Dual beam irradiation of nanostructured FeCrAl oxide dispersion strengthened steel

Nanostructured ferritic oxide dispersion strengthened (ODS) alloy is an ideal candidate for fission/fusion power plant materials, particularly in the use of a first-wall and blanket structure of a next generation reactor. These steels usually contain a high density of Y-Ti-O and Y-Al-O nanoparticles, high dislocation densities and fine grains. The material contains nanoparticles with an average diameter of 21 nm and was treated by several cold rolling procedures, which modify the dislocation density. Structural analysis with HRTEM shows that the chemical composition of the initial Y₂O₃ oxide is modified to perovskite YAlO₃ (YAP) and Y₂Al₅O₁₂ garnet (YAG). Irradiation of these alloys was performed with a dual beam irradiation of 2.5 MeV Fe⁺/31 dpa and 350 keV He⁺/18 appm/dpa. Irradiation causes atomic displacements resulting in vacancy and self-interstitial lattice defects and dislocation loops. Extended SRIM calculations for ODS steel indicate a clear spatial separation between the excess vacancy distribution close to the surface and the excess interstitials in deeper layers of the material surface. The helium atoms are supposed to accumulate mainly in the vacancies. Additionally to structural changes, the effect of the irradiation generated defects on the mechanical properties of the ODS is investigated by nanoindentation. A clear hardness increase in the irradiated area is observed, which reaches a maximum at a close surface region. This feature is attributed to synergistic effects between the displacement damage and He implantation resulting in He filled vacancies. Fine He cavities with diameters of a few nanometers were identified in TEM images.     

双束强辐照对氧化物弥散强化铁素体型合金强化相稳定性影响

采用双束复合辐照装置,研究了He存在条件下,强辐照对长期时效后的ODS合金中强化相(Y2O3)的稳定性和辐照损伤特征的影响。实验结果表明:双束强辐照下,ODS合金中强化相不稳定,发生聚集长大并造成附近基体中Ti、Y浓度增高,导致空洞尺寸和空洞肿胀增加,并对这一结果从理论上进行解释。     

Radiation response of a 9 chromium oxide dispersion strengthened steel to heavy ion irradiation

Ferritic-martensitic (FM) alloys are expected to play an important role as cladding or structural components in Generation IV systems operating in the temperature range 350-700 °C and to doses up to 200 dpa. Oxide dispersion strengthened (ODS) ferritic-martensitic steels have been developed to operate at higher temperatures than traditional FM steels. These steels contain nanometer-sized Y-Ti-O nanoclusters as a strengthening mechanism. Heavy ion irradiation has been used to determine the nanocluster stability over a temperature range of 500-700 °C to doses of 150 dpa. At all temperatures, the average nanocluster size decreases but the nanocluster density increases. The increased density of smaller nanoclusters under radiation should lead to strengthening of the matrix. While a reduction in size under irradiation has been reported in some other studies, many report oxide stability. The data from this study are contrasted to the available literature to highlight the differences in the reported radiation response.     

Microstructure and mechanical properties of friction stir welded oxide dispersion strengthened alloy

Effect of friction stir welding (FSW) on microstructure and mechanical properties of oxide dispersion strengthened alloy MA956 was investigated. Fine grained microstructure was developed in the processed region with slight particle coarsening. Tensile behavior of the processed material was compared with that of the as-received material at room temperature. Results indicated that significant grain refinement during FSW compensated for the reduced particle strengthening contribution and enhanced tensile strength by 145 MPa without loss in ductility. Further analysis indicated a good agreement between experimentally measured yield strength and the strength calculated by Pythagorean superimposition of strengthening contribution in FSWed material.     

Microstructural investigation of the stability under irradiation of oxide dispersion strengthened ferritic steels

Some fuel pin cladding made from a ferritic steel reinforced by titanium and yttrium oxides were irradiated in the French experimental reactor Phénix. Microstructural examination of this alloy indicates that oxides undergo dissolution under irradiation. This irradiation shows the influence of dose and, in a smaller part, of temperature. In order to better understand the mechanisms of dissolution, three ferritic steels reinforced by Y₂O₃ or MgO were irradiated with different charged particles. Inelastic interactions induced by 1 MeV He ion irradiation do not lead to any modification, neither in their chemical composition, nor in their spatial and size distribution. In contrast, isolated Frenkel pairs created by electron irradiation lead to significant oxide dissolution with a radius decrease proportional to the dose. Moreover, the comparison between irradiation with ions (displacements cascades) and electrons (Frenkel pairs only) shows the importance of free point defects in the dissolution phenomena.     

Irradiation creep of oxide dispersion strengthened (ODS) steels for advanced nuclear applications

Ferritic oxide dispersion strengthened steels with different microstructure were in-beam creep tested in a temperature range from 300 to 500 °C. Irradiation was by He-ions. Elongation was determined as a function of stress and irradiation damage rate. Damage was investigated by transmission electron microscopy. A thorough analysis of the loops developing during irradiation creep did not show any dependence of orientation or size on the direction of the applied stress. At 400 °C radiation induced segregation was found (most probably an iron aluminide) which had no effect on irradiation creep. No pronounced influence of microstructure or dispersoid size on the irradiation creep behavior was detected. Irradiation creep compliance of PM2000 with dispersoids of about 30 nm diameter were found to differ little from material with dispersoids of only 2-3 nm diameter. This is in contrast to thermal creep where dislocation-obstacle interactions are extremely important. An assessment of the technical relevance of irradiation creep in advanced nuclear systems is presented.     

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.     

Creep behavior of oxide dispersion strengthened 8Cr-2WVTa and 8Cr-1W steels

Microstructures and creep behavior of two martensitic oxide dispersion strengthened (ODS) steels 8%Cr-2%W-0.2%V-0.1%Ta (J1) and 8%Cr-1%W (J2) with finely dispersed Y₂Ti₂O₇ have been investigated. Creep tests have been carried out at 670, 700 and 730 °C. Creep strength of J1 is stronger than that of any other ODS martensitic steels and the hoop strength of the ferritic ODS steel cladding. At the beginning of creep test, shrinkage was frequently observed for J1. This is one of the reasons for high creep strength of J1. The δ-ferrite, which is untransformed to austenite at hot isostatic press and hot rolling temperatures, was elongated along the rolling direction, and volume fraction of δ-ferrite in J1 is larger than J2. Although the elongated δ-ferrite affects the anisotropy of creep behavior, the extent of anisotropy in J1 is not so large as that of the ferritic ODS steel.     

Defect and void evolution in oxide dispersion strengthened ferritic steels under 3.2 MeV Fe ion irradiation with simultaneous helium injection

In an attempt to explore the potential of oxide dispersion strengthened (ODS) ferritic steels for fission and fusion structural materials applications, a set of ODS steels with varying oxide particle dispersion were irradiated at 650°C, using 3.2 MeV Fe⁺ and 330 keV He⁺ ions simultaneously. The void formation mechanisms in these ODS steels were studied by juxtaposing the response of a 9Cr–2WVTa ferritic/martensitic steel and solution annealed AISI 316LN austenitic stainless steel under the same irradiation conditions. The results showed that void formation was suppressed progressively by introducing and retaining a higher dislocation density and finer precipitate particles. Theoretical analyses suggest that the delayed onset of void formation in ODS steels stems from the enhanced point defect recombination in the high density dislocation microstructure, lower dislocation bias due to oxide particle pinning, and a very fine dispersion of helium bubbles caused by trapping helium atoms at the particle–matrix interfaces.     

Development of oxides dispersion strengthened steels for high temperature nuclear reactor applications

By introducing a dispersion of nanosized yttrium oxides particles into a steel matrix, the upper temperature limit in mechanical creep strength can be enhanced in temperature by 100 K at least. Production routes for the production of a new class of oxides dispersion strengthened (ODS) steels are investigated within this work. Preliminary results obtained when doping pure iron matrix phase with two types of yttrium oxides (Y₂O₃) nanoparticles (commercial as well as laboratory fabricated nanopowder) are presented. The twofold purpose of this work is firstly to obtain a comparative analysis between the commercial and the laboratory fabricated Y₂O₃ nanopowder used to produce the doped iron, and secondly to demonstrate the feasibility of new production route by observing the nanostructure of the first test batches with pure iron. Observations are carried out with transmission electron microscopy (TEM) to determine the size distribution of the particles in the powder, while glow discharge optical emission spectroscopy (GDOES) and high resolution-scanning electron microscopy (HR-SEM) are used to analyze the chemical composition and the homogeneity of the produced doped iron. It is demonstrated, that even with small size particles nanopowder fabricated in the laboratory, the distribution is fairly homogeneous compared to the one obtained with a relatively large particles commercial nanopowder, confirming the feasibility of the new production route.     

Simulation of irradiation hardening of Zircaloy within plate-type dispersion nuclear fuel elements

Within plate-type dispersion nuclear fuel elements, the metal matrix and cladding attacked continuously by fast neutrons undergo irradiation hardening, which might have remarkable effects upon the mechanical behaviors within fuel elements. In this paper, with the irradiation hardening effect of metal materials mainly considered together with irradiation growth effect of the cladding, the three-dimensional large-deformation constitutive relations for the metal matrix and cladding are developed. The method of virtual temperature increase in the previous studies is further developed to model the irradiation swelling of fuel particles; the method of anisotropic thermal expansion is introduced to model irradiation growth of the cladding; and a method of multi-step-temperature loading is proposed to simulate the coupling features of irradiation-induced swelling of the fuel particles together with irradiation growth of the cladding. Above all, based on the developed relationship between irradiation growth at certain burnup and the loaded virtual temperatures, with considering that certain burnup corresponds to certain fast neutron fluence, the time-dependent constitutive relation due to irradiation hardening effect is replaced by the virtual-temperature-dependent one which is introduced into the commercial software to simulate the irradiation hardening effects of the matrix and cladding. Numerical simulations of the irradiation-induced mechanical behaviors are implemented with the finite element method in consideration of the micro-structure of the fuel meat. The obtained results indicate that when the irradiation hardening effects are introduced into the constitutive relations of the metal matrix and cladding: (1) higher maximum Mises stresses for certain burnup at the matrix exist with the equivalent plastic strains remaining almost the same at lower burnups; (2) the maximum Mises stresses for certain burnup at the cladding are enhanced while the maximum equivalent plastic strains are reduced; and (3) the maximum first principal stresses for certain burnup at the matrix or the cladding are lower than the ones without the hardening effect, and the differences are found to increase with burnup; and the variation rules of the interfacial stresses are similar.     

Depinning transition of a dislocation line in ferritic oxide strengthened steels

The dynamics of an edge dislocation in a medium with random oxide dispersoid particles acting as pinning centres is analysed. The dislocation line undergoes a depinning transition, where the order parameter is the dislocation line velocity v, which increases from zero for driving external resolved shear stresses τ beyond to a threshold value τ_c, known as the critical resolved shear stress. The critical stress is obtained by means of statistical analysis of the motion of a single dislocation in its glide plane, using overdamped, discrete dislocation dynamics simulations.     

A review of refractory metal alloys and mechanically alloyed-oxide dispersion strengthened steels for space nuclear power systems

Mechanical and thermo-physical properties of refractory metal alloys and mechanically alloyed (MA)-oxide dispersion strengthened (ODS) steels are reviewed and their potential for use in space nuclear reactors is examined. Preferable refractory alloys for use in liquid metal and gas-cooled space reactors include Nb-1%Zr, PWC-11, Mo-TZM, Mo-xRe where x varies from 7% to 44.5%, T-111 and ASTAR-811C. These alloys are heavy, difficult to fabricate, and are not readily available. The advantages of the MA-ODS alloys are: (a) their strength at high temperatures (>1000 K), which decreases slower with temperature than those of niobium and molybdenum alloys; (b) relatively lightweight and less expensive; (c) low swelling and no embrittlement with exposure to high-energy neutrons (>0.1 MeV) up to 10²⁷ n/m²; and (d) high resistance to oxidation and nitration. The few data available on compatibility of MA-ODS alloys with alkali liquid metals up to 1100 K are encouraging, however, additional tests at typical temperatures (1000-1400 K) in space nuclear reactors are needed. The anisotropy of MA-ODS alloys when cold worked, and particularly rolled into tubes, should not hinder their use in space nuclear power systems, in which operation pressure is either near atmospheric or as high as 2 MPa, but joints weldability is an issue.     

Effects of stress on radiation hardening and microstructural evolution in A533B steel

Bent specimens of A533B steel (0.16 wt% Cu) were irradiated at 290 °C to 1 dpa with 6.4 MeV Fe³⁺ ions. Calculated tensile stresses at the irradiated surface were set to 0, 250, 500 and 750 MPa. The specimens were subjected to hardness measurements, transmission electron microscopy (TEM) observations and three-dimensional atom probe (3DAP) analysis. The radiation-induced hardening decreased with increasing stress to 500 MPa which was near the yield strength. TEM and 3DAP results showed that well-defined dislocation loops and solute clusters were formed. The diameter of dislocation loops increased and the number density decreased when the stress was applied, whereas the diameter and number density of solute clusters decreased. The hardening was mainly attributed to solute cluster formation. Application of tensile stress would control hardening by suppressing the solute cluster nucleation and growth.     

Clarification of strain limits considering the ratcheting fatigue strength of 316FR steel

The effect of ratcheting on fatigue strength was investigated in order to rationalize the strain limit as a design criterion of commercialized fast reactor systems. Ratcheting fatigue tests were conducted at 550 °C. Duration of the ratchet straining was set for a certain number of strain cycles taking the loading condition of fast reactors into account, and the number of cycles for strain accumulation was defined as the ratchet-expired cycle. Fatigue lives decrease as the accumulated strain by ratcheting increases. Mean stress increased during the ratcheting cycle and its maximum value depended on the accumulated strain and the ratchet-expired cycle. Fatigue life reduction was negligible when the maximum mean stress was less than 25 MPa, corresponding to an accumulated strain of 2.2%. Accumulated strain is limited to 2% in the present design guidelines and this strain limit is considered effective to avoid reducing fatigue life by ratcheting. Microcrack growth behaviors were also investigated in these tests in order to discuss the life reduction mechanisms in ratcheting conditions.     

Effects of dose rate change under irradiation on hardening and microstructural evolution in A533B steel

Iron-ion irradiations were carried out for 0.09wt%Cu A533B steel specimens at 290°C to investigate effects of dose rate change during irradiation; the irradiations consisted of the base irradiation (with an unchanged dose rate) and an additional one with changed dose rates from 1 to 50 times that of the base one. Nano-indentation hardness measurements showed that the increase in hardness was higher for lower dose rate of the base irradiation. A similar trend was identified during the additional irradiation. Transmission electron microscope (TEM) and three-dimensional atom probe (3DAP) analyses were carried out for the quantitative characterization of defect features. Mn/Ni/Si/Cu-enriched clusters and dislocation loops were observed in all specimens. The increase in hardness mainly depended on the formation of the solute atom clusters. The square root of the volume fraction of the solute atom clusters provided a good correlation with the increase in hardness. The effects of dose rate and dose rate change during irradiation were explained by the formation of solute atom clusters.     

An experimental investigation of cyclic hardening of 316 stainless steel and of 2024 aluminium alloy under multiaxial loadings

This paper is concerned with the experimental behavior of a 316 stainless steel and a 2024 aluminium alloy at room temperature and under complex nonproportional strainings in tension-torsion. The basic features of this behavior are underlined and their interactions emphasized. It is observed that the response of these materials under general loading paths is a balance between hardening and softening occuring respectively when the nonproportionality of the straining path is increased or decreased.