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Waterside corrosion of zirconium alloys in nuclear power plants PDF

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XA9846730 IAEA-TECDOC-996 Waterside corrosion of zirconium alloys in nuclear power plants INTERNATIONAL ATOMIC EMEBOY AGENCY I) January 1998 \J, 1 The IAEA does not normally maintain stocks of reports in this series. However, microfiche copies of these reports can be obtained from INIS Clearinghouse International Atomic Energy Agency Wagramerstrasse 5 P.O. Box 100 A-1400 Vienna, Austria Orders should be accompanied by prepayment of Austrian Schillings 100, in the form of a cheque or in the form of IAEA microfiche service coupons which may be ordered separately from the INIS Clearinghouse. The originating Section of this publication in the IAEA was: Nuclear Fuel Cycle and Materials Section International Atomic Energy Agency Wagramer Strasse 5 P.O. Box 100 A-1400 Vienna, Austria 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 NEXT PAGE(S) left BLANK 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. 10

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The effect of pH already noted in Chapter 1 appears to be general for all alkaline for planned future operation of 24 month cycles, German workers are up in BWR plants", Water Chemistry of Nuclear Reactor Systems 6, British Nuclear Energy LE ZIRCONIUM, Proprietes -Microstructures.
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