Significant Irradiation Hardening Research Articles

Post-irradiation examination of commercial tantalum alloys following neutron irradiation

Post irradiation examination of two commercial tantalum alloys (based on Ta-8 %W) revealed significant irradiation hardening following neutron irradiation in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Tensile samples of commercial Ta alloys T-111 and ASTAR-811C were irradiated to a total neutron fluence (E > 0.1 MeV) in the range (2–4)x1021 n/cm2 (0.39–0.75 dpa) at an irradiation temperature near 900 °C. Tensile testing conducted at room temperature and at 800 °C revealed significant irradiation hardening under all test conditions. Samples that were tensile tested at room temperature exhibited brittle failure, whereas ductility was maintained for elevated temperature testing. The irradiation hardening and embrittlement noted for both alloys was attributed to irradiation-induced dislocation loop formation as confirmed through scanning transmission electron microscopy analysis. Corresponding energy dispersive x-ray spectroscopy revealed radiation-induced segregation of Hf in these loop populations. This Hf enrichment is shown to result in eventual co-precipitation of elongated (Hf,O)-rich precipitates along the dislocation loops. These results show that irradiation hardening is enhanced via irradiation-enhanced precipitation at elevated irradiation temperatures. Consequently, the minimum recommended operating temperature window for Ta alloys should be increased to account for observed hardening/embrittlement at irradiation temperatures as high as 900 °C (1,173 K).

Ion-irradiation induced hardening behavior of zirconium alloys: A combination of experimental and theoretical study

This study aimed to investigate the microstructure information and mechanical behavior of ion-irradiated zirconium (Zr) alloys through a combination of experimental measurements and theoretical analysis. Experimental procedures involved Au3+ irradiation of Zr-1.5Sn-0.2Fe-0.1Cr and Zr-1.0Sn-1.0Nb-0.1Fe at 3 displacements per atom (dpa) and 30 dpa. Transmission electron microscopy analysis revealed the formation of precipitates such as Zr(Fe, Cr)2 in Zr-1.5Sn-0.2Fe-0.1Cr, as well as β-Nb and Zr(Nb, Fe)2 in Zr-1.0Sn-1.0Nb-0.1Fe. These precipitates exhibited a transformation from crystalline state to partial or complete amorphization after ion-irradiation. Furthermore, macroscopic mechanical properties of the Zr alloys were characterized by nano-indentation tests, which revealed a significant indentation size effect (ISE) and irradiation hardening behavior. To provide a theoretical understanding of the depth-dependent hardness observed in ion-irradiated Zr alloys, a mechanistic model was developed to address the different expansion rates of the plastic zone in the irradiated and unirradiated region. The results indicated that the ISE phenomenon could be attributed to the reduction in the density of geometrically necessary dislocations (GNDs) caused by the expansion of the plastic zone. In addition, the irradiation hardening behavior primarily resulted from the contribution of irradiation-induced defects. With increasing indentation depth, the dominant hardening mechanisms changed from the contribution of irradiation-induced defects, GNDs and statistically stored dislocations (SSDs) in the irradiated region to SSDs in the unirradiated substrate.

Tensile properties of modified 316 stainless steel (PNC316) after neutron irradiation over 100 dpa

ABSTRACT The effects of fast neutron irradiation on tensile properties of 20% cold-worked modified 316 austenitic stainless steel (PNC316), which was improved swelling resistance and creep strength, claddings and wrappers for fast reactors were investigated. PNC316 claddings and wrappers were irradiated in the experimental fast reactor Joyo at irradiation temperatures between 400°C and 735°C to fast neutron doses ranging from 21 to 125 dpa. The post-irradiation tensile tests were carried out at room and irradiation temperatures. Elongations of PNC316 measured by the tensile tests were maintained at an engineering level, although the material incurred significant irradiation hardening and softening depending on irradiation temperature. The maximum swelling of PNC316 wrappers was about 2.5 vol.% at irradiation temperature between 400°C and 500°C up to 110 dpa. Japanese 20% cold-worked austenitic steels, PNC316 and 15Cr-20Ni, had sufficient ductility and work-hardenability even after above 10 vol.% swelling, while they had very weak plastic instabilities.

Effect of temperature history on the irradiation behavior of vanadium alloy irradiated with the MARICO-II rig in a fast reactor, JOYO

The irradiation behavior of vanadium alloys was examined to evaluate the effect of temperature history during reactor operation. Irradiation experiments were performed in a fast reactor, Joyo, with an irradiation-temperature-controlled irradiation rig, MARICO-II. In-core irradiation rigs that control the irradiation temperature and are not influenced by reactor power were used. Irradiation conditions with or without the MARICO-II rig were performed. Significant differences in irradiation hardening and swelling behavior in undersized vanadium binary alloys were visible between temperature-controlled and conventional rigs. The temperature history affects the microstructural evolution, especially the nucleation and growth of voids during startup and shutdown procedures. No difference was observed in the vanadium alloys with oversized solute atom because the vacancy migration was suppressed by tight binding between oversized solute atoms and vacancies during irradiation. V–4Cr–4Ti–(Si,Al,Y) alloys showed significant irradiation hardening at high-temperature irradiation even though the contribution of dislocation and precipitates for irradiation hardening was insufficient to satisfy the measured irradiation hardening. The significant irradiation hardening in the V–4Cr–4Ti–(Si,Al,Y) alloys may be caused by transmission-electron-microscope-invisible clusters including minor elements (Si, Al, Y) that were formed at a high temperature during irradiation.

Effect of neutron irradiation on ductility of tungsten foils developed for tungsten-copper laminates

Severe plastic deformation of tungsten (W) is known to be an efficient way to reduce its inherently high ductile-to-brittle transition temperature (DBTT), what is essential for its use in components of a fusion reactor. Thin rolled W foils possess superior mechanical behaviour at room temperature (RT), as demonstrated in previous works. It was then proposed to expand the beneficial mechanical properties of the foil to bulk by fabricating tungsten-copper (W-Cu) laminate composites, which can be used for structural applications. Neutron irradiation in HFIR resulted in embrittlement of the laminate already after 0.016dpa, with the W foil determining the composite behaviour.In this work, for the first time, we investigate the effect of neutron irradiation on individual W foil, and determine the resulting DBTT shift with the help of cantilever bend tests, using bulk W and the W-Cu composite as reference. The W foil and the bulk samples were irradiated to 0.15dpa at 400 °C in the BR-2 reactor in Mol (Belgium). We also hypothesise that diffusion of Cu atoms into W could modify the response to irradiation in these materials. We substantiate it with complementary density functional theory (DFT) ab initio calculations to analyse the Cu-vacancy and Cu-self-interstitial interaction, which helps to elucidate co-alignment of the fluxes of point defects and Cu solutes in W matrix.Irradiated foil was found to retain its ductility at RT. No significant irradiation hardening or DBTT shift were detected in the irradiated W foil compared to the bulk W. The different irradiation effect on embrittlement in individual foils and in the laminate may be attributed to the irradiation-assisted diffusion of Cu solutes in W foil, which could form intermetallic phases and affect the accumulation of lattice defects.

The microstructure evolution and irradiation hardening in 15Cr-ODS steel irradiated by helium ions

Oxide dispersion strengthened (ODS) steels with excellent irradiation swelling resistance and high temperature creep resistance are regarded as one of the priority candidate materials for the first wall/blanket materials in fusion reactors. Microstructure evolution and irradiation hardening in 15Cr-ODS steel irradiated by helium ions at R.T. and 400 °C respectively have been studied by transmission electron microscopy and nanoindentation technique. Three main oxides were identified, including high-density dispersed Y-Ti-O nanoclusters, tens of nanometers Y2TiO5 and a few hundred nanometers Y2Ti2O7 particles with the largest size but the least density. Helium bubbles were homogeneously distributed in the matrix in both irradiated specimens and the aggregation state of bubbles near GBs was related to the nature of GB. The average size of bubbles increased and the density decreased in the steel irradiated at elevated temperature. A defect-depleted zone was observed near a GB with a Y2TiO5 precipitate imbedded and a large helium bubble was present at the interface of the Y2TiO5 particle/GB. It suggests that the GBs combined with Y2TiO5 can enhance the aggregation of helium bubbles since GBs can act as defect sinks as well as diffusion channels and Y2TiO5 has a high ability of trapping helium atoms. Significant irradiation hardening was obtained at both irradiation temperatures. The relationship between microstructure and the irradiation hardening was discussed.

Irradiation hardening of stainless steel model alloy after Fe-ion irradiation and post-irradiation annealing treatment

Microstructural observation with atom probe tomography, transmission electron microscopy, and mechanical examination with a nano-indenter were performed on stainless steel model alloy that had been irradiated with Fe ions and subjected to post-irradiation annealing treatment (PIA). Significant irradiation hardening occurred in the as-irradiated specimen and irradiation hardening recovery occurred at 400 °C of cumulative PIA, Tiny black dots, Frank loops and Ni-Si clusters formed in the as-irradiated specimen and their recovery was observed with the cumulative PIA. A classic dispersed-barrier hardening model was used to evaluate irradiation hardening connected to microstructural information by applying the density and size of defect variants. Ni-Si clusters and Frank loops contributed substantially to the increase of irradiation hardening, whereas the black dots contributed little for the stainless steel model alloy. The barrier strength factor of Ni-Si clusters, αSC, was obtained as 0.029 and 0.015 for two different obstacle strength model, which suggests that an individual Ni-Si cluster is a weak barrier obstacle to mobile dislocation on the dispersed-barrier hardening model, but that the dense formation of Ni-Si clusters controls irradiation hardening behavior. A quantitative analysis was performed of the increase in strength from the Ni-Si clusters that was caused by coherency strains, and an irradiation hardening mechanism for the stainless steel, including the solute atom clusters, is discussed.

Effects of proton irradiation on microstructure and mechanical properties of nanocrystalline Cu–10at%Ta alloy

Duplex nanocrystalline Cu–10at%Ta is attractive for extreme environments because of its grain size stability at high homologous temperatures, owing to dispersed Ta nanoclusters. This study is an initial investigation into the microstructural and mechanical stability of Cu–10at%Ta under irradiation. Proton irradiation is carried out to 1 displacement per atom (dpa) at 500 °C. The Cu–10at%Ta exhibits grain size stability throughout irradiation and resistance to the formation of irradiation-induced defects. However, Ta nanoclusters undergo ballistic dissolution. Nanoindentation reveals there is no significant irradiation hardening, consistent with the relatively irradiation-stable microstructure. Ballistic dissolution of Ta nanoclusters is discussed in the context of Orowan strengthening.

Effects of Proton Irradiation on Microstructure and Mechanical Properties of Nanocrystalline Cu-10at%Ta Alloy

Duplex nanocrystalline Cu-10at%Ta is attractive for extreme environments because of its grain size stability at high homologous temperatures, owing to dispersed Ta nanoclusters. This study is an initial investigation into the microstructural and mechanical stability of Cu-10at%Ta under irradiation. Proton irradiation is carried out to 1 displacement per atom (dpa) at 500°C. The Cu-10at%Ta exhibits grain size stability throughout irradiation and resistance to the formation of irradiation-induced defects. However, Ta nanoclusters undergo ballistic dissolution. Nanoindentation reveals there is no significant irradiation hardening, consistent with the relatively irradiation-stable microstructure. Ballistic dissolution of Ta nanoclusters is discussed in the context of Orowan strengthening.

In-situ high-energy X-ray characterization of neutron irradiated HT-UPS stainless steel under tensile deformation

The tensile deformation behavior of a high-temperature, ultrafine-precipitate strengthened (HT-UPS) stainless steel was characterized in-situ with high-energy X-ray diffraction at 20 and 400 °C. The HT-UPS samples were neutron irradiated to 3 dpa at 400 °C. Significant irradiation hardening and ductility loss were observed at both temperatures. Lattice strain evolutions of the irradiated samples showed a strong linear response up to near the onset of the macroscopic yield, in contrast to the unirradiated HT-UPS which showed a pronounced non-linear behavior well below the macroscopic yield. While the room-temperature diffraction elastic moduli in the longitudinal direction increased after irradiation, the 400 °C moduli were similar before and after irradiation. The evolution of the {200} lattice strain parallel to the loading axis (ε{200}L) showed unique characteristics: in the plastic regime, the evolution of ε{200}L after yield is temperature-dependent in the unirradiated specimens but temperature-independent in the irradiated specimens; and the value of ε{200}L at the yield is an irradiation-sensitive, temperature-independent parameter. The evolution of ε{200}L corresponds well with the dislocation density evolution, and is an effective probe of the deformation-induced long-range internal stresses in the HT-UPS steel.

Micro mechanical testing of candidate structural alloys for Gen-IV nuclear reactors

Ion irradiation is often used to simulate the effects of neutron irradiation due to reduced activation of materials and vastly increased dose rates. However, the low penetration depth of ions requires the development of small-scale mechanical testing techniques, such as nanoindentation and microcompression, in order to measure mechanical properties of the irradiated material. In this study, several candidate structural alloys for Gen-IV reactors (800H, T91, nanocrystalline T91 and 14YWT) were irradiated with 70 MeV Fe9+ ions at 452 °C to an average damage of 20.68 dpa. Both the nanoindentation and microcompression techniques revealed significant irradiation hardening and an increase in yield stress after irradiation in austenitic 800H and ferritic-martensitic T91 alloys. Ion irradiation was observed to have minimal effect on the mechanical properties of nanocrystalline T91 and oxide dispersion strengthened 14YWT. These observations are further supported by line broadening analysis of X-ray diffraction measurements, which show a significantly smaller increase in dislocation density in the 14YWT and nanocrystalline T91 alloys after irradiation. In addition, good agreement was observed between cross-sectional nanoindentation and the damage profile from SRIM calculations.

Effect of irradiation hardening on brittle fracture performance of box-shaped blanket fabricated by F-82H

Fusion reactor blankets designed for DEMO are based on a water cooled ceramic breeder. The unstable fracture of the box-shaped blanket would occur by the superimposition of three factors as follows: (1) it is the reduction of fracture toughness for structural materials, such as F-82H, caused by the significant irradiation hardening, (2) it is the formation of cracks which act as the initiation point of the fracture, and (3) it is the excessive tensile load caused by the inner pressure to In-Box loss of coolant accident (LOCA). In the present study, application of Weibull stress criterion to fracture assessment is to investigate the effect of irradiation hardening on the brittle fracture performance of box-shaped blanket fabricated by F-82H. Both structural analysis of the box-shaped blanket simulating In-Box LOCA and deformation behavior analysis of irradiated toughness specimens were conducted by elastic-plastic finite element analysis (FEA). Weibull stress values calculated from FEA of irradiated toughness specimens were higher than those from FEA of irradiated structural components. It, therefore, is considered that the irradiation hardening of structural materials would not promote the brittle fracture at the corner of rectangular coolant tubes of the blankets caused by the load of In-Box LOCA.

The effects of ion irradiation on the micromechanical fracture strength and hardness of a self-passivating tungsten alloy

An ultra-fine grained self-passivating tungsten alloy (W88-Cr10-Ti2 in wt.%) has been implanted with iodine ions to average doses of 0.7 and 7 dpa, as well as with helium ions to an average concentration of 650 appm. Pile-up corrected Berkovich nanoindentation reveals significant irradiation hardening, with a maximum hardening of 1.9 GPa (17.5%) observed. The brittle fracture strength of the material in all implantation conditions was measured through un-notched cantilever bending at the microscopic scale. All cantilever beams failed catastrophically in an intergranular fashion. A statistically confirmed small decrease in strength is observed after low dose implantation (−6%), whilst the high dose implantation results in a significant increase in fracture strength (+9%), further increased by additional helium implantation (+16%). The use of iodine ions as the implantation ion type is justified through a comparison of the hardening behaviour of pure tungsten under tungsten and iodine implantation.

Evaluation of irradiation hardening and microstructure evolution under the synergistic interaction of He and subsequent Fe ions irradiation in CLAM steel

Sequential dual-ion irradiation is a useful technique for experimental exploration on the synergistic effects of ion accumulation and cascade damage. In this research, the helium-ion accumulation concomitant with displacement damage induced by helium and iron ions irradiation in China low active martensitic (CLAM) steel was studied by ion-irradiation, transmission electron microscopy and nano-indentation technique. The helium bubbles formed under continuous implantation of helium ions at room temperature and the helium dose forming observed bubbles resolved by TEM clearly in CLAM steel is around 0.7 × 1017–1.0 × 1017 He+/cm2 under single 100 keV He+ irradiation. Significant irradiation hardening was observed in the samples with various ion- and dose-irradiations. Positive correlation between the hardening increment and the helium-ion dose has been established in single helium-ion irradiated samples. On the other hand, the subsequent Fe-ion irradiation greatly promoted the formation and growth of helium bubbles as well as dislocation loops in sequential dual-ion irradiated samples. No significant contribution of the subsequent Fe-ion irradiation on the hardness increment was found for the sequential dual-ion irradiated samples. It is suggested that the defects recombination, the combination effects of size and density of defects contribute to the degree of irradiation hardening. The calculated hardness increment based on dispersion strengthening model and the experimental microstructure analysis followed the same trend as the experimental nano-indentation data.

Combined Effect of Irradiation and Temperature on the Mechanical Strength of Inconel 800H and AISI 310 Alloys for In-Core Components of a Gen-IV SCWR

Inconel 800H and AISI 310 alloy samples were exposed to Fe4+ ions to simulate neutron irradiation damage, and then annealed at 400°C and 500°C to study the kinetics of thermal recovery of the irradiation damage. The increase in hardness with ion irradiation and the decrease in hardness due to thermal recovery were recorded. Our findings suggest that under thermal and neutron irradiation conditions envisaged for the Canadian Gen-IV SCWR concept, both alloys will experience significant irradiation hardening; however, this will be concurrently negated by even more rapid thermal recovery of the irradiation damage.

Effect of post-weld heat treatment and neutron irradiation on a dissimilar-metal joint between F82H steel and 316L stainless steel

A dissimilar-metal joint between F82H steel and 316L stainless steel was fabricated by using electron beam welding (EBW). By microstructural analysis and hardness test, the heat-affected zone (HAZ) of F82H was classified into interlayer area, fine-grain area, and coarse-carbide area. Post-weld heat treatment (PWHT) was applied to control the hardness of HAZ. After PWHT at 680°C for 1h, neutron irradiation at 300°C with a dose of 0.1dpa was carried out for the joint in Belgian Reactor II (BR-II). Compared to the base metals (BMs) and weld metal (WM), significant irradiation hardening up to 450HV was found in the fine-grain HAZ of F82H. However, the impact property of F82H-HAZ specimens, which was machined with the root of the V-notch at HAZ of F82H, was not deteriorated obviously in spite of the significant irradiation hardening.

Radiation effects on microstructure and hardness of a titanium aluminide alloy irradiated by helium ions at room and elevated temperatures

A 45XD TiAl alloy possessing a lamellar microstructure was irradiated using 5MeV helium ions to a fluence of 5×1021ionm−2 (5000appm) with a dose of about 1dpa (displacements per atom). A uniform helium ion stopping damage region about 17μm deep from the target surface was achieved by applying an energy degrading wheel. Radiation damage defects including helium-vacancy clusters and small helium bubbles were found in the microstructure of the samples irradiated at room temperature. With increasing irradiation temperature to 300°C and 500°C helium bubbles were clearly observed in both the α2 and γ phases of the irradiated microstructure. By means of nanoindentation significant irradiation hardening was measured. For the samples irradiated at room temperature the hardness increased from 5.6GPa to 8.5GPa and the irradiation-hardening effect reduced to approximately 8.0GPa for the samples irradiated at 300°C and 500°C.

Effect of alloying elements on irradiation hardening behavior and microstructure evolution in BCC Fe

Ion irradiations with 6.4MeV Fe3+ were performed on pure-Fe, Fe–1at.% Cr, Fe–1at.% Mn, and Fe–1at.% Ni to a nominal damage of 1dpa, at a damage rate of 1×10−4dpa/s, at irradiation temperatures of 473, 563, and 673K. After irradiations at 473 and 563K, Fe–1Mn and Fe–1Ni showed significant irradiation hardening, which was due to irradiation induced dislocation loops in high density. In pure-Fe, the dislocation loops were localized in the vicinity of dislocations, while those in Fe alloys were distributed rather homogeneously. This can be interpreted in terms of the interaction between alloying element and dislocation strain field. Irradiation at 673K resulted in the formation of voids in pure-Fe, Fe–1Cr, and Fe–1Mn. We found that chromium suppressed void swelling.

Dose dependence of irradiation hardening of binary ferritic alloys irradiated with Fe3+ ions

The effects of alloying elements on irradiation hardening behavior of various Fe-based binary alloys irradiated with Fe3+ ions were investigated by using the nano-indentation technique. Ion-irradiations with 6.4MeV Fe3+ were performed on Pure-Fe, Fe–1Cr, Fe–1Mn, Fe–1.5Mn, Fe–1Ni, Fe–1Cu, and Fe–1Mo to nominal displacement damage levels of 0.1, 0.3 and 1dpa, at a damage rate of 1×10−4dpa/s at 290°C. Significant irradiation hardening was observed in the Fe–1Mn, Fe–1.5Mn and Fe–1Ni as well as in the Fe–1Cu, but not in the Pure-Fe, Fe–1Cr and Fe–1Mo, after irradiation to 1dpa at 290°C. The dose dependence of the irradiation hardening was well explained for all the alloys by the saturation fitting model. Microstructural observation suggested that irradiation induced matrix defects in Fe–Mn binary alloy promoted the irradiation hardening.

Neutron Irradiation Hardening of Fe-Based Binary Alloys

We investigated mechanical properties of neutron irradiated Fe based binary alloys in order to extract roles of each alloying element in reactor pressure vessel (RPV) steels on irradiation hardening and annealing recovery behavior. Materials used were Pure-Fe, Fe-1Cr, Fe-1Mn, Fe-1Ni, Fe-1Cu and Fe-1Mo in at.%. Neutron irradiations were carried out at various irradiation doses from 0.3 to 8.5 × 1019 n/cm2 ( > 1.0 MeV) at 290 °C. Irradiation hardening of Fe-1Cu showed a tendency of saturation at a low dose. Irradiation hardening of Pure-Fe and the other binary alloys increased with increasing in irradiation dose. Especially, Fe-1Mn irradiated over 4.3 × 1019 n/cm2 showed significant irradiation hardening which is comparable to Fe-1Cu. However, the post-irradiation annealing recovery behavior of the irradiation hardening in Fe-Mn showed one-stage recovery at around 450 °C, which was completely different from the two-stages recovery behavior of Fe-1Cu.