Influence of SHI irradiation on the formation of buried SiC

Silicon carbide (SiC) precipitates buried in Si(1 0 0) substrates were synthesized by ion implantation of 50 keV and 150 keV C⁺ ions at different fluences. Two sets of samples were subsequently annealed at 850 °C and 1000 °C for 30 min. Fourier transform infrared (FTIR) spectroscopy studies and X-ray diffraction (XRD) analysis confirmed formation of β-SiC precipitates in the samples. Ion irradiation with 100 MeV Ag⁷⁺ ions at room temperature does not induce significant change in the precipitates. It could be interpreted from the FTIR observations that ion irradiation may induce nucleation in Si + C solution created by ion implantation of C in Si. Modifications induced by swift heavy ion irradiation are found to be dependent on implantation energy of C⁺ ions.     

Microstructural evolution in medium copper low alloy steels irradiated in a pressurized water reactor and a material test reactor

A533B steels containing 0.12% and 0.16% Cu were irradiated to 3×10²³ and 6×10²³ n/m² (E>1 MeV) at 290 °C in a pressurized water reactor (PWR) and a material test reactor (MTR). Microstructural changes were examined using atom probe, small angle neutron scattering, field emission gun scanning transmission electron microscopy and post-irradiation annealing (PIA) coupled with positron annihilation (PA) and hardness testing (Hv). Cu rich precipitates had a Cu enriched core with surrounding Ni, Mn and Si rich region and the atomic composition was Fe–(7–16)Cu–(2–8)Mn–(0–4)Ni–(0–4)Si. The size and number density of Cu rich precipitates and the residual Cu concentration in matrix were almost saturated at above 3×10²³ n/m². Low flux irradiation in PWR produced slightly larger precipitates of a lower density with a higher Cu concentration in the precipitates. PIA (PA and Hv) examination showed that vacancy type matrix defects after PWR irradiation were more stable and more effective for hardening than those after MTR irradiation.     

Evolution of the nanostructure of VVER-1000 RPV materials under neutron irradiation and post irradiation annealing

A high nickel VVER-1000 (15Kh2NMFAA) base metal (1.34 wt% Ni, 0.47% Mn, 0.29% Si and 0.05% Cu), and a high nickel (12Kh2N2MAA) weld metal (1.77 wt% Ni, 0.74% Mn, 0.26% Si and 0.07% Cu) have been characterized by atom probe tomography to determine the changes in the microstructure during neutron irradiation to high fluences. The base metal was studied in the unirradiated condition and after neutron irradiation to fluences between 2.4 and 14.9 × 10²³ m⁻² (E > 0.5 MeV), and the weld metal was studied in the unirradiated condition and after neutron irradiation to fluences between 2.4 and 11.5 × 10²³ m⁻² (E > 0.5 MeV). High number densities of ∼2-nm-diameter Ni-, Si- and Mn-enriched nanoclusters were found in the neutron irradiated base and weld metals. No significant copper enrichment was associated with these nanoclusters and no copper-enriched precipitates were observed. The number densities of these nanoclusters correlate with the shifts in the ΔT_{41 J} ductile-to-brittle transition temperature. These nanoclusters were present after a post irradiation anneal of 2 h at 450 °C, but had dissolved into the matrix after 24 h at 450 °C. Phosphorus, nickel, silicon and to a lesser extent manganese were found to be segregated to the dislocations.     

Evidence for neutron irradiation-induced metallic precipitates in model alloys and pressure-vessel weld steel

We have used positron Doppler-broadening spectroscopy to examine a series of neutron-irradiated model alloys (1 × 10²³ n/m², E>0.5 MeV) and 73W-weld steel (to 1.8 × 10²³ n/m², E>1 MeV. The copper, nickel and phosphorus content of the model alloys was systematically varied. The samples were examined in the as-irradiated state and after post-irradiation isochronal anneals to temperature up to 600 °C. By following the S and W parameters, and especially by plotting the results in (S,W) space, we can infer that the damage is a combination of irradiation-induced metallic precipitates and vacancy-type defect clusters. Samples with either high Cu or with a combination of high Ni and medium Cu (and the pressure-vessel weld steel) showed evidence for both irradiation-induced metallic precipitation, and vacancy-type clusters, while samples without either high Ni or high Cu showed predominantly evidence of annihilations in vacancy-type clusters. These results are discussed in terms of embrittlement models.     

Surface morphology of the Ar ion-irradiated Ag/Si(1 1 1) system

In the present study, a 500 Å thin Ag film was deposited by thermal evaporation on 5% HF etched Si(1 1 1) substrate at a chamber pressure of 8×10⁻⁶ mbar. The films were irradiated with 100 keV Ar⁺ ions at room temperature (RT) and at elevated temperatures to a fluence of 1×10¹⁶ cm⁻² at a flux of 5.55×10¹² ions/cm²/s. Surface morphology of the Ar ion-irradiated Ag/Si(1 1 1) system was investigated using scanning electron microscopy (SEM). A percolation network pattern was observed when the film was irradiated at 200°C and 400°C. The fractal dimension of the percolated pattern was higher in the sample irradiated at 400°C compared to the one irradiated at 200°C. The percolation network is still observed in the film thermally annealed at 600°C with and without prior ion irradiation. The fractal dimension of the percolated pattern in the sample annealed at 600°C was lower than in the sample post-annealed (irradiated and then annealed) at 600°C. All these observations are explained in terms of self-diffusion of Ag atoms on the Si(1 1 1) substrate, inter-diffusion of Ag and Si and phase formations in Ag and Si due to Ar ion irradiation.     

In-reactor uniaxial tensile testing of pure copper at a constant strain rate at 90 °C

Recently, annealed specimens of pure copper have been tensile tested in a fission reactor at a damage rate of 6 × 10⁻⁸ dpa/s with a constant strain rate of 1.3 × 10⁻⁷ s⁻¹. The specimen temperature during the test was about 90 °C. The stress response was continuously recorded as a function of irradiation time (i.e. displacement dose and strain). The experiment lasted for 308 h. During the dynamic in-reactor test, the specimen deformed and hardened homogeneously without showing any sign of yield drop and plastic instability. However, the specimen yielded a uniform elongation of only about 12%. The preliminary results are briefly described and discussed in the present note.     

Structural stability of Cu nanocrystals in SiO2 exposed to high-energy ion irradiation

Cu nanocrystals (NCs) were synthesized in SiO₂ by ion implantation and thermal annealing. Annealing at two different temperatures of 950 °C and 650 °C yielded two different nanocrystal size distributions with an average diameter of 8.1 and 2.5 nm, respectively. Subsequently the NCs were exposed to 5.0 MeV Sn³⁺ ion irradiation simultaneously with a thin Cu film as a bulk reference. The short-range atomic structure and average NC diameter was measured by means of extended X-ray absorption fine structure (EXAFS) spectroscopy and small angle X-ray scattering (SAXS), respectively. Consistent with the high regeneration rate of bulk elemental metals, no irradiation induced defects were observed for the reference, whereas the small NCs (2.5 nm) were dissolved as Cu monomers in the matrix. The latter was attributed to irradiation-induced mixing of Cu, Si and O based on dynamic binary collision simulations. For the large NCs (8.1 nm) only minor structural changes were observed upon irradiation, consistent with a more bulk-like pre-irradiation structure.     

Cluster formation, evolution and size distribution in Fe-Cu alloy: Analysis by XAFS, XRD and TEM

Fe-Cu alloys containing 1.3 at.% copper were studied as model systems for cluster formation in reactor pressure vessel steels. The samples were annealed at 775 K for different times and subsequently analyzed using X-ray absorption fine structure spectroscopy at the Cu-K-edge, X-ray diffraction and transmission electron microscopy. The results show that copper cluster formation might occur even with short annealing times. These clusters of about 1 nm size can switch easily from bcc iron-like structures to fcc copper, if the local copper concentration is high enough. While a short annealing time of 2.5 h at 775 K maintains a good dilution of copper in the bcc iron matrix, annealing for 312 h leads to large fcc copper precipitates. A linear combination analysis suggests that in the sample annealed 8 h, copper clusters are mostly formed with the same structure as the matrix. A co-existence of bcc and fcc clusters is obtained for 115 h of annealing. Transmission electron microscopy indicates the presence of precipitates as large as 60 nm size for an annealing time of 312 h, and X-ray diffraction provided complementary data about the clusters size distributions in all of the four samples.     

Optical behaviour of Er doped rutile by ion implantation

Rutile single crystals were implanted at room temperature with fluences of 5 × 10¹⁵ Er⁺/cm² ions with 150 keV energy. Rutherford backscattering/channeling along the 0 0 1 axis reveals complete amorphization of the implanted region. Photoluminescence reveals the presence of an optical centre close to the intra-ionic emission of Er³⁺ in the as-implanted samples. After annealing at 800 °C in air no changes were observed in the aligned RBS spectrum. On the contrary, annealing in reducing atmosphere (vacuum) induces the epitaxy of the damage layer. These results are unexpected, since for implantations of other ions under the same conditions, epitaxial recrystallization of the damage region occurs at this temperature. On the other hand, photoluminescence studies show the presence of new Er-related optical centres with high thermal stability in the samples annealed under oxidizing conditions. Annealing at 1000 °C in vacuum leads to the complete recrystallization of the damaged region. At this temperature a large fraction of Er segregates to the surface.     

Irradiation induced growth of CoSi2 precipitates in Si at 650°C: An in situ study

Crystalline Si samples were implanted at 350°C with 50 keV Co⁺ ions to a fluence of 10¹⁵ Co cm⁻². Small CoSi₂ precipitates were formed. We studied the precipitate growth, via in situ transmission electron microscopy, under irradiation with 100 keV Si ions at 650°C. We deduce the precipitate growth processes involved. Irradiation-induced (or enhanced) Ostwald ripening is the main growth mechanism. We also find an instability of the B-type precipitates, which leads to their transformation into A-type precipitates above a critical size. These preliminary results show that direct comparisons with kinetic Monte Carlo modelling of the precipitate growth is at hand.     

Release of tritium from boron carbide irradiated with reactor neutrons

The release of tritium from irradiated boron carbide in a pure Ar atmosphere was investigated between 500 and 900°C. The sintered B₄C samples with densities between 75 and 95% of the theoretical density were irradiated with reactor neutrons with total neutron doses up to 5 × 10²⁰/cm². Effective diffusion coefficients, D_eff, were derived from the release data using the model “diffusion out of a sphere”. D_eff decreases by about 3 orders of magnitude with increasing total neutron dose, levels off at about 10¹⁸n/cm² and increases at very high doses ( > 10²⁰ n/cm²). The decrease in the tritium mobility is attributed to the radiation defects formed in the B₄C. The activation energy of 210 ± 30 kJ/mol for the tritium diffusion in the irradiated B₄C is much higher than the value found for unirradiated material. D_eff depends also very strongly on the density of the sintered material.     

Effect of neutron-irradiation on the spinodal decomposition of Fe-32% Cr model alloy

A comparison has been performed between the extent of decomposition found in an Fe-32% Cr alloy that was both neutron-irradiated at 290°C and thermally aged for the equivalent time at the same temperature. In addition, these results are compared to data obtained from a series of specimens isothermally aged at 470 and 500°C. These atom probe results indicate that neutron irradiation has a significant effect on both the kinetics of decomposition and the morphology of the chromium-enriched ′ phase. The results are consistent with the neutron irradiation significantly changing the location of the phase field in the phase diagram.     

Defect evolution during ion bombardment of amorphous Si probed by in situ conductivity measurements

In this work we use in-situ conductivity measurements during ion irradiation as a sensitive probe of the defect structure of amorphous Si. Electronic transport in amorphous Si occurs by hopping at the high density ( 10²⁰ cm⁻³ eV⁻¹) of deep lying localized states introduced by the defects in the band gap. In-situ conductivity measurements allow to follow directly the defect generation and annihilation kinetics during and after ion bombardment of the material. Amorphous Si layers, patterned to perform conductivity measurements, were annealed at 500°C in order to reduce the defect density by about a factor of 5. Defects were subsequently reintroduced by high energy ion irradiation at different temperatures (77–300 K). During irradiation the conductivity of the layer increases by several orders of magnitude and eventually saturates. Turning off the beam results in a decrease of the conductivity by a factor of 2 in times as long as a few hours even at 77 K. The effects of different ions (He, C, Si, Cu, and Au) and different ion fluxes (10⁹–10¹² ions/cm² s) on these phenomena have been explored. These data give a hint on the mechanisms of defect production and annihilation and demonstrate a strong correlation between electrical and structural defects in amorphous silicon.     

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.     

Post-irradiation mechanical properties of three EUROFER97 joints

Within the development of reduced activation ferritic martensitic (RAFM) steels as prominent structural materials for future fusion reactors, EUROFER97 has recently emerged in Europe as the reference material for the DEMO design. In order to characterise the in-service performance of EUROFER97 as structural material, it is important to assess the properties of welded joints, particularly under irradiation. In the present paper, three EUROFER97 joints (two diffusion welds and one TIG weld) have been irradiated in the BR2 reactor of SCK-CEN at 300 °C up to 1.8 dpa and subsequently characterised for tensile, impact and fracture toughness properties. Comparisons of the results are provided with base EUROFER97 (both unirradiated and irradiated under similar conditions) and, where available, with properties measured on the joints in the unirradiated condition. The post-irradiation mechanical behaviour of both diffusion joints (“laboratory” and “mock-up”) appears similar to that of the base material; therefore, diffusion joining looks a very promising technique. On the other hand, the properties of the TIG joint are affected by the lack of a post-weld heat treatment, which causes the material from the upper part of the weld to be significantly worse than that of the lower region.     

A swelling model for stoichiometric SiC at temperatures below 1000°C under neutron irradiation

A simple phenomenological model for the saturation swelling below 1000°C of neutron-irradiated silicon carbide (SiC) is presented in this paper. Under fast neutron irradiation, SiC is known to undergo volumetric expansion (swelling) which quickly saturates at a fast fluence of approximately 10²⁵ n/m² for irradiation temperatures below 1000°C. A previous model due to Balarin attributes swelling to lattice dilation as a result of single point defects. We show in this paper that the experimentally observed linear temperature dependence of saturation swelling can be explained in terms of the formation and growth of small interstitial clusters, resulting directly from collision cascades initiated by energetic neutrons. These loops grow by absorption of mobile carbon interstitials and their composition is subject to stoichiometry constraints, requiring absorption of slower silicon interstitials. Because of cascade re-solution events, the density of loops decreases sharply with temperature as a result of overlap of cascades with larger size loops at higher temperatures. The average radius of these loops increases with temperature. Volumetric swelling is shown to obey a linear temperature dependence as a consequence of the strong decrease in density and the simultaneous increase in average radius, and to saturate with fluence. The model is shown to be consistent with experimental observations. In the temperature range below 500–600°C, swelling seems to be dominated by single point defects, or defect clusters containing only a few atoms, in accordance with the explanation offered by Balarin.     

The effects of large concentrations of helium on the mechanical properties of neutron-irradiated stainless steel

Tensile and creep properties have been determined on specimens of type 316 stainless steel irradiated in the High Flux Isotope Reactor in the range 380 to 785°C. Irradiation of type 316 in this reactor partially simulates fusion reactor irradiation, with displacement damage levels up to 120 dpa and helium contents up to 6000 appm achieved in two years. Samples irradiated in the annealed condition to about 100 dpa and 4000 appm helium showed an increased yield strength between 350 and 600°C and, except at 350°C, a reduced ultimate tensile strength compared with values for the unirradiated material. Samples irradiated in the 20%-cold-worked condition showed decreases in both yield and ultimate tensile strengths at all test temperatures. The irradiated samples of both annealed and cold-worked material exhibited little strain hardening, and total elongations were small and became zero,for tests at 650° C. Tensile tests at 575°C and creep-rupture tests at 550°C showed strong effects of fluence on strength and ductility for helium contents above about 30 appm. Optical metallography showed extensive carbide precipitation at all temperatures and precipitation of a second phase, believed to be sigma, at the higher temperatures.     

Effect of nickel content of 9Cr-1MoVNb and 12Cr-1MoVW steels on the aging and irradiation response of impact properties

Irradiating nickel-containing alloys in a mixed-spectrum reactor can simulate both transmutation helium and displacement damage expected in a fusion reactor first wall. Impact properties of 9Cr-1MoVNb and 12Cr-1MoVW steels doped with nickel were determined in the as-heat-treated, thermally aged, and irradiated conditions to determine if nickeldoping affects the behavior. The irradiation was carried out in a fast-spectrum reactor which produces only an insignificant amount of helium during irradiation, thereby evaluating the effect of nickel alone. Only limited property changes resulted from thermal aging or irradiation to 12 dpa at 450 to 550°C. Irradiation of the 12Cr-1MoVW steel at 390°C produced severe degradation of impact properties. Nickel additions affected the unirradiated material properties, but subsequent radiationinduced changes were similar. The results indicate that nickel doping and subsequent irradiation in a mixed spectrum reactor is a viable method for simulating irradiation effects in a fusion reactor first wall.     

Influence of post irradiation annealing on the mechanical properties and defect structure of AISI 304 steel

The effect of post irradiation annealing on the mechanical properties and the radiation induced defect structure was investigated on stainless steel, of type AISI 304, that was irradiated up to 24 dpa in the decommissioned Chooz A reactor. The material was investigated both in the as-irradiated state as well as after post irradiation annealing. In the as-irradiated specimen the typical radiation induced defects were found as well as γ′-precipitates (Ni₃Si). Annealing at 400 °C had almost no effect on the radiation induced defects, but annealing at 500 °C resulted in the immediate unfaulting of the Frank loops. As to the mechanical properties, annealing at 400 °C did not strongly affect the yield strength and the ductility of the material, although the fraction of intergranular fracture during slow strain rate tensile tests under pressurised water reactor conditions, was significantly reduced. Annealing at 500 °C did reduce the yield strength and restored substantially the ductility and the strain hardening capability of the material. The microstructure investigated by transmission electron microscopy correlates to the mechanical test results. It was found that the observed defect changes after post irradiation annealing provide a reasonable explanation for the observed changes of the mechanical properties.     

The effects of low dose rate irradiation and thermal aging on reactor structural alloys

As part of the EBR-II reactor materials surveillance program, test samples of fifteen different alloys were placed into EBR-II in 1965. The surveillance (SURV) program was intended to determine property changes in reactor structural materials caused by irradiation and thermal aging. In this work, the effect of low dose rate (approximately 2 × 10⁻⁸ dpa/s) irradiation at 380–410°C and long term thermal aging at 371°C on the properties of 20% cold worked 304 stainless steel, 420 stainless steel, Inconel X750, 304/308 stainless weld material, and 17-4 PH steel are evaluated. Doses of up to 6.8 dpa and thermal aging to 2994 days did not significantly affect the density of these alloys. The strength of 304 SS, X750, 17-4 PH, and 304/308 weld material increased with irradiation. In contrast, the strength of 420 stainless steel decreased with irradiation. Irradiation decreased the impact energy in both Inconel X750 and 17-4 PH steel. Thermal aging decreased the impact energy in 17-4 PH steel and increased the impact energy in Inconel X750. Tensile property comparisons of 304 SURV samples with 304 samples irradiated in EBR-II at a higher dose rate show that the higher dose rate samples had greater increases in strength and greater losses in ductility.