J. I. Cole

The stability of 9Cr-ODS oxide particles under heavy-ion irradiation

Abstract An oxide-dispersion–strengthened (ODS) martensitic steel 9Cr-ODS was irradiated with 5-MeV Ni ions at 500°C at a dose rate of 1.4 × 10–3 dpa/s to doses of 5, 50, and 150 dpa. The ODS steel has been designed for use in higher-temperature energy systems. However, the radiation effects are not fully characterized, particularly to high doses. Dense dislocations, precipitates, and yttrium-titanium oxide particles dominated the microstructure of 9Cr-ODS for both the unirradiated and irradiated cases with no dislocation loops observed. No voids were detected for doses up to 150 dpa. The average size of the oxide particles, whose size is approximately described by a lognormal distribution, slightly decreased with dose from ~12 nm for the unirradiated case to ~9 nm at 150 dpa. The decrease in size follows a square root of dose dependency, indicating the effect is radiation induced. The decrease in size is not expected to have a detrimental effect on high-temperature strength, even to extremely high dose.

Behavior of Irradiated Type 316 Stainless Steels under Low-Strain-Rate Tensile Conditions

The effects of lower strain rate on the tensile behavior of 12 % cold-worked type 316 stainless steels irradiated in the EBR-II reactor under low-dose-rate and moderate temperature conditions were investigated. Tensile tests were carried out at a strain rate of 1 × 10−7/s. Post-test fractography and microstructural characterization were also performed. Irradiation temperature and dose appeared to have the greatest effect on hardening and ductility loss, whereas dose rate appeared to have less apparent effects. In conjunction with earlier work performed on the same material at a strain rate of 1 × 10−3/s, there was no significant effect of strain rate on tensile behavior under the irradiation conditions examined. For fracture behavior, the material after irradiation exhibited typical ductile fracture during both high and low-strain-rate tests.

Properties of 20% cold-worked 316 stainless steel irradiated at low dose rate.

To assess the effects of long-term, low-dose-rate neutron exposure, tensile, hardness, and fracture properties were measured and microstructural characterization performed on irradiated 20% cold-worked Type 316 stainless steel. Samples were prepared from reactor core components retrieved from the EBR-II reactor following final shutdown. Sample locations were chosen to cover a dose range of 1-56 dpa at temperatures from 371-390 C and dose rates from 0.8-3.3 x 10{sup -7} dpa/s. Irradiation caused hardening, with the ultimate tensile strength (UTS) reaching about 800 MPa near 20 dpa and appearing to saturate at higher doses. The yield strength (YS) follows approximately the same trend as the ultimate tensile strength. At higher dose, the difference between the UTS and YS decreases, suggesting the work-hardening capability of the material is decreasing with increasing dose. The hardness and yield strength increases occur roughly over the same range of dose. While the material retained respectable ductility at 20 dpa, the uniform and total elongation decreased to <1 and <3%, respectively, at 47 dpa. Fracture in the 30 dpa specimen is mainly ductile but with local regions of mixed-mode failure, consisting mainly of dimples and microvoids. The fracture surface of the higher-exposure 47 dpa specimen displays more brittlemorexa0» features. Changes in yield strength predicted from the microstructural components are roughly consistent with the measured changes in yield strength.«xa0less

Swelling and Microstructural Evolution in 316 Stainless Steel Hexagonal Ducts Following Long-Term Irradiation in EBR-II

Swelling behavior and microstructural evolution of 12% cold-worked 316 SS hexagonal ducts following irradiation in the outer rows of EBR-II is described. Immersion density measurements and transmission electron microscopy (TEM) examination were performed on a total of seven irradiation conditions. The samples were irradiated to temperatures between 375 and 430 C to doses between 23 and 51 dpa and at dose-rates ranging from 1.3 x 10{sup -7} to 5.8 x 10{sup -7} dpa/s. Dose-rates and temperatures approach conditions experienced by a variety of components in pressurized water reactors (PWRs) and those which may be present in future advanced reactors designs. TEM analysis was employed to elucidate the effect of radiation on the dislocation, void and precipitate structures as a function of irradiation conditions. A moderate dose-rate effect was observed for samples which were irradiated at dose-rates differing by a factor of two. Lower dose-rate samples contained voids of larger diameter and typically swelled more in the bulk. The dislocation and precipitate structure was not visibly influenced by a dose-rate decrease.

Radiation-induced segregation and void swelling in 304 stainless steel.

. Void swelling and radiation-induced segregation have been measured in 304 stainless steel. Samples were irradiated in the outer regions of the EBR-11reactor where displacement rates of 2.OX10-8and 6.6 xl 0-8dpds are comparable to those in pressurized water reactor components. Samples were inadiated at temperatures from 371-390°C to total doses of up to 20 dpa. Void swelling reached a maximum of 2 0/0 at 20 dpa. Nickel enrichment and chromium dep~etion of up to of 20 at% and 12 atO/Orespectively were measured. Both void swelling and radiation-induced segregation were dependent on dose rate, increasing as the dose rate decreased. Grain boundary compositions were measured both near and in areas free of precipitates. The presence of a precipitate significantly changes the grain boundary compositions near the precipitate.

The Irradiated Microstructure of Ferritic-Martensitic Steel T91 and 9Cr-ODS

A ferritic steel T91 and an oxide dispersion strengthened (ODS) martensitic steel 9Cr-ODS were irradiated with 5 MeV Ni ions at 500 C at a dose rate of 1.38x10{sup -3} dpa/s to doses of 5, 50 and 150 dpa. Both alloys are iron-based with 9Cr and have been designed for use in higher temperature energy systems. However, the radiation effects on these two alloys are not well characterized. For T91, the irradiated microstructure was dominated by tangled dislocation and precipitates, similar to the unirradiated condition except the presence of large dislocation loops of type a . The microstructure of alloy 9Cr-ODS for both the unirradiated and irradiated cases was dominated by dense dislocations, precipitates and yttrium oxides particles and no dislocation loops were observed. The average size of yttrium oxides particles slightly decreased with dose from 11.8 nm for the unirradiated to 9.1 nm at 150 dpa. No voids were detected for both alloys up to a dose of 150 dpa.

Effect of Zr on the Irradiated Microstructure and Hardening in 304 Stainless Steel

ABSTRACT Model alloys of 304 Stainless Steels (SS) (Fe-18Cr-9.5Ni-1.75Mn) and 304 SS+Zr (Fe-18Cr-9.5Ni-1.75Mn+0.04Zr and Fe-18Cr-9.5Ni-1.75Mn+ 0.16Zr) were irradiated with 3.2 MeV protons to a dose of 1.0 dpa at 400°C. Following irradiation, the microstructure was characterized. The number density, defect size, and size distributions for faulted loops and voids were determined. Swelling for each irradiation condition was calculated based on the void measurements. The effect of Zr addition on the irradiated microstructure and hardening is clearly demonstrated. The number density of defects decreased with the Zr addition while the size change of faulted loops and voids is less pronounced. Radiation hardening was reduced by Zr addition. Void swelling is decreased with Zr addition. The reduction in void density and swelling may be caused by the enhanced recombination of defects at oversized Zr solute atoms, suppressing the vacancy super saturation and therefore directly suppressing void nucleation. The reduction in loop density is believed due to the enhanced point defects recombination caused by oversized solute Zr.

The Influence of Pre-Irradiation Heat Treatments on Thermal Non-Equilibrium and Radiation-Induced Segregation Behavior in Model Austenitic Stainless Steel Alloys

The effect of pre-irradiation heat treatments on thermal non-equilibrium grain boundary segregation (TNES) and subsequent radiation-induced grain boundary segregation (RIS) is studied in a series of model austenitic stainless steels. The alloys used for this study are based on AISI 316 stainless steel and have the following nominal compositions: Fe-16Cr-13Ni-1.25Mn (base 316), Fe-16Cr-13Ni-1.25Mn-2.0Mo (316 + Mo) and Fe-16Cr-13Ni-1.25Mn-2.0Mo-0.07P (316 + Mo + P). Samples were heat treated at temperatures ranging from 1100 to 1300 C and cooled at 4 different rates (salt brine quench, water quench, air cool and furnace cool) to evaluate the effect of annealing temperature and quench rate on TNES. The alloys were than processed with the treatment (temperature and cooling rate) that resulted in the maximum Cr enrichment. Alloys with and without the heat treatment to enrich the grain boundaries with Cr were characterized following irradiation to 1 dpa at 400 C with high-energy protons in order to understand the influence of alloying additions and pre-irradiation grain boundary chemistry on irradiation-induced elemental enrichment and depletion profiles. Various mechanistic models will be examined to explain the observed behavior.

Mechanical properties of 20% cold-worked 316 stainless steel irradiated at low dose rate.

To assess the effects of long-term, low-dose-rate neutron exposure on mechanical strength and ductility, tensile properties were measured on irradiated 20% cold-worked Type 316 stainless steel. Samples were prepared from reactor core components retrieved from the EBR-II reactor following final shutdown. Sample locations were chosen to cover a dose range of 1-47 dpa at temperatures from 371-385 C and dose rates from 0.8-2.8 x 10{sup -7} dpa/s. These dose rates are about one order of magnitude lower than those of typical EBR-II in-core experiments. Irradiation caused hardening, with the yield strength (YS) following approximately the same trend as the ultimate tensile strength (UTS). At higher dose, the difference between the UTS and YS decreases, suggesting the work-hardening capability of the material is decreasing with increasing dose. Both the uniform elongation and total elongation decrease up to the largest dose. Unlike the strength data, the ductility reduction showed no signs of saturated at 20 dpa. While the material retained respectable ductility at 20 dpa, the uniform and total elongation decreased to <1 and <3%, respectively, at 47 dpa. Fracture in the 30 dpa specimen is mainly ductile but with local regions of mixed-mode failure, consisting of dimples and microvoids. The fracture surface of the higher-exposure 47 dpa specimen displays significantly more brittle features. The fracture consists of mainly small facets and slip bands that suggest channel fracture.The hardening in these low-dose-rate components differs from that measured in test samples irradiated in EBR-II at higher-dose-rate. The material irradiated at higher dose rate loses work hardening capacity faster than the lower dose rate material, although this effect could be due to compositional differences.

The effect of bulk composition on swelling and radiation-induced segregation in austenitic alloys.

Changes in bulk composition are known to affect both radiation-induced segregation and microstructural development, including void swelling in austenitic stainless steel. In this work, three alloys (designations corresponding to wt%) have been studied: Fe-18Cr-8Ni alloy (bulk composition corresponding to 304 stainless steel), Fe-18Cr-40Ni (bulk composition corresponding to 330 stainless steel), and Fe-16Cr-13Ni (bulk composition corresponding to 316 stainless steel). Following irradiation with high-energy protons, the change in hardness and microstructure (void size distribution and grain boundary composition) due to irradiation was investigated. Increasing the bulk nickel concentration decreases void swelling, increases matrix hardening, and increases grain boundary chromium depletion and nickel enrichment. The analysis shows that decreases in lattice parameter and shear modulus due to radiation- induced segregation (RIS) correlate with decreased void swelling and a decreased susceptibility to irradiation assisted stress corrosion cracking (IASCC). Traditional thinking on IASCC assumed RIS was a contributing factor to cracking. It may, however, be that properly controlled RIS can be used to mitigating cracking.