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U.S. ATOMIC ENERGY COMMISSION ORNL-TM-

306.1.

E INFLUENCE OF TITANIUM, ZIRCONIUM, AND HAFNIUM APplTlONS ON THE RESISTANCE OF MODIFIED HASTELLOY N TO IRRADlATlON DAMAGE AT HIGH TEMPERATURE - PHASE I

H. E . McCoy, Jr.

NOTICE This document contains informotion of o preliminary nature and was prepared primarily for internal use ot the Oak Ridge Notional Laboratory. It is subiect to revision or correction ond therefore does not represent a fino1 report.

- , -I

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 i t s use would not infringe privately owned rights.

ORNL-TM- 3064

Contract No. W-7405-eng-26 METALS AND CEXAMICS D I V I S I O N

INFLUE3CE OF TITANIUM, ZIRCONIUM, AND HAFNIUM ADDITIONS ON THE RESISTANCE OF MODIFIED HASTELLOY N TO IRRADIATION DAMAGE AT HIGH TEMPERATURF: - PHASE I H. E. McCoy, Jr.

JAN1 ARY 1 9 7 1

t E G A L 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, completeness or usefulness of any Information, apparatus, product or process disclosed, or represents that its use would not infringe privately owned rights.

OAK RIDGE NATIONAL LABORATORY Oak Ridge, T e n n e s s e e operated by UNION CARBIDE CORPORATION f o r the U.S. ATOMIC ENERGY COMMISSION

U a

iii

comms Page n

............................. ........................... Experimental Details . . . . . . . . . . . . . . . . . . . . . . . Test Materials . . . . . . . . . . . . . . . . . . . . . . . . I r r a d i a t i o n Conditions . . . . . . . . . . . . . . . . . . . . Testing Procedure . . . . . . . . . . . . . . . . . . . . . . Experimental Results . . . . . . . . . . . . . . . . . . . . . . . Alloys Containing Titanium . . . . . . . . . . . . . . . . . . Alloys Containing Zirconium . . . . . . . . . . . . . . . . . Alloys Containing Hafnium . . . . . . . . . . . . . . . . . .

100

................ ......................

124 I35

Abstract Introduction

Alloys Containing No Additions Discussion of Results

............................. .........................

Summary Acknowledgments

1 1

2 4 5 5 5 50

143 143

INFLUENCE OF TITANIUM, ZIRCONIUM, AND HAFNIUM ADDITIONS ON "HE RESISTANCE OF MODIFIED HGTELLOY N TO IRRADIATION DAMAGE AT HIGH TEMPERATURE - PHASE I H. E. McCoy, Jr.

ABSTRACT

The influence of small additions of T i , Z r , and H f on the mechanical properties of a modified Hastelloy N with t h e nominal composition Ni-12% M0-'7$ 0 4 . 2 % Mn4.05$ C i s described i n t h i s report. It deals s p e c i f i c a l l y with t e s t r e s u l t s from numerous, small, laboratory melts and several 100-lb melts from commercial vendors. Additions of T i , Zr, and Hf had b e n e f i c i a l e f f e c t s on the properties of t h e a l l o y both unirradiated and a f t e r irradiation. I r r a d i a t i o n tempera t u r e had a marked e f f e c t upon the properties of a l l alloys investigated. Generally, good properties were observed when t h e i r r a d i a t i o n temperature was 650째C or l e s s and poor when t h e temperature was 700째C or higher. We a t t r i b u t e d t h i s large e f f e c t of i r r a d i a t i o n temperature t o coarsening of the carbide structure a t t h e higher temperature.

INTRODUCTION

Previous studies'-4

showed Hastelloy N (Ni-16% M0-7$ C r - 4 $ Fe-O.OSq& C )

susceptible t o a type of damage produced by i r r a d i a t i o n a t high temperatures that results in reduced stress-rupture properties and fracture ductility.

One approach t o solving t h i s problem i s t h a t of making s l i g h t

changes i n the chemistry o

he alloy.

composition t o Ni-12$

14.2% Mn4.05$ C t o obtain an a l l o y t h a t

Mo-'7

We i n i t i a l l y modified t h e base

IH. E. McCoy and J. R. r, "Stress-Rupture Properties of b r a diated and Unirradiated Hastelloy N Tubes," Nucl. Appl. 4(2), 96-104 (1968).

2H. E. McCoy, "Variation of Mechanical Properties of Irradiated Hastelloy N with S t r a i n Rate," J. Nucl. Mater. 3l( - l), 67-85 (1969). *t

6.1

3H. E. Hastelloy N 4H. E. Hastelloy N

McCoy, A n Evaluation of t h e Molten-Salt Reactor Experiment Surveillance Specimens F i r s t Group, ORNL-TM-1997 (1967). McCoy, A n maluation of t h e Molten-Salt Reactor Experiment Surveillance Specimens - Second Group, ORNLTM-2359 (1969).

was free of t h e s t r i n g e r s of t h e MgC type of carbide (where M i s t h e metallic component) c h a r a c t e r i s t i c a l l y found i n standard Hastelloy N.

We then made a number of laboratory melts with various additions of T i , 22, and H f , since they form very strong, stable diborides* and should

t i e up t h e boron as compounds r a t h e r than allow it t o segregate t o t h e grain boundaries. The more random d i s t r i b u t i o n of t h e boron should reduce t h e radiation damage by reducing t h e concentration a t grain boundaries of helium produced by t h e 1°B(n,a) reaction. The r e s u l t s of t e n s i l e and creep t e s t s on these experimental alloys show t h a t t h e resistance t o i r r a d i a t i o n damage i s improved by t h e addit i o n of T i , Zr, or Hf.

Results of tests on t h e first small (100-lb) heats

of commercial alloys that contained these same a l l o y additions a l s o indi-

cate improved properties. EXPERIMENTAL DETAILS

T e s t mterials b

Our alloys w e r e nonconsumably arc melted i n an argon atmosphere

f r o m melting stock of commercialpurity.

A s t a r t i n g charge of about

2 l b was consolidated and melted several times.

The charge was then

placed i n another arc-melting furnace with a hearth f o r drop casting. The a l l o y was again melted and drop cast i n t o a 1-in.-dim X 6-in.-long ingot. ule:

The ingot was swaged t o 1/4-in.-diam

rod by t h e following sched-

from 1 in. t o 3/4 i n . a t ll77"C, from 3/4 i n . t o 7/16 in. a t 8 7 l o C ,

and from 7/16 in. t o 1/4 in. a t ambient temperature. The commercial melts were melted by vacuum induction and c a s t i n t o s i x 4-in.-diam

ingots t h a t weighed about 16 l b each.

They were i n i t i a l l y

forged a t 1177°C; t h e f i n a l working was done a t room temperature.

The

f i n a l product was small 1/2-in.-thick p l a t e s with 40% cold work.

The chemical analysis of each a l l o y used i n t h i s study i s given i n Table 1. ?

5B. Post, "Refractory Binary Borides," p. 340 i n €!oron, MetalloBoron Campounds and Boranes, ed. by R. M. Adams, Interscience, New York, 1964.

Table 1. Chemical Compositions of Alloys Alloy Number

Composition Mo

100 101 102 103 104 105 106 107 75 76 21545 108 109 110 111 112 146

12.16 12.16 12.10 12.37 11.98 11.10 11.82 11.95

u.7

12.0 12.4 12.4 12.2 12.4 12.4 12.5 12.9 12.5 12.7 12.3

21554 21555 152 153 154

155 156 67-503 67-504 21546

11.8 11.8 12.0 11.9 12.5 12.6

12.7 11.3 12.4

6.96 6.97 7.03 6.98 6.97 6.84 7.02 7.00 7.94 7.88 7.18

0.20 0.20 0.20 0.20 0.20 0.20 0.20 0.20 0.20 0.20 0.29 0.21 0.21 0.19 0.21 0.23 0.23 0.2 0.16 0.16 0.21 0.16 0.14

0.071 0.056 0.089 0.077 0.10 0.089 0.061 0.082 <0.05 <0.05 0.034 0.065 0.081 0. 12 0.069 0.069

7.2 7.2 6.7 6.6 7.5 6.7 6.7 7.4 7.2 6.2 7.0 6.9 6.9 7.0 7.8 7.5 7.2

0.12 0.12 0.08 0.12 0.16

0.1 0.097 0.065 0.03 0.02 0.01 0.03 0.03 0.15 0.12 0.05

(wt Si

(PPd B

< 0.005 < 0.005 < 0.005 < 0.005 < 0.005 < 0.005 < 0.005 < 0.005 < 0.05 < 0.05

0.052

< 0.01

< 0.03 < 0.03 < 0.03 < 0.03 < 0.03 < 0.03 < 0.03 < 0.03

0.02 0.02 0.02 0.02 0.02 0.01 0.01 0.01

0.015

0.050

0.01 0.10 0.28 0.53 0.75

0.7 0.2 0.01 0.01 0.02 0.01 0.4

< 0.005 < 0.005 < 0.005 < 0.005 < 0.005

< < <

0.005 0.01 0.01 0.008 0.03 0.019 0.01 0.01 0.01 0.01 0.02 0.009

0.057 0.045 0.055 0.050

0.049 0.050

0.056 0.062 0.117 0.056 0.060 0.065 0.065 0.065 0.060 0.069 0.065 0.052

0.054 0.062 0.058

0.052 0.056 0.06 0.07 0.05

0.11 0.39 0.31 0.55 0.66 0.81 1.04 0.99 1.01 0.49 0.001 0.01 0.01

< < < < 0.01 < 0.01

0.01 0.01 0.003 0.003 < 0.01 < 0.01 < 0.01 < 0.01 < 0.01 0.06 < 0.01 0.1

Other

1.18

< < <

0.90 0.94 0.35 0.05 0.02 0.01 0.01 0.01 0.03 0.01 0.01 0.005

Re=O.02 Re=O.3 Hf =O .16 Hf=O.06 H f =O .lo Hf=O.I 3 Hf=O .43 Hf=O. 08 Hf=0.50

2 2 0.2

0.2 1

0.3 2

Weight of Melt (lb) 2 2

2 2 2 2 2 2 2 2 100 2 2 2 2 2 2 2 100 100 2 2 2 2 2 100 100 100

4 The rods were fabricated i n t o t h e t e s t specimens sham I n Fig. 1. Our work with t h i s specimen has shown t h a t t h e rest r e s u l t s a r e reproduc-

i b l e and agree q u i t e w e l l with those obtained f o r more massive specimens. However, there i s s u f f i c i e n t s t r e s s concentration t o cause some of t h e more b r i t t l e specimens t o break i n t h e radius a t the end of t h e gage length.

ORNL-OWG 67-30131

L 1 '/e in. _ I Fig. 1. Specimen Used f o r Tests of Mechanical Properties.

I r r a d i a t i o n Conditions The results presented here were obtained from specimens irral iated i n several experiments. reactors:

These experiments were carried out i n three

t h e Engineering Test Reactor (ETR) a t Idaho Falls, Idaho,

and t h e Oak R i d g e Research Reactor (ORR) and Molten S a l t Reactor Experiment (MSRE) a t Oak Ridge, Tennessee.

A core f a c i l i t y was used i n t h e

ETR where the thermal and fast (> 1MeV) fluxes were each 3.2 X

neutrons cm-2 sec'l

and t h e fluence was 6

lo2'

neutrons/cm2.

loi4 The ETR

experiments were uninstrumented, and t h e design temperatures were e i t h e r below 150째C o r 600

100째C.

Melt wires included i n these experiments

indicated t h a t t h e operating temperatures were within t h e desired range. The experiments i n the ORR were performed i n a poolside f a c i l i t y where t h e peak fluxes were 6 X 1013 neutrons cm'* 5 X

neutrons

sec'l

(> 2.9 MeV).

sec'l

(thermal) and

The experiments were run f o r

W .

e i t h e r one or two cycles s o t h a t t h e thermal fluence was 2 t o 5 X neutrons/cm2 and the fast fluence was 2 t o 4 x

lo1'

lo2'

neutrons/cm2.

These

experiments were instrumented, and t h e temperatures were controlled a t i

650 t o 871째C. The experiments i n t h e MSRE were run i n t h e center of t h e

core, where the thermal f l u x was 4.1 x 1OI2 neutrons cm-2 sec'l

and t h e

fast (> 1.22 MeV) f l u x was 1.0 X 10l2 neutrons sec'l. The samples were exposed t o a noncorrosive fluoride s a l t environment f o r 4-800 h r a t 645

1 0 째 C and received a thermal fluence of 1.3 x lo2'

a fast fluence of 3.1

neutrons/cm2 and

lo1' neutrons/cm2. Since we could not see any consistent e f f e c t s of thermal-neutron fluence over t h e small range X

encountered i n these experiments, we present our data i n t h i s paper as having been obtained a t a common fluence. Testing Procedure The creep-rupture t e s t s a f t e r i r r a d i a t i o n were run i n lever-arm creep machines i n t h e hot c e l l s a t t h e Oak Ridge National Laboratory. The s t r a i n was measured by an extensometer with rods attached t o t h e upper and lower specimen grips.

The r e l a t i v e movement of these two rods

was measured by a l i n e a r d i f f e r e n t i a l transformer, and the transformer

s i g n a l was recorded.

The accuracy of t h e s t r a i n measurements i s about

50.15, considering the e f f e c t s of ambient temperature variations i n t h e

hot c e l l , mechanical vibrations, e t c .

This accuracy i s considerably

better than t h e specimen-to-specimen reproducibility t h a t one would

expect f o r t h i s alloy.

The systems f o r measuring and controlling temper-

a t u r e combine t o give a temperature uncertainty of about 51$. The t e n s i l e t e s t s were run on Instron Universal Testing Machines. The s t r a i n measurements were taken from t h e crosshead t r a v e l .

EXPERIMENTAL RESULTS

f I

taining Titanium The alloys with additions of titanium f e l l i n t o three groups.

The

f i r s t group, alloys 100 through 107, contained from l e s s than 0 . 0 1 t o

1.04s T i . These alloys were of t h e nominal base camposition of Ni-l2$ Mo7s Cr-Q.2$ C. Typical microstructures of these alloys a f t e r a 1-hr anneal a t 1177°C are shown i n Fig. 2. The a l l o y s were a l l f r e e of

intermetallic p r e c i p i t a t e s and contained only a small amount of carbide p r e c i p i t a t e s . Generally, t h e amount of carbide increased and t h e grain s i z e decreased with increasing titanium concentration. group, alloys 75 and 76, contained

1s T i and were made t o explore t h e

influence of varying carbon content. alloys are shown i n Fig. 3.

The second

Typical microstructures of these

The t h i r d material, a 100-lb commercial

a l l o y from Special Metals Corporation, designated heat 21545, contained Typical microstructures of this a l l o y a r e shown i n Fig. 4 .

0.49s T i .

This a l l o y had a n m k a l composition q u i t e close t o t h a t of a l l o y 104,

and t h e microstructures of t h e two alloys after a 1-hr anneal a t 1177°C were quite similar except t h a t t h e commercial a l l o y contained patches of p r e c i p i t a t e . Close examination showed these regions t o be quite high i n molybdenum.

Some of

the p r e c i p i t a t e s w e r e carbides of the MC type and

others w e r e almost pure molybdenum.

The grain s i z e a f t e r annealing f o r 1

100 h r a t 871°C was q u i t e fine, and copious quantities of carbide prec i p i t a t e s were present. Alloys 100 through 107 were tested under a v a r i e t y of conditions. Tensile tests were run a t 650°C on samples t h a t had been annealed 1 h r

a t 1177°C and aged 1000 h r a t 650°C t o duplicate t h e thermal h i s t o r y of irradiated samples.

The r e s u l t s of these t e s t s a r e summarized i n

Table 2. The variation of t h e t e n s i l e and y i e l d strengths with titanium content i s sham i n Fig. 5. titanium.

The y i e l d strength increased with increasing

The t e n s i l e strength varied i n a poorly defined manner:

a s t r a i n rate of 0.05 min-l, it increased systematically with increasing titanium, and a t a rate of 0.002 m i n - l it reached a mximum a t 0.55% T i . The variation of t h e f r a c t u r e s t r a i n with titanium content i s shown i n Fig. 6.

A t a s t r a i n r a t e of 0.05 min’l,

t h e addition of O.ll$ T i g r e a t l y

increased t h e f r a c t u r e s t r a i n ; f u r t h e r additions decreased t h e f r a c t u r e strain.

A t a s t r a i n r a t e of 0.002 mino’, t h e f r a c t u r e s t r a i n improved

g r e a t l y with t h e addition of O.ll$ T i , and f u r t h e r additions caused a gradual improvement.

c-d

Y-72230

...

I_c

.. c

Fig. 2. Photomicrographs of Alloys That Contain Various Amounts of Titanium. The basic composition is Ni-l2$ M0-7$ Cr-O.2$ W . 0 5 $ C. Etchant: glyceria regia. lOOX. Reduced 38%.

Fig. 3. Photomicrographs of Alloys (a) 75 and (b) 76. The nominal chemical compositions are Ni-124 M0-7$ C r 4 . 2 5 Mn-1% T i - O . O S $ C and T i - O . l O $ C, respectively. Etchant: Ni-12$ Mo-7$ C1-0.24 -1.05 glyceria regia. 1OOx.

10 Table 2.

Alloy Nrrmber

Tensile Propertiek of Several Unirradiateda Titanium-Modified Alloys a t 650째C

Specimen Number

Strain Rat e (min-l)

Stress, psi Ultimste Yield Tensile

ation

Elo

Reduction in &ea

(%I

100 100

1906 1910

0.05 0.002

101 101

1933 1926

0.05 0.002

x 103 x 103 27.1 77.4 30 73.7 27.9 82.6 27.4 75.2

102 102

1943 1952

0.05 0.002

31.1 30.9

87.1 78.4

51.2 31.4

53.2 37.5

42.7 32.8

103 103

1962 1968

0.05 0.002

31.5 28.9

87.7 69.4

51.4 24.5

54.0 37.5

47.3 33.8

104 104

1987 1984

0.05

31.1 35.4

87.5 87.9

51.0 28.2

53.0 36.7

42.4 28.8

105 105 106 106 107 107

2007 2OW 2023 2026

34.2 34.3 34.8 33.9 36.4 33.7

93.6 84.1 100.7 79.8

51.0 31.9

52.8 40.3 47.3 37.8 46.1 38.5

45.0 32.4

0.002 0.05 0.002 0.05 0.002

2039

0.05

2a1

0.002

99.2

75.1

43.6 27.8 62.1 33.0

44.2 29.7 64.8 37.3

34.2 25.6 44.2 29.2

45.4 26.0 43.8 22.3

40.3 35.1 36.4 32.5

a Annealed 1hr a t U77"C and aged lo00 hr a t 65OOC.

0.2

0.4 46 0.0 4.0 TITANIUM co(TENT I r t W

t.2

4.4

Fig. 5 . Influence of Titanium Content on t h e Yield and Ultimate Tensile Strengths a t 650째C a f t e r Annealing 1 h r a t 1177째C and Aging 1000 hr a t 650째C.

. i

11 I

ORNL-DWG 69-14397R

65 0

STRAIN RATE=O.WZmlh’

ANNEALED Ihr AT 1177.C AGE0 ( O W hr AT 650%

-E 50 z 0

z e 45 3 Y

25 0.2

0.4

0.6

0.8 1.)

1.0

1.2

TITANIUM CONTENT ( w t

Fig. 6. Influence of Titanium Content on the Tensile Fracture S t r a i n a t 650°C of Ni-l2$ M0-7$ Cr-O.2$ Mn-0.05$ C Alloys. Alloys 100 through 107 were i r r a d i a t e d a t 650°C t o a thermal fluence of 2.5 x 1020 neutrons/cm2.

The r e s u l t s of t e n s i l e t e s t s a t 650°C a f t e r

i r r a d i a t i o n a r e summarized i n Table 3 .

The fracture s t r a i n s measured a t

650°C a r e p l o t t e d as a function of titanium content i n Fig. 7. The data exhibit considerable variation, but show no systematic variation w i t h titanium concentration.

The r e s u l t s of these t e s t s a r e analyzed f u r t h e r

i n Fig. 8, where t h e r a t i o s of t h e various t e n s i l e properties of t h e irradiated and unirradiated alloys are compared.

The yield s t r e s s i s increased

by i r r a d i a t i o n , and the t e n s i l e s t r e s s and f r a c t u r e s t r a i n a r e decreased. However, t h e r a t i o s are only moderately dependent upon t h e titanium level. Another s e t of samples of alloys 100 through 107 was i r r a d i a t e d a t

50°C t o a thermal fluence of 9 s i l e t e s t s of these sampl parison of these r e s u l t s ’.

loi9 neutrons/cm2.

The r e s u l t s of ten-

0 and 760°C a r e given i n Table 4. Com-

s e i n Table 3 f o r t h e samples i r r a d i a t e d

t o a higher fluence a t a higher temperature reveals that t h e f r a c t u r e s t r a i n s a r e higher f o r t h e samples i r r a d i a t e d a t 50°C.

The fracture

Table 3.

Tensile Properties of Several Titanium-Modified Alloys Tested at 650°C after Irradiationa

Alloy

Specimen

Number

Strain (min..l) Rate ~

Stress, psi Yield ultimate Tensile ~

Elongation $ uniform Toial

Reduction in Area

($1

x 103

100 100

1908 1912

0.05

31.5 26.3

66.3 48.8

26.3

0.002

13.4

26.6 l4.1

23.3 13.1

101

1924

101

1925

0.05 0.002

35.5 36.9

68.9 57.2

23.2 10.7

23.7 10.9

19.1 12.5

102 102

1949 1951

0.05 0.002

35.1 37.7

74.8 67.8

33.7 18.5

35.1 19.2

28.8

103

1978

0.002

35.3

65.9

21.9

22.8

18.8

104

2006

0.002

39.0

65.2

u.5

15.4

15.6 31.6

14.9

105

1989

0.002

35.5

69.7

21.9

22.7

106

2021

0.002

39.2

65.1

12.0

12.6

14.4

107

2051

0.002

39.3

71.6

35.9

16.3

13.0

aAnnealed 1 hr at ll77"C before irradiation at 650°C to a thermal fluence of 2.5 x lo2' neutrons/cm2.

MMIL-DWG 69-14396R

STRAIN RAl€-O.O02min-' ANNEALEDfhrATt177.C

IRRAMATION CONMTIONS TEMPERATURE 65YC TIME1loo0 hr THERMAL FLUENCE= 2 . 5 1 1 9 rmtrons/crn2

TITANIUM CONTENT ( w t Ye)

Fig. 7. Influence of T i t a n i u m Content on t h e Tensile Fracture Crc-O.05$ C Alloys after Irradiation. S t r a i n a t 650°C of N i - I 2 4 M-7%

LJ t

ORNL-DWG 69-1439517

1.4

1.2

1.0

0.4

0.2

I 0

A ULTIMATE STRESS 0

0.4

0.2

FRACTURE STRAIN 0.6

0.8

1.0

1.2

TITANIUM CONTENT ( w t % )

Fig. 8 . Ratios of €he Unirradiated and Irradiated Tensile Properties of Alloys Modified with Titanium. Tested at 650°C at a Strain Rate of 0.002 min-1.

Table 4 .

Tensile Propertiesa of Several Titanium-Modified

Alloys Tested at 650 and 760°C after Irradiationb ALLOY

Number

Specimen Number

Stress, psi UTensile lt~te x io3

100 101 102 103 106 105 106 107

1907 1941 1944 1975 1992 2003 2025 2W3

100 101 102 103 104 105 106 1W

1905 1940 1948 1977 1995 2016 2024 2056

in Area Elongation, $ Reduction Uniform Total (5)

x 103

Test Temperature 650°C 26.4 60.3 27.6 62.0 31.3 71.4 30.7 74.9 29.8 71.8 34.4 76.1 33.4 80.8 39.6 89.3 Test Temperature 760°C 48.8 27.4 51.0 27.6 53.9 30.6 53.8 26.8 54.4 30.0 56.7 31.8 57.0 30.6 31.5 58.9

21.3 21.6 22.8 23.8 25.1 23.5 26.0 23.4

21.9 22.1 23.4 24.4 25.7 23.8 26.4 23.8

24.0 18.3 19.4 27.2 33.2 22.8 34.5 21.1

8.4 8.1 8.6 9.3 8.9 8.8 9.4 9.3

9.5 9.2 8.9 10.3 10.0 11.2 12.8 13.2

9.3 5.6 9.4 10.4 8.1 8.6 11.6 12.1

'Strain rate = 0.002 min-1. bAnnealed 1 hr at ll77"C before irradiation at 50°C to a thermal fluence of 9 x 1 0 1 ~neutronslcm'.

s t r a i n s measured a t 650°C a f t e r i r r a d i a t i o n ranged from 22 t o 26% but e

However, a t a t e s t

did not vary systematically with titanium content.

temperature of 760"C, t h e f r a c t u r e s t r a i n did show a s l i g h t dependence on titanium level; it varied from 9.54 with no titanium present t o U.24 with 1.04% T i present. Several unirradiated samples of alloys 100 through 107 were creep t e s t e d at 650°C. The results of these t e s t s are summarized i n Table 5 and p l o t t e d i n Figs. 9 through ll. The r e s u l t s of stress-rupture t e s t s i n Fig. 9 show a marked improvement i n rupture l i f e a t a given s t r e s s l e v e l with increasing titanium content.

The results are somewhat

scattered, and there are some questions about t h e exact location of t h e l i n e f o r each heat. The measurements of minimum creep rate i n Fig. 10 show t h a t titanium decreased t h e creep r a t e .

A l l of these tests were

run without an extensometer, and t h e precision of t h e measurements was not very good. However, the data indicate that t h e creep rate did not vary appreciably with titanium content from 0.11 t o 1.04%T i (alloys 101 and 107).

The data a l s o indicate t h a t titanium concentration may have

no influence on t h e creep r a t e a t low stress levels.

The f r a c t u r e s t r a i n s

are sham i n Fig. 11as a function of the minimum creep rate. additions d e f i n i t e l y improve t h e f r a c t u r e s t r a i n .

Titanium

The f r a c t u r e s t r a i n s

of alloys 106 and 107, which contained 0.81 and 1.04%T i , respectively,

were almost independent of creep rate; t h e other a l l o y s exhibited decreasing f r a c t u r e s t r a i n with decreasing creep rate. Several of these alloys were heat t r e a t e d t o simulate irradiated samples by anneals of 1hr a t 1177°C and 1000 hr a t 650°C. The results of tests on these samples are summarized i n Table 6. A comparison of these results with those i n Figs. 9 through 11f o r samples t h a t were annealed 1 h r at 1177°C indicates haw these a l l o y s respond t o aging.

The

rupture l i f e , minimum creep rate, and f r a c t u r e s t r a i n were a l l increased by t h i s aging treatment. Alloys 100 through 107 were i r r a d i a t e d a t 650°C and creep tested

a t t h e same temperature after i r r a d i a t i o n . The results of these tests are given i n Table 7. A l l of these samples w e r e t e s t e d a t a s t r e s s l e v e l of 32,400 p s i . The rupture l i v e s and f r a c t u r e s t r a i n s are shown as funct i o n s of titanium content i n Fig. 12. The rupture l i f e was increased I

15 Table 5.

Alloy Number

S t r e s s Rupture Life (psi) (hr)

Minimum Creep R a te ($/hr)

Elongation

Reduction in Area

($1

(4)

100 100 100 100 100

1915 1917 1909 19l4 1921

5725 6099

5554

6108 6172

32.4 27

101 101 101 101 101

1930 1937 1929 1934 1932

5753 5692 6203 6203 6104

70 55

102 102 102 102

1958 195 0 1961 1956

5751 5726 6100 5669

103 103 103 103

1979 1976 1966 1980

5747

5687

55 47

157.4

541.5

0.45 0.029 0.010 0.0082

104 104 104 104

1998 1991 1982 1990

6101 5566 5959 5638

3.5 41.6 2718.9 1001.2

0.88 0.060 0.0033 0.0066

32.3 18.8 13.9

20.8 14.0 21.4 19.9

105

2018 2001 2017 2000

5754

i.7

18.5

94.0 179.7

2031 2029 2022 2032 2034

15.9 15.5 16.6 25.0 21.3 21.9 22.1 21.4

13.7

106 106 106 106 106

107 107

2054 2046 2055 2040

5698 6089 5586 5750 5691 5658 5988 6107 5760

0.25 0.019

23.4

105 105 105

Specimen Test Number Number

x 103 55

107 107 -

Creep-Rupture Properties of Unirradiated" TitaniumModified Alloys a t 650째C

6087 5663

47 40 32.4 70

55 36.5 40

3.4 20.8

47.4 185.1

827.8 0.6

23.4 52.1 292.1

792.8 3.3 32.4 162.4 506.5

3.2 69.9

476.6 11.6

95.4

47 40 70

5644 6106

5633

158.0 372.9 832.4 5.5 322.6 375.6 1490.5

20.8 12.5 9.5 6.2 4.0

0.43 0.55 0.088 0.0077 0.0022

22.5 14.5 12.5 6.3 4.9

2.9 0.064 0.050 0.0060 0.0037

34.5

28.3

21.0 16.0 14.1 10.5

21.3 14.0 13.2

0.56 0.047 0.016 0.0076

28.5 20.0 15.8 13.7 28.3 18 ..5 14.0 15.6

0.015 0.0078 0.20 0.042 0.038 0.015 0.014 1.22 0.028 0.022 0.010

Annealed 1 h r a t 1177째C before t e s t i n g .

22.5 20.3 23.1 26.5

12.8 22.8 15.4 16.9 14.7 22.8 13.9 11.6

18.4

19.6 13.8 19.9 20.0 17.6

18.5 21.2 19.1 20.3 23.1

27.3

ORNL-DWG 69- I4394

(xi031

V 3

--8

W v)

100

eo.01

102

0.39

n 100

10'

io4

102

RUPTURE TIME (hr)

Fig. 9. Stress-Rupture Properties at 650째C of Several Alloys Containing Titanium. A l l samples annealed 1 h r a t 1177째C before t e s t i n g .

ORNL-DWG

69 -14393

0 100

0.ot

At01

0.11

to2

rn 103 o 104

(05 A 406 + 107

io-'

0.39 0.31 0.55 0.66 0.8t t.04

COO

MINIMUM CREEP RATE (%/hr)

Fig. 10. Creep Properties a t 650째C of Several Alloys Containing Titanium. A l l samples annealed 1hr a t 1177째C before t e s t i n g .

17 ORNL-DWG 69-14392

a 20

ti W

a a

IL 4 0

-40-3

lo-'

too

40'

MINIMUM CREEP RATE (%/hr)

Fig. 11. Fracture S t r a i n s a t 650°C of Several Alloys Containing Titanium. A l l samples annealed a t 1177°C before testing;

Table 6. Creep Properties of Several Ageda Titanium-Modified Alloys a t 650°C

Alloy Number

Specimen Test Number Number

Stress (psi)

Rupture Life

Minimum Creep Elorgation Rate ($1

Reduction in Area

($1

(hr)

(&/hr)

5.9

0.61 0.0007 0.12 0.021 O.MO 0.0051 0.21 0.0046 0.095 0.0054 0.119 0.0033 0.14 0.0039

23.0 1.0

19.2 0

24.5 19.2

21.9

37.5 10.3 36.4 6.2 36.0 5.8 38.0 5.9

29.7 7.2 25.6 3.3 26.9 3.3 32.4 4.5 35.0 4.8

0.027

42.2

0.004

6.0

x io3

bpi

2035 2027

573% 5779

2052 2047

5654b 5784

1922 1919

101 101

1936

102 102 103 103 104 104 105 105 106 106

107 LUi

1935 1960 1959 1973 1972 1999 1994 2013 2011

5723b 5630 5724,, 5783 570% 5780 572gb 5757 5730b 5758 5728b 5785

100 100

'Samples

55 17.8

1007 48.6 55 32.4 1250.7 88.0 55 32.4 1659.1 54.0 55 32.4 1006.9 55 101.2 32.4 1006.9 55 66.6 32.4 1704.5 107.5 55 32.4 1655.8 42.7 902.2 52.4 1704.5

45.0 7.5

annealed 1 hr a t 1177°C and aged 1000 hr at 650°C.

bTest discontinued before f a i l u r e .

12.5

35.0 4.8

Table 7.

Alloy Number

Creep Properties of Irradiateda Allbys That Contain Various Amounts of Titanium

Specimen Number

Test Number

Titanium Content (.wt $1

Rupture Life (hr1

Minimum Creep

Elongation

( 4)

Reduction in Area

($1

True Fracture Strain

($1

100

1916

R- 146

< 0.01

105.4

0.0099

1.1

5.70

5.90

101

1928

R- U.7

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- U.8

0.39

1327.3

0.0034

5.8

4.63

4.70

104

2005

R- U.9

0.55

U6.3

0.0048

8.7

5.86

6.00

105

1983

R-151

0.66

L540.3

0.0046

9.2

107

2042

R- 153

1.04

2928.8

0.0026

12.8

I r r a d i a t e d a t 650째C t o a thermal fluence of 2.5 32,400 psi.

lo2'

neutrons/cm*.

Tested a t 650째C and

bDia&eter not measured a f t e r t e s t .

4600 1400

1200

2 4000 2

800 r UNIRRADIATED, . 5

STD HASTELLOY N

600

I I 1.5'2

IRRADIATION TEMPERATURE=65OTTHERMAL FLUENCE = 2.5 X !Om neutrons/cm* TEST TEMPERATURE=656"C STRESS=32,350psi

400

2oo

-1.2

-1.1 n

-0

0.4

0.2

0.3 0.4 0.5 0.6 0.7 TITANIUM CONTENT ( w t %)

0.8

0.9

4.0

Fig. 12. Influence of Titanium on t h e Creep Properties of Ni-12% McA"7 CI-0.058 C a f t e r Irradiation. Tested a t 650째C a t a s t r e s s l e v e l of 32,400 p s i . by a f a c t o r of 25 and t h e f r a c t u r e s t r a i n by a f a c t o r of 10 by t h e addit i o n of 1.04% T i .

These r e s u l t s are compared w i t h those of the unirra-

diated control samples i n Table 8.

This comparison i s fl-aught with t h e

uncertainties footnoted i n Table 8, but it indicates t h a t i r r a d i a t e d samples containing 0.55% T i o r more have rupture l i v e s t h a t a r e about 0.7 times the unirradiated value, minimum creep rates t h a t are about equal t o that of t h e unirradiated material, and f r a c t u r e s t r a i n s t h a t are about 0.7 times t h e

Alloys 100 through fluence of 9 x

were a l s o i r r a d i a t e d a t 5 0 째 C t o a thermal

W 9neutrons

of these t e s t s a r e given i n

and creep-rupture t e s t e d . e 9.

These samples were t

f e r e n t s t r e s s levels than those used i n the experiment described previously (Table 7).

However, a rough comparison shows t h a t t h e d i f f e r e n t

i r r a d i a t i o n temperatures (650 vs 50째C) and t h e d i f f e r e n t fluences (25 vs 9 x lo1' neutrons/cm2) had l i t t l e e f f e c t on t h e t e s t r e s u l t s .

Both

experiments show c l e a r l y t h e increasing rupture l i f e and f r a c t u r e s t r a i n w i t h increasing titanium level.

The combined results of the t e n s i l e and creep t e s t s allow a determination of t h e variation of t h e f r a c t u r e s t r a i n with s t r a i n r a t e a t 650째C.

Table 8,

Comparison of the Creep Properties a t 32,400 p s i and 650째C of Unirradiated and Irradiated Titanium-Modified Alloys

Unirradiat ed Content

(5)

< 0.01

Creep

Rupture Life

Irradiated

Fracture Strain

($1

185.1a

Rupture Life

Creep

Fracture Strain

($1

Life

Creep Rate

Strain

6.3'

105.4

0.0099

1.1

0.57

1.3

0.17

0.021

19.2

169.9

0.0084

1.6

0. u

0.40

0.08

0.0046

12.5'jb

570.4

0.0077

5.2

0.30

1.7

0.42

0.70 0.76 0.81 0.75

0 62

0.47

0 89

0.70

14 0.65

0.74

0.0077a

0.11

1250.7

0.31

1900b

0.39

190OaYb 0.0051

12.5~"~

1327.3

0.00%

5.8

0.55 0.66

1900aJb

0.0054

12,5'jb

U6.3

0.0048

8.7

1900ajb

0.0033

l2.5'jb

1540.3

0.0046

9.2

1.04

3900ajb

O.O(u.0

2laYb

2928.8

0.0026

12.8

Ratio ( Irradiated: Unirradiated) Rupture Fracture

lu 0

0.61

a Values from waged samples. bExtrapolated.

Table 9. Creep Properties of Irradiated

Alloy Number

100

101

103 102 lor, 105 106 107

Specimen Number

1911 1927

1974 1957 1981

2008 2030 2044

Test Number

R-257 R-2 R-286 R-282 R-287 R-290

R-288 R-289

Titanium Content (wt

< 0.01 0.11 0.31

0.39 0.55 0.66 0.81 1.04

Alloys That Contain Various Amounts of Titanium

Test Temperature (OC)

650 650 650 650 650 650 650 650

Stress

RLife ~ture

(psi)

(hr)

x 103 40 40 40 40

47 47 47 47

Minimum Creep , Rate ( $/hr )

68.9

0.019

226.1 114.0 255.4 114.8 200.9 318.9 374.0

0.0037 0.033

0.010 0.025

0.0078 0.0072 0.0073

a Annealed 1 hr a t 1177째C before irradiation a t 50째C t o a thermal fluence of 9 x

lo1'

Elongat ion

(4)

1.79 1.55 3.07 5.04 5.25 3.89 8.25 7.55 neutrons/cm2.

1u I-'

22 "he data f o r samples i r r a d i a t e d a t 650°C t o a fluence of 2.5

lo2'

neutrons/cm2 were used t o construct Fig. 13. The s t r a i n rate i s a cont r o l l e d parameter i n a t e n s i l e t e s t and was used d i r e c t l y i n t h i s figure. The s t r e s s i s controlled i n a creep test, and t h e r e s u l t i n g minimmi creep rate was used i n Fig. 13.

The l i n e s on t h i s graph simply j o i n t h e

data points, and t h e i r slopes have l i t t l e significance.

The most impor-

t a n t observation i s t h a t t h e addition of titanium causes a greater improvement i n t h e f r a c t u r e s t r a i n a t low s t r a i n r a t e s than a t high

strain r a t e s .

ORNL-DWG 6944391

i3 25 5

$ I

$ 4s 0

LL K

0 40-3

40-'

400

to'

STRAIN RATE (X/hr)

Fig. 13. Variation of' Fracture S t r a i n w i t h S t r a i n Rate a t 650°C f o r Several Titanium-Bearing A l l 0 s Irradiated a t 650°C t o a T h e m 1 Fluence of 2.5 x lo2' neutrons/cm

The results presented thus far f o r alloys 100 through 107 were obtained a t a t e s t temperature of 650°C f o r samples t h a t were i r r a d i a t e d

a t 650°C or lower.

The results i n Table 10 i l l u s t r a t e t h e e f f e c t s of

varying t h e temperatures a t which specimens were i r r a d i a t e d and tested. One group of samples was irradiated and t e s t e d a t 760°C.

These r e s u l t s

show a general improvement i n properties with increasing titanium level. The stress-rupture r e s u l t s f o r t h i s series of tests a r e compared i n Fig.

Ict with those f o r t e s t s on samples i r r a d i a t e d and t e s t e d a t 650°C.

Several samples were irradiated a t 50°C and t e s t e d a t 760°C.

These were

tested under d i f f e r e n t conditions, but t h e trend of increasing rupture

Table 10. Creep Properties of Alloys That Contain Various Amounts of Titanium After Irradiationa a t Various Temperatures Alloy Number

Specimen Number

Titanium Test . Content Number ( w t $)

Temperature, "C IrraTest diation

Stress (Psi)

Rupture Life

Minimum Creep

Elongation

( $1

(h)

x 103 2.5 x 3791 3801

100 101 102 103 104

104 105 106 107 107

3831 3841 3855 3851

R-455 R-454 R-453 R-452 R-451 R-763 R-450 R-458 R-784 R-449

< 0.01 0.11 0.39 0.31 0.55

0.66 0.81 1.04

lo2'

neutrons/cm2 (thermal)

657 760 760 760 760 760 774 774 760 760

760 760 760 760 760 650 760 760 650 760

12.5 12.5 12.5 12.5 12.5 25.5 12.5 12.5 30 12.5

I5 17.5

1.0 0.8 18.1 12.4 340.8 203.3 271.6

49%4 l0OOb 242 38

0.10 0.10 0.0252 0.0333 0.0086 0.004.2 0.0066 0.0196 0.0238 0.0039 0.01135 0.0638

0.21 0.20 0.71 0.66 5.72 1.36 3.05 5.40 1.58 5.19 4.88 3.92

15.1 50.8 358.8 316.3 914. 2b 734.5 4.8

0.037 0.020 0.016 0.011 0.015 0.016 0.19

0.83 1.67 11.9 11.3 30.2 25.1 2.7

159.0

9 x lo1' neutrons/cm2 (thermal) 1913 1918 1986 1985 2045 2035

100 100

104 104 107 107

R-254c R-308 R-255c R- 309 R-261c R- 3 10

< 0.01 < 0.01 0.55 0.55 1.04 1.04

50 50 50 50 50 50

a Annealed 1 h r a t 1177째C before i r r a d i a t i o n . bAnnealed 1000 h r a t 650째C a f t e r i r r a d i a t i o n . C

Load increased a t t h e intervals indicated.

760 760 760 760 760 760

I5 I5 15 I5 I5 I5 20

, TITANIUM CONTENT (ut X I

Fig. 14. Effect of Titanium Content on Variation of Creep Propert i e s a f t e r &radiation.

l i f e and elongation i s t h e same as f o r samples i r r a d i a t e d a t 760°C.

The

only property that seems appreciably d i f f e r e n t i s t h e f r a c t u r e s t r a i n : higher values were obtained f o r samples i r r a d i a t e d a t 50°C.

The most

dramatic effect of i r r a d i a t i o n temperature i s demonstrated by t h e combination of i r r a d i a t i o n a t 760°C and t e s t i n g a t 650°C (compare t h e r e s u l t s i n Tables 7 and l o ) .

The general e f f e c t is t h a t t h e higher irra-

diation temperature r e s u l t s i n a shorter rupture l i f e , a higher minimum creep r a t e , and a lower f r a c t u r e s t r a i n .

For example, a l l o y 107 was

i r r a d i a t e d a t 650°C and t e s t e d a t 32,400 p s i and 650°C. was 2928.8 hr, t h e minimurn creep r a t e was 0.0026$/hr,

The rupture l i f e

and t h e f r a c t u r e

s t r a i n was 12.8%. The same a l l o y i r r a d i a t e d a t 760°C and t e s t e d a t 30,000 p s i and 650°C had a rupture l i f e of only 49.4 hr, a minimum creep

r a t e of O.O24$/hr,

and a f r a c t u r e s t r a i n of 1.6s.

25 Alloys 75 and 76 were similar i n composition t o a l l o y 107, and only limited t e s t i n g has been carried out.

The r e s u l t s of t e n s i l e t e s t s a r e

sumarized i n Table 11. The r e s u l t s f o r t h e unirradiated samples agree

w e l l with those f o r a l l o y 107 (Table 2). Alloy 75 contains 0.062% C, and a l l o y 76 contains 0.117% C; therefore, t h e trends of higher strength and lower f r a c t u r e s t r a i n f o r a l l o y 76 a r e reasonable.

Irradiation

increased strength and decreased f r a c t u r e s t r a i n (Table 11). The f r a c t u r e s t r a i n s agree reasonably well with those f o r a l l o y 107 i r r a d i a t e d a t 650°C (Table 3) and 50°C (Table

4). The samples annealed f o r 100 hr a t 871°C

had a very small grain size, and the f r a c t u r e s t r a i n s were l m e r f o r

these samples.

The higher carbon content of a l l o y 76 resulted i n s l i g h t l y

lower f r a c t u r e s t r a i n s i n t e n s i l e tests. Table 11. Tensile Properties of Irradiated and Unirradiated Alloys 75 (0.99% T i , 0.062% C) and 76 (1.01$ T i , 0.117$ C) a t 650°C Alloy Number

Specimen Before Number Irradiation

Strain Rate

-ts Tensile

(min-1)

ReducElongation, % tion in Uniform Total Area (5 )

x 103 x 103 Unirradiated

75 75 76 76

1197 1l83 1300 1184

I21 121 121 121

75 75 75 75 76 76 76 76

1010 1011 1118 1119 1012 1013 1121 1122

121 121

16 16 121 121 16 16

89.3 0.05 31 0 002 31.4 86.4 0.05 39.3 100.3 0.002 40.5 92.6 Irradiatedb 99.4 0.05 89.5 90 0.002 90.4 0.05 61.6 78 0.002 60.8 66.5 0.05 89.5 102.3 95.3 91.8 0 002 76.8 0.05 62.6 0.002 61.4 65.7

a Annealing designations:

45.0 32.2 34.6 25.1

45.8 33.2 34.8 28.7

36.3 25.9 26.9 24.2

16.0 5.1 8.9 6.4 13.0 5.0 7.1 5.2

17.2 6.1 8.9 6.4 13.2 5.3 7.1 5.2

20.5 11.7 13.2 10.2 19.1

11.7 8.6 8.6

221 = annealed 1 h r a t 1177°C. 16 = annealed 100 h r a t 871°C.

bIrradiated at 150°C t o a thermal fluence of 5.8

lo2'

neutrons/cm*.

26 Several samples of a l l o y s 75 and 76 were creep t e s t e d a t 650°C; t h e r e s u l t s of these t e s t s a r e summarized in Table 12.

The properties of the

U I

unirradiated a l l o y a r e q u i t e camparable with those f o r a l l o y 107 (1.04% Ti) Same samples were i r r a d i a t e d a t 760°C and

t h a t a r e given i n Table 5.

t e s t e d a t 650°C; t h e t e s t r e s u l t s a r e given i n Table 12.

Both of these

alloys had extremely good properties under these conditions.

A comparison

with t h e r e s u l t s f o r heat 107 (l.04$ T i ) i n Table 10 reveals t h e superio r i t y of a l l o y s 75 and 76.

We have no explanation f o r t h i s observation.

Table 12. Creep Properties of I r r a d i a t e d and Unirradiated Alloys 75 (0.995 Ti, 0.0625 C) and 76 (1.015 Ti, 0.1174 C) a t 650°C Alloy Number

Specimen Test Number Number

S t r e s s Rctpture Life (psi)

Minimum Reduction Creep Elongation in Area Rate ( 5) (5 ) ($/hr)

x 103 Unirradiat eda

75 75 76

8293 8341 8292

5276 5458 5275

55 40 55

128.5 1919.9 92.1

0.0325 0.0075 0.0880

21.9 33.3 15.6

0.0442 0.0061 0.0540 0.0059

13.1 23.6 19.6 22.0

18.4 26.3 22.7

Irradiateday

75 75 76 76

10266 10265 10269 10268

R-888 R-855 R-889 R-852

47 35 47 35

248.1 2669.2 276.0 2688.4

a Annealed 1h r a t 1177°C. bIrradiated a t 760°C t o a thermal fluence of 3 x

lo2'

neutrons/cm2.

The f i r s t connnercialheat of a titanium-modified a l l o y was a 100-lb vacuum-melted a l l o y from Special Metals Corporation, designated 21545. This a l l o y was t e s t e d as received (405 cold work), a f t e r annealing 100 hr

a t 871°C ( f i n e g r a i n size, Fig. A), and a f t e r annealing 1h r a t 1177°C (coarse grain s i z e , Fig. 4).

The r e s u l t s of t e n s i l e t e s t s on these

samples a r e summarized i n Table 13. The f r a c t u r e s t r a i n s a t a s t r a i n r a t e of 0.05 min-l a r e shown in Fig. I5 as a function of t e s t temperature. The f r a c t u r e s t r a i n of t h e m a t e r i a l as received was about 105 a t 25°C and

27 Table 13. Tensile Properties of Unirradiated Alloy 21545 (0.49% T i ) Anneala Number

3432 345 1 3447 3435 2531 2521 3436

3444 3446 2530 3425 3429 2523

3448 3445 I565 25 19 1566

I544 1545 1567 I549 1550

2544 2543 2520 1765 1766 2540

2542 2546 2454

2547 25 22 2541

2455 2456 2457 2536 25 35 25 37

Before

Test

121 I21 121 121 121 121 121 121 121 121 121 121 121 121 121 000

000 000 000 000 000 16 16 16 16 16 16

l6 16 16 16 16 16 16 16 16 16 16

46 46 46

Test S t r a i n TemperRate atme (min-l) ( o c )

0.05 0.05 0.05 0.05 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.002 0.05 0.002 0.05 0.05 0.05 0.05 0.002 0.05 0.05 0.05 0.05 0.05 0.05 0.05

0.002

25 200

427 500 550 550 600 600 650 650 760 760 760 800 800 25

25 427 650 650 760

25 427 500 550 600 650

700

Stress, p s i Tensile

(4) x 103 39.4 31.8 29.4 28.1 29.1 30.5

26.4 27 24.9 22.1 25.1

760

0.002 0.05 0.002 0.05 0.002 0.002 0.002

800 871 871 982 600 650 700

x 103 109.4 99.2 95.6 91.1 93.4 94.1 85.9

74.8 76

67.4

24.7

67.6 50.8

23.1

45.4

23.7 23.2

59 41.3 170.4 171.6 147.1 131.6 116.8 93.4 122.9 102.2 103.6 99.8 89 87.6

159.7 164.2 131.5 119 4 113.4 92.7

54.2 37 43.3 42.5 40.4 37.9 2.2 0.9 41.4 41.6

38 0.002

ReducElongation, $ tion in Uniform Total Area

37 36.9 30 31.8 17.6 17.9

26.4 26.5 26.5

73 70.7 59.6 56 50.4 37.3 44.. 9 30 32.1 17.6 17.9 76.9 74.6 56.2

68.9 70.9 73.1

78.9 72.5

72.5 74.9

78.8 81.9

57.5

75.2

53.3 43.9 52.0

65.5 70.9 41.8 41.6 21.2 31.0 12.0 10.1 20.5

67.9 75.3 43.0 43.5 22.0 33.2 19.2 25.2

8.5

18.6 11.9 10.7

8.5 6.3 11.4 8.0

3.4 3.0

48.7

27.7

15.2 11.8

8.9 24.9

49.8 42.1 43.9 30.0

52.0 54.0 48.1 45.9 31.5

39.9

46.1

19-4 21.2

48.3 41.8 46.0 53.5 56.2 53.3 57.8 47.7

12.3 11.7 13.0 1.5 11.0 1.1 1.3 1.0 0.8

63.7 58.1 58.6

57.4 47.7 59.8

45.8 35.5

47.6 38.5

15.6

34.3

37.5 38.2 30.7 21.8 20.4 26.5 22.6

22.7 17.5

37.8 28.8 20.9 13.2 41.9 56.6

42.3 42.2 35.0

27.3 37.2 53.2 42.3 28.3

58.3 74.4 68.1 70.2 29.8 78.1 41.6 64.6 32.6 31.0 31.0

a Annealing designations: 121 = annealed 1 hr at 1177째C; 000 = as received (40% cold work); 16 = annealed 100 hr a t 871째C; 46 = annealed 1 h r a t 1177째C plus 1000 h r a t 650째C.

. ORNL-DWG

69-44390

E50

fa 40 W

a 2

$ 30 LL

100

200

j ~

300

400

500

700

600

800

900

1OOO

TEST TEMPERATURE ("C)

Fig. 15. Fracture Strains of Alloy 21545 (0.49% Ti) i n Tensile Tests a t a Strain Rate of 0.05 min-l.

j i

increased gradually with increasing temperature.

The material annealed

f o r 100 h r a t 871°C had a fracture s t r a i n of 52% a t 2 5 ° C with a minimum

d u c t i l i t y of 32$ at 600°C.

t u r e s t r a i n was 75% a t temperatures between 25 and 600°C and then dropped

After an anneal of 1h r a t 1177"C, the frac-

precipitously with increasing temperature.

Three samples were annealed

! ~

f o r 1 h r a t ll77"C and aged f o r lo00 h r a t 650°C.

This anneal did not

change the strength properties appreciably, but did increase the fracture strain.

Alloy 1U (0.554 T i ) was given the same anneal before testing,

and a comparison between t h e r e s u l t s f o r t h i s a l l o y i n Table 2 and those f o r heat 21545 (0.49% Ti) i n Table 13 shows t h a t the commercial alloy i s s l i g h t l y weaker and t h a t both alloys have comparable fracture s t r a i n s . The r e s u l t s of t e n s i l e t e s t s on irradiated samples of heat 21545 a r e given i n Table U. The r e s u l t s f o r samples irradiated a t 650°C or below were used i n constructing Fig. 16 f o r a s t r a i n r a t e of 0.002 min'l The samples t h a t were irradiated as received and those irradiated a f t e r annealing 100 h r a t 871°C had comparable fracture s t r a i n s . annealed f o r 1h r a t 1177°C were more ductile.

The samples

A comparison with the

Table l4. Tensile Properties of Alloy 21545 (0.49$ T i ) a f t e r Irradiationa Specimen Number

1554 I553 1551 I552 2524 2525 2548 2550 5953 5954 1560 I557 I558 1559 1561

AnnealU Before Test

000 000 000 000 121 121 121 121 3-21 121 16 16 16 16 16

Temperature, "C I r r a d i a t i o n Test

600 600 . 650 650 590 590 600 600 760 760 600 600 600 600 600

550 600 650 650 650 650 650 760 650 760 550 600 650 650 760

Strain Rate (min-l)

0.002 0.002 0.05 0.002 0.05 0.002 0.002 0.002 0.002 0.002 0.002 0.002 0.05 0.002 0.002

Stress, p s i Tensile

x 103

111.6 110.1 107.0 107.7 29.0 28.9 30.0 29.2 22.4

134.9 127.9 124.9 113.3 64.4 58.8 61.3 51.7 22.4 15.8 85.3 69.1 71.5 62.4 37.8

15.8 42.9 39.3 38.2 38.8 37.8

Elongation, $ Uniform Total

11.5 11.2 12.5 5.1 36.0 21.6 21.1 10.3 1.0 0.7 12.1 13.3 u.7 9.7 1.4 ~~~

Irradiated t o a thermal fluence of 2 t o 5 X

bAnnealing designations:

lo2'

neutrons/cm2.

121 = annealed 1 h r a t 1177"C, 000 = as received (40% cold work), 16 = annealed 100 hr a t 871째C.

15.1 13.4 15.1 9.5 37.0 21.9 21.4 13.6 1.4 0.7 12.4 13.4 l5.1 9.7 4.1

Reduction i n Area ( $1 14.? 11.7 16.2 11.7 32.0 17.1 18.3 19.7 3.2 0.81 20.5 8.6 16.2 13.2 7.1

N Q

30 ORNL-DWG 69-14389

AS RECEIVED A IO0 hr AT 871.C a t hr AT 1177OC 0

550

600

0, A, 0 0, A, S

650 700 750 TEST TEMPERATURE ('C)

UNIRRADIATED IRRADIATED

800

850

Fig. 16. Fracture Strains of Alloy 21545 (0.494 Ti) i n Tensile Tests a t a S t r a i n Rate of 0.002 min'l.

r e s u l t s f o r a l l o y 104 (0.554 Ti) i n Tables 3 and 4 shows t h a t t h e commercial a l l o y was more d u c t i l e and s l i g h t l y weaker.

Two samples i n

Table U were i r r a d i a t e d a t 760°C. The samples had much lower f r a c t u r e s t r a i n s than those i r r a d i a t e d a t 650°C or below, indicating again a very strong e f f e c t of i r r a d i a t i o n temperature. Samples of heat 2 W 5 were subjected t o creep t e s t s i n several annealed conditions; the r e s u l t s of these t e s t s a r e presented i n Table 15 , J

f o r unirradiated samples and i n Table 16 f o r i r r a d i a t e d samples.

The

stress-rupture properties a t 650°C i n the unirradiated condition a r e shown i n F i g . 17. The as-received material had t h e longest rupture l i f e a t a given s t r e s s l e v e l and t h e lowest f r a c t u r e s t r a i n .

Annealing a t

871°C produced a material with f i n e grain s i z e t h a t f a i l e d i n much shorter times with f r a c t u r e s t r a i n s of about 504. Annealing f o r 1hr a t

1177°C gave longer rupture l i v e s with intermediate f r a c t u r e s t r a i n s of about 30%. Aging t h i s material at 650 or 760°C before t e s t i n g had only minor e f f e c t s on t h e rupture l i f e or the f r a c t u r e s t r a i n . However, close

6.i .

Table 15.

Creep Properties of Unirradiated Alloy 21545 (0.49% T i )

Specimen Test Number Number

Before Test

(psi)

RY?ture Life

(hr)

Minimum Creep Elongation Rate ($1

($/hr)

Reduction i n Area

($1

x 103 Test Temperature 650°C

1564 1546 3441

3404 5250 3406 5251 3416 5244 3418 5246 3413 5247 1822 1771 1773 1768 2539 2538 2528 5534 3400 2533 5508 5509 5510 5511

5512 5513

5535 5578 6294 6295 6417 6296 6418 6297 6419 6298 6420 6299 6421 5564 5409 5580 54.49 6167 6161 7716 7715 7718 7717, 6384b 638!jb 6386b 6387b 638gb 6428

000 000

121 121 121 121

121 121 121 121 121 121 121 816 16 16 16 46 46

l-48 l48 156 156 103

70 55 70 62 60 55 52 44 47 40 37 35 30 40 55 40 30 40 32.4 40 55 40 55 70

103

103 103 103 103

55 47 40 32.4

174.1 1317.8 0 17.7 26.8 39.9 170.9 197.6 466.1 564.2 1478.6 1720.9 4383.2 24.8.1 20.1 242.6 1171.5 893.7 2077.6 708.9 26.7 527.4 33.7 1.9 5.5 15.7 50.3 177.2 438.8

0.0093 0.0009 0.288 0.062 0.148 0.015 0.037 0.011 0.015 0.0035 0.0063 0.0013 0.0688 1.37 0.120 0.017 0.0129 0.0041 0.0318 0.625 0.0334 0.570 4.53 2.50 0.50 0.165 0.050

l4.1 18.8 41.1 27.2 27.0 22.6 24.8 26.3 26.3 31.3 34.6 30.1 22.2 34.4 52.5 59.4 56.7 38.7 30.8 33.3 34.4 36.1 32.1 39.2 46.1 47 .A 43.9 55.3 40.1

20.5 31.2 32.6 26.7 23.5 17.0 27.2 27.7 22.6 29.7 31.7 30.0 26.7 20.5 47.3 56.6 51.5 32.1 24.7 34.2 41.4 31.8 30.6 33.1

34.3 65.9 45.1 50.3

Test Temperature 760°C 3424 3431 3422

7041 7042 6358

121 121 121

25 20

28.7 71.1 280.0

0.519 0.213 0.060

37.8 39.4 36.2

32.6 35.0 35.1

1= annealed 1h r a t 1177°C; 000 = as Annealing designati received (40$ cold work); 816 = annealed 8 h r a t 871°C; 16 = annealed 100 hr a t 871°C; 46 = annealed 1h r a t 1177°C plus 1000 hr a t 650°C; I48 = annealed 1hr a t 1177°C plus 1000 hr a t 760°C; 156 = annealed 1 h r a t ll77"C, 510 h r a t 550°C, 1500 hr a t 76OoC, and 1 hr a t 1177°C; 103 = annealed 100 h r a t 871"C, 5600 hr i n fluoride salt a t 650°C.

bMolten S a l t Reactor Experiment surveillance controls.

32 Table 16.

Creep Properties of Alloy 21545 (0.49$ T i ) a f t e r I r r a d i a t i o n

hd a

Test

5235 2558 2554 2549 5237 5242 2527 3402 5236 2553 2526 2529 5238 3426

34u 5240 5239 3409 3397 1767 I563 1769 I562 1770 1774 1772

R-676 R-415 R-328 R-325 R-653 R-657 R-258 R-909

R-644 R-249 R-232 R-348 R-654 R-771 R-9Ub

R-84% R-817

R- 127

2451 2452 2453 9100 9099 9102 9103 3430 3233 2556 2555 5234

R-3% R-329 R-652

2449 X50

a d

a a d

I21 I21 I21 I21 121 I21 I21 I21 I21 I21 121 I21 I21 121 I21 I21 I21 I21 I21

R-352 R- I25 R-353 R-354 R-416 R-421 R-346 R-378 R-347 R-362 R-419 R-420 R-317' R-3UC R-319' R-323' R-829 R-906

2448

Anneala Irradiation Stress &fore Temperature Irradiation ("c)

16 l6

16 l6 l6 l6

16 16 l6 l6 l6

16 16 15 l6 l6 l6 L13

u3 121 121 I21 121 I21 121 22 22

Minimum Creep Elongation Reduction in Area ( $1 ( $1

mture LFfe

x io3 Test Temperature 650°C 650 60 12.7 63 8.6 600 55 24.2 600 600 264.8 47 50 65.0 650 50 178.0 650 47 87.4 590 47 33.5 654 44 285.0 650 40 780.6 600 40 453.8 590 40 261.2 590 40 574.6 650 27 ll.6 760 27 6.1 760 650 27 355.5 650 27 109.4 27 1531.8 704 21.5 223.4 760 40 4.9 650 40 72.6 600 35 43.4 650 600 32.4 356.9 650 30 lll.7 650 27 430.8 25 1056.5 650 150 55 3.5 47 16.1 I50 40 66.4 I50 35 193.9 I50 30 596.0 150 27 721.8 I50 650 47 2.8 40 13.1 650 32.4 51.1 650 650 27 l24.1 27 7.8 760 21.5 108.4 760 Test Temperature 760°C 650 25 351.0 600 20 127.4 650 20 200.2 25 0.50 760 760 20 1.5 760 11 94.0 12 101.0 760 760 11 296.0

0.445 0.237 0.114 0.022 0.023 0.oll 0.071 0.258 0.0040 0.0091 0.010 0.017 0.016 0.019 0.056 0.0081 0.0070

0.373 0.077 0.039 0.019 0.019 0.0053 0.0056 1.59 0.40 0.13 0.037 0.0075

0.0059 0.46 0.085 0.013 0.0089 0.024 0.0052 0.0058 0.0328 0.037 0.50 0.42 0.0% 0.040 0.020

7.0 3.1 4.0 10.4 16.8 8.0 10.0 13.2 18.9 16.5 11.3 8.2 13.4 0.33 0.42 4.3 1.3 8.77.

28.6 19.6 23.2 18.3 23.8 19.5

2.0 6.0 3.4 7.3 3.3 3.9 6.5 6.1 7.8 9.9 16.2 9.9 6.9 1.6 2.7 3.4 3.7 0.32 1.3

4.8

18.1 u.3 13.8 0.66 1.0 4.7 5.2

31.1 26.3 17.3 22.8 5.4 ll.6 9.1 5.2

6.4 4.0 4.8 10.7 10.6 ll. 1 12.9 9.9 11.4 10.9

13.5 6.3 7.3 12.9 13.3

7.0

a Annealing designations: I21 annealed 1 hr at ll77"Cj 16 = annealed 100 hr at 871°C; 113 = annealed 1 hr at l l W C plus 100 hr at 650°C; 22 = annealed 1 hr at ll77"C plus 100 hr at 871°C. bAnnealed 5000 hr at 760°C after irradiation. %olten Salt Reactor Experiment surveillance specimens exposed to MSRE core for 5600 hr at 650°C. %. G. arma an, personal cammunication, 1970. E.

33 OWL-DWG 69-14388

AS RECEIVED ANNEALED IOOhr AT 871°C 0 ANNEALED Ihr AT 4477.C. W O h r AT 65OoC . LED Ihr ATI177°C,

4 0

I / I

IO n 400

IO'

103

104

RUPTURE TIME (hr 1

Fig. 17. Creep-Rupture Properties of Unirradiated Alloy 21545 (0.49% T i ) a t 650°C. The number by each datum point designates t h e fracture strain. examination of t h e data i n Table 15 reveals t h a t t h e aging treatment of 1000 h r a t 760°C increased t h e minimum creep r a t e .

The r e s u l t s of creep-rupture t e s t s a t 650°C a f t e r i r r a d i a t i o n a r e shown i n Figs. 18 through 20 f o r material annealed 1 h r a t 1177°C. The i r r a d i a t i o n temperatures a r e indicated by each datum point.

A t irradia-

t i o n temperatures of 650°C or below, t h e rupture l i f e was reduced only a

s m a l l amount, and t h e minimum creep r a t e was unchanged.

A t an irradia-

t i o n temperature of 760°C and t e s t temperature of 65OoC, the rupture

life decreased at least one order of magnitude, and the minimum creep r a t e increased one order of magnitude.

The corresponding f r a c t u r e

s t r a i n s shown i n Fig. 20 a r e 3%and higher f o r the samples i r r a d i a t e d

a t 650°C o r below and about 0.5% f o r samples i r r a d i a t e d a t 760°C. Tests R-841 and R-817 (Table 16) demonstrate t h e a b i l i t y of annealing a t 760°C a f t e r i r r a d i a t i o n t o degrade t h e properties a t 650°C. This observation

ctural change during annealing at 760°C i s responsible f o r the poor properties after irradiation. A comparison of t h e results f o r heat 21545 i n Figs. 18 through 20 with those i n Fig. 14 f o r a l l o y 1% (0.55% T i ) shows good q u a l i t a t i v e agreement. The r e s u l t s of postirradiation creep t e s t s on material having a f i n e grain s i z e (annealed a t 871°C) a r e a l s o given i n Table 16. The f r a c t u r e supports t h e premise t h a t a

. c

34 ORNL-DWG 69-14387

(xi031

U rl

fn fn

E 30 fn

0 400

10'

102 RUPTURE TIME (hr)

Fig. 18. Stress-Rupture Properties of Alloy 21545 (0.49%~T i ) a t 650째C. A l l material annealed 1 h r a t 1177째C. Number by each datum point designates t h e i r r a d i a t i o n temperature i n degrees Celsius.

ORNL-DWG 69-14386

600

760

40-4

MINIMUM CREEP RATE (%/hr)

Fig. 19. Creep Properties a t 650째C of Alloy 21545 (0.49$ T i ) . A l l material annealed 1 h r at 11'77째C. Number by each datum point designates i r r a d i a t i o n temperature i n degrees Celsius.

35 ORNL-DWG

STRAIN RATE (%/hr)

Fig. 20. Fracture S t r a i n of Alloy 2l545 (0.49% T i > a t 650°C a f t e r Irradiation. A l l material annealed 1hr a t 1177°C. Number by each datum point designates the i r r a d i a t i o n temperature i n degrees Celsius. s t r a i n s f o r these samples were generally lower than those f o r the mate-

r i a l annealed 1 h r a t 1177°C. material a r e shown i n Fig. 21.

The stress-rupture properties of t h i s In addition t o t h e rupture l i f e being

low f o r t h e fine-grained, unirradiated material, the rupture l i f e was reduced a t l e a s t 50% by i r r a d i a t i o n . Some of t h e fine-grained material of heat 21545 was included i n the surveillance program f o r t h e MSRE.

The detailed r e s u l t s of these t e s t s

ORN L- D W

( X to3)

-14384

--.-840 In

4 8

10'

to2

to3

to4

RUPTURE TIME (hr)

Fig. 21. Stress-Rupt Properties of Alloy 21545 (0.49% T i ) a t 650°C. Annealed 100 hr a t 871°C t o produce a f i n e grain size. Number by each datum point indicates t h e i r r a d i a t i o n temperature i n degrees Celsius.

36 were reported previously6 and w i l l be discussed only b r i e f l y i n t h i s These samples were exposed t o molten fluoride s a l t f o r 5600 h r

report.

a t 650°C while receiving a thermal-neutron fluence of 4.1 neutrons/cm2. i n Fig. 22.

lo2'

The stress-rupture properties of these samples a r e shown The unirradiated control samples were exposed t o barren

f'uel s a l t f o r an equivalent length of time and show a s l i g h t reduction in rupture l i f e (compare Figs. 17 and 22).

I r r a d i a t i o n decreased t h e

rupture l i f e about an order of magnitude, and t h e f r a c t u r e s t r a i n s ranged from 1.6 t o 3.7$.

Since t h e metal was q u i t e compatible with the

salt, we a t t r i b u t e t h e changes t o the long thermal anneal and t h e neutron exposure. Several creep t e s t s were run at 760°C on both t h e unirradiated and i r r a d i a t e d materials.

The r e s u l t s of these t e s t s a r e given i n Tables 15

and 16, and t h e stress-rupture properties a r e shown graphically i n Fig. 23.

The rupture l i f e a t 760°C was not reduced by i r r a d i a t i o n a t

650°C or beluw, but was reduced d r a s t i c a l l y by i r r a d i a t i o n a t 760°C. The data i n Table 16 a l s o show t h a t t h e higher i r r a d i a t i o n temperature increased the minimum creep r a t e and decreased t h e f r a c t u r e s t r a i n . The microstructures of several samples were examined a f t e r t e s t i n g . Typical photomicrographs of alloys 100, 102 (0.39% T i ) , 1%(0.55% Ti), and 107 (1.04% Ti) after extended aging a t 650°C a r e shown i n Fig. 24. The carbide p r e c i p i t a t i o n along grain boundaries and within t h e matrix was l i g h t in a l l o y 100, heavier i n a l l o y 102, and s t i l l heavier f o r

alloys 104 and 107.

The f r a c t u r e of a l l o y 100 a f t e r creep t e s t i n g a t

40,000 p s i and 650°C i s shown i n Fig. 25.

Numerous intergranular cracks

were present, and the f r a c t u r e was intergranular.

The short time of

about 50 hr a t temperature was not s u f f i c i e n t t o cause carbide precipitation.

The f r a c t u r e of a l l o y 104 (0.55$ T i ) a f t e r t e s t i n g a t 40,000 p s i

and 650°C i s shown in Fig. 26.

Details of i q o r t a n c e a r e t h e carbide

s t r i n g e r s , general carbide precipitation, intergranular cracking, and a predominantly intergranular fracture.

This sample was a t temperature

f o r loo0 hr, and copious amounts of carbide p r e c i p i t a t e formed.

Photo-

micrographs of t h e f r a c t u r e of a l l o y 107 (1.04% T i ) a r e shown i n Fig. 27.

6H. E. McCoy, Jr., An Evaluation of t h e Molten-Salt Reactor Experiment Hastelloy N Surveillance Specimens - Second Group, ORNL-TM 2359 (1969).

ORNL-0%

69-14385

Fig. 22. Stress-Rupture Properties of Alloy 21545 (0.49% T i ) a t 650°C. A l l material annealed 100 hr a t 871°C and exposed t o fluoride s a l t f o r 5600 h r a t 650°C. Number by each datumpoint indicates f r a c t u r e strain. ~

ORNL- DWG 69- 44382

i0-4

100

10' 102 RUPTURE TIME (hr)

I 03

Fig. 23. Stress-Rupture Properties of Alloy 2U45 (0.49% T i ) a t A l l material annealed 1hr a t 1177°C. Number by each datum point indicates the i r r a d i a t i o n temperature i n degrees Celsius. 760°C.

Fig. 24. Microstructures of Several Titanium-Modified Alloys Annealed 1 h r at 1177OC, Aged 1000 hr at 65OoC, and Stressed 1000 hr at 650째C at 32,400 psi. Etchant: glyceria regia, 500X. Reduced 22,5$.

Fig. 25. Microstructures of Alloy LOO after Annealing 1h r a t 1177째C and Testing a t 40,000 psi and 650째C. (a) Fracture,. 1OOX; (b) typical, 5 0 0 ~ . Etchant, glyceria regia.

Fig. 26. Photomicrographs of A l l o y 106 (0.554 T i ) a f t e r Annealing 1 h r a t 1177째C and Stressing a t 40,000 p s i at 650째C. (a) Fracture, 1OOx; (b) typical, 500~. Etchant, glyceria regia.

Fig. 27. Photomicrographs of Alloy 107 (1.04% T i ) a f t e r Annealing 1h r a t 1177째C and Testing a t 650째C and 40,000 p s i . (a) Fracture, lOOX; (b) typical, 500X. Etchant, glyceria regia.

The sample was stressed a t 40,000 p s i a t 650°C and fractured a f t e r

1490 hr.

The grain s i z e was q u i t e small, and there were large quantities

of p r e c i p i t a t e present.

The f r a c t u r e was predominantly intergranular.

Alloy 21545 (0.49% Ti) was examined metallographically a f t e r being stressed a t 40,000 p s i at 650°C. The f r a c t u r e was predominantly i n t e r granular (Fig. 28).

The sample was a t temperature 566 hr, and some very

f i n e p r e c i p i t a t i o n occurred i n t h e matrix and a t grain boundaries.

The

s t r i n g e r s of p r e c i p i t a t e were present i n t h e annealed material before t e s t i n g (Fig. 4). Alloy 100 (< O.Ol$ Ti) was examined after i r r a d i a t i o n and creep t e s t i n g a t 650°C. Typical microstructures a r e shown i n Fig. 29. sample f a i l e d a f t e r only 1.1% s t r a i n (Table 7).

This

There were a f e w i n t e r -

granular cracks, but t h e deformation seemed r e s t r i c t e d l a r g e l y t o t h e

I l

zone near t h e fracture. Some f i n e carbides were formed a t t h e grain boundaries and i n the matrix. The microstructure of a l l o y 104 (0.55% Ti) i s sham i n Fig. 30 a f t e r i r r a d i a t i o n and creep t e s t i n g a t 650°C. This a l l o y f a i l e d after 10.0% s t r a i n ; t h e f r a c t u r e was l a r g e l y intergranular.

Alloy 107 (1.04% Ti) failed after ll.1$ s t r a i n .

The microstructure,

shown i n Fig. 31, consists of an intergranular f r a c t u r e with some carbide precipitation.

Alloy 21545 (0.49% Ti) was examined after exposure t o t h e MSRE core and control f a c i l i t y .

The material was annealed a t 871°C before exposure

t o give a f i n e grain s i z e (Fig. 4). The samples were then exposed t o fluoride salt a t 650°C f o r 5600 hr.

Some of t h e samples were exposed t o

barren f u e l salt f o r controls, and others were i r r a d i a t e d i n t h e MSRE core.

Microstructures of an irradiated sample t h a t was t e s t e d a t 25°C

are s h a m i n Fig. 32.

The fracture was transgranular.

There was some

edge cracking due p a r t i a l l y t o s l i g h t changes i n composition near t h e surface.

The comparative control sample i s shown i n Fig. 33.

t u r e was transgranular, and t h e r e was less edge cracking.

The frac-

Another irra-

diated sample was t e s t e d a t 650°C; t y p i c a l photomicrographs a r e shown i n Fig. 34.

The f r a c t u r e was intergranular, and t h e r e were a few i n t e r -

granular cracks near t h e fracture.

There were some edge cracks.

The

control scmq>le shown i n Fig. 35 was more ductile, with 40.8$ s t r a i n compared with only 6.2% f o r t h e irradiated sample shown i n Fig. .%. higher s t r a i n led t o extensive intergranular cracking.

The

Fig. 28. Photomicrographs of Alloy 21545 (0.49$ T i ) after Annealing 1 h r a t 1177째C and Testing a t 650째C and 40,000 p s i . (a) Fracture, 1OOX; (b) typical, 5 0 0 ~ . Etchant: glyceria regia.

. .

Fig. 29. Photomicrographs of A l l o y 100 (< 0.015 T i ) a f t e r Annealing 1hr a t ll77"C, I r r a d i a t i o n f o r 1000 hr a t 650째C t o a Thermal Fluence of 2.5 X lo2' neutrons/cm*, and Testing a t a S t r e s s of 32,400 p s i a t 650째C. Fractured i n 105.4 hr with 1.1% strain.. ( a ) Fracture, loox; (b) typical, 5OOX. Etchant: aqua regia.

Fig. 30. Photomicrographs of Alloy 1% (0.55% T i ) Annealed 1h r a t ll??"C, Irradiated f o r 1000 h r a t 650째C t o a Thermal Fluence of 2.5 X lo2' neutrons/cm2, and Stressed a t 32,400 p s i a t 650째C. Fractured i n 1446.3 h r with 10.0% s t r a i n . ( a ) Fracture, 1OOX; (3)typical, 500X. Etchant: aqua regia.

Fig. 31. Photomicrographs of Alloy 107 (1.04% Ti) Annealed 1hr a t ll77"C, I r r a d i a t e d f o r 1000 hr a t 650째C t o a Thermal Fluence of 2.5 X lo2' neutrons/cm2, and Stressed a t 32,400 p s i a t 650째C. Fractured i n 2497.2 hr with 11.1% s t r a i n . (a) Fracture ( f l a t t e n e d end caused by handling), 1OOX; (b) f r a c t u r e (blunted t i p caused by handling), 500x.

W *

F i g . 33. Photomicrograph of the Fracture of a Titanium-Modified Hastelloy N Surveillaace Sample (Heat 2l!j45) Tested at 25째C at a Strain Rate of 0.05 min'l. Exposed to a static fluoride salt for 5500 hr at 650째C before testing. Note the shear fracture and the absence of edge cracking. 1OOx. Etchant: glyceria regia.

Fig. 34. Photomicrographs of a Titanium-Modified Hastelloy N Surveillance Sample (Heat 21545) Tested a t 650째C a t a Strain Rate of 0.002 min'l. Exposed i n t h e MSRE core f o r 5500 h r a t 650째C t o a thermal fluence of 4 . 1 X lo2' neutrons/cm2. Failed a f t e r 6.2$ s t r a i n . ( a ) Fracture, 1OOx; (b) edge of sample about 1/2 in. from fracture, 1OOX; ( c ) edge of sample showing edge cracking, 5 0 0 ~ . Oxide formed during t h e t e n s i l e t e s t . Etchant: aqua regia. Reduced 10%.

Fig. 35. Photomicrograph Sharing a Portion of t h e B a c t u r e of a Titanium-Modified Hastelloy N Surveillance Sample (Heat 21545) Tested a t Failed a f t e r 40.8% s t r a i n . 650째C a t a S t r a i n Rate of 0.002 min'l. Exposed t o static fluoride s a l t f o r 5500 hr a t 650째C before t e s t i n g . lOOx. Etchant: glyceria regia. Alloys Containing Zirconium Three groups of alloys were studied t h a t contained various amounts of zirconium.

A s e r i e s of 2-lb laboratory melts containing fram l e s s

than 0.03 t o 1.18%Zr was made from v i r g i n melting stock.

These alloys

are designated 100 and 108 through 112; t h e i r chemical compositions are given i n Table 1. Typical microstructures of these alloys a f t e r a l-hr anneal at 1177째C a r e shown i n Fig. 36.

The addition of t h e first O . l $ Zr

( a l l o y 108) markedly decreased t h e grain size.

Further additions up t o

t h e mstximum 1.18%Zr caused only s l i g h t additional refinement i n grain size.

The second group consisted of two alloys t h a t contained 1% Zr

with additions of 0.1 and 1.04 Re.

The t h i r d s e r i e s of a l l o y s consisted

of two 100-lb heats from Special Metals Corporation t h a t contained 0.05 and 0.35% Zr, designated heats 21555 and 21554, respectively.

The a l l o y

Fig. 36. Typical Photomicrographs of Alloys 100 and 108 Through 112. Annealed 1 h r a t 1177OC. ( a ) Alloy 100 (< 0.03% Zr), (b) alloy 108 (0.10% Z r ) , ( c ) a l l o y 109 (0.28% Zr), (d) a l l o y 110 (0.53$ Z r ) , ( e ) a l l o y ll1 (0.75% Z r ) , and ( f ) a l l o y 112 (1.18%Z r ) . loox. Etchant: glyceria regia. Reduced 42.5%.

52 with t h e higher zirconium content from Special Metals Corporation was i n i t i a l l y forged a t 1177°C and cracked profusely.

A second a l l o y was

melted and forged a t 1066"C, and a good product was obtained.

This

experience gave t h e first indication t h a t zirconium caused t h e a l l o y t o be hot-short and t h a t it would l i k e l y have poor weldability.

Typical

photomicrographs of these two alloys are shown i n Figs. 37 and 38.

received, t h e material contained 40% cold work; annealing a t 871°C produced a banded s t r u c t u r e with a f i n e grain s i z e and numerous carbides. The banding i s l i k e l y due t o the working pattern, but t h e r e was a l s o

some variation i n t h e density of t h e carbides. Alloys 100 and 108 through ll2 were t e s t e d t o determine t h e i r tens i l e properties.

Samples of each a l l o y were annealed 1 h r a t 1177°C and aged f o r 1000 hr a t 650°C t o match t h e thermal h i s t o r y of samples t h a t

w e r e irradiated. The r e s u l t s of t e n s i l e tests on t h e unirradiated samples a r e given i n Table 17. The variation o f t h e strength parameters

a t 650°C with zirconium content i s shown i n Fig. 39. The yield strength was not affected appreciably u n t i l t h e zirconium l e v e l exceeded 0.3$, but it increased markedly above t h i s level.

The ultimate t e n s i l e strength

increased about 10% with t h e addition of 0.1% Zr, but very l i t t l e f u r t h e r strengthening occurred as t h e zirconium content increased.

In t h e

absence of aging, t h e strength was not affected by zirconium content. The fracture s t r a i n s are shown i n Fig. 40 as a function of zirconium content.

A very d i s t i n c t maximum f o r aged material occurred a t a concentra-

t i o n of about 0.3% Zr.

The f r a c t u r e s t r a i n was almost doubled by t h e

addition of even 0.14 Zr. ,

The fracture s t r a i n a t t h e lower s t r a i n rate

was lower, a trend t h a t is usually observed.

I n unaged material, a

maximum occurred i n t h e f r a c t u r e s t r a i n f o r zirconium concentrations of 0.5 t o 0.74. The same experimental alloys were i r r a d i a t e d a t 650°C t o a thermal fluence of 2.5

lo2'

neutrons/cm2.

The results of t e n s i l e t e s t s on

these samples a f t e r i r r a d i a t i o n are summarized i n Table 18, and t h e variations of t h e fracture s t r a i n s with zirconium content a r e shown i n Fig. 41.

The p o s t i r r a d i a t i o n results a l s o show a d u c t i l i t y maximum a t

about 0.34

The results f o r t h e i r r a d i a t e d and unirradiated samples

are compared i n Fig. 42, where t h e r a t i o s of t h e various t e n s i l e properties

W P

bd c

Fig. 37. Typical Photomicrograph of Alloy 21555 (0.05% Zr) Annealed 1 hr a t 1177째C. loox. Etchant: glyceria regia.

Fig. 38. Typical Photomicrograph of Alloy 21554 (0.35% Z r ) Annealed 1hr a t 1177째C. lOOx. Etchant: glyceria regia.

Ld 4

Table 17.

Alloy

Nmber

Specimen Number

Tensile Properties of Several Unirradiated Zirconium-Modified A l l o y s a t 650°C Strain Rate (Illin-’)

Stress, p s i Ultimate Tensile

ReducElongation, $ tion i n Uniform Total Area

Heat Treat-

($1

merit

x 103 108 109

3870 3880 3890 3900 3909

110

0.05 0.05 0.05 0.05 0.05

25.7 29.4 25.9 26.7 26.5

x 103 68.3 78.6 75.6 75.5 69.6

28.6 33.7 42.8 42.3 32.4

29.8 34.7 44.8 44.6 34.4

23.7 26.8 30.6 39.9 33.3

a a a a a

46.1 47.3 54.8 52.5 49.0 48.9 45.6 35.6 26.9 22.6

48.9

42.6 47.2 50.9 47.3 53.1 48.7 53.2 47.5 28.1 25.8

lll 112 108

2084

0.05

28.9

86.4

108

2078

0.002

29.2

86.2

109 109

2061 2062 2105 2ll4 2132 2w 2U6 21ct8

0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002

29.1 29.4 43.8 44.6 50.5 52.9 51.2 55.2

85.9 84.7 85.8 87 92.2 80.9 89.1 89.3

UO ll0 111 ll1

ll2 ll2

50.0

57.7 56.2 56.8 51.9 50.3 44.4 28.8 23.3

b b b b b b b b

aAnnealed 1hr a t 1177°C. bAnnealed 1hr a t 1177°C plus lo00 hr a t 650°C.

. 1

55 ORNL-DWG 69-4438W2

B.i

(X 103)

-9

.u)

20 a

STRAIN RATE = 0.05 min-',

AGED

o STRAIN RATE 50.002 min-1, AGED A

STRAIN RATE =0.05 min-l, UNAGED

0 0

0.2

0.4 0.6 0.8 1.0 ZIRCONIUM CONTENT ( U t %)

1.2

1.4

Fig. 39. 'Influence of Zirconium Content on Tensile Properties a f t e r Annealing 1h r a t 1177°C and Aging 1000 h r a t 650°C.

Table 18.

Tensile Properties of Several Zirconium-Modified Alloys a t 650°C a f t e r Irradiationa

Alloy Number

Specimen Number

108 108 109 110 ll0

2075 2087 2074 2109 2116 2u1 2147 2152

Strain

Stress, psi

Rate

(dnal)Yield

x 103

111 1

. W

112 112

0.05 0.002 0.002 0.05 0.002 0.002 0.05 0.002

33.9

ultimate Tensile

x 103 72.6

unifFm

Elo ation, 4 Total

28.0 22.1

24.5

29.0 23.5 29.8

16.5 13.4

18.5 17.1

8.5 14.9 7.1

Reduct ion

in Area

(8)

10.0 17.2

18.7 18.9 28.5 17.9 13.6 12.1 13.4

7.7

11.9

a Annealed 1 h r a t 1177°C before i r r a d i a t i o n a t 650°C t o a thermal fluence of 2.5 x lo2' neutrons/cm2.

-....

___

........

. . .._._..____........____...._..___.I__._-________....

ORNL-DWG 69-14380R3

ORNL-OWG 69-14379R

3 20

3 45

;

5a b

I-

3 I-

I-

;35 O

30 STRAIN RATE ~ 0 . 0 5mln-', AGED o STRAIN RATE 10.002 rnin-', AGED A . STRAiN RATE =0.05 min-', UNAGED

0 0

20 0

0.2

0.4

0.6

0.8

4.0

1.2

0.2

0.4 0.6 0.8 ZIRCONIUM CONTENT (w?%)

4.0

4.2

1.4

ZIRCONIUM CONTENT (ut%)

Fig. 40. Influence of Zirconium Content on Fracture Strain. The unaged samples were annealed 1hr a t 1177째C and t h e aged samples were annealed 1hr a t 1177째C and aged 1000 h r a t 650째C before testing.

Fig. 41. Influence of Zirconium Content on Fracture S t r a i n a t 650째C a f t e r Irradiation. Annealed hr at u77~c before irradiation.

LiJ

ORNL- DWG 69- 44378R

YIELD STRESS

‘

1.0

1.2

0.8

wl-

+ a

4 8 n u

2E cr

0.6

0.2

0.4 0.6 0.8 ZIRCONIUM CONTENT ( w t % )

Fig. 42. Effects of Zirconium on t h e Unirradiated and Irradiated Tensile Properties a t 650°C and a S t r a i n Rate of 0.002 min-l. a r e a l s o shown.

The yield strength was increased s l i g h t l y by i r r a d i a t i o n

i n alloys t h a t contained 0.10 and,0.28$ Zr (alloys 108 and 109) and decreased s l i g h t l y i n t h e other alloys. was decreased about 30% i n each alloy.

The ultimate t e n s i l e strength The fracture s t r a i n was reduced

about 50% by i r r a d i a t i o n i n alloys t h a t contained 0.3% Z r o r less and reduced about 70%f o r alloys t h a t contained more zirconium. Samples of alloys 100 and 108 through 112 w e r e also i r r a d i a t e d a t

43°C t o a thermal fluence of 9 x lo1’ neutrons/cm*.

These samples were tested a t 650 and 760°C; the r e s u l t s a r e given i n Table 19. The frac-

t u r e s t r a i n s a r e sham i n Fig 43 as a function of zirconium concentrat i o n . The peak i n f r a c t u r e s t r a i n a t 650°C with zirconium concentration

s t i l l occurs, but l e s s sharply and a t a d i f f e r e n t concentration than observed previously (Fig. 41).

The f r a c t u r e s t r a i n s a r e a l s o higher i n

t h e experiment a t lower temperature.

We do not know whether t o a t t r i b u t e

these differences t o t h e

ion i n i r r a d i a t i o n temperature or i n

fluence.

a n t i a l evidence indicates t h a t t h e irra-

However, t h e c

d i a t i o n temperature i s t h e most important factor.

Recall t h a t t h e yield strengths of t h e aged control samples varied from 28,000 t o 55,000 p s i

58 Table 19.

Alloy Number

Tensile Properties of Several ZirconiumModified Alloys a f t e r Irradiationa Stress, p s i

Specimen Number

Yield

x 103 108 109 110 ll1

112

Tensile

Elongation, $ ’ Uniform Total

($1

x 103

Test Temperature 650°C 18.9 31.5 62.9 34.7 37.8 88.8 38.6 33.1 79.3 33.6 33.3 80.8 80 29.9 33.8

2086 2064 2u3 2139 2156

Reduction i n Area

19.2 38.9 40.5 37.4 31.5

17.5 24.0 34.9 33.8 21.8

10.0 10.2 11.6 12.4

12.3 25.6 25.6 30.7

15.5

18.0

9.8 28.3 27.3 34.2 17.5

Test TemDerature 760°C 108 109 110 1ll u2

2088 2058 2ll1 2128

27.4 29.8 27.7 32.2

2153

52.3 58.3 56.7 56.2 61.8

a Annealed 1 h r a t 1177bC, irradiated a t 50°C t o a thermal fluence of 9 X loi9 neutrons/cm2 and tested a t a s t r a i n r a t e of 0.002 min‘l. ORK-WG

69-44377R

-x 2z 25 GI $2c

Y 45 II.

a4 0.6 0.8 1.0 ZIRCONIUM CONTENT (Ut %I

1.2

1.4 a

Fig. 43. Fracture Strains of Several Zirconium-Bearing Alloys a f t e r Irradiation a t 43°C t o a Thermal Fluence of 9 X lo1’ neutrons/cm*.

and varied systematically with zirconium content (Fig. 39).

Irradiation

a t 650°C d i d not change these values markedly (Table l8). The yield strengths of the alloys i r r a d i a t e d a t 43°C a r e a l l about 33,000 p s i and Thus, t h e

do not vary detectably with zirconium content (Table 19).

thermal treatment seems necessary t o bring about the changes characteri s t i c of alloys with various concentrations of zirconium. Some of t h e samples i r r a d i a t e d a t 43°C were t e s t e d a t 760°C.

The.

f r a c t u r e s t r a i n varied markedly with zirconium content (Fig. 4 3 ) .

How-

ever, t h e peak was s h i f t e d t o a higher zirconium l e v e l than was noted a t

a t e s t temperature of 650°C. Alloys 100 and 108 through 112 were a l s o creep-rupture tested.

The

r e s u l t s of several t e s t s on material annealed f o r 1 hr a t 1177°C a r e shown i n Table 20 and i n Figs. 44 through 46.

The stress-rupture proper-

ties, shown i n Fig. 44, illustrate t h e potent e f f e c t of zirconium i n

increasing t h e rupture l i f e a t a given l e v e l of s t r e s s .

The rupture l i f e

of a l l o y 109 (0.28% Z r ) seems anomalously high, but t h e other alloys followed t h e trend of increasing rupture l i f e with increasing zirconium content.

The creep-rupture strength, shown i n Fig. 45, was a l s o

affected by the zirconium concentration. Again, a l l o y 109 was t h e only exception t o t h e trend of increasing strength with increasing zirconium content.

Zirconium improved t h e f r a c t u r e s t r a i n (Fig. 4 6 ) .

The addition

of only 0.1% Zr ( a l l o y 108) improved t h e f r a c t u r e s t r a i n , compared t o t h a t of a l l o y 100, several hundred percent a t l o w s t r a i n rates.

Higher

concentrations of zirconium improved the f r a c t u r e s t r a i n a t high s t r a i n

rates, but brought about l i t t l e @ddedimprovement a t low s t r a i n r a t e s . Some samples were aged f o r 1000 hr a t 650°C t o correspond t o t h e thermal treatment of irradiated samples.

The r e s u l t s of creep-rupture

tests on these samples are given i n Table 21. . A comparison with t h e properties of t h e unaged samples i n Table 20 shows t h a t t h i s aging treatment did not a f f e c t t h e properties appreciably. Alloys 100 and 108 through 112 were i r r a d i a t e d t o a thermal fluence z

of 2.5

lo2'

neutrons/cm2 a 650°C and subjected t o creep-rupture tests

a t 650°C a f t e r i r r a d i a t i o n . The r e s u l t s of these tests a r e given i n Table 22. These samples were stressed a t 32,400 p s i f o r about 675 hr

Table 20.

Creep Properties a t 650째C of Several Alloys That Contain Various Concentrations of Zirconiuma

Minim Alloy Number

Specimen Test Number Number

Stress mture ufe Creep (psi)

x 100 100 100 100 100 108 108 108 108 109 109 109 109 109 110 110 110 110 111 111 111

112 112 112 112

112

1915 1917 1909 19U 1921 2077 2082 2091 2094 2068 2066 2057 2072 2065 2115 2106 2117 2112 2130 2135 2126 2159

2151 2U5 2163 2158

5725 6099 5554 6108 6172 5689 5991 5721 6162 5778 5712 5740 5987 5909 5746 5673 6096 5665 5771 5688 6091 5752 5908 5666 5641 6166

Elongation

($1

Reduct ion i n Area

(8)

io3

55 47 40 32.4 27 55 47 40 32.4 70 55 51 47 40 70 55 47 40 70 55 47 70 62 55 51 47

3.4 20.8 47.4 185.1 827.8 U.1 178.7 348.5 2137.3 30.7 178.1 358.5 851.2 U74.3 12.5 200.7 283.0 709.0 26.9 192.8 589.4 39.2 169.5 537.4 673.1 1254.8

0.43 0.55 0.088 0.0077 0.0022 0.16 0.025 0.018

0.0059 0.128 0.015 0.013 0.0075 0.0042 0.138 0.028 0.030 0.0080 0.174 0.027 0.014 0.181 0.038 0.011 0.010 0.0056

22.5 14.5 12.5 6.3 4.9 17.5 18.8 25.0 31.3 32.4 32.0 31.0 34.1 46.9 39.3 35.9 38.2 18.8 35.0 30.9 36.6 38.2 34.3 30.5 37.5 32.0

20.8 12.5 9.5 6.2 4.0 U.8 15.9 21.4 28.4 28.3 31.0 36.0 34.8 35.3 26.9 31.1 38.6 25.6 26.9 29.7 33.7 31.6 27.2 26.9 28.8 29.9

a Annealed 1 h r a t 1177째C before testing.

61 ORNL-DWG 69-14376 ( ~103)

5 40

v) W v)

30 ALLOY

NO. 20

-0

100

108 A 109 A ($0 10

-0 8

H1 112

10'

i I

102

(03

RUPTURE TIME (hr)

Fig. 44. Stress-Rupture Properties a t 650째C of Several Alloys That Contain Various Concentrations of Zirconium. A l l material was annealed 1h r a t 1177째C before t e s t i n g . * ORNL-DWG 69-14375

-'g -

E 30 .

A 109

CONTENT(%)

0.28

40-3 .-

40-2

10-'

MINIMUM CREEP RATE (%&r)

Fig. 45. Influence of Zirconium on Creep Rates a t 65OOC. alloys annealed 1h r a t 1177째C before t e s t i n g .

All

62 ORNL-OW 69-44574

-E”

9 40 Iv)

a a 30

3 !

2 LL

10-3

100

10-2 4 0-‘ MINIMUM CREEP RATE (%/hr)

F i g . 46. Fracture Strains During Creep Tests a t 650°C f o r Alloys That Contain Various Amounts of Zirconium. A l l samples annealed 1hr a t 1177°C before testing.

Table 21.

Alloy Number

Stress Properties a t 650°C of Several Ageda Alloys That Contain Zirconium

Specimen Test Number Number

Stress (psi)

Rupture Life

Minimum Creep

Elongation

Reduction in Area

(4)

23.0 1.0 40.6 15.0 38.4 3.5 37.5

19.2 0

x 103 100 100

108 108 109 109 110 110 111

lll ll2 ll2 ~

1922 1919 2090 2085 2071 2063 2123 2120 2 u 2137 2154 2U9

5723b 5630 5680b 5911 5727b 6092 568% 6086 5674b 6144 5637b 5629

55 17.8 55 32.4 55 32.4 55 32.4 55 32.4 55 32.4

5.9

1007 280.9 1000 152.7 1000 197.7 1002.2 305.5 1002.2 530.1 1000.6

0.61 0.0007 0.c42 0.0096 0.068 0.0044

0.098

38.8 9.8

44.7 35.0

0.0012

0.052 0.0011 0.030 0.0016

29.7 3.1 34.4 3.1

28.3 0.03 29.6 0.01

a Annealed 1 h r a t 1177°C and aged 1000 h r a t 650°C before testing. bDiscontinued before failure.

Table 22-

Creep Properties at 650째C of Irradiated' Alloys t h a t Contain Various Amounts of Zirconium

Stress,b p s i Alloy Number

Specimen Test Number Number

32,400 Zirconium Content Minimum (w-t 79 Creep Elongation

($1

40,000 Minimum Rupture Creep Life (hr)

Elongation

(46)

Reduction In Area

True Fracture Strain

(8)

( 9)

< 0.03

0.0099

1.1

R- 163

0.10

0.0094

0. U 7

53.1

7.3

19.0

21.1

2070

R- 171

0.28

0.0034

8.2 2.3

0.025

301.3

10.8

17.2

18.9

110

2121

0.53

0.0013

0.97

0.0094

429.5

5.9

11.1

11.7

111

2UO

R-172 Rt 195

0.75

0.0037

3.3

0.035

127.1

5.7

13.9

15.0

112

2155

R- 199

1.18

0.0033

5.1

0.080

46.5

3.9

6.8

7.0

100

1916

R-U6

108

2076

109

A l l samples annealed 1 h r a t 1177째C before i r r a d i a t i o n a t 650째C t o a thermal fluence of 2.5 neutrons/cm2.

lo2'

b A l l tests except R-U6 were i n i t i a l l y s t r e s s e d a t 32,400 p s i f o r about 675 hr and then stressed a t 40,000 p s i . Test R-U6 f a i l e d while loaded a t 32,400 p s i a f t e r 105.4 hr.

64 and then stressed a t 40,000 p s i u n t i l f a i l u r e .

(This increase i n s t r e s s

l e v e l was necessary t o cause f a i l u r e i n a reasonable time.)

The rupture

hj L

l i f e a t t h e higher s t r e s s i s sham i n Fig. 47 as a function of zirconium content.

The s t r a i n during t h e first 675 h r a t 32,400 p s i i s given i n

parentheses, and t h e s t r a i n a t t h e higher s t r e s s l e v e l i s sham by each 600

500

ORNL-OWG 67-352

E S T CONDITIONS I TEMPERATURE= 650째C 675 tw AT o = 32.400 psi INDICATED TIME AT 40.000 psi

IRRADIATIONCONDITIONS 5*9(0-97) TEMPERATURE= 6 W C

400

THERMAL FLUEN== 2.5 102' neutrons/cm2

0.2

0.4

0.6

0.8

1.0

1.2

ZIRCONIUM CONTENT (ut%)

Fig. 47. Influence of Zirconium Content on the Creep-Rupture Properties a t 650째C a f t e r Irradiation. A l l samples annealed a t ll77"C before i r r a d i a t i o n . point.

The rupture l i f e increased rapidly with zirconium content up t o a maximum f o r an a l l o y containing 0.53s Zr. The a l l o y s containing less zirconium had b e t t e r f r a c t u r e s t r a i n s under these test conditions; however, a l l alloys containing after i r r a d i a t i o n . given i n Fig. 48.

0.15

Zr or more had excellent fracture s t r a i n s

The minimum creep r a t e s f o r these same samples a r e The creep rates a t both s t r e s s levels exhibited a

d e f i n i t e minimum f o r t h e a l l o y containing 0.53s Zr.

The creep r a t e s f o r

t h e unirradiated control samples are a l s o shown i n Fig. 4 8 a t t h e

32,400 p s i stress level.

The minimum creep r a t e was not affected detect-

ably by i r r a d i a t i o n except f o r t h e a l l o y s containing 0.75 and 1.18%Z r . A second i r r a d i a t i o n experiment exposed samples t o a thermal-neutron

fluence of 9 X

loi9

neutrons/cm2 a t 50째C.

These r e s u l t s are summarized

65 ORNL-DW

6944373R

6’

-ae

\ W

n W

6 2

0.4 0.6 0.0 4.0 ZIRCONIUM CONTENT ( W t 70)

0.2

4.2

4.4

Influence of Zirconium Content on the Creep Rates a t 650°C.

Fig. 48.

A l l samples annealed 1 hr a t 1177°C before irradiation a t 650°C.

i n Table 23.

Because of the different s t r e s s levels involved, the

r e s u l t s a r e d i f f i c u l t t o compare with those presented previously.

How-

ever, the following points can be established by cross-plotting results: Table 23.

Creep Properties a t 650°C of Irradiateda Alloys That Contain Various Amounts of Zirconium

Alloy Number

Specimen Test Number Number

Zirconium Content

(wt

Stress (psi)

Rypture Life

(hr)

Minimum Creep Elongation ( $1 Rate ($/hr)

x 103 100

108 108 109 109 110 111 112 112

1911 2080

R-257 R-283

< 0.03

2059

R-285

0.28

2110 2136 2161

R-292 R-293 R-307

0.53 0.75

0.10

1.18

40 40 55b

47b 55 47 47 47 55b

68.9 314.9 5.1 158.6 43.0 453.3 7l4.6 905.7 74.5

0.019 0.014 0.96 0.010 0.116 0.0063 0.0076 0.0028 0.059 ~

1.79 8.03 5.05 2.42 5.26 73 2 15.5 5.32 5.24

a Annealed 1h r a t 1177°C before irradiation a t 50°C t o a thermal fluence of 9 x 1 0 1 ~neutrons/cm2-.

bLoad increased during t e s t .

66 (1)t h e rupture l i f e increased and t h e creep r a t e decreased continuously

with increasing zirconium (Table 23); (2) t h e rupture l i v e s of these

samples were greater than those of samples i r r a d i a t e d a t 650°C (Table 22);

(3) t h e minimum creep rates were lower than those observed before irradiation (Fig. 45); and (4)t h e f r a c t u r e s t r a i n s were a l l q u i t e good. Tests were run a t 760°C on several samples; the r e s u l t s a r e shown i n Table 24. For samples irradiated and t e s t e d a t 76OoC,t h e minimum creep rate decreased with increasing zirconium (Fig. 49). The f r a c t u r e s t r a i n seemed t o go through a ntzXirmrm i n alloys t h a t contained 0 . 1 t o

0.28% Zr.

Test samples of a l l o y 108 (0.1% Zr) indicated t h a t t h e irra-

diation temperature had very l i t t l e e f f e c t when t h e t e s t temperature was

760°C. A t e s t sample of t h i s same composition a l s o indicated that annealing a t 650°C after i r r a d i a t i o n a t 43°C had very l i t t l e e f f e c t on t h e properties. A s i n g l e sample of a l l o y 112 (1.18%Z r ) was irradiated a t 760°C and t e s t e d a t 650°C. "he rupture life a t 30,000 psi was only

444.9 h r compared w i t h more than 675,hr f o r a sample stressed a t 32,400 p s i after i r r a d i a t i o n a t 650°C (Table 22).

However, t h e f r a c t u r e -

s t r a i n was q u i t e good f o r both samples. Two other alloys were made t h a t contained a nominal l$Zr.

These

two alloys, designated 146 and 147, a l s o contained 0.1 and 1% Re, respect i v e l y , an addition t h a t was made i n an attempt t o improve t h e weldability.

These alloys were creep t e s t e d i n the i r r a d i a t e d and unirradiated

conditions; t h e r e s u l t s are given i n Tables 25 and 26, and t h e s t r e s s rupture properties are shown graphically i n Fig. 50.

The stress-rupture

properties of the unirradiated samples a r e about t h e same f o r both heats and quite similar t o those shown i n Fig. 44 f o r a l l o y 108 (1.18%Zr).

Thus, rhenium does not appear t o be a very i n f l u e n t i a l addition; the properties are b a s i c a l l y those of an a l l o y containing 1$Zr. The single t e s t of an unirradiated sample of each heat t h a t had been aged f o r 1000 h r a t 760°C indicates an i n s t a b i l i t y i n t h i s alloy. This anneal decreased t h e rupture l i f e and increased t h e minimum creep r a t e .

Irra-

d i a t i o n a t 650°C reduced t h e rupture l i f e and f r a c t u r e s t r a i n when samples were tested a t 650°C (Fig. 50).

I r r a d i a t i o n a t a higher temper-

a t u r e further reduced rupture l i f e , but t h e f r a c t u r e s t r a i n remained reasonably high.

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

rnprnt.nr~.

Neutron

x 100 108

108 108 109

3791 2084 2093 3861 3871

R-455 R-284,, R-291

R-448 R-456'

< 0.03

657

0.10 0.10 0.10 0.28

43 750 750

760 760 760 760 760

1020

2.5 0.9 0.9 2.5 2.5

110

3881

R-447'

0.53

760

2.5

ll1

3891

R-459'

0.75

740

760

2.5

112

3901

R-457'

1.18

740

760

2.5

112

3910 ~

R-8U

1.18

760

aAnnealed 1 hr at 1177째C before irradiation. bAnnealed 1000 hr at 650째C after irradiation. C Stress increased during test as indicated.

650

2.6

Rupti

x 103 12.5 15 I5 12.5 12.5 I5 17.5 12.5 I5 17.5 12.5 17.5 12.5 I5 17.5 30

1.0

468.1 380.9 665.2 1009.6 310.4

7.9 1007.4 262.4 167.6 1296.3 308.6 1008.8 310.4 218.4 444.9

0.10 0.037 0.049 0.026 0.0094 0.062 0.37 0.0084 0.014 0.038 0.0047 0.043 0.0013

0.013 0.041 0.017

0.21 36.1 28.3 35.3

14.4 23.8 3.9 11.1 3.8 16.6

10.4 19.2 2.5 4.2 12.4 10.1

68 ORNL-DWG 69-(4372R

404 5

zw 10-2 5 K a w

$ 40-3 z

0.2

0.4

0.6

0.8

4.2

4 .O

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 irradiated at 760°C to a thermal fluence of 2.5 x lo2’ neutrons/cm2.

ORNL-DWG 69-44371

--5

TO 30%

I-

4T08%!

OPEN SYMBOLS-UNIRRADIATED CLOSED SYMBOLS-IR

0 (00

103

40’

104

RUPTURE L I F E (hr)

Fig. 50. Stress-Rupture Properties at 650°C of Alloys U.6 and 147. A l l samples annealed 1 h r at 1177°C before testing. Numbers by each point designate the irradiation temperature.

. hd

Table 25.

Specimen Number

Test

Creep Properties of Irradiated and Unirradiated Alloy 146

Anneala Before

Number Irradiation

Temperature, "C Irradiation Test

Thermal Neutron Fluence ( neutrons/cm2 )

x 1020 3515 6647 I21 35 16 6648 121 35 17 6649 I21 3504 6136 I21 3503 6135 121 35 11 7769 148 3518 6650 321 35 19 6651 121 3509 R-678 121 35 10 R- 656 I21 3508 R- G8 121 3522 R-915 121 3523 R- 916 121 3526 R- 928 121 35 12 R-801 121 3528 R-935 I21 3513 R- 830 121 3506 R-552 121 a Annealing designations:

650 650 650 696 721 696 760 721 760 760

650 650 650 650 650 650 760 760 650 650 650 650 650 650 650 650 650 760

2.6 2.6 2.6 2.6 2.6 2.6 2.6 2.6 2.6 2.4

Stress

Rupture Life

Elongation

(4)

Reduction i n Area ( 46)

x io3 70 63 55 47 40 47 25 20 55 47 47 47 47 40 40 35 30

36.1 90.1 311.3 865. 5b 1003.2 110.5 352.2 984.1 38.4 177.4 42.7 7.9 2.7 153.5 59.0 226.3 306.3 380.1

121 = annealed 1 hr a t 1177OC, 148 = annealed 1 h r a t 1177째C plus 1000 h r a t 760째C.

bTest- discontinued before failure.

Creep

0.U 0.037 0.012 0.0049 0.0023 0.16 0.016 0.0039 0.043 0.021 0.046 0.20 1.0 0.024 0.071 0.022 0.016 0.0211

29.6 24.9 20.1 22.3 5.4 20.9 32.2 29.1 2.8 4.9 2.3 1.8 3.0 4.9 4.8 7.2 5.4 10.6

24.7 22.2 20.6 17.2 2.4 15.8 32.8 51.9

Table 26.

Specimen Test Number Number

Creep P r o p e r t i e s of I r r a d i a t e d and Unirradiated Alloy Ut7

Anneala Before Temperature, "C Irradiation Irradiation Test

ThWItEi.1

Neutron Fluence (neutrons /an2 )

x 4031 6652 I21 4032 6653 I21 6654 4033 121 4045 6116 I21 6117 4044 I21 4038 7770 U8 6655 4034 I21 4035 6656 I21 4028 R- 679 I21 . 4030 R- 687 I21 4027 R- 647 121 4037 121 R-917 405 1 R-802 I21 4041 R- 934 121 405 2 R-832 I21 4049 R-553 121 a Annealing designations:

650 650 650 727 760 727 760 685

650 650 650 650 650 650 760 760 650 650 650 650 650 650 650 760

1020

2.6 2.6 2.6 2.6 2.6 2.6 2.6 2.4

Stress (Psi)

Rupture Life (hr)

($1

Reduct ion i n Area

($1

x 103 70 63 55 47 40 47 25 20 55 47 47 47 40 35 30 15

51.8 106.2 461.1 U 8 9 . 5b U7.8 53.8 453.3

mo.1 65.2 508.1 164.2 15.0 136.4 274.1 804.4 1082.1

121 = annealed 1hr a t 1177"C, = annealed 1hr a t 1177째C plus 1000 hr a t 760째C. bTes t discontinued before failure.

Minimum Creep Elongation

0.069 0.038 0.10 0.0054 0. o o u 0.19 0.018 0.0039 0.0089 0.0068 0.0095 0.29 0.027 0.017 0.0078 0.011

26.2 23.3 22.1 25.3 4.2 12.7 55.4 29.2 1.4 4.8 3.2 3.5 4.3 5.8 8.2 15.6

20.0 16.6 17.8 21.5 2.6 u.5 37.7 34.1

The two 100-lb commercial melts, heat 21555 (0.05% Z r ) and 21554

(0.35% Z r ) were evaluated i n both t e n s i l e and creep tests.

The t e n s i l e

results a r e shown i n Tables 27 and 28. The t e s t s on heat 21555 a f t e r i r r a d i a t i o n showed t h a t t h e coarse-grained material had suqerior ductili t y and that t h e fine-grained material had higher yield strength.

The

t e n s i l e results f o r heat 21554 (0.35% Z r ) a r e summarized i n Table 28, and t h e f r a c t u r e s t r a i n f o r the fine-grained material i s sham i n Fig. 51. The behavior i s similar t o t h a t observed f o r standard Hastelloy N, but a t 760°C the d u c t i l i t y of t h e zirconium-modified a l l o y was much lower. The strong dependence of f r a c t u r e s t r a i n on s t r a i n r a t e a t 760°C was a l s o quite unusual where the fracture s t r a i n was higher a t t h e lower s t r a i n rate.

After irradiation, t h e f r a c t u r e s t r a i n of t h e fine-grained

material was generally reduced, and t h e reduction was larger a t higher t e s t temperatures (Fig. 52).

The r a t i o s of the various properties i n

t h e i r r a d i a t e d and unirradiated alloys a r e s h m i n Fig. 53.

The yield

and ultimate t e n s i l e strengths were not affected appreciably by irradiat i o n . The elongation remained high with increasing temperature and dropped a t t e s t temperatures above 600°C. The high r a t i o a t 760°C i s due t o t h e very low f r a c t u r e s t r a i n of t h e unirradiated sample.

The

reduction i n area follows t h e normal p a t t e r n f o r standard Hastelloy N. Further analysis o f t h e r e s u l t s i n Table 28 shows the e f f e c t s of various annealing treatments.

Several samples were annealed f o r 100 hr

a t 871°C followed by 1000 hr at 650°C. The strengths were lower and t h e f r a c t u r e s t r a i n s higher than those of t h e samples held a t 650°C f o r a

much longer time.

Two samples were annealed f o r 1 h r a t 1177°C and then

f o r 1000 h r a t 760°C. This treatment gave good strength and d u c t i l i t y characteristics.

Some samples were i r r a d i a t e d t h a t . h a d heat treatments

other than those already discussed.

The s t r a i n s a f t e r i r r a d i a t i o n were

higher f o r samples annealed f o r 1hr a t 1177°C than f o r those annealed

a t 871°C. Irradiated and unirradiated samples of heats 21555 (0.05% Z r ) and s

2l554 (0.35% Zr) were t e s t e d i n creep. The creep r e s u l t s a t 650°C f o r heat 21555 a r e given i n Tables 29 and 30, and the stress-rupture propert i e s a r e sham i n Fig. 54.

In t h e unirradiated conditions, the optimum

rupture l i f e was obtained a f t e r a 1-hr anneal a t 1177°C. Further aging

Table 27.

Specimen Before Number Irradiat ion

Tensile Properties of Irradiated and Unirradiated Alloy 21555 (0.05$ Z r )

Temperature, "C Irradiation Test

Neutron Fluence (neutrons/c&)

Strain Rat e (min-l)

x 1020 2488 2493 2494 2496 2478 2479 2480 2481 2464 2465

36 36 36 36

U.8 U.8 121 121 16 16

540 540 540 540

600 650 650 700 25 650 650 650 650 650

2.5 2.5 2.5 2.5

0.002 0.05 0.002 0.002 0.05 0.002 0.05 0.002 0.05 0.002

psi

Elongation

($1

Tensile

x 103

24.4 24.4 25.2 22.8 43.9 27.3 26.6 27.1 36.8 37.7

4.1 74.2 66.8 54.6 116.1 66.9 63.6 60 60.2 48.2

50.2 55.0 33.0 20.2 54.3 30.7 43.3 27.5 11.3

52.5 59.0 42.6 40.8 56.9 33.7 45.0 28.3 11.9 6.0

5.8 ~~

a Annealing designations:

36 = 148 = 121 = 16 =

100 h r a t 871째C plus 1000 h r a t 65OoC, 1 h r a t 1177째C plus 1000 h r at 760"C, 1 h r a t 1177OC, 100 h r a t 871째C.

Reduct ion i n Area

~~~~~~~

37.6 40.1 35.2 41.3

44.2 31.5 33.8 20.5 12.6 7.9

73 Table 28.

. .

Tensile Properties of Irradiated and Unirradiated Alloy 21554 (0.35% Z r )

Anneala ThermalStrain Tewerature, "C Neutron Specimen Before Rate Yield Irradiation Test Fluence Number Irradiation (neutrons/cm2) (min-l)

x 1020

. c

121 36 36 36 36 36 36 u8 u8 16 16

9ll2

9l31 9I.U 9U2 9ll5 91U 9ll6 9121 9ll7 9122 9l.U 9123 9ll9 9124 9120 9125 2405

25 200 400 400 450 450 500 500 550 550 600 600 650 650 760 760 850 850 25 200

104 104 1oZ 104 104 104 104 104 104 104 104 104 104 104 104 104 104 104 121 121 I21 121 121 121 121 121 I21 121

5543 5544 5545 5553 5546 5554 5547 5555 5548 5556 5549 5557 5550 5558 5551 5559 5552 5560 10094 10095 10096 10097 10098 10099 10103 10100 10104 10101 10102 10105 2431 2436 2534 2437 2438 2439 10092 10093 9llo 9lll

400

I21

16 16 16 16 l6 16 l6 16 l6

16 16 16 16 l6 16 121 121 16 l6

650 650 650 650 650 650 650 650 650 -650 650 650 650 650 650 650 650 650 570 575 550 550

500 550 600 600 650 650 700 760 760 600 650 650 650 650 700 25 650 25 200

4.1 4.1

400

4.1

400 450 450 500 500 550 550 600 600 650 650 760 760 850 850 650 650 650 650

4.1 4.1 4.1 4.1 4.1 4.1 4.1 4.1 4.1 4.1 4.1 4.1 4.1 4.1 4.1

2406 2395 2396 a Annealing designations: 16 = 36 = l48 = 121 = 104 =

2.5 2.5 2.5 2.5

x 103 67.7 60 54.7 54.2 53 55.9 51.4 58.6 58.7 59.8 53.4 56.8 46.4 55.8 43.7 37.9 36.9 22 37.5 33.2 29.8 25.6 26.9 27.4 26.2 25.7 25.9 25.8 24.7 24 39.8 26.5 26 27.2 25.5 25.2 47.3 30.1 64.2 61 0.05 52.7 0.002 64.6 0.05 56 0.002 57.4 0.05 56.9 0.002 50.5 0.05 60.2 0.002 43.6 0.05 52.7 0.002 50.3 0.05 0.002 48.1 0.05 42.4 0.002 37.6 0.05 34.8 0.002 3.8 0.05 28.6 0.002 30.5 0.05 36.3 0.002 38.7 0.05 0.05 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.05 0.05 0.05 0.05 0.05 0.002 0.05 0.002 0.05 0.05 0.002 0.002 0.05 0.05 0.002 0.002 0.002 0.05 0.002 0.05 0.05

Psi

x 103 128.7 LU.9 109.2 111.8 107.2 l10.5 106.9 lll.4 106.9 101.3 102 91 87.5 75.5 59 37.9 37.7 22 102.2

Elongation, k Reduction Uniform Total in Area ( $1

42.3 39.9

47.0 42.9

40.7

44.4

45.8 41.3 42.7 42.5 39.9 38.7 26.4 37.0 22.7 27.6 U.3 4.0 1.6 7.5 100.0 72.6 64.0 71.8 79.9 75.8 74.3 55 .7 65.8 45.5 47.0 38.0 ll.5 29.2 51.6 50.8 43.5 41.0 18.7 53.5 37.3

48.3 44.3 48.1 44.8 45.3 43.2 28.0 30.3 25.0 30.4 46.6 8.1

47.4

91.7 94.1 80.3

43.7 42.5 39.5 39.8 41.8 20.4 43.4 44-5 22.5 31.0 15.7

56.5 46.9 76.3 65.3 77.8 83.7 78.3 78.0 57.1 68.5 46.5 48.7 40.2 33.5 46.8 60.6 54.4 49.6 45.7 34.4 55.1 47.3 47.5 46.0 46.2 45.2 42.3 45.7 43.4 45.6 47.5 23.0 32.0 16.8

65.6 55.4 37.9 38.6 12.1 63.1 56.6 60.2 50.9

10.9 6.6 1.7 3.1 5.1 43.0 28.4 12.0 6.3

ll.2 6.8 5.4 6.4 10.4 48.8 32.6 12.4 6.5

100.1 96.8 84.4 89.8 89.1 81 81.2 72.2 72.5 66 45.5 81 71.3 78.3 69.1 65.5 56.7 ll4.5 68.9 122.1 l16.1 109.1 l12.6 u0.2 107.9 109.9 l19.7

107.7

100 hr at 871"C, 100 hr a t 871°C plus loo0 hr a t 650°C, l h r at 1177°C plus loo0 hr at 760°C, 1 hr at 1177"C, 100 hr at 871°C plus 5600 hr at 650°C.

44.1

56.33 59.80 46.82 53.59 54.56 48.26 49.79 43.69 42.73 24.70 31.44 24.18 31.88 63.15 64.27 69.45 69.70 60.94 62.6 46.3 61.8 50.7 54.5 49.6 47.2 44.1 32.6 38.9 23.6 29.9 48.5 42.5 35.1 31.4 39.2 31.2 41.7 42.9 39.08 51.90 2.96

36.26 33.24 33.94 33.48 25.54 36.52 21.M 28.09 18.42 25.54 12.39 10.89

4.74 6.27 13.88 31.9 34.0 14.3 8.8

CRNL-DWG 68-776t

1 0 0 2 0 0 3 0 0 4 0 0 5 0 0 6 0 0 7 0 0 8 0 0 9 0 0 TEST TEMPERATURECCI

Fig. 51. Variation of Ductility with Temperature f o r Control Specimens of Zirconium-Modified Hastelloy N (Heat 21554). Annealed 100 h r a t 871째C and aged 5600 hr a t 650째C i n barren s a l t .

ORNL-DWG 68-7762

60 50

-t9 I-

ds 30 w

2 20

P 10

0 0

100

200

300

400

500

600

700

800

900

TEST TEMPERATURE ("C)

Fig. 52. Variation of D u c t i l i t y with Temperature f o r Molten S a l t Reactor Experiment Surveillance Specimens of Zirconium-Modified Hastelloy N (Heat 21554). Thermal fluence was 4 . 1 X lo2' neutrons/cm2. Samples annealed 100 hr a t 871째C before i r r a d i a t i o n over a period of 5600 hr a t 650째C.

Table 29. /

Specimen, Number

Test

Creep Properties of Unirradiated Alloy 21555 (0.05% Z r )

Anneala Before Test

Test Temperature ("C)

Stress

Rupture Life

(psi1

(hr)

Minimum Creep

Elongation ( 4)

Reduction in Area

( 8)

x 103 2500 2506 2506 2507 2501 2509 2498 2497 2461 2459 2460 2492 2490 2505 2499 25 12

6312 6313 63U 6315 6316 6317 5863 6 163 7669 7492 7668 5859 6 164 7282 6357 7281

'Annealing

121 121 121 121 121 121 37 37 U8

U8 148 36 36 121

121 121

designations:

65 0 65 0 650 650 650 650 650 650 65 0 650 650 650 65 0 760 760 760

70 62 55

47 40 35 40 32.4 62

47 40 47 32.4 30 20 15

3.5 17.0 62.5 231.3 451.5 1115.6 511.2 1673.8 3.4 77.1 272.9 77.9 1028.9

14.4 183.8 527.3

1.49 0.39 0. 12 0.026 0.0085 0.0065 0.0153 0.0050 3.59 0.24 0.062 0.34 0.028 0.64 0.087 0.023

121 = annealed 1h r a t 1177"C, 37 = annealed 1 h r a t 1177째C plus 1000 h r a t 65OoC, 148 = annealed 1 h r a t 1177째C plus 1000 h r a t 760"C, 36 = annealed 100 h r a t 871째C plus 1000 h r a t 650째C.

37.3 32.5 28.1 39.7 34.7 31.8 41.5 34.1 31.4 33.8 29.8 60.0 65.9 38.4 34.6 26.8

36.1 35.3 26.1 33.2 32.0 32.8 37.7 23.1 38.3 37.8 37.5 56.3 51.0 39.1 27.1 26.9

76 ORNL-DWG

68-7763

4.2

0.2

TOTAL ELONGATION

REDUCTION I N AREA

STRAIN RATE

I \

o YIELD STRESS 0. U L T I M A T E S T R E S S

400

200

700

800

900

= 0.05 min-'

300

400

500

600

T E S T T E M P E R A T U R E (OC

Fig. 53. Comparison of the Tensile Properties of Irradiated and Unirradiated Zirconium-Modified Hastelloy N (Heat 21554). Thermal fluence was 4.1 X lo2' neutrons/cm2. Annealed 100 hr a t 871°C and aged or irradiated over a period of 5600 h r a t 650°C. Tested a t a s t r a i n r a t e of 0.05 min-l.

Table 30.

Specimen Number

Creep Properties of Irradiateda Alloy 21555 (0.05% Z r )

Test Number

Annealb Before Irradiat ion

Stress (Psi)

Life (h)

Minimum Creep

Elongat ion

($1

x 103 Test Temperature 650°C 2483 2482 2467 2466 2469

R-381 R-259 R- 349 R-268 R-350

121 I21 16 16 16

47 40 47 40 35

106.4 410.9 0 20.0 45.7

0.032 0.0089

6.7 10.0

0.11 0.032

2.6 3.0

Test Temperature 760°C 0.036 2485 I5 110.1 R-227 I21 0.0095 2484 349.5 121 12.5 R-355 0.0096 16 15 R- 237 34.2 2472 0.027 2471 12 123.1 16 R-351 a Irradiated a t 540°C t o a thermal neutron fluence of 2.5 neutrons/cm2. bAnnealing designations:

7.2 7.7 5.0 6.8 X

lo2' I

I21 = annealed 1hr a t 1177"C, 16 = annealed 100 hr a t 871°C. i

77 ORNL-DWG 69- 14370

b I

o ANNEALED I hr AT 4477°C A 0

ANNEALED I h r AT 1177OC. 1000 hr AT 65OOC ANNEALED ihr AT 1177"C1 1000 hr AT 760°C

0 ANNEALED 100 hr AT 874°C ,IO00 hr AT 65OOC OPEN SYMBOLS UNIRRADIATED CLOSED SYMBOLS IRRADIATED

n " IO0

Fig. 54.

$0'

102 RUPTURE TIME (hr)

lo3

to4

Stress-Rupture Properties a t 650°C of Alloy 21555 (0.05'$ Z r ) .

f o r 1000 h r a t 650°C produced no detectable effect, but aging f o r 1000 h r

a t 760°C decreased t h e time t o f a i l u r e .

Annealing a t 871°C t o produce a

f i n e grain s i z e resulted i n shorter rupture l i v e s .

Irradiated samples

had been annealed a t 1177 and 871°C before irradiation.

As shown i n

Fig. 54, i r r a d i a t i o n reduced t h e rupture l i f e of t h e l a r g e r grained mater i a l by a f a c t o r of about 2 and t h a t of the f i n e r grained material by a f a c t o r of about 10. The minimum creep rat

f heat 21555, shown

affected appreciably by irradiation.

i n Fig. 55, was not

The lowest minimum creep rate was

obtained by annealing a t 1177°C; aging a t 650°C had no effect; aging a t 760°C increased t h e creep rate; and annealing a t 871°C t o give a small grain s i z e produced a high creep rate. The r e s u l t s of creep t e s t s on heat 21554 (0.35$ Z r ) are given i n Tables 3 1 a n d 32. Fig. 56.

The stress-rupture praperties a t 650°C a r e shown i n

The same general trends with regard t o heat treatment were

observed f o r t h i s heat and f o r t h e heat with lower zirconium content (heat 21555).

The optimum rupture l i f e occurred a f t e r a 1-hr anneal a t

. -

ORNL- DWG 69 (4369

v) v)

I hr AT I177OC

ANNEALED 1 hr AT 4177OC. lo00 hr AT 65OOC 0 ANNEALED 1 hr AT 1177OC. 1000 hr AT 760OC 0 ANNEALED 1 0 0 hr AT 871OC ,1000 hr AT 65OOC OPEN SYMBOLS UNIRRADIATED CLOSED SYMBOLS IRRADIATED

0 to-3

Fig. 55.

10-2

IO-’ MINIMUM CREEP RATE (’70/hr)

100

10’

Creep Rates at 650°C of Alloy 21555 (0.05$ Zr).

Table 31.

Specimen Number

Creep Properties of Unirradiated Alloy 21554 (0.35% Z r )

Test Number

m Before eala Test

Stress (psi)

Rqture Life

Minimum Creep Elongation Rate

Reduct ion in Area

(5)

0.03l.3 0.20 0.13 0.056 0.018

49.4 38.1 31.6 26.8 19.7 20.5 54.9 46.0 36.5 63.0 60.8 35.5 55.1

43.8 43.1 26.7 27.7 28.0 12.8 41.8 44.0 33.4 50.6 50.9 51.9 42.4

8.38 3.20 1.01 0.37 0.12 0.36 0.11 0.032

60.6 50.0 55.8 38.3 69.5 75.1 63.1 56.1

48.8 50.1 45.0 58.3 56.6 56.0 50.0 50.8

30.3 32.9 28 .? 61.7

25.9 30.7 31.3 45.0

x 103 Test Temperature 650°C 2442 2443 2444 2445 2446 2441 10086 10084 10085 2435 2491 2434 2433

6306 6307 6308 6309 6310 5913 7667 7491 7666 5862 5858 5857 6168

I21 121 121 121 121 37 148 .U 8 148 36 36 36 36

70 62 55 47 40 40 62 47 40 47 40 40 32.4

5535 5536 5537 5538 5539 5540 5541 5542

6399 6398 6397 6389 6390 6425 6426 6427

104 104 104 104 104 104 1% 104

70 63 55 47 40 47 40 32.4

7.3 32.0 94.4 395.2 803.8 627.7 5.5 157.5 597.4 128.3 226.3 324.2 1390.9 3.3 8.3 28.2 58.1 228.7 84.1 244.0 709.8

1.38 0.20 0.062 0.018 0.0071 1.38 2.45 0.u

Test Temperature 760°C 2421 2415 2427 2404 a

7286 6356 7285 6056

121 121

121

Annealing designations:

30 20

35.6 354.1 1073.0 85.0

0.33

0.W7 0.0123 0.35

l21= annealed 1 h r a t 1177"C, 37 = annealed 1hr a t 1177°C plus 1000 hr a t 650"C, 148 = annealed 1hr a t 1177°C plus 1000 hr a t 760°C, 36 = annealed 100 hr a t 871°C plus 1000 hr a t 650"C, 104 = annealed 100 hr a t 871°C plus 5600 hr a t 65OoC, 16 = annealed 100 h r a t 871°C'.

Table 32.

Specimen

Tes$

Number

Creep Properties of Irradiated Alloy 21554 (O.35$ Z r )

Anneala Before Irradiation

Irradiation Temperature ("C)

ThermalNeutron Fluence ( neutrons/cm2 )

Stress

x 1O2O

x 103

Rupture Life

Minimum Creep

Elongation

($1

Test Temperature 650°C

2424 2407

R- 382 R-479 R-260

2420 2398 2397 2402 9127 9126 9128 9129

R-265 R-269 R-356 R- 316 R-311 R-318 R-322

2408

R-480

121 121 121 121 16 16 16 61 61 61 61

590 650 590 650 550 550 550 650 650 650 650

2.5 2.5 2.5 2.5 2.5 2.5 2.5 4.1 4.1

4.1 4.1

47 47 40 40 47 40 35 47 40 32.4 27

u0.2 111.3 336.3 279.0 3.9 76.7 163.7 11.1 19.3 65.4 204.9

0.024 0.034 0.0036 0.0032 0.35 0.051 0.0068 0.34 0. l.4 0.013 0.0088

4.75 5.05 2.50 12.6 1.51 4.74 3.43 4.57 3.99 2.36 3.45

20 I5 l5 20

71.5 264.1 9.1 5.5 13.3 222.8

o.ol.4 0.0094 0.092 0.36 0.10 0.0092

1.95 6.52 1.24 2.95 2.62 5.62

Test Temperature 760°C 2409 2410 2412 2403 2401 2400

R-364 R-239 R-470 R- 174 R-236 R-357

I21 121 121 16 16 16

590 590 760 760 566 560

2.5 2.5 2.5 2.3 2.5 2.5

I5 12.5

a Annealing designations: 121 = annealed 1hr a t 1177°C; 16 = annealed 100 h r a t 871°C; 6 1 = annealed 200 h r a t 871°C.

03 0

81 ORNL-DWG 69- 14368

bi c

AT CC77OC, COO0 hr AT 65OOC AT C177OC. 1 0 0 0 hr AT 760°C 0 ANNEALED 100 hr AT 87CoC, 1 0 0 0 hr AT 65OOC 0 ANNEALED 100 hr AT 87C°C, 5600 h r AT 65OOC OPEN SY M EOLS UNIRRADIATED CLOSED SYMBOLS-IRRADIATED

0 COO

Fig. 56.

102 RUPTURE TIME (hr)

Stress-Rupture Properties a t 650°C of Alloy 21554 (0.35% Zr).

1177°C. I r r a d i a t i o n reduced t h e rupture l i f e of t h e material annealed a t 1177°C by a f a c t o r of 2 t o 3 and t h a t of the fine-grained material by a f a c t o r of about 10.

The variation of t h e minimum creep r a t e with heat

treatment and i r r a d i a t i o n i s sham i n Fig. 57.

The lowest minimum creep

r a t e was observed f o r the material annealed a t 1177°C. i r r a d i a t i o n on t h e minimum creep rate was evident.

No e f f e c t of

The f r a c t u r e s t r a i n s are shown as a f%mction of minimum creep r a t e

i n Fig. 58 f o r material annealed a t 1177°C and tested i n t h e unirradiated condition a t 650°C.

The r e s u l t s f o r heat 21554 (0.35% Z r ) show a c l e a r

trend of decreasing f r a c t u r e s t r a i n with decreasing creep rate.

The

results f o r heat 21555 (0.05% Zr) show a similar trend, but the data seem t o f a l l on two separate l i n e segments. A l l of t h e f r a c t u r e s t r a i n s are quite high and f a l l largely i n t h e range of 20 t o 40%. The results i n Tables 29 and 31 show t h a t t h e samples of both alloys t h a t were annealed a t 871°C t o give a f i n e grain s i z e had f r a c t u r e strains of 40 t o 60%.

82 ORNL-OWG 69-1'

W i

2oll 0 ANh A ANNEALED I hr AT'l177"C. 4000 hr AT 650% 0 ANNEALED lhr AT 1177.C. IO00 hr AT 760% 0 ANNEALED 1 0 0 hr AT 87PC. I000 hr AT 650% 0 ANNEALED IOOhr AT 871OC. 5600 hr AT 65OOC

0 10-3

OPEN SYMBOLS-UNIRRADIATED CLOSED SYMBOLS-IRRADIATED

10-2

IO-' MINIMUM CREEP RATE, %/hr

I00

10'

Fig. 57. Creep Properties a t 650째C of Alloy 21554 (0.35% Zr).

ORNL-DWG 69-14366

5 30

Li w

ga 20

E IL

n 40-2

16' MINIMUM CREEP RATE (%/hr)

Fig. 58. Fracture Strains a t 650째C of Alloys 21554 (0.35% Z r ) and 21555 (0.055 a). A l l samples annealed 1hr a t 1177째C before t e s t i n g .

bj z

The f r a c t u r e s t r a i n s of i r r a d i a t e d samples of heats 21554 and 21555 a r e shown i n Fig. 59 as a function of s t r a i n rate.

The r e s u l t s generally

follow a trend of increasing f r a c t u r e s t r a i n with decreasing creep r a t e , although one point d e f i n i t e l y deviates from t h i s trend.

I W

a ?

21554 A 21555 0

ORNL-DWG 69-t4365

I I

0 10-~

lliiL 10-2

10-~

10-1

MINIMUM CREEP RATE (%/hr)

Fig. 59. Fracture Strains a t 650°C of Alloys 21554 (0.35% Z r ) and 21555 (0.05% Zr) after Irradiation. A l l samples annealed 1hr a t 1177°C before i r r a d i a t i o n . The results f o r a t e s t temperature of 760°C are fragmentary (Tables 29 thrhugh 32).

The stress-rupture c h a r a c t e r i s t i c s a r e s h a m

i n Figs. 60 and 6 1 f o r heats 21555 (0.05% Zr) and 21554 (0.05% Zr). Rupture l i f e was reduced by irradiation, but t h e samples annealed a t 1177°C before i r r a d i a t i o n had longer rupture l i v e s than those annealed

a t 871°C.

The f r a c t u r e s t r a i n s were reduced by irradiation, and t h e

r e s u l t s f o r heat 21554 (Table 32) indicate t h a t t h e reduction was greater t

as t h e i r r a d i a t i o n temperature increased. r e examined metallographically. Several t e s t e d sample

Typical

photomicrographs of several alloys t h a t contained zirconium are shown i n Fig. 62.

These samples were annealed 1 h r a t 1177"C, aged 1000 hr

ORNL-DWG 6944364

4 400

I I I111111to' I I I11111!02 1 I I I111111to3 I RUPTURE TIME (hr)

Fig. 60. Stress-Rupture Properties of Alloy 21555 (0.05% Z r ) at 760째C. 3 -.

ORNL-DWG 6944363

8 6 CLOSEDSYMBOLS IRRADIATED AT INDICATED TEMPERATURES

I I

1'111

too

IO'

102 RUPTURE TIME (hr)

103

Fig. 61. Stress-Rupture Properties of Alloy 21554 (0.35% Z r ) at 760째C.

a t 65OoC, and stressed f o r 1000 hr a t 650°C.

Alloy 108 (0.10% Zr) had

a very f i n e carbide p r e c i p i t a t e t y p i c a l of t h a t observed i n a l l o y 100 (Fig. 24), which did not contain zirconium.

I n a l l o y 109 (0.28% Zr) t h e

p r e c i p i t a t e was coarser and located predominantly along the grain bound-

aries.

Alloys ll0 (0.53% Zr) and lll (0.75% Zr) contained a very f i n e

p r e c i p i t a t e t h a t could not be resolved a t a magnification of 5 0 0 ~ . Alloy l l 2 (1.189 Zr) had an even higher concentration of p r e c i p i t a t e (Fig. 63).

Figure 63(c) shows that the p r e c i p i t a t e was acicular.

The

grain boundaries had a denuded zone, and t h e r e was a denuded region assoc i a t e d with t h e oxide a t t h e surface.

The p r e c i p i t a t e was extracted and

found by x-ray d i f f r a c t i o n t o have a l a t t i c e parameter of 4.68 A, equal t o t h a t f o r zirconium carbide.

We a l s o strained some samples of

a l l o y 112 a t 25°C a f t e r they had been aged a t 650°C t o produce t h e p r e c i p i t a t e . The f r a c t u r e s t r a i n s were i n excess of 50$, thus indicating t h a t t h i s microstructure was not b r i t t l e . Some of t h e aged samples were t e n s i l e t e s t e d a t 65OOC (Table 17). The f r a c t u r e of a l l o y 112 after t e s t i n g a t 650°C at a s t r a i n r a t e of

0.002 min-l i s shown i n Fig. 64.

The fracture s t r a i n was 23.3$,

and t h e

f r a c t u r e was transgranular. The f r a c t u r e of a l l o y 110 (0.53% Zr), which was t e s t e d i n creep a t

65OoC, i s shown i n Fig. 65.

This a l l o y was a t temperature 709 hr, and

small quantities of p r e c i p i t a t e are present i n t h e microstructure.

This

p r e c i p i t a t e cannot be resolved a t l o w magnification and appears as dark spots.

There a r e numerous intergranular cracks, and t h e f r a c t u r e i s

intergranular.

Typical photomicrographs of a l l o y 112 (l.B$ Zr) a f t e r

creep t e s t i n g a t 650°C a r e shown i n Fig. 66.

This sample was a t temper-

ature 1255 hr, and large quantities of p r e c i p i t a t e a r e present i n t h e microstructure.

There are numerous intergranular cracks, and t h e frac-

ture has both intergranular and transgranular components. Typical fractures of a l l o y 21555 (0.05% Zr) a f t e r t e s t i n g a t 650°C and 32,400 p s i are shown i n Fig. 67. When t h i s a l l o y was annealed 100 hr a t 650°C before testing, the grain s i z e was q u i t e small. This sample was strained 65.9% before it fractured (Table 29), predominantly transgranularly [Fig. 67(a)].

Annealing f o r 1hr a t 1177°C and t e s t i n g

i n creep resulted i n t h e microstructure shown i n Fig. 67( b)

The

.a.

U I

Fig. 64. Photomicrographs of Alloy 112 (1.18% Z r ) Annealed for 1 h r at ll77"C, Aged for 1000 hr at 65OoC, and Tensile Tested at 650째C at a Strain Rate of 0.002 min-l. (a) Fracture, 1OOX; (b) typical, 500x.

Fig. 65. Photomicrographs of Alloy 110 (0.53% 2 2 ) Annealed 1 hr at (a) Fracture, 1OOX; (b) typical, 5 0 0 ~ . Etchant: glyceria regia.

1 1 7 7 째 C and Tested at 6 5 0 째 C and 40,000 psi.

92 f r a c t u r e was s t i l l largely intergranular, and t h e f r a c t u r e s t r a i n was

31.8%. Annealing a t 1177°C and aging f o r 1000 hr a t 650°C resulted i n a f r a c t u r e s t r a i n of 34.1% (Table 29).

The microstructure contained

only a f e w small precipitates, and t h e fracture was s t i l l l a r g e l y transgranular.

Alloy 21554 (0.35% Z r ) was subjected t o a similar annealing

sequence.

After annealing f o r 100 h r a t 871°C and 1000 h r a t 650°C and

t e s t i n g a t 650°C t h e f r a c t u r e shown i n Fig. 68(a) was obtained.

This

a l l o y o r i g i n a l l y had stringers present (Fig. 38), and numerous cracks formed around these s t r i n g e r s during t e s t i n g .

The f r a c t u r e was trans-

granular, and t h e fracture s t r a i n was 55.1% (Table 31). 1 h r a t 1177°C produced a coarser grain size.

Annealing f o r

The f r a c t u r e shown i n

Fig. 68(b) was mixed intergranular and transgranular, and t h e f r a c t u r e s t r a i n was about 20%. The approximately lo00 hr a t temperature during t h e creep t e s t was s u f f i c i e n t t o produce some precipitation.

Another sample was annealed 1h r a t 1177°C and aged f o r 1000 hr before testing.

This sample failed with about 20% s t r a i n .

The f r a c t u r e was mixed trans-

granular and intergranular [Fig. 68(a)].

Copious quantities of very f i n e

p r e c i p i t a t e formed. Alloy 108 (0.10% Zr) was a l s o examined metallographically after

As shown i n Fig. 69, a very f i n e precipit a t e was formed t h a t was q u i t e similar t o t h a t sham i n Fig. 62 f o r an i r r a d i a t i o n and creep t e s t i n g .

unirradiated sample.

Alloy 110 (0.53% Zr) had more p r e c i p i t a t e a f t e r

i r r a d i a t i o n than t h e control sample (compare Figs. 70 and 62).

As shown

i n Fig. 70 there was a denuded zone along the boundaries and some banding of t h e precipitate, l i k e l y due t o inhomogeneous d i s t r i b u t i o n of zirconium i n t h e alloy.

The f r a c t u r e was largely intergranular.

Alloy 112

(1.18% Zr) had profuse quantities of t h e precipitate; t h e microstructure, however, was q u i t e similar t o t h a t f o r t h e unirradiated sample (compare Figs. 71 and 63).

The sample could not be etched heavily enough t o show

t h e grain boundaries without o b l i t e r a t i n g t h e p r e c i p i t a t e .

The f r a c t u r e

was mixed intergranular and transgranular.

Typical photomicrographs of a l l o y 21554 (0.35% Zr) are shown i n Figs. 72 through 75.

The sample shown i n Fig. 72 was irradiated and

t e s t e d a t 25°C and f a i l e d transgranularly a f t e r s t r a i n i n g 47.5$. comparable control sample i s s h a m i n Fig. 73.

The

This sample failed

Fig. 69. Photomicrographs of Alloy 108 (0.109 Zr) Annealed 1 h r at ll77"CY Irradiated at 650째C for 1000 hr to a Thermal-Neutron Fluence of 2.5 X lo2' neutrons/cm*, and Creep Tested at 65OOC. Stressed at 32 400 psi for 675 hr (8.25 strain) and 40,000 psi for 53 hr (8.3% strain). (aj Fracture, blunted end caused in handling, 1OOX; (b) typical, 500X. Etchant: glyceria regia.

r a t 1177"C, Irradiated a t 650째C and Creep Tested a t 650째C. 430 h r (5.9% Strairl). ( a ) Fracture, t : glyceria regia. Reduced 24$.

,..

. _ . . .

.... . .

._. ....

. .-.. -

... . ..__ ..- .

. .,_

Fig. 71. Photomicrographs of Alloy 112 (1.18% Z r ) Annealed 1 hr at 1177째C, Irradiated at 650째C for 1000 hr to a Thermal-Neutron Fluence of 2.5 X lo2' neutrons/cm*, and Creep Tested at 650째C. Stressed at 32,400 psi for 677 hr (5.8$'Strain) and 40,000 psi for 47 hr (3.9% Strain). (a) Fracture, 1OOX; (b) fracture, 500~;and (c) typical, lOOOX. Etchant: glyceria regia. Reduced 24%.

F i g . 73. Photomicrograph of t h e Fracture of a Zirconium-Modified Hastelloy N Sample (Heat 21554) Tested a t 25째C a t a S t r a i n Rate of 0.05 min'l. Exposed t o a static fluoride salt f o r 5500 h r a t 650째C before t e s t i n g . Note t h e shear f r a c t u r e and t h e absence of edge cracking. 100~. Etchant: glyceria regia.

100

Fig. 75. Photomicrograph Showing a Portion of t h e Fracture of a Zirconium-Modified Hastelloy N Sample (Heat 21554) Tested a t 650°C a t a S t r a i n Rate of 0.002 min’l. Exposed t o s t a t i c f l u o r i d e salt f o r 5500 hr a t 650°C before t e s t i n g . 1OOx. Etchant: glyceria regia. transgranularly after straining 47.0%.

An i r r a d i a t e d sample t e s t e d at

650°C is shown i n Fig. 74; it fractured intergranularly a f t e r 11.2% strain.

The comparable control sample, shown i n Fig. 75, strained 46.6%

before f a i l i n g .

There a r e numerous intergranular separations, and the

f r a c t u r e i s largely intergranular. Alloys Containing Hafnium

Two groups of alloys were used t o study t h e e f f e c t s of hafnium content on t h e mechanicalproperties.

The f i r s t group was composed of

f i v e laboratory melts that contained frm 0.06 t o 0.434 H f (designated alloys 152 through 156), and t h e second group consisted of two 100-lb commercial melts t h a t contained 0.08 and 0.509 H f (designated alloys 67-503 and 67-504, respectively).

The chemical compositions of

these alloys a r e given i n Table 1, and t y p i c a l photomicrographs are shown i n Figs. 76 through 80.

Alloy 153 (Fig. 76) contained only 0.06% H f

and had t h e l a r g e s t grain size of t h e group.

Alloys 154 (0.10’$H f ) and

Fig. 76. Photomicrographs of Alloys Modified with Hafnium Annealed 1 hP a t 1177째C. ( a ) Alloy 153 (0.06% Hf); (b) a l l o y I54 (0.10% Hf) and ( c ) a l l o y 155 (0.13% Hf). loox. Etchant: glyceria regia. Reduced 18%.

102

Fig. 77. Photomicrographs of Alloy 152 Modified w i t h 0.16% H f . Annealed 1h r a t 1177째C. ( a ) lOOX and ( b ) 5 0 0 ~ . Etchant: glyceria regia.

103

Fig. 78. Photomicr Annealed 1 hr at 1177째C. regia.

Alloy 156 Modified with 0.43$ Hf'. lOOX and (b) 500~. Etchant:

&ceria

Fig. 79. Typical Photomicrographs of Heat 67-503 (0.08$ Hf) Annealed 1 hr at 1177째C. Longitudinal section. (a) lOOX and (b) 5 0 0 ~ . Etchant: glyceria regia.

105

Fig. 80. Typlcal Ph 1h r a t 1177째C. Irongitud glyceria r e g i a !

rographs of Heat 67-504 (0.504 H f ) Annealed (a) lOOX and b) 500~. Etchant:

106 155 (0.134 H f ) had much smaller grain sizes.

Alloy 152 contained

0.164 H f , and the grain s i z e was refined by copious quantities of carbide precipitates a t grain boundaries.

When the hafnium content was increased

t o 0.434, as i n alloy 156, the precipitation became more general with precipitation a t grain boundaries and i n the matrix. laboratory melts. (Figs. 79 and 80). The r e s u l t s of creep-rupture t e s t s a t 650째C on unirradiated and irradiated samples of alloys 152 through 156 a r e given i n Tables 33 and The stress-rupture properties for the unirradiated

Table 33.

Stress Properties a t 650째C of Several Hafnium-Modified Alloysa

Minimum Specimen Test Stress Rrxpture Life Creep Elongation Number Number (psi) R a te (5) (hr) ($/hr)

Alloy Number

Reduction i n Area

(5)

x 103 152 152 152 152 152 153 153 I53 153 153 I53

2979 2964 2967 2965 2966 6554 a53 2864 2863 6545 6544 6559 6556 6567 6566 8495 2873 2874 6642 6643 2878 6650

I54 154 I54 154 155 I55 155 156 156 156 156 a

7506 6537 6538 6539 6540 7507 6658 6535 6536 6687 6686 6771 6689 7509 7508 7554 6531 6532 6692 6693 6694 7510

35 70 55 47 40 35 70 55 47 40 55, 70 40 55 47 70 55 47 70 55 47 35

1906.3 3.4 21.1 67.4 234.4 216.5 0.3 6.6 76.7 114.2 7.4 0.2 86.0 5.0 41.0 0 33.6 90.2 1.9 19.0 257.5 1210.5

0.0015

1.23 0. U 6 0.051 0.018 0.0050 7.69 0.194 0.023 0.013

0.223 3.5 0.023 0.188 0.0352 0.112

0.043 1.25 0.152 0.0214 0.Oll

The two commercial

alloys had larger grain sizes than t h e i r counterparts i n the series of

34, respectively.

15.0 23.4 14.1 12.2 11.7 9.3 41.8 20.4 16.2 11.9 22.6 32.0 7.8 19.7 13.1

48.4 U.0 13.2 22.4 14.7 18.5 22.2

A l l materials annealed 1 h r a t 1177째C before testing.

20.5 22.3 15.6 13.3 ll.4 11.3 29.3 17.4 18.2

11.8 18.0 26.5 5.5 24.8 18.7 37.3 17.9 U.0 22.0 11.9 15.6 20.3

107 Table 34.

Alloy Number

Creep Properties a t 650째C of Irradiateda Alloys Containing Various Amounts of Hafnium

Specimen Test Number Number

Hafnium Content Stress (h8 ) (Psi)

Rupture Life

(hr)

Minimum Creep Elongation Rate ($1

($/hr)

x 103

1916 2961 2963 2962 2860 2861 2862 2865 2867 2866 2872 2871 2870 2877 2875 2876

100 I52 I52 I52 153 I53 I53 I54 I54 I54 I55 I55 I55 I56 I56 I56

R- 34-6 R-373 R- 208 R-371 R-212 R-222 R-383 R-374 R- 188 R- 372 R- 200 R-221 R-238 R- 251 R- 190 R-213

0.16 0.16 0.16 0.06 0.06 0.06 0.10 0.10 0.10 0.13 0.13 0. I 3 0.43 0.43 0.43

32 ..4 47 40 32.4 40 35 32.4 47 40 32.4 40 35 35 47 40 35

105.4 10.7 87.3 459.5 35.6 202.2 224.8 7.1 41.5 138.3 63.1 221.9 350.9 80.8 395.1 1707.2

0.0099 0.240 0.044

0.0068 0.053 0.0065 0.010 0.253 0.0548 0.0115 0.0385 0.0080 0.0084 0.101 0.0277 0.0064

1.1 3.u 4.28 5.u 2.38 3.07 3.32 2.25 2.70 3.09 2.78 1.93 2.85 10.45 15.75 u.20

A l l materials annealed 1 h r a t 11'77째C before irradiation a t 650째C t o a thermal fluence of 2.3 X lo2' neutrons/cm*.

alloys a r e shown i n Fig. 81.

The r e s u l t s for alloy 100, which contained

no detectable hafnium, a r e used for comparative purposes.

The general

observation i s t h a t the rupture l i f e increased with increasing hafnium content up t o about 0.16s. The stress-rupture properties of the irradiated alloys shown i n Fig. 82 were even more dependent upon the hafnium content than those of the unirradiated material.

In t h i s condition,

alloy 156 (0.43s Hf) had far superior stress-rugture properties. The minimum creep rat sham i n Fig. 83.

The a l l

lloys 100 and 152 through 156 are

t contained hanfium had lower creep

rates, but the magnitudes showed no clear dependence upon the concentrat i o n between 0.06 and 0.43% Hf.

Irradiation caused a s l i g h t increase

i n the creep r a t e of mos

s; alloy 156 (0.43s Hf) seemed unaffected. The fracture s t r a i n s of the alloys i n the unirradiated condition a r e

shown i n Fig. 84.

Generally, the alloys followed the trend of decreasing

108 ORNL-DWG 69-t4362

ALLOY CONTENT (%)

NO.

40°

to3

io2

40'

4 O4

RUPTURE TIME (hr)

Fig. 81. Stress-Rupture Properties a t 650"C.of Several A l l o y s T h a t Contain Hafnium. All samples annealed 1 h r a t 1177°C before t e s t i n g . ORNL

-14361

'00

too

to'

'02

to3

4 o4

RUPTURE LIFE (hi)

Fig. 82. Creep-Rupture Properties a t 650°C of Irradiated A l l o y s That Contain Various Concentrations of Hafnium. A l l samples annealed 1 h r a t 1177°C before i r r a d i a t i o n a t 650°C.

109 OWL-DWG 69-14360

(X4031

-.-In"0

l n K w

Y NO.

Hf CONTENT

'04

0 40-3

'0-2

40 '

MINIMUM CREEP RATE (%/hrl

Fig. 83. Creep Rates a t 650째C of Alloys That Contain Various Concentrations of Hafnium. A l l samples annealed 1 h r a t 1177째C.

ORNL-DWG 69-44359

ALLOY NO.

100 152 153 i54 155 456

I-

io 5

~, 40'

io+ MINIMUM CREEP RATE (%/hr)

Fig. 84. Fracture Strains a t 650째C of Alloys That Contain Various Amounts of Hafnium. A l l samples annealed 1 h r a t 117'7째C before testing.

fracture s t r a i n with decreasing creep rate.

Alloy 156 (0.434 H f ) had

s l i g h t l y superior d u c t i l i t y , p a r t i c u l a r l y a t t h e lower s t r a i n rates. The f r a c t u r e s t r a i n s a f t e r i r r a d i a t i o n a r e shown i n Fig. 85. Alloy I56 (0.434 €E)had high f r a c t u r e s t r a i n s , and a l l o y 152 (0.16$ Hf) had frac-

W 1

ture s t r a i n s somewhat-improved over those of t h e other alloys. ORNL-DWG 69- I4358

9 6 4

io-' MINIMUM

CREEP

400

4 0' c

RATE (%/hr)

Fig. 85. Fracture S t r a i n s a t 650°C of I r r a d i a t e d Alloys That Cont a i n Various Amounts of Hafnium. A l l samples annealed 1 h r a t 1177°C before i r r a d i a t i o n a t 650°C. The commercial alloys, 67-503 (0.084 Hf) and 67-504 (0.50$ Hf) were subjected t o various t e s t s .

are given i n Table 35.

The r e s u l t s of t e n s i l e t e s t s on a l l o y 67-503

The results on i r r a d i a t e d t e s t s shared c l e a r l y

t h a t t h e anneal of 1 h r a t 1 1 7 7 ° C before t e s t gave superior f r a c t u r e s t r a i n s a t 650°C. (0.50% Hf').

MSRE.

More detailed t e s t i n g was carried out on heat 67-504

"his a l l o y was included i n t h e surveillance program f o r t h e

Samples were exposed t o a molten f l u o r i d e f u e l s a l t i n the core

f o r 9800 h r a t 650°C and received a thermal-neutron fluence of 5.3 X neutrons/cm*.

lo2'

Control samples were exposed t o s t a t i c barren s a l t f o r

9800 hr a t 650°C.

Some t e s t results on these samples a r e presented i n i

Table 35.

Tensile Properties of Irradiated and Unirradiated Alloy 67-503 (0.08$ H f ) Temperature,

Specimen Number

Before Irrad i a t ion

Irrad i a t ion ~~

Strain Stress, p s i

O C

Test

Neutron Fluence (neutrons/cm2)

Rate (min-1)

816 816 816

121 121 323 816 816 816 121 121 123

( $1

x 1020 6195 6196 6197 6193 6194 4999 6229 6234 4941 4974 4949 4966

Yield

Elongation, $ Reduction Ultimate ~ ~ Uniform ~ Total ~ in i Area l

650 650 760 760 650 760

a Annealing designations: 1hr at 126OOc.

650 650 760 650 760 760 650 650 760 760 650 760

2.0 2.0 2.4 2.4 2.4 2.4

0.05 0.002 0.002 0.002 0.002 0.002 0.002 0.05 0.002 0.002 0.002 0.002

io3

48 50.6 37.1 24.3 31.2 27.7 33.4 35.1 8.6 u.3 26.9 7.8

io3

92.4 74.5 37.8 55.8 45.3 41.2 38.9 46.2 8.7 u.7 40.,2 7.8

28.1 14.6 1.5 21.7 6.3 5.5 3.7 6.0 1.7 1.6 8.0 0.7

33.7 33.8 24.9 22.9 9.5 6.7 4.4 7.0 1.9 1.7 8.6 0.8

27.1 31.5 22.7 17.4 12.9 4.7 6.7 14.1 6.9 3.2 10.0 2.1

816 = annealed 8 hr a t 871째C; 121. = annealed 1h r a t 1177째C; 123 = annealed

Table 36; d e t a i l s of these t e s t s were discussed p r e ~ i o u s l y . ~The fract u r e s t r a i n s a t a s t r a i n rate of 0.05 min-l are shown i n Fig. 86 as a f'unction of t e s t temperature.

The material i n t h e unirradiated condition

was characterized by a high f r a c t u r e s t r a i n of about 50$ t h a t decreased

gradually with increasing temperature above 650°C.

I r r a d i a t i o n caused

a general decrease i n the f r a c t u r e s t r a i n ; t h e decrease was l a r g e s t a t t e s t temperatures of 25°C and above 600°C. The decrease a t 25°C was l i k e l y associated with carbide p r e c i p i t a t i o n and has been observed f o r many other modified and standard heats.7 The decrease a t high temperatures i s due t o helium produced during i r r a d i a t i o n . Similar r e s u l t s f o r a l a r e r s t r a i n r a t e of 0.002 min'l a r e shown i n Fig. 87. The f r a c t u r e s t r a i n s were lower than those observed a t t h e higher s t r a i n r a t e .

Addi-

t i o n a l results given i n Table 36 show t h a t t h e higher anneal of 1 h r a t 1177°C resulted i n b e t t e r f r a c t u r e s t r a i n s a f t e r i r r a d i a t i o n than t h e anneal of 8 hr a t 8 7 1 ° C .

The samples irradiated i n the MSRE f o r 9800 hr

had s l i g h t l y higher f r a c t u r e s t r a i n s than those i r r a d i a t e d i n t h e ORR

f o r about 1100 hr.

This may have resulted f r o m t h e d i f f e r e n t thermal

h i s t o r i e s o r from t h e d r a s t i c a l l y d i f f e r e n t flux spectra.

7H. E. McCoy, A n Evaluation of t h e Molten-Salt Reactor Experiment Hastelloy N Surveillance Specimens - T h i r d Group, ORNL-TM-2647 (1970). RNL-LWG

69-14357

I00

200

300

400 500 TEST TEMPERATURE

600

700

800

900

6) i

Fig. 86. Variation of Fracture S t r a i n s of Alloy 67-504 with Test Temperature a t a S t r a i n Rate of 0.05 min-l. Annealed 1hr a t 1177°C. Irradiated a t 650°C.

Table 36.

Tensile Properties of Irradiated and Unirradiated Alloy 67-504 (0.50$ Hf)

x 103 x 103

x 1020 5058 5046 5057 5064 5061 5061 5065 5053 5055 5 w 5065 5042 5067 5048 5066 5047 5060 5056 6255 6256 6257 6258 6259 6260 6265 6262 6266 6263 6264 6267 5086

5oe7

5088 5089 5090 5091 5092 5093 5079 5078 5085 5084 5077 5076 5075 5074 5073 5072 6238 6239 4945 4927 4986 4970

I51 I51 I51 I51 I51 I51 I51 l51 I51 151 I51 I51 I51 I51 I51 I51 I51 I51 121 121 121 121 I21 I21 121 121 I21 121 121 121 I21 121 121 121 I21 I21 121 I21 I21 121 I21 121 121 121 I21 121 121 I21 816 816 816 I21 I21 I23

25 200 400 Yt00 500 500 550 550 600 600 650 650 700 700 760 760 850 850 25 200 400 500 550 600 600 650 650

700 650 650 650 659 650 650 650 650 650 650 650 650 650 650 650 650 650 650 650 650 760 650 760 760

760 760 25 200 400 400

500 500 550 550 600

600 650 650 700 700 760 760 850 850 650 650 760 650 760 760

5.3 5.3 5.3 5.3 5.3 5.3 5.3 5.3 5*3 5.3

5.3 5.3 5.3 5.3 5.3 5.3 5.3 5.3 2.02.0 2.0 2.4 2.4 2.4

0.05 0.05 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.05 0.05 0.05 0.05 0.05 0.002 0.05 0.002 0.05 0.05 0.002 0.05 0.05 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.002 0.05

0.002 0.002 0.002 0.002

67.5 59.7 45.5 53.4 43.1 61.3 45.3 49.2 45.4 45.8 43.3 44.9 53.3 42.1 44.2 42.7 34.9 31.5 37.4 33.1 27.1 30 27.8 25.5 24.4 25.6 23.9 25.7 24 35.8 102 46.4 41 41.8 40.2 41.6 39.6 38 39.7 40.2 32.9 34.7 35 35.8 34.9 37 32.6 25.4 34.8 34.5 18.5 30.4 u.5 9.3

133 ll9 105 123 99.9 116 102 99.7 97.7 91.3 94.3 80.2 94.6 68.1 75.4 50.8 47.8 32.4 106.7 106.2 88 96.6 91.2 87.4 71.4 83 63.6 76.2 65.8 42 ll9 106 97.6 99 93 91.2 92 78.2 82.6 69.6 70 63.9 65.9 56.2 57.6 42.3 40.8 26 50.4 54.5 18.6 50.4 u.5 9.3

50.2 43.5 46.2 46.2 50.6 44.5 47.9 47.2 49.1 41.9 48.8 30.2 30.8 16.2 20.3 6.0 7.7 2.9 66.6 68.7 77.1 67.3 79.0 73.6 45.7 63.9 33.1 46.7 33.5 6.3 26.0 43.7

44.0 41.8 42.8 42.2 46.0 28.2 35.3 23.7 30.0 21.9 23.2 ll.9 12.8 4.3 4.8 2.1 9.2 10.3 2.4 12.8 1.1 1.2

52.3 44.0 47.2 47.9 52.1 46.1 49.4 48.5 50.9 43.7 50.9 41.2 39.5 45.8 43.5 38.9 38.7 34.8 73.2 71.3 80.1 73.9 82.3 79.6 47.1 66.9 34.2 49.3 35.5 ll.3 27.0 44.4 44.5 42.9 43.3 43.2 46.7 29.9 35.7 24.8 30.7 23.1 24.2 U.6 13.8 6.7 6.6 4.4 9.9 12.0 2.8 16.8 1.1

1.3

43.7 41.6 36.4 34.2 41.3 33.5 36.9 39.6 36.7 37.3 37.4 38.1 36.0 40.6 38.4 41.8 35.5 34.3 73.7 52.2 60.6 50.0 62.5 56.3 34.7 45.1 26.9 32.4 25.7 16.4 33.5 31.9 37.2 36.4 37.9 28.2 36.5 21.2 33.5 17.2 31.5 17.0 24.5 13.1 28.0 11.9 9.4 11.5 u.2 19.4 4.9 7.0 6.0 5.1

aAnnealing designations: I51 = annealed 1h r a t 1177째C plus 9800 hr a t 650째C; I21 = annealed 1h r a t ll77'C; 816 = annealed 8 hr a t 871째C; 123 = annealed 1hr a t 1260째C.

OKNL-OWG

69-14356

z I

$30 K

k K W

2 20

.,0

100

200

300

400 500 600 TEST TEMPERATURE PC)

700

000

900

Fig. 87. Variation of Fracture Strain of Alloy 67-504 with Test Temperature a t a S t r a i n Rate of 0.002 en-’. Annealed 1h r a t 1177°C. Irradiated a t 650°C. The results of creep-rupture t e s t s on heat 67-503 are given i n The stress-rupture properties are shown graphically i n

Table 37.

Fig. 88, and t h e creep c h a r a c t e r i s t i c s are sham i n Fig. 89.

Irradia-

t i o n reduced t h e rupture l i f e more a t 760°C than a t 650°C. I r r a d i a t i o n a t 650°C had no e f f e c t on t h e minimum creep rate, but i r r a d i a t i o n a t

760°C increased t h e minimum creep rate (Fig. 89). Results of creep-rupture tests f o r heat 67-504 (0.50% H f ) a r e summarized i n Table 38. program.

Data a r e included fkomthe MSRE surveillance

The stress-rupture c h a r a c t e r i s t i c s a r e s h a m i n Fig. 90.

This

heat had a rupture l i f e eight t o t e n times higher than t h a t of heat 67-503

(0.084 H f ) .

The samples aged i n fluoride salt f o r 9800 h r had s l i g h t l y

shorter rupture l i v e s than those annealed 1 h r at 1177°C before t e s t . I r r a d i a t i o n a t 650°C reduced t h e rupture l i f e , but less f o r t h e samples i r r a d i a t e d i n t h e MSRE than f o r those i r r a d i a t e d i n t h e ORR.

Irradia-

t i o n a t 760°C i n t h e ORR d r a s t i c a l l y decreased t h e rupture l i f e .

The

minimum creep rates f o r these samples a r e shown graphically i n Fig. 91. The creep strength of heat 67-504 is about t e n times t h a t of heat 67-503. The minimum creep r a t e was unaffected by aging a t 650°C f o r 9800 hr i n

static s a l t or by short-term (1000-hr) i r r a d i a t i o n a t 650°C. Irradiat i o n a t 650°C i n t h e MSRE increased t h e creep rate, and i r r a d i a t i o n a t

760°C d r a s t i c a l l y increased the creep r a t e .

Table 37. Specimen

Test

Creep Properties of Irradiated' Annealb Before Irradiation

Number

Temperature, "C Irradiation

Test

and Unirradiated Alloy 67-503 (0.084 Hf) Stress (Psi)

RY?twe Life

(hr)

Minimum Creep R a te (&/hr)

Fracture Strain ( $)

Reduct ion i n Area

24.6 u.3 9.9 8.3 9.2 8.4, 2.8 3.2 3.7

26.0

(4)

x 103 6179 4957 4958 6178 6177 6176 495 0 4952 4955 4975

4978 4979 4967 4969 4942 4944 4954 4977 4968 4943

75u 6257 6256 7513 7512 7511 R-493 R-505 R-537 R-500 R-515 R-547 R-499 R-541 R- 5 10 R-539 R-513 R-508 R-533 R-527

121

121 121 121 121 121 121 121 121 121 121 121 123 123 816 816 121

I21 123 816

65 0 650 650 760 760 760 760 760 760 760 65 0 760 760 760

650 65 0 650 65 0 760 760 65 0 650 650 650 650 650 650 65 0 650 65 0 760 760 760 760

47 40

5.6 68.3

150.1

0.26 0.046 0.019 0.0%5

30 20

498.1

41.1

0.15

15 47 40

128.1

0.039 0.45

47 35 30

47 40 40 35

15 12.5 12.5 12.5

4.1 32.6 45.6 0 0.4 2.2 0 0 2.6 3.3 36.8 7.6 0.5 8.1

0.071 0.073 0.35 0.036

0.77

0.22 0.096 0.029 0.082 0.053 0.021

0.87 0.51

0.11

1.7 0.41 0.07 0.68

Irradiated a t indicated temperatures t o a thermal fluence of 2.4 X lo2' neutrons/cm2. bAnnealing designations: -121= annealed 1h r a t 1177째C; 123 = annealed 1h r a t 1260째C; 816 = annealed 8 h r a t 871째C.

18.0 11.6 17.5 5.1 16.8

116 ORNL-DWG 6944355

t0-'

to3

to' RUPTURE LIFE (hr)

Fig. 88. Stress-Fhrpture Properties a t 650째C of Alloy 67-503. A l l samples annealed 1h r a t 1177째C before t e s t or irradiation. Number by each datum point indicates i r r a d i a t i o n temperature. ORNL-DWG 69-14354

2 30 W

ti 20

0 a-3

10-1 10-2 MINIMUM CREEP RATE (%/hr)

too

Fig. 89. Creep Rates a t 650째C of Alloy 67-503. A l l samples annealed 1hr a t 1177째C. Numbers by each datum point indicate irradiat i o n temperature.

117

Table 38.

Specimen Test Number Number

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

Anneala Before IrraIrradiation diation

Test

ThermalNeutron Stress Fluence (Psi) (neutrons/cm21

x 1020 6247 6245 4991 4936 4933 6243 5050 5043 5049 5062 6249 4929 4928 4932 4987 4989 5083 5080 5081 5082 4971 4973 4946 4948

5m 5 M 4990 4930 4988 4947 4572

7432 7431 6255 6254 6253 7430 7228 7227 7226 7225 7516 R-518 R-507 R-548 R-509 R-543 R-733 R-715 R-708 R-732 R-495 R- 544 R-511 R-540 R-952 R-953 R-542 R-519 R-517 R-492 R-504

121 121 I21 121 I21 121 I51 I51 I51 I51 I21 121 121 121 121 121 121 121 I21 121 123

I23 816 816 121 I21 I21 I21 I21

121 I23

650 650 650 760 760 650 650 650 650 760 760 760 760 650 650 760 650 760 760 760

650 650 650 650 650 650 650 650 650 650 760 650 650 650 650 650 650 650 650 650 650 650 650 650 650 650 760 760 760 760 760

2.4 2.4 2.4 2.4 2.4 5.3 5.3 5.3 5.3 2.4 2.4 2.4 2.4 0.25 0.25 2.4 2.4 2.4 2.4 2.4

Life (&)

x 103 1.3 70 17.7 63 327.9 55 425.5 47 876.9 40 1236.5 40 5.5 70 108.3 55 329.2 47 10L4.6 40 2 U .4 20 17.0 55 87.9 47 261.2 40 0.3 47 40 1.3 59.6 55 lA1.7 47 467.2 40 32.4 1643.8 0 47 0.1 40 4.6 47 19.9 40 0.6 55 1.0 55 8.1 17.5 287.7 I5 E' 74.7 9.3 I5 12.5 0.3

(5,~)

3.74 0.16 0.029 0.0068 0.0030 0.0034 3.80 0.l8 0.049 0.020 0.088 0.36 0.060 0.017 7.38 0.43 0.091 0.027 0.oll 0.0025 0.17 0.020 0.089 0.015

0.042 0.076 0.35

Fracture Reduction Strain in rea

(5)

(%I

32.9 31.2 27.2 28.7 16.3 26.5 43.7 41.7 39.4 40.8 34.8 13.1 7.6 5.9 2.2 0.60 7.0 6.8 6.9 5.9

33.4 27.7 37.8 19.7 20.1 27.3 43.9 39.6 43.3 48.0 27.2

0.68 1.8 0.93 8.6 10.6 0.94 8.0 3.9 1.6 0.33

aAnnealing designations: 121 = annealed 1 hr at 1177'C; I51 = annealed 1 hr at ll77"C plus 9800 hr at 650째C; I23 = annealed lhr at 1260째C; 816 = annealed 8 hr at 871째C.

6RNL-DWG 69-14353

w .

- 0 ANNEALED Ihr AT 1177% A ANNEALED l t r AT 1177.C. 9800 hr AT 650.C

OPEN SYMBOLS -UNIRRADIATED

100

m II

I I

103

102

(01

RUPTURELIFE (hr)

Fig. 90. Stress Rupture a t 650째C of Alloy 67-504. datum point indicates irradiation temperature.

ORNL-DWG

Number by each

(xt~3)

5 a

ul ul W

E 3 0

0 10-3

ANNEALED l h r AT 1177% ANNEALED lhr AT 1177%. 9800 hr AT 650% SYMBOLS-UNIRRADIATED D SYMBOLS- IRRADIATED

16'

40'

MINIMUM CREEP RATE W/hr)

Fig. 91. Creep Rates a t 650째C of Alloy 67-504. datum point indicates irradiation temperature.

Number by each

119

The f r a c t u r e s t r a i n s of heats 67-503 and 67-506 i n t h e unirradiated condition a t 650°C a r e shown i n Fig. 92. Heat 67-506 (0.505 H f ) had much higher f i a c t u r e s t r a i n s than heat 67-503 (0.08% H f ) , p a r t i c u l a r b

a t t h e luwer creep r a t e .

ORNL-DWG 69- I4351

-z sp

t;, W

I-

YLL IL

0 (0-3

10-2 (0'00 MINIMUM CREEP RATE (% /hr)

10'

Fig. 92. F r a c t u r e S t r a i n s a t 650°C of Unirradiated Alloys 67-503 and 67-504. A l l sarnples annealed 1 h r a t 1177°C. The r a t h e r sparse data f o r heats 67-503 and 67-504 at.76O"C a r e pre5

sented i n Tables 37 and 38, and the stress-rupture properties a r e p l o t t e d i n Fig. 93.

In t h e unirradiated condition, a l l o y 67-504 had a rupture

l i f e about f i v e times greater than t h a t of a l l o y 67-503.

Irradiation a t

ORNL-OW 69-14350

! ,

! I

I l & J i ! I

Fig. 93. Stress-Rrrpture Properties a t 760°C of Alloys 67-503 and 67-504. A l l samples annealed 1hr at 1177°C. Number by each datum point indicates t h e i r r a d i a t i o n temperature.

I20 650°C reduced t h e rupture l i f e , but i r r a d i a t i o n a t 760°C reduced it much

The f r a c t u r e s t r a i n s i n t h e i r r a d i a t e d condition a r e shown i n

more.

Fig. 94.

Q t

The f r a c t u r e s t r a i n s of heat 67-503 were 3 t o 4$ when irra-

diated and t e s t e d a t 760°C.

Irradiation a t 760°C and t e s t i n g a t e i t h e r

650 o r 760°C resulted i n lower f r a c t u r e s t r a i n s .

Heat 67-504 generally

had f r a c t u r e s t r a i n s of 6 t o 8% when i r r a d i a t e d a t 650°C and t e s t e d a t 650 o r 760°C.

Irradiation a t 760°C resulted i n f r a c t u r e s t r a i n s as low

as 0.6%. Several of the samples were examined metallographically a f t e r t e s t ing. Typical photomicrographs of a sample of heat 67-504 a f t e r exposure t o a noncorrosive fluoride salt f o r 9800 h r a t 650°C and testing a t 25°C a r e shown i n Fig. 95.

The f r a c t u r e was p a r t i a l l y intergranular, although

t h e bulk grains elongated greatly. annealed material shown i n Fig. 80.

Also more carbides formed than i n t h e A t 65OoC, t h e grains s t i l l showed ORNL-DWG 69-(4349

io-2

IO-'

f00

f 0'

MINIMUM CREEP RATE (%/hr)

Fig. 94. 67-504.

Postirradiation Fracture Strains of Heats 67-503 and I

Fig. 95. Photomicrographs of Sample of Heat 67-504 (0.50% Hf) Annealed 1 h r at 1177OC, Exposed 9800 hr to Fluoride Salt a% 650째C, and Tested in Tension at 25째C. (a) Fracture, 1OOX; (b) typical unstressed, lOOX. Etchant: glyceria regia.

I22 \

extensive elongation, and there are numerous intergranular cracks, although t h e f i n a l f r a c t u r e was mixed intergranular and transgranular (Fig. 96). After irradiation, t h e fkacture of-heat 67-504 a t 25째C (Fig. 97) was more intergranular than t h a t of the unirradiated material (Fig. 95). A t y p i c a l f r a c t u r e of heat 67-503 a f t e r creep t e s t i n g a t

650째C i s sham in Fig. 98. There are numerous intergrarlular cracks, and t h e fracture i s predominantly intergranular. Heat 67-504 (Fig. 99) exhibited fewer intergranular cracks and a f r a c t u r e t h a t was mixed i n t e r granular and transgranular.

Fig. 96. Photomicrograph of t h e Fracture of a Sample of Heat 67-504 (0.50% H f ) Annealed 1hr a t 1177"C, Exposed 9800 h r t o Fluoride S a l t a t 650"C, and Tested i n Tension a t 650째C and a S t r a i n Rate of 0.002 min-l.

1oox.

w i

I23

Fig. 97. Fracture of a Sample of Heat 67-504 (0.43% H f ) Annealed 1hr a t 1177"C, Irradiated i n t h e Molten Salt Reactor Experiment t o a Thermal Fluence of 5.3 X lo2' neutrons/cm2 over a period of 9800 h r a t 650"C, and Tested i n Tension a t 25°C. Etchant: glyceria regia. 1OOX.

Fig. 98. Fracture of a Sample of Heat 67-503 (0.08% €If)Annealed 1 h r a t 1177°C and Creep Tested a t 650°C and 40,000 p s i . Etchant: glyceria regia. 1 0 0 ~ .

I24

Fig. 99. Fracture of a Sample of Heat 67-504 (0.50% H f ) Annealed 1 hr a t 1177°C and Creep Tested a t 650°C and 40,000 p s i . Etchant: glyceria regia. 1 0 0 ~ .

Alloy Containing No Additions The r e s u l t s f o r a small, 2-lb laboratory melt ( a l l o y 100) t h a t contained no additions of T i , Zr, o r H f were presented above. One small, 100-lb commercial a l l o y (heat 21546) was a l s o procured. This a l l o y was received i n t h e cold-worked condition.

Samples were t e s t e d i n t h i s con-

d i t i o n and a f t e r annealing a t 871 and 1177°C. of t h e material a r e sham i n Figs. 100 and 101.

Typical photomicrographs The microstructure a f t e r

annealing a t 871°C was characterized by bands of p r e c i p i t a t e and f i n e

grains.

The bands were caused by carbide p r e c i p i t a t i o n and retarded

r e c r y s t a l l i z a t i o n i n these areas. recrystallized.

The material outside these bands was

Annealing f o r 1hr a t 1177°C (Fig. 101) completed

recrystallization.

Some s t r i n g e r s were s t i l l present, although none

a r e v i s i b l e i n Fig. 101.

These s t r i n g e r s were q u i t e similar t o those

found i n heat 21545 and appeared t o be m e l t e d molybdenum.

Fig. 100. Photomicro hs of Alloy 21546 Annealed 100 hr at 871째C. The material was cold worked 40qd before annealing. (a) lOOX and (b) 5 0 0 ~ . Etchant: glyceria regia.

126

Fig. 101. Photomicrographs of Alloy 21546 Annealed 1 hr at ll77OC. The material was cold worked 40% before annealing. (a) l O O X and (b) 500~. Etchant: glyceria regia.

I27 The r e s u l t s of t e n s i l e t e s t s on t h i s material are summarized i n Table 39.

The f r a c t u r e s t r a i n s a r e shown as a function of temperature

i n Fig. 102.

Up t o about 700"C, t h e material annealed 1h r a t 1177°C

had t h e highest f r a c t u r e s t r a i n and the cold-worked material had t h e lowest f r a c t u r e s t r a i n .

Above 700"C, the d u c t i l i t y of t h e cold-worked Chly samples as received and as annealed

material increased markedly.

100 h r a t 871°C were irradiated, and these were a l l t e s t e d a t 600°C and above.

I r r a d i a t i o n decreased t h e f r a c t u r e s t r a i n of a l l samples.

The r e s u l t s of creep t e s t s on t h i s material a r e given i n Table 40.

In t h e unirradiated condition, the material as received had t h e highest strength and the material annealed a t 871°C had t h e lowest strength. The strength of material annealed 1hr a t 1177°C.f e l l intermediate, and

several i l l u s t r a t i o n s follow t o compare the properties of i r r a d i a t e d and unirradiated material annealed 1 hr a t 1177°C. The stress-rupture proper-

t i e s a t 650°C are shown i n Fig. 103.

The i r r a d i a t i o n temperature was an

important factor; t h e higher t h e i r r a d i a t i o n temperature, t h e shorter was t h e rupture l i f e . I

The two s e t s of data seem t o converge, indicating

t h a t a t very low levels of stress there would be no e f f e c t of i r r a d i a t i o n on t h e rupture l i f e .

The creep r a t e s a r e shown i n Fig. 104 f o r 650°C.

A s long as the i r r a d i a t i o n temperature was about 65OoC, there was no

difference between t h e creep r a t e s of i r r a d i a t e d and unirradiated mate-

rial.

A t i r r a d i a t i o n temperatures of 700°C and above, t h e postirradia-

t i o n creep rate was increased.

There was a t l e a s t one datum point t h a t

disagreed with t h i s general trend.

These curves a l s o converged, again

indicating common behavior a t l o w stresses. The f r a c t u r e s t r a i n s

i r r a d i a t e d samples t e s t e d a t 650°C a r e

s h m as a function of minimum creep rate i n Fig. 105.

A l l of t h e sam-

p l e s irradiated above 700°C had f r a c t u r e s t r a i n s of 1.1t o 0.2%.

Irra-

d i a t i o n a t about 650°C resulted i n f r a c t u r e s t r a i n s of 1.1t o 5.1%.

The

exact dependence of t h e f r a c t u r e s t r a i n on s t r a i n r a t e cannot be determined from t h e available data, and t h e indicated curve i s influenced by cr

. b,

results on standard Hastelloy N ( r e f . 8 ) .

8H. E. McCoy, "Variation of Mechanical Properties of Irradiated 67-85 (1969). Hastelloy N with S t r a i n Rate," J. Nucl. Mater. 2(1),

Table 39.

Tensile Properties of Irradiated and Unirradiated Alloy 21546

x 1020 10220 10221 10223 10222 10224 10225 10226 10227 10228 u40

U41 U52 1453 U51 u57 U58 u59 1749 1750 1746 1747 1752 1753 1754 u38 1439 1436 1437 m 7 U46

OOO

m m OOO

m OOO

l6 l6 l6 16 l6 l6 16 OOO OOO OOO OOO

lkc2 lkc5

u48

I43

600

l6 16 l6 l6 l6 I6

l444

25 200 260 400 550

I21 121 121 I21 121 121 121 121 I21 OOO

600

600 600 600 600 600 600

650 650 600 600

650 700 760 25 427 650 650 760 871 871 982 25 427 650 650 760 871 871 550 600 650 650 550 600 650 650 650 650 760

3.5 3.5 3.5 3.5 3.5 3.5 3.5 1.0 1.0 3.5 3.5

0.05 0.05 0.05 0.05 0.05 0.05 0.05 0.05 0.05 0.05 0.05

x io3 38.7 32.6 31.4 30.2 27.1 27.5 23.7 24.4 24 151.9 127.3

0.05

ll9.5

0.002

109.1 76.4 31.5 19 17.9 51.7 40.9 44.3 38.8 37.8 27.7 18.2 m.2 105.3 106.8 103.1 41.9 43.7 38.8 38.6 38.4 45 33.5

0.05 0.05 0.002 0.05 0.05 0.05 0.05 0.002 0.05 0.05 0.002 0.002 0.002 0.05 0.002 0.002 0.002 0.05 0.002 0.05 0.002 0.002

x io3 107.5 99.7 100.9 97.8

90.5 87.1 69.8 64.7 60.4 164.2 U4.2 U2.7 U.8 78.4 31.7 19 17.9 l19.4 105.2 95.5 69 51.5 28.8 18.2 127.9 ll5

l16.6 106.8 74.8 65.6 a.2 55 67.5 56.7 33.8

69.6 69.8 71.5 73.0 73.4 66.0 46.3 35.1 25.0 6.8 12.1 '7.7 3.4 2.6 1.4 1.0 0.9 49.2 47.9 27.8 19.0 3.6 10.0 0.9 6.0 4.4 5.7 3.5 U.8 8.2 10.3 7.3

u.5 5.6 1.8

73.7 73.8 76.4 79.3 78.3 69.5 48.3 36.1 25.9 13.4 15.5

75.1 63.3 68.0 64.8 59.5 47.2 32.1 26.9 20.0 53.3 32.7

33.6

18.8

8.2 32.3 46.7 48.1 60.8 53.5 52.4 30.0 35.1 7.5 53.5 45.3 6.4 4.5 6.2 3.7 15.1 8.4 10.5 7.4 u.7 5.7 4.3

16.3 51.1 58.4 44.0 60.6 72.4 46.2 31.2 64.7 68.2 69.1 51.2 7.1 4.0 8.6 4.0 16.2 10.2 ll.7 12.9 14.7 8.6 7.1

'Annealing designations: 121 = annealed 1 hr at ll77"C; OOO = as received (4% cold work); 16 = annealed 100 hr at 871째C.

I29

Table 40.

Creep Properties of Irradiated and Unirradiated Alloy 21546

pTa-

Anneala erature o c Specimen Test Before 2 Nmber Number IrraTest diation diation

4056 4057 4058 4059 10219 10218 1455 l454 1456 1751 1748 1745 1826 10231 10229 5988 4060 4061 4065 5981 5980 5979 5974 5965 5975 5966 5982 5967 5969 5985 5976 5968 5971

5985 5972 4066

4067 5970 5973 4068 1450 U49

6300 6301 6302 6303 7288 7289 5528 5527 5408 5531 5536 5532 5435 7521 7517 R-598 R-294 R-482

R-481 R-635 R-604 R-642 R-602 R-605 R-607 R-609 R-6-43 R-638 R-639 R-600 R-631 R-632 R-637 R-600 R-6-40 R-296 R-468 R-634 R-6-41 R-469 R-120 R-W

121 121 121 I21 121 121 OOO OOO

16 16 16 816 121 121 121 121 121 121 121 121 121 121 121 121 121 121 121 121 121 121 121 I21

121 121 121 121 121 121 121 16 16

235 650 657 650 666 666 766 754 732 760 704 666 732 816 499 760 832 871 499 799 746

746 832 799 746 600 600

mermalNeutron Stress Rupture Life Fluence (psi) (neutrons/cm2) x 1020 x 103 62 55 47 40 35 30

650 650 650 650 650 650 650 650 650 650 650 650 650 760 760 650 650 650 650 650 650 650 650 650 650 650 650 650 650 650 650 650

2.4 2.5 2.5 2.5 2.4 2.4 2.4 2.4 2.4 2.4 2.4 2.4 2.4 2.4 2.4 2.4 2.4

650

2.4

760 760 760 760 760 760 760 650 650

2.4 2.4 2.5 2.5 2.4 2.4 2.5 3.5 3.5

70 55

47 70 55 47

40 15 12.5 47 47

40 40 40 35 35 35 35 30 30 27 25 25 25 21.5 21.5 17 25 I5 15 12.5 12.5 10 10

40 32.4

5.3 18.5 63.0 187.5 256.2 561.6 11.7 92.7 223.6 1.1 6.9 31.7 28.8 221.3 414.7 84.5 10.1 0 9.2 62.0 20.1 0.4 0 0.4 1.2 1.1 727.8 46.6 ll.1 9.4 68.3 69.6 1157.1

9.4 14.3 0.5 3.6 45.7 U0.9 33.8 3.3

97.7

T z 0.56 0.22 0.074 0.026 0.013 0.0080

0.31 0.025 0.OU 18.0 2.87 0.83 0.20 0.08 o.ou.2 0.OU 0.41

Fracture Reduction Strain in Area ( %) (%I

25.6 21.3 18.1 19.4 l3.0 ll.l 12.5 12.5 7.5 34.4 46.9 46.9 8.5 16.3 u.4 2.3 5.1

0.019 0.033 0.042 2.1

2.4 2.8 1.1 0.9

L3 0.4

0.6 0.8 0.7 3.1

0.E

0.0030 0.0065 0.015 0.26 0.0039 0.0028

0.4 0.4 3.2 0.53 0.24

O.OOO6

1.1

0.26

3.1 1.3

0.061

0.060

o,ol&

0.7 0.4 0.8

0.0098 0.036 0.40 0.038

1.5 1.1 1.3 4.0

0.12

ahnealing designations: 121 = annealed 1 hr at 1177째C; OOO = as received (40g cold work) 16 = anuealed 100 hr at 871째C; 816 = annealed 8 hr at 871째C.

24.9 21.3 16.6 15.6 16.2 18.7 15.5 l4.0 6.3 31.0 33.7 39.6 9.4 15.8 12.7

I30 ORNL-DWG 69-44348

6.' b

0 0

100

200

300

400 500 600 TEST TEMPERATURE (OC)

700

800

900

f000

Fig. 102. Variation of the Flracture Strain with Test Temperature for Alloy 21546 at a Strain Rate of 0.05 min'l.

Fig. 103. Stress-Rupture Properties at 650째C of Alloy 21546. A l l samples annealed 1 h r at 1177째C before testing. Number by each datum point indicates irradiation temperature.

ORNL-DWG 69-14346

hsld c

Et 10-1

I00

MINIMUM CREEP RATE (%/hr)

Fig. 104. Creep Rate a t 650°C of Alloy 21546. A l l samples annealed 1 h r a t 1177°C before testing. Number by each datum point indicates i r r a d i a t i o n temperature.

ORNL-DWG 69-44345

--s-”

z4 a a I-

m w 3

I-

u E* t

0 to-4

10-3

10-‘ MINIMUM CREEP RATE (%/hr)

Fig. 105. Fracture Strains a t 650°C of Irradiated Alloy 21546. A l l samples annealed 1h r a t 1177°C before t e s t i n g . Number by each datum point indicates i r r a d i a t i o n temperature.,

132 Some samples were tested a t 760°C; the r a t h e r sparse r e s u l t s are given i n Table 40, and t h e stress-rupture data are shown i n Fig. 106. There was an indication t h a t i r r a d i a t i o n a t 746°C may result i n poorer stress-rupture properties a t 760°C than i r r a d i a t i o n a t 799 t o 832°C. The minirmrm creep rate a t 760°C (Fig. 107) showed t h a t i r r a d i a t i o n did not a l t e r t h e creep rate d r a s t i c a l l y .

However, there was again t h e ten-

dency f o r t h e t h r e e samples i r r a d i a t e d a t 746°C t o show t h e greater effect.

The f r a c t u r e s t r a i n s a t 760°C were a l l q u i t e low (0.4 t o 1.5$

s t r a i n ) except f o r t h e sample irradiated a t 499°C (3.1$ s t r a i n ) .

ORNL-DWG 69-!4344

400

40-'

IRRADIATED

io'

ro3

RUPTURE LIFE (hr)

Fig. 106. Stress-Rupture Properties a t 760°C of Alloy 21546. A l l samples annealed 1h r a t 1177°C. Number by each datum point indicates i r r a d i a t i o n temperature.

4 x IO4

o UNIRRADIATED

MINIMUM CREEP RATE (%jhr)

Fig. 107. Creep Rates a t 760°C of Alloy 21546. Material annealed, 1 h r a t U77"C. Number by each datum point indicates i r r a d i a t i o n temperature.

I33 Some insight into the effect of irradiation temperature was gained by transmission electron microscopy.

Typical microstructures for two samples irradiated at 666 and 766째C are sham in Figs. 108 and 109.

Fig. 108. Transmission Electron Photomicrograph of Alloy 21546 af'ter Irradiation at 666째C for about 1100 hr to a Thermal-Neutron Fluence of 2.4 x 1020 neutrons/cm*. 50,OOOX. ,

w e

Fig. 109. Transmission Electron Photomicrograph of Alloy 21546 af'ter Irradiation at 766째C for about lloo hr to a Thermal-Neutron Fluence of 2.4 X lo2' neutrons/cm*. 50,000~.

A%er i r r a d i a t i o n a t 666OC, t h e grain boundq4es were completely lined % ,*-- . with f i n e p r e c i p i t a t e p a r t i c l e s , and there were a l s o numerous precipin-

t a t e s within t h e grains.

M t e r i r r a d i a t i o n a t 7 6 6 째 C (Fig. log), there

were fewer r e l a t i v e l y coarse p r e c i p i t a t e s along t h e grain boundaries, and only random coarse p r e c i p i t a t e s were present within the grains.

The

p r e c i p i t a t e s i n both samples were i d e n t i f i e d as carbides of the M2C type i n which t h e metal component was primarily molybdenum.

Thus, t h e poorer

p o s t i r r a d i a t i o n properties a r e associated with t h e very coarse p r e c i p i t a t e . A t y p i c a l f r a c t u r e of heat 2U46 a t 6 5 0 째 C i s sham i n Fig. 110.

There was extensive intergranular cracking, and t h e f r a c t u r e was e n t i r e l y intergranular.

The random s t r i n g e r s a r e obvious i n t h i s f i e l d and may have

been instrumental i n causing t h e f r a c t u r e a t t h i s p a r t i c u l a r location.

Fig. 110. m a c t u r e of a Sample of Heat 21546 Annealed 1 h r a t 1 1 7 7 째 C and Tested a t 40,000 p s i and 65OOC. Etchant: glyceria regia.

DISCUSSION OF RESULTS Fl

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

l a t e some of t h e most important observations.

136 1. A modified base composition of Hastelloy N was developed: Ni-12% M0-7$

C?.-0.2$ W . 0 5 $ C.

This a l l o y i s free of t h e carbide

s t r i n g e r s present i n standard Hastelloy N, but is s t i l l very susceptible t o embrittlement by i r r a d i a t i o n . 2. Additions of T i , Zr, or H f t o t h i s base composition improved resistance t o i r r a d i a t i o n embrittlement.

Small heats showed a general

ilnprovement with increasing additions of titanium and hafnium.

Similar

work with additions of zirconium shpwed t h a t an a l l o y containing 0.53% Zr had t h e minimum creep rate and maximum rupture l i f e .

Other concentra-

t i o n s of zirconium resulted i n higher fracture s t r a i n s . 3.

Commercial melts (100 l b ) were obtained with nominal additions

of 0.5% each of T i , Zr, and Hf.

The properties of these alloys were

comparable with those of t h e small, laboratory melts ( 2 l b )

4. Properties were very dependent upon t h e i r r a d i a t i o n temperature. I r r a d i a t i o n a t about 650°C resulted i n good properties, but i r r a d i a t i o n

a t 700°C and above gave r e l a t i v e l y poor properties.

Transmission electron microscopy revealed that the e f f e c t of

i r r a d i a t i o n temperature was associated with t h e p r e c i p i t a t i o n of coarser carbides a t the higher temperature.

During i r r a d i a t i o n a t 65OoC, f i n e

carbides precipitated, whereas above 700°C r e l a t i v e l y coarse carbides

formed

The optimum properties after i r r a d i a t i o n r e s u l t from annealing

f o r 1 h r at 1177°C before i r r a d i a t i o n .

The fine-grained microstructure

obtained by annealing at low temperatures such as 871°C was not as strong o r as ductile, and annealing above 1177°C decreased t h e f r a c t u r e s t r a i n of irradiated and unirradiated samples.

The general trend revealed by metallographic examination a f t e r

t e s t i n g was that additions of T i , Zr, and H f reduced t h e frequency of cracking a t grain boundaries. Fractures i n these alloys i n creep a t 650°C a r e predominantly intergranular but usually have some transgranular components. Same comparisons are i n order t o show t h e r e l a t i v e properties of these a l l o y s and standard Hastelloy N.

For t h i s comparison, we used

only t h e commercial alloys with 0.5% nominal additions of T i , Zr, and Hf.

All r e s u l t s are f o r samples annealed 1hr a t 1177°C.

The t e n s i l e

137

and y i e l d strengths are shown i n Fig. 111. Alloys 21546 and 21545 (0.49% Ti) had very similar properties; t h e observed strengths were s l i g h t l y less than those measured f o r standard Hastelloy N.

The strength

of heat 67-504 (0.50% Hf) was t h e highest of a l l alloys studied.

The

f r a c t u r e s t r a i n s of these samples a r e shown i n Fig. 112. Alloys 21546, 21545 (0.49% T i ) , 21554 (0.35% Zr), and 67-504 (0.50% Hf) have similar Heat 67-504 (0.50% H f ) a f t e r aging f o r 9800 hr a t 650째C had

properties. \

fracture s t r a i n s lower than those of t h e other modified alloys when t e s t e d

below about 650째C and higher above t h i s temperature.

Standard Hastelloy N

had lower f r a c t u r e s t r a i n s than any of t h e modified alloys. ORNL-DWG 69-44342

4 40

I20

100

-g u)

I-

0 0

100

200

300

400 500 600 TEMPERATURE ('C)

700

800

900

Fig, 111. Tensile Properties of Several Commercial Melts of Modif i e d Alloys. A l l material annealed 1hr a t 1177째C. The r e l a t i v e stress-rupture properties of these alloys a t 650째C

are shown i n Fig. 113. The rupture l i v e s a t 40,000 p s i range from 170 t o 1100 hr; a l l o y 21546 had the shortest rupture life, and a l l o y 67-504 I

B.I

(0.50% H f ) had t h e longest.

Minimum creep r a t e i s a b e t t e r measure of

138 ORNL-DWG 69-44344

rd r

z 50 t-

A 24545 (0.49 % Ti) A 21546

67-504 (AGED 9000 hr AT 6 5 0 W 24554 (0.35 7% Zr) 0 STANDARD HASTELLOY N 67-504 (0.50 % Hf)

400

200

400 500 600 TEST TEMPERATURE CC)

300

700

000

900

Fig. 112. Variation of Tensile Fracture S t r a i n with Test Temperat u r e f o r Several Commercial Melts of Modified Alloys a t a S t r a i n Rate of 0.05 min-l. All material annealed 1hr a t ll77"C.

ORNL-DWG 69- I4340

v) v) W

30 20 IO

to'

102

to3

RUPTURE L I E (hr)

Fig. 113. Stress-Rupture Properties a t 650째C of Several Commercial Melts of Modified Alloys. A l l material annealed 1h r a t 1177째C before testing.

I39 strength than rupture l i f e ; these data a r e shown i n Fig. l l 4 . 1

40,000 p s i , t h e minimum creep r a t e varied from 0.03$/hr f o r a l l o y 21546 t o O.O02$/hr f o r a l l o y 67-504 (0.50% H f ) .

The addition of hafnium

increased strength t h e most, by far, but the three alloys modified with H f , Zr,

and T i were a l l stronger than standard Hastelloy N.

There was

not much difference between t h e creep strengths of standard Hastelloy N and t h e modified base a l l o y 21546.

The f r a c t u r e s t r a i n s of t h e unirra-

diated modified alloys a r e shown i n Fig. 115. A l l of t h e alloys except a l l o y 21545 (0.49% T i ) followed t h e trend of decreasing fracture s t r a i n with decreasing creep r a t e .

A t high s t r a i n r a t e s , a l l o y 21554 (0.35% Zr)

had t h e highest f r a c t u r e s t r a i n . (0.49% T i ) had superior d u c t i l i t y .

A t low s t r a i n r a t e s , a l l o y 21545

Alloy 21546 and standard Hastelloy N

had the lowest f r a c t u r e s t r a i n s and did not d i f f e r appreciably from each other. The stress-rupture properties determined a t 650째C a f t e r i r r a d i a t i o n a r e compared i n Fig. 116.

After i r r a d i a t i o n a t 650째C, alloys 21554

ORNL-DWG 6944339

t.0-2 lo-' MINIMUM CREEP RATE (?&/hr) c

lo0

40'

Fig. 114. Creep Rates at 650째C of Several Commercial Melts of Modified Alloys. A l l material annealed 1h r a t 11'77째C before t e s t i n g .

UO ORNL-DWG 69-44338

10-4

10-3

40-2

a-‘

roo

MINIMUM CREEP RATE (%/k)

Fig. 115. Comparison of t h e Fracture Strains a t 650°C of Several Commercial Melts of-Mcsdified Alloys. A l l material annealed 1hr a t 1177°C.

ORNL- DWG 69- (4337R’

ro3,

60 t

u) u)

IO 0

IO-‘

IO0

-a’

10~

lo4

RUPTURE LIFE (hr)

Fig. 116. Postirradiation Stress-Rupture Properties a t 650°C of Several Commercial Melts of Modified Alloys. A l l material annealed 1 h r a t 1177°C before i r r a d i a t i o n .

(0.354 Zr), 21545 (0.494 Ti), and 67-504 (0.50$ H f ) had equally good t

properties.

A l l three of t h e modified alloys shuwed improvement over

t h e standard Hastelloy N shown f o r reference.

The base modified 1

a l l o y 21546 had properties q u i t e similar t o those of standard Hastelloy N.

141

63 a

Standard Hastelloy N t h a t contained about 0.5% S i i s not s e n s i t i v e t o i r r a d i a t i o n temperature, and the indicated l i n e holds f o r i r r a d i a t i o n a t 760°C (ref. 9).

Heats 21546 and 21545 (0.49%’Ti) both had poor stress-

rupture properties a f t e r i r r a d i a t i o n a t 760°C.

The stress-rupture proper-

t i e s of a l l o y 67-504 (0.50$ H f ) f e l l intermediate between those of these two alloys and those of standard Hastelloy N. The p o s t i r r a d i a t i o n f r a c t u r e s t r a i n s f o r i r r a d i a t i o n and t e s t i n g

a t 650°C a r e summarized i n Fig. 117.

These curves a r e based on r a t h e r

sparse data and should be considered as schematics. some improvement over standard Hastelloy N.

Alloy 21546 shows

Alloy 21545 (0.49% T i ) i s

generally b e t t e r than standard Hastelloy N, but s t i l l has a d i s t i n c t d u c t i l i t y minimum of only about 3%. The sparse data f o r heats 21554 (0.35% Z r ) and 67-504 (0.50%Hf) indicate t h a t these heats had improved fracture strains.

The presence of a d u c t i l i t y minimum has been estab-

lished f o r standard Hastelloy N and f o r heat 21545, but there a r e too

few data t o e s t a b l i s h whether t h e other alloys conform t o t h i s pattern.

ORNL-DWG 69- 44336

MINIMUM CREEP RATE f%/hr)

Fig. 117. Postirradiation Fracture Strains of Several Commercial Melts of Modified Alloys Irradiated and Tested a t 650°C. A l l material annealed 1 h r a t 1177°C before i r r a d i a t i o n .

u.2 Many of t h e properties of these alloys a r e l i k e l y determined by t h e

carbide precipitate.

As already shown, a l l o y 21546 exhibited a change

i n properties a f t e r i r r a d i a t i o n t h a t correlated w e l l with changes i n t h e carbide s t r u c t u r e (Figs. 108 and 109). Further work by Gehlbach and Cook1' shuwed t h a t T i , Hf, and Zr promoted t h e formation of t h e MC type of carbides. The normal type of carbide f o r Hastelloy N i s I$C when t h e s i l i c o n content i s l o w and MgC when t h e s i l i c o n content i s high.ll work of Gehlbach and Cook1'

The

a l s o shuwed t h a t l o w levels (about 0.55) of

Ti, Hf, and Zr can r e s u l t i n an MC type of carbide a t 650째C and a rela-

t i v e l y coarse I$C type a t 760째C.

Thus, t h e strong e f f e c t of i r r a d i a t i o n

temperature on these alloys i s associated with t h i s t r a n s i t i o n i n carbide types. The mechanism whereby these p r e c i p i t a t e s improve t h e unirradiated creep properties a l s o l i k e l y has t o do with t h e carbide precipitates.12

I S , t h e creep strength i s increased with t h e addition of Ti, Zr, and H f . These e f f e c t s are larger than anticipated f'rom solidsolution strengthening, p a r t i c u l a r l y when one considers t h a t most of t h e alloying additions are l i k e l y t i e d up as carbides. Thus, t h e strengthen-

As shown i n Fig.

ing observed i s l i k e l y due t o t h e f i n e l y dispersed carbide phase.

These

p r e c i p i t a t e s a l s o l i k e l y impede crack propagation and account f o r t h e improved f r a c t u r e s t r a i n s (Fig. 115). In i r r a d i a t e d material, t h e f u r t h e r complication of helium trans-

muted f o r t h e 'OB e x i s t s .

Our reasoning i n adding T i , H f , and Z r was t o

t i e t h e boron up i n dispersed p r e c i p i t a t e s . The only measure t h a t we have of t h e location of boron i s t o assume t h a t it behaves l i k e t h e carbon. Thus, each carbide p r e c i p i t a t e contains some boron. The finely 'OR. E. Gehlbach and S . W. Cook, Metals and Ceramics Div. Ann. Progr. Rept. June 30, 1969, ORNL-4470, p. 187.

'lR. E. Gehlbach and H. E. McCoy, Jr., "Phase I n s t a b i l i t y i n Hastelloy N," pp. 346-366 i n International Symposium on S t r u c t u r a l Stab i l i t y i n Superalloys, Seven Springs, Pennsylvania, September 4 . 4 , 1968, V o l . 11. Available from Dr. John Radavich, AIME High-Temperature Alloys Committee, Micromet Laboratories, West Lafayette, Indiana. 1 2 C . E. Sessions, Influence of Titanium on t h e High-Temperature Deformation and Fracture Behavior of Some Nickel Based Alloys, ORNL-4561 1July 1970).

u.3 divided MC type of carbide gives a b e t t e r dispersion and would be expected t o give b e t t e r properties.

Thus, t h e q u a l i t a t i v e observation

t h a t t h e f i n e MC type of p r e c i p i t a t e s a r e associated with good proper3

t i e s and t h e coarse M2C type with poor porperties seems q u i t e reasonable. The f u r t h e r r o l e of these p r e c i p i t a t e s i n i n h i b i t i n g propagation of cracks may a l s o be important.

All of t h e modified alloys a r e s l i g h t l y susceptible t o aging during long-term annealing.

These changes a r e q u i t e small and a r e associated

with p r e c i p i t a t i o n of carbides.

The general e f f e c t s of aging a r e t h a t

t h e minimum creep rate and f r a c t u r e s t r a i n increase and t h e rupture l i f e \

decreases.

Because of t h e difference i n t h e type of carbide p r e c i p i t a t e d

a t 650 and 760째C i n many of these alloys, aging.at 760째C seems t o produce l a r g e r changes i n properties.

The p r e c i p i t a t i o n of copious quanti:

t i e s of Z r C i n a l l o y s t h a t contain l e s s than 0.5% Zr i s q u i t e dramatic, but t h e changes i n properties a r e minimal.

SUMMARY A

This f i r s t phase of t h i s study of t h e e f f e c t s of additions of T i , i

Zr, and Hf t o a base a l l o y of Ni-l2$ M&7$

CI-0.24 Mn4.05% C d e a l t with

small, laboratory melts and t h e f i r s t 100-lb commercial melts.

The

r e s u l t s show t h a t these t h r e e elements improve t h e creep strength and

fracture s t r a i n i n unirradiated a l l o y s . Tests on samples i r r a d i a t e d a t 650째C o r less show that t h e rupture l i v e s a t a given s t r e s s l e v e l and t h e fracture s t r a i n s are improved markedly. I r r a d i a t i o n a t 760째C r e s u l t s i n very poor properties. The good properties are associated with t h e formation of a very f i n e MC

ype of carbide and t h e poor properties with a very coarse &C type of carbide. ACKNOWLEDGMENTS The author i s g r a t e f u l t o several technicians who assisted i n c1

running t e s t s and collecting data: N. 0. Pleasant, and B. C

lliams.

H. W. Kline, J. Feltner, B. McNabb, The laboratory melts were melted

and fabricated by C. E. Dunn and J. N. Hix.

The metallographic samples

l.44 were prepared and photographed by H. R. Tinch and E. Lee.

The drawings

were prepared by t h e Graphic A r t s Department and t h e manuscript by t h e Metals and Ceramics Division Reports Office. The author a l s o thanks J. R. Weir, A. C. Schaffâ&#x20AC;&#x2122;hauser, and C. E. Sessions f o r t h e i r reviews of t h i s report.

The transmission

electron photomicrographs and t h e phase-identification r e s u l t s were obtained by R. E. Gehlbach.

145 ORNLTM-3064 4

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