Irradiation creep in Type 316 stainless steel and US PCA with fusion reactor He/dpa levels

Irradiation creep was investigated in type 316 stainless steel (316 SS) and US Fusion Program PCA using a tailored spectrum of the Oak Ridge Research Reactor in order to achieve a He/dpa value characteristic of a fusion reactor first wall. Pressurized tubes with stresses of 20 to 470 MPa were irradiated at temperatures of 330, 400, 500, and 600/sup 0/C. It was found that irradiation creep was independent of temperature in this range and varied linearly with stress at low stresses, but the stress exponent increased to 1.3 and 1.8 for 316 SS and PCA, respectively, at higher stresses. Specimens of PCA irradiated in the ORR and having helium levels up to 200 appM experienced a 3 to 10 times higher creep rate than similar specimens irradiated in the FFTF and having helium levels below 20 appM. The higher creep rates are attributed to either a lower flux or the presence of helium. A mechanism involving interstitial helium-enhanced climb is proposed. 17 refs

Irradiation creep in the US fusion alloy PCA at fusion He/dpa levels

Irradiation creep has been studied in fast reactors where little helium is generated and in research reactors where very high levels of helium are generated in stainless steels. A summary of the conclusions made from the viewgraph information of this paper follows: the creep rate is 4 to 10 times higher for the fusion He/dpa ratio compared to the breeder reactor results; higher creep rate can only result from differences in the neutron spectrum; a higher level of irradiation creep could be beneficial in relieving swelling stresses, but detrimental in causing deformation under primary loads in the structure. (LSP

Fatigue performance of HFIR-irradiated Nimonic PE-16 at 430/sup 0/C

Nimonic PE-16 was irradiated in the HFIR to 6 to 9 dpa and 560 to 1000 at. ppM He at 430/sup 0/C. Postirradiation fatigue tests revealed a reduction in fatigue life by about a factor of 10 at 430/sup 0/C. In contrast to AISI type 316 stainless steel, no endurance limit was observed. All irradiated specimens exhibited some intergranular fracture with an increasing tendency toward cleavage-like intragranular fracture for low strain ranges

Comparison of the irradiated tensile properties of a high-manganese austenitic steel and type 16 stainless steel

The USSR steel EP-838 is a high-manganese (13.5%), low-nickel (4.2%) steel that also has lower chromium and molybdenum than type 316 stainless steel. Tensile specimens of 20%-cold-worked EP-838 and type 316 stainless steel were irradiated in the High Flux Isotope Reactor (HFIR) at the coolant temperature (approx. 50/sup 0/C). A displacement damage level of 5.2 dpa was reached for the EP-838 and up to 9.5 dpa for the type 316 stainless steel. Tensile tests at room temperature and 300/sup 0/C on the two steels indicated that the irradiation led to increased strength and decreased ductility compared to the unirradiated steels. Although the 0.2% yield stress of the type 316 stainless steel in the unirradiated condition was greater than that for the EP-838, after irradiation there was essentially no difference between the strength or ductility of the two steels. The results indicate that the replacement of the majority of the nickel by manganese and a reduction of chromium and molybdenum in an austenitic stainless steel of composition near that for type 316 stainless steel has little effect on the irradiated and unirradiated tensile properties at low temperatures

Effect of helium on irradiation creep in an austenitic stainless steel

The first wall of a fusion reactor will be subjected to a flux of high energy neutrons which will result in the formation of atomic displacements and their associated phenomena, coupled with the formation of helium via (n,..cap alpha..) reactions. The purpose of the experiments described here is to determine the irradiation creep properties of several austenitic stainless steels under irradiation conditions that simulate fusion reactor conditions in terms of both damage rate and helium generation rate (He:dpa ratio)

Feasibility of correlating V-Cr-Ti alloy weld strength with weld chemistry. CRADA final report

The mechanical properties of refractory metals such as vanadium are determined to a large extent by the interstitial impurities in the alloy. In the case of welding, interstitial impurities are introduced in the welding process from the atmosphere and by dissolution of existing precipitates in the alloy itself. Because of the necessity of having an ultra-pure atmosphere, a vacuum chamber or a glove box is necessary. In the V-Cr-Ti system, the titanium serves as a getter to control the concentration of oxygen and nitrogen in solid solution in the alloy. In this project the secondary ion mass spectrometry (SIMS) technique was used to detect, measure, and map the spacial distribution of impurity elements in welds in the alloy V-4Cr-4Ti. An attempt was then made to correlate the concentrations and distributions of the impurities with mechanical properties of the welds. Mechanical integrity of the welds was determined by Charpy V-notch testing. Welds were prepared by the gas-tungsten-arc (GTA) method. Charpy testing established a correlation between weld impurity concentration and the ductile to brittle transition temperature (DBTT). Higher concentrations of oxygen resulted in a higher DBTT. An exception was noted in the case of a low-oxygen weld which had a high hydrogen concentration resulting in a brittle weld. The concentrations and distributions of the impurities determined by SIMS could not be correlated with the mechanical properties of the welds. This research supports efforts to develop fusion reactor first wall and blanket structural materials

Tensile properties and swelling of 20%-cold-worked Type 316 stainless steel irradiated in HFIR

Immersion density and elevated-temperature tensile properties were determined on 20%-cold-worked type 316 stainless steel irradiated in the HFIR at approximately 285, 370, 470, 560, and 620/sup 0/C to fluences of 1.8 to 6.2 x 10/sup 26/ neutrons/m/sup 2/ (> 0.1 MeV), which resulted in displacement damage levels of 16 to 54 dpa and helium concentrations of 900 to 3300 at. ppM. Tensile tests were done at temperatures near the estimated irradiation temperatures. Swelling increased with increasing irradiation temperature to a maximum of 2.1% at 620/sup 0/C. Irradiation at the lowest temperature (285 and 370/sup 0/C) increased the strength. At the higher irradiation temperatures, the strength decreased during irradiation. Ductility generally reflected the strength behavior: an increase in strength resulted in a descrease in ductility. When the present data are compared with previously published data, there is good agreement with one exception. Previous experiments showed a large decrease in ductility when irradiated at 600/sup 0/C and tested at 575/sup 0/C that was not observed in the present tests. There was also good agreement between HFIR-irradiated steel and literature data for EBR-II-irradiated steel