Materials challenges in nuclear energy
S.J.Zinkle a ,⇑,G.S.Was b
a
Oak Ridge National Laboratory,P.O.Box 2008,Oak Ridge,TN 37831,USA
b
Nuclear Engineering and Radiological Sciences Department,University of Michigan,Ann Arbor,MI 48109,USA
Abstract
Nuclear power currently provides about 13%of electrical power worldwide,and has emerged as a reliable baseload source of elec-tricity.A number of materials challenges must be successfully resolved for nuclear energy to continue to make further improvements in reliability,safety and economics.The operating environment for materials in current and proposed future nuclear energy systems is summarized,along with a description of materials used for the main operating components.Materials chal
lenges associated with power uprates and extensions of the operating lifetimes of reactors are described.The three major materials challenges for the current and next generation of water-cooled fission reactors are centered on two structural materials aging degradation issues (corrosion and stress corrosion cracking of structural materials and neutron-induced embrittlement of reactor pressure vessels),along with improved fuel system reliability and accident tolerance issues.The major corrosion and stress corrosion cracking degradation mechanisms for light-water reactors are reviewed.The materials degradation issues for the Zr alloy-clad UO 2fuel system currently utilized in the majority of commercial nuclear power plants are discussed for normal and off-normal operating conditions.Looking to proposed future (Gen-eration IV)fission and fusion energy systems,there are five key bulk radiation degradation effects (low temperature radiation hardening and embrittlement;radiation-induced and -modified solute segregation and phase stability;irradiation creep;void swelling;and high-temperature helium embrittlement)and a multitude of corrosion and stress corrosion cracking effects (including irradiation-assisted phe-nomena)that can have a major impact on the performance of structural materials.Ó2012Acta Materialia Inc.Published by Elsevier Ltd.All rights reserved.
Keywords:Nuclear materials;Radiation effects;Stress corrosion cracking;Structural alloys (steels and nickel base);Nuclear fuels
1.Introduction
Access to reliable,sustainable and affordable energy is viewed as crucial to worldwide economic prosperity and stability [1,2].Nuclear fission energy has emerged over the past 40years to become a reliable baseload source of clean and economical electrical energy.As of 2011,there were 435nuclear reactors in operation worldwide,produc-ing 370GW e of electricity [3].Another 108units or 108GW e are forthcoming (under construction or on order),for a total of 543units and 478GW e of electrical capacity.The largest producer of power from nuclear energy is the USA,with 104commercial reactors licensed to operate at 65sites,producing a total of 103GW e of
electricity.These provided just under 20%of the nation’s total electric energy generation and more than 30%of worldwide nuclear generating capacity.Worldwide,nuclear energy provides about 13%of the electrical demand [1].Given that nuclear power has very low carbon emission [2]and that energy generation currently accounts for 66%of worldwide greenhouse gas emissions [4],nuclear energy is considered an important resource in managing atmospheric greenhouse gases and associated climate change [1].
The core of a nuclear reactor presents an exceptionally harsh environment for materials due to the co
mbination of high temperature,high stresses,a chemically aggressive coolant and intense radiation fluxes.Many of the features that make reactors attractive from a physics perspective (e.g.high specific power,self-sustaining reaction)exert high operational burdens on structural materials.For example,
1359-6454/$36.00Ó2012Acta Materialia Inc.Published by Elsevier Ltd.All rights reserved./10.1016/j.actamat.2012.11.004
⇑Corresponding author.Tel.:+18655765785.
E-mail address:v (S.J.Zinkle).
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Acta Materialia 61(2013)
735–758
the recoverable energy from each235Ufission reaction is $200MeV,which is about eight orders of magnitude per atom higher than typical chemical reactions.As a result, typical power densities in commercial nuclear reactor cores are$50–75MW th mÀ3,which is nearly two orders of mag-nitude higher than the average power density in the boiler furnace of a large-scale coal power plant.This intense pro-duction of heat is accompanied by the generation of ener-getic neutrons(which serve to sustain thefission reaction) and gamma radiation,which can degrade materials by dis-placement damage and radiolysis processes,respectively. Recent activities to extend the operating lifetime of current water reactors,to develop advancedfission reactor con-cepts with greater functionality and capability,and the coming emergence of fusion energy represent even greater demands on materials[5–8].
1.1.Types of nuclearfission reactors
The predominant reactor design worldwide is the pres-surized water reactor(PWR),accounting for two-thirds of the installed capacity,followed by boiling water reactors (BWRs)at21%and heavy-water reactors at14%of installed capacity,respectively(Table1)[3].All of these water-cooled reactors use ceramic fuel pellets consisting of UO2or otherfissile actinide oxides to generate heat. The ceramic pellets are stacked inside of long Zr alloy tubes (fuel cladding)that transfer the nuclear heat toflowing water coolant and serve as the primary barrier containing the volatile radioactivefission byproducts.The remaining 5%of installed nuclear energy comes from gas-cooled reac-tors,graphite-moderated reactors and liquid metal cooled reactors(Table1).
The vast majority of the reactors listed in Table1are classified as Generation II reactors[9],which were designed in the1960s and predominantly achieved initial commercial operation from the1970s through the1990s.These reactors are distinguished from Generation I designs(1950s-60s), which were early commercial prototype and demonstration reactors,and Generation III reactors,designed in the1990s to incorporate significant advances in safety and economics [9].Generation III reactor construction for the past decade has been centered in Asia,with a few units recently built in Europe.The current generation of light-water reactors (LWRs),Generation III+,include still further advance-ment in economics and safe
ty,such as passive heat removal systems.There are a total of108Generation III and Generation III+reactors on order or under construction around the world,and of those,89are PWRs.
Given the high representation of PWRs and BWRs in the world’sfleet,materials issues in these two types of reac-tors are of greatest interest.And of the many materials in a reactor,those that experience the most extreme conditions (stress,corrosion,and radiation)are most important for maintaining plant safety and reliability.Fig.1shows a schematic of the major components in the primary and secondary circuits of a PWR[10].Pressurized water ($15.5MPa)in the primary circuit enters the reactor core at$275°C,picks up heat from the reactor core with a core exit temperature of$325°C,and transfers the heat across the U-tubes in the steam generator to water at a lower pressure.This water turns to steam that powers the tur-bine,and is condensed and recirculated.Fig.1also lists the alloys used throughout the primary and secondary cir-cuits,all of which are in contact with high-temperature water and are subject to significant mechanical stress. Alloys inside(and including)the reactor vessel are also subject to varying levels of radiation,which produces displacement damage and radiolytic decomposition of the coolant water.Major pressure boundary components (reactor pressure vessel,pressurizer,steam generator, steam lines,turbine and condenser)are made of either low carbon or low alloy steel.Austenitic stainless steels (Types304,304L,316,316L,321,347)dominate the core structural materi
als,as well as serving for cladding(308SS and309SS)on the inside surface of the reactor pressure vessel and pressurizer.Higher strength components such as springs and fasteners are made of nickel-base alloys. Vessel penetrations and steam generator tubes are made of nickel-base alloy690(previously alloy600,which was found to provide insufficient resistance to stress corrosion cracking).Condenser tubes are generally made of titanium or stainless steel.The selection of nickel-base alloys and austenitic stainless steels for core internals and the steam generator tubes is driven by the need for good aqueous corrosion resistance at high temperatures.These alloys have low corrosion rates due to the formation of chro-mium-bearing spinels that form adherent,high-density protective surface layers that grow very slowly at operating temperatures.
Table1
Power reactors by type,worldwide[3].
Reactor type#Units Net MW e#Units Net MW e#Units Net MW e
(in operation)(forthcoming)(total)
Pressurized light-water reactors(PWR)267246555.18993,014356339569.1 Boiling light-water reactors(
BWR)8478320.6680569086376.6 Gas-cooled reactors,all models178732.01200188932.0 Heavy-water reactors,all models5125610.0851125930722.0 Graphite-moderated reactors,all models1510219.0001510219.0 Liquid-metal-cooled reactors,all models1560.0410*******.0 Totals435369996.7108107,896543477894.7 736S.J.Zinkle,G.S.Was/Acta Materialia61(2013)735–758
The main difference between PWRs and BWRs is that the latter consists of a single water circuit designed for boil-ing to occur in the core with steamflowing directly to the turbine,which eliminates the steam generator and pressur-izer found in the PWR.The operating temperatures are comparable for both reactor types($300°C),with compa-rable stress and radiation environments.As such,most of the structural alloys are very similar between the two reac-tor types.The main difference is in the zirconium alloys used as fuel rod cladding,with BWR fuel cladding opti-mized for corrosion resistance in higher oxygen potentials and PWR fuel cladding optimized for resistance to hydro-gen absorption in the low potential environment of the core.Typical zirconium alloy cladding materials used in BWR and PWR reactors are summarized in Table2. Differences in oxygen potential result in significant impacts on the stress corrosion degradation of materials through-out the water circuit in both reactor types,as will be discussed in Section2.1.
The last reactor design that is in significant use world-wide is the pressurized heavy water reactor(PH
WR),the most prevalent version being the CANDU(CANadian Deuterium Uranium)reactor.This reactor uses heavy water as the moderator and primary coolant,transferring heat to light water via a steam generator.The key charac-teristic of this reactor is the use of deuterium as a modera-tor,for which neutron absorption is low enough to permit the use of natural(unenriched)uranium,thus bypassing the need for expensive enrichment facilities.A major differ-ence in materials in this system vs.LWRs is the use of Zr–Nb pressure tubes that house the Zircaloy-clad fuel and the high pressure D2O.These tubesfit into Zircaloy-4calan-dria tubes that pass through a thin walled stainless steel calandria vessel,which also contains the low
temperature Table2
Summary of typical commercial zirconium alloys used as cladding in PWRs and BWRs.
Reactor type Zr alloy composition Thermomechanical treatment
BWR Zircaloy-2(1.5%Sn–0.15%Fe–0.1%Cr–0.05%Ni)Recrystallized
PWR Zircaloy-4(1.5%Sn–0.2%Fe–0.1%Cr)Cold-worked and stress relief anneal PWR ZIRLO(1–2%Nb–1%Sn–0.1%Fe)Quench and temper/stress relief anneal PWR M5(1%Nb)Recrystallized
D2O moderator.Thus,zirconium alloys play a larger role as pressure boundary materials in PHWRs than they do in LWRs.
Most reactors in the USA and elsewhere in the world were completed in the1970s and1980s,and today the aver-age age of thefleet is over30years.Fig.2shows the world-wide distribution of nuclear power plants classified by years of commercial operation[11].Since the original license period in the USA is40years,many reactor opera-tors are seeking license renewal to allow them to operate the plants for an additional20years.To date,73of the 104operating commercial reactors in the USA have received license extensions with another13applications under review,and a key question is how long can these reactor pressure中文
plants be safely,reliably and economically operated.The limiting factor is whether critical materials can continue to maintain their integrity beyond60years[5].These mate-rials include reactor components,concrete,cables and bur-ied piping.So the lifetime of the current reactorfleet is ultimately governed by the performance of materials.
1.2.Major materials degradation modes in nuclear energy systems
In addition to satisfying standard materials design crite-ria based on tensile properties,thermal creep,cyclic fatigue and creep-fatigue,structural materials for current and pro-posed future nuclear energy systems must provide adequate resistance to two additional overarching environmental
There arefive key bulk radiation degradation effects (low temperature radiation hardening and embrittlement; radiation-induced and-modified solute segregation and phase stability(including amorphization);irradiation creep;void swelling;and high-temperature helium embrit-tlement)[8,12–16],and a multitude of corrosion and stress corrosion cracking effects in water-cooled reactors[13,17–22]and proposed advanced reactors utilizing other cool-ants[23–26](including irradiation-assisted phenomena) that can have a huge impact on the performance of struc-tural materials in nuclear energy systems.The amount of radiation damage produced in materials from exposure to neutrons cre
ated by the nuclear energy reactions is quanti-fied by the international standardized parameter[27,28]of displacements per atom(dpa);a displacement damage value of1dpa means that,on average,each atom has been displaced from its lattice site once.
Neutron irradiation can produce pronounced hardening at low and intermediate irradiation temperatures due the production of high densities of nanoscale defect clusters (dislocation loops,helium bubbles,etc.),which serve as obstacles to dislocation motion.This hardening is generally accompanied by a reduction in tensile elongation and frac-ture toughness.The radiation hardening and reductions in elongation and fracture toughness typically emerge at dam-age levels above$0.1dpa and are generally most pro-nounced for homologous irradiation temperatures below 0.35T M,where T M is the absolute melting temperature [26,29–35].Fig.3shows an example of the effect of moder-ate neutron displacement damage levels on the engineering stress–strain curve for austenitic stainless steel[36]and a8–9%Cr-tempered martensitic steel[35]at250°C.Both materials exhibit significant radiation-induced increases in yield and ultimate tensile stress,large reductions in elonga-tion(particularly uniform elongation)and decreased strain hardening capacity.The reductions in elongation and strain hardening capacity have been attributed toflow dislocation channeling)[37–44]and strain hardening exhaustion[29–31]mechanisms.In addition to the decreased el
ongation,neutron irradiation at low tem-perature also generally produces a decrease in fracture toughness.Fig.4summarizes some of the fracture tough-ness data for Types304and316austenitic stainless steels following irradiation at LWR-relevant conditions near 250–350°C[32,36,45–48].The fracture toughness decreases rapidly with increasing irradiation dose,and approaches a value near50MPa m1/2after5–10dpa.The reduction in fracture toughness can be of particular concern for body-centered cubic materials such as ferritic/martensitic steels if the ductile to brittle transition temperature is shifted to temperatures above cold or warm standby temperatures. The potential for neutron radiation-induced embrittlement of reactor pressure vessel steels has been intensively inves-tigated due to its importance for public safety[49].
At intermediate temperatures(homologous tempera-tures>0.3T M),the increased mobility of the radiation defects produces a diverse range of potential microstruc-
Age distribution of the world’s commercial nuclear power reactors
December2011[11].
738S.J.Zinkle,G.S.Was/Acta Materialia61(2013)735–758
precipitation in austenitic stainless steel for temperatures
as low as300°C[50].Void swelling(due to nucleation and growth of the supersaturation of vacancies produced by irradiation)is characterized by an initial low-swelling transient regime at low doses(during the void nucleation and initial growth phase),followed by a steady-state swell-ing regime where the volumetric swelling increase is pro-Fig.5.Precipitate phases observed in Type316austenitic stainless steel after neutron irradiation as a function of temperature and dose.Partially shaded data points at temperatures<400°C denote the presence of c phase and solid data points are for either G and related phases or an unidentified phase[50].

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