4.01

Radiation Effects in Zirconium Alloys

F. Onimus and J. L. Be´chade
Commissariat a` l’Energie Atomique, Gif-sur-Yvette, France

ß 2012 Elsevier Ltd. All rights reserved.

4.01.1

Irradiation Damage in Zirconium Alloys

4.01.1.1
4.01.1.1.1
4.01.1.1.2
4.01.1.1.3
4.01.1.2
4.01.1.2.1
4.01.1.2.2
4.01.1.2.3
4.01.1.3
4.01.1.3.1
4.01.1.3.2
4.01.1.3.3
4.01.1.3.4
4.01.1.3.5
4.01.1.4
4.01.1.4.1
4.01.1.4.2
4.01.2

4.01.2.1
4.01.2.1.1
4.01.2.1.2
4.01.2.1.3
4.01.2.1.4
4.01.2.2
4.01.2.3
4.01.3
4.01.3.1
4.01.3.1.1
4.01.3.1.2
4.01.3.2
4.01.3.2.1
4.01.3.2.2
4.01.3.3
References

Damage Creation: Short-Term Evolution
Neutron–zirconium interaction
Displacement energy in zirconium
Displacement cascade in zirconium
Evolution of Point Defects in Zirconium: Long-Term Evolution
Vacancy formation and migration energies
SIA formation and migration energies
Evolution of point defects: Impact of the anisotropic diffusion of SIAs
Point-Defect Clusters in Zirconium Alloys
hai Dislocation loops
hai Loop formation: Mechanisms
hci Component dislocation loops
hci Loop formation: Mechanisms

Void formation
Secondary-Phase Evolution Under Irradiation
Crystalline to amorphous transformation of Zr-(Fe,Cr,Ni) intermetallic precipitates
Irradiation effects in Zr–Nb alloys: Enhanced precipitation
Postirradiation Mechanical Behavior
Mechanical Behavior During Tensile Testing
Irradiation hardening: Macroscopic behavior
Irradiation hardening: Mechanisms
Post-yield deformation: Macroscopic behavior
Post-yield deformation: Mechanisms
Effect of Postirradiation Heat Treatment
Postirradiation Creep
Deformation Under Irradiation
Irradiation Growth
Irradiation growth: Macroscopic behavior
Irradiation growth: Mechanisms
Irradiation Creep
Irradiation creep: Macroscopic behavior
Irradiation creep: Mechanisms
Outlook

Abbreviations
BWR
CANDU
DAD
EAM
EID
FP-LMTO

Boiling-water reactor

Canadian deuterium uranium
Diffusion anisotropy difference
Embedded atom method
Elastic interaction difference
Full-potential linear muffin-tin orbital

2

GGA
hcp
HVEM
LDA
MB
MD
NRT

2
2
2
2
4
4
4
6
7
7
8
9
9
10

10
10
13
14
14
14
14
16
16
17
18
19
19
19
21
24
24
25
26
27

Generalized gradient approximation
Hexagonal close-packed
High-voltage electron microscope
Local density approximation
Many body
Molecular dynamics
Norgett–Robinson–Torrens

1



2

Radiation Effects in Zirconium Alloys

PKA
PWR
RXA
SANS
SIA
SIPA
SIPA-AD
SIPN
SRA
TEM
Tm
UTS
YS

Primary knocked-on atom
Pressurized water reactor
Recrystallization annealed
Small-angle neutron scattering
Self interstitial atom
Stress-induced preferential absorption
Stress preferential induced nucleationanisotropic diffusion
Stress preferential induced nucleation
Stress-relieved annealed
Transmission electron microscopy

Melting temperature
Ultimate tensile strength
Yield stress

4.01.1 Irradiation Damage in
Zirconium Alloys
4.01.1.1 Damage Creation: Short-Term
Evolution
4.01.1.1.1 Neutron–zirconium interaction

Zirconium alloys are used as structural components
for light and heavy water nuclear reactor cores
because of their low capture cross section to thermal
neutrons and their good corrosion resistance. In a
nuclear reactor core, zirconium alloys are subjected
to a fast neutron flux (E > 1 MeV), which leads to
irradiation damage of the material. In the case of
metallic alloys, the irradiation damage is mainly due
to elastic interaction between fast neutrons and atoms
of the alloy that displace atoms from their crystallographic sites (depending on the energy of the incoming neutron) and can create point defects without
modifications of the target atom, as opposed to
inelastic interactions leading to transmutation, for
instance. During the collision between the neutron
and the atom, part of the kinetic energy can be transferred to the target atom. The interaction probability
is given by the elastic collision differential cross section1,2 which depends on both the neutron kinetic
energy and the transferred energy.3 For a typical fast
 Þ of
neutron of 1 MeV, the mean transferred energy ðT
 % 22keV. For low value of the
the Zr atomglide processes occur at even higher

stress. For very high stress, close to the YS, dislocation channeling occurs.
For cold-worked zirconium alloys, such as SRA
Zircaloy or cold-worked Zr–2.5Nb alloy,163 the SIPA
mechanism on the initial dislocations is a likely mechanism for irradiation creep. However, according to
Holt,171 the creep anisotropy of cold-worked zirconium alloys computed from the SIPA mechanism
assuming only hai type dislocations is not in agreement with the experimental anisotropy. The anisotropy computed from the climb-plus-glide mechanism
assuming 80% prism slip and 20% basal slip is in good
agreement with the experimental anisotropy, demonstrating that climb-plus-glide mechanism is probably
the effective mechanism. It should also be pointed out
that, since dislocations climb toward grain boundaries
or toward other dislocations, recovery of the initial
dislocation network occurs. In order to maintain a
steady-state creep rate, multiplication of dislocations
should also occur either via loop coalescence or via
dislocation sources, as discussed previously.
It should also be pointed out that, as there is a
coupling between swelling and irradiation creep in
stainless steel,181 we could assume a coupling
between growth and irradiation creep to occur in
zirconium alloys due to the effect of the stress on
the partitioning of point defects.134,162 Nevertheless,
the simple assumption of two separable deformation
components has proved to hold correctly for the
results given in the literature.163,180

4.01.3.3 Outlook
Concerning damage creation and point-defect cluster formation, improvement in the knowledge of
anisotropic diffusion of SIAs as well as better understanding of the microstructure of vacancy and interstitial hai loops and basal hci vacancy loops (origin of
the loop alignment, origin of the corduroy contrast



Radiation Effects in Zirconium Alloys

for instance) has to be aimed at. Multiscale modeling
approaches coupled with fine experimental analyses
of the irradiation microstructure (high-resolution
TEM, synchrotron radiation analyses, tomography
atom probe, etc.) should bring new insight concerning
the previous points mentioned but also elements in
order to propose modeling of the microstructure
evolution during irradiation: for instance, origin of the
alignments of Nb precipitates, stability of b-Nb precipitates, etc.
Concerning the mechanical behavior of Zr alloys
after irradiation, multiscale modeling of the postirradiation deformation with a better understanding of
the dislocation channeling mechanism and understanding of its effects on the postirradiation mechanical behavior are needed.
Moreover, better understanding of the postirradiation creep deformation mechanisms is also needed
using multiscale modeling.
The last point concerns the deformation mechanisms under irradiation. In that field, the basic questions
are still without answers: What are the irradiation creep
deformation mechanisms? What are the coupling
between the deformation under irradiation and the
thermal creep and growth? Progress has to be made
especially using in situ deformation devices under
irradiation, coupled with modeling approaches. (See
also Chapter 1.01, Fundamental Properties of
Defects in Metals; Chapter 2.07, Zirconium Alloys:
Properties and Characteristics and Chapter 5.03,
Corrosion of Zirconium Alloys).

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