Aydogan, Eda. Effect of interstitial elements on the irradiation response of HT9 tempered ferritic/martensitic steels

Principal Investigator
First Name:
Last Name:
Sabanci University
Assistant Professor
Team Members:
Name: Institution: Expertise: Status:
Stuart Maloy Los Alamos National Laboratory Nuclear Materials, Radiation Effects, Alloys Other
Eda Aydogan Sabanci University Ferritic/martensitic alloys, nanostructured ferritic alloys, transmission electron microscopy Faculty
Experiment Details:
Experiment Title:
Effect of interstitial elements on the irradiation response of HT9 tempered ferritic/martensitic steels
Describe the work that you are proposing in detail. Please include as many specifics as possible (e.g., dose, dose rate, ion energy, types of ions, beam line x-ray energy, irradiation temperature, analysis temperature, atmosphere, etc.):
• Samples o Electropolished 3 mm TEM samples having a thickness less than 100 µm to minimize electromagnetic effects during the experiments. We will have 6-9 HT9 ferritic/martensitic samples having low and high nitrogen contents. • Preparation o TEM samples are prepared from bulk HT9 samples having low and high N content. 3 mm discs are cut by electrical discharge machining and mechanically polished to 50-100 µm thickness. They are then electropolished using a 5% perchloric acid + 95% methanol solution. Detailed dislocation studies will be performed at Los Alamos National Laboratory and Sabanci University before and after the irradiations at Argonne National Laboratory. • Instrument Requested o IVEM-Tandem facility at Argonne National Laboratory. o In-situ TEM/Irradiation will be performed on the TEM samples. • Imaging Conditions o Imaging will be performed with 200 or 300 keV electron beam energy under two-beam TEM imaging conditions, tilted from a known zone axis. Diffraction conditions will be determined before the irradiation. To keep the same diffraction condition, we will take diffraction patterns intermittently. • Irradiation Conditions o The samples will be irradiated with 1 MeV krypton ions. For this purpose, double tilt high temperature holder is requested. The flux requested is about ~5 x 1015 ion.cm-2. Irradiation temperatures will range from 300 °C to 600 °C. • Procedure o 6-9 samples will be studied. Imaging as a function of fluence (for understanding the defect evolution and radiation induced precipitation as a function of ion dose at various temperatures) The sample will be inserted into the TEM. Using bright field and diffraction, a grain being close to a low-index zone axis will be determined. After bringing it to its zone axis, sample will be tilted to a two-beam condition for the in-situ irradiations. For the high temperature irradiations, the video will be captured before irradiation while heating to observe the change in microstructure. Once the specific temperature is reached, irradiation will start. To keep the same diffraction condition throughout irradiation, diffraction patterns will be taken intermittently (after each 0.5 or 1 dpa). Irradiations up to ~10 dpa will be performed. The magnification during irradiations will be ~50kx. To observe the small defects, it will be increased time to time during irradiations. Final dose after 10 dpa should be around ~5 x 1015 ion.cm-2. Irradiations will be conducted at 300 °C, 450 °C and 600 °C on low and high N HT9 alloys. Therefore, total of 6-9 samples are planned to be irradiated. Detailed characterization on the defect and radiation induced particle size and densities after irradiation will be performed at LANL and Sabanci University. Correlation between damage, second phase particle and nitrogen content will be made.
Technical Abstract
There is a worldwide interest in nuclear energy due to the increase in the world’s population and the desire to reduced greenhouse gasses from the burning of fossil fuels. Therefore, fission reactors are being investigated to meet the need for clean sources of energy. However, next generation advanced reactors are expected to operate at extreme conditions such as high temperatures and neutron damages as well as corrosive environments. It has been discovered that the ferritic steels having a bcc structure have higher swelling tolerance along with a lower thermal expansion compared to fcc austenitic steels. Therefore, tempered martensitic steels with bcc structure are one of the best candidates for next generation reactors because of their high defect sinks of submicron size lath structure, smaller dislocation bias and higher self-diffusion coefficient. Still, once they reach the steady state swelling regime, they swell with the rate of 0.2%/dpa. Radiation damage resistance in metals is directly correlated with the alloy composition, therefore microstructure. Initially, interstitial elements were reported to deteriorate radiation resistance of the alloys; however, later, swelling resistance has been reported to be improved with the interstitial content. It is still unknown the effect of interstitials on the radiation response of the materials. In this project, tempered ferritic/martensitic HT9 steels produced in various nitrogen contents will be investigated by in-situ irradiations. Ex-situ irradiations up to ~10 dpa at 300 °C have been conducted at Los Alamos National Laboratory (LANL) to investigate the radiation induced hardening, dislocation loop and radiation induced precipitate formation. Therefore, in order to investigate the evolution of the defects in the deformed materials at the early stages of the irradiation, in-situ ion irradiations will be performed. 3-mm foils for transmission electron microscopy (TEM) will be prepared by electropolishing and initial TEM characterization will be performed at LANL and Sabanci University. In-situ irradiations at the Intermediate Voltage Electron Microscopy (IVEM) will be performed using heavy ion irradiations (Kr) at 300 °C, 450 °C and 600 °C up to ~10 dpa. Further detailed characterization will be performed at LANL and Sabanci University. Ultimately, both ex-situ irradiations and low dose in-situ irradiations will reveal the effect of interstitial elements on the radiation resistance of the materials and defect evolution mechanisms. This research will provide a fundamental understanding on the effect of interstitial content on dislocation loop evolution (size, density, type, etc.) and radiation induced particle (G-phase) formation. Application of this understanding will deliver insights on the optimization of the compositions of the cladding materials for next generation advanced reactors. We ask for 10 days of beam time at IVEM as we have 6 to 9 low and high N alloys which will be irradiated to ~10 dpa at various temperatures. The expected period to run this project is 6 months starting from September 2019.