Radiation-induced Precipitation Research Articles

Analytical model of radiation-induced precipitation at the surface of dilute binary alloy

Growth of precipitate layer at the foil surface of an undersaturated binary alloy under uniform irradiation is treated analytically. Analytical expressions for the layer growth rate, layer thickness limit and final component concentrations in the matrix are derived for coherent and incoherent precipitate–matrix interfaces. It is shown that the high temperature limit of radiation-induced precipitation is the same for both types of interfaces, whereas layer thickness limits are different. A parabolic law of the layer growth predicted for both types of interfaces is in agreement with experimental data on γ ′-phase precipitation at the surface of Ni–Si dilute alloys under ion irradiation. Effect of sputtering on the precipitation rate and on the low temperature limit of precipitation under ion irradiation is discussed.

Radiation-induced precipitation in V–(Cr,Fe)–Ti alloys irradiated at low temperature with low dose during neutron or ion irradiation

Effects of neutron irradiation damage on mechanical properties and microstructure were studied for V–4Cr–4Ti–0.1Si, V–3Fe–4Ti–0.1Si and vanadium binary alloys with three fluences between 0.01 and 0.1 dpa for 220°C and 340°C. Irradiation hardening appeared in V–(Cr,Fe)–Ti alloys at an early stage of irradiation. Microstructural analysis showed that the irradiation hardening in V–(Cr,Fe)–Ti alloys is caused by radiation-induced precipitates (RIPs). In order to investigate the behavior of RIPs, ion irradiations of highly purified vanadium alloys were performed. It is likely that the nucleation and growth processes in vanadium alloys containing titanium are independent of the impurity concentration during ion irradiation.

Features of radiation damage of vanadium and its alloys at a temperature of 330–340°C

Microstructural changes of vanadium alloys after irradiation at 340°C to 12 dpa in the BOR-60 reactor in 7Li environment is analyzed. Materials are vanadium and its alloys V–3Ti, V–3Fe, V–6Cr, V–4Cr–4Ti, V–5Cr–10Ti, V–6Cr–1Zr–0.1C. Void formation was observed in the binary alloys V–3Fe, V–3Ti and V–6Cr. It is shown that three–four-fold increase in V–4Cr–4Ti yield stress is produced by the formation of dislocation loops (DLs) and fine radiation-induced precipitates (RIPs) with a density of 1.7×1017cm−3. It is expected that embrittlement of the welds will be worse because density of DLs and RIPs is 1.4–1.6 times higher. Besides, invisible coherent or semi-coherent RIPs are formed in the fusion zone. Elemental maps of the rupture surface of irradiated V–4Cr–4Ti are presented.

Microstructural examination of V–(3–6%)Cr–(3–5%)Ti irradiated in the ATR-A1 experiment

Microstructural examination results are reported for four heats of V–(3–6%)Cr–(3–5%)Ti irradiated in the ATR-A1 experiment to ∼4 dpa at 200°C and 300°C to provide an understanding of the microstructural evolution that may be associated with degradation of mechanical properties. Fine precipitates were observed in high density intermixed with small defect clusters for all conditions examined following the irradiation. The irradiation-induced precipitation does not appear to be affected by preirradiation heat treatment or composition.

Intergranular and intragranular phosphorus segregation in Russian pressure vessel steels due to neutron irradiation

Russian reactor pressure vessel steels have been studied in three conditions: initial, irradiated and annealed. It has been established that irradiation induces both intergranular as well as intragranular phosphorus segregation. Fractographic studies demonstrated that brittle intergranular and ductile intergranular fracture surfaces of Charpy specimens appear as a result of intergranular and intragranular segregation, respectively. Transmission electron microscope (TEM) studies have revealed radiation-induced precipitates on interface boundaries to which intragranular phosphorus segregation occurs. Auger electron spectroscopy (AES) has been applied to detect phosphorus enrichment of fracture surfaces in the regions of brittle and ductile intergranular fractures.

Microstructural Evolution of Ni-doped Ferritic/martensitic Steels Irradiated in the HFIR

ABSTRACTThe microstructures of reduced-activation ferritic/martensitic steels, 9Cr-2WVTa and 9Cr-2WVTa doped with 2% Ni, irradiated at 400oC up to 12 dpa in the High Flux Isotope Reactor (HFIR), were investigated by transmission electron microscopy. Specimens were tempered at two different temperatures in order to investigate the effects of tempering on microstructural evolution during irradiation. Before irradiation, the lath width of Ni-doped 9Cr-2WVTa was somewhat narrower and the dislocation density tended to be higher compared with 9Cr-2WVTa. Dislocation density of specimens tempered at 750°C was lower than that tempered at 700°C. In all steels, precipitates on grain and/or lath boundaries were mainly M23C6, and there were a few TaC along dislocations in the matrices. Irradiation-induced cavities were observed in all the steels. The cavity number density of the Ni-doped 9Cr-2WVTa was higher than that of 9Cr-2WVTa due to the higher concentration of helium; however, swelling in each steels was < 0.1% because cavity sizes were so small. There was no difference of cavity number density between the steels tempered at 700°C and 750°C, but the mean size of the cavities in the steels tempered at 750°C was larger than that tempered at 700°C. Irradiation-induced a0<100> and (a0/2)<111> type dislocation loops were observed in all steels; number density and mean diameter of a0<100> type loops was higher and larger than that of (a0/2)<111> type loops. There was a tendency for the number density of loops in Ni-doped 9Cr-2WVTa to be slightly higher than that in 9Cr-2WVTa. In addition, the mean size of loops in the steels tempered at 750°C was larger than for those tempered at 700°C, while there was not much difference of number density between them. In the steels doped with Ni, irradiation-produced precipitates, identified as M6C(η)-type carbide, were found in the matrices. In this experiment, the change in tensile properties and the δDBTT of the 9Cr-2VWTa-2Ni was greater than for 9Cr-2WVTa. The microstructural evidence suggests that these differences in mechanical behavior are related to the formation of radiation-induced precipitates and cavities in the Ni-doped alloy; further analysis of the data using barrier hardening models is in progress.

高純度鉄中の不純物銅原子の照射誘起析出

高純度鉄中の不純物銅原子の照射誘起析出

Radiation-induced precipitation of nickel ion in aqueous solution saturated with hydrogen gas

Abstract Ni 2+ ion is proved to precipitate in a solution saturated with H 2 by the aid of 60 Co γ -ray irradiation. The amount of Ni precipitated increases in proportion to the total dose, where the G (Ni) is accounted for as 0.29. A log–log plot has revealed that precipitation increases linearly with dose rate under less than 1 kGy/h, followed by a subsequent levelling off. In regard to the effect of initial [Ni 2+ ], the increase in [Ni 2+ ] has incurred a reduction of precipitation rate of Ni.

Reversal from Depletion to Enrichment of Solute Elements in Radiation-Induced Segregation at Grain Boundaries

Ni ++ ion irradiations of a Nb-stabilized steel of type 1.4981 were performed at 400, 500, and 600°. Dislocation loop evolution with increasing dose and radiation-induced segregation at grain boundaries were observed by transmission electron microscopy. At low doses the typical radiation-induced segregation behaviour at grain boundaries, as commonly observed in steels and Fe-Ni-Cr alloys, is confirmed. At high doses the segregation behaviour at grain boundaries was found to be completely inverted, i.e. elements which were radiation-enriched at the grain boundaries at low doses were now depleted and vice versa. This is attributed to the formation and growth of radiation-induced precipitates with increasing dose. The precipitates were only found after irradiation at 500 and 600°C and to doses higher than 5 dpa. They are rich in Ni, Si, and Nb. The effect of inverse radiation-induced segregation at grain boundaries could possibly be useful for solving the problems arising from radiation-assisted sensitization and stress corrosion cracking.

Microstructural examination of irradiated V–(4–5%)Cr–(4–5%)Ti

Microstructural examination results are reported for two heats of V–(4–5%)Cr–(4–5%)Ti irradiated in the EBR-II X530 experiment to 4.5 dpa at ∼400°C to provide an understanding of the microstructural evolution that may be associated with degradation of mechanical properties. Fine precipitates were observed in high density intermixed with small defect clusters for all conditions examined following the irradiation. The irradiation-induced precipitation does not appear to be affected by preirradiation heat treatment at 950–1125°C. There was no evidence for a significant density of large (diameter > 10 nm) dislocation loops or network dislocations. Analytical investigations successfully demonstrated that the precipitates were enriched in titanium, depleted in vanadium and contained no nitrogen. These results are discussed in terms of future alloy development options.

Swelling behavior of V–Fe binary and V–Fe–Ti ternary alloys

V–Fe binary alloys with different Fe concentrations, i.e., V–1, 3 and 5 at.% Fe, and V–5% Fe alloy added with 1, 3 and 5 at.% of Ti were irradiated in EBR-II at 380–615°C to about 11 dpa. TEM observation was performed after irradiation. A systematic increase in cavity size was observed with increasing iron concentration in the binary alloys, especially at 510°C and 615°C irradiation. On the other hand, the density of cavities decreased with increasing iron concentration and irradiation temperature. Maximum swelling in V–Fe system occurred between 500°C and 600°C and the amount of swelling was up to 30% at a damage level of 11 dpa. The alloy containing only 1% Fe already showed substantial swelling. The effect of titanium addition to the swelling was very remarkable. One atomic percent of titanium addition to V–5 at.% Fe significantly suppressed cavity formation, and 3 at.% of titanium addition entirely suppressed swelling. There seems to be a threshold titanium concentration for suppression of swelling in V–5 at.% Fe. Radiation-induced precipitation of titanium oxide may be one reason why titanium additions suppress the swelling in vanadium alloys. Homogeneous titanium oxide precipitates were not observed so that the titanium in solution is more likely to be playing an important role for suppression of swelling.

Neutron irradiation effects in fusion or spallation structural materials: Some recent insights related to neutron spectra

A review is presented of recent insights on the role of transmutation in the development of radiation-induced changes in dimension or radiation-induced changes in physical or mechanical properties. It is shown that, in some materials and some neutron spectra, transmutation can significantly affect or even dominate a given property change process. When the process under study is also sensitive to displacement rate, and especially if it involves radiation-induced segregation and precipitation, it becomes much more difficult to separate the transmutation and displacement rate dependencies. This complicates the application of data derived from “surrogate” spectra to predictions in other flux-spectra environments. It is also shown in this paper that one must be sensitive to the impact of previously-ignored “small” variations in neutron spectra within a given reactor. In some materials these small variations have major consequences.

Small angle neutron scattering investigations of the microstructure of VVER-440-type reactor pressure vessel steel after irradiation at 60°C

The formation of point defects and precipitates after neutron irradiation of VVER-440-type reactor pressure vessel steel was investigated by small angle neutron scattering experiments. Irradiation at 60°C increased the number of point defects, decreased the precipitates, which already exist in the unirradiated state, and formed a new type of fine-scaled precipitates. Post-irradiation annealing near the operational temperature of the nuclear power plants (270°C) provoked a slight decrease of the content of point defects and of the irradiation-induced precipitates. The content of these precipitates which are also present in the unirradiated state did not change by annealing.

Effects of heat treatments and neutron irradiation on microstructures and physical and mechanical properties of copper alloys

Tensile specimens of CuAl 2O 3, CuCrZr and CuNiBe alloys were given different heat treatments and then irradiated with fission neutrons at 250°C to a dose level of ≅ 0.3 dpa. Both unirradiated and irradiated specimens were tensile tested at 250°C. The microstructure and electrical resistivity were determined in the unirradiated as well as irradiated conditions. The post-deformation microstructure and fracture surfaces were also investigated. The main effect of the bonding thermal cycle heat treatment was a slight decreased in the strength of CuCrZr and CuNiBe alloys. The strength of CuAl-25, on the other hand, remained almost unaltered. The post-irradiation tests at 250°C showed a severe loss of ductility in the case of CuNiBe alloy. The irradiated CuAl-25 and CuCrZr specimens, on the other hand, exhibited a reasonable amount of uniform elongation. The results are briefly discussed in terms of thermal and irradiation stability of precipitates and particles and irradiation-induced segregation, precipitation and recovery of dislocation microstructure.

Effects of Neutron Irradiation on Tensile Properties in High-Purity Fe–Cr Alloys

The tensile properties of high- and low-purity Fe–9, –18 and –30Cr alloys irradiated by neutrons up to a dose of 5 × 1024 n/m2 (E >1 MeV) at 613, 673, or 763 K have been examined. The yield strength and the ultimate strength are increased and the elongation is decreased by irradiation. The enhancement of these strengths due to the irradiation has a tendency to increase with chromium and impurity content. Large stress drops are often observed, especially at 763 K, in stress–strain curves of high-purity and high-chromium-content alloys except for Fe–9Cr alloys. Irradiation-induced precipitates, with 2% larger interplanar spacings than the α′-phase, on dislocation loops are more easily formed in the specimens of higher chromium content and higher purity. The precipitates are formed even in the irradiated Fe–9Cr alloy of high purity. The stress drop behaviour during the tensile tests is predominant in the specimens of higher chromium content and higher purity.

Microstructures of deformed VTiCrSi type alloys after neutron irradiation

The alloy of V5Ti5Cr1SiAl,Y (nominal composition, weight percentage) was developed to improve oxidation properties and high temperature strength, and has been studied as one of the candidates for fusion applications. This alloy showed low swelling properties and enough tensile ductility after neutron irradiation to high fluence levels. The dislocation microstructures after tensile deformation and defect microstructures in the neutron-irradiated alloy to high fluences were studied. Irradiation was conducted in the Materials Open Test Assembly of the Fast Flux Test Facility (FFTF/MOTA-2A) at 406°C to 46 dpa and the deformation microstructures were examined by transmission electron microscopy. Slip dislocations were developed inhomogeneously in the specimen deformed at ambient temperature after neutron irradiation. Dislocation loops contributed mainly to hardening of the alloy after irradiation; however, cavities and radiation-induced precipitates did not so much.

Irradiation-induced precipitation in direct-aged alloy 625

Direct-aged alloy 625 (DA-A625) is a precipitation-hardened nickel-base superalloy with good mechanical properties and excellent corrosion resistance. Its superior stress corrosion cracking (SCC) performance makes it a candidate material for applications in light water reactors where resistance to irradiation-assisted SCC (IASCC) is important. This alloy derives its strength from the intragranular precipitation of fine DO22 – ordered γ” precipitates. These precipitates have a disc-like morphology and are crystallo-grahically related to the γ matrix such that the [001] axis of the precipitates are oriented parallel to <100> directions in the matrix. This alloy is generally solution-annealed then aged within the temperature range ~600 to 750°C. In direct-aging, the alloy is immediately aged at ~ 660°C for 80 h following hot-working. In addition to the γ° precipitates, intergranular M23C6, M7C3, M6C, or MC carbides can form in this material.

Evolution of long-range order and composition for radiation-induced precipitate dissolution.

Disordering and dissolution of L${1}_{2}$ ordered \ensuremath{\gamma}\ensuremath{'} precipitates under irradiation at temperatures between room temperature and 623 K are investigated by means of transmission electron microscopy and field-ion microscopy with atom probe. The combination of both experimental techniques allows us to follow the disordering process as well as chemical decomposition of the precipitates with atomic resolution. During room-temperature irradiation and for increasing irradiation fluence, the concentration profiles across the precipitates show a broadening of the \ensuremath{\gamma}/\ensuremath{\gamma}\ensuremath{'} interface. The experimentally obtained depth profiles can be interpreted assuming a dissolution process of the concentration inhomogeneities due to ballistic transport only. A correlation analysis of the experimental data yields a mixing diffusion coefficient of ${\mathit{D}}_{\mathrm{mix}}$/K=(0.${75}_{\mathrm{\ensuremath{-}}0.4}^{+0.2}$) ${\mathrm{nm}}^{2}$ ${\mathrm{dpa}}^{\mathrm{\ensuremath{-}}1}$. Depending on irradiation temperature, two dissolution regimes are observed. For a displacement rate of ${10}^{\mathrm{\ensuremath{-}}3}$ dpa ${\mathrm{s}}^{\mathrm{\ensuremath{-}}1}$, the precipitates first disorder and then dissolve in a disordered state at temperatures below 540 K, while disordering and dissolution occur simultaneously at temperatures between 540 and 623 K. These results demonstrate that disordering of the precipitates is not necessarily required for the dissolution. The results are in accordance with recent theoretical predictions for the dissolution mechanism of ordered precipitates under irradiation. \textcopyright{} 1996 The American Physical Society.

In Situ Ion Beam Research In Argonne's Intermediate Voltage Electron Microscope

AbstractSince Fall 1995, a state-of-the-art intermediate voltage electron microscope (IVEM) has been operational in the HVEM-Tandem Facility with in situ ion irradiation capabilities similar to those of the HVEM of the Facility. A 300 kV Hitachi H-9000NAR is interfaced to the two ion accelerators of the Facility, with a demonstrated point-topoint spatial resolution for imaging of 0.25 nm with the ion beamline attached to the microscope. The IVEM incorporates a Faraday cup system for ion dosimetry with measurement aperture 6.5 cm from the TEM specimen, which was described in Symposium A of the 1995 MRS Fall Meeting. The IVEM is now employed for a variety of in situ ion beam studies ranging from low dose ion damage experiments with GaAs, in which damage zones individual displacement cascades are observed, to implantation studies in metals, in which irradiation-induced noble gas precipitate mobility is studied in real time. In this presentation, the new instrumentation and its specifications will be described briefly, several basic concepts relating to in situ experiments in transmission electron microscopes will be summarized and examples of in situ experiments will be presented which exploit the experimental capabilities of this new user facility instrumentation.

Radiation-Induced α' Phase Formation on Dislocation Loops in Fe-Cr Alloys During Electron Irradiation

Radiation-induced precipitates on dislocation loops in low and high purity Fe-9, -18 and -50 % Cr alloys were examined under electron irradiation in a high voltage electron microscope operated at 1 MV. Two types of dislocation loops on {100} planes with a Burgers vectors and on {111} planes with a/2 are formed in high purity Fe-Cr alloys. However, only a type loops are formed in low purity alloys, i.e. where carbon concentration is greater than about 60 wt.ppm. The growth rate of the loops formed in Fe-9 and -18Cr alloys is faster than that of Fe-50Cr alloy. All of the loops formed at temperatures ranging from 573 K to 973 K, even in the (σ+ α + α') phase regions in thermal equilibrium, have moire fringe contrast due to precipitates identified as α' phase. The contrast is very weak in the Fe-50Cr alloy because of coarse distribution of precipitates, while the α' phase formed in low purity alloys has somewhat larger interplanar spacings that are modified by interstitial impurities. The α' phase formation on the loops can be understood by radiation-induced segregation of chromium and interstitial impurity atoms.