iMORNL-3281,UC-80 —Reactor Technology

TID-4500 (17th ed.)

REVIEW OF ZIRCALOY-2 AND ZIRCALOY-4

PROPERTIES RELEVANT TO N.S. SAVANNAH

REACTOR DESIGN

C. L. Whitmarsh

I

OAK RIDGE NATIONAL LABORATORY

operated by

UNION CARBIDE CORPORATION

for the

U.S. ATOMIC ENERGY COMMISSION

?o

Printed in USA. Prico ULZ5 Available from the

Office of Technical Services

U. S, Deportment of Commerce

Washington 25, D. C.

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ORNL-3281

UC-80 - Reactor TechnologyTID-4500 (17th ed.)

Contract No. W-7405-eng-26

Reactor Division

REVIEW OP ZIRCALOY-2 AND ZIRCALOY-4 PROPERTIES

RELEVANT TO N.S. SAVANNAH REACTOR DESIGN

C. L. Whitmarsh

Date Issued

OAK RIDGE NATIONAL LABORATORY

Oak Ridge, Tennesseeoperated "by

UNION CARBIDE CORPORATION

for the

U. S. ATOMIC ENERGY COMMISSION

CONTENTS

Page

ABSTRACT v

1. INTRODUCTION 1

2. DESIGN CONSIDERATIONS 2

Environment 2

Zircaloy-2 Composition 3

Zircaloy-4 Composition 4

3. CORROSION 4

Water and Steam Corrosion 4

Kinetics 5

Factors Affecting Corrosion 9

Fretting 10

Hydride Effect 12

Hydrogen Solubility 12

Hydrogen Absorption 13

Hydrogen Diffusion 14

Mechanical Properties 16

Predicted Pickup of Hydrogen During Reactor Service 18

4. PHYSICAL PROPERTIES 20

5. MECHANICAL PROPERTIES 23

Tensile Properties 23

Modulus of Elasticity and Poisson's Ratio 29

Impact Properties 30

Fatigue Properties 32

Creep Properties 34

Torsional Properties 38

Fabricability 39

6. NUCLEAR PROPERTIES 40

7. RADIATION EFFECTS 41

Mechanical Properties 42

Hydrogen Pickup 47

Other Effects 47

In-reactor Failure 48

in

8. HAZARDS OF METAL-WATER REACTIONS 48

9. OTHER ZIRCONIUM BASE ALLOYS 52

REFERENCES 53

APPENDIX 59

IV

ABSTRACT

A literature search covering all available technology on Zircaloy

was conducted with specific emphasis on environmental conditions similar

to those present in the reactor core of the N.S. SAVANNAH. Information

on mechanical properties, corrosion properties, radiation effects, and

other pertinent subjects was collected.

Out-of-pile data relevant to Zircaloy-2 structural components in the

N.S. SAVANNAH core show a yield strength of -20 000 psi for annealed ma

terial, a creep rate of <10~5^/hr at 15 000 psi, and corrosion of only a

few mils in 20 years. The principal effects of radiation exposure are

increased mechanical strength and decreased uniform elongation. Brittle

fracture does not occur, however, because reductions in area permit local

plastic deformation. The corrosion rate is increased by radiation expo

sure (less than a factor of 5), but even at the increased rate metal loss

is insignificant. The possibility of a violent metal-water reaction, al

though remote, can be significant and must be evaluated for specific cases.

Except for hydrogen pickup, Zircaloy-4 exhibits essentially the same

properties as Zircaloy-2, as anticipated from the composition similarity.

Excessive hydrogen content, with resulting precipitation of ZrH2, causes

embrittlement of Zircaloy-2. The pickup of hydrogen by Zircaloy-2 is

about three times that of Zircaloy-4.

On the basis of the available information, Zircaloy-4 appears to be

adequate for use as the material of construction for the fuel element con

tainers in the advanced core for the N.S. SAVANNAH. Zircaloy-4 is recom

mended over Zircaloy-2 because of its lower hydrogen pickup. Recommenda

tions for 20-year service life cannot be made with complete confidence,

however, because of the necessary extrapolations of data. Creep, corrosion,

and irradiation data must all be extrapolated by factors of 10 to 15 to

obtain 20-year performance predictions. Existing data give no indication,

however, that any insurmountable problems will be encountered in these

areas for the application of interest.

The possibility of hairline cracks in Zircaloy-2 being propagated to

major defects by long-term radiation exposure deserves special mention.

Hairline cracks sometimes occur during fabrication of the metal that are

not always detected by present inspection methods. Effects of radiation

on these cracks are unknown, but the possibility of propagation to fail

ure has caused some designers to change from Zircaloy-2 to stainless steel

for fuel element cladding.

Zircaloy-2 is presently being used in various reactors as part of

the fuel matrix, fuel element cladding, and structural components.

Zircaloy-4, although not yet in use as a reactor material because of its

newness, has been recommended as the pressure-tube material in the de

sign of the CANDU reactor. Considerable experimental work pertinent to

reactor applications is being performed at various locations on zirconium-

base alloys.

VI

REVIEW OF ZIRCALOY-2 AND ZIRCALOY-4 PROPERTIES

RELEVANT TO N.S. SAVANNAH REACTOR DESIGN

C. L. Whitmarsh

1. INTRODUCTION

Preliminary considerations for reducing fuel-cycle costs in reactors

of the N.S. SAVANNAH type1 have indicated the feasibility of replacing

the stainless steel fuel element container with a container of a lower-

neutron-absorption material in order to reduce the required enrichment

or to increase the core life or both. Since neutron economy is one of

the more limiting design features, the list of applicable container ma

terials was quickly reduced to magnesium, aluminum, beryllium, and zir

conium or their alloys. Other criteria, such as mechanical strength,

corrosion resistance, fabricability, and available technology, further

reduced the list to Zircaloy-2 or Zircaloy-4. A literature search was

then conducted to collect the available information on the Zircaloys

pertinent to reactor design. The information thus collected is presented

in this report. Although most of the data reported here were obtained

from tests of Zircaloy-2, it is generally believed that the results, other

than those for hydrogen pickup, pertain equally well to Zircaloy-4 be

cause of the strong composition similarity between these alloys.

Existing reactors that use Zircaloy as fuel cladding or structural

components include the HRE-2, Dresden, Shippingport, submarine thermal

and advanced types, PRTR, NRU, and NRX. Those in advanced design or con

struction stages include the HWCTR, CVTR, NPD, and CANDU. Experimental

work on Zircaloy is being done primarily at Westinghouse's Bettis Plant,

Chalk River, Savannah River, Hanford, and Knolls Atomic Power Laboratory.

The standard reference for zirconium technology is the book, The

Metallurgy of Zirconium, which was published in 1955.2 Zircaloy-2 pro

perties were reviewed with respect to reactor technology and were pre

sented in a Hanford report in 1959.3 Additional data are scattered through

out the classified and unclassified literature.

2. DESIGN CONSIDERATIONS

Environment

A principal function of the fuel element container in the N.S.

SAVANNAH reactor is to channel cooling water through the core. The con

tainer, or can, also provides passageways for cruciform control rods, as

shown in Fig. 1. Mechanical and thermal stresses are of concern because

deflection of the container walls might interfere with control-rod move

ment. Deflection of the container walls is also dependent on the pres

sure differential of the cooling water between the second- and third-pass

channels and the modulus of elasticity of the material. Since the con

tainer is a structural component of the reactor, long-term usage is speci

fied. In this case, 20 years is the expected lifetime. Long-term use

enhances the importance of such factors as corrosion resistance and the

influence of radiation exposure on mechanical properties. The peak ex

posure to neutrons of energies greater than 1 Mev during a 20-year period

will be ~1023 neutrons/cm2.

ALUMINA FILLER

.'.IM1.'.; ZIRCALOY

'• ,' \ , -CONTROL ROD FOLLOWER

PER PHERAL FERRULE

!0663

',06666**666000 w

}00&Q:' 1 ){ n a x

SPACER FERRULE

-SPACER BAR(FLOW BLOCK)

FUEL ELEMENT

CONTAINER

UNCLASSIFIED

ORNL-LR-DWG 43889R

u

Fig. 1. Detail of Fuel Cell in the N.S. SAVANNAH Reactor,

A list of pertinent operating conditions is presented below:

Primary loop

Pressure, psig 1750

Water temperature, °F

Core inlet 496

Core outlet 520

Maximum 546

Water velocity, first pass, ft/sec 9.66

Water chemistry

pH range 7.5 to 8.5

Hydrogen content, ppm 3.6

Total maximum allowable solids, ppm 5

Maximum oxygen content 0.05

Maximum chlorine content, ppm 0.1

Zircaloy-2 Composition

Zircaloy-2 is a complex alloy of sponge zirconium with the follow

ing major components and impurities specified for reactor-grade material:3

Weight

Element

Major Components

Sn 1.2-1.7

Fe 0.07-0.20

Cr 0.05-0.15

Ni 0.03-0.08

Tmpur:ity Maxima

Al 0.0075

B 0.00005

C 0.0270

Cd 0.00005

Co 0.0020

Cu 0.0050

H 0.0025

Hf 0.0200

Pb 0.0130

Mg 0.0020

Mn 0.0050

N 0.0080

Si 0.0120

Na 0.0020

Ti 0.0050

W 0.0100

U (total) 0.00035U-235 0.0000025

It was developed specifically for nuclear reactor applications in a high-

temperature water environment. The alloying agents — tin, iron, chro

mium, and nickel — were added to sponge zirconium to neutralize the detri

mental effect on corrosion resistance of the impurities — nitrogen, alu

minum, and carbon — and for their strengthening effect. The low neutron-

absorption cross section of pure zirconium was not increased significantly

by these alloying materials.

Zircaloy-4 Composition

Zircaloy-4 was developed from Zircaloy-2 with the principal aim of

reducing the tendency to pick up hydrogen. Thus, the same composition

specifications are applicable, except for nickel, which is limited to

a maximum of 0.007^, and iron, the range of which is reduced to 0.12 to

0.1$fo.

3. CORROSION

Iodide-grade zirconium has excellent corrosion resistance in high-

temperature, high-pressure water; however, small amounts of impurities,

such as those present in commercially available sponge zirconium, reduce

the corrosion resistance drastically. Since iodide-grade zirconium is

both difficult and expensive to produce and fabricate, Zircaloy with

sponge zirconium as its major constituent represents a practical approach

to the problem.

Water and Steam Corrosion

Corrosion in water or steam produces a protective oxide film on the

surface of Zircaloy-2 by the reaction

Zr + 2H20 -> Zr02 + 2H2 .

Two distinct corrosion periods have been observed, pretransition and post-

transition. The pretransition period is characterized by a decreasing

corrosion rate and an adherent, protective film with a shiny black ap

pearance. The posttransition period is characterized by a constant but

greater corrosion rate and a less protective film that has a white flaky

appearance. The time required to reach transition is both temperature

and time dependent.

Kinetics

Pretransition film formation may be represented on a log-log plot

by a straight line:

log W = log k + n log t ,

where W is the weight gain per unit area, log k is the ordinate inter

cept, n is the slope, and t is the exposure time. The corrosion rate

during this period is thus seen to be time dependent:

dW , ,n-x— = nktdt

After transition, the corrosion reduces to a linear function; and, conse

quently, the corrosion rate is constant. Data2^;5 on Zircaloy-2 cor

rosion in water and steam at various temperatures are summarized in

Table 1 and Fig. 2. (The tests were made in static, neutral water and

in steam.) Additional data5 on the temperature dependence of the time

to transition are presented in Fig. 3. Rate-constant data for the post-

transition period (linear corrosion rate) are presented in Fig. 4.

In the extrapolation of these data to examine long-term effects,

corrosion during the pretransition period was found to be negligible,

except at the lower temperatures. For long-term corrosion prediction,

linear corrosion data were extrapolated as shown in Fig. 5. Both sets

of equations given in Table 1 were plotted for 750, 680, and 600°F; data

0^

Table 1. Corrosion of Zircaloy-2 in Water and in Steam2 >*">5

Pretransition PeriodTransition

PeriodPosttransition Period

Temperature

(°F)Corrosion

n • r, 4- a Weight m.Corrosion Rate „ . „ Time

GainaCorrosion8-

550

600

680

750^

600

680

750

572

log W=0.50 +0.30 log t |^ =0.95 t-0"70

log W = 0.74 + 0.26 log t — = 1.4 t-0-74

log W=0.76 +0.38 log t |^ =2.2 t-0-62

log W=1.10 +0.32 log t |H =4.0 t-0-68

dW

dt0.02c

w is in mg/dm2; t in days.

Steam at 1500 psig.

"Average rate during pretransition period.

34

34

41

40

1150 W - 34 = 0.065(t - 1150)

112 W - 34 = 0.37(t - 112)

41 w - 41 = 1.27(t - 41)

W = 0.02 t + 11

W = 0.03 t + 15

W = 1.4 t + 15

2000

Corrosion

Ratea

dW

dt= 0.065

^=0.37dt

dW

dt

dW

dt

1.27

= 0.02

dt

^ = 1.4dt

dW

dt0.03

<]

1000

UNCLASSIFIED

ORNL-LR-DWG 56352

REF 2

500 /REF 4

//

200

E 100

X

/' //T^

o>

E

<

t- "i" ^^^>

C5

S 20

10

-[£r"rT i

ZK>^^

6« 2>00^fc

5b0„, nN^>-

5

I 2 5 10 20 50 100 200

TIME (days)

500 1000

Fig. 2. Corrosion of Zircaloy-2 in Water andSteam.

UNCLASSIFIED

ORNL-LR-DWG 56353

TEMPERATURE (°F)

680 600

• /

•

9

</r(°R)

(xio4)

Fig. 3. Transition Time for Corrosion of Zircaloy-2 in Water andSteam.

03

20

E ,-ID LU

\ F-o> <

E a

10

a. 0.5

0.2

0.05

0.02

0.01

900

UNCLASSIFIED

ORNL-LR-DWG 56354

TEMPERATURE (°F)

800 680 620 550

\' 'EF: KAPL-2086

CI ASSIFIFr

\ p p 2 Fig 1

STEAM DATA ABOVE

pbi

^-ZIRCALOY- 2OR

,_ ZIRCALOY *•

\

\I I N

9

Vr (°R)

10 (xlO*)

Post-Transition Corrosion Rate of Zircaloy - 2 andZircaloy-4 in Water and Steam.

Fig. 4. Posttransition CorrosionRate of Zircaloy-2 and Zircaloy-4 inWater and in Steam.

8000

7000

3000 4000

TIME (days)

UNCLASSIFIED

ORNL-LR-DWG 56355

5000 6000 7000

Fig. 5. Predicted Long-Term Corrosion of Zircaloy-2 and Zircaloy-4 in Water and in Steam.

for 572 and 536°F are estimates made at Chalk River5>6 based on extrapola

tion of rate-constant data. It should be noted that data from Lustman2

were extrapolated from maximum test times of 1000 days. The assumption

was made that spalling of the corrosion film does not occur. Film thick

ness can be calculated from weight gain by use of the expression

X = 0.00252 W ,

where the film thickness X is in mils and the weight gain W is in mg/dm2

Metal penetration can be calculated from weight gain by use of the ex

pression

X = 0.00158 W ,m '

where the metal penetration X is in mils.r m

Factors Affecting Corrosion

The effects of various factors related to corrosion resistance are

summarized below.

Chemical Impurities. Impurities such as nitrogen, carbon, and alu

minium have a pronounced deleterious effect on the corrosion resistance

of Zircaloy-2. Specific effects of various impurities are treated in

detail in ref. 2.

Cold Work. Various studies2}7 have indicated that the cold work

effects are minor relative to other factors, for example, fabrication

methods, physical and chemical abnormalities, etc.

Heat Treatment. Cooling rates of greater than 90°F/min through the

range 1850 to 1470°F are considered necessary to minimize possible seg

regation of phases and a resultant corrosion rate increase.8;9 Annealing

at 1650 to 1850°F results in 20 to 80$ higher corrosion rates in 680°F

water or 750°F steam than annealing at lower temperatures. Work at the

Bureau of Mines3 has indicated that corrosion resistance is relatively

insensitive to heat treating temperatures below approximately 1470°F.

Fabrication. Instances of poor corrosion behavior of Zircaloy-2,

manifest by the appearance of "stringer"-type corrosion, have been traced

to improper arc-melting practice and heat treatment.10 Two types of

stringers — gas void and intermetallic — can occur near the metal sur

face and increase corrosion in that area.

Contamination. Contamination which would affect the water-corrosion

behavior of Zircaloy-2 may result from fabrication and processing, in

spection and corrosion-evaluation testing, and the reactor service en

vironment. In particular, fluorides introduced in etching procedures

are know to markedly increase the water-corrosion rate, apparently through

galvanic attack.3 Reactor cooling-water impurities could be important,

but no specific data are available.

Reactor Environment. Although many investigators have reported

that no significant effects on the corrosion rate of Zircaloy-2 in high-

temperature water have been observed that are attributable to radiation

flux, more recent data14' have indicated otherwise. Examination of Zircaloy-2

sheaths on irradiated fuel elements at Chalk River gave some evidence that

corrosion rates under irradiation may be higher by, at most, a factor of

5. Later results5 for Zircaloy-2 coupons tended to confirm an increased

rate, although the difference was much less than the factor of 5.

Water Velocity. No effect of water velocity in the ranges of in

terest on the corrosion rate of Zircaloy-2 in water has been observed.

Fretting

Fretting corrosion has been defined as the attrition of metal be

tween two contacting surfaces, in the absence of a corrosive environment,

as a result of small-amplitude vibrations under small loads.2 Although

few quantitative data are available, several general trends have been

observed. 5 Generally, in fretting, the relative velocity and amplitude

of motion are smaller than in normal wear processes; and the product formed

by fretting normally does not escape, since the two surfaces are never

out of contact. However, there is no sharp dividing line between fretting

and wear. Experiments have indicated the more important variables to

be contact pressure, number of cycles, amplitude of relative motion,

10

frequency, material hardness, temperature, coefficient of friction, and

environment. Most of these variables are interdependent, and care should

be used in treating any one as a separate factor.

Fretting of Zircaloy-2 probably occurs by the removal of metallic

particles which subsequently oxidize to form an abrasive powder. The

abrasive action of the powder is then regarded as being the more severe

cause of wear. Feng and Rightmire16-'17 assumed that fretting begins with

ordinary metal transfer and wear. Later accumulation of trapped wear

particles, which cannot escape because of the small amplitude of relative

motion, shifts the wear action gradually into abrasive action until,

finally, the major portion of the damage is due to abrasion.

Results of fretting tests at Bettis15 may be summarized as follows:

1. Fretting weight losses between Zircaloy surfaces were less than

those found between mild steel surfaces. (However, mild steel has poor

resistance to fretting corrosion.)

2. Fretting weight losses were less when tests were made in water

than in oxygen or air, particularly in the tests of Zircaloy against

stainless steel.

3. Fretting weight losses of Zircaloy against stainless steel were

a factor of 10 greater than in the case of Zircaloy-Zircaloy couples.

4. Slip versus fretting weight loss was not a linear function;

weight losses increased rapidly with relative slip of greater than 3 mils.

5. The effect of number of cycles and applied load on fretting was

similar to that found for mild steel; that is, it was nearly linear after

initial periods of greater effect.

Experience at Chalk River in the HRX reactor5 has indicated that

fretting of Zircaloy-2 is of little importance. During three years of

operation, a thick-walled (692 mils) Zircaloy-2 pressure tube was ex

amined internally eight times. The tube was found in very good condition

despite frequent mechanical damage to the surface film caused by the

movement of fuel charges. At least two reasonably authenticated instances

of fretting have occurred. In all instances the affected regions have

subsequently been found to be covered with a continuous black film. Thin

samples taken from the inner surface of the tube have a measured corrosion

11

of approximately 0.014 mg/dm2/day. (This is somewhat lower than that

predicted in Table 1.)

Hydride Effect

The primary source of hydrogen for the formation of ZrH2 is the cor

rosion reaction between zirconium and water. Absorbed hydrogen at a

concentration of 1000 ppm appears to have no effect on Zircaloy-2 cor

rosion resistance,18 but 10 to 50 ppm hydrogen is known to affect notch

sensitivity.3 Precipitation of the hydride phase in Zircaloy-2 increases

notch sensitivity and reduces ductility to such a degree that reactor

designers are now considering the use of Zircaloy-4 to minimize hydrogen

pickup. Based on information obtained at Chalk River,19 hydriding will

be the limiting factor for Zircaloy-2 pressure-tube service life in the

D20 moderated NPD and CAEDU reactors. Failure of two ZIrcaloy-2-clad U02

fuel rods similar to PWR Core 1 blanket fuel elements in a high-pressure

loop in the NRX reactor was attributed to absorption and thermal diffusion

of hydrogen in the Zircaloy-2.20

A comprehensive coverage of the effect of hydrogen on zirconium alloys

is presented in ref. 6.

Hydrogen Solubility

Solubility data6 for hydrogen in the alpha-phase of Zircaloy-2 are

plotted in Fig. 6. A least-squares fit of the data is represented by

C0 = 8.50 X 104 exp (-7600/RT) ,

where C0 is the terminal solid solubility of hydrogen in ppm by weight,

7600 represents the difference in partial molar heats of solution of

hydrogen in the hydride and alpha phases, R is the universal gas constant

(1.986 cal/mole-°K), and T is the absolute temperature in °K. Solubility

data from other sources are presented graphically in ref. 21.

12

5000

2000

COz>

UJo

oceQ

>I

tooo

500

200

100

UNCLASSIFIED

ORNL-LR-DWG 56356

• •

\V

\s»

\\

\

\.

\VN

\

504 5 6 7

Vr (°R)

9 (xKT

Fig. 6. Hydrogen Solubility in the Alpha-Phase of Zircaloy-2.

Hydrogen Absorption

In addition to the hydrogen initially contained in Zircaloy-2

(specification, <25 ppm), the principal sources of hydrogen in the re

actor environment are the radiolytic dissociation of water and the cor

rosion reaction between Zircaloy-2 and water. Various investigators4>

estimate that approximately 30$ of the hydrogen produced in the corrosion

reaction forms ZrH2, whereas others22 have found that up to 100$ of the

corrosion hydrogen is absorbed. Other studies5>23 indicate that a linear

relationship exists between hydrogen pickup and weight gain for a given

alloy (Fig. 7).

Hydrogen absorption in Zircaloy-2 during water corrosion is depend

ent on the hydrogen concentration in the water.6 Although the data are

somewhat scattered, a definite increase in the amount of hydrogen picked

up during corrosion occurs as the concentration of dissolved hydrogen is

increased. Some reported data24 indicate, however, that the increase in

hydrogen absorption is almost negligible at the N.S. SAVANNAH reactor con

dition of 25 to 35 cm3 of hydrogen per kg of water (-3 ppm). Other results

indicate that hydrogen absorption is sensitive to the pH of the water.25

13

UNCLASSIFIED

ORNL-LR-DWG 56357

20 40 60 80

WEIGHT GAIN (mg/dm2)100

Fig. 7. Hydrogen Pickup in Zircaloy-2 and Zircaloy-4 during Corrosion in Water.

Ammoniated water showed lower values than neutral or lithiated water

(Table 2). Sensitivity to heat treatment has also been shown, but results

are sufficiently scattered and scarce that predictions cannot be made

reliably.

Alloy composition has a significant effect on hydrogen pickup. As

the result of several investigations,6>23;26 nickel was determined to

be the principal hydrogen getter in Zircaloy-2. This led to the develop

ment of Zircaloy-4 (nickel-free Zircaloy-2), which has essentially the

same properties as Zircaloy-2 but a lower tendency to pick up hydrogen.

Hydrogen Diffusion

The diffusivities of hydrogen in alpha-phase zirconium and Zircaloy-2

have been determined by various investigators21'27-'28 and are presented

14

Table 2. Effect of Water Condition and Heat-Treatment on HydrogenPickup During Corrosion of Zircaloy-2 in 680°F Water

Heat Treatment

Time

(hr)Temperature

(°c)

Method

of

Cooling

1/2 775 Furnace

cooled

24 800 Furnace

cooled

1 850 Furnace

cooled

1 900 Furnace

cooled

24 900 Furnace

cooled

1 950 Furnace

cooled

1 1000 Furnace

cooled

1 850 Mercury

quenched

1 900 Mercury

quenched

1 950 Mercury

quenched

1 1000 Mercury

quenched

Hydrogen Pickup ($ of theoretical)

Water at

pH 7

Ammoniated

Water at

pH 9.0

Lithiated

Water at

pH 10.5

32.8 ± 4.13 23.2 ± 5.49 31.2 ± 3.62

18.4 ± 3.1 14.4 ± 1.24 21.6 ± 5.00

28.8 ± 5.54 23.2 ± 3.92 36.0 ± 5.46

32.8 ± 3.72 28.0 ± 7.93 40.8 ± 6.37

38.4 ± 2.53 28.0 ± 7.28 36.0 ± 12.5

24.8 ± 7.20 37.6 ± 11.98 28.0 ± 8.24

32.8 ± 9.26 28.0 ± 6.93 44.0 ± 9.50

27.2 ± 8.62 18.4 ± 3.39 28.0 ± 14.29

27.2 ± 6.47 20.8 ± 4.84 17.6 ± 8.46

21.6 ± 3.65 28.0 ± 4.34 17.6 ± 4.80

21.6 ± 6.57 23.2 ± 7.14 13.6 ± 4.06

in Fig. 8. Thermal diffusion occurs when a temperature gradient exists

in a Zircaloy-2 component and hydrogen concentrates in the cooler region.

In situations where large thermal gradients are present, diffusion of

hydrogen in Zircaloy-2 becomes a serious problem, as indicated by fuel

rod failures in loop experiments which were attributed to hydrogen em-

brittlement along the outer surface of the rods. Calculations of con

centration factors based on thermal diffusion theory have been made and

15

5x10'

UNCLASSIFIED

ORNL-LR-DWG 56358

TEMPERATURE (°C)

400

2.0 (x10=)

Fig. 8. Diffusivity of Hydrogenin Zircaloy-2.

16

plotted for various tempera

tures and thermal gradients.6

Work at KAPL4 has shown that

irradiation of Zircaloy-2-clad

fuel elements causes a redistri

bution of the hydrogen by dif

fusion, with concentration at

the coolest section, but the

redistribution is by no means

complete; that is, a hydride

gradient exists in the cladding

after irradiation. Diffusion

of hydrogen as the result of

stress and activity gradients has also been observed29 and, although not

clearly defined as yet, may make a significant contribution to hydrogen

redistribution in Zircaloy-2.

Mechanical Properties

The primary concern with hydrogen pickup in Zircaloy-2 is its effect

on impact strength, ductility, and, to a lesser degree, tensile strength.

Notched-bar impact properties of zirconium-base alloys are sensitive to

hydrogen content in that the transition temperature can be related to the

amount of hydride precipitate present.6 Quenching from above the hydrogen

solubility temperature lowers the transition temperature and increases

the transition range. Experimental data30 indicating decreased impact

resistance with Increased hydride precipitate are presented In Fig. 9.

Correlation between elongation and hydrogen content31 shows a definite

decrease in elongation with increasing hydrogen content (Fig. 10). Ten

sile properties are not so dependent on hydrogen content. Data31 from

Zircaloy-2 specimens with various heat treatments showed the tensile

strength to be essentially independent of hydrogen concentration up to

approximately 500 ppm; beyond this the strength was severely reduced (Fig.

11). Screening studies at KAPI34 indicated only small effects of hydrogen

on strain cycling and creep properties, and essentially no effect on

„ 100

UNCLASSIFIED

ORNL-LR-DWG 56359

-200 -100 0 100 200 300 400 500

TEMPERATURE (°C)

Fig. 9. Hydride Effect on Impact Properties of Zircaloy-2.

Fig.

UNCLASSIFIED

ORNL-LR-DWG 56360

REF: CRMet-849, p 50, Fig.17

NORMAL ZIRCALOY-2

O AS FABRICATED

a CORROSION TESTED

• IRRADIATED

NICKEL-BONDED ZIRCALOY-2

• AS FABRICATED

a CORROSION TESTED

• IRRADIATED

"B"o o oN

fD a \

\ • ft \

-«s.•9

B\

. "\~°\ \ ^

^

^-u-\

10 10 10

HYDROGEN CONTENT(ppm)

10. Effect of Hydrogen on Elongation of Zircaloy-2.

17

Fig.

(X10°)

80

•S 70

x 60

I 50Ld

CE

1- 40en

^ 30CO

S 20

10

0

UNCLASSIFIED

ORNL-LR-DWG 56361

REF: CRMet -849, p 50 , Fig. 16

NORMAL ZIRCALOY-2 NICKEL-BONDED ZIRCALOY-2

a AS FABRICATEDO AS FABRICATED

a CORROSION TESTED

• IRRADIATED

a CORROSION TESTED

• IRRADIATED

xx-roV^fr-

Ua*=L==£-=4%=*%—-°- \

-\\V

\ "\

V \

^«

^° 10' I02 103 104HYDROGEN CONTENT (ppm)

11. Effect of Hydrogen on Tensile Strength of Zircaloy-2.

10u

tensile properties at 600°F for hydrogen levels up to 500 ppm. Effects

of simultaneous hydrogen pickup and neutron irradiation have not yet been

clearly assessed, although limited data suggest no significant change In

expected behavior.

These data obviously are not extensive enough to reliably predict

tolerable hydrogen levels for long-term reactor operations. It appears,

however, that an upper limit of 500 ppm of hydrogen could be suggested

for wrought Zircaloy-2.4 Chalk River designers,5'19 taking a very con

servative approach, established the hydrogen solubility limit in Zircaloy-2

as a maximum tolerance level for the design of pressure tubes for D20-

moderated natural-uranium reactors. Their contention was that any pre

cipitated hydride would have a deleterious effect on the properties of

Zircaloy-2.

Predicted Pickup of Hydrogen During Reactor Service

Data on hydrogen pickup by Zircaloy-2 is generally in an unsatis

factory state,5 as indicated by the nonreproducibility of experimental

measurements and inadequacy of experimental methods. Investigators still

18

do not have a fundamental understanding of the mechanism of hydrogen pickup.

Any prediction can be no more than an extrapolation of available data,

and one must bear this in mind when evaluating the following information.

Calculated hydrogen pickup data* for Zircaloy-2 and Zircaloy-4

plates under design conditions of the N.S. SAVANNAH reactor are plotted

in Fig. 12. The assumption was made that the plates would be corrosion

^Detailed results are included in the Appendix (Table 2)

700

10 15 20 25

IN-REACTOR SERVICE (yeors)

UNCLASSIFIED

ORNL-LR-DWG 56362

Fig. 12. Prediction of Hydrogen Pickup by Zircaloy-2 and Zircaloy-4Plates in the Core of the N.S. SAVANNAH Reactor.

19

tested prior to use, although this contribution to hydrogen pickup was

almost negligible. The assumed Zircaloy-2 surface temperature of 572°F

was somewhat higher than the design temperature of 546°F but was one for

which data were available. Cool-region concentration of hydrogen was

neglected because of the very small thermal gradients existing in the

reactor. Corrosion rates of 0.02 mg/dm2/day were used for the pretransi

tion period and 0.03 mg/dm2/day for the posttransition period.5

Corrosion was considered to occur on both sides of the plates which

make up the fuel element container. Hydrogen pickup was estimated to be

30$ of the theoretical amount produced in the corrosion reaction during

corrosion testing and 100$ during exposure in the reactor. Out-of-pile

corrosion-rate data were used to calculate available hydrogen. This

method gives values in agreement with measured values of hydrogen content

in fuel sheathing.14 Zircaloy-4 data were calculated on the basis that

its hydrogen pickup5 was one-third that of Zircaloy-2.5

4. PHYSICAL PROPERTIES

Hot-rolled, vacuum-annealed Zircaloy has a density32 of 6.570 ± 0.0006

g/cm3 at 20°C and a melting point of 3360°F. Data33 on the coefficient of

thermal expansion as a function of temperature are presented in Fig. 13

(see also Table 1 of Appendix). The thermal expansion may be calculated

from the expression

L = L0[0.99983 + (5.628 X 106)T + (l.58l X 10"9)T2] ,

and then the coefficient of expansion corresponding to a particular tem

perature T can be obtained from the following equation:

^ dL (5.62 X 10-6) + (3.162 X 10"9)T ,L0 &P L0 dT

where T is in °C. The constants in these expressions were obtained for

25$ cold-rolled and annealed Zircaloy-2 plate. No anisotropy was de

tected, although work on zirconium34 had indicated otherwise.

20

< 4.0

'o

o 3.9CO

<Q.X

w 3.8

<

5en

£ 3.7F-

Li_O

F-

z 3.6LU

O

Li_Li_LU

o 3.5

3.4

2 3.3100 200 300 400 500 600

TEMPERATURE (°F)

700

UNCLASSIFIED

ORNL-LR-DWG 56363

800 900 1000

Fig. 13. Coefficient of Thermal Expansion of Zircaloy-2 as a Functionof Temperature.

Data on electrical resistivity are sensitive to specimen history.

The electrical resistivity of Zircaloy-2 at room temperature has been

reported2'4 as 29 X 10"6 ohm-in. and as 25 X 10"6 ohm-in. Temperature,

cold work, and impurity effects have not been well established.

Data on thermal conductivity are presented in Fig. 14. Considerable

variation exists in the data from the sources cited.2,3^33

No data on the specific heat of Zircaloy-2 were found; however, the

small alloy content would not be expected to effect a large deviation

from the values for zirconium. The data for zirconium are presented in

Fig. 15.

Two allotropic forms occur in Zircaloy-2: an alpha phase (close-

packed hexagonal) up to approximately 1480°F and a beta phase (body-

centered cubic) which is stable above approximately 1770°F to the melt

ing point, 3360°F. These transition temperatures are sensitive to heat

ing and cooling rates, and thus may vary as much as 50 to 100°F. Although

data are meager, there are reports of dimensional instability resulting

from thermal cycling through the transition range.2-*35

21

9.2

8,8

t 8-4

>-

|7.6oz>Q

I 72<

Scr

w 6.8

F-

UNCLASSIFIED

ORNL-LR-DWG 56364

*^

• REF 2

•

1 A REF 3 jO REF 33

--- ,^-A-A

1

o ^—

'^\0 100 200 300 400 500 600 700 800 900 1000

TEMPERATURE (°F)

Fig. 14. Thermal Conductivity of Zircaloy-2 as a Function of Temperature .

22

UNCLASSIFIED

ORNL-LR-DWG 56365

0.088

0.084

0.080

0.076

0.072

"s^—

-

u_

<rs

,

LJ

0_

/

!

200 400 600 800

TEMPERATURE (°F )

1000 1200 1400

Fig. 15. Specific Heat of Zirconium as a Function of Temperature.

5. MECHANICAL PROPERTIES

The mechanical properties of Zircaloy-2 are considered reasonably

good in a high-temperature water environment. These properties, are, how

ever, quite sensitive to specimen history. Even small variations in the

exact details of the fabrication procedure can affect the preferred ori

entation and the resulting mechanical properties.36 Impurities — oxygen,

nitrogen, and hydrogen — also significantly affect mechanical properties,

although the mechanism and degree are not always known. Radiation effects

on mechanical properties are discussed in Section 7.

Tensile Properties

Zircaloy-2 shows the typical stress-strain relationship of nonferrous

metals (Fig. 16). Since tensile properties are dependent on specimen

history, a so-called "base-annealed" condition (annealed at 1382°F in

vacuum for approximately 20 hr, and furnace cooled in vacuum) is used here

as a reference condition.

(xlCT)

100

80

60

UNCLASSIFIED

ORNL-LR-DWG 56366

/ ' ^ 50 7o COLD WORK

/1

^ __ ANNEALED

40

20

0.10 0.15

STRAIN (in./in.)

0.20 0.25

Fig. 16. Typical Stress-Strain Relationship for Zircaloy-2,

23

Averaged data showing the effect of temperature on tensile prop

erties36-42 and elongation are plotted in Figs. 17 and 18. Anisotropy

is clearly shown, with yield strength and ultimate strength generally

being affected in opposite direction.

A statistical study42 was performed on tensile-strength data from

a large number of samples of more than 100 different ingots. The spread

of the data, which was reported as a standard deviation, was attributed

primarily to measurement errors, heat treatment effects, and specimen

heating time. As-welded data were also reported because of their im

portance in stress analysis. The data from this study are presented in

Table 3.

Table 3. Tensile Strength of Zircaloy-2

Longitudinal Specimen Transverse Specimen As-Welded Specimen

At Room

TemperatureAt 600°F

At Room

TemperatureAt 600°P

At Room

TemperatureAx, 600CP

'iield strength at 0.2% 38 900 15 000 56 200 19 800 59 600 24 800

offset, psi

Standard deviation, psi 5 600 1 300 5 400 2 300 5 700 2 600

Ultimate Strength, psi 68 400 33 100 65 400 29 400 78 400 36 000

Standard deviation, psi 4 600 2 100 4 700 1 300 6 000 3 000

The tensile strength of Zircaloy-2 is increased (with consequent

reduction in ductility) when the material is cold worked, as shown in

Figs. 19 and 20. The da.ta39 A3~A5 were, in general, obtained from speci

mens with similar histories. Ratios were also calculated3 from averaged

data to show tne trend of tensile and ductility properties as a function

of cold work and temperature (Figs. 21 and 22). One study37 indicated

that a temperature of approximately 800°F was necessary before any signifi

cant loss of strength induced by cold work occurred with time (up to

500 hr). Initial cold working tended to reduce anisotropy in Zircaloy-2

plate and tubes.5 Hot-rolled plates were shown to have isotropic tensile

properties at 20fo cold work. Further cold working increased anisotropy

24

90,000

70,000

60,000

•S 50,000

ft 40,000

30,000

20,000

10,000

UNCLASSIFIED

ORNL-LR-DWG 56367

0 200 400 600 800 1000 1200

TEMPERATURE (°F)

Fig. 17. Effect of Temperatureon Longitudinal Tensile Properties ofAnnealed Zircaloy-2.

90,000

80,000

UNCLASSIFIED

ORNL-LR-DWG 56368

Ref: Zr DATA FILE, BULLETIN NO.5,

CARBORUNDUM METALS CO.

70,000

60,000

w^ 50,000

CO

K 40,000

30,000

20,000

YIELD STRENGTH^10,000

0

50

0 200 400 600 800 1000 1200

TEMPERATURE (°F)

Fig. 18. Effect of Temperatureon Transverse Tensile Properties ofAnnealed Zircaloy-2.

2.0

or

lu z

or h

2 Qo z

uo

< Q 160^ LU I-"Ll -I— <

c/> lu

</> lu 1.4

<CD

1.2

UNCLASSIFIED

ORNL-LR-DWG 56369

REF: HW-60908 ,p.63,Fig.

1

4

< •

i HYIELD

— STRENGTH

/? , o

^ .-,0

t

ULTIMATESTRENGTH

// c1

10 20 30 40

COLD WORK (%)

50 60 70

Fig. 19. Effect of Cold Work on the Ultimate and Yield Strength ofZircaloy-2 at Room Temperature.

m

o

lu zor oor r_

1.2

or o

z ° 0.6

or

0.4

0.2

\

6O

20

o

UNCLASSIFIED

ORNL-LR-DWG 56370

REF: HW-60908, p. 63, Fig. 14 A

REDUCTION OF AREA

30 40

COLD WORK (%)

50

TOTAL. ELONGATION

60 70

Fig. 20. Effect of Cold Work on the Elongation and Reduction of Areaof Zircaloy-2 at Room Temperature.

26

ro

H P c+ tl)

fjq CO

ro

o d-

H-

O H)

N H-

O P H O VI I !V>

P tn p o c+

H-

O P O o O H P^

« O tv P p p*

F3

n> i

DU

CT

ILIT

Y(F

RA

CT

ION

RE

FE

RR

ED

TO

BA

SE

-AN

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ED

CO

ND

ITIO

NA

TS

AM

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TE

MP

ER

AT

UR

E)

—C

O

O O

O ro

O ino C

O

//

-TO

REDUCTIONOFAREA,10-|COLDWORK/

-1-

1 4-

\

>/

r~1

m/

C c

3[

3|

>1

-H1

O\

C5

\n

—V

3\

o

c o

O

C C r r

\

WOR

\7

K

p p ^ F3

cd*x

j0

H-

WO

QC

D•

i-i

pro

c+

H£

•4 C

D•

F3

CD

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p en H-

H CD

o

O O

y CD

4 c+

H-

-k

CD

C)

en

-1

m

o

o5

Hj

um

N3

3>

(T)

-1

o4

c-

oO

Tl

Pm

H O0

•^n

1•—

'<

D

ro

o o

p (73

P ^rJ

o

P

o o

o c+

H-

O PIM o

Oo

Hd

o O H P"

s:

o 4 fV

TE

NS

ILE

ST

RE

SS

(FR

AC

TIO

NR

EF

ER

RE

DT

OB

AS

E-

AN

NE

AL

ED

CO

ND

ITIO

NA

TS

AM

ET

EM

PE

RA

TU

RE

)

but in the opposite direction; that is, the longitudinal yield strength

became greater than the transverse yield strength.

Heating of both hot-rolled and annealed Zircaloy-2 in the beta region

(>1770°F) causes roughly a 10% loss in tensile strength, a 20$ increase

in yield strength, a 40$ loss in elongation, and a 50% loss in reduction

in area.46 Heating within the alpha-beta range (-1480 to -1770°F) is

reported to have no effect on strength, although the ductility decreases

with increasing temperature. Quenching from the beta region, which is

typical of welding, provides increased yield strength. There is some

indication that Zircaloy-2 is susceptible to age hardening,3 with re

sulting strength increase and ductility decrease.

Oxygen, nitrogen, and hydrogen have an influence on tensile properties

of Zircaloy-2. Hydrogen effects were discussed in Section 3. Both oxygen

and nitrogen cause increased strength and decreased ductility 7 (Fig. 23).In fact, variations in tensile test results from several investigators

are often due to oxygen contamination of the test specimens.

Modulus of Elasticity and Poisson's Ratio

Modulus of elasticity data from several sources (refs. 3, 33, 37,

-41, 48, 4-9) are presented in Fig. 24. Reasonably good linearity with

temperature was established; the scatter was attributed to effects of

specimen history. Anisotropy of the elastic properties was evident. The

few data available on Poisson's ratio for Zircaloy-2 are presented in

Table 4.

Table 4. Elastic Properties of Zircaloy-2

28

Specimen

Condition

Annealed

Cold worked

Test

Temperature

(°C)

27

150

27

150

Modulus

of

Elasticity

X 106

13.9-14.3

13.3-13.8

14.5

13.9-14.2

Poisson's

Ratio

0.368-0.380

0.425-0.460

0.382-0.406

0.392-0.432

or

LU

LU

or

<

oCC0.

Soor

<

-120

NITROGEN CONTENT (wt%)

0.02 0.04 0.06 0.08

0.2 0.3 0.4

OXYGEN CONTENT (wt%)

UNCLASSIFIED

ORNL-LR-DWG 56373

0.12

0.6

Fig. 23. Effect of Oyxgen and Nitrogen on the Tensile Properties ofZirconium.

Impact Properties

Impact properties are of particular importance because of the pos

sibility of rapid crack propagation similar to that observed in ferritic

steels. Various studies5 have indicated that catastrophic crack propaga

tion will not occur in Zircaloy-2 at room temperature or above. Further

more, indications are that the decrease in energy to fracture with de

creasing temperature in impact tests is associated with the nucleation of

small cracks within the material rather than the rapid propagation of one

crack.

29

12

10

REF

49 ♦

CONDITION

48 0 STATIC

33 A LONGITUDINAL

33 A TRANSVERSE

3 V LONGITUDINAL

3 • TRANSVERSE

37 • ANNEALED

37 o COLD WORKED

41 • TRANSVERSE

41 • LONGITUDINAL

UNCLASSIFIED

ORNL-LR-DWG 56374

0 100 200 300 400 500 600 700 800 900 1000 1100

TEMPERATURE (°F)

Fig. 24. Modulus of Elasticity of Zircaloy-2 as a Function of Temperature .

Data on hardness and impact strength41'50'51 of Zircaloy-2 are listed

in Table 5 and plotted on Figs. 25 and 26. Beta quenching tends to im

prove the impact strength of Charpy V-notch specimens of Zircaloy-2.

30

Table 5. Impact Data on Zircaloy-2

Temperature

(°F)Specimen

Condition

Room As cast

Room Beta quenchedRoom Beta quenched

392 Beta quenched

Hardness

(DPH*)

161

180.4

DPH = Diamond-Pyramid Hardness,

Charpy V-NotchImpact Energy

(ft-lb)

22.5

51

240

200

160

120

200 400 600 800 1000

TEMPERATURE (°F)

UNCLASSIFIED

ORNL-LR-DWG 56375

1200 1400 1600 1800

Fig. 25. Diamond Point Hardness of Zircaloy-2 as a Function of Temperature.

Ref: Zr DATA FILE, BULLETIN NO. 5,

CARBORUNDUM METALS CO. :

LONGITUDINAL SPECIMEN

VERTICAL NOTCH

UNCLASSIFIED

ORNL-LR- DWG 56376

TRANSVERSE SPECIMEN

HORIZONTAL NOTCH

0 100 200

TESTING TEMPERATURE PC)

300 400

Fig. 26. Charpy Impact Data on Zircaloy-2.

31

Both oxygen and nitrogen would be expected to cause embrittlement, al

though quantitative data are not available. The effect of hydrogen con

tent was discussed in Section 3.

Fatigue Properties

Fatigue-test data32>A0>52~54 show that Zircaloy-2 is notch sensitive.

This effect is illustrated in Fig. 27, which also indicates the similarity

between the fatigue strengths of Zircaloy-2 and Zircaloy-4. Modification

of notch geometry in the range Kip = 3 to 9 does not affect fatigue strength

for lifetimes less than 106 cycles as shown in Fig. 28; KT is the theoretical stress concentration factor. The strain-fatigue properties of

Zircaloy-2 are relatively insensitive to temperature, orientation, and

stress concentration magnitude, as shown in Fig. 29.

(xlCT

40

30

20

10

UNCLASSIFIED

ORNL-LR-DWG 56377

Ref WAPD- ZH-23, p. 5, Fig.1

I

0

I

•-

•

~~ • ZIF

• ZIF

A ZIF

A ZIF

CALOY-2,

CALOY-4,

)CAL0Y-2,

JCALOY-4,

UNNOTCF

UNNOTCF

notche:

NOTCHEC

ED, KT=1ED, KT=, /fr = 3

), A>=3

1

1

5

I *-

^^ A

—• A f^r___

Kr IS THEOR ETICAL STRESS CONCENTRATION Ft \CT0R

10D 2 5 10°

CYCLES TO FAILURE

10'

Fig. 27. Reverse-Bend Stress-Fatigue Data for Base-Annealed Zircaloy-2

and Zircaloy-4 Tested at 600°F.

32

(X 10°)

10J 10°

CYCLES TO FAILURE

UNCLASSIFIED

ORNL-LR-DWG 56378

Fig. 28. Results of Reverse-Bend Fatigue Tests at 600°F and ZeroMean Stress on Base-Annealed Zircaloy-2.

10'

P 10u

O HOLE R.T. KT= 2.5• HOLE 600°F KT=2.5D "U" NOTCH R.T. KT=4.0• "U" NOTCH 600°F K?= 4.00 "V" NOTCH R.T. KT=>6.0^ "V" NOTCH 600°F KT=>6.0

HEE-NOTE- DATA BASED ON V4 inI ,— GAGE LENGTH _,_

10 ' 103

UNCLASSIFIED

ORNL-LR-DWG 56379

TRANSVERSE TO

ROLLING DIRECTION

HOLE R.T.

HOLE 600°F

"U" NOTCH R.T.

"U" NOTCH 600°F

"V" NOTCH R.T.

"V" NOTCH 600°F

KT = 2.5

KT = 2.5

KT = 4.0

KT = 4.4

KT = > 6.0

Kr = > 6.0

PARALLEL

ROLLING Dl

-T-t+H—-^r=R-IIHH i

ZIRCALOY-2-ROLLED (BASE ANN. CONDITION)

NOTCHED SPECIMENS (FLAT) .

.NO HC2 528 —• [ i

p. B-22,1SSIFIED

Tl WF TESTS, UNNOTCHED

M-TEMPERATURE TESTS, UNNOTCHED

105

CYCLES TO FAILURE

107

TO

RECTION

Fig. 29. Strain-Fatigue Properties of Zircaloy-2.

33

Effects of heat treatment on fatigue properties are indicated in

Table 6. The strain-fatigue properties of welded Zircaloy-2 are unaf

fected by base annealing in comparison with as-welded unnotched speci

mens at 600°F; however, stress-fatigue data for the same specimens show

somewhat better properties for the as-welded specimens.53 Room-tempera

ture tests indicate no difference between as-welded and base-annealed

specimens.

Table 6. Fatigue Properties of Zircaloy-2 at 600°F

Endurance Limit (psi) ,. Notch-SensitivitySpecimen _ ^T Index AtCondltion Unnotched Notched Ratl° 108 Cycles^

X 103 X 103

Beta quenched and 27.5 11.0 0.80 0.60alpha annealed

Base annealed 28.0 9.0 0.87 0.85

Fatigue ratio is the ratio of the unnotched fatigue strength tothe tensile strength.

Notch-sensitivity index = (Kf - l)/(Kt - l), where Kf = ratio ofunnotched fatigue strength to notched fatigue strength and K-p, = ratio oftheoretical maximum stress at notch to nominal stress.

Combined static and alternating stresses tend to decrease the over

all life of Zircaloy-2. This effect is shown on a modified Goodman dia

gram for a notched specimen with K^ = 9 in Fig. 30.

Creep Properties

The creep characteristics of Zircaloy-2 tend to be sensitive to ther-o/; Tr7 *^Q ^ ^

mal, mechanical, and chemical factors in the specimen history, >'.> >

as shown in Figs. 31 and 32. (In the figures in this section the parameters

high and low oxygen content refer to material derived from atmospheric-

melted ingots and from vacuum-melted ingots, respectively. Typical

atmospheric-melted Zircaloy contains 1100 to 1200 ppm of oxygen.) A

good summary of experimental results is available in ref. 3.

34

(x103)

25

Ref:WAPD-MRK-3, p.B-24,

B-3, Classified

4 = m

UNCLASSIFIED

ORNL-LR-DWG 56380

ZIRCALOY-2 BASE-ANNEALED

"V"-N0TCHED SPECIMENS, KT =

BENDING FATIGUE TESTS AT 600°F,HEAT 0M-160

10 15 20

MEAN STRESS (psi

25 30 (x103)

Fig. 30. Modified Goodman Diagram Showing the Combined Static andBending Fatigue Test Results at 600°F for Zircaloy-2.

5

UNCLASSIFIED

ORNL-LR-DWG 56381

Z» LONGITUDINAL TO ROLLING DIRECTIO

-A TRANSVERSE TO ROLLING DIRECTION

N 15°F

6— • 7

—A • — 1

4

1—— •—'. •

u-— •_400"F-J

-600°F —1

—i^:

2

4

•. ^?-*'• 700°F- ^_^^i

A'

A^—-

900° F "l'^"*

^kZZ^"

10"' 10

CREEP RATE (%/hr)

Fig. 31. Secondary Creep Rate of Zircaloy-2.

Primary creep strain in annealed Zircaloy-2 is almost entirely

stress sensitive, whereas secondary creep rates appear to be mainly tem

perature sensitive. The major mode of deformation in creep is apparently

by slip. In the case of 15$ cold-worked Zircaloy-2, the primary creep

strain is both stress and temperature sensitive, as shown in Fig. 33.

35

ON

60

50

'5 40

30

20

10

400

UNCLASSIFIED

ORNL-LR-DWG 56382

Ref: HW-60908, p. 82, Fig. 23

.COLD WORKED (10- 25%), LOWOXYGEN CONTENT i

500 600 700 800

TEMPERATURE (°F)

900 1000

Fig. 32. Effect of Oxygen on An

nealed and Cold-Worked Zircaloy-2 at aCreep Rate of 1 X 10-4 $/hr.

500

200

UNCLASSIFIEDORNL-LR-DWG 58521

100

REF: HW-60908, p. 80, FIG 21

V ANNEALED, 450°F, LOW 0

O ANNEALED, 550 °F, LOW 0

• ANNEALED, 650°F, LOW 0

A 15 7» CW, 550 °F, LOW 0J 15% CW, 650 °F, LOW 0

ANNEALED, 700 °F, HIGH 0

15% CW. 482°F. HIGH 0

•

D

• 15% CW fifi?°F HIGH 0

1

7 ir '

•/ •/ // •

»P •r i*i

/

i

—-/r.^ j- • rf a*1 / a P.. SO* •••/ _i,-A_A l-A .' /

/ kr^ - yf V

J x *>' •''" •'-// X y /2

* /

/ y

<or

LU

oro

>-or

<

50

20

10

0.5

0.2

0.1

10 20 30 40

STRESS (psi)

60 (x 103)

Fig. 33. Primary Creep Strain of Zircaloy-2.

Considerable work on creep testing of cold-worked Zircaloy-2 has been

performed in conjunction with reactor development at Chalk River5 and

Battelle37 which indicates that, at temperatures less than 650°F, cold

work of up to 25% essentially doubles the creep strength of Zircaloy-2.

Within the range 13 to 25%, the degree of cold working appears to have

very little effect on the creep rate of the cold-worked Zircaloy.

For long-term operation, it is necessary to know the time for onset

of the third stage of creep ($3). On the basis of present data, this

cannot be confidently done.3 Available information indicates a linear

relation between log 03 and stress for a given stress and test material,

as shown in Fig. 34. Impurity content and temperature are major variables.

Other data5 indicate that third-stage creep in 13 to 25%o cold worked

Zircaloy-2 at 300°C and 16 500 psi occurs at approximately 2%> total plastic

strain and that first-stage

creep represents approximately

10M

10°

10'

10

UNCLASSIFIED

ORNL-LR-DWG 56383

Ref: HW-60908, p. 84, Fig. 24

[O ANNEALED I 'GROUP A^»ANNEALED> HIGH OXYGEN CONTENT -

[a ANNEALEDj-0- B{-^-- --) LOW OXYGEN CONTENT

••««*»> ^ !& COLD ^S H'GH °*YGEN C0N™T

50 (X103)

Fig. 34. Time for Onset of Third-Stage Creep in Zircaloy-2.

0.08$ total plastic strain.

Thus, for a secondary creep

rate of 5 X 10"6$/hr (typical

for N.S. SAVANNAH design con

ditions), approximately 47 years

would be required to reach

third-stage creep.

Even though not considered

completely reliable, long-term

creep predictions3 have been

made that are based on experi

mental data (Fig. 35). The

principal uncertainties are

the previously mentioned third-

stage creep and the possibility

of recurrence of primary creep

under cyclic temperature, stress,

and radiation conditions of in-

reactor service. Data have

37

(xlO3

50

40

co 30

20

10

400

UNCLASSIFIED

ORNL-LR-DWG 56384

Ref: HW-60908, p.85, Fig. 25

500

O 15 7o COLD WORK, HIGH OXYGEN CONTENT

• 157o COLD WORK, LOW OXYGEN CONTENT

T ANNEALED,LOW OXYGEN CONTENT

600 700

TEMPERATURE (°F)

>10 years

800 900

Fig. 35. Long-Term Creep Predictions for Zircaloy-2,

been obtained in relaxation tests36 that indicate a 15 to 50% increase in

creep rate at 900°F.

Torsional Properties

Data obtained on the torsional properties56 of Zircaloy-2 are pre

sented in Table 7. One effect not evident is the lower hydrogen content

of the base-annealed material. (Base-annealing removes hydrogen from

the material; for example, 50 ppm in a hot-rolled specimen would be re

duced to 10 ppm.)

38

Table 7. Torsional Properties of Zircaloy-2

Specimen

Condition

Torsional Torsional Torsional Yield

Elastic

Modulus

Proportional

Limit (psi)Stress at 0.2$Offset (psi)

X 106

Base annealed 5.24 17 850 29 000

Base annealed 5.29 19 450 32 100

Hot rolled 5.22 13 550 29 500

Hot rolled 5.22 13 700 28 300

Fabricability

Zircaloy-2 has been fabricated by almost all the known metal-working

operations, although those operations requiring elevated temperatures

must be performed in an inert atmosphere to prevent oxygen and nitrogen

contamination. Extrusion, drawing, casting, and welding have been suc

cessfully done with Zircaloy-2. Welding techniques have been developed57

that do not require expensive gas-chamber setups. Obtaining a good weld

between Zircaloy-2 and stainless steel is difficult, however, because

of the large difference between the respective coefficients of thermal

expansion. Riveting is apparently a feasible joining method, but it has

the possible disadvantage of creating a more susceptible corrosion area.

The principal defect resulting from fabrication is the occlusion of

a gas void or intermetallic compound near the metal surface; such a region

is susceptible to excessive corrosion. This is usually called a "stringer"

defect. The practice of vacuum-melting ingots has greatly reduced the

incidence of "stringers."

Cracks have been observed in the cladding of newly fabricated fuel

elements, but the exact source of the trouble has not been determined.58

Occurrence of such defects is not always detected by the eddy-current

method of Inspection, and in certain cases the use of the more expensive

ultrasonic inspection procedure is being considered. No completely sat

isfactory nondestructive inspection method has been developed for detecting

defects less than 5 mils deep. In some cases,59 the cladding material is

39

being changed from Zircaloy-2 to stainless steel because of the possi

bility of major defects propagating from long-term radiation exposure of

these undetected hairline cracks.

During the latter phases of fabrication of the Shippingport reactor

fuel tubes,60 the over-all yield from ingot to finished mill product was

45$, and acceptability of the tubes was 98$.

6. NUCLEAR PROPERTIES

An important feature of Zircaloy-2 is its low thermal-neutron-

absorption microscopic cross section, which is reported to be 0.22 to 0.24

barn.32'41 Pure zirconium has a 0.l8-barn cross section; 0.015 to 0.02

barn is contributed by the alloying agents, and the remainder comes from

impurities in the metal. Cross-section dependence on energy is discussed

in ref. 61.

Natural zirconium has an atomic number of 40 and is composed of five

isotopes, the average atomic weight being 91.22 (Table 8). For neutron

energies less than 2 Mev, all reactions are (n,5) reactions, whereas at

higher energies, (n,p) reactions can occur.62 Neutron absorption yields

the stable isotopes Zr91 and Zr92 and the unstable isotopes Zr93, Zr95,

and Zr97, which undergo beta decay accompanied by gamma emission (Table

9).63 A few data are available on capture gamma rays64 which indicate

40

Table 8. Isotopic Composition of Natural

Zirconium

AbundanceIsotope

($)

Zr90 51.5

Zr91 11.2

Zr92 17.1

Zr94 17.4

Zr96 2.80

Table 9. Decay Data on Zirconium Isotopes63

T , Decay TT ,„ T._ Radiation and Energy8-Isotope ,, -, Half-Life /.. -,

* Mode (Mev)

Zr93 P,7 9 X 105 years p"(0.063)

Zr95 p,y 65 days P~(0.37,0.84),7(0.72,0.23)

Zr97 P,7 17 hours P"(1.91), 7(0.75)

Beta energies are maximum values; the average energyis approximately 40$ of the maximum energy.

that in approximately 18$ of the neutron captures in natural zirconium,

a gamma ray with an energy of 6 to 8 Mev is emitted.

7. RADIATION EFFECTS

Mechanical property changes have been the principal effects observed

upon neutron irradiation of Zircaloy. Although data are not available

for reliable predictions for a fast-neutron* dose of 1023 neutrons/cm2,

trends have been indicated by tests at lower exposures. Irradiations to

fast-neutron doses of up to 6 X 1021 neutrons/cm2 have caused an increase

in strength and a decrease in ductility. The data are scattered, how

ever, and consequently precise correlations with integrated flux are not

possible. Existing data have indicated saturation effects only on the

properties of hardness and electrical resistivity.

Most of the mechanical property data presented in this section were

obtained in pressurized-water reactors60;65 with a water temperature of

approximately 475°F and a fast flux of 0.75 to 5.11 X 1014 neutrons/cm2-sec,When available, fast-flux values are reported, because thermal-neutron

effects in metals are insignificant in comparison with those caused by

fast neutrons. Thus far, all in-reactor experience has been with Zircaloy

as fuel cladding.

*In this report, fast neutrons are defined as those having an energyof 0.65 Mev or greater.

41

Mechanical Properties

The stress-strain characteristics of Zircaloy-2 are altered by fast

neutrons in a manner similar to the effect of cold-work (Fig. 36). The

yield strength increases faster than the ultimate strength; the total

elongation is reduced; and the uniform elongation is reduced to almost

zero, as shown in Figs. 37 through 40. At an integrated flux level of

approximately 1021 neutrons/cm2, the yield strength and ultimate strength

are essentially equal. The data show, however, that sizeable reductions

in area (~20 to 40$) can still occur after a dose of 6 X 1021 neutrons/cm2

(Fig. 41). Thus, Zircaloy-2 appears to retain the ability to deform

locally in a plastic manner rather than by brittle fracture, in spite of

the reduction in uniform elongation.

Other postirradiation mechanical tests65 have shown hardness to be

dependent on total flux level (Fig. 42). Strain-fatigue tests60 showed

42

-l^l

UNCLASSIFIED

ORNL-LR-DWG 56385

Ref: WA PD-TM-184, p.15,F g. 2, Classified

i \

80

70

IRRADIATED (6.0 x102' neutrons/cm2)

^^\-^NONIRRAC IATED

60 /f'4 r \

50 ! 1 " ~ I— — -/

40

' ; SPECIMENS TESTED ATROOM TEMPERATURE

1

12 16 20 24

NOMINAL STRAIN (7.)

28 32

Fig. 36. Typical Stress-Strain Curve for Irradiated Zircaloy-2.

a reduction of only about 15$ in the bending stress of notched and un

notched 0.100-in.-thick plates after a fast-neutron exposure of 5.0 X 1021

neutrons/cm2. Data on the fatigue properties of irradiated Zircaloy-2

are presented in Table 10.

Table 10. Fatigue Properties of Irradiated Zircaloy-2

Unirradiated Specimens Irradiated Specimens6

tt 4. i_ j Notched TT , •, , NotchedUnnotched Tr _ Unnotched

Kt - 3 Kt = 3

15 500 21 500 13 000Failure stress at 106 24 000cycles, psi

Kf, strength reductionfactor0

q, notch sensitivityfactord

1.55

0.275

1.60

0.30

Fast-neutron exposure of approximately 5.0 X 1021 neutrons/cm2.

K, = theoretical stress-concentration factor.

Ratio of unnotched stress to notched stress.c

(Kf - l)/(Kt - 1).

(xtO

UNCLASSIFIED

ORNL-LR-DWG 56386

F 60

WAPD-TM-184, p. 22

. 19, Classified

TEMPERATURE OF TESTS

• ROOM TEMPERATURE

O 400°F

A 600°F

2 3 4 5

TOTAL INTEGRATED FAST FLUX (neutrons/cm2)

(x102<)

Fig. 37. Effect of Fast-Neutron Exposure on the 0.2$ Offset YieldStrength of Zircaloy-2.

43

(X10~

100

UNCLASSIFIED

ORNL-LR- DWG 56387

2 3 4 5 6 (X10 )

TOTAL INTEGRATED FAST FLUX (neutrons/cm2)

Fig. 38. Effect of Fast-Neutron Exposure on the Ultimate TensileStrength of Zircaloy-2.

40

UNCLASSIFIED

ORNL-LR-DWG 56388

1

<TEMPERATURE OF TEST

• ROOM TEMPERATURE

pi 30 o 400° F _i

2O

A 600°F

F-<

2 20

3LU

[ r 400° AND 600° Fi —r-i<

F-

R 10| -*-™« IEMPERATURE " <)

n

REF: WAPD-TM-184

i

, p 23, FIG. 21 , CLASSIFIED I! 1

2 3 4 5

TOTAL INTEGRATED FAST FLUX (neutrons/cm2)(X10'

Fig. 39. Effect of Fast-Neutron Exposure on the Total Elongationof Zircaloy-2.

Preliminary in-pi]e creep data4 indicate that the creep rate may

be somewhat reduced during neutron exposure. No effect has been observed

up to a fast-neutron exposure of 6 X 1021 neutrons/cm2 on the modulus of

elasticity.65 Impact properties are apparently dependent on a combina

tion of hydrogen content and flux exposure, although Charpy V-notch tests

showed no change in the brittle-to-ductile fracture behavior of Zircaloy-2

specimens containing 8 ppm of hydrogen that were irradiated to a fast-

neutron exposure of 1.9 X 1019 neutrons/cm2 in water at 150 to 200°C. A

44

20

E is'2 (O

<

O

0 10_J

LU

cr

0Li.

i 53

UNCLASSIFIED

ORNL-LR-DWG 56389

REF

\\ \

\\oA^8

WAPD-TM-184, p 23 , Fig. 22 ,

' TEMPERATURE• ROOM TEMPE

O 400° F

A 600° F

CLASSIFIED

OF TEST

RATURE

•

ROOM TEMPERATURE •

0 • * 400° AND 600" F J5A2 3 4 5

TOTAL INTEGRATED FAST FLUX (neutrons/cm2)( X10'

Fig. 40. Effect of Fast-Neutron Exposure on the Uniform Elongationof Zircaloy-4.

70 r

TEMPERATURE OF TEST

• ROOM TEMPERATURE

o 400°F

• 600°F J

UNCLASSIFIED

ORNL-LR-DWG 56390

TOTAL INTEGRATED FAST FLUX (neutrons /cm')

Fig. 41. Effect of Fast-Neutron Exposure on the Reduction in Areaof Zircaloy-2.

transition temperature increase of 54 to 63°F over that for the unirra

diated control specimen was found for a specimen with 130 ppm of hydro

gen. 66>67 More recent data5 indicate, however, that irradiation to doses

of up to 6 X 1021 neutrons/cm2 have no significant effect on the impact

properties of Zircaloy-2.

45

_,1(4)

"(6)

UNCLASSIFIED

ORNL-LR-DWG 56391

(6) fe^.

ROOM-TEMPERATURE TESTS

I INDICATES THE RANGE OF VALUES0 INDICATES THE NUMBER OF VALUES REPRESENTED

Ref: WAPD-TM-184, p. 22, Fig.18, CLASSIFIEDI I I L_

1 2 3 4 5 6 (X10 )

TOTAL INTEGRATED FAST FLUX (neutrons / cm2 )

Fig. 42. Effect of Total Fast-Neutron Exposure on the Hardness ofZircaloy-2.

Work at Chalk River5 has shown that (l) irradiation at temperatures

up to 716°F does not increase the rate of recovery of cold-worked

Zircaloy-2, (2) although the ductility of annealed Zircaloy-2 is consid

erably greater than that of the cold-worked material prior to irradiation,

there is little difference after irradiation, and (3) irradiation at 716°F

gives a small increase in mechanical properties in comparison with those

of material held at 716°F without irradiation. Additional data suggest

that Zircaloy-2 can be used at temperatures up to 752°F before thermal

softening becomes the dominant factor with respect to mechanical strength.

Postirradiation annealing has indicated that irradiation-induced

effects on hardness and tensile properties can be reduced considerably

by annealing, the annealing temperature and time being dependent on ir

radiation conditions.65 Recovery of properties by annealing appeared

to be insignificant at temperatures below 536°F, whereas the recovery was

quite pronounced at 716°F. The dependence of strength on exposure tem

perature3 indicates that nominal effects occur in the temperature range

240 to 450°F at ~3 X 1020 neutrons/cm2 (Fig. 43).

Reliable comparison of test data obtained by different investigators

is difficult because of the strong influence of specimen history on

mechanical properties. In addition, flux and energy-spectra data are

hard to obtain, and generally several experimental conditions differ in

46

100

60

40

100

UNCLASSIFIED

ORNL-LR-DWG 56392

TOTAL INTEGRATED FAST FLUX =~3>

1020 neutrons/cm2Ref: HW-60908, p. 89, Fig. 26

YIELD STRENGTH

200 300 400 500

EXPOSURE TEMPERATURE (°F)

600 700

Fig. 43. Effect of Irradiation Temperature on the Room-TemperatureTensile Properties of Zircaloy-2.

the various studies. A theoretical treatment of radiation-damage models

and considerable experimental information are presented in ref. 68.

Hydrogen Pickup

The effect on hydrogen pickup represents an indirect effect of ir

radiation exposure on mechanical properties, since the resulting hydrogen

embrittlement decreases the ductility and increases the notch sensitivity

of Zircaloy-2. This was discussed previously in Section 3. Zircaloy-2-

clad UO2 fuel rods have shown greater hydrogen pickup during reactor ser

vice than would be expected from out-of-pile experiments.25 Whether this

is attributable to irradiation, increased corrosion, or dissolved hydro

gen in the water has not been determined.

Other Effects

In-pile exposure of Zircaloy-2 components in water-moderated reactors

is usually accompanied by the formation of a surface layer consisting

primarily of iron oxide and water, with lesser amounts of other corrosion

products, usually referred to as crud. In-pile corrosion of the Zircaloy-2

and crud formation are definitely interrelated, although most of the iron

comes from corrosion of steel components in the reactor. Crud settles

out in low-velocity regions in the reactor's primary system and causes

47

objectionable situations by fouling heat transfer surfaces and creating

high activity levels. Long-lived crud activity is primarily due to Co58and Co60, which are formed by activation of Ni58 and Co59, respectively.Although these materials are present in the Zircaloys, the principal source

in existing reactors is usually stainless steel, Stellite, or Haynes No.

25 alloy components in the reactor.

The corrosion rate of Zircaloy-2 has not been observed to be signifi

cantly affected by irradiation by some investigators,11-13'24 but morerecent information5;14 has indicated a one- to fivefold increase in the

corrosion rate (see Sec. 3).

Buildup of fission products will occur in the primary system cool

ing water of a reactor as the result of uranium impurity in the Zircaloy

(nominal, 1.6 ppm; maximum, 3.5 ppm). Absorption of neutrons by U238results in the formation of fissionable Pu239 and Pu241. The activity

levels are not high enough, however, to restrict reactor operation.

In-reactor Failure

Fuel elements consisting of UO2 pellets clad with Zircaloy-2 have

been observed to have defects in the cladding after irradiation. Three

failures occurred in 1000 fuel segments irradiated in the VBWR and GETR

as part of a test program. The segments that failed had longitudinal

cracks along the high heat-flux portion of the cladding. Postirradiation

examination revealed severe hydride formation in the vicinity of the

cracks and heavy oxidation of the cladding internal surface. The possi

bility exists that defects of this type might rapidly self-propagate and

eventually lead to disintegration of the fuel element.

Several explanations for these failures have been proposed; for ex

ample, undetected fabrication defects, such as stringers, fission-product

damage, stress-corrosion cracking, hydrogen embrittlement, etc.

8. HAZARDS OF METAL-WATER REACTIONS

Although the probability of occurrence is very remote during re

actor operation, under certain conditions zirconium can react with water

48

(or steam) with explosive violence. Such conditions could occur if the

core overheated as the result of a loss-of-coolant accident. The rate

of heat generation from the Zr-H20 reaction is dependent on the heat of

reaction, rate of reaction, specific area of metal, mixing reactants,

total quantity of metal, and subsidiary heat, such as from combustion of

product hydrogen. From the standpoint of reactor environment, the most

important variable is the metal-particle size. Heat release from massive

metal reaction (ordinary corrosion in water) is insignificant, whereas

vaporized metal can react explosively. Core meltdown could produce par

ticles of sizes in the dangerous range, and, depending on core geometry

and heat transfer characteristics, the reaction could be self-propagating.

The problem has been studied by several investigators69"71 and all seem

to agree that, although the possibility of a Zr-H20 reaction is very re

mote, the extent of possible damage resulting therefrom warrants a care

ful study of each reactor.

Thermodynamic data72 on the reaction

Zr(liquid) + 2H20(steam) —> Zr02 + 2H2

are given in Table 11 and Fig. 44. This reaction can be represented by73

AW =k(T) tn(T) ,

k(T) = A exp (-Q/RT) ,

where

AW = oxygen weight gain, g/cm2,

k(T) = rate constant,

t = time,

n(T) = temperature dependent exponent,

A = constant,

Q = activation energy, cal/mole,

R = gas constant, cal/°K-mole,

T = absolute temperature, °K.

Calculated values of corrosion are presented in Fig. 45.

49

(X 10^40

30

20

10

UNCLASSIFIED

ORNL-LR-DWG 56394

! //

//

OXYGEN WEIGHT GAIN, ISW =k (r)/"m, /WHERE / IS TIME //

//BOSTROM'S DATA Ref. 72 //

~~ EXTRAPOLATIONS

-

/

/ »-

// /

y/

/

XX /

/ /

>s

--"'

1.00

0.80

0.60 5

0.40 ~

0.20

800 1200 1600

TEMPERATURE (°K)

2000

Fig. 44. High-Temperature Metal-Water Reaction Curves for Zircaloy.

Table 11. Thermodynamic Data of the Liquid Zirconium-Steam Reaction69^72

Temperature AHy,k(T) x 104 n(T)

°K °C °F cal/mole cail/g of Zr

1400 6.31 0.462

1500 1227 2241 -144 320 -1582

1600 1327 2421 -143 970 -1578 9.6* 0.552

1700 1424 2601 -143 660 -1575

1800 1527 2781 -143 280 -1571 15.2 0.66

1900 1627 2961 -142 990 -1568

2000a 1727 3141 -142 680 -1564 25.0 0.784

2100 1827 3321 -142 352 -1561

2200 1927 3501 -142 024 -1557 43.6 0.945

aFor temperatures above 2000°K, the data are extrapolated on thebasis that AEL decreases uniformly by 3.28 cal/mole-"K.

50

<

Fig

UNCLASSIFIED

ORNL- LR- DWG 56393

17, 300

1730

173

17.3

1.73

0.173

o_l

<

0.0173

CORROSION TIME (sec)

. 45. Oxidation Constants for Zirconium.

For precise evaluation of hazards associated with Zr-H20 reaction,

the present state of knowledge is inadequate. The principal deficiencies

are in the determination of the metal particle size and the understanding

of the heat transfer mechanism.69 However, calculations based on assumed

models for pressurized-water reactor accidents have been performed.70'7

In no case did the added heat of the metal-water reaction significantly

alter the over-all situation. Even if core meltdown occurred, the prob

ability of an extensive Zr-H20 reaction would be very low because heat trans

fer out of the core would be sufficient to inhibit the reaction. In a test

at Chalk River74 a Zircaloy-2-clad U02 fuel element was intentionally melted

in a pressurized-water loop and only a limited reaction was observed.

51

9. OTHER ZLRCONIUM-BASE ALLOYS

Development work on zirconium-base alloys is being performed by

many investigators.19;5 j75 The primary aim of most work to date has

been to increase mechanical strength while maintaining corrosion resist

ance and low-neutron-absorption properties. Although still in the devel

opment stage, the binary zirconium-niobium and the ternary zirconium-tin-

niobium alloys appear to be the most promising. For the application con

sidered here, however, the limited technology available on these alloys

prohibits their use.

Zircaloy-4 was developed to provide an alloy with greater resistance

to hydride formation and its resulting embrittlement. Information pre

sently available5'40 indicates no significant differences between the

properties of Zircaloy-4 and Zircaloy-2 at temperatures up to 572°F, ex

cept with respect to hydride pickup. The hydrogen pickup of Zircaloy-4

is substantially lower than that of Zircaloy-2 (Fig. 6). Kass and

Kirk23'26'76 report only 10 to 15$ corrosion hydrogen pickup by Zircaloy-4

as compared with 40 to 55$ by Zircaloy-2. Elimination of nickel (deter

mined to be a hydrogen getter) from Zircaloy-2 produced the desired ef

fects. Thus, Zircaloy-4 contains 1.2 to 1.7$ Sn, 0.12 to 0.18$ Fe, 0.05

to 0.15$ Cr, and the balance Zr, with a maximum of 0.007$ Ni. Since

nickel is insoluble in alpha zirconium, this change would be expected to

cause no effect on mechanical properties; and this has been found to be

true.

52

REFERENCES

1. G. E. Kulynych, Nuclear Merchant Ship Reactor, Final Safeguards

Report, Volume I, Description of the N.S. SAVANNAH, BAW-1164

(June 1960).

2. B. Lustman and F. Kerze, The Metallurgy of Zirconium, McGraw-Hill,

New York, 1955.

3. G. E. Zima, A Review of the Properties of Zircaloy-2, HW-60908

(Oct. 14, 1959).

4. R. J. Alllo, Behavior of Zirconium-Base Alloys and Hafnium in Long-

Lived Cores, KAPL-2088 (May 5, 1960). Classified

5. E. C. W. Perryman, A Review of Zircaloy-2 and Zircaloy-4 Properties

Relating to the Design Stress of CANDU Pressure Tubes, CRMet-937

(June 1960).

6. G. J. Biefer, L. M. Howe, and A. Sawatsky, Hydrogen Pick-Up In

Zirconium Alloys, A Review of Data up to June 1, 1959, CRMet-849

(September 1959).

7. Quarterly Progress Report - Fuels Development Operation for January,

February, and March, 1957, HW-49803 (Apr. 15, 1957). Classified

8. S. H. Bush, Properties of Zirconium and Zircaloy-2 and Alternate

Reactor Structural Materials for the DPR, HW-33248 RD (Sept. 30, 1954).

9. Bettis Technical Review, Reactor Metallurgy, WAPD-BT-2 (July 1957).-^——^—————— — — ••- — — • f

10. Bettis Technical Review, Reactor Metallurgy, WAPD-BT-10 (October 1958).

11. G. E. Galonian et al., Effect of Radiation on the Corrosion of

Metallic Materials in 580°F Water, KAPL-M-GEG-4 (Nov. 11, 1955).

12. B. Cox, The Effect of Solution Composition and Radiation on the High

Temperature Aqueous Corrosion of Zirconium and its Alloys, HARD(C)P 44

(July 1957).

13. J. K. Dawson, B. Cox, R. Murdock, and R. G. Sowden, "Some Chemical

Problems of Homogeneous Aqueous Reactors," Proceedings of Second United

Nations International Conference on the Peaceful Uses of Atomic Energy,

Geneva, 1958, Vol. 7, p. 234, United Nations, New York, 1958.

14. W. Evans and F. H. Krenz, The X-2 (t) Test, Examination of Hydrogen

Pick-Up in Specimen LR, Exp-NRX-2904 (February 1960).

53

15. Bettis Technical Review, Reactor Metallurgy, WAPD-BT-15 (September

1959).

16. I-Ming Feng and B. G. Rightmire, "An Experimental Study of Fretting,"

Proc. Inst. Mech. Eng. 170, 1055 (1956).

17. I-Ming Feng and B. G. Rightmire, "The Mechanism of Fretting,"

Lubrication Eng. 9, 134 (1953).

18. F. H. Krenz, G. J. Biefer, and N. A. Graham, "Chalk River Experience

with Zircaloy-2 and Aluminum-Nickel-Iron Alloys in High-Temperature

Water," Proceedings of Second United Nations International Conference

on the Peaceful Uses of Atomic Energy, Geneva, 1958, Vol 5, p 241—8,

United Nations New York, 1958.

19. AECL Industrial Symposium on Equipment Manufacturing and Development

Problems for Nuclear Power Systems, Chalk River, Ontario, April 19

and 20, 1960, AECL-990.

20. D. A. Vaughan and J. R. Bridge, "High Temperature X-Ray Diffraction

Investigation of the Zr-H System," J. Metals, 528-31 (May 1956).

21. A. Sawatzky, The Diffusion and Solubility of Hydrogen in the a

Phase of Zircaloy-2, CRMet-826 (February 1959).

22. K. M. Goldman and D. E. Thomas, Hydrogen Pick-Up during Corrosion

Testing of Zirconium and Zircaloy-2, WAPD-MM-184 (Feb. 19, 1953).

23. W. W. Kirk, "Corrosion and Hydrogen Pick-Up of Zircaloy-2 and

Zircaloy-4," pp 8—14 in Zirconium Highlights, WAPD-ZH-21 (December

1959).

24. J. D. Eichenberg, R. M. Lieberman, and F. P. Mrazik, Irradiation of

U02 Fuel Rods - The X-l-L Experiment, WAPD-208 (April 1960).

25. S. Kass, Report of October 1954 Meeting of the Zirconium Alloy

Corrosion Committee, WAPD-MM-713 (Nov. 30, 1954).

26. S. Kass and W. W. Kirk, "Corrosion and Hydrogen Evaluation of Nickel-

Free Zircaloy-2," pp 1-A- in Zirconium Highlights, WAPD-ZH-15 (March

1959).

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54

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44. A. B. Burgess and W. S. Kelly, The Effect of Neutron Irradiation at

100 C and 200 C on the Mechanical Properties of Cold-Worked Zircaloy-2,

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56

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58

APPENDIX

Table 1. Coefficient of Thermal Expansionfor Zircaloy-2a

Instantaneous Avf^e CoefficientTemperature Coefficient of of Expansion from

3F) Expansion )°F to Indicated

(in./in./°F) Temperature' ' J (in./in./°F)

X 10"6 X 10"6

3.24

3.33

3.42

3.50

3.59

3.67

212 3.30

392 3.48

572 3.65

752 3.82

932 4.00

1112 4.18

a

From ref. 34.

59

ONo

Table 2. Predicted Hydrogen Pickup rjy Zircaloy-2 Plates8,

Plat.e

.ess

Zircaloy-2Surface

Density

(mg/dm2)

Hydrogen from

Corrosion

Test

Total

Initial

Hydrogen

(ppm)

Hydrogen PickupTotal Hydrogi

(ppm)sn

Thickn10 yr 20 ;yr 30 yr

mils cm mg/dm2 ppm Mg/dm2 ppm Mg/dm2 ppm Mg/dm2 ppm 10 yr 20 yr 30 yr

X 106

75 0.919 0.126 0.375 3 28 22.3 177 49.8 395 77 611 205 423 639

100 0.254 0.167 0.375 2 27 22.3 134 49.8 298 77 461 161 325 488

125 0.318 0.209 0.375 2 27 22.3 107 49.8 238 77 368 134 265 395

140 0.356 0.234 0.375 2 27 22.3 95 49.8 213 77 329 122 240 356

150 0.382 0.251 0.375 1 26 22.3 89 49.8 198 77 307 115 224 333

163 0.415 0.273 0.375 1 26 22.3 82 49.8 182 77 282 108 208 308

Both surfaces of plate are exposed to water; Zircaloy-2 surface temperature = 572°Pj Zircaloy-2 corrosion rate(per surface) = 0.02 mg/dm2/day (pretransition); Zircaloy-2 corrosion rate (per surface = 0.03 mg/dm2/day (posttransition);corrosion transition time = 2000 days; initial hydrogen content of 25 ppm; water pH of 6.5—7.5 and hydrogen content of3.6 ppm.

ORNL-3281

UC-80 — Reactor TechnologyTID-4500 (17th ed.)

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26. V. 0. Haynes 60. A. M. Weinberg27. N. E. Hinkle 61. C. L. Whitmarsh

28. G. H. Jenks 62. M. L. Winton

29. W. H. Jordan 63-64. Central Research Library30. s. I. Kaplan 65-67. OREL - Y-12 Technical Library31. 0. H. KLepper Document Reference Section

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