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C o n t r a c t No. W-7405-erg-26 METALS AND CERAMICS DIVISION

AN EVALUATION OF THE MOLTEN-SALT REACTOR EXPERIMENT HASTELLOY N SURVEILLANCE SPECIMENS - FOURTH GROW

H. E. McCoy, Jr.

MARCH 1971

LDGAL NOTIC

rs, or their employees, implied, or assumes any €or the accuracy, cominformation, apparatus,

OAK R I E E NATIONAL LABORATORY Oak R i d g e , Tennessee

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

-c

iii

CONTENTS Page .

.............................. 1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 Experimental Details . . . . . . . . . . . . . . . . . . . . . . . . 2 Surveillance Assemblies . . . . . . . . . . . . . . . . . . . . 2 Materials . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 T e s t Specimens . . . . . . . . . . . . . . . . . . . . . . . . . 5 I r r a d i a t i o n Conditions . . . . . . . . . . . . . . . . . . . . . . . 5 Testing Techniques . . . . . . . . . . . . . . . . . . . . . . . 7 Experimental Results . . . . . . . . . . . . . . . . . . . . . . . . 8 Visual and Metallographic Examination . . . . . . . . . . . . . 8 Mechanical Property Data . Standard Hastelloy N . . . . . . . . 17 Mechanical Property Data . Modified Hastelloy N . . . . . . . . 35 Metallographic Examination of Test Samples . . . . . . . . . . . 57 Discussion of Results . . . . . . . . . . . . . . . . . . . . . . . 72 Summary and Conclusions . . . . . . . . . . . . . . . . . . . . . . 87 Acknowledgments . . . . . . . . . . . . . . . . . . . . . . . . . . 87 Abstract

AN EVALUATION OF THE MOLTEN-SALT REACTOR EXPERIMENT HASTELLOY N SURVEILLANCE SPECIMENS FOURTH GROUP

H. E. McCoy, Jr. ABSTRACT

Two heats of standard Hastelloy N were removed from the core of t h e MSRE a f t e r 22,533 hr a t 650째C and exposed t o a thermal fluence of 1.5 x 1O2I neutrons/cm* and a fast fluence > 50 kev of 1.1X loz1 neutrons/cm2. The mechanical propert i e s have systematically deteriorated with increasing fluence. However, t h e cha e i n roperties i s due t o t h e helium produced by t h e 8B(n,CX77Li transmutation and can be reduced by changes i n chemical composition. Some of these modified heats have been exposed t o the core of t h e MSRE and show improved resistance t o irradiation. The corrosion of t h e Hastelloy N has been l a r g e l y due t o t h e s e l e c t i v e removal of chromium. The r a t e s of removal a r e much as predicted from t h e measured diffusion rate of chromium. Other s u p e r f i c i a l structure modifications have been observed, b u t ' t h e y l i k e l y result from carbide p r e c i p i t a t i o n along s l i p bands t h a t were formed during machining.

INTRODUCTION

The Molten-Salt Reactor Experiment (MSRE) i s a single region reac-

t o r t h a t i s fueled by a molten fluoride s a l t (65 LiF-29.1 BeF2-5 ZrF40.9 UF4, mole $), moderated by unclad graphite, and contained by Hastelloy N (Ni-16 Mo-7 C I Y : Fe-O.05 C, w t $)

The details of t h e reac-

t o r design and construction can be found elsewhere.' neutron environment wou

rials I

- graphite and Ha

We knew t h a t t h e

roduce some changes i n t h e two s t r u c t u r a l mate-

Although we were very confident of

t h e compatibility of thes

e r i a l s with t h e fluoride salt, we needed

t o keep abreast of t h e

e development of corrosion problems within

lR. C. Robinson, M3RE Design and Operations Report, Pt. 1, Descript i o n of Reactor Design, ORNETM-728 (1965).

z t h e reactor i t s e l f .

For these reasons, we developed a surveillance

program t h a t would allow us t o follow the property changes of graphite

and Hastelloy N specimens as t h e reactor operated.

The reactor went c r i t i c a l on June 1, 1965.

After many small prob-

lems were solved, normal operation began i n May 1966. groups of surveillance samples. groups were

.e;

We removed four

The r e s u l t s of t e s t s on t h e first three

This report deals primarily with t h e r e s u l t s

of t e s t s on samples removed with t h e fourth group.

The fourth group

included two heats of standard Hastelloy N used i n fabricating t h e MSRE and three heats with modified chemistry t h a t had b e t t e r mechanicalpropert i e s a f t e r i r r a d i a t i o n and appear a t t r a c t i v e f o r use i n f u t u r e molten-

salt reactors. The respective h i s t o r y of each l o t was (1) standard Hastelloy N, annealed 2 hr a t 900°C and exposed t o t h e USRE core f o r 22,533 hr a t 650°C t o a thermal fluence of 1.5 X

neutrons/cm2, (2) two heats of modified Hastelloy N, annealed 1hr a t 1177°C and exposed

t o the MSRE core f o r 7244 h r a t 650°C t o a thermal fluence of 5.1

neutrons/cm2, and (3) a single heat of modified Hastelloy N, annealed f o r 1hr a t 1177°C and exposed t o t h e USRE c e l l environment of

+ 2 to

59 0 2 f o r 17,033 hr a t 650°C t o a thermal fluence of 2.5 X loi9 neutrons/cm2. The r e s u l t s of t e s t s on these materials w i l l be presented i n d e t a i l , and some comparisons w i l l be made with t h e data f r o m the groups removed previously. EXPERIMENTAL DETAIIS

Surveillance Ass emblies The core surveillance assembly5 was designed by W. H. Cook and others, and t h e d e t a i l s have been reported previously.

The specimens

2H. E. McCoy, An Evaluation of t h e Molten-Salt Reactor Experiment Hastelloy N Surveillance Specimen - First Group, ORNL-TM-1997 (1967). 3H. E. McCoy, An Evaluation of t h e Molten-Salt Reactor Experiment Hastelloy N Surveillance Specimen - Second Group, ORNLTM-2359 (1969).

“H. E. McCoy, An Evaluation of t h e Molten-Salt Reactor Experiment Hastelloy N Surveillance Specimen Third Group, ORNL-TM-2647 (1970). 5W. H. Cook, MSR Program Semiann. Progr. Rept. Aug. 31, 1965, ORNG3872, p. 87.

Bd %

2' P

a r e arranged i n three stringers.

Each s t r i n g e r i s about 62 i n . long and consists of two Hastelloy N rods and a graphite section made up of

various pieces t h a t a r e joined by pinning and tongue-and-groove j o i n t s . The Hastelloy N rod has periodic-reduced sections 1 l/8 i n . long by

1/8 in. i n diameter and can be cut i n t o small t e n s i l e specimens a f t e r it i s removed from t h e reactor. Three s t r i n g e r s a r e joined together so t h a t they can be separated i n a hot c e l l and reassembled with one or more new s t r i n g e r s f o r r e i n s e r t i o n i n t o t h e reactor. The assembled s t r i n g e r s f i t i n t o a perforated Hastelloy N basket t h a t i s inserted i n t o an a x i a l p o s i t i o n about 3.6 i n . from t h e core center l i n e . A control f a c i l i t y i s associated with t h e surveillance program.

It u t i l i z e s a "fuel salt" containing depleted uranium i n a s t a t i c pot t h a t i s heated e l e c t r i c a l l y .

The temperature i s controlled by t h e MSRE

computer so that t h e temperature matches t h a t of t h e reactor.

Thus, these specimens a r e exposed t o conditions similar t o those i n t h e react o r except f o r t h e s t a t i c salt and t h e absence of a neutron flux. There i s another surveillance f a c i l i t y f o r Hastelloy N located outside t h e core i n a v e r t i c a l p o s i t i o n about 4.5 in. from t h e vessel. These specimens a r e exposed t o t h e c e l l environment of N2 + 2 t o 5% 0 2 . Materials The compositions of t h e two heats of standard Hastelloy N a r e given i n Table 1. These heats were a i r melted by t h e S t e l l i t e Division of Union Carbide Corporation.

Heat 5085 was used f o r making t h e c y l i n d r i c a l

portion of t h e reactor v e s s e l and heat 5065 was used f o r forming t h e top and bottom heads.

These materials were given a m i l l anneal of 1 h r a t

1177째C and a f i n a l anneal of 2 hr a t 900째C a t l O R N L after fabrication. The chemical compositions of t h e three modified a l l o y s are given i n Table 1. The modifications i n composition were made p r i n c i p a l l y t o improve t h e a l l o y ' s resistance t o r a d i a t i o n damage and t o bring about i

general improvements i n t h e f a b r i c a b i l i t y , weldability, and d u c t i l i t y .

6H. E. McCoy and J. R. Weir, Materials Development f o r Molten-Salt Breeder Reactors, ORNLTM-l854 (1967).

4 ,-

LiJ

Table 1. Chemical Analysis of Surveillance Heats

Heat 5065

Heat 5085

Content, w t 4 Heat 7320 Heat 67-551

Heat 67-504

7.3

7.2

7.0

6.94

3.9

3.5

< 0.05

0.02

0.05

16.5

16.7

12.4

12.2

0.065

0.052

12.0 0.059

0.60

0.58

0.03

0.02

0.07 0.010

0.08

0. I5

0.01

0.03

0.02

0.04

6.001

0.03

0.55

0.22

0.004

0.07 0.67 0.20 0.0043

S Al

0.007 0.01

0.01

< 0.05 0.17

< 0.02 0.002

0.004

0.003

0.02

0.15

< 0.01

0.65

0.028

0.12 < 0.001 0.0006

0.12 0.01

< 0.002 < 0.05

0.003

0.002 0.03

< 0.02

1.1

0.01

0.02

0.01

0.0016

0.0093

0.001

0.0004

< 0.0001

0.011

0.013

0.0002

0.0003

Hf B

< 0.1 < 0.1 0.0024

< 0.002

< 0.05

0.03

< 0.01

0.01 0.50

0.0038

0.00002

0.00003

0.0002

Alloys 67-551 and 67-504 were small, 100-lb heats made by t h e S t e l l i t e Division of Union Carbide Corporation by vacuum melting. finished t o 1/2 in. p l a t e by working at 870째C.

They were

We cut s t r i p s 1/2 in. by

1/2 in. from t h e p l a t e s and swaged them t o l/Ct-in.-diam rod.

!bo sec-

t i o n s of rod were welded together t o make 62-in.-long rods f o r fabricating t h e samples.

The rods were annealed f o r 1h r a t 1177째C i n argon

and then t h e reduced sections were machined.

Heat 7320 was a 5000-lb

melt made by t h e Materials Systems Division of Union Carbide Corporation.

Part of t h e heat was fabricated by t h e vendor t o 5/16-in.-diam rod and

was s i n t e r l e s s ground t o obtain t h e needed 1/4 in. stock.

The material

was annealed 1h r a t 1177째C and then t h e reduced sections were machined.

, k ,-

Test Specimens &.

The surveillance rods inside t h e core a r e 62 i n . long and those

E r) 'II

outside t h e v e s s e l a r e 84 i n . long.

They both a r e 1/4 i n . i n diameter with reduced sections 1/8 i n . i n diameter by 1 1/8 i n . long. After removal from t h e reactor, t h e rods a r e sawed i n t o small mechanical prope r t y specimens having a gage section 1/8 i n . i n diameter by 1 l/8 i n . long. The f i r s t rods were machined as segments and then welded together, but we described previously an improved technique i n which we use a milling c u t t e r t o machine t h e reduced sections i n t h e rod.3

This tech-

nique i s quicker, cheaper, and requires l e s s handling of t h e r e l a t i v e l y f r a g i l e rods than t h e previous method of making t h e rods i n t o segments. The standard Hastelloy N rods were machined and welded together and t h e modified alloys were prepared by milling.

' .

IRRADLATION CONDITIONS

The i r r a d i a t i o n conditions f o r t h e various groups of surveillance specimens t h a t have been removed are summarized i n Table 2.

The reac-

t o r operated from June 1965 u n t i l March 1968 w i t h a single change of f u e l salt i n which t h e r e was a 33% enrichment of 235U- After t h i s period of operation t h e uranium i n t h e f u e l was stripped by fluorination and replaced with 233U ( r e f . 7).

This charge of s a l t was used u n t i l t h e

present group of samples was removed.

Referring again t o Table 2, t h e

standard Hastelloy N i n t h e core was exposed t o both salts and t h e modif i e d Hastelloy N i n t h e core was exposed only t o t h e l a t t e r salt.

The same s a l t has been used i n t h e control f a c i l i t y throughout operation. The specimens outside the core (designated "vessel" specimens)

were exposed t o t h e c e l l envi

nment of

+ 2 t o 5% 02.

7P. N. Haubenreich and J. R. Engle, "Experience with t h e Molteng( 2), 118 (1970). S a l t Reactor Experiment, " Nucl. Appl. Technol. -

e 1

Table 2. ,

Sumnary of Exposure Conditions of Surveillance Samples‘

Group 1 Core Group 2, Haetelloy N Vessel Standard Core Standard Haetelloy Modified

Date inserted Date removed‘ Megawatt-hour on EiLsRE a t time of insertion Megawatt-hour on EiLsRE at time of r e m 1 Temperature, ‘C Time a t temperature, hr Peak fluence, neutrons/cm2 Thermal (< 0.876 ev) Epithermal (> 0.876 ev) (> 50 kev) (> 1.22 MeV) (> 2.02 MeV) Heat Designations

Group 3, HasteUoy N Core Core Vessel Standard Modified Standard

Group 4, Hastelloy N Core Core Vessel Standard Modified Modified

9/8/65 7/28/66 0.0066

9/13/66 5/9/67 8682

8/24/65

9/13/66

6/5/67

6/5/67 0

4/3/68 8682

4/3/68 36,247

8/24/65 5/7/68 0

9/13/66 6/69 8682

4/ID/68 6/69 72,441

5/7/68 6/69 36,247

8682

36,247

72,441

72,4441

72,441

92,805

650 f 10 482800

550 f 10 5500

650 f 10

u,m

15,289

650 f 10 9789

650 f 10 20,789

650 f 10 22,533

650 f 10 7244

650 f 10 17,033

1.3 x 1019 2.5 X 1019 2.1 x 10’’ 5.5 X 10l8 3.0 X 10l8

9.4 x 1020 2.8 X 1021 8.5 x lozo 2.3 X lozo 1.1 X IDzo

5.3 X 10’’ 1.6 x loz1 4.8 x lozo 1.3 X 10’’ 0.7 X lozo

2.6 X loi9 5.0 x U)l9 4.2 x 1019 1.1 X lo1’ 6.0 X ID18

1.5 X loz1 3.7 x loz1 1.1 x loz1 3.1 x 1020 1.5 X lozo

5.1 9.1 1.1 0.8 0.4

5065 5085

5065 5m5

67-502 67-504

5065 5085

7320 67-551

1.3 x IDzo 4.1 3.8 x lozo 1.2 1.2 x 1020 3.7 3 . 1 x 1019 1.0 1.6 X lo1’ 0.5

5081 5085

x 1020 x lo2’ x 1020 x 1020

x lozo

21545 2U54

aInformetion ccrmpiled by R. C. Steffy, Reactor Division, OWL, July 1969. Rwieed for full-paver operation a t 8 Mw.

lozo x mZo x io20 x lozo x lozo X

2.5 3.9 3.3 8.6 3.5

1019 1019 1019

x x 10l8 x 10l8

67-504

Testing Techniques 5

The laboratory creep-rupture t e s t s of unirradiated control specimens were run i n conventional creep machines of t h e dead-load and leverarm types.

The s t r a i n was measured by a d i a l indicator that showed the

t o t a l movement of t h e specimen and p a r t of t h e load t r a i n .

The zero

s t r a i n measurement was taken immediately a f t e r t h e load was applied. The temperature accuracy was +0.75$,

t h e guaranteed accuracy of t h e

Chromel-P-Alumel thermocouples used. The p o s t i r r a d i a t i o n creep-rupture tests were run i n lever-arm machines t h a t were located i n hot c e l l s .

The s t r a i n was measured by an

extensometer with rods attached t o t h e upper and lower specimen grips. The r e l a t i v e movement of these two rods was measured by a l i n e a r d i f f e r e n t i a l transformer, and t h e transformer s i g n a l was recorded.

The

accuracy of t h e s t r a i n measurement i s d i f f i c u l t t o determine.

The exten-

someter (mechanical and e l e c t r i c a l portions) produced measurements t h a t could be read t o about +0.02$ s t r a i n ; however, other f a c t o r s (temperat u r e changes i n t h e c e l l , mechanical vibrations, etc. ) probably combine t o give an o v e r a l l accuracy of &0.1$1s t r a i n .

This i s considerably b e t t e r

than t h e specimen-to-specimen reproducibility t h a t one would expect f o r r e l a t i v e l y b r i t t l e materials.

The temperature measuring and c o n t r o l

systemwas t h e same as t h a t used i n t h e laboratory with only one exception.

I n t h e laboratory, t h e control system was s t a b i l i z e d a t t h e

desired temperature by use of a recorder w i t h an expanded scale. t e s t s i n t h e hot c e l l s , t h c o n t r o l l e r without t h e aid

In t h e

o n t r o l point was established by s e t t i n g t h e t h e expanded-scale recorder.

This e r r o r

and t h e thermocouple accuracy combine t o give a temperature uncertainty of about kl$. The t e n s i l e t e s t s were run on Instron Universal Testing Machines. The s t r a i n measurements were taken from t h e crosshead t r a v e l and genera l l y a r e accurate t o

*z$ s t r a i n .

The t e s t environment

air i n a l l cases.

Metallographic examina-

t i o n showed t h a t t h e depth of oxidation was small, and we f e e l t h a t the! environment did not appreciably influence t h e t e s t results.

8 EXPERIMENTAL RESULTS

V i s u a l and Metallographic Examination

W. H. Cook was i n charge of t h e disassembly of the core surveillance fixture.

As shown i n Fig. 1, t h e assembly was i n excellent mechan-

i c a l condition when removed.

The Hastelloy N samples were more

discolored than noted previously; however, surface marking such as numbers were r e a d i l y visible.

The detailed appearance of t h e s t r i n g e r

has been described previously by Cook.8

The Hastelloy N surveillance

rods located outside t h e core were oxidized, but t h e oxide was tenacious. Metallographic examination of t h e Hastelloy N s t r a p s t h a t held t h e graphite and metal together revealed intergranular cracks. A t y p i c a l

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

0-3872,

Fig. 1. Overall V i e w of USRE Surveillance Assembly Removed A f t e r Run 18. Parts of t h i s assembly had been exposed t o t h e salt f o r 22,533 h r a t 650째C. The center portion i s graphite and t h e long rods are Hastelloy N.

crack i s s h m i n Fig. 2 and extends t o a depth of about 3 mils.

similar s t r a p on t h e modified samples which had been i n t h e reactor f o r

7244 hr had cracks t o a depth of about 1.5 mils.

These s t r a p s are about

0.020 i n . t h i c k and they l i k e l y encountered some deformation while being removed.

However, t h e cracks were quite uniformly spaced on both sur-

faces of the straps, and t h e i r general appearance attests t o a general corrosion t h a t rendered t h e grain boundaries extremely b r i t t l e .

Examina-

t i o n of unirradiated control straps f a i l e d t o reveal a similar type of cracking.

Fig. 2. Typical Microstructure of a Hastelloy N (Heat 5055) After Exposure t o t h e MSRE: Core f o r 22,533 hr a t 650째C. This materialwas used f o r s t r a p s f o r t h e surveillance assembly. As polished. 500X. These observations l e d t o t h e examination of tabs from t h e surveillance stringers.

Small sections were cut from t h e centers of t h e

Hastelloy N surveillance rods.

Typical photomicrographs of heat 5065

after exposure t o t h e core f o r 22,533 hr are shown i n Fig. 3 .

The unetched view i n Fig. 3(a) shows t h e surface layer t h a t led t o the discolored appearance and a single grain boundary t h a t is v i s i b l e .

Much

of t h e surface layer looks metallic, but t h i s i s d i f f i c u l t t o judge on

Fig. 3. Typical Photomicrographs of Hastelloy N (Heat 5065) Exposed t o t h e MSRE Core f o r 22,533 h r a t 65OOC. ( a ) As polished. (b) Etchant: aqua regia. 500~.

such a t h i n f i l m .

The etched view i n Fig. 3(b) reveals some carbide

p r e c i p i t a t i o n near t h e surface and grain boundaries t h a t are generally Heat 5085 was exposed an i d e n t i c a l time and

lined with carbides.

t y p i c a l photomicrographs a r e shown i n Fig. 4.

The unetched view i n

Fig. 4(a) shows a modified grain bouhdary s t r u c t u r e t o a depthiof about

2 mils. Etching [Fig. 4(b)] reveals a grain boundary network of carbides. The g r a i n boundaries near t h e surface seem t o etch d i f f e r e n t l y f r o m t h e r e s t of t h e sample, but l i t t l e more can be said. We interpreted these observations as being indicative of some corrosion and performed one further crude experiment t o reveal the depth of t h i s attack. One t e n s i l e sample had been cut too short f o r testing, and we bent t h e remaining portion i n a vise. The s k l e was then sectioned and examined metallographically; the resulting photomicrographs a r e shown i n Fig. 5. The tension s i d e cracked t o a depth of about 4 mils whereas the compression s i d e did not crack.

Both sides etched abnormally t o a

depth of about 4 mils. ZI

Samples of heats 5065 and 5085 t h a t were exposed t o the s t a t i c barr e n s a l t i n t h e control f a c i l i t y were examined.

Figure 6 shows t y p i c a l

photomicrographs of heat 5065 a f t e r exposure f o r 22,533 hr.

There i s

some surface roughening, but no structure modification near t h e surface such as t h a t shown i n Fig. 3 f o r t h e sample *om t h e reactor.

Likewise,

heat 5085 (Fig. 7) showed some surface e f f e c t s t h a t were minor compared with i t s i r r a d i a t e d counterpart i n Fig. 4. Thus, t h e r e i s l i t t l e doubt that t h e samples i n t h e core experienced some modifications, apparently t o a depth of 3 t o 4 mils.

This a l t e r s

up t o about l 2$ of t h e sample cross section and can be expected t o influence t h e mechanicalproperties. We s h a l l dwell further on t h i s very important subject l a t e r i n t h i s report. A sample of heat 5085 f r o m t h e core was examined by transmission

electron microscopy.

This sample had received s u f f i c i e n t thermal fluence

t o transmute about 97$ of the I 0 B t o helium. obvious i n Fig. 8.

Another point of concern was t h e formation of voids

i n the laaterial due t o fast neutrons. c

bles and dislocations were present.

The helium bubbles are

No defects other than helium bub-

u Q

Fig. 4. Typical Photomicrographs of Hastelloy N (Heat 5085) Exposed t o t h e MSRE Core for 22,533 hr a t 650째C. (a) As polished. (b) Etchant: glyceria regia. 5OOx.

Fig. 6. Typical Photomicrographs of Heat 5065 After l&osure S t a t i c Barren Fuel S a l t f o r 22,533 h r a t 650째C. (a) As polished. (b) Etched. Etchant: glyceria regia. 500X.

to f

. 1

Fig. 7. Typical Photomicrographs of Heat 5085 Af'ter Exposure t o S t a t i c Barren Fuel S a l t for 22,533 hr a t 65OOC. (a) As polished. (b) Etched. Etchant: glyceria regia. 500x.

Fig. 8 . Transmission Electron Micrograph of Hastelloy N (Heat 5085) Exposed t o t h e MSRE Core f o r 22,533 h r a t 650째C. Irradiated t o a thermal neutron fluence of 1.5 X 10" netrtrons/cm2 and a f a s t neutron fluence of 3.1 X ' ' 0 1 neutrons/cm2 (> 1.22 MeV). The white spots a r e helium bubbles t h a t are located on a twin boundary. 25,OOOX. Reduced 16%.

Mechanical Property Data

- Standard Hastelloy N

Two heats of standard Hastelloy N were exposed t o t h e MSRE core environment f o r 22,533 h r a t 650째C and received a thermal neutron fluence of 1.5 X 1021 neutrons/cm2.

Similar control samples were exposed t o

s t a t i c barren f u e l s a l t f o r a corresponding length of time.

The results

of t e n s i l e t e s t s on heat 5085 a r e summa;rized i n Tables 3 and 4 f o r unirradiated and i r r a d i a t e d samples, respectively.

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

are shown as a function of test temperature i n Fig. 9. With increasing temperature t h e unirradiated samples exhibit f i r s t an increase i n frac-

ture s t r a i n , then a sharp decrease, and then an increase.

The irra-

diated samples follow t h e same general p a t t e r n except f o r t h e absence

Table 3.

Specimen Number

Results of Tensile Tests on Control Samples of Heat 5085&

Test Temperature

("C>

(min-1)

stress^ Ultimate

Yield Tensile

Elongation, $ Uniform Total

Reduction i n Area

( 4)

0.05

48,100 108,500 39,300 101,400 35,400 94,900 35,200 98,500 34,000 92,800 35,800 90,100 32,700 88,900

40.5 44.7 48.2 48.6 48.0 37.5 51.6

40.6 45.1 48.9 49.0 48.8 38.2 52.2

32.8 33.4 39.3 36.3 40.0 28.8 37.1

True Fracture

Strain

($1 40 41 50 45

10409 10408 10407 10406 10405 10416

25 200 400 400 500 500 550

10417

550

0.002

36,500

79,500

29 -4

30.1

24.1

10418 10420 10421 10422 10404 1W12 10423 10424 10426 10425

600 600 650 650 650 650 760 760 850 850

0.05

33,400 35,200 32,600 34,300 35,700 30,000 33,900 31,600 33,600 24,100

82,200 69,000 75,200 65,600 64,500 59,700 64,900 52,000 50,600 26,300

37.3 25.7 31.0 23.3 24.9 15.2 25.6 12.2 10.0 4.2

37.9 26.9 32.0 23.9 25.5 22.2 27.5 27.0 33.0 36.7

28.6 19.1 28.9 23.9 19.7 21.2 19.7 31.6 37.1 25.3

34 21 34 27 22 24 22 38 46 29

10410

Strain Rate

0.05 0.05 0.05 0.002 0.05 0.002

0.002 0.05 0.002 0.0002 0.000027 0.05 0.002 0.05 0.002

aAnnealed 2 hr at 900째C; exposed t o a static v e s s e l of barren Ate1 salt f o r 22,533 hr at 650째C.

-51

34 46

Table 4 .

Specimen Number 8806 8805 88od 8803 8802 8801

8812 8813 88U 8816 8817 8818 8800 8799 8819 8820 8821 8822

Postirradiation Tensile Properties of Hastelloy N (Hea, 5085)'

Test Temperature ("C)

25 200 400 400 500 500 550

Stress, p s i

Strain

Rate (min-1)

Yield Tensile

53,900 44,800 0.05 41,600 0.002 41,300 0.05 40,800 39,400 0.002 37,900 0.05 36,200 0.002 0.05 36,700 0.002 39,000 0.05 36,400 0.002 37,200 0.002 35,200 0.000027 32,900 0.05 30,700 0.002 35,100 0.05 32,100 22,200 0.002

89,000 85,600 80,900 80,100 73,200 61,100 62,600 49,100 56,200 48,500 50,800 42,800 37,700

0.05 0.05

550

600 600 650 650 650 650 760 760 850 850

34,500

36,300 35,100 32,700 22,200

Elongation, $ Uniform Total 22.0 26.4 30.2 28.4 24.8 ll.7 14.4 5.8 9.8 4.5

8.8 4.7 2.1 1.2 6.1 1.7 2.3 1.6

in Area ($1

True Fracture Strain

($1

22.1 27.0 30.5 29.2 25.0 12.4

19.6 26.4 24.3 26.8 23.2 13.9

22 31 28 31 26

U.9

14.3

6.2 10.5 4.9 9.3 5.0 2.4 1.4 6.4 2.0 2.3 1.7

9.1 13.3 6.7 9.1 5.8 2.8 6.1 6.4 2.5 2.1 2.6

15 10

U 7 10 6 3 6 7 3 2 3

a Annealed 2 hr at 900째C prior t o insertion i n reactor. Irradiated t o a thermal fluence of 1.5 X 1021 neutrons/cm2 over a period of 22,533 hr a t 650째C.

I 7o

STRAIN RATE UJIRRAM4TEO

RRACWTEO

cmin-9 0.05 0.002

TEST TEMPERATURE FC)

Fig. 9. Fracture S t r a i n s of Hastelloy N (Heat 5085) After Removal from t h e MSFE and from t h e Control F a c i l i t y . A l l samples annealed 1hr a t 900째C before i r r a d i a t i o n f o r 22,533 h r a t 650째C t o a thermal fluence of 1.5 X loz1 neutrons/cmz.

,.-.

19 of a d u c t i l i t y increase a t high temperatures. I

However, the levels of

t h e f r a c t u r e s t r a i n s are lower f o r the i r r a d i a t e d material over t h e e n t i r e temperature range.

For both materials t h e f r a c t u r e s t r a i n

decreases with decreasing s t r a i n r a t e .

Another difference i n behavior

between t h e i r r a d i a t e d and unirradiated samples i s t h a t a t t h e highest t h e f r a c t u r e s t r a i n begins i t s precipitious drop

s t r a i n rate (0.05 min")

a t a lower temperature f o r t h e i r r a d i a t e d material. Further c h a r a c t e r i s t i c s of t h e e f f e c t s of i r r a d i a t i o n on t h e tens i l e properties of heat 5085 a r e apparent when t h e r a t i o s of t h e irraThe yield

diated and unirradiated properties a r e compared (Fig. lo).

s t r e s s i s frob 10 t o 20% higher f o r the irradiated material a t t e s t t e m peratures up t o 760째C.

The ultimate t e n s i l e s t r e s s i s about 20% lower

f o r t h e i r r a d i a t e d material and drops even further as t h e t e s t tempera-

ture i s increased above 500째C.

The f r a c t u r e s t r a i n s of t h e i r r a d i a t e d

samples a r e about 50% of those of the unirradiated samples up t o about 5OO0C, above which t h e reduction i s even greater.

ORNL- DWG 70 729i

1.4

t.2

5 '

0.4

0.2

0 0

. bi

100

200

300

400 500 600 TEST TEMPERATURE ( 'C)

700

800

900

Fig. lo. Comparison of t h e Tensile Properties of Control and Irradiated Hastelloy N (Heat 5085). Samples i r r a d i a t e d t o a fluence of 1.5 X 10" neutrons/cm2 over a period of 22,533 hr a t 650째C. Tested a t a s t r a i n rate of 0.05 min'l.

20 . The r e s u l t s of t e n s i l e t e s t s on t h e samples of heat 5065 a r e summarized i n Tables 5 and 6. The f r a c t u r e s t r a i n s of these sanqles a r e s h a m i n Fig. 11as a function of t e s t temperature.

hd t

The r e s u l t s a r e

q u i t e similar t o those shown i n Fig. 9 f o r heat 5085. A notable except i o n i s t h e much higher fracture s t r a i n s a t low temperatures of heat 5065

after irradiation. The r a t i o s of t h e i r r a d i a t e d and unirradiated tens i l e properties a r e shown i n Fig. 12. The yield s t r e s s was not a l t e r e d appreciably by irradiation. The ultimate t e n s i l e s t r e s s was decreased about 10% by i r r a d i a t i o n a t low t e s t temperatures and up t o 35% a t high temperatures.

Similarly, the f r a c t u r e s t r a i n was reduced about 10% a t

low temperatures, and t h i s reduction progressed t o about 85% a t high t e s t temperatures. The progressive change i n t h e fracture s t r a i n w i t h increasing fluence i s i l l u s t r a t e d i n Fig. I3 f o r heat 5085.

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

has been reduced over the e n t i r e range of t e s t temperatures investigated. A similar trend was noted a t a slawer s t r a i n r a t e of 0.002 min-l

(Fig.

U),although t h e absolute values of t h e f r a c t u r e s t r a i n s were

lower than noted i n Fig. I3 a t t h e higher s t r a i n r a t e of 0.05 min'l. These samples have experienced various holding times a t 650째C and some of t h e property changes can be a t t r i b u t e d t o thermal aging.

The frac-

t u r e s t r a i n s f o r several s e t s of control samples of heat 5085 a r e compared i n Fig. l5. A t low t e s t temperatures and at 760째C and above t h e f r a c t u r e s t r a i n seems t o show a progressive decrease with increasing holding time a t 650째C. A t intermediate temperatures t h e behavior i s more complex.

A t a t e s t temperature of 65OoC, t h e r e s u l t s indicate a

progressive decrease i n f r a c t u r e s t r a i n up t o an exposure time of

15,289 hr and then an increase w i t h f u r t h e r aging time. Heat 5065 was not included i n one set of surveillance samples, but there are s u f f i c i e n t data t o follow t h e property changes with fluence and aging time a t 650째C. The f r a c t u r e s t r a i n i s shown i n Fig. 16 as a function of temperature f o r various fluences.

A t low t e s t temperatures,

excluding t h e one apparently anomalous point, t h e f r a c t u r e s t r a i n was

a c t u a l l y higher f o r i r r a d i a t i o n t o fluences of 1.3 and 2.6 X lo1' neutrons/cm*.

Higher fluences reduced t h e f r a c t u r e s t r a i n at low t e s t

temperatures, but not t o values as law as noted i n Fig. I3 f o r heat 5085.

21 Table 5. Specimen Number

Results of Tensile Tests on Control S w l e s of Heat 5065'

Test Temperature (OC)

10384 10388 10392 10383 10391 10378 10395 10375 10377 10393 10382 10394 10397 1MOO 10385 10398 10390 10379

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

Strain Rat e (min-1) 0.05 0.05 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.0002 0. oooO27 0.05 0.002 0.05 0.002

Stress, p s i

61,200 44,000 42,600 43,900 42,200 45,800 42,500 43,800 41,800 39,400 41,500 39,200 41,800 42,900 32,900 41,100 35,500 24,800

El0

ation

Tensile

Unifz

126,500 107,900 102,000 106,000 99,400 98,700 98,800 84,600 93,800 76,700 81,900 71,700 71,200 60,800 70,500 42,100 49,500 25,100

48.5 47.5 48.9 47.5 48.1 45.5 48.2 25.5 42.0 25.6 26.4 23.1

18.8 7.5 24.3 5.7 8.7 2.9

Tkal 49.8 49.4 50.8 47.7 49.6 46.2 49.9 26.1 43.0 26.126.8 23.6 19.6 19.7 27.3 39.6 38.1 56.5

a Annealed 2 h r a t 900째C prior t o insertion i n the reactor. vessel of barren fuel s a l t for 22,533 hr a t 650째C.

Reduction True in Area Fracture Strain ( 4) ( 4) 36.77 40.16 45.31 39.30 37.54 34.08 38.M 27.51 34.08 34.29 23.8.: 20.89 17.88 24.57 22.85 45.48 45.31 47.15

46 51 60 50 47 42 48 32 42 42 27 23 20 28 26 61 60 64

Exposed t o a s t a t i c

r I

Table 6.

Specimen Nwnber 8833 8832 8831 8830 8829 8828 8839 8840 8841 8842 8843 8W 8827 8826 8845 8846 8847 8848 L

Postirradiation Tensile Properties of Hastelloy N (Heat 5065)& Test Temperature

Strain Rate (min-i)

25 200

0.05 0.05

400

0.05

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

0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.0002 O.OOO27 0.05 0.002 0.05 0.002

Stress, psi Yield 56,700 46,000 43,700 43,200 43,600 39,800 41,200 37,800 40,500

Ultimate Tensile

109,500 99,900 94,300 91,300 82,100 63,200 68,700 57,700 60,100 41,000 50,400 39,700 52,100 40,000 42,800 34,300 36,100 28,400 36,900 37,800 41,400 35,200 35,200 35,400 35,400 20,300 20,300

Reduction Elongation, $ Uniform Total 42.5 42.8

42.8 44.6

39.0

39.3

35.7 29.1 10.9 15.5 7.1 8.6 4.3 6.0 3.0 2.1

37.2 29.5 n.3 15.7 7.2 9.0 4.5 6.2 3.1 2.2 1.2 2.7 1.3 1.6 1.0

1.1

2.5 1.3 1.5 1.0

Area

3.32 4.44 31.22 27.31 22.98 10.70 16.83 8.93 9.30 7.40 7.37 4.89 3.30

Fracture True

Strain 3 5 37 32 26 11

18 9 10 8 8 5 3

1.1

3.33 2.86 2.22 0.16

3 3 2 0

a Annealed 2 hr a t 900째C prior t o insertion i n the reactor. Irradiated t o a thermal fluence of 1.5 X lo2' neutrons/cm2 over a period of 22,533 hr a t 650째C.

22 ORNL- DWG m-7292

W *

50 Y

3 - 40 t a a

k30

920 lo

0 0

f00

200

300

400

500

600

700

800

900

TEST TEMPERATURE (OC)

Fig. ll. f i a c t u r e Strains of Hastelloy N (Heat 5065) After Removal f r o m t h e MSRE and f'romthe Control F a c i l i t y . A l l samples annealed 2 h r a t 900째C before i r r a d i a t i o n f o r 22,533 hr a t 650째C t o a fluence of 1.5 X loz1 neutrons/cmz.

ORNL-DWG 70-7293

1.2

1.0

W 0

s6 0.8 2

e 0.6 2

z= 0

0.4

sa 0.2

n "

(00

200

300

400 500 600 TEST TEMPERATURE PC)

700

800

900

Fig. 12. Comparison of t h e Tensile Properties of Control and Irradiated Hastelloy N (Heat 5065). Samples i r r a d i a t e d t o a fluence of 1.5 X 10" neutrons/cm' over a period of 22,533 hr a t 650째C. Tested a t a s t r a i n rate of 0.05 m i n - l .

23 80

400

200

300

70 c

ORNL- OWG 70- 7294

ANNEALED 2 h r AT900°C 1.3 x 40'9 neulrons/cm2, 4400 hr 0 2.6 x neutrons/cm2, 20,789 hr v 1.3 x 40' neutrons/cm2, 4800 hr 0 9.4 x1OZo neutrons/cm2, 45,289 h i 0 4.5 X 40" neutrons/cm2. 22.533 hr 0

2 50

I-

* 40 W a

3 Io

E 30 20

n ., 0

400 500 600 TEST TEMPERATURE (OC)

700

000

900

Fig. 13. Fracture Strains of Hastelloy N (Heat 5085) After Irrad i a t i o n t o Various Thermal Fluences i n t h e MSRE. Tested a t a s t r a i n rate of 0.05 rnin-l.

LL w

5 20 3

* (00

200

ORNL-WG 69-4467R

22,533

300

400 500 600 TEST TEMPERATURE C C )

700

800

9€Jcl

Fig. Ik. Postirradiation Tensile Properties of Hastelloy N (Heat 5085) After Exposure t o Various Neutron Fluences. Tested a t a s t r a i n rate of 0.002 min-1.

..0

400

2 0 0 3 0 0 4 0 0 5 0 0 6 0 0 7 0 0 8 0 0 9 0 0 TEST TEMPERATURE ('c)

F i g . 15. Variation of t h e Tensile Properties of Hastelloy N (Heat 5085) with Aging Time i n Barren Fuel S a l t a t 650째C. Tested a t a strain rate of 0.05 min-l.

ORNL- DWG 70- 7295

-E

3 40 t; e W

Ea 30 e

0 0

(00

200

300

400 500 600 TEST TEMPERATURE CC)

700

800

900

Fig. 16. Variation o f t h e Postirradiation Tensile Properties of Hastelloy N (Heat 5065) with Thermal Neutron Fluence.

A t t e s t temperatures above 550째C t h e f r a c t u r e s t r a i n decreases progres-

s i v e l y with increasing fluence.

A s shown i n Fig. 17 some of these

changes i n f r a c t u r e s t r a i n can be a t t r i b u t e d t o thermal aging a t 650째C. Except a t t e s t temperatures above 75OoC, t h e f r a c t u r e s t r a i n i s lowest f o r the material aged 15,289 h r and i s improved a f t e r aging 22,533 hr. The changes i n properties a t l o w t e s t temperatures a r e l e s s than those f o r heat 5085 (Fig. 15).

ORNL- DWG 70- 7296

W a

30 a a

n v

400

200

300

400 500 600 TEST TEMPERATURE CC)

700

000

900

Fig. 17. Effects of Thermal Aging a t 650째C on t h e Tensile Propert i e s of Hastelloy N (Heat 5065) a t a S t r a i n Rate of 0.05 min-l. A s discussed previously i n t h i s s e r i e s of reports, the changes i n

f r a c t u r e s t r a i n a t low temperatures were not expected.

Although t h e f r a c t u r e s t r a i n s have not reached values below 2046, we a r e s t i l l i n t e r ested i n i t s progression. Fig. 18.

Our experience t o date i s summarized i n

If t h e noted e f f e c t s were due s i m p l y t o thermal aging, then

t h e r e s u l t s should c o r r e l a t correlation does not e x i s t .

i t h t h e time a t 650째C.

Obviously such a

Where data a r e available f o r p a i r s of irra-

diated and unirradiated samples, the i r r a d i a t e d sample has the lower

26 70

ORNL-OWG 70-7297

I HEAT

IRRADIATED

UNIRRADIATED

12 16 TIME AT 650T (hr)

(Xl03)

Fig. l8. Variation of t h e Fracture S t r a i n a t 25째C with Annealing T h e and Thermal Fluence. fracture strain. embrittlement.

This indicates t h a t i r r a d i a t i o n has a r o l e i n t h e

We previously showed t h a t t h e d u c t i l i t y could be

recovered by an anneal of 8 hr a t 8 7 1 째 C and concluded t h a t t h e changes

m u s t be associated with carbide precipitation.

The higher susceptibil-

i t y of heat 5085 t o t h i s type of embrittlement is not understood, since heat 5065 a c t u a l l y has a higher carbon concentration (Table 1, p. 4). The data i n Fig. 18 defy extrapolation, so one cannot conclude whether t h e room temperature embrittlement i s l i k e l y t o become worse.

In these discussions of t e n s i l e properties we have emphasized t h e changes i n f r a c t u r e s t r a i n and made l i t t l e mention of strength changes. Some pertinent data a r e summrized i n Table 7 f o r heat 5085.

The sam-

p l e s were t e s t e d a t 25 and 650째C and include three h i s t o r i e s : annealed, (2) thermally aged, and (3) irradiated.

(1)as The changes i n yield

strength a r e l i k e l y significant, but t h e main point i s that they a r e

9H. E. McCoy, An Evaluation of t h e Molten-Salt Reactor Experiment Hastelloy N Surveillance Specimen- Third Group, ORNLTIG2647 (19701, p. 17.

27 Table 7. Comparison of t h e Tensile Properties of Heat 5085 Before and After Irradiationa Heat Treatment

Ultimate Tensile Stress, p s i macture Strain, 4 Yield Stress, p s i a t 25OC a t 650°C at a o c at 6 5 0 " ~ at 25°C a t 650°C

Annealed 2 hr a t 900°C

51,500

29,600

120,900

75,800

53.1

33.7

Annealed 2 hr at 900°C and aged 22,533 hr at 650°C

48,100

32,600

108,500

75,200

40.6

32.0

Annealed 2 hr a t 900"C, irradiated f o r 22,533 h r a t 650°C t o a thermal fluence of 1.5 X 1 O 2 l neutrons/cm2

53,900

36,400

89,000

50,800

22.1

9.3

S t r a i n r a t e of 0.05 min'l.

small considering t h e long thermal h i s t o r y of 22,533 hr.

The ultimate

t e n s i l e strength i s reduced due t o the lower f r a c t u r e s t r a i n and t h e resulting f a c t t h a t t h e t e s t i s interrupted before it reaches the high ultimate strength noted f o r t h e as-anneaJ-ed material. a r e s h a m i n Table 8 f o r heat 5065.

Similar r e s u l t s

The yield strength i s increased by

aging, but the main e f f e c t s a r e on t h e fracture s t r a i n s and t h e ultimate t e n s i l e strengths. Table 8. Comparison of t h e Tensile Properties of Heat 5065 Before and Af'ter Irradiationa

Heat Treatment

Yield Stress, p s i

Fracture Strain, $

a t 25°C a t 650°C at 25°C at 6 5 0 " ~ a t 25°C a t 650°C

Annealed 2 hr a t 900°C Annealed 2 h r a t 900°C and 61,200 aged 22,533 hr a t 650°C Annealed 2 hr a t 900°C, irradiated f o r 22,533 hr a t 650°C t o a thermal fluence of . 1.5 X 1 O 2 I neutrons/cm2

ultimate Stress, Tensile psi

56,700

aS t r a i n r a t e of 0.05 min-l.

32,100 41,500

126,400 126,500

81,200 81,900

55.3 49.8

34.3 26.8

39,700

109,500

52,100

42.8

6.2

28 The r e s u l t s of creep-rupture t e s t s on heats 5085 and 5065 from the fourth group of surveillance samples a r e sumarized i n Tables 9 and 10.

u *

The r e s u l t s a r e most i n t e r e s t i n g when they a r e campared with those obtained on previous groups of surveillance samples so t h a t t h e progress i v e changes i n properties can be noted.

The stress-rupture properties

of heat 5085 a r e shown i n Fig. 19. The rupture l i v e s of t h e last group of samples (1.5

lo2' thermal neutrons/cm2) a r e not detectably d i f f e r -

ent from those irradiated previously t o a thermal fluence of 9.4 X 1020

I n t h e unirradiated condition, the rupture l i v e s f o r the samples aged 22,533 hr are s l i g h t l y l e s s than i n t h e as-annealed condineutrons/cm2.

t i o n , but greater than those observed f o r samples aged f o r l e s s e r times. Table 9.

Test Number

Specimen Number

Creep-Rupture Tests on Heat 5085 a t 650째C

Stress (psi)

Rupture Ufe

(hr)

Rupture

Strain

( 4)

Reduction i n Area

Minimum Creep

23.4 28.5 9.4

( 4)

7978

10406

Unirradiat eda 55 000 22.1 21.0

7897

10415

40,OOO

7896' 7895b

1mU. 10413

2187.1

5.40

7894b 7893b

10412 104U

32,400 27,000 21,500

2064.9

2.44

17,OOO

2000.0

2.10 1.1

0.288 0.0299 0.0045 0.0026 0.ooll

0.0007

500.1 2500.7

24.5

12.7

Irradiated'

R-969

8811

40,000

0.6

1.1

0.78

R- 958

8810

32,400

31.6

0.54

R-956

8809

27 ,000

R- 957 R- 964

8808

21,500

60.2 362.0

0.57 0.63

0.0067 0.0055 0.0010

8807

17 ,000

426.4

0.69

0.0004

a Annealed 2 hr a t 900째C, exposed t o s t a t i c barren f u e l s a l t f o r 22,533 hr a t 650째C.

bDiscontinued p r i o r t o f a i l u r e . C

Annealed 2 hr a t 900째C, exposed t o MSRE core f o r 22 533 h r a t 65OoC, received thermal fluence of 1.5 X lo2' neutrons/cm

29 Table 10. Creep-Rupture Tests on Heat 5065 a t 650째C

Test Number

7978 7892 78937, 7890b 788gb 7888

Specimen Number

104.06 10388 10387 10386 10385 10384

Stress (psi)

Rupture Life

Rupture Strain

( 4)

Reduction i n Area ( 4)

Unirradiat eda 22.1 21.0 55,000 40,000 651.2 36.3 2144.3 44.7 32,400 8.8 2l87.1 27,000 2373.3 3.6 21,500 2000.0 2.8 17,000

23.4 29.9 37.7 7.4 2.2 0

Minimum Creep

0.34 0.0331 0.0096 0.0040

0.0013

Irradiat edc

R-966 R-962 R- 954 R-960 R- 963

8838 8836 8836 8835 8834

40,000 32,400 27,000 21,500 17,000

0 22.7 40.5 46.2 1567.3

0.0075 0.0061 0.0051 0.0004

0.25 0.44 0.47 0.86

aAnnealed 2 h r a t 900째C, exposed t o s t a t i c barren f u e l s a l t f o r 22,533 h r a t 650째C. bDiscontinued p r i o r t o f a i l u r e . C

Annealed 2 h r a t 9OO"C, exposed t o MSRE core f o r 22 533 h r a t 650"C, received thermal fluence of 1.5 X 1021 neutrons/cm 1

The minimum creep r a t e s a r e shown as a function of s t r e s s i n Fig. 20 f o r heat 5085.

A s noted previously, neither i r r a d i a t i o n nor

aging has a detectable e f f e c t on the creep r a t e . The data deviate from t h e l i n e shown i n Fig. 20 a t both extremes. A t high s t r e s s e s , t h e irrad i a t e d s a q l e s f a i l a t such l o w s t r a i n s t h a t a minimum creep r a t e i s not established over long enough eriod f o r measurement.

A t low s t r e s s

l e v e l s t h e data deviate due to t h e semilog p l o t . The f r a c t u r e s t r a i n i s t h e parameter most a f f e c t e d by i r r a d i a t i o n , and a v a r i a t i o n of t h i s parameter with minimum creep r a t e i s shown i n Fig. 2 1 f o r heat 5085 a t tion o f t h e fracture str Tensile and creep t s t r a i n r a t e s and s t r e s s e s .

There has been a continual deterioraincreasing thermal fluence. e run a t 650'C a t s e v e r a l d i f f e r e n t

e fracture strains fromthese t e s t s are

p l o t t e d together i n Fig. 22 although d i f f e r e n t parameters a r e controlled

30 ORNL-OWG 69-

t 0

o 1.3

x 1019

A 2.6 x 1019 o 9.3 x 4OZo 0 9.4 x 1020 0 1.5 x (02'

I 1.1111111 fo-'

400

neutrons/cm2 (MSRE) AN<

20,709 4000 (5,289 22.533

3-5 x 40"

neutrons/cm2 (ORR)

I 11111111

io'

402

103 .-

' RUPTURE TIME (hi)

Fig. 19. Postirradiation Stress-Rupture Properties of MSRE Surveillance Specimens (Heat 5085) a t 650째C.

ORNL-DWG 69-4474R2

--I

9 v)

c v)

0 40-3

40-2 io-' MINIMUM CREEP RATE (%

/ hr)

$00

40'

Fig. 20. Minimum Creep Rate of Hastelloy N (Heat 5085) Surveillance Specimens from the MSRE a t 650째C.

_-.

31 ORNL-DWG 6 9 - 4 4 7 2 R

io-4

10-3

40-2 lo-‘ MINIMUM CREEP RATE (%/hr)

I00

4 0’

Fig. 21. Variation of Fracture S t r a i n w i t h S t r a i n Rate f o r Hastelloy N (Heat 5085) Surveillance Specimens a t 650°C.

ORNL-DWG 70-7298

I I I111111 UNIRRADIATED

(1’

30 -

s-”

e--

2 8 V a a LL

40-2

to‘

lo-’

I02

io3

STRAIN RATE ( % / h t )

Fig. 22. Fracture S t r a i n a t 650°C of Heat 5085 i n the Unirradiated and Irradiated Conditions. Irradiated t o a thermal fluence of 1.5 X lo2’ neutrons/cm2 over a period of 22,533 h r a t 650°C. i n t h e two types of tests.

The most s t r i k i n g feature i s . t h e marked

dependence of t h e f r a c t u r e s t r a i n of t h e irradiated samples on t h e s t r a i n r a t e and t h e r a t h e r weak dependence of the f r a c t u r e s t r a i n s of t h e unirradiated samples on the s t r a i n r a t e .

”he i r r a d i a t e d samples a t t h i s

32 high fluence l e v e l show a general decrease i n f r a c t u r e s t r a i n with decreasing s t r a i n r a t e , with a possible s l i g h t increase i n f r a c t u r e

s t r a i n a t very l o w s t r a i n rates. The s e n s i t i v i t y of t h e fracture s t r a i n of heat 5085 a t 650째C t o

t h e helium content i s i l l u s t r a t e d i n Fig. 23 f o r t h r e e s t r a i n r a t e s .

This material contains 38 ppm B (Table 1, p. 4) t h a t can y i e l d an equiva l e n t amount of helium when transmuted.

(Several factors contribute t o

t h e relationship that 1ppm of' natural boron by weight leads t o 1.1ppm He on an atomic basis.)

The points shown i n Fig. 23 a l l come

from surveillance samples from t h e lSRE.

The two higher s t r a i n rates

are obtained by t e n s i l e t e s t i n g , and t h e f r a c t u r e s t r a i n decreases with increasing helium content.

The 0.1% s t r a i n rate i s t h e creep r a t e that

corresponds t o t h e lowest fracture s t r a i n (Fig. 21).

The f r a c t u r e

s t r a i n drops abruptly with the presence of l p p m He and then decreases gradually with increasing helium content. The creep properties of heat 5065 are i l l u s t r a t e d i n a s e r i e s of

graphs similar t o t h a t j u s t presented f o r heat 5085.

The stress-rupture

properties a t 650째C a r e shown i n Fig. 24 f o r heat 5065. The rupture ORNL-OWG 69-7297R2

Fig. 23. Variation of t h e Fracture S t r a i n with Calculated Helium Content and S t r a i n Rate f o r Hastelloy N (Heat 5085) a t 650째C.

33 ORNL-DWG 69-4474R

100

10'

102

to4

RUPTURE TIME (hr)

Fig. 24. Stress-Rupture Properties of MSRE Surveillance Specimens (Heat 5065) a t 650째C. times are equivalent f o r t h e samples i r r a d i a t e d t o the two highest fluences.

The rupture times a r e a l s o quite camparable with those s h a m

i n Fig. 19 f o r heat 5085.

The unirradiated samples t h a t were aged f o r

22,533 h r a t 650째C i n s t a t i c barren f u e l salt have longer rupture l i v e s than t h e as-received material. The minimum creep r a t e s are shown i n Fig. 25 and show a lack of s e n s i t i v i t y t o any of t h e variables being studied. The f r a c t u r e s t r a i n i s shown as a f'unction of s t r a i n r a t e i n Fig. 26. The f r a c t u r e s t r a i n decreases with increasing fluence up t o t h e two highest fluence levels, where t h e s t r a i n s are about equivalent. This heat of material shows a d u c t i l i t y minimum with t h e f r a c t u r e s t r a i n increasing s l i g h t l y with decreasing creep r a t e except f o r t h e samples showing t h e lowest fluence. The t e n s i l e and creep t e s t r e s u l t s f o r heat 5065 have been combined i n Fig. 27.

These r e s u l t s again show the marked dependence of the frac-

ture s t r a i n on t h e s t r a i n rate f o r t h e i r r a d i a t e d material.

A comparison

with t h e similar p l o t f o r heat 5085 (Fig. 22) reveals some s l i g h t differences i n t h e f r a c t u r e s t r a i n s of t h e two heats, but shows generally

34 70

60 c

-nn 8 40

-0

8 a

5 20

40-2

io-'

too

MINIMUM CREEP RATE (%/hr)

F i g . 25. Minimum Creep Rate of Hastelloy N (Heat 5065) Surveillance Samples from the NRE at 650째C.

ORNL-DWG

ERMAL FLUENCE

69-4476R

a 4 a $ I

u !E

0 to-'

46'

40to-' MINIMUM CREEP RATE (%/hr)

ioo

io'

Fgg. 26. Variation of Fracture Strain with Strain Rate for HasteUoy N (Heat 5065) Surveillance Specimens at 650째C.

35 ORNL-DWG 70-7299R

e, 5

a a

a 2

I I I11111 IRRADIATED

I-

TENSILE TESTS

. 0 CREEP TESTS

___~_ --e

-0-

c’

I ~~

2-4h----

4’

/’ I

I I1111

/* e**

++*

ro3

Fig. 27. Fracture Strains a t 65OOC of Heat 5065 i n t h e Unirradiated and t h e Irradiated Conditions. Irradiated t o a thermal fluence of 1.5 X 1O2I neutrons/cm’ over a period of 22,533 hr a t 650°C. Heat 5065

t h a t t h e two heats respond t o i r r a d i a t i o n q u i t e similarly.

contains about 23 ppm B (average of t h e four values i n Table 1, p. 41, and t h e v a r i a t i o n of t h e f r a c t u r e s t r a i n with helium content i s sham The behavior i s quite similar t o that noted i n Fig. 23 f o r

i n Fig. 28. heat 5085.

Thus, t h e lower boron (and hence helium) concentration i n

heat 5065 r e s u l t s i n about t h e same properties as t h e higher boron l e v e l

i n heat 5085. Mechanical P Two heats of m o d i 6244 hr a t 650°C and rece neutrons/cm2 (Table 2, p titanium

- heat

Data -Modified Hastelloy N

loy N were exposed t o t h e MSRE core f o r

a thermal neutron fluence of 5.1 X lo2’ Both of these alloys were modified with

7320 contained 0.65% and heat 67-551 contained 1.1$.

The t e n s i l e properties of heat 7320 a r e summarized i n Tables 11 and

l2 f o r t h e control and i r r a d i a t e d samples, respectively.

The fracture

0.2

0.5

HELIUM CONTENT (ppm)

F i g . 28. Variation of Fracture S t r a i n a t 650째C with Calculated Helium Content and S t r a i n Rate f o r Heat, 5065. Table ll. Results of Tensile Tests on Control Samples of Heat 7320a Test Specimen TemperNumber ature (OC)

9464 9482 9470 9481 9472 9456 9467 9466 9458 9476 9478 9461 9475 9474 9477 9480

9471 9457

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

Strain Rate

Stress, psi Ultimate

(dn-1) Yield

0.05 0.05 0.05 0.002

Tensile

68,800 122,600 60, OOO 1W7,600 56,600 103,300 54, 100 96,000 0.05 99,200 54,500 0.002 53,600 96,700 0.05 51,700 95,500 0.002 49,400 85,300 0.05 52,500 91,900 54,800 80,500 0.002 0.05 49,300 78,600 0.002 47,700 72,800 0.0002 50,800 67,900 0. oo0027 47,800 59, loo 50,100 71,500 0.05 44,200 45, 100 0.002 0.05 45,000 48,200 0.002 27,800 27,500

Elongation, $ Uniform Total 48.2 40.6 46.9 45.2 47.9 45.1 49.2 36.7 41.4 18.2 28.4 14.5 12.3 5.6 12.0 4.2 5.6 1.6

49.7 41.9 49.1 46.2 51.1 46.2 52.0 37.7 45.5 19.7 30.6 15.5 13.7 9.6 13.1

17.2 10.2 10.1

Reduction

in ($) Area 40.83 43.75 43.09 40.64 44.38 41.21 45.39 32.99 41. 11 23.35 36.67 19.46 U .40 13.88 ll.69 19.07 12.39 ll.97

True Fracture Strain ($1 52 58 56 52 59 53 61 40 53 27 46 22 16 15 12

21 13 13

aAnnealed 1hr a t ll77"C and exposed t o a vessel of s t a t i c barren fuel s a l t f o r 7244 hr a t 650째C.

h-,

Table 12.

Postirradiation Tensile Properties of Hastelloy N (Heat 7320)&

9221 9222 9223 9224 9225 9226 9215 92U 9213 9212 9211 9210 9227 9228 9209 9208 9207 9206

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

0.05 0.05 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.0002 0.000027 0.05 0.002 0.05 0.002

69,800 116,600 59,000 102,900 54,100 93,500 58,200 94,100 54,900 89,100 55,400 83,300 61,100 88,200 61,000 74,700 60,800 81,700 61,700 69,800 67,400 74,900 57,100 68,500 49,500 58,200 37,500 49,400 56,700 63,700 48,700 48,900 37,900 38,200 24,800 24,800

43.8 40.7 40.5 38.0 39.1

31.1 26.8 9.2 18.4

4.3 8.0 4.3 4.3 2.0 3.8 1.6 1.6 1.0

45.2

35.48

41.7 41.6 38.7 41.1 32.5 27.6 11.1 19.6 5.5 9.5 5.1 4.7 2.9 4.0 3.0 2.2 1.6

35.69 33.09 31.67 36.97 26.61 26.26 21.27 27.74 11.83 7.10 9.08 7.54 3.00 6.I5 6.15 7.25 2.23

&Annealed 1h r a t 1177°C. Irradiated t o a thermal fluence of 5.1 neutrons/cm2 over a period of 7244 hr a t 650°C.

s t r a i n s a r e s h m as a function of t e s t temperature i n Fig. 29.

44 44 40 38 46 31 30 24 32 I3 7 10 8 3 6 6 8 2

lo2'

Irra-

d i a t i o n reduces t h e f r a c t u r e s t r a i n over the e n t i r e range of t e s t temperatures, but t h e magnitude of the decrease i s much greater a t elevated temperatures.

The sharp drop i n fracture s t r a i n with increasing temper-

a t u r e occurs a t a lower temperature f o r t h e i r r a d i a t e d material.

The

r a t i o s of t h e i r r a d i a t e d and unirradiated properties are shown i n Fig. 30.

The y i e l d s t r e s s i s higher f o r the i r r a d i a t e d material over

the t e s t temperature range of 550 t o 760°C.

The ultimate t e n s i l e strength

i s about t h e same f o r t h e i r r a d i a t e d and unirradiated material but sham

a gradual decline with increasing t e s t temperature.

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

decreased due t o irradiation, with the reduction being about 80s a t 850°C. Some idea of t h e t e n s i l e properties of heat 7320 due t o aging and i r r a d i a t i o n can be obtained from Table 13. The yield strengths a t 25 and 650°C are increased by aging; at a t e s t temperature of 650°C a f u r t h e r P

increase occurs due t o i r r a d i a t i o n .

The changes i n ultimate t e n s i l e

38 ORNL-DWG 70-7501

-0

H)o

200

300

400 500 600 TEST TEMPERATURE ('C)

700

800

900

Fig. 29. Fracture S t r a i n s of Heat 7320 After Removal from t h e MSRE and from t h e Control Facility. A l l samples annealed 1h r a t 1177째C before i r r a d i a t i o n t o a thermal fluence of 5.1 X lo2' neutrons/cm* over a period of 7% h r at 650째C.

1.4

f.2

0.2

100

200

300

400 500 600 TEST TEMPERATURE E)

700

800

900

Fig. 30. Comparison of Unirradiated and Irradiated Tensile Pr t i e s of Heat 7320 After I r r a d i a t i o n t o a Thermal Fluence of 5.1 X 10 neutrons/cm* Over a Period of 7244 hr at 650째C. Tested a t a s t r a i n r a t e of 0.05 min-l.

TF-

39 Table 13. Comparison of the Tensile Properties of Heat 7320 Before and A f t e r Irradiationa

Heat Treatment

Yield Stress, p s i a t 25°C a t 650°C

Annealed 1 hr a t 1177°C

38,500 Annealed 1hr a t ll77"C, 68,800 aged 7244 hr a t 650°C Annealed 1hr a t 1177"C, 69,800 irradiated f o r 7244 h r a t 650°C t o a thermal

'Itimate Stress, p s i

f i a c t u r e Strain, a t 25°C a t 650°C

a t 25°C

a t 650°C

25,600 49,300

107,300 122,600

72,700 78,600

75.4 48.1

53.5 30.6

67,400

116,600

74,900

45.2

9.5

fluence of 5.1 X lo2' neutrons/cm2

a Tested a t a s t r a i n r a t e of 0.05 min'l.

strength a r e r a t h e r small and l i k e l y within experimental error.

The

f r a c t u r e s t r a i n i s decreased by aging and by irradiation; t h e magnitude of the decrease i s much l a r g e r at t h e t e s t temperature of 650°C. The r e s u l t s of t e n s i l e t e s t s on the control and irradiated samples

of heat 67-551 are given i n Tables i

U- and 15, respectively.

The f r a c t u r e

s t r a i n i s shown as a flxnction of t e s t temperature i n Fig. 31.

The frac-

ture s t r a i n a t low temperatures i s not influenced by irradiation, but above 400°C t h e f r a c t u r e s t r a i n i s decreased by i r r a d i a t i o n .

However,

t h e f r a c t u r e s t r a i n s a t high temperatures a r e higher than those f o r e i t h e r heat 7320 (Fig. 29) o r t h e standard a l l o y (Figs. 9 and 11, pp. 18 and 22, respectively).

The r a t i o s of t h e i r r a d i a t e d and unirradiated

t e n s i l e properties a r e shown i n Fig. 32.

The yield stress i s generally

about 25% lower f o r t h e i r r a d i a t e d material.

The u l t h a t e t e n s i l e

strength i s reduced only about lO$ by i r r a d i a t i o n when t e s t e d below 55OoC, but i s reduced up t o 30%as t h e t e s t temperature i s increased.

The frac-

ture s t r a i n i s unaffected by i r r a d i a t i o n up t o test temperatures of 500°C; above t h i s temperature the

cture s t r a i n decreases progressively with

increasing t e s t temperatur

Some values of t h e t e n s i l e properties a t

25 and 650°C a r e given i n Table 16.

Aging a t 650°C seems t o increase

t h e yield strength a t both 25 and 650°C; however, t h e strength of t h e P

i r r a d i a t e d material i s quite close t o t h a t of t h e as-annealed material.

The ultimate t e n s i l e strength i s increased o n l y s l i g h t l y by aging and

Table l-4. Results of Tensile Tests on Control Samples of Heat 6?-55la Test Temperature

Specimen Number

( O C 1

9403 9419 9422 9417 9506 9423 9402 9418 9416 9428 9410 9407 9426 94x7 9424 9421 9425 9420

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

'Annealed 65OOC.

CI,

Strain Rat e

Stress, p s i

(min-1)

0.05 0.05 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.05 , 0.002 0.0002 0.000027 0.05 0.002 0.05 0.002

Tensile U1tihate 62,400 54,100 52,300 51,400 48,100 50,400 45,200 47,600 42,800 46,200 41,900 42,700 46,000 39,900 45,600 43,000 41,300 25,500

120,200 108,700 104,300 101,700 89,900 92,100 94,100 80,400 87,800 79,600 84,600 76,300 74,800 60,900 78,000 45,200 5 1,000 26,200

Elongation, $I Uniform Total 50.3 47.5 48.7 50.4 50.4 42.0 52.0 26.3 46.8 23.1 40.4 27.3 22.3 11.9 21.6

5.5 6.2 2.3

52.3 50.1 50.4 51.5 53.6 43.6 53.7 27.9 49.8 25.3 42.6 31.5 30.2 21.7 28.8 36.2 23.6 22.4

Reduct ion i n Area

True Fracture

($1

41.59 45.22 46.66 43.84 46.58 36.67 41.78 31.67 43.09 26.37 37.07 25.42 28.87 23.23 28.87 33.98 26. I& 18.81

Strain 54 60 63 58 63 46 54 38 56 31 46 29 34 26 34 42 30 21

1 hr a t 1177째C p r i o r t o exposure t o a vessel of s t a t i c barren f u e l s a l t f o r 7244 h r a t

Table 15. Spech e n Number

9167 9168 9169 9175 9176 9172 9161 9160 9159 9158 9157 9156 9173 9174 9155 9154 9153 9152

Test Temperature

("C) 25 200 400 400 5 00 500 550 550 600 600 650 650 650 650 760 760 850 850

Postirradiation Tensile Properties of Hastelloy N (Heat 67-551)" Strain Rat e (min-l)

0.05 0.05 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.05 0.002 0.0002 0.000027 0.05 0.002 0.05 0.002

Stress, p s i Ultimate Tensile

49,600 39,800 39,200 40,400 37,000 37,000 34,500 39,500 37,000 41,900 31,300 31,200 35,100 34,800 29,700 29,200 26,700 22,300

107,800 94,700 91,800 88,100 83,400 67,900 80,400 57,100 71,300 61,800 56,600 53,300 57,100 51,300 51,500

38,400 34,700 22,700

Reduction in Area

Elongation, '$

True Fracture

Uniform

Total

( 8)

Strain

50.6 50.0

51.0 53.2 50.7 52.6 52.9

36.87 42.15 38.44 39.69 38.34 29.21 34.92 23.35 39.59 22.68 28.76 19.20 L3.74 12.5 8.79

46 55 49 51

48.7 51.3 52.1 25.8

42.8 14.7 31.9 18.0 20.7 16.0 12.3 7.5 13.8

5.7 4.8 2.2

27.4 44.5 17.0 33.9 19.2 22.6 16.9 14.5 12.9 14.8 9.2 6.5 6.2

'Annealed 1 hr a t 1177째C. Irradiated t o a thermal fluence of 5 . 1 X ' 0 1 period of 7244 h r a t 650째C.

10.74 7.54 5.24

(8)

48 35 43 27 50 26 34 21 15 13 9 11 8

neutrons/cm2 over a

Fig. 31. Fracture Strains of Heat 67-551 After Removal from the MSRE and from t h e Control Facility. All samples annealed 1hr a t 1177째C before i r r a d i a t i o n t o a thermal fluence of 5 . 1 X lo2' neutrons/cm2 over 7244 h r a t 650째C.

ORNL-OWG

70-7304

1.2

4.0

0.2

(00

200

300

400 5-90 600 TEST TEMPERATURE CC)

700

800

900

Fig. 32. Comparison of Unirradiated and Irradiated Tensile Propert i e s of Heat 67-551 After I r r a d i a t i o n t o a Therm1 Fluence of 5.1 X lo2' neutrons/cm2 over a period of 7244 hr at 650째C. Tested a t a s t r a i n r a t e of 0.05 min-l.

Table 16.

Comparison of the Tensile Properties of Heat 67-551 Before and After Irradiationa

Heat Treatment

Ultimate Tensile Yield Stress, p s i Stress, p s i Fracture Strain, $ a t 25°C at 650°C at 25°C at 6 5 0 " ~ a t 25°C a t 650°C

Annealed 1hr a t 1177°C

44,600

26,900

113,600

78,900

79.6

57.3

Annealed 1h r a t 1177"C, aged 7244 hr a t 650°C

62,400

41,900

120,200

84,600

52.3

42.6

Annealed 1h r a t 1177"C, aged 7244 hr at 650"C, irradiated t o a thermal fluence of 5.1 x neutrons/cm2

49,600

31,300

107,800

56,600

51.0

22.6

&Tested a t a s t r a i n r a t e of 0.05 min-l.

decreased by irradiation.

A t 25°C t h e f r a c t u r e s t r a i n i s reduced from

79.6% t o 52.3% by aging, and i r r a d i a t i o n does not cause any flxrther

change.

A t 650°C t h e f r a c t u r e s t r a i n i s decreased by aging and decreased

f'urther by i r r a d i a t i o n . The creep-rupture properties of heat 7320 a r e summarized i n Table 17.

These results are compared i n Figs. 33, 34, and 35 with those

f o r unirradiated samples t h a t were given a p r e t e s t anneal of 1 h r a t 1177°C. The stress-rupture properties i n Fig. 33 show t h a t t h e aging treatment given t h e controls, 7244 h r a t 75O9C, increased t h e rupture The i r r a d i a t e d samples f a i l e d a f t e r longer times than did the unirradiated samples that w e r e s i m p l y annealed 1 hr at 1177"C, but life.

ruptured i n shorter times than the control samples.

However, the minimum

creep r a t e s shown i n Fig, 34 seem t o f a l l within a common s c a t t e r band f o r a l l conditions. Thus, neither i r r a d i a t i o n nor aging seems t o have a detectable e f f e c t on t h e minimum creep r a t e . The f r a c t u r e s t r a i n is shown as a function of s t r a i n rate i n Fig. 35.

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

the unirradiated samples vary from about SO$ i n a t e n s i l e test t o about

lO$ i n a long-term creep test. The i r r a d i a t e d samples have lower f'ract u r e s t r a i n s t h a t vary from 9.5$ f o r t h e f a s t e s t t e n s i l e t e s t t o 2 t o 3$ f o r creep tests.

Table 17.

Test Number

Creep-Rupture Tests on Heat 7320 a t 650°C

Specimen Number

Unirradiated 7276 10329 7277 7279 7278 10252 7280

7013 7425 70l.4 7016 7015 7424 7017

Stress (psi)

Lify( 4 )

ture Strain

(hr)

- Annealed 55,000 50,000 47,000 43,000 40,000 35,000 30,000

Minimum Creep

Reduct ion i n Area

(7 8

1 h r a t 1177°C p r i o r t o t e s t

4.7 11.5 49.9 103.4 384.6 1602.1 1949.5

28.7 21.2 21.8

18.1 16.3 27.3 10.9

29.6 28.3 31.3 20.5 17.6 30.6 19.9

0,190

32.9 26.3 29.3

0.25 0.045 0.019 0.015 0.0006

0.125 0.0194 0.0069 0.0059 0.0056 0.0017

Unirradiated Controls‘ 7885 7991 7886 7887b 7884

9462 9468 9463 9465 9460

55,000 47,000 40,000 40,000 32,400

32.5 224.1 653.7 501.9 2420.2

U.2 15.3 17.9 7.9 6.9

21.4 4.8

Irradiat edc

R-1151 R- 1016 R-951 R-955 R- 967 R-950

‘

9204 9216 9217 9219 9218 9220

63,000 63,000 55,000 47,000 40,000 32,400

1.7 4.0 12.3 95.1 329.9 2083.3

12.2d 2.0 2.7 2.3 3.2 3.1

1.2 0.44 0.20 0.018 0.0074 0.0058

a Annealed 1 h r a t 1 1 7 7 ° C and exposed t o a vessel of s t a t i c barren f u e l s a l t f o r 7244 hr a t 650°C. bDiscontinued p r i o r t o f a i l u r e . 5.1

‘Annealed 1h r a t 1177°C. Irradiated t o a thermal fluence of 10” neutrons/cm2 over a period of 7244 h r a t 650°C.

%est loaded so t h a t s t r a i n on loading was included. t e s t s did not include t h i s s t r a i n .

A l l other

10'

402

RUPTURE TIME lhrl

Stress-Rupture Properties of Heat 7320 a t 650째C.

Fig. 33.

ORNL-OWG 70-7306

-a .)

0 0

Y) u)

IRRADIATED, S.lX4&

to-'

to-2

to-'

COO

MINIMUM CREEP RATE (%/hr)

Fig. 34.

Creep Properties a t 650째C of Heat 7320.

46 _.

ORNL- DWG 70-7307

I I IRRADIL

10-3 2

10-2 2

lo-'

loo

(0'

(02

103

STRAIN RATE (%/hr)

Fig. 35. Fracture Strains a t 650째C of Heat 7320 i n t h e Unirradiated and Irradiated Conditions. Irradiated t o a thermal fluence of 5 . 1 x lo2' neutrons/cm2 over a period of 7244 hr at 650OC. The r e s u l t s of creep-rupture t e s t s on heat 67-551 with 1.1% T i are given i n Table 18.

The stress-rupture properties of t h e c o n t r o l and

irradiated samples a r e compared i n Fig. 36 f o r 650째C.

The i r r a d i a t e d

samples generally f a i l i n s l i g h t l y shorter times than t h e unirradiated samples.

The minimum creep'rate s e a s t o be unaffected by irradiation,

although t h e r e a r e two data points t h a t seem t o be anomalous (Fig. 37). The f r a c t u r e s t r a i n s of t h i s a l l o y (Fig. 38) a r e higher i n both t h e unirradiated and i r r a d i a t e d conditions than those f o r heat 7320. The f r a c t u r e s t r a i n s of t h e unirradiated samples varied over t h e range of

43 t o 25% and those of t h e i r r a d i a t e d samples varied from 22.6 t o 5.84 over t h e range of s t r a i n r a t e s studied. One heat of modified Hastelloy N containing 0.49% H f (heat 67-506) was exposed t o t h e c e l l environment f o r 17,033 h r a t 650째C.

fluence was 2.5 X 1019 neutrons/cm*.

The.therma1

This same heat of material was

included previously i n our surveillance program and was exposed t o a

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

Test Number

7872 7871

Specimen Number

Stress (psi)

Rupture Life Strain

( 8)

(hr)

Reduction i n Area ( $)

Unirradiated, Annealed 1 h r at 1177°C before t e s t i n g 55,000 U.2 29.5 33.2 6889 6884 47,000 U3.8 33.6 26.0

Minimum Creep

0.32 0.071

Unirradiated Controlsa 7882

7880 7883 7881

9409 9404 94u. 9408

55,000 47,000 40,000 32,400

44.7 271.2 956.2 2034.9

25.5 29.1 30.4 25.2

28.5 29.6 25.5 29.9

0.190 0.067 0.018 0.0053

Irradiated R- l I 5 O R- 1036 R-949 R-961 R-948 R- 947

9150 9151 9 162 9165 9164 9163

63,000 63,000 55,000 47,000 40,000 32.400

0.9 0.6 50.5

20.8 2.2 5.8

117.1

8.3

746.9 U85.6

9.1 12.5

2.2 2.6 0.079 0.058 0 0079 0 0053

a Annealed 1 h r a t 1177°C p r i o r t o exposure t o a vessel of s t a t i c barren fuel s a l t f o r 7244 h r a t 650°C.

5.1

bAnnealed 1h r a t 1177°C. Irradiated t o a thermal fluence of lo2' neutrons/cm2 over a period of 7244 h r a t 650°C.

Test loaded t o include s t r a i n on loading. include t h i s s t r a i n . thermal fluence of 5.3

lo2'

A l l other t e s t s did not

neutrons/cm* while being a t temperature i n

t h e core f o r 9789 h r ( r e f . 10). I n t h e group of samples presently being discussed, there were two rods of heat 67-504.

Two heats of material

should have been involved, but p o s t i r r a d i a t i o n chemical analysis revealed t h a t both rods were made of t h e same material.

This was not discovered,

u n t i l after many of t h e samp es were tested, so we have several tests under duplicate conditions.

loH. E. McCoy, An Evaluation of t h e Molten-Salt Reactqr Experiment Hastelloy N Surveillance Specimen Third Group, ORNETM-2647 (1970).

I I

"2 n

Q n

"2 n

PI0 ard 0

9 8

i r

5 "f

I %

alo

h aO

49 ORNL- DWG 70- 7340

0-3 2

40-2

lo-'

5 1 8 2 STRAIN RATE (%/hr)

TENSILE TESTS CREEP TESTS

10'

102

Fig. 3 8 . Comparison of t h e Fracture Strains of Unirradiated and Irradiated Heat 67-551 a t 650°C. Irradiated t o a thermal fluence of 5 . 1 X lo2' neutrons/cm2 over a period of 7244 h r a t 650°C. The r e s u l t s of the p o s t i r r a d i a t i o n t e n s i l e t e s t s on heat 67-5W

a r e given i n Table 19.

The p o s t i r r a d i a t i o n f r a c t u r e s t r a i n s a r e shown

i n Fig. 39 as a function of temperature and s t r a i n r a t e .

The f r a c t u r e s t r a i n generally decreases with increasing temperature above about 400°C and with decreasing s t r a i n r a t e . those reported previously1'

The r e s u l t s from the present t e s t s and

show some very s t r i k i n g changes i n flracture

s t r a i n with varying h i s t o r i e s (Fig. 40).

The samples were a l l i n i t i a l l y

annealed 1hr a t 1177°C before being given t h e treatment indicated i n Fig. 40. I n t h e as-annealed

ondition, t h e a l l o y has a fracture s t r a i n

of 70 t o 804 up t o about 6OO0C, where t h e f r a c t u r e s t r a i n drops precipA

itiously.

Aging f o r 9789 hr a t 650°C reduced the f r a c t u r e s t r a i n at

lower temperatures and increased it above about 700°C.

However, t h e

f r a c t u r e s t r a i n s a r e i n t h e range of 40 t o 50$ over t h e e n t i r e range of

temperatures studied.

I r r a d i a t i o n t o a thermal fluence of 2.5

neutrons/cm2 over a period of 17,033 hr i n N2

lo1'

2 t o 5% 02 resulted i n

Table 19. Postirradiation Tensile Properties of Hastelloy N (Heat 67-504)& , Test Specimen TemperNuniber ature ("C)

5 u 51k7 5llO 5162 5109 5u5 5108 5 M 5107 5u3 5106 5l42 5u7 5153 5 m 5154 5096 53.64 5120 5156 5128 5121 5155 5122 5158 5105 5l41 5UO 5104 5098 5123 5159 5l24 5160 5102 5138 5103 5139

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

Strain Rate

Yield

Elongation, $ Uniform Total

(min-1)

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

0.05 0.002 0.002 0.05 0.05 0.002 0.002 0.05 0.05 0.05 0.002 0.002 0.0002

O.OOO2 -0.m27 -0.m27

0.05 0.05 0.05 0.002 0.002 0.05 0.05 0.002 0.002

50,300 112,200 50,700 112,800 41,000 96,400 40,900 95,300 36,700 90,800 37,500 90,800 36,700 91,800 41,100 97,900 40,600 92,700 40,900 94,000 41,600 90,500 33,800 78,600 39,700 74,900 34,700 77,800 44,600 62,000 33,200 59,900 31,900 74,500 32,900 70,800 36,400 49,600 30,900 49,500 39,900 55,500 38,800 59,600 30,200 46,200 52,200 54,500 31,700 44,800 32,200 40,800 30,700 38,800 30,200 37,100 30,700 37,200 27,OOO 38,600 33,300 40,100 28,800 38,OOO 33,300 33,300 28,200 34,600 27,300 29,700 27,500 30,900 21,100 21,100 23,700 23,700

45.9 54.6 43.0 45.7 49.8 52.8 52.2 52.7 45.7 49.6 44.5 41.1 34.5 39.7 11.9 20.5 34.7 32.9 8.9 12.6 24.0 U.6 12.8 2.2 u.2

7.4 7.8

4.4 4.8 9.0 6.7 9.6 1.2 4.2 2.9 3.2 1.0 1.0

48.1 56.6

44.8 47.3 52.0 54.8 54.7 53.7 47.2 50.6 46.2 42.6 36.5 43.5 15.1 21.5 36.9 34.0 10.9 15.7 25.8 16.5 15.8 6.2 u.3 9.8 8.8 6.5 6.5 23.7 8.7 13.5 2.7 5.6 4.3 4.5 1.9 2.5

Reduction i n Area (4 )

34.63 36.54 38.34 46.66 36.15 34.84 45.71 33.31 32.23 39.97 39.40 36.00 31.45 37.94 16.02 20.38 26.02 35.69 20.38 23.35 22.06 19-15 10.17 3.51 ll.54 13.52 6.65 5.1 5.4 18.94 10.17 13.19 4.63 6.00 2.40 7.10 0.64 0.48

aAnnealed 1 hr a t ll77"C. Irradiated t o a thermal fluence of 0.25 neutrons/cm2 over a period of 17,033 hr at 650째C.

1020

Rue Fracture Strain

($1 43 45 48 63 45 43 61 41 39 51 50 45 38 48 17 23 30 44 23 27 25 21 11 4 12 15 7 5 6 21 11

u 5 6 2

7 1

51 ORNL-OWG 70-734t

'ii 40

c z

w 30 3 cc

c 0 a (L

0.05

A 0.002 a

0.0002

400

200

300

400 500 600 TEST TEMPERATURE ("C)

700

800

900

Fig. 39. Fracture Strains of Heat 67-504 After I r r a d i a t i o n t o a Thermal Fluence of 2.5 X 10'' neutrons/cm2 f o r 17,033 h r a t 650째C.

ORNL-DWG 70-7312

6 ~ 0 0 7 0 0 8 0 0 9 0 0

TEST TEMPERATURE (TI

Fig. 40. Comparison of t h e Fracture Strains of Heat 67-504 i n Various Conditions When Tested at a S t r a i n Rate of 0.05 min'l.

f r a c t u r e s t r a i n s of about 50% up t o 50O0C, above which t h e fracture s t r a i n dropped progressively with increasing temperature. t o a thermal fluence of 5.3

lo2'

Irradiation

neutrons/cm2 a t 650°C i n f u e l s a l t

over a period of 9789 h r resulted i n a low f r a c t u r e s t r a i n a t 25°C (which may be anomalous), f r a c t u r e s t r a i n s of about 45% up t o 500°C, and decreasing fracture s t r a i n s with increasing t e s t temperature.

The most

s t r i k i n g feature i s t h a t above 550°C t h e f r a c t u r e s t r a i n s a r e lower f o r

material i r r a d i a t e d i n N2 + 2 t o 5% 02 f o r 17,033 h r t o a fluence of 2.5 x loi9 neutrons/cm2 than f o r those i r r a d i a t e d i n f u e l salt f o r 9789 h r t o a fluence of 5.3 X

lo2* neutrons/cm2.

Some values of t h e t e n s i l e properties a t 25 and 650°C are given i n Table 20. The yield stress a t both t e s t temperatures was increased by aging and by i r r a d i a t i o n .

The ultimate t e n s i l e s t r e s s a t 25°C was

increased by aging and by irradiation; a t 650°C it was increased by aging, but decreased.by i r r a d i a t i o n .

We have already discussed the

changes i n the f r a c t u r e s t r a i n .

Table 20.

Comparison o f t h e Tensile Properties of Heat 67-504 After Various Treatmentsa

Heat Treatment

Yield Stress, p s i

a t 25°C a t 650°C

Ultimate Stress,. p- s i at 25°C at 650°C

Fracture Strain,

a t 25°C

a t 65OOC

73.2

65.9

52.3

50.9

Annealed 1hr a t ll77"C

37,400

25,600

105,000

Annealed a t U77"C and aged 978gb hr a t 650°C

67,500

43,300

133,000

83,000 94,300

Annealed a t UWOC, i r r a - 102,OOO diated tob a t h e m 1 fluence of 5.3 x neutrons/cm2 over 9789 h r ( s a l t environment)

32,900

ll9,300

70,000

27.0

30.7

Annealed a t ll77"C, irradiated t o a therm a l fluence of 2.5 X lof9 neutrons/cm2 Over 17,033 hr (N2 + 2 t o 55 92 environment)

38,800

ll2,OOO

59,600

48.1

16.5

50,300

BTested a t a s t r a i n r a t e of 0.05 min-1. bData from: H. E. McCoy, Jr., A n Evaluation of the Molten-Salt Reactor Experiment 1 m l . Hastelloy N Surveillance Specimens Third Group, ORNLTM-2647 (

53 The r e s u l t s of creep-rupture t e s t s on heat 67-504 a r e summarized i n Table 21.

These r e s u l t s and those obtained previously1'

t o prepare Figs. 41, 42, and 43.

were used

The stress-rupture properties i n

Fig. 4 1 show t h a t t h e rupture l i f e was not changed appreciably by aging,

a t l e a s t f o r 9789 h r a t 650째C.

The stress-rupture l i f e was reduced at

l e a s t two orders of magnitude by t h e lower fluence and only a f a c t o r of 2 by t h e higher fluence.

The minimum creep r a t e s i n Fig. 42 show t h a t

t h i s property was not changed a s much as t h e rupture l i f e .

Aging f o r

9789 h r increased t h e minimum creep, but samples i r r a d i a t e d over t h e same period had a lower creep r a t e than t h e aged samples.

The samples

i r r a d i a t e d t o t h e lower fluence had a creep r a t e an order of magnitude higher than t h a t of the as-annealed material.

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

' H . E. McCoy, An Evaluation of t h e Molten-Salt Reactor Experiment Hastelloy N Surveillance Specimen Third Group, ORNETM-264'7 (1970).

Table 21.

Test Number

Creep-Rupture Tests on Heat 67-504 a t 650째C

Specimen Number

Stress (psi)

Rupture Life Strain (b) (4)

Minimum Creep

I r r a d i a t e d a t 650째C t o a thermal fluence of 2.5 X loi9 neutrons/cm2 R-952 R-953 R-959 R-965 R-971 R- 1017 R- 968 R- 1033 R- 1019

5112 5U8 5113 5u9 51U 5150 5115 5137 5116

55,000 55,000 47,000 47,000 40,000

0.6 1.0 7.6 3.3 u.5 33.8 235.5 268.7 l175.4

6.2 10.9 4.1 3.8 4.2 2.7

9.88 6.6 0.83 0.52 0.42 0.050 0.0071 0.0066 0.0016

32.9 31.2 27.2 28.7 16.3

3.7 0.16 0.029 0.0068 0.0030

8.6 10.6 12.1

Annealed 1h r at 1177째C 7432 7431 6255 6254 6253

6247 6245 4991 4936 4933

0,000 3,000 5,000 7,000 40,000

1.3 17.7 127.9 425.5 876.9

54 mNL-DWG 70-731s

RVPNRE TIME (hr)

Fig. 41.

Stress-Ftupture Properties at 650째C of Heat 67-504.

ORNL-DWG 70-7313

rn m W

E30 rn

'h' 0

F i g . 42. Creep-Rupture Properties at 650째C of Heat 67-504.

55 ORNL-OWG 70-73!4R

io-‘

io-’

400

40’

SOLID P OPEN POINTS CREEP o ANNEALED i hr AT 4f77OC A ANNEALEDANDAGED 9789 hr AT 65OOC ANNEALED,THERMAL FLUENCE OF 5.3XtOZ0 neutrons/cm* 0 ANNEALED.THERMAL FLUENCE OF 2.5xiOt9 neutrons/cm

io2

STRAIN RATE (70)

Fig. 4 3 . Comparison of the Fracture Strains of Unirradiated and Irradiated Heat 67-504 a t 65OOC. [Data f o r samples aged f o r 9789 h r a t 650°C and f o r those i r r a d i a t e d t o 2.5 X lo1’ neutrons/cm2 from: H. E. McCoy, An Evalwtion of t h e Molten-Salt Reactor Experiment Hastelloy N Surveillance Specimen - Third Group, ORNLTM-2647 (1970). 1

shown as a f’unction of stra

rate in Fig. 43. Aging increased t h e fracture s t r a i n of t h e unirradiated material. A f t e r i r r a d i a t i o n t h e material i

i r r a d i a t e d t o t h e lower fluence had t h e lowest f r a c t u r e s t r a i n . Several samples of heat 67-504 were subjected t o tests t h a t were interrupted.

The r e s u l t s of these t e s t s a r e summarized i n Table 22.

strained continuously a t 650°C a t a s t r a i n r a t e of 0.05 min-l, t h e

Table 22.

Type of Testa

Results of Interrupted Tensile Tests on Heat 67-504

Test Temperature ("C>

A B

C D E A

650 650 650 650 650

66.9 100.4 109.0 104.0 95.4

760 760

35.5

- Run uninterrupted

- Strained

Fracture Strain, $ Unirradiat ed Irradiated

96.0

16.5

25.8 8.7 23.7

a t a s t r a i n rate of 0.05 min-l.

5$, held a t temperature f o r 2 min, cycle

repeated t o failure. C

- Strained

54, held a t temperature f o r 10 min, cycle

repeated t o f a i l u r e . D - Strained 55, held at temperature f o r 30 min, cycle repeated t o f a i l u r e .

E Strained 5$, held a t temperature f o r 60 min, cycle repeated t o f a i l u r e . F - Strained 34, held a t temperature f o r 60 min, cycle repeated t o f a i l u r e .

material had a f r a c t u r e s t r a i n of 66.98 i f unirradiated and 16.5$ i f irradiated.

If t h e material was s t r a i n e d 5 4 a t 650째C then annealed 1h r

a t 650째C and t h i s p a t t e r n continued u n t i l failure, the unirradiated samp l e failed with 95.44 s t r a i n and t h e i r r a d i a t e d sample with 25.84 s t r a i n . A sequence of t e s t s was performed on unirradiated samples a t 650째C i n which t h e annealing time was vaxied from 2 t o 60 min; t h e r e s u l t s show t h a t t h i s variable had l i t t l e e f f e c t on t h e f r a c t u r e s t r a i n . A t a t e s t temperature of 760째C t h e fracture s t r a i n was 35.5$ f o r an unirradiated sample and 8.7% f o r an irradiated material. Samples were a l s o t e s t e d by s t r a i n i n g 3$, annealing 1h r a t 76OoC, s t r a i n i n g 3$, and repeating t h i s sequence u n t i l f a i l u r e . The fracture strains were 96.0 and 23.7% f o r t h e unirradiated and irradiated samples, respectively. These tests have

57 demonstrated t h a t there i s not, f o r each t e s t temperature, a unique s t r a i n a t which f a i l u r e occurs; rather, t h e fracture s t r a i n depends upon t h e loading history. Metallographic Examination of Test Samples The t e n s i l e samples that were t e s t e d a t 25°C a t a s t r a i n r a t e of 0.05 rnin'l

and a t 650°C a t a s t r a i n r a t e of 0.002 min-l were examined.

An irradiated sample of heat 5085 from the core that was t e s t e d a t 25°C i s shown i n Fig. 4.4.

The fracture i s primarily intergranular.

There i s a high frequency of edge cracking t h a t extends t o a depth of about 4 mils. The microstructure i n the 4-mil region near the surface etches differently. A t a 650°C t e s t temperature (Fig. 45) the f r a c t u r e i s intergranular with

no evidence of p l a s t i c deformation of t h e individual grains except adjacent t o t h e fracture.

This sample a l s o had t h e modified structure and

edge cracks t h a t extend 10 mils i n t o t h e sample.

The control samples of

heat 5085 t h a t were annealed f o r 22,533 h r a t 650°C i n s t a t i c barren f u e l

s a l t a r e shown i n Figs. 46 and 4'7. t e s t e d a t 25°C.

The sample sham i n Fig. 46 was

The f r a c t u r e i s predominantly intergranular, but there

are no edge cracks.

There i s a very t h i n layer near the surface where

t h e s t r u c t u r e i s modified s l i g h t l y .

The mLcrostructure i s generally

characterized by large &C-type carbides, many of which cracked during deformation, and a network of almost continuous carbides along the grain and twin boundaries.

intergranular.

When t e s t e d a t 650°C (Fig. 47) the f r a c t u r e i s

There is some edge cracking, but not as frequently as

noted i n Fig. 45 for t h e i r r a d i a t e d sample.

Another f a c t o r t h a t should

be considered i n comparing Figs. 45 and 47 i s t h a t t h e i r r a d i a t e d sample i n Fig. 45 failed after s t r a i n i n g only 5 4 (Table 4, p.

a),whereas

the

unirradiated sample i n Fig. 47 strained 24% (Table 3, p. 17) before fracturing.

The higher s t r a i n should have increased t h e number and depth of

edge cracks.

The observed

t h e exposure t o t h e reactor

o s i t e trend indicates some influence of ironment on the fracture characteristics

near t h e surface.

Heat 5065 was a l s o exposed t o t h e core f o r 22,533 h r a t 650°C. The microstructure of a sample t e s t e d at 25°C i s sham i n Fig. 48. The

% X

Fig. 45. Photomicrographs of a Hastelloy N (Heat 5085) Sample Tested a t 650째C After Bei %Exposed t o t h e Core of t h e MSRE for 22,533 hr a t 650째C and Irradiated t o a Thermal Fluence of 1.5 X 10 neutrons/cm*. (a) Fracture, etched. lOOX. (b) Edge, as polished. loox. ( c ) Edge, as polished. 500X. ( d ) Edge, etched. 1OOX. Etchant: glyceria regia. Reduced 27%.

Fig. 46. Typical Photomicrographs of a Hastelloy N (Heat 5085) Sample Tested a t 25째C After Being Exposed t o S t a t i c Barren Fuel S a l t f o r 22,533 h r a t 650째C. ( a ) Fracture, etched. 1OOX. (b) Edge near fiacture, etched. 1OOX. ( c ) Representative unstressed structure, etched. 500x. Etchant: glyceria regia. Reduced 24.54.

Fig. 47. Typical Photomicrographs of a Hastelloy M (Heat 5085) Sample Tested a t 650째C ( S t r a i n Rate 0.002 min"') After Being Exposed t o S t a t i c Barren Fuel Salt for 22,533 hr a t 650째C. (a) Fracture. lOOX. (b) Edge near fracture. 1OOX. ( c ) Representative unstressed structure. 500X. Etchant: glyceria regia. Reduced 25$.

f r a c t u r e is mixed transgranular and intergranular. The same s t r u c t u r a l modification and edge cracking that were noted i n heat 5085 (Fig. 4 4 ) a r e a l s o present i n heat 5065. shown i n Fig. 49.

A sample of heat 5065 t e s t e d a t 650°C i s

The f r a c t u r e i s intergranular with l i t t l e evidence

of deformation of t h e adjacent grains.

The microstructure a t t h e edge

i s modified t o a depth of a b o u t 4 mils, and cracks extend t o a depth of about 10 mils.

A control sample t e s t e d a t 25°C i s shown i n Fig. 50.

The f r a c t u r e i s mixed transgranular and intergranular, and the grains

are elongated i n t h e direction of stressing.

There i s no edge cracking.

There a r e copious amounts of carbides, both o f t h e large primary carbides and t h e f i n e r carbides formed during t h e long annealing treatment

a t 650°C. The microstructure of a control sample of heat 5065 t e s t e d a t

650°C (Fig. 51) shows an intergranular f r a c t u r e and some edge cracking. Again t h e frequency of cracking i s l e s s than t h a t noted f o r t h e irradiated material shown i n Fig. 49.

Heat 7320 was exposed t o t h e MSRE core f o r 7244 h r a t 650°C. Typicalmicrographs of a sample t e s t e d a t 25°C a r e shown i n Fig. 52.

The

f r a c t u r e i s predominantly transgranular and t h e f l a w l i n e s and t h e elongated grains attest t o t h e deformation of t h e matrix.

There are some

edge cracks t o a depth of about 3 mils, but the microstructure i s not modified a t t h e sample surface.

Microstructures of an i r r a d i a t e d sample

tested a t 650°C a r e shown i n Fig. 53.

The f r a c t u r e is intergranular.

There a r e edge cracks t h a t extend t o a depth of about 10 mils, but t h e microstructure is not modified near the surface. A control sample t h a t was t e s t e d a t 25°C i s sham i n Fig. 54.

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

with the grains being elongated i n t h e direction of stressing. no edge cracking.

There i s

The unstressed microstructure gives a h i n t of the

very f i n e matrix p r e c i p i t a t i o n t h a t i s present i n t h i s alloy.

Photo-

micrographs of a control sample t e s t e d a t 650°C are shown i n Fig. 55. The fracture i s intergranular, and t h e r e a r e no edge cracks such as were

noted i n t h e i r r a d i a t e d sample (Fig. 53). Heat 67-551 was expos

e MSRE core f o r 7244 h r a t 650°C. The

f r a c t u r e a t 25°C (Fig. 56)

ed transgranular and intergranular, and

grains have deformed extensively.

The microstructure near the surface

i s not altered, but t h e r e a r e edge cracks t h a t extend t o a depth of about

..-

...

Exposed Fig. 49. Photomicrographs of a Hastelloy N (Heat 5065) Sample Tested a t 650째C After Bei 3 1 t o t h e Core o f t h e MSRF: f o r 22,533 h r a t 650째C and Irradiated t o a Thermal Fluence of 1.5 x 10 neutrons/cm2. (a) Fracture, etched. lOOX. (b) Edge, as polished. lOOX. ( c ) Edge, as polished. 5 0 0 ~ . ( d ) Edge, etched. 1OOX. Etchant: glyceria regia. Reduced 26.5%.

(sd

Fig. 50. Typical Photomicrographs of a Hastelloy N (Heat 5065) Sample Tested a t 25째C After Being Exposed t o S t a t i c Barren Fuel S a l t f o r 22 533 hr a t 650째C. (a) Fracture. lOOX. (b) Edge near fracture. lOOX. ( c j Representative unstressed structure. 500X. Etchant: glyceria regia. Reduced 22%.

.Fig. 51. Typical Photomicrographs of a Hastelloy N (Heat 5065) Saniple Tested at 650째C (Strain Rate of 0.002 min-l) After Being Exposed to Barren Fuel Salt for 22,533 hr at 650째C. (a) Eracture. 1OOx. (b) Edge near fracture. 1OOX. (c) Representative unstressed structure. 500X. Reduced 21.59.

Fig. 52. Photo&crographs of a Modified Hastelloy N (Heat 7320) ing Exposed t o t h e MSRE Core f o r 7244 h r Sample Tested a t 25째C Aft a t 650째C and Irradiated t luence of 5.1 x 1 O 2 O neutrons/cm2. (a) Fracture, etched. 1OOX. (b) Edge, as polished. 1OOX. ( c ) Edge, etched. lOOX. Etchant: glyceria regia. Reduced 22%.

Fig. 53. Photomicrographs of a Modified Hastelloy N (Heat 7320) Sample Tested a t 650째C ( S t r a i n Rate, 0.002 min'l) After Being Exposed t o the MSRE Core for 7244 hr a t 650째C and Irradiated t o a Fluence of 5.1 X lo2' neutrons/cm2. (a) Fracture, etched. 1OOX. (b) Edge, as polished. 1OOX. ( c ) Edge, etched. 1OOX. Etchant: glyceria regia. Reduced 224.

Fig. 54.

Photomicrographs of a Modified Hastelloy N (Heat 7320)

Sample Tested a t 25째C After Being Exposed t o Static Barren Fuel Salt f o r 7244 hr a t 650째C. (a) Fracture. (b) Edge. 1OOX. ( c ) Typical unstressed microstructure. 5 0 0 ~ . Etchant: glyceria regia. Reduced 22.5$.

loox.

Fig. 55. Photamicrographs of a Modified Hastelloy N (Heat 7320) Sample Tested a t 650°C (Strain Rate, 0.002 min-l) After Being Exposed t o S t a t i c Barren F’uel S a l t f o r 7244 h r a t 650°C. (a) Fracture. lOOX. (b) Edge near fracture. lOOX. (c) Typical unstressed microstructure. 500X. Etchant: gylceria regia. Reduced 23%.

Fig. 56. Photomicrographs of a Modified Hastelloy N (Heat 67-551) Sample Tested a t 25째C After Being Exposed t o t h e MSRE Core for 7244 hr a t 650째C and Irradiated t o a Fluence of 5.1 X lo2' neutrons/cm2. (a) Fracture, etched. lOOX. (b) Edge, as polished. lOOX. ( c ) Edge, etched. lOOX. Etchant: glyceria regia. Reduced 21%.

72 2 mils.

A t a t e s t temperature of 650°C t h e fracture i s mixed intergran-

ular and transgranular (Fig. 57).

The microstructure near t h e edge i s

not modified, but edge cracks extend t o a depth of about 8 mils. A cont r o l sample t h a t was t e s t e d a t 25°C is shown i n Fig. 58. The f r a c t u r e

i s l a r g e l y transgranular, and t h e r e are no edge cracks nor s t r u c t u r a l modifications. The high magnification view shows the f i n e carbide prec i p i t a t e s t h a t form during t h e long thermal anneal. A t 650°C t h e fracture i s intergranular with numerous intergranular cracks scattered throughout t h e sample (Fig. 59). A few cracks are near t h e surface, but these a r e t o be expected i n l i g h t of t h e 31.5% s t r a i n (Table U, p. 40) t h a t occurred i n t h i s sample before failure. Heat 67-504 was exposed t o t h e MSRE c e l l environment of N2 + 2 t o 5% 02 f o r 17,033 h r a t 650°C and had an oxide film of 1 t o 2 mils:

when

t e s t e d a t 25°C (Fig. 60) t h e fracture was mixed transgranular and i n t e r granular, although m o s t of the deformation occurred within the grains.

There was no edge cracking.

was primarily intergranular.

15 mils.

When tested a t 650°C (Fig. 61) t h e f r a c t u r e There were edge cracks t o a depth of

Since there were no controls t o compare with these samples, it

i s d i f f i c u l t t o say whether t h e frequency and depth of cracking a r e g r e a t e r than would be expected. DISCUSSION OF RESULTS The mechanical property changes of t h e standard Hastelloy N i n both t h e irradiated and unirradiated conditions have followed very regular trends throughout i t s exposure i n t h e MSRE and i n t h e control f a c i l i t y . Heats 5065 and 5085 have been used throughout t h e surveillance program. Exposure t o t h e static barren fuel salt up t o 15,289 h r brought about a gradual reduction of t h e tensile fracture s t r a i n s i n both heats; f'urther exposure up t o 22,533 h r caused a s l i g h t improvement over t h e s t r a i n s observed after 15,289 h r ( F i g s . 15 and 17, pp. 24 and 25).

These changes

were small campared with those observed a f t e r i r r a d i a t i o n , and we attribu t e them t o t h e p r e c i p i t a t i o n of grain-boundary carbides. These changes i n t e n s i l e properties occurred without detectable change i n t h e creep strength a t 650°C ( F i g s . 19 and 25, pp. 30 and 3 4 ) .

. Fig. 57. Photomicrographs of a Modified Hastelloy N (Heat 67-551) Sample Tested a t 650째C ( S t r a i n Rate, 0.002 min-') After Being Exposed t o t h e MSRE Core for.7244 h r a t 650째C and Irradiated t o a Fluence of 5.1 X lo2' neutrons/cm2. (a) Fracture, etched. 1OOX. (b) Edge, as polished. lOOX. (c) Edge, etched. lOOX. Etchant: glyceria regia. Reduced 20%.

Fig. 58. Photomicrographs of a Modified Hastelloy N (Heat 67-551) Sample Tested a t 25째C After Being Exposed t o S t a t i c Barren Fuel S a l t f o r 7244 h r a t 650째C. (a) Fracture. lOOX. (b) w e . lOOX. ( c ) Typical unstressed microstructure. 500X. Etchant: glyceria regia. Reduced 20.5%.

' U

hs of a Modified Hastelloy N (Heat 67-551) Fig. 59. Photomicrog Sample Tested a t 650째C ( S t r a i n Rate, 0.002 min'l) After Being Exposed t o S t a t i c Barren Fuel S a l t f o r 7244 hi. a t 650째C. (a) Fracture. 1OOX. (b) Edge near f'racture. lOOX. ( c ) Typical unstressed microstructure. 500X. Etchant: glyceria regia. Reduced 20.5$.

4 4

Fig. 61. Photomicrographs of a Modified Hastelloy N (Heat 67-504) Sample Tested a t 650째C ( S t r a i n Rate, 0.002 rnin-l) Following Exposure t o the MSRE Cell Environment f o r 17,033 hr and Irradiated t o a Fluence of 2.5 x lo1' neutrons/cm2. (a) Fracture, etched. lOOX. (b) Edge, a s polished. lOOX. ( c ) Edge, etched. lOOX. (d) Edge, as polished. 500X. Etchant: glyceria regia. Reduced 25%.

78 I r r a d i a t i o n of both heats caused a general decrease of t h e f r a c t u r e s t r a i n w i t h increasing thermal fluence a t test temperatures of 25 t o

b) 6

850°C. The f r a c t u r e s t r a i n s a t low temperatures are lower f o r heat 5065. We attribute t h e reduction i n f r a c t u r e s t r a i n a t low temperatures t o prec i p i t a t i o n of carbides and demonstrated previously t h a t it can be We a t t r i b u t e the embrit-

recovered by annealing t o coarsen t h e carbide.12

tlement a t high temperatures t o t h e helium formed by thermal transmutat i o n of l 0 B t o form ‘He and 7 L i . by annealing.

This embrittlement cannot be recovered

The presence of helium i n the material influences t h e

creep properties a t 650°C by reducing t h e rupture l i f e and t h e s t r a i n a t

fracture; the creep rate i s unaffected (Figs. 19 through 26, pp. 30 through 3 4 ) . The e f f e c t s on t h e rupture l i f e are greater a t higher s t r e s s levels and decrease t o insignificant below about 10,000 p s i .

mum operating stress13 i n t h e MSRE i s 6000 p s i . )

(The maxi-

The changes i n t h e

f i a c t u r e s t r a i n are of utmost importance and are s l i g h t l y dependent upon the thermal fluence (Figs. 2 1 and 26).

Figure 23 shows more c l e a r l y the

s e n s i t i v i t y of heat 5085 t o the presence of helium and how t h i s sensitivi t y a t 650°C depends upon the s t r a i n rate.

1.3

A thermal fluence of only

lo1’ neutrons/cm2 produced about 1ppm (atomic) He and reduced t h e

f r a c t u r e s t r a i n from 30 t o 2s when the material was crept at a r a t e of O.ls/hr.

The s t r a i n rate results i n t h e lowest f r a c t u r e s t r a i n s with

slower rates r e s u l t i n g i n s l i g h t l y higher values.

Figure 28, p. 36,

shows that t h i s same general behavior holds f o r heat 5065 although t h e amounts of helium produced a r e a c t u a l l y lower. received a thermal fluence of about 4

The MSRE vessel has

10” neutrons/cm2, and Figs. 23

and 28 indicate t h a t the fracture s t r a i n s w i l l not drop much more with continued operation. fluence of about 2

The control rod thimbles have received a thermal 1021 neutrons/cm2.

They should be extremely b r i t t l e ,

but are normally subjected t o only a s l i g h t compressive s t r e s s . The greater s t r a i n rate s e n s i t i v i t y of the i r r a d i a t e d Hastelloy N

at 650°C i s i l l u s t r a t e d i n Figs. 22 and 27 f o r heats 5085 and 5065, 12H. E. McCoy, An Evaluation of t h e Molten-Salt Reactor Experiment Hastelloy N Surveillance Specimen - Third Group, ORNL-TM-2647 (1970). I3R. B. Briggs,

ORNL, privdte cammunication.

. W

19,

respectively.

material varies from 32 t o 219, a decrease of 349, and t h e i r r a d i a t e d

For heat 5085 t h e f r a c t u r e s t r a i n of t h e unirradiated

material varies from 9.3 t o 0.5$, a decrease of 959. i

For heat 5065 t h e

corresponding reductions a r e 359 f o r the unirradiated material and 97% f o r t h e i r r a d i a t e d material. The question of primary concern i s whether t h e Hastelloy N has corroded more r a p i d l y during the previous surveillance period than previously.

We have at l e a s t t h r e e parameters or phenomena that can be

observed o r measured t h a t l i k e l y a r e dependent upon t h e corrosion r a t e . The f i r s t measure i s t h e chemical change of t h e s a l t during reactor operation.

Since the chromium observed i n t h e s a l t can hardly come

fkom anywhere except t h e Hastelloy N, we cannot question t h i s parameter as a measure of t h e corrosion r a t e .

t h e microstructure.

Second, we can observe changes i n

Figures 2, 3, and 4 , pp. 9, 10, and 12, show typi-

c a l changes, but these observations a r e d i f f i c u l t t o i n t e r p r e t without additional information. i

Third, t h e mechanical properties themselves may

be a l t e r e d o r t h e crack patterns t h a t develop may be useful indications

of the e f f e c t s of corrosion on t h e mechanical properties. I

The property

changes themselves a r e not very useful i n t h i s case because we have only matched sets of control samples and surveillance samples.

The control

samples a r e exposed t o a static barren f u e l s a l t t h a t does not match the corrosiveness of t h e MSRE f u e l c i r c u i t .

The surveillance samples a r e

exposed t o a neutron fluence and t o a more corrosive s a l t c i r c u i t .

The

i r r a d i a t i o n e f f e c t s predominate, so the effects of t h e added corrosion a r e not large enough t o see

observing t h e mechanical property changes.

Thus t h e r e a r e several measures of corrosion i n t h e MSRE, but they a r e a l l subject t o i n t e r p r e t a t i o n .

Because of t h e importance of t h i s sub-

j e c t , l e t us review t h e p e r t i n e n t f a c t s and observations. 1. Before the last period of operation, t h e salt was removed from t h e MSRF, and processed

remove t h e uranium and replace it with 233U.

This processing l e f t t h

a l t fairly oxidizing, and it was necessary t o i a l b y adding beryllium metal. The chromium

a d j u s t t h e oxidation

content of the s a l t was only 40 ppm when placed i n t h e reactor and rose i

over a few weeks t o a l e v e l of about 100 ppm.

This apparent'corrosion

codld be accounted f o r by uniformly removing t h e chromium from t h e

80 Hastelloy N t o a depth of 0.3 m i l .

The t o t a l increase of chromium i n

t h e salt during t h e e n t i r e MSFE operation would require uniform chromium extraction t o a depth of 0.4 m i l .

Hawever, t h e diffusion rate along t h e

g r a i n boundaries can be about lo6 t h e r a t e through t h e bulk grains,14 and it i s more l i k e l y that chromium be removed t o g r e a t e r depths along t h e g r a i n boundaries. 2.

Electron microprobe examination of the sample i n Fig. 5 , p. 13,

revealed a d e f i n i t e chromium gradient near t h e surface and a surface chromium concentration of nearly zero.

A diffusion coefficient of

cm2/sec f o r chromium was computed based on t h e shape of t h e

p r o f i l e which i s i n excellent agreement with t h e value of 2 x

cm2/sec

measured by Grimes et a1.I5 (Fig. 62). 3.

Complete loss of chromium from Hastelloy N does not r e s u l t i n

t h e type of microstructural modification t h a t was observed, not does it cause embrittlement.16 We made laboratory melts containing from 0 t o 9% C r w i t h t h e standard Hastelloy N base composition.

The mount of car-

bide p r e c i p i t a t e increases and t h e g r a i n s i z e decreases as t h e chromium cont e n t increases.

The a l l o y is weaker i n creep a t 650째C without chromium,

but t h e f r a c t u r e s t r a i n s a r e not dependent upon t h e chromium concentration.

4. Similar microprobe scans have been made on control sanrples where t h e modified s t r u c t u r e i s a l s o present.

The resolution i s much b e t t e r

on these samples where the background r a d i a t i o n i s not present.

The gradients measured i n a sample that had been i n the c o n t r o l f a c i l i t y f o r 15,289 hr a t 650째C are shown i n F i g . 63.

t h e i r o n i s enriched near the surface.

The chromium i s depleted and These chemical modifications

extend only a short distance i n t o the material.

These changes would not

be expected t o produce t h e microstructural changes. 14W. R. Upthegrove and M. J. Sinnot, "Grain-Boundary Self-Diffusion of Nickel," Trans. Am. SOC. Metals 50, - 1031 (1958). 15W. R. Grimes, G. M. Watson, J. H. DeVan, and R. B. Evans, "RadioTracer Techniques i n t h e Study of Corrosion by Molten Fluorides," pp. 559-574 i n Conference on t h e Use of Radioisotopes i n t h e P G s i c a l Sciences and Industry, September 6 1 7 , 1960, Proceedings, Vol. 111, International Atomic Energy Agency, Vienna, 1962.

16H. E. McCoy, Influence of Various Alloying Additions on t h e Strength of Nickel-Base Alloys (report i n preparation).

81 ORNL- OWG 70-4933

hd d

SMEASURED (D= 4x cm2/sec)-

CALCULATED( D = 2 x

cm2/sec)

DISTANCE FROM SURFACE (mils)

Fig. 62. Chromium Gradient i n Hastelloy N Sample Exposed t o t h e MSRE Core f o r 22,533 hr. ORNL-OWG 70-7659 12

gci a I-

W z

Y 4

0 0

Fig. 6 3 . Concentrat t o S t a t i c Barren Fuel Sa 650°C. 5.

40 15 20 DISTANCE FROM SURFACE ( p )

radients i n Hastelloy N (Heat 5085) Exposed t h e MSRE Control Vessel f o r 15,289 h r a t

The Hastelloy N straps t h a t held t h e surveillance assembly

together f o r 22,533 h r had

ntergranular surface cracks t o a depth of

about 3 mils (ref. 17) (see Fig. 2, p. 9 ) .

Straps t h a t had been i n the

reactor f o r 7244. hr had cracks t o a depth of 1.5 mils.

The s t r a p s are

0.020 i n . t h i c k and should not be stressed during reactor operation, but

. c-d

17W. H. Cook, MSR Program Semiann. F’rogr. Rept., Aug. 31, 1969, ORNL4449, pp. 165-168.

82 they were handled considerably during disassembly.

Thus one cannot con-

clude whether t h e cracks were formed during service o r i n handling a f t e r -

wards. straps

The microstructure was not modified near t h e surface of these

A surface microstructural modification has been observed spas-

modically during t h i s e n t i r e program.

The photomicrographs of t h e

various l o t s of heat 5085 that were examined a r e presented i n Fig. 64. The modified microstructure has been present t o some degree i n a l l samp l e s including t h e i r r a d i a t e d and t h e control samples.

It i s d i f f i c u l t

t o say t h a t t h e modification has g r a m any worse.

We have been able t o produce a microstructural modification

q u i t e similar t o t h a t shown i n Fig. 65 by s i n t e r l e s s grinding Hastelloy N and then annealing it f o r long periods of time i n argon a t 650째C (Fig. 65).

Thus, t h e layer may w e l l be associated with cold working and the manner i n which t h i s alters carbide deposition near t h e surface. 8.

There i s a d e f i n i t e tendency f o r t h e samples removed from t h e

MSRE t o show progressively more edge cracking during p o s t i r r a d i a t i o n

t e s t i n g as t h e time of exposure increases.

As shown i n Fig. 66 f o r sam-

p l e s s t r e s s e d a t 25"C, t h e frequency of edge cracks'increases w i t h exposure, but t h e maximum depth is constant a t about 4 mils.

The control

samples do not show much edge cracking. 9.

Microprobe scans on i r r a d i a t e d samples f a i l e d t o r e v e a l any

f i s s i o n products i n t h e sample.

However, t h e i n a b i l i t y t o analyze s o l e l y

t h e material i n a g r a i n boundary makes it impossible t o conclude t h a t no f i s s i o n products are entering t h e metal.

We plan t o dissolve progres-

s i v e layers from t h e sample surfaces f o r chemical analysis t o answer t h i s quest ion. The observations on t h e alloys of modified Hastelloy N a r e q u i t e

important.

Heats 7320 (0.5% T i ) and 67-551 (1.1% T i ) were exposed t o

t h e MSRE core f o r 724.4 h r a t 650째C.

The properties of heat 7320 were

not as good as we have observed f o r other 0.5% Ti-modified a l l o y s irradiated a t 650째C.

However, t h e properties a f t e r i r r a d i a t i o n i n t h e MSRE

are equivalent t o those measured a f t e r i r r a d i a t i o n i n t h e ORR i n a helium environment.

The properties of heat 67-551 are outstanding.

Both of

NRADIATLD

Y-Ppy?

4.m k 8

22.533 k

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

Fig. 65. Photomicrograph of Hastelloy N (Heat 5085) R o d That Was Annealed 1 h r a t 1177OC, Sinterless Ground 5.2 mils, and Annealed f o r 4370 hr a t 650째C i n Argon. 500X. Etchant: glyceria regia.

Fig. 66. Samples of Hastelloy N (Heat 5085) Stressed a t 25째C. The control samples were exposed t o s t a t i c barren f u e l s a l t f o r the indicated time and t h e irradiated samples were exposed t o t h e core of t h e MSFE.

these heats seem t o demonstrate some small property changes due t o therm a l aging.

These changes are not large, but m u s t be followed t o longer

times t o make sure that they do not become large. These two modified heats did not show a s t r u c t u r a l modification near t h e surface, but did exhibit edge cracking when t e s t e d .

The fre-

quency of edge cracks i s much higher than noted f o r the controls. One of t h e most conf'using observations was the poor mechanical properties of heat 67-504 from outside t h e core.

This heat was pre-

viously18 exposed t o t h e core f o r 9789 h r t o a thermal fluence of

5.3 X

lo2* neutrons/cm*.

The present group of samples outside t h e core

ISH. E. McCoy, An Evaluation of t h e Molten-Salt Reactor Experiment Hastelloy N Surveillance Specimen - Third Group, ORNETM-2647 (1970).

85 was exposed t o N2 + 2 t o 5% 02 f o r 17,033 h r and received a thermal f

fluence of 2.5 x loi9 neutrons/cm2.

The l a t t e r group had poorer mechan-

i c a l p r o p e r t i e s than t h e group exposed t o t h e higher fluence. J

The

rupture l i f e a t a given stress l e v e l was l e s s (Fig. 41, p. 54), t h e minimum creep r a t e higher (Fig. 42, p. 541, and t h e fracture strain lower (Fig. 43, p. 5 5 ) .

The photomicrographs i n Figs. 60 and 61, pp. 76 and 77, /

show same surface oxidation but no other unusual features.

Two heats

should have been used, heat 67-502 (2%W + 0.5$ T i ) and heat 67-504

(0.5% H f )

0.5$ Hf,

but no tungsten was present.

Postirradiation chemical analyses showed the presence of Hence, we have concluded t h a t

only heat 67-504 was included in these tests. Our surveillance program has given us t h e opportunity t o look a t s e v e r a l alloys of modified canposition.

The creep properties of these

heats a r e compared with those o f standard Hastelloy N i n Fig. 67.

The

curves were drawn through only four or f i v e data points i n each case,

so s l i g h t differences i n slope a r e not s i g n i f i c a n t . The rupture l i v e s o f t h e modified alloys a r e within a f a c t o r of 2. The rupture l i v e s of t h e irradiated, modified heats are about equivalent t o those of unirradiated standard Hastelloy N.

The creep rates, shown i n Fig. 67(b), a r e

l w e r f o r t h e modified alloys by as much as a f a c t o r of 3.

and 67-504 have t h e lowest creep rates.

strains are shown i n Fig. 67(c).

Heats 67-502

The p o s t i r r a d i a t i o n f r a c t u r e

Quite a range of f r a c t u r e s t r a i n s

r e s u l t e d with a m i n i m of about 0.5% f o r standard Hastelloy N and a minimum of about 7%f o r heat 67-551.

Heats 67-502 and 67-504 had frac-

t u r e strains of 6.4%; heat 7320 was not much b e t t e r than standard

Hastelloy N with o n l y 2.8%. None of t h e modified alloys have shown any adverse corrosion behav-

ior i n the salt. Thus, it would appear t h a t we have several alloys t h a t a r e s u i t a b l e f o r use i n future reactors t h a t operate a t 650째C. However, as discussed previously, these new alloys a r e very s e n s i t i v e t o i r r a d i a t i o n temperature and the proposed 700째C operating temperature of a breeder i s too high f o r these a l l o y s . l g

Present work offers encouragement t h a t an a l l o y t h a t is stable a t 700째C can be developed by adding

I9H. E. McCoy et a l . , MSR Program Semiann. Progr. Rept. Aug. 31,

1969, ORNE44.49,

p. l84.

0 (DY

rnm

FRACTURE STRAIN ( %)

P 0

STRESS (4000 psi)

lJl

STRESS (4000psi) 0

---

.-.

is,

87 l a r g e r amounts of titanium o r combined amounts of these elements along t

with niobium and hafnium.

SUMMARY AND CONCLUSIONS The heats of Hastelloy N used i n fabricating t h e MSRE have s h a m a systematic deterioration of mechanical properties w i t h increasing neutron fluence.

The material exposed f o r t h e longest period of time i n t h e

core has reached a thermal fluence of 1.5 x 1021 neutrons/cm2 and a fast fluence (> 50 kev) of 1.1X

neutrons/cm2.

These values are q u i t e

close t o those anticipated f o r f u t u r e reactors with a 30-year design life.

The d u c t i l i t y of t h e material was too low, but t h e microstructure

was f r e e of irradiation-induced voids and defects other than helium bub-

bles.

Several heats of t h e modified alloys have been exposed t o the MSRE

and these have b e t t e r p o s t i r r a d i a t i o n properties.

They a l s o seem t o

have good corrosion resistance. n

The standard Hastelloy N removed from t h e core shows some evidence of corrosion.

The corrosion seems generally t o be due t o t h e s e l e c t i v e

removal of chromium, as predicted by prenuclear tests.

Some observations

t h a t have not been explained adequately are (1)t h e presence of grainboundazy cracks i n t h e s t r a p s t h a t held p a r t s of t h e surveillance assembly together, (2) t h e modified microstructure near the surface, and

(3) t h e formation of intergranular cracks originating from t h e surface

when i r r a d i a t e d materials a r e strained. One of t h e modified alloys, heat 67-504, was exposed t o the c e l l environment.

This heat had been previously exposed t o t h e MSRE.

The

fluence was higher i n t h e core, but t h e p o s t i r r a d i a t i o n properties were superior t o those of t h e material exposed t o t h e c e l l environment.

presently have no explanation for t h e observed behavior.

The author i s indebte

many people f o r a s s i s t i n g i n t h i s study:

W. H. Cook and A. Taboada f o r design of t h e surveillance assembly and i n s e r t i o n of t h e specimens; W. H. C o o k and R. C. S t e f f y f o r measurements

of flux; J. R. Weir, Jr., R. E. Gehlbach, and C. E. Sessions f o r review of t h e manuscript; E. J. Lawrence and J. L. Griff'ith f o r assembling t h e surveillance and control specimens i n the f i x t u r e ; P. Haubenreich and t h e MSRE Operation S t a f f f o r t h e extreme care w i t h which they inserted and removed t h e surveillance specimens; E. M. King and t h e Hot C e l l Operation S t a f f f o r developing techniques f o r cutting long rods i n t o individual specimens, determining specimen straightness, and assistance i n running creep and t e n s i l e t e s t s ; B. C. W i l l i a m s , B. McNabb, and H. W. Kline f o r running tensile and creep tests on surveillance and control specimens; J. Feltner f o r processing t h e t e s t data; H. R. Tinch and N. M. Atchley

f o r metallography of t h e control and surveillance specimens; Frances Scarboro of The Metals and Ceramics Division Reports Office f o r preparing t h e manuscript; and t h e Graphic Arts Department f o r preparing the d r a m s .

d A

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u5.

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188. 189.

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