Numerical simulation of void nucleation in irradiated metals

An atomistics-based theory for void nucleation has been used to calculate — for the first time — terminal void number densities for irradiated nickel, type 316 stainless steel, and the Ni-base alloy, PE-16. Both the absolute magnitudes and temperature dependences of the void number densities are in agreement with experiment. The void nucleation parameter, ψ, which governs spontaneous void nucleation was evaluated for the three materials; the results are in agreement with experiment. The critical gas content for rapid void growth was calculated for PE-16 and type 316 stainless steel, and was found to increase from about 10 helium atoms at the lower end of the void swelling range to some 10⁴ atoms at the upper end. The theory was also found to predict re-nucleation of a new distribution of voids after a drop in temperature during irradiation.     

Void nucleation, growth, and coalescence in irradiated metals

A novel computational treatment of dense, stiff, coupled reaction rate equations is introduced to study the nucleation, growth, and possible coalescence of cavities during neutron irradiation of metals. Radiation damage is modeled by the creation of Frenkel pair defects and helium impurity atoms. A multi-dimensional cluster size distribution function allows independent evolution of the vacancy and helium content of cavities, distinguishing voids and bubbles. A model with sessile cavities and no cluster–cluster coalescence can result in a bimodal final cavity size distribution with coexistence of small, high-pressure bubbles and large, low-pressure voids. A model that includes unhindered cavity diffusion and coalescence ultimately removes the small helium bubbles from the system, leaving only large voids. The terminal void density is also reduced and the incubation period and terminal swelling rate can be greatly altered by cavity coalescence. Temperature-dependent trapping of voids/bubbles by precipitates and alterations in void surface diffusion from adsorbed impurities and internal gas pressure may give rise to intermediate swelling behavior through their effects on cavity mobility and coalescence.     

Shielding calculations for ships carrying irradiated nuclear fuel

A number of ships have been constructed to carry irradiated fuel from Japan to the U.K. and France, for reprocessing. About 20 transport flasks may be carried on each voyage. Permanent shielding must be provided on the ships to ensure that no member of the crew receives an annual dose greater than a specified limit. As the fuel is of varying type and radiation history, and as flasks of differing designs are used, many shielding calculations are needed.There are a number of difficulties in making shielding calculations for the ships. The geometry is complex, dimensions are large and considerable air spaces are involved. The paper considers possible methods of calculation. The line-of-sight method is chosen for most of the calculations, for both γ-radiation and neutrons. The basic data which is used in the calculations is described.As the methods of calculation are somewhat approximate, it is necessary to provide confirmation that they are sufficiently accurate. Validation has been provided in two ways. First, measurements have been made on board one of the ships, Pacific Crane, and these have been checked against calculation. Second, a simplified model of the flasks and ship has been set up, and calculations checked against more sophisticated methods. Results of the validation checks are presented, and it is shown that adequate accuracy is achieved.     

Isochronal recovery of fast neutron irradiated metals

The isochronal recovery of Al, Ni, Cu, Ag. Au, Pt, Fe, Mo and Co, irradiated at 4.5K with fast neutrons has been studied resistometrically. For the face-centered cubic elements, Al, Ni, Cu, Ag, and Au, the magnitude of stage I recovery is inversely proportional to the atomic mass of these elements, which is in excellent agreement with recent calculations on the spatial distribution of lattice defects in fast neutron irradiated metals. The total irradiation-induced resistivity and hardness of neutron spectra have more of an influence on the details of the recovery stages than does the initial residual resistivity and the initial resistivity ratio. Stage II recovery, at least in Cu, can be explained best by the rearrangement and/or annihilation of defects that are released from the stress fields of the vacancy clusters that exist in fast neutron irradiated metals.     

Percolative mass transport in heavily irradiated metals

Computer simulated dark-field electron micrographs at atomic resolutions have been generated by calculating the diffuse elastic scattering distribution of short range order objects with the important point being that images are formed from regions of reciprocal space that do not contain Bragg reflections of the perfect crystal. Interpretation of these images in terms of atom positions and atom correlations becomes straightforward and it is a simple matter to distinguish between real structural information and image artifacts produced by the phase contrast mechanism in the electron optical imaging process. In this paper images were calculated under a variety of microscope conditions for a [111] split crowdion interstitial in tungsten which included up to 182 atoms of the surrounding strain field. The effect of specimen orientation, microscope objective lens defocus and the contribution of atoms lying in different shells around the defect have been considered. To aid in image interpretation accompanying diffraction patterns have been computed for different specimen orientations which show either the perfect crystal Bragg diffraction pattern or the diffuse scattering distribution produced by the crowdion defect.     

Positron lifetime calculations of defects in fusion irradiated beryllium

Numerical quantum-mechanical positron lifetime calculations were performed for mono-vacancies, di-vacancies, tri-vacancies and small nano-voids containing helium and hydrogen in neutron irradiated beryllium. Helium and hydrogen atoms in the sample after the irradiation are considered as atoms forming interstitial O-type loops. Spherical clusters of vacancies are included in the calculations as a reference. It was found that the presence of He and H atoms significantly changes the positron lifetime in irradiated beryllium. A correlation between the positron lifetime and mutual position of vacancies in nano-voids and interstitial loops was established.     

The nucleation and early growth of interstitial dislocation loops in irradiated materials

A hierarchy of rate equations is solved to describe the homogeneous nucleation of interstitial dislocation loops in irradiated materials. Calculations for graphite and M316 stainless steel have been performed. The concept of nucleation time is examined and a procedure is adopted which gives a useful criterion for defining the end of the nucleation period. Calculations have been performed which demonstrate the effects of temperature, dose rate and network-dislocation density on the nucleation and final concentration of interstitial loops. The assumption that di-interstitial atom pairs are stable against thermal dissociation is examined and found to be appropriate for the conditions used in this work.     

On the low-temperature nucleation and growth of bubbles by helium bombardment of metals

A model for the nucleation and growth of helium bubbles at low temperatures is described. The model characteristics are based on recently published results of helium desorption experiments and on atomistic calculations, which indicate that mass transport is achieved by punching out and migration of self interstitials. Simulations are performed up to a helium content of 30 at % resulting in the prediction of the bubble density, bubble volume, fraction of helium in bubbles and point defect concentration. The results show good agreement with experiment and a possible mechanism of blistering is discussed.     

The influence of pre-injected helium on void nucleation and growth in FV548 steel irradiated with 1 MeV electrons

A study has been made of the effect of helium pre-injection on the void populations produced in solution-treated specimens of FV548 austenitic steel irradiated with 1 MeV electrons in the temperature range 450–650°C. The results suggest that helium atom clusters, created during pre-injection, act as void nuclei during subsequent irradiation, although the helium cluster-void transition is limited at high irradiation temperatures by a rapid temperature-dependent increase in the critical void nucleus size. In most cases the void swelling rates at 600 and 650°C decreased with increasing void number density. This was not the case in those specimens in which helium had been pre-injected at elevated temperatures, suggesting that void swelling is an extremely sensitive function of the balance between the void and dislocation sink strengths.     

X-ray microtomography analysis and Molecular Dynamics calculations to understand the dynamic damage process in metals

Dynamic ductile damage has been created by a laser driven shock. The study of these processes consists of predicting it and analysing spatial distributions of pores observed in the target. The challenge is to use the Molecular Dynamics to calculate the propagation of shock waves in ultra thin sheets of metal as well as to compare results with ESRF’s X-ray microtomography.     

Pore nucleation

Conclusions The analysis carried out here allows the problem of pore nucleation in irradiated materials to be divided into two: one problem on the nucleation of pores of minimum size which depends on the structure of the material and the form of the radiation and another on the growth of these pores to a size which depends on the properties of the supersaturated solid solutions of vacancies and interstitials and at which the pores become viable.In the one-component case S(x) has a quite pronounced maximum which makes it easy to calculate the nucleation rate by using Eq. (7) [1, 13]. In the case under consideration the behavior of S(x) in the vicinity of the maximum leads to condition (15) within the framework of which the nucleation rate can be calculated. The analytic expressions obtained for S(x_c) permit the conclusion that interstitials have an exceptionally strong effect on the nucleation rate of viable pores.Translated from Atomnaya Énergiya, Vol. 45, No. 4, pp. 276–280, October, 1978.     

Impact of nuclear irradiation on helium bubble nucleation at interfaces in liquid metals coupled to permeation through stainless steels

The impact of nucleating gas bubbles in the form of a dispersed gas phase on hydrogen isotope permeation at interfaces between liquid metals, like LLE, and structural materials, like stainless steel, has been studied. Liquid metal to structural material interfaces involving surfaces, may lower the nucleation barrier promoting bubble nucleation at active sites. Hence, hydrogen isotope absorption into gas bubbles modelling and control at interfaces may have a capital importance regarding design, operation and safety.He bubbles as a permeation barrier principle is analysed showing a significant impact on hydrogen isotope permeation, which may have a significant effect on liquid metal systems, e.g., tritium extraction systems. Liquid metals like LLE under nuclear irradiation in, e.g., breeding blankets of a nuclear fusion reactor would generate tritium which is to be extracted and recirculated as fuel. At the same time that tritium is bred, helium is also generated and may precipitate in the form of nano bubbles.Phenomena modelling is exposed and implemented in openFROM^® CFD tool for 0D to 3D simulations. Results for a 1D case show the impact of a He dispersed phase of nano bubbles on hydrogen isotopes permeation at an interface. In addition, a simple permeator simulation, consisting in a straight 3D pipe is exposed showing the effect of a He dispersed gas phase on hydrogen isotope permeation through different stainless steels. Results show the permeation reduction as a function of the interface area covered by He bubbles.Our work highlights the effect of gas bubble nucleation at interfaces and the importance of controlling these phenomena in nuclear technology applications.