Corrosion of Zr Alloys in Nuclear Powerplants (IAEA TECDOC-684)

177 Pages·1993·21.408 MB·English

IAEA-TECDOC-684
Corrosionf o zirconium alloys
in nuclear power plants
p
INTERNATIONAL ATOMIC ENERGY AGENCY
UZr^
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CORROSION FOZ IRCONIUM ALLOYSN I NUCLEAR POWER PLANTS
IAEA, VIENNA, 1993
IAEA-TECDOC-684
ISSN 1011-4289
Printed by the IAEA in Austria
January 1993
FOREWORD
Current trends towards extended burnup, increased outlet
temperatures and plant life extension in nuclear power facilities require
the solution of a large number of problems related to the reliability of
the materials used, especially in key components such as fuel cladding. In
particular, the corrosion of metals and alloys under irradiation poses
special pe rnhotub clrleoemfas r industrt yAp. resee nhattb, iliotty
predice tht long term behaviourf o materiala snir adiation environmentsi
rather limit eeumdd aina loltya cf okk nowledg foed etailed mechanisms.
For zirconium alloys, it is clear that there is a need for more
feedback between experiments on the factors influencing in-reactor
corrosion, experiments desigo nrtee vdeal micromechanistic procdesnsae s
attemptso t modele ht overall behaviourf o fuel claddinge ht ni reactor
core.
o Timpr ruuoonvd eerstand finocog rrosion mechanisms under
irradiation of zirconium alloys, to collect information systematically and
to identify areas where further experimentatis noien eden di1, 9e 8hI9tA EA
initiata es dpecial project ewphiattr hticipatf ieoox nperts from Canada,
France, Japan, USA and the former USSR.
This technical document is the result of two years of joint
investigationsn I.ve hirt eafowp idly evolving mechanistic understanding
of the phenomena in this field, the document presents a series of snapshots
of current nisdipeea csific f a esrwotehhuao dte slyrer eltaoheav ttan t
problem. Any attempt to present an agreed upon micromechanistic hypothesis
that explai enhots verall phenomena must await further detailed
investigations. Througe hohttuee tx ahttu, thors have endeavouorted
indicate criticalr u boga asnpiis c knowles idh gotepIe. d that this will
stimulate experimental studies in just those areas where further data are
most urgently required.
ehITAEA wisho e eethaexs tuhpt trs lh ettlofsiharos a s nokts
report. The IAEA staff member responsible for this publication was
A. Nechaev of the Division of Nuclear Fuel Cycle and Waste Management.
EDITORIAL NOTE
In preparine Ignh e tptethhrrien stsaf mstoi, o asrntteaao frfli faA ltomic Energy Agency have
mounted and paginated the original manuscripts and given some attention to presentation.
The views expressed do not necessarily reflect those of the governments of the Member States or
organizations under whose auspices the manuscripts were produced.
n tih eiTssu hbe ook of particular designatio fnocs ountrir eotse rritories dt ooineymns paly
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 mentionf o specific companief o rsot heir productsr o brand names doet osn implynya
endorsementr o recommendatioe nht npo are htt foI AEA.
CONTENTS
1. INTRODUCTION ....................................................................................... 7
2. CORROSION MECHANISM IN THE ABSENCE OF IRRADIATION ..................... 9
2.1. Uniform oxide formation ........................................................................ 9
2.1.1. Introduction ................................................................................. 9
2.1.2. Oxidation kinetics ......................................................................... 13
2.1.3. The mechanism of pretransition oxide growth ...................................... 22
2.1.4. Mechanismf o oxide breakdown .......................................................72.
2.1.5. Post-transition growth ....................................................................13
2.2. Non-uniform (nodular) oxide formation ...................................................... 33
2.2.1. Nodular oxide formation ...............................................................4.3.
2.2.2. Mechanism of nodule formation ........................................................ 36
2.2.3. Simulating nodular corrosion in high temperature water .......................... 38
2.3. Hydrogen absorption .............................................................................. 38
2.3.1. Absorpf thoioyn ds ro.a.gg.e..n ................................................9..3..... .
2.3.2. Hydrogen uptake during oxidation ..................................................... 47
2.3.3. Hydrogen absorption via metallic contacts ........................................... 54
3. RADIATION EFFECTS ON THE CORROSION OF ZIRCONIUM ALLOYS ............ 57
3.1. Influef norcae diat nizooinr conium alloy corrosion phenomena ...................7.5.....
3.1.1. Radiation type, intensity, spectral shift, duration ................................... 57
3.1.2. Coolant chemistry ......................................................................... 58
3.1.3. Alloy composition ......................................................................... 58
3.1.4. Metallurgical condition ................................................................... 58
3.1.5. Surface condition .......................................................................... 58
3.1.6. Temperature ..............................................................................9.5.
3.1.7. Dissimilar metal effects .................................................................. 59
3.1.8. Heat flux .............................................................................9.5..... .
3.2. Effectsf o alloy composition ....................................................................95
3.3. Uniform oxidatiof noz irconium alloys under irradiation ...............................0.6..
3.3.1. Oxidation in irradiated low-oxygen coolants ......................................... 62
3.3.1.1. Oxidation and hydriding of fuel cladding ................................. 62
3.3.1.2. Oxidation and hydriding of reactor pressure tubes ...................... 64
3.3.1.3. Oxidatd ionhany dridinn gtie st reactor loops ........................6.6...
3.3.1.4. Thick-film effen clitos w-oxygen coolants ...........................7.6....
3.3.2. Oxidatio nnii rradiated oxygenated coolants ......................................8.6..
3.3.2.1. Oxidation and hydriding of fuel cladding ................................. 68
3.3.2.2. Oxidation and hydriding of pressure tubes ............................... 68
3.3.2.3. Oxidation and hydriding in test reactor loops ............................ 69
3.4. Localized oxidation and hydriding ............................................................. 70
3.4.1. Nodular oxide formation ................................................................. 70
3.4.2. Alloy element segregation .........................................................1.7.....
3.4.3. Dissimilar metal effects .................................................................. 71
3.4.4. Other localized effects .................................................................... 71
3.5. Significancef o radiation effecto st zirconium alloy behavioun ri service ............2.7.
4. FACTORS AFFECTING IRRADIATION CORROSION ....................................... 73
4.1. Irradiation damage ................................................................................ 73
4.1.1. Fast neutron damage in metals ......................................................... 73
4.1.2. Displacement damn aogiteh er structures ....................................4..7...... .
4.1.3. Efff eoicrt radiat niomon icrostructures .........................................6.7....
4.2. Local radiation chemistry ..................................................................2..8....
4.2.1. Radiolys ehitb sni ulk water ...........................................................28.
4.2.2. Radiolysis near meta elp hostu rnerfisa rcsoeus rroundy mebd etal8 8oxid .e.s
4.2.3. "Thick oxide film effects" ............................................................. 90
4.3. Crud deposition and heat transfer effects ..................................................... 92
4.3.1. PWR crud deposition ..................................................................... 92
4.3.2. WWER crud deposition .................................................................. 109
4.3.3. BWR crud deposition ..................................................................... 109
4.3.4. Effects of heat transfer ................................................................... Ill
4.4. Metallurgical and chemical variables .......................................................... 116
4.4.1. Fabrication variables ...............................................................6..1..1...
4.4.2. Behaviour of alloying additions ........................................................ 119
4.4.3. Electrochemical effects .........................................................1..2...1.... .
4.4.4. Effects of surface treatment of zirconium alloy components ..................... 125
. 5 PRESENT SE MTHATETC FUHOSA NISTIC UNDERSTAENHTD IFNOG
EFFECTSF O IRRADIATION .......................................................................921
5.1. Current understanding of the out-reactor oxidation mechanism .......................... 129
5.1.1. Mobile species ......................................................................9..2..1...
5.1.2. Evolution of oxide morphology ........................................................ 131
5.1.3e .hT developmend tna naturef o oxide porosity .....................................9.31
5.1.4. Oxide barrier layers ...................................................................... 142
5.1.5. Effect of some variables on the oxide structure ..................................... 144
5.2. Empirical correlationf ose ffecf tois rradiation ...........................................7.41..
5.2.1. Development of irradiation corrosion models ....................................... 149
5.2.2. Micromechanismsr of in-reactor corrosion ..........................................65.1
5.2.3. Present status of mechanistic studies .................................................. 157
5.2.4. Recommendations for future work ..................................................... 157
6. CONCLUSIONS .....................................................................................9.51..
APPENDIX .................................................................................................... 161
REFERENCES ................................................................................................ 163
CONTRIBUTORS TO DRAFTING AND REVIEW .................................................. 177
1. INTRODUCTION
e Fbhaatisl iuncri em saterials u nsnieu dclear power plants
continue to be costly and insidious, despite increasing industry vigilance
to catch failures before they degrade safety. For instance, the overall
costs to the US industry from materials problems could amount to as much as
US $10 billion annually. Moreover, estimates indicate that the cost of
1
a pipe f ana uinclulie reh naeuron dprsleaidn ttimes greatere hthta n
cost of a similar failure in a coal-fired plant [1].
So the problem exists and is demanding more attention from utilities
and regulators. According to a recent study on "Materials Science &
Engineeringe ht rof 1990s": "Resumptionf o commercializationf o nuclear
power will require solving many problems related to reliability of
materials for reactors... Particular requirements are (1) a non-destructive
measuree ht fo conditionf o materials (surfaced na bulk) thats i preferably
on-line and continuous and (2) a physically based mechanistic model of
materie aelhn vtbier hoannvmeiinotu "r[1].
elahstt c e ocnfohcnlottuesx I itnon cod rf ramnoleolsatoiayo l sns
under e ipbra rroattdiic austlaiahor lny emphasized. Dese mphaitnt ye years
of experid einnnacv eestigatin otinh sis fielo dcn o tmhsmeiorn e agreement
on either the importance or precise mechanisms of irradiation effects.
Consequently, ta h lesarif ce ock apac rioptfry edictif ooln ong-term
materials behaviour in radiation environment. At the same time, the modern
trends in nuclear power development (extended burnup, increased outlet
temperature, increased lithium concentration, plant life extension,
radiation control, etc.) urgently require an answer to the question: are
the existing materials capablef o guaranteeinga high levef lo reliability
in improved schemes of fuel utilization, or under improved plant operation
conditions, or under extension of the nuclear power plant operation over
the design lifetime [2,3]? As was stated at the IAEA meeting on
fundamental aspectf osc orrosiof onZ r-based alloyn iws ater reactor
environment (Portland, USA, 1989), "clarification of irradiation effects is
the i ftoteo mpmost importance".
Zirconium alloys have been establishe sda vital materialn siw ater
reactor technology, functioning as fuel cladding, pressure tubes, fuel
channels (boxes) and fuel spacer grid materials. Performance of the
zirconium alloysn i services ah generally been satisfactory,t ub there have
been occasional problems, including fuel cladding and pressure tube
degradation and failure. However, a major motivation for this review of
zirconium alloy oxidation and hydriding behaviour under irradiation
invol evhaetns ticipatee dhct o nnretoeifdn uing satisfactory performance
of the zirconium alloys over the next several decades. An element of this
expected se ehprtov tisecine tr iomafol re severe operating requirements,
notably a major trend toward longer residence of some fuel types in reactor
to achieve higher burnups; also some t RrWcePon o ohdtnlis iag nhHtepsr
and hydrogen additions to BWR coolants. Anticipating that from time to time
reactor operators may want to make other service modifications or may be
faced with interpretatiof no unexpected phenomenat ,i seems importanott
develop an improved understanding of the phenomena that have emerged in
some thirty yearf so zirconium alloy utilization. While hetb asic trendnsi
corrosion behavioe ulbra rsogeetel my consiste edhnettv ,elao pmfeon t
mechanistic understanding that provida se osundr ibonatfsei rspretation
and prediction remains only fragmentary. This review intendso t summarize
briefly the key phenomena, the status of the mechanistic understanding, and
to identify the gaps in our understanding and establish a basis for future
1 bi l1li= 1o 0n9.
initiatives to advance this understanding of radiation effects on zirconium
alloy corrosion.
n iAmpoo rattnat ne itccsh ionpniteasete dequf eanoccte isons
before they cause serious degradatf izooi nrconium alloy reactor
components. This strongly suggese thve tsah stl yfusoet ematic
consideratio enht fo consequencesf o changesn i reactor operation no fuel
cladd dnoiatnh ger zirconium alloy components. Sometimes operational
consideran tici oeonrnfas lir coteF.x ample, increasing lithium contenntis
PWR coolants to minimize crud transport at some point begins to jeopardize
the oxidation resistance of zirconium alloys. Early recognition of the
problemd nsas electi ao sfnou itable compromin sacea voia d periofod
expensive retrenchment.
This review includes only information available in the open
literate ushroTeu .r fcdoea sta include:
- oxidatd ihnoayn driding data from fuel cladding, non-fueled
reactor components, and irradiated unfueled specimens;
- relevant data from unirradiated specimens, from specimens exposed
in-t roeuuatcb-tod o far-nlcsoaor e f,rom specimens n tleiasbtoe rdatory
d lanouaotpo sclaves operated with thd ewnramatae lr chemistry parameters
similar to those in-flux;
- supporting data on water radiolysis and radiation damage.
Much of the existing oxidation and hydriding data base has been
developed from examina ftioiror andiated zirconium alloy materials
(specimens, pressure tubes, BWR channels, spacer materials, water tubes)
that operated at low heat transfer rates. To the extent that these data can
be shown to correlate with phenomena on fuel cladding, operating under
elevated heat transfer conditions, they expae nhbtd asr ioifs nterpreting
observations on fuel cladding. Conversely, the fuel cladding results offer
potential contrn iiabnu tteoigortna sted data base s taihpap tlicable over
a wide range of heat transfer conditions. At the same time, it is important
to identify and understand oxidation and hydriding phenomena that result
uniquely from heat transfer effects.
Some alloy systems seem to be pressed to near their operational
lir mroiefta sctor serv nisico eme regimesA .better understandf iaonl gloy
yallow aod piemrpautmriinotnya le fifamepcrtosveme elnetmse nwtithout
major initiatives to qualify completely different alloy systems. Nuclear
reactor operating regimes includa e rangf owe ater chemistries, including
hydrogenated, low oxygen (e.g. PWR, WWER), marginally low oxygen (e.g.
PHd WoRnx)ay ,genated conditions (e.g. BWR, RBMK). Some phenodmnean a
oxidation kinetice rsua niquo ets ome regimes n.I some cases radiatiosni
clearly e iohnxtvio dlnavitei don kinetin coIst .her cases, effef chote sat
transfer, coolant species (e.g. Li+) and radiation remain a matter of
controversy.
These are some examples of the need for a more fundamental
understanding of zirconium alloy behaviour, with the opportunity for
practical applications. Thus, theree ra important practical stimulidna
much sr cfooufpr ether underse tehaftnf defif coniot gr sradiatnioo n
Zr-alloys (and other materials usedn i nuclear installations)y b careful
experimentation. Moreover, these studies will need to address the effect of
irradiation on all components of heterogeneous systems: the metal, the
oxide and the environment, and especially those processes recurring at the
interphases between these components.
The present review is aimed at providing specialists with some
systematic information on the subject and with important considerations on
they ek itemsr of further experimentation.
8
2. CORROSION MECHANISM IN THE ABSENCE OF IRRADIATION
2.1. UNIFORM OXIDE FORMATION
2.1.1. Introduction
Bece hattuh fseore modyne ahmZtri f.coOs system (Figl.l2a. 1t)on,
the oxygen that reacts forms oxide, some dissolves in the matrix [4].
e rmae r3.e2Fl idyng aue r2x.ea2sm ples from amoe hnmtga ny microhardness
plots available fre ohtlm iterature.
100 -
_99
150 _
- 800
200 _ / ^*Zr+0-05°2^Zr°0.1 a
3
5
A!
- 900
. 1OOO
-20
250 .
- 1100
1200
300 -
1300
500 1000 1500 2000 2500k
FIG. 2.1. Darken and Gurry plot of thermodynamics ofZr/O and other relevant systems [4].
The fraction dissolving depends on the balance between the kinetics
of oxidation and the kinetics of dissolution [5]. Thus, since the
dissolution process seo evtma sry much less from alo latol yloy ethhta n
oxidation process, rapidly oxidizing alloys should have shallower diffusion
zones than slowly oxidizing alloys at all temperatures, although the
abilio ttmy easure these diffusion zw ootlne etmasp eratur seivs ery
limited.y nA discontinuitye ht ni oxidation kinetics will result eitherni
growth or diminution of the oxygen diffusion zone.
e fhrTactf iooox nygen dit swsoeonll vlsi ink gnor waonc ,cepted,
evet na high temperatures whe sir hteii ghest because heta ctivation energy
for dissolution is higher than the activation energy for oxide growth
(Table 2.1) [5]. wol tAt emperatures neithere ht fraction dissolvestid ron
distribus twiieol nl known. Dissoe llbuos t ckiiaonolo ttniwa zne d
i ~6e 0het0 x°ttuCbe nf ott his localizatt owin eosiln l known.
Preferential dissolution along grain boundaries has been demonstrated at
~600°C by the nuclear reaction:
18o_P,n_18ß ( 118 min) 18 (Fig. 2.4)
F > n