Neutron irradiation hardening and microstructure changes in Fe-Mn binary alloys

Irradiation hardening and microstructure changes in Fe-Mn binary alloys were investigated after neutron irradiation at 290 °C and up to 0.13 dpa. Significant irradiation hardening comparable to that of Fe-1 at.%Cu alloy was observed in Fe-1 at.%Mn alloy. Manganese increases the number density of dislocation loops, which contributed to the observed irradiation hardening. Manganese serves as a nucleus of the loop by trapping interstitial atoms and clusters, preventing 1D motion of the loops.     

Welding-induced microstructure in austenitic stainless steels before and after neutron irradiation

The effects of neutron irradiation on the microstructure of welded joints made of austenitic stainless steels have been investigated. The materials were welded AISI 304 and AISI 347, so-called test weld materials, and irradiated with neutrons at 300 °C to 0.3 and 1.0 dpa. In addition, an AISI 304 type from a decommissioned pressurised water reactor, so-called in-service material, which had accumulated a maximum dose of 0.35 dpa at about 300 °C, was investigated. The microstructure of heat-affected zones and base materials was analysed before and after irradiation, using transmission electron microscopy. Neutron diffraction was performed for internal stress measurements. It was found that the heat-affected zone contains, relative to the base material, a higher dislocation density, which relates well to a higher residual stress level and, after irradiation, a higher irradiation-induced defect density. In both materials, the irradiation-induced defects are of the same type, consisting in black dots and Frank dislocation loops. Careful analysis of the irradiation-induced defect contrast was performed and it is explained why no stacking fault tetrahedra could be identified.     

Irradiation damage of nickel in a high-voltage electron microscope

Electron irradiation damage in high-purity annealed and 20% deformed nickel has been studied using a high-voltage electron microscope (HVEM) operating at 650 kV. The effects of temperature of irradiation, electron dose and cold work on point-defect clustering in general and void formation in particular have been investigated. Both faulted and unfaulted dislocation loops were observed during irradiation at 240 to 500°C; multilayer dislocation loops were observed at the higher temperatures. Voids exhibited a cubic shape at low dose with a nearly homogeneous distribution in annealed and an inhomogenous distribution in 20% deformed nickel. The average void size for annealed nickel was larger than that for 20% deformed nickel and the void growth rate was found to be higher for annealed nickel. In annealed nickel, the void concentration increased up to ≈14 dpa and then decreased, while in 20% deformed nickel it increased up to ≈35 dpa. Swelling was considerably reduced by cold work compared to annealed nickel. These observations are discussed with emphasis on the role of dislocation density in the nucleation and growth of voids and swelling.     

Proton irradiation study of GFR candidate ceramics

This work investigated the microstructural response of SiC, ZrC and ZrN irradiated with 2.6 MeV protons at 800 °C to a fluence of 2.75 × 10¹⁹ protons/cm², corresponding to 0.71-1.8 displacement per atom (dpa), depending on the material. The change of lattice constant evaluated using HOLZ patterns is not observed. In comparison to Kr ion irradiation at 800 °C to 10 dpa from the previous studies, the proton irradiated ZrC and ZrN at 1.8 dpa show less irradiation damage to the lattice structure. The proton irradiated ZrC exhibits faulted loops which are not observed in the Kr ion irradiated sample. ZrN shows the least microstructural change from proton irradiation. The microstructure of 6H-SiC irradiated to 0.71 dpa consists of black dot defects at high density.     

Thermal desorption of helium from defects in nickel

Helium atoms, introduced into materials by helium plasma or generated by the (n, α) nuclear reaction, have a strong tendency to accumulate at trapping sites such as vacancy clusters and dislocations. In this paper, the effects of dislocations, single vacancies and vacancy clusters on the retention and desorption of helium atoms in nickel were studied. Low energy (0.1-0.15 keV) helium atoms were implanted in nickel with vacancies or dislocations without causing any displacement damage. He atoms, interstitial-type dislocation loops, and vacancy clusters were also introduced with irradiation damage by 5.0 keV helium ions. Helium thermal desorption peaks from dislocations, helium-vacancy clusters and helium bubbles were obtained by thermal desorption spectroscopy at 940 K, in the range from 900 to 1370 K, and at 1500 K, respectively. In addition, a thermally quasi-stable state was found for helium-vacancy clusters.     

Ar~+辐照Hastelloy C276显微结构演化多尺度模拟研究及实验验证

采用分子动力学结合团簇动力学研究了Hastelloy C276Ni基合金在Ar+辐照(室温,约10dpa)下的显微结构演化机理,开发了多尺度模拟程序Radieff,利用Radieff模拟了在Ar+辐照下C276中间隙位错环和孔洞的形核、长大过程。在武汉大学串列加速器-离子注入机-透射电镜一体化联机装置上开展了115keV Ar+辐照C276验证实验,采用一体化联机透射电镜观察了辐照缺陷尺寸及形貌。不同辐照剂量下位错环尺寸模拟结果与实验结果吻合很好。     

Dislocation loop evolution under ion irradiation in austenitic stainless steels

A solution annealed 304 and a cold worked 316 austenitic stainless steels were irradiated from 0.36 to 5 dpa at 350 °C using 160 keV Fe ions. Irradiated microstructures were characterized by transmission electron microscopy (TEM). Observations after irradiation revealed the presence of a high number density of Frank loops. Size and number density of Frank loops have been measured. Results are in good agreement with those observed in the literature and show that ion irradiation is able to simulate dislocation loop microstructure obtained after neutron irradiation.Experimental results and data from literature were compared with predictions from the cluster dynamic model, MFVIC (Mean Field Vacancy and Interstitial Clustering). It is able to reproduce dislocation loop population for neutron irradiation. Effects of dose rate and temperature on the loop number density are simulated by the model. Calculations for ion irradiations show that simulation results are consistent with experimental observations. However, results also show the model limitations due to the lack of accurate parameters.     

Radiation stability of ZrN under 2.6 MeV proton irradiation

Zirconium nitride is a promising alternative material for the use as an inert matrix for transuranic fuel, but the knowledge of the radiation tolerance of ZrN is very limited. We have studied the radiation stability of ZrN using a 2.6 MeV proton beam at 800 °C. The irradiated microstructure and hardening were investigated and compared with annealed samples. A high density of nano-sized defects was observed in samples irradiated to doses of 0.35 and 0.75 dpa. Some defects were identified as vacancy-type pyramidal dislocation loops using lattice resolution imaging and Fourier-filter image processing. A very slight lattice expansion was noted for the sample with a dose of 0.75 dpa. Hardening effects were found for samples irradiated to both 0.35 and 0.75 dpa using Knoop indentation.     

Microstructural stability of an HT-9 fuel assembly duct irradiated in FFTF

To explore whether the known resistance of fully tempered HT-9 to neutron-induced phase instability and void swelling are maintained under realistic time-dependent reactor operating conditions, the radiation-induced microstructure of an HT-9 ferritic/martensitic hexagonal duct was examined following a 6-year irradiation campaign of a fuel assembly in the Fast Flux Test Reactor Facility (FFTF). Microscopy examination was conducted on specimens irradiated to 4 dpa at 505 °C, 28 dpa at 384 °C and 155 dpa at 443 °C where quoted temperatures are the average operating temperatures over the lifetime of the duct.The dislocation and phase microstructure were observed to remain relatively unchanged at 4 dpa at 505 °C, but significant microstructural changes were observed to have occurred at 28 and 155 dpa and 384 and 443 °C respectively. At these doses the microstructures have experienced precipitation and formation of interstitial loops. In addition, void swelling had occurred at 155 dpa with an average swelling of ∼0.3%, although some local areas swelled as much as 1.2%. In general it appears that this alloy retains its swelling resistance under typical reactor operation conditions up to 155 dpa.     

Reaction kinetics analysis of damage evolution in accelerator driven system beam windows

Reaction kinetic analysis was used to estimate the damage evolution in window materials of 800 MWth accelerator driven system (ADS). Parameters were fitted to F82H of the STIP-II experiment at 673 K and EC316LN of the STIP-I experiment at 626 K. In F82H, the concentration of bubbles was almost constant and the bubble size increased, while the concentration of interstitial type dislocation loops increased and their size was constant between 2.1 × 10⁻³ and 210 dpa. EC316LN showed almost the same behavior. Swelling increased almost linearly with irradiation dose above 0.21 dpa between 673 K and 773 K.     

Irradiation damage in 304 and 316 stainless steels: experimental investigation and modeling. Part I: Evolution of the microstructure

Irradiation damage in three austenitic stainless steels, SA 304L, CW 316 and CW Ti-modified 316, is investigated both experimentally and theoretically. The density and size of Frank loops after irradiation at 320 and 375 °C in experimental EBR II, BOR-60 and OSIRIS reactors for doses up to 40 dpa are characterized by TEM. The evolution of the initial dislocation network under irradiation is evaluated. A cluster dynamics model is proposed to account quantitatively for the experimental findings.     

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 = 1 MeV) W, W-3Re, W-3Os and 0.15 dpa neutron-irradiated (E > 1 MeV) 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.     

Understanding silicon-rich phase precipitation under irradiation in austenitic stainless steels

Atom probe samples have been Fe⁺ ion irradiated at different doses (from 0.5 to 10 dpa) and different temperatures (between 300 and 400 °C) in order to understand the mechanism of formation, under irradiation, of Si-rich phases in austenitic stainless steels. Atom probe results show the presence of Si-enriched clusters which can also be enriched in Ni and depleted in Cr. Number densities of solute clusters can be linked to number densities of dislocation loops already observed by transmission electron microscopy in a previous work. This suggests that solute clusters are formed by heterogeneous precipitation on dislocation loops. Furthermore, the evolution of the composition of solute clusters as a function of the irradiation temperature is consistent with a radiation-induced mechanism. Results are also compared with previous results obtained after neutron irradiation at lower dose rate (in term of dpa s⁻¹). The comparison is, here again, consistent with the radiation-induced mechanism. Thus, Si-rich clusters may be formed by radiation-induced segregation to dislocation loops. Results also show that Si is probably dragged to sinks via the interstitial mechanism.     

Defect structure of irradiated PH13-8Mo steel

PH13-8Mo bolts, which are considered for use in the ITER reactor, were irradiated up to doses of 0.5, 1 and 2 dpa. The microstructure was investigated with transmission electron microscopy and its evolution is discussed with reference to the mechanical properties. PH13-8Mo is a precipitation hardened martensitic steel, but a large amount of austenite has been observed as well. The precipitation hardening results from the formation of small coherent NiAl precipitates in the martensite phase. Their size, size distribution and density are found to be unaffected by neutron irradiation. The dislocations in the martensite phase are mainly a/2〈1 1 1〉 type screw dislocations, whereas in the austenite phase mainly a/2〈1 1 0〉 type screw dislocations are present. The line dislocation structure did not change during irradiation, but small irradiation induced defects were observed. Using the Orowan model, it is argued that the latter are responsible for the irradiation hardening.     

Precipitate coarsening,redistribution and renucleaton during irradiation of Ni-6.35 wt.% Al

Precipitation during irradiation was investigated using 3.2-MV ⁵⁸Ni⁺ ions incident on Ni-6.35 wt.% Al at 650°C. The resultant γ' precipitate morphologies were observed using TEM. A radiation-enhanced coarsening regime is quantitatively demonstrated at doses ?20 dpa. The enhanced coarsening regime terminates by preferential interaction of dislocations with larger precipitates. Segregation of nickel to dislocations causes these large particles to dissolve and leads to renucleation of precipitates and un increase in the density of small particles. Consistent with this, maxima of γ' size as a function of dose are observed at 650 and 550°C.Biconvex lens shape, precipitate-free zones were observed and their growth kinetics were followed to 15 dpa. The zone diameter grows linearly at a rate of 0.21 nm/s which is consistent with dislocation loop growth. The precipitate-free zones are also caused by nickel segregation to dislocations, increase in size until maxima γ' mean size occurs, and then are obliterated due to γ' renucleation within the zone. The presence of a precipitate-free zone at the irradiated surface is confirmed. Mechanisms for the various phenomena are described.     

Swelling and microstructure of austenitic stainless steel ChS-68 CW after high dose neutron irradiation

Austenitic stainless steel ChS-68 serving as fuel pin cladding was irradiated in the 20% cold-worked condition in the BN-600 fast reactor in the range 56-84 dpa. This steel was developed to replace EI-847 which was limited by its insufficient resistance to void swelling. Comparison of swelling between EI-847 and ChS-68 under similar irradiation conditions showed improvement of the latter steel by an extended transient regime of an additional ∼10 dpa. Concurrent with swelling was the development of a variety of phases. In the temperature range 430-460 °С where the temperature peak of swelling was located, the principal type of phase generated during irradiation was G-phase, with volume fraction increasing linearly with dose to ∼0.5% at 84 dpa. While the onset of swelling is concurrent with formation of G-phase, the action of G-phase cannot be confidently ascribed to significant removal from solution of swelling-suppressive elements such as silicon. A plausible mechanism for the higher resistance to void swelling of ChS-68 as compared with EI-847 may be related to an observed higher stability of faulted dislocation loops in ChS-68 that impedes the formation of a glissile dislocation network. The higher level of boron in ChS-68 is thought to be one contributor that might play this role.     

The role of irradiated microstructure in the localized deformation of austenitic stainless steels

Localized deformation has emerged as a potential factor in irradiation-assisted stress corrosion cracking of austenitic stainless steels in LWR environments and the irradiated microstructure may be a critical factor in controlling the degree of localized deformation. Seven austenitic alloys with various compositions were irradiated using 2-3 MeV protons to doses of 1 and 5 dpa at 360 °C. The irradiated microstructure consisting of dislocation loops and voids was characterized using transmission electron microscopy. The degree of localized deformation was characterized using atomic force microscopy on the deformed samples after conducting constant extension rate tension tests to 1% and 3% strain in argon. Localized deformation was found to be dependent on the irradiated microstructure and to correlate with hardening originating from dislocation loops. Dislocation loops enhance the formation of dislocation channels and localize deformation into existing channels. On the contrast, voids mitigate the degree of localized deformation. The degree of localized deformation decreases with SFE with the exception of alloy B. Localized deformation was found to have similar dependence on SFE as loop density suggesting that SFE affects localized deformation by altering irradiated microstructure.     

Emulation of neutron irradiation effects with protons: validation of principle

This paper presents the results of the irradiation, characterization and irradiation assisted stress corrosion cracking (IASCC) behavior of proton- and neutron-irradiated samples of 304SS and 316SS from the same heats. The objective of the study was to determine whether proton irradiation does indeed emulate the full range of effects of in-reactor neutron irradiation: radiation-induced segregation (RIS), irradiated microstructure, radiation hardening and IASCC susceptibility. The work focused on commercial heats of 304 stainless steel (heat B) and 316 stainless steel (heat P). Irradiation with protons was conducted at 360 °C to doses between 0.3 and 5.0 dpa to approximate those by neutron irradiation at 275 °C over the same dose range. Characterization consisted of grain boundary microchemistry, dislocation loop microstructure, hardness as well as stress corrosion cracking (SCC) susceptibility of both un-irradiated and irradiated samples in oxygenated and de-oxygenated water environments at 288 °C. Overall, microchemistry, microstructure, hardening and SCC behavior of proton- and neutron-irradiated samples were in excellent agreement. RIS analysis showed that in both heats and for both irradiating particles, the pre-existing grain boundary Cr enrichment transformed into a ‘W' shaped profile at 1.0 dpa and then into a ‘V' shaped profile between 3.0 and 5.0 dpa. Grain boundary segregation of Cr, Ni, Si, and Mo all followed the same trends and agreed well in magnitude. The microstructure of both proton- and neutron-irradiated samples was dominated by small, faulted dislocation loops. Loop size distributions were nearly identical in both heats over a range of doses. Saturated loop size following neutron irradiation was about 30% larger than that following proton irradiation. Loop density increased with dose through 5.0 dpa for both particle irradiations and was a factor of 3 greater in neutron-irradiated samples vs. proton-irradiated samples. Grain boundary denuded zones were only observed in neutron-irradiated samples. No cavities were observed for either irradiating particle. For both irradiating particles, hardening increased with dose for both heats, showing a more rapid increase and approach to saturation for heat B. In normal oxygenated water chemistry (NWC) at 288 °C, stress corrosion cracking in the 304 alloy was first observed at about 1.0 dpa and increased with dose. The 316 alloy was remarkably resistant to IASCC for both particle types. In hydrogen treated, de-oxygenated water (HWC), proton-irradiated samples of the 304 alloy exhibited IG cracking at 1.0 dpa compared to about 3.0 dpa for neutron-irradiated samples, although differences in specimen geometry, test condition and test duration can account for this difference. Cracking in heat P in HWC occurred at about 5.0 dpa for both irradiating particles. Thus, in all aspects of radiation effects, including grain boundary microchemistry, dislocation loop microstructure, radiation hardening and SCC behavior, proton-irradiation results were in good agreement with neutron-irradiation results, providing validation of the premise that the totality of neutron-irradiation effects can be emulated by proton irradiation of appropriate energy.     

The effect of neutron irradiation on mechanical properties of YAG laser weldments using previously irradiated material

The objective of this study is to make clear the effect of neutron irradiation on mechanical properties of laser weldments using irradiated material. This estimation is necessary for the application to joining coolant piping of the ITER blanket. Irradiation testing was performed at Japan Material Testing Reactor (JMTR). On the irradiation condition for weldments using irradiated material, fast neutron fluence was 1.4 × 10²⁴ n/m², which corresponds to a displacement damage rate of 0.26 displacement per atom (dpa) and irradiation temperature 200 °C. The results of this study show that tensile properties of all weldments changed into that of base material by the effect of neutron irradiation. The results of hardness tests show that irradiation hardening at an irradiation damage dose of 0.3 dpa is almost same as that at irradiation damage 0.6 dpa. It is concluded that irradiated weldments using irradiated material were moved toward irradiated base material on tensile and hardness properties up to 0.6 dpa. On the other hand, tensile properties of base material were changed by the effect of neutron irradiation up to about 0.3 dpa, and with much less change from 0.3 dpa to 0.6 dpa. It is inferred that the effect of neutron irradiation of SS316LN-IG almost saturated up to 0.3 dpa.     

Microstructural analysis of an HT9 fuel assembly duct irradiated in FFTF to 155 dpa at 443 °C

The majority of data on the irradiation response of ferritic/martensitic steels has been derived from simple free-standing specimens irradiated in experimental assemblies under well-defined and near-constant conditions, while components of long-lived fuel assemblies are more complex in shape and will experience progressive changes in environmental conditions. To explore whether the resistance of HT9 to void swelling is maintained under more realistic operating conditions, the radiation-induced microstructure of an HT9 ferritic/martensitic hexagonal duct was examined following a six-year irradiation of a fuel assembly in the Fast Flux Test Reactor Facility (FFTF). The calculated irradiation exposure and average operating temperature of the duct at the location examined were ∼155 dpa at ∼443 °C. It was found that dislocation networks were predominantly composed of (a/2)<1 1 1> Burgers vectors. Surprisingly, for such a large irradiation dose, type a<1 0 0> interstitial loops were observed. Additionally, a high density of precipitation occurred. These two microstructural characteristics may have contributed to the rather low swelling level of 0.3%.