The response of austenitic steels to radiation damage

181

Journal of Nuclear Materials 122 & 123 (1984) 181-190 North-Holland, Amsterdam

THE RESPONSE OF AUSTENITIC A. F. ROWCLIFFE

STEELS TO RADIATION

DAMAGE*

and M. L. GROSSBECK

Structural Materials Group, Metals and Ceramics Division Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 USA

Austenitic stainless steels are prominent contenders as first wall and blanket structural Properties affecting the performance of this materials for early fusion power reactors. class of alloys in the fusion irradiation environment, such as swelling, tensile elongation, These properties and irradiation creep, fatigue, and crack growth, have been identified. the effects of neutron irradiation on them are discussed in this paper. Emphasis is placed on the present status of understanding of irradiation effects.

1.

location and cavity sink strengths,

INTRODUCTION Breeder

generated effects

reactor materials

an extensive

in austenitic

ticularly

comparatively

about the effects which

stainless

little is kriown

of irradiation

in

enVirOntWItS

in a

In this paper we describe the

reactor.

effects

of irradiation

on three important

of properties

(cavity swelling

sile strength

and ductility,

eventually

accelerates

regime characterized

steel (316

generate helium at rates expected

fusion

ical changes and precipitate

have

steels, par-

on type AISI 316 stainless

However,

SS).

programs

data base on irradiation

sets

and creep, ten-

dose.

formation,

swelling

into a rapid swelling

by a linear dependence

This behavior

Flux Test Facility

microchem-

is exemplified

on

by the Fast

(FFTF) first core heats of For all terrperatures in the

316 SS (Fig. 1). range 500-650°C

the low-swelling,

regime, extends

out to -40 dpa (1 x lo22 n/cm2

or transient

Swelling then accelerates until a z 5 dpa). 35,,,,, I I I I , ,

fatigue and crack

growth) and summarize what is known about the effects 2.

of enhanced

SWELLING

helium

generation.

BEHAVIOR

The cavity swelling phenomenon nated radiation decade.

has domi-

SWELLING 96

damage studies for the past

Breeder

reactor data at fluences

about -100 dpa are now available and 304 stainless

of

on types 316

steels and at lower fluences

on the simpler Fe-Cr-Ni

ternary

alloys.l-4

These data have recently 'allowed a comprehensive description function complex

of void swelling

of both te~erature alloys,

swelling

first, and follows dose.

However,

in austenitics and fluence.

develops

a power-law

as a In

slowly at

dependence

as a result of changes

on

in dis-

NEUTRON FLUENCE IE ,O 1 MsVl

FIGURE 1 Swelling of 20%-cold-worked AISI 316 stainless steel irradiated in EBR-II. Four separate heats of FFTF first core steel are shown [H.R. Brager and F.A. Garner, 3. Mater. 117 (1983) 15+76.

%Research sponsored by the Office of Fusion Energy, U.S. Department W-7405-eng-26 with the Union Carbide Corporation.

0022-3 115/84/$03.00 @ Elsevier Science Publishers (North-Holland Physics Publishing Division)

B.V.

of Energy, under contract

182

A.F. Rowcliffe.

rate of approximately Swelling

O.G%/dpa

is achieved.

any reduction

The transient

in swelling

decreased

belaw 500°C.

swelling

is

also by increasing minor alloying

elements

behavior

alloys.

the concentrations

In alloys containing

Ni, the transient

of certain

ternary

15% Cr and l&19%

and the subsequent

rate approaches

l%/dpa

(ref. 3).

regime increases

progressively

above 510°C.

swelling

with temperature

increases

is decreased

or the nickel content

This new perspective ior of austenitic

stainless

steels indicates

structural

and microchemical

trol the transition linear swelling knowledge

range.

to understand

information

behavior

damage rate.9

The only

on helium effects at rele-

of 316 SS heats irradiated

in EBR-II

and in HFIR (where the He:dpa ratio is higher than that expected

in a fusion reactor by a fac-

The most recent evidence

tor of 4 to 5).

(ERR-II) to -60 lo-swelling

swelling

(HFIR) may extend or shorten the

regime, depending (Fig. 2) (ref. 10).

of bubbles,

segregation

the linear swelling

differences

low-

with high number of radiation-

(RIS), and delayed conversion A rapid transition

into

regime is associated

with

RIS-affected

rapid conversion

upon the heat An extended

inhibition

of bubbles to voids.

extensive

indi-

the He:dpa ratio from -0.5

regime is associated

densities

precipitation

and the

of bubbles to matrix voids.

voids and

An example of these

in cavity distribution

is shown in

rates

Fig. 3 for a single heat of 20%-cold-worked

It is

(CW) type 316 SS irradiated

in both the HFIR

changes which conto the

regime, and to apply this

m - HFIR O-EBR-il

of alloys which

.

will resist this transition. The possible

and precipi-

the micro-

from low-swelling

to the development

available

precipitate-assisted

regime

high-swelling

will ensue over a wide temperature therefore

is increased.5

or transient

has ended, then unacceptably

important

content

on the swelling behav-

that once the lo-swelling

in complex alloys are not repro-

induced

in the duration of the low-

regime occur when the chromium

of the segregation

behavior

chemistry

The extent of the low-swelling

Remarkable

dose dependence tation

cates that increasing

regime is very short (-12 dpa)

over the range 400-51O"C, linear swelling

and

such as Ti, Si, and P.

occurs in Fe-Cr-Ni

damage

vant damage rates is from corrparisons of the

The onset of the linear

regime is delayed by cold working

of arrstenitic steels to radiation

duced at the increased

regime extends to progres-

sively higher doses as the temperature

Similar

/Response

values in excess of 30% have been

observed without rate.

M.L. Grossbeck

effects of increasing

dpa ratio are of major concern

the He:

in the develop-

- HFlR

ID0

HEAT)

(DO HEAT) (N-LOT)

CAVITY VOLUME FRACTION (%1

ment of alloys for fusion reactor applications. Strong effects

of helium injection

tion levels on cavity nucleation demonstrated (refs. 6,7). provide

in heavy-ion

irradiations

While dual-ion

of 316 SS

beam irradiations FLUENCE

a powerful

damage phenomena, with neutron

or preinjec-

have been

means of studying a quantitative

irradiation

behavior

by surface-related

phenomena

stitial effects.8

Further,

(dRo>

radiation

correlation is confounded

and injected interthe temperature

and

FIGURE 2 Swelling of 20%-cold-worked 316 SS. Data from two heats and irradiations from two reactors are shown to conpare effects of He:dpa and heat them istry (P. J. Maziasz, this volume).

A.F. Rowcliffe, M.L. Grossbeck /Response

183

of austenitic steels to radiation damage 4).

(Fig.

Titanium

is thought to influence

swelling

behavior

residual

gases are strongly

unavailable

thus

nuclei;

in several ways:

for stabilization

(b) the cold-work

is stabilized

of void

dislocation

and maintained

and (c) in environments

(a) the

gettered and are

structure

to higher fluences;

with high He:dpa ratio

the interaction

between helium atoms and the MC

particle-matrix

interface

strongly

modifies the

helium bubble distribution.14

FIGURE 3 Microstructures of 20%-cold-worked 316 SS irradiated in EBR-II and HFIR. [P. J. Maziasz, J. Nucl. Mater. 108&109 (1982) 35W34.J

and EBR-II to approximately A quantitative structures reactors

comparison

produced

behavior

on segregation

20 SWELLING (=%) 15

in the two

and precipitation

in ref. 10.

It is not

or not the bubble-dominated

microstructure -60 dpa.

25

for the effects of cavity

are presented

known whether

of the cavity micro-

by irradiation

and evidence

number density

equal damage levels.

30

can be maintained

However,

in HFIR beyond

theory and modeling

that the transition

0

suggest

to a linear swelling

0

25

50 FLUENCE

regime

75

425

100

(dpa)

may occur more rapidly at the intermediate He:dpa

ratios relevant to fusion reactors."

The balance of evidence because lifetime would

of components

be restricted

tures >450°C. component Although

lifetimes swelling

fabricated

regime, the

from 316 SS

to 40-50 dpa for tempera-

Swelling

range 300-45O"C, discern

suggests that

of the limited low-swelling

is unlikely to limit

at temperatures

certainly

below 300°C.

occurs over the

there are insufficient

any effects

regime.

The limited amount of published

titanium

data indicates

breeder

that the addition

or niobium to the austenitic

steels prolongs

data to

of He:dpa ratio on the

extent of the low-swelling

reactor

FIGURE 4 Swelling behavior of 20%-cold-worked 316 SS and titanium-modified variations irradiated in EBR-II at 50&6OO"C. [After B. A. Chin et al., Nucl. Technol. 57 (1982) 426-35.1

the low-swelling

of

stainless

regime12313

Figure 5 compares

swelling behavior

CW 316 SS (N-lot) with developmental worked

PCA, a titaniummodified

of 20%-

25%-cold-

austenitic

stain-

less steel with a corrposition as shown in Table

I.

Both materials were irradiated

in the

HFIR at 600°C to a maximum fluence of about 40 dpa and helium levels as high as 3000 at. ppm. The 316 SS has clearly emerged sient regime.

In the

from

PCA, however,

the

tran-

the small

184

A.F. Rowcliffe,

M.L. Grossbeck

/Response

FIGURE 5 Swelling of 20%-cold-worked 316 SS steel (N-lot) and path A PCA (25%-cold-worked) (after P. J. Maziasz and D. N. Braski, this volume).

Table

I.

Alloy ComDosition

of austenitic

steels to radiation

damage

FIGURE 6 Fine bubbles in titanium-modified austenitic stainless steel irradiated to 40 dpa by dual-ion bombardment (20 at. ppm He/dpa) at 677°C [E. H. Lee, N. H. Packan, and L. K. Mansur, J. Nucl. Mater. 117 (1983) 123-331.

(wt %) number of helium atoms per bubble is lowered.

316 MFE reference Cr Ni MO

17.0 12.4 2.1

Mn Si Ti

heat

1.7 0.67 <0.05

Thus, rmch higher doses are required before the

S P Fe

0.018 0.037 Bal

Path A PCA Cr Ni MO

14.0 16.2 2.3

Mn Si Ti

1.8 0.4 0.24

S P Fe

0.003 0.01 Bal

number of helium atoms per bubble is sufficiently

high for conversion

attached to MC particles

conversion

to bias-driven

Dual-ion

voids.

beam studies on titanium-modified

steels have demonstrated generated

that when helium is

at 20 at. ppm/dpa,

sion of helium bubbles

a very fine disper-

is nucleated

tion with dislocation-nucleated (Fig. 6) (ref. 15). sink density,

Because

long-range

coupled with defects

increasing

of the very high

is prevented,

Another

in associa-

MC particles

diffusion

opment of coarse precipitate is suppressed.

have resisted

of solutes and the devel-

phases through RIS

important benefit of

the scale of bubble nucleation

that, for a given helium content,

is

the average

voids

It is possible that other types of

precipitates

with different

facial

properties

refining

surface or inter-

may be even more effective

the microstructural

and microchemi-

cal changes which are the precursors swelling

regime.17

range of austenitic

of a linear

In the U.S. Alloy Develop-

ment for Irradiation

designed

in

the scale or helium bubble nucleation

and delaying

bubbles

to bias-driven

to occur.16

Performance

program,

a

steels with compositions

to promote fine stable particle

sions is being studied.

Parallel

disper-

irradiations

in the FFTF and the HFIR are being used to determine generation 3.

the effects of an enhanced rate on swelling

MECHANICAL

PROPERTIES

In investigating radiation

environment

ties of materials, ations

helium

resistance.

the effects of the fusion on the mechanical

proper-

there are two basic consider-

in determining

the properties

to be

A.F. Rowcliffe, M.L. Grossbeck /Response

investigated identify which

a simple test that yields

mechanistic

information

I

data from

properties

to

the fusion reactor structure will be senThe tensile test was selected on the

sitive.

basis of the first criterion. second category studies

Properties

are often discovered

in the

in design

and are to some extent a function

specific

reactor design or the specific

of the

class of

alloy.

Certain properties

because

they appear to be critical to several

designs

and because there is general agreement

on the necessity Examples diation

for experimental

creep, fatigue,

large specimen importance

measurements.

properties

are irra-

fracture

Because

in this paper.

size required

of fracture

toughness

toughness

of the

and the lesser

Tensile

in austenitic

is often derived from

well established

irradiation

of 20%-CW 316 SS are

and the property changes corre-

lated with changes

in the microstructure.ls

and test temperatures

the increase

in yield

dose saturates

For

below -5OO"C,

strength with increasing

at -15 dpa.

For higher tempera-

tures, the 20%-CW 316 SS softens initially approaches

a constant

after -10 dpa. Significant peratures

value of yield

See, for example,

ductility

in excess of 8% at 575°C for material

but

stress

3000 at. ppm He, as shown in Fig. 8 (ref. 21).

to doses of 40-

50 dpa.

Early tests of HFIR-irradiated

revealed

tensile

ductilities

material

diated to 75 dpa and containing

perature

irra-

4300 at. ppm

More recent tests on the MFE reference

data from the fast reactor

and irradiation

temperature dependence

on test tern

temperature.

At a test

of 575"C, there is a very weak of tensile properties

diation temperature

on the irra-

when it is increased beyond

the test temperature.

Even though there is evi-

dence to indicate that the HFIR irradiation peratures

tern

were higher than stated,z* the weak on irradiation

temperature

allows a

valid corrparison of tensile properties made.

in

good agreement with the HFIR data.

program to separate the dependence

Comparison

to be

of data for test temperatures

of 350, 450, and 575°C shows that both ductility and strength

of only a few

tenths of a percent at 575°C for material

showing

There are sufficient

dependence

Fig. 7.

is retained at all tem-

after irradiation

irradiated

approximately

Also shown in Figs. 7,8 are curves from tests of

EBR-II

behavior

properties

total elongations

the FFTF first core heat of 316 SS irradiated

The effects of breeder reactor irradiation on the tensile

heat of 316 SS have exhibited

to about 50 dpa and containing

other tests such as tensile tests.18 3.1

FIGURE 7 Yield strength of 20%-cold-worked 316 SS (data points are for MFE reference heat irradiated in HFIR). The curves are for FFTF first core steel irradiated at 575 and 650°C.

crack growth, and frac-

Only the first three of these

will be addressed

steels,

have been selected

of such mechanical

ture toughness.

He.20

1200

can be extracted.

The other is to identify specific which

One is to

and the types of tests.

185

of austenftic steels to radiation damage

behavior

of HFIR-irradiated

mate-

rial are similar to those found on other heats of 316 SS irradiated

in fast reactors where only

trace amounts of helium are present.

However,

186

A.F. Rowcliffe, M.L. Grossbeck /Response

of austenitic steels to radiation damage

Total elongation of 20%-cold-worked 316 SS (data points are for MFE reference heat irradiated in HFIR). The curves are for FFTF first core steel irradiated at 575 and 650°C. The data point designated by x is for material tested at 675°C.

very limited data show ductilities rial irradiated

-1% for mate-

Except for temperatures

Irradiation

breeder

material

fluence

creep in

A transient

(Fig. 9).

linear dependence

period

thermal

creep.

The dependence

correction

for temperature

activation

enthalpy

on flux has been in the

Their data (after dependence

using an

of 0.114 eV derived from pro-

data)zs are shown in Fig. 10.

From this figure, creep rate appears to be an inverse function

of damage rate.

little information generation

There is very

on the effects of helium

rate on irradiation

relevant

/

,

,

,

,

T’430.C

9 110”

n/m’,

40

EaO.4 MM

data at the present time are from

program

of irradiation

and Mosedale

Fast Reactor.24

ton irradiation

,

FIGURE 9 Creep rate for 20%-cold-worked AISI 316 stainless.steel (FFTF first core heat) given by Puigh et al., ASTM STP 782, 1982, pp. 108-21.

There is an

in Fig. 9 are due mostly to

studied by Lewthwaite Dounreay

,

PARAMETER VALUES - STRESS (MPal

FLUENCE

pressurized

The rapid creep rates at the

higher temperatures

,

012345676

creep rate

creep rate on stress and a very weak dependence upon temperature.

,

20X-CW 316 SS has

by a slowly increasing

increasing

essentially

material.

creep

been well characterized.23

with

t

,

behav-

is similar to

of irradiation

reactor-irradiated

is followed

,

in excess of 600°C and

that of fast-reactor-irradiated

The phenomenon

4

(Fig. 8).

low strain rates, the tensile

ior of HFIR-irradiated

3.2

6,

in the HFIR (30 dpa, 2000 at.

ppm He) at 575°C and tested at 675°C

especially

2

creep.

The nest

Path A PCA irradiated

experiment tailored

tubes of both 316 SS and the MFE in the ORR in an

where the neutron spectrum

is

to achieve the fusion reactor He:dpa

ratio typical Figure

of fusion reactor irradiation.

11 shcws data from an interim exami-

nation at 5 dpa where a helium level of about 50 at. ppm was achieved. is apparent

reactor data from another first core).

At 330°C no difference

between the ORR data and the fast heat of 316 SS (FFTF

At 500°C the ORR data fall signif-

icantly below the fast reactor data.

However,

at this early stage of the experiment

this dif-

ference

in behavior

could equally well be

related to differences higher exposures

in heat chemistry.

and control experiments

Both using

187

A.F. Rowcliffe, ML. Grossbeck / Response of austenitic steels to radiation damage an area of concern

in cyclic operating

tokamak

Even though the design trend is

reactors.

toward progressively

longer burn times or even

steady state, fatigue

is still a concern because

of startup and shutdown ruption events.

as well as plasma dis-

Again, the data most relevant

to a heliu~producing

irradiation

are from irradiations

in the HFIR.

20%CW

to damage levels as high

316 irradiated

environment Specimens

of

as 15 dpa and levels of helium up to 900 at. ppm irradiated

at 430 (ref. 26) and 55O'C (ref. 27)

have been examined diation

reduction

in fatigue

(Fig. 12); hwever, FIGURE 10 Irradiation creep strain rate per unit stress per dpa in cold-worked AISI 316 stainless steel showing flux dependence. The data were adjusted for temperature using an activation enthalpy of 0.114 eV [data from G. W. Lewthwaite and D. Mosedale, J. Nucl. Mater. 90 (1980) 205-153.

the same steel irradiated

in fast reactors are

in progress. 3.3

recognized

life by a factor of 3 to 10

life is observed

(Fig. 13).

raising the test temperature irradiation fatigue

is a property

temperature

100°C above the

to 650°C fails to reduce

life over the unirradiated

material.zs

these tests at low doses have not

revealed

any significant

on fatigue

example,

in

Even

Although

serious

that was clearly

The 43O'C irra-

at 55O*C, no reduction

effect of irradiation

life, it has been shown that under

other testing conditions,

Fatigue properties

Fatigue

fatigue

in fatigue.

and test terrperature data exhibit a

helium may have

effects on mechanical

properties.

Batra et al.29 demonstrated

For

a frequency

early in fusion reactor research as

-.-

PCA 25%

-

TYPE 316 STAINLESS

C.W.

EFFECTIVE

STEEL

STRESS

(MPa)

FIGURE 11 Irradiation creep in austenitic stainless steels irradiated in the ORR in a spectrally tailored experiment designed to achieve the fusion He:dpa value. The cross-hatched regions show fast reactor data for FFTF first core 316 reported by Puigh et al., ASTM-STP-782. (a) 330°C and (b) 500°C.

helium become more significant rate is lowered.

as the strain

The sensitivity

of the fatigue

test to helium effects would be increased by interjecting fatigue

a tensile

cycle.

hold period into the

This mode of testing would also

be mOre relevant to the operation

of a tokamak

fusion reactor than the conventional sinusoidal

sawtooth or

fatigue cycle.

Oata on fatigue crack growth rates are fimited to very low levels of displacement damage. FIGURE 12 Fatigue life of 20%-cold-worked 316 SS irradiated in the HFIR (6-15 dpa, ZOO-900 at. ppm He) at 430°C and tested at 430°C.

The highest dose data available

of Michel and Smith3" EBR-II to -11 dpa. irradiation

No significant

However. the 20%-CW 316 SS

an irradiation-induced

increase in

growth rate by about a factor of 10 when

similarly 593OC.

irradiated

specimens were tested at

Nhen a hold period was introduced

the fatigue cycle, an acceleration growth rate was observed and 20X-W effects

conditions.

produced

in an acceleration

dependence

on the fatigue

life of 326 SS pre-

injected with 800 at. ppm He at 600°C; a reduction in fatigue life by a factor of -50 occurred when the test frequency was reduced from 5 to The work of Van der Schaaf et al.30

0.5 Hz.

demonstrated

that irradiation

very low fluences tility

resulted

of

AISf

304 t0

in reductions

in due-

in low strain rate creep tests at 6OO"C+

They ascribed enhanced

this ductility

intergranular

these results

loss to

crack growth.

helium-

Clearly,

indicate that the effects of

of crack

There are no data on the

of helium on crack gro.+th. However,

toughness

4.

into

for both the annealed

is expected that any reduction

FIGURE 13 Fatigue life of 20%-cold-worked 316 SS irradiated in the HFIR (g-15 dpa, 40~00 at. ppm He) at 550°C and tested at 5!WC.

effects of

and 20% C@ 316 SS irradiated

and tested at 427°C.

crack

in

on crack growth rate were observed

for both annealed

exhibited

is that

for 316 SS irradiated

it

in fracture

by irradiation

would

result

of fatigue crack growth.'"

SUMMARY I,

20X-W

The transient

swelling

regime for

316 SS is affected by helium generation

rate and may be extended upon heat chemistry.

or shortened

Present

that cavity swelling will probably application 40-50

limit the

temperatures

The titanium-edified

greatly

improved swelling

breeder

reactor conditions.

is strong evidence ~00~3000

suggests

of this material to fluences

dpa for operating 2.

depending

evidence

of

>3OO"C.

steels exhibit

resistance

under

Furthermore,

there

that helium con~~ntratjons

ppm can be a~~o~dated

in a disper-

sion of fine bubbles attached to MC particles.

of

189

A. F. Rowefiffe, ML. Grossbeck /Response of atrsteniticsteels to radiatiotadamage it is thought

that further

steels could result maintain levels

in microstructures

a low swelling

approaching

high He:dpa 3.

devefopment

of these which

rate mode to fluence

100 dpa in environments

with

ratios.

Although

the irradiation

creep behav-

ior of both 316 SS and titaniummodified is well characterized, generation

5.

rate on creep have yet to be

6.

The tensile ductility

SS is not seriously diation

of 20%-CW 316

degraded by neutron irra-

up to -50 dpa at temperatures

even when high levels of helium At these helium

generated* embrittlement peratures

is expected

G575'C

(-3GOO ppm) are

levels, serious

to occur at tern

Fatigue

life of 20%~CW 316 SS is not

seriously

inpaired by irradiation

43O-550°C

even in the presence of >900 at. ppm

It is recommended

testing

typing

in austen-

steels would be to interject

hold periods

discussions

into the fatigue cycle.

10.

for

and Frances Scarboro for

the manuscript*

The assistance

is also gratefully

11.

of

D.N. Braski and R. L. Kfueh in reviewing manuscript

9,

that a more appropriate

The authors wish to thank P.J. Maziasz helpful

8.

to -I5 dpa at

mode to assess helium effects

itic stainless tensile

7”

>600°C.

5.

He.

4.

steels

the effects of helium

determined. 4.

3.

the

appreciated.

12"

REFERENCES 1.

2,

W.J.S. Yang and F.A. Garner, "Relationship Between Phase Development and Swelling of AISI 316 During Temperature Changes," p. 186206 in: Proc. Svmo. on Effects of Radiation on Materials, S&ttsdale, AZ, 1982, ASTM-STP-782. B.G. Makenas, J.F. Bates, and J.W. Gost, '"Swelling Behavior of 20% CW 316 Stainless Steel Cladding Irradiated with and Without Adjacent Fuel," p. 1%34 in: Proc. Symp. on Effects of Radiatian on Materials, Scottsdale, AZ, 1982, ASTM-STP-782.

13.

14.

15.

F.A. Garner

and W.G. Wolfer,

"'Recent

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