Influence of silicon on swelling and microstructure in Russian austenitic stainless steel EI-847 irradiated to high neutron doses

Void swelling and microstructural development of niobium-stabilized EI-847 austenitic stainless steel with a range of silicon levels were investigated by destructive examination of fuel pin cladding irradiated in three fast reactors located in either Russia or Kazakhstan. The tendency of void swelling to be progressively reduced by increasing silicon concentration appears to be a very general phenomenon in this steel, whether observed in simple, single-variable experiments on well-defined materials or when observed in multivariable, time-dependent irradiations conducted on commercially produced steels over a wide range of irradiation temperatures, neutron spectra and dpa rates. The role of silicon on microstructural development is expressed both in the solid solution via its influence on dislocation and void microstructure and via its influence on formation of radiation-induced phases that in turn alter the matrix composition. Surprisingly, increases in silicon level in this study do not accelerate the formation of silicon-rich G-phase, but act to increase the formation of Nb (C,N) precipitates. Such precipitates are known to be associated with delayed void swelling.     

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.     

Swelling and radiation creep of austenitic corrosion-resistant steel irradiated by neutrons in wide dose and temperature ranges

The results of a study of the swelling and in-reactor creep of EI-847, EP-172, and ChS-68 austenitic steel after irradiation in materials science assemblies in the range 330–700°C and damaging dose 20–96 dpa are presented. The temperature dependences of the volume change of steel were obtained from measurements of the diameter of unloaded ampuls. It is shown that the swelling of the steel increases linearly with increasing tangential stress. The modulus of in-reactor creep in the interval 410–630°C for the steel investigated in the cold-deformed state varies in the range (0.5–3)·10^–6 MPa^–1·dpa^–1. For lower and higher temperatures, the creep modulus increases to (5–8)·10^–6 MPa^–1·dpa^–1.     

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%.     

Void swelling in a 9Cr ferritic/martensitic steel irradiated with energetic Ne-ions at elevated temperatures

In this work the void swelling behavior of a 9Cr ferritic/martensitic steel irradiated with energetic Ne-ions is studied. Specimens of Grade 92 steel (a 9%Cr ferritic/martensitic steel) were subjected to an irradiation of ²⁰Ne-ions (with 122 MeV) to successively increasing damage levels of 1, 5 and 10 dpa at a damage peak at 440 and 570 °C, respectively. And another specimen was irradiated at a temperature ramp condition (high flux condition) with the temperature increasing from 440 up to 630 °C during the irradiation. Cross-sectional microstructures were investigated with a transmission electron microscopy (TEM). A high concentration of cavities was observed in the peak damage region in the Grade 92 steel irradiated to 5 dpa, and higher doses. The concentration and mean size of the cavities showed a strong dependence on the dose and irradiation temperature. Enhanced growth of the cavities at the grain boundaries, especially at the grain boundary junctions, was observed. The void swelling behavior in similar 9Cr steels irradiated at different conditions are discussed by using a classic void formation theory.     

The influence of silicon and helium on the microstructural stability of ion-irradiated incoloy DS alloys

In order to test their void swelling behaviour under irradiation, several alloys based on the solid-solution nickel alloy Incoloy DS (18Cr-38Ni-Fe) with additions of 0.05, 0.43, 0.92 and 2.24 wt% Si have been studied using 4 MeV helium and 46 MeV nickel ion irradiation in the Harwell VEC. For irradiations of 60 dpa with 10 appm He the void swelling decreased from ˜ 0.9% to negligible levels with increasing silicon content. After irradiation to 90 dpa following injection with 10 appm He the 2.24% Si alloy showed <1.0% swelling at an apparent peak swelling temperature of 625° C. This alloy was subsequently irradiated to check the swelling response with concentrations of helium and hydrogen appropriate to fusion-reactor conditions. Following irradiation to 60 dpa after 1000 appm He injection the swelling peak was shifted to 575° C where a swelling maximum of 4% was observed. At 625° C with 1000 appm He alone, swelling was 2.0% compared with 1.2% in samples injected with 1000 appm He +1000 appm H. This small reduction in swelling was associated with a higher cavity (bubble) concentration in the hydrogen implanted sample. Fine-scale precipitation of Ni₃Si(γ'), η-carbide and G-phase was observed after irradiation together with helium bubbles attached to the η- and G-phase precipitates. The precipitation and void swelling was significantly greater in irradiated samples containing 1000 appm He than in those with 10 appm and irradiated to 90 dpa. It is concluded that although the Incoloy DS alloy possibly has a potential for fission-reactor core applications it has little to commend it for fusion-reactor use where the high swelling response, microstructural instability and likely long-term induced activation arising from the higher nickel content are clearly undesirable factors.     

Dose and Temperature Dependence of Void Swelling in Electron Irradiated Stainless Steel

In a high voltage electron microscope, solution treated Type 316 stainless steel was electron-irradiated at temperatures in the range of 370–630°C to a dose of about 30 dpa. The swelling (ΔV/V) induced by the irradiation beyond about 5 dpa is well described by an empirical equation, ΔV/V=A(dpa) ⁿ , under constant void and dislocation densities. With increasing irradiation temperature, the fluence exponent n increases and the pre-exponent term A decreases. At 550°C irradiation, the fluence exponent takes the value of 1.5 due to the diffusion-limited void growth. The value of n larger than 1.5 at higher temperature (>550°C) is attributable to the surface reaction-limited void growth. The smaller value of n for the low temperature (?500°C) irradiation appears to arise from the dislocation assisted vacancy diffusion. The peak swelling temperature of the specimen irradiated to 30 dpa is about 570°C, which shifts to a higher temperature with increase in electron dose.     

Very high swelling and embrittlement observed in a Fe–18Cr–10Ni–Ti hexagonal fuel wrapper irradiated in the BOR-60 fast reactor

The highest void swelling level ever observed in an operating fast reactor component has been found after irradiation in BOR-60 with swelling in Kh18H10T (Fe–18Cr–10Ni–Ti) austenitic steel exceeding 50%. At such high swelling levels the steel has reached a terminal swelling rate of 1%/dpa after a transient that depends on both dpa rate and irradiation temperature. The transient duration at the higher irradiation temperatures is as small as 10–13 dpa depending on which face was examined. When irradiated in a fast reactor such as BOR-60 with a rather low inlet temperature, most of the swelling occurs above the core center-plane and produces a highly asymmetric swelling loop when plotted vs. dpa. Voids initially harden the alloy but as the swelling level becomes significant the elastic moduli of the alloy decreases strongly with swelling, leading to the consequence that the steel actually softens with increasing swelling. This softening occurs even as the elongation decreases as a result of void linkage during deformation. Finally, the elongation decreases to zero with further increases of swelling. This very brittle failure is known to arise from segregation of nickel to void surfaces which induces a martensitic instability leading to a zero tearing modulus and zero deformation.     

Microstructural evolution in nickel alloy C-276 after Ar ion irradiation

Irradiation damage in nickel alloy C-276 irradiated by 115 keV argon ions at room temperature with irradiation doses from 0.28 to 82.5 dpa has been investigated by transmission electron microscopy. Nano-scale black spot damage appeared at a dose higher than 0.83 dpa. Large interstitial-type dislocation loops were observed at the dose of 8.25 dpa. Both the density of dislocation loops and the density of network dislocations grew significantly with the increase of irradiation dose. However, the density of network dislocations declined at the dose of 27.5 dpa. But the total of dislocation density (density of dislocation lines plus density of dislocation loops) kept increasing and no signs of saturation were seen in the dose range explored. The results showed that the nickel alloy C-276 had good performance in delaying the development of black spots, dislocations and dislocation loops. However, original grains have formed into subgrains at the dose of 82.5 dpa, meaning that the grains in C-276 lost its structural integrity at doses higher than 82.5 dpa.     

Neutron-induced swelling and embrittlement of pure iron and pure nickel irradiated in the BN-350 and BOR-60 fast reactors

Pure iron and nickel were irradiated in the range of 2-15 × 10⁻⁷ dpa/s at 345-650 °C to very high neutron exposures in two fast reactors, BOR-60 and BN-350, to study void swelling and changes in mechanical properties of these two metals. Both nickel and iron swell in this temperature range with the maximum swelling rate at ∼500 °C in nickel, but possibly at ?350 °C for iron. It also appears that the swelling rate in nickel and possibly in iron may be dependent on the dpa rate, increasing with decreasing dpa rate. The evolution of mechanical properties of the two metals is quite different. The differences reflect the fact that b.c.c. iron is subject to a low-temperature embrittlement arising from a shift in ductile-brittle transition temperature, while f.c.c. nickel is not. Nickel, however, exhibits high temperature embrittlement, thought to arise from the collection of transmutant helium gas at the grain boundaries. Iron is not strongly affected by transmutation since it generates much less helium during equivalent irradiation.     

HVEM irradiation damage in heat affected zone of welded SUS304 steel

The heat affected zone (HAZ) of a welded SUS304 steel has been irradiated in an 1250 kV high voltage electron microscope at 673 K and up to 5.4 dpa (displacements per atom) to study the effect of electron irradiation on microstructure. The dislocation loop density of initial irradiation state increased with electron irradiation dose. Void size, void number density and void swelling increased and then saturated gradually with irradiation dose. The depletion of Cr and the enrichment of Ni at the grain boundary were also recognized by EDS analysis in the HAZ of welded SUS304 steel.     

Mechanical properties of high-nickel alloys EP-753 and РЕ-16 after neutron irradiation to 54 dpa at 400-650 °С

Short-term mechanical properties and void swelling were investigated for high-nickel alloys РЕ-16 and three compositional variants of Russian alloy EP-753 and in various starting conditions after side-by-side irradiation in the BN-350 fast reactor at 400, 500, 600 and 650 °С to 54 dpa. For both alloys irradiation resulted in significant hardening and ductility reduction dependent on their chemical composition and initial heat treatment. At test temperatures equal to the irradiation values both alloys exhibited a high level of strength and satisfactory ductility. In the test temperature range of 550-650 °С the phenomenon of high-temperature irradiation embrittlement was observed.     

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.     

Deformation mode maps for tensile deformation of neutron-irradiated structural alloys

The deformation microstructures of neutron-irradiated nuclear structural alloys, A533B steel, 316 stainless steel, and Zircaloy-4, have been investigated by tensile testing and transmission electron microscopy to map the extent of strain localization processes in plastic deformation. Miniature specimens with a thickness of 0.25 mm were irradiated to five levels of neutron dose in the range 0.0001-0.9 displacements per atom (dpa) at 65-100 °C and deformed at room temperature at a nominal strain rate of 10⁻³ s⁻¹. Four modes of deformation were identified, namely three-dimensional dislocation cell formation, planar dislocation activity, fine scale twinning, and dislocation channel deformation (DCD) in which the radiation damage structure has been swept away. The modes varied with material, dose, and strain level. These observations are used to construct the first strain-neutron fluence-deformation mode maps for the test materials. Overall, irradiation encourages planar deformation which is seen as a precursor to DCD and which contributes to changes in the tensile curve, particularly reduced work hardening and diminished uniform ductility. The fluence dependence of the increase in yield stress, ΔYS = α(?t)ⁿ had an exponent of 0.4-0.5 for fluences up to about 3 × 10²² n m⁻² (∼0.05 dpa) and 0.08-0.15 for higher fluences, consistent with estimated saturation in radiation damage microstructure but also concurrent with the acceleration of gross strain localization associated with DCD.     

Irradiation creep relaxation of void swelling-driven stresses

Swelling-driven-creep test specimens are used to measure the compressive stresses that develop due to constraint of irradiation void swelling. These specimens use a previously non-irradiated 20% CW Type 316 stainless steel holder to axially restrain two Type 304 stainless steel tubular specimens that were previously irradiated in the US Experimental Breeder Reactor (EBR-II) at 490 °C. One specimen was previously irradiated to fluence levels in the void nucleation regime (9 dpa) and the other in the quasi-steady void growth regime (28 dpa). A lift-off compliance measurement technique was used post-irradiation to determine compressive stresses developed during reirradiation of the two specimen assemblies in Row 7 of EBR-II at temperatures of 547 °C and 504 °C, respectively, to additional damage levels each of about 5 dpa. Results obtained on the higher fluence swelling-driven-creep specimen show that compressive stress due to constraint of swelling retards void swelling to a degree that is consistent with active load uniaxial compression specimens that were irradiated as part of a previously reported multiaxial in-reactor creep experiment. Swelling results obtained on the lower fluence swelling-driven creep specimen show a much larger effect of compressive stress in reducing swelling, demonstrating that the larger effect of stress on swelling is on void nucleation as compared to void growth. Test results are analyzed using a recently proposed multiaxial creep-swelling model.     

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.     

Cobalt content and vacancy swelling of ChS-68 steel

Analysis of the state of fuel-element claddings before and after operation in a fast reactor revealed a statistically significant factor correlating with the vacancy swelling of 06Kh16N15M2G2TFR (ChS-68) austenitic steel. This factor, which depends on the quantitative composition of the alloying and impurity elements in steel melts, explains the reason for the large (more than a factor of 2) differences in the size changes of spent fuel elements in BN-600 reactor fuel assemblies and makes it possible to predict the swelling resistance of different batches of casing tubes made of ChS-68 steel. __________ Translated from Atomnaya énergiya, Vol. 102, No. 3, pp. 163–168, March, 2007.     

Radiation-induced reduction in the void swelling

Vacancy voids have been produced in Ni by 1.2 MeV Cr ion irradiation at 873 K up to the ion fluence of 10²¹ m⁻². Subsequent irradiation of specimens containing voids at 798 and 723 K has resulted in the reduction of the void size and number density. Accordingly, the void swelling has decreased by a factor of ∼5. The experimental results are explained in the framework of an original model taking into account the interaction of voids with radiation-induced excitations of atomic structure such as focusing collisions and long-propagating self-focusing breathers.     

Proton irradiation emulation of PWR neutron damage microstructures in solution annealed 304 and cold-worked 316 stainless steels

Solution annealed (SA) 304 and cold-worked (CW) 316 austenitic stainless steels were pre-implanted with helium and were irradiated with protons in order to study the potential effects of helium, irradiation dose, and irradiation temperature on microstructural evolution, especially void swelling, with relevance to the behavior of austenitic core internals in pressurized water reactors (PWRs). These steels were irradiated with 1 MeV protons to doses between 1 and 10 dpa at 300 °C both with or without 15 appm helium pre-implanted at ∼100 °C. They were also irradiated at 340 °C, but only after 15 appm helium pre-implantation. Small heterogeneously distributed voids were observed in both alloys irradiated at 300 °C, but only after helium pre-implantation. The pre-implanted steels irradiated at 340 °C exhibited homogenous void formation, suggesting effects of both helium and irradiation temperature on void nucleation. Voids developed sooner in the SA304 alloy than CW316 alloy at 300 and 340 °C, consistent with the behavior observed at higher temperatures (>370 °C) for similar steels irradiated in the EBR-II fast reactor. The development of the Frank loop microstructure was similar in both alloys, and was only marginally affected by pre-implanted helium. Loop densities were insensitive to dose and irradiation temperature, and were decreased by helium; loop sizes increased with dose up to about 5.5 dpa and were not affected by the pre-implanted helium. Comparison with microstructures produced by neutron irradiation suggests that this method of helium pre-implantation and proton irradiation emulates neutron irradiation under PWR conditions.