Search

Austenitic 316 stainless steel flux thimble tubes (FTTs) removed from a commercial pressurized water reactor (PWR) were re-irradiated using nickel ions to evolve the irradiated microstructure to higher damage levels than those achieved in reactor. The microstructures of the neutron-plus-ion irradiated samples were compared with those of neutron irradiated only samples at the same doses to evaluate the effectiveness of ion irradiation to extend the damage range and match that created in reactor. Ion irradiations effectively captured, both qualitatively and semi-quantitatively, the evolution of dislocation loops, nanocavities, nanocluster size and density, and radiation-induced segregation (RIS) with dose. Only Ni-Si clusters were observed in ion irradiated FTT samples while Ni-Si-Mn clusters that were frequently observed in reactor irradiated samples above ∼41 dpa were absent in ion irradiated samples probably due to ballistic mixing at the high ion irradiation damage rate. Two samples were ion irradiated to ∼160 dpa to predict the irradiated microstructure that may develop due to an increase of the damage accumulation by extending a PWR life from 40 years to 60 years. The extended ion irradiation results indicated that significant microstructure changes were unlikely to occur even to very high doses near common PWR relevant temperature conditions. Overall, the microstructure resulting from ion irradiation following neutron irradiation matched that of neutron to the same dose reasonably well in most cases. [ABSTRACT FROM AUTHOR]

Abstract Thirteen alloys including high- and low-strength nickel-base alloys, austenitic stainless steels, and ferritic alloys were irradiated using 2 MeV protons to a damage level of 2.5 dpa at 360 °C and assessed for their susceptibility to irradiation assisted stress corrosion cracking (IASCC) in both BWR normal water chemistry (NWC) and PWR primary water. Cracking susceptibility was highest for high strength nickel-base alloys, followed by the low strength nickel-base alloys and then the low strength iron-base alloys. Cracking in the nickel-based alloys was worst in normal water chemistry, which was reversed for the iron-based alloys. In general, cracking correlated with the degree of microstructure changes, though no single feature could be linked to cracking. IGSCC occurred in both the unirradiated and irradiated conditions in high strength nickel-base alloys with susceptibility being considerably higher following irradiation. In all cases, slip was planar, and the degree of slip localization correlated with the probability of IG crack initiation. Low strength nickel-base alloys showed the same dependence on environment as high strength alloys but were considerably less susceptible to IASCC initiation. Among the low strength iron-base alloys, alloy 800 was most susceptible to IASCC initiation in both BWR NWC and PWR primary water, which also correlated with grain boundary chromium depletion and silicon segregation. Across all alloys, cracking correlated with both the degree of localized deformation and the hardness in the irradiated condition. The agreement is expected as increased hardening also correlates with localized deformation, which is likely a necessary, though insufficient condition for cracking. [ABSTRACT FROM AUTHOR]

Abstract The irradiation-induced microstructure and irradiation-assisted stress corrosion cracking (IASCC) behavior of additively manufactured (AM) 316L stainless steels produced by laser powder bed fusion were evaluated for the first time. Irradiation-induced dislocation loops, voids, and γ′ precipitates were observed in all processing conditions following 2.5 dpa at 360 °C. The cell structure and dense dislocation walls in the stress-relieved AM materials recovered and showed signs of recrystallization following irradiation. Anisotropy in both tensile property and IASCC susceptibility were observed in the stress-relieved AM 316L stainless steel due to the printing texture. The hot-isotropic pressed AM 316L had better radiation tolerance and lower IASCC susceptibility than the stress-relieved AM 316L and conventionally forged 316L. Therefore, post-printing hot-isotropic pressing (HIP) is recommended for enhancing radiation tolerance and IASCC performance in nuclear applications as it eliminates the anisotropic mechanical behavior and IASCC susceptibility associated with the printing texture. Highlights • First irradiated microstructural characterization and IASCC evaluation of 3D printed 316L SSs. • Printed texture induced anisotropic mechanical and cracking behavior of additively manufactured 316L SSs. • Post-printing hot-isotropic pressing is critical for improving IASCC resistance of 3D printed 316L SS. [ABSTRACT FROM AUTHOR]

The role of dislocations on ion irradiation-induced void formation is studied in a high-entropy alloy (HEA) NiCoFeCrPd. Despite previous observations that show high-entropy alloys are swelling resistant due to a high defect recombination rate, the swelling is enhanced with increasing density of pre-existing dislocations at low strain levels that shortened transient duration before the onset of void swelling. Under certain irradiation conditions, a high density of dislocations may carry the material closer to the sink-dominated regime. Compared to another HEA NiCoFeCrMn, NiCoFeCrPd has a smaller loop size and higher loop density due to the stronger lattice distortion. Swelling behavior in a newly designed high-entropy alloy NiCoFeCrPd has been studied for the first time. Void swelling is enhanced with increasing density of pre-existing dislocations in NiCoFeCrPd. [ABSTRACT FROM AUTHOR]

Twelve commercial-grade austenitic alloys based on the Ni-Cr-Mo-Fe quaternary system were irradiated using 2 MeV protons at 360 °C to a damage level of 2.5 displacements per atom (dpa). Long-range ordered (Pt 2 Mo-type) precipitation under proton irradiation was observed, for the first time, in alloys C22, 625, 625P, 625D, 725, and 690. No relevant short-range ordering was observed. These irradiation-enhanced long-range ordered precipitates are coherent with the matrix despite their irregular shape. Of the potential influences on long-range ordering (Ni:Cr, and Ni:(Cr + Mo) ratios, Mo, and iron concentration), Fe content was the strongest by far. The volume fraction of LRO decreases with increasing Fe content by virtue of its role as a stabilizer of the disordered FCC phase, thus reducing the energy savings from ordering. The observed effects of Fe on long-range ordering show qualitative agreement with predictions from thermodynamic modeling. Although solid state diffusion kinetics dominate long-range ordering under purely thermal conditions, ordering under irradiation here (∼ 2.5dpa) is controlled by the thermodynamic driving force. Proton irradiation thus offers a unique approach for studying the low temperature phase transformation in a thermodynamically favored, but kinetically constrained condition. [ABSTRACT FROM AUTHOR]

Compositionally-graded austenitic 316L stainless steel (SS) samples with five different Hafnium (Hf) concentrations (up to 1 wt.%) were additively manufactured by directed energy deposition and then were irradiated using 5 MeV Fe2+ ions to 50 displacements per atom (dpa) at 500, 550, and 600 °C, respectively. A rastering beam was used to ensure homogenous irradiations for the large areas of ∼1.40 cm2. Composition-dependent void evolution was evaluated. Void size, void number density, and void swelling all tend to decrease with increasing Hf concentration at all three temperatures. When the nominal Hf concentration increases to 1 wt.% in the doped additively manufactured (AM) 316L SS, the void swelling is over an order of magnitude lower than that in pure AM 316L. Atom probe tomography results showed that about 0.14 wt.% Hf dissolved in the matrix at the 1 wt.%. Hf nominal concentration. The suppression of void swelling by Hf addition shows qualitative agreement with the numerical calculation based on the vacancy trapping mechanism, indicating that the oversized Hf reduces the steady state vacancy concentration, and hence the incubation period of swelling is extended, resulting in a dramatic difference in void swelling. Other factors including grain boundaries, dislocation and dislocation cells, Hf-rich particles, and delta ferrite are secondary to this mechanism. This work shows that additive manufacturing enabled microalloying is promising for developing void swelling resistant materials. [ABSTRACT FROM AUTHOR]

• The temperature-dependent void swelling of additively manufactured 316L SSs was investigated for the first time. • The void swelling behaviors of additively manufactured 316L and cold-worked 316L SSs were systematically compared. • The dislocations trapped at cell walls and dislocation networks induced by cold work probably functioned differently in void swelling. We studied the void swelling in stress-relieved (SR) additively manufactured (AM) 316L stainless steel (SS) in comparison with hot-isostatic pressed (HIP) AM 316L counterparts with identical chemistry. To understand the role of dislocation cell structure in SR AM 316L SS, we also included cold-worked (CW) 316L SS with various CW levels for dislocation networks. The irradiation damage of 30 displacements per atom (dpa) in AM and CW 316L SSs were compared after 5 MeV Fe2+ ion irradiations at 500 °C, 550 °C, and 600 °C, respectively. To investigate the temperature and dose dependence, the AM 316L SSs were further irradiated to 100 dpa at 400 °C and 600 °C. Results show that the void welling in SR AM 316L SS is more severe than in HIP AM 316L SS, in terms of a higher void density and a deeper void nucleation depth through all the examined conditions. This is ascribed to a more extended incubation period in HIP AM 316L SS. Beyond the incubation, both samples proceeded at a similar swelling rate. Dislocation networks from cold-working can effectively suppress void swelling, while dislocation cell structures containing entangled dislocations on cell walls cannot. The distinct responses indicate the potential influence of dislocation configuration on void swelling. Despite some relevant data are available, our understandings of radiation damage in AM 316L SSs are still limited. [ABSTRACT FROM AUTHOR]

Additive manufacturing (AM) or rapid prototyping has become a crucial tool for reducing both cost and time while increasing efficiency in qualifying structural materials for reactor use. In this study, directed energy deposition (DED) was used to develop an as-built 316 L stainless steel sample with three regions of increasing Hf-dopant to study the effects of Hf on the irradiation response of the material. Morphological and microstructural changes were analyzed before and after 2 MeV proton irradiation at 360 °C to a damage of 2.5 dpa at ∼ 10 µm below the surface. The addition of Hf effectively suppressed radiation-induced damage (dislocation loops, radiation-induced segregation) due to enhanced point defect recombination. The radiation damage seen in the as-built sample was further compared to a thermo-mechanically treated counterpart of the same fabrication and irradiation parameters which was found to behave superiorly. The increased radiation resistance of this material may be attributed to the as-built microstructure, which includes undissolved Hf particles, delta ferrite grains and cellular sub-grain boundaries that can hinder defect motion. [ABSTRACT FROM AUTHOR]

The role of proton irradiation in the general oxidation of 316 L stainless steel in simulated PWR primary water after 1000 h immersion was clarified by comparing the microstructural and microchemical features of oxides formed on both the irradiated and un-irradiated regions of identical grains. Hence, the interference from differences in crystallographic orientation can be ruled out. Interestingly, the average inner oxide thickness on the irradiated region is significantly thinner than that on the un-irradiated region. The enhanced resistance to oxidation on the irradiated region originates from the more protective inner oxide layer and the formation of a continuous Ni-rich zone near the oxide/matrix interface. The inner oxide formed on irradiated region is less deficient in cation due to the enhanced diffusion of Cr from matrix by the irradiation-induced defects, thus making a better diffusion barrier. Meanwhile, the formation of continuous Ni-rich zone near the oxide/matrix interface, which is facilitated by the Ni-rich defects serving as nuclei, fast generation of vacancy at the interface and suppression of the outward diffusion of surplus Ni, can diminish the available space for the growth of inner oxide. The limited oxidation space and high Cr content at the oxide/matrix interface in the irradiated region result in the formation of rocksalt oxide while spinel oxide is formed near the interface in the un-irradiated region. [Display omitted] [ABSTRACT FROM AUTHOR]