Department of Structural Materials
We investigate the behavior of materials exposed to energetic particle irradiation. The work contributes to the program NUSAFE (Nuclear Waste Management, Safety and Radiation Research) of the Helmholtz Association.
Neutron irradiation provokes the formation and long-term evolution of nm-scale defects such as dislocation loops and solute atom clusters. These defects give rise to hardening accompanied by a reduced fracture resistance of reactor pressure vessel steels of running nuclear power plants. Materials for advanced reactor concepts will be exposed to higher operation temperatures and higher neutron doses. The overall objectives of our research are to identify the mechanisms of irradiation-induced damage in structural materials and to assess the resulting changes of the mechanical properties.
We work on two main directions:
- In the case of running nuclear power plants, the work is focused on long-term irradiation effects in reactor pressure vessel steels.
- Our work in the field of advanced reactor concepts is dedicated to ferritic/martensitic Cr-steels, oxide dispersion strengthened (ODS) steels and the emerging class of high-entropy alloys.
The new insight substantially contributes to the scientific background for the safety assessment of nuclear reactors. The work is embedded in the Euratom projects SOTERIA, MATISSE and M4F. A close cooperation with the Fundamentals and Simulation Group provides additional insight via atomistic simulation.
- Mechanical testing of irradiated materials
- Characterization at the nm length scale
- Ion irradiation to emulate neutron irradiation effects
Using mini-CT specimens for the fracture characterization of ferritic steels within the ductile to brittle transition range: a review
The use of mini-CT specimens for the fracture characterization of structural steels is currently a topic of great interest from both scientific and technical points of view, mainly driven by the needs and requirements of the nuclear industry. In fact, the long-term operation of nuclear plants requires accurate characterization of the reactor pressure vessel materials and evaluation of the embrittlement caused by neutron irradiation without applying excessive conservatism. However, the amount of material placed inside the surveillance capsules used to characterize the resulting degradation is generally small. Consequently, in order to increase the reliability of fracture toughness measurements and reduce the volume of material needed for the tests, it is necessary to develop innovative characterization techniques, among which the use of mini-CT specimens stands out. In this context, this paper provides a review of the use of mini-CT specimens for the fracture characterization of ferritic steels, with particular emphasis on those used by the nuclear industry. The main results obtained so far, revealing the potential of this technique, together with the main scientific and technical issues will be thoroughly discussed. Recommendations for several key topics for future research are also provided.
Keywords: mini-CT; ductile-to-brittle transition range; reference temperature; master curve
Metals 13(2023), 176
|Name||Bld./Office||+49 351 260|
|Dr. Eberhard Altstadt||801/P151||2276||e.altstadthzdr.de|
|Dr. Cornelia Kaden||801/P102firstname.lastname@example.org, c.heintzehzdr.de|
|Name||Bld./Office||+49 351 260|
|Dr. Frank Bergner||801/P150||3186||f.bergnerhzdr.de|
|Dr. Paul Chekhonin||801/P146||2149||p.chekhoninhzdr.de|
|Dr. Andreas Ulbricht||801/P146||3155||a.ulbrichthzdr.de|