Assessment of Initial Test Conditions for Experiments to Assess Irradiation Assisted Stress Corrosion Cracking Mechanisms

Irradiation-assisted stress corrosion cracking is a key materials degradation issue in today s nuclear power reactor fleet and affects critical structural components within the reactor core. The effects of increased exposure to irradiation, stress, and/or coolant can substantially increase susceptibility to stress-corrosion cracking of austenitic steels in high-temperature water environments. . Despite 30 years of experience, the underlying mechanisms of IASCC are unknown. Extended service conditions will increase the exposure to irradiation, stress, and corrosive environment for all core internal components. The objective of this effort within the Light Water Reactor Sustainability program is to evaluate the response and mechanisms of IASCC in austenitic stainless steels with single variable experiments. A series of high-value irradiated specimens has been acquired from the past international research programs, providing a valuable opportunity to examine the mechanisms of IASCC. This batch of irradiated specimens has been received and inventoried. In addition, visual examination and sample cleaning has been completed. Microhardness testing has been performed on these specimens. All samples show evidence of hardening, as expected, although the degree of hardening has saturated and no trend with dose is observed. Further, the change in hardening can be converted to changes in mechanical properties. The calculated yield stress is consistent with previous data from light water reactor conditions. In addition, some evidence of changes in deformation mode was identified via examination of the microhardness indents. This analysis may provide further insights into the deformation mode under larger scale tests. Finally, swelling analysis was performed using immersion density methods. Most alloys showed some evidence of swelling, consistent with the expected trends for this class of alloy. The Hf-doped alloy showed densification rather than swelling. This observation may be related to the formation of second-phases under irradiation, although further examination is require

Assessment of Initial Test Conditions for Experiments to Assess Irradiation Assisted Stress Corrosion Cracking Mechanisms

Irradiation-assisted stress corrosion cracking is a key materials degradation issue in today s nuclear power reactor fleet and affects critical structural components within the reactor core. The effects of increased exposure to irradiation, stress, and/or coolant can substantially increase susceptibility to stress-corrosion cracking of austenitic steels in high-temperature water environments. . Despite 30 years of experience, the underlying mechanisms of IASCC are unknown. Extended service conditions will increase the exposure to irradiation, stress, and corrosive environment for all core internal components. The objective of this effort within the Light Water Reactor Sustainability program is to evaluate the response and mechanisms of IASCC in austenitic stainless steels with single variable experiments. A series of high-value irradiated specimens has been acquired from the past international research programs, providing a valuable opportunity to examine the mechanisms of IASCC. This batch of irradiated specimens has been received and inventoried. In addition, visual examination and sample cleaning has been completed. Microhardness testing has been performed on these specimens. All samples show evidence of hardening, as expected, although the degree of hardening has saturated and no trend with dose is observed. Further, the change in hardening can be converted to changes in mechanical properties. The calculated yield stress is consistent with previous data from light water reactor conditions. In addition, some evidence of changes in deformation mode was identified via examination of the microhardness indents. This analysis may provide further insights into the deformation mode under larger scale tests. Finally, swelling analysis was performed using immersion density methods. Most alloys showed some evidence of swelling, consistent with the expected trends for this class of alloy. The Hf-doped alloy showed densification rather than swelling. This observation may be related to the formation of second-phases under irradiation, although further examination is require

In situ measurement and modelling of the growth and length scale of twins in α -uranium

Twinning is an important deformation mechanism in crystalline materials. Twins nucleate, then grow, forming a discrete pattern of deformation bands with a certain thickness. The mechanism for twin growth is not completely understood. We present a simple phase field model which captures the physics of twin growth and reproduces the details observed during in situ electron backscatter diffraction experiments on α -uranium. The interaction between residual dislocations at twin interfaces and mobile dislocations in untwinned regions increases the stress needed for twin growth. This is described by a nonlocal term in the proposed constitutive equations: the nucleation stress of a twin increases proportionally to the twin phase field in a neighborhood. Competition between slip and twinning favors the nucleation of a new twin rather than the growth of a preexisting twin. The phase field model is coupled with a crystal plasticity finite element solver that includes the plastic deformation due to twinning. Simulations are able to reproduce the number of twins and their thicknesses as a function of the strain observed during in situ electron backscatter diffraction experiments. The stress-strain curve is also reproduced. This comparison allows the values of the nucleation stress and the interaction strength between residual and mobile dislocations to be found. It also gives an estimation of the density of residual dislocations that is consistent with the observed twin thickness. This model can be applied to understand microstructural effects in materials using twinning as a strengthening mechanism

Revealing How Alkali Cations Affect the Surface Reactivity of Stainless Steel in Alkaline Aqueous Environments.

Stainless steel is a ubiquitous structural material and one that finds extensive use in core-internal components in nuclear power plants. Stainless steel features superior corrosion resistance (e.g., as compared to ordinary steel) due to the formation of passivating iron and/or chromium oxides on its surfaces. However, the breakdown of such passivating oxide films, e.g., due to localized deformation and slip line formation following exposure to radiation, or aggressive ions renders stainless steel susceptible to corrosion-related degradation. Herein, the effects of alkali cations (i.e., K+, Li+) and the interactions between the passivated steel surface and the solution are examined using 304L stainless steel. Scanning electrochemical microscopy and atomic force microscopy are used to examine the inert-to-reactive transition of the steel surface both in the native state and in the presence of applied potentials. Careful analysis of interaction forces, in solution, within ≤10 nm of the steel surface, reveals that the interaction between the hydrated alkali cations and the substrate affects the structure of the electrical double layer (EDL). As a result, a higher surface reactivity is indicated in the presence of Li+ relative to K+ due to the effects of the former species in disrupting the EDL. These findings provide new insights into the role of the water chemistry not only on affecting metallic corrosion but also in other applications, such as batteries and electrochemical devices

Revealing How Alkali Cations Affect the Surface Reactivity of Stainless Steel in Alkaline Aqueous Environments

Stainless steel is a ubiquitous structural material and one that finds extensive use in core-internal components in nuclear power plants. Stainless steel features superior corrosion resistance (e.g., as compared to ordinary steel) due to the formation of passivating iron and/or chromium oxides on its surfaces. However, the breakdown of such passivating oxide films, e.g., due to localized deformation and slip line formation following exposure to radiation, or aggressive ions renders stainless steel susceptible to corrosion-related degradation. Herein, the effects of alkali cations (i.e., K+, Li+) and the interactions between the passivated steel surface and the solution are examined using 304L stainless steel. Scanning electrochemical microscopy and atomic force microscopy are used to examine the inert-to-reactive transition of the steel surface both in the native state and in the presence of applied potentials. Careful analysis of interaction forces, in solution, within ≤10 nm of the steel surface, reveals that the interaction between the hydrated alkali cations and the substrate affects the structure of the electrical double layer (EDL). As a result, a higher surface reactivity is indicated in the presence of Li+ relative to K+ due to the effects of the former species in disrupting the EDL. These findings provide new insights into the role of the water chemistry not only on affecting metallic corrosion but also in other applications, such as batteries and electrochemical devices

Insights from Microstructure and Mechanical Property Comparisons of Three Pilgered Ferritic ODS Tubes

International audienceThree oxide dispersion strengthened alloys were fabricated into thin-walled (~500 µm wall thickness) tubes and characterized using x-ray, electron microscopy, and atom probe tomography methods. The three iron-based alloys included the 14%Cr alloy 14WYT, the 12%Cr alloy OFRAC, and a 10%Cr-6%Al alloy CrAZY. Each tube was subjected to a different thermal history during the pilgering process, which allowed for a detailed comparison between varying grain structures and alloy compositions. Atom probe tomography and energy-filtered transmission electron microscopy (TEM) comparisons showed good agreement in precipitate distributions, which matched predicted values using state-of-the-art nanoprecipitate coarsening models. The grain size, precipitate dispersion characteristics, and dislocation densities were then used to estimate yield strengths that were compared against room temperature axial and ring-pull tensile test data. For all three alloys, axial tensile specimens exhibited high tensile strength (>1 GPa) and reasonable plastic strains (10-17%). Ring tensile specimens, conversely, showed limited ductility (~1%) with similar strengths to those measured in the axial orientation. The strengthening models showed mixed agreement with experimentally measured values due to the highly anisotropic microstructures of all three ODS tubes. These results illustrate the need for future model optimization to accommodate non-isotropic microstructures associated with components processed using rolling/pilgering approaches