Deformation mode maps for tensile deformation of neutron-irradiated structural alloys

The deformation microstructures of neutron-irradiated nuclear structural alloys, A533B steel, 316 stainless steel, and Zircaloy-4, have been investigated by tensile testing and transmission electron microscopy to map the extent of strain localization processes in plastic deformation. Miniature specimens with a thickness of 0.25 mm were irradiated to five levels of neutron dose in the range 0.0001-0.9 displacements per atom (dpa) at 65-100 °C and deformed at room temperature at a nominal strain rate of 10⁻³ s⁻¹. Four modes of deformation were identified, namely three-dimensional dislocation cell formation, planar dislocation activity, fine scale twinning, and dislocation channel deformation (DCD) in which the radiation damage structure has been swept away. The modes varied with material, dose, and strain level. These observations are used to construct the first strain-neutron fluence-deformation mode maps for the test materials. Overall, irradiation encourages planar deformation which is seen as a precursor to DCD and which contributes to changes in the tensile curve, particularly reduced work hardening and diminished uniform ductility. The fluence dependence of the increase in yield stress, ΔYS = α(?t)ⁿ had an exponent of 0.4-0.5 for fluences up to about 3 × 10²² n m⁻² (∼0.05 dpa) and 0.08-0.15 for higher fluences, consistent with estimated saturation in radiation damage microstructure but also concurrent with the acceleration of gross strain localization associated with DCD.     

Analysis of tensile deformation and failure in austenitic stainless steels: Part I - Temperature dependence

This paper describes the temperature dependence of deformation and failure behaviors in the austenitic stainless steels (annealed 304, 316, 316LN, and 20% cold-worked 316LN) in terms of equivalent true stress-true strain curves. The true stress-true strain curves up to the final fracture were calculated from tensile test data obtained at −150 to 450 °C using an iterative finite element method. Analysis was largely focused on the necking and fracture: key parameters such as the strain hardening rate, equivalent fracture stress, fracture strain, and tensile fracture energy were evaluated, and their temperature dependencies were investigated. It was shown that a significantly high strain hardening rate was retained during unstable deformation although overall strain hardening rate beyond the onset of necking was lower than that of the uniform deformation. The fracture stress and energy decreased with temperature up to 200 °C and were nearly saturated as the temperature came close to the maximum test temperature 450 °C. The fracture strain had a maximum at −50 to 20 °C before decreasing with temperature. It was explained that these temperature dependencies of fracture properties were associated with a change in the dominant strain hardening mechanism with test temperature. Also, it was seen that the pre-straining of material has little effect on the strain hardening rate during necking deformation and on fracture properties.     

Determination of fatigue curve parameters at cyclic strain limited loading according to the mechanical characteristics of power energy structural materials

Over 280 structural materials used in the Russian nuclear power energy industry were tested at Kaunas University of Technology, as commissioned by St. Petersburg Central Research Institute of Structural Materials in the period of 1970-2000. Alloyed structural steels, stainless steels and metals of their welded joints with different types of thermal treatment were under research in the conditions of symmetric low cycle tension-compression (N ≤ (1-2) × 10⁴) at room and elevated (200-350 °C) temperatures. During these experiments the characteristics of monotonic tension, cyclic stress-strain curves and low cycle fatigue curves, which are expressed by linear dependence of the total strain ? on the number of cycles N in co-ordinates log ? − log N, parameters C_1exp and m_1exp were determined. The range of cycle strain ? when this low cycle fatigue curve may be used is determined. The definition of the approached values of low cycle fatigue curves parameters C_1cal and m_1cal at room temperature by the mechanical characteristics of alloyed structural steels and their weld metals, stainless steels and their weld metals is analyzed in this work. The new relationships ln C_1exp − m_1exp among low cycle fatigue curves parameters, obtained by analyzing experimental data of 22-48 materials in each group, are determined. These relationships were used for determination of dependencies of low cycle parameters C_1cal and m_1cal by the mechanical characteristics of analyzed groups of materials. From theoretically and experimentally determined equations by using calculated values C_1cal and m_1cal dependencies for revision of Coffin-Manson fatigue curves parameters C and m were proposed. These parameters for the cyclically hardening or softening materials are calculated by estimating the equivalent value of the cycle plastic strain range. The proposed analytical dependencies make it possible to more exactly calculate the number of cycles N ≤ N₀ (N₀ = 10⁶) before the fatigue crack initiation for the structural materials of analyzed groups under cyclic strain limited loading by using mechanical characteristics of materials.     

Influence of post irradiation annealing on the mechanical properties and defect structure of AISI 304 steel

The effect of post irradiation annealing on the mechanical properties and the radiation induced defect structure was investigated on stainless steel, of type AISI 304, that was irradiated up to 24 dpa in the decommissioned Chooz A reactor. The material was investigated both in the as-irradiated state as well as after post irradiation annealing. In the as-irradiated specimen the typical radiation induced defects were found as well as γ′-precipitates (Ni₃Si). Annealing at 400 °C had almost no effect on the radiation induced defects, but annealing at 500 °C resulted in the immediate unfaulting of the Frank loops. As to the mechanical properties, annealing at 400 °C did not strongly affect the yield strength and the ductility of the material, although the fraction of intergranular fracture during slow strain rate tensile tests under pressurised water reactor conditions, was significantly reduced. Annealing at 500 °C did reduce the yield strength and restored substantially the ductility and the strain hardening capability of the material. The microstructure investigated by transmission electron microscopy correlates to the mechanical test results. It was found that the observed defect changes after post irradiation annealing provide a reasonable explanation for the observed changes of the mechanical properties.     

Hoop strength and ductility evaluation of irradiated fuel cladding

The objective of this study is to evaluate the hoop-directional mechanical properties comprising strength such as yield strength and ultimate tensile strength as well as mechanical ductility such as uniform elongation and total elongation. Therefore, in this paper, the ring tensile tests were performed in order to evaluate the mechanical properties of high burn-up fuel cladding under a hoop loading condition in a hot cell. The tests were performed with Zircaloy-4 nuclear fuel cladding whose burn-up is approximately 65,000 MWd/tU in the temperature range of room temperature to 800 °C. All the experiments were carried out at a constant strain rate of 0.01/s.On the basis of the ring tensile tests for a high burn-up Zircalay-4 cladding, the following conclusions were drawn. Firstly, the mechanical properties are abruptly degraded beyond 600 °C, which corresponds to a design-basis accident condition such as a RIA. Secondly, the un-irradiated fuel cladding showed ductile fracture behaviors such as 45° shear type fracture, cup and cone type fracture, cup and cup type fracture and chisel edge type fracture. While the high burn-up Zircalay-4 cladding showed a brittle fracture behavior even at the high temperatures (e.g. over 600 °C) which are achievable during a RIA. Thirdly, in the case of the high burn-up Zircalay-4 cladding, the strength, ductility and the energy to break are strongly dependent on the material property itself which are degraded by oxidation and hydriding during an operation rather than the temperature. Fourthly, hydride rim formation in the vicinity of metal-oxide interface can play an important role in the degradation of the mechanical properties for high burn-up fuel cladding.     

Tensile fracture characteristics of nanostructured ferritic alloy 14YWT

High temperature tensile fracture behavior has been characterized for the nanostructured ferritic alloy 14YWT (SM10 heat). Uniaxial tensile tests were performed at temperatures ranging from room temperature to 1000 °C in vacuum at a nominal strain rate of 10⁻³ s⁻¹. Comparing with the existing oxide dispersion strengthened (ODS) steels such as Eurofer 97 and PM2000, the nanostructured alloy showed much higher yield and tensile strength, but with lower elongation. Microstructural characterization for the tested specimens was focused on the details of fracture morphology and mechanism to provide a feedback for process improvement. Below 600 °C, the fracture surfaces exhibited a quasi-brittle behavior presented by a mixture of dimples and cleavage facets. At or above 600 °C, however, the fracture surfaces were fully covered with fine dimples. Above 700 °C dimple formation occurred by sliding and decohesion of grain boundaries. It was notable that numerous microcracks were observed on the side surface of broken specimens. Formation of these microcracks is believed to be the main origin of the poor ductility of 14YWT alloy. It is suggested that a grain boundary strengthening measure is essential to improve the fracture property of the alloy.     

Failure of Zircaloy cladding under transverse plane-strain deformation

Experiments have been performed to examine the ductility of Zircaloy 4 cladding tubes under conditions of near plane-strain deformation in the hoop direction (transverse to the tube axis) at temperatures of 25 and 300°C and at strain rates of 10⁻³ and 10² s⁻¹. To conduct these experiments, a specimen configuration was designed in which near plane-strain deformation is achieved, and a test methodology was established to determine two failure conditions: the limit strain at the onset of localized necking and the fracture strain. Experiments performed on cold-worked stress relieved material using the transverse plane-strain specimen geometry indicate major differences in failure behavior from that observed in uniaxial tension, although both test conditions result in failure by a localized necking process. The experimental results also indicate that while plane-strain fracture strains increase with temperature between 25 and 300°C, at a given temperature they are insensitive to strain rate. The limit strains at localized necking also increase with temperature but only at the high 10² s⁻¹ strain rate. Finally, the failure data indicate a strong sensitivity to surface flaws, as predicted by localized necking theory.     

Effect of corrosion potential on the corrosion fatigue crack growth behaviour of low-alloy steels in high-temperature water

The low-frequency corrosion fatigue (CF) crack growth behaviour of different low-alloy reactor pressure vessel steels was characterized under simulated boiling water reactor conditions by cyclic fatigue tests with pre-cracked fracture mechanics specimens. The experiments were performed in the temperature range of 240-288 °C with different loading parameters at different electrochemical corrosion potentials (ECPs). Modern high-temperature water loops, on-line crack growth monitoring (DCPD) and fractographical analysis by SEM were used to quantify the cracking response. In this paper the effect of ECP on the CF crack growth behaviour is discussed and compared with the crack growth model of General Electric (GE). The ECP mainly affected the transition from fast (‘high-sulphur’) to slow (‘low-sulphur’) CF crack growth, which appeared as critical frequencies ν_crit = f(ΔK, R, ECP) and ΔK-thresholds ΔK_EAC = f(ν, R, ECP) in the cycle-based form and as a critical air fatigue crack growth rate da/dt_Air,crit in the time-domain form. The critical crack growth rates, frequencies, and ΔK_EAC-thresholds were shifted to lower values with increasing ECP. The CF crack growth rates of all materials were conservatively covered by the ‘high-sulphur’ CF line of the GE-model for all investigated temperatures and frequencies. Under most system conditions, the model seems to reasonably well predict the experimentally observed parameter trends. Only under highly oxidizing conditions (ECP ? 0 mV_SHE) and slow strain rates/low loading frequencies the GE-model does not conservatively cover the experimentally gathered crack growth rate data. Based on the GE-model and the observed cracking behaviour a simple time-domain superposition-model could be used to develop improved reference CF crack growth curves for codes.     

Elastic and plastic properties of βZr at room temperature

Room temperature elastic and plastic properties of a single phase β_Zr have been studied by in-situ neutron diffraction compression testing. The measured macroscopic Young’s modulus is ∼60 GPa and the yield strength is ∼500 MPa. Dislocation slip is the major mode of plastic deformation. An Elasto-Plastic Self-Consistent (EPSC) model was used to interpret the experimental results and was shown to be effective in extracting the single crystal properties from the polycrystalline data. The single crystal elastic constants of the β-phase are determined as: C₁₁ = 145.9 ± 2.6 GPa, C₁₂ = 117.4 ± 2.5 GPa and C₄₄ = 29.8 ± 0.2 GPa. The calculated elastic modulus of 〈1 0 0〉, 〈1 1 0〉, 〈1 1 1〉, 〈2 1 1〉 and 〈3 1 0〉 directions was ∼41.2, 66.2, 82.9, 66.2 and 47.7 GPa, respectively. Pencil glide on the {110}, {112} and {123} planes was used in the EPSC model and gave a good simulation to the early part of the plastic deformation. The average β-phase strain is best represented by the peak average method, while in cases where only a limited number of diffraction peaks are available, the {211} grain family is a good candidate for estimation of the average β-phase strain.     

Analysis of tensile deformation and failure in austenitic stainless steels: Part II - Irradiation dose dependence

Irradiation effects on the stable and unstable deformation and fracture behavior of austenitic stainless steels (SSs) have been studied in detail based on the equivalent true stress versus true strain curves. An iterative finite element simulation technique was used to obtain the equivalent true stress-true strain data from experimental tensile curves. The simulation result showed that the austenitic stainless steels retained high strain hardening rate during unstable deformation even after significant irradiation. The strain hardening rate was independent of irradiation dose up to the initiation of a localized necking. Similarly, the equivalent fracture stress was nearly independent of dose before the damage (embrittlement) mechanism changed. The fracture strain and tensile fracture energy decreased with dose mostly in the low dose range <∼2 dpa and reached nearly saturation values at higher doses. It was also found that the fracture properties for EC316LN SS were less sensitive to irradiation than those for 316 SS, although their uniform tensile properties showed almost the same dose dependencies. It was confirmed that the dose dependence of tensile fracture properties evaluated by the linear approximation model for nominal stress was accurate enough for practical use without elaborate calculations.     

Mechanical properties of irradiated ODS-EUROFER and nanocluster strengthened 14YWT

Irradiations to 1.5 dpa at 300-750 °C were conducted to investigate the changes in mechanical properties of an advanced nanocluster strengthened ferritic alloy, designated 14YWT, and an oxide dispersion strengthened ferritic alloy ODS-EUROFER. Two non-dispersion strengthened variants, 14WT and EUROFER 97, were also irradiated and tested. Tensile results show 14YWT has very high tensile strengths and experienced some radiation-induced hardening, with an increase in room temperature yield strength of 125 MPa after irradiation, while results for ODS-EUROFER show a 275 MPa increase following irradiation. Master curve fracture toughness analysis show 14YWT has a cryogenic T_o reference temperatures before and after irradiation of about −188 and −176 °C, respectively, and upper-shelf K_JIc values between 175 and 225 MPa√m. The favorable fracture toughness properties and resistance to radiation-induced changes in mechanical properties observed for 14YWT are attributed to a fine grain structure and high number density of Y-Ti-O nanoclusters.     

Tensile and fracture toughness properties of neutron-irradiated CuCrZr

Tensile and fracture toughness properties of a precipitation-hardened CuCrZr alloy were investigated in two heat treatment conditions: solutionized, water quenched and aged (CuCrZr SAA), and hot isostatic pressed, solutionized, slow-cooled and aged (CuCrZr SCA). The second heat treatment simulated the manufacturing cycle for large components, and is directly relevant for the ITER divertor components. Specimens were neutron irradiated at ∼80 °C to two fluences, 2 × 10²⁴ and 2 × 10²⁵ n/m² (E > 0.1 MeV), corresponding to displacement doses of 0.15 and 1.5 displacements per atom (dpa). Tensile and fracture toughness tests were carried out at room temperature. Significant irradiation hardening and plastic instability at yield occurred in both heat treatment conditions with a saturation dose of ∼0.1 dpa. Neutron irradiation slightly reduced fracture toughness in CuCrZr SAA and CuCrZr SCA. The fracture toughness of CuCrZr remained high up to 1.5 dpa (J_Q > 200 kJ/m²) for both heat treatment conditions.     

Proton irradiation effect on microstructure, strain localization and iodine-induced stress corrosion cracking in Zircaloy-4

The radiation-induced microstructure, strain localization, and iodine-induced stress corrosion cracking (I-SCC) behaviour of recrystallized Zircaloy-4 proton-irradiated to 2 dpa at 305 °C was examined. <a> type dislocation loops having 1/3〈1 1  0〉 Burgers vector and a mean diameter and density of, respectively, 10 nm and 17 × 10²¹ m⁻³ were observed while no Zr(Fe,Cr)₂ precipitates amorphization or Fe redistribution were detected after irradiation. After transverse tensile testing to 0.5% macroscopic plastic strain at room temperature, almost exclusively basal channels were imaged. Statistical Schmid factor analysis shows that irradiation leads to a change in slip system activation from prismatic to basal due to a higher increase of critical resolved shear stresses for prismatic slip systems than for basal slip system. Finite element calculations suggest that dislocation channeling occurs in the irradiated proton layer at an equivalent stress close to 70% of the yield stress of the irradiated material, i.e. while the irradiated layer is still in the elastic regime for a 0.5% applied macroscopic plastic strain. Comparative constant elongation rate tensile tests performed at a strain rate of 10⁻⁵ s⁻¹ in iodized methanol solutions at room temperature on specimens both unirradiated and proton-irradiated to 2 dpa demonstrated a detrimental effect of irradiation on I-SCC.     

Stress ratchetting under tension and cyclic torsion

Investigations have been carried out to find the effect of interposed alternating twist between two tensile loadings on the plastic strain accumulation in tension. From the experiments carried out on aluminium and steel it was observed that interposing an alternating twist of ± θ between two successive tensile stressings increased the plastic strain on the next reloading. However, the rate of increase in the plastic strain accumulation decreases in the subsequent alternate twisting and reloading which may be the result of strain hardening of the material, introduced due to alternate cycling. Assuming a general yield function, a relation is developed for the ratchetting strain under this programme of loading, namely, tension and cyclic torsion.     

Uniaxial creep response of Alloy 800H in impure helium and in low oxygen potential environments for nuclear reactor applications

Studies were conducted on the creep behavior of Alloy 800H in impure helium and in a 1%CO-CO₂ environment. At relatively low applied stresses and at low temperatures, the presence of methane in helium reduced the rupture strain significantly while increasing the rupture life relative to the behavior in pure helium. The degradation in rupture strain is due to the occurrence of cleavage fracture in the He + CH₄ environment; this explanation is also supported by high activation energy (Q = 723 kJ/mol) for creep in He + CH₄. At higher applied stresses and also at higher temperatures, creep-rupture behavior in He and He + CH₄ was similar. Creep response in pure He and in CO-CO₂ follows a dislocation climb-controlled power-law behavior whereas that in He + CH₄ has a different behavior as indicated by the high stress exponent (n = 9.8). The activation energy for creep in pure He was 391 kJ/mol and in CO-CO₂ was 398 kJ/mol, and appeared to be independent of stress in both environments. On the other hand, in He + CH₄, the activation energy (Q = 723 kJ/mol) seems to be dependent on stress.     

Mechanical properties of fluorite-related oxides subjected to swift ion irradiation: Pyrochlore and zirconia

The modifications of the mechanical properties of related-fluorite oxides (cubic zirconia [c-ZrO₂] and pyrochlores [Gd₂(Ti_1−xZr_x)₂O₇ with x = 0.5 and x = 1]) induced by swift heavy ion irradiation are investigated. Polycrystalline pellets of both materials were irradiated at room temperature with 940 MeV Pb or 870 MeV Xe ions at the GANIL accelerator in Caen at fluences ranging from 2 × 10¹¹ to 10¹³ cm⁻². Residual macroscopic stresses induced by irradiation were determined using X-ray diffraction and the sin²ψ method. The microhardness and the fracture toughness of irradiated samples were studied by Vickers micro-indentation. Amorphization occurs in Gd₂TiZrO₇ and not in Gd₂Zr₂O₇ and c-ZrO₂. The mechanical behavior of materials is found to be closely related to the residual stresses induced in the surface layer by irradiation. Compressive stresses are generated in c-ZrO₂ and Gd₂TiZrO₇ (leading to an increase of fracture toughness), whereas tensile stresses (inducing a large decrease of fracture toughness) are observed in Gd₂Zr₂O₇ due to the lattice contraction related to a pyrochlore fluorite→transition.     

Irradiation and annealing effects on delamination toughness in carbon/epoxy composite

Gamma irradiation to various doses (4.8-27.2 MGy) was performed on unidirectional carbon fiber/epoxy resin composite plates. Unidirectional composite coupons irradiated to various doses were annealed at 180 and 250 °C, in vacuum. The strain energy release rate G_IC, as a measure of delamination fracture toughness, was determined by Mode I fracture testing on double cantilever beam coupons. The glass transition temperature (T_g) of the tested coupons matrices was determined in DMA tests. The effects of irradiation and annealing on G_IC values - the mean values of 10 propagation points (G_IC,mean) and that of fracture initiation (G_IC,init) - were established. These values were analyzed as a function of irradiation dose and annealing temperatures, having in mind glass transition temperature values changes, as well as the possible mechanisms and phenomena of irradiation and annealing.     

Tensile-shear correlations obtained from shear punch test technique using a modified experimental approach

Shear punch testing has been a very useful technique for evaluating mechanical properties of irradiated alloys using a very small volume of material. The load-displacement data is influenced by the compliance of the fixture components. This paper describes a modified experimental approach where the compliances of the punch and die components are eliminated. The analysis of the load-displacement data using the modified setup for various alloys like low carbon steel, SS316, modified 9Cr-1Mo, 2.25Cr-1Mo indicate that the shear yield strength evaluated at 0.2% offset of normalized displacement relates to the tensile YS as per the Von Mises yield relation (σ_ys = 1.73τ_ys). A universal correlation of type UTS = mτ_max where m is a function of strain hardening exponent, is seen to be obeyed for all the materials in this study. The use of analytical models developed for blanking process are explored for evaluating strain hardening exponent from the load-displacement data. This study is directed towards rationalizing the tensile-shear empirical correlations for a more reliable prediction of tensile properties from shear punch tests.     

Li (i = 1-3) subshell X-ray production cross sections and fluorescence yields for some elements with 56 ? Z ? 68 at 22.6 keV

The L_k (k = l, α, β_1,4, β_3,6, β_2,15,9,10,7, γ_1,5 and γ_2,3,4) X-ray production (XRP) cross sections have been measured for six elements with 56 ? Z ? 68 at 22.6 keV incident photon energy using the EDXRF spectrometer. The incident photon intensity, detector efficiency and geometrical factors have been determined from the K X-ray yields emitted from elemental targets with 22 ? Z ? 42 in the same geometrical setup and from knowledge of the K XRP cross sections. The L₁ and L₂ subshell fluorescence yields have been deduced from the present measured L_k XRP cross sections using the relativistic Hartree-Fock-Slater (HFS) model based photoionization cross sections. The present deduced ω₁ (exp) values have been found to be, on an average, higher by 15% and 20% than those based on the Dirac-Hartree-Slater (DHS) model and the semi-empirical values compiled by Krause, respectively, for elements with 60 ? Z ? 68.