Other chapters in this volume present more details on the fundamental nature and aspects ofthe primary damage state in irradiated metals and alloys, and on the detailed effects of irradiation on mechanical properties’ behavior. This chapter simply highlights some changes in microstructure caused by fission or fusion reactor neutron irradiation, and the changes in properties that they cause in 300 series austenitic stainless steels, to facilitate easy comparison to the unirradiated behavior properties described above. Various other sections of this volume deal in far more detail with the effects of irradiation in various kinds of alloys.

In LWRs at 20-250 °C, the interstitials migrate freely to sinks, while the vacancies or their clusters are relatively immobile, so this has been termed the ‘low-temperature regime’ of microstructural evolu­tion in austenitic stainless steels.14 In sodium-cooled FBRs, temperatures are not lower than the sodium coolant, so they are typically 300-350 °C or above, which is termed the ‘intermediate-temperature regime,’ and both vacancy and interstitial defects can migrate to sinks. Transmutation-produced helium atoms are another form of primary radiation damage that varies with reactor environment (high in LWR and magnetic fusion reactor (MFR) systems, low in FBR systems). Thermal neutrons produce helium in austenitic stainless steels from boron atoms directly, and by a two-step reaction with nickel atoms.17 He/dpa ratios for LWR systems can be very high, over 100appm He/dpa, while in mixed-spectrum fission reactors used for radiation-effects studies on materials, the ratios vary from 1 to 70 appm He/dpa. MFRs, with 14MeV neutrons from D-T fusion reactions, have linear He/dpa ratios of about 14 appm dpa-1 in a stainless steel first-wall component. The FBR reactors with mainly fast fission neutron spectra produce very low He/dpa ratios of 0.1-0.5 appm He/dpa in austenitic stainless steels. Irradiated mate­rials properties data discussed in the remainder of this section is mainly from LWR or mixed-spectrum fission reactor facilities used to study irradiation effects for MFR applications, so they have relatively high He/dpa ratios as well as a wide range of irradia­tion temperatures.

The major effects of irradiation in mixed — spectrum fission reactors, such as Oak Ridge Research Reactor (ORR) or High Flux Isotope Reactor (HFIR), on mechanical properties in the low-temperature regime are dramatic hardening (increased YS) and reduced ductility in SA and 316 and Ti-modified 316 stainless steels, and more modest hardening and ductility reduction in 20-25% CW steels. The increased YS for irradiated SA steels are illustrated in Figure 10.1 The SA stainless steels have 250-300MPa YS in the unirradiation condition, and 50% or more total elongation at room temperature and up to 250-300 °C, but irradiation increases the YS to 600-800 MPa or more, and reduces ductility to 10% or less. However, the fracture mode in this irradiation temperature regime still remains ductile.8,9,19 After irradiation, 20-25% CW steels have YS of 800-1000 MPa, and less ductility, but still retain ductile fracture. This is an important feature to note, and despite transmutation-produced helium levels of 1000-2000 appm, they do not embrittle, because helium and vacancy complexes are immobile in this temperature regime. However, most tensile test­ing results are in vacuum or air, and radiation-induced sensitization in water is not found after irradiation at 20-200 °C, but does become an embrittling factor to consider for irradiation above 300 °C.2

Irradiation-induced hardening of austenitic stain­less steels at room temperature to <250 °C is caused by the microstructural changes produced by irradia­tion in this low-temperature regime. Effects of alloy composition are small in this regime, but the effects of processing condition prior to irradiation (SA or 20-25% CW) are very large. Both SA and 25% CW steels, like 316 or Ti-modified 316, have very dense dispersions of ‘black-spot’ interstitial loops (2-4 nm diameter) uniformly within the grains,2,1 ,21 as illu­strated for 25% CW Ti-modified steel in Figure 11. However, the SA steels also have larger (10-50 nm) diameter Frank (faulted) interstitial loops and no network dislocations, whereas the 25% CW steels

have a recovered dislocation network and virtually no large Frank loops (Figure 12). These microstructural effects directly reflect the fact that interstitial defects are main point defects migrating freely to sinks in this temperature regime. Large Frank loops cannot

nucleate and grow until the concentration of network dislocations is below some critical concentration. This also affects mechanical behavior, because the ‘black-dot’ and larger Frank loops are sessile until they unfault, whereas the network dislocations can
climb and glide in response to stress or as they absorb point defects.

Radiation-induced microstructural changes are definitely different at 300 °C and above. In the dislo­cation structure, the ‘black-dot’ loop damage clearly observed at 200-250 °C is absent at 300-330 °C, and the dislocation structure consists of larger Frank loops and networks that add up to a fairly high total dislocation density.14,21,22 There is now also a cavity component of the microstructure, with nanoscale helium bubbles visible at 300-330°C, and larger voids and helium bubble visible at 400 °C, after irra­diation at high He/dpa ratios in mixed-spectrum reactors (ORR, HFIR), or just voids in FBR irradia­tions at 350-400 °C.12,14,22 The appearance of cavities is a clear indication that vacancy or vacancy clusters and complexes (and helium atoms) are migrating in this temperature regime.

Tensile properties of austenitic stainless steels irradiated at 300 °C and above reflect the micro­structural changes, particularly the dislocation compo­nent of the microstructure. This higher temperature regime in austenitic stainless steels is marked by stronger and more complex temperature and dose dependencies of all the microstructural components, including precipitation and micro/nano-compositional changes.14,22 The YS declines from the 800 MPa values at 300 °C to values of about 400 MPa at 500 °C (Figure 10), which approach the YS of un­irradiated steels, because all components of the radiation-induced microstructure coarsen, and dislo­cation density falls by several orders of magnitude. Ductility can vary significantly, but is generally higher (>10%) at 400-500 °C, but not as high as that of unir­radiated materials. However, the effects also depend on He/dpa ratio. For FBR irradiations (low He/dpa ratio, <50dpa), total elongation can be good even at 600­650 °C, but for irradiations in mixed-spectrum reactors such as ORR or HFIR (high He/dpa ratio, >20dpa), ductility becomes very low above 500 °C, with almost no ductility and brittle grain-boundary fracture at 600 °C due to severe grain-boundary helium embrit­tlement (>500-1000 appm He). For more detailed information on tensile properties after irradiation, see Chapter 1.04, Effect of Radiation on Strength and Ductility of Metals and Alloys.

Microstructural changes produced by irradiation at temperatures of 400 °C and above manifest the intense effects of radiation-induced solute segrega­tion (RIS), which drive nonequilibrium flows and buildups of solute-atoms to sinks (bubbles, voids, dis­location loops and networks, and grain boundaries), because they are coupled to the point defect flows. Such changes are important to note because prolonged aging at <550 °C produces little or no change to the as-fabricated microstructure. Undersized atoms, such as Ni and Si, strongly couple to interstitial defects, and migrate with them to all sinks. Inverse — Kirkendall effects cause fast-diffusing elements such as Cr and Mo to migrate away from sinks with vacancy fluxes diffusing toward them, whereas slow — diffusing Ni atoms build up at such vacancy sinks. The original austenitic solid-solution alloy phase then unmixes after prolonged irradiation into differ­ent kinds of micro/nano-alloys (Figure 7). Regions around the point-defect sinks (voids, loops, and grain boundaries) become enriched in Ni and Si, while the remaining alloy left behind in between such sinks is rich in Cr, and poor in Si and Ni.12 The different micro/nano-alloy regions become unstable as dose increases, and then transform into various precipitate phases, most of which are radiation-induced or mod­ified relative to the natural thermal precipitation that would form in austenitic alloy during aging at higher temperatures (550-650 °C).12,14,15 The most obvious Ni — and Si-rich radiation-induced phase is Ni3Si g0, which forms abundantly in reactor-irradiated SA 316, as shown in Figure 13, but would not form at all in thermally aged SA 316 (Figure 12).15 Another extreme effect of such RIS, found in some FBR — irradiated steels, is the actual decomposition of the austenite parent phase into austenite shells around voids and other sinks, and ferrite regions in between.15 These effects tend to maximize at about 450-550 °C, and then all diminish with increasing irradiation tem­perature. At 650-700 °C, RIS effects are nearly gone and are replaced by basically thermal-aging effects with slightly enhanced kinetics due to radiation — enhanced diffusion.

In addition to the irradiation-produced mechani­cal properties described above, irradiation in this higher temperature regime also causes void/cavity swelling to occur. Void swelling is caused by the biased (or preferred) flows of interstitial and vacancy defects to different sinks, with more vacancies flow­ing to cavities (helium bubbles and voids) and more interstitials flowing to Frank loops and/or radiation — induced precipitates. An example of precipitation — enhanced void swelling in a SA 316 + Ti steel irradiated in ORR at 500 °C to 11 dpa (200 appm He) is shown in Figure 14; clearly the larger voids are directly associated with RIS-induced G-phase (Mn6Ni16Si7) silicide particles. Formation of such voids is the direct cause of volumetric swelling in

G-phase

50 nm

Figure 14 Transmission electron microscopy of radiation — induced voids in SA PCA steel irradiated in ORR at 500 °C to 11 dpa. Largest voids have G-phase particles attached. Reproduced from Maziasz, P. J. J. Nucl. Mater. 1989, 169, 95-115.

reactor-irradiated steel, with an example of swelling of SA 316 steel as a function of dose for FBR irradia­tion at 420 °C shown in Figure 152 For more detailed information on swelling, see Chapter 4.02, Radiation Damage in Austenitic Steels. Such void swelling is generally observed in various FBR or mixed-spectrum reactor environments at 400-650 °C. If very high concentrations of helium bubbles, disloca­tions, or precipitates become the dominant sinks for

point defects, then all the radiation-induced point defects recombine at those sinks (critical radius for void growth becomes very large), and both void swelling and RIS are suppressed.12,14,15,22,24 Such delayed void swelling is seen for dense dispersions of Ti-rich MC carbide particles and dislocation networks
in CW Ti-modified 316 steel (D9 or prime candidate alloy, PCA), as also shown in Figure 15. Very high concentrations of helium bubbles suppress void swelling at 300 °C in mixed-spectrum reactors such as ORR or HFIR (high He/dpa), but as those bubbles coarsen with increased temperature, void swelling is observed, particularly at 500-600 °C. For FBR irra­diations (low He/dpa), void swelling will abate at 650-700 °C, with only tiny helium bubbles being visi­ble at grain boundaries at high doses. However, in HFIR (high He/dpa) cavity swelling due to very large helium bubbles can still be 8-10% or more above 650 °C, with large grain-boundary cavities.12

One aspect of RIS effects on mechanical proper­ties worth noting and highlighting is RIS causing grain-boundary sensitization. For specimens of 25% CW Ti-modified 316 irradiated in ORR at 330 and 400 °C to 6-7 dpa, electrochemical testing to detect grain-boundary sensitization revealed grain­boundary grooving only at 400 °C, suggesting RIS — induced sensitization due to lower Cr at the grain boundaries.20 Similar ORR irradiation of SA 316 at 400 °C shows ductile fracture when tested in vacuum, but very brittle intergranular fracture when tested in oxygenated water at 300 °C (Figure 16), suggesting severe RIS-induced sensitization. In LWR systems, concerns about irradiation-assisted stress-corrosion­cracking (IASCC) at about 300 °C and <10 dpa are important for extended service,25 and these data sup­port such concerns and their connection to RIS. For more detailed information on IASCC, see Chapter 5.12, Material Performance in Supercritical Water.