Comprehensive nuclear materials 4 06 radiation effects in refractory metals and alloys

Comprehensive nuclear materials 4 06 radiation effects in refractory metals and alloys Comprehensive nuclear materials 4 06 radiation effects in refractory metals and alloys Comprehensive nuclear materials 4 06 radiation effects in refractory metals and alloys Comprehensive nuclear materials 4 06 radiation effects in refractory metals and alloys Comprehensive nuclear materials 4 06 radiation effects in refractory metals and alloys Comprehensive nuclear materials 4 06 radiation effects in refractory metals and alloys

K J Leonard

Oak Ridge National Laboratory, Oak Ridge, TN, USA

Published by Elsevier Ltd.

4.06.4.2 Irradiation-Induced Swelling and Physical Property Changes in Mo and

4.06.5.1 Introduction and Irradiated Properties Database for W and W Alloys 2064.06.5.2 Irradiation-Induced Swelling and Physical Property Changes in W and W Alloys 206

NERVA Nuclear experiment for rocket vehicle

UWMAK-III University of Wisconsin Madison

fusion reactor concept

181

conditions In addition, significant issues related

to low-temperature irradiated mechanical property

degradation at even low neutron fluences restrict the

use of refractory metals Protection from oxidizing

environments also restricts their use, unless suitable

protection or a liquid metal coolants is used

Much of the early research on refractory metal

alloys was centered on applications in aerospace as

well as cladding and structural materials for fission

reactors, with particular emphasis on space reactor

applications Reviews concerning the history of

these programs and the development of many of

the alloys whose irradiated properties are discussed

cancellations and reintroduction of new mission

cri-teria for these space reactor programs, the macri-terials

database shows similar waves in the gains of

intellec-tual knowledge regarding refractory alloy and

irra-diated property behavior Unfortunately, as seen in

the subsequent sections of this chapter, much of the

irradiated property database for refractory metals

con-sists of scoping examinations that show little overlap

in either material type, metallurgical conditions (i.e.,

grain size, impurity concentrations, thermomechanical

treatments), radiation conditions (i.e., spectra, dose

and temperature), or postirradiation test conditions

or methods

The irradiation behavior of body-centered cubic

(bcc) materials is known Irradiation-induced swelling

because of void formation in the material lattice is

displace-ment damage levels up to 50 dpa (displacedisplace-ments per

atom), but typical values for fission reactor

sensitivity to swelling, for example, rhenium additions

to molybdenum or tungsten These levels of swelling

are manageable through the appropriate engineering

design of components

The generation of dislocation loops and point

defects provide significant irradiation-induced

strengthening or hardening of refractory metals and

alloys This in turn creates reductions in the ductility

and fracture toughness of the material This is most

mobility is reduced The increase in the yield

strength of the material because of the

irradiation-induced defects can exceed the fracture strength

of the material, leading to brittle behavior These

degradations in material property can begin to

n cm2,

and increase in severity with dose

As irradiation temperatures increase, dislocation loopand void sizes increase, whereas their densities arereduced, providing improvements in ductility, though

at a reduced strength of the material At high enoughtemperatures, recovery of properties to levels close tothat of the unirradiated values is possible, thoughchanges in material properties may be further influ-enced by microstructural changes such as segregation

or precipitate formation of solute and transmuted cies or recrystallization, which can lead to further dete-rioration of properties Detailed information on the

1.03, Radiation-Induced Effects on Microstructure,

and Ductility of Metals and Alloys In general, the use

of refractory alloys in radiation environments is not

new research work, particularly on molybdenumand its alloys, has shown that control over interstitialelement contamination levels, grain size, and morphol-ogy, as well as the introduction of oxide dispersionstrengthening, can lead to improvements in the low-temperature irradiation behavior This is discussed indetail in this chapter

The following sections of this chapter deal withthe irradiated properties database of niobium, tanta-lum, molybdenum, and tungsten, as well as theiralloys While vanadium may sometimes be consid-ered a refractory metal, its melting temperature isconsiderably lower than that of the other materialsmentioned However, its radiation effects database

is considerable and well advanced relative to somerefractory metals and it is therefore discussed sepa-

alloys is particularly thin, especially involving ture toughness properties, irradiation creep effects,and combined radiation effects with high thermome-chanical loads such as those experienced in plasmafacing components or spallation target materials.Where needed, a comparison of the unirradiated andirradiated properties of a material is given

and Nb AlloysThe push for higher operating temperatures in tur-bine engines, as well as in reactor designs for bothterrestrial and space applications, has frequently

governed the periodic scientific examinations of

refractory alloys The historical examination of Nb

and its alloys is typical of this, with early studies of

the irradiation properties exploring the potential uses

of these alloys in fusion energy and fission type space

reactor While these alloys have favorable properties,

such as elevated temperature capability and

compati-bility with liquid alkali metals for energy applications,

and attractive physical properties such as thermal

conductivity, much of the work on Nb and Nb-base

alloys has examined the nonirradiated properties

It is worth putting into perspective the relatively

small commercial market for niobium-base alloys

Approximately, 75% of all niobium metal is used

as minor alloying additions in steel, and only 1–2%

is produced in the form of niobium-base alloys

The total market for niobium-base alloys in the

(100 metric tonnes

used in 1961 for the SNAP-50 reactor program

alone, and substantial additional quantities were

used for other research projects such as the NERVA

(Nuclear Experiment for Rocket Vehicle

vari-ous space reactor programs, dozens of alloys were

examined, with several brought to near-commercial

production However, today only the Nb–1Zr and

C-103 (Nb–10Hf–1Ti) alloys remain commercially

available for use in the sodium vapor lamp and rocket

or turbine engine exhaust nozzles

Nb–1Zr has historically been considered the

only niobium-base alloy with a sufficiently mature

database (mechanical properties including thermal

creep, chemical compatibility, fabrication, and

weld-ing knowledge) to be considered a near-term

high ductility and good weld characteristics, the alloy

shows less-than-desirable thermal creep strength at

elevated temperatures compared to other refractory

alloys Though the C-103 alloy has greater short-term

elevated temperature strength than that of Nb–1Zr,

its long-term properties show no improvement over

alloy with a significant radiation effects database

Despite the periodic programmatic interest in the

use of Nb and Nb-base alloys, no clear fundamental

study of the irradiated properties for a specific

appli-cation has been performed or completed Much of the

data available on the irradiated properties is scattered

and easily spans a time frame of several decades,

which can lead to misinterpretations of results on

the basis of either the limited scientific knowledge

of the time, lack of understanding of the sensitivity

of properties on impurity concentrations, or agingeffects Radiation effects data are limited to theexamination of swelling and tensile properties, with

no information regarding fracture toughness or diation creep performance

irra-The following sections deal with radiation effects

on the properties of Nb and Nb–1Zr specifically.While some initial scoping examinations havebeen performed on other Nb-base alloys, these arerelatively inconsequential and based on the less-than-desirable ductility, thermal stability, or welding cap-abilities of these alloys

and Nb-Base AlloysLike all group V transition metals, the affinity of Nbfor, and its ability to dissolve, high concentrations ofinterstitial atoms such as hydrogen, oxygen, nitrogen,and to a lesser extent, carbon can strongly influencethe properties of the metal through defect–impurityinteractions Hydrogen, carbon, and oxygen impuritieshave a strong effect on the tensile Ductile–brittletransition temperature (DBTT) of pure Nb (reviewed

), with hydrogen levels near 10 ppmincreasing the DBTT to 173 K and over 273 K at

carbon were less severe, but influential at levels

of 100 ppm and greater The effect of nitrogen onembrittlement also appears to be as severe as that ofoxygen, though some uncertainty exists as to whether

The effect of interstitial impurities on the irradiatedproperties of Nb and Nb-base alloys is significantand has been examined, though the overall databasefor irradiated properties is limited

The interplay between the radiation-created defectsand the interstitial impurity elements was investi-gated by Igata et al.16

for pure Nb (70 wppm oxygen

(E > 1 MeV) at temperatures below 413 K and irradiation annealed up to 973 K Increases in yieldstrength over the as-irradiated values followingannealing were measured at 473 and 673 K, attrib-uted to the interplay of the defect clusters trappingoxygen and nitrogen atoms, respectively Above 773 K,

post-no difference between the annealed and as-irradiatedyield stress was observed

impurities in irradiated Nb through positron

annihi-lation studies In high-purity Nb, vacancy clustering

within the collision cascades is observed, starting as

low as 160 K, with vacancy migration peaking around

250 K, but in materials with higher hydrogen content,

the vacancy migration stage shifts to temperatures

swelling, irradiation creep, and helium embrittlement

through processes involved in (n,a) reactions or

examined the effect

of He and its interaction with vacancies in pure Nb,

leading to the development of bubbles through

a-irradiated specimens At temperatures between

623 and 1023 K, bubble growth occurs through the

addition of He atoms and vacancies, followed by

migration and coalescence at higher temperatures,

eventually leading to the annealing out of the He

The irradiation-induced swelling of pure Nb

gen-erally appears at temperatures between 673 and 1323 K

limits are not clearly defined and are based on the very

considerable scatter in the literature, possibly

reflect-ing the influence of impurity concentrations and

dif-ferences in irradiation conditions and microstructural

that void concentration increased four to seven times

for a fourfold increase in flux for the same total fluence

This produced a reduction in void size with flux and

therefore a reduction in the total swelling

of oxygen and substitutional binary alloy additions on

and morphology were found to be dependent on

temperature, oxygen concentration, and the type of

substitutional alloy addition The average void

diam-eter was found to increase with temperature as well as

oxygen up to 0.02 at.% Higher oxygen

concentra-tions resulted in a decrease in void diameter to

0.1 at.% O, above which void diameters showed no

significant changes The number density of voids was

found to decrease with temperature, but increase

the number density showed no significant change

As the volume fraction of swelling (DV/V) is

propor-tional to both the void number and the cube of the

void diameter, the volume fraction is observed toincrease with temperature and oxygen concentration

the volume fraction above 0.1 at.% The dependence

examina-tion revealed an ordering of the voids into a type structure in the material irradiated at 1050 K

oxygen The higher temperature of the maximumswelling as compared to the neutron irradiation data

is believed to be associated with the higher

though the higher impurity levels may also provide

an influence

The effect of dilute (2.4 at.%) substitutionalalloy addition on the swelling of 0.06 at.% oxygen-

increase through the addition of Ta, but decreasedwith increasing effectiveness by the addition of Ti,

Zr, V, and Hf The addition of the reactive alloyingelements to Nb suppresses void formation throughthe gettering of interstitial impurities that act as void

unaffected by the addition of Ni or Fe The

Figure 1 The dependence of void volume fraction ( DV/V )

in 3 MeV58Ni+ion-irradiated Nb on the concentration of oxygen and dilute solute additions Reproduced from Loomis, B A.; Gerber, S B J Nucl Mater 1983, 17, 224–233.

Swelling in Nb–1Zr has been examined, though

only scattered data are available in the examination of

temperature and flux dependence The available

lack of data on the temperature range in which peak

swelling appears The swelling data shown in the

figure were measured through electron microscopy,

and

interpretation and measurement error, may account

for the scatter associated with the lower

and Garner

et al.27

indicates that irradiation-induced swelling is

dependent on the thermomechanical history of the

material In that material, cold-working followed by

solution anneal and aging exhibited swelling, while

material not given the preirradiated cold-working

showed some densification The changes in density

of the material are dependent on the phase-related

transformations involving precipitation

Swelling in Nb–1Zr appears to be centered over a

more narrow temperature range than in Nb, with a

peak near 1073 K that is higher than that of the pure

metal While the addition of Zr to Nb appears to delay

nucleation of voids to higher temperatures, the voids

that form are of larger size than those appearing in

pure Nb under comparable conditions For example,

n cm2 (E > 0.1MeV) at 1063 K, the diameter, concentration, andvolume fraction of voids in Nb–1Zr was 57.5 nm,

under similar conditions, the same void parameters in

While void formation and swelling in Nb and

and within engineering limits, even for high neutronexposures >10 dpa.3

The addition of Ti to Nb wasfound to increase void resistance and has been found

to suppress void formation in V at concentrations as

ele-ments and Nb in the C-103 alloy may suggest agreater void formation resistance than in pure Nband Nb–1Zr

Irradiated Nb and Nb AlloysLittle coverage of the changes in mechanical proper-ties following irradiation has been given to Nb and

Nb alloys, with the majority of the data for tures below 800 K Some preliminary experimentalwork on the irradiated mechanical properties of Nb

are not commercially produced and have shownindications of thermal aging instabilities, leading to

mechanical properties of these alloys show similarradiation hardening as in the pure metal, but withmechanical properties more sensitive to thermalaging conditions The bulk of the irradiated mechan-ical properties data is for the Nb–1Zr alloy as well asthe pure metal, and is covered in this review.The irradiated mechanical properties of Nb andNb–1Zr are strongly governed by irradiation temper-ature, determining whether the mechanical propertiesare controlled by dislocation loops or a combination ofloops and voids in the microstructure As cavity for-mation can be delayed or suppressed by higher irradi-ation temperatures in Nb–1Zr, mechanical propertycomparisons between the alloy and the base metal willreflect their irradiated microstructure For Nb and

pure metal contains both dislocation loops and voids,

compar-ison of the tensile properties of Nb and Nb–1Zr

The irradiated strength of both materials shows anincrease in tensile strength above the unirradiated

Figure 2 Swelling as a function of irradiation temperature

and dose for neutron-irradiated Nb–1Zr from available

literature compiled by Powell et al.24and Watanabe et al.25

condition, with Nb–1Zr showing a greater sensitivity

to irradiation As the mechanical properties of Nb–1Zr

are dominated by the dislocation loop structures, yield

instability is observed in the material, leading to the

elongation, though total elongation near 10% is still

achieved The yield instability is associated with

dislo-cation channeling, in which deformation dislodislo-cations

will create defect-free channels along their slip plane,

following the annihilation of the loop structures This

occurs only after enough applied stress is achieved to

overcome the obstacles, but the net effect is a

nonuni-form plastic denonuni-formation through channels that allow

for the movement of deformation dislocations at

reduced stress

The irradiated Nb samples whose properties are

loops, voids that limit dislocation channeling by

providing added obstacles to deformation, resulting

in some measure of uniform elongation and work

hardening upon yielding The microstructure

depen-dence on the tensile properties can best be illustrated

at 328 and 733 K The higher irradiation temperature

results in the development of microstructural voids

and thus the significant differences in the tensile

curves The lower irradiation temperature results in

dislocation channeling following yield and the

associated work softening during necking to failure ataround 11% total elongation While the higher irradi-ation temperature sample was irradiated to a highertotal fluence, the effect of dose is observed only on the

Figure 3 Comparison of tensile properties between Nb and Nb–1Zr tested under similar irradiation conditions.

Reproduced from Wiffen, F W In Refractory Alloy Technology for Space Nuclear Power Applications, CONF-8308130; Cooper, R H., Jr, Hoffman, E E., Eds.; Oak Ridge National Laboratory: Oak Ridge, TN, 1984; pp 252–277.

4

200

0 0

400 600

800

8 Elongation (%)

relative strength increase over the unirradiated

condi-tion The higher irradiation temperature produced

voids in the microstructure, providing additional

obsta-cles to deformation and higher uniform elongations

and modest work hardening

Little is known with regard to the aging properties

of Nb–1Zr or the combined thermal and radiation

effects The addition of 1 wt% Zr to Nb creates a

dispersion-strengthened alloy, in which the Zr

com-bines with interstitial impurities creating fine

preci-pitates throughout the material The development of

these fine precipitates on aging at 1098 K can increase

the tensile strength between 50 and 100 MPa over the

annealed condition and provide an effective

strength-ening greater than that observed through modest

irradiation31(Table 1)

Irradiation of Nb–1Zr to 0.9 dpa at 1098 K showed

a modest increase in yield and ultimate tensile

strength to 135 and 192 MPa, respectively, over the

annealed condition This increase in tensile strength

either through aging or irradiation results in a

cor-responding decrease in uniform elongation from 15%

to 3.5% and total elongation from 25% to 15%

Aging at temperatures above 1098 K produces little

effective hardening as the precipitates coarsen in the

1398 K of Nb–1Zr showed only a modest increase

in the yield strength over the aged and annealed

specimens, though ultimate tensile strength and

elon-gation were unchanged or less Irradiation to 1.88 dpa

at 1223 K resulted in weaker tensile properties

compared to the 0.9 dpa sample, believed to be due

to further precipitate coarsening The time underirradiation conditions for the 1.88 dpa sample wasnear 1100 h and produced similar tensile properties

as that of the aged-only material

As discussed in the preceding paragraphs, theirradiated properties of Nb and Nb–1Zr are governed

by their microstructure and are influenced by ature, displacement damage rate, and neutron spec-trum The tensile properties of neutron-irradiatedNb–1Zr for damage levels between 0.1 and 5 dpa(Horak et al.34

below 800 K, a large increase in the tensile strengthfrom irradiation is observed with the correspondinglow uniform elongations At higher temperatures,uniform elongation increases because of the presence

of voids in the microstructure However, the data

remain-ing low up to 1100 K, while radiation hardenremain-ing isrelatively moderate, suggesting that impurities arethe source of the reduced elongation values

No irradiated fracture toughness data exist for Nb

or Nb–1Zr, though comparisons can be made from thelarger irradiated vanadium alloy database, in whichfracture toughness embrittlement becomes a concernwhen tensile strength exceeds 600–700 MPa and there-

How-ever, if a conservative value is assigned to the critical

lower than that observed in vanadium alloys),

Table 1 Tensile property comparison illustrating the effects of aging and irradiation on the mechanical properties

of Nb–1Zr

Test/aged/irradiated

temperature (K)

Yield strength (MPa)

Ultimate tensile strength (MPa)

Uniform elongation (%)

Total elongation (%)

fracture toughness becomes a concern at

ten-sile strength above 800 K is close to the unirradiated

values, uniform elongation values remain low until

con-servative approach towards engineering design needs

to be taken with this alloy

The mechanical properties of irradiated refractory

alloys can be influenced by the formation of He

grain boundary formation of bubbles and the eventual

embrittlement of the material Some scoping

investiga-tions on the effect of He on the irradiated mechanical

investigated the high-temperature mechanical

conducted at 1273 and 1473 K, no significant effect of

He on the strength or ductility of Nb–1Zr was observed

for samples containing 2–20 appm He Later analysis

of the creep ductility reductions was found to be

dependent on the observed precipitate phase

develop-ment through the pick-up of oxygen during

found no significant effect on ductility up to 80 appm

around 1% between test temperatures of 723 and

in undoped material; this is believed to be due to the

formation of He bubbles in the grains of the material

acting similar to voids in generating obstacles todislocation channeling In general, no detrimentaleffects on mechanical properties were reported foraccelerator-injected He between 1273 and 1473 K for

Ta and Ta AlloysTantalum and its alloys have historically been exam-ined for high-temperature nuclear applications, par-ticularly in the various space reactor programs Forreasons similar to those of Nb and its alloys, variousalloying combinations of Ta were examined, particu-larly in the late 1950s to 1960s Much of this effortemphasized the development of solid solution (W and

Re additions) and dispersion-strengthened (Hf tion) alloys While Ta-alloys pay a penalty in higherdensity over, for example, Nb, and decreases the lowtemperature density-compensated strength to com-parable values on Nb-base alloys The higher melting

retention above 1000 K and in density-compensatedcreep strength.12,41

Early work on substitutional solid

led to limited examination of this alloy for irradiationenvironments The improved strengthening by addi-tion of a maximum of 10 wt% allows the retention of

5 10 15 20 25 30 35

However, the use of Ta–10W in space reactor

appli-cations where liquid alkali coolants are considered

was unacceptable because of the lack of oxide

getter-ing elements such as Hf that form stable

dispersion-strengthened structures The T-111 (Ta–8%W–2%

Hf) alloy, with its demonstrated compatibility with

liquid alkali metals and improved strength over pure

Ta while retaining ductility and weldability, has been

a lead candidate alloy in space reactor systems

been made on the Ta–10W and T-111 alloys, the

irradiation properties database is very small

Irra-diated mechanical property behavior follows typical

bcc alloys in which radiation hardening effects

including limit ductility appear and are expected at

of Ta and Ta-Base Alloys

Swelling data for Ta and its alloys are limited to a few

experimen-tally observed through TEM examination of material

estimation of the bulk swelling taken from

microstruc-tural void size density data of that study is shown in

Figure 6 Void concentrations in the material were

highest at the peak swelling temperature and decreased

with higher irradiation temperature with an associated

increase in cavity size Ordering of the voids at the peak

swelling condition was reported to occur along the{110} planes in the bcc structure A subject of consid-erable theoretical debate, the mechanisms of voidordering that have appeared in bcc and fcc metals

struc-tures in the microstructure of the higher temperatureirradiated Ta appear as the size of the voids increase,

micro-structural analysis correlate well with the immersion

which an empirical equation for percent swelling as a

The broader width of the swelling peak as a function

of irradiation temperature for the calculation

errors in the accurate irradiation temperature ofthese early measurements Experimental evidence

of decreased swelling at higher fluences was reported

and attributed to the tation of Ta to W, resulting in a shift in the latticeconstant Similar effects have been more closelyexamined in Mo and TZM alloys, and attributed toimpurity segregation at void surfaces leading to shrink-age of the voids.53

transmu-Swelling measurements in Ta–10W and T-111alloys are limited specifically to work by Wiffen,

(E > 0.1 MeV), no swelling in T-111 was observed,though a possible densification of up to 0.36% mayhave occurred as evidenced in length measurements

In companion irradiations to that of pure Ta already

(E > 0.1 MeV) at temperatures between 698 and

postirradiation examination involving TEM analysis.The microstructure of the irradiated Ta–10W con-tained fewer voids than the companion Ta samples,with a lower swelling assumed in the Ta–10W alloy

Figure 6 Swelling data for pure Ta measured through

microstructural void density measurements by Wiffen46and

from immersion density measurements by Bates and Pitner.47

4.06.3.3 Mechanical Properties of

Irradiated Ta and Ta-Base Alloys

The overall mechanical property data for irradiated

Ta and Ta-base alloys are very limited, with most

In general, the behavior of Ta and its alloys is similar

to that of other bcc materials in that radiation

harden-ing is observed with significant reductions in

As is discussed in this section, the addition of solute

strengthening elements creates an increased sensitivity

to radiation hardening of the material In addition to

the lack of high-temperature irradiation behavior,

impact and fracture toughness data for irradiated Ta

and Ta alloys are also limited

As with all refractory metals, the mechanical

be-havior of pure Ta is highly dependent on the

impu-rity levels in the material This may explain the

observed differences between the work of Brown

et al.54

of 800 MeV proton

While chemical analysis quantifying the purity of

Ta was not reported in the former, irradiation to

0.26 dpa resulted in a yield strength increase from

350 to 525 MPa over the unirradiated value with a

corresponding drop in ductility below 2% Flow

insta-bility following yield was characteristic of samples

high-purity Ta irradiated to 0.6–11 dpa tested at room

temperature and 523 K showed similar increases in

tensile strength, while the uniform elongation

re-mained near 8% following irradiation to 0.6 dpa or

higher.55

The tensile properties of neutron-irradiated Ta

first, irradiation to 0.13 dpa (where irradiation to

though no significant loss in ductility occurred overthe unirradiated control However, work softeningfollowing the yield drop was observed

Irradiation to higher displacement doses in pure

temperature limitation of Ta Following irradiation to1.97 dpa at 663 K, yield and ultimate tensile strengthsincreased to near 600 MPa with a corresponding drop

elonga-tion near 10% The observed plastic instability, tributed to the lack of uniform elongation followingyielding, resulted from dislocation channeling Somerecovery of ductility is observed following irradia-tion to 913 K, which correlates with temperaturesapproximating the maximum swelling temperature(Figure 6) and a change in the dominating microstruc-tural features influencing deformation behavior in

along with the irradiated properties of T-111, whichare discussed later

investi-gated tensile behavior as a function of fluence forpure Ta, Ta–1W, and Ta–10W, establishing deforma-tion mode maps for pure Ta and Ta–1W that outlinethe conditions in which brittle failure and uniformand unstable plastic deformation occur Following

progressive hardening and gradual loss in ductilityare observed in the tensile properties of pure Ta,leading to a near doubling of the yield stress by

400

600 800

pp 131–140 and (b) Reproduced from Chen, J.; Ullmaier, H.; Floßdorf, T.; et al J Nucl Mater 2001, 298, 248–254.

An early onset of necking or plastic instability was

observed in Ta at doses above 0.0004 dpa The lower

elongation strains in the pure Ta compared with the

is believed to be due to the

The introduction of 1 wt% W resulted in a

near-30% increase in unirradiated strength over pure Ta

(Figure 9(b)) The Ta–1W alloy showed greater

sen-sitivity to radiation hardening than the pure metal

The tensile properties as a function of dose were

similar to those of the pure Ta However, above

0.004 dpa, plastic instability becomes more

predomi-nant in the Ta–1W alloy and occurs immediately

following yielding For Ta–1W irradiated from 0.7 to

7.5 dpa in a mixed proton and neutron irradiation from

the same study, hardening was saturated with little

Macroscopic deformation mode maps produced

graphi-cal way of predicting the performance of a material in

an irradiation environment The deformation mode

plastic instability stress were directly obtained fromtensile data, while the fracture stress was calculatedthrough a linear strain hardening model for neckingdeformation, assuming that during instable deforma-tion, the strain hardening rate remains unchanged and

is approximately the plastic instability stress The ture and plastic instability stresses are independent of

materials studied The fracture strength decreases withdose if the material becomes embrittled, for example,through interstitial segregation or secondary phaseprecipitation at grain boundaries, though this was notobserved in their work The yield strength is highlydose dependent, though the yield stress was signifi-cantly lower than the fracture strength in Ta–1W,suggesting that the material may show limited ductil-ity to even higher displacement doses The effect ofincreasing test temperature for each material furtherincreases the boundaries for uniform deformationbehavior This increase was found to be greater inpure Ta

Ultimate

Yield

Yield

Yield Yield

Figure 8 Comparison of tensile properties of neutron-irradiated Ta and T-111 Uniform elongations of <0.3% for the

663 K irradiations are not shown in the figure Reproduced from Wiffen, F W In Refractory Alloy Technology for Space Nuclear Power Applications, CONF-8308130; Cooper, R H., Jr, Hoffman, E E., Eds.; Oak Ridge National Laboratory: Oak Ridge, TN, 1984; pp 252–277.

The room temperature unirradiated tensile strength

of Ta–10W is nearly double the value of the Ta–1W

and triple that of pure Ta in the material investigated

sensitivity in radiation hardening over the pure metal

(Figure 9(c)) This sensitivity is also clearly apparent

at higher irradiation temperatures near 673 K, as

shown in the comparison of tensile curves that were

(Figure 11) Near room temperature irradiation of

Ta–10W to the mixed proton and spallation neutron

25.2 dpa showed prompt necking following yielding

doses between 2 and 7.5 dpa, with near-zero ductility

observed at 25.2 dpa Fast neutron irradiation studies of

observed brittle failureafter 0.13 dpa in materials irradiated and tested near

600 K Less than 5% total elongation was measured

following 1.97 dpa irradiation at 700 K, despite a near

doubling of the yield stress over the unirradiated rial Limited ductility was also observed following2.63 dpa exposure in materials irradiated and tested

mate-at 1073 K, with a yield strength increase from 240 to

315 MPa over the unirradiated control While temperature embrittlement following exposure to0.13 dpa was reported in the neutron-irradiated mate-

concentra-tions on the behavior of these materials may bemore influential than the irradiation spectrum.Similar to Ta and Ta–10W, very limited data exist

on the irradiated properties of T-111 The most

ultimate tensile strengths are observed following

at 688 and 913 K The increase in radiation hardening

is substantially greater than that observed in pure

Ta irradiated under similar conditions Yield and

25.2 dpa

7.5

0 200 400 600 800 1000

Elongation (%)

4.4 0.7 2.0

ultimate tensile strengths of around 1250 MPa are

reported for irradiation at 688 K, with uniform and

Irradiation at 913 K improves uniform and total

values represent more than a 50% reduction in

ductil-ity over the unirradiated values No known irradiated

property data for T-111 exist for temperatures above

913 K As tensile strengths of both Ta–10W and T-111

and are well above the stresses that produce brittlebehavior in vanadium alloys for which more data areavailable, it is likely that these Ta alloys are embrittled

irradiated materials database including fracture ness data for Ta and Ta alloys irradiated near andabove 1000 K is much needed to ascertain the uppertemperature limitations However, based on this pre-liminary data, temperatures below1000 K may need to

tough-be avoided for Ta and Ta-base alloys

room temperature and 623 K have been performed

to evaluate the performance of T-111 and Ta–10W

low-dose irradiations produced little change in thetensile properties of the two alloys Some variations

in the total elongation were observed in T-111, whichmay be related to the distribution and make-up of theHf-rich compounds in the material as well as theeffects of radiation Thermal stability of T-111 can

be an issue, as a brittle behavior following 1100 h

precipi-tation of Hf-rich compounds along grain boundaries

It is not known how the combination of long-termthermal aging under irradiation affects the structure–property relationships or how the detrimental pre-cipitation of the interstitial elements with Hf can becontrolled

Plastic instability region Plastic instability region

Fracture region

Elastic region Elastic region

Uniform plasticity (523 K) Uniform plasticity (523 K)

Uniform plasticity (Trm)

Uniform plasticity (Trm)

0.0 0 500

Dose (dpa) (b)

1

Figure 10 Deformation mode map of (a) pure Ta and (b) Ta–1W for room temperature irradiations, illustrating fracture, plastic instability, uniform plasticity, and elastic regions as a function of stress and displacement dose The increases in the uniform plasticity region for temperatures of 523 K are superimposed Reproduced from Byun, T S.; Maloy, S A J Nucl Mater 2008, 377, 72–79.

Figure 11 Comparison of the radiation hardening of Ta

and Ta–10W irradiated at 673 K to displacement doses

of <0.39 dpa Adapted from Ullmaier, H.; Casughi,

F Nucl Instr Methods Phys Res B 1995, 101, 406–421.

4.06.4 Molybdenum and Mo-Base

Alloys

Mo Alloys

Molybdenum and its alloys are the perennial

candi-dates for refractory metal alloy use in irradiation

environments, due in part to their high melting

temperature (2896 K), good thermal properties,

high-temperature strength, and lower induced

radioactivity (as compared to tantalum) The density

lower than that of Ta and W, though greater than Nb

But like other refractory metal alloys, Mo can

pres-ent difficulties in fabrication, low-temperature

duc-tility, and low-temperature embrittlement from

radiation damage The TZM (Mo–0.5%Ti–0.1%

Zr) and Mo–Re alloys were examined as part of the

SP-100 and JIMO/Prometheus space reactor

pro-grams, respectively, and offer additional benefits of

improved high-temperature strength over the pure

examined for plasma facing and diverter components

in fusion reactor designs due to the relatively low

sputter yield, high thermal conductivity, and thermal

In addition, because of these benefits, Mo has also been

examined for use as a grazing incident metal mirror in

As in all other refractory metals, the mechanical

properties are influenced by impurity

concentra-tions, particularly through grain boundary

weaken-ing However, improvements in Mo ductility are

achievable through grain refinement, impurity

con-trol, and the addition of Re or reactive elements such

as Ti and Zr An upper limit to the acceptable level of

C was also found to improve grain boundary strength

Low-carbon arc-cast molybdenum (LCAC-Mo) is

one such example, in which oxygen impurities are

Higher levels of C will result

in reduced fracture toughness, unless additional

reac-tive alloy additions are present in the alloy The TZM

alloy also incorporates a small level of carbon to

produce Ti- and Zr-carbide strengthening

Improvements in ductility and toughness through

the ‘rhenium effect’ have been observed in Mo for

VIa metals are alloyed with elements from Group

phe-nomenon range from enhanced mechanical twinning,

reduced resistance to dislocation glide, reduction of

oxygen at grain boundaries, and increased interstitial

the initial work that had suggested a maximum

inconclusive because of inadequate control of O and

C impurity levels in the earlier studies Higher centration alloys with 40–50 wt% Re have also beenexamined for use in the radiation environments.Alloys with Re concentrations up to 41–42% are

Com-mercially available alloys include Mo–41Re andMo–47.5Re (sometimes referred to as Mo–50Re).Recently, introduction of oxide dispersion strength-ened (ODS)-Mo through the incorporation of lantha-

alloys show great resistance to recrystallization andhigh-temperature deformation while maintaining lowductile-to-brittle transition temperatures (DBTT)

The radiation effects database for Mo and itsalloys is limited to scattered scoping examinations,which show little overlap in the experimental vari-ables such as material purity, alloying level, materialthermomechanical history, irradiation conditions, andpostirradiation test conditions Where available, infor-mation on the physical and mechanical propertychanges to LCAC-Mo, TZM, Mo–Re alloys, andODS-Mo will be reviewed

Physical Property Changes in Mo andMo-Base Alloys

Two earlier reviews of the irradiation-induced erties of Mo and TZM have been presented as part of

known swelling data on irradiated Mo is contained inthese reviews, with the majority of data for irradia-

swelling data available are considerably scattered,with little coherence to examinations on the swelling

as a function of temperature or dose

Swelling in Mo is expected to begin around 573–

Maximum swelling in pure Mo remains below 4%

50 dpa, with peak swelling at irradiation tures near 900 K Attempts at consolidating thereported swelling data as a function of irradiationtemperature through normalizing the fluences proved

tempera-to be inaccurate in determining the upper bound

for irradiated Mo as a function of dose and

8
1022

Stubbins et al.88

between 1173 and 1393 K up to 50 dpa remained

below 4%, while irradiations between 1523 and

1780 K were near 10%

Void ordering has been observed in both

examined the irradiation and material conditions

that contribute to void ordering Irradiation

tempera-tures near 700 K delineate the lower boundary

tem-perature for void lattice formation at irradiations

above 20 dpa At lower doses, void lattice formation

was not observed The void superlattice constant,

mea-sured as the distance between void centers along the

<100> direction in the material, is found to increase

of 3–4% on the development of the void lattice ture, based on an attainment of an equilibrium ratio of

longer observed, leading to the high values of swelling

The onset of void growth in neutron-irradiatedmaterial appears to be accelerated in cold-workedmaterials compared to annealed materials, reaching

a maximum in swelling at doses near 40 dpa for peratures below 873 K and 20 dpa at higher tempera-

void shrinkage, with swelling values approachingthose of annealed materials Void shrinkage has also

presum-ably due to the segregation of transmuted species

at the void surfaces, making them more attractivefor interstitials

Irradiation-induced swelling in TZM has been

temper-ature dependence as the pure metal The fluence andtemperature dependence of swelling of TZM was

and Gelles et al.,94

Irradiation temperatur

e (K)

Garner and Stubbins 89 (neutron)

Stubbins et al.88 (ion) Evans 53 (neutron)

Lee et al.87 (neutron)

Brimhall et al.90 (neutron)

Brimhall et al.90 (ion)

Figure 12 Irradiation-induced swelling ( DV/V ) as a function of irradiation temperature and displacement damage (dpa) for pure Mo The irradiation source is marked in the key Reproduced from Lee, F.; Matolich, J.; Moteff, J J Nucl Mater.

1976, 62, 115–117; Evans, J H J Nucl Mater 1980, 88, 31–41; Stubbis, J F.; Moteff, J.; Taylor, A J Nucl Mater 1981, 101, 64–77; Garner, F A.; Stubbins, J F J Nucl Mater 1994, 212–215, 1298–1302; Brimhall, J L.; Simonen, E P.; Kissinger,

H E J Nucl Mater 1973, 48, 339–350.

results from the latter shown in Figure 13 Peak

data are limited to irradiation temperatures below

923 K Only limited data are available on direct

com-parisons between TZM and pure Mo, with Bentley

and TZM alloys and 0.6% swelling in pure Mo under

the same irradiation conditions Similarly, 4%

swelling was observed in TZM and 3% in pure Mo

In examining Mo and TZM of different

or greater swelling in TZM compared to Mo

n cm2(E > 0.1 MeV) at

823 and 873 K However, in the materials irradiated

at 723 K for the same fluence, the TZM alloy showed

lower swelling, except in the carburized condition

The Ti and Zr atoms not tied up as carbides are

assumed to have played a role in reducing void size

in the material at the lower temperature

There is little information on the swelling

behav-ior of Mo–Re alloys Measured swelling of 0.44% in

n cm2 (E > 0.1MeV) at temperatures which rose during irradiation

Mo–Re alloys, radiation-induced segregation (RIS)

and transmutation can lead to precipitation of

equilibrium or nonequilibrium phases, which can bedetrimental to mechanical properties This is examined

in the next section

Electrical resistivity changes to 5.4 dpa irradiated

using single crystal samples Increases in resistivity

of 10–14% and 92–110% were measured at diation test temperatures of 298 and 77 K, respec-tively The largest resistivity changes were measured

postirra-in the [100] direction A residual 10% postirra-increase postirra-inresistivity was measured following annealing above

Mo rapidly increases between 0.01 and 0.1 dpa

tempera-ture resistivity of 10–12% were reported following0.5–1.2 dpa irradiation at 543 K, and 3.3–5.3% after1.4–2.4 dpa at 878 K At irradiation temperatures

1208 K, little (<3%) to no net increase in resistivitywas observed for irradiations up to 3.3 dpa This isreflected in the higher mobility of vacancies and

20

1100 1000 900

T

irr (K) 800 700 600 0.0 0.5 1.0 1.5

40 60 80

dpa

Figure 13 The swelling dependence on temperature and fluence for neutron-irradiated TZM Adapted from Gelles, D S.; Peterson, D T.; Bates, J F J Nucl Mater 1981, 103–104, 1141–1146; Evans, J H J Nucl Mater 1980, 88, 31–41.