| Nuclear energy,a high-quality clean energy will play an important role in the future energy consumption structure.The priority on the development of nuclear power technology has been included in the long-term energy development strategy for China’s energy.316 LN stainless steel has been widely used in the third-generation pressurized water reactor and the fourth-generation sodium-cooled fast reactor nuclear power plants due to its excellent high temperature mechanical properties.The structures in nuclear power plants are frequently subjected to repeated mechanical load and cyclic thermal stress due to the occurrence of substantial thermal transients during its long-term service.Consequently,thermomechanical fatigue failure has become one of the most important reasons that threaten the integrity of nuclear power structure.Therefore,a comprehensive and systematic study on the uniaxial/multiaxial high temperature and thermomechanical fatigue properties of the domestic nuclear-grade 316 LN stainless steel with high nitrogen content was carried out based on the self-developed tensiontorsional multiaxial thermomechanical fatigue testing system.Two parameters with clear physical meaning,i.e.number of stress serrations and the maximum stress drop of serrations were proposed,which can be used to evaluate the strength of dynamic strain aging(DSA)quantitatively.The DSA evolution behavior of 316 LN stainless steel with high nitrogen content under high temperature low-cycle fatigue loading was systematically studied.It was found that the DSA action intensity presented a significant loading direction and loading history dependence,and its microscopic mechanisms were revealed.The cyclic hardening,softening and continuous hardening behaviors of 316 LN stainless steel under high temperature fatigue loadings were explained from the perspective of internal stress components which were determined by the KuhlmannWilsdorf-Laird method based on the concept of Cottrell hysteresis decomposition.In addition,different microscopic characterization techniques,such as optical microscopy,scanning electron microscopy,electron backscattered diffraction and transmission electron microscopy were utilized to provide sufficient microscopic evidence for the interpretation of cyclic stress response based on stress decomposition.The change of dominant damage mechanism led to the difference in the relationship between the isothermal fatigue life at 550 ℃(IF-550 ℃)and the thermomechanical fatigue(TMF)life within the temperature range of 350℃– 550℃in different strain ranges.DSA played the dominant role in fatigue failure at the relatively high strain amplitudes(>0.4%),correspondingly,the TMF life was about two times of the IF-550℃ life.High temperature oxidation became the dominant damage mechanism when the strain amplitude was low(≤0.4%),which resulted in an approximately equal fatigue life in IF-550℃ and TMF tests.Further,the result of shorter TMF life comparing to the IF-550℃ life was likely to occur when the strain amplitude is small enough(<0.2%).The fatigue design curve of domestic nuclear grade 316 LN stainless steel with0.112% nitrogen content at 550℃ was determined,which showed that unconservative result was quite possible to happen if the fatigue design curves provided by ASME-NH and RCC-MRx were directly used in the structural design of the fourth generation sodium cooled fast reactor.The parameter of ‘weighting factor coefficient for the shear damage’ was proposed to reasonably evaluate the proportion of shear damage in the total multiaxial fatigue damage.The CXH-tension model modified by the above-mentioned parameter achieved satisfied results for the uniaxial/multiaxial isothermal and thermomechanical fatigue life prediction consistently.All the life data fell in the scatter band of 1.5 times and 2 times,respectively,and the predicted results were conservative,namely the modified CXH-tension model met the requirements of engineering application. |
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