Irradiation-induced Dislocation Loops Research Articles

Microstructure and fracture behavior of F82H steel under different irradiation and tensile test conditions

Specimens of martensitic steel F82H were irradiated to doses ranging from 10.7 dpa/850 appm He to 19.6 dpa/1740 appm He at temperatures between 165 and 305 °C in the second experiment of SINQ Target Irradiation Program (STIP-II). Tensile tests were conducted at different temperatures and various fracture modes were observed. Microstructural changes including irradiation-induced defect clusters, dislocation loops and helium bubbles under different irradiation conditions were investigated using transmission electron microscopy (TEM). The deformation microstructures of tensile tested specimens were carefully examined to understand the underlying deformation mechanisms. Deformation twinning was for the first time observed in irradiated martensitic steels. A change of deformation mechanism from dislocation channeling to deformation twinning was observed when the fracture mode changed from rather ductile (quasi-cleavage) to brittle (intergranular or cleavage and intergranular mixed).

Effect of heavy ion irradiation on microstructural evolution in CF8 cast austenitic stainless steel

The microstructural evolution in ferrite and austenitic in cast austenitic stainless steel (CASS) CF8, as received or thermally aged at 400 °C for 10,000 h, was followed under TEM with in situ irradiation of 1 MeV Kr ions at 300 and 350 °C to a fluence of 1.9 × 1015 ions/cm2 (∼3 dpa) at the IVEM-Tandem Facility. For the unaged CF8, the irradiation-induced dislocation loops appeared at a much lower dose in the austenite than in the ferrite. At the end dose, the austenite formed a well-developed dislocation network microstructure, while the ferrite exhibited an extended dislocation structure as line segments. Compared to the unaged CF8, the aged specimen appeared to have lower rate of damage accumulation. The rate of microstructural evolution under irradiation in the ferrite was significantly lower in the aged specimen than in the unaged. This difference is attributed to the different initial microstructures in the unaged and aged specimens, which implies that thermal aging and irradiation are not independent but interconnected damage processes.

Effects of Hydrogen-Ion Irritation on the Microstructure and Hardness of Fe-0.2wt.%V Alloy

Microstructural and hardening changes of Fe-0.2wt.%V alloy and pure Fe irradiated with 100 keV hydrogen ions at room temperature were investigated. It was found that dislocation density varies dramatically after irradiation, ranging from dislocation free to dense areas with tangled and complex dislocation configuration. As the irradiated Fe-0.2wt.%V samples were annealed at 773 K, the irradiation-induced dislocation loops disappear, while many small precipitates with enriched C distribute in the matrix. Some large precipitates with enriched V were also observed. The hardness of Fe-0.2wt.%V alloy and pure Fe increases after irradiation, which ascribes to the formation of dislocation loops in the irradiated specimens. Compared with pure Fe, the size of the irradiation-introduced dislocation loops in Fe-0.2wt.%V alloy decreases and the density increases, the change of the hardness also decreases. Keywords—Irradiation, Fe-0.2wt.%V alloy, microstructures, hardness.

Helium and hydrogen irradiation induced hardening in CLAM steel

The irradiation hardening behavior of low activation martensitic steel was investigated by nano-indentation measurements combined with transmission electron microscopy analysis. The steel was irradiated by single-(He+) and sequential-(He+ plus H+ subsequently) ions at 773K respectively. The continuous hardness (H) vs. depth (h) profiles at shallow depths were given by the continuous stiffness measurement (CSM) which can be analyzed according to Nix–Gao model and film/substrate system hardness model. The nano-indentation hardness results showed that both single-(He+) and sequential-(He+ plus H+ subsequently) irradiated steel exhibited obvious hardening and the hardening level of single-(He+) was higher than that of sequential-(He+ plus H+ subsequently) one. The bulk-equivalent hardness H0irr of the ion irradiated region was obtained by fitting the nano-indentation data using Kasada method. The microstructure analysis indicates that sequential-ion irradiation induced dislocation loops with larger size and smaller number density compared to single-ion irradiation at 773K. The irradiation induced hardening behavior in the steels subjected to different ion irradiations can be explained by the dispersed barrier hardening model which was consistent with the nano-indentation hardness results.

Interaction of He and He–V clusters with self-interstitials and dislocations defects in bcc Fe

The understanding of helium effects in synergy with radiation damage is crucial for the development of structural steels for fusion applications. Recent investigations in ultra-pure iron, taken as a basic model, have shown a drastic impact of dual beam (helium and iron) exposure on the accumulation of radiation-induced dislocation loops in terms of strong bias towards a0/2〈111〉 loops, while a0〈100〉 loops are mostly observed upon iron ion beam. In this work we perform a series of atomistic studies to rationalize possible mechanisms through which He could affect the evolution of microstructure and bias the population of a0/2〈111〉 loops. It is shown that He atoms are dragged by gliding a0/2〈111〉 loops. This strongly suppresses loop diffusivity and in turn it prohibits the mutual interaction of a0/2〈111〉 loops, being prerequisite of the formation of a0〈100〉 loops, as well as it reduces the disappearance of a0/2〈111〉 loops at sinks. A scenario for the microstructural evolution in the single- and dual-beam conditions is discussed.

Convoluted dislocation loops induced by helium irradiation in reduced-activation martensitic steel and their impact on mechanical properties

Helium irradiation induced dislocation loops in reduced-activation martensitic steels were investigated using transmission electron microscopy. The specimens were irradiated with 100keV helium ions to 0.8dpa at 350°C. Unexpectedly, very large dislocation loops were found, significantly larger than that induced by other types of irradiations under the same dose. Moreover, the large loops were convoluted and formed interesting flower-like shape. The large loops were determined as interstitial type. Loops with the Burgers vectors of b=〈100〉 were only observed. Furthermore, irradiation induced hardening caused by these large loops was observed using the nano-indentation technique.

Dislocation dynamics simulations of interactions between gliding dislocations and radiation induced prismatic loops in zirconium

The mechanical behavior of Pressurized Water Reactor fuel cladding tubes made of zirconium alloys is strongly affected by neutron irradiation due to the high density of radiation induced dislocation loops. In order to investigate the interaction mechanisms between gliding dislocations and loops in zirconium, a new nodal dislocation dynamics code, adapted to Hexagonal Close Packed metals, has been used. Various configurations have been systematically computed considering different glide planes, basal or prismatic, and different characters, edge or screw, for gliding dislocations with 〈a〉-type Burgers vectors. Simulations show various interaction mechanisms such as (i) absorption of a loop on an edge dislocation leading to the formation of a double super-jog, (ii) creation of a helical turn, on a screw dislocation, that acts as a strong pinning point or (iii) sweeping of a loop by a gliding dislocation. It is shown that the clearing of loops is more favorable when the dislocation glides in the basal plane than in the prismatic plane explaining the easy dislocation channeling in the basal plane observed after neutron irradiation by transmission electron microscopy.

Characterization of dislocation loops in CeO2 irradiated with high energy Krypton and Xenon

In order to fully characterize the structure of dislocation loops in CeO2, ion irradiations have been performed at 800 °C on CeO2 single-crystal thin films individually using 1 MeV Kr ions and 150 keV Xe ions, both to a dose of 5 × 1014 ions/cm2. Post-irradiation TEM examination, diffraction contrast imaging and high-resolution transmission electron microscopy (HRTEM), has confirmed that the irradiation-induced dislocation loops in CeO2 were Frank loops, having an interstitial nature with {1 1 1} habit planes and a 1/3 Burgers vector. Dislocation loops were confirmed to be a stacking fault in nature through TEM observations of the interference fringes inside the loop periphery.

Effects of carbide precipitate on the mechanical properties and irradiation behavior of the low activation martensitic steel

Abstract Precipitates may be induced during the processing, heat treatment and reactor operating and will play a key role in determining the properties of materials. The effects of precipitates on the mechanical properties and irradiation behavior of the low activation martensitic steel have been investigated using transmission electron microscopy combined with tensile test. Two types of China low activation martensitic (CLAM) specimens (with and without Si) were chosen as the model materials. The significant microstructural difference between the two CLAM specimens was the distribution of fine (Ta, W, V) C precipitates which can strengthen materials. Nano-indentation test revealed that the fine (Ta, W, V) C precipitates promoted by adding silicon can also decrease the irradiation hardening ratio of CLAM steel because the MC precipitates with high interface-to-volume ratio could decrease the density of the dislocation loop induced during irradiation. The interaction between precipitates and irradiation induced dislocation loops was discussed.

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.

In situ Evidence of Defect Cluster Absorption by Grain Boundaries in Kr Ion Irradiated Nanocrystalline Ni

Significant microstructural damage, in the form of defect clusters, typically occurs in metals subjected to heavy ion irradiation. High angle grain boundaries (GBs) have long been postulated as sinks for defect clusters, like dislocation loops. Here, we provide direct evidence, via in situ Kr ion irradiation within a transmission electron microscope, that high angle GBs in nanocrystalline (NC) Ni, with an average grain size of ~55 nm, can effectively absorb irradiation-induced dislocation loops and segments. These high angle GBs significantly reduce the density and size of irradiation-induced defect clusters in NC Ni compared to their bulk counterparts, and thus NC Ni achieves significant enhancement of irradiation tolerance.

Evolution of weld metals nanostructure and properties under irradiation and recovery annealing of VVER-type reactors

The results of VVER-440 steel Sv-10KhMFT and VVER-1000 steel SV-10KhGNMAA investigations by transmission electron microscopy, scanning electron microscopy, Auger-electron spectroscopy and mechanical tests are presented in this paper. The both types of weld metals with different content of impurities and alloying elements were studied after irradiations to fast neutron (E>0.5MeV) fluences in the wide range below and beyond the design values, after recovery annealing procedures and after re-irradiation following the annealing. The distinctive features of embrittlement kinetics of VVER-440 and VVER-1000 RPV weld metals conditioned by their chemical composition differences were investigated. It is shown that the main contribution into radiation strengthening within the design fluence can be attributed to radiation-induced precipitates, on reaching the design or beyond design values of fast neutron fluencies the main contribution into VVER-440 welds strengthening is made by radiation-induced dislocation loops, and in case of VVER-1000 welds – radiation-induced precipitates and grain-boundary phosphorous segregations. Recovery annealing of VVER-440 welds at 475°C during 100h causes irradiation-induced defects disappearance, transformation of copper enriched precipitates into bigger copper-rich precipitates with lower number density and leads to almost full recovery of mechanical properties followed by comparatively slow re-embrittlement rate. The recovery annealing temperature of VVER-1000 welds was higher – 565°C during 100h – to avoid temper brittleness. The annealing of VVER-1000 welds leads to almost full recovery of mechanical properties due to irradiation-induced defects disappearance and decrease in precipitates number density and grain-boundary segregation of phosphorus. The re-embrittlement rate of VVER-1000 weld during subsequent re-irradiation is at least not higher than the initial rate.

Effect of 16.3 dpa neutron irradiation on fatigue lifetime of the RAFM steel EUROFER97

Low cycle fatigue specimens of the reduced-activation martensitic/ferritic steel EUROFER97 were neutron irradiated at 250 °C up to an accumulated dose of 16.3 dpa. After irradiation, the specimens were push–pull fatigue tested under strain–controlled conditions at 250 °C to determine the impact of irradiation on lifetime, fracture behavior, and microstructure. The typical cyclic softening of martensitic/ferritic steels was observed. Furthermore, a considerable increase of lifetime after irradiation and subsequent cycling at lower strain amplitudes was remarkable. This behavior was attributed to the homogeneous distribution of stable irradiation-induced dislocation loops and small precipitates acting as barriers for the cyclic motion of dislocations, thereby influencing substantially crack initiation and crack network formation. While in the un-irradiated material push–pull fatigue sweeps the dislocations to the boundaries, a significant fraction of dislocations was fixed at irradiation-induced defects after irradiation and fatigue testing.

Copper indium diselenide: crystallography and radiation-induced dislocation loops

Copper indium diselenide (CIS) is a prime candidate as the absorber layer in solar cells for use in extraterrestrial environments due to its good photovoltaic efficiency and ability to resist radiation damage. While CIS-based devices have been tested extensively in the laboratory using electron and proton irradiation, there is still little understanding of the underlying mechanisms which give rise to its radiation hardness. To gain better insight into the response of CIS to displacing radiation, transmission electron microscope samples have been irradiated in situ with 400 keV Xe ions at the Intermediate Voltage Electron Microscope facility at Argonne National Laboratory, USA. At room temperature, dislocation loops were observed to form and grow with increasing fluence. These loops have been investigated using g · b techniques and inside/outside contrast analysis. They have been found to reside on {112} planes and to be interstitial in nature. The Burgers vector were calculated as b = 1/6 ⟨221⟩. The compositional content of these interstitial loops was found to be indistinguishable from the surrounding matrix within the sensitivity of the techniques used. To facilitate this work, experimental electron-diffraction zone-axis pattern maps were produced and these are also presented, along with analysis of the [100] zone-axis pattern.

Transmission electron microscopy study on neutron irradiated pure iron and RPV model alloys

The radiation induced microstructure was examined by Transmission Electron Microscopy in Fe, FeCu, FeMnCuNi, FeMnNi and a Reactor Pressure Vessel steel that were neutron irradiated to 0.026, 0.051, 0.10 and 0.19 dpa at 300 °C. The effect of dose and composition on defect accumulation and microstructure evolution was investigated. The damaged microstructure consisted in the presence of dislocation loops of interstitial type. The presence of voids was also studied in pure iron. Results on density, size and Burgers vector of radiation induced dislocation loops showed that the evolution of the interstitial component of the neutron irradiation induced microstructure was strongly affected by the presence of solutes such as Cu, Mn and Ni. Density and size increased with increasing dose in all the materials, while the effect of solutes is clearly to decrease the size of defects compared to pure iron. It has been observed that, for the same irradiation dose, the defect size decreases as the material becomes more complex, with the extreme case of the RPV steel where no defects were observed at any of the irradiation doses studied.

Dynamic observations of heavy-ion damage in Fe and Fe–Cr alloys

We give an overview and summary of recent in-situ heavy-ion irradiation experiments on Fe and Fe–Cr alloys carried out on the Argonne IVEM Facility at irradiation temperatures up to 500 °C. Several new and unexpected observations were made. At low doses the contrast of new irradiation-induced dislocation loops sometimes developed over time intervals as long as 0.2 s, many orders of magnitude longer than expected for a process of cascade collapse. In addition at temperatures ⩽300 °C, “hopping” of 1/2<1 1 1> loops was induced by the ion or electron beams, especially in UHP Fe. At high doses complex microstructures developed in all materials, involving the formation of large interstitial loops. At 300 °C and RT these loops had Burgers vectors of type b = 1/2<1 1 1> and large shear components. At 500 °C only edge loops with b = <1 0 0> were produced.

Effects of transmutation elements on the defect structure development of W irradiated by protons and neutrons

A series of W–Re–Os alloys were fabricated by arc melting for investigating the effects of transmutation elements of tungsten on the defect structure development. Transmutation electron microscopy has been used to investigate the defect structure for proton-irradiated (E=1MeV) W, W–3Re, W–3Os and 0.15dpa neutron-irradiated (E>1MeV) W–5Re–3Os and W, W–3Re, W–5Re and W–26Re. The irradiation-induced voids and dislocation loops which directly cause the irradiation hardening were observed. The results show the combination of W with Re or Os effectively restrains irradiation damage since the number density and radius of both voids and dislocation loops remarkably decrease with increasing Re or Os content.

Effects of Nickel Addition on Microstructural Evolution and Mechanical Properties of Reduced Activation Martensitic Steels Irradiated in the ATR-A1

Validity of nickel isotope tailoring (NIT) method to be able to generate a large amount of transmutation helium in 9%Cr-2%W reduced activation ferritic (RAF) steel for fusion reactor structural material was investigated and discussed through mechanical properties and microstructural evolution after fission neutron irradiation. Miniature tensile and Charpy impact specimens of RAF steels and 1% nickel added RAF steel were irradiated in the ATR-A1 to evaluate irradiation hardening and shift in ductile-brittle transition temperature (� DBTT). The amount of transmutation helium in all the RAF steels with and without nickel addition was calculated as only about 0.6 appm. After irradiation at 621 K, theDBTT for the 1% nickel added steel was similar for the JLF-1. After irradiation at 543 K, however, theDBTT for the 1% nickel added steel was significantly larger than that for the JLF-1. Microstructure observations revealed that irradiation-induced dislocation loops in the 1% nickel added steel were finer and denser than in the RAF steel without nickel addition, suggesting that the nickel addition to the RAF steels directly affected nucleation and growth processes of dislocation loops and enhanced irradiation hardening and embrittlement. Therefore, there is a limit in the NIT method as a simulation method for understanding effects of helium on irradiation embrittlement of RAF steels. The RAF steels used in the present study were JLF-1, Mod.JLF-1/LSM and Mod.JLF-1/LSM/Ni. (The ''LSM'' means low concentration of Si and Mn compared with the JLF-1.) The chemical compositions and the heat treatments of these steels are shown in Table 1. Mod.JLF-1/LSM/Ni contained 1 weight percentage of natural nickel. Neutron irradiation was carried out in the two sub-capsules of the ATR-A1, AS4 and AS16, in the Advanced Test Reactor (ATR) at Idaho National Engineering and Environ- mental Laboratory for 132.9 Effective Full Power Days (EFPDs). 11) Irradiation temperature and displacement dam- age were estimated for the two sub-capsules as at 543 K to 2.2 dpa and 621 K to 3.4 dpa. These sub-capsules were equipped with gadolinium shields of 1.7 mm in thickness to reduce the thermal neutron flux in order to suppress trans- mutation effects. The amount of transmutation helium was also suppressed due to the thermal shields. Greenwood et al. calculated the concentration of transmutation helium as 0.6 appm in Mod.JLF-1/LSM/Ni and 0.5 appm in the other RAF steels without nickel addition, respectively. 12)

Microstructure of irradiated ferritic/martensitic steels in relation to mechanical properties

A review of the microstructure of irradiated reduced activation ferritic/martensitic steels is presented, with a focus on F82H. Because of its resistance to irradiation induced hardening, swelling and low irradiation induced temperature shift in the ductile to brittle transition, this class of steels is a candidate structural material in future fusion reactors. The microstructure induced by irradiation is investigated in order to identify the key elements to deformation mechanisms. The study is focussed on F82H irradiated between 0.5 and 9.2 dpa at temperatures between 250 and 310 °C. Irradiation induced dislocation loops, when resolvable in the TEM, are known to have a Burgers vector a 0 〈1 0 0〉 , while for the smallest visible defects, or ‘black spots,’ there is still an uncertainty on their type. Black spot damage, possibly due to dislocation loops or precipitates, is investigated. It is attempted to quantitatively relate the dislocation loops and the precipitates to the irradiation induced hardening.

Retention and desorption of implanted deuterium and high-Z plasma facing materials

Thermal desorption after D 2 + irradiation from high-Z materials (W and Mo) has been compared with radiation induced microstructural evolution to investigate the details of retention and desorption of the implanted deuterium and underlying microscopic mechanism of the processes. The experimental results indicate that residual impurities act as major trapping sites for implanted deuterium independently of the ion energy for relatively low dose. The trapped deuterium is desorbed in two steps between 470 K and 660 K for high purity Mo, and 400 K and 650 K for high purity W. In the case of Mo damaged heavily at room temperature (1 × 10 22 ions/m 2), small cavities trap deuterium and give a large desorption stage at around 640 K. Radiation induced dislocation loops also trap the implanted deuterium effectively.