The migration of single interstitials and small interstitial clusters in ox-zirconium at different temperatures has been analysed using a molecular dynamics model with a many-body interatomic potential. The migration exhibits a strong anisotropy. The defects are very mobile (with Em = 0. 01 eV) along <1120> directions in the basal plane, and this motion is dominant for the single interstitial at low temperature and the di- and tri-interstitials at all temperatures. Above about 500 K, the single interstitial exhibits 2-D and 3-D motion, but Em for non-basal motion is about 0.133 eV. These results point to important consequences for the behaviour of defects formed by displacement cascades in irradiated zirconium.

The damage produced in α-zirconium at 100 K by displacement cascades with energy up to 20 keV has been investigated by MD simulations. In agreement with modelling of fcc and bcc metals, the defect production efficiency in zirconium is well below the NRT estimate. The number and size of clusters, both vacancy and interstitial, are increased by increasing PKA energy, and clusters containing up to 25 interstitials and 30 vacancies were formed by 20 keV cascades. Most interstitial clusters have dislocation character with perfect Burgers vectors of the form 1/3<1120>, but a few metastable clusters are formed and are persistent over the timescale of MD simulations. Collapse of the 30-vacancy cluster to a faulted loop on the prism plane was found to occur over a period of more than 100 ps

Whenever an alloy changes composition because of metal to metal transmutations, as is the case in Ag-In-Cd alloys undergoing (n, −) actions in the control rods of Pressurised Water Reactors, unique features may develop.

A simple model shows that in a two components alloy more than two phases can coexist under appropriate irradiation conditions i.e. temperature, which scales diffusivities and solubilities, and flux, which scales the transmutations rate; because second phase precipitation occurs at the grain boundaries, the grain size in the underlying matrix is found to be one important parameter of the process.

The above ideas are illustrated on a multicomponent alloy, the thermodynamics and the kinetics of which are treated at the same level of sophistication in a mean field approximation.

Gas desorption and positron annihilation techniques have been employed to study the evolution of nitrogen associated defects in nitrogen irradiated metals: Fe, Ni, Mo and W. Nitrogen in these metals has a rather high affinity to vacancy type defects. The results obtained for low irradiation dose show that substitutional nitrogen (NV; with V=vacancy) is formed. The nitrogen vacancy complex dissociates at temperatures ranging from 350 K for Ni to 900K for Mo and 1100 K for W. At high doses defects are formed which can be characterized as nitrogen saturated vacancy clusters. These defects, as observed by helium probing, disappear during annealing for nickel at 800K, and for Mo at 1100 K. The direct observation of the desorbing nitrogen for nickel and molybdenum reveals a very fast desorption transient at the dissociation temperature of the clusters. This is the characteristic desorption transient of a small nitride cluster, e.g., by shrinkage with constant rate. For iron the nitrogen desorption is complicated because of a general background that continuously rises with temperature. With the positron beam technique depth information was obtained for defects in iron and the defect character could be established with the help of the information provided on annihilation with conduction and core electrons of the defect trapped positrons.

SEM observations have shown that irradiation induced interaction of the aluminum cladding with uranium silicide reactor fuels strongly affects both fission gas and fuel swelling behaviors during fuel burn-up. We have used ion beam mixing, by 1.5 MeV Kr, to study this phenomena. RBS and the 27 A1( p, γ) 28 Si resonance nuclear reaction to was used to measure radiation induced mixing of Al into U3Si and U3Si2 after irradiation at 300γ;C.

Initially U mixes into the Al layer and Al mixes into the U3 Si. At a low doses, the Al layer is converted into Ual4 type compound while near the interface the phase U(Al93 Si. 07 )3 grows. Under irradiation, Al diffuses out of the Ual4 surface layer, and the lower density ternary, which is stable under irradiation, is the final product. Al mixing into U3 Si2 is slower than in U3 Si, but after high dose irradiation the Al concentration extends much father into the bulk. In both systems Al mixing and diffusion is controlled by phase formation and growth. The Al mixing rates into the two alloys are similar to that of Al into pure uranium where similar aluminide phases are formed.

Vapor-deposited Zr-Fe multilayered thin films with various wavelengths and of overall composition either 50% Fe or Fe-rich up to 57 % Fe were either irradiated with 300 keV Kr ions at temperatures from 25K to 623 K to fluences up to 2 × 1016 cm−2, or simply annealed at 773K in-situ in the Intermediate Voltage Electron Microscope at Argonne National Laboratory. Under irradiation, the final reaction product is the amorphous phase in all cases studied, but the dose to amorphization depends on the temperature and on the wavelength. In the purely thermal case (annealing at 773 K), the 50–50 composition produces the amorphous phase but for the Fe-rich multilayers the reaction products depend on the multilayer wavelength. For small wavelength, the amorphous phase is still formed, but at large wavelength the Zr-Fe crystalline intermetallic compounds appear. These results are discussed in terms of existing models of irradiation kinetics and phase selection during solid state reaction.

The evolution of the microstructure of solution annealed austenitic stainless steels irradiated by neutrons at temperatures less than 300°C is dominated by black spots and Frank loops. This leads to hardening and a degradation of ductility and work hardening accompanied by a loss of fracture toughness. Microscopic flow localization is attributed to the annihilation of loops by their interaction with moving dislocations. The resulting loss of work hardening can lead to flow localization on a macroscopic scale as well in the vicinity of notches, holes and cracks. The combination of macro and microscopic flow localization may have important implications to the failure modes available to structures and to irradiation-assisted stress corrosion cracking.

Microstructural evolution and deformation behavior of austenitic stainless steels are evaluated for neutron, heavy-ion and proton irradiated materials. Radiation hardening in austenitic stainless steels is shown to result from the evolution of small interstitial dislocation loops during lightwater-reactor (LWR) irradiation. Available data on stainless steels irradiated under LWR conditions have been analyzed and microstructural characteristics assessed for the critical fluence range (0.5 to 10 dpa) where irradiation-assisted stress corrosion cracking susceptibility is observed. Heavy-ion and proton irradiations are used to produce similar defect microstructures enabling the investigation of hardening and deformation mechanisms. Scanning electron, atomic force and transmission electron microscopies are employed to examine tensile test strain rate and temperature effects on deformation characteristics. Dislocation loop microstructures are found to promote inhomogeneous planar deformation within the matrix and regularly spaced steps at the surface during plastic deformation. Twinning is the dominant deformation mechanism at rapid strain rates and at low temperatures, while dislocation channeling is favored at slower strain rates and at higher temperatures. Both mechanisms produce highly localized deformation and large surface slip steps. Channeling, in particular, is capable of creating extensive dislocation pileups and high stresses at internal grain boundaries which may promote intergranular cracking.

The radiation-induced microstructure of austenitic stainless steel was investigated using proton irradiation. Samples of Fe-20Cr-24Ni and Ni-18Cr-9Fe were irradiated using 3.2 MeV protons at a dose rate of 7×10−6 dpa/s between 300°C and 600°C. The irradiation produced a microstructure consisting of dislocation loops and voids in both alloys. The number density and size of dislocation loops and voids are strong functions of irradiation temperature with the number density decreasing and the size increasing with temperature for both defects. The dose dependence of dislocation loop density saturated around 1.0 dpa for Fe-20Cr-24Ni and 0.3 dpa for Ni-18Cr-9Fe. The microstructure evolution in the nickel base alloy is faster than that in Fe-20Cr-24Ni. The dose and temperature dependence of proton irradiated Fe-20Cr-24Ni closely follow that of neutron irradiated austenitic stainless steels. The changes in yield strength due to irradiation were estimated from nano indenter hardness measurements and compared to calculation using a dispersed barrier hardening model. Results were in close agreement with each other. Yield strength changes as a function of dose for proton irradiated Fe-20Cr-24Ni are similar to those resulting from neutron irradiation of austenitic stainless steels.

We present experimental results on the microstructure evolution of dual-beam irradiated (300 keV heavy ions plus 15 keV helium ions) ferritic-martensitic steels (Manet, DIN 1.4926, F82H mod). The helium bubble morphology as well as microchemistry of the alloys are investigated by means of transmission electron microscopy and field-ion microscopy with atom probe. The alloys were irradiated to fluences up to 50 dpa and implanted with helium up to a concentration of 1 at.% at the temperatures of 723 K and 773 K. The damage and implantation rates variied from 2.5 · 10−3 dpa/s to 2.5·10−2 dpals and from 0.5 appmls to 5 appm/s, respectively. Size and number density of helium bubbles is found to be rate dependent. Smaller implantation rates produce larger helium bubbles and smaller bubble number densities. Regions of local enrichment of alloy elements, typically 5 nm in size, containing chromium (up to 40 at.%), silicon, and nickel are detected. Number densities of helium bubbles and of regions of chromium enrichments are comparable and lie between 1023/m3 and 1024/m3. Possible extrapolation of the present ion irradiations to spallation source and fusion reactor conditions is shortly adressed.

Nanoscale Cu rich precipitates (CRPs) are widely believed to be the dominant hardening feature resulting in severe embrittlement in irradiated reactor pressure vessel (RPV) steels. However, this view has recently been challenged by interpretations of atom probe field ion microscopy (APFIM) measurements that describe the dominant nanofeatures as dilute solute atmospheres (DSAs). The practical impact of these differing views is very significant. This work compares and contrasts the CRP versus DSA descriptions to a wide variety of pertinent data. Mechanical property trends as well as small angle neutron scattering (SANS) and field emission scanning transmission electron microscopy (FEGSTEM) measurements support the presence of CRPs. CRPs are also consistent with the fundamental thermodynamic and kinetic laws. However, standard theory cannot provide the atomic level resolution needed to fully understand the nanofeatures. Therefore, a new Lattice Monte Carlo (LMC) atomistic method is used to simulate the complex chemical structures of the CRPs. The LMC method unifies the SANS/FEGSTEM and APFIM data within a well founded physical framework.

In this paper we review the advances in understanding ofmicrostructural evolution during irradiation in pressure vessel steels. A particular focus is the benefits that arise from combining microstructural techniques on characterising the microstructure, especially the copper clusters. In addition to experimental techniques the importance of developing microstructurally based models linking the microstructural features to mechanical properties behaviour will be demonstrated. This paper will consider the recent advances in understanding in this area, and highlight the potential benefits that can be gained from a thorough understanding of the materials response to irradiation.

We have used positron annihilation lifetime spectroscopy to study the development of damage and annealing behavior of neutron-irradiated reactor pressure-vessel steels. We irradiated samples of ASTM A508 nuclear reactor pressure-vessel steel to fast neutron 172 fluences of up to 1017 n/cm2, and we examined these samples using positron annihilation lifetime spectroscopy (PALS) to study the effects of neutron damage in the steels on positron lifetimes. Non-irradiated samples show two positron lifetimes: a 110 ps component corresponding to annihilations in the bulk material, and a 165 ps lifetime corresponding to annihilations in dislocation defects. The irradiated samples show an additional lifetime component of 300 ps in the PAL spectra and an increase in the proportion of annihilations with a 165 ps lifetime, suggesting that vacancies and vacancy clusters are present in the material after room temperature irradiation. The samples were then annealed to temperatures ranging from 210° C to 450° C. The positron lifetimes introduced by neutron damage disappear after annealing the samples at 280° C.

The magnetic-to-nuclear scattering intensity ratio observed in the scattering from copper-rich precipitates in irradiated pressure vessel steels is much smaller than the value of 11.4 expected for a pure copper precipitate in iron. A model for precipitates in pressure vessel steels which matches the observed scattering typically incorporates manganese, nickel, silicon and other elements and it is assumed that the precipitate is non-magnetic. In the present work consideration is given to the effect of composition gradients and ferromagnetic penetration into the precipitate on the small angle scattering cross section for copper-rich clusters as distinquished from conventional precipitates. The calculation is an extension of a scattering model for micelles which consist of shells of varying scattering density. A discrepancy between recent SANS scattering experiments on pressure vessel stells was found to be related to applied magnetic field strength. The assumption of cluster structure and its relation to atom probe FIM findings as well as the effects of insufficient field for magnetic saturation is discussed.

The small, coherent BCC precipitates of copper that form during fast neutron irradiation of ferritic steels are an important component of in-service irradiation hardening. Many-body interatomic potentials for the Fe-Cu alloy system have been developed and used to simulate the atomic structure of the ½<111> screw dislocation in both pure a-iron and the metastable BCC phase of copper. In iron, the core has the well-known 3-fold form of atomic disregistry. In BCC copper, however, the core structure depends on the lattice parameter. At the metastable equilibrium value, the core is similar to that in iron, but as the lattice parameter is reduced, as in a precipitate, the core becomes delocalised by transformation of the copper. Simulation of dislocated crystals containing precipitates shows that the extent of this effect depends on precipitate size. The energy changes indicate a significant dislocation pinning effect due to this dislocation-induced precipitate transformation process.

In order to probe the possibility of simulating irradiation effects with swift ions and obtaining some insight about the production rate of freely migrating defects in iron, we studied the kinetics of copper precipitation in FeCu1.34at% irradiated at 300°C with 202 MeV O- ions and 2.5 MeV electrons. The precipitation kinetics are studied in situ (GANIL for O- ions and Van de Graaff for electrons) by measuring the electrical resistivity at 30°C to reduce the phonon component. We found that, in both cases, the resistivity versus fluence (in dpa) curves are identical. This would imply either that the relatively small cascades induced by 202 MeV O- ions (compared with those produced by neutrons) are very efficient in producing freely migrating point defects or that the energy deposition by electronic excitations plays an important role. We show that this latter possibility is not relevant for 202 MeV O18 ions.

In low copper steels for nuclear reactor pressure vessel, point defect clustering seems to play an important role in hardening. In order to study the hardening component which results from the clustering of freely migrating point defects, we irradiated, in a high voltage electron microscope, Fe, the alloys Fe0.13%Cu and Fe0.014%P, alloys and the alloy Fel.5%Mn0.8%Ni0.1%Cu0.01%P, the composition of which is close to the matrix of pressure vessel steels. We studied the nucleation of dislocation loops and their growth velocity. We find out that copper and phosphorus have no effect on the vacancy migration energy but that this parameter decreased significantly in the complex alloy. The main point is certainly that loops are nucleated in the complex model alloy up to 500°C while no loop appears above 300°C in Fe and in FeCu. FeP shows an intermediate behavior.

Two heats of reactor pressure vessel steel type A533-B-1 differing in their Cu content (0.07 and 0.14 wt.-%) were irradiated to a fluence of 1* 1018 neutrons/cm2 at temperatures of 60, 121, and 288 °C, respectively. Their microstructure has been studied by SANS, SAXS, and PAS in the asirradiated and post-irradiation annealed stages and compared to the behaviour of unirradiated material and the results are discussed.

The production of small, coherent BCC precipitates of copper during fast neutron irradiation of ferritic steels and their apparent stability at peak hardening size are important phenomena governing in-service irradiation embrittlement. An MD model employing manybody interatomic potentials for the Fe-Cu alloy system has been developed and used to simulate the effect of displacement cascades on both dilute random solid solutions of copper and small coherent BCC copper precipitates in an oa-iron matrix. Within the timescales that can be modelled, cascades caused only small changes to the distribution of copper atoms in both the solid solution and the precipitate studies. Vacancies are observed to bind to copper solute atoms and also to accumulate in precipitates near the interface, where they are mobile. Interstitials are mobile in the solid solution matrix and two mechanisms of interstitial entrapment by precipitates are indicated.

The microstructural and microchemical evolution of an Alloy X-750 heat under neutron irradiation has been studied in order to understand the origin of irradiation-assisted stress corrosion cracking. Both clustering of point defects and radiation-induced segregation (RIS) at interfaces and boundaries have been observed. Although no significant changes in the precipitate structure were observed, boundaries exhibited additional depletion of Cr and Fe and enrichment of Ni.