M.S. Anand

Diffusion of carbon in stainless steels

Abstract Using residual activity technique, diffusion of carbon-14 in 304, 347 and 316 steels has been studied in the temperature range of 450–1200°C. The temperature dependence of diffusivity could be expressed as Dc/304s. steel = 6.18 exp (−44 610/RT); Dc/347s. steel = 0.35 exp (−40 140/RT); Dc/316s. steel = 0.19 exp (−37 400/RT). The activation energy for the diffusion of carbon in steels has been explained on the basis of alloying effect of the constituents in steels. Furthermore, thermodynamic calculations for the free energy of formation of various carbides have been carried out to predict the carbon pick up by stainless steels as Cr23C6 from uranium carbide in a uranium carbidestainless steels system. In order to study the segregation of carbon along the grain boundaries of stainless steels, in the present investigation autoradiographie technique has been employed. The results indicate that the segregation of carbon is due to the preferential precipitation of Cr23C6 along the grain boundaries of stainless steels.

Diffusion of chromium in Inconel-600

Abstract The lattice and grain boundary diffusion of 51Cr in Inconel-600 (nickel base alloy) has been studied using Gruzins residual activity technique in the temperature range of 673–1523K. The lattice and grain boundary diffusivities can be expressed as D(1073–1523K) = (1.60 ± 0.30) × 10 −4 exp ( −q RT )m 2/s , q = 277.67 ± 4.21 and D gb σ(673–1073K) = (4.23 ± 0.80) × 10 −14 exp ( −q RT ) m3/s, q = 179.54 ± 3.60.δ is the grain boundary width and The activation energies are given in kJ/mol and R in joule per degree per mole. The effect of chromium and iron on the diffusivity in the alloy has been discussed on the basis of bond strength.

Defect production rates in metals by reactor neutron irradiation at 4.6 K

Abstract The absolute differential neutron energy spectrum for the low-temperature irradiation facility (TTB) in the Research Reactor Munich (FRM) was determined by a multiple foil activation technique. For a large number of elements radiation damage parameters were calculated. Experimental data for initial resistivity damage-rates in various metals were obtained and compared with the calculations, as well as with results from other irradiation facilities.

Effect of alloying elements on recovery and damage rates in zirconium

Abstract Pure Zr and Zr-X (where X = Cr, Fe, Ce and Sn ) alloys have been irradiated to a fast neutron fluence of 1.8 × 10 22 nm −2 at 4.6 K. The radiation induced resistivity has been measured in situ as a function of dose. It was observed that compared to pure zirconium the radiation induced resistivity and hence the initial damage accumulation rate increases in the Zr-Sn alloy whereas it was practically unchanged by the other alloy elements. The irradiated samples were also subjected to isochronal annealing in the temperature range of 20–400 K and different annealing stages in pure zirconium and its alloys were studied. It was concluded that tin acts as a good trap for interstitials in stage I and for vacancies in stage III.

Self- and impurity diffusion in zirconium-manganese alloys

Abstract Self- and impurity diffusion in Zr-Mn alloys (Mn∼0–2 at. %) have been studied using the sectioning method in the temperature range 1173–1473 K. With an increase of the Mn content in the alloys, the diffusion parameters (D 0 and Q) for 95Zr diffusion in these alloys varies from 0·31 × 10−8 to 3·36 × 10−8 m2/s and 105·25 to 125·34 KJ/mole, respectively, while for 54Mn diffusion, they vary from 5·38 × 10−7 to 0·08 × 10−7 m2/s and 140·65 to 104·62 KJ/mole. Thus the addition of manganese to zirconium results in the enhancement of self-diffusivities while retarding impurity diffusion. The impurity correlation factor calculated using Le Claires model for body-centred-cubic crystal is 0·42. The results are consistent with a vacancy mechanism of diffusion.

Surface damage of V-Zr alloy by helium ions

Abstract Helium ion irradiation of V-2 wt% Zr alloy was carried out at various energies in the MeV range. Blisters and exfoliations were observed using SEM at higher doses. SIMS analysis was used to estimate qualitatively the retention of helium after annealing at 700 K. Hardness measurements were carried out on the bombarded area to determine the effects of retained gases on mechanical properties.

Damage rate measurements at 4.6 K and recovery studies in zirconium and zirconium-tin alloys

Abstract Pure Zr and Zr–Sn (0.015 at. % to 2.3 at. % Sn) allOays were irradiated at 4.6 K to a fast neutron fluence of 2.5 × 1022 neutrons m−2 and the radiation-induced resistivity was measured in situ as a function of dose. Compared to pure Zr the initial damage rate was higher in Zr—Sn alloys, whereas the saturation resistivity was found to remain almost constant. The samples were subjected to isochronal annealing in the temperature range of 30—509 K and the different recovery stages studied. The addition of tin suppressed the recovery in stage I and stage III.

Low temperature (4.6 K) fast neutron irradiation of zirconium and zircaloys 2 and 4: Dose and recovery studies

Abstract Zirconium Ziicaloy-2 and Zircaloy-4 were irradiated with fast neutrons to a fluence of 1 × 10 22 n / m 2 at 4.6 K. Dose curves [ d ( Δp )/ d (∅ t ) versus Δp ] were plotted for all the materials. Isochronal annealing of these irradiated materials was carried out from 18 to 323 K, showing defect peaks at 110–120 K and 235–237 K. Migration energy E M was determined for stage III and was 0.28 eV for zirconium and 0.35 eV for Zircaloy-2. An attempt has been made to explain these results.

Fast neutron irradiation of A-203 steel — recovery studies

Abstract A-203 steel has been irradiated by fast neutrons to fluences of 1 × 10 22 and 1.25 × 10 22 n / m 2 . The isochronal annealing studies have been carried out from 333 to 913 K. The results show a recovery peak at ~ 385 K, which has been explained to arise due to migration of carbon and carbon-vacancy complexes. After this stage the resistivity instead of decreasing starts increasing. This has been explained on the basis of increase of short range ordering on annealing. The influence of nickel and carbon on the recovery behaviour has also been discussed.

Interaction of oxygen with radiation-induced defects in dilute vanadium alloys

Abstract Isochronal annealing studies of both cold-worked and neutron-irradiated vanadium, vanadium-2 at.% niobium and vanadium-2 at.% aluminium alloy, were carried out. A stage was observed for all the materials at ~430 K. Recovery in vanadium-2 at.% aluminium was comparatively less than pure vanadium or vanadium-2 at.% niobium. Results indicate that oxygen is the migrating specie in this stage.