Mikhail A. Sokolov

The mechanical properties of 316L/304L stainless steels, Alloy 718 and Mod 9Cr-1Mo after irradiation in a spallation environment

Abstract The Accelerator Production of Tritium (APT) project proposes to use a 1.0 GeV, 100 mA proton beam to produce neutrons via spallation reactions in a tungsten target. The neutrons are multiplied and moderated in a lead/aluminum/water blanket and then captured in 3 He to form tritium. The materials in the target and blanket region are exposed to protons and neutrons with energies into the GeV range. The effect of irradiation on the tensile and fracture toughness properties of candidate APT materials, 316L and 304L stainless steel (annealed), modified (Mod) 9Cr–1Mo steel, and Alloy 718 (precipitation hardened), was measured on tensile and fracture toughness specimens irradiated at the Los Alamos Neutron Science Center accelerator, which operates at an energy of 800 MeV and a current of 1 mA. The irradiation temperatures ranged from 50°C to 164°C, prototypic of those expected in the APT target/blanket. The maximum achieved proton fluence was 4.5×10 21 p / cm 2 for the materials in the center of the beam. This maximum exposure translates to a dpa of 12 and the generation of 10 000 appm H and 1000 appm He for the Type 304L stainless steel tensile specimens. Specimens were tested at the irradiation temperature of 50–164°C. Less than 1 dpa of exposure reduced the uniform elongation of the Alloy 718 (precipitation hardened) and Mod 9Cr–1Mo to less than 2%. This same dose reduced the fracture toughness by 50%. Approximately 4 dpa of exposure was required to reduce the uniform elongation of the austenitic stainless steels (304L and 316L) to less than 2%. The yield stress of the austenitic steels increased to more than twice its non-irradiated value after less than 1 dpa. The fracture toughness reduced significantly by 4 dpa to ∼100 MPa m1/2. These results are discussed and compared with results of similar materials irradiated in fission reactor environments.

Embrittlement of reduced-activation ferritic/martensitic steels irradiated in HFIR at 300°C and 400°C

Abstract Miniature tensile and Charpy specimens of four ferritic/martensitic steels were irradiated at 300°C and 400°C in the high flux isotope reactor (HFIR) to a maximum dose of ≈12 dpa. The steels were standard F82H (F82H-Std), a modified F82H (F82H-Mod), ORNL 9Cr–2WVTa, and 9Cr–2WVTa–2Ni, the 9Cr–2WVTa containing 2% Ni to produce helium by (n,α) reactions with thermal neutrons. More helium was produced in the F82H-Std than the F82H-Mod because of the presence of boron. Irradiation embrittlement in the form of an increase in the ductile–brittle transition temperature (ΔDBTT) and a decrease in the upper-shelf energy (USE) occurred for all the steels. The two F82H steels had similar ΔDBTTs after irradiation at 300°C, but after irradiation at 400°C, the ΔDBTT for F82H-Std was less than for F82H-Mod. Under these irradiation conditions, little effect of the extra helium in the F82H-Std could be discerned. Less embrittlement was observed for 9Cr–2WVTa steel irradiated at 400°C than for the two F82H steels. The 9Cr–2WVTa–2Ni steel with ≈115 appm He had a larger ΔDBTT than the 9Cr–2WVTa with ≈5 appm He, indicating a possible helium effect.

Deuterium trapping at defects created with neutron and ion irradiations in tungsten

The effects of neutron and ion irradiations on deuterium (D) retention in tungsten (W) were investigated. Specimens of pure W were irradiated with neutrons to 0.3 dpa at around 323 K and then exposed to high-flux D plasma at 473 and 773 K. The concentration of D significantly increased by neutron irradiation and reached 0.8 at% at 473 K and 0.4 at% at 773 K. Annealing tests for the specimens irradiated with 20 MeV W ions showed that the defects which play a dominant role in the trapping at high temperature were stable at least up to 973 K, while the density decreased at temperatures equal to or above 1123 K. These observations of the thermal stability of traps and the activation energy for D detrapping examined in a previous study (≈1.8 eV) indicated that the defects which contribute predominantly to trapping at 773 K were small voids. The higher concentration of trapped D at 473 K was explained by additional contributions of weaker traps. The release of trapped D was clearly enhanced by the exposure to atomic hydrogen at 473 K, though higher temperatures are more effective for using this effect for tritium removal in fusion reactors.

Recent progress in small specimen test technology

Small specimen test technology (SSTT) has enabled, the development of fusion materials by efficiently using available irradiation volumes. The technology has also evolved in anticipation of the construction and operation of a high-energy neutron source for development and verification of an engineering database for materials for fusion power reactors. Work to date has brought SSTT to a robust state of maturity. SSTT specimens and techniques now routinely serve as the foundation for a number of ongoing and planned experimental programs. Moreover, the need to use small specimens has given rise to the development of new approaches to fracture assessment, such as the master curves-shifts method. Nonetheless a wealth of opportunities exists to further develop new and very innovative SSTT methods. not only for characterizing standard mechanical properties but also to enable both large matrix single variable experiments and highly controlled basic mechanism studies. This paper reviews briefly the recent progress on developing a more science-based SSTT, including some future opportunities. The importance and utility of applying a variety of quasinon-destructive evaluations to a single specimen and closely integrating. finite element. simulations and fundamental models of deformation and fracture are emphasized

Effect of chromium, tungsten, tantalum, and boron on mechanical properties of 5-9Cr-WVTaB steels

Abstract The Cr–W–V–Ta reduced-activation ferritic/martensitic steels use tungsten and tantalum as substitutes for molybdenum and niobium in the Cr–Mo–V–Nb steels that the reduced-activation steels replaced as candidate materials for fusion applications. Studies were made to determine the effect of W, Ta, and Cr composition on the tensile and Charpy properties of the Cr–W–V–Ta; steels with 5%, 7%, and 9% Cr with 2% or 3% W and 0%, 0.05%, or 0.10% Ta were examined. Boron has a long history of use in steels to improve properties, and the effect of boron was also examined. Regardless of the chromium concentration, the steels with 2% W and 0.05–0.1% Ta generally had a better combination of tensile and Charpy properties than steels with 3% W. Boron had a negative effect on properties for the 5% and 7% Cr steels, but had a positive effect on the 9% Cr steel. When the 5, 7, and 9Cr steels containing 2% W and 0.05% Ta were compared, the tensile and Charpy properties of the 5 and 9Cr steels were better than those of the 7Cr steel, and overall, the properties of the 5Cr steel were better than those of the 9Cr steel. Such information will be useful if the properties of the reduced-activation steels are to be optimized.

Charpy Impact Properties of Reduced-Activation Ferritic/Martensitic Steels Irradiated in HFIR up to 20 dpa

ABSTRACT The effects of irradiation up to 20 dpa on the Charpy impact properties of reduced-activation ferritic/martensitic steels (RAFs) were investigated. The ductile-brittle transition temperature (DBTT) of F82H-IEA shifted up to around 323K. TIG weldments of F82H showed a fairly small variation on their impact properties. A finer prior austenite grain size in F82H-IEA after a different heat treatment resulted in a 20K lower DBTT compared to F82H-IEA after the standard heat treatment, and that effect was maintained even after irradiation. Helium effects were investigated utilizing Ni-doped F82H, but no obvious evidence of helium effects was obtained. ORNL9Cr-2WVTa and JLF-1 steels showed smaller DBTT shifts compared to F82H-IEA.

A potential new ferritic/martensitic steel for fusion applications

Abstract The A-21 steel is an Fe–Cr–Co–Ni–Mo–Ti–C steel that is strengthened by a fine distribution of titanium carbide (TiC) precipitates formed by thermomechanical treatment. Transmission electron microscopy of the A-21 reveals a high number density of small TiC particles uniformly distributed in the matrix. Below ≈600 ∘ C, the strength of A-21 is less than the average value for conventional Cr–Mo or reduced-activation ferritic/martensitic steels. However, the strength is greater above 600°C. The Charpy impact properties of A-21 are comparable to those of the conventional and reduced-activation steels. Due to the fine TiC particles in the matrix, the creep-rupture properties of A-21 are superior to those of conventional Cr–Mo or reduced-activation Cr–W steels. Although the composition of the A-21 is not applicable for fusion because of the cobalt, the innovative production process may offer a route to an improved steel for fusion.

Role of Scale Factor During Tensile Testing of Small Specimens

The influence of scale factor (tensile specimen geometry and dimensions) on mechanical test results was investigated for different widely used types of small specimens (SS-1, SS-2, SS-3, and SS-J3) and a set of materials. It was found that the effect of scale factor on the accurate determination of yield stress, ultimate tensile stress, and uniform elongation values was weak; however, clear systematic differences were observed and should be accounted for during interpretation of results. In contrast, total elongation values were strongly sensitive to variations in specimen geometry. Modern experimental methods like digital image correlation allow the impact of scale factor to be reduced. Using these techniques, it was shown that true stress true strain curves describing strain-hardening behavior were very close for different specimen types. The limits of miniaturization are discussed, and an ultra-miniature specimen concept was suggested and evaluated. This type of specimen, as expected, may be suitable for SEM and TEM in situ testing.

Atom probe tomography of radiation-sensitive KS-01 weld

An atom probe tomography characterization has been performed on a neutron-irradiated (fluence = 0.8 × 1019 n cm−2 (E > 1 MeV)) high copper (0.37%), high manganese (1.64%), high nickel (1.23%) and high chromium (0.47%) KS-01 test weld. This weld exhibited a high sensitivity to neutron irradiation. Atom probe tomography revealed that there was an unusually high supersaturation of copper in the matrix after the stress relief treatment, which resulted in a high number density (3 × 1024 m−3) of Cu–Mn–Ni-enriched precipitates after neutron irradiation. Their average size and composition were estimated to be ⟨lg⟩ = 2.6 ± 0.5 nm and Fe-17.0 ± 9.7 at% Cu, 31.9 ± 13.8% Ni, 31.7 ± 11.8% Mn. Phosphorus clusters and a Cr-, Mn-, Ni-, Cu-, C-, N-, Si- and Mo-enriched atmosphere, possibly associated with a dislocation, were also observed in the neutron irradiated material. Nickel, manganese, silicon, phosphorus and carbon segregation to a grain boundary were observed in the unirradiated condition. The microstructural and mechanical response to irradiation was consistent with other lower solute level steels.

Applicability of the Fracture Toughness Master Curve to Irradiated Highly Embrittled Steel and Intergranular Fracture

The Heavy-Section Steel Irradiation Program at Oak Ridge National Laboratory has evaluated a submerged-arc (SA) weld irradiated to a high level of embrittlement and a temper embrittled base metal that exhibits significant intergranular fracture relative to representation by the Master Curve. The temper embrittled steel revealed that the intergranular mechanism significantly extended the transition temperature range up to 150°C above To. For the irradiated highly embrittled SA weld study, a total of 21 1T compact specimens were tested at five different temperatures and showed the Master Curve to be nonconservative relative to the results, although that observation is uncertain due to evidence of intergranular fracture.