Development of microstructure and irradiation hardening of Zircaloy during low dose neutron irradiation at nominally 358 °C

Journal of Nuclear Materials 418 (2011) 46–61

Contents lists available at SciVerse ScienceDirect

Journal of Nuclear Materials

journal homepage: www.elsevier .com/locate / jnucmat

Development of microstructure and irradiation hardening of Zircaloyduring low dose neutron irradiation at nominally 358 �C

B.V. Cockeram a,⇑, R.W. Smith a, K.J. Leonard b, T.S. Byun b, L.L. Snead b

a Bettis Laboratory, Bechtel Marine Propulsion Corporation, West Mifflin, PA 15122-0079, USAb Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6138, USA

a r t i c l e i n f o

Article history:Received 26 March 2011Accepted 4 July 2011Available online 22 July 2011

0022-3115/$ - see front matter � 2011 Elsevier B.V. Adoi:10.1016/j.jnucmat.2011.07.006

⇑ Corresponding author. Tel.: +1 412 476 5647; faxE-mail address: [email protected] (B.V. Cock

a b s t r a c t

Wrought Zircaloy-2 and Zircaloy-4 were neutron irradiated at nominally 358 �C in the high flux isotopereactor (HFIR) at relatively low neutron fluences between 5.8 � 1022 and 2.9 � 1025 n/m2 (E > 1 MeV). Theirradiation hardening and change in microstructure were characterized following irradiation using ten-sile testing and examinations of microstructure using Analytical Electron Microscopy (AEM). Small incre-ments of dose (0.0058, 0.11, 0.55, 1.08, and 2.93 � 1025 n/m2) were used in the range where thesaturation of irradiation hardening is typically observed so that the role of microstructure evolutionand hai loop formation on irradiation hardening could be correlated. An incubation dose between5.8 � 1023 and 1.1 � 1024 n/m2 was needed for loop nucleation to occur that resulted in irradiation hard-ening. Increases in yield strength were consistent with previous results in this temperature regime, andas expected less irradiation hardening and lower hai loop number density values than those generallyreported in literature for irradiations at 260–326 �C were observed. Unlike previous lower temperaturedata, there is evidence in this study that the irradiation hardening can decrease with dose over certainranges of fluence. Irradiation induced voids were observed in very low numbers in the Zircaloy-2 mate-rials at the highest fluence.

� 2011 Elsevier B.V. All rights reserved.

1. Introduction

The microstructure of Zircaloy-2 and Zircaloy-4 consistsprimarily of a hexagonal Zr-phase containing Sn, Fe, Cr, Ni forZircaloy-2, or Sn, Fe, and Cr for Zircaloy-4 [1–6]. Zircaloy-2 and Zir-caloy-4 both have Sn in solid solution and negligible solubility forFe, Cr, and Ni in the hexagonal phase. Both Zircaloy-2 and Zircaloy-4 also contain fine intermetallic precipitates nominally 0.4 lm indiameter that are primarily Laves (hcp Zr(Fe,Cr)2 and cubicZr(Fe,Cr)2) and Zintl (bct Zr2(Fe,Ni)) phases in the case of Zirca-loy-2 and laves phases in Zircaloy-4. During irradiation Zircaloyis known to exhibit changes in microstructure, increases instrength, and decreases in uniform elongation as functions of irra-diation temperature and fluence [7–14]. Plasticity in unirradiatedZircaloy is generally known to result from slip on the prism planes(f10 �10g), but slip has been observed on the basal ({0 0 0 1}) andpyramidal planes (f10 �11g or f10 �12g) at high temperatures andin irradiated material [1,7]. Due to its hexagonal crystal structure,irradiation of Zircaloy results in an anisotropic distribution of loopswith a high number density of hai loops formed on the prismplanes (f10 �10g) and vacancy hci loops formed on the basal planes({0 0 0 1}) starting at fluences greater than 3 � 1025 n/m2

ll rights reserved.

: +1 412 476 5779.eram).

(E > 1 MeV). One difference between Zircaloy and other metals isthat the hai loops have been observed to exist as both vacancy orinterstitial type [1,7,11–14], while loops in other metals are gener-ally of a single type for any given orientation. The increase in thenumber density of hai loops generally reaches a steady-state levelat a neutron fluence of about 1 � 1025 n/m2 (E > 1 MeV) that alsocorresponds to the fluence where the largest increases in tensilestrength are observed [1]. This suggests hai loops have the domi-nant contribution to the irradiation hardening. Because basal slipis active in irradiated Zircaloy, hci loop nucleation may also havean effect on slip. Additional processes of microstructural evolutionduring irradiation to higher fluences such as the change in precip-itate structure to an amorphous state, dissolution of alloying ele-ments from the precipitates into the matrix, and reconfigurationof the hai loops may also contribute to the hardening.

Some attempts have been made to quantify the size and num-ber densities of hai and hci loops in the literature [1,7,11–14]. Morerecent efforts to characterize the microstructure of irradiatedZircaloy have been primarily focused on the change in the struc-ture of the precipitates and dissolution of precipitates to betterunderstand the influence of irradiation on corrosion resistance[3–6,15–18]. Although some data has been collected over a varietyof dose levels and irradiation temperatures, variations in materialand irradiation conditions between different investigations com-plicates the task of accurately assessing the effects of dose andirradiation temperature independently. The work reported herein

http://dx.doi.org/10.1016/j.jnucmat.2011.07.006

mailto:[email protected]

http://dx.doi.org/10.1016/j.jnucmat.2011.07.006

http://www.sciencedirect.com/science/journal/00223115

http://www.elsevier.com/locate/jnucmat

B.V. Cockeram et al. / Journal of Nuclear Materials 418 (2011) 46–61 47

is a concerted effort to quantify the hai and hci loop size and num-ber density and the resulting irradiation hardening for the sameZircaloy-2 and Zircaloy-4 heats over a range of low fluences priorto hai saturation where the changes in microstructure are the mostsignificant.

2. Materials and experimental procedure

The nominal compositions for the Zircaloy-4 and Zircaloy-2used in both the alpha-annealed and beta-treated conditions aregiven in Table 1. Alpha-annealed consists of an equiaxed grainstructure and has texture in the direction of wrought processing[2,19]. Beta-treated has the typical b to a basketweave Wid-manstätten or lath type microstructure within prior beta grainboundaries that is macroscopically random [2,19]. Tensile speci-mens (SSJ type) were machined in the transverse orientation(16 mm long, 0.25 mm thick, and 4 mm wide with a gauge sectionnominally 5 mm long and 1.2 mm thick) where the length of thetensile specimen was perpendicular to the rolling direction. Aftermachining, the tensile specimens were laser engraved and thenpickled at room temperature using a solution of nitric acid (30–39 vol.%), hydro fluoric (HF) acid (0.5–3 vol.%) and water to removeabout 25–100 lm total thickness.

The irradiations were performed at nominally 358 �C to fluencesbetween 0.058 and 29.3 � 1024 n/m2 (E > 1 MeV) in the High FluxIsotope Reactor (HFIR) in the hydraulic tube. The capsules eitherwere translated through the reactor for low dose exposures orplaced in the fixed peripheral target tube position (PTP) using theconditions given in Table 2. Passive SiC temperature monitors wereused to determine the specimen temperatures. The zirconium dpalevels were determined based on a previously reported conversion[8].

Tensile testing was performed at room-temperature at an actu-ator displacement rate of 0.508 mm/min (strain rate = 0.1 min�1)in accordance with ASTM E8 procedures [20]. Engineering stress/strain values were determined from the load and crosshead dis-placement record with no correction made for the compliance of

Table 1Nominal chemical composition of wrought Zircaloy-4 used for both alpha-annealed and b

Element Zr Sn Fe

Zircaloy-4 Bal 1.49 0.21Alpha-annealed Zircaloy-2 Bal 1.57 0.14Beta-treated Zircaloy-2 Bal 1.59 0.14

Table 2Summary of the irradiation conditions for the Zircaloy in-cascade HFIR irradiations.

Irradiationcapsule

Design specimenirradiation temperature(�C)

True irradiation temperature fortemperature monitor analysis (�C)

SCI-1 343 358 ± 20SCI-2SCI-3SCI-4SCI-5

Notes: 1. The target irradiation temperature was the calculated tensile specimen temperwithin ±20 �C for irradiations at 343 �C. 2. Capsules SCI-1, SCI-2, and SCI-3 were irradiatedflux of 3.2 � 1018 n/m2 s (E > 1 MeV). These capsules were irradiated for portions of Cycles2, and SCI-3, respectively. The capsules SCI-4 and SCI-5 were irradiated mainly in the fixeof irradiation was performed in the hydraulic tube (HT-2) at the slightly higher flux of 3.SCI-5, the MW days and hours of operation for each capsule and cycle are as follows: (1) c(338.2 MW days and 95.5 h), (3) capsule SCI-3 in cycled 415 and 416 (1690.9 MW days1015.3 h), and (5) capsule SCI-5 in cycles 414, 415, 416, 417, and 418 (9939.7 MW daysanalysis of the temperature monitors.

the load train for the stress–strain curves. Discs nominally 3 mmin diameter were punched from the tab of the tested tensile spec-imens for Transmission Electron Microscopy (TEM) of the irradi-ated structure. The TEM discs were mechanically thinned andthen electropolished using a 10 vol.% HClO4 in ethanol solution at5 �C @ less than 100 mA. A CM200-FEG instrument was used forTEM analysis of the loop structures using two-beam and weakbeam imaging conditions at a range of magnifications. EDS analysisof precipitates was also performed in STEM mode to characterizethe change in precipitate structure during irradiation. The thick-ness of each location was determined using the Kossel-Möllenstedtfringe spacing technique in the zero order Laue zone convergentbeam electron diffraction pattern for a given g-vector imagedunder a two-beam condition.

3. Results and discussion: post-irradiation tensile testing

The yield strength, change in yield strength, total elongation,and uniform elongation values determined from post-irradiatedtensile testing are provided in Fig. 1. These values were extractedfrom the stress–strain curves shown in Fig. 2.

3.1. Yield strength and irradiation hardening

The values for yield strength and change in yield stress (DrYS)used as a measure of irradiation hardening for alpha-annealedand beta-treated Zircaloy-4 and Zircaloy-2 all exhibit an increasewith increasing dose at low fluences (0.058–1.1 � 1024 n/m2) tovalues that are generally consistent with literature data, seeFig. 1a. Irradiation hardening may also be determined using Ulti-mate Tensile Strength (UTS) values, but hardening is determinedin terms of yield strength in this work so that comparisons canbe made with literature data and models. Calculation of irradiationhardening using UTS results in values that are not consistent withthose determined using the yield strength values in this work. Thisis because yielding with low uniform elongation was not observedfor many of the post-irradiated tensile results. The yield strength

eta-treated, alpha-annealed Zircaloy-2, and beta-treated Zircaloy-2 (weight%).

Cr Ni Hf C O

0.11 <.0035 <0.008 0.0095 0.140.11 0.06 0.004 0.0070 0.150.11 0.06 0.005 0.0034 0.11

HFIR cyclenumber

Neutron flux (�1018 n/m2 s, E P 1 MeV)

Neutron fluence(�1024 n/m2,E P 1 MeV]

Zirconium(dpa)

415 3.2 0.058 0.013415 3.2 1.10 0.25415 and 416 3.2 5.5 1.27415, 416, 417 2.9–3.2 10.6 2.45414, 415,416, 417, 418

2.9–3.2 29.3 6.76

ature objective for the irradiation test. The irradiation temperatures were generallyin the hydraulic tube only (HT-2) at the same position to give the constant nominal415 and 416 for nominal times of 5.03 h, 95.5 h, and 477.4 h, for capsules SCI-1, SCI-d target positions in TRRH-1 at a nominal flux of 2.9 � 1018 n/m2 s, but some period2 � 1018 n/m2 s. Taking the nominal flux of 2.9 � 1018 n/m2 s for capsules SCI-4 andapsule SCI-1 in cycle 415 (17.83 MW days and 5.03 h), (2) capsule SCI-2 in cycle 415and 477.4 h), (4) capsule SCI-4 in cycles 415, 416, and 417 (3595.9 MW days andand 2806.5 h). 3. Actual specimen irradiation temperatures were determined from

Yiel

d St

reng

th [M

Pa]

300

350

400

450

500

550

600

650

700

750

800

alpha Zircaloy-2beta Zircaloy-2alpha Zircaloy-4beta Zircaloy-4Zircaloy Literature (260C to 326C)Zircaloy Literature (326C to 450C)

(a)

Fluence [X 1024 n/m2, E > 1 MeV] Fluence [X 1024 n/m2, E > 1 MeV]

Fluence [X 1024 n/m2, E > 1 MeV] Fluence [X 1024 n/m2, E > 1 MeV]

(b)

YS, C

hang

e in

Yie

ld S

tren

gth

[MPa

]

0

50

100

150

200

250

300

350

400

450

alpha Zircaloy-2beta Zircaloy-2alpha Zircaloy-4beta Zircaloy-4Ziraloy Literature (260C to 326C)Zircaloy Literature (326C to 450C)

(c)

Tota

l Elo

ngat

ion

[%]

0

5

10

15

20

25

30

alpha Zircaloy-2beta Zircaloy-2alpha Zircaloy-4beta Zircaloy-4Zircaloy Literature (260C to 326C)Zircaloy Literature (326C to 450C)

(d)

0 5 10 15 20 25 30 0 5 10 15 20 25 30

0 5 10 15 20 25 30 0 5 10 15 20 25 30

Uni

form

Elo

ngat

ion

[%]

0

1

2

3

4

5

6

7

8

9

10alpha Zircaloy-2beta Zircaloy-2alpha Zircaloy-4beta Zircaloy-4Zircaloy Literature (260C to 326C)Zircaloy Literature (326C to 450C)

Fig. 1. Comparison of post-irradiated tensile data for Zircaloy irradiated at nominally 358 �C in compared with literature results for Zircaloy for irradiations at 260–326 �C[1,2,21–29] and 326–450 �C [30,31]: (a) 0.2% yield strength, (b) irradiation hardening data (DrYS = rYS(irad) � rYS(non-irad)), (c) total elongation, and (d) uniform elongation.

48 B.V. Cockeram et al. / Journal of Nuclear Materials 418 (2011) 46–61

values for alpha-annealed Zircaloy-2 are at the high end of therange of literature values, while the yield strengths for beta-treatedZircaloy-2 are at the low end of the range of values. Irradiation tofluences greater than 1.1 � 1024 n/m2 shows two distinctly differ-ent trends for Zircaloy-2 and Zircaloy-4 that apply to both the al-pha-annealed and beta-treated conditions for either alloy. Theyield strength and irradiation hardening for Zircaloy-4 remains rel-atively flat at fluences between 1.1 � 1024 n/m2 and 29.3 � 1024 n/m2 with no apparent systematic increase or decrease in yieldstrength being observed within the scatter of data. The variationsin yield strength observed in the results for alpha-annealed andbeta-treated Zircaloy-4 show some indication that a maxima existsbelow 5.5 � 1024 n/m2 followed by a sharp drop that is graduallyrecovered as the highest fluence is achieved, but it is difficult to re-solve this from possible data scatter. Results for Zircaloy-4 at flu-ences between 5.5 � 1024 n/m2 and 29.3 � 1024 n/m2 are 25–45%below the range of values reported in the literature for irradiationsat 260–326 �C, but are consistent with data reported for tests con-ducted at higher irradiation temperatures ((326 �C to 450 �C) inFig. 1).

Both alpha-annealed and beta-treated Zircaloy-2 follow a muchdifferent path than Zircaloy-4 with respect to neutron fluence, suchthat a much greater increase in yield strength occurs out to a flu-ence of 5.5 � 1024 n/m2 that is followed by a gradual decrease in

yield strength with continued dose to 29.3 � 1024 n/m2. Althoughthe path for the change in yield strength with respect to fluenceis very different for Zircaloy-4 and Zircaloy-2, both alloys have acomparable strength level by the time the higher fluence of29.3 � 1024 n/m2 is achieved. Higher strengthening was observedfor alpha-annealed Zircaloy-2 than for beta-treated Zircaloy-2,and the maximum yield strength for both Zircaloy-2 materialswas significantly higher than for either alpha-annealed or beta-treated Zircaloy-4 over the fluence range tested.

The higher irradiation temperature (nominally 358 �C) em-ployed in these irradiations is expected to produce the lowerpost-irradiated yield strength and irradiation hardening values rel-ative to the Zircaloy literature data for irradiation at temperaturesbetween 260 and 326 �C. The post-irradiated yield strength valuesand irradiation hardening results for Zircaloy irradiated at temper-atures between 326 and 450 �C [30] are shown in Fig. 1 to be with-in the range of values determined in this work for irradiations at358 �C. This result is also confirmed by ring tests measured fromcommercial Zircaloy-2 tubes irradiated at nominally 347 �C [31]where little irradiation hardening was observed with a calculatedultimate strength of 540 MPa that was within the range of tensilestrength values in this work (509.5–648.9 MPa). It is recognizedthat the ring test results were obtained from testing at 343 �C[31], but these results are analyzed and compared with the

(a)

Strain [%]

Stre

ss [M

Pa]

0

100

200

300

400

500

600

700

800

alpha Zr2: non-irrad alpha Zr2: 0.058X1024 n/m2

alpha Zr2: 1.1X1024 n/m2




Non-Irradiated

0.058X1024 n/m2

29.3X1024 n/m2

1.1X1024 n/m2

5.5X1024 n/m210.6X1024 n/m2

(b)

Strain [%]

Stre

ss [M

Pa]

0

100

200

300

400

500

600

700

800

beta Zr2: non-irrad beta Zr2: 0.058X1024 n/m2

beta Zr2: 1.1X1024 n/m2

beta Zr2: 5.5X1024 n/m2

beta Zr2: 10.6X1024 n/m2

beta Zr2: 29.3X1024 n/m2

Non-Irradiated

0.058X1024 n/m2

29.3X1024 n/m2

1.1X1024 n/m2

5.5X1024 n/m2

10.6X1024 n/m2

(c)

Strain [%]

Stre

ss [M

Pa]

0

100

200

300

400

500

600

700

800

alpha Zr4: non-irrad alpha Zr4: 0.058X1024 n/m2





Non-Irradiated

0.058X1024 n/m2 29.3X1024 n/m2

1.1X1024 n/m2

5.5X1024 n/m2

10.6X1024 n/m2

(d)

Strain [%]

0 2 4 6 8 10 12 14 16 18 20 22 0 2 4 6 8 10 12 14 16 18 20 22

0 2 4 6 8 10 12 14 16 18 20 22 0 2 4 6 8 10 12 14 16 18 20 22

Stre

ss [M

Pa]

0

100

200

300

400

500

600

700

800

beta Zr4: non-irrad beta Zr4: 0.058X1024 n/m2

beta Zr4: 1.1X1024 n/m2

beta Zr4: 5.5X1024 n/m2

beta Zr4: 10.6X1024 n/m2

beta Zr4: 29.3X1024 n/m2

Non-Irradiated

0.058X1024 n/m2

29.3X1024 n/m21.1X1024 n/m2

5.5X1024 n/m2

10.6X1024 n/m2

Fig. 2. Summary plot of engineering stress–strain curves for the non-irradiated and irradiated testing of: (a) alpha-annealed Zircaloy-2, (b) beta-treated Zircaloy-2, (c) alpha-annealed Zircaloy-4, and (d) beta-treated Zircaloy-4. The stress–strain curves were determined from the load versus actuator displacement with no correction made forcompliance. The curves were purposely off-set from the origin by strain with respect to increasing fluence.


assumption that the increase in yield strength (Dr) for Zircaloy isessentially independent of temperature.

3.2. Change in stress–strain curves and ductility following irradiation

The total and uniform elongation values for alpha-annealed andbeta-treated Zircaloy-4 and Zircaloy-2 are shown in Fig. 1c and d tobe similar and generally within the range of values reported in theliterature, with the exception of higher total elongation results re-ported for 326–450 �C irradiations. The exception is the uniformelongation values at higher neutron fluence are larger than the lit-erature data for Zircaloy irradiated at nominally 260–326 �C. Mostof the literature data for Zircaloy for 260–326 �C irradiations exhi-bit a prompt yield point at low total strain due to flow localizationthat results from dislocation channeling with very low uniformelongation values being measured at neutron fluences of10 � 1024 n/m2 or greater [1,2]. Higher values of uniform elonga-tion were also reported for Zircaloy-2 irradiated at higher temper-atures (326–450 �C) [30], which is consistent with the results fromthis work. The reason for the higher values of total elongation ob-served for the Zircaloy-2 irradiated at 326–450 �C in the literature[30] is not readily apparent. It is interesting that while the lower

irradiation temperature data show a fairly clear decrease in uni-form elongation with dose, there are fluence regimes for this workin which the higher irradiation temperature data display no de-crease in uniform elongation within the data scatter for Zircaloy-4 materials or even an increase in values for the Zircaloy-2materials.

Fig. 2a shows that for alpha-annealed Zircaloy-2 material, thestress–strain curve develops a pronounced yield point with lowuniform elongation at irradiation fluences between 1.1 and10.6 � 1024 n/m2. Low values of uniform elongation and evidenceof flow localization with less pronounced yield points are also ob-served to develop for beta-treated Zircaloy-2 at fluences between5.5 and 10.6 � 1024 n/m2. This is consistent with the behavior ofZircaloy irradiated at 260–326 �C reported in literature [1,2,21–33]. Interestingly as dose is increased beyond this level any evi-dence of a yield point or flow localization is completely eliminatedfor both alpha-annealed and beta-treated Zircaloy-2. This dimin-ishing of the yield point is also consistent with the higher uniformelongation and lower irradiation hardening observed at the fluenceof 29.3 � 1024 n/m2 for both alpha-annealed and beta-treated Zir-caloy-2. Fig. 2c and d shows that there is little evidence of anyupper yield point on either alpha-annealed or beta-treated

Fig. 3. SEM fractography of irradiated tensile specimens: (a) alpha-annealed Zircaloy-4 irradiated at 29.3 � 1024 n/m2, (b) beta-treated Zircaloy-4 irradiated at 29.3 � 1024 n/m2, (c) inclined view of betatreated Zircaloy-4 irradiated at 29.3 � 1024 n/m2, (d) alpha-annealed Zircaloy-2 irradiated at 10.6 � 1024 n/m2, (e) inclined view of alpha-annealed Zircaloy-4 irradiated at 10.6 � 1024 n/m2, and (f) betatreated Zircaloy-2 irradiated at 29.3 � 1024 n/m2.


Zircaloy-4 at this irradiation temperature over the range of flu-ences studied in this work. This lack of upper yield point for al-pha-annealed and beta-treated Zircaloy-4 is consistent with thehigher values of uniform elongation and lower values of irradiationhardening observed over this fluence range. Upper yield pointsonly appear in samples where the yield stress itself exceeds600 MPa, and suggests source hardening dominates friction hard-ening for this material condition.

Representative post-irradiated SEM fractography shows in Fig. 3that a ductile failure mode was observed for all of the post-irradi-ated tensile specimens for both alpha-annealed and beta-treatedZircaloy-4 and Zircaloy-2. The overall mechanism for ductile fail-ure is shown to be similar for all materials. For the alpha-annealedmaterials (Fig. 3a and d), the fracture mode consists of dimples andtearing ridges inhomogeneously distributed on the fracture surfacethat separate smoother regions to give the general appearance of

larger dimples. The fracture surface for the beta-treated samplesalso consists of fine voids and tearing ridges that are separatedby smoother regions generally having the same shape as the acic-ular lath shaped grains. High magnification images show the tear-ing ridges present in both alpha-annealed and beta-treatedZircaloy-2 and Zircaloy-4 can consist of both fine and larger voids.The same general types of features were observed for the non-irra-diated materials. For non-irradiated Zircaloy-4 in either condition,these fracture features have been shown to result from void nucle-ation, growth and coalescence that initiates from second phaseprecipitates, grain boundaries, and slipbands [2,19,34–36]. Theseresults suggest the mechanism leading to final failure of the358 �C irradiated alpha-annealed and beta-treated Zircaloy-2 andZircaloy-4 is similar to that observed for non-irradiated materials.

The fractures for the irradiated specimens are observed inFig. 3c and e to generally occur along planes that are orientated

Table 3Summary of TEM characterization results for specimens taken from post-irradiated tensile specimens following testing for alpha-annealed Zircaloy-4 following irradiation testingat nominally 358 �C to nominal fluences of 0.058, 1.1, 5.5, 10.6, and 29.3 � 1024 n/m2 (E > 1 MeV), respectively. Each line is the average for each specimen.

Capsule Neutron fluence[�1024 n/m2

(E > 1 MeV)]

hai loops hci loops

Averagediameter ± st.deviation (nm)

Max–mindiameter(nm)

Average loop numberdensity ± st. deviation/max–min(#/cm3)



Loop number density(#/cm3)

SCI-1 0.058 N/A N/A N/A N/A N/A N/ASCI-2 1.1 8.18 ± 2.95 17.15–1.26 3.34 � 1015 ± 1.44 � 1015/

6.73–1.78 � 1015N/A N/A N/A

SCI-3 5.5 11.33 ± 4.99 28.93–3.24 1.89 � 1015 ± 1.29 � 1015/3.65–0.78 � 1015

N/A N/A N/A

SCI-4 10.6 5.70 ± 2.40 16.02–1.65 12.21 � 1015 ± 7.79 � 1015/23.40–5.69 � 1015

N/A N/A N/A

SCI-5 29.3 9.65 ± 5.09 34.45–2.48 3.42 � 1015 ± 1.67 � 1015/6.02–1.95 � 1015

136.7 ± 67.2 289.0–67.2 1.37 � 1013 ± 2.55 � 1012/1.63–1.12 � 1013

Note: 1. ‘‘–‘‘ means the data were not measured. 2. N/A means the respective hai or hci loops were not observed.

Table 4Summary of TEM characterization results for specimens taken from post-irradiated tensile specimens following testing for beta-treated Zircaloy-4 following irradiation testing atnominally 358 �C to nominal fluences of 0.058, 1.1, 5.5, 10.6, and 29.3 � 1024 n/m2 (E > 1 MeV), respectively. Each line is the average for each specimen.


(E > 1 MeV)]

hai loops hci loops



Average loop numberdensity ± st. deviation/max–min(#/cm3)





13.9–8.89 � 1015N/A N/A N/A

SCI-3 5.5 14.60 ± 7.35 54.00–4.26 1.11 � 1015 ± 0.24 � 1015/1.39–0.75 � 1015

48.4 ± 27.0 54.00–4.26 <2.02 � 1012

SCI-4 10.6 10.47 ± 4.75 29.62–2.42 2.89 � 1015 ± 0.95 � 1015/3.95–1.75 � 1015

171.2 ± 76.6 309.6–49.8 6.60 � 1012 ± 7.64 � 1012/12.0–1.19 � 1012

SCI-5 29.3 10.09 ± 4.75 33.51–3.45 2.14 � 1015 ± 0.54 � 1015/2.76–1.47 � 1015

159.2 ± 89.7 337.7–69.4 8.04 � 1012 ± 1.11 � 1013/2.09–0.11 � 1013

SCI-5 29.3 8.02 ± 3.28 24.85–2.53 4.67 � 1015 ± 1.50 � 1015/6.35–3.16 � 1015

[3] N/A N/A

Note: 1. ‘‘–‘‘ means the data were not measured. 2. N/A means the respective hai or hci loops were not observed. 3. Conclusive evidence of hci loops was observed, butquantitative measurements were not performed.


about 45� to the tensile orientation where the shear stress is max-imized. The observed fine slip bands could be evidence of disloca-tion channeling. Dislocation channeling is known to occur inirradiated Zircaloy when the strain is localized on shear planes thathave been swept clear of barriers to slip [1,32,33]. These welldefined planes are not present in some of the beta-treatedZircaloy-4 and Zircaloy-2 specimens such as shown in Fig. 3c. Frac-ture in beta-treated Zircaloy generally occurs along lath bound-aries due to void coalescence from the Laves phases along theseboundaries [2,19,34–36], and this may disrupt the planar-typefracture mode in irradiated material. The second phases presentin alpha-annealed Zircaloy of either chemistry are inhomoge-neously distributed at grain boundaries and within grains so thatvoid nucleation, growth, and coalescence occurs more randomlythroughout the microstructure and results in a more homogeneousdistribution of dimples and tearing ridges. The more homogeneousdistribution of precipitates may result in a more uniform distribu-tion of slip bands that results in a tendency for more well defineddislocation channels for these materials, as shown in Fig. 3e.

4. TEM examinations of irradiated microstructure

A summary of the loop size and number densities from the TEMexamination of alpha-annealed and beta-treated Zircaloy-2 andZircaloy-4 materials is provided in Tables 3–6, respectively. Repre-sentative TEM images for Zircaloy-2 and Zircaloy-4 are provided in

Figs. 4 and 5, respectively. Plots of the average hai loop size andnumber density (ND) as a function of fluence are given in Fig. 6,while the hai loop size distributions as a function of fluence areshown in Fig. 7. Although hai loops were frequently observed, hciloops were generally only observed at the higher fluences andthe hci loops were non-uniformly distributed with a low ND atthese fluences. No change in the crystal structure or amorphizationof the second phase precipitates was observed over the range ofirradiation fluences evaluated in this work. Zircaloy-4 was ob-served to primarily contain both hexagonal and cubic Laves type((Fe,Cr)2Zr) precipitates, while Zircaloy-2 was observed to containboth Laves type ((Fe,Cr)2Zr) and Zintl type (Zr2(Fe,Ni)) precipitates[3–16]. The discussion given in the following sections is organizedby irradiation fluence.

4.1. Lowest fluence irradiations (0.058 � 1024 n/m2)

For irradiation to the lowest fluence of 0.058 � 1024 n/m2,Figs. 4a and 5a show that neither hai nor hci type dislocation loopswere observed in any specimens examined. This is consistent withthe results from literature [7–13] for irradiations at slightly lowertemperatures where hai and hci dislocation loops were also notobserved for irradiations at low fluences of about 0.1–0.2 �1024 n/m2. If dislocation loops or point defect clusters were pres-ent, then they were below the size detection limit for the TEM(nominally <0.4 nm) and could not be resolved. While the lack of

Table 5Summary of TEM characterization results for specimens taken from post-irradiated tensile specimens following testing for alpha-annealed Zircaloy-2 following irradiation testingat nominally 358 �C to nominal fluences of 0.058, 1.1, 5.5, 10.6, and 29.3 � 1024 n/m2 (E > 1 MeV), respectively. Each line is the average for each specimen.

Capsule Neutron Fluence[�1024 n/m2

(E > 1 MeV)]

hai loops hci loops



Average loop numberdensity ± st. deviation/max–min (#/cm3)





3.35–2.05 � 1015N/A N/A [3]

SCI-3 5.5 7.17 ± 2.99 20.21–1.99 8.25 � 1015 ± 2.16 � 1015/10.1–4.86 � 1015

N/A N/A N/A

SCI-4 10.6 7.46 ± 2.53 14.56–2.53 3.60 � 1015 ± 0.44 � 1015/4.00–3.13 � 1015

169.3 ± 42.69 268.2–111.6

2.93 � 1012 ± 2.43 � 1012/5.34–0.48 � 1012

SCI-5 29.3 21.53 ± 15.99 82.69–4.39 3.81 � 1014 ± 1.02 � 1014/4.79–2.62 � 1014

175.6 ± 105.5 432.0–59.6 2.21 � 1012 ± 1.81 � 1012/4.85–1.02 � 1012

Note: 1. ‘‘–‘‘ means the data were not measured. 2. N/A means the respective hai or hci loops were not observed. 3. The observation of hci loops was possible, but the evidencewas inconclusive.

Table 6Summary of TEM characterization results for specimens taken from post-irradiated tensile specimens following testing for beta-treated Zircaloy-2 following irradiation testing atnominally 358 �C to nominal fluences of 0.058, 1.1, 5.5, 10.6, and 29.3 � 1024 n/m2 (E > 1 MeV), respectively. Each line is the average for each specimen.


(E > 1 MeV)]

hai loops hci loops



Average loop numberdensity ± st. deviation/max–min (#/cm3)



Loop number density (#/cm3)

SCI-1 0.058 N/A N/A N/A N/A N/A N/ASCI-2 1.1 9.24 ± 3.74 31.15–3.05 1.29 � 1015 ± 0.43 � 1015/1.86–

0.94 � 1015N/A N/A N/A

SCI-3 5.5 9.30 ± 3.29 18.77–3.01 5.45 � 1015 ± 0.74 � 1015/6.06–4.51 � 1015

[3] N/A [3]

SCI-4 10.6 7.92 ± 3.22 22.73–2.28 3.66 � 1015 ± 0.68 � 1015/4.27–2.50 � 1015

N/A N/A N/A

SCI-5 29.3 18.18 ± 8.41 65.09–6.36 5.31 � 1014 ± 3.75 � 1014/10.9–3.03 � 1014

220.5 ± 84.7 378.0–146.0

7.76 � 1012 ± 3.90 � 1012/10.5–4.98 � 1012

Voids (cavities)

Average diameter ± st. deviation (nm) Max–min diameter (nm) Average void number density ± st. deviation/max–min (#/cm3)

SCI-5 29.3 16.69 ± 8.21 49.57–3.07 2.43 � 1013 ± 7.96 � 1012/3.59–1.77 � 1013

Note: 1. ‘‘–‘‘ means the data were not measured. 2. N/A means the respective hai or hci loops were not observed. 3. Conclusive evidence of hci loops was observed, butquantitative measurements were not performed.


observable loops is not surprising in the case of Zircaloy-2 andbeta-treated Zircaloy-4 where little to no irradiation hardeningwas seen at this low dose, visible loops were also not observedin the alpha-annealed Zircaloy-4 specimen which did show a moresubstantial change in yield strength at this low dose.

4.2. Low fluence irradiations (1.1 � 1024 n/m2)

For irradiation to the higher fluence of 1.1 � 1024 n/m2, the for-mation of hai type dislocation loops was observed for all materialsin Figs. 4b and 5b. The formation of hci type loops was observed inalpha-annealed Zircaloy-2, but the numbers were not large enoughto determine a number density. The average size of the hai-typeloops at this dose is shown in Fig. 6 to range from 6.1 to 9.2 nm,increasing in size from beta-treated Zr-4 to alpha-annealed Zr-2to alpha-annealed Zr-4 to beta-treated Zr-2. The number densityof these loops ranged from 1.3 � 1015 #/cm3 to 3.3 � 1015 #/cm3

for beta-treated and alpha-annealed Zircaloy-4 and for alpha-an-nealed Zircaloy-4, but increased to 1.1 � 1016 #/cm3 for beta-trea-ted Zircaloy-4.

4.3. Medium fluence irradiations (5.5 � 1024 n/m2)

All of the precipitates were observed to be crystalline, and notransformation to amorphous features was observed. A distribu-

tion of hai loops are observed for all materials in Figs. 4c and 5c.The slight increase in hai loop size for alpha-annealed and beta-treated Zircaloy-4 resulting from irradiation to higher fluenceshown in Figs. 6 and 7 is balanced by a decrease in the loop numberdensity, which may explain why a relatively small change in irra-diation hardening was observed for alpha-annealed and beta-trea-ted Zircaloy-4 at fluences between 1.1 and 5.5 � 1024 n/m2. Thefirst clear evidence of hci dislocation loops was observed forbeta-treated Zircaloy-4.

In contrast to the Zircaloy-4 behavior, in the case of alpha-an-nealed and beta-treated Zircaloy-2 there was no apparent changein the hai dislocation loop size following irradiation between lowto medium fluence, but an increase in the hai loop number densityby about a factor of 4 was observed. The hai loop sizes for alpha-an-nealed and beta-treated Zircaloy-2 are slightly smaller than for Zir-caloy-4, but the hai loop number densities in Zircaloy-2 are morethan a factor of 4 larger than for Zircaloy-4 at this fluence. The sig-nificant increase in the hai loop number density observed for Zirca-loy-2 is consistent with the increase in irradiated yield strengthshown in Fig. 1.

4.4. High fluence irradiations (10.6 � 1024 n/m2)

At this exposure, the hai loops observed for both alpha-annealedand beta-treated Zircaloy-4 in Fig. 4d are shown in Figs. 6 and 7 to

Fig. 4. Representative TEM images of loops from alpha-annealed and beta-treated Zircaloy-4 following irradiation at nominally 358 �C to fluences between 0.058 � 1024 n/m2

and 29.3 � 1024 n/m2 (E > 1 MeV): (a) alpha-annealed Zircaloy-4 at dose of 0.058 � 1024 n/m2, (b) alpha-annealed Zircaloy-4 following a fluence of 1.1 � 1024 n/m2, (c) beta-treated Zircaloy-4 after a dose of 5.5 � 1024 n/m2, (d) alpha-annealed Zircaloy-4 after a fluence of 10.6 � 1024 n/m2, (e) alpha-annealed Zircaloy-4 following a fluence of29.3 � 1024 n/m2, and (f) hci loops indicated by arrows in alpha-annealed Zircaloy-4 following a dose of 29.3 � 1024 n/m2.


exhibit a reduction in loop diameter and an increase in hai loopnumber density. The hai loop sizes observed for the Zircaloy-2samples in Fig. 5d are shown in Figs. 6 and 7 to exhibit little changeover this range of dose. A factor of 1.5–2 decrease in the hai loopnumber density was observed for Zircaloy-2 going from the fluenceof 5.5 � 1024 n/m2 to 10.6 � 1024 n/m2. This decrease in hai loopnumber density with no significant change in hai loop size is con-sistent with the slight decrease in irradiation hardening observedfor both alpha-annealed and beta-treated Zircaloy-2 at these flu-ences, see Fig. 1.

The hci loops that had been observed in beta-treated Zircaloy-4at 5.5 � 1024 n/m2 increased in both size and number density withcontinued irradiation to 10.6 � 1024 n/m2. The formation of ha + ciloops was observed in many cases for alpha-annealed and beta-treated Zircaloy-4, indicating there could be tendency for the hciloops to react with hai loops to produce the ha + ci configuration.The formation of hci loops was also observed for alpha-annealedZircaloy-2 to high enough numbers that could be quantified.Although hci loops were observed for beta-treated Zircaloy-2 atlower fluences of 5.5 � 1024 n/m2, no hci loops were observed atthe higher fluence of 10.6 � 1024 n/m2. In all cases the formationof hci loops was observed to be very localized and non-uniform,such that the examination of many areas would be needed to bet-ter quantify the hci loop sizes and number density. The hci loop

number density values are taken from localized regions, and repre-sent a maximum value that would need to be better quantified byadditional measurements.

4.5. Highest fluence irradiations (29.3 � 1024 n/m2)

Irradiation to the highest fluence 29.3 � 1024 n/m2 resulted inabout a factor of 2 increase in hai loop size with a factor of 8–10decrease in hai loop number density for both alpha-annealed andbeta-treated Zircaloy-2, see Fig. 5e. In contrast, little change inhai loop size is observed in the Zircaloy-4 materials with either aND decrease (alpha-annealed) or ND increase (beta-treated), seeFig. 4e. All of the precipitates were observed to be crystalline,and no transformation to amorphous features was observed. Irradi-ation to higher fluences may be required for the precipitates to be-come amorphous at the higher irradiation temperature of 358 �C.Some evidence of iron diffusing out of the Laves Zr(Fe,Cr)2 precip-itates was observed, but the gradients were small (on the order of2–4 atom% composition difference) in most cases and not all pre-cipitate types exhibited evidence of elements diffusing out. No dif-fusion of elements was observed for the Zr2(Fe,Ni) precipitates thatare present in Zircaloy-2, which is consistent with literature data[3–18]. While precipitates becoming amorphous and elements dif-fusing out of precipitates is reported for irradiations at 260–326 �C,

Fig. 4 (continued)


the lack of the precipitates becoming amorphous and slight move-ment of precipitate elements into the matrix is consistent with lit-erature data for irradiation at higher temperatures that bound the358 �C irradiation temperature in this work [3–18]. The greatesttendency for the precipitates to become amorphous is reportedin the literature to occur at irradiation temperatures of nominally260–326 �C. Some enrichment of Sn (1.5–2.5%) and Fe (0.9–1.5%)was observed for Zircaloy-4 and Zircaloy-2 at the grain boundariesin some cases with a slight enrichment of 0.7–1.4% Ni also ob-served for Zircaloy-2, while some grain boundaries did not exhibitevidence of enrichment.

Clear evidence of hci dislocation loops was observed for both al-pha-annealed and beta-treated Zircaloy-4 and Zircaloy-2 in Figs. 4fand 5f, respectively. An increase in the hci loop size and hci loopnumber density was observed at the higher fluences for alpha-an-nealed and beta-treated Zircaloy-4. However, the number densityof the hci dislocation loops remains relatively low, and the hci loopdistribution is inhomogeneous at these fluences. The hci loops gen-erally had the same type of trend as observed for the hai loops, withthe hci loop sizes for alpha-annealed and beta-treated Zircaloy-2being larger than for alpha-annealed and beta-treated Zircaloy-4,but the hci loop number densities for Zircaloy-2 are lower thanfor Zircaloy-4. It is interesting to note that clear evidence of hci loop

nucleation was observed at the same fluence where evidence of Feand Cr diffusion from the Laves Zr(Fe,Cr)2 precipitates wasobserved.

Cavities or voids are observed to form at a very low numberdensity for beta-treated Zircaloy-2 in Fig. 5g, but void forma-tion is not observed for Zircaloy-4 or alpha-annealed Zircaloy-2. The voids are on the order of the same size as the hai loopsobserved for beta-treated Zircaloy-2, but these cavities exhib-ited a 3-dimensional characteristic that indicates these are in-deed voids (or cavities). Some of the voids are observed to befaceted, while faceting was less obvious for other voids. Thesevoids were randomly distributed, very localized, and their for-mation was not observed to be associated with any specifictype of feature. The voids were not uniformly distributed inbeta-treated Zircaloy-2, and measurement of this low voidnumber density is associated with a lot of variability and error.A search of available literature indicate that void formation istypically not observed for irradiated Zircaloy, but inhomogenousor patchy distributions of voids have been reported in a fewcases for Zircaloy irradiated at temperatures of 326–500 �C[13,18,37,38]. Since the void formation was very non-uniform,additional characterization is needed to better characterize thevoids.

Fig. 5. Representative TEM images of loops from alpha-annealed and beta-treated Zircaloy-2 following irradiation at nominally 358 �C to fluences between 0.058 � 1024 n/m2

and 29.3 � 1024 n/m2 (E > 1 MeV): (a) alpha-annealed Zircaloy-2 after dose of 0.058 � 1024 n/m2, (b) hai loops observed in alpha-annealed Zircaloy-2 following a fluence of1.1 � 1024 n/m2, (c) beta-treated Zircaloy-2 following a fluence of 5.5 � 1024 n/m2, (d) alpha-annealed Zircaloy-2 following a fluence of 10.6 � 1024 n/m2, (e) alpha-annealedZircaloy-2 after a fluence of 29.3 � 1024 n/m2, (f) images of hci loops for alpha-annealed Zircaloy-2 following a dose of 29.3 � 1024 n/m2, and (g) images of cavities observed inbeta-treated Zircaloy-2 at a fluence of 29.3 � 1024 n/m2.


4.6. Fluence dependence for hai loop evolution

The increase in loop number density and size has been reportedin the literature for molybdenum, stainless steels, and other mate-rials to have a (fluence)1/n or (/t)1/n dependence where n = 2, 4, or 6[1,7–18,26–32,39,40]. Assuming the increase in hai loop diameterand number density with fluence for Zircaloy-2 and Zircaloy-4 alsofollow a (/t)1/n dependence, fits were made to the change in hailoop size and number density with fluence using an incubation

term ð/tÞxwx

oþð/tÞx

h i� �that varies from 0 to 1, where wo and x are mate-

rial constants. Fits for the change in hai loop diameter (D) and num-ber density (ND) as a function of fluence were made to thefollowing equations.

D ¼ ð/tÞx

wxo þ ð/tÞx

� �� ½aþ bð/tÞ1=n� ð1Þ

ND ¼ ð/tÞx

wxo þ ð/tÞx

� �� ½c þ dð/tÞ1=n� ð2Þ

where a, b, c, and d are material constants obtained from regres-sion fits to the TEM data in Tables 3–6. The best fit that

represented the fluence dependence for these limited datasetswas found for n = 6. The fits shown in Fig. 6 provide a basis forcomparison of the change in hai loop D and ND as a function offluence. Comparisons are also made with the available literaturedata for the hai loop size and number density [7–18] categorizedby the irradiation temperature. In general, for Zircaloy irradiationsat temperatures less than 316 �C the hai loop number densities aremuch higher and the hai loop sizes reported in the literature aremuch smaller than for the results reported here for irradiationat 358 �C. For Zircaloy irradiations at temperatures greater than316 �C, the literature data for hai loop size and number densityare generally within the range of the results reported here. Thelower hai loop number density and larger hai loop size observedat the higher irradiation temperatures are likely due to the greatermobility of point defects which promotes loop growth over con-tinued loop nucleation [1,7–18,26–32]. There is an overall changein the evolution of hai loops with irradiation temperature. Satura-tion of the hai loop number density is observed at fluences of10 � 1024 n/m2 for irradiations at temperatures less than 316 �C.Two different behaviors are observed for the hai loops in Figs. 6and 7 in this work for the irradiation of Zircaloy-4 and Zircaloy-2 at 358 �C with a general saturation at a lower fluence of

Fig. 5 (continued)


1.1 � 1024 n/m2 with changes discussed in the following section.For Zircaloy-2 a peak in hai loop number density is observed ata fluence of 5.5 � 1024 n/m2 followed by a decrease in numberdensity out to the fluence of 29.3 � 1024 n/m2.

The evolution of the hai-loop population, plotted in Figs. 6 and7, displays a distinct difference in behavior between Zircaloy-2and Zircaloy-4. All Zircaloy-2 and Zircaloy-4 specimens were irra-diated in the same capsule environment. Material variability is onepossible explanation, but consistent results and trends are ob-served for duplicate testing performed among the material typeswith relatively low amounts of data scatter. For both the alpha-annealed and beta treated Zircaloy-2, average loop diameter in-creases rapidly with dose and then remains relatively constantout to 10.6 � 1024 n/m2. Between the dose of 10.6 and29.3 � 1024 n/m2, there is a significant increase in loop diameter.The loop number density increases with dose to a maximum at5.5 � 1024 n/m2 and then drops off with continued irradiation, asthe loop population shifts to larger diameter and widens. The max-imum hai loop number density could be between 1.1 and5.5 � 1024 n/m2. The behavior of both the alpha-annealed andbeta-treated samples of this Zircaloy-2 material appear quite sim-ilar. The increasing hai loop number density between fluences of1.1 and 5.5 � 1024 n/m2 followed by decreases at fluences between5.5 and 29.3 � 1024 n/m2 n/m2 is also reflected by the peak in the

inverse hai loop spacing shown in Fig. 8, which is defined byffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiðDÞðNDÞ

p.

The average diameter of the Zircaloy-4 loops for both alpha-an-nealed and beta-treated material also increase rapidly at the initialdose of 1.1 � 1024 n/m2 where hai loops are observed to the fluenceof 5.5 � 1024 n/m2, but exhibit more dose dependence below10.6 � 1024 n/m2 as compared to Zircaloy-2. Relatively littlechange in hai loop size occurred over the final increment of fluencebetween 10.6 and 29.3 � 1024 n/m2. The number density in thebeta treated Zircaloy-4 peaks quite early in the irradiation historyand then drops to lower values that do not differ much at fluencesgreater than 5.5 � 1024 n/m2. This suggests the peak number den-sity for beta-treated Zircaloy-4 may occur between a fluence of 1.1and 5.5 � 1024 n/m2. A peak is also observed for alpha-annealedZircaloy-4 at the intermediate level fluence of 10.6 � 1024 n/m2,with a substantial decline observed over the final interval of dosesuch that the number density values at the fluence of29.3 � 1024 n/m2 is similar to that observed at the fluence of1.1 � 1024 n/m2. The changes in the hai loop size for the Zircaloy-4 materials are generally balanced by opposing changes in numberdensity, such that the net effect is the inverse hai loop spacing, asdefined by

ffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiðDÞðNDÞ

p, is shown in Fig. 8 to be relatively constant

with fluence between the dose of 1.1 of 29.3 � 1024 n/m2. By theend of the irradiation history, the Zircaloy-2 material develops

Fluence [ X 1024 n/m2, E > 1 MeV]

Dia

met

er [n

m]

0.0

5.0

10.0

15.0

20.0

25.0

Alpha Zircaloy-2Beta Zircaloy-2Alpha Zircaloy-4Beta Zircaloy-4Fit - alpha-Zr2 - (φt)1/6

Fit - beta-Zr2 - (φt)1/6

Fit - alpha-Zr4 - (φt)1/6


Zircaloy Literature T < 316CZircaloy Literature T > 316C

Fluence [ X 1024 n/m2, E > 1 MeV]

0 5 10 15 20 25 30

0 5 10 15 20 25 30

Num

ber D

ensi

ty [#

/cm

3 ]

2.0e+15

4.0e+15

6.0e+15

8.0e+15

1.0e+16

1.2e+16

1.4e+16

Alpha Zircaloy-2Beta Zircaloy-2Alpha Zircaloy-4Beta Zircaloy-4Fit - alpha-Zr2 - (φt)1/6


Fit - alpha-Zr4 - (φt)1/6


Zircaloy Literature T < 316CZircaloy Literature T > 316C

(a)

(b)

Fig. 6. Summary plot of average hai loop measurements determined from TEMcharacterization as a function of dose for alpha-annealed and beta-treated Zircaloy-4 and Zircaloy-2 irradiated at nominally 358 �C with a curve fit to the data andcomparisons to literature data [7–14]: (a) hai loop diameters with a fit to

D ¼ ð/tÞxwx

oþð/tÞx

h i� ½aþ bð/tÞ�1=n with n = 6, and (b) hai loop number density values with

a fit to ND ¼ ð/tÞxwx

oþð/tÞx

h i� ½c þ dð/tÞ�1=n

h iwith n = 6.


larger hai loops than Zircaloy-4, but considerably fewer of them.From the data presented in Tables 3–6, it appears that in eachmaterial the most significant decrease in hai-loop density occursduring the interval in which hci-loops are first observed in quanti-fiable numbers.

Although the basic chemistries of the materials studied here aresimilar, the precipitate populations of alpha and beta treated Zirca-loy-2 and Zircaloy-4 do differ from each other. Only laves Zr(Fe,Cr)2

precipitates are observed for Zircaloy-4 that more readily give Feand Cr solute into the matrix during irradiation. Zircaloy-2 has alower fraction of laves Zr(Fe,Cr)2 precipitates but also has ZintlZr2(Fe,Ni) precipitates that do not change in composition duringirradiation, which may result in less solute release into the matrixduring irradiation. It is expected that the evolution of these parti-cles with irradiation, particularly in regard to the release of speciessuch as Fe and Cr into the matrix, may have a significant effect onthe behavior or point defects and loops. The late in life growth dis-played by the Zircaloy-2 loops, that is apparently absent in Zirca-loy-4, may be correlated to greater propensity for Fe and Crresolution from the higher fraction of laves precipitates presentin Zircaloy-4, providing solute pinning or otherwise impeding theabsorption.

5. Loop hardening models

Irradiation hardening is generally described in terms of thestress needed to start a mobile dislocation moving from rest ornucleate the fresh dislocations needed to support plasticity (sourcehardening) and the stress required to keep the dislocation movingthrough the material (friction hardening). Source hardening is usu-ally attributed to the pinning of dislocations by small point defectclusters or solute atoms. Larger microstructural features influencedislocation motion by increasing the friction stress through eitherlong range (network dislocations) or short range (voids, loops, pre-cipitates) interactions. Models for irradiation hardening due to ex-tended point defect configurations, such as dislocation loops orvoids, generally focus on short range friction hardening causedby a dispersed array of hard barriers to dislocation motion[39,40]. The increase in the yield stress (Dry) required to movethe dislocations in this model is given by the following generalexpression [39]

Dry ¼ aMGbffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiðDÞðNDÞ

pð3Þ

where M is the Taylor factor (taken here as 2 with the assumptionthat slip occurred in grains that were optimally orientated for slip),G is the shear modulus, b the Burgers vector, ND the loop numberdensity and D the loop diameter (previously identified as the in-verse hai loop spacing term). An average Taylor factor for randomlyorientated grains has been calculated to be 3.06 for fcc metals and2.65–3.06 for bcc metals [39,41,42], but 2 is selected as a boundingvalue in this work, recognizing that a lower M will tend to result inhigher a values. The parameter a is used to describe the strength ofthe individual barriers. For barriers that are too hard to be pene-trated, and can only be overcome by dislocation bow (Orowan hard-ening), such as voids and second phase precipitates, a = 1 [39]. Forsofter barriers that can be cut by the dislocation, a < 1. Values ofa for dislocation loops have been reported to range from 0.25 to0.5 [39] with the value of a = 0.25 being more typical in some cases[40].

For alpha-annealed Zircaloy-4, beta-treated Zircaloy-4, and al-pha-annealed Zircaloy-2, measured irradiation hardening values(Dr) ranging from 5.96 to 23.0 ksi were observed in Fig. 1b afterthe lowest fluence of 0.058 � 1024 n/m2, but defects that could pro-duce such hardening, such as hai or hci loops, were not resolvedduring the TEM examination (<0.4 nm). Small point defect clustersor depletion zones [43] could potentially produce such hardening.A correlation between the hai loop distributions and irradiationhardening was observed at the fluences between 1.1 and29.3 � 1024 n/m2. The trend for the inverse hai loop spacing(ffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiðDÞðNDÞ

p), which is the material parameter used to calculate

the irradiation hardening in Eq. (3), is shown in Fig. 8 to duplicatethe same trends observed for the measured irradiation hardeningshown in Fig. 1b. For alpha-annealed and beta-treated Zircaloy-4,a large initial increase is observed at a fluence of 1.1 � 1024 n/m2

followed by little change at the higher fluences. For alpha-annealedand beta-treated Zircaloy-2, a large increase is observed at a flu-ence of 1.1 � 1024 n/m2 that peaks at a fluence between 1.1 and5.5 � 1024 n/m2 and is then followed by a decrease at fluencesbetween 5.5 and 29.3 � 1024 n/m2.

Eq. (3) can be applied to the loop data for number density andaverage diameter in Table 3–6 in order to compare model predic-tions to the measured hardening values. Assuming the hai loopsprovide the strengthening given by a standard dislocation loophardening model with a typical value of a = 0.25 (Eq. (3)) withG = 32.3 GPa and b = 3.23 Å [1], the calculated values are generallya factor of 1.3–11.0 lower than for the measured irradiation hard-ening values. The significantly lower calculated strength suggeststhe hai loops are either much stronger barriers than indicated by

Loop Size [nm]

Inte

rval

Num

ber D

ensi

ty [#

/cm

3 ]

0.0

2.0e+14

4.0e+14

6.0e+14

8.0e+14

1.0e+15

1.2e+15

1.4e+15

1.6e+15

1.8e+15

2.0e+15

2.2e+15

Alpha-Zr4: 1.1X1024 n/m2




Loop Size [nm]

Inte

rval

Num

ber D

ensi

ty [#

/cm

3 ]

0.0

2.0e+14

4.0e+14

6.0e+14

8.0e+14

1.0e+15

1.2e+15

1.4e+15

1.6e+15

1.8e+15

2.0e+15

2.2e+15

Beta-Zr4: 1.1X1024 n/m2

Beta-Zr4: 5.5X1024 n/m2

Beta-Zr4: 10.6X1024 n/m2

Beta-Zr4: 29.3X1024 n/m2

Beta-Zr4: 29.3X1024 n/m2

Loop Size [nm]

Inte

rval

Num

ber D

ensi

ty [#

/cm

3 ]

0.0

2.0e+14

4.0e+14

6.0e+14

8.0e+14

1.0e+15

1.2e+15

1.4e+15

1.6e+15

1.8e+15

2.0e+15

2.2e+15





Loop Size [nm]

0 2 4 6 8 10 12 14 16 18 20 22 24 26 28 30 32 0 2 4 6 8 10 12 14 16 18 20 22 24 26 28 30 32

0 5 10 15 20 25 30 35 40 45 50 55 60 0 5 10 15 20 25 30 35 40 45 50 55 60

Inte

rval

Num

ber D

ensi

ty [#

/cm

3 ]

0.0

2.0e+14

4.0e+14

6.0e+14

8.0e+14

1.0e+15

1.2e+15

1.4e+15

1.6e+15

1.8e+15

2.0e+15

2.2e+15

Beta-Zr2: 1.1X1024 n/m2

Beta-Zr2: 5.5X1024 n/m2

Beta-Zr2: 10.6X1024 n/m2

Beta-Zr2: 29.3X1024 n/m2

(a) (b)

(c) (d)

Fig. 7. Measured hai loop size distributions where the number of loops is defined in terms of a loop number density by dividing by interval hai loop number by the totalvolume used for measurement. The results are determined from TEM characterization of alpha-annealed and beta-treated Zircaloy-4 and Zircaloy-2 following irradiation atnominally 358 �C to nominal fluences of 0.058, 1.1, 5.5, 10.6, and 29.3 � 1024 n/m2 (E > 1 MeV) for: (a) alpha-annealed Zircaloy-4, (b) beta-treated Zircaloy-4, (c) alpha-annealed Zircaloy-2, and (d) beta-treated Zircaloy-2.


these standard loop hardening models, or fine clusters smaller thanthe detection limit for TEM are present that dominate the harden-ing. For the literature data for Zircaloy irradiation temperaturesexceeding 316 �C [1,7–18,26–32], the hai loop number densityand sizes are shown in Fig. 6 to be similar to those in this work,and the calculated irradiation hardening values are within therange of those determined in this work. Using literature data forthe higher hai loop number density and lower hai loop size ob-served at the lower irradiation temperatures less than 316 �Cshown in Fig. 6 [1,7–18,26–32], the calculated loop hardening val-ues are still a factor of 3.4–6.1 lower than the experimentally mea-sured irradiation hardening values. The finer hai loop size andnumber density observed at irradiation temperatures less than316 �C shown in Fig. 6 [1,7–18,26–32] results in a higher calculatedirradiation hardening value than obtained in this work that is with-in the range of differences in experimentally measured irradiationhardening values.

If the hai dislocation loops are treated as perfectly hard barriersby assigning a = 1 as is generally done with voids, Fig. 9a and cshows that an improved correlation exists between the measuredand calculated irradiation hardening values due to hai loops usingEq. (3). In this case the ratio of measured versus predicted valuesranges from 0.31 to 2.5, with an average of 1.32 and standard devi-ation of 0.66. Changes in the barrier strength of loops due to thepresence of solutes and interstitials has been shown to be potenteffect for materials such as vanadium [44]. Thus, part of the reasonfor the small differences in the calculated and measured irradiationhardening for the irradiated Zircaloy-2 and Zircaloy-4 could result

from the possibility of the influence of solute atoms on the barrierstrength of the loops, but further investigation is needed to verifythis theory. With few exceptions, the measured data is somewhatover predicted with a = 1, but this is to be expected since loopsmay not be perfectly hard impenetrable obstacles. Fig. 9a alsoshows that the calculated irradiation hardening values also followthe same trend as determined for the measured hardening valueswith initial hardening that peaks between a fluence of 1.1 and5.5 � 1024 n/m2 and then a decrease in values at higher fluencesfor alpha-annealed and beta-treated Zircaloy-2, and an initial hard-ening with little change observed between fluences of 1.1 and29.3 � 1024 n/m2 for Zircaloy-4. For the literature data shown inFig. 6, a closer match between the measured and calculated irradi-ation hardening values was determined using Eq. (3) with theassumption of a = 1.

One effect to consider is the fact that the hai-loops exist in a dis-tribution that may not be accurately represented by using the aver-age diameter and total number density. The calculation can berefined to sum over the various loop sizes but in doing so it isimportant to employ the proper rule for superposition. Performingthe summations over the loop distributions in Fig. 7 in accordancewith the expression from Eq. (3) [37]

Dry ¼ aMGbffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiXi¼bins

ðDÞðNDÞr

ð4Þ

where a fractional summation is performed for the void size distri-bution. The size distributions for the hai loops given in Fig. 7 aregenerally close to normal and symmetric, and are generally not

Fluence [X 1024 n/m2, E > 1 MeV]0 5 10 15 20 25 30

inve

rse

<a>

loop

spa

cing

[cm

-1]

0.0

1.0e+4

2.0e+4

3.0e+4

4.0e+4

5.0e+4

6.0e+4

7.0e+4

8.0e+4

9.0e+4

1.0e+5

1.1e+5

alpha Zircaloy-2beta Zircaloy-2alpha Zircaloy-4beta Zircaloy-4

Fig. 8. Summary plot of measured inverse hai loop spacing determined by averagehai loop values as a function of fluence for alpha-annealed and beta-treatedZircaloy-2 and Zircaloy-4 irradiated at nominally 358 �C.


significantly skewed towards smaller or larger sizes. For a = 1, thehai loop hardening values determined from the average size andnumber density (Eq. (3)) are only slightly lower than for thosedetermined from the size distribution (Eq. (4)) being 15.5% largerto 0.6% larger with the average being 4.7% higher and a standarddeviation of 2.6.

Fluence [X 1024 n/m2, E > 1 MeV]0 5 10 15 20 25 30

YS, C

alcu

late

d Va

lue

[MPa

]

0

50

100

150

200

250

300

350

400

450


YS, Calculated Value [MPa]

YS, M

eas.

Cha

nge

in Y

ield

Str

engt

h

0

50

100

150

200

250

300

350

400

450

alpha Zircaloy-2beta Zircaloy-2alpha Zircaloy-4beta Zircaloy-4Perfect Fit Line

0 50 100 150 200 250 300 350 400 450

[MPa

]

(a) (

(c) (

Fig. 9. Comparison of calculated irradiation hardening values determined for hai loopsannealed and beta-treated Zircaloy-2 and Zircaloy-4 irradiated at nominally 358 �C (as-ionly, (b) calculated hardening as a function of fluence for hai + hci loop hardening usingloops, and (d) plot of measured versus calculated hardening using hai + hci loops.

Fig. 9c shows that a closer match between measured and calcu-lated irradiation hardening is generally observed for beta-treatedZircaloy-2 and Zircaloy-4 than for the respective alpha-annealedcondition. The more uniform texture and uniform lath grain struc-ture with less variability of location for the second phase precipi-tates for the beta-treated microstructure likely contributes to lessvariability in the mechanical properties. Beta-treated Zircaloy hasbeen shown to exhibit less variation in fracture toughness and ten-sile strength results due to the mechanism of void nucleation,growth, and coalescence that initiates at the lath boundaries[2,19]. The localization of precipitates primarily at lath boundariesfor beta-treated Zircaloy, which are sites for the void, nucleation,growth, and coalescence mechanism that results in fracture, is be-lieved to be the reason for the more uniform mechanical proper-ties. The fractography results in this work show that irradiatedbeta-treated and alpha-annealed exhibit the same void nucleation,growth, and coalescence mechanism that is observed for non-irra-diated materials. The fracture mechanism resulting in more uni-form mechanical properties for beta-treated may explain thecloser match to model results than for alpha-annealed.

A final refinement to the hardening calculation can be made byincluding the effect of the hci-loops. While these loops appear insignificantly smaller number densities than the hai-loops, their in-creased size and larger Burger’s vector (b = 5.15 Å for hci loops) willenhance their contribution to hardening. Adding the hci-loops via asummation procedure with the assumption that their individual

Fluence [X 1024 n/m2, E > 1 MeV]

YS, C

alcu

late

d Va

lue

[MPa

]

0

50

100

150

200

250

300

350

400

450


0 5 10 15 20 25 30

YS, Calculated Value [MPa]0 50 100 150 200 250 300 350 400 450

0

50

100

150

200

250

300

350

400

450

alpha Zircaloy-2beta Zircaloy-2alpha Zircaloy-4beta Zircaloy-4Perfect Fit Line

YS, M

eas.

Cha

nge

in Y

ield

Str

engt

h[M

Pa]

b)

d)

only and hai + hci loops using the summation void hardening equation for alpha-rradiated condition): (a) calculated hardening as a function of fluence for hai loopsthe summation method, (c) plot of measured versus calculated hardening using hai


strength is comparable are considered using a superposition ap-proach by a linear sum of squares value recommended for barriersthat have a similar strength [37], which yields

Dry ¼ffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiahaiMGbhai

ffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiðDhaiÞðNDhaiÞ

q� �2þ ahciMGbhci

ffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiðDhciÞðNDhciÞ

q� �2r

ð5Þ

An alternative form is considered where a linear addition of thehai and hci loop strengthening is considered as the maximum ex-tent of calculated irradiation hardening contribution due to thecombined effects of hai and hci loop hardening, and is consideredas a bounding value for calculated irradiation hardening

Dry ¼ ahaiMGbhaiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiðDhaiÞðNDhaiÞ

qþ ahciMGbhci

ffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiðDhciÞðNDhciÞ

qð6Þ

The calculated hardening values were the same as the hai loophardening (0% difference) at lower doses where hci loops werenot observed, or a maximum of 42% higher at higher fluenceswhere hci loops were observed, see Fig. 9b and d. For the summa-tion approach (Eq. (6)), the calculated values for hardening usinghai and hci loops are a factor of 0.31–2.5 difference than for themeasured irradiation hardening with an average of 1.11 and stan-dard deviation of 0.65 that is generally closer to the measured val-ues than for the hardening that results from consideration of thehai loops only (Eq. (3)) or the use of the superposition approach(Eq. (5)). Plots comparing the calculated irradiation hardeningdetermined for the hai loops (Eq. (3)) in Fig. 9c with those for theadditional consideration of the hardening contributions from thehci loops using Eq. (6) shows in Fig. 9d that inclusion of hci loophardening results in a slightly improved match between the mea-sured and calculated irradiation hardening. These comparisons ofcalculated hardening indicate that the hci loops may have an effecton irradiation hardening that may not be insignificant. Since basalslip is known to become important for irradiated Zircaloy[1,7,32,33], the formation of hci loops could have some effect onthe irradiation hardening.

6. Summary and conclusions

The increase in yield strength resulting from irradiation harden-ing was investigated for alpha-annealed and beta-treated Zircaloy-4 and Zircaloy-2. Irradiations were performed at nominally 358 �Cto various levels of fast fluence between 0.058 and 29.3 � 1024 n/m2. Irradiation hardening (Dr) was consistent with previous workperformed at the higher irradiation temperatures consistent withthis work, but considerably less than observed for literature datafor irradiations between 260 and 326 �C. Materials initially hardenvery quickly with dose to about 1.1 � 1024 n/m2, and then display areduced dose dependence. While Zircaloy-4 exhibited some scatterin the higher dose dependence of yield stress, the Zircaloy-2 sam-ples displayed a more definite decrease in hardening with increas-ing dose beyond 5.5 � 1024 n/m2. Irradiation at nominally 358 �C isshown in this work to consistently result in a 25–45% lower yieldstrength, less increase in tensile strength and less irradiation hard-ening than for literature data obtained at irradiation temperaturesof 260–326 �C. Furthermore, the decrease in uniform elongationdue to flow localization is much less in this work than observedfor the irradiation of Zircaloy at temperatures of nominally 260–326 �C. Less irradiation hardening has also been reported in the lit-erature [30] for the irradiation of Zircaloy at 326–450 �C than forirradiations at nominally 260–326 �C. These results indicate thehigher irradiation temperature of 358 �C results in lower irradia-tion hardening and less decrease in uniform elongation and ten-dency for flow localization.

TEM examinations of the microstructure for alpha-annealedand beta-treated Zircaloy-2 and Zircaloy-4 show that hai loopsare observed to form between an incubation fluence of 0.058–1.1 � 1024 n/m2, but the trends were very different for Zircaloy-4and Zircaloy-2 at fluences between 1.1 and 29.6 � 1024 n/m2. ForZircaloy-4, saturation to a relatively constant hai loop size andnumber density is observed at a low fluence of 1.1 � 1024 n/m2 thatis followed by either a small increase or decrease in hai loop sizethat is countered by a corresponding decrease or increase in hailoop number density so the net effect is that little change in thehai loop spacing occurs at higher fluences. This is consistent withthe irradiation hardening results observed for alpha-annealedand beta-treated Zircaloy-4 where the initial increase in irradiationhardening at a fluence of 1.1 � 1024 n/m2 is followed by a muchsmaller change in irradiation hardening with dose at fluences be-tween 1.1 and 29.3 � 1024 n/m2. For alpha-annealed and beta-trea-ted Zircaloy-2 an increase in hai loop number density with littlechange in hai loops size is observed at fluences between 1.1 and5.5 � 1024 n/m2 that is followed by an increase in hai loop sizeand corresponding decrease in hai loop number density betweenfluences of 5.5 and 29.3 � 1024 n/m2. This is consistent with the in-crease and then decrease in yield strength and irradiation harden-ing observed for both alpha-annealed and beta-treated Zircaloy-2.The hai loop size observed in the literature for Zircaloy at irradia-tion temperatures less than 316 �C were much smaller than ob-served in this work, and this finer loop size and higher loopnumber density is consistent with the lower irradiation hardeningmeasured in this work for the 358 �C irradiations. These resultsshow that irradiation at the higher temperature of 358 �C resultsin a larger hai loop size and lower hai loop number density that re-sults in lower irradiation hardening than observed for irradiationsat lower temperatures of 260–326 �C.

Irradiation of alpha-annealed and beta-treated Zircaloy-2 andZircaloy-4 at 358 �C in this work to the lowest fluence of0.058 � 1024 n/m2 results in measurable irradiation hardening,but TEM examinations of microstructure showed that no loopscould be observed that might explain these change in properties.Use of an Orowan hardening model with the TEM results for al-pha-annealed and beta-treated Zircaloy-2 and Zircaloy-4 at flu-ences between 1.1 and 29.3 � 1024 n/m2 gives calculatedhardening results that are correlated with the measured irradiationhardening values. The calculated irradiation hardening results alsomatch the trends in measured irradiation hardening as a functionof fluence for each material. Differences in calculated and mea-sured irradiation hardening values are likely the result of the for-mation of point defect or solute clusters smaller than the sizethat can be detected by TEM, or a change in the barrier strengthterms for the loops due to the segregation of alloying elementsto the loops during irradiation. The impact of such segregationon loop hardening, if present, also needs to be understood andquantified.

The irradiation hardening is primarily the result of hai loops,and the coarser distribution of hai loops formed at the higher irra-diation temperature of nominally 358 �C is the reason for the lowerirradiation hardening than observed for Zircaloy irradiated at tem-peratures of 260–326 �C. The formation of hci loops is observed atthe highest fluences in this work (29.3 � 1024 n/m2), but the hciloop number densities are relatively low and an inhomogeneousdistribution of hci loops is observed that is inherent to the lownumber density of these features being measured. A slightly bettermatch between the calculated and measured irradiation hardeningis obtained if hci loops are considered to have an effect on irradia-tion hardening with the hai loops. Basal plane slip is reported to oc-cur for irradiated Zircaloy so hci loops may have some effect onirradiation hardening, but additional results from higher fluenceirradiations where the accumulation of a higher number density


of hci loops is observed are needed to better understand the role ofhci loop formation on irradiation hardening at the irradiation tem-perature of 358 �C.

Acknowledgments

This work was supported by USDOE. The authors are grateful forthe review and comments provided by J.E. Hack and B.F.Kammenzind. Thanks also to the following ORNL personnel fortheir contributions in completing irradiations and testing (A.W.Williams and T.S. Byun). Irradiations were carried out in the HighFlux Isotope Reactor, a Department of Energy Office of Science UserFacility. Thanks to L.T. Gibson and M.S. Meyers for their work onTEM sample preparation. TEM examination was performed at theORNL User Center sponsored by the Division of Materials Scienceand Engineering, DOE.

References

[1] R. Adamson, B. Cox, ZIRAT-10 Special Topics Report, Impact of Irradiation onMaterial Performance, ANT International, Sweden, October, 2005.

[2] P.H. Kreyns, W.F. Bourgeois, C.J. White, P.L. Carpentier, B.F. Kammenzind, D.G.Franklin, in: E.R. Baradley, G.P. Sabol (Eds.), Zirconium in the Nuclear Industry:Eleventh International Symposium, ASTM STP 1295, American Society forTesting and Materials, 1996, pp. 758–782.

[3] M. Griffiths, R.W. Gilbert, V. Fidleris, R.P. Tucker, R.B. Adamson, J. Nucl. Mater.150 (1987) 159–168.

[4] W.J.S. Yang, R.P. Tucker, B. Cheng, R.B. Adamson, J. Nucl. Mater. 138 (1986)185–195.

[5] W.J.S. Yang, J. Nucl. Mater. 158 (1988) 71–80.[6] K. Kakiuchi et al., J. Nucl. Technol. 43 (2006) 1031–1036.[7] G.J.C. Carpenter, D.O. Northwood, J. Nucl. Mater. 56 (1975) 260–266.[8] N. Hashimoto, T.S. Byun, Mater. Sci. Forum. 561–565 (2007) 1769–1772.[9] K. Farrell, T.S. Byun, N. Hashimoto, J. Nucl. Mater. 335 (2004) 471–486.

[10] M. Griffiths, R.A. Holt, A. Rogerson, J. Nucl. Mater. 225 (1995) 245–258.[11] C. Regnard, et al. Proceeding of the 13th International Symposium on

Zirconium in the Nuclear Industry, ASTM STP 1423, 2002, p. 384.[12] D.O. Northwood et al., J. Nucl. Mater. 79 (1979) 379–394.[13] M. Griffiths, J. Nucl. Mater. 159 (1988) 190–218.[14] R.W. Gilbert, K. Farrell, C.E. Coleman, J. Nucl. Mater. 84 (1979) 137–148.[15] R. Bajaj, B.F. Kammenzind, D.M. Farkas, B-T-3351, DOE/OSTI, Oak Ridge, TN,

2001.

[16] M. Griffiths, R.W. Gilbert, G.J.C. Carpenter, J. Nucl. Mater. 150 (1987) 53–66.[17] M. Griffiths, J. Nucl. Mater. 170 (1990) 294–300.[18] M. Griffiths, R.W. Gilbert, C.E. Coleman, J. Nucl. Mater. 159 (1988) 405–416.[19] B.V. Cockeram, K.S. Chan, J. Nucl. Mater. 393 (2009) 387–408.[20] Standard Test Methods for Tension Testing of Metallic Materials, ASTM E8-01,

American Society for Testing and Materials, Philadelphia, PA, 2001.[21] A.L. Bement, J.C. Tobin, R.G. Hoagland, ASTM STP-380, June, 1964.[22] A.L. Bement, Radiation Damage in Hexagonal Close-Packed Metals and Alloys,

BNWL-SA-236, June 1, 1965.[23] B. Lustman, M.L. Bleiberg, E.S. Byron, J.N. Chirigos, J.C. Goodwin, G.J. Salvaggio,

Nucleonics 19 (1) (1961) 58–63.[24] D.G. Hardy, ASTM STP-484, June 1970.[25] H.H. Klepfer, C.N. Spalaris, Nuclear Metallurgy, vol. VIII, AIME, New York, 1960.[26] S.B. Wisner, R.B. Adamson, Nucl. Eng. Des. 185 (1998) 33–49.[27] P. Morize, J. Baicry, J.P. Mardon, in: R.B. Adamson, L.F.P. Van Swam (Eds.),

Zirconium in the Nuclear Industry: Seventh International Symposium, ASTMSTP 936, American Society for Testing and Materials, West Conshohocken, PA,1987, pp. 101–119.

[28] R.S. Ishimoto, T. Kubo, O. Kubota, Development of new zirconium alloys forhigh burnup fuel, in: Proceedings of TopFuel 2003, Wurzburg, Germany, 2003.

[29] S.T. Mahmood, D.M. Farkas, R.B. Adamson, Zirconium in the Nuclear Industry:Twelfth International Symposium, ASTM STP 1384, American Society forTesting and Materials, West Conshohocken, PA, 2000, pp. 139–169.

[30] S.B. Wisner, G.H. Henderson, R.P. Tucker, R.B. Adamson, Mechanical Propertiesof Zircaloy Irradiated in the EBR-II, GEAP-25163-10, Appendix B, 1984.

[31] T. Torimaru, T. Yasuda, M. Nakatsuka, J. Nucl. Mater. 238 (1996) 169.[32] F. Onimus, I. Monnet, J.L. Bechade, C. Prioul, P. Pilvin, J. Nucl. Mater. 328 (2004)

165–179.[33] M.S. Wechsler, Dislocation Channeling in Irradiated and Quenched Metals, The

Inhomogeneity of Plastic Deformation, ASM, Metals Park, OH, 1973.[34] G. Bertolino, G. Meyer, J.P. Ipiña, J. Nucl. Mater. 322 (2003) 57–65.[35] A.H. Yawny, J.P. Ipiña, J. Test. Eval. 31 (2003) 413–422.[36] T.J. Walker, J.N. Kass, Zirconium in the Nuclear Applications, ASTM STP 551,

American Society for Testing and Materials, 1974, pp. 328–354.[37] M. Griffiths, J. ASTM Int. 5 (2009) 19–26.[38] A. Jostons, P.M. Kelly, R.G. Blake, K. Farrell, ASTM STP 683, American Society for

Testing and Materials, West Conshohocken, PA, 1979, p. 46.[39] G.S. Was, Fundamentals of Radiation Materials Science, Springer, New York,

2007. pp. 581–640.[40] D.R. Olander, Fundamental Aspects of Nuclear Reactor Fuel Elements,

Technical Information Center, US-DOE, ERDA, 1976.[41] R.E. Stoller, S.J. Zinkle, J. Nucl. Mater. 283–287 (2000) 349–352.[42] U.F. Kocks, Metall. Trans. 1 (1970) 1121–1143.[43] C.C. Dollins, WAPD-TM-1305, dated March 1, 1978.[44] R. Bajaj, M.S. Wechsler, Metal Trans. 7A (1976) 351–358.

Documents

Development of microstructure and irradiation hardening of Zircaloy during low dose neutron irradiation at nominally 358 °C