Peter Trampus

Low Cycle Fatigue Behavior of VVER-440 Reactor Pressure Vessel Steels at Isothermal Condition

The fatigue life of the structural materials 15H2MFA and 08H18N10T of VVER-440 reactor pressure vessel under completely reserved total strain controlled tests were investigated. The measurements were carried out with isothermal condition at 260°C on GLEEBLE 3800 servo-hydraulic thermal mechanical simulator. The isothermal low cycle fatigue results were evaluated with the plastic strain based Coffin-Manson law, and plastic strain energy based model as well. It was shown that both methods are able to predict the fatigue life of reactor pressure vessel steels accurately.

Investigation of Local Corrosion Degradation Developed on a Pipeline System in Service Period

Corrosion degradation was observed in a nuclear power plant spent fuel cooling system. A systematic and comprehensive investigation program was developed which was negatively influenced by the limit of sampling (contaminated material). Corrosion tests, mechanical and microstructural investigations were carried out and also microbiological effect was examined. Major contributors to the degradation were identified.

Acoustic Barkhausen Effect Observed in Various Steels

Carrying out fatigue testing of reactor vessel material 15H2MFA acoustic emission sensors were applied to follow changes. It is shown, that observed bursts can be explained only with appearance of acoustic Barkhausen Effect (ABE). Interesting source localization is shown during heat treatment and consecutive stress test, which can be explained acoustic emission due to material transition from martenzit phase to bainite phase. Observed ABE opens the way to apply it in industry using magnetic stresses to provoke acoustic response for characterization of the state of the magnetic materials.

Micro Structural Aspects of Unstable Crack Propagation in Ferritic Steels

Fracture of ferritic reactor pressure vessel steels is determined by temperature and specimen size. In low temperature fracture occurs by cleavage, with elevation of temperature and decrease of specimen size the ratio of ductile tearing prior to cleavage increases and, in the end, fracture will have a ductile character. There is a temperature interval where both fracture mechanisms are present. Hypothesis of the traditional (global) assessment method has been that the shift of the fracture toughness reference curves due to neutron irradiation is equal with the shift of the ductile-to-brittle transition temperature measured by the Charpy impact test. Local assessment approaches based on the „weakest link” statistical theory recommend the direct measurement of fracture mechanics parameters to structural integrity evaluation. The paper describes the main features of both global and local approaches, analyzing the micro structural aspects of the most known micromechanical models of local approach and drawing conclusions concerning their applicability.

Reactor Pressure Vessel Integrity in Light of the Evolution of Materials Science and Engineering

Structural integrity of the reactor pressure vessel of pressurized water reactors is one of the key safety issues in nuclear power operation. Integrity may be jeopardized during operational transients. The problem is compounded by radiation damage of the vessel structural materials. Structural integrity assessment as an interdisciplinary field is primarily based on materials science and fracture mechanics. The paper gives an overview on the service induced damage processes and associated changes of mechanical properties, the prediction of degradation and the assessment of the entire component against brittle fracture with a special focus on how the evolution of materials science and engineering has contributed to reactor vessel structural integrity assessment.