Plane strain ductility of zircaloy-4 tubes as dependent on structure and texture

Zircaloy-4 tubes were cold rolled 50 or 80% to a series of textures normal for regular tube production and finish annealed at 510 and 575°C. Two different types of ring expansion specimens giving local transverse deformation under a plane strain condition were developed. Plane strain tests were performed out-of-reactor at room temperature and 350°C, after which the plane strain ductility was measured as local transverse elongation. Both at room temperature and at 350°C tubes recrystallized at 575°C have higher plane strain ductility than tubes partially recrystallized at 510°C. The plane strain ductility is independent of the texture over the texture range of practical interest for regularly produced Zircaloy tubes. This seems to be a consequence of the fact that the stress ratio $\text{σ}{}_{\mathrm{\theta }}\text{: σ}{}_{\mathrm{z}}$ in plane strain loading varies with the texture in such a way that the degree of texture softening during deformation is the same for all canning tube textures.     

The effect of cold working on creep rupture properties for helium-injected austenitic stainless steel

In order to study the effect of helium on the high temperature embrittlement of stainless steel, helium $\text{(7.5 × 10}{}^{\mathrm{?6}}\text{atomic fraction)}$ was injected into cold-worked stainless steels by using a cyclotron. At 650°C, it appeared that the reduction in creep-rupture strength due to helium was larger as cold-working was increased, but a loss of rupture elongation was less for a particular degree of cold-working. The 10% cold-worked material showed particularly good creep-rupture properties in the presence of helium. The loss of ductility was more pronounced in the creep test than in the tensile test.     

Examination of the plastic properties of zircaloy-4 at elevated temperatures in air atmosphere

Results are reported about the examination of the influence of test temperatures and sample anisotropy upon the yield stress, the strain rate sensitivity and the ductility of Zry-4 sheet-type tensile specimens. Most of the investigations were performed in air atmosphere in the temperature range between 750 and 1050°C at cross-head speeds ranging over three orders of magnitude. Superplastic deformation behaviour was observed at temperatures between 860 and 1000°C for a strain rate of the order of 10^?4 s^?1. For samples deformed under these conditions in air atmosphere the strain rate sensitivity index $\text{m}{}_{\mathrm{a}}$ has shown to depend sensitively upon strain. Although no influence of anisotropy upon the yield stress was observed above 550°C, in the superplastic region $\text{m}{}_{\mathrm{a}}$ revealed a strong dependence upon sample orientation. For higher temperatures even in air the total elongation increased with temperature up to 850°C. At the highest temperature due to severe take-up of oxygen the samples failed brittle. The apparent activation energy for superplastic flow agrees very well with the energy of grain boundary diffusion in α-Zr.     

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.     

Tensile properties of annealed type 316 stainless steel after EBR-II irradiation

The effects of fast reactor neutron irradiation on the tensile properties of annealed Type 316 stainless steel were determined over a wide range of irradiation temperatures and reactor fluences. These effects are described as a function fluence and temperature to $\text{7 × 10}{}^{22}\text{n/cm}{}^2$ ($\text{E}{}_{\mathrm{n}}\text{> 0.1}\text{MeV}$) at 430 to 820 °C. The usual flow stress increase and ductility decrease were observed with increasing neutron fluence. Strengthening decreases continuously from 480 °C to ≈ 700 °C with no hardening at or above 760 °C. Elongation values increase with temperature in the 430–540 °C range and generally decrease with temperature above 540 °C.     

The transformation behavior of a U-16.4 at % Nb-5.6 at % Zr alloy

The isothermal transformation kinetics of a γ-phase U-16.4 at % Nb-5.6 at % Zr alloy have been studied by resistometric techniques between 130 and 590° C. Three distinct regions of transformation were observed. Above 420° C the transformation proceeded by nucleation and growth and progressed to the equilibrium $\text{α + γ}{}_2$ phases. Aging at temperatures between 270 and 400° C resulted in the appearance of a transition $\text{α}{}^{\mathrm{″}}$ monoclinic phase that remained stable after even 7 days of aging at 380° C. Aging below 250° C produced a tetragonal $\text{γ}{}^0$ phase. A negative temperature coefficient of resistance is observed at temperatures below 400° C when the material is quenched from a 900° C anneal. It is suggested that 400° C represents an athermal γ to $\text{γ}{}^{\mathrm{<mtext>S</mtext>}}$ transformation and that the negative resistance coefficient results from a continual change of lattice parameters of $\text{γ}{}^{\mathrm{<mtext>S</mtext>}}$ with decreasing temperature.     

The tensile and corrosion behaviour of ordered Zr3Al-based alloys

Polyphase alloys based on ordered Zr₃Al (Ll₂ type) have been deformed in tension at 20 to 600°C and exposed to moist air (300°C), pressurized water (300°C) and atmospheric pressure steam (400°C). Certain alloys display a uniform elongation of 25 to 30% before failure, an ultimate tensile strength at 300°C of 1120 MN/m² (160 000 psi), and an oxidation rate of 1.2 mg/dm² d day in atmospheric steam at 400°C. For good ductility a requirement is the absence (or at least a fine distribution) of Zr₂Al, an intermetallic phase which may remain following the transformation $\text{Zr + Zr}{}_2\text{Al → Zr}{}_3\text{Al}$. Similarly, a requirement for corrosion resistance is the absence (or at least a fine distribution) of the highly corrodible α-Zr(Al) phase. At low temperatures yield appears to be controlled by long-range, internal stresses. Failure occurs suddenly, suggesting that a critical fracture stress exists.     

Influence of temperature on tearing resistance of welds for PWR vessels

The effect of temperature on ductile fracture toughness of three narrow gap SAW welds and one MMAW weld (SA 508 Cl.3 base metal) was investigated using 25 mm thick CT specimens. Chemical analyses, tensile and Charpy V tests were also performed. Two methods of toughness characterization (partial unloadings or interrupted tests) were used at 20–43°C and service temperature (293°C). Values of J at initiation, and after a moderate propagation were considered and compared. At a given temperature, properties of the four welds were fairly similar. A conservative estimate of the toughness reduction factor, associated with a temperature increase from 43°C to 293°C, is $\text{J}{}^{293}\text{?}\text{J}{}^{43}\text{1.6}$. Fracture surfaces were examined, showing a pattern of patches with dimples separated by areas of smoother surfaces. This fracture surface appearance can be related to the weld microstructure. Two models were tried for predicting the change of fracture toughness with temperature through the effect of this latter parameter on tensile properties. The characteristic distance model of ductile fracture provides a satisfactory estimate of the temperature effect on toughness.     

Inelastic behaviour of 214Cr-1Mo steel under plasticity-creep interaction condition: An interim report of the Bench Mark project by the subcommittee on inelastic analysis and life prediction of high temperature materials,JSMS

Some of the interim results of the Bench Mark Project by the Subcommittee on the Inelastic Analysis and Life Prediction of High Temperature Materials, JSMS, is presented. The purpose of the present bench mark study is to review and evaluate the inelastic constitutive models relevant to material response under the plasticity-creep interaction.By specifying normalized and tempered $\text{2}\text{1}\text{4}\text{Cr-1Mo}$ steel at 600°C, sixteen bench mark problems of four categories are first established: (I) tensile stress-strain relations and creep curves, (II) material response under mixed modes of plastic and creep loading, (III) ratcheting and deformation under program loads, and (IV) cyclic deformation behaviour under the combination of different strain rates. Then, the outline of seventeen inelastic constitutive models of nine types discussed in this project is presented. Finally, the interim results of these bench mark tests are compared with the corresponding predictions of the constitutive models to evaluate their accuracy in simulating the actual behaviour of the material.     

Nickel ion bombardment of type 304 stainless steel: Comparison with fast reactor swelling data

Annealed Type 304 stainless steel containing 15 atomic ppm of helium has been bombarded with 5 MeV nickel ions at 525°C to 700°C. A pronounced swelling peak occurs at 625°C, compared to a swelling peak temperature of about 475°C in reactor. TEM measurements of void swelling at 625°C as a function of ion dose show a swelling of almost 40% at 124 dpa without evidence of saturation. Measurements of gross swelling of the ion-bombarded material by a new step-height method provide information that is in good agreement with TEM data, and can be extended to larger swellings. The step-height results indicate a swelling of over 90% at 290 dpa at 625°C. The ion-produced swelling agrees well with in-reactor data when the two are compared at the respective peak swelling temperatures, and the void concentrations and average void diameters are comparable for the two cases. The high ion dose results are used to guide extrapolation of reactor data to higher fluences, leading to the following predictions for swellings at the peak swelling temperature in reactor: 18% swelling at $\text{1× 10}{}^{23}\text{n/cm}{}^2$ (fast), 50% at $\text{2 × 10}{}^{23}$, and 80% at $\text{3 × 10}{}^{23}$.     

The irradiation creep of nickel and AISI 321 stainless steel during 4 MeV proton bombardment

An apparatus has been developed to study the creep of thin metal specimens under tensile stress during bombardment by 4 MeV protons from the Harwell Van de Graaff Accelerator. The specimen is held in a helium atmosphere and the proton beam reaches it through a thin metal window at the end of the accelerator beam line. The proton beam passes through the thin (25 μm) specimen, losing ~1.5 MeV in the process (most of which contributes to heating the specimen) and creating almost uniform radiation damage at the rate of $\text{(1–10) × 10}{}^{\mathrm{?7}}$ displacements per atom per second (dpa s^?1). The specimen temperature is monitored by infra-red pyrometry and controlled to ± 0.2°C by additional DC heating via the infra-red pyrometer output to compensate for ion beam fluctuations. The irradiation creep strain of the specimen is continuously measured with a sensitivity of $\text{5 × 10}{}^{\mathrm{?6}}$ by a linear variable differential transformer. Irradiation times up to about 100h with reasonable beam stability are possible. Results are presented of the irradiation creep behaviour of pure Ni and both solution treated and cold-worked AISI 321 stainless steel bombarded in the temperature range 400–600°C under tensile stresses in the range 20–250 MPa.     

The effect of fast neutron irradiation upon the fatigue-crack propagation behavior of two austenitic stainless steels

The effect of fast neutron irradiation ($\text{454° < T}{}_{\mathrm{irr}}\text{< 477° C}$) to a fluence of $\text{9 × 10}{}^{21}\text{n/cm}{}^2$ ($\text{E > 0.1 MeV}$) on the fatigue-crack growth behavior was investigated for annealed Type 304 and 20% coldworked Type 316 stainless steels using linear-elastic fracture mechanics techniques. Irradiation to this fluence had little or no effect upon the crack growth behavior of annealed Type 304 at a test temperature of 427° C, nor upon the behavior of 20% cold-worked Type 316 at test temperatures of 427° C and 538° C. Irradiation to this fluence did tend to decrease crack growth rates slightly, relative to unirradiated material, in annealed Type 304 at a test temperature of 538° C.     

Thermal conductivity of neutron-irradiated pyrolytic β-silicon carbide

Samples of pyrolytic β-silicon carbide deposited at 1400 °C (grain size $\text{～ 1 μ}\text{m}$) and at 1750 °C (grain size $\text{～ 3 μ}\text{m}$) were irradiated with fast neutrons to 2.7-7.7 × 10²¹ n/cm² ($\text{E > 0.18 MeV}$) at 550 °–1100 °C. Irradiation reduced the room-temperature thermal conductivity from $\text{～0.15}\text{cal/cm}\text{·}\text{sec}\text{· °}\text{C}$ to $\text{～ 0.02}\text{cal/cm}\text{·}\text{sec}\text{· °}\text{C}$ after irradiation at 550 °C and to $\text{～ 0.05}\text{cal/cm}\text{·}\text{sec}\text{· °}\text{C}$ for an irradiation temperature of 1100 °C. The thermal conductivity of unirradiated samples decreased with increasing measurement temperature, while that of the irradiated samples was much less temperature dependent. No difference in behaviour was found between the samples with $\text{～ 1 μ}\text{m}$ grain size and the samples with $\text{～ 3 μ}\text{m}$ grain size.     

The effect of carbon on 2.25 Cr-1 Mo steel: (I). Microstructure and tensile properties

The microstucture, hardness, and the tensile properties of 2.25 Cr-1 Mo steel with 0.009, 0.030, 0.120, and 0.135 wt % C were determined on steels in the annealed (furnace-cooled from 927°C), normalized (air-cooled from 927°C), and normalized-and-tempered conditions. As annealed, the microstructure was primarily proeutectoid ferrite with spherical carbides and pearlite, the amounts increasing with increasing carbon content. During normalization ($\text{7}\text{8}$ in rods or 1 in plates were heat-treated), granular bainite formed: 1 to 2 and 15 to 20% bainite (remainder proeutectoid ferrite) for the 0.009 and 0.030 wt % C steels, respectively; the 0.120 and 0.135 wt % C steels were entirely bainite. On tempering, carbides precipitated. In all heat-treated conditions, there was little difference in the room temperature hardness of the 0.009 and 0.030 wt % C steels and between the 0.120 and 0.135 wt % C steels. Tensile tests from 25 to 565°C indicated that strength depends on microstructure, which is determined by carbon content.     

Post-irradiation ductility loss of FeNibase precipitation hardenable alloys

Several γ′- and γ′/γ″-strengthened Fe-Ni-base superalloys have shown near-zero ductility after neutron irradiation to fluences of $\text{～ 4 × 10}{}^{22}\text{n/cm}{}^2$, $\text{E > 0.1 MeV}$, at 500 to 650°C. The ductility loss is most pronounced in solution-treated or in solution-treated and aged specimens tested at 110°C above the irradiation test temperature. Failed specimens exhibit brittle intergranular fractures. Microstructural examination of the embrittled specimens showed that continuous or semi-continuous coatings of γ′ formed at grain boundaries during the irradiation. In some cases, the grain boundaries were also decorated by small bubbles, thought to be transmutation-induced helium or contained trace elements such as sulphur and phosphorus. All of these grain boundary alterations are attributed to radiation-induced solute segregation. Microanalyses of the γ′ coatings indicate that Ni, Al, Si, Ti and Nb had segregated to grain boundary sinks in irradiated FeNiCr based alloys. Nonequilibrium segregation of helium and trace impurities is also considered likely. The role of radiation-induced segregation in the embrittlement phenomenon is consistent with the observation that introducing a high density of dislocation sinks by cold-working reduces γ′ formation at the grain boundaries and reduces the ductility loss. The embrittlement is attributed to concurrent strengthening within grains by irradiation-induced γ′ precipitation and brittle cleavage failure of grain boundary precipitates.     

Nickel-ion bombardment of annealed and cold-worked type 316 stainless steel

Bombardment with high doses of 5 MeV nickel ions has produced swellings as high as 90% and 60%, respectively, in annealed and 20% cold-rolled Type 316 steels. The steels contained 15 ppm of cyclotron-injected helium. Swellings were determined by both transmission electron microscopy and by a step-height method that measures the total swelling integrated along the ion path. The swelling in annealed Type 316 has a pronounced peak in the vicinity of 625°C, which is about 155°C higher than the peak swelling temperature in-reactor. The magnitudes of the swelling, void densities and void sizes produced in annealed Type 316 by nickel ions and in-reactor at the respective peak swelling temperatures are similar and it is concluded that the nickel ion bombardments provide an acceptable simulation of in-reactor behavior. Using the high dose ion results to guide extrapolation of presently available EBR-II data to higher fluences leads to the prediction that the swelling of annealed Type 316 steel at the peak swelling temperature will reach 40% at $\text{2 × 10}{}^{\mathrm{p23}}\text{n/cm}{}^2\text{(E > 0.1 MeV)}$ in EBR-II core, and 70% at $\text{3 × 10}{}^{23}\text{n/cm}{}^2$. These fluences in EBR-II correspond to 155 and 230 dpa respectively. Twenty percent reduction by cold-rolling reduces the ion produced swelling by 35% at 625°C and by 50% at 575°C.     

Pic de relaxation associe aux interstitiels intrinseques dans le zirconium deforme a basse temperature

Internal-friction measurements carried out at three widely different frequencies (≈1, 16 and 350 Hz) have allowed the detailed study of a damping peak in zirconium deformed at low temperature and not allowed to warm up beyond ?68°C. The values obtained for the activation energy and the frequency factor (respectively $\text{0,27 ± 0.02}$ eV and $\text{10}{}^{\mathrm{13\pm 0.8}}\text{sec}{}^{\mathrm{?1}}$), the narrowness of the peak — which allows it to be defined in terms of two simple relaxation times — and the fact that disappearance of this peak at ?68°C is not directly consistent with a dislocation-pinning mechanism, lead to the postulate of a relaxation mechanism involving point defects of a paired oxygen self-interstitial type formed during low-temperature deformation.