Radiation induced segregation (RIS) is the preferential migration of atoms in a point defect flux (i.e. vacancies and self – interstitials) directed away from or towards point defect sinks. The flow of defects to sinks (such as surfaces, grain boundaries and dislocations) results in the loss or enrichment of local elements leading to non – equilibrium segregation. As this segregation is driven by the flux of radiation produced defects to sink, it is fundamentally different form thermal segregation process.
Different species diffuses at different rates. Species with slow diffusing rate are enriched while species with faster diffusing rate are depleted. The directions of segregation are dependent upon the atomic volume of the solute. The undersized atoms (e.g. Ni in austenitic steel) migrate towards while oversized atoms (e.g. Cr) migrate away from sinks. RIS thus can produce large compositional change on a local scale. In austenitic stainless steel, phenomena of RIS can be well understood by two major mechanisms which are solute concentration dependent and occur at various times. These are:
1. Interstitial association Segregation (or Solute – defect bindings)
This mechanism is based on the formation of interstitial – solute complex in low solute concentration alloys. The self – interstitial binds with an undersized solute atoms such as Si and P, and if their (interstitial – solute complex) migration energy is less than dissociation energy, they become mobile. The complex diffuses towards the sink where interstitial gets eliminated (Figure 1-21a). Thus, a concentration gradient is established causing enrichment of the solute elements around the sinks.
2. Inverse Kirkendall effect
Vacancies produced in the damage cascade require exchanging positions with atoms in lattice in order to diffuse to sink. They preferentially exchange with faster diffusing species which results in the depletion of faster diffusing species (such as Cr in austenitic stainless steels) at the sinks. This is counterbalanced by the enrichment of slower diffusing species (such as Ni) at sinks. The motion of atoms is opposite to that of vacancies (Figure 1-21b).
Inverse Kirkendall effect occurs for interstitial as well. In the case of interstitial, the motion of atoms is in same direction as that of interstitial. The faster diffusing species in this case results in enrichment.
Figure 1-21 : Schematics showing the flow of defects and changes in composition at sink for a) Interstitial association Segregation b) Inverse Kirkendall Segregation mechanisms of irradiation induced segregation [6].
SEGREGATION OF ALLOYING ELEMENTS
In austenitic steels, inverse Kirkendall mechanism effectively explains the observed major element segregation [71 - 73]. Neutron irradiation of austenitic stainless steel at LWR operating temperatures results in redistribution of the major alloying elements and segregation of impurities. Depletion of chromium and iron and enrichment of nickel at grain boundaries has been observed in 304 and 316 stainless steels after neutron irradiation. RIS increases with dose and saturates at 3 – 5 dpa at irradiation temperature of 300 °C. A typical RIS profile for Cr, Ni and Fe at the grain boundary of a neutron irradiated sample is presented in Figure 1-22. A basic characteristic of RIS profiles is their narrowness (typically on the order of 5 – 10 nm at the grain boundaries). Figure 1-23 shows grain boundary chromium depletion and nickel enrichment in austenitic stainless steel as a function of dose. Jacob et al. [74, 75] reported to observe Ni enrichment of 1.75 times the bulk and Cr depletion to 0.75 – 0.85 times the bulk level in a commercial purity 304 SS irradiated to 4 – 5 dpa. Chromium imparts corrosion resistance to the grain boundaries and its significant depletion could lead to intergranular stress corrosion cracking of the material.
In addition to major alloying element, segregation of minor alloying elements such as depletion of Mn, Mo and enrichment of Si and P has been reported as well. The extent of segregation is dependent on matrix composition. Increase in Ni matrix content enhances the enrichment of Ni and depletion of Cr, while increase in Cr matrix content diminishes it. Addition of Mo and P reduces the segregation as well.
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Figure 1-22 : Concentration Profile plotted as a function of distance from grain boundary observed in a baffle former bolt taken from Tihange PWR and irradiated to a dose of 10 dpa [50, 76].
Figure 1-23 : Concentration profile of a) Cr b) Ni of grain boundary in irradiated stainless steel plotted as a function of dose [76, 77].
Several studies have reported to observe the segregation at the grain boundaries, however, only few reported to observe the segregation on voids and loops. Kenik and Hojou [78] were the first to observe Ni and Si enrichment on both sides of the loop on edge – on position. Fukuya et al. [60] also studied the segregation at edge on Frank loops (Figure 1-24) as a function of dose in CW 316 flux thimble tubes irradiated in a PWR and observed no significant segregation at the loops. They used spot analyses (using TEM with energy dispersive X ray spectroscopy, EDS) to estimate the segregation. The problem with this technique is the undesired strong signal from the matrix making the estimation of the real segregation at the loops difficult. However, few studies have shown Ni and Si segregation at the loops using Atom Probe Tomography (APT) analysis on ion irradiated austenitic stainless steels.
Figure 1-24 : Enrichment of Cr, Ni and Si from matrix level at edge on Frank loops in PWR irradiated CW 316 SS [60].
Temperature and flux are the primary factors controlling RIS. For a given neutron fluence (LWR relevant), at low temperatures (< 80 °C), mobility of the defects is low. While at high temperatures (> 500 °C), recombination dominates. RIS is low for both of these temperature ranges. RIS dominates at intermediate temperature which corresponds to the LWR operating conditions [33, 79]. Decrease in dose rate shift the temperature dependence of RIS to lower temperatures. Lower dose rate implies a lower point defect generation rate which increases the probability of finding a sink over recombination resulting in higher segregation. This argument is in agreement with the results obtained by Allen et al. [80]. They witnessed greater chromium depletion and nickel enrichment for samples irradiated at lower displacement rate.
PRECIPITATION
In addition to producing local chemical composition changes, migration of alloying element to sinks can also lead to phase change or acceleration of phase formation. If the solute enrichment caused by RIS exceeds the solubility limit of alloying elements at the defect sinks, precipitation of the second phase occurs. Indeed, the precipitates form in austenitic steels during irradiation can be classified in three categories, namely Radiation enhanced (/retarded) phases, radiation modified phases and radiation induced phases [50]. Radiation enhanced (/retarded) phases includes the thermal phases (τ (M23C6), η (M6C),
MC, laves, σ, χ) which are present in the material before irradiation. They have the same composition after irradiation but their abundance is accelerated (/retarded) by irradiation. This means that they are present in the irradiated material at temperatures where they are not present under thermal aging.
Radiation modified phases are the phases that have same crystal structure after irradiation as the corresponding phase formed during aging, but their composition is different (i.e.
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wt. % of their constituting elements is different) for the two conditions. These are η (M6C), laves, M2P.
Radiation induced phases are the phases that are unique to the irradiation conditions such as G (M6Ni16Si7), (Ni3Si), MP, M2P, M3P. These phases are not observed during aging
treatment at any temperatures.
Figure 1-25 : a) Temperature and dose regime where precipitation is observed in SA 316 irradiated in Fast neutron fission reactor [49]. b) BF TEM image of observed in a CW 316 baffle former bolt irradiated at 8.5 dpa at 300 °C in Tihange – 1 PWR [57].
The phase evolution in stainless steel is sensitive to several factors such as alloy composition, irradiation conditions etc. Zinkle et al. [49] reported that precipitation in austenitic steel generally occurs for irradiation temperature range of 400 – 800 °C and for doses over 1 – 10 dpa (Figure 1-25a). They also reported that the observation of radiation modified or induced phases during irradiation temperatures of 450 – 600 °C indicates higher segregation during irradiation and hence, is a sign of poor radiation resistance whereas presence of radiation enhanced (/retarded) phases is an indicator of radiation resistant microstructure.
In contradiction, Hashimoto et al. [81] witnessed presence of low density precipitates formed in stainless steel irradiated at 200 °C. Observation of γ’ (Figure 1-25b) in a CW 316 baffle-former bolt irradiated at a temperature of 300 °C in PWR to a dose of 8.5 dpa has been stated by Thomas et al. [82]. In addition to radiation enhanced (/retarded) phases, Renault et al. [55] also detected the presence of radiation induced phases in SA 304L irradiated in PHENIX Fast Breeder Reactor to a dose of 36 dpa at 390°C. They also
reported to observe γ’ attached to cavities. Other studies have also reported to observe γ’ at dislocation loops and cavities in stainless steels irradiated at higher temperature ( > 380 °C) and to higher doses (> 20 dpa) [50].