The effect of carbon on 2.25 Cr-1 Mo steel: (II). Creep—rupture properties

The creep-rupture properties of normalized-and-tempered 2.25 Cr-1 Mo steel with 0.009, 0.030, 0.120, and 0.135 wt % C were determined at 454, 510, and 565°C; the results show an effect of carbon that depends on temperature. At all three temperatures, the 0.120 and 0.135 wt % C steels have similar strengths, and at 565°C are considerably stronger than the 0.009 and 0.030 wt % C steels, which also have similar strengths. At 454 and 510°C, however, the properties for the 0.030 wt % C steel are superior to those of the 0.009 wt % C steel and approach those for the 0.120 and 0.135 wt % C steels. These observations can be explained in terms of the type of carbide precipitation reactions that occur in 2.25 Cr-1 Mo steel.     

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.     

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.     

The effects of warm work on some mechanical and fabrication properties of uranium

Uranium plates have been warm rolled at 250°C to reductions of up to 90%. Subsequent tensile tests at ambient temperature have shown a progressive increase in fracture stress and ductility with increased deformation. No evidence of anistropy in tensile properties was observed for material unidirectionally rolled to 80% reduction at 250°C. Cupping tests have been performed on 0.048 in (1.2 mm) uranium sheet warm rolled up to 90% reduction. Tests were made at R.T. and 200° C on both the as-rolled and the annealed conditions. Values of $\text{R}$, the average width strain to thickness strain ratio, were determined during the tensile tests. No correlation was observed between the determined limiting drawing ratios and $\text{R}$ values. However, good cupping behaviour was shown to be a characteristic of material with a high tensile ductility.     

A mechanism for in-reactor oxidation of zirconium alloys from oxide electrical characteristics: (III). Zr-2.5 wt % Nb

A molten nitrate/nitrite eutectic has been used for studying the electrical characteristics of oxides on heat-treated Zr-2.5 wt % Nb in two conditions: quenched from 850°C or quenched, cold-worked and aged at 500°C. Oxides were grown in either water or steam, with or without oxygen, out-of- or in-reactor (fast flux $\text{1.5 × 10}{}^{13}\text{n · cm}{}^{\mathrm{?2}}\text{· s}{}^{\mathrm{?1}}$). Tafel analysis of current-voltage curves yielded oxide growth rates that were generally in reasonable agreement with weight gain data. Radiation suppressed the oxidation rate and the ionic and electronic oxide conductances, by inducing oxide recrystallisation and precipitating niobium from solid solution in the metal. The latter was not a factor for quenched, cold-worked and aged material whose most important corrosion resistant property was that its oxide did not become highly water permeable in reactor environments containing free oxygen. In this the alloy's behaviour was in sharp contrast to that of Zircaloy-2 and other alloys.     

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.     

Corrosion of various ferritic steels in an isothermal sodium loop system

Ferritic chromium-molybdenum steels with chromium contents of 1 wt% up to 12 wt% have been exposed for 8370 h to flowing sodium at 550°C. The oxygen content of the sodium was 6–7 ppm by weight. Weight measurements, carbon analyses and metallographic examinations were carried out. The low chromium steels show weight loss and decarburisation. The high chromium steels show weight gain and carburisation. The crossover point is at about 5 wt% Cr. The composition at the utmost surface (<10 μm) of the various steels tends to about 8 wt% chromium, about 2 wt% nickel and 0.02–0.09 wt% carbon. Sodium chromite crystals were present on the steels with a chromium content of 5 wt% or more. At the exposed surface of the $\text{2}\text{1}\text{4}$ wt% chromium steel sodium chromite crystals were found locally.     

The corrosion of some zirconium alloys under radiation in moist carbon dioxide-air mixtures

In-reactor experiments are reported in which Zircaloy-2 and $\text{Zr-2}\text{1}\text{2}$ wt % Nb alloys were exposed to moist carbon dioxide-air mixtures at 300°C for periods up to ≈ 800 days. For Zircaloy-2 the corrosion was significantly enhanced by the reactor radiation but the percentage hydrogen pick-up was reduced. For $\text{Zr-2}\text{1}\text{2}$ wt % Nb alloy the effect of radiation on the corrosion rate ranged from a slight suppression to a significant enhancement, depending on the metallurgical condition of the alloy, but there appeared to be no effect on the percentage hydrogen pick-up. The effect of improving the purity of the corrodent was also studied in the absence of radiation. The results are used to predict that the corrosion and hydrogen pick-up of typical reactor pressure tubes exposed to moist carbon dioxide-air mixtures are unlikely to cause a significant deterioration of their mechanical properties.     

Effects of stress on swelling in 316 stainless steel

Stress was found to increase the magnitude of irradiation-induced swelling in 316 stainless steel. Measurement of the densities of pressurized tube specimens, irradiated at temperatures of ~ 430–475°C to peak fluences of $\text{~ 9 × 10}{}^{22}\text{n/cm}{}^2$ ($\text{E > 0.1 MeV}$) in EBR-II, has indicated increased swelling in both the annealed and 20% cold worked conditions of this alloy. Swelling in the annealed specimens was observed to increase linearly with hoop stress up to ~ 20 ksi (130 MPa), whereupon further increases in stress resulted in reduced swelling. Swelling in the cold worked material was linear with stress up to levels of ~ 28 ksi (193 MPa).     

Effect of cyclotron-injected helium on void-formation characteristics of type 304 stainless steel

Samples of Type 304 stainless steel were injected with helium by cyclotron bombardment to concentrations ranging between $\text{1.1 × 10}{}^{\mathrm{?7}}$ and $\text{1 × 10}{}^{\mathrm{?4}}$ ppma. Following cyclotron injection, the samples were given a variety of heat treatments prior to insertion in EBR-II for irradiation at 450 °C to a total dose of $\text{1 × 10}{}^{21}\text{n/cm}{}^2$. Samples that were not heat treated or that were annealed at 650 °C following cyclotron injection formed few voids and dislocation loops after EBR-II irradiation. This behavior is apparently due to the precipitate clusters that were formed during the helium injection. These precipitates were analyzed by electron microscopic techniques and found to have spherically symmetric strain fields that were of interstitial character. Samples that were annealed at 760 °C following cyclotron injection formed a larger number density of both voids and dislocation loops than did the control sample after EBR-II irradiation. The void volume also exceeded that of the control. Clustering of the dislocation loop population near grain boundaries and precipitate particles was observed in the control and low helium concentration samples.     

Thermal stability of the superalloys Inconel 625 and Nimonic 86

The thermal stability of Inconel 625 and Nimonic 86, as received, cold worked (10, 20, and 40%), and solution treated, was investigated in the temperature range 500–900°C. The annealing times varied from 0.3 (0.03) to 100 days.Precipitation hardening and recovery (recrystallisation) takes place in cold worked material, beginning after shorter times in cold worked material than in as received material. The temperature interval for precipitation hardening is extended in Nimonic 86, due to cold working, from about 500–600°C to about 450–700°C. It is possible to suppress or retard the precipitation hardening in solution treated Inconel 625 and Nimonic 86 by fast cooling after solution annealing. Hardness was measured at room temperature with five different loads, so that the parameters $\text{k}$ and $\text{n}$ from Meyer's-law, and the Brinell hardness number (for $\text{F ÷ D}{}^2\text{=30}$) could be determined.The lattice contraction of Inconel 625 due to ageing was investigated with X-ray measurements. The change of intensities of the diffractometer traces due to recovery was also determined.     

Heat treatment effects on the tensile properties of annealed 2.25 Cr-1 Mo steel

The effect of heat treatment on the tensile properties of annealed 2.25 Cr-1 Mo steel was investigated. Detailed tensile properties were determined from 25 to 593°C and strain rates between 2.67 × 10^?6 and 6.67 × 10^?3/s on steel plates (from a single heat) given three different heat treatments. For all heat treatments, dynamic strain-aging peaks were observed between 200 and 400°C. The peak height for the annealed material that was cooled fastest was largest and occurred at the highest temperature. The dynamic strain aging was concluded to be the result of interaction solid solution hardening in the proeutectoid ferrite and involves interactions between molybdenum and carbon atoms or atom clusters with dislocations. The difference in the dynamic strain-aging effects for the different heat treatments was explained in terms of the precipitation reactions that occur during the heat treatment.     

The kinetics of helium release from irradiated samples of austenitic steels OKh16N15M3B and OKh16N15M3BR

Samples of a heat-resistant austenitic steel (both boron free and with 0.005% by weight boron added) were irradiated in the active zone of Reactor WWR-M with a neutron flux of $\text{4 × 10}{}^{13}\text{n/cm}{}^2\text{· s (E ? 3 MeV)}$ and $\text{6 × 10}{}^{13}\text{n/cm}{}^2\text{· s (E ? 0.03 eV)}$ at fixed temperatures in the range of 600°C to 850°C. The neutron fluence was 6 to $\text{8 × 10}{}^{19}\text{n/cm}{}^2$. For the study of the helium release kinetics the samples were annealed with a constant heat-up rate. The experimental equipment and procedure were described previously. Most of the helium (70 to 90%) is released at temperatures of 1420 to 1490°C. This is connected with the melting of the samples. Helium release during the annealing in the range 20 to 1300°C is observed in the form of seven peaks on the curve of gas-release rate. Attempts to interpret the helium diffusion mechanism of each stage is given by determination of the activation energies and other data. Irradiation in a helium atmosphere leads to penetration of helium into the samples, which is proportional to the external gas density. A significant intensification of the generation of helium cavities occurs at irradiation temperatures of 750 to 850°C.     

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 deposition of uranium-sesquicarbide

Deposition of U₂C₃ from an aic-melted mixture of UC + UC₂ ($\text{7.0 ± 0.1}$ wt % carbon) was examined primarily by the change of electrical resistivity as a function of time at fixed temperatures: 1370, 1440, 1575 and 1650°C. The rate constant of the reaction UC + UC₂ → U₂C₃ was investigated in detail. Growth with one constant dimension, which has been named layer growth, was predominate at 1370°C and during the early stages of growth at higher temperatures. Various growth schemes, which are generally involved, yield various activation energies. The rateconstant of thereaction was obtained as $\text{k}{}_{\mathrm{n}}\text{= 1.7 × 10}{}^{\mathrm{?2}}\text{exp(?n × 24.67/RT)}$ sec^?1, where $\text{n}$ depends on growth scheme and is usually a number less than 3. The activation energy of 24.67 kcal/mol in the above equation was obtained from rate constants with similar $\text{n}$ values ($\text{n ≈ 3}$) at 1440 and 1575°C.     

Carburization and tensile behavior of alloy 800 in liquid sodium

The carbon transfer has been analyzed in the Alloy 800/sodium/stainless steel system by determining the carbon-uptake of Alloy 800 foils, which were exposed in liquid sodium of known carburizing potential. p ]Under equilibrium conditions between 650 and 550°C the measured total carbon concentrations in the Alloy 800 labs were found to be roughly related to the carbon activities of the sodium environment by the equation already stated for the $\text{18Cr-}\text{8}\text{10}$ Ni stainless steels, extrapolated to the chemical composition of the Ni-rich austenitic alloy. However the Alloy 800 was not found to undergo any decarburization in low-carbon activity environments. The carbon diffusion kinetics was determined as a function of temperature, it was found to be similar to that reported for the AISI-304 type of steel. p ]The effects of sodium exposure on microstructural and mechanical properties of several Alloy 800 heats were examined at 550°C as a function of the active carbon concentration in the sodium and of the $\text{Ti}\text{C}$ ratio in the alloys.     

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.     

Observations of the microstructure of a 8% manganese stainless steel (ICL 016) before and after irradiation with 46 MeV nickel ions

Results are given of a preliminary investigation of the microstructure of a commercial Mn 8%/Cr 19%/Ni 7% austenitic steel (ICL 016) before and after irradiation with 46 MeV nickel ions. Pre-irradiation phases observed were Cu-rich precipitates (d ~ 10 nm) and α-MnS phase. A surface-localised ferromagnetism observed after annealing or irradiation was found to be due to α'-martensite formed as a result of an increase in the $\text{γ/ga'}$ transformation temperature due to evaporation of austenising elements such as Mn.Ion irradiation to 60 dpa at 625°C resulted in void-swelling of ~ 7% in solution-treated alloy containing 10 appm He. whereas swelling of ~ 1.8% occurred in the absence of helium. Irradiation also resulted in the formation of thin lath-like precipitates and the coarsening of the Cu precipitates. The results indicate that this manganese-containing alloy has an average swelling response when helium is present, with an indication that swelling can be reduced by pre-ageing at 700°C. In the ST or STA condition the alloy does not seem to offer any advantage in terms of void-swelling over other Fe-Cr-Ni austenitic steels currently favoured for LMFBR applications. The swelling sensitivity of the alloy to helium and the tendency to induced surface ferromagnetism indicate the need for further study before selecting this type of alloy for use in fusion reactors.     

Dynamic strain-ageing of A203D nuclear structural steel

The present investigation deals with the dynamic strain-ageing behaviour of a nuclear structural steel, designated ASTM A203 grade D, in tempered martensitic and ferritic-pearlitic microstructural conditions. The serrated stress-strain curves, characteristic of this phenomenon have been observed in the temperature range 100–200°C, with nominal strain rates varying from $\text{1.33 × 10}{}^{\mathrm{?5}}$ to $\text{6.66 × 10}{}^{\mathrm{?4}}\text{/s}$. It has been noted that dynamic strain-ageing causes a sharp rise in ultimate strength and work-hardening rate, a marked decrease in ductility and a negative strain-rate sensitivity of the flow stress. In this temperature range, the yield stress also increases with increasing temperature but the rise in ultimate stress is much greater than the rise in yield stress. The temperature and strain-rate dependence of the onset of serrations yields an activation energy of 63 kj/mol (15 kcal/mol), which suggests that the process is controlled by interstitial diffusion, probably of nitrogen, in ferrite. It appears that microstructure does not have any strong influence on the changes in mechanical properties of this steel during dynamic strain-ageing.