J.H. Evans

Significant differences in defect accumulation behaviour between fcc and bcc crystals under cascade damage conditions

Abstract Various aspects of irradiation-induced damage production and accumulation in fcc and bcc metals and alloys are examined. Results on the evolution of defect morphology in individual cascades, the global evolution of planar (clusters/loops) and three-dimensional defect agglomerates (voids) and the temperature dependence of defect accumulation in the form of loops and voids in fcc and bcc metals are compiled and compared. The comparison demonstrates that the damage accumulation in the form of clusters/loops and voids and their dependence on irradiation temperature are dramatically different in fcc and bcc metals. These differences may arise because of significant differences in the number of surviving defects (in cascades) and intracascade clustering of vacancies and self-interstitial atoms (SIAs) that get established already during the thermal spike phase of cascades. The implications of these differences are discussed. It is suggested that theoretical treatments of the damage accumulation in fcc and bcc metals must explicitly include the details of intracascade clustering of vacancies and SIAs, for example, the ratio of clustered to non-clustered vacancies and SIAs.

Effects of neutron irradiation on microstructure and deformation behaviour of mono- and polycrystalline molybdenum and its alloys

The influence of neutron irradiation on microstructural evolution and mechanical properties of mono- and polycrystalline molybdenum and its alloys has been investigated. Tensile specimens and 3 mm diameter discs of monocrystals of pure molybdenum and Mo-5%Re were irradiated with fission neutrons at similar to 320 K to displacement doses in the range 5.4 x 10(-4) to 1.6 x 10(-1) dpa (NRT) in the DR-3 reactor at Riso National Laboratory. For comparison, polycrystalline specimens of Mo-5% Re and TZM were also irradiated together with the monocrystalline specimens. Both unirradiated and irradiated specimens were tensile tested at 295 K. Post-irradiation microstructures were quantitatively characterized using a transmission electron microscope (TEM). Fracture surfaces were examined in a scanning electron microscope (SEM). The results of tensile testing as well as of transmission and scanning microscopy are presented and discussed in terms of intracascade clustering of self-interstitial atoms and the role of one-dimensional glide of these clusters in controlling microstructural evolution and the resulting mechanical properties

Microstructure and mechanical behaviour of TZM and Mo-5% Re alloys irradiated with fission neutrons

Abstract The response of microstructural and mechanical properties of TZM and Mo-5% Re alloys to neutron irradiation are reported. Irradiations were carried out at five temperatures between 323 and 723 K to a dose level of ∼ 0.16 dpa. The resulting microstructures consisted of a high density of small loops in both alloys with the addition of characteristic loop rafting in TZM. There was a marked sensitivity of microstructure to temperature. Void formation was found in both TZM and Mo-5% Re after the 623 and 723 K irradiations. The irradiations induced a significant increase in Vickers hardness in all cases but with no obvious dependence on irradiation temperature. Even at the present low neutron dose, large increases in tensile strength and a drastic decrease in ductility were observed. These results are discussed in terms of source hardening of the matrix within grains rather than obstacle hardening. Although the fracture processes are concentrated at the grain boundaries, these are not due to grain boundary embrittlement.

Effect of neutron irradiation on microstructure and tensile properties of TZM and Mo-5% Re alloys

Abstract The response of two molybdenum alloys (e.g. TZM and Mo-5% Re) to fission neutron irradiation has been investigated. These alloys were irradiated in the DR-3 reactor at Riso National Laboratory with a flux of ∼ 2.5 × 10 17 n/m 2 /s ( E > 1 MeV) at ∼ 320 and 373 K toa fluence level of ∼ 1.5 × 10 24 n/m 2 ( E > 1 MeV), corresponding to ∼ 0.16 dpa (NRT). The microstructure of these materials was investigated before and after irradiation. Both unirradiated and irradiated specimens of the two alloys were tensile tested at ∼ 295 and 373 K. The fracture surfaces of both alloys in unirradiated as well as irradiated conditions were investigated by scanning electron microscopy. The main conclusion of the present investigation is that even this low dose irradiation causes a tremendous increase in tensile strength and a drastic decrease in elongation.

Gas release processes for high concentrations of helium bubbles in metals

Abstract Although many aspects of helium behaviour in metals relevant to fusion reactor technology have been studied, relatively little information is available on the kinetics of high temperature release of helium precipitated into bubbles. Recently, based on the many observations showing that bubble coarsening during annealing is induced initially in regions nearest thermal vacancy sources (e.g. grain boundaries or surfaces), a qualitative model has been suggested in which the vacancy gradient between the vacancy source and initially overpressurised bubble concentrations must lead to directed bubble migration up the gradient towards the source. It is clear that this gives a potential for gas release, either directly for surfaces, or indirectly for grain boundaries. The present paper extends work on this mechanism by discussing quantitative aspects of bubble movement in vacancy gradients and using computer simulations to provide information on the parameters of importance in the gas release mechanism. Among these are the helium content and the associated local swelling, while release kinetics and temperature are primarily controlled by self-diffusion parameters.

The role of directed bubble diffusion to grain boundaries in post-irradiation fission gas release from UO2: a quantitative assessment

Abstract In the analysis of post-irradiation fission gas release experiments on UO 2 , there have been difficulties in understanding the processes involved in the important step of intragranular fission gas transfer to the grain boundaries. Recently, a new model has been introduced, based on the strong directed bubble migration that must occur in the vacancy gradient induced between the grain boundary and overpressurised bubble concentrations within a grain during annealing. The present paper extends this work significantly by using numerical calculations to simulate these processes. This approach has allowed quantitative predictions to be made within the model, showing that large effects are possible, and providing information on the parameters of importance. Among these parameters are the accumulated fission gas content and the induced swelling adjacent to grain boundaries, while gas release kinetics are predominantly controlled by self-diffusion parameters. In addition, the paper discusses several examples in which there is good agreement between the results of the model and experimental data taken from the literature.

Inert gas release from metals and UO2 during high temperature annealing: the role of thermal vacancies

Abstract It has recently been suggested that if two well established phenomena - that of thermal vacancy production at grain boundaries or surfaces, together with bubble migration up vacancy gradients - were combined, then the large resulting enhancement of bubble diffusion toward the grain boundaries could explain a long-standing difficulty in understanding fission gas release in UO 2 during high temperature annealing. Although this paper partly reiterates this new approach, and includes its application to a set of recent results, the main purpose is to extend the approach to results on inert gas behaviour in metals where the coincidence of gas release and swelling during annealing corroborate the idea that thermal vacancies play an essential and important role in gas release. Of further importance are new results for silicon where the lack of krypton release can be attributed to the low thermal vacancy concentrations in this material.

Post-irradiation fission gas release from high burn-up UO2 fuel annealed under oxidising conditions

Abstract In post-irradiation fission gas release experiments on UO 2 fuel, large increases in effective diffusion coefficients have been reported when anneals have been carried out under oxidising conditions. These conditions affect in some way the processes involved in the diffusive phase of gas release associated with the important step of intragranular fission gas transfer to the grain boundaries. The purpose of this paper is to extend a recent model involving fission gas bubble movement to propose a possible explanation. A central part of the model concerns the importance of grain boundaries as a dominant vacancy source. Under normal conditions, boundary concentrations are due to thermal vacancy production but under oxidising conditions, it is suggested that if fast oxygen diffusion takes place along grain boundaries then the subsequent oxidation reaction at the boundary has the potential to cause a huge enhancement of the local vacancy level. As will be demonstrated, an acceleration in the rate of fission gas bubble movement to the grain boundaries will result. A comparison of model calculations with a few selected literature results suggests that the mechanism has more than sufficient strength to give the large effects observed.

Helium desorption studies on vanadium and V5Ti and V3Ti1Si alloys and their relevance to helium embrittlement

Abstract This paper investigates the trapping of helium in pure vanadium and Vue5f85Ti, Vue5f83Tiue5f81Si alloys using thermal helium desorption spectroscopy (THDS). The implantation of helium has been carried out with energies varying from 50 eV subthreshold implantation up to 3 keV with irradiation temperatures between 300 and 1000 K. On a separate set of samples of the same alloys, tensile measurements were performed at 773, 873, and 973 K. The samples used for the tensile measurements were pre-irradiated with neutrons up to 6.4 dpa. Helium was injected by cyclotron irradiation. The differences observed in the desorption spectra for pure vanadium and the alloys after different irradiation conditions are discussed with respect to their mechanical properties. The role of interstitially dissolved impurities in helium-vacancy clustering and helium trapping was investigated both experimentally and by Monte-Carlo simulations of clustering processes. By matching the calculation results to the THDS results the interstitially dissolved fraction of impurities is estimated.

Comments on “Behaviour of inert gas bubbles under chemical concentration gradients” by G.P. Tiwari

Abstract The motion of inert gas bubbles induced by thermal vacancy gradients has previously been used by the present authors to understand gas bubble release in UO 2 and metals. This approach has been recently questioned by Tiwari. In the present letter, a critical discussion of his viewpoint is presented, together with an analysis of the important experimental results of Marachov et al. There appears to be good evidence for the disputed effect.