Materials challenges in nuclear energy

Available online at www.sciencedirect.com

Acta Materialia 61 (2013) 735–758 www.elsevier.com/locate/actamat

Materials challenges in nuclear energy S.J. Zinkle a,⇑, G.S. Was b b

a Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831, USA 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 electricity. 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 challenges 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 UO2 fuel system currently utilized in the majority of commercial nuclear power plants are discussed for normal and off-normal operating conditions. Looking to proposed future (Generation 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 hightemperature helium embrittlement) and a multitude of corrosion and stress corrosion cracking effects (including irradiation-assisted phenomena) that can have a major impact on the performance of structural materials. Ó 2012 Acta 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 40 years to become a reliable baseload source of clean and economical electrical energy. As of 2011, there were 435 nuclear reactors in operation worldwide, producing 370 GWe of electricity [3]. Another 108 units or 108 GWe are forthcoming (under construction or on order), for a total of 543 units and 478 GWe of electrical capacity. The largest producer of power from nuclear energy is the USA, with 104 commercial reactors licensed to operate at 65 sites, producing a total of 103 GWe of ⇑ Corresponding author. Tel.: +1 865 576 5785.

E-mail address: [email protected] (S.J. Zinkle).

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 combination 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/$36.00 Ó 2012 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved. http://dx.doi.org/10.1016/j.actamat.2012.11.004

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the recoverable energy from each 235U fission reaction is 200 MeV, 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 50–75 MWth m3, which is nearly two orders of magnitude higher than the average power density in the boiler furnace of a large-scale coal power plant. This intense production of heat is accompanied by the generation of energetic neutrons (which serve to sustain the fission reaction) and gamma radiation, which can degrade materials by displacement damage and radiolysis processes, respectively. Recent activities to extend the operating lifetime of current water reactors, to develop advanced fission reactor concepts with greater functionality and capability, and the coming emergence of fusion energy represent even greater demands on materials [5–8]. 1.1. Types of nuclear fission reactors The predominant reactor design worldwide is the pressurized water reactor (PWR), accounting for two-thirds of the installed capacity, followed by boiling water reactors (BWRs) at 21% and heavy-water reactors at 14% of installed capacity, respectively (Table 1) [3]. All of these water-cooled reactors use ceramic fuel pellets consisting of UO2 or other fissile actinide oxides to generate heat. The ceramic pellets are stacked inside of long Zr alloy tubes (fuel cladding) that transfer the nuclear heat to flowing water coolant and serve as the primary barrier containing the volatile radioactive fission byproducts. The remaining 5% of installed nuclear energy comes from gas-cooled reactors, graphite-moderated reactors and liquid metal cooled reactors (Table 1). The vast majority of the reactors listed in Table 1 are classified as Generation II reactors [9], which were designed in the 1960s and predominantly achieved initial commercial operation from the 1970s through the 1990s. These reactors are distinguished from Generation I designs (1950s-60s), which were early commercial prototype and demonstration reactors, and Generation III reactors, designed in the 1990s 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 safety, such as passive heat removal systems. There are a total of 108 Generation III and Generation III+ reactors on order or under construction around the world, and of those, 89 are PWRs. Given the high representation of PWRs and BWRs in the world’s fleet, materials issues in these two types of reactors 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. 1 shows a schematic of the major components in the primary and secondary circuits of a PWR [10]. Pressurized water (15.5 MPa) 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 turbine, and is condensed and recirculated. Fig. 1 also lists the alloys used throughout the primary and secondary circuits, 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 (Types 304, 304L, 316, 316L, 321, 347) dominate the core structural materials, as well as serving for cladding (308SS and 309SS) 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 alloy 690 (previously alloy 600, 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 chromium-bearing spinels that form adherent, high-density protective surface layers that grow very slowly at operating temperatures.

Table 1 Power reactors by type, worldwide [3]. Reactor type

# Units (in operation)

Net MWe

Pressurized light-water reactors (PWR) Boiling light-water reactors (BWR) Gas-cooled reactors, all models Heavy-water reactors, all models Graphite-moderated reactors, all models Liquid-metal-cooled reactors, all models

267 84 17 51 15 1

246555.1 78320.6 8732.0 25610.0 10219.0 560.0

89 6 1 8 0 4

Totals

435

369996.7

108

# Units (forthcoming)

Net MWe

# Units (total)

Net MWe

93,014 8056 200 5112 0 1016

356 90 18 59 15 5

339569.1 86376.6 8932.0 30722.0 10219.0 2076.0

107,896

543

477894.7

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Fig. 1. Schematic of the primary and secondary circuits of a pressurized water reactor and materials of construction (courtesy of R.W. Staehle) [10].

The main difference between PWRs and BWRs is that the latter consists of a single water circuit designed for boiling to occur in the core with steam flowing directly to the turbine, which eliminates the steam generator and pressurizer found in the PWR. The operating temperatures are comparable for both reactor types (300 °C), with comparable stress and radiation environments. As such, most of the structural alloys are very similar between the two reactor types. The main difference is in the zirconium alloys used as fuel rod cladding, with BWR fuel cladding optimized for corrosion resistance in higher oxygen potentials and PWR fuel cladding optimized for resistance to hydrogen absorption in the low potential environment of the core. Typical zirconium alloy cladding materials used in BWR and PWR reactors are summarized in Table 2. 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 Section 2.1. The last reactor design that is in significant use worldwide is the pressurized heavy water reactor (PHWR), 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 characteristic of this reactor is the use of deuterium as a moderator, 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 difference 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 tubes fit into Zircaloy-4 calandria tubes that pass through a thin walled stainless steel calandria vessel, which also contains the low temperature

Table 2 Summary of typical commercial zirconium alloys used as cladding in PWRs and BWRs. Reactor type

Zr alloy composition

Thermomechanical treatment

BWR PWR PWR PWR

Zircaloy-2 (1.5% Sn–0.15% Fe–0.1% Cr–0.05% Ni) Zircaloy-4 (1.5% Sn–0.2% Fe–0.1% Cr) ZIRLO (1–2% Nb–1% Sn–0.1% Fe) M5 (1% Nb)

Recrystallized Cold-worked and stress relief anneal Quench and temper/stress relief anneal Recrystallized

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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 the 1970s and 1980s, and today the average age of the fleet is over 30 years. Fig. 2 shows the worldwide distribution of nuclear power plants classified by years of commercial operation [11]. Since the original license period in the USA is 40 years, many reactor operators are seeking license renewal to allow them to operate the plants for an additional 20 years. To date, 73 of the 104 operating commercial reactors in the USA have received license extensions with another 13 applications under review, and a key question is how long can these plants be safely, reliably and economically operated. The limiting factor is whether critical materials can continue to maintain their integrity beyond 60 years [5]. These materials include reactor components, concrete, cables and buried piping. So the lifetime of the current reactor fleet 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 criteria based on tensile properties, thermal creep, cyclic fatigue and creep-fatigue, structural materials for current and proposed future nuclear energy systems must provide adequate resistance to two additional overarching environmental degradation phenomena: radiation damage and chemical compatibility. Since the chemical compatibility issues (corrosion, stress corrosion cracking, etc.) are largely dependent on the specific coolant and engineering application, these issues are discussed in the relevant sections on lightwater reactors (2.2.2) and advanced reactor concepts (3.2 and 4).

Fig. 2. Age distribution of the world’s commercial nuclear power reactors as of December 2011 [11].

There are five 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 embrittlement) [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 coolants [23–26] (including irradiation-assisted phenomena) that can have a huge impact on the performance of structural materials in nuclear energy systems. The amount of radiation damage produced in materials from exposure to neutrons created by the nuclear energy reactions is quantified by the international standardized parameter [27,28] of displacements per atom (dpa); a displacement damage value of 1 dpa 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 fracture toughness. The radiation hardening and reductions in elongation and fracture toughness typically emerge at damage levels above 0.1 dpa and are generally most pronounced for homologous irradiation temperatures below 0.35TM, where TM is the absolute melting temperature [26,29–35]. Fig. 3 shows an example of the effect of moderate neutron displacement damage levels on the engineering stress–strain curve for austenitic stainless steel [36] and a 8– 9% Cr-tempered martensitic steel [35] at 250 °C. Both materials exhibit significant radiation-induced increases in yield and ultimate tensile stress, large reductions in elongation (particularly uniform elongation) and decreased strain hardening capacity. The reductions in elongation and strain hardening capacity have been attributed to flow localization (e.g. dislocation channeling) [37–44] and strain hardening exhaustion [29–31] mechanisms. In addition to the decreased elongation, neutron irradiation at low temperature also generally produces a decrease in fracture toughness. Fig. 4 summarizes some of the fracture toughness data for Types 304 and 316 austenitic 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 near 50 MPa m1/2 after 5–10 dpa. The reduction in fracture toughness can be of particular concern for bodycentered 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 investigated due to its importance for public safety [49]. At intermediate temperatures (homologous temperatures >0.3TM), the increased mobility of the radiation defects produces a diverse range of potential microstruc-

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Fig. 4. Fracture toughness of Types 304 and 316 austenitic stainless steels following irradiation at LWR-relevant conditions near 250–350 °C [32,36,45–48].

Fig. 3. Effect of neutron irradiation to 3 dpa on the engineering stress– strain curves for (a) solution annealed Type 316LN austenitic steel and (b) 8–9% Cr-tempered martensitic steel at 250 °C (based on Refs. [36] and [35], respectively).

tural evolutions. The most important radiation degradation phenomena at intermediate temperatures are radiation-induced solute segregation (and associated radiation-induced or -modified precipitation), void swelling, irradiation creep and anisotropic growth. Fig. 5 summarizes the numerous radiation-induced phases that can be induced in initially single-phase austenitic stainless steel as a result of localized radiation-induced solute segregation processes during neutron irradiation [50]. Initial investigations indicated that radiation-induced precipitation was limited to temperatures above 400 °C [51,52], but recent long-term experiments have observed radiation-induced precipitation in austenitic stainless steel for temperatures as low as 300 °C [50]. Void swelling (due to nucleation and growth of the supersaturation of vacancies produced by irradiation) is characterized by an initial low-swelling

Fig. 5. Precipitate phases observed in Type 316 austenitic stainless steel after neutron irradiation as a function of temperature and dose. Partially shaded data points at temperatures <400 °C denote the presence of c0 phase and solid data points are for either G and related phases or an unidentified phase [50].

transient regime at low doses (during the void nucleation and initial growth phase), followed by a steady-state swelling regime where the volumetric swelling increase is pro-

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portional to the dose [12,16,53–55]. Typical post-transient steady-state swelling rates in irradiated metals are 0.2– 1% dpa1, which would produce unacceptable volumetric swelling in structural components exposed to high neutron doses. Therefore, research has focused on identifying mechanisms that extend the low-swelling transient regime and delay the onset of the steady-state swelling regime [55,56]. Irradiation creep [12,53,57–60] and irradiation growth [58–61] can cause substantial dimensional changes in addition to changes due to void swelling. Irradiation growth is mainly an issue in anisotropic crystallographic systems, such as hexagonal close-packed materials; for this phenomenon, volume is conserved but pronounced anisotropic expansion in one crystallographic direction (and shrinkage in another direction) can occur due to preferential nucleation of defect clusters, such as dislocation loops, on certain crystallographic habit planes. Materials for nuclear energy systems that exhibit irradiation growth include graphite and pure metals or alloys based on zirconium and beryllium. The amount of deformation from irradiation creep is typically proportional to the applied stress and irradiation exposure, with a steady-state creep compliance coefficient of 0.5 to 1  106 MPa1 dpa1 for ferritic and austenitic steels, respectively [12]. Another consequence of irradiation creep is that it can induce undesirable stress relaxation of bolts or springs. Fig. 6 shows the measured stress relaxation for neutron irradiated Inconel X750 springs [62]. Nearly complete relaxation of the initially applied stress on the springs occurred after an irradiation dose of 20 dpa at 400 °C. At high temperatures (above 0.5–0.6TM) the efficient annealing of lattice defects produces recovery of most of the radiation damage. One notable exception is associated with the transmutant He produced from (n, a) reactions within the material. The helium can diffuse to grain boundaries, where it can form large bubbles that weaken the grain boundary strength and cause dramatic reductions in the total elongation [63–66]. This phenomenon of high-temperature helium embrittlement may restrict the upper operat-

ing temperature of materials in nuclear energy systems to temperatures significantly lower than what would be established by thermal creep strength considerations. 2. Materials challenges in current commercial fission reactors 2.1. Operating environment for materials in existing LWRs Materials in LWRs are exposed to a variety of conditions. In the following, the operating environment for normal, extended life and transient conditions are summarized. Used fuel disposition issues, while important, are not discussed in this paper. 2.1.1. LWR materials under normal operating conditions Core materials include both fuel materials and structural components. The fuel consists of UO2 pellets in the shape of right circular cylinders with length and diameter of approximately 1 cm each, loaded into 3–4 m long zirconium alloy fuel tubes (cladding), which are grouped into fuel assemblies containing control rods or blades. In BWRs, the assemblies generally contain approximately 100 fuel rods arranged in a square array. Each BWR assembly is encased in a square zirconium alloy tube or fuel channel approximately 12 cm on a side. The control rods are clustered into cruciform-shaped control blades that pass between assemblies clustered in groups of four, as illustrated in Fig. 7. There are typically 700–800 fuel assemblies in a BWR. PWRs contain fewer (200), but larger (21 cm on a side) assemblies, containing up to 300 fuel rods of slightly smaller diameter than those in BWRs. The PWR

1.2 Inconel X750 375-415°C

Stress Relaxation

1

-8

-8

3x10 to 7x10 dpa/s

0.8 0.6 0.4 0.2 0

0

5

10

15

20

25

Dose (dpa) Fig. 6. Stress relaxation (normalized to the initial applied stress) for Inconel X750 springs irradiated in the EBR-II fast fission reactor [62].

Fig. 7. Fuel assemblies and control blade used in a boiling water reactor (image credit GE).

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control rods are distributed throughout the square lattice and are connected to each other to form a control rod cluster, as shown in Fig. 8. In both cases, control rods consist of stainless steel tubes filled with boron carbide for neutron absorption. There is no channel box around a PWR assembly, so cross-flow of water between assemblies is possible. The PWR and BWR cores are typically operated nonstop for 18–24 months between refueling operations. Low and intermediate burn-up fuel assemblies are typically moved to different positions in the core during refueling outages to provide optimized fuel management, with a total core residence period of 3–4 fuel cycles (i.e. a third to a quarter of the fuel assemblies are removed each refueling cycle) until they achieve typical cumulative burn-up levels of 40–60 GW days per metric ton of uranium (GWd MTU1), corresponding to fissions in 4.2–6.4% of the original uranium atoms. Non-fuel core components consist of major structures such as the core shroud (BWR) or the baffle–former assembly (PWR), and smaller components such as bolts, springs, support pins and clips. The core shroud in a BWR is a cylindrical barrel, open at both ends, that surrounds the fuel assemblies. Water from the condenser mixes with water recirculated from the core between the shroud near the top of the vessel and is channeled down the annulus formed by the shroud and the reactor pressure vessel (RPV), then up through the fuel assemblies. In a PWR, the baffle–former assembly plays a similar role in forcing the incoming water down the annulus formed between the baffle–former assembly and the RPV inner diameter and up through the fuel channels to remove heat from fission. Moving out from the core, additional key components are the control rod drive mechanisms and housing, and vessel head penetrations consisting of welded austenit-

Fig. 8. Fuel assembly and control rod cluster used in a pressurized water reactor (image credits: Commissariat a` l0 e´nergie atomique and Westinghouse).

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ic stainless steel or nickel-base nozzles in the reactor head (PWR) or bottom (BWR), and the RPV. The RPV serves as both a pressure barrier and a containment barrier for the radioactive fission products produced during the nuclear fission reaction, and thus plays a key role in reactor safety. The RPV is typically constructed from carbon and low alloy ferritic steels with 1– 2% Mn, 0.5–1% Ni, 0.5% Mo and 0.15–0.4% Si [67], and has a typical wall thickness of 20 cm. Older LWR pressure vessels were constructed from rolled plates that were welded to form a large cylinder, whereas newer vessels are formed from ring forgings in order to eliminate welds in the vessel “beltline” region closest to the center of the reactor core. The top and bottom heads are usually constructed from low alloy steel forgings, and are welded to the central cylindrical vessel (or bolted and gasketed in the case of the upper head). The internal surface of the RPV is typically clad with 5–10 mm of an austenitic stainless steel to provide corrosion compatibility with the reactor coolant. Multiple penetrations for coolant flow and instrumentation are made through the pressure vessel. The fast neutron flux is three to four orders of magnitude lower at the RPV compared to core internal structures [67], but it is still of sufficient intensity to cause radiation hardening, which could lead to fracture toughness embrittlement. It is important to maintain adequate levels of fracture toughness for a wide variety of operational conditions, including normal operation, cold shutdown for refueling and other maintenance, and postulated transient accident scenarios such as pressurized thermal shock in PWRs that would introduce cold water into the reactor vessel while the vessel is at operating pressure and temperature (creating large thermal stresses and potential for crack propagation). Materials utilized in LWR cores must withstand simultaneous application of mechanical stress, neutron irradiation and corrosion due to hot water or steam (see rows 1 and 2 of Table 3). Temperatures of core components are in the range 275–288 °C in BWRs and 290–320 °C in PWRs. The BWR environment is characterized by an electrochemical potential (ECP) in the range of 150 mV relative to the standard hydrogen electrode, or 150 mVSHE, due to a combination of boiling of the water in the core and radiolysis. PWRs operate at a lower potential (<500 mVSHE) by virtue of the addition of hydrogen at a level of 35 cc of H2 per kg of water (3 ppm) to scavenge radiolysis products and lower the corrosion potential. PWR primary water also contains 1000 ppm B as boric acid (H3BO3) added for reactivity control and 2–4 ppm Li as LiOH added for pH control. The lower ECP is better for both corrosion and stress corrosion cracking of core materials and is possible in a PWR due to the lack of boiling. In addition to controlling pH, boron impacts the formation of solid corrosion products (CRUD) and corrosion of fuel cladding, as well as reactivity control of the reactor, and is briefly described in Section 2.2. Stresses on fuel and core components come from a variety of sources, including thermal expansion, high velocity water

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Table 3 Reactor core environment and materials for light water reactors and advanced fission reactor concepts [7]. System

Sodium fast reactor – SFR Lead fast reactor – LFR Molten salt reactor – MSR

Pressure (MPa)

Tin/ Tout (°C)

Neutron spectrum, maximum dose (dpa)

Fuel

Water – single phase

16

290/ 320

Thermal, 80

UO2 (or MOX)

Water – two phase

7

280/ 288

Thermal, 7

Supercritical water

25

290/ 600

Helium

7

Helium, supercritical CO2 Sodium

Cladding

Structural materials In-core

Out-of-core

Zirconium alloy

Stainless steels, nickelbased alloys

Stainless steels, nickel-based alloys

UO2 (or MOX)

Zircaloy

Stainless steels, nickelbased alloys

Stainless steels, nickel-based alloys

Thermal, 30, fast, 70

UO2

Same as cladding options, plus low swelling stainless steels

F-M, low-alloy steels

600/ 1000

Thermal, <20

UO2, UCO

F-M (12Cr, 9Cr, etc.) (Fe– 35Ni–25Cr–0.3Ti), Incoloy 800, ODS, Inconel 690, 625, and 718 SiC or ZrC coating and surrounding graphite

Graphites, PyC, SiC, ZrC, vessel: F-M

7

450/ 850

Fast, 80

MC

Ceramic

0.1

370/ 550

Fast, 200

F-M or F-M ODS

Lead or lead– bismuth

0.1

600/ 800

Fast, 150

MOX or U– Pu–Zr or MC or MN MN

Ni-based superalloys, 32Ni–25Cr– 20Fe–12.5W–0.05C, Ni–23Cr–18W– 0.2C, F-M w/thermal barriers, lowalloy steels Ni-based superalloys, 32Ni–25Cr– 20Fe–12.5W–0.05C, Ni–23Cr–18W– 0.2C, F-M w/therm barriers Ferritics, austenitics

Molten salt, for example: FLiNaK

0.1

700/ 1000

Thermal, 200

Salt

Refractory metals and alloys, Ceramics, ODS, vessel: F-M F-M ducts, 316SS grid plate

High-Si F-M, ODS, ceramics, or refractory alloys Not applicable

High-Si austenitics, ceramics, or refractory alloys Ceramics, refractory metals, Mo, Ni-alloys, (e.g., INOR-8), graphite, Hastelloy N

High-Mo, Ni-based alloys (e.g., INOR-8)

Abbreviations: F-M, Ferritic–martensitic stainless steels (typically 9–12 wt.% Cr); ODS, oxide dispersion-strengthened steels (typically ferritic–martensitic); MC, mixed carbide (U,Pu)C; MN, mixed nitride (U,Pu)N; MOX, mixed oxide (U,Pu)O2.

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Pressurized water reactor – PWR Boiling water reactor – BWR Supercritical water cooled reactor – SCWR Very high temperature reactor – VHTR Gas fast reactor – GFR

Coolant

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flow, residual stresses due to welding, and stresses due to radiation-induced volume expansion or distortion. The unique environmental element of a reactor core is radiation. Fission results in several different types of radiation that affect materials in different ways. Principal radiation types contributing to material degradation are the fission products, neutrons and gamma rays. Fission products consist of high-energy (100 MeV) elements of sizable mass (generally between 90 and 150 atomic mass units) that result directly from the fission process. These elements are created as highly charged ions that deposit their energy within 10 lm of their origin. As such, except for those that are born within this distance of the fuel pellet surface, fission product damage is confined to the fuel. The fission process releases both neutrons and gammas. Neutrons are created with an energy of 2 MeV and slow down via collisions with the coolant and structural components. Neutrons are the primary source of radiation damage to core materials and fuel cladding and assembly components. Typical displacement damage exposures to the fuel cladding (replaced every 5 years) are about 15 dpa. The cumulative displacement damage in core internal structures can approach 80 dpa after 40 years. The displacement damage rate decreases rapidly with increasing distance from the core due to neutron moderation (lower neutron energy resulting from energy loss via collisions with the coolant and core materials); the displacement damage levels in the RPV wall for a PWR are typically 0.05 dpa after 40 years of operation, and the corresponding damage in a BWR vessel wall can be up to an order of magnitude lower. This displacement damage can result in significant temperature- and dose-dependent changes to the microstructure (formation of dislocation loops, precipitate formation/dissolution, void formation, radiationinduced segregation, etc.) [16], which affects mechanical properties (strength/hardness, ductility, fracture toughness and embrittlement, creep, fatigue). When combined with the environment, high temperature and stress, additional modes of degradation occur such as irradiation-assisted stress corrosion cracking, corrosion fatigue and environmentally enhanced fracture toughness degradation. The gamma radiation field is intense and extends throughout the core, aided by (n, c) reactions in structural components. While atom displacement by gamma rays is of minor consequence, the main importance of gamma rays is in their heating and changes to water chemistry. Gamma heating can elevate temperatures in thicker components close to the fuel (such as baffle–former plates and bolts) by as much as 60 °C above the water temperature. Gamma rays also induce radiolysis of the water and create a number of radicals that elevate the corrosion potential in the core. Corrosion potential is the critical element governing stress corrosion cracking of core materials at elevated temperature. Beyond the materials contained within the reactor vessel, the major components affected by the water chemistry environment include piping, turbine rotors and blades, the condenser and, in PWRs, the pressurizer and steam gener-

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ator. Historically, the main materials degradation problems have been intergranular stress corrosion cracking (IGSCC) of BWR piping and of steam generator tubes and in vessel penetrations in PWRs. IGSCC of BWR pipes and steam lines made of 304 stainless steel was due to a combination of weld-induced residual stresses and sensitization of grain boundaries caused by heat treatment of high carbon steels in the temperature region in which chromium carbides formed rapidly on the boundaries, depleting them of chromium and making them susceptible to attack. IGSCC of Alloy 600 occurred in steam generators, on both the primary and secondary sides, and was driven by a susceptible microstructure and the creation of crevices on the secondary side in which crevice chemistry was favorable for intergranular attack. 2.1.2. Life extension and power uprates Due to the initial high capital cost for construction of nuclear power plants relative to the cost of the fuel and other operating expenses, the levelized cost of electricity for LWRs is dominated by the amortized original cost of construction; the annualized costs associated with the fuel and operating and maintenance costs in a new nuclear power plant are estimated to contribute about 20% of the levelized cost of nuclear electricity [68]. This factor, along with the high capacity factor of LWRs demonstrated during the past decade, has led to significant interest in extending the operational licenses of nuclear power plants beyond their initial term (typically 40 years). Extension of reactor operating licenses for an additional 20 years means that reactor components will be required to maintain their integrity for a period that is 50% longer than the initial 40-year license. This increase in operational life introduces a wide range of potential materials aging issues that must be considered as part of the renewal license process [5,69]. While the effect of increased irradiation exposure is dependent on the component, the increase in operating life by 50% means that displacement damage at the bottom of the top guide in a BWR may be >50 dpa, while the shroud will acquire a damage level 100 times lower. High-fluence components such as baffle bolts will reach damage levels exceeding 100 dpa. Fig. 9 shows a rough approximation of the fluence (damage) levels associated with various component failures (top) in both BWRs and PWRs or with microstructure/property changes (bottom) and the impact of a 20year increase in operating lifetime. Life extension will increase the maximum expected damage level on the components receiving the highest fluence, and it will also elevate the damage level on the balance of components proportionately. Additional life extensions beyond 20 years are also being contemplated and the fluence (damage) level coinciding with three 20-year life extensions is shown for comparison. In addition to life extension activities, a second approach to leveraging the existing capital assets of a nuclear power plant is to make modifications to the operating parameters (e.g. coolant flow rate) and/or changes to existing equipment (e.g. turbines) that enable high power levels to be achieved. These power uprate requests require detailed

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Fig. 9. Neutron fluence (E > 1 MeV) and corresponding displacement damage levels and the corresponding failure modes or changes in microstructure/properties of BWR and PWR components and materials. Dashed lines refer to the maximum expected fluence/dose in PWR core components after the designated time period.

safety analysis review. Since 1977, a total of 139 power uprates ranging from 0.4 to 18% have been approved by the US Nuclear Regulatory Commission, resulting in 6 GWe additional generating capacity. The major impacts of the power uprate on the reactor materials include slightly higher operating temperature and higher neutron flux and fluence to permanent core internal structures. A potential consequence of higher damage levels from power uprates and/or life extension is the appearance of new or unanticipated degradation modes. One such mode is void swelling. The formation and growth of voids was discovered in the development of materials for fast reactors [70], where core temperatures range from 350 to 550 °C and damage rates are approximately 10 times that in thermal reactors. Since void formation in stainless steels under fast reactor irradiation conditions is pronounced only for irradiation temperatures of 400–650 °C [52], void swelling was not initially considered an issue for LWR core materials, where maximum core component temperatures are below this temperature range. However, as predicted by void swelling models [71] and recently confirmed by long-term LWR irradiation experiments [72], lower damage rates can be more damaging than high dose rates as they induce void formation at lower fluences and also extend the void swelling regime to lower temperatures. Further, as noted in Section 2.1.1, gamma heating causes the temperature in thicker components to increase up to 380 °C, pulling them into the range where void swelling readily occurs in austenitic stainless steels at LWR-relevant dose rates [52,73]. Baffle–former bolts in PWRs have developed voids at damage levels as low as 7.5 dpa [13]. If the plates fastened by the bolt also swell, they will apply a stress to the bolts, creating conditions that are ripe for irradiation assisted stress corrosion cracking (IASCC). To date, there have been several instances of cracked baffle–former bolts caused by IASCC [20]. There are a variety of life extension and power uprate issues for ex-core materials systems used in nuclear power

plants. Many of the materials systems, such as piping and heat exchangers, can be considered as replaceable (albeit expensive) components and therefore are not directly influenced by life extension considerations. However, improved predictive knowledge of failure mechanisms can lead to more economical maintenance and replacement schedules, while also improving plant worker and public safety. For the RPV, the primary effect of life extension and power uprate scenarios is an increase in the cumulative neutron fluence to the vessel. As will be discussed in Section 2.2.3, there are significant uncertainties in the long-term embrittlement behavior of irradiated RPV steels due to uncertainties in potentially synergistic, late-emerging nucleation and growth of solute–defect cluster complexes. There are also uncertainties in the maximum useful lifetime of a variety of other non-replaceable (or difficult to replace) materials systems in nuclear power plants. For example, there are nearly 1000 km of power, control and instrumentation cables in a typical nuclear power plant. Considering that electrical cables are known to be susceptible to age-related degradation and shorting (due to moisture, heat, etc.) in applications ranging from electrical appliances to vehicles, homes and factories, which are typically less hostile environments than nuclear power plants, there is interest in developing a suite of non-destructive evaluation techniques to investigate the current performance and expected lifetime of cables [74]. Similarly, the concrete used for safety-related structures in LWRs (e.g. primary containment dome and base slab) is susceptible to a variety of environmental degradation mechanisms that act on the cement matrix and steel reinforcement rods. Research topics of highest interest include long-term degradation mechanisms (>50 years, including ionizing radiation effects), improved noninvasive inspection techniques and improved repair methods [75]. 2.2. Materials degradation challenges in current LWRs Although there are many potential areas of concern for materials in LWRs, the demonstrated highly reliable operating performance of commercial reactors for the past 10– 15 years suggests that most of these issues are tractable by appropriate materials selection and engineering design. As described in the following, three specific materials challenges are considered to be of highest importance: (i) exploration of potential further improvements in fuel reliability and operational burn-up limits under normal operating conditions, and safety under transient accident conditions; (ii) corrosion and stress corrosion cracking in reactor components; and (iii) RPV integrity, particularly for life extension scenarios. 2.2.1. Fuel system challenges during normal operations to high burn-up Reliable operation of the fuel assemblies for extended time periods (several cycles, each consisting of 18– 24 months of continuous operation) in an extremely harsh

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radiation environment is of central importance for the economics of nuclear power. Fuel failures (breach of the cladding, allowing escape of some of the gaseous and volatile fission products into the primary coolant, or other damage to the fuel that prevents normal operation) can be managed for a few fuel cladding failures in an operating reactor due to the presence of coolant clean-up systems, but it is undesirable due to increased radiation exposure to personnel, possible leakage of radioactive material, increased inspection and fuel assembly replacement activities during refueling outages, and possible premature shutdown for refueling (if too many fuel failures occur, due to limits on allowable radioactivity in the primary coolant). The fuel concept based on monolithic UO2 fuel encased in Zr alloy cladding has undergone remarkable improvements in reliability since it was first developed in the 1950s. The primary advantages associated with Zr for fuel cladding are very low parasitic absorption of neutrons (thereby requiring less initial isotopic enrichment of the fuel to achieve a given burn-up level), and good fabricability and strength; this is countered by relatively poor oxidation behavior in hot water and steam, and anisotropic properties due to the hexagonal close-packed crystal structure. A series of alloying additions have resulted in commercial Zr alloys that have good resistance to oxidation during normal nuclear reactor operations. Alternative cladding options, such as steels, have higher parasitic neutron absorption, and may be susceptible to stress corrosion cracking under certain water chemistry environments in nuclear reactors. Many first-generation LWRs constructed in the 1960s used either austenitic stainless steel (Types 304, 316, 347) or zirconium alloys for the cladding. The stainless steel cladding experienced severe intergranular cracking issues in the BWR coolant environment, due to the highly oxidizing high-temperature steam in combination with radiation hardening that fostered stress corrosion cracking, whereas the Zr alloy cladding was observed to function with adequate reliability [76,77]. Stainless steel cladding generally exhibited better performance than Zr alloy cladding in first-generation PWRs (fuel pin defect rate of 0.01% vs. 0.1–0.3% for Zr alloy cladding). However, steady improvement of the Zr alloy cladding performance during the 1960s and early 1970s along with the superior (low) parasitic absorption of neutrons by Zr led to nearly universal adoption of Zr alloy cladding by the early 1970s [76,77]. Cladding performance has continued to improve over the past 30 years, with fleet-averaged fuel pin failure rates of 1  104 in 1980, 2  105 in 1990 and 3  106 in 2010 [78,79]. Considering that there are 50,000 fuel pins in a LWR, the current fuel reliability means that most reactors now routinely operate without any cladding breaches. This achievement is even more remarkable when one recognizes that the average fuel burn-up more than doubled during the time period of 1980–2010. As a result of this improved fuel performance, along with improvements in other reactor operations such as utilization of online maintenance and improvements in steam generator reliability in

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PWRs, the capacity factor for fission reactors in the USA has steadily increased from 60% in 1980 to 90% for the past 10 years. Fig. 10 shows the average fuel burn-up at discharge for US BWR and PWR power plants and the fleet-averaged capacity factors from 1977 to 2010 [78– 80]. Since a wide range of phenomena can induce fuel failure, a multifaceted approach involving cladding composition, water chemistry and reactor operation modifications has collectively contributed to improved fuel performance [81]. In the 1960s and early 1970s, the relatively poor fuel performance was largely controlled by internal hydriding of the zirconium alloy cladding due to excessive moisture in the ceramic UO2 fuel pellets; this was solved by improved fuel pellet fabrication and use of moisture getters in the fuel plenum in the reactor. In the 1980s and 1990s, as fuel reliability improved, debris fretting of the cladding became a significant problem; debris filters were added to reactors and improved fuel maintenance procedures to minimize introduction of debris were successfully implemented. Numerous water chemistry changes were made in an attempt to balance corrosion issues in different coolant loop components. In the primary circuit of PWRs, Zn was added to suppress steam generator corrosion and stress corrosion cracking and to reduce overall system corrosion, Li was added for pH control (to maintain constant pH as B was added for reactivity control at beginning of reactor cycles) and the pH was increased from 6.9 to 7.4 in a series of steps to control CRUD deposits on the surfaces of the fuel cladding [79]. In BWRs, noble metal additions and H were added to reduce stress corrosion cracking. As burn-ups increased, pellet–cladding interactions became more prominent due to increased fuel swelling. The closure of the initial gap between the fuel pellets and the cladding can produce a variety of mechanical stress effects as well as multiple interactions with aggressive fission products. One

Fig. 10. Summary of US LWR burn-up and capacity factors [78–80].

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successful approach has been to utilize multiple-layered cladding (e.g. utilizing a soft Zr liner as a compliant barrier liner between the fuel and cladding) [81,82]. The fuel pin diameter and cladding thickness were also decreased to reduce heat flux and facilitate high burn-up (along with operational changes that limited the reactor ramp rate during start-up) [81], with current pellet diameters of 10 mm and cladding thickness of 0.6 mm for BWRs and PWRs [83]. The current major fuel cladding degradation phenomena for high burn-ups [79,84] include grid-to-rod fretting in PWRs and cladding corrosion/oxidation/CRUD deposition and pellet-cladding interactions in both BWRs and PWRs [85]. Grid-to-rod fretting is induced by coolantflow-induced vibration of the fuel cladding against the horizontal grid assembly spacer. At high burn-ups the spring force between the cladding and grid assembly is relaxed due to irradiation creep [12]. Grid-to-rod fretting can be reduced by providing additional stiffening of the fuel assembly and by design changes to minimize fuel rod or assembly vibration, as well as to suppress cross flow and jetting (grid-to-rod fretting is not a major issue in BWRs due to the presence of channel boxes that restrict flow between fuel assemblies). Looking to the future, there is interest in continuing to increase the fuel burn-up values while maintaining or improving the reactor capacity factor and continuing to reduce the fuel failure rate toward zero (individual fuel pin failure probability of <1  107). Decreased overall fuel cycle costs, improved operational flexibility and reduced volume of radioactive waste are all benefits associated with increasing burn-up [81]. Grid-to-rod fretting challenges may be more pronounced in reactors approved for power uprates due to increased coolant flow rates [85], and higher burn-ups would produce additional irradiation creep relaxation of the springs in the grid assembly, which could exacerbate the fretting issue. Advanced thermal hydraulic modeling and other analyses would be useful to determine if new designs for the spacer grid might reduce the magnitude of the fretting. Distortions in the fuel bundles, PWR control rod guide tubes and BWR channel plates at high burn-ups (including bowing of constrained components) due to anisotropic irradiation growth of the zirconium alloy could impede the motion of control rods or blades and therefore need to be carefully evaluated. Finally, oxidation of the cladding at very high burn-ups may require the development of a new generation of ultra-high oxidation-resistant claddings. Oxidation introduces two potential degradation mechanisms: loss of sound metal for the structural function of the cladding and the possible introduction into the cladding of hydrogen formed as a result of the reduction of steam during the oxidation process (including the formation of hydride precipitates, which may serve as internal stress concentrators). An additional concern associated with hydrogen pick-up is the potential for reorientation of the hydride precipitates from circumferential to radial direction as a result of spent fuel handling procedures after the fuel is withdrawn from the

reactor [86]. This reorientation of the hydride precipitates could lead to cladding failure in used fuel during handling and transport, with a consequential effect on the cost for safe management of used fuel. Current international safety regulations for loss of coolant scenarios specify that the oxidized cladding should be less than 15–17% of the wall thickness, which corresponds to an oxide thickness of about 100 lm. This limits historical cladding materials such as Zircaloy-4 to burn-up levels below 50– 60 GWd MTU1, whereas more recent oxidation-resistant Zr alloys, such as optimized ZIRLO and M5/AXIOM, appear to exhibit suitable in-pile oxidation behavior for burn-ups much higher than 70 GWd MTU1 [87–89]. The slightly higher temperatures associated with power uprate activities should pose an additional challenge, since the oxidation rate will be correspondingly increased due to the higher temperature. The fuel pellet experiences the most severe temperatures, radiation damage and chemical transmutation environment of any component in a nuclear reactor. The cumulative amount of radiation damage approaches 1000 dpa for a typical fuel pellet, with 5% or more of the original uranium atoms converted into 10 at.% or more fission products [90]. The microstructure of the initially monolithic polycrystalline UO2 fuel undergoes dramatic changes during high burn-up irradiation due to the intense displacement damage and chemical transmutations, along with high power densities and the relatively low thermal conductivity of the UO2 fuel that produce temperature gradients >1000°C cm1. Of particular interest is the formation at burn-up levels above 50 GWd MTU1 of nanoscale polygonized grains in the outer “rim” region of the fuel pellet, where the burn-up levels are highest and the temperature is lowest [90–92]. There was initially concern that this new high burn-up structure might exhibit inferior fission gas retention, thermal conductivity or other poor fuel properties. However, recent results indicate that the high burn-up structure generally exhibits behavior comparable or favorable to the lower burn-up microstructure [90]. Good fuel behavior has been observed in test irradiations for fuel irradiations up to 83 GWd MTU1 [93]. Exploratory research is also being performed on a variety of alternative fuel forms that are a significant departure from monolithic sintered UO2, including inert matrix fuels [94], metallic or ceramic matrix microencapsulated article fuels [95], and nanocrystalline oxide fuels [96], which may offer some advantages in achieving very higher burn-ups. 2.2.2. Corrosion and stress corrosion cracking (SCC) in structural materials The temperature range of the water coolant in LWRs is generally between 275 and 325 °C, and spans the saturated (BWR) to sub-cooled (PWR) regimes. Although LWRs have been in operation for over 50 years, corrosion remains a significant concern that will become even more important with age. Corrosion occurs in all of the major systems exposed to a water environment, including the reactor core,

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steam generator, turbine, condenser and piping, valves and fittings, and in a wide variety of alloys such as carbon and low alloy steels used in piping and turbine components, stainless steel used in core internals and primary flow circuits and the condenser, nickel-base alloys in the steam generator and in reactor vessel penetrations and welds, and zirconium alloy fuel cladding [17,19,20,97]. Early corrosion problems stemmed from “epidemics” that were generally precipitated by improper water chemistry control; ingress of chlorides that induced pitting in steam turbine discs and blades, and pitting and stress corrosion cracking in stainless steel, or poor secondary side pH control that resulted in wastage and crevice corrosion in steam generator tubes, or poor microstructure or alloy chemistry control; high corrosion rates of zirconium fuel cladding, stress corrosion cracking of stainless steel BWR piping due to sensitization or weld knife-line attack [17]. More recently, corrosion degradation has emerged in the forms of stress corrosion cracking of stainless steel steam lines and nickel-base steam generator tubes and reactor vessel penetrations, flow assisted corrosion in low alloy steels, and nodular corrosion, shadow corrosion, CRUDinduced localized corrosion and fretting of zirconium alloy fuel cladding [18]. Still more recently, irradiation has emerged to play an increasingly important role in irradiation-assisted stress corrosion cracking (IASCC) and irradiation-accelerated corrosion [21,22]. As plants age, the most important corrosion issues will center about stress corrosion cracking, and the accelerating role of irradiation in both IASCC and corrosion. A recent example of corrosion-induced degradation occurred in 2002 at the DavisBesse reactor (PWR), when stress corrosion cracking of the weld metal between the vessel head and a control rod drive housing on the vessel head caused coolant to leak onto the head. Evaporation of water on the hot surface resulted in a concentrated boric acid solution that resulted in corrosion to such an extent as to create a hole in the head nearly the size of a soccer ball. The 10 mm thick stainless steel weld metal liner prevented a loss-of-coolant accident. 2.2.2.1. Stress corrosion cracking. Today, the major stress corrosion cracking issues are in the nickel-base alloys used in steam generators and in vessel penetrations and stainless steels used in the reactor core. Since steam generator tubes comprise some 75% of the surface area of the primary circuit in contact with the coolant, their performance is critical to that of the reactor. Fig. 11 shows that, while many degradation modes of alloy 600 steam generator tubes exist, SCC has been the dominant failure mode over the past 25 years [98]. SCC of nickel-base alloys is very sensitive to composition and water chemistry. Fig. 12 shows the propensity for cracking of austenitic alloys as a function of nickel content in both pure water and 0.1% NaCl; this cracking can either be transgranular (TGSCC) or intergranular (IGSCC) [10]. Note that SCC in chloride is at a maximum at the extremes in nickel content, but SCC in

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pure water occurs only at high nickel concentrations. Unfortunately, alloy 600 used originally in steam generators and vessel penetrations contains approximately 78% Ni and therefore is highly susceptible to SCC even in pure water. Neither microstructure modification (thermal treatment) nor water chemistry control has been able to mitigate SCC in alloy 600. Staehle and Gorman identified seven different SCC modes defined by their potential–pH combinations [97]. In fact, SCC in alloy 600 appears to track the Ni–NiO stability line, in which cracking is at a maximum at the line and drops off both above and below it (Fig. 13) [10]. Since this range of potential and pH spans those typically achievable in service, control of SCC in alloy 600 has proved very difficult. As a result, alloy 600 components have been replaced by alloy 690, which has a nominal composition of 60Ni–30Cr–10Fe and exhibits much higher resistance to SCC in pure water, chlorides and alkaline solutions. However, historically, alloy 600 was found to be difficult to crack in pure high-temperature water in the laboratory, prompting the belief that it would be resistant in service. In fact, even after years in operation, few incidents of cracking were reported. Over time, the incubation period was discovered to be of the order of 11–12 years, and eventually, all plants began to experience cracking. This has led to widespread replacement by steam generators constructed from alloy 690. The concern today is that the superior resistance of alloy 690 to SCC may be due to a longer incubation period and that, in fact, it may eventually begin to crack. Current efforts are focused on trying to determine if there are microstructures, processing routes or water chemistries through which alloy 690 is susceptible to SCC. Recently, it was found that a single cold rolling operation to 20–30% reduction in thickness increases the crack growth rate in pure water by over a factor of 100 [99]. Similar research activities are in progress to identify susceptibility to crack initiation. 2.2.2.2. Irradiation-assisted stress corrosion cracking. IASCC of austenitic stainless steels and some nickel-base alloys has presented a significant problem in ensuring the integrity of LWR core components. IASCC is a generic challenge as it apparently cuts across all LWR designs and materials. Table 4 shows that IASCC has been observed in at least four water reactor designs, 11 different alloys and dozens of components. The specific effects of irradiation on IASCC are classified into two categories: water chemistry and microstructure [21,22]. Water chemistry effects include radiolysis and its effects on corrosion potential, and the effects of corrosion potential on IASCC. Microstructure effects include radiation-induced segregation, irradiated microstructure, swelling and creep, and H and He generation. Leading mechanisms proposed to explain the roles of radiation in the SCC process are: radiolysis and crack tip strain rate, grain boundary chromium depletion, irradiation hardening, localized deformation and radiation-induced solute segregation of minor elements [21,22].

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Fig. 11. Failure modes of mill-annealed Alloy 600 steam generator tubes in US PWRs over a 38-year period [98].

Fig. 12. SCC severity of austenitic alloys as a function of nickel content in pure water and 0.1% sodium chloride (courtesy of R.W. Staehle [10]).

Irradiation causes a significant change in local composition near grain boundaries and other defect sinks [3,4]. The enrichment of nickel and silicon and the depletion of chromium can affect the susceptibility to IASCC, especially under oxidizing conditions. Irradiation also alters the microstructure and, under LWR conditions, faulty dislocation loops represent the primary irradiation-induced microstructure defect. The loops impede the motion of dislocations, resulting in an increase in the yield strength of stainless steels by as much as a factor of five. Radiation hardening correlates with IASCC propensity, and also induces highly localized deformation in the form of dislocation channels, which could contribute to IASCC [100,101]. Irradiation also induces creep that can relax macroscopic stresses, and can also enhance local dynamic deformation. Other factors, such as swelling and formation of new phases, may enhance IASCC at high fluence. With the

many effects of irradiation that overlap spatially and temporally, more work is needed to identify their roles in the mechanism(s) of IASCC and develop a comprehensive prediction methodology. In particular, the evolution of localized deformation in an irradiation damage microstructure and the emergence of phases at high dose are the areas that require a better understanding to ensure the structural integrity of core components to lives beyond 40 or 60 years. The mechanism of IASCC remains a major outstanding issue in the degradation of core components in LWRs of all types. In considering the attributes of an IASCC-resistant austenitic alloy, the following are likely beneficial: high Ni and Cr contents, low Si content, absence of brittle oxide and nitride inclusions, high coincident site lattice fraction of grain boundaries, low connectivity of high-angle grain boundaries and grain boundary coverage by chromium

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Fig. 13. Modes of SCC in Alloy 600 affected by environmental chemistry (courtesy of R.W. Staehle [10]). Regimes in the figure are: AcSCC: acidic-induced SCC; AkSCC: alkaline-induced SCC; HPSCC: high potential-induced SCC; LPSCC: low potential-induced SCC; AkIGC: alkaline-induced intergranular corrosion; PbSCC: lead-induced SCC and Sy-SCC: sulfide-induced SCC.

carbides. Finally, ferritic or ferritic–martensitic alloys should be considered due to their inherently greater resistance to radiation effects and IGSCC compared to ironand nickel-base austenitic alloys [102].

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2.2.2.3. Irradiation-accelerated corrosion. Another major issue in LWRs operated to high doses involves the interplay of radiation and corrosion in irradiation-accelerated corrosion. The mechanism of this process is unknown, but it has been shown to have significant impacts on corrosion rates. For example, zirconium irradiated in-reactor in moist carbon dioxide–air mixtures had oxygen weight gains of more than five times that in the unirradiated state [103]. In-reactor corrosion rates of zirconium alloys were found to be 10 times greater than those conducted outof-reactor, and part of the difference was attributed to greater permeability of the oxide irradiated in-reactor [104]. More recently, Lewis and Hunn [105] found that proton irradiation of a 316 stainless steel foil in room-temperature water for 4 h produced an oxide that was 20 times thicker than the unirradiated control. Further, old data on in-reactor exposure of Zircaloy-2 [106] revealed ten-fold increases in the oxide weight gain and a strong, linear dependence on neutron flux. These data suggest that displacement damage to the solid during corrosion produces a significantly greater effect than that due to radiolysis alone. 2.2.3. Reactor pressure vessel integrity issues The two major degradation issues for RPV steels are embrittlement associated with hardening from radiationinduced solute–defect clusters, and corrosion and stress corrosion cracking phenomena; the corrosion and stress corrosion cracking issues have been briefly summarized in Section 2.2.2.

Table 4 Summary of observed IASCC issues in LWR components [21,22]. Component

Material

Reactor type

Possible sources of stress

Fuel cladding Fuel cladding Fuel cladding* Fuel cladding ferrules Neutron source holders Instrument dry tubes Control rod absorber tubes Fuel bundle cap screws Control rod follower rivets Control blade handle Control blade sheath Control blades Plate type control blade Various bolts** Steam separator dryer bolts** Shroud head bolts** Various bolts Guide tube support pins Jet pump beams Various springs Various Springs Baffle former bolts Core shroud Top guide

304 SS 304 SS 20%Cr25%Ni/Nb 20%Cr25%Ni/Nb 304 SS 304 SS 304/304L/316L SS 304 SS 304 SS 304 SS 304 SS 304 SS 304 SS A-286 A286 600 X-750 X-750 X-750 X-750 718 316 SS cold work 304/316/347/L SS 304 SS

BWR PWR AGR SGHWR BWR BWR BWR BWR BWR BWR BWR PWR BWR PWR and BWR BWR BWR BWR and PWR PWR BWR BWR and PWR PWR PWR BWR BWR

Fuel swelling Fuel swelling Fuel swelling Fabrication Welding and be swelling Fabrication B4C swelling Fabrication Fabrication Low stress Low stress Low stress Low stress Service Service Service Service Service Service Service Service Torque, differential swelling Weld residual stress Low stress (bending)

* **

Cracking in AGR fuel occurred during storage in spent fuel pool. Cracking of core internals occurred away from high neutron and gamma fluxes.

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Radiation embrittlement is of greatest concern for PWR pressure vessels; the larger vessel diameter in BWRs allows more moderation of fast neutrons in the water between the core and vessel, resulting in 3 to 10 times lower fast neutron fluxes compared to PWR pressure vessels [67]. The fundamental mechanisms associated with radiation hardening and accompanying embrittlement of RPV steels are generally understood, including the pronounced deleterious effects associated with Cu, P and Ni solute additions due to the formation of nanoscale precipitates that produced significant matrix hardening [49,67,107]. A series of improved mechanistic models for radiation embrittlement have been developed over the past 25 years that provide a clear description of the complex and synergistic effects associated with solute (Cu, Ni, P), temperature, dose and irradiation flux [49,108]. Fig. 14 shows an example of the predicted increase in the ductile-to-brittle transition temperature (DBTT) of RPV steel containing high levels of Cu or Ni solute [49]. For the high Cu, medium Ni solute case, a high density of Cu-rich precipitates forms rapidly at low dose and causes significant embrittlement for exposure levels corresponding to a few years of operation of a PWR pressure vessel. A relatively small amount of additional embrittlement occurs after the initial rapid embrittlement phase due to the lack of additional Cu solute. For the high-Ni case, predicted slow nucleation rates of the Ni-rich precipitates result in an extended transient regime with relatively low embrittlement, followed by pronounced embrittlement associated with the copious formation of Mn–Ni–Si precipitates. The full impact of such “late blooming phases” on possible accelerated hardening and embrittlement of reactor vessel steels near the end of their design life is the subject of active current research, and could become a major impediment to additional future extensions in the operating lifetime of PWR power plants.

Fig. 14. Predicted increase in the radiation-induced DBTT of RPV steel for two solute concentrations [49].

Pressure vessel embrittlement issues that need further research include [49,67,108]: (i) the effects of high dose, long operation lifetimes and irradiation flux on hardening and embrittlement; (ii) impact of heat-to-heat variability and the quantitative accuracy of surrogate materials such as surveillance specimens; (iii) effect of Ni concentration on the formation and embrittlement impact of Ni-rich “late blooming phases”; (iv) quantitative validity of the fracture toughness master curve concept (is the curve shape universal, correlation between Charpy impact and fracture toughness tests, consequences of intergranular fracture, etc.); (v) impact of size (constraint) effects and other phenomena on the toughness correlation between pre-cracked Charpy (or smaller) surveillance specimens on possible introduction of biased estimates of reference fracture toughness; (vi) correctly accounting for the large attenuation in neutron fluence and displacement damage dose (a factor of 5) that occurs through the wall thickness of reactor vessels; (vii) development of improved-fidelity modeling and microstructural analysis to achieve better understanding of the roles and synergies of the various key experimental parameters; (viii) potential for phosphorus segregation to induce intergranular fracture; (ix) thermal annealing and reirradiation research to investigate feasibility and quantify optimized conditions for periodic vessel annealing to mitigate radiation embrittlement; and (x) investigation of long-term (>50 years) thermal aging effects on the microstructure and properties of low alloy steels. Finally, research on neutron radiation embrittlement of weldments (including the impact of various post weld heat treatments) should be continued. 2.3. Materials challenges during off-normal events There are two major transient operational scenarios that have been considered for design basis safety analyses used to define operational safety limits for commercial reactors: reactivity-initiated accidents (RIAs) and loss-of-cooling accidents (LOCAs). RIA events are an unwanted increase in fission rate (reactor power), such as might occur from an unintended prompt ejection of a control rod [109]. The RIA scenario would produce a rapid increase in fuel power and temperature, which could lead to potential failure of fuel cladding and release of radioactivity into the primary coolant. Due to degradation of the cladding mechanical properties with increasing exposure, safety authorities have established design limits for the maximum allowable energy deposition in the fuel, which decrease with increasing burn-up. The RIA event would be relatively short in duration since the negative coefficient of reactivity in LWRs would quickly lead to a reduction of the excess reactivity and primary coolant flow would continue throughout the event. The LOCA scenarios are based on the knowledge that fission reactors continue to generate substantial amounts of decay heat for days and weeks after they are shut down, due to a variety of radioactivity decay processes in the fuel and core materials; for example, the

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residual heat relative to the full-power reactor value is about 6% at a time of 1 s after a reactor scram and 0.5% 1 day after shutdown. Considering that commercial reactors typically create over 3000 MW of thermal power at full power, these residual heat values represent significant heat removal challenges for a LOCA scenario. Without adequate heat removal, the increase in core temperature can lead to cladding rupture and release of radioactivity into the primary coolant. The LOCA event could last for hours or days at elevated temperature, depending on the severity of the loss of cooling capability. Whereas zirconium alloys have evolved to achieve very impressive fuel performance under normal operating conditions in water-cooled nuclear reactors (Section 2.2.1), the behavior of UO2 monolithic fuel with Zr alloy cladding under accident scenarios is far from optimized. Hydrogen uptake and radiation-induced hardening can lead to reduced ductility in Zr alloy cladding that can impair the resistance to cladding failure associated with pellet– cladding mechanical interactions during an RIA event (expansion of the fuel pellet due to rapid heating) [110]. If pronounced clad breaching occurs, the possibility exists for significant fragmentation and dispersal of the ceramic UO2 fuel pellets. For LOCA scenarios, the rapid increase in Zr alloy oxidation rate with increasing temperature will degrade the mechanical properties of the cladding and could lead to cladding breach and/or fracture, which in turn could lead to reduced flow of primary coolant within the fuel assemblies in the core due to coolant channel blockage by cladding fragments. The high heat of oxidation for Zr can also make a large contribution to the core heating; considering that a typical 1000 MWe LWR core contains 30,000 kg of cladding, 55 MW h of heat would be released if it were completely oxidized (note: BWR cores also contain nearly a comparable amount of Zr in the channel boxes). This could set up an autocatalytic reaction where the heat from Zr oxidation drives a temperature increase in the core, leading to more rapid oxidation and heat generation. Finally, the oxidation of Zr by steam leads to the production of potentially explosive hydrogen gas (1200 kg if all of the cladding was oxidized by steam). The shortcomings of the Zr alloy/UO2 fuel system to severe accident conditions such as occurred at the Fukushima Dai-ichi nuclear power plant in Japan following the 2011 earthquake and tsunami has led to increased interest in improving the safety of nuclear reactors to rare but credible accident scenarios [111]. Considering that the key functions of the fuel system in an accident scenario are to maintain core cooling capability and to minimize or prevent dispersal of fuel and fission products, there are three major potential approaches to design LWR fuel systems with improved accident tolerance. (i) Utilize cladding options with reduced reaction kinetics with high-temperature steam; reduction of the cladding high-temperature oxidation rate by one or more orders of magnitude compared to alloys such as Zircaloy would nearly eliminate the contribution of heat input from

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oxidation and would proportionally reduce the generation of hydrogen. Possible approaches in this category include utilization of oxidation-resistant coatings or development of new alloys with significantly improved oxidation resistance (either Zr-based or new alloys). (ii) Utilize fuel cladding with improved high-temperature mechanical properties and resistance to hydrogen embrittlement (high tensile and short-term creep strength, good thermal shock resistance, high melting temperature). (iii) Utilize new fuel forms with lower operating temperature, higher margins to fuel melting, enhanced retention of fission products (e.g. fuel microencapsulation or gettering techniques), and/or reduced fuel dispersal probability compared to UO2. Since fuel systems are inherently a replaceable commodity, it would be possible to implement new fuels with incremental improvements in accident tolerance (using one or more of the approaches outlined above) as long as the geometry and performance under normal operations do not have a negative impact on the reactor operations. In addition to the potential changes to the fuel system, there are several other core materials systems that would be considered for improvements if robust accident tolerant fuels were successfully developed. For example, the AgInCd control rods currently used in many PWRs have a relatively low melting temperature (900 °C). Similarly, the control blades in many BWRs, which utilize B4C pellets inside stainless steel tubes, begin to have a eutectic Fe–B reaction at 1170 °C. Higher-temperature material options for neutron control should be pursued. 3. Materials challenges in future fission reactor concepts Construction is currently in progress worldwide on several so-called Generation III and Generation III+ LWR power plants that are designed for improved efficiency, passive safety and economics. To a large extent, these reactors represent an evolutionary design change utilizing materials systems that are similar to current (Generation II) LWRs, and therefore the materials challenges facing the new reactors will be comparable to those faced in existing reactors. Another class of light-water-cooled reactors under consideration would utilize in-factory construction techniques and new designs with high emphasis on passive safety to construct small (50–300 MWe) nuclear power plants; some of the materials challenges with these small modular reactors (SMRs) are discussed in the following section. Finally, a brief discussion on the materials challenges for Generation IV reactor concepts will be provided. 3.1. Materials challenges for proposed light-water small modular reactors Due to the high capital cost for constructing a large (1000 MWe) nuclear power station, a variety of designs for small modular reactors (<300 MWe) have been proposed [112]. In order to compete against the traditional economies of scale, which favor large reactor size (per

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MWe), the proposed small reactors would utilize modular construction techniques in factories followed by shipping of major components by rail or barge for final construction at the reactor site. The conceptual designs include both light water and other coolants [112,113]. The materials issues facing the advanced reactor concepts (coolants other than light water) are qualitatively similar to those for conceptual Generation IV systems discussed in Section 3.2. Many of the specific SMR materials issues are uncertain due to the preliminary nature of their engineering designs at this time. For example, it is uncertain whether there will be any significant modification in the water chemistry for the SMRs compared to what has been utilized in existing LWRs. For the light-water-cooled designs that are being proposed for construction in the near future (within 10 years), most of the materials issues are expected to be comparable to those faced by Generation II and Generation III reactors. One key issue will be to determine if the manufacturing processes used in the factory assembly provide any advantage (or disadvantage) in material performance compared to standard fabrication techniques. For example, advanced manufacturing techniques such as additive manufacturing could potentially reduce the fabrication costs of specialized pump components and other devices, but it would be important to compare the microstructure and performance of the materials to conventional wrought material under prototypic operating environments. One of the key design changes in many of the light-water small modular reactors is the use of integral containment (i.e. the steam generator for a PWR is located within the RPV). In some concepts, the primary water flows across the outside of the tubes, resulting in a reversal in primary and secondary sides of conventional steam generators used in all operating PWRs. Integral containment has a significant safety advantage in that the possibility of a large-pipe LOCA is eliminated. However, considering that nuclear reactors still occasionally experience stress corrosion cracking issues with steam generator tubing, and that the evolution of water chemistry control is strongly coupled to the steam generator geometry, the inversion of primary and secondary sides of a tubed steam generator and the limited accessibility of the steam generator for routine inspection may be a source of problems unless further improvements in steam generator reliability are achieved.

ation as part of the first phase of the collaborative Generation IV program in 2002 [114]. The objective was to identify concepts that had one or more of the following attributes: increased efficiency, generation of process heat to drive chemical processes such as the production of hydrogen, increased safety and reduction in waste generation. The concepts finally selected were the supercriticalwater-cooled reactor (SCWR), the sodium fast reactor (SFR), the lead fast reactor (LFR), the very-high-temperature reactor (VHTR), the gas fast reactor (GFR) and the molten salt reactor (MSR). Table 3 summarizes the basic characteristics of each of these reactor types and the materials proposed for the various major components [7]. Note that all designs call for higher operating temperatures and radiation doses, placing a higher burden on the integrity of materials. To allow operation at much higher temperatures, advanced Generation IV reactor concepts utilize different coolants, including water in the supercritical state, liquid metals such as sodium and lead–bismuth, molten salts and high-pressure helium gas. The materials challenges for the Generation IV reactor concepts come about because of the very high fuel temperatures, the intense radiation flux and coolant compatibility issues. Thus, the fuel, the cladding, the structural materials, the reactor vessel and the interaction of these materials with the coolants present the greatest challenges to new, more robust nuclear reactor concepts for the twenty-first century. Structural materials used in the cores of advanced reactors will face unprecedented combinations of temperature, radiation dose and stress. As shown in Fig. 15, a common feature of all advanced designs is a high operating temperature compared to current LWRs. Another unique feature is the simultaneous presence of intense knock-on displacement damage by the fission neutrons. Almost all of the Generation IV concepts call for radiation damage levels that exceed those of LWR experience. One additional proposed

3.2. Proposed next-generation (Generation IV) fission reactor concepts 3.2.1. Brief overview of the six Generation IV concepts Over the past 10 years, the United States Department of Energy and the Generation IV International Forum have explored six particularly appealing advanced reactor concepts as potential next-generation (Generation IV) nuclear power systems [114,115]. These concepts were selected from hundreds of ideas submitted to the US DOE by scientists and engineers worldwide, during a broad canvassing oper-

Fig. 15. Temperature and dose requirements for in-core structural materials for the operation of the six proposed Generation IV advanced reactor concepts, the traveling wave reactor and fusion reactor concepts. The dimensions of the colored rectangles represent the ranges of temperature and displacement damage for each reactor concept.

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concept, the traveling wave reactor (TWR), would require fuel cladding integrity to damage levels approaching 600 dpa. The need for more robust materials extends to the pressure vessel, the primary safety structure for most reactor designs. 3.2.2. Materials-limiting phenomena for Generation IV concepts The high temperatures and damage levels of all of the designs described in Table 3 will accelerate the corrosion and oxidation kinetics and open new pathways for materials degradation. The six Generation IV concepts and the TWR can be grouped into three general categories according to their positions in Fig. 15, as well as the nature of the coolant. The SFR, LFR, MSR and the TWR will all operate at elevated temperatures and very high damage levels, and will utilize either liquid metal or molten salt as the coolant. The GFR and VHTR will experience lower damage levels but even higher temperatures, and will use the relatively innocuous He as the working fluid. The SCWR is a unique third category in that it is the only water-cooled reactor among the Generation IV designs; it will experience lower damage levels relative to other concepts (comparable damage to current LWRs), but conversely will operate at high temperature and very high pressure, where corrosion and stress corrosion cracking issues may become paramount. 3.2.2.1. High-temperature, high-dose fission concepts. The challenges associated with the high-temperature, high-dose operating environment of the first group (SFR, LFR, MSR, TWR) will place increased emphasis on strength, creep and creep-fatigue behavior in addition to fracture toughness at low temperature. While chemical interaction between coolant and structural materials will present some degradation challenges, the major concern is the very high radiation damage levels expected in core components. At such high damage levels, the major degradation modes are likely to be driven by void swelling and phase stability. Void swelling occurs at homologous temperatures of 0.35– 0.55TM, which for steels (325–650 °C) overlaps the temperature range of the reactor core for these four high-temperature, high-dose concepts. Fig. 16 compares the void swelling behavior (obtained from immersion density measurements) for Type 304L [116], 316 [12] and a Ti-modified (D9) [12] austenitic stainless steel and 9–12% Cr-tempered ferritic/martensitic steels [12,117–119] following irradiation at 400–550 °C to high doses in a fast fission reactor spectrum. In all cases, the void swelling behavior consists of an initial low-swelling transient regime followed by a high swelling rate regime (approaching 1% dpa1 for the austenitic steels [12]). Although the maximum allowable volumetric swelling for structural applications is design dependent, void swelling levels >5% are generally unacceptable based on typical engineering design considerations. Severe embrittlement has also been observed in irradiated austenitic steels when the volumetric swelling is >10% [120].

Fig. 16. Comparison of the volumetric void swelling behavior of Type 304L [116], 316 [12] and a Ti-modified (D9) [12] austenitic stainless steel and 9–12% Cr-tempered ferritic/martensitic steels [12,117–119] following irradiation at 400–550 °C to high doses in a fast fission reactor spectrum.

Many years of research and development of austenitic alloys have managed to extend the low-swelling transient regime by utilizing swelling resistant microstructures such as the fine TiC precipitates in Ti-modified austenitic steels, but the delay of steady-state void growth is not sufficient to avoid significant void swelling during operation to the high doses contemplated for several Generation IV reactor concepts (cf. Figs. 9 and 10). As such, more radiation-resistant alloys, such as 2.5–12% Cr bainitic–ferritic–martensitic steels are being considered for high-dose core internal and RPV applications. Exacerbating the problem is phase instability at high doses due to radiation-induced or enhanced solute segregation [121,122] and ballistic dissolution of precipitates [122] by energetic displacement cascades. Irradiation can nucleate or dissolve phases, changing the solute composition of the matrix and enhancing void growth [123]. Further, dissolution of particles added to increase the strength of the alloy results in softening and compromises high-temperature strength and creep. For example, c0 matrix precipitates that provide strength to nickel-base alloys used in high-temperature applications are unstable under irradiation [121,122,124–126]. Further, radiation can induce the formation of brittle phases along grain boundaries and other defect sinks that can reduce ductility and degrade fracture toughness [124,125,127]. However, some types of second-phase particles have demonstrated good stability to high-dose neutron irradiation and can provide strength at high temperature while ameliorating radiation damage effects. For example, controlled additions of Ti and P to austenitic Fe–Cr–Ni alloys has been demonstrated to produce fine dispersions of TiC or M2P (M = Fe, etc.) precipitates that provide dramatic improvement in void swelling resistance after high-dose (100 dpa) irradiation compared to standard Fe–Cr–Ni alloys [51,128–130]. Similarly, the nanometer-sized

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(Y,Ti,O)-rich particles in oxide-dispersion-strengthened alloys appear to exhibit good stability under irradiation and provide significant strength advantages over ferritic– martensitic alloys up to high temperatures [131–135]. An added benefit of nanometer-sized oxides is their role in healing radiation damage. The high particle density presents a very large surface area for point defect trapping, promoting self-healing via recombination [56] and thus keeping the net accumulated radiation damage at a low value. 3.2.2.2. Very high-temperature gas-cooled concepts. The structural materials challenges become magnified considerably when moving from medium-temperature designs to very-high-temperature designs in which materials must withstand temperatures approaching 1000 °C (GFR, VHTR). Corrosion and oxidation of alloys are unavoidable at these temperatures due to the very rapid kinetics. In all cases, the challenge is to develop coolant–materials systems that result in the formation of protective and self-healing films to ensure the longevity of the structures for the life of the reactor. At the extreme operating temperatures envisioned for gas-cooled reactor concepts, graphite and ceramic composites are the leading candidates for structural materials [136]. Along with the numerous engineering design issues associated with utilization of low-ductility materials in a complex high-power energy system, the property degradation associated with neutron displacement damage poses particular challenges. The anisotropic response of graphite to neutron displacement damage due to its hexagonal close-packed crystal structure requires the use of specially manufactured “nuclear” grades of graphite to achieve the desired component lifetimes [136]. For components that are subject to relatively high displacement damage exposures or engineering stresses, ceramic composites must be used instead of graphite. Nevertheless, metals are still the most viable materials for heat transfer components such as heat exchangers. In this concept, high-temperature helium gas will pass through an intermediate heat exchanger, where it will transfer heat to a secondary coolant. Such temperatures require the use of nickel-based alloys rich in chromium (about 22 wt.%) and strengthened by additions of Mo, Co and W (for example, Inconel 617 and Haynes 230Ò) [137]. The helium inevitably contains parts per million (ppm) levels of CO, CO2, H2, H2O, and CH4 as impurities, which arise mainly from reactions between the hot graphite core and in-leakage of O2, N2 and water vapor from seals and welds, and degassing of reactor materials such as fuel, thermal insulation and in-core structural materials [138,139]. Depending on the impurity concentration, temperature and alloy composition, the impurities react with the metallic surfaces of the heat exchanger resulting in oxidation, oxide reduction, carburization and decarburization. Chromium oxide is stable at oxygen partial pressures above a

critical value and reduces at partial pressures below this value. Similarly, chromium carbide is stable above a critical carbon activity, and decarburization is expected to occur below the critical value. Oxidation, decarburization and carburization can degrade the mechanical properties of the alloy; for example, oxidation reduces the load-bearing cross-section of the component and internal oxide precipitates act as the preferential crack initiation sites [140] near the surface of the alloy, which can, potentially, decrease the creep and fatigue life of the alloy. A significant reduction in the creep-rupture ductility of alloy 800H [141], alloy 617 [142] and Hastelloy X [143] has been reported in a carburizing environment in comparison to pure helium and air environments; Fig. 17 shows an example of the effect of oxygen and methane impurities on the creep rupture behavior of alloy 617 (note the longer creep lifetime but lower tertiary creep regime for the methane impurity condition). A coarse and semi-continuous film of carbides forms along the grain boundaries during carburization, and this may act as preferential crack initiation and propagation paths, and could decrease the operating life of the alloys. Grain boundary migration and sliding has been identified as the dominant creep deformation mechanism in the candidate alloys, such as alloy 617 at 1000 °C [144,145], and the dissolution of carbides due to decarburization may lead to significant loss of the creep strength. Therefore, a detailed knowledge of the oxidation mechanisms and rates of microstructure degradation is important to estimate the lifetime of the component and define mitigation strategies for improved oxidation performance of alloys. However, it is unlikely that unprotected alloys can maintain their integrity at 1000 °C without protection by coatings or barrier layers. Development of protective layers without compromising thermal conductivity is perhaps the most important major challenge for structural materials for the VHTR environment. 3.2.2.3. The SCWR concept. The SCWR is unique in that it is the only water-cooled concept among the Generation IV designs. Maintaining water in the supercritical state requires very high pressures and exposes the entire circuit

Fig. 17. Effect of methane and oxygen impurities in helium on the creep rupture behavior of Alloy 617 at 843 °C [142].

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to water temperatures that are well above the existing knowledge base. Until recently, processes such as IGSCC and IASCC in supercritical water have been relatively unexplored. Recent research has shown that alloys that are susceptible to these degradation modes in LWR environments also show susceptibility in SCW, though at higher rates due to the higher temperatures [146]. Alloys considered for structural components fall broadly into two classes: austenitic stainless steels and nickel base alloys are resistant to general corrosion, but susceptible to IGSCC and IASCC, whereas ferritic–martensitic alloys are resistant to SCC but generally exhibit higher rates of oxidation. For iron-based systems, corrosion resistance is associated with the ability of iron- and chromium-based surface oxides to function as barriers to the transport of reactants (oxygen and metal ions). These oxides are the same as those typically formed in other oxidizing environments, such as steam, under less extreme conditions. The main issues related to the behavior of these oxides as protective layers in SCW are similar to those for advancedsteam-cycle fossil energy plants [147]. However, wall thicknesses of critical core components such as fuel cladding and coolant tubes are an order of magnitude smaller than boiler tubes in fossil plants, placing a greater burden on the development of thin protective oxide layers. While oxide growth rates and product morphological details are specific to the oxygen content of the fluid, the temperature and the steel composition, and possibly other factors [148], the oxide structure on steels after exposure to SCW is similar to what is observed under steam conditions for ferritic and ferritic–martensitic (F-M) steels [147]. Because it is generally accepted that chromia-containing spinels are better permeation barriers to cations (metal) and anions (oxygen, OH, etc.) relative to iron oxides [149], it is the underlying oxide layer that can proffer the best corrosion resistance in the SCW environment. This has been observed in recent work on steels under nuclear SCW conditions [148,150,151] in terms of increasing corrosion resistance with increasing chromium content of the alloy. The most daunting challenge for materials in the SCW environment is resistance to SCC and IASCC. While nickel-base alloys and austenitic stainless steels are very resistant to corrosion in SCW, they are most susceptible to SCC. Intergranular stress corrosion cracking occurs readily in high-purity, deaerated SCW at 400 °C and above in both austenitic stainless steels and nickel-base alloys. Cracking severity increases exponentially with temperature in both stainless steels and nickel-base alloys [152]. Over this same temperature range of 400–600 °C, ferritic–martensitic alloys are resistant to SCC [151,153,154]. Furthermore, SCC of susceptible alloys is known to be exacerbated by persistent radiation damage of the metal [155]. Radiation effects on IGSCC are only now being investigated for SCW conditions, yet results show that irradiation significantly increases the extent of SCC in stainless steels

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and nickel-base alloys. Proton irradiations of type 316L stainless steel and Ni-based alloy 690 showed a significant increase in intergranular cracking relative to the unirradiated cases. The increased cracking could not be attributed to radiation-induced segregation or hardening alone, so combinations of factors or other defect mechanisms must be at play [156]. Both the density of cracks and the crack depth increased over the unirradiated case following irradiation to 7 dpa and testing in SCW at 400 °C. One set of data exists on the effect of neutron irradiation on cracking in SCW, where an austenitic stainless steel was irradiated to doses of over 40 dpa and showed extreme embrittlement [157]. Under the same irradiation and testing conditions, ferritic–martensitic alloys were found to be resistant to cracking. 4. Conclusions The continued utilization of nuclear energy systems for worldwide baseload electricity offers a number of materials research challenges. The high reliability of current lightwater fission reactors (e.g. 90% average capacity factor by US reactors for the past decade) demonstrates the high reliability of this energy source under normal operating conditions. Planned extensions in the operating lifetime for reactors are being supported by accompanying materials R&D to investigate corrosion and neutron-induced materials degradation phenomena. The three major materials challenges for continued safe, reliable and cost-effective utilization of water-cooled nuclear reactors for electricity production are development of improved understanding of the synergistic fundamental mechanisms responsible for corrosion and stress corrosion cracking degradation of austenitic steels and nickel base alloys, development of a truly predictive understanding of the multi-physics phenomena responsible for radiation hardening and degradation in ductility and fracture toughness of complex structural alloys (in particular RPV steels), and nuclear fuels innovations including investigation of further improvements in the reliability of LWR fuel systems (can an additional order of magnitude improvement be achieved on top of the three orders of magnitude improvement accomplished over the past 40 years?), as well as exploration of new LWR fuel systems with improved accident tolerance without reducing the favorable performance and reliability features achieved by current fuels under normal operating conditions. The eventual development of advanced Generation IV fission reactor systems is directly linked to successful resolution of several daunting materials challenges associated with the higher radiation damage levels and/or operating temperatures for these concepts and aggressive coolants, all of which are detrimental to the reliability of structural materials. The materials research challenges that need to be successfully resolved for fusion energy systems are even more daunting due to the high concentrations of H and He gases that will be produced in the materials as a result of the high-energy fusion neu-

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tron spectrum. A significant practical issue that is a barrier to the development of structural materials for Generation IV fission and fusion energy concepts is the limited worldwide capability for high-flux materials irradiations; there are no high-intensity fusion neutron irradiation facilities, and the number of fast fission test reactors is dwindling. Acknowledgements The authors would like to acknowledge Roger Staehle for graciously providing several figures, and Jeremy Busby for comments on the draft manuscript. We also thank Robin Grimes (Imperial College) for helpful comments on fuels and materials for nuclear reactors, and Bo Cheng (Electric Power Research Institute), Kurt Terrani and Lance Snead (ORNL), and Kemal Pasamehmetoglu (Idaho National Laboratory) for input on considerations for accident tolerant fuels. References [1] [2] [3] [4] [5]

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