Comprehensive nuclear materials 1 04 effect of radiation on strength and ductility of metals and alloys

Comprehensive nuclear materials 1 04 effect of radiation on strength and ductility of metals and alloys Comprehensive nuclear materials 1 04 effect of radiation on strength and ductility of metals and alloys Comprehensive nuclear materials 1 04 effect of radiation on strength and ductility of metals and alloys Comprehensive nuclear materials 1 04 effect of radiation on strength and ductility of metals and alloys Comprehensive nuclear materials 1 04 effect of radiation on strength and ductility of metals and alloys

Metals and Alloys

M L Grossbeck

University of Tennessee, Knoxville, TN, USA

ß 2012 Elsevier Ltd All rights reserved.

A 1 Lowest equilibrium temperature at which

the austenite phase exists in steel

appm Atomic parts per million

ASTM ASTM International

ATR Advanced Test Reactor, Idaho Falls, ID,

USA

bcc Body-centered cubic

BR2 Belgian Reactor-2, Mol, Belgium

DBTT Ductile-brittle transition temperature

dpa Displacements per atom

EBR-II Experimental Breeder Reactor-II, Idaho

Falls, ID, USA

fcc Face-centered cubic

FFTF Fast Flux Test Facility, Richland, WA,

USA

HFBR High Flux Beam Reactor, Brookhaven,

Upton, NY, USA

HFIR High Flux Isotope Reactor, Oak Ridge,

TN, USA

HFR High Flux Reactor, Petten, The

Netherlands

JPCA Japanese Prime Candidate Alloy

LMFBR Liquid Metal Fast Breeder Reactor LWR Light Water Reactor

ORR Oak Ridge Research Reactor, Oak Ridge,

TN, USA PCA Prime candidate alloy, adopted by the US

Fusion Program in mid-1970s ppm Parts per million

Unirr Unirradiated

The most commonly considered mechanical ties of metals and alloys include strength, ductility,fatigue, fatigue crack growth, thermal and irradiationcreep, and fracture toughness All these properties areimportant in the design of a structure that is to experi-ence an irradiation environment While determiningthe mechanical properties of irradiated materials, ten-sile properties, typically yield strength, ultimate tensilestrength, uniform elongation, total elongation, andreduction of area are the most commonly consideredbecause they are usually the simplest and the leastcostly to measure In addition, the tensile propertiescan be used as an indicator of the other mechanical

proper-99

properties Space in a reactor or in an accelerator target

is often so limited that the larger specimens required

for fatigue and fracture toughness testing are not

prac-tical; consequently, the number of specimens that can

be irradiated is so small that a meaningful test matrix

is not possible Shear punch testing of 3-mm

diame-ter disks, typically used as transmission microscopy

specimens, was developed to address the problem of

irradiation space Although much information can be

obtained from shear punch testing, the tensile test

remains the most reliable indicator of strength and

ductility For these reasons, the tensile test is usually

the first mechanical test used in determining the

irradiated properties of new materials This chapter

addresses the tensile strength and ductility of alloys

Hardening

Irradiation introduces obstacles to dislocation motion,

which results in plastic deformation, in the form of

defects resulting from atomic displacement and from

transmutation products Small Frank loops and defect

clusters, known as black dots, large Frank loops (about

an order of magnitude larger), precipitates, and cavities

(either voids or bubbles) contribute to hardening in an

irradiated alloy Frank loops unfault and eventually

contribute to the network dislocation density

Precipi-tates are certainly present in the unirradiated alloy,

but additional precipitation results from the

segrega-tion of elements during irradiasegrega-tion and from the

irradiation-induced changes that shift the

thermody-namic stability of phases Transmutation production of

new elements in the alloy can also result in the

forma-tion of new precipitates The producforma-tion of insoluble

species, most importantly helium, also results in

pre-cipitation, especially in the form of bubbles

Defects are divided into two classes: long range

and short range Short-range obstacles are defined as

those that influence moving dislocations only on the

same slip plane as opposed to long-range obstacles,

which impede dislocation motion on slip planes not

containing the obstacle.1 Coherent precipitates and

large loops are long-range obstacles, but for this

analysis, only network dislocations will be considered

as long-range obstacles, a reasonable simplification

from observations As recommended by Bement,2the

contributions from short-range obstacles are added

directly,

DFTS¼ DFLRþ DFSR ½1

where the quantities ineqn [1]are total stress, range contribution to stress, and short-range contribu-tion to stress The contributions from the short-rangeobstacles are added in quadrature as follows3:ðDFSRÞ2¼ ðDFSMloopÞ2þ ðDFLGLoopÞ2

long-þ ðDFPRECIPÞ2þ ðDFCAVITYÞ2 ½2where the term on the left represents the contributionfrom all short-range obstacles, and the terms on theright represent the stress contributions from smallloops, large loops, precipitates, and cavities, eithervoids or bubbles

The contribution to hardening by network cations may be expressed by

dislo-tnet¼ aGbpffiffiffiffiffird ½3where tnetis the increment in shear stress, G is theshear modulus, b is the Burgers vector, and rd isthe dislocation density The constanta is dependentupon the geometry of the dislocation configurationand is usually determined experimentally However,Taylor has calculateda to be between 0.15 and 0.3,4and Seeger has determined the value to be 0.2, incor-porating the assumption of a random distribution ofdislocation directions.5 Short-range defects such assmall and large Frank loops and precipitates aretreated as hard impenetrable obstacles where disloca-tions bow around them by the Orowan mechanism.The stress increment is expressed by

Dt ¼ GbpffiffiffiffiffiffiffiffiffiffiffiNd =b

½4where N is the defect density and d is the diameter.The constantb ranges between 2 and 4 as suggested

by Bement2or 6 as suggested by Olander.6Voids andbubbles are also treated as hard obstacles using thesame expression Precipitates and bubbles have beenobserved in austenitic stainless steels to nucleate andgrow together.7In this case, the bubbles and precipi-tates are considered as one obstacle where the hard-ening increment is calculated assuming rod geometryusing a treatment by Kelly expressed by8:

Bubble-precip¼0:16Gb

ffiffiffiffiffiffiNdp

1pffiffi6 3

ffiffiffiffiffiffiNd

ffiffiffi6

pd3b

½5

where the parameters are the same as foreqn [4].From the previous discussion, it can be inferred thatbecause the nature of the irradiation-induced defectsdetermines the degree of hardening, and because thenature, size, and density of defects is a strong function oftemperature, radiation strengthening will be a strongfunction of irradiation temperature.Figure 1illustrates

strengthening from individual types of defects as a

function of irradiation temperature for the austenitic

stainless steel PCA.7

As can be seen from Figure 1, the black dot

damage characteristic of low temperatures vanishes

at temperatures over 300C as Frank loops emerge

Bubbles and precipitates also become major

contri-butors to hardening above 200C

Tensile behavior is determined by the

irradiation-induced defect structure previously discussed

Aus-tenitic stainless steels will again be used for the

example since they are typical of fcc alloys and in

many respects to other alloys (see Chapter 2.09,Properties of Austenitic Steels for Nuclear Reac-tor Applications and Chapter 4.02, RadiationDamage in Austenitic Steels) The behavior ofother example classes of alloys will be discussed inlater sections of this chapter The tensile behaviorcharacteristic of austenitic stainless steels is shown in

Figure 2, where yield strength is plotted as a function

of fluence and displacement level.9 Saturation instrength is clear with the saturation time becomingshorter as irradiation temperature is increased Attemperatures above about 500C, saturation is evi-dent, but in this case, strength decreases

This decrease is a result of recovery of the worked microstructure of the 20% cold-worked type

cold-316 stainless steel presented in Figure 2.Figure 3

0 0.2 0.4 0.6 0.8 1

Black dots Frank loops Bubbles-precipitates Network dislocation

Figure 1 Relative contribution to strengthening from irradiation-induced defects in the austenitic stainless steel, PCA, irradiated to 7 dpa in the Oak Ridge Research Reactor Reproduced from Grossbeck, M L.; Maziasz, P J.; Rowcliffe, A F.

J Nucl Mater 1992, 191–194, 808.

1100 1000 900 800 700 600 500 400 300 200 100

371 427 483 593 649 704 760 816

371⬚C 427⬚C

483⬚C

538⬚C 593⬚C

Neutron fluence (n cm −2) (E > 0.1 MeV)

649 ⬚C 704⬚C

760⬚C 816⬚C

100

120 Test

Irrad temp.» Test Temp.

Strain rate ~ 4 ´ 10 -5S-1

Symbol Temp ( ⬚C)

140 160

Figure 2 Yield strength of 20% cold-worked type 316 stainless steel irradiated in the EBR-II Reproduced from

Fish, R L.; Cannon, N S.; Wire, G L In Effects of Radiation on Structural Materials; Sprague, J A., Dramer, K., Eds.; ASTM: Philadelphia, PA, 1979; ASTM STP 683, p 450 Reprinted, with permission, from Effects of Radiation on Structural Materials, copyright ASTM International, West Conshohocken, PA.

shows yield strength resulting from the recovery of a

cold-worked dislocation structure and the generation

of a radiation-induced microstructure, resulting in a

saturation strength independent of the initial

condi-tion of the alloy.10Again, it is seen that the approach

to saturation is faster with increasing temperature,

with saturation achieved between 5 and 10 dpa at 538

and 650C, but 15–20 dpa is necessary to achieve

saturation at 427C

Saturation is observed in yield strength curves for

fluences as high as 9 1022

n cm2in a fast reactor(45 dpa), but more recent data show a hint of soft-

ening above 50 dpa,11,12 and other fast reactor data

have shown a reduction in strength even for

dis-placements below 50 dpa, as shown in Figure 4.13

This could result from coarsening of the

microstruc-ture or depletion of interstitial elements from the

matrix due to precipitation This effect is also

observed in martensitic steels irradiated to high dpa

levels in the FFTF, but in this class of alloys, recovery

of the martensitic lath structure is also a factor.12

However, even in austenitic steels, it is difficult

to attribute such softening with certainty to an

irradiation effect because of the strong influence

of irradiation temperature on strength.14 Indeed,uncertainties in irradiation temperature are an inher-ent difficulty in neutron irradiation experiments

This discussion has used neutron irradiations forillustration purposes Reactors provide an effectiveinstrument for achieving high neutron exposuresunder conditions relevant for most nuclear applica-tions However, reactor irradiations suffer from manydifficult-to-control and, sometimes, uncontrolledvariables The neutron energy spectrum is responsi-ble for large differences in irradiation effects betweendifferent reactors

The mechanism of atomic displacement is wellunderstood.15 With a known neutron energy spec-trum, neutron atomic displacements can be calcu-lated as a function of fluence for a given reactor.Transmutation of elements in the material understudy, which is a strong function of neutron spectrum,results in wide variation in some mechanical proper-ties This is of particular importance in applyingfission reactor results to fusion In a fusion device,helium and hydrogen will be generated through (n,a)and (n,p) reactions in nearly all common structuralmaterials Hydrogen has a very high diffusivity inmetals so that an equilibrium concentration will be

Figure 3 Yield strength of type 316 stainless steel

irradiated in the EBR-II Reproduced from Garner, F A.;

Hamilton, M L.; Panayotou, N F.; Johnson, G D J Nucl.

Mater 1981, 103 & 104, 803.

800

700 650 600 550 500 450

750

Figure 4 Strength properties of 20% cold-worked type

316 stainless steel irradiated in EBR-II Reproduced from Allen, T R.; Tsai, H.; Cole, J I.; Ohta, J.; Dohi, K.;

Kusanagi, H Effects of Radiation on Materials; ASTM: Philadelphia, PA, 2004; ASTM STP 1447, p 3 Reprinted, with permission, from Effects of Radiation on Structural Materials, copyright ASTM International, West Conshohocken, PA.

established at a level that is believed to be benign.16

By contrast, helium is insoluble in metals, segregating

at grain boundaries and other internal surfaces and

discontinuities

Although helium is produced in all nuclear

reac-tors, the thermal spectrum is responsible for the

highest concentrations The largest contributors to

helium in a thermal reactor are boron and nickel by

the following reactions:

ing elements but only at ppm levels Nickel is a

major constituent of many alloys and a minor

con-stituent of still others The two nickel reactions

constitute a two-step generation process for

helium, which starts slowly and accelerates as

59

Ni builds up in the alloy, limited only by the

supply of 58Ni, which for practical purposes is

often unlimited In austenitic alloys, the high flux

isotope reactor (HFIR) has generated over

4000 appm He in austenitic stainless steels The

generation rate is so high that multistep absorber

experiments have been conducted to reduce the

helium generation rate to that characteristic of

fusion reactors, 12 appm He per dpa in austenitic

stainless steels.17 (see Chapter 1.06, The Effects

of Helium in Irradiated Structural Alloys)

Other transmutation products may also

compli-cate reactor irradiation studies Examples are the

transmutation of manganese to iron by the

V The first reaction leads to loss of an

alloy constituent, and the second leads to doping

with an extraneous element However, neither of

these reactions has been shown to significantly affect

mechanical properties of steels.18

Helium remains the most studied transmutation

product, and it can have profound effects on tensile

properties, especially at high temperatures

Experi-ments have been conducted in various reactors

throughout the world to assess the effects of helium

on mechanical properties of alloys.19 An interesting

result is that helium has little effect on strength

This is illustrated in Figure 5where a comparison

has been made between austenitic steels irradiated

in Rapsodie, a fast spectrum reactor, and steels

irradiated in HFIR, a mixed-spectrum reactor with

a very high thermal flux The saturation yieldstrength of all alloys remains within a single scatterband.20,21

The tramp impurity elements sulfur and phorus have significantly high (n,a) cross sections athigh energies, as shown in Figure 6 Although thecross section for phosphorus is large only at energiescharacteristic of fusion, a boiling water reactor pro-duces 500 appm He from sulfur and 40 appm He fromphosphorus in eight years of operation An LiquidMetal Fast Breeder Reactor (LMFBR) can produce

phos-100 times these concentrations All these elementsare expected to enhance embrittlement when seg-regated to grain boundaries, but it remains to bedetermined which is more detrimental, helium, sul-fur, or phosphorus

Tensile ductility is a more vulnerable parameter thanstrength to radiation effects since it tends to be veryhigh in unirradiated austenitic stainless steels and isoften reduced to quite low levels by irradiation It isalso of more concern since strengthening, althoughnot reliable due to its slow initiation, is usually abeneficial change In contrast, embrittlement isalways detrimental Like strength, ductility exhibitssaturation with increasing fluence, although thebehavior is significantly more complex than that ofstrength The general trends in type 316 stainlesssteel are shown in Figure 7 for material irradiated

in the EBR-II These data are for the same specimensfor which the yield strength was shown inFigure 2.9Fast reactor data are used here to avoid the compli-cation of helium effects Once stabilization of thedislocation microstructure is achieved, a smoothcurve approaching an apparent saturation is observed.More information can be gleaned from ductilitydata if they are viewed in terms of irradiation andtest temperature.Figure 822shows total tensile elon-gation for a series of irradiated austenitic alloys at

a displacement level of 30 dpa in both annealedand cold-worked conditions The room temperatureductility exceeds 10%, but it decreases rapidly withincreasing temperature up to approximately 300Cand then exhibits the expected increase with tem-perature observed for unirradiated alloys Beyond

500C, ductility again decreases with an onset ofintergranular embrittlement resulting from heliumintroduced through transmutations in the thermalflux of the HFIR

US PCA SA + 800 ⬚C, 8 h HFIR

316 20% cw EBR-II JPCA 15% cw HFIR JPCA SA HFIR 400

Typical yield strength values of unirradiated solution annealed austenitic stainless steel

Figure 5 Saturation yield strength as a function of temperature for austenitic alloys irradiated in Rapsodie, EBR-II, and high flux isotope reactor showing similar saturation strength Reproduced from Grossbeck, M L.; Ehrlich, K.; Wassilew, C.

J Nucl Mater 1990, 174, 264.

Uniform elongation, the elongation at the onset of

plastic instability, or necking, appears to be most

sensi-tive to the effects of irradiation and, in general, is

less dependent on specimen geometry than other

para-meters such as total tensile elongation The low values

of uniform elongation are often cause for great

con-cern, which is usually justified However, it should be

borne in mind that if stresses remain below the yield

stress of a metal, elongation becomes a secondary cern As long as limited plastic deformation relieves thestress that produced it, a structure remains intact.The high level of irradiation strengtheningobserved at temperatures below 300C, which isdue to black dot defect clusters and small loops,also results in low ductility throughout this tempera-ture range Small helium bubbles and helium-defect

con-17 16 15 14 13 12 11 10 9 8 7 6 5 4 3 2 1 0

Test Symbol Temp ( ⬚C)

371 ⬚C

538 ⬚C 593⬚C

649 ⬚C

704 ⬚C

760 ⬚C

816 ⬚C

Neutron fluence (n cm −2) (E > 0.1 MeV)

Figure 7 Total elongation of 20% cold-worked type 316 stainless steel irradiated in EBR-II Reproduced from

Fish, R L.; Cannon, N S.; Wire, G L In Effects of Radiation on Structural Materials; Sprague, J A., Dramer, K., Eds.; ASTM: Philadelphia, PA, 1979; ASTM STP 683, p 450 Reprinted, with permission, from Effects of Radiation on Structural Materials, copyright ASTM International, West Conshohocken, PA.

J

*

* C

Figure 8 Total elongation as a function of irradiation and test temperature for fast (EBR-II) and mixed-spectrum

(high flux isotope reactor) reactor irradiation.

clusters also contribute to hardening and reduction

in ductility, but this form of helium embrittlement

is not related to the severe intergranular

embrit-tlement that is observed above 500C Both these

effects are apparent in Figure 9 where uniform

elongation for an extensive set of austenitic alloys

irradiated in thermal and fast spectrum reactors is

shown.11 The specimens irradiated in the fast

spec-trum (<5 appm He) exhibit consistently higher

duc-tility than the mixed-spectrum reactor specimens

(500–1000 appm He) even at this low displacement

level, especially above 600C, where helium

embrit-tlement is certain to control

A similar pattern is exhibited at 30 dpa where a

very limited uniform elongation characteristic of

lower temperatures is apparent After a restoration

of ductility above 400C, ductility again decreases

above 500C due to the onset of intergranular

helium embrittlement Differences in alloy behavior,

especially in the case of titanium-modified alloys

somewhat clouds the understanding of helium

embrittlement observed in Figure 10.11 However, at

50 dpa, where helium levels exceed 4000 appm, thetrend becomes clear with the fast reactor specimensshowing uniform elongations several times largerthan those observed in mixed-spectrum reactors(Figure 11).11 What is less expected is the recovery

of ductility at 50C at 50 dpa compared to theresults at 30 dpa This irradiation annealing effecthas also been observed at 230C by Ehrlich, wherestrength of the alloy 1.4988 decreased continuouslyfrom 10 to 30 dpa.20 Results from an experiment inthe Oak Ridge Research Reactor (ORR), where thespectrum was tailored to produce a ratio of He perdpa characteristic of a fusion reactor, show similarlow levels of uniform elongation for cold-workedalloys at low temperatures, but high uniform elonga-tions were observed in annealed type 316 stainlesssteel at 60C This high ductility was drasticallyreduced between 200 and 330C before the micro-structure characteristic of higher temperatures becameeffective.21

700 Symbols

316 20% CW EBR-II 316 ANN EBR-II

US PCA 25% CW ORR

US PCA 25% CW HFR

316 20% CW DO HFIR

Figure 9 Uniform elongation as a function of irradiation and test temperature at a displacement level of 10 dpa.

The trend curves are for type 316 stainless steel and PCA Reproduced from Grossbeck, M L.; Ehrlich, K.; Wassilew, C.

J Nucl Mater 1990, 174, 264.

1.04.6 Effect of Test Temperature

An interesting phenomenon is observed when

irra-diated alloys are tested at temperatures different

from the irradiation temperature Figure 12 shows

total elongation data from cold-worked type 316

stainless steel irradiated to displacement levels of

48–63 dpa in the FFTF, where elongation is plotted

against the increment of the test temperature abovethe irradiation temperature.23 Although there issignificant scatter in the data, the elongations below1% obtained by test temperatures about 100Cabove the irradiation temperature are cause forconcern This phenomenon has also been observed

in higher nickel alloys The cause of this non remains elusive, pending further testing with

US PCA 25% CW ORR

US PCA 25% CW HFR

316 20% CW DO HFIR

Figure 10 Uniform elongation of austenitic stainless steels irradiated in fast and thermal reactors to a displacement level

of 30 dpa Severe helium embrittlement is shown at 600C Reproduced from Grossbeck, M L.; Ehrlich, K.; Wassilew, C.

78 dpa

J 316 20% CW HFIR

JPCA SA HFIR JPCA 15% CW HFIR

316 20% CW EBR-II

Figure 11 Uniform elongation of austenitic stainless steels irradiated to 50 dpa in high flux isotope reactor (HFIR) and

78 dpa in EBR-II showing embrittlement from helium generated in the mixed-spectrum reactor, HFIR Reproduced from Grossbeck, M L.; Ehrlich, K.; Wassilew, C J Nucl Mater 1990, 174, 264.

different holding times at various temperatures

before tensile testing Migration of interstitial solutes

to moving dislocations is a candidate mechanism for

this phenomenon

1.04.7.1 Introduction

The class of ferritic–martensitic alloys with

chro-mium concentrations in the range of 9–12% has

attracted interest in the fast reactor programs because

of its radiation resistance, in particular, very low

swelling and low irradiation creep Alloys such as

Sandvik HT-9 (12Cr1Mo.6Mn.1Si.5W.3V)) and

other alloys of this class were irradiated in the

EBR-II,24 in research reactors25 and with heavy

ions.26The quantitative results from the ion

irradia-tions in this class of alloys and the low neutron

absorption cross section led to inclusion of ferritic

alloys into the fast reactor alloy development

pro-grams, in particular in the United States in the

mid-1970s The radiation resistance has been

con-firmed to displacement levels of 70 dpa.12,14

Further interest in this class of alloys was initiated

by the fusion reactor programs in Europe and the

United States when the necessity for low neutron

activation structural materials was realized Further

research on martensitic alloys by fusion programs inEurope, the United States, and Japan led to the devel-opment of low-activation alloys by replacing elementsthat result in long-term activation products Molyb-denum and niobium, both of which result in long-lived activation products, were replaced by tungstenand tantalum This research led to radiation-resistantalloys with a fracture toughness superior to that of thecommercial alloys even in the unirradiated condi-tion.27The compositions of representative members

of this class of alloys referred to in this chapter arepresented inTable 1 An excellent review of irradia-tion behavior of this class of alloys has been published

by Klueh and Harries.27 Details of the metallurgy ofmartensitic alloys appears inChapter4.03, FerriticSteels and Advanced Ferritic–Martensitic Steels.1.04.7.2 Tensile Behavior

Unlike the tensile behavior of fcc metals, where there

is a smooth increase in strength as plastic tion proceeds and work hardening progresses, bccmetals typically exhibit a load drop almost immedi-ately following the onset of plastic deformation.Interstitial solutes such as carbon in steels effectivelylock dislocations leading to a longer period of elasticdeformation after which generation of new disloca-tions results in a load drop, or yield point, until

deforma-32 28 24 20 16 12 8 4 0

terminated by the work hardening mechanism of

dislocation interaction.28

Upon irradiation, the load drop is frequently

masked by an early termination of work hardening,

leading to very low values of uniform elongation

This behavior is evident even at displacement levels

below 0.01 dpa and is illustrated inFigure 13 fromresearch presented at the Second Atoms for PeaceConference in 1958.29 Extreme irradiation hardeningand severe plastic instability are clearly illustrated bythis early research More recent alloys with morecareful control of impurities and controlled processing

Table 1 Nominal or typical compositions of ferritic–martensitic alloys cited

Steel type Designation Composition (wt%)

have led to ferritic alloys with appreciable work

hard-ening even at high displacement levels Figure 14

shows tensile curves for the 9Cr–2Mo–1Ni steel,

JFMS, neutron irradiated and tested at room

temper-ature Uniform elongations of several percent are

evident, a reasonable value for irradiated steels.12

A plot of yield stress as a function of displacement

damage level is shown inFigure 15for low-activation

ferritic alloys irradiated in fast reactors,30 and plots

of yield stress and total elongation are shown in

Figure 16for fast and mixed-spectrum reactors.31Unlike the austenitic alloys, the martensitic alloysrapidly reach a peak in strength then soften withfurther irradiation followed by near saturation instrength beginning at about 30 dpa Total elongationfollows a corresponding pattern, demonstrating

1400 1200 1000 800 600 400 200

Figure 14 Stress–strain curves for JFMS alloy irradiated in FFTF to 44 dpa at temperatures of 373–427C and tested at

25C Reproduced from Maloy, S A.; Toloczko, M B.; McClellan, K J.; et al J Nucl Mater 2006, 356, 62.

800

700

600

500 400

300

200 100

70

Figure 15 Yield stress as a function of displacement level for martensitic alloys irradiated in FFTF or EBR-II.

Reproduced from Kohno, Y.; Kohyama, A.; Hirose, T.; Hamilton, M L.; Narui, M J Nucl Mater 1999, 271 & 272, 145.