Ion Irradiation Damage Of The Candidate Materials For Super-critical Water-cooled Reactors And Fusion Reactors

The irradiation damage in four candidate materials for Super-critical Water-cooled Reactor irradiated with ions was studied in present paper, which included Nickel-base Alloy C-276and718, Ferritic/martensitic Steel P92and Austenitic Steel HR3C. At last, irradiation-induced precipitation behavior in Super-clean Reduced Activation Martensitic Steel (SCRAM steel) irradiated with ions was studied in this paper, which is one of the candidate materials of Fusion Reactor.1. Irradiation damage in nickel-base alloy C-276irradiated by argon ions at room temperature and elevated temperatures with irradiation doses from0.28dpa to82.5dpa and irradiated at500℃with300keV self-ions (Ni+) to a peak displacement damage of4.5dpa has been investigated by TEM. Nano-scale black spot appeared at low dose (0.83dpa-2.75dpa). Large dislocation loops were observed at the dose of6dpa-27.5dpa. When the irradiation dose reached to82.5dpa, original grains transformed into subgrains whose mean size were determined with TEM and GIXRD. Irradiation damage in surface morphology of C-276alloy was analyzed by AFM. High density dislocation loops were observed in C-276alloy irradited with Ni ions, and the Burgers vectors of dislocation loops were predominantly (a/3)<111> and (a/2)<110>.2. Evolution of γ’ and γ" precipitates in nickel-base alloy718irradiated with argon ions at elevated temperature has been investigated via TEM. The results showed that the y’and y" precipitates became disorered. The reason that the state of order was completely destroyed may be the role of argon ion irradiation induced defects.3. Ferritic/martensitic P92steel was irradiated with Ar ion beams at room temperature and elevated temperatures. Evolution of the matrix, the carbides and surface morphology has been investigated by TEM. The precipitate periphery (the matrix/carbide interface) became amorphous under irradiation at low temperature (<290℃). However, the amorphization phenomena did not occur when the irradiation temperatures rose to390℃and550℃. High density small carbide precipitates were observed in matrix irradiated to room temperature/34.5dpa,290℃/12dpa and390℃/7dpa. The Ar-bubbles induced by irradiation were observed at550℃. EDX analysis revealed that segregation of Cr and W and depletion of Fe in carbides occurred under irradiation. SEM and AFM showed that a large number of nanometer-sized hillocks were formed in the surface irradiated with Ar ions at550℃, and the surface roughness increased.4. The microstructural evolution in argon ion irradiated austenitic stainless steel HR3C at290℃and550℃were examined via TEM. High density dislocation loops were observed in HR3C steel at290℃. The mean size of dislocation loops grew and the density of loops decreased with the increase of irradiation dose. The selected-area electron diffraction patterns of precipitates undoubtedly indicated that the precipitates were the M23C6, carbides at550℃. It is found that M23CV, precipitates nucleated on dislocation loops in austenitic stainless steel readily at high temperature, then grew and coarsened with the irradiation dose increasing.5. Two SCRAM steels were selected for present study:SCRAM-6(9Cr2W0.25V) and SCRAM-9(9Cr2W0.25V-Ti). Irradiation-induced precipitation behavior at300℃in single (Fe+) and dual-beam (Fe++H+) ion-irradiated specimen has been investigated by using TEM and EDX. The EDX analysis indicated that the precipitates in ion-irradiated SCRAM-6steel were chromium-rich M23C6carbides and the main precipitates in the Ti-added steel SCRAM-9were titanium-rich Ti(N, C) precipitates. It is, thus, believed that the Ti-added SCRAM-9steel had good performance in delaying the increasing of precipitates. The hydrogen effect on irradiation-induced precipitation behavior has been also discussed in present work. The implantation of hydrogen atoms could accelerate the precipitation of the minor precipitates.