2.07

Zirconium Alloys: Properties and Characteristics

C. Lemaignan
Commissariat a` l’E´nergie Atomique, Grenoble, France

ß 2012 Elsevier Ltd. All rights reserved.

2.07.1

Introduction

217

2.07.2
2.07.3
2.07.3.1
2.07.3.1.1
2.07.3.1.2
2.07.3.1.3
2.07.3.2
2.07.3.3
2.07.3.4
2.07.3.4.1
2.07.4
2.07.4.1
2.07.4.2
2.07.5
2.07.5.1
2.07.5.2

2.07.5.3
2.07.5.4
References

Physical Properties
Alloy Processing
Nuclear Grade Zr Base Metal
Ore decomposition
Hf purification and removal
Reduction to the metal
Alloy Melting
Forging
Tube Processing
Crystallographic texture development
Alloys
Alloying Elements and Phase Diagrams
Industrial Alloys
Mechanical Properties Before Irradiation
Strength and Ductility
Mechanical Properties in Temperature and Creep
Hydrogen Embrittlement and Other H Effects
Prospectives

218
219
219
219
219
220
220

220
221
221
222
222
227
228
228
229
229
230
231

Abbreviations
ASTM American Society for Testing Materials
BWR
Boiling water reactor
CANDU Canada Deuterium Uranium (heavy water
pressurized reactor)
DHC
Delayed hydride cracking
DSA
Dynamic Strain Aging
hcp
Hexagonal closed packed
HPUF
Hydrogen pick-up fraction
LOCA
Loss of coolant accident
MIBK

Methyl-isobutyl-ketone
PWR
Pressurized water reactor
RBMK Reaktor Bolshoy Moshchnosti Kanalniy
(pressure tube power reactor)
RIA
Reactivity induced accident
SPP
Second phase particle
TEM
Transmission Electron Microscope
TM
Transition metal
VVER
Voda-Voda Energy Reactor (PWR of
Russian design)

2.07.1 Introduction
Zirconium (Zr) exhibits a physical property of
uppermost importance with respect to the design
of in-core components of thermal neutron power
reactors: it has a very low thermal neutron capture
cross-section, and its alloys exhibit good engineering
properties. For an improvement in neutron efficiency
of the water-cooled reactors, the development of
industrial-type Zr-based alloys started as early as
the beginning of the nuclear reactor design, and is
still continuing. The engineering properties of Zr
and Zr alloys are therefore widely studied. Information exchanges and reviews are available in various
sources; for example, the International Atomic

Energy Agency issued reviews on Zr alloys for
nuclear applications. For more detailed, up-to-date
information, the reader is referred to a recent one,1 or
to the proceedings of the symposia on ‘Zr in the
nuclear industry,’ organized at 2–3 year intervals
by ASTM.2

217


218

Zirconium Alloys: Properties and Characteristics

It was found early that Zr is naturally mixed in its
ore with its lower companion of the periodic table,
hafnium, the latter being a strong neutron absorber.
Purification of Zr from Hf contamination is therefore
mandatory for nuclear applications. The development of the industrial alloys has been performed
following the classical route: searching for elements
of significant solubility that would improve the engineering properties, without too much impact on
the nuclear ones. Tin, niobium, and oxygen are the
main alloying elements, with minor additions of
transition metals (TMs) (Fe, Cr, and Ni). Heat treatments aiming at homogeneous solid solutions, phase
transformations, and precipitation control allow optimizing the structure of the alloys. In addition, the
thermomechanical history of the components strongly
impacts their behavior, via the formation of a crystallographic texture, because of the anisotropy linked to the
hexagonal crystallography of Zr at low temperature.
A few Zr alloys are commonly used for structural
components and fuel cladding in thermal neutron

reactors. Zircaloy (Zry)-4 is used in pressurized
water reactors (PWRs) and Zircaloy-2 in boiling
water reactors (BWRs). The heavy water-moderated
CANDU reactors, as well as the Russian VVER or
RBMK reactors, use Zr–Nb alloys. New alloys are
designed based on variants of the Zr–1% Nb, with
small additions of Fe and sharp control of minor
additions (M5®), or variants of the quaternary alloys,
such as Zirlo® and E635. More complex alloys with
other types of alloying elements are also being tested
in power plants, but the actual experience accumulated on these alloys is too low to consider them as
commonly accepted, from an industrial point of view.
Fuel claddings are made out of Zry-2 or Zry-4.
Those tubes have different geometries, depending on
reactor design. In PWR’s, the fuel cladding rods are
4–5 m long and have a diameter of 9–12 mm for a
thickness of 0.6–0.8 mm. BWR fuel rods are usually
slightly larger. The design is similar for the Russian
VVER, with Zr–1%Nb. In CANDUs, the fuel bundles
are shorter (0.5 m) and the cladding is thinner (0.4 mm)
in order to collapse very fast on the UO2 pellets.
Structural components of zirconium alloys are the
guide tubes, the grids, and the end plates that maintain the components of the fuel assemblies. They
have to maintain the structural integrity at the stress
levels corresponding to normal or accidental operations. In addition, they should have very low corrosion rates in the hot, oxidizing coolant water. In
BWRs, each assembly is surrounded by a Zircaloy-2
channel box that avoids cross-flow instabilities of

the two-phase coolant. Their geometrical stability is
a mandatory requirement for the neutron physics

design of the core.
In the case of CANDUs and RBMKs, the moderator is separated from the coolant water. The coolant
water in contact with the fuel rods is contained in
pressure tubes, usually made of Zr–Nb alloys. They
are large components (L $ 10 m, [ $ 30 cm, and
e $ 5 mm), with a design life expected to match the
reactor life, that is, tens of years, with only minor
corrosion and creep deformation.

2.07.2 Physical Properties
Natural zirconium has an atomic mass of 91.22 amu,
with five stable isotopes (90Zr : 51.46%, 91Zr: 11.23%,
92
Zr: 17.11%, 94Zr: 17.4%, and 96Zr: 2.8%). The
depletion of the most absorbing isotope (91Zr, with
sa $ 1.25 Â 10À28 m2) would increase further the interest of using Zr alloys in reactors, but would clearly be
economically inefficient. The cross-section for elastic
interaction with neutrons is normal, with respect to
its atomic number (sdiff $ 6.5 barn). Despite its high
atomic mass, the large interatomic distance in the hcp
crystals lead to a limited specific mass of 6.5 kg dmÀ3.
The thermophysical properties correspond to standard metals: thermal conductivity $22 W mÀ1 KÀ1
and heat capacity $280 J kgÀ1 KÀ1, that is, close to
3R per mole.
Below 865  C, pure Zr has an hcp structure, with
a c/a ratio of 1.593 (slightly lower than the ideal
1.633). The lattice parameters are a ¼ 0.323 nm and
c ¼ 0.515 nm.3 The thermal expansion coefficients
show a strong anisotropy, with almost a twofold difference between the aa and ac coefficients (respectively
5.2 and 10.4 Â 10À6 KÀ1).4 This anisotropic behavior of

the thermal expansion induces internal stresses due to
strain incompatibilities: After a standard heat treatment of 500  C, where the residual stresses will relax,
cooling down to room temperature will result in internal stresses in the range of 100 MPa, depending on
grain-to-grain orientations. The modulus of elasticity
is also anisotropic, but with lower differences than for
thermal expansion (Ea ¼ 99 GPa, and Ec ¼ 125 GPa).5
For industrial parts, the values recommended are
close to a $ 6.5 Â 10À6 KÀ1 and E $ 96 GPa. The temperature evolution of the elasticity constants is
unusual: the elasticity is strongly reduced as the temperature increases ($5% per 100 K).6,7 This abnormal
behavior is specific to the hcp metals of the IV-B row
of the periodic table.8


Zirconium Alloys: Properties and Characteristics

219

tons per year, out of which only 5% is processed
into zirconium metal and alloys.
The processing of Zr alloy industrial components
is rather difficult because of the high reactivity of
the Zr metal with oxygen. It consists of several steps
to obtain the Hf-free Zr base metal for alloy preparation: decomposition of the ore to separate Zr
and Si, Hf purification, and Zr chloride or fluoride
reduction.
50 mm

2.07.3.1.1 Ore decomposition

Three different processes are currently used for the

Zr–Si separation:
Figure 1 Microstructure of a b-quenched Zr alloys, with
a-platelets of four different crystallographic orientations
issued from the same former b-grain.

At 865  C, Zr undergoes an allotropic transformation from the low temperature hcp a-phase to the bcc
b-phase. On cooling, the transformation is usually bainitic, but martensitic transformation is obtained for
very high cooling rates (above 500 K sÀ1). The bainitic
transformation occurs according to the epitaxy of
the a-platelets on the old b-grains, as proposed by
Burgers9,10: (0001)a // {110}b and h1120ia // h111ib.
Among the 12 different possible variant orientations
of the new a-grains, only a few are nucleated out of a
given former b-grain during this transformation to
minimize the internal elastic strain energy. This process leads to a typical ‘basket-weave’ microstructure
(Figure 1). As a result, a b-quenching does not
completely clear out the initial crystallographic texture
that had been induced by the former thermomechanical processing.11,12 Although the alloying elements
present in the Zr alloys change the transformation
temperatures, with a 150  C temperature domain in
which the a- and b-phases coexist, the crystallographic
nature of the a-b transformation is equivalent to that of
pure Zr. Specific chemical considerations (segregations
and precipitations) will be described later.
The melting of pure Zr occurs at 1860  C, significantly above the melting temperature of other
structural alloys, such as the structural or stainless
steels. At high pressures, (P > 2.2 GPa) a low-density
hexagonal structure is observed, known as the o-phase.

2.07.3 Alloy Processing

2.07.3.1

Nuclear Grade Zr Base Metal

The most frequently used ore is zircon (ZrSiO4), with
a worldwide production of about 1 million metric

 In alkali fusion, where the zircon is molten in a
NaOH bath at 600  C, the following reaction takes
place:
ZrSiO4 þ 4NaOH ! Na2 ZrO3 þ Na2 SiO3 þ 2H2 O
Water or acid leaching allows the precipitation of
ZrO2.
 The fluo-silicate fusion:
ZrSiO4 þ K2 SiF6 ! K2 ZrF6 þ 2SiO2
It produces a potassium hexafluorozirconate which,
reacting with ammonia, leads to Zr hydroxide.
 The carbo-chlorination process is performed in a
fluidized bed furnace at 1200  C. The reaction
scheme is the following:
ZrO2 ðþSiO2 þ HfO2 Þ þ 2C þ 2Cl2
! ZrCl4 ðþSiCl4 þ HfCl4 Þ þ 2C
The controlled condensation of the gaseous tetrachloride allows the separation of Zr and Si, but not
of Hf from Zr.
2.07.3.1.2 Hf purification and removal

The processes described above separate Si from
Zr, but the Zr compounds remain contaminated
with the initial Hf concentration. The high neutron
capture cross-section of Hf (sa $ 105 barn, compared

to 0.185 barn for Zr) requires its suppression in
Zr alloys for nuclear application. Two major processes are used for this step: the MIBK-thiocyanate
solvent extraction and the extractive distillation of
tetrachlorides.
 In the first case, after reaction of zirconyl chloride
(ZrOCl2), obtained by hydrolysis of ZrCl4, with
ammonium thiocyanate (SCN-NH4), a solution
of hafnyl-zirconyl-thiocyanate (Zr/Hf )O(SCN)2
is obtained. A liquid–liquid extraction is performed with methyl-isobutyl-ketone (MIBK,


220

Zirconium Alloys: Properties and Characteristics

name of the process). Hf is extracted into the
organic phase, while Zr remains in the aqueous
one. Hf-free ZrO2 is obtained after several other
chemical steps: hydrochlorination, sulphation, neutralization with NH3, and calcination.
 In the dry route, after the transformation of
zircon into its chloride ZrCl4, through the carbochlorination process, Zr and Hf are separated using
a vapor phase distillation, at 350  C, within a
mixture of KCl-AlCl3, where the liquid phase is
enriched in Zr, and the vapor in Hf.
2.07.3.1.3 Reduction to the metal

The final step to obtain metallic Zr of nuclear grade
is to reduce the Hf-free Zr compounds that have been
obtained by the previous steps. Two processes are to
be considered at an industrial scale: the Kroll process

and the electrolysis.
 In the Kroll process, the Zr metal is obtained by
the reduction of ZrCl4 in gaseous form by liquid
magnesium, at about 850  C in an oxygen-free
environment. The following reaction occurs:
ZrCl4 ðgÞ þ 2MgðlÞ ! MgCl2 ðlÞ þ ZrðsÞ
After distillation of the remaining Mg and MgCl2,
under vacuum at 950  C, sintering of the Zr agglomerate at 1150  C gives the metallic sponge cake.
 After wet chemical chemistry, the reduction of the
ZrO2 obtained by the MIBK process is often performed by electrolysis. It is realized with the mixed
salt K2ZrF6 dissolved in NaCl or KCl at 850  C under
inert gas, with stainless steel cathode on which Zr is
deposited, and chlorine evolution at the graphite
anode. This route is mainly used in the Russian
Federation, the names of the Russian alloys starting
with an ‘E,’ referring to electrolytic processing.
High purity Zr can be obtained by the Van Arkel
process. It consists of reaction of Zr with iodine at
moderate temperature, gaseous phase transport as
ZrI4, and decomposition of the iodide at high temperature on an electrically heated filament. The iodine
released at the high temperature side is used for the
low temperature reaction in a closed loop transport
process, according to the following scheme:
Zr + 2 I2 => ZrI4 => Zr + 2 I2
250–300 ЊC

1300–1400 ЊC

This source of metallic Zr (called ‘iodide Zr’) is used
in Russia in addition to Zr obtained by the electrolytic


process for the melting of the alloys (typically 30%
‘iodide Zr’ in the first electrode to be melted).
2.07.3.2

Alloy Melting

Whatever the processing route followed for the production of Zr metal, the sponge or the chips obtained
by scrapping out the electrodes are the base products
for alloy ingot preparation. The melting of the alloys
is performed using the vacuum arc remelting (VAR)
process. This process is specific to highly reactive
metals such as Zr, Ti, or advanced superalloys.
For industrial alloy preparation, an electrode is
prepared by compaction of pieces of base metal fragments (sponge or scraps) with inclusion of the alloying elements. Typically, the elements to be added are
the following: O (in the form of ZrO2 powder), Sn,
Nb, Fe, Cr, and Ni to the desired composition. In
addition, a strict control of minor elements, such as C,
N, S, and Si, is ensured by the producers, at concentrations in the range of 30–300 ppm, according to their
requirements to fulfill the engineering properties.
A few specific impurities are strictly controlled for
neutron physics reasons: Cd and Hf due to their
impact on neutron capture cross-section, U for
the contamination of the coolant by recoil fission
fragments escaping from the free surface of the cladding, and Co for in-core activation, dissolution transport, contamination, and g-irradiation.
The compact stack is melted in a consumable
electrode electric vacuum furnace with water chilled
Cu crucible. Electromagnetic fields are often used for
efficient stirring of the liquid pool and reduced segregations. After three to four melts, the typical
dimensions of the final ingots are 0.6–0.8 m diameter

and 2–3 m length, that is, a mass of 4–8 tons.
2.07.3.3

Forging

Industrial use of Zr alloys requires either tube- or
plate-shaped material. The first step in mechanical
processing is forging or hot rolling in the b-phase, at a
temperature near 1050  C, or at lower temperatures
in the a þ b range or even in the upper a range. The
high oxidation kinetics of Zr alloys in air at high
temperatures restricts the high temperature forging
process to thick components, that is, with minimum
dimensions larger than 10 cm, at least. Final dimensions after forging correspond to 10–25 cm diameter
for billets and 10 cm for slabs.
A b-quenching is usually performed at the end of
the forging step. This heat treatment allows complete


Zirconium Alloys: Properties and Characteristics

dissolution of the alloying elements in the b-phase
and their homogenization above 1000  C, followed by
a water quench. During the corresponding bainitic
b to a transformation, the alloying elements are
redistributed, leading to local segregations: O and
Sn preferring the middle of the a-platelets, while
the TMs (Fe, Cr, and Ni) and Nb are being rejected
to the interface between the platelets.13 These segregations lead to plastic deformation strains highly
localized at the interplatelet zones for materials having

a b-quenched structure (heat-affected zones, welds, or
b-quenched without further thermomechanical processing). As described later, this b-quench controls the
initial size distribution of the precipitates in Zircaloy,
and further recovery heat treatments should be performed below the b–a transus only.
2.07.3.4

Tube Processing

For seamless tube production, first a hot extrusion is
performed in the temperature range of 600–700  C.
For pressure tube fabrication, this step is followed by
a single cold drawing step and a final stress relieving
heat treatment. For cladding tubes, the extrusion
produces a large extruded tube (‘Trex’ or ‘shell’), of
50–80 mm in diameter and 15–20 mm in thickness,
which is further reduced in size by cold rolling on
pilger-rolling mills.
After each cold working step of plate or tube
material, an annealing treatment is mandatory to
restore ductility. It is usually performed in the range
of 530–600  C to obtain the fully recrystallized
material (RX). The resultant microstructure is an
equiaxed geometry of the Zr grains with the precipitates located at the a-grain boundaries or within the
grains. The location of the precipitates at the grain
boundaries is not due to intergranular precipitation
but because they pin the grain boundaries during
grain growth (Figure 9). These different heat treatments contribute to the control of the cumulative
annealing parameter to be described below. For better
mechanical properties of the final product, the temperature of the last annealing treatment can be
reduced to avoid complete recrystallization. This is

the stress-relieved (SR) state, obtained with final heat
treatment temperature of 475  C, which is characterized by elongated grains and a high density of dislocations, and by relief of the internal stresses,
leading to a greater ductility than cold-worked materials. It is mostly applied to the PWR claddings, while
for BWRs, a complete recrystallization is performed
at 550–570  C.

221

2.07.3.4.1 Crystallographic texture
development

Two plastic deformation mechanisms are operating
during low temperature deformation of the Zr alloys:
dislocation slip and twinning. As reviewed by
Tenckhoff,14 the most active deformation mechanism
depends on the relative orientation of the grain in the
stress field.
Dislocation slip occurs mostly on prism plane with
an a Burgers vector. It is referred to as the {1010}
h1210i, or prismatic, system. The total strain imposed
during mechanical processing of the Zr alloys cannot,
however, be accounted for only with this single type
of slip, as the different orientations of the crystal
would only give two independent shear systems.
At high deformations, and as the temperature is
increased, (c þ a) type slip is activated on {1121} or
{1011} planes. These are the pyramidal slip systems,
having higher resolved shear stresses (Figure 2).
Different twinning systems may be activated
depending on the stress state: for tensile stress in

the c-direction, {1012} h1011i twins are the most
frequent, while the {1122} h1123i system is observed
when compression is applied in the c-direction.
The resolved shear stresses of the twin systems have
been shown to be higher than the one necessary for
slip, but due to the dependence of the Schmid factor
on orientation, twinning is activated before slip,
for some well-oriented grains. Therefore, there are
five independent deformation mechanisms operating
in each grain, and thus the von Mises criterion for
grain-to-grain strain compatibility is fulfilled.
At the large strains obtained during mechanical
processing, steady-state interactions occur between
the twin and slip systems that tend to align the
basal planes parallel to the direction of the main
deformation.15,16 For cold-rolled materials (sheets or
tubes), the textures are such that the majority of the

Figure 2 The two Burgers vectors (a and c þ a) for strain
dislocations in Zr alloys, and the two slip planes (prismatic
and pyramidal) in hcp a-Zr.


222

Zirconium Alloys: Properties and Characteristics

AD

AD


90Њ

90Њ

80Њ

80Њ

60Њ

60Њ

30Њ

30Њ

5

5

1
1

2

3

4 5


6

3 0Њ

TD
6 54 3

2

2

fR = 0.64
(a)

35
3

4

4Њ
0Њ

TD
3

4

fR = 0.55
(b)


Figure 3 h0001i Pole figure of two cladding tubes with slightly different mechanical processing routes.

grains have their c-axis tilted 30–40 away from the
normal of the foil or of the tube surface toward the
tangential direction, as can be seen in the h0001i pole
figure of a cladding tube (Figure 3).
During tube rolling, the spread of the texture can
be reduced by action on the ratio of the thickness to
diameter reductions (Q factor): a reduction in thickness higher than the reduction in diameter gives a
more radial texture, that is, a texture with the c poles
closer to the radial direction.16
After cold processing, the h1010i direction is parallel to the rolling direction, and during a recrystallization heat treatment a 30 rotation occurs around the
c-direction and the rolling direction is then aligned
with the h1120i direction for most of the grains.

2.07.4 Alloys
2.07.4.1 Alloying Elements and
Phase Diagrams
Like any metal, pure Zr exhibits rather poor engineering properties. To improve the properties of a
given metal, the metallurgical engineering procedures are always the same: It consists in finding additions, any species of the periodic table could be
considered, with significant solubility, or heat treatments producing new phases that could improve the
properties. The relative solubility of the various alloying elements in the a- and b-phases is therefore one
basis for the choice of additions, as well as for developing the heat treatments, for microstructure control.
For the nuclear applications, neutron physics
requirements restrict the possibilities, by rejection

of the isotopes having high interaction cross-sections,
or isotopes that would transmute to isotopes of high
capture cross-section or having high irradiation
impact (Co). Elements such as Hf, Cd, W, and Co

have therefore not been considered for alloy developments. With low nuclear impact, O, Sn, and Nb
have been selected (Al and Si having also low nuclear
impacts were not retained because of degradation in
corrosion resistance), while other TMs (Fe, Cr, Ni,
etc.) can be accepted up to limited concentrations
(below 0.5% total).
The additions have to improve the engineering
properties. The main properties to be improved are
the corrosion behavior in hot water and the mechanical strength (yield stress, ductility, and creep). As
described below, Sn and Nb are added for corrosion
resistance, and elements forming secondary phases
(Nb and Fe, Cr, and Ni) or solid solutions are also
used for increasing the mechanical properties.
Last, the microstructure obtained after the thermomechanical processing should not change without
control under irradiation. Therefore, hardening
obtained by precipitation or strain hardening can be
considered only if the irradiation-induced evolution
of the initial microstructure will be compensated by
the development of irradiation-induced microstructural defects. In this respect, the evolution of precipitates in Zircaloys is of high importance for
corrosion behavior and geometrical integrity. These
points are discussed in Chapter 5.03, Corrosion of
Zirconium Alloys and Chapter 4.01, Radiation
Effects in Zirconium Alloys.
Most of the binary phase diagrams with Zr are
already known and many ternary or higher-level


Zirconium Alloys: Properties and Characteristics

diagrams of industrial interest are now known.17 The

need for a better control of the processing of the
current alloys and the aim of finding new alloys and
structures without too much experimental work have
been a driving force for the modern trend in numerical simulation for material science. It is now also
possible to extrapolate the binary data to multicomponent systems. In that respect, a thermodynamic
database for Zr alloys, called ZIRCOBASE, has
been developed under the Calphad methodology.18
This database contains 15 elements and is frequently
updated. The most complex ternary or quaternary
phase diagrams available are optimized or computed
using this database, and, in the case of missing basic
thermodynamic data, with the contribution of ‘abinitio’ computations.19 The phase diagrams presented
in this review were obtained according to this
procedure.
Oxygen is highly soluble in the a-phase, and stabilizes at high temperature (Figure 4). Oxygen has
to be considered as an alloying element. This use of
oxygen for strengthening is rare in metallurgy, compared to the use of nitrogen. However, the use of
nitrogen for strengthening would severely deteriorate
the corrosion resistance, and nitrogen is removed as
much as possible. The purpose of oxygen additions
is to increase the yield strength by solution strengthening, without degradation of the corrosion resistance. The O content is not specified in the ASTM
standards, but usually it is added to concentrations in
the range of 600–1200 ppm, and this has to be agreed
between producer and consumers. High O concentrations (O > 2000 ppm) reduce the ductility of the
alloys; therefore, O additions above 1500 ppm are
not recommended. In addition, O atoms interact
with the dislocations at moderate temperatures,

223


leading to age-strengthening phenomena in temperature ranges depending on strain rate.20 The oxygen
in solid solution in a-zirconium is an interstitial in
the octahedral sites. In the Zr–O system, the only
available stable oxide is ZrO2. A monoclinic phase
is stable at temperatures up to about 1200  C,
above which it transforms to a tetragonal structure.
The impact on corrosion of the different phases of
ZrO2, according to temperature and pressure is discussed in Chapter 5.03, Corrosion of Zirconium
Alloys.
Tin tends to extend the a-domain, and has a
maximal solubility in the hcp Zr of 9 wt% at 940  C
(Figure 5). It was originally added at concentrations
of 1.2–1.7% to increase the corrosion resistance,
especially by mitigating the deleterious effects of
nitrogen. The amount of Sn needed to compensate
the effect of 300 ppm of N is about 1% of Sn. However, in N-free Zr, Sn has been observed to deteriorate the corrosion resistance. Therefore, the modern
trend is to reduce it, but only slightly, in order to
maintain good creep properties.21
Iron, chromium, and nickel, at their usual concentrations, are fully soluble in the b-phase (Figure 6).
However, in the a-phase, their solubilities are very
low: in the region of 120 ppm for Fe and 200 ppm for
Cr at the maximum solubility temperature.22 In the
pure binary systems, various phases are obtained:
ZrFe2 and ZrCr2 are Laves phases with cubic or
hexagonal structure, while Zr2Ni is a Zintl phase
with a body-centered tetragonal C16 structure.
These precipitates are called the Second Phase Particles (SPPs).
In the Zircaloys, the Fe substitutes for the
corresponding TM and the intermetallic compounds
found in Zircaloy are Zr2(Ni,Fe) and Zr(Cr,Fe)2.

2000

g-ZrO2

1800

b-Zr
1500
a-Zr
1000

aЈ

500
aЈЈ®
1

0

10

a2ЈЈ

20

¬ aЈЈ
4
3 ¬ aЈЈ

30

40
50
Atomic percent oxygen

Figure 4 Zr–O binary phase diagram.

b-Zr

1400

b-Zr + Zr5Sn3

1200
1000
800

a-Zr

Zr4Sn

b-ZrO2

2000

Temperature (ЊC)

1600

a-ZrO2


Temperature ( ЊC)

L

L

2500

600
400
200

60

70

0

5

10
15
20
Atomic percent tin

Figure 5 Zr–Sn binary phase diagram.

25

30



224

Zirconium Alloys: Properties and Characteristics

2000
1800

Temperature (ЊC)

1600

L

1400

b-Zr

1200
1000
800
¬

FeZr3

FeZr2

600 ¬ a-Zr
400

200
0

(a)

5

10

15
20
25
Atomic percent iron

1800

¬

g-Cr2Zr

1600

1600

b-Zr

Temperature (ЊC)

¬


1400

a-Cr2Zr

¬

Temperature (ЊC)

40

1200
1000

¬ a-Zr

1400

L

¬ b-Zr

1200
1000
800

600

600 ¬ a-Zr

400


400

200

NiZr2

L

1800

200
0

(b)

35

2000

2000

800

30

10

20


30

40

50

60

70

Atomic percent chromium

0

(c)

5

10

15

20

25

30

35


40

Atomic percent nickel

Figure 6 Zr-rich site of the Zr-transition metal binary phase diagram: (a) Zr–Fe, (b) Zr–Cr, (c) Zr–Ni.

The formation of these precipitates, and more
complex ones in industrial alloys, is analyzed in
detail for the control of the corrosion behavior of
the Zircaloys. Indeed, a strong correlation has been
observed between precipitate size distributions and
corrosion kinetics, the behavior being opposite for
BWRs and PWRs. A better uniform corrosion resistance is obtained for Zircaloys used in PWRs if they
contain large precipitates, while better resistance to
the localized forms of corrosion is seen in BWRs in
materials that have finely distributed small precipitates.23,24 With an increase in the particle diameters
from 0.05 to 0.1 mm or higher, the in-pile corrosion of
Zircaloy cladding diminishes appreciably. However,
nodular corrosion may occur in BWR cladding with a
further increase in the particle diameters above about
0.15 mm25 (Figure 7).
Due to the low solubility of the transition metals
(Fe, Cr and Ni) in the Zr matrix, coarsening of the
precipitates, after the last b-quench, occurs at very

low rates, during the intermediate annealing heat
treatments, following each step of the rolling process.
Therefore, the precipitate growth integrates the thermal activation times of each recovery, and their temperatures and durations can be used to control the
size of SPPs. This integrated coarsening activation
time is referred as the ‘A ’ or ‘SA ’ parameter.

The A-parameter calculates the integral of the
activation processes for the different anneal durations
and temperatures. The annealing parameter is
defined as A ¼ Si (ti exp (ÀQ/RTi), where ti is the
time (in hours) of the ith annealing step, at temperature Ti (in K); Q/T is the activation temperature of
the process involved. The activation energy for the
process should have been taken as the one controlling
the coarsening, that is, the diffusion. However, as the
early studies were undertaken with the aim of
improving the corrosion resistance, an unfortunate
practice has been induced to take 40 000 K as the
value of Q/T. A more correct value would be


Zirconium Alloys: Properties and Characteristics

2000

10

1600
Temperature (ЊC)

5
3
BWR
2

Relative corrosion rate


L

1800

In-reactor

PWR

1400

(b-Zr, b-Nb)

1200
1000
800
600

1

¬ a-Zr

400

0.8

0.02

200
0


Out-of-pile
30
10

0.02

20

40
60
Atomic percent niobium

80

100

Figure 8 Zr–Nb binary phase diagram.

500 ЊC/16 h

3
1

225

350 ЊC/ 1a
0.1
Average diameter of precipitates (mm)

0.8


Figure 7 Effect of precipitate size on the corrosion
kinetics of Zircaloys. Reproduced from Garzarolli, F.;
Stehle, H. Behavior of structural materials for fuel and
control elements in light water cooled power reactors, IAEA
STI/PUB/721; International Atomic Energy Agency: Vienna,
1987; p 387.

32 000 K, which fits very well with the recrystallization kinetics. The influence of the A-parameter
on the corrosion of Zircaloy is discussed in more
detail in Chapter 5.03, Corrosion of Zirconium
Alloys. High resistance to uniform corrosion in
PWR is obtained for the A-parameter close to
(1.5–6.0) Â 10À19 h. In BWR, the A-parameter value
for the Zircaloy-2 cladding in BWR has to be in
the range (0.5–1.5) Â 10À18 h (Figure 7).25 This
corresponds to precipitates larger than 0.18 mm.
The SA approach has been developed for the
Zircaloys and is clearly not applicable for other
alloys, such as the Zr–Nb alloys.
Niobium (columbium) is a b-stabilizer that can
extend the bcc domain to a complete solid solution
between pure Zr and pure Nb at high temperatures
(Figure 8). A monotectoid transformation occurs at
about 620  C and around 18.5 at.% Nb. The solubility of Nb in the a-phase is maximal at the monotectic
temperature, and reaches 0.65%.

Water b-quenching of small pieces leads to the
precipitation of a0 martensite supersaturated in Nb.
Tempering at intermediate temperature results

in b-Nb precipitation within the a0 needles and
subsequent transformation of a0 into a. When
quenching is performed from an a þ b region,
a uniform distribution of a- and b-grains is
obtained, and the Nb-rich b-phase does not transform. By aging at temperatures in the range of 500  C,
the metastable Nb-rich b-phase can be decomposed
into an hcp o-phase. This gives a sharp increase
in mechanical strength because of the fine microstructure obtained by the b-o transformation.26
In the usual form of the Zr–2.5% Nb, the cold work
condition after a þ b extrusion and air-cooling, the
microstructure consists of Zr grains with layers of
b-Nb rich phase (close to eutectoid composition).
Owing to the affinity of Fe for the b-phase, most of
this element is found in the minor b-grains. These
b-grains are metastable and decompose, upon aging,
to a mixture of a-Zr and pure b-Nb. The Nb dissolved in the a-hcp Zr phase is itself metastable and
the irradiation-induced precipitation of the supersaturated Nb solid solution is believed to be the origin
of the improvement in corrosion resistance under
irradiation of these alloys.27
In the case of Zr–1% Nb used for VVER and
RBMK, or M5® in PWRs, the concentration of Nb
in the Zr matrix after processing corresponds to the
maximum solubility near the monotectoid temperature, which is higher than the solubility at the service
temperature. Owing to the slow diffusion of Nb, the
equilibrium microstructure cannot be obtained thermally. However, the irradiation-enhanced diffusion
allows precipitation of fine b-Nb needles in the
grains after a few years in reactors.28


226


Zirconium Alloys: Properties and Characteristics

Sulfur has recently been observed to be extremely
efficient in improving the creep resistance, even at
concentration as low as 30–50 ppm. This chemical
species, formerly not considered as important, is now
deliberately added during processing to reduce the
scatter in behavior and to improve the high temperature mechanical properties.29 The efficiency of such
low concentrations on the creep properties has been
explained by the segregation of the S atoms in the core
of the dislocations, changing their core configurations.
It does not affect the corrosion properties.30
In the case of complex alloys, other thermodynamical interactions are expected and intermetallic
compounds including three or four chemical elements are observed. The chemistry and the crystallography of these phases may be rather complex.

Two examples will be given of the complex structure
and behavior of these intermetallics.
 For the Zr–Cr and Zr–Ni binary alloys, the stable
forms of the second phase are Zr2Ni or ZrCr2.
These phases are effectively the ones observed in
the Zircaloys, with Fe substituting for the
corresponding TM. Therefore, the general formulae of the intermetallic compounds in Zircaloys are
Zr2(Ni,Fe) and Zr(Cr,Fe)2. The crystal structure of
the Zr(Cr,Fe)2 precipitates is either fcc (C15) or
hcp (C14), depending on composition and heat
treatment. Both structures are Laves phases, with
characteristic stacking faults as seen in Figure 9.
The equilibrium crystallographic structure is
dependent upon the Fe/Cr ratio, cubic below 0.1

and above 0.9, and hexagonal in the middle. Under
irradiation, these precipitates transform to amorphous state and release their Fe in the matrix,
with strong impact on corrosion behavior under
irradiation.31
 In the Zr–Nb–Fe ternary, other intermetallic compounds can be observed (Figure 10): the hexagonal Zr(Nb,Fe)2 phase and the cubic (Zr,Nb)4Fe2.32
Although of apparent similar composition, the two
phases are indeed different: Nb can substitute Fe in
the hexagonal phase, while it will substitute Zr in
the cubic phase. In these alloys, due to the slow
diffusion of Nb, metastable phases are often present
and the equilibrium microstructure after industrial
heat treatments may be far from the stable one.
Therefore, the final microstructure is strongly
dependent on the exact thermomechanical history.

3 mm

200 nm
Figure 9 Microstructure of recrystallized Zry-4: Zr(Fe,Cr)2
precipitates in the Zr(Sn–O) matrix (TEM at two different
scales).

0.1
α-Zr + cub

Fe (wt%)

0.08

α-Zr

+ cub
+ hex

α-Zr +
hex

β-Zr + hex
β-Nb + hex + cub
Hex + cub
β-Zr phase boundary
Domain limit
Domain limit

α-Zr + Zr3Fe
+ cub

0.06

0.04

α-Zr + β-Nb
+ hex

α-Zr + β-Zr (metastable)
+ β-Nb + hex

0.02
α-Zr + β-Nb
α-Zr


0
0

0.2

0.4

0.6
Nb (wt%)

0.8

Figure 10 Zr-rich corner of the Zr–Nb–Fe ternary phase diagram at 580  C.

1

1.2


Zirconium Alloys: Properties and Characteristics

In addition, the low solubility of these elements at
operating temperatures drastically reduces the diffusion kinetics and requires more than a year to
reach equilibrium at 450  C, in the absence of
irradiation.33
Other minor constituents are often found in the
form of precipitates. Among them are the carbide fccZrC and silicides or phosphides of various stoichiometries (Zr3Si, ZrSi2, ZrP, and Zr3P) that act as
nucleation sites for the b ! a-phase transformation
during quenching and, therefore contribute to control the a-platelets thickness and density.


2.07.4.2

Industrial Alloys

The zirconium alloys in use today for nuclear applications are limited in number: besides pure Zr, only four
alloys are currently listed in the ASTM standards for Zr
ingots for nuclear applications (ASTM-B350). Those
are shown in Table 1. The first three are used for
cladding and structural materials, such as guide tubes
and channel boxes in PWRs and BWRs and structural
materials in CANDU reactors, while grade R 60904 is
used exclusively in pressure tubes for CANDU reactors. For cladding tubes, only Zircaloy-2 and -4 are
listed in the applicable standard (ASTM B-811).
Alloys of more recent use such as ZIRLO®, M5®,
E110, or E635 are now of common use in light water
reactor cladding, but are not considered for ASTM
designation in the near future. Due to the limited
market of cladding tubes for nuclear reactors, and the
small number of tube producers or fuel vendors, the
exact chemistry, processing routes, or mechanical
properties are usually agreed mutually between the
contracting parties.
Historically, the first Zircaloy was conceived in
the United States as a 2.5% Sn alloy. Owing to its
poor long-term corrosion behavior, the tin content
was reduced roughly by a factor of 2. A fortuitous
Table 1

Zircaloy-2
Zircaloy-4

M5
E110
Zr 2.5Nb
E125
Zirlo
E635

227

contamination of one melt by stainless scraps showed
the drastic improvement induced by small additions
of Fe, Cr, and Ni, the constituents of the austenitic
stainless steels. Systematic composition variations to
optimize the alloy introduced the Zircaloy-2. The
capture of a significant amount of hydrogen by the
alloy during corrosion was attributed to the presence
of nickel. Its replacement by an equivalent amount
of iron and chromium led to the Zircaloy-4.34
Zr–Nb alloys were developed in Canada, Russia,
and the United States, with initial high Nb concentrations (up to 4%). For the claddings of BWRs,
they showed poor behavior and the Zr–Nb alloys
development was stopped soon in the United States.
Zr–2.5Nb was quite satisfactory for the pressure
tubes, due to its low hydrogen pick-up during operation, and the engineering optimization of this alloy
was continued in Canada and Russia. It remains the
reference alloy for pressure tubes, in CANDUs.
Zr–1% Nb has been developed for cladding in these
countries and behaved very satisfactorily in VVER.
A renewal of interest in such alloys in the western
world in the 1990s led to the development of a

Zr–1% Nb alloy, with controlled additions of Fe and
S. The M5® alloy is now of regular use in PWRs, with
excellent corrosion resistance, compared to the
former Zry-4.35
‘Quaternary’ alloys were conceived as a mixture of
the Zircaloys and the Zr–Nb alloys, hoping to conserve the specificity of each of them in addition to the
different alloying elements:
 Niobium for the resistance to hydrogenation
during corrosion
 Tin for the corrosion resistance by reducing the
dependence on deleterious impurities
 Iron also for corrosion resistance by mitigating the
dependence on the coolant temperature.
The results appear to be in-line with the expectations
for these alloys, which can be considered as variants

Composition of the Zr alloys of industrial use (concentration in wt% or ppm)
ASTM

Sn

R 60802
R 60902

1.2–1.5
1.2–1.7

R 60904
1
1.2


Nb

0.8–1.2
1
2.5–2.7
2.5–2.6
1
1

Fe

Cr

Ni

O

0.07–0.2
0.18–0.24
<500 ppm
100 ppm
<650 ppm

0.1
0.1

0.05

0.12

0.1–0.14
0.11–0.16
0.05–0.07
0.12–0.15
0.04–0.07
0.09–0.12
0.05–0.07

0.1
0.35

S

10–35 ppm


228

Zirconium Alloys: Properties and Characteristics

of the composition: Zr–1% Nb–1% Sn–TM alloys.
The multicomponent alloys, namely, E635 and
Zirlo® are, respectively, used in the cores of VVER
and PWR.36,37 Both show low corrosion and very
limited irradiation growth. Irradiation growth refers
to the dimensional changes at constant volume of an
unstressed material under irradiation.38 The growth
phenomenon is induced by the anisotropic clustering
or disappearance of the point defects created by
irradiation. The stability under irradiation of the

precipitates present in these alloys appears to be the
origin of such good dimensional stability (Chapter
4.01, Radiation Effects in Zirconium Alloys).39

2.07.5 Mechanical Properties Before
Irradiation
2.07.5.1

Strength and Ductility

The mechanical properties of the Zr alloys are
strongly dependent on several parameters such as
composition, texture, and metallurgical state. For
practical purposes, the properties at 300–400  C are
most important and room temperature behavior is
used mostly for comparison.
At room temperature, in the annealed state, pure,
oxygen-free Zr has a low yield strength of about
150 MPa. However, the yield strength can be
increased by solution strengthening, using oxygen,
tin, or niobium. Tin causes only a small increase in
tensile strength, but is efficient for creep improvement; Nb increases both yield and creep strength.40
Table 2

By contrast, the addition of 1000 ppm of oxygen
increases the yield strength to 300 MPa. As a result
of this, Zircaloys have minimal yield strengths in the
range of 250–300 MPa and the Zr–2.5% Nb alloy,
about 300 MPa. As in other metals, reduction in
grain size is also used to obtain higher strength,

leading to the request of a grain index of 7 or finer
for standard products.
For all those materials, the ductility remains high,
above 20%. Cleavage is never observed in Zr and Zr
alloys, even at liquid nitrogen temperature. Additional strength is obtained by cold working, allowing
the increase of the yield strength above 400–500 MPa.
This is followed by a final stress-relief heat treatment
to restore ductility without drastic reduction in
strength, by relaxing the internal stresses.
Finally, the texture itself can increase alloy
strength by changing the Schmid factor for slip or
twinning. This can be observed by the differences in
strength between the axial and transverse directions.
In addition, due to the distorted shape of the yield
locus, and consequently of the orientation of the
strain vector, strain is also anisotropic.
Typical values of the strength of industrial alloys in
various metallurgical states are given in Table 2.
Due to the absence of cleavage, the toughness
properties of Zr alloys are not a matter of concern
for unirradiated materials. Indeed, difficult measurements have obtained values of KIC well above
50 MPa m1/2. However, during irradiation, two processes deteriorate the ductility of the alloys:

Typical values of the mechanical properties of industrial Zr alloys

Material

Zry-2 or-4
Stress relieved
Zry-2 or-4

Recrystallized
M5
Recrystallized
E110
Recrystallized
ZIRLO
Recrystallized
E635
Recrystallized
E125
Annealed
Zr 2.5Nb
CANDU

Heat
treatment ( C)

350–375  C

Room temperature
sY (MPa)

UTS (MPa)

Ductility (%)

sY (MPa)

UTS (MPa)


Ductility (%)

480  C 4 h

570

770

18

360

450

20

500  C 4 h

370

550

30

150

250

45


550  C 2 h

320

480

35

130

250

40

570  C 2.5 h

250

400

45

110

210

55

580  C 4 h


420

560

15

280

340

15

600  C 4 h

300

480

30

150

270

35

550  C 2 h

350


530

25

200

350

30

350

510

15

Cold worked

UTS, Ultimate Tensile Strength.


Zirconium Alloys: Properties and Characteristics

 The first one is the formation of numerous fine
dislocation loops that harden the materials and
induce basal dislocation slip and localized strains.
The macroscopic effect is a highly localized
strain and highly reduced engineering ductility at
fracture.41,42 The details of this ductility degradation
induced by the irradiation are described in Chapter

4.01, Radiation Effects in Zirconium Alloys.
 The second mechanism of ductility degradation
is linked to the impact of hydrogen trapped by
the alloys during operation. The corresponding
mechanisms are described in Section 2.07.5.3.
2.07.5.2 Mechanical Properties in
Temperature and Creep
The thermal properties of Zr alloys are given in
Table 3. The heat capacity increases with temperature, about 10% for 300 K. The anisotropy in any
property is indeed decreasing with temperature, as
the anisotropy in thermal expansion drives the hcp
cell of Zr toward the ideal c/a ratio. In addition to the
anisotropy of the thermal expansion coefficient, the
elastic weakening in temperature has also to be
considered.
As for any alloy, an increase in temperature results
in a decrease in strength, but the evolution is not
uniform, and a plateau in strength is observed near
200–400  C. Known as the dynamic strain aging
(DSA), it corresponds to the interaction of oxygen
atoms with the dislocations. In this temperature
range, oxygen atoms diffuse at rates commensurable
with dislocation glide and hinder their motion.
At higher temperatures, they are too mobile to
affect the creep rates drastically.
Typical creep deformation rates are expressed by
equations such as


ÀQ

e_ ¼ Asn exp
RT
Due to the transition induced by the oxygen in DSA,
large discrepancies are observed in the creep
Table 3

229

behavior of different alloys.43 For conditions leading
to strain rates below 10À9 sÀ1 (i.e., low temperatures
and/or stresses), a low stress coefficient is observed:
n $ 1–2. Grain boundary sliding may be the mechanism involved in these conditions. For high creep
rates, stress exponents larger than 4–6 have been
reported. The mechanisms considered are complex,
with predominance of dislocation glide controlled by
local climb.44 Similarly, the activation energy was
measured to be as low a 40 kJ molÀ1 for the low
temperature regime, and 2 or 3 times larger for the
high strain rates.
Most of the metallurgical parameters affect the
creep rate. The effect of alloying elements on creep
properties is different from their effect on tensile
strength. Oxygen improves creep resistance, especially at low temperatures, but its effect is small
compared to improvement in yield strength at room
temperature. Despite its detrimental effect on corrosion, tin is maintained at significant levels in
Zircaloys as an efficient alloying element to improve creep resistance.45 For similar alloys with the
same structure, creep behavior is similar; for example, Zircaloy-2 and Zircaloy-4 have similar creep
strengths for the same thermomechanical processing.
However, the metallurgical state of the material also
influences the creep mechanisms; although SR material has higher tensile strength, its high dislocation

density induces a two- to threefold increase in creep
rate, compared to the RX state. Completely dissolved
hydrogen increases the creep rate, while precipitated
hydrides harden the alloys.46 For additions as low
as 50 ppm, sulfur drastically increases the creep
strength, and is now an alloying element for specific
alloys.47
2.07.5.3 Hydrogen Embrittlement and
Other H Effects
During oxidation of Zr alloy components in reactors
or in autoclaves, the reduction of water by the Zr
alloy follows the general reaction scheme:

Thermal properties of Zr alloys
Average

Heat capacity (RT )
Expansion coefficient
Modulus of elasticity
Room temperature
400  C
RT, Room temperature.

À1

a-direction

c-direction

5.2 Â 10À6 KÀ1


10.4 Â 10À6 KÀ1

99 GPa
82 GPa

125 GPa
105 GPa

À1

280 J kg K
8.5 Â 10À6 KÀ1 (cladding tubes)
90 GPa (cladding tubes)


230

Zirconium Alloys: Properties and Characteristics

2H2 O þ Zr ! ZrO2 þ 4H



The reduction of the water molecules at the
coolant-oxide interface releases four hydrogen

atoms as radicals H . They are chemically adsorbed
at the tips of the oxide pores and their evolution
controls the behavior of this chemical species. Most

of the H atoms recombine, creating hydrogen molecules that escape and dissolve into the coolant.
A limited amount can ingress in the oxide and
migrate to the metallic matrix, where H is soluble,
or interact with Zr to form hydrides (see Chapter
5.03, Corrosion of Zirconium Alloys). The fraction of the hydrogen that is trapped in the Zr alloy is
called the hydrogen pick-up fraction (HPUF). For
Zry-2, it is in the range of 30–60%, and for Zry-4
a lower HPUF is observed (15–25%), while the
Zr–Nb alloys show the lowest one (4–10%). In
order to reduce H pick-up, care should be taken to
avoid the alloys catalyzers of the hydrogen molecule
dissociation, such as Ni and Pt. This is the main
reason for suppression of that element in Zircaloy-4:
on removing Ni, HPUF is generally below 15% for
standard PWR fuel cladding. Due to the strong
variation of the solubility of H with temperature
($200 ppm at 350  C, but near 1 ppm at RT), hydrogen reacts with Zr to precipitate as hydrides as the
alloy is cooled down.
One of the consequences of hydrogen ingress
into Zr is the delayed hydride cracking (DHC).48
This high temperature mechanism involves the
ingress of hydrogen into a Zr-alloy component, its
migration up the stress or thermal gradients, and
its concentration in the regions of low temperature
or higher tensile stress. When the local concentration
exceeds the terminal solid solubility, the hydride
phase precipitates.49 At operation temperatures, a
quasicontinuous crack growth is observed, whose
rate depends on the hydrogen content, on the structure of the alloy, and on its crystallographic texture.50
The failure of CANDU pressure tubes by this mechanism added pressure to the R&D in this field.51

At lower temperatures, corresponding to fuel
handling and transport, the dissolved hydrogen precipitates as d hydrides that have very brittle behavior.
The hydrides are brittle below 200  C and crack
when the stresses are high enough; failure of the
components at low temperatures occurs because of
percolation of the broken hydride platelets. Depending on the geometry and spatial distribution of the
hydrides, very low ductilities can be observed.
A slight increase in temperature would make the

hydrides ductile and the mechanical behavior of
the claddings returns to normal above 200  C.52,53
Hydrogen increases the creep strength of SR
Zry, whatever be the H content. However, for
RX alloys, H that gets completely dissolved enhances
the creep rate, while precipitated hydrides harden
the alloys.54 This difference in behavior is connected
to the effect of hydrides in inhibiting the thermal
recovery. As described earlier, sulfur at very low
concentrations drastically increases the creep strength,
and even the yield strength.47,55
Hydrogen in the cladding is also claimed to
increase the corrosion rate due to the presence of an
outer rim of hydrides,56 and to deteriorate the behavior of the cladding during accident sequences, such as
reactivity-induced accident (RIA), in which brittle
hydrides drastically reduce the strain to failure,57–60
or loss of coolant accident (LOCA), where the cladding that is softened by hydrogen creeps faster and
fails at lower temperatures than it would with lower
hydrogen contents,61 or mixtures of these effects.
2.07.5.4


Prospectives

Although the use of Zr alloys has proven to be a good
engineering solution, several properties are still subject to poor scientific understanding of their origins.
Among them, deformation mechanisms (critical shear
stresses on the different shear and twin systems),
irradiation behavior (point defects interactions with
alloying elements and mobilities), and corrosion
mechanisms at atomic scales clearly need further
basic scientific work.
For industrial purposes, the lot-to-lot variations
in properties observed in many instances require larger
margins in design and consequently reduce the efficiency of the power plants. Detailed analysis of the
origins of these variations could be highly cost effective.
Last, the current development of computational
materials science is a major opportunity for in-depth
understanding and forecasting of the mechanisms of
irradiation damage under irradiation. In that direction, multiscale modeling is particularly suited to
nuclear materials. Zr alloys having no other sizable
application than in the nuclear industry would
strongly benefit from these new R&D techniques.

Acknowledgments
Special thanks are given to P. Barberis, A. Motta, and
N. Dupin for their efficient support during the
writing process of this chapter.


Zirconium Alloys: Properties and Characteristics


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