

| | (%/br)

| ‘ x 103
6179 - 7514 121 o | 650 55 5.6 0.26 24.6 26.0
4957 6257 121 650 47 68.3 0.046 14.3 - 18.0
4958 6256 121 650 40 150.1 0.019 9.9 11.6
6178 - 7513 .. 121 . - 650 30 498.1 0.0045 8.3 17.5
6177 7512 - 121 - ' 760 20 41.1 0.15 9.2 5.1
6176 7511 . 121 760 15 - 128.1 0.039 8.4 16.8
4950  R493 121 - 650 650 47 4ol 0.45 2.8 |
4952 R-505 . 121 650 650 40 32.6 0.071 3.2
4955 ~ R-537 121 650 650 35 45.6  0.073 - 3.7
4975 - R-500. 121 ' 760 650 47 0
4978 . R=515" 121 ' 760 650 35 0.4 0.35 0.77
4979 ~ R-547 121 760 650 30 2.2 0.036 0.11
4967 - R=499 123 - 760 650 47 0
4969 R-541 123 760 650 40 0
4942 - R-510 816 760 650 40 2.6 0.22 0.87
4944 R-539 816 760 650 35 3.3 0.096 0.51
4954+ R-513 121 650 760 15 36.8 0.029 1.7
4977  R-508 121 760 760 12.5 7.6 0.082 0.41
4968 @~ R-533 123 - 760 760 12.5 0.5 0.053 0.07
4943 = R-527 816 760 760 12.5 8.1 0.021 0.68

GTT

 

®Irradiated at indicated temperatures to a thermal fluence of 2.4 X 1020 neutrons/em?.

bAnnealing designations: 121 = annealed 1 hr at 1177°C; 123 = annealed 1 hr at 1260°C;
816 = annealed 8 hr at 871°C.
 

116

ORNL-DWG 69-14355

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

(x103)
ol
\
50--750 65;»0 "\\‘ .
4 - -
\ ‘-h..\"‘ \"\
40 Py * o
N N )

_ N N 650 AN
: 7609 N N
° N \
- - 7§0 \
9 30 - ™
w N
@
-
wn

20

10 o UNIRRADIATED

¢ |RRADIATED
0 ‘ ! I | '
101 100 ot 102 103

RUPTURE LIFE (hr}
Fig. 88. Stress-Rupture Properties at 650°C of Alloy 67-503. All

samples annealed 1 hr at 1177°C before test or irradiation. Number by
each datum point indicates irradiation temperature.

ORNL-DWG 69-14354

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

- (x103) I
50 - -
650 //
40 -
i /
650 4 760
- P L J
a 1
~ V] 760
@ 30 ] s
w
@
-
wn
20
10 o UNIRRADIATED:
 IRRADIATED
o " 1 0
10-3 102 10 : 10

MINIMUM CREEP RATE (%/hr)

Fig. 89. Creep Rates at 650°C of Alloy 67-503. All samples _
annealed 1 hr at 1177°C. Numbers by each datum point indicate irradia-
tion temperature. . ‘
a3

&

 

»

 

117

Table 38. Creep Properties of Irradiated and Unirradiated
Alloy 67-504 (0.50% Hf)

 

 

Anneal® o Thermal- Minimum
Specimen Test Before Temperature, C erature, C Neutron Stress Rupture Creep Fracture Reduction
Irra- : Life Strain in Ares
Number  Number Irra- . Test Fluence (psi) (hr) Rate (%) (4)
diation diation {neutrons/cm?) (%/nr)
X 1029 x 103

6247 7432 121 650 70 1.3 . 3.7 32.9 33.4
6245 7431 121 : 650 63 17.7 0.16 31.2 27.7
4991 6255 121 650 55 127.9 0.029 27.2 37.8
4936 6254 121 650 47 425.5 0.0068 28.7 19.7
4933 6253 121 650 40 . 876.9 0.0030 16.3 20.1
6243 7430 121 650 40 1236.5 0.0034 26.5 27.3
5050 7228 151 650 70 5.5 3.80 43.7 43.9
5043 7227 151 650 55 108.3 0.18 41.7 39.6
5049 7226 151 650 47 329.2 0.049 39.4 43.3
5062 7225 151 650 - A0° 1014.6 0.020 40.8 48.0
6249 7516 121 760 - 20 213.4 0.088 34.8 27.2
4929 R-518 121 650 650 2.4 55 17.0 0.36 13.1

4928 R-507 121 650 650 2.4 47 87.9 0.060 7.6

4932 R-548 121 650 650 2.4 40 261.2 0.017 5.9

4987 R-509 121 760 650 2.% 47 0.3 7.38 2.2

4989 R-543 121 760 650 ‘2. 40 1.3 0.43 0.60

5083 R-733 121 650 650, 5.3 55 59,6 0.091 7.0

5080 R-715 121 650 650 5.3 &7 181.7 0.027 6.8

5081 R-708 121 650 650 5.3 40 467.,2 0.011 6.9

5082 R-732 121 650 650 5.3 . 32.4 1643.8 0.0025 5.9

4971 R-495 123 760 650 2.4 47 0

4973 R-544 123 760 - 650 2.4 40 0.1 . 0.68

4946 R-511 816 760 €50 2.4 47 4.6 0.17 1.8

4948 R-540 816 760 650 2.4 40 19.9 0.020 0.93

5112 R-952 121 650 650 0.25 ©.55 0.6 8.6

5148 R-953 121 650 650 0.25 55 1.0 10.6

4990 R-542 121 760 - 760 2.4 7.5 8.1 0.089 0.9

4930 R-519 121 650 760 2.4 15 287.7 0.015 8.0

4988 R-517 121 760 760 2.4 15 7477 0.042 3.9

4947 R-492 121 760 760 2.4 15 9.3 0.076 1.6

4972 R-504 123 - 760 760 2.4 12.5 0.3 0.35 0.33

 

®Annealing designations: 121 = ennealed 1 hr at 1177°C; 151 = annealed 1 hr at 1177°C plus
9800 hr at 650°C; 123 = annealed 1 hr at 1260°C; 816 = annealed 8 hr at 871°C. '
 

118

 

ORNL-DWG 69-14353
(x10%) Oy

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

-‘ \ o . ]
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= X 650‘ A o)
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= 40 T~ o760 Nol Tal [1IM o
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o« "N
£ 2 650*
w 30 © ANNEALED 1hr AT 1477°C
A ANNEALED 1he AT 1477°C, 9800 hr AT 650°C
I OPEN SYMBOLS - UNIRRADIATED
20 CLOSED SYMBOLS-IRRADIATED
? 10
o _
10! 10° 10! 102 103

RUPTURE LIFE {hr)

 

Fig. 90. Stress Rupture at 650°C of Alloy 67-504. Number by each
datum point indicates irradiation temperature. ‘

ORNL~DWG 69-14352

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

3 .
{xt0°) - -
L
- - ’-‘
o
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a| 48650 J 4
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=
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T a0 o4 LT 650 .
2 7 LY
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£
20 o ANNEALED 1hr AT H77°C
A ANNEALED 1hr AT 1477°C, 9800 hr AT 650°C
OPEN SYMBOLS—UNIRRADIATED
0 CLOSED SYMBOLS— IRRADIATED
1°

010'3 w2 ! 10"

MINIMUM CREEP RATE (Yo/hr}

Fig. 91. Creep Rates at 650°C of Alloy 67-504. Number by each
datum point indicates irradiation temperature.

 

 

 
 

 

119

cfij The fracture strains of heats 67-503 and 67-504 in the unirradiated
condition at 650°C are shown in Fig. 92. Heat 67-504 (0.50% Hf) had
much higher fracture strains than heat 67-503 (0.08% Hf), particularly

. at the lower creep rate.

ORNL-DWG 69- 14351

50

40

30

20

FRACTURE STRAIN (%)

10

 

0
103 10°2 101 100 to!
MINIMUM CREEP RATE (% /hr)

Fig. 92. Fracture Strains at 650°C of Unirradiated Alloys 67-503
and 67-504. All samples annealed 1 hr at 1177°C.

»

The rather sparse data for heats 67-503 and 67-504 at [760°C are pre-
w sented in Tables 37 and 38, and the stress-rupture properties are plotted
in Fig. 93. 1In the unirradiated condition, alloy 67-504 had a rupture
life about five times greafier than that of alloy 67-503. Irradiation at

ORNL~-DWG 69-14350
5 x 104 ,

   

— ‘0, A UNIRRADIATED
&, & JRRADIATED

2 11 : 760

Tflms

ot

 STRESS (psi)

5x10° \
w0 e o 10° 10
" RUPTURE TIME (hr)

0

' Fig. 93. Stress-Rupture:Properties‘at 760°C of Alloys 67-503 and
67-504. All samples annealed 1 hr at 1177°C. Number by each datum
* point indicates the irradiation temperature.
 

 

120

650°C reduced the rupture life, but irradiation at 760°C reduced it much
more. The fracture strains in the irradiated condition are shown in

Fig. 94. The fracture strains of heat 67-503 were 3 to 4% when irra-
diated and tested at 760°C. Irradiation at 760°C and testing at either
650 or 760°C resulted in lower fracture strains. Heat 67-504 generally
had fracture strains of 6 to 8% when irradiated at 650°C and tested at
650 or 760°C. Irradiation at 760°C resulted in fracture strains as low
‘as 0.6%.

| Several of the samples were examined metallographically after test-
ing. Typical photomicrographs of a sample of heat 67-504 after exposure
to a noncorrosive fluoride salt for 9800 hr at 650°C and testing at 25°C
are shown in Fig. 95. The fracture was partially intergranular, although
the bulk grains elongated greatly. Also more carbides formed than in the
annealed material shown in Fig. 80. At 650°C, the grains still showed

ORNL-DWG €69-14349

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

14 F T P T 1T 17T% i | LN BLE) lll
o 67-503
s 67-504 650
o L OPEN SYMBOLS TESTED AT 650°C
< CLOSED SYMBOLS TESTED AT 760°C
IRRADIATION TEMPERATURE SHOWN
BY EACH SYMBOL
10 [ 1]
=
z 8 6504 650
= I !‘“
3 650,MSRE & . 3650, MSRE
x Pl 650, MSRE
K 6 -8 650,MSRE 2650
<
@
w
4
b 650
76041 | p650 -
l o650
2 .ij ] 76041
6504 s 760
7604|
h 760
o 760,| | | |F760 [760°[ |7 |
103 10-2 107 - 400 10!

MINIMUM CREEP RATE (% /hr)

Fig. 94. Postirradiation Fracture Strains of Heats 67-503 and
67-504. o
 

 

 

 

h

&

 

 

121

Y-92935

 

 

 

 

 Fig. 95. Photomicrographs of Sample of Heat 67-504 (0.50% Hf)
Annealed 1 hr at 1177°C, Exposed 9800 hr to Fluoride Salt at 650°C, and
Tested in Tension at 25°C. (&) Fracture, 100x; (b) typical unstressed,
100X. Etchant: glyceria regia.
 

122

-extensive elongation, and there are numerous intergranular cracks,

 

although the final fracture was mixed intergranular and transgranular
(Fig. 96). After irradiafion, the fracture of -heat 67-504 at 25°C

(Fig. 97) was more intergranular than that of the unirradiated material
(Fig. 95). A typical fracture of heat 67-503 after creep testing at
650°C is shown in Fig. 98. There are numerous intergranular cracks,

and the fracture is predominantly intergranular. Heat 67-504 (Fig. 99)
exhibited fewer intergranular cracks and a fracture that was mixed inter-

granular and transgranular.

 

 

 

 

,Fig. 96. Photomicrograph of the Fracture of a Sample of Heat 67-504
(0.50% Hf) Annealed 1 hr at 1177°C, Exposed 9800 hr to Fluoride Salt at
650°C, and Tested in Tension at 650°C and a Strain Rate of 0.002 min-1.
100x. o : '

 
123

 

Fig. 97. TFracture of a Sample of Heat 67-504 (0.43% Hf) Annealed
1 hr at 1177°C, Irradiated in the Molten Salt Reactor Experiment to a
Thermal Fluence of 5.3 X 10?0 neutrons/cm? over a period of 9800 hr at
650°C, and Tested in Tension at 25°C. Etchant: glyceria regia. 100x.

3

&

 

o Y-89473

¥

 

* Fig. 98. Fracture of 8 Sample of Heat 67-503 (0.08% 'Hf) Annealed
-\ 1 hr at 1177°C and Creep Tested at 650°C and 40,000 psi. Etchant:
glyceria regia. 100X.
 

 

 

 

124

 

 

 

Fig. 99. Fracture of a Sample of Heat 67-504 (0.50% Hf) Annealed
1 hr at 1177°C and Creep Tested at 650°C and 40,000 psi. Etchant:
glyceria regia. 100X,

Alloy Containing No Additions

The results for a small, 2-1b laboratory melt (alloy 100) that con-
tained no'additions of Ti, Zr, or Hf were presented above. One small,
100-1b commercial alloy (heat 21546) was also procured. This alloy was
recéived in the cold-worked condition. Samples were tested in this con-
dition and-after annealing at 871 and 1177°C. Typical photomicrographs
of.thérmaterial are shown in Figs. 100 and 10l. The microstructure after
annealing ét 871°C was characterized by bands of precipitate and fine
grains. The bands were caused by carbide precipitatioy and retarded
recryétallizatiOn in these areas. The material outside these bands was
recrystallized. Annealing for 1 hr at 1177°C (Fig. 101) completed
recryStéllizétiOn;_ Some stringers were still present, although none
are visible in Fig. 101. These stringers were quite similar to those

found in heat 21545 and appeared to be unmelted molybdenum.
®)

N

125

 

 

 

| Fig. 100. Photomicrographs of Alloy 21546 Annealed 100 hr at 871°C.
The material was cold worked 40% before annealing. (a) 100X and (b) 500x.
_V'Etchant:  glyceria regia. : '
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Y-71901

   
 

()

Fig. 101. Photomicrographs of Alloy 21546 Annealed 1 hr at 1177°C.
The material was cold worked 40% before annealing. (a) 100x and (b) 500x.
Etchant: glyceria regia.
¥

"

»

127

The results of tensile tests on this material are summarized in
Teble 39. The fracture strains are shown as a function of temperature
in Fig. 102. Up to about 700°C, the material annealed 1 hr at 1177°C
had the highest fracture strain and the cold-worked material had the
lowest fracture strain. Above 700°C, the ductility of the cold-worked

material increased markedly. Only samples as received and as annealed

‘100 hr at 871°C were irradiated, and these were all tested at 600°C and

above. Irradiation decreased the fracture strain of all samples.

The results of creep tésts on this material are given in Table 4O.
In the unirradiated condition, the material as received had the highest
strength and the material annealed at 871°C had the lowest strength.
The strength of material énneéled 1 hr at 1177°C fell intermediate, and
several illustrations follow to compare the properties of irradiated and
unirradiated material annealed 1 hr at 1177°C. The stress-rupture proper-
ties at 650°C are shown in Fig. 103. The irradiation temperature was an
important factor; the higher the irradiation temperature, the shorter wa.s
the rupture life. The two sets of data seem to converge, indicating
that at-very'iow levels.of stress there would be no effect of irradiation
on the rupture life. The creep rates are shown in Fig. 104 for 650°C.
As.long as the irradiation temperature was about 650°C, there was no
difference between the creep rates of irradiated and unirradiated mate-
rial. At irradiation temperatures of 700°C and ébove, the postirradia-
tion créep rate was increased. There was at least one datum point that

disagreed with this general trend. These curves also converged, again

"indicating'common behavior at low stresses.

The fracture strains of irradlated samples tested at 650°C are
showvn as a function of mlnimum creep rate in Fig. 105. All,of the sam-
ples irradiated above 700°C had fracture strains of 1.1 to 0.2%. Irra-

‘diation at about 650°C resulted in fracture strains of 1.1 to 5.1%. The

exact dependence of the fracture strain on strain rate cannot be deter-

mined from the avallable data, and the indicated curve is influenced by

" results on standard Hastelloy N (ref. 8).

 

8H. E. McCoy, "Varlatlon of Mechanical Pr0pert1es of Irradiated
Hastelloy N with Strain Rate," J. Nucl. Mater. 31(1), 67-85 (1969).
128

Table 39. Tensile Properties of Irradiated and Unirradiated Alloy 21546

 

 

 

 

 

 

 

Anneal? ‘ Thermal- '
Specimen Before  lemperature, °C Neutron s;r:.in —% Elongation, % Ridugfiim
Number Irra- Irra- Test Fluence a%¢y Yield € Uniform Total ~Tygy o
diation diation (neutrons/cm?) (min=?) Tensile (%)
x 1020 x 10> x 103
10220 121 25 0.05 38.7 107.5 69.6 73.7 75.1
10221 121 7 200 0.05 32.6 99.7 69.8 73.8 63.3
10223 . 121 - 260 0.05 31.4 100.% 71r.5 76.4 68.0
10222 121 400 0.05 30.2 97.8 73.0 79.3 64.8
10224 - 121 550 0.05 27.1 90.5 3.4 78.3 59.5
10225 121 600 0.05 27.5 87.1 66.0 69.5 47.2
10226 121 650 0.05 23.7 €9.8 46.3 48.3 32.1
10227 121 ' 700 0.05 24 .4 64.7 35.1 36.1 26.9
10228 121 760 0.05 24 60.4 - 25.0 25.9 20.0
1440 000 25 0.05 151.9 164.2 6.8 13.4 53.3
1441 000 427 0.05 127.3 144.2 12.1 15.5 32.7
1452 000 650 0.05 119.5 132.7 R 13.6 18.8
1453 000 650 0.002 109.1 114.8 3.4 8.2 16.3
1451 000 760 0.05 76.4 T78.4 2.6 32.3 51.1
1457 000 871 0.05 31.5 31.7 1.4 46.7 58.4
1458 000 871 0.002 19 19 1.0 48.1 44,0
1459 000 982 0.05 — 17.9 17.9 0.9 60.8 60.6
1749 16 25 - 0.05 51.7 119.4 49.2 53.5 72.4
1750 16 427 0.05 40.9  105.2 47.9 52.4 46.2
1746 16 650 0.05 44,3 95.5 27.8 30.0 31.2
1747 16 650 0.002 38.8 69 19.0 35.1 64.7
1752 16 760 0.05 37.8 51.5 3.6 7.5 68.2
1753 16 871 0.05 27.7 28.8 10.0 53.5 69.1
1754 16 871 0.002 18.2 18.2 0.9 45.3 51.2
1438 000 600 550 3.5 0.002 115.2 127.9 6.0 6.4 7.1
1439 000 600 600 3.5 0.002 105.3 115 b.te 4.5 4.0
1436 000 600 650 3.5 0.05 106.8 116.6 5.7 6.2 g.é
1437 000 600 650 3.5 0.002 103.1 106.8 3.5 3.7 4.0
1447 16 600 550 3.5 0.002 41.9 74.8 14.8 15.1 16.2
’ 1446 16 600 600 3.5 0.002 43.7 65.6 8.2 8.4 10.2
1444 16 600 - 650 3.5 0.05 38.8 64 .2 10.3 10.5 = 1.7
1443 16 €650 650 1.0 0.002 38.6 55 7.3 7.4 12.9
1442 16 650 650 1.0 0.05 38.4 €7.5 13.5 13.7 - 14.7
1445 16 600 650 3.5 0.002 45 56.7 5.6 5.7 - 8.6
1448 16 600 760 3.5 0.002 33.5 33.8 1.8 4.3 7.1

 

®pnnealing designations: 121 = annealed 1 hr at 1177°C; 000 = as received (40% cold work);
16 = annealed 100 hr at 871°C. :
29

 

»

L 1

129

Table 40. Creep Properties of Irradiated and Unirradiated Alloy 21546

Anneal® o " Thermal- Minimum

 

Specimen Test Before Exgg‘;er_axir_g,__c © Neutron Stress mfl?;w:re Creep F‘;‘:c:zre Ridu:f‘l:n
Number Number Irra- dietion  Test Fluence {psi) (br) Rate f%)n n(%)e
diation (neutrons/cm?) (%/nr)
) x 1020 x 103
4056 6300 121 © . 650 - 62 5.3 0.56 25.6 24.9
4057 6301 121 + . 650 o 55 13.5 0.22 21.3 21.3
4058 6302 121 650 47 63.0 0.074 3.1 16.6
4059 6303 121 , 650 40 187.5 0.026 19.4 15.6
10219 7288 121 650 35 256,2 0.013 13.0 16.2
10218 7289 121 - 650 30 56l1.6 0.0080 1.1 18.7
1455 5528 00 : 650 70 1.7 - - 0.31 12.5 15.5
1454 5527 000 650 . 55 92.7 0.025 12.5 14.0
1456 5408 000 650 47 223.6 0.014 7.5 6.3
1751 5531 16 650 70 3.1 18.0 34.4 31.0
1748 5536 16 _ 650 55 6.9 2.87 46.9 33.7
1745 5532 . 16 - - 650 47 31.7 0.83 46.9 39.6
1826 - 5435 816 650 40 28.8 .20 8.5 9.4
10231 7521 121 760 15 221.3 0.08 16.3 15.8
10229 7517 121 760 . 12.5 4147 0.0142 13.4 12.7
5988 R-598 121 © 235 - 650 2.4 47 84.5 0.011 2.3
4060 R-294 121 650 - 650 - 2.5 47 10.1 0.41 5.1
4061 R-482 121 657 650 2.5 40 0
4065 R-481 121 650 650 2.5 40 9.2 0.019 2.4
5981 R-635 121 666 650 2.4 40 62.0 0.033 2.8
5980 R-604 121 666 - 650 2.4 35 20.1 0.042 1.1
5979 R-642 121 - 766 650 2.4 35 0.4 2.1 0.9
5974 R-602 121 7547 . 650 2.4 35 0
5965 R-605 121 732 650 2.4 35 0.4 L3 0.6
5975 R-607 121 760 650 - 2.4 . 30 1.2 0.44 0.8
5966 R-609 121 - 704 - 650 2.4 30 1.1 0.15 0.7
5982 R-643 121 666 -650 2.4 27 727.8 0.0030 3.1
5967 R-638 121 732 650 2.4 25 46.6 0.0065 0.4
5969 R-639 121 816 650 2.4 25 11.1 0.015 0.4
5985 R-600 121 499 650 2.4 25 9.4 0.26 3.2
5976 R-631 - 121 760 650 2.4 21.5 68.3 0.0039 0.53
5968 R-632 121 832 650 2.4 21.5 69.6 0.0028 0.24
5971 R-637 121 871 650 2.4 17 1157.1 0.0006 1.1
- 5985 R-600 121 499 - 760 2.4 25 9.4 0.26 3.1
5972 R-640 121 799 _ 760 2.4 15 4.3 0.061 . 1.3
4066 R-296 121 = 746 760 2.5 15 0.5 0.2 0.7
4067 R~-468 121 746 760 2.5 12.5 3.6  0.060 0.4
5970 R-634 121 © 832 - 760 2.4 12.5 " 45.7 . 0.014 0.8
5973 . R-641 121 799 C760° . 2.4 100 1109  0.0098. 1.5
- 4068 R-469 121 = 746 C 780 2.5 10 - 13.8 0.036 1.1
1450 R-120 16 600 650 3.5 40 3.3 0.40 . 1.3
1449 ‘R-113 16 - 600 ~. 650 3.5 32.4 o7.7 0.038 4.0

 

®pnnealing designations: 121

annealed 1 hr at 1177°C; 000 = as received (40% cold work)
16 = annealed 100 hr at 871°C; 816 8 o '

_annealed 8 hr at 871°C.
 

130

 

ORNL-DWG 69-14348

 

 

 

 

 

 

 

 

 

 

 

 

80
[
___-—-—"

70

60 -
el —— \ 7
£ 50 [
Z \ \
=
=
9 40 © 4 hr AT #77°C \
w 4 COLD WORKED \
g 0 100 hr AT 874°C
< 30
£ f

CLOSED SYMBOLS, IRRADIATED

»

\
S \

N

OPEN SYMBOLS , UNIRRADIATED \

8

 

 

A
/
)

 

 

 

 

 

 

 

 

 

 

 

 

0

 

Y 100 200 300 400 500 600 700 800 900 1000
: TEST TEMPERATURE (°C)

Fig. 102. Variation of the Fracture Strain with Test Temperature
for Alloy 21546 at a Strain Rate of 0.05 min~?!.

ORNL-DWG 69-1434T7

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

(x103)
\\b‘
»
60 .\q
50 &5 N
> h| & 235
h,
S 40e657 650 €66
a
- 766 \
vy
a ® 753 T~ 732 b 650 N
x by
| .
30 T04 o®
“ 760 T N
asg da~—1._| 1% NI
S T, I
760% n o
20 a2 L {{1] a7
el @
[
0 © UNIRRADIATED
 IRRADIATED
0
A0 10° 10! . {02 103

RUPTURE TIME (hr)

Fig. 103. Stress-Rupture Properties at 650°C of Alloy 21546. All
samples annealed 1 hr at 1177°C before testing. Number by each datum
point indicates irradiation temperature.
&

131

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

p " ORNL-DWG 69-14346
(x403) ’ l — -
| /’/
60 o UNIRRADIATED At o1
o IRRADIATED | . T 17
50 o1 /|
2357 650 [T/
* , :
. o . 1
§40 650 setls
I A
v 732 766
= 30 ||| 1] 704 | L-H
? °3° ’77::5 816 ™ i
(el e
4
20 ‘832"760
87
7 J,/’
10
0

1074 103 - 10-2 107 100 10!
'MINIMUM CREEP RATE (% /hr}

Fig. 104. Creep Rate at 650°C of Alloy 21546. All samples annealed
1 hr at 1177°C before testing. Number by each datum point indicates
irradiation temperature.

ORNL-DWG 6914345

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

6

5 6508,
&
z 4
a
o
|_
n 3 666“ A .499
2‘5" ' W\ | leee6
5 [ oxae. 4650
Q 235-\
re . N

N\ /
, - \__,4 ‘
p871 RN - b~ |
1 ‘366 - + .766
760 . _ - - e704 e [760 o
o 8324 | 732 | | | l
1074 10°3 . 402 - 10~! 400 10!

MINIMUM CREEP RATE (%/hr)

Fig. 105, Fracture Stra:.ns a.t 650°C of Irradlated Alloy 21546
- All samples amnnealed 1 hr at 1177°C before 'best:a.ng Number by each datum
point indicates 1rradia.t10n temperature. / _ _
 

 

132

Some samples were tested at 760°C; the rather sparse results are
given in Table 40, and the stress-rupture data are shown in Fig. 106.
There was an indication that irradiation at 746°C may result in poorer
stress-rupture properties at 760°C than irradiation at 799 to 832°C.

The minimum creep rate at 760°C (Fig. 107) showed that irradiation did
not alter the creep rate drastically. However, there was again the ten-
dency for the three samples jrradiated at 746°C to show the greater
effect. The fracture strains at 760°C were all quite low (0.4 to 1.5%
strain) except for the sample irradiated at 499°C (3.1% strain).

ORNL-DWG 69-14344
5x10?

© UNIRRADIA
® IRRADIATED

STRESS {psi)

w0t

 

5x10°
107! 1o° 10’ 102 103
RUPTURE LIFE (hr)

Fig. 106. Stress-Rupture Properties at 760°C 6f Alloy 21546. All
samples annealed 1 hr at 1177°C. Number by each datum point indicates
irradiation temperature.

4 ORNL-DWG 69-14343
4x10

© UNIRRADIATED
& [RRADIATED

STRESS (psi)

10t

 

3

5x40

10~3 1072 g~ 10°
MINIMUM CREEP RATE (%/hr)

Fig. 107. Creep Rates at 760°C of Alloy 21546. Material annealed-

1 hr at 1177°C. Number by each datum point indicates irradiation
temperature. .
 

133

Some insight intOfthe;effect of irradistion temperature was gained
by transmissionrelectron'micrOSCOPy. Typical microstructures for two
samples irradiated at 666 and 766°C are shown in Figs. 108 and 109.

 

‘Fig. 108. Transmission Electron Photomicrograph of Alloy 21546
after Irradiation at 666°C for about 1100 hr to a Thermal-Neutron Fluence
of 2.4 X 1029 neutrons/em?®. 50,000x.
 

 

 

YE-10195

 

- Fig. 109. Transmission Electron Photomicrograph of Alloy 21546
after Irradiation at 766°C for about 1100 hr to & Thermal-Neutron Fluence
of 2.4 x 10%° neutrons/em®. 50,000x.

»
e

135

 

After irradiation at 66620 the grain boundar;eS'were completely lined

   

 

with fine prec1p1tate particles, and there wefe also numerous precipi-

tates within the grains. After irradiation at 766°C (Fig. 109), there

were fewer relatively coarse precipitates along the grain boundaries,

and only random coarse precipitates were present within the grains. The

precipitates in both samples were identified -as carbides of the MxC type

in which the metal component was primarily molybdenum. Thus, the poorer

postirradiation properties are associated with the very coarse precipitate.
A typical fracture of heat 21546 at 650°C is shown in Fig. 110.

There was extensive intergranular cracking, and the fracture was entirely

intergranular. The random stringers are obvious in this field and may have

been instrumental in causing the fracture at this particular location.

 

 

‘Fig. 110. Fracture.offeiSample'of Heat 21546 Annealed 1 hr at
1177°C and Tested at 40,000 psi and 650°C. Etchant: glyceria regia.

BISCUSS ION OF RESULTS

A large‘amount of data has been presented and we should recapitu-

late some of the most important observations.
 

 

136

1. A modified base composition of Hestelloy N was developed:
Ni-12% Mo—7% Cr-0.2% Mn—0.05% C. This alloy is free of the carbide
stringers present in standard Hastelloy N, but is still very susceptible
to embrittlement by irradiation. | |

2. Additions of Ti, Zr, or Hf to this base composition improved
resistance to irradiation embrittlement. Small heats showéd a general
improvement with increasing additions of titanium and hafnium. Similar
work with additions 6f zirconium showed that an alloy containing 0.53% Zr
had the minimum creep rate and maximum rupture life. Other concentra-
tions of zirconium resulted in higher fracture strains.

3. Commercial melts (100 1b) were obtained with nominal additions
of 0.5% each of Ti, Zr, and Hf. The properties of these alloys were
comparable with those of the small, laboratory melts (2 1b).

4. Properties were very dependent upon the irradiation temperature.
Irradiation at about 650°C resulted in good properties, but 1rrad1at10n
at 700°C and above gave relatively poor properties.

5. Transmission electron microscopy revealed that the effect of
irradiation temperature was associated with the precipitation of coarser
carbides at the higher temperature. During irradiation at 650°C, fine
carbides precipitated, whereas above 700°C relatively coarse éarbides
formed.

6. The optimum properties after irradiation result from annealing
for 1 hr at 1177°C before irradiation. The fine-grained microstructure
obtained by annealing at low temperatures such as 871°C was not as strong
or as ductilé, and annealing above 1177°C decreased the fracture strain
of irradiated and unirradiated samples.

_.7.4 The general trend revealed by metallographlc examination after
testlng was that additions of Ti, Zr, and Hf reduced the frequency of
cracking at grain boundaries. Fractures in these alloys in creep at
650°C are predominantly intergranular but usually have some transgranular
components. |

Some cqmparisons.are in order to show the relative properties of
these alloys and standard Hastelloy N. For this comparison, we uséd
only the commercial alloys with 0.5% nominal additions of Ti, Zr, and
Hf. All results are for samples annealed 1 hr at 1177°C. The tensile
 

L3

137

and yield strengths are shown in.Fig.'lll., Alloys 21546 and 21545

(0.49% Ti) had very similar properties; the observed strengths were
slightly less than those measured for standard Hastelloy N. The strength
of heat 67-504 (0.50% Hf) was the highest of all alloys studied. The
fracture strains of these samples are shown in Fig. 112. Alloys 21546,
21545 (0.49% Ti), 21554 (0.35% 2Zr), and 67-504 (0.50% Hf) have similar
properties. Heat 67-504 (0.50% Hf) after aging for 9800 hr at 650°C had
fracture strains lower than those ofrthé other modified alloys when tested
below about 650°C and higher above this temperature. Standard Hastelloy N

had lower fracture strains fhan any of the modified alloys.

ORNL-DWG 69-14342

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

(x103) ' l I |
, a 21546
A 21545 .
140 * 57-504 (0.50 % Hf), AGED
.\ - 9800 hr AT 650°C
o 67-504
¢
120 I~ o 21554
ULTIMATE \ ¢ STANDARD HASTELLOY N
é\ \" l .
100 g \S\L__ S ——— *.__
. . L TT——0-a e _
= . i~ \\\
‘J—,’ 80 AN X
Ll ' g e
5 ol NN
" 60 \\ . a 2\
0 ‘aa\\\\\ { \\
YIELD b i & o . .\
40 A \
-\g . ° ) %
o — % .
. - e \.ir o _&‘—IH—A
20 '
0 100 200 300 . 400 500 - 600 700 800 900

TEMPERATURE (°C)

Fig. 111. Tensile'Prfipérties'of Several Commercial Melts of Modi-
fied Alloys. All material annealed 1 hr at 1177°C.

The relative stress-rfipture properties of these élloys at 650°C

‘are shown in Fig. 113.:-The;ffipture lives at 40,000 psi range from 170

to 1100 hr; alloy 21546 had the shortest rupture life, and alloy 67-504
(0.50% Hf) had the longest. Minimum creep rate is a better measure of
 

138

ORNL-DWG 69-14344

 

 

 

 

 

 

 

 

  

 

 

 

 

 

 

 

 

 

 

 

.90
/E‘L\.\
80 . ] 0:—. \
3.—_—____—-11:_—-&"" | b ‘ A\
‘ ,
70 |— ' _ \
_ ® )
¢
: _ \ \21554 (0.35% Zr)
§6° 67-504 (0.50 % Hf)
Z = 67-504, AGED . }& \\!
§ 50 - _ ~ " - 2 -
’u.) & i
E g
>
x ¢ 2 N
30 L& 21545 (0.49 % Ti) _ | 21545(0.49% Ti)
a 21546 21546 \ 4
s 67-504 (AGED 9800 hr AT 650°C) 4
20 |-° 21554 (0.35 % Zr)
o STANDARD HASTELLOY N -
® 67-504 (0.50 % Hf)
10
o

 

 

 

 

 

 

 

 

 

 

 

0 100 200 300 400 500 €00 700 800 900
TEST TEMPERATURE (°C) ‘

Fig. 112. Variation of Tensile Fracture Strain with Test Tempera-
ture for Several Commercial Melts of Modified Alloys at a Strain Rate of
0.05 min“!. All material annealed 1 hr at 1177°C.

ORNL~DWG 69- 14340

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

(x10%) IR T T TV TTIm
-...:%._——— 21554 (0.35 % Zr)
NSNS i 11 T
60 , : {1 67-504 (0.50 % Hf) T
\ -\\Eu{: P o
N TN ~.,%<21545 (0.49 % Ti)
I
50 NN
N T \\i\
— . Ny \ \\\\
> \\\ \\:‘
& 40 7 N NN
-
N
- NN
u N ™N
& 30 Pt
& LAY ™ I
STANDARD HASTELLOY NN "%\
! ;
20 21546—’)\
10
o
100 10! 102 103 104

RUPTURE LIFE (hr)

'Fig. 113. Stress-Rupture Properties at 650°C of Several Commercial
Melts of Modified Alloys. All material annealed 1 hr at 1177°C before
testing.
)

139

strength than rupture life; these data are shown in Fig. 114. At
40,000 psi, the minifium creep rate varied from 0.03%/hr for alloy 21546
to 0.002%/hr for alloy 67-504 (0.50% Hf). The addition of hafnium
increased'strength the most,,by far, but the three alloys modified with
Hf, Zr, and Ti were all stronger than standard Hastelloy N. There was
not much difference between the creep strengths of standard Hastelloy N
and the modified base alloy 21546. The fracture strains of the unirra-
diated modified alloys are shown in Fig. 115. All of the alloys except
alloy 21545 (0.49% Ti) followed the trend of decreasing fracture strain
with decreasing creep rate. At high strain rates, alloy 21554 (0.35% Zr)
had the highest fracture strain. At low strain rates, alloy 21545
(0.49% Ti) had superior ductility. Alloy 21546 and standard Hastelloy N
had the lowest fracture strains and did not differ appreciably from each
other.

The stress-rfipture properties determined at 650°C after irradiation

are compared in Fig. 116. After irradiation at 650°C, alloys 21554

ORNL-DWG 69-14339

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

(x403)
| =7 ’,7
60 . . ; /7;/ /, ]
ot Y
' A7 "‘7//5‘; DlARD HASTELLO
21554 10.35% 20 7] Lt AN LLOY N
NS ] /
50 : - i
AT 0T | LT
A A
67-504(0.50% Hf) ] pra

™ s
3 40 §” —
o ) / “/ . .
"(3 / r” 4
@ 30 |- Y _A /
E 21545(049% i 7 FH

| LA

20 ‘ AL

o+
11 21546
10 ' '
0 L _
Io"4 o - t0—'_’, e _i'o.z 1o \o° 1o

- - MINIMUM CREEP RATE {%/hr)

Fig. 114. Creep_Réftes_ at 650°C of Several Commercial Melts of
Modified Alloys. All material annealed 1 hr at 1177°C before testing.
 

 

140

ORNL-DWG 69-14338

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

50
i
e
A

40 7
= 21545 (0.49% Ti) | 1 67-504 (0.50% Hf) |
é 30 > ~ | - Y ot =t J A
‘5 | -7 ] —", i
g 20 H a : _{fi T
5 21554(0.35% Zr) 41 T
S ‘ 2
& stanoaro (| | | ||

10 } HASTELLOY N H=—1F 5z et

w04 1073 102 0! 10° .0

MINIMUM CREEP RATE (%/hr)

Fig. 115. Comparison of the Fracture Strains at 650°C of Several
Commercial Melts of Modified Alloys. All material annealed 1 hr at 1177°C.

ORNL~-DWG 69— 14337R

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

" [T 1
—21545 ° 21554 (0.35% Zr) ||
80 N a 21546
§7|—504—a\§\
STANDARD NN
50 < _HASTELLOY N N
\\ \.\. h,.:.co\
"-\\ P~ 1 \
% ) M | RN
=% “Gswr [k S~ IRRADIATED AT
g \___-“ g b e .,._.'. N.\ ssooc
uw moil N
= 30 \ ~<H] 11 <
\ 1
IRRADIATED AT | T+t >~
760°C n .,
20 : T
N21546 ]
21545 (0.49%Ti)
10
o .
10" 10° o' 10° 10° 10

'RUPTURE LIFE (hr)

Fig. 116. Postirradiation Stress-Rupture Properties at 650°C of
Several Commercial Melts of Modified Alloys. All material annealed 1 hr
at 1177°C before irradiation. :

(0.35% Zr), 21545 (0.49% Ti), and 67-504 (0.50% Hf) had equally good
properties. All three of the modified alloys showed improvement over
the standard Hastelloy N shown for reference. The base modified

alloy 21546 had properties quite similar to those of standard Hastelloy N.

e
.

141

_Standard Hastelloy N that contained about 0.5% Si is not sensitive to

irradiation temperature, and the indicated line holds for irradiation at
760°C (ref. 9). Heats 21546 and 21545 (0.49% Ti) both had poor stress-
rupture properties after irradiation at 760°C. The stress- -rupture proper-
ties of alloy 67-504 (0.50% Hf) fell intermediate between those of these
two alloys and those of standard Hastelloy N.

The postirradiation fracture strains for irradiation and testing
at 650°C are summarized in Fig. 117. These curves are based on rather

sparse data and should be considered as schematics. Allcy 21546 shows

- some improvement over standard Hastelloy N. Alloy 21545 (0.49% Ti) is

generally better than standard Hastelloy N, but still has a distinct
ductility minimum of only about 3%. The sparse data for heats 21554
(0.35% 2r) and 67-504 (0.50% Hf) indicate that these heats had improved
fracture strains. The presence of a ductility minimum has been estab-
lished for standard Hastelloy N and for heat 21545, but there are too
few data to establish whether the other alloys conform to this pattern.

 

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

 

ORNL-DWG 69—14336

-
N

\( 21545 (0.49% Ti)

 

o 21854
® 67-504 (0.50 Hf)
10 o=
g . - B
E .
28
-
wn
o
S 6
-
Q
<1
m.
*a
2 _
"HASTELLOY N |
0 —* y

MINIMUM CREEP RATE (%/ hr)

.. Fig. 117. PostirradlatiOn Fracture Strains of Several Commercial
 Melts of Modified Alloys Irradiated and Tested at 650 C. All material
annealed 1 hr at 1177°C before irradiation. S '
 

 

 

142

Many of the properties of these alloys are likely determined by the
carbide precipitate. As already shown, alloy 21546 exhibited a change:
in properties after irradiation that correlated well with changes in the
carbide structure (Figs. 108 and 109). Further work by Gehlbach and
Cook!® showed that Ti, Hf, and Zr promoted the formation of the MC type
of carbides. The normal type of carbide for Hastelloy N is MéC when the
silicon content is low and M¢C when the silicon content is high.ll The
work of Gehlbach and Cook!® also showed that low levels (about 0.5%) of
Ti, Hf, and Zr can result in an MC type of carbide at 650°C and a rela-
tively coarse M,C type at 760°C. Thus, the strong effect of irradiation
temperature on these alloys is associated with this transition in carbide
types. |

- ' The mechanism whereby these precipitates improve the unirradiated
creep properties also likely has to do with the carbide precipitates.l?
As shown in F:Lg 114, the creep strength is increased with the addition
of Ti, Zr, and Hf. These effects are larger than anticipated from solid-
solution strengthening, particularly when one considers that most of the
alloying additions are likely tied up as carbides. Thus, the strengthen-
ing observed is likely due to the finely dispersed carbide phase. These
precipitates also likely impede crack propagation and account for the
improved fracture strains (Fig. 115).

In irradiated material, the further complication of helium trans-
muted for the 19B exists. Our reasoning in adding Ti, Hf, and Zr was to
tie the boron up in dispersed precipitates. The only measure that we
have of the location of boron is to assume that it behaves like the car-

"bon. Thus, each carbide precipitate contains some boron. The finely

 

10R. E. Gehlbach and S. W. Cook, Metals and Ceramics Div. Ann.
Progr. Rept. June 30, 1969, ORNL-4470, p. 187.

11R. E. Gehlbach and H. E. McCoy, Jr., "Phase Instability in
Hastelloy N," pp. 346366 in International Symposium on Structural Sta-
bility in Superalloys, Seven Springs, Pennsylvania, September 4—6, 1968,
Vol. IT. Available from Dr. John Radavich, AIME High-Temperature Alloys
Committee, Micromet Laboratories, West Lafayette, Indiana.

12¢, E. Sessions, Influence of Titanium on the High-Temperature
Deformation and Fracture Behavior of Some Nickel Based Alloys, ORNI~-4561
(July 1970).
.

»

143

divided MC type of carbide‘gives 8 better dispersion and would be
expected to give better properties. Thus, the qualitative observation
that the fine MC type of precipitates are associated with good proper-
ties and the coarse M,C type with poor porperties seems quite reasonable.
The further role of these prec1p1tates in inhibiting propagatlon of
cracks may also be important. ' - -

All of the mod1f1ed_alloy5'are slightly susceptible to aging during
long-term annealing. These changes are quite small and are associated
with precipitation of carbides. The general effects of aging are that
the minimum creep rate and fracture strain increase and the rupture life
decreases. Because of the difference in the type of carbide precipitated
at 650 and 760°C in many of these alloys, aging-at 760°C seems to pro-
duce larger changes in propefties. The precipitation of copious quanti-
ties of ZrC in alloys that contain less than 0.5% Zr is quite dramatic,

but the changes in properties are minimal.
SUMMARY

This first phase of this study of the effects of additions of Ti,
Zr, and Hf to a base alloy of Ni-12% Mo—7% Cr-0.2% Mn~0.05% C dealt with
small, laboratory melts and the first 100-1b commercial melts. The
results show that these three elements improve the creep strength and
fracture strain in unirradiated alloys. Tests on samples irradiated at
650°C or less show that the rupture lives at a given stress level and
the fracture strains are improved‘markedly. Irradiation at 760°C results

~in very poor proPertles_ The good properties are assoclated with the

formation of a very fine MCitype of carbide and the poor propertles with

a very coarse MoC type of carbide.
.ACKNOWLEDGME_NTS

' The author is grateful to several technlcians who a581sted 1n '

‘-runnlng tests and collectlng data' -H. W. Kline, J. Feltner, B. McNabb

N. O. Pleasant, and B. C. Wllll&nB The laboratory melts were melted
and fabricated by C. E. Dunn and J. N. Hix. The metallographic samples
 

 

 

144

were prepared and photographed by H. R. Tinch and E. Lee. The drawings
were prepared by the Graphic Arts Department and the manuscript by the
Metals and Ceramics Division Reports.Office. _ |

The author also thanks J. R. Weir, A. C. Schaffhauser, and
C. E. Sessions for their reviews of this report. The transmission
electron photomicrographs and the phase-identification results were
obtained by R. E. Gehlbach.
o,

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45.
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47.

49.
50.

51.

52.
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C.

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G. Asquith, Atomies International

F. Cope, RDT, SSR, AEC, Oak Ridge National Laboratory
B. Deering, Bla.ck and Vea.tch Post Office Box 8405 s

Kansas City, Mo. 64114

A.

H.

R. DeGrazia, AEC, Washington .
M. Dieckamp, Atomics International

David Elias, AEC, Washington

A.
J.

Giambusso, AEC, Washington
E. Fox, AEC, Washington

F. D. Haines, AEC, Washington

D.

G. Harman, Westinghouse Tampa Division, Nuclear Energy

Systems, 6001 S. Westshore Blvd., Tampa, Fla. 33616

C.
W.

E. Johnson, AEC, Washington
L. Kitterman, AEC » Washington

Kermit Laughon, AEC, OSR, Oak Ridge National Labora.tory

C.
T.
A.
J.

International Nickel Company, Inc., Huntington, W. Va.

D.
G.
D.
T.
D.
T.

L. Matthews, AEC, OSR, Oak Ridge National Laboratory
W. McIntosh, AEC, Washington

B. Martin, Atom:l.cs International

M. Martin, Huntington Alloy Products Da.v:.s:l.on, The

G. Mason, Atomics International

W. Meyers, Atomies Internationsal :
E. Reardon, AEC, Canoga Park Area Office
C. Reuther, AEC, Washington

R. Riley, AEC, Washington

K. Roche, Stellite Division, Cabot Corporation, 1020 W.

Park Ave., Kokomo, Ind. 46901

H.
M.
dJ.
T.

W.

M. Roth, AEC, Oak Ridge Operations
Shaw, AEC, Washington

M. Simmons, AEC, Washington

G. Schleiter, AEC, Washington

L. Smalley, AEC, Washlngton

Earl Q. Smith, Black and Veatch, Post Off:l.ce Box 8405,
Kansas City, Mo. 64114

S.
E.
D.
R.
A.
Ww.
M.
R.

R. Stamp, AEC, Canoga Park Area Office

E. Stansbury, The University of Tennessee

K. Stevens, AEC, Washington

F. Sweek, AEC, Washington

Taboada, AEC, Washington

M. Thomas, Allvac, Post Office Box 759, Monroe, N C.
J. Whitman, AEC, Washington

F. Wilson, Atomics International

25720

28110

Iaboratory and University Division, AEC, Oak Ridge Opera.t:l.ons
Division of Technical Information Extension