Mandrel diameter effect on ring-pull testing of nuclear fuel cladding
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36 MATERIALS SCIENCE↗
Engineering topics
Eftink, Benjamin P.
Publications and source records attributed to Eftink, Benjamin P..
Mechanical creep testing and characterization of Cr coated Zircaloy
Zirconium based cladding has been developed for use in light water reactor fuel applications due to the material’s desirable low neutron scattering cross section. To help improve corrosion resistance in accident conditions, such as a loss of coolant accident (LOCA), a chromium coating can be applied to the outer diameter of the Zircaloy tube. Cladding strength at reactor relevant temperatures is important for the structural performance. Stress relaxation creep testing was performed on Cr coated Zircaloy-4 to measure strain rate as a function of stress for 350 and 650°C. These were performed at both 5 ± 1 % and 7.5 ± 1 % strain. The lower temperature provides results near normal operating temperature, while the higher temperature provides results relevant to off-normal conditions where recrystallization will take place. Characterization was also performed to analyze the structure of the Cr coating and Zircaloy-4 as well as the interface between the two.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Additive manufacturing of ferritic/martensitic steel with improved high temperature strength
Techniques for using additive manufacturing (AM) to fabricate creep resistant ferritic/martensitic steel with improved high temperature strength are described. AM processing may be performed on Grade 91 steel powder. Beam powers from about 221 W to about 270 W may be used. Traverse rates from about 675 mm/s to about 825 mm/s may be used. Heat inputs ranging from about 55.7 J/mm3 to about 83.2 J/mm 3 may be produced. Creep resistant ferritic/martensitic steel, produced according to the present disclosure, has improved strain yield strength and ductility as compared to wrought steel.
Lienert, Thomas J.↗
High-Throughput Nanoindentation Mapping of Additively Manufactured T91 Steel
Here, this work aims to adapt nanoindentation mapping combined with a k-means algorithm as a high-throughput technique to study the nano-scale spatial changes in mechanical properties for a heterogeneous material. This technique can also classify the individual data points based on their properties. Hundreds to thousands of indents were performed on additively manufactured T91 at room temperature, 300°C, 400°C, and 500°C across a square area with a side length of 120 μm to 400 μm. From this data, the hardness and reduced modulus at each point could be calculated and mapped. Using k-means clustering, we were able to arrange the data into three or four clusters corresponding roughly to the ferritic and martensitic phases as well as one or two intermediate clusters sampling both the phases. The hardness of these two phases appears to be quite stable as a function of temperature. Nanoindentation mapping and the k-means algorithm can therefore be used to rapidly assess the feasibility of heterogeneous materials under extreme conditions, such as nuclear reactor steels.
36 MATERIALS SCIENCE↗
The influence of nitrogen and nitrides on the structure and properties of proton irradiated ferritic/martensitic steel
The 12Cr1MoWV (wt%) ferritic/martensitic steel HT9 is a candidate material for fuel cladding in advanced nuclear reactors, such as the Versatile Test Reactor currently under development. As such, understanding the relationship between microstructure and mechanical properties in the context of irradiation environments for these steels is critical. N content, and more specifically interstitial N, has been hypothesized to be detrimental to irradiated properties at lower temperatures (less than 0.3T m ) to a total of 6 dpa; however, in this work at a dose of 1 dpa the irradiated microstructure was improved with added N, leading to less irradiation hardening. Three variants of HT9 were irradiated with 1.5 MeV protons to a dose of 1 dpa at 300°C. The HT9 variants included Low (10 ppm), Mid (190 ppm), and High (440 ppm) N alloys that were otherwise nearly identical. Changing the N content had a variety of effects on the irradiated defect structures. As N content increased, the average dislocation loop diameter decreased, while the number density of loops increased. Additionally, extensive Ni clustering was observed on dislocations and interfaces. The Mid and High N specimens exhibited significantly less hardening (ΔHV≃100) relative to the Low N specimen (ΔHV≃160). The decrease in hardening is attributed to vanadium carbonitride acting as a sink for Ni clusters that would otherwise form on dislocations. Under the irradiation conditions used, these results suggest increasing the N content in HT9 may have a desirable effect on the irradiated structure and properties at the dose studied, as well as the swelling resistance at higher doses. In other words, N content appears to be a powerful tool for tailoring the self-interstitial atom cluster mobility in F/M steels for different temperature and dose applications.
36 MATERIALS SCIENCE↗
Tensile properties and microstructure of additively manufactured Grade 91 steel for nuclear applications
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99 GENERAL AND MISCELLANEOUS↗
Comparison of mechanical properties measured on multiple scales
FeCrAl alloys are being developed for accident tolerant fuels for current light water reactors. Specifically as fuel claddings with enhanced safety. These alloys are being developed to have excellent corrosion resistance in light water reactor coolant environments and structural integrity for longer durations in the case of a loss of coolant scenario where time is critical for safety. Tensile, shear punch, and Vickers hardness testing was performed on neutron irradiated FeCrAl alloys with three variations in composition. Testing was done at room temperature for comparison of the three techniques. FeCrAl alloys in this report were irradiated at Oak Ridge National Laboratory (ORNL) in the High Flux Isotope Reactor (HFIR) to 7 dpa and at 320 °C.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Proton irradiation and characterization of additively manufactured 304L stainless steels
Irradiations were performed with 1.5 MeV protons to 0.6 dpa at 40–150 °C on additively manufactured (AM) 304L stainless steel and the changes in microstructure and mechanical behavior after irradiation were compared to wrought 304L stainless steel. Interestingly, all microstructural and hardness results after irradiation suggest the samples evolve toward a similar state, despite significant differences in the unirradiated microstructures and hardness values. A TEM and nanoindentation-based investigation of before and after proton irradiation at 40–150 °C is presented. Results are interpreted in terms of initial dislocation content, dislocation structures, and microstructural and chemical homogeneity.
36 MATERIALS SCIENCE↗
Shear Punch Testing of Neutron-Irradiated HT-9 and 14YWT
Shear punch mechanical testing was performed on neutron-irradiated HT-9 and 14YWT alloys, comparing the traditional ferritic/martensitic steel with the advanced oxide dispersion strengthened nanostructured ferritic alloy. The samples were irradiated at various temperatures between 380°C and 520°C. In HT-9, on increasing the irradiation temperature, the 1% offset shear stress and maximum shear stress decreased. The advanced 14YWT alloy was less sensitive to irradiation temperature with respect to irradiation hardening. However, it still exhibited irradiation hardening at each temperature.
36 MATERIALS SCIENCE↗
Capacitive Discharge Resistance Welding for ODS Steel Cladding: Weld Properties and Radiation Resistance
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Lear, Calvin Robert↗
Computer generated three-dimensional models of microstructural features based on stereomicroscopy
3D information may be extracted from two 2D images by capturing a first image of a sample at a first orientation. The sample may be titled at a second or different orientation, resulting in a second image of the titled sample to be captured. Third dimension of information may be extracted from the images.
71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
Characterization of 14YWT & ODS Base Metals
Nanostructured ferritic alloys (NFA) with oxide dispersion strengthening are being developed for next generation reactor applications. Applications such as transmutation of long lived isotopes in used fuels require cladding materials that can withstand high irradiation doses, potentially in the hundreds of displacements per atom (dpa). Currently, however, there are no joining processes that retain the strength and fine distribution of oxide precipitates characteristic of ODS alloys. As part of a larger effort to adapt a solid state capacitive discharge welding technique for use with ODS cladding, presented will be characterization of the initial materials including 14YWT and MA956 ODS alloys.
36 MATERIALS SCIENCE↗
Shear Punch Testing of BOR60 Irradiated HT-9 and 14YWT
Ferritic/martensitic (FM) steels and nanostructured ferritic alloys (NFA) are being developed for next generation reactor applications. Applications such as transmutation of long lived isotopes in used fuels require cladding materials that can withstand high irradiation doses, potentially in the hundreds of displacements per atom (dpa). The FM steels in this report include HT-9 steel, which is a Fe-12Cr steel with additions of Mo, Mn, Ni, V and W. An oxide dispersion strengthened (ODS) 14YWT is also presented, which has a distribution of nano-scale oxide particles throughout the microstructure. Shear punch testing was performed on both steels. The material was neutron irradiated at the BOR60 reactor in Russia and transferred to LANL for shear punch mechanical testing. These results are complementary to a previous report, also from a BOR60 irradiation.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Tensile Testing of Generation II FeCrAl Tubing
FeCrAl tubing is being developed as an improved accident tolerant cladding material for light water reactor fuels. A very promising generation FeCrAl alloy, C26M, is not being produced in long tubing for future reactor testing. In this report, the results of two FeCrAl tubes are presented. Tubes from two different heats of C26M are tested. Tensile tests were conducted in air at room temperature, 300 °C, and 600 °C.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Deformation response of cube-on-cube and non-coherent twin interfaces in AgCu eutectic after dynamic plastic compression
For this research, Split-Hopkinson pressure bar dynamic compression experiments were conducted to determine the defect/interface interaction dependence on interface type, bilayer thickness and interface orientation with respect to the loading direction in the Ag-Cu eutectic system. Specifically, the deformation microstructure in alloys with either a cube-on-cube orientation relationship with {111} Ag ||{111} Cu interface habit planes or a twin orientation relationship with {$\overline{3}13$} Ag ||{$\overline{1}12$} Cu interface habit planes and with bilayer thicknesses of 500 nm, 1.1 µm and 2.2 µm were probed using TEM. The deformation was carried by dislocation slip and in certain conditions, deformation twinning. The twinning response was dependent on loading orientation with respect to the interface plane, bilayer thickness, and interface type. Twinning was only observed when loading at orientations away from the growth direction and decreased in prevalence with decreasing bilayer thickness. Twinning in Cu was dependent on twinning partial dislocations being transmitted from Ag, which only occurred for cube-on-cube interfaces. Lastly, dislocation slip and deformation twin transfer across the interfaces is discussed in terms of the slip transfer conditions developed for grain boundaries in FCC alloys.
36 MATERIALS SCIENCE↗
Deformation response of AgCu interfaces investigated by in situ and ex situ TEM straining and MD simulations
The mechanisms of strain transfer across Ag/Cu interfaces were determined by a combination of in situ and ex situ TEM straining experiments and molecular dynamics simulations. Minimizing the magnitude of the Burgers vector of the residual dislocation generated in the interface was the dominant factor for determining the outcome of dislocation and deformation twin interactions with both non-coherent twin and cube-on-cube interfaces. This included the unexpected finding, due to the loading condition, of deformation twin activation in the Cu layer due to the intersection of deformation twins in Ag with the interface. As a result, deformation twin nucleation in Ag from the non-coherent twin interfaces was also explained by a Burgers vector minimization argument.
36 MATERIALS SCIENCE↗
Synergies Between ' and Cavity Formation in HT-9 Following High Dose Neutron Irradiation
Candidate cladding materials for advanced nuclear power reactors including fast reactor designs require materials capable of withstanding high dose neutron irradiation at elevated temperatures. One candidate material, HT-9, through various research programs have demonstrated the ability to withstand significant swelling and other radiation-induced degradation mechanisms in the high dose regime (>50 displacements per atom, dpa) at elevated temperatures (>300 C). Here, high efficiency multi-dimensional scanning transmission electron microscopy (STEM) acquisition with the aid of a three-dimensional (3D) reconstruction and modeling technique is used to probe the microstructural features that contribute to the exceptional swelling resistance of HT-9. In particular, the synergies between ' and fine-scale and moderate-scale cavity formation is investigated.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Bulk texture evolution of nanolamellar Zr–Nb composites processed via accumulative roll bonding
We report that it was recently demonstrated that bulk two-phase 50/50 Zr–Nb nanolayered composites with 90 nm individual layers can be fabricated from an initial coarse-layered composite with 1 mm layers via the severe plastic deformation process of accumulative roll bonding. During the deformation, the Zr phase retained its hcp crystal structure and the Zr–Nb interface remained sharp. Here we use a combination of neutron diffraction and dislocation-based polycrystal plasticity constitutive modeling to assess the evolution of texture and deformation mechanisms over a four order-of-magnitude range in layer thickness. The phase textures in the nanocomposite strongly deviate from that of Zr or Nb rolled in monolithic form, becoming highly peaked and intense. The model suggests that texture development in the Nb phase is associated with multiple slip and contributions from both {1 1 2} <1 1 0> slip and {1 1 0} <1 1 0> slip. Lastly, in the Zr phase the model suggests that the texture develops due to a predominance of prismatic and basal slip.
36 MATERIALS SCIENCE↗