Comprehensive nuclear materials 5 08 irradiation assisted stress corrosion cracking ,

Comprehensive nuclear materials 5 08 irradiation assisted stress corrosion cracking , Comprehensive nuclear materials 5 08 irradiation assisted stress corrosion cracking , Comprehensive nuclear materials 5 08 irradiation assisted stress corrosion cracking , Comprehensive nuclear materials 5 08 irradiation assisted stress corrosion cracking , Comprehensive nuclear materials 5 08 irradiation assisted stress corrosion cracking , Comprehensive nuclear materials 5 08 irradiation assisted stress corrosion cracking ,

5.08 Irradiation Assisted Stress Corrosion Cracking P L Andresen GE Global Research Center, Schenectady, NY, USA G S Was University of Michigan, Ann Arbor, MI, USA ß 2012 Elsevier Ltd All rights reserved 5.08.1 Introduction 177 5.08.2 5.08.2.1 5.08.2.2 5.08.3 5.08.3.1 5.08.3.2 5.08.4 5.08.4.1 5.08.4.2 5.08.4.2.1 5.08.4.2.2 5.08.4.2.3 5.08.4.3 5.08.5 References Irradiation Effects on SCC: Laboratory and Plant Data Individual Effects of Radiation on IASCC Service Experience Irradiation Effects on Water Chemistry Radiolysis and Its Effect on Corrosion Potential Effects of Corrosion Potential on IASCC Irradiation Effects on Microchemistry and Microstructure Radiation-Induced Segregation Microstructure, Radiation Hardening, and Deformation Irradiated microstructure Radiation hardening Deformation mode Radiation Creep and Stress Relaxation Summary 180 180 183 187 187 189 190 190 194 194 196 198 201 202 202 Abbreviations AES AGR BWR CT CW DPA FEGSTEM FWHM HWC HWR IASCC IGSCC NWC PWR RIS RH SCC SFE SGHWR SS SSRT STEM Auger electron spectroscopy Advanced gas cooled-reactor Boiling water reactor Compact type (specimen) Cold work Displacements per atom Field emission gun TEM Full-width – half-max (for profiles) Hydrogen water chemistry (in BWRs) Heavy water reactor Irradiation-assisted stress corrosion cracking Intergranular stress corrosion cracking Normal water chemistry (in BWRs) Pressurized water reactor Radiation-induced segregation Radiation hardening Stress corrosion cracking Stacking fault energy Steam generating heavy water reactor Stainless steel Slow strain rate test Scanning transmission electron microscopy TEM TG Transmission electron microscopy Transgranular 5.08.1 Introduction Nuclear power accounts for about 17% of the world’s electricity production, and the rapid expansion in nuclear power throughout the world will necessitate that they operate with high reliability and safety Stress corrosion cracking (SCC) has occurred in all water cooled reactors, including boiling-water reactors (BWRs) and pressurized-water reactors (PWRs), with a greater incidence in unirradiated, out-of-core components, especially between 1970 and 1990 As these materials, component designs, and water chemistries have improved, an increasing percentage of cracking problems has occurred in irradiated components While irradiation-assisted stress corrosion cracking (IASCC) has been observed since early plant operation, increasing operating time and fluence has led to an increased incidence of cracking Setting aside zircaloy fuel cladding and pressure vessel steels, most irradiated core components consist of austenitic stainless steels and nickel-base alloys 177 178 Irradiation Assisted Stress Corrosion Cracking exposed to environments that span oxygenated to hydrogenated water at $270–340  C The core of a nuclear reactor is an extreme environment consisting of high-temperature water, imposed stresses and strains, and an intense radiation field that affects the water chemistry, stress, and microstructure of the core materials (Figure 1) For background, the reader is referred to Chapter 1.03, RadiationInduced Effects on Microstructure; Chapter 1.05, Radiation-Induced Effects on Material Properties of Ceramics (Mechanical and Dimensional), and Chapter 1.07, Radiation Damage Using Ion Beams for more detailed treatments of radiation effects on materials Initially, the affected components were primarily small components (bolts, springs, etc.) or components designed for replacement (fuel rods, control blades, or instrumentation tubes) However, in the last $20 years, IASCC has been observed in structural components (e.g., PWR baffle bolts and BWR core shrouds and top guides) Extensive literature exists for SCC under unirradiated conditions, and the basic factors and dependencies are well defined and reasonably well modeled for austenitic stainless steel and nickel alloys (e.g., Alloys 600 and its weld metals).1–9 A complete consensus on the underlying mechanism of cracking has not emerged although the well-behaved continuum in crack growth rate response versus material/ composition (including from stainless steels to nickel Solution renewal rate to crack-tip Stress Δf Anionic transport Oxide rupture rate at crack-tip Environment Microstructure g-field Crack tip f [A]–, pH Passivation rate at crack-tip Grain boundary denudation Hardening Relaxation N-fluence Segregation Figure Schematic of the primary engineering parameters that effect stress corrosion cracking – stress, microstructure, and environment – and the underlying crack tip processes that control stress corrosion cracking The primary ways in which radiation affects stress corrosion cracking is also shown: segregation, hardening, relaxation, and radiolysis Radiolysis can increase the corrosion potential, which in turn increases the potential gradient (Á’) and the crack tip potential ’, anion concentration [A], and pH alloys), water chemistry, temperature, and radiation suggests that a common crack growth mechanism is operative.8–13 Our understanding has evolved from the view that SCC occurs under very specific and unique conditions to the view that a continuum in response exists.8–13 With steady improvement in laboratory and plant detection of SCC, it is clear that SCC occurs under a wide range of conditions and also at a wide range of growth rates Figures and show examples of the effect of environment (corrosion potential and water purity), material condition (sensitized vs cold-worked), and stress (stress intensity factor) on SCC growth rates; the solid curves are the predicted response.8,9,12,13 SCC occurs even at low corrosion potential (Figure 2), and thus the behavior in BWRs and PWRs is linked, with the primary differences being dissolved H2, temperature, and the dissolved ion chemistry (B and Li are added to PWR primary water; see Chapter 5.02, Water Chemistry Control in LWRs).8–10,14 Of these, temperature has a universal effect, variations in dissolved H2 are particularly important in nickel alloys, and B/Li has little or no effect in deaerated water.9,14 Early plant (Figure 4) and laboratory (Figure 5) observations showed that the same basic dependencies existed for unirradiated and irradiated stainless steels, and that increasing fluence produces a well-behaved increase in SCC susceptibility (Figure 6) Figure shows a strong effect of water purity for both unirradiated and irradiated BWR components, and Figure shows a very similar response to corrosion potential to that in Figure Thus, it was proposed that radiation enhances SCC primarily in four ways: segregation, hardening, relaxation, and radiolysis (Figure 1) The neutron fluence where these processes have an effect is shown in Figure 7, along with the current end-of-life fluence for various BWR and PWR components The primary radiation effects on materials operate in a similar range of fluences, and thus their individual contributions can be difficult to distinguish An example of their interaction in altering SCC growth rate is shown in the prediction of cracking of a weld in a BWR core shroud (Figure 8) in which the individual effects are plotted along with the resulting crack length versus time While many of the enhancements in SCC susceptibility from irradiation dose (neutron fluence) have been well established, it remains possible that additional factors will emerge at high fluences (e.g., >30 displacements per atom (dpa)) Intergranular (IG) SCC is promoted in austenitic stainless steels above a ‘threshold’ fluence Irradiation Assisted Stress Corrosion Cracking 179 1.0E–05 25 mm CT specimen Furnace sensitized; 15 C cm–2 288 ЊC water ; 0.1–0.3 mS cm–1 Constant load ; 25 Ksi in1/2 Crack growth rate (mm s–1) 42.5 μin h–1 10–7 10 1.0E–06 11 Crack growth rate (mm s–1) 10–6 Sensitized 304 stainless steel 30 MPa m1/2, 288 ЊC water 0.06–0.4 μS cm–1, 0–25 ppb SO4 filled triangle = constant load open squares = ‘gentle’ cyclic 14 14.2 μin h–1 Theoretical curves ααα μS cm–1 0.3 0.2 0.1 10–8 200 ppb O2 500 ppb O2 2000 ppb O2 304 Stainless steel Screened round robin data - highest quality data - corrected corr potential - growth rates corrected to 30 MPa m1/2 42.5 28.3 14.2 –1 μin h 1.0E–07 GE pledge predictions 30 MPa m1/2 0.5 2000 ppb O2 Ann 304SS 200 ppb O2 0.25 1.0E–08 0.1 12 0.06 μS cm–1 Hydrogen water chemistry β Normal water chemistry (ex-core) –600 –400 –200 +200 Corrosion potential (mVshe) +400 30 MPa m1/2 1.0E–09 –0.6 –0.5 –0.4 –0.3 –0.2 –0.1 0.1 Corrosion potential (Vshe) 1.0E–05 Sensitized 304 stainless steel 30 MPa m1/2, 288 ЊC water 0.06–0.4 μS cm–1, 0–25 ppb SO4 SKI round robin data filled triangle = constant load open squares = ‘gentle’ cyclic 200 ppb O2 500 ppb O2 2000 ppb O2 β 10–9 –1 0.06 μS cm Industry mean 0.2 0.3 0.4 dpa 304SS Crack growth rate (mm s–1) 1.0E–06 316L (A14128, square) 304L (Grand gulf, circle) nonsensitized SS 50% RA 140 C (black) 10% RA 140 C (gray) 1.0E–07 20% CW A600 42.5 28.3 14.2 μin h–1 20% CW A600 GE pledge predictions 30 MPa m1/2 Sens SS 0.5 2000 ppb O2 Ann 304SS 200 ppb O2 0.25 1.0E–08 –1 0.1 0.1 μS cm Means from analysis of 120 lit sens SS data 0.06 μS cm–1 0.06 μS cm–1 GE pledge predictions for Unsens SS (upper curve for 20% CW) 1.0E–09 –0.6 –0.5 –0.4 –0.3 –0.2 –0.1 0.1 Corrosion potential (Vshe) 0.2 0.3 0.4 Figure Stress corrosion cracking growth rate versus corrosion potential for stainless steels tested in high-purity water at 288  C containing 2000 ppb O2 and 95–3000 ppb H2 Dissolved O2 strongly influences corrosion potential, which in turn affects crack chemistry and growth rate of sensitized stainless steels (two graphs at left) as well as cold-worked stainless steels and Alloy 600 (large rectangular symbols on right graph) and irradiated stainless steel (large triangular symbols) Cold-worked or irradiated materials have an elevated yield strength, which causes an increase in growth rate at both low and high potential RA, Reduction in area; CW, Cold work 180 Irradiation Assisted Stress Corrosion Cracking 10–5 Stress intensity (ksi in1/2) 10 20 30 40 60 80 10–3 Sens 304 stainless steel 288 ЊC water 10–4 10–7 NRC disposition line 10–5 * 10–8 * Theory 15 C cm–2, –50 mVshe 0.5 ms cm–1 Theory –2 15 C cm , –50 mVshe 0.2 ms cm–1 10–9 10–6 * Theory 15 C cm–2, –200 ® –500 mVshe 0.2 ms cm–1 10–10 Crack growth rate (in h–1) Crack growth rate (mm s–1) 10–6 10–7 10 20 30 40 60 80 Stress intensity (MPa m1/2) Figure Effect of stress intensity factor on stress corrosion cracking growth rate for sensitized stainless steel exposed in various water chemistries at 288  C (Figures and 6) This occurs in oxygenated (e.g., BWR) water above 2–5  1020 n cm–2 (E > MeV), which corresponds to about 0.3–0.7 dpa, and depends on the stress, water chemistry (especially, sulfate and chloride), and other factors Attempts to reproduce the same level of IG cracking in inert environments have been unsuccessful, confirming that it is an environmental cracking phenomenon, not simply a change in the mechanical properties and response of the material in an inert environment Cracking in hydrogenated water (i.e., BWR hydrogen water chemistry (HWC) or PWR water) is typically observed at roughly a 4 higher fluence than in oxidizing water, with IASCC enhanced at elevated temperature (Figure 9) For both BWR and PWR conditions, the same basic dependencies exist for unirradiated and irradiated materials It is important to distinguish the results of different kinds of SCC testing Crack-growth testing typically uses fracture mechanics specimens, commonly a compact type (CT) specimen It has the significant advantage of continuous, online monitoring of crack length versus time, usually by employing an electric current, potential drop technique It can provide a resolution of about mm and can accurately characterize the inherent resistance to crack advance (i.e., beyond the ill-defined initiation stage) as well as the dependencies on corrosion potential, stress intensity factor, etc Smooth specimen tests, whether by constant load, constant displacement, or slow strain rate (SSR), are simpler tests to perform, but represent some combination of initiation and growth SSR tests impose failure and can overstate or understate the SCC susceptibility Constant load or displacement tests generally require periodic interruption for examination, and ‘initiation’ can be microscopic cracks or complete failure Work over the last 25 years has enabled many aspects of IASCC phenomenology to be explained and predicted on the basis of the experience with intergranular stress corrosion cracking (IGSCC) of nonirradiated stainless steel in high-temperature water environments This continuum approach has successfully accounted for radiation effects on water chemistry and its influence on electrochemical corrosion potential However, all radiation-induced microstructural and microchemical changes that promote IASCC are neither fully known nor fully reproducible in similar materials Well-controlled data from well-characterized irradiated materials remain insufficient due to the inherent experimental difficulties and financial limitations Many of the important metallurgical, mechanical, and environmental aspects that are believed to play a role in the cracking process are illustrated in Figure Only persistent material changes are required for IASCC to occur, but in-core processes such as radiation creep and radiolysis also have an important effect on IASCC 5.08.2 Irradiation Effects on SCC: Laboratory and Plant Data 5.08.2.1 Individual Effects of Radiation on IASCC IASCC can be categorized into radiation effects on the water chemistry (radiolysis) and on the material/stress, and the accepted definition of IASCC encompasses cases where either factor is dominant (low-fluence materials tested in water undergoing radiolysis, or preirradiated materials tested without an active radiation flux) Radiation dose rate in rads/h is often used in radiolysis, and in neutrons per square centimeter (n cm–2) or displacements per atom in materials In light water reactors (LWRs), dpa corresponds Frequency of SCC initiation increases dramatically with increasing conductivity 0.2 Best fit 1.2 1.0 0.8 0.6 Threshold conductivity for SCC initiation increases as level of sensitization decreases Low carbon SS High carbon SS Nonsensitized low carbon SS 0.2 0 0.2 (b) 1.0 0.9 Sensitized high carbon SS 0.4 0.6 Plant average conductivity (ms cm–1) 0.4 0.6 Plant average conductivity (ms cm–1) * 0.8 Frequency of IGSCC initiation increases with plant conductivity 0.7 0.6 0.5 * 0.4 * Experienced substantial high conductivity excursions not reflected in average value 0.3 0.2 * 0.1 (c) 181 1.4 Upper bound 0.4 % with IGSCC/on-line months (a) 1.5 1.4 1.3 1.2 1.1 1.0 0.9 0.8 0.7 0.6 0.5 0.4 0.3 0.2 0.1 0 % with IGSCC/on-line months % with IGSCC/on-line month Irradiation Assisted Stress Corrosion Cracking 0.1 0.2 0.3 0.4 0.5 0.6 0.7 Plant average conductivity (ms cm–1) Figure The effects of average plant water purity shown in field correlations of the core component cracking behavior for (a) stainless steel intermediate and source range monitor dry tubes, (b) creviced stainless steel safe ends, and (c) creviced Inconel 600 shroud head bolts, which also shows the predicted response versus conductivity Adapted from Brown, K S.; Gordon, G M In Proceedings of 3rd Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors; The American Institute of Mining, Metallurgical, and Petroleum Engineers (AIME): New York, NY, 1988; pp 243–248 to $7  1020 n cm–2 when counting neutrons with E > MeV, or $1.5  1021 n cm–2 for E > 0.1 MeV The primary effects of radiation on materials8,15–22 include microcompositional effects (grain boundary chemistry) and microstructural changes (formation of dislocation loops, voids, precipitates, and the resulting changes hardening and deformation mode) In terms of their effect on IASCC, the primary effects of radiation are the following:  Radiolysis of water, in which a variety of short- and long-lived radicals and species are produced There is no evidence that the specific species formed are important, and indeed their effect on cracking appears to be fully captured by their overall effect on the corrosion potential of the material  Radiation-induced segregation (RIS), which produces an enrichment in some species (e.g., Ni and Si) at grain boundaries and other defect sinks, and a depletion in other species (e.g., Cr) Even though the distance over which RIS occurs is very limited (a few nanometers), studies of unirradiated materials have shown that the narrow profiles can affect SCC.23,24  Radiation hardening (RH), which results from radiation damage and the creation of vacancy and interstitial loops, which impede dislocation motion Once a few dislocations move along a given slip plane, they clear the ‘channel’ of most of these barriers, and subsequent dislocation occurs primarily in these channels  Radiation creep relaxation, which reduces constant displacement stresses such as in bolts or associated with weld residual stress During active irradiation, radiation creep can promote dynamic strain, and thereby SCC  Swelling, which occurs to a limited extent at temperatures above $300  C, but can be sufficient to produce reloading of components such as PWR baffle former bolts Swelling occurs differently in different materials, and is delayed in cold-worked materials Stresses due to swelling are balanced by 182 Irradiation Assisted Stress Corrosion Cracking Data of Jacobs( ) and Kodama( ) Postirrad SSRT 2–3 ϫ 10–7 s–1 288 ЊC Comm purity 304SS( ) and 316SS( ) 42 ppm O2-sat’d vs 0.02 ppm O2 100 80 »2 100 40 »42 ppm O2 60 %IGSCC % IGSCC fracture Data shifted right by Init grain boundary Cr enrichment »0.2 0.02 –0.05 fC, Vshe 20 –0.4 1019 –0.2 0.2 1020 1021 Neutron fluence (n cm–2 ) (E > MeV) 1022 Percentage of spot welds inspected with IGSCC Figure Dependence of irradiation-assisted stress corrosion cracking on fast neutron fluence as measured in slow strain rate tests at 3.7  10À7 sÀ1 on preirradiated type 304 stainless steel in 288  C water The effect of corrosion potential via changes in dissolved oxygen is shown at a fluence of %2  1021 n cm–2 Reproduced from Jacobs, A J.; Hale, D A.; Siegler, M Unpublished Data on SCC of Irradiated SS in 288  C Water and Inert Gas; GE Nuclear Energy: San Jose, CA, 1986 SSRT, Slow strain rate test PWR control BWR core BWR end rod failures (IASCC) component of life failures (IASCC) 100 80 1020 BWR creviced control blade sheath 60 Threshold fluence for IGSCC ≈5 ϫ 1020 n cm–2 40 0.1 0 0.2 0.4 0.6 0.8 1.2 1.4 Neutron fluence (n cm–2 ϫ 1021) (E > MeV) Figure Dependence of irradiation-assisted stress corrosion cracking on fast neutron fluence for creviced control blade sheath in high-conductivity boiling-water reactors Reproduced from Gordon, G M.; Brown, K S In Proceedings of 4th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors; NACE: Houston, TX, 1990; pp 14-46–14-62 radiation creep relaxation, but the resulting stress can be sufficient to cause IASCC  Other microstructural changes, such as precipitation or dissolution of phases in materials While there is no clear evidence that such changes affect IASCC response, this may only reflect the limited PWR end of life PWR life extension 1021 1022 1023 Neutron fluence (n cm–2) (E >1 MeV) Irradiation dose (dpa) 10 Significant changes in grain boundary composition, alloy strength, and ductility 20 PWR baffle bolt failures (IASCC) 100 Onset of significant void swelling and possible embrittlement Figure Neutron fluence effects on irradiation-assisted stress corrosion cracking susceptibility of type 304SS in boiling-water reactor environments Reproduced from Bruemmer, S M.; Simonen, E P.; Scott, P M.; Andresen, P L.; Was, G S.; Nelson, J L J Nucl Mater 1999, 274, 299–314 characterization and IASCC studies that has been performed on high fluence materials The individual (segregation, hardening, and creep, Figure 10) and composite (SCC, Figures 5–9) effects of radiation increase with dose in much the same manner, which makes the isolation of, and attribution to, individual contributions difficult The dislocation loop microstructure is closely tied to radiation Irradiation Assisted Stress Corrosion Cracking 30 20 30 25 Effect of rad segregation 20 Effect of rad hardening 15 10 Depth 10 Stress intensity (K) Effect of stress relaxation 0 100 200 300 Time (month) 25 Cr Cr depletion 20 Loop line length SCC 15 10 500 400 Figure Predicted effect of radiation segregation, radiation hardening, and radiation creep relaxation on a boiling-water reactor core shroud, where the through-wall weld residual stress profile is the primary source of stress Less aggressive water chemistry (corrosion potential and water purity) would result in less crack advance early in life, which would give a greater opportunity for radiation creep relaxation The leak depth is the wall thickness of the shroud While radiation hardening continues to increase the yield strength, its effect on crack growth is reduced (see Figure 21(a)) EPR, Electrochemical potentiokinetic repassivation (a test for sensitization) Fraction of IG cracking area Hardness Arbitrary units Crack depth (mm) Leak depth 38.1-mm thick 304SS, two-sided weld 0.75 Vshe, 0.15 μS cm–1 EPR0 = 10.8 C cm–2 (0.050% C) Flux = ϫ 1019 n cm–2 year K, segregation, hardening, or relaxation 40 30 183 CP-304SS 3.2 MeV protons 360 ЊC 0 Dose (dpa) Figure 10 Schematic diagram showing the dose dependence of key irradiated microstructure features (radiation-induced segregation and dislocation microstructure) and radiation hardening along with stress corrosion cracking susceptibility Reproduced from Was, G S In Proceedings of 11th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors; American Nuclear Society (ANS): La Grange Park, IL, 2004; pp 965–985 100 80 60 40 In PWR primary water CW316, 340 ЊC CW316, 325 ЊC CW316, 290 ЊC Type 304, 325 ЊC irradiation Subsequent sections focus on the possible mechanisms by which water chemistry, RIS, microstructure, hardening, deformation mode, and irradiation creep – individually or on concert – may affect IASCC * * Specimen broke at the pin hole 20 * * * 1.0E + 20 1.0E + 21 1.0E + 22 Fluence (n cm–2) (E > 0.1 MeV) 1.0E + 23 Figure Percentage intergranular stress corrosion cracking versus fluence for cold-worked type 316 stainless steel tested at various temperatures Despite the low potential environment of pressurized-water reactor primary water, at high fluence (especially at higher temperature) there is significant susceptibility to stress corrosion cracking hardening and both increase with dose until saturation occurs by $5 dpa RIS also increases with dose and tends to saturate by $5 dpa Although dependent on both metallurgical and environmental parameters, IASCC generally occurs at doses between 0.5 dpa (for BWRs) and 2–3 dpa (for PWRs), which encompasses the steeply rising portion of the curves in Figure that describe the changes in materials properties with 5.08.2.2 Service Experience The IASCC service experience extends over 50 years, and the early observations projected an accurate view of the important characteristics and dependencies, and pointed to the ‘assisted’ nature of radiation in enhancing SCC As with other forms of SCC, early observations suggested that a growing incidence with time and neutron fluence should be expected IASCC was first reported in the early 1960s7,8,15–22,25–36 and involved intergranular cracking of stainless steel fuel cladding, with the initiation of multiple cracks occurring from the water side By contrast, mostly ductile, transgranular cracking was observed in postirradiation mechanical tests performed in inert environments at various strain rates and temperatures Grain boundary carbide 184 Irradiation Assisted Stress Corrosion Cracking precipitation was rarely observed although preexisting thermal sensitization was present in some cases A correlation between time-to-failure and stress level was reported, with failure occurring first in thinwalled rods with small fuel-to-cladding gaps, where fuel swelling strains were the largest The highest incidence of cracking occurred in peak heat flux regions, corresponding to the highest fluence and the greatest fuel–cladding interaction (highest stresses and strains.) Similar stainless steel cladding in PWR service exhibited fewer instances of intergranular failure At that time, off-chemistry conditions or stress rupture were often considered to be the cause of PWR failures In the last 30ỵ years, a growing number of other stainless steel (and nickel alloy) core components have exhibited IASCC, including neutron source holders in 1976 and control rod absorber tubes in 1978 Instrument dry tubes and control blade handles and sheaths, Figure 5, which are subject to very low stresses, are also cracked, generally in creviced locations and at higher fluences.15,18,20,37,38 These initial failures in the most susceptible components were Table followed by more numerous incidents of IASCC in the past 20 years, perhaps most notably in BWR core shrouds8,15,16,39 and PWR baffle bolts.19,40,41 A summary of reported failures of reactor internal components is listed in Table and demonstrates that IASCC is not confined to a particular reactor design, material, component, or water chemistry For example, stainless steel fuel cladding failures were reported years ago in commercial PWRs and in PWR test reactors.15,22,25,29–36,42 At the West Milton PWR test loop, intergranular failure of vacuum-annealed type 304 stainless steel fuel cladding was observed31 in 316  C ammoniated water (pH 10) when the cladding was stressed above yield Similarly, IASCC was observed in creviced stainless steel fuel element ferrules in the Winfrith steam generating heavy water reactor (SGHWR),43,44 a 100 MWe plant in which light water is boiled in pressure tubes, where the coolant chemistry is similar to other boiling-water reactor designs The 20% Cr–25% Ni–Nb stainless steel differs from type 304 primarily in Ni and Nb content, as well as in its lower sulfur (%0.006%) and phosphorus (%0.005%) contents The ferrules were designed for Irradiation-assisted stress corrosion cracking service experience Component Material Reactor type Possible sources of stress Fuel cladding Fuel cladding Fuel claddinga 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 boltsb Steam separator dryer boltsb Shroud head boltsb Various bolts Guide tube support pins Jet pump beams Various springs Various springs Baffle former bolts Core shroud Top guide 304SS 304SS 20% Cr–25% Ni–Nb 20% Cr–25% Ni–Nb 304SS 304SS 304/304L/316L SS 304SS 304SS 304SS 304SS 304SS 304SS A-286 A-286 600 X-750 X-750 X-750 X-750 718 316SS cold work 304/316/347/L SS 304SS 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) a Cracking in AGR fuel occurred during storage in spent fuel pond Cracking of core internal occurs away from high neutron and gamma fluxes AGR, Advanced gas-cooled reactor b Irradiation Assisted Stress Corrosion Cracking a 5-year exposure during which the peak fast neutron flux is 2–3  1013 n cm–2 s (E > 1.5 MeV), yielding a peak fluence over years of 3–5  1021 n cm–2 The similarity of IASCC in BWRs and PWRs was also noted in swelling tube tests performed in the core44,45 on a variety of commercial and high-purity heats of types 304, 316, and 348 stainless steel and Alloys X-750, 718, and 625 Swelling was controlled by varying the mix of Al2O3 and B4C within the tubes; the latter swells as neutrons transmute B to He Nominally identical strings of specimens were inserted into the core in place of fuel rods Historically, the oxidizing potential in a PWR core is lower than in BWRs, but in the last decade, most BWRs employ an electrocatalytic technology called NobleChem™(46–48) to create a low corrosion potential Some early investigators attributed PWR cracking to low ductility stress rupture (of course, this mechanism would apply equally to BWRs) A few laboratory studies reported small amounts of intergranular cracking of irradiated stainless steels in SSR tests in %300  C inert environments49 although in many related experiments50,51 no intergranular failure was found Small amounts of intergranular cracking in inert tensile tests are not surprising, and since the early 1990s, the plant and laboratory IASCC data show that cracking is environmentally assisted and follows a well-behaved continuum in response over ranges in fluence, corrosion potential, temperature, stress, etc.8,15–17 Factors that distinguish PWRs from BWRs include their higher operating temperature, %10 higher maximum neutron fluence in core structural components, higher hydrogen fugacity, and borated– lithiated water chemistry (including the possibility of localized boiling and thermal concentration cells in crevices from gamma heating, which could lead to aggressive local chemistries) The possible role of RIS of Si may be especially important in accounting for the limited difference in SCC response at high potential (BWR) versus low potential (PWR) at high fluence.52,53 Brown and Gordon37,38 (Figure 4) accumulated and analyzed data for cracking in Alloy 600 shroud head bolts (first observed in 1986) as well as stainless steel safe ends (first observed in 1984) and in-core instrumentation tubes (first observed in 1984) with a focus on components that were creviced, a factor known to exacerbate cracking.7,37,38 The highest radiation exposure occurred for the intermediate range and source range monitor (IRM/SRM) dry tubes, which contain flux monitors housed in thin-walled, 185 annealed stainless steel tubes Cracking initiated in the crevice between the spring housing tube and the guide plug at fluences between 0.5 and 1.0  1022 n cm–2 (E > MeV) Wedging stresses from the thick oxide observed in the crevice were implicated, since other (applied and residual) stresses were negligible The primary variable from plant to plant is the average coolant conductivity, which correlates strongly with cracking incidence (Figure 4(a)–4(c)) Each point in Figure represents inspection results for one BWR plant, and data are normalized using reactor operating time (i.e., percentage of components with intergranular cracks divided by the online exposure time) The scatter in Figure 4(a) was attributed to variations in fluence and specific ion chemistry, as well as limitations in the resolution of underwater visual inspection Scatter can also result from short-term excursions in conductivity, which is not adequately reflected in the average, as identified in Figure 4(c) Correlations between IASCC and conductivity were also reported for cracking in shroud head bolts (Figure 4(c)) and creviced safe ends (Figure 4(b)) The strong influence of conductivity on cracking of stainless steel has also been shown in laboratory tests and plant recirculation piping, where predictive modeling8,12,13,15,54,55 has been compared to field data on the operational time required to achieve a detectable crack (typically, 10% of the wall thickness) Preliminary prediction of the shroud head bolt cracking response8 also provides reasonable agreement with observation (Figure 4(c)) High-strength, nickel-base alloy components15,22 have also experienced IASCC (Table 1), with many incidents in lower radiation flux regions (e.g., where the end-of-life fluence is below %5  1019 n cm–2) such as cracking of Inconel X-750 jet pump beams in BWRs Inconel X-750 cracking has also occurred extensively in PWR fuel hold-down springs, which attain an end-of-life fluence of 1–10  1021 n cm–2; it is proposed that cracking has been aggravated by vibrational stresses (corrosion fatigue) The effects of irradiation on IASCC in high-strength, precipitationhardened nickel-base alloy components as well as in stainless steels have not been characterized BWR core shrouds8,15,16,39 and PWR baffle bolts40,41 are the two most common examples of IASCC although susceptibility clearly exists in other areas, such as control blade components, fuel components, the BWR top guide, etc SCC in the BWR core shroud occurs almost exclusively near the welds (both circumferential and vertical), and 186 Irradiation Assisted Stress Corrosion Cracking initiation is observed from both the inside (ID) and outside (OD) surfaces (the shroud separates the upward core flow from the downward recirculation flow that occurs in the annulus between the shroud and the pressure vessel) This large-diameter welded ‘pipe’ has little active (DP) loading, and its susceptibility arises primarily from weld residual stresses and weld shrinkage strains.56–58 Cracking is observed in both low fluence and moderate fluence areas, and the extent of the enhancement in SCC susceptibility by irradiation is limited because, while RH and RIS occur, radiation creep also relaxes the weld residual stress SCC predictions for a BWR core shroud that account for the damaging effect of RIS and RH and the beneficial effect of radiation creep relaxation are shown in Figure and illustrate the complexity of the interactions of these phenomena in the evolution of cracking Predictions also indicate that, if SCC does not nucleate early in life (e.g., below 0.5 dpa), for example, from high coolant impurity levels or severe surface grinding, susceptibility tends to decrease with fluence in the shroud welds because of radiation-induced creep relaxation (although many shroud welds are in very low flux areas) The last decade has also seen extensive failures of PWR baffle bolts40,41 although large plant-to-plant and heat-to-heat differences are observed Most baffle bolts are fabricated from type 316 stainless steel cold-worked to %15% to increase their yield strength The complex baffle former structure exists in a PWR because their fuel does not have a surrounding ‘channel,’ so the baffle former structure must conform closely to the geometry of the fuel to provide well-distributed water flow The baffle former plates are usually made from annealed material, typically type 304 stainless steel Because of their proximity to the fuel, very high fluences can develop – up to $80 by the end of the original design life The high gamma flux produces significant heating in the components, in some instances estimated at ỵ40  C, especially in designs where the PWR coolant does not have good access to the bolt shank While the heat-to-heat variations are not understood, it is clear that plants that load-follow (and therefore undergo power level changes and thermal cycles) are much more prone to baffle bolt cracking Another aggravant is the thermal gradient and possible boiling (resulting in altered water chemistry) in the shank area of the bolt if the coolant access is restricted However, primary factors must be the very sizeable stress relaxation that occurs early in life (e.g., during the first dpa), followed by preferential radiation swelling of the annealed baffle plates over the cold-worked baffle bolts, which will cause reloading The dynamic equilibrium between swelling and radiation creep, which determines the ‘reloading’ stress in the bolt, is likely a complex function of many parameters, including local neutron flux, temperature, baffle plate geometry, and composition The number of IASCC incidents continues to grow, and there can be no question that many LWR components are susceptible Strategies to mitigate IASCC (e.g., NobleChem™(46–48)) and manage IASCC (e.g., by showing some IASCC could be tolerated, installing mechanical restraints to mitigate the impact of IASCC in BWR shrouds, or selectively inspect and replace baffle bolts) have been successful IASCC field experience has led to the following trends and correlations:  Intergranular cracks associated with radiation effects on solution-annealed stainless steel were once thought to occur only at fluences above %0.3  1021 n cm–2 But significant intergranular cracking in BWR core shrouds (which not have thermal sensitization) occurs over a broad range of fluences, showing that a true fluence threshold does not exist.15,16 The observations of SCC in unirradiated, unsensitized stainless steel (with or without cold work) also undermine the concept of a threshold fluence below which no SCC occurs This also holds for thresholds in corrosion potential, water impurities, temperature, etc.8,54,59,60  SCC susceptibility is affected by fluence in a complex fashion SCC in BWR shrouds and PWR baffle bolts does not always correlate strongly with fluence, and one important reason for this is that radiation creep produces relaxation of the stresses from welding and in bolts The need to account for many changing factors is necessary in interpreting and predicting SCC  Most early incidents involved high stresses or dynamic strains, but cracking has since been observed at quite low stresses at high fluences and longer operating exposure Laboratory and field data indicate that IASCC occurs at stresses below 20% of the irradiated yield stress, and at stress intensities below 10 MPa m1/2  Extensive laboratory and field data show that corrosion potential is a very important parameter, with its effect being consistent from low to high fluence, except in some high fluence materials and/or Irradiation Assisted Stress Corrosion Cracking 191 250 12.59 2.7 × 10−8 mm s–1 12.53 150 To ppb O2@1508 h Crack length (mm) 12.55 1.3 × 106 mm s–1 12.51 12.49 100 Constant K Dissolved oxygen (ppb) 200 12.57 50 CT#2 - irradiated 304SS, dpa 21 MPa m1/2, 220 ppb O2, 288 ЊC, 20 ppb SO4 12.47 12.45 1475 1495 1515 1535 1555 1575 1595 Time (h) Figure 13 Crack length versus time for Type 304 stainless steel irradiated to dpa and tested in at 21 MPa m1/2 in water at 288  C A large reduction in crack growth rate is observed as the dissolved O2 and corrosion potential are decreased Reproduced from reproduced from Andresen, P L.; Ford, F P.; Higgins, J P.; et al In Proceedings of ICONE-4 Conference; ASME International: New York, NY, 1996 Cv Ci Jv JA Ji JB JB (a) JA (b) CB CA0,B CA (c) Figure 14 Schematic illustration of radiation-induced segregation in a binary 50A–50B system showing (a) the development of the vacancy concentration profile by the flow of vacancies to the grain boundary balanced by a equal and opposite flow of A and B atoms, but not necessarily in equal numbers, (b) the development of the interstitial concentration profile by the flow of interstitials to the grain boundary balanced by a equal and flow of A and B atoms migrating as interstitials, but not necessarily in equal numbers, and (c) the resulting concentration profiles for A and B Reproduced from Was, G S Fundamentals of Radiation Materials Science: Metals and Alloys; Springer: Berlin, 2007 Enrichment and depletion can also occur by association of the solute with the interstitial flux The undersized species will enrich, and the oversized species will deplete.80 The magnitude of the buildup/depletion is dependent upon several factors such as whether a constituent migrates more rapidly by one defect mechanism or another, the binding energy between solutes and defects, the dose, dose rate, and the temperature RIS profiles are also characterized by their narrowness, often confined to within 5–10 nm of the grain boundary, as shown in Figure 15 for an irradiated stainless steel Segregation is a strong function of irradiation temperature, dose, and dose rate (Figure 16) Segregation peaks at intermediate temperatures since a lack of mobility suppresses the process at low temperatures, and back-diffusion of segregants minimizes segregation at high temperature (where defect concentrations approach their thermal equilibrium values) For a given dose, a lower dose rate results in a greater amount of segregation At high dose rates, the high defect population results in increased recombination which reduces the number of defects that are able to diffuse to the grain boundary Figure 16 shows the interplay between temperature and dose rate for an austenitic stainless steel RIS occurs in the intermediate temperature range, and this range rises along the Irradiation Assisted Stress Corrosion Cracking 22 18 JEOL 2010F 0.75 nm probe 16 14 12 10 Ni Si P –20 –15 –10 –5 10 15 Distance from grain boundary (nm) 20 Figure 15 Compositional profiles across grain boundaries obtained by dedicated scanning transmission electron microscopy from a low-strain, high-purity 348 stainless steel swelling tube specimen irradiated to 3.4  1021 n cm–2 in water at 288  C in a boiling-water reactor Composition profiles were measured using a field-emission gun scanning transmission electron microscope Reproduced from Jacobs, A J.; Clausing, R E.; Miller, M K.; Shepherd, C M In Proceedings of 4th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors; Cubicciotti, D., Ed.; NACE: Houston, TX, 1990; pp 14–21 temperature scale with increasing dose rate to compensate for the higher recombination rate In Fe–Cr–Ni alloys, the vacancy exchange (inverse Kirkendall) mechanism successfully explains the observed major element segregation.84,85 Studies have shown that nickel segregates to grain boundaries while chromium and iron deplete The directions of segregation are consistent with an atomic volume effect in which the subsized solute migrates preferentially with the interstitial flux, and the oversized solute participates preferentially in the vacancy flux The results are also consistent with the diffusivity of the solutes in Fe–Cr–Ni, in which Ni is the slow diffuser, Cr is the fast diffuser, and Fe is intermediate In commercial austenitic stainless steels, chromium depletes at grain boundaries and nickel enriches, while iron can either deplete or enrich according to the magnitude of the diffusion coefficient relative to the other solutes.88 RIS increases with neutron dose in LWRs and saturates after several ($5) displacements per atom in the 300  C temperature range Figure 17 shows grain boundary chromium depletion for austenitic Homologous temperature (T/Tm) Cr Si or P concentration (wt%) Cr or Ni concentration (wt%) 20 1105 0.8 0.7 Back diffusion of vacancies 933 0.6 761 589 0.5 0.4 Radiation-induced segregation 416 LWR 0.3 peak flux region 0.2 0.1 10–8 244 Recombination of vacancies and interstitials 10–7 10–6 10–5 10–4 –1 Radiation flux (dpa s ) 72 Temperature for g stainless steel (ЊC) 192 –101 10–3 Figure 16 Dependence of radiation-induced segregation on homologous temperature and dose rate for austenitic stainless steels stainless steels as a function of dose.89–97 As the slowest diffusing element, nickel becomes enriched at the grain boundary Since iron depletes in 304 and 316 stainless steels, the nickel enrichment makes up for both chromium and iron depletion and can reach very high levels up to $30 wt% Minor alloying elements and impurities also segregate and have been implicated in the IASCC process Mn and Mo strongly deplete at the grain boundary under irradiation,98 but neither is believed to be a significant factor in IASCC Minor alloying or impurity elements such as Si and P also segregate under irradiation Silicon strongly enriches at the grain boundary to as much as ten times the bulk (0.7–2.0 at.%) composition in the alloy99 and can be important in IASCC Phosphorus is present at much lower concentrations and is only modestly enriched at the grain boundary because of irradiation.83,98 Phosphorus tends to segregate to the grain boundary following thermal treatment, which reduces the amount of additional segregation to the grain boundary during irradiation, making the contribution due to irradiation difficult to detect.98 Undersized solutes such as C, B, and N should also segregate, but there is little evidence of RIS, due in part to the difficulty of measurement Another potential segregant is helium, produced by the transmutation of 10B The mobility of He is low at LWR core temperatures, but the opportunity for accumulation at the grain boundary is increased by segregation of B to the boundary Overall, the behavior of these minor elements under irradiation is not well understood Irradiation Assisted Stress Corrosion Cracking Fast neutron fluence (E > MeV) ϫ 1025 n m–2 Grain boundary Cr concentration (wt%) 26 304 (82) 304 (13) 304 (91) 304 (92) 304 (93) 316 (82) 316 (94) 348 (91) 24 22 20 10 12 14 HP 304 (95) CP 304 (95) CP 316 (95) CP 304 (17) CP 316 (17) HP 316 (17) CP 304 Protons 96 CP 316 Protons 96 18 16 14 12 10 10 Dose (dpa) 15 20 Figure 17 Dose dependence of grain boundary chromium concentration for several 300-series austenitic stainless steels irradiated at a temperature of about 300  C Adapted from Asano, K.; Fukuya, K.; Nakata, K.; Kodama, M In Proceedings of 5th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors; Cubicciotti, D., Simonen, E P., Gold, R E Eds.; American Nuclear Society (ANS): LaGrange, IL, 1992; p 838; Jacobs, A In Proceedings of 7th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors; NACE: Houston, TX, 1995; p 1021; Jacobs, A J.; Wozadlo, G P.; Nakata, K.; et al In Proceedings of 6th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors; Gold, R E., Simonen, E P., Eds.; The Minerals, Metals, and Materials Society (TMS): Warrendale, PA, 1993; p 597; Kenik, E A J Nucl Mater 1992, 187, 239; Nakahigashi, S.; Kodama, M.; Fukuya, K.; et al J Nucl Mater 1992, 179–181, 1061; Jacobs, A J.; Clausing, R E.; Miller, M K.; Shepherd, C M In Proceedings of 4th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors; Cubicciotti, D., Ed.; NACE: Houston, TX, 1990; pp 14–21; Walmsley, J.; Spellward, P.; Fisher, S.; Jenssen, A In Proceedings of 7th International Symposium on Environment Degradation of Materials in Nuclear Power System—Water Reactors; NACE: Houston, TX, 1997; p 985; Was, G S.; Busby, J T.; Gan, J.; et al J Nucl Mater 2002, 300, 198 Oversize solutes can affect the microchemistry or microstructure of the alloy, thereby altering the IASCC susceptibility They are believed to affect RIS by acting as vacancy traps, thereby increasing the effective recombination of vacancies and interstitials and thus reducing RIS Kato et al.100 conducted electron irradiations of several stainless steels at temperatures of 400–500  C up to 10 dpa Results showed that some solutes (Zr and Hf) consistently produced a large suppression of radiation-induced chromium 193 depletion, while others resulted in less suppression or suppression at only certain temperatures Fournier et al.101 conducted irradiation of 316 containing Hf or Pt using 3-MeV protons (400  C) and 5-MeV Ni ions (500  C) Ni irradiations showed little effect of the oversize impurity in reducing grain boundary chromium depletion (Cr depletion increased in the case of Hf), but proton irradiation showed a significant suppression of RIS of chromium at low dose (2.5 dpa) with the effect diminishing at higher (5.0 dpa) dose Pt had a smaller effect on Cr Ti and Nb similarly produced little change in the grain boundary chromium concentration after irradiation with 3.2-MeV protons to 5.5 dpa at 360  C In Zr-doped 304SS, there were no consistent results of suppression of grain boundary chromium after 3.2-MeV proton irradiation to 1.0 dpa at 400  C.102 Neutron irradiation at very low dose (0.5 dpa) shows a small effect of Ti and Nb on grain boundary Cr.103 In all, the data on the effect of oversize solutes on RIS of chromium are inconsistent RIS is understandably implicated in IASCC of stainless steels, especially in oxidizing environments, in part because of the effect of thermal sensitization in extensive data from lab and plant operational experience.15–20,37,38,104,105 As shown in Figure 17, grain boundary chromium depletion during irradiation can be severe Figure 18(a) shows a correlation between grain boundary chromium level and IGSCC susceptibility in stainless steels where the grain boundary depletion is due to thermal sensitization.106 Much data have been accumulated to support the role of chromium depletion as an agent in IGSCC of austenitic alloys in oxidizing conditions Numerous studies show that, as the grain boundary chromium level decreases, intergranular SCC increases Typical chromium-depleted zone widths are of the order 100–300 nm full width at half-maximum (FWHM), providing a significant volume of depleted material adjacent to the grain boundary Figure 18(b) shows a similar correlation between grain boundary chromium level and IASCC susceptibility as measured by the percentage IG cracking on the fracture surface during SSR experiments A major difference between Cr depletion profiles resulting from RIS and those due to precipitation reactions is that the width of the RIS profiles can be as much as orders of magnitude smaller, typically 5–10 nm There is large scatter in the data that makes a direct correlation difficult to support, and differences in testing conditions undoubtedly contribute 194 Irradiation Assisted Stress Corrosion Cracking %IGSCC in slow strain rate test 100 80 60 e° = ϫ 10–7 S–1 ␧° = ϫ 10–6 S–1 40 Alloy 600 SSR tests, 23 ЊC sulfuric acid 20 Type 304SS SSR tests, 288 ЊC ppm O2 water (a) 10 12 14 16 18 20 Minimum grain boundary chromium concentration (wt%) 100 HP 316 SS with single solute additions108 3xx SS17 80 %IG 60 40 20 10 (b) 12 14 16 18 20 Grain boundary Cr content (wt%) irradiation However, it should be noted that these alloys were not irradiated, and this difference may be important in the relevance of such experiments to IASCC Using 1.5–5% Si stainless steels of both standard (e.g., 304L) base composition and synthetic irradiated grain boundary composition, Andresen has observed significantly increased growth rates, reduction in the benefit of lowering corrosion potential, and very little effect of stress intensity factor between 27 and 13 ksi in1/2 The data on impurity segregation effects on IASCC remain inconclusive Extensive experiments have been conducted to isolate the effect of particular impurities such as S, P, C, N, and B in IASCC, but none have yielded unambiguous results Sulfur has not been found to segregate under irradiation, and, while P thermally segregates to a significant extent, irradiation-induced P segregation is small in comparison C, N, and B cannot be measured in STEM; N and B are very difficult to identify in AES; and C is a common contaminant Overall, it has been a challenge to establish a link between impurity element segregation and IASCC in austenitic stainless steels 22 Figure 18 Effect of grain boundary Cr content on intergranular stress corrosion cracking for (a) sensitized stainless steel and Alloy 600 and (b) irradiated stainless steels Slow-strain-rate tests are tests in which the specimen is monotonically strained versus time Reproduced from Bruemmer, S M.; Was, G S J Nucl Mater 1994, 216, 348 Among the minor alloy elements, only Si is known to segregate to high levels, and Si segregation is correlated with IASCC Experiments by Busby et al.107,108 on a high-purity 316 base alloy doped with wt% Si showed severe IASCC in NWC and in primary water after irradiation to 5.5 dpa at 360  C STEM measurements of grain boundary Si confirm levels up to wt% Past studies comparing Auger electron spectroscopy (AES) and scanning transmission electron microscopy (STEM) results have shown that the actual concentration of Si at the grain boundary plane may be as high as 15–20 wt% Though the electron beam probe in STEM is very small, the measurement underpredicts the concentration at the grain boundary by a factor of to Yonezawa et al.109–112 and Li et al.113 have provided extensive evidence to show that increased Si in stainless steel results in increased IGSCC in alloys tailored to imitate the composition of grain boundaries under 5.08.4.2 Microstructure, Radiation Hardening, and Deformation 5.08.4.2.1 Irradiated microstructure The microstructure of austenitic stainless steels under irradiation changes rapidly at LWR service temperatures Point defect clusters (called ‘black dot damage’ when electron optics could not resolve the details) begin to form at very low dose, dislocation loops and network dislocation densities evolve with dose over several displacements per atom, and the possibility exists for the formation and growth of He-filled bubbles, voids, and precipitates in core components in locations exposed to higher dose and temperatures.114–120 Below 300  C, the microstructure is dominated by small clusters and dislocation loops Near 300  C, the microstructure contains larger faulted loops plus network dislocations from loop unfaulting and cavities at higher doses The primary defect structures in LWRs are vacancy and interstitial clusters and Frank dislocation loops The clusters are formed during the collapse of the damage cascade associated with primary and secondary atom collisions after an interaction with a high-energy particle The larger, faulted dislocation loops nucleate and grow as a result of the high mobility of interstitials The loop population grows in size and number density until absorption of vacancies and Irradiation Assisted Stress Corrosion Cracking interstitials equalize, at which point the population has saturated Figure 19 shows the evolution of loop density and loop size as a function of irradiation dose during LWR irradiation at 280  C Note that saturation of loop number density occurs very quickly, by $1 dpa, while loop size continues to evolve up to $5 dpa The specific number density and size are dependent on irradiation conditions and alloying elements, but the loop size rarely exceeds 20 nm and densities are of the order of  1023 m–3 The hardening process and IASCC susceptibility are influenced by small defects The traditional view that small defect clusters are predominantly faulted interstitial loops and vacancy clusters121 may be inaccurate Analysis of recent postirradiation annealing experiments by Busby et al.122 and Simonen et al.123 suggests that there are at least two types of defects with different annealing characteristics: vacancy and interstitial faulted loops, each with different annealing kinetics The step change in hardness as a function of annealing time suggests that the density of vacancy loops is perhaps much higher than previously believed, and higher than the density of interstitial loops.122 Above 300  C, voids and bubbles may begin to form, aided by the increased mobility of point defects at the higher temperature The dislocation structure will evolve into a network structure as larger Frank loops unfault The reduction in the sink strength of the dislocation loops aids in the growth of voids and bubbles While their size and number density increase with temperature, the dislocation microstructure continues to be the dominant microstructure component over the temperature range expected for LWR components (