Waterside corrosion of zirconium alloys in nuclear power plants

312 Pages·2004·8.21 MB·English

XA9846730
IAEA-TECDOC-996
Waterside corrosion of
zirconium alloys in
nuclear power plants
INTERNATIONAL ATOMIC EMEBOY AGENCY
I)
January 1998
\J,
1
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WATERSIDE CORROSION OF ZIRCONIUM ALLOYS IN
NUCLEAR POWER PLANTS
IAEA, VIENNA, 1998
IAEA-TECDOC-996
ISSN 1011-4289
©IAEA, 1998
Printed by the IAEA in Austria
January 1998
FOREWORD
Technically the study of corrosion of zirconium alloys in nuclear power reactors is a very active
field and both experimental work and understanding of the mechanisms involved are going through
rapid changes. As a result, the lifetime of any publication in this area is short. Because of this it has been
decided to revise IAEA-TECDOC-684 — Corrosion of Zirconium Alloys in Nuclear Power Plants —
published in 1993. This updated, revised and enlarged version includes major changes to incorporate
some of the comments received about the first version.
Since this review deals exclusively with the corrosion of zirconium and zirconium based alloys
in water, and another separate publication is planned to deal with the fuel-side corrosion of zirconium
based fuel cladding alloys, i.e. stress corrosion cracking, it was decided to change the original title to
Waterside Corrosion of Zirconium Alloys in Nuclear Power Plants.
The rapid changes in the field have again necessitated a cut-off date for incorporating new data.
This edition incorporates data up to the end of 1995; including results presented at the 11 International
Symposium on Zirconium in the Nuclear Industry held in Garmisch-Partenkirchen, Germany, in
September 1995.
The IAEA wishes to express its thanks to all the authors, both of this updated review and of
IAEA-TECDOC-684 on which it was based. The IAEA staff member responsible for this publication
was I.G. Ritchie of the Division of Nuclear Power and the Fuel Cycle.
EDITORIAL NOTE
In preparing this publication for press, staff of the IAEA have made up the pages from the
original manuscripts as submitted by the authors. The views expressed do not necessarily reflect
those of the IAEA, the governments of the nominating Member States or the nominating
organizations.
Throughout the text names of Member States are retained as they were when the text was
compiled.
The use of particular designations of countries or territories does not imply any judgement by
the publisher, the IAEA, as to the legal status of such countries or territories, of their authorities
and institutions or of the delimitation of their boundaries.
The mention of names of specific companies or products (whether or not indicated as
registered) does not imply any intention to infringe proprietary rights, nor should it be construed
as an endorsement or recommendation on the part of the IAEA.
CONTENTS
1. INTRODUCTION 9
2. METALLURGY OF ZIRCONIUM ALLOYS 11
2.1. Processing 11
2.2. Microstructure 12
2.2.1. Pure zirconium 12
2.2.2. Alloys and alloying elements 12
2.3. Heat treatments and resultant microstructure 19
2.4. Deformation and texture 23
3. OXIDATION THEORY 27
3.1. Microcryistalline nature of the oxide 28
3.2. Electrical resistivity of zirconia 29
3.3. Effects of electric fields on the oxidation kinetics 29
3.4. Effect of impurities and alloying elements 34
4. CORROSIUM IN THE ABSENCE OF IRRADIATION 37
4.1. Introduction 37
4.2. Uniform oxide formation 37
4.2.1. Oxidation kinetics 40
4.2.2. Pre-transition oxidation mechanism 57
4.2.3. Mechanism of oxide breakdown on the Zircaloys 67
4.2.4. Mechanism of oxide breakdown in Zr-Nb alloys 78
4.2.5. Post-transition growth 78
4.3. Non-uniform (nodular) oxide formation 84
4.3.1. Nodular oxide formation 85
4.3.2. Mechanism of nodule formation 88
4.3.3. Simulating nodular corrosion in high temperature water 90
5. HYDROGEN ABSORPTION 91
5.1. Hydrogen absorption mechanism 91
5.1.1. Hydrogen uptake during corrosion 92
5.1.2. Absorption of hydrogen gas 104
5.1.3. Hydrogen absorption via metallic contacts Ill
5.1.4. Hydrogen uptake during cathodic polarisation 114
5.2. Effects of hydrogen content on oxidation 116
6. FACTORS AFFECTING THE CORROSION OF ZIRCONIUM ALLOYS IN REACTORS . . 124
6.1. Alloy compositions for nuclear applications 124
6.1.1. Alloy types 124
6.1.2. Alloy development programmes 126
6.2. Metallurgical variables 136
6.2.1. Precipitate size 136
6.2.2. Influence of quenching conditions 145
6.2.3. Influence of final annealing 145
6.2.4. Influence of cold work and deformation sequence 150
6.2.5. Initiation of nodular corrosion in BWR materials 150
6.2.6. Effect of metallurgical conditions on the corrosion of Zr-Nb alloys 150
6.3. Surface conditions 152
6.4. Coolant chemistry 154
6.4.1. PWR chemistry 155
6.4.2. BWR chemistry 161
6.4.3. WWER chemistry 162
6.4.4. PHWR (CANDU) chemistry 164
6.5. Effect of temperature 164
6.5.1. High temperature oxidation of Zircaloy alloys 165
6.5.2. High temperature oxidation of Zr-l%Nb alloys 165
6.6. Effect of heat flux 165
7. MODELLING OF IN-REACTOR CORROSION OF ZIRCONIUM
ALLOY FUEL CLADDING 170
7.1. Introduction 170
7.2. Calculation of oxide-metal interface temperatures 171
7.2.1. Single phase coolants 171
7.2.2. Two phase coolants 173
7.2.3. Oxide thermal conductivity 174
7.3. Semi-empirical models for Zircaloy corrosion in PWRs 175
7.3.1. Generic formulation for semi-empirical models 178
7.3.2. Individual models of simple generic form 179
7.3.3. Individual models incorporating additional effects 188
7.4. Mechanistic models 189
7.4.1. Cox's model 189
7.4.2. Russian models for Zr-l%Nb cladding 191
7.5. Summary of PWR corrosion modelling 195
8. IRRADIATION EFFECTS ON CORROSION 198
8.1. Irradiation damage 198
8.1.1. Fast neutron damage in the metals 198
8.1.2. Displacement damage in other structures 199
8.1.3. Effect of irradiation on microstructures 203
8.2. Radiation chemistry 212
8.2.1. Radiolysis in the bulk water 212
8.2.2. Radiolysis near metal surfaces or in the pores surrounded by
metal oxides 218
8.2.3. "Thick oxide film effects" 221
8.2.4. Localised corrosion and dissimilar metals 224
8.3. Crud deposition and heat transfer effects 225
8.3.1. PWR crud deposition 225
8.3.2. WWER crud deposition 236
8.3.3. BWR crud deposition 236
8.4. Metallurgical and chemical variables 238
8.4.1. Behaviour of alloying additions 238
8.4.2. Electrochemical effects 239
8.5. Corrosion of Zr-l%Nb cladding 242
9. PRESENT STATUS OF THE MECHANISTIC UNDERSTANDING 249
9.1. Current understanding of the out-reactor oxidation mechanism 249
9.1.1. Mobile species 249
9.1.2. Evolution of oxide morphology 250
9.1.3. The development and nature of oxide porosity 256
9.1.4. Oxide barrier layers 261
9.1.5. Effect of some variables on the oxide structure 264
9.2. Empirical correlations of effects of irradiation 265
9.2.1. Development of irradiation corrosion mechanisms 266
9.2.2. Open questions on micromechanisms for in-reactor corrosion 277
9.2.3. Present status of mechanistic studies 278
9.2.4. Recommendations for future work 278
APPENDIX 279
REFERENCES 281
BIBLIOGRAPHY 311
LIST OF CONTRIBUTORS 313
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1. INTRODUCTION
The original version of this TECDOC [1] was written at a time when major programmes on
fuel cladding improvement were under way in most countries with active nuclear power programmes,
but few of the results of these programmes had been published. The references on which this first
version was based were cut off essentially prior to the Portland IAEA Conference [2], whose
Proceedings were not then available, and the Kobe Zirconium Conference [3] respectively in
September 1989 and November 1990, although a few references to these meetings were subsequently
added. The contents of this version, therefore, rapidly became dated. The original version had been
targeted at the relatively limited audience of those professionals actively working on some aspect of
the research and development of corrosion resistant zirconium alloys, but in practice a large fraction
of the demand came from those involved in the nuclear fuel cycle at the utility level. This has been
taken into account in the new version.
Zirconium alloys continue to be the major structural materials employed within the fuelled
region of all water cooled nuclear power reactors. Thus, they are invariably used as fuel cladding, fuel
channels (boxes, wrappers), pressure tubes and calandria tubes and often as fuel spacer grids. Other
structural metals appear in this region of the reactor core mainly as minor components such as grid
springs and garter springs (spacers between pressure and calandria tubes in CANDUs). The
performance of zirconium alloys in service has been generally satisfactory, although the pressures to
achieve higher fuel burnups and higher reactor thermal efficiencies have pushed the historically used
alloys to the limits of their capabilities. Evidence that these limits were being reached was the
primary driving force for the major new alloy development programmes already mentioned. A further
driving force has been the acknowledgement that debris fretting had become the primary cause of fuel
failures, and that primary failures from this cause could lead to unexpectedly severe secondary
failures, especially for zirconium barrier cladding developed to protect against pellet-cladding
interaction (stress-corrosion cracking) failures as a primary defect mechanism.
In PWRs, therefore, there is a general desire to reduce oxidation rates in order to achieve
higher fuel burnup and rating. However, because of the temperature feedback loop (section 7. 2. 3.) at
the end of life, the corrosion rate (and the associated hydrogen uptake rate) accelerates rapidly. Other
factors may also increase the corrosion rate under these conditions, including the precipitation of
hydrides (section 5. 2.), dissolution of precipitates and the concentration of lithium hydroxide. There
is a need to understand the potential effects of concentrating lithium hydroxide under these conditions
because they are linked to the ability to reduce circuit activation, and hence personnel radiation
exposures, that could result from the use of increased LiOH concentrations. In BWRs, the infrequent
secondary degradation failures that led to serious operational consequences as a result of rapid
increases in off-gas radiation levels, are also the target of a major research and development effort, p-
quenched cladding amongst other changes has eliminated serious episodes of nodular corrosion
induced (Crud Induced Localised Corrosion-CILC) failures, but a reduction in end-of-life uniform
oxide thickness is still a desirable objective.
As in any system where the consequences of minor changes in materials or operating
conditions can have major impacts on the economics of the system if they lead to forced outages, it is
vitally important that the consequences of any changes be thoroughly explored and understood.
Decisions on whether to make operational changes (e.g. increased Li) can often be beset with
conflicting requirements which have to be balanced before a decision can be made. It is hoped that
this review will provide sufficient background and information on the factors controlling zirconium
alloy corrosion and hydrogen uptake in-reactor to permit such decisions to be made on a sound basis.
The revised format of the review now includes:
• Introductory chapters on basic zirconium metallurgy and oxidation theory;
• A revised chapter discussing the present extent of our knowledge of the corrosion mechanism
based on laboratory experiments;
• A separate and revised chapter discussing hydrogen uptake;
• A completely reorganised chapter summarising the phenomenological observations of
zirconium alloy corrosion in reactors;
• A new chapter on modelling in-reactor corrosion;
• A revised chapter devoted exclusively to the manner in which irradiation might influence the
corrosion process;
• Finally, a summary of our present understanding of the corrosion mechanisms operating in
reactor.
Although much new information has become available in the last five years, there are still
blocks of data that have not been linked together in an understandable manner. Thus, much of the
early corrosion data was obtained from non-heat transfer specimens in in-reactor loops, whereas
virtually all of the recent in-reactor data comes from high heat flux fuel cladding. Only minor
amounts of recent data come from non-heat transfer surfaces such as oxide thicknesses on plena,
spacer grids, pressure tubes, water rods or guide tubes. As a result, it remains difficult to extrapolate
conclusions drawn from the early loop tests to the behaviour of current fuel cladding or pressure
tubes.
Great strides have been made recently in delineating the impact of variations in fabrication
route and of careful control of impurity and alloying additions on the in-reactor behaviour of fuel-
cladding. As a result most fuel vendors have moved to some version of "optimised" Zircaloy
cladding, as precursor to the introduction of new cladding alloys lying outside the range of the
Zircaloy specifications. The introduction of such new alloys has been greatly facilitated by the
demonstration of both the production and satisfactory performance of duplex cladding tubes. These
are in the form of duplex tubes ~90% of the wall thickness of which is standard Zircaloy-4, with the
outer -10% of the tube made of the new alloy. This requires similar technology to that which puts
unalloyed (or low alloyed) zirconium barriers on the inside of fuel cladding tubes for BWR
applications. The advantage of this duplex tube technology is that alloys that could not be considered
for fuel cladding use in a monotube form, because of inadequate, or inadequately known, mechanical
properties, can be introduced in the form of duplex tubes with minimal regulatory limitations.
Another area where major changes have been apparent since the original review was written
is in the availability of much evidence on the behaviour of Zr-l%Nb cladding in KOH/ammonia or
hydrazine water chemistries typical of Russian designed reactors. This information has been
incorporated wherever possible to provide a comparison with the observations on Zircaloy-4 in LiOH
water chemistry. The low oxide thicknesses still present on Zr-l%Nb cladding after high burnup in
KOH/ammonia water chemistry (where thermal hydraulic conditions have been comparable to those
in a high temperature PWR, i.e. T >345°C with sub-channel boiling) call for some comparative
Mt
testing of Zircaloy-4 under these conditions so that any contribution of LiOH to current in-reactor
experience can be properly evaluated.
This revision of the review should increase its value to a wider range of readership than was
aimed for in the original.
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