Post Irradiation Tensile Performance of Fe-Cr Base Alloys This proposal describes the use of x-ray diffraction experiments at the Advanced Photon Source to study the evolution of tensile properties of ferritic alloys following irradiation at the ATR. The samples consist of a selection of model, commercial, and developmental FeCr alloys, the primary choice for reactor fuel cladding and structural applications in advanced reactor designs. The samples are in minitensile form, presently undergoing irradiation inside the ATR. Once removed from the ATR the samples will be loaded into interchangeable tensile testing jaws and placed inside a sealed glovebox containing the tesile testing stage. The specially constructed glovebox will allow the tensile stage to be mounted on the MRCAT beamline, maintaining the appropriate geometry for x-ray diffraction, as well providing containment for the irradiated samples. The measurements will utilize a two-dimensional area detector to record the diffraction pattern. This will allow bulk stress and strain components to be measured in situ during the tensile tests, as well as monitoring the development of plasticity-induced voids which lead to flow localization (necking). Diffraction data will be available for analysis as it is collected, and the full data set available for analysis immediately after the conclusion of the beamtime. The period of performance is therefore limited to the constraints of the sample irradiation and beamtime award schedule, with additional time required only for the analysis of experimental data. With a favorable outcome, these experiments could be extended to further radiation doses or other systems of interest. If successful, these studies will complement information obtained from higher dose irradiations in the Phenix fast reactor. These experiments will extend current experimental and modeling activities, including the use of the viscoplastic selfconsistent (VPSC) polycrystal approach to understanding the tensile behavior of alloy systems as a function of irradiation dose, as well as finite element analysis of deformation and structural loads. Ultimately, this will improve our understanding of the deformation mechanisms of critical importance to performance and lifetime prediction of the next generation of nuclear reactor components.

Post Irradiation Tensile Performance of Fe-Cr Base Alloys



National nuclear infrastructure faces two problems which limit expansion of the reactor fleet. First is the sustainability of the current generation of reactors, hampered by a lack of understanding the processes associated with irradiation assisted stress corrosion cracking. Second, next generation reactor and advanced fuel cycle designs call for operating temperatures up to 1000 °C, where hydrogen generation from water becomes thermodynamically favorable. Atomistic models are sought for the study of mechanical properties under irradiation, eventually leading to a choice of materials for the next generation of reactor materials.



Ferritic FeCr alloys are the primary choice for advanced fuel cladding and structural applications. The need to understand deformation and fracture mechanisms in irradiated materials is of critical importance to structural integrity and lifetime prediction of nuclear reactor components. However, a lack of understanding the contributions from structural heterogeneity has limited the development of predictive capabilities. A better understanding of the physical mechanisms responsible for flow localization must be developed.



We seek to utilize x-ray diffraction to study tensile deformation in situ on several model, commercial, and developmental FeCr-base alloys to understand the joint influences of irradiation, temperature, and microstructure on tensile deformation. By monitoring stress and strain components during tensile tests, as well as the development of plasticity-induced voids which lead to flow localization (necking), these experiments will extend current modeling activities, as well as complement ongoing related studies. In doing so, they will provide a major new basis for mechanical properties modeling, and form a well coordinated set of irradiation performance data applicable to multiple alloy systems. All of which are required for constructing a validated model for predicting irradiation performance in reactor materials.