Void Swelling in Solution Treated, 20% Cold-Worked and 20% Cold-Worked+Aged Ti-Modified 316 Steel

Void swelling in 0.045%Ti-modified 316 stainless steel was investigated by 200-keV C⁺ ion bombardment. Three metallurgical conditions, i.e. solution treated, 20% cold-worked and 20% cold-worked + aged at 898 K for 1,000 h, were compared. Solution treated materials showed considerably low swelling, while the other two conditioned materials showed much higher swelling. The swelling for 20% cold-worked + aged materials was the highest. In solution treated condition, a number of fine-scale TiC precipitates appeared uniformly in the irradiated matrix. The precipitates and also solute Ti and C are considered to suppress void nucleation by vacancy trapping mechanism. In 20% cold-worked condition, high density dislocation networks were observed being heavily decorated dy TiC precipitates, leading to void swelling increase as a consequence of reduced dislocation sink strength and of depletion of solute Ti and C in matrix. In 20% cold-worked + aged condition, as a result of solute depletion by precipitation of various type precipitates and partial recovery of cold-worked dislocation structure, an enhancement of void swelling occurred. The 0.045% Ti-modified 316 steel was found to be more swelling resistant than type 316 steel.     

The effect of damage rate on void growth in metals

The rate theory formulation of void growth was utilized to analyze the effects of damage rate on metal swelling. In particular, the swelling behavior of 316 SS was modeled as a function of temperature, over a range of displacement damage rates between 10^?6dpa/s and 10^?3dpa/s. Detailed analysis of the rate processes for point defect annihilation, migration, and loss to sinks indicated that small vacancy loops limit void growth at high damage rates. The reduction of void growth rates by vacancy emission from voids was found to be shifted towards higher temperature at higher displacement rates. In effect, the peak swelling temperature as well as the upper cutoff temperature for swelling are increased as the displacement rate is increased. The influence of constant or rate dependent nucleation conditions on the final swelling was investigated and it was shown that the initial microstructure before the growth stage essentially determines the peak swelling temperature. When appropriate empirical expressions for void and loop densities were used, the final peak swelling temperature shift agrees reasonably well with experimental data.     

Surface Effect on Void Swelling Behavior of Stainless Steel

The nucleation and growth of voids have been observed successively in different thickness specimens of Type 316 stainless steel electron-irradiated at 550°C in a high voltage electron microscope. In a bulk representative 1.5 μm thick specimen, the void number density increases rapidly and saturates during the initial stage of irradiation and then decreases with following dose by void coalescence. The swelling increases proportionally with (dpa)^1-5 up to about 30 dpa. In a thin specimen, of 0.4 μm thickness, on the other hand, the void number density increases continuously with dose up to about 25 dpa. The swelling of the thin specimen showed a tendency to saturation due to the disappearance of voids at the specimen surfaces. The difference in swelling behavior between the 1.5 and 0.4;μm thick specimens can be ascribed to the different effects of the specimen surfaces, which serve as a dominant sink for both radiation-produced point defects and dislocations.     

Void Swelling of Solution-Annealed Type 316 Stainless Steel under Long Time Proton Irradiation

Solution-annealed type 316 stainless steel was irradiated by 150 keV proton to a dose of about 6 dpa at the irradiation temperature ranging 450–700°C. To examine the effect of aging during irradiation, the present proton irradiation was carried out for about 25 h at a low dose rate of 7×10–^?5dpa/s. The specimens without He preinjection showed much smaller void swelling than those preinjected with He to the content of 10 at.ppm. Similarly to the case of neutron irradiations, the void swelling in the He preinjected specimens showed the temperature dependence with double peaks, and the peak swelling temperatures were about 550 and 650°C. In these specimens with He preinjection. void number density decreased and average void diameter increased with the increase of irradiation temperature in the range of 450–600°C, but these trends were reversed between 600 and 650°C. The volume of the grain boudary M₂₃C6 precipitates increased with the increase of irradiation temperature from 600 to 700°C, and it was concluded that the decrease of soluble carbon due to the precipitation of M₂₃C6 caused the second swelling peak at 650°C.     

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.     

Effect of Solute Titanium on Void Swelling in 316 Stainless Steel

To investigate the solute Ti effect on void swelling in stainless steels, the well-annealed 316 stainless steels modified with various amounts of Ti were electron-irradiated in a high voltage electron microscope (HVEM). It was found that the swelling decreased dramatically with increasing Ti content up to 0.25w/0. This strong dependence of swelling on Ti content arises mainly from the changes in void number density, although the void growth rate also decreases with increase in Ti content. The dependence of the void number density on Ti content is interpreted in terms of the vacancy trapping effect of Ti, which decreases the steady state free vacancy concentration and results in the effective suppression of the void nucleation rate.     

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.     

Defect development in neutron irradiated type 316 stainless steel

The effects of fast neutron irradiation on the defect development in unstressed solution treated Type 316 stainless steel were investigated by transmission electron microscopy. The irradiation conditions investigated covered the fluence range from 0.75 to $\text{5.1 × 10}{}^{22}\text{n/cm}{}^2$ ($\text{E > 0.1}\text{MeV}$) and temperatures from 380 to 850°C. Empirical equations were developed relating the void volume, void number density, mean void size, Frank faulted loop diameter, Frank loop number density and dislocation density with the neutron fluence and irradiation temperature. Void nucleation changes from homogeneous at low irradiation temperature (? 400°C) to heterogeneous at higher temperatures in that voids are preferentially associated with irradiation induced rod shaped precipitates. The void number density decreases while the void diameter increases with irradiation temperature. The total faulted loop line length per unit volume and dislocation density increases with fluence and decreases with temperature. The Frank loop diameter increases and number density decreases with temperature. The range of temperature in which Frank faulted loop formation occurs decreases with neutron fluence.     

Void formation and precipitation during electron-irradiation in austenitic stainless steels modified with Ti,Zr and V

Solution-annealed type 316 stainless steels modified with one or two elements of titanium, zirconium and vanadium were electron-irradiated up to a dose of about 50 dpa at temperatures of 773 to 823 K in a high voltage electron microscope. Addition of 0.3 wt% of titanium or zirconium to 316 steel remarkably reduce the void density at 823 K, compared with that in the standard 316 steel. Conversely, addition of 0.3 wt% of vanadium enhances void formation at 823 K; the void density is two orders of magnitude higher than that in the 316 steel. The enhancement is related to the radiation-induced V-carbides in the early stage of irradiation. Further addition of 0.15 wt% titanium to the V- and Zr-modified steels completely suppresses void formation up to 50 dpa at 823 K, because fine Ti-carbides precipitated along dislocations beyond about 10 dpa change dislocation bias to reduce the vacancy supersaturation rate. A beneficial effect of zirconium on void formation is disappeared by the addition of 0.15 wt% vanadium to the Zr-modified steel irradiated at 773 to 823 K.     

The nucleation and early growth of interstitial dislocation loops in irradiated materials

A hierarchy of rate equations is solved to describe the homogeneous nucleation of interstitial dislocation loops in irradiated materials. Calculations for graphite and M316 stainless steel have been performed. The concept of nucleation time is examined and a procedure is adopted which gives a useful criterion for defining the end of the nucleation period. Calculations have been performed which demonstrate the effects of temperature, dose rate and network-dislocation density on the nucleation and final concentration of interstitial loops. The assumption that di-interstitial atom pairs are stable against thermal dissociation is examined and found to be appropriate for the conditions used in this work.     

Voids resulting from fast neutron irradiation of a stabilized stainless steel

A 16 Cr-13 Ni-niobium stabilized stainless steel (Type 1.4988) was irradiated as fuel pin cladding in the DFR reactor. The irradiation temperatures ranged from 300 to 650 °C. Neutron fluences from 3.3 to 4.0 n/cm²$\text{(E > 0.1 MeV)}$ were achieved. Swelling behavior was studied in this material by transmission electron microscopy, immersion density determinations and diametral change measurements.Void formation was observed in the temperature range 360–610 °C. Void concentration values at lower temperatures (< 480 °C) are comparable to those cited for the unstabilized stainless steels AISI 304 and 316, however, they decrease more strongly with increasing irradiation temperatures. The mean and maximum diameters show a pronounced maximum at about 530 °C. The decrease of the void diameters at higher temperatures can be explained by the nature of the cavities observed. Annealing experiments with specimens irradiated at 610 °C have revealed a bubble or a mixed void-bubble character for these cavities.Swelling in type 1.4988 was found to be lower than that reported previously for types 304 and 316 at comparable irradiation and material conditions. This is especially pronounced at higher temperature. Diameter changes of the pin at $\text{T >- 610 °C}$ cannot be explained by void-formation.     

Ferrite formation in neutron-irradiated austenitic stainless steel

Magnetic measurements were carried out on type 316, 321 and three modified heats of 316 austenitic stainless steels that had been irradiated to high fluences (1 ? 8 × 10²²n/cm², E > 0.1 MeV) in EBR-II at temperatures ranging from 450–700°C. Most of the specimens showed increases of magnetization after exposure to the reactor environment that can be attributed to formation of numerous small ferrite particles. The amount of ferrite formed during irradiation is a function of alloy composition as well as irradiation temperature and fluence. Specimens with low molybdenum concentrations had a greater ferrite content than specimens with the normal molybdenum content of type 316 stainless steel. A modified heat of type 316 with 0.23 wt% Ti had lower levels of ferrite under given irradiation conditions than the other heats. Some particles with diffraction patterns corresponding to the ferrite phase were found in an irradiated type 321 stainless specimen, but none were observed in the type 316 stainless specimens.     

Void Swelling in Electron Irradiated High Purity Fe-Cr-Ni Austenitic Alloys

High-purity Fe-Cr-Ni austenitic alloys corresponding to commercial Type-316 stainless steel and Hastelloy-X were used to investigate the void swelling mechanism of the austenitic steels. The alloys were irradiated with 1 MeV electrons in a high voltage electron microscope in the temperature range of 300~600°C to a total dose of about 30 dpa. Low void swelling in Ni-base alloy is attributed to both low void number density and small void size. Void embryos in Fe-base alloy are stabilized by strain field arised from Ni solute segregation around the void surface. The stabilization does not occur in Ni-base alloy, which results in extremely low void number density at high temperatures (>500°C). Higher void growth rates in Fe-base alloy than in Ni-base alloy are attributable to large climbing rate of dislocations produced during the irradiation.     

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.     

Swelling of type-316 stainless steel at high fluences in EBR-II

A test to measure swelling induced by fast neutron irradiation in unstressed specimens of type-316 stainless steel has completed irradiation in the EBR-II reactor. Results are reported and discussed which describe the swelling as a function of neutron fluence, temperature of irradiation and extent of cold work in the alloy. Density determinations showed swellings of up to 15% $\text{ΔV}\text{V}{}_{\mathrm{f}}$ for 20% cold worked type-316 stainless steel at a neutron fluence level of 1.4 × 10²³n/cm², E > 0.1 MeV (70 dpa). The peak swelling temperature range was 550°C–600°C regardless of the extent of cold working. Increasing the cold work level reduced the swelling and tended to broaden the swelling temperature peak. Transmission electron microscopy (TEM) investigations showed that cold working had reduced the average void sizes compared to those observed in the solution annealed material.     

Temperature effects on the mechanical properties of candidate SNS target container materials after proton and neutron irradiation

This report presents the tensile properties of EC316LN austenitic stainless steel and 9Cr-2WVTa ferritic/martensitic steel after 800 MeV proton and spallation neutron irradiation to doses in the range 0.54-2.53 dpa at 30-100 °C. Tensile testing was performed at room temperature (20 °C) and 164 °C. The EC316LN stainless steel maintained notable strain-hardening capability after irradiation, while the 9Cr-2WVTa ferritic/martensitic steel posted negative hardening in the engineering stress-strain curves. In the EC316LN stainless steel, increasing the test temperature from 20 to 164 °C decreased the strength by 13-18% and the ductility by 8-36%. The effect of test temperature for the 9Cr-2WVTa ferritic/martensitic steel was less significant than for the EC316LN stainless steel. In addition, strain-hardening behaviors were analyzed for EC316LN and 316L stainless steels. The strain-hardening rate of the 316 stainless steels was largely dependent on test temperature. A calculation using reduction of area measurements and stress-strain data predicted positive strain hardening during plastic instability.     

Removal of polonium from stainless steel surface contaminated by neutron irradiated lead-bismuth eutectic

Lead-bismuth eutectic (LBE) has good characteristics for the coolant and/or target of various nuclear systems, but it also has a problem of polonium contamination. In the study, baking experiment was performed to remove polonium contamination on type 316 stainless steel plate that was originated from neutron irradiated LBE. The contaminated type 316 stainless steel plate was baked in a vacuum condition at various temperatures from 200°C to 600°C. In the previous preliminary study, the effect of short time baking was investigated. In the study, the effect of long time baking was investigated. The detail of the experimental method was also described. The result of long time baking experiment showed that the baking method was effective for removal of polonium from stainless steel surface contaminated by neutron irradiated LBE, if the baking was performed at 500°C and higher in a vacuum condition. The obtained result was consistent to the previous preliminary study.     

Effect of mobile helium on void nucleation in materials during irradiation

The steady-state rate of void nucleation is calculated for irradiated materials containing mobile helium. At the low displacement rates typical of a fast-breeder reactor a concentration of less than 10^?10 atom fraction helium can cause a 10²⁰ increase in nucleation rate. The helium is less effective at the high displacement rates typical of accelerator experiments, but can increase the void-nucleation rate by 10⁴ at a helium concentration of 10^?8. The calculated void-nucleation rates for low displacement rates and without helium are too low to explain the void number densities observed in breeder-reactor irradiated materials. Therefore, void nucleation in reactor environments is helium-assisted. Accelerator experiments intended to simulate void nucleation under reactor conditions must be carefully designed to observe gas-assisted rather than homogeneous void nucleation.     

Grain boundary engineering of austenitic steel PNC316 for use in nuclear reactors

Austenitic stainless steel PNC316 was subjected to grain boundary engineering (GBE). It was found that the grain boundary engineered PNC316 (PNC316-GBEM) had a coincidence site lattice (CSL) fraction of 86% and that the network of random grain boundaries was perfectly divided by the CSL boundaries. The thermal stability and the void swelling behavior of PNC316-GBEM were investigated by means of SEM and TEM analyses. After thermal aging at 973 K for 100 h, structural changes were observed neither in the grain boundary networks of PNC316-GBEM nor in another sample of PNC316-GBEM subjected to 20% additional cold rolling, PNC316-GBEM20%CW. PNC316-GBEM showed a higher void swelling rate than as-received PNC316 (PNC316-AS). However, with additional 20% cold rolling after GBE, the void swelling rate decreased to as low as that of PNC316-AS.     

The effects of cold working and pre-irradiation heat treatment on void formation in neutron-irradiated type 316 stainless steel

The effect of fast-neutron irradiation on void formation in Type 316 stainless steel having undergone specific thermalmechanical treatments was investigated by transmission electron microscopy. The study showed that, for irradiation at the three lower temperatures (420, 475 and 580°C): (1) the void volume decreased with increasing cold work; (2) the reduction in swelling was due to a decrease in both void-number density and void size; (3) the decrease in void size with increasing cold-work level was enhanced at higher irradiation temperatures; (4) cold working from 0 to 10% decreased the voidnumber density, and void volume, more than in the range from 10 to 20%; (5) void formation in the 20% cw steel which had been heat treated 100 h at 650°C before irradiation was similar to that of the solution-treated steel. The temperature dependence of swelling of the cold-worked material was different from that of the solution-treated steel. Irradiation at 650°C resulted in a larger void volume in the cold-worked material than for irradiation at 475 or 580°C. The effects of cold work and irradiation temperature on void growth are consistent with the predictions of a diffusion-controlled model.