Point Defect Cluster Formation in Iron Displacement Cascades Up to 50 keV

The results of molecular dynamics displacement cascade simulations in iron at energies up to 50 keV and temperatures of 100, 600, and 900K are summarized, with a focus on the characterization of interstitial and vacancy clusters that are formed directly within the cascade. The fraction of the surviving point defects contained in clusters, and the size distributions of these in-cascade clusters have been determined. Although the formation of true vacancy clusters appears to be inhibited in iron, a significant degree of vacancy site correlation was observed. These well correlated arrangements of vacancies can be considered nascent clusters, and they have been observed to coalesce during longer term Monte Carlo simulations which permit short range vacancy diffusion. Extensive interstitial clustering was observed. The temperature and cascade energy dependence of the cluster size distributions are discussed in terms of their relevance to microstructural evolution and mechanical property changes in irradiated iron-based alloys

Modeling of Microstructure Evolution in Austenitic Stainless Steels Irradiated Under Light Water Reactor Conditions

A model for the development of microstructure during irradiation in fast reactors has been adapted for light water reactor (LWR) irradiation conditions (275 {approximately} 325 C, up to {approximately}10 dpa). The original model was based on the rate-theory, and included descriptions of the evolution of both dislocation loops and cavities. The model was modified by introducing in-cascade interstitial clustering, a term to account for the dose dependence of this clustering, and mobility of interstitial clusters. The purpose of this work was to understand microstructural development under LWR irradiation with a focus on loop nucleation and saturation of loop density. It was demonstrated that in-cascade interstitial clustering dominates loop nucleation in neutron irradiation in LWRS. Furthermore it was shown that the dose dependence of in-cascade interstitial clustering is needed to account for saturation behavior as commonly observed. Both quasi-steady-state (QSS) and non-steady-state (NSS) solutions to the rate equations were obtained. The difference between QSS and NSS treatments in the calculation of defect concentration is reduced at LWR temperature when in-cascade interstitial clustering dominates loop nucleation. The mobility of interstitial clusters was also investigated and its impact on loop density is to reduce the nucleation term. The ultimate goal of this study is to combine the evolution of microstructure and microchemistry together to account for the radiation damage in austenitic stainless steels

Modeling of Microstructure Evolution in Austenitic Stainless Steels Irradiated Under Light Water Reactor Conditions

A model for the development of microstructure during irradiation in fast reactors has been adapted for light water reactor (LWR) irradiation conditions (275 {approximately} 325 C, up to {approximately}10 dpa). The original model was based on the rate-theory, and included descriptions of the evolution of both dislocation loops and cavities. The model was modified by introducing in-cascade interstitial clustering, a term to account for the dose dependence of this clustering, and mobility of interstitial clusters. The purpose of this work was to understand microstructural development under LWR irradiation with a focus on loop nucleation and saturation of loop density. It was demonstrated that in-cascade interstitial clustering dominates loop nucleation in neutron irradiation in LWRS. Furthermore it was shown that the dose dependence of in-cascade interstitial clustering is needed to account for saturation behavior as commonly observed. Both quasi-steady-state (QSS) and non-steady-state (NSS) solutions to the rate equations were obtained. The difference between QSS and NSS treatments in the calculation of defect concentration is reduced at LWR temperature when in-cascade interstitial clustering dominates loop nucleation. The mobility of interstitial clusters was also investigated and its impact on loop density is to reduce the nucleation term. The ultimate goal of this study is to combine the evolution of microstructure and microchemistry together to account for the radiation damage in austenitic stainless steels

The impact of mobile point defect clusters in a kinetic model of pressure vessel embrittlement

The results of recent molecular dynamics simulations of displacement cascades in iron indicate that small interstitial clusters may have a very low activation energy for migration, and that their migration is 1-dimensional, rather than 3-dimensional. The mobility of these clusters can have a significant impact on the predictions of radiation damage models, particularly at the relatively low temperatures typical of commercial, light water reactor pressure vessels (RPV) and other out-of-core components. A previously-developed kinetic model used to investigate RPV embrittlement has been modified to permit an evaluation of the mobile interstitial clusters. Sink strengths appropriate to both 1- and 3-dimensional motion of the clusters were evaluated. High cluster mobility leads to a reduction in the amount of predicted embrittlement due to interstitial clusters since they are lost to sinks rather than building up in the microstructure. The sensitivity of the predictions to displacement rate also increases. The magnitude of this effect is somewhat reduced if the migration is 1-dimensional since the corresponding sink strengths are lower than those for 3-dimensional diffusion. The cluster mobility can also affect the evolution of copper-rich precipitates in the model since the radiation-enhanced diffusion coefficient increases due to the lower interstitial cluster sink strength. The overall impact of the modifications to the model is discussed in terms of the major irradiation variables and material parameter uncertainties

Microstructural evolution in fast-neutron-irradiated austenitic stainless steels

The present work has focused on the specific problem of fast-neutron-induced radiation damage to austenitic stainless steels. These steels are used as structural materials in current fast fission reactors and are proposed for use in future fusion reactors. Two primary components of the radiation damage are atomic displacements (in units of displacements per atom, or dpa) and the generation of helium by nuclear transmutation reactions. The radiation environment can be characterized by the ratio of helium to displacement production, the so-called He/dpa ratio. Radiation damage is evidenced microscopically by a complex microstructural evolution and macroscopically by density changes and altered mechanical properties. The purpose of this work was to provide additional understanding about mechanisms that determine microstructural evolution in current fast reactor environments and to identify the sensitivity of this evolution to changes in the He/dpa ratio. This latter sensitivity is of interest because the He/dpa ratio in a fusion reactor first wall will be about 30 times that in fast reactor fuel cladding. The approach followed in the present work was to use a combination of theoretical and experimental analysis. The experimental component of the work primarily involved the examination by transmission electron microscopy of specimens of a model austenitic alloy that had been irradiated in the Oak Ridge Research Reactor. A major aspect of the theoretical work was the development of a comprehensive model of microstructural evolution. This included explicit models for the evolution of the major extended defects observed in neutron irradiated steels: cavities, Frank faulted loops and the dislocation network. 340 refs., 95 figs., 18 tabs

From Molecular Dynamics to Kinetic Rate Theory: A Simple Example of Multiscale Modeling

Radiation damage formation in iron has been investigated using the method of molecular dynamics simulation. The MD simulations have been used to determine primary defect production parameters for cascade energies up to 50 keV at temperatures from 100 to 900K. The energy dependence of these parameters has been used to determine appropriate neutron-energy-spectrum averaged damage production cross sections for various irradiation environments. Two applications of these effective cross sections are discussed. The first is an evaluation of neutron energy spectrum effects in commercial fission reactor pressure vessels. The second example deals with the use of these cross sections in the source term of a kinetic model used to predict void swelling and microstructural evolution. The simulation of the primary damage event by MD involves times less than 100 ps and a size scale of a few tens of nm, while the kinetic simulation encompasses several years and macroscopic sizes. This use of the MD results to develop an improved source term for rate theory modeling provides a simple example of multiscale modeling

Microstructural characterization of selected AEA/UCSB model FeCuMn alloys

A set of 22 model ferritic alloys was purchased as part of a collaborative research program by the AEA Harwell Laboratory and the University of California at Santa Barbara. Nine of these alloys were selected by the Oak Ridge National Laboratory for use in a series of ion irradiation experiments investigating dispersed barrier hardening. These nine alloys contain varying amounts of copper, manganese, titanium, carbon, and nitrogen. The alloys have been characterized by transmission electron microscopy in the as-received condition to provide a baseline for comparison with the irradiated specimens. A description of the microstructural observations is provided for future reference. This summary focuses on the type and size distributions of the precipitates present; grain size and dislocation measurements are also included