The ageing processes of simplified nuclear glasses under β-irradiation have been investigated on zirconium-bearing aluminoborosilicate glasses by micro Raman spectroscopy. No dose rate effect has been detected between 2.103 and 2.104 Gy/s for a 3.107 Gy integrated β dose. Important structural differences appear between the volume and the surface, with an increase of the polymerisation and a better defined danburite-like ring structure in the bulk relative to the surface. Molecular 02 is created during glass ageing, mostly close to the sample surface. Both processes indicate an important reorganisation of the glassy matrix during β-irradiation.

We have studied by electron microscopy the copper-rich precipitates in an Fe-1.3wt%Cu model alloy irradiated with neutrons to doses of 8.61 × 10−3dpa and 6.3×10−2 dpa at a temperature of-270°C. In the lower dose material a majority (ca. 60%) of the precipitates visible in high-resolution electron microscopy were twinned 9R precipitates of size ∼ 2−4 nm, whilst ca. 40 % were untwinned. In the higher dose material, a majority (ca. 75%) of visible precipitates were untwinned although many still seemed to have a 9R structure. The average angle α between the herring-bone fringes in the twin variants was measured as 1250, not the 1290 characteristic of precipitates in thermally-aged and electron-irradiated material immediately after the bcc-→9R martensitic transformation. We argue that these results imply that the bcc-→9R transformation of small (< 4 nm) precipitates under neutron irradiation takes place at the irradiation temperature of 270°C rather than after subsequent cooling. Preliminary measurements showed that precipitate sizes did not depend strongly on dose, with a mean diameter of 3.4 ± 0.7 nm for the lower dose material, and 3.0 ± 0.5 rim for the higher dose material. This result agrees with the previous assumption that the lack of coarsening in precipitates formed under neutron irradiation is a consequence of the partial dissolution of larger precipitates by high-energy cascades.

Anomalous small-angle x-ray scattering (ASAXS) experiments were performed at energies near the Mn-K and Fe-K absorption edges on an Fe-0.9 Cu wt.%-1.0 Mn wt.% alloy subjected to either annealing at 450°C, 24 hrs or irradiation with 10 MeV electrons to 0.5 mdpa at 300°C. A Guinier analysis of the net scattered intensity revealed scatterers with radii of 16.5.16.8 Å and 16.4.17.3 Å for the irradiated and annealed foils, respectively. The variation in net scattered intensity with photon energy in the annealed sample was consistent with scattering contrast variations expected for Cu-rich, Cu-Mn precipitates. Minimal variation in net intensity (scattering contrast) with energy determined that Cu-Mn precipitates were not detected in the irradiated sample.

Over 1200 measurements of grain boundary composition and microstructure have been made on 14 different austenitic Fe-Cr-Ni alloys following proton irradiation in the temperature range 200-600°C and in the dose range 0.1-3.0 dpa. Grain boundary composition measurements revealed that Cr depletes at grain boundaries, Ni enriches and Fe can either enrich or deplete depending on alloy composition. Analysis of temperature and composition dependence of RIS revealed that the magnitude and direction of grain boundary segregation depends on alloy composition because the value of migration enthalpy differs among the alloy constituents, and diffusivities of the alloy constituents are composition-dependent. The dose dependence of segregation revealed ordering in Ni-base alloys and temperature dependence was used to show that RIS occurs by vacancy exchange rather than an interstitial binding mechanism. The dependence of segregation on composition is consistent with all known, relevant neutron data.

Changes in mechanical and corrosion properties caused by the development of radiation-induced microstructures have relevance to the aging and lifetime extension of light water reactors (LWR‘s). However, much of the current data related to microstructural development in irradiated metals are generated from studies carried out at much higher dose-rates than encountered in LWR‘s. An opportunity exists to study the influence of low dose-rate irradiation on microstructural development for a variety of structural and surveillance materials extracted from the experimental breeder reactor EBR-lI. In this study, irradiated 304 stainless steel hexagonal “hex” duct material is examined in order to compare microstructures in the dose-rate range of 10−7 - 10−9 dpa/sec. The samples, taken from the reflector locations in EBR-II, experienced a total dose between 10 and 12 dpa at a temperature of ∼375 °C. Transmission electron microscopy (TEM) analysis reveals that there is a moderate dose-rate effect on microstructural development for samples irradiated in the range of 2 × 10−8 to 4 × 10−8. dpa/sec, however a substantial dose-rate effect exists between dose-rates of 2 × 10−8 and 1 × 10−9 dpa/sec Results detail the development of the microstructure in terms of radiation-induced cavities, dislocations, and precipitates.

A model for the development of microstructure during irradiation in fast reactors has been adapted for light water reactor (LWR) irradiation conditions (275 ∼ 325 °C, up to ∼ 10 dpa). The original model was based on the rate-theory, and included descriptions of the evolution of both dislocation loops and cavities. The model was modified by introducing in-cascade interstitial clustering, a term to account for the dose dependence of this clustering, and mobility of interstitial clusters. The purpose of this work was to understand microstructural development under LWR irradiation with a focus on loop nucleation and saturation of loop density. It was demonstrated that in-cascade interstitial clustering dominates loop nucleation in neutron irradiation in LWRs. Furthermore it was shown that the dose dependence of in-cascade interstitial clustering is needed to account for saturation behavior as commonly observed. Both quasi-steady-state (QSS) and non-steady-state (NSS) solutions to the rate equations were obtained. The difference between QSS and NSS treatments in the calculation of defect concentration is reduced at LWR temperature when in-cascade interstitial clustering dominates loop nucleation. The mobility of interstitial clusters was also investigated and its impact on loop density is to reduce the nucleation term. The ultimate goal of this study is to combine the evolution of microstructure and microchemistry together to account for the radiation damage in austenitic stainless steels.

The pre-existing segregation at grain boundaries in two austenitic stainless steels has been investigated by atom probe field ion microscopy and analytical electron microscopy. In addition, the effect of radiation-induced segregation on the near-grain-boundary composition has been studied by analytical electron microscopy. Pre-existing enrichment of Cr, Mo, B, C and P and depletion of Fe and Ni near grain boundaries has been observed. Significant affinity between Mo and N in both alloys is indicated by the detection of MoN2+` molecular ions during field evaporation. The pre-existing segregation is modified by radiation-induced segregation resulting in Ni and Si enrichment near the boundary as well as depletion of chromium adjacent to the boundary resulting in a “W-shaped” Cr profile.

Radiation-induced segregation (RIS) has been identified as a potential contributor to irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels in reactor core components. The occurrence of grain boundary segregation prior to irradiation influences both the shape and magnitude of RIS profile development during subsequent irradiation. In an effort to better understand the impact of this pre-irradiation enrichment on RIS profile development, the evolution of grain boundary Cr segregation profiles with irradiation dose has been characterized. Commercial purity and high-purity austenitic stainless steels with different initial levels of grain boundary Cr have been irradiated with neutrons (at 275°C) or protons (at 360-400°C) to doses up to ∼5 dpa. Grain boundary composition profiles were measured before and after irradiation using scanning transmission electron microscopy with energy dispersive xray spectroscopy (STEM-EDS). The initial enrichment of Cr is shown to delay radiation-induced Cr depletion and produce a “W-shaped” profile at low irradiation doses. Further irradiation causes the central peak of the W to decrease, eventually resulting in the classical “V-shaped” depletion profile. Possible mechanisms for the pre-irradiation enrichment and its evolution into a “W-shaped” profile will be discussed.

The stability of Cu precipitates in an Fe-1.3wt%Cu alloy under 300 keV Fe+ion irradiation has been investigated using transmission electron microscopy and high-resolution electron microscopy. The irradiations were carried out between room temperature and 550°C at displacement rates of 103 to 10−2 dpa(s)−1 to fluences of up to 30 dpa. Copper precipitates were found to keep their shape but decrease in size under all irradiation conditions. The results are discussed within the framework of a competitive process between irradiation induced ballistic destruction of precipitates by cascades and irradiation-enhanced precipitation.

Irradiation hardening is an issue of practical importance as it relates to the remanent life and the nature of failure of reactor components exposed to displacement-producing radiation. Understanding these phenomena requires studies of fundamental microstructural mechanisms of hardening. In this paper, we analyze the tensile behavior of pressure vessel steels (A212B and A350) irradiated by neutrons and electrons. The results show that the post-yield true stress/true strain behavior can provide fingerprints of the different hardening effects that result from irradiation by the two particles, which suggests correspondingly different hardening microstructures for the two particles. Microstructurally-based models for irradiation-induced yield strength increases, combined with a model for strain hardening, are used to make predictions of the different effects of irradiation by the two particles on the entire flow curve that agree well with data.

On a variety of pressure-vessel (PV) steels, we have observed changes in the average positron lifetime with increasing (near end-of-life) neutron fluences. Samples were irradiated at reactor-temperature and subjected to post-irradiation annealing, and they were examined using positron annihilation lifetime spectroscopy (PALS). The measured average positron lifetimes in high-temperature (2900 C-300° C) irradiated PV steels decrease with increasing neutron damage up to fluences of 8.5×1018 cm−2 and increase again at higher fluences. Annealing of high-fluence, 300° C irradiated ASTM A508 PV steel samples produces an initial decrease in average positron lifetimes with increasing annealing temperatures of up to 400° C, followed by an increase in average positron lifetime with higher annealing temperatures, when samples were annealed in successive 24-hour steps. A sample of weld steel, irradiated to 2.2×1019 cm−2 at 290° C, shows similar behavior in which the minimum lifetime occurs at ≈ 450° C. These trends are similar to those seen in previous studies performed on VVER and other ferritic steels.

The effect of post-irradiation annealing (PIA: at 573-958 K for 0.1-1000 h) on the intergranular S and P segregation has been examined in ferritic alloys doped with Cu and/or P, all containing residual S. In a Cu-doped alloy showing no P segregation, the rate and amount of segregated S were promoted more by PIA than thermal annealing. Intergranular P enrichment was observed during an early PIA stage at lower temperatures in P-containing alloys. The intergranular P desegregation and S segregation concurrently occurred at higher PIA temperatures. By applying a modified McLean theory, it is shown that the S segregation in the post-irradiation annealed Cu-doped alloy is controlled by vacancy-enhanced diffusion while the P segregation in P-containing alloys subjected to lower temperature PIA proceeds via mobile P-interstitial complexes. Based on a Gaussian relaxation model, the P desegregation during prolonged PIA at higher temperatures is assisted by the formation and migration of vacancies.

A model is presented that estimates the effect of radiation damage on stress corrosion cracking (SCC), resulting from microstructural evolution induced by irradiation. The model is based on the Ford-Andresen film-rupture and slip-dissolution model and on the observation that the crack propagation rate increases as the material hardens. The model relates the neutron and gamma exposure to the irradiation hardening and consequent increase in the crack tip strain rate. A chemical rate theory model was employed to describe the evolution of microstructure in stainless steel leading to radiation hardening. The increase in yield stress can be correlated to an increase in the crack tip strain rate, and thus to SCC behavior. This approach was used to evaluate the effects of irradiation hardening on core shroud cracking. The results show that hardening can have a substantial effect on core shroud cracking.

I is generally recognized that radiation induced sensitization plays an important role in initiating irradiation assisted stress corrosion cracking (IASCC) of austenitic stainless steels in reactor core internal of light water reactor. However, the synergism between radiation sensitization and prior thermal sensitization is unclear. This situation is likely to occur in most welded core internal structures subjected to neutron irradiation. In this study, the effect of prior thermal treatment on radiation sensitization were investigated on proton irradiated Type 304 stainless steel (SS) of initially as-received (AR) and thermal-sensitized (SEN) conditions. The Cr depletion profiles were measured by field emission gun transmission electron microscopy/ energy dispersive spectroscopy (FEGTEM / EDS), and were calculated by a radiation induced segregation (RIS) model.

The different initial conditions were input in the RIS model calculations. For the asreceived condition, the initial Cr profile was modeled by a uniform concentration distribution. For the initially thermal-sensitized condition, the wider Cr depletion profile measured by FEGTEM / EDS was input as the initial condition. The results showed that radiation sensitization is characterized by a very narrow Cr depleted zone. The Cr content at grain boundary tends to be lower as radiation dose increases. Comparing with the non-sensitized (asreceived) specimens with the same dosage, the grain boundary Cr content without prior sensitization is higher than that with sensitization pre-treatment. The deeper grain boundary Cr concentration of irradiated thermally sensitized sample is induced not only from proton irradiation effect, but also resulted from the pre-existing Cr depletion.

Both microstructural effects (radiation hardening) and microcompositional effects (radiationinduced segregation (RIS)) have been identified as potential contributors to irradiation-assisted stress corrosion cracking (IASCC). However, the importance of each in IASCC is unclear. In an effort to isolate their roles, post-irradiation annealing of proton-irradiated 304L stainless steel samples has been studied. Model simulations of post-irradiation annealing at intermediate temperatures (400-650°C) indicate that microstructural features such as dislocation loops are removed faster than RIS. Simulations also predict that there exist time-temperature combinations that will significantly reduce the dislocation loop population while leaving the grain boundary segregation essentially unaffected. Ultra-high purity (UHP) 304L stainless steel samples have been irradiated with 3.2 MeV protons at 360°C to 1.0 dpa and then annealed in-situ as thinned TEM disks in the temperature range of 500°C- 625°C for times between 20 and 60 minutes. RIS and dislocation loops were characterized before and after annealing in the same areas of each specimen. Post-irradiation anneals at 500TC for 45 minutes or 60 minutes resulted in no appreciable change in dislocation loop population or RIS. Annealing at 600°C for 20 minutes., or 40 minutes and at 625°C for 40 minutes, resulted in decreasing dislocation loop densities and increasing loop size with increasing annealing time or temperature, while the amount of RIS did not change significantly from the pre-annealed condition.

Interactions between point defects and major solute strongly influence grain boundary concentrations during heat treatment, irradiation and annealing of austenitic stainless steels. Previous approaches to nonequilibrium segregation emphasize only elastic defect-solute interactions. The present evaluation of nonequilibrium concentrations at grain boundaries indicates chemical interactions unique to solution annealing and cooling during thermal nonequilibrium segregation (TNES). Subsequent to TNES, radiation-induced segregation and post-irradiation annealing are modeled and compared with measured changes in grain boundary composition. The latter two mechanisms are controlled by exchanges between vacancies and major solute such as Cr.

As part of the shutdown of the EBR-II reactor, structural materials were retrieved to analyze the effect of long term, low dose rate irradiation. In this work, the effect of low dose rate (10 to 10−9 dpa/s) irradiation on grain boundary and void surface chemistry is analyzed. These dose rates are comparable to those in light water reactor structural components. The components were irradiated at 375-379°C, temperatures near the highest temperatures experienced in pressurized water reactors. Radiation-induced segregation (RIS) was measured on samples taken from 304 stainless steel hex ducts irradiated to doses between 10 and 12 dpa. Radiation-induced segregation is shown to vary with dose rate, with measured grain boundary chromium concentrations reaching as low as 5 at. % and nickel concentrations reaching as high as 33 at. %. For some radiation conditions, significant grain boundary precipitation occurs, possibly leaving components susceptible to environmental attack.

The influence of the irradiation temperature Tirr on the development of disordered zones produced at displacement cascades in Ni3A1 by heavy-ion irradiation with 50 keV Ta+ and 300 keV Ni+ ions has been investigated. The normalised number density (yield) of disordered zones for 300 keV Ni+ irradiation showed a sharp fall between Tirr= 373 K and 573 K. For 50 keV Ni+ irradiation there was a similar fall between 573 K and 673 K. The mean diameters of the disordered zones produced by 300 keV Ni+ ions decreased by about 2 nm between room temperature and 573 K, and there was a tendency for larger zones to become more regular in shape. For 50 keV Ta+ ions, a similar trend was observed between 573 K and 873 K. An annealing experiment confirmed that disordered zones produced at lower temperatures were stable up to a temperature of about 673 K, showing that these trends cannot be due to thermal annealing of disordered zones. The experimental results are consistent with an increased tendency for reordering at the peripheries of disordered zones, due to the increased lifetimes of thermal spikes at higher irradiation temperatures.

A low-temperature in situ ion-irradiation and annealing experiment has been performed by TEM in copper. Most defect clusters which persisted through an anneal to 120 K showed no size changes within the resolution (0.5 nm) of a new weak-beam sizing technique. Of 55 defects measured under a range of weakly diffracting conditions, 7 showed measurable size decreases while 3 showed size increases. We argue that these clusters are likely to be of vacancy and interstitial nature, respectively. Also on annealing to 120 K a fraction of about 25% of the clusters formed by irradiation with 600 kV Cu+ ions at 20 K disappeared, while a similar number of clusters appeared in different locations.

In copper which was irradiated by fission neutrons at 300°C in a temperature controlled rig to 0.0003 dpa with the damage production rate of 6 × 10−5 dpa/hr, the number of voids exceeds the number of gas atoms of hydrogen and helium which were generated by the transmutation reaction. The number density of voids was almost the same in as-received and gas-atomremoved copper which was neutron irradiated to 0.0003 dpa. These results suggest that voids can be formed by clustering of only vacancies. In 5 MeV Ni-ion irradiated copper, voids could not be observed in specimens which were irradiated between 200°C and 450°C from 0. 1 dpa to 30 dpa with the damage production rate of 6 dpa/hr. Voids were observed only in copper which was Ni-ion irradiated at 500°C. The difference of void formation on neutron-irradiation and ionirradiation is discussed. It is concluded that the production rate of damage is an important factor on the void formation in irradiated metals.