Radiation-induced Precipitation Research Articles

APT-studies of phase formation features in VVER-440 RPV weld and base metal in irradiation-annealing cycles

This paper presents a thorough APT study of VVER-440 RPVs WM and BM samples cut out from the inner RPV surface. The features of phase formation (number densities, sizes and composition) in VVER-440 RPV steels under primary irradiation, recovery annealing, re-irradiation, recovery re-annealing and subsequent accelerated re-irradiation are identified to justify its lifetime extension up to 60 years by recovery re-annealing.It is shown that primary and re-irradiation is accompanied by formation of radiation-induced Cu-P-based precipitates whereas recovery annealing leads to their partial dissolution and coarsening of the undissolved ones. At this, for both WM and BM under primary annealing, Cu dissolves in the matrix, while under re-annealing Cu almost doesn't return to the matrix. In WM P fully dissolves in the matrix under annealing while in BM there is gradual matrix P depletion under operation (including recovery annealing) due to formation of P grain boundary segregation. Under the third irradiation cycle, Cu almost doesn't contribute to precipitation: in WM radiation-induced precipitates are P-based, enriched with Si, Ni and Mn, while in BM there are Si-based precipitates, enriched with Ni and Mn. In BM number density of precipitates is lower than in WM during all operation stages that causes its lower radiation embrittlement.

Radiation-induced bcc-fcc phase transformation in a Fe[sbnd]3%Ni alloy

The issue of neutron irradiation embrittlement of Reactor Pressure Vessel steels must be considered for Nuclear Power Plant life extension. This phenomenon partly arises from the existing interactions between dislocations and nanometric clusters composed of Cu, P, Si, Mn and Ni. The latter alloying element, playing a key role in the evolution of solute enriched clusters under irradiation, is the focus of this publication. To assess the effect of Ni on microstructure evolution under irradiation, particle accelerator based experiments were conducted. An under-saturated Fe3at.%Ni alloy was irradiated with self-ions, at 673 K, up to ∼1.2 dpa. Then, the microstructural damage was characterized, at the atomic scale, by conventional Transmission Electron Microscopy, Scanning Transmission Electron Microscopy coupled to Energy Dispersive X-ray Spectroscopy and Electron Energy Loss Spectroscopy, while chemical features were investigated by Atom Probe Tomography. Informations obtained by combining these coupled techniques provide evidence for the formation of a FCC phase, containing 25 at.%Ni, which can be either the disordered γ phase or the ordered L12 type Fe3Ni phase. The metastable or stable state of this FCC phase is discussed in the light of what is known from the literature. It is the first time that this BCC-FCC phase transformation is observed in an under-saturated α-FeNi alloy and this likely occurred via a Radiation Induced Precipitation (RIP) mechanism. Ni atom segregation is observed on cavities, dislocation lines and dislocation loops. The latter constitute nuclei for precipitates, leading to the formation of an additional segregation site for Ni: the precipitate FCC - matrix BCC nearly coherent interface. Similar mechanisms are argued to be operating also in high Ni RPV steels under neutron irradiation.

Radiation tolerance of commercial and advanced alloys for core internals: a comprehensive microstructural characterization

Thirteen austenitic stainless steels, nickel-base alloys, and ferritic alloys were irradiated using 2 MeV protons at 360 °C to a damage level of 2.5 displacements per atom (dpa). Comprehensive microstructural characterization was performed for irradiation-induced features, including dislocation loops, voids, precipitates, and radiation induced segregation (RIS). Dislocation loops formed in all alloys except 14YWT, while voids were observed in alloys 316 L, 310, C22, and 14YWT. Irradiation-induced formation of γ′ precipitates was observed in alloys 316 L, 310, 800, and 690; the irradiation-enhanced, long-range ordered Ni2Cr phase (Pt2Mo-type) was observed in alloys 690, C22, 625, 625Plus, 625DA, and 725; and G-phase was observed in alloy T92. No irradiation-induced precipitates were observed in alloys X750, 718 or 14YWT. Precipitation of the γ′ phase can be understood through segregation and clustering of Si, Al, and Ti. Overall, austenitic stainless steels are generally susceptible to irradiation damage in the form of loops, voids, precipitates, and RIS. Ni-base alloys have this same type of dislocation loops and RIS behaviors but are more resistant to void swelling. Ferritic alloys showed better resistance to loop formation, void swelling and irradiation-induced precipitation. From the degree of irradiation-induced microstructural change, alloy T92 was identified as the most radiation resistant among these alloys.

Understanding irradiation-induced nanoprecipitation in zirconium alloys using parallel TEM and APT

We investigate nano-scale irradiation-induced precipitation in a ZrSnFeCrNi-alloy (Zircaloy-2) by combining atom probe tomography (APT) for chemical detail with scanning transmission electron microscopy (STEM) and high resolution energy dispersive X-ray (EDX) spectroscopy for wider context and complimentary and correlative TEM diffraction techniques for crystallographic relationships. We find that Fe and Cr-rich nano-rods precipitate in Zircaloy-2 following proton irradiation at 350 °C to a low dose of ∼2 dpa. The long-axis of the nano-rods are aligned in a direction 12–15° from the Zr matrix 〈0001〉, align in the basal plane and are of width 1.5–5 nm. Smaller rods are of APT-determined composition Zr4(Fe0.67Cr0.33), tending towards Zr3(Fe0.69Cr0.31) as the rod volume increases to > ∼400 nm3, in agreement with STEM-EDX determination of composition resembling that of Zr3Fe with Cr replacing some of the Fe. The Fe/Cr ratio has been shown to increase with distance from the nearest partially-dissolved Zr(Fe,Cr)2 phase particle. The nucleation of nano rods has implications for macroscopic irradiation-induced deformation phenomena, irradiation-induced hardening and the evolution of dislocation loops and other defects.

The Effect of Reducing Irradiation Temperature on the Structure and Radiation Embrittlement Mechanism of RPV Steels

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Study of Current State of Templates Cut Out of VVER-440 Reactor Pressure Vessel

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First-principles study of solvent-solute mixed dumbbells in body-centered-cubic tungsten crystals

Tungsten (W) is considered as a promising candidate for plasma-facing materials for future nuclear fusion devices, and selecting optimal alloying constituents is a critical issue to improve radiation resistance of the W alloys as well as to improve their mechanical properties. We conducted in the current study a series of first-principles calculations for investigating solvent-solute mixed dumbbells in W crystals. The results suggested that titanium (Ti), vanadium (V), and chromium (Cr) are favorable as solutes for W alloys from irradiation-effect perspectives because these elements are expected to promote vacancy-interstitial recombination without causing radiation-induced precipitation that reduces ductility of irradiated materials.

Atomistic study of the hardening of ferritic iron by Ni-Cr decorated dislocation loops

The exact nature of the radiation defects causing hardening in reactor structural steels consists of several components that are not yet clearly determined. While generally, the hardening is attributed to dislocation loops, voids and secondary phases (radiation-induced precipitates), recent advanced experimental and computational studies point to the importance of solute-rich clusters (SRCs). Depending on the exact composition of the steel, SRCs may contain Mn, Ni and Cu (e.g. in reactor pressure vessel steels) or Ni, Cr, Si, Mn (e.g. in high-chromium steels for generation IV and fusion applications). One of the hypotheses currently implied to explain their formation is the process of radiation-induced diffusion and segregation of these elements to small dislocation loops (heterogeneous nucleation), so that the distinction between SRCs and loops becomes somewhat blurred. In this work, we perform an atomistic study to investigate the enrichment of loops by Ni and Cr solutes and their interaction with an edge dislocation. The dislocation loops decorated with Ni and Cr solutes are obtained by Monte Carlo simulations, while the effect of solute segregation on the loop's strength and interaction mechanism is then addressed by large scale molecular dynamics simulations. The synergy of the Cr-Ni interaction and their competition to occupy positions in the dislocation loop core are specifically clarified.

Irradiation evolution of Cu precipitates in Fe1.0Cu alloy studied by positron annihilation spectroscopy

The Fe1.0Cu alloy was irradiated by 2 MeV Fe ions to a peak dose of 0.06dpa, 0.45dpa and 1.35dpa, respectively. The irradiation evolution of Cu precipitates in Fe1.0Cu alloy was studied at various depths by positron annihilation spectroscopy. The values of the S parameters directly indicate the formation of vacancy clusters. The S-W relationship showed that CunVm complexes (n > m) dominated the microstructure in the track region. Vacancy clusters and CunVm complexes (n < m) were the main defect types in the damage region. There finding were supported by the results of the performed theoretical calculations. The coincidence Doppler broadening showed that the Cu precipitates formed easily at a low irradiation dose, and that the irradiation-induced Cu precipitation was enhanced by the increase of the irradiation dose in the track region. We showed that the Cu precipitates finally reach a steady-state morphology under multiple active processes.

Fe13+ ion irradiation-induced M2X precipitate in P92 steel at 700°C up to 1.62 dpa

Microstructure of P92 ferritic/martensitic steel irradiated under 3.5 MeV Fe13+ ion beams with a damage rate of ∼1 × 10−4 at 700 °C to 1.62 displacements per atom (dpa) was investigated using transmission electron microscopy (TEM). M23C6 and MX precipitates were the dominant phases in the unirradiated samples. Irradiation induced high density Cr-rich Cr2(C,N) precipitates were identified in irradiated samples using TEM. The nucleation of the Cr-rich Cr 2(C,N) precipitates in the irradiated P92 steel at elevated temperature were discussed in detailed.

Mechanisms of radiation embrittlement of VVER-1000 RPV steel at irradiation temperatures of (50–400)°C

This work summarizes and analyzes our recent research results on the effect of irradiation temperature within the range of (50–400)°C on microstructure and properties of 15Kh2NMFAA class 1 steel (VVER-1000 reactor pressure vessel (RPV) base metal). The paper considers the influence of accelerated irradiation with different temperature up to different fluences on the carbide and irradiation-induced phases, radiation defects, yield strength changes and critical brittleness temperature shift (ΔTK) as well as on changes of the fraction of brittle intergranular fracture and segregation processes in the steel. Low temperature irradiation resulted solely in formation of radiation defects – dislocation loops of high number density, the latter increased with increase in irradiation temperature while their size decreased. In this regard high embrittlement rate observed at low temperature irradiation is only due to the hardening mechanism of radiation embrittlement. Accelerated irradiation at VVER-1000 RPV operating temperature (∼300 °C) caused formation of radiation-induced precipitates and dislocation loops, as well as some increase in phosphorus grain boundary segregation. The observed ΔTK shift being within the regulatory curve for VVER-1000 RPV base metal is due to both hardening and non-hardening mechanisms of radiation embrittlement. Irradiation at elevated temperature caused more intense phosphorus grain boundary segregation, but no formation of radiation-induced precipitates or dislocation loops in contrast to irradiation at 300 °C. Carbide transformations observed only after irradiation at 400 °C caused increase in yield strength and, along with a contribution of the non-hardening mechanism, resulted in the lowest ΔTK shift in the studied range of irradiation temperature and fluence.

Specific Features of Structural-Phase State and Properties of Reactor Pressure Vessel Steel at Elevated Irradiation Temperature

This paper considers influence of elevated irradiation temperature on structure and properties of 15Kh2NMFAA reactor pressure vessel (RPV) steel. The steel is investigated after accelerated irradiation at 300°C (operating temperature of VVER-1000-type RPV) and 400°C supposed to be the operating temperature of advanced RPVs. Irradiation at 300°C leads to formation of radiation-induced precipitates and radiation defects-dislocation loops, while no carbide phase transformation is observed. Irradiation at a higher temperature (400°C) neither causes formation of radiation-induced precipitates nor provides formation of dislocation loops, but it does increase the number density of the main initial hardening phase—of the carbonitrides. Increase of phosphorus concentration in grain boundaries is more pronounced for irradiation at 400°C as compared to irradiation at 300°C due to influence of thermally enhanced diffusion at a higher temperature. The structural-phase changes determine the changes of mechanical properties: at both irradiation temperatures irradiation embrittlement is mainly due to the hardening mechanism with some contribution of the nonhardening one for irradiation at 400°C. Lack of formation of radiation-induced precipitates at T = 400°C provides a small ΔTK shift (17°C). The obtained results demonstrate that the investigated 15Kh2NMFAA steel may be a promising material for advanced reactors with an elevated operating temperature.

Radiation response of alloy T91 at damage levels up to 1000 peak dpa

Ferritic/martensitic alloys are required for advanced reactor components to survive 500–600 neutron-induced dpa. Ion-induced void swelling of ferritic/martensitic alloy T91 in the quenched and tempered condition has been studied using a defocused, non-rastered 3.5 MeV Fe-ion beam at 475 °C to produce damage levels up to 1000 peak displacements per atom (dpa). The high peak damage level of 1000 dpa is required to reach 500–600 dpa level due to injected interstitial suppression of void nucleation in the peak dpa region, requiring data extraction closer to the surface at lower dpa levels. At a relatively low peak damage level of 250 dpa, voids began to develop, appearing first in the near-surface region. With increasing ion fluence, swelling was observed deeper in the specimen, but remained completely suppressed in the back half of the ion range, even at 1000 peak dpa. The local differences in dpa rate in the front half of the ion range induce an “internal temperature shift” that strongly influences the onset of swelling, with shorter transient regimes resulting from lower dpa rates, in agreement not only with observations in neutron irradiation studies but also in various ion irradiations. Swelling was accompanied by radiation-induced precipitation of Cu-rich and Si, Ni, Mn-rich phases were observed by atom probe tomography, indicating concurrent microchemical evolution was in progress. In comparison to other ferritic/martensitic alloys during ion irradiation, T91 exhibits good swelling resistance with a swelling incubation period of about 400 local dpa.

Migration of rhenium and osmium interstitials in tungsten

Tungsten is expected to be a promising plasma-facing material for future fusion devices, but radiation-induced precipitation (RIP), which leads the material to hardening, is a concern at their practical use. One of the keys to accurate prediction of the emergence of RIP is migration of solute atoms, rhenium and osmium, that are produced by nuclear transmutation through irradiation. We conduct a series of numerical simulations using an atomic kinetic Monte Carlo method and investigate the migration of these solute atoms in the form of tungsten–rhenium and tungsten–osmium mixed dumbbells, considered to be the most efficient ”carriers” of the solute atoms. We find that the low rotation energy barrier of these mixed dumbbells leading to three-dimensional migration greatly influences their diffusivities. The result also suggests that, although these dumbbells have three-dimensional motion, one cannot simply reduce their migration behavior to that of vacancy-like spherical objects.

A phase field model for segregation and precipitation induced by irradiation in alloys

A phase field model is introduced to model the evolution of multicomponent alloys under irradiation, including radiation-induced segregation and precipitation. The thermodynamic and kinetic components of this model are derived using a mean-field model. The mobility coefficient and the contribution of chemical heterogeneity to free energy are rescaled by the cell size used in the phase field model, yielding microstructural evolutions that are independent of the cell size. A new treatment is proposed for point defect clusters, using a mixed discrete-continuous approach to capture the stochastic character of defect cluster production in displacement cascades, while retaining the efficient modeling of the fate of these clusters using diffusion equations. The model is tested on unary and binary alloy systems using two-dimensional simulations. In a unary system, the evolution of point defects under irradiation is studied in the presence of defect clusters, either pre-existing ones or those created by irradiation, and compared with rate theory calculations. Binary alloys with zero and positive heats of mixing are then studied to investigate the effect of point defect clustering on radiation-induced segregation and precipitation in undersaturated solid solutions. Lastly, irradiation conditions and alloy parameters leading to irradiation-induced homogeneous precipitation are investigated. The results are discussed in the context of experimental results reported for Ni–Si and Al–Zn undersaturated solid solutions subjected to irradiation.

Deuterium ion irradiation induced precipitation in Fe–Cr alloy: Characterization and effects on irradiation behavior

A new phase was found to precipitate in a Fe–Cr model alloy after 58keV deuterium ion irradiation at 773K. The nanoscale radiation-induced precipitate was studied systematically using high resolution transmission electron microscopy (HRTEM), image simulation and in-situ ultrahigh voltage transmission electron microscopy (HVEM). B2 structure is proposed for the new Cr-rich phase, which adopts a cube-on-cube orientation relationship with regard to the Fe matrix. Geometric phase analysis (GPA) was employed to measure the strain fields around the precipitate and this was used to explain its characteristic 1-dimensional elongation along the 〈100〉 Fe direction. The precipitate was stable under subsequent electron irradiation at different temperatures. We suggest that the precipitate with a high interface-to-volume ratio enhances the radiation resistance of the material. The reason for this is the presence of a large number of interfaces between the precipitate and the matrix, which may greatly reduce the concentration of point defects around the dislocation loops. This leads to a significant decrease in the growth rate.

84 MeV C-ions irradiation effects on Zr–45Ti–5Al–3V alloy

Newly developed Zr–45Ti–5Al–3V alloy were irradiated by 84MeV carbon ions with doses of 4*1015ions/cm2 and 12*1015ions/cm2, respectively. XRD, SEM, TEM, SAD and tensile tests were performed to study the microstructural evolution and mechanical properties modification upon high energy carbon ion irradiation. XRD patterns show no phase change while the diffraction peak position and intensity vary with irradiation doses. Tensile tests verify monotonic change of alloy strengths and elongations upon irradiation. Microstructure observations of the irradiated samples reveal the irradiation-induced precipitation of (Zr,Ti)3C2, which was believed contributing to the alloy hardening. Superlattice was discovered by the SAD patterns of original and irradiated samples and the high energy C-ions implantation was demonstrated to promote the disorder–order transition by introducing lattice defects.

Neutron irradiation effects on tungsten materials

To understand the irradiation response of tungsten in the ITER and a DEMO-like reactor, irradiation effects on microstructure development, hardening and electrical resistivity of pure W and W-Re-Os alloys are studied using fission reactor irradiation up to about 1–1.5dpa in the temperature range from 400 to 800°C. Microstructural development and hardening behavior are summarized here based on the irradiation data. Voids are the major damage structure in pure W, but Re addition clearly suppressed void formation. Irradiation hardening was also suppressed in lower Re content alloys. The hardening was caused by irradiation-induced precipitation of WRe (σ-phase) and WRe3 (χ-phase), and the hardening behavior strongly depended on the neutron spectrum of the irradiation field. On the basis of these results, a damage structure development prediction map in fusion devices is suggested.

Precipitate behavior in self-ion irradiated stainless steels at high doses

To study radiation-induced precipitation at high doses, solution annealed 304L SS and cold worked 316 SS were irradiated to 46 and 260dpa at 380°C using 5MeV Fe++ and the radiation-induced precipitates were examined using atom probe tomography. Ni/Si-rich clusters were observed in all examined conditions. G-phase precipitates were observed in 316 SS at 46dpa but only appeared in 304L SS at 260dpa. Using the neutron irradiation to 46dpa at 320°C as a reference, the temperature shift for cold worked 316 SS appeared to be smaller than that of solution annealed 304L SS, probably due to the high density of dislocations, which served as defect sinks and mitigated the effect of high dose rate.

γ-radiation effect on dicyclohexano-18-crown-6 aqueous solution in the presence of metal cations

Abstract γ-radiation effect on dicyclohexano-18-crown-6 (DCH18C6) aqueous solution in the presence of metal cations, such as K+ and Sr2+, was investigated, and some pale yellow precipitates were obtained at doses above 100 kGy. It was found that the complexation interaction of metal cations could accelerate the conversion of DCH18C6 into oligomers and the formation of precipitates, but it did not affect the formation mechanism of precipitates. Furthermore, the salting-out effect of metal cations played an important role in the precipitation. In order to inhibit the radiation-induced precipitation of DCH18C6 in water phase, nitric acid with a concentration of more than 1 M could be added to the system. The methods and results for analyzing radiation-induced products reported in this work are useful to understand the radiation chemistry behavior of DCH18C6 which is used as the extractant in the separation of long half-life radionuclides from spent nuclear fuel.