The effects of silicon and titanium on void swelling and phase stability in 12Cr-15Ni austenitic alloys irradiated with 46 MeV nickel ions

The structure of 20% cold-worked 12 wt%Cr-15 wt%Ni austenitic alloys with Si and Ti contents in the ranges 0.14-1.42 and <0.02-0.27 wt%, respectively, has been investigated by TEM following irradiation with 46 MeV Ni⁶⁺ ions to 60 dpa at 525, 575 or 625° C in the Harwell VEC. Increasing the Si content progressively reduced the void concentrations and swelling to zero at all three temperatures and the addition of Ti further reduced the void concentration and swelling. The higher Si alloys were structurally unstable during irradiations at 525 and 575°C, when γ'(Ni₃Si) particles, M₂₃X₆ type precipitates and grains of a bcc phase were formed. Small amounts of the γ' and bcc phases were observed following irradiation at 625° C, the principal precipitates being of the M₂₃X₆ type. Voids were present in both austenite and bcc phases in the lower Si alloys and segregation of an unidentified fcc phase was observed around voids. It was concluded that all the void nucleation occurred in the austenite phase and that the bcc phase resulted from the partial transformation of the austenite to α'-martensite during cooling from the irradiation temperatures. This occurred as a consequence of the M_s(α') transformation temperature being raised to well above ambient owing to elemental depletions of the alloy matrices resulting from precipitation and segregation.     

Radiation induced homogeneous precipitation in undersaturated solid-solutions

TEM study of 1 MeV electron irradiated Al 1.9 at% Zn solid solution shows that Zn precipitates form, under irradiation, at temperatures well above the Zn solvus temperature outside irradiation. The corresponding upward shift of this temperature is dose rate dependent. This new example of radiation-induced precipitation exhibits unexpected features, which are not accounted for by the available models: (1) no correlation exists between the location of the precipitates and that of the point defects sinks; (2) the precipitation of incoherent β phase with atomic volume smaller than that of the matrix, and of coherent G.P. zones both occurs; (3) the size of the incoherent β precipitates saturates at large dose. A general mechanism for solute concentration fluctuations under irradiation is proposed which qualitatively accounts for the formation of coherent G.P. zones and for the nucleation of solute clusters with more complex structures. A reanalysis of Russell's model [16] for the growth of incoherent precipitates shows that it may qualitatively account for the observed behavior of the β phase precipitates.     

Dissolution of precipitates in Al-Ge during 200 keV Al + ion irradiation

The effect of recoil dissolution on precipitate stability during irradiation has been studied in an Al-Ge alloy. Two-phase samples of Ge precipitates in an Al matrix have been irradiated at ? 100°C with 200 keV Al⁺ ions to doses between 6 and 122 dpa. Transmission electron microscopy observations have shown that individual precipitates are replaced by families of precipitates as a result of low-temperature irradiation and warming to room temperature. The microstructure at room temperature following low-temperature irradiation is thought to be determined by the amount of cascade mixing of solute atoms and solvent atoms during irradiation and the degree of re-nucleation in the solute-enriched region surrounding the original precipitate site during warming to room temperature.     

Void swelling of an oxide dispersion strengthened ferritic alloy in a high voltage electron microscope

An oxide dispersion strengthened ferritic alloy with nominal composition Fe-13Cr-3.5Ti-1.5Mo-2TiO₂ and a cast alloy with a composition close to that of the matrix of the oxide dispersion strengthened alloy are irradiated in a high voltage electron microscope in the temperature range 380–550°C. The alloys are doped with 0–30 ppm helium. For alloys containing 10 ppm He a peak swelling temperature at 450°C is found. A maximum swelling of 1.1% is found at an irradiation dose of 20 dpa. In the absence of He no swelling is found in the temperature range 430–470°C. The swelling rate is highest at the onset of swelling. The results obtained here are quite similar to those for some ferritic steels such as FV607, EM 12 and HT9, except for the influence of He and for the dose dependence.     

Zur beständigkeit des vanadiums und der legierung V-3Ti-15Nb in flüssigem lithium BEI 700°C

Unalloyed vanadium was deoxidised at 700°C in liquid lithium; deep cavities were formed in a zone 100 μm wide around grain boundaries. The alloy V-3Ti-15Nb proved to be stable against deoxidation and there was no grain-boundary attack by lithium. Both vanadium and the vanadium alloy however became carbonised, when unstabilised stainless steel was also immersed in the lithium (as was the case in these experiments). The interior of the steel capsule was observed to have become decarbonised. Lithium scarcely attacks vanadium alloys at all at 700°C, provided the transport of oxygen and carbon is prevented.     

Mechanical properties and microstructure in neutron-irradiated nickel-based alloys and stainless steels for supercritical water-cooled-reactor fuel cladding

ABSTRACT

To investigate the irradiation behavior of mechanical properties and microstructural changes of commercial Ni-based alloys and improved stainless steels, a neutron-irradiation experiment was performed at the Joyo reactor, and post-irradiation examinations with tensile tests and TEM observations were carried out. The room-temperature tensile tests showed that all specimens that were irradiated at 485°C exhibited significant hardening and ductile behavior, especially in alloy 625. The irradiation hardening of all specimens irradiated at 668°C was less than that of specimens irradiated at 485°C. The fine-grained stainless steel, T3 and the Zr-added stainless steels, H1 and H2 showed good mechanical-property performance with keeping ductility after neutron irradiation. Most alloys and steels showed ductile behavior on the fracture surface except for alloy 625 specimen. The TEM observations showed that a high density of tangled dislocations and irradiation-induced defect clusters formed in the stainless steels and Ni-based alloys irradiated at 485°C. At 668°C, the material microstructures coarsened and their dislocation density decreased significantly. Long rod-like precipitates of Zr(Cr, Fe) compounds formed in the H1 and H2 steels that were modified with Zr. The yield stress drop of T3 steel in tensile stress was observed and is caused by grain-size coarsening at an irradiation of 668°C.     

Observations of the microstructure of a 8% manganese stainless steel (ICL 016) before and after irradiation with 46 MeV nickel ions

Results are given of a preliminary investigation of the microstructure of a commercial Mn 8%/Cr 19%/Ni 7% austenitic steel (ICL 016) before and after irradiation with 46 MeV nickel ions. Pre-irradiation phases observed were Cu-rich precipitates (d ~ 10 nm) and α-MnS phase. A surface-localised ferromagnetism observed after annealing or irradiation was found to be due to α'-martensite formed as a result of an increase in the $\text{γ/ga'}$ transformation temperature due to evaporation of austenising elements such as Mn.Ion irradiation to 60 dpa at 625°C resulted in void-swelling of ~ 7% in solution-treated alloy containing 10 appm He. whereas swelling of ~ 1.8% occurred in the absence of helium. Irradiation also resulted in the formation of thin lath-like precipitates and the coarsening of the Cu precipitates. The results indicate that this manganese-containing alloy has an average swelling response when helium is present, with an indication that swelling can be reduced by pre-ageing at 700°C. In the ST or STA condition the alloy does not seem to offer any advantage in terms of void-swelling over other Fe-Cr-Ni austenitic steels currently favoured for LMFBR applications. The swelling sensitivity of the alloy to helium and the tendency to induced surface ferromagnetism indicate the need for further study before selecting this type of alloy for use in fusion reactors.     

Void Swelling of Solution-Annealed Type 316 Stainless Steel under Long Time Proton Irradiation

Solution-annealed type 316 stainless steel was irradiated by 150 keV proton to a dose of about 6 dpa at the irradiation temperature ranging 450–700°C. To examine the effect of aging during irradiation, the present proton irradiation was carried out for about 25 h at a low dose rate of 7×10–^?5dpa/s. The specimens without He preinjection showed much smaller void swelling than those preinjected with He to the content of 10 at.ppm. Similarly to the case of neutron irradiations, the void swelling in the He preinjected specimens showed the temperature dependence with double peaks, and the peak swelling temperatures were about 550 and 650°C. In these specimens with He preinjection. void number density decreased and average void diameter increased with the increase of irradiation temperature in the range of 450–600°C, but these trends were reversed between 600 and 650°C. The volume of the grain boudary M₂₃C6 precipitates increased with the increase of irradiation temperature from 600 to 700°C, and it was concluded that the decrease of soluble carbon due to the precipitation of M₂₃C6 caused the second swelling peak at 650°C.     

Effects of an alloying element on a c-component loop formation and precipitate resolution in Zr alloys during ion irradiation

It is important to clarify the mechanisms of the dislocation loop formation, dissolution of precipitates to understand the degradation behavior of the fuel cladding tubes in light water reactors (LWR) under neutron irradiation. In this study, 3.2 MeV Ni ion irradiation was carried out at 400°C on Zircaloy-2 and two types of model alloys with and without Fe (Zr-1.5Sn-0.3Fe and Zr-1.5Sn). To understand the effects of hydrogen, 60 and 300 ppm pre-injected Zircaloy-2 samples were also irradiated. The microstructure was observed with a conventional transmission electron microscopy. Additionally, the dissolution of precipitates and the enrichment of the alloying element due to irradiation were analyzed using a spherical aberration (Cs)-corrected high-resolution analytical electron microscope. After ion irradiation at 400°C, the dissolution of Fe-enriched precipitates and the c-component dislocation loops were observed in the region of peak ion damage. Observations by STEM-EDS showed that Fe atoms were enriched in the c-component dislocation loops. On the contrary, the c-component dislocation loops were detected in Fe-containing alloys (Zircaloy-2 and Zr-1.5Sn-0.3Fe alloy) but were not in the Zr-1.5Sn alloy. These results indicate that the dissolution of Fe-enriched precipitates and the enhanced formation of c-component dislocation loops are essential for the degradation of LWR fuel cladding under irradiation.     

The influence of neutron irradiation in FFTF on the microstructural and microchemical development of Mo-41Re at 470-730 °C

Specimens of Mo-41 wt% Re irradiated in the fast flux test facility (FFTF) experience significant and non-monotonic changes in density that arise first from radiation-induced segregation, leading to non-equilibrium phase separation, and second by progressive transmutation of Re to Os. As a consequence the density of Mo-41Re initially decreases and then increases thereafter. Beginning as a single-phase solid solution of Re and Mo, irradiation of Mo-41 wt% Re over a range of temperatures (470-730 °C) to 28-96 dpa produces a high density of thin platelets of a hexagonal close-packed (hcp) phase identified as a solid solution of Re, Os and possibly a small amount of Mo. These hcp precipitates are thought to form in the alloy matrix as a consequence of strong radiation-induced segregation to Frank loops. Grain boundaries also segregate Re to form the hcp phase, but the precipitates are much bigger and more equiaxed in shape. Although not formed at lower dose, continued irradiation at 730 °C leads to the co-formation of late-forming chi-phase, an equilibrium phase that then competes with the preexisting hcp phase for rhenium.     

A SANS investigation of the irradiation-enhanced α–α′ phases separation in 7–12 Cr martensitic steels

Five reduced activation (RA) and four conventional martensitic steels, with chromium contents ranging from 7 to 12 wt%, were investigated by small angle neutron scattering (SANS) under magnetic field after neutron irradiation (0.7–2.9 dpa between 250 and 400 °C). It was shown that when the Cr content of the b.c.c. ferritic matrix is larger than a critical threshold value (∼7.2 at.% at 325 °C), the ferrite separates under neutron irradiation into two isomorphous phases, Fe-rich (α) and Cr-rich (α^′). The kinetics of phase separation are much faster than under thermal aging. The quantity of precipitated α^′ phase increases with the Cr content, the irradiation dose, and as the irradiation temperature is reduced. The influence of Ta and W added to the RA steels seems negligible. Cold-work pre-treatment increases slightly the coarsening of irradiation-induced precipitates in the 9Cr–1Mo (EM10) steel. In the case of the low Cr content F82H steel irradiated 2.9 dpa at 325 °C, where α^′ phase does not form, a small irradiation-induced SANS intensity is detected, which is probably due to point defect clusters. The α^′ precipitates contribute significantly to the irradiation-induced hardening of 9–12 wt% Cr content steels.     

Tensile behavior of irradiated 12Cr-1MoVW steel

The effect of irradiation on the tensile properties of 12Cr-1MoVW steel given two different normalized-and-tempered heat treatments was determined for specimens irradiated in the Experimental Breeder Reactor-II (EBR-II) at 390. 450, 500 and 550°C to ~13 dpa. Tensile tests were conducted at the irradiation temperature and room temperature on the irradiated specimens, as-heat-treated specimens, and as-heat-treated specimens thermally aged 5000 h at the irradiation temperature.Irradiation at 390°C increased the 0.2% yield stress and ultimate tensile strength above the strength of the unaged and aged controls for both heat-treated conditions. The effect was greater for the steel heat treated with the shorter austenitization and tempering times. The increased strength was accompanied by a slight decrease in ductility. After irradiation at 450, 500 and 550°C the yield stress, ultimate tensile strength, and ductility were similar to the as-heat-treated and the thermally aged controls. Strength changes were discussed in terms of microstructural changes observed in other studies.     

Void Swelling and Precipitation in Ti-Modified Austenitic Fe-Ni-Cr Alloys

Irradiation temperature dependence of void swelling, and the effects of aging before irradiation on void swelling and on precipitation were examined in three Ti-modified austenitic steels called No. 1 (Fe-14Ni-16Cr-1.8Mn-2.4Mo-0.1Nb-0.1Ti-0.9Si-0.07C), No. 2 (Fe-14Ni-15Cr-1.6Mn-2.6Mo-0.24Ti-1.0Si-0.06C) and No. 3 (Fe-25Ni-15Cr-1.6Mn-2.4Mo-0.4Ti-1.0Si-0.06C). After irradiation to 10 dpa with 150–200 keV proton, the cold worked No. 1 showed swelling peak of 0.4% at 823 K, the cold worked No. 2 0.15% at 773 K and the cold worked No. 3 0.08% at 723 K. Sample Nos. 2 and 3 which were aged at 923 K for 5.4 Ms (1,500 h) showed larger void swelling because of higher void number density than the unaged ones after the irradiation to 20–40 dpa at 923 K. The amount of intragranular TiC precipitates in the irradiated specimens did not vary significantly with preirradiation treatments. Voids were mostly attached to eta (M₆C) phase which was produced during irradiation. The number density of eta phase produced during the irradiation in aged specimens was much higher than that in unaged ones. This is thought to give a main reason why void swelling in aged specimens was larger than that in unaged ones.     

Irradiation damage in proton irradiated palladium-iron solid solutions

Transmission electron microscopy was used to investigate the irradiation damage, in particular irradiation induced precipitation (IIP), in Pd-base alloys containing 2, 8, 12 and 18 at % Fe. The specimens were irradiated mainly using 400 keV protons at a current density of 0.16 μA/mm² over the temperature range 110 to 750°C. A few samples containing 2 and 8% Fe were also irradiated using 3 MeV NiP⁺ ions. The irradiation microstructure of the proton irradiated alloys consists of dislocation loops over the temperature range 110 to 550°C and voids up to 650°C in all the alloys. IIP of Pd₃Fe was observed only in the Pd-18% Fe alloy between 110 and 500°C, irradiated to a dose of 0.9 dpa. Pd₃Fe was associated with dislocation loops, voids and grain boundaries. IIP was not observed in the Pd-2,8 and 12% Fe alloys proton irradiated to the same dose, nor to a higher dose of 1.5 dpa. It was also not observed in the 2 and 8% Fe alloys irradiated at 600 and 700°C by 3 MeV Ni⁺ ions.The absence of IIP in the more dilute alloys is attributed to the fast back diffusion of Fe atoms, which is due to the high mobility of vacancies in these alloys. This causes the Fe concentration at the sinks to remain below the solubility limit. Therefore, even though Fe is an undersized solute, the size effect alone is not sufficient for the production of IIP at point defect sinks in most Pd-Fe alloys. It is proposed that IIP can occur only when the alloy concentration is high enough to minimize the rate of back diffusion, which depends not only on the vacancy mobility but also on the concentration gradient near point defect sinks.     

Modification of the precipitate interface under irradiation and its effect on the stability of precipitates

Ni-alloys containing coherent γ'-précipitates of different lattice mismatch and an A1-Mg-Si alloy with semicoherent β' -precipitates were irradiated with self-ions of 25 0 keV energy. The microstructural changes as a function of irradiation dose, ranging from 2 to 398 dpa, and temperature, ranging from 323 to 998 K, were followed by transmission electron microscopy. It was found that irrespective of the initial interface, the alloy chemistry or the irradiation condition, the precipitate interface is considerably modified under irradiation. Further, the interfacial dislocations, evolved either during preaging or irradiation, obstructed the dissolution, and their loss marked the beginning of rapid dissolution starting from the surface of precipitates. Based on these observations, a hypothesis is proposed to account for the interfacial modification and to explain the stabilizing effect of the interfacial dislocations during the radiation-induced dissolution.     

Elevated-temperature tensile properties of irradiated 214 Cr-1 Mo steel

The effect of neutron irradiation on the tensile properties of normalized-and-tempered $\text{2}\text{1}\text{4}$ Cr-1 Mo steel was determined for specimens irradiated in Experimental Breeder Reactor II (EBR-II) at 390 to 550°C. Two types of unirradiated control specimens were tested: as-heat-treated specimens and as-heat-treated specimens aged for 5000 h at the irradiation temperatures. Irradiation to approximately 9 dpa at 390° C increased the strength and decreased the ductility compared to the control specimens. Softening occurred in samples irradiated and tested at temperatures of 450, 500, and 550 °C; the amount of softening increased with increasing temperature. The tensile results were explained in terms of the displacement damage caused by the irradiation and changes in carbide precipitates that occur during elevated-temperature exposure.     

Elevated-temperature tensile properties of irradiated 9 Cr-1 MoVNb steel

Normalized-and-tempered 9 Cr-1 MoVNb steel tensile specimens were irradiated in the Experimental Breeder Reactor-11 (EBR-11) at 390, 450, 500, and 550°C to ～2.1 and 2.5 × 10²⁶ neutrons/m² (> 0.1 MeV), which produced displacement damage levels of ～10 and 12 dpa, respectively. Tensile tests were conducted at the irradiation temperature and at room temperature. In addition to the irradiated specimens, as-heat-treated specimens and as-heat-treated specimens thermally aged at the irradiation for 5000 h were also tested.Thermal aging had no effect on the unirradiated tensile properties. Irradiation at 390°C increased the 0.2% yield stress and the ultimate tensile strength above those of the unirradiated control specimens. The ductility decreased slightly. After irradiation at 450, 500, and 550°C, the tensile properties were essentially the same as the unirradiated values. The hardening at 390°C was attributed to the dislocation and precipitate structure formed during the irradiation. The lack of hardening at 450°C and higher correlates with an absence of an irradiation-induced damage structure.     

Observation and analysis of damage structure in Al and Al/Mg (N4) alloy after irradiation with 100 and 400 keV aluminium ions

A comparative TEM study has been made of ion irradiation damage structure in pure aluminium, commercial aluminium (grade 1100) and in a modified N4 (Al/2.95% Mg) alloy of the type used in the construction of the calandria of the Winfrith prototype SGHW Reactor. Atom displacements equivalent to many years neutron irradiation were simulated by bombardment with 100 and 400 keV Al⁺ ions to doses of up to 200 dpa at temperatures between 30 and 250°C. Dynamic observations of damage formation were made during irradiation with 100 keV ions in a linked heavy-ion accelerator/200 keV electron microscope, and further results were obtained by 400 keV Al⁺ ion bombardment in a Cockcroft-Walton accelerator. Dislocation structure and voids were seen in aluminium irradiated with 100 and 400 keV A1⁺ at temperatures between 30 and 250°C. Void swelling of 8.7% at 104 dpa was a maximum at 1&#x0303;50°C in type 1100 aluminium. No voids were found at temperatures μ 250°C. No voids were seen in the Al/Mg N4 alloy after bombardments up to 200 dpa with 100 keV A1⁺, and 150 dpa with 400 keV Al⁺ at temperatures between 50 and 170°C. The void-resistant property is consistent with observations in the USA of neutron-irradiated 5052 Al alloy which has a similar magnesium content to the modified N4 alloy. The 1100 alloy and N4 results have been analysed using the rate theory of swelling. The absence of voids in the N4 alloy indicates an effective vacancy annihilation mechanism, which possibly occurs at small precipitates formed during irradiation.     

Effect of irradiation on the deformation of copper and copper-aluminium alloys

The effect of low-temperature irradiation (<70°C) on the deformation of pure copper, Cu-0.09 wt% Al alloy, and Cu-0.19 wt% Al alloy was investigated. Tensile specimens were prepared from single crystals of various orientations, and were exposed to a neutron fluence of 6.85 ×10¹⁹ n/cm²(E>1 MeV) at a temperature less than 70°C. All irradiated and unirradiated specimens were pulled. A large increase in critical shear stress due to irradiation was observed; the increase was smaller in Cu-0.19 wt% Al alloy than in pure copper and Cu-0.09 wt% Al alloy. Ultimate shear stress and shear strain were less influenced by irradiation. Yield points were observed in all irradiated specimens. The yield drop was large in irradiated pure copper, and decreased with the increase of aluminium content in copper-aluminium alloys. All unirradiated specimens showed a high work-hardening coefficient (n) in the beginning of the deformation, followed by a lower value. By irradiation, the first value drastically decreased, while the second remained nearly constant. Shear stress and shear strain were influenced by crystal orientation.     

Corrosion characteristics of Hastelloy N alloy after He+ ion irradiation

With the goal of understanding the invalidation problem of irradiated Hastelloy N alloy under the condition of intense irradiation and severe corrosion, the corrosion behavior of the alloy after He⁺ ion irradiation was investigated in molten fluoride salt at 700 °C for 500 h. The virgin samples were irradiated by 4.5 MeV He⁺ ions at room temperature. First, the virgin and irradiated samples were studied using positron annihilation lifetime spectroscopy (PALS) to analyze the influence of irradiation dose on the vacancies. The PALS results showed that He⁺ ion irradiation changed the size and concentration of the vacancies which seriously affected the corrosion resistance of the alloy. Second, the corroded samples were analyzed using synchrotron radiation micro-focused X-ray fluorescence, which indicated that the corrosion was mainly due to the dealloying of alloying element Cr in the matrix. Results from weight-loss measurement showed that the corrosion generally correlated with the irradiation dose of the alloy.