Abstract
The reduction of grain size down to several tens or hundreds of nanometers leads to the enhancement of radiation resistance of metals. Based on this approach, the aim of the Labex EMC3 (Energy Materials and Clean Combustion Center) project "Naninox" is (1) to study the stability of the microstructure of a nanostructured 316 stainless steel under ion irradiation and (2) to link between this microstructure and the properties (corrosion resistance and the microhardness) of the steel (thanks to a better irradiation resistance, a better corrosion resistance and higher mechanical properties after irradiation are expected in the ultra-fine grained stainless steel). Ultrafine grained 316L austenitic stainless steel samples have been produced by high pressure torsion (HPT) at 430°C and then ion irradiated in Jannus facilities (CEA Saclay) at 450°C and 5 displacements per atoms (dpa). Their microstructure is characterized before and after irradiation by atom probe tomography, X-ray diffraction and transmission electron microscopy. Corrosion behavior in NaCl solution is tested and nano-indentation tests are performed. The first results obtained by atom probe tomography described in this paper indicate that the microstructure of ultrafine grain 316 austenitic stainless steel is more stable under irradiation than the microstructure of a coarse grain 316 austenitic stainless steel.
Highlights
Austenitic 316L type stainless steel is used as baffle bolts to join the baffle plates and former frames in the pressurized water reactor (PWR) internal structures
From the atom probe tomography studies of the annealed sample, the grain boundaries were found to be enriched with Cr, Si, P, C and Mo
Cr levels at various grain boundaries range from 23 to 30%. One such Atom probe tomography (APT) reconstruction is shown in figure 1a, which shows the level of grain boundary enrichment of these elements
Summary
Austenitic 316L type stainless steel is used as baffle bolts to join the baffle plates and former frames in the pressurized water reactor (PWR) internal structures. These bolts undergo severe environmental conditions of stress, temperature and irradiation during the service. During periodic maintenance of the reactor internal structures, it is occasionally observed that some of these bolts have developed cracks due to the irradiation assisted stress corrosion cracking (IASCC) This degradation process is assisted by the radiation induced segregation or depletion of solutes at various defect structures including grain boundaries, dislocations, etc.
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