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Frazer, David

Publications and source records attributed to Frazer, David.

52 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

UNIRRADIATED

This is the mechanical testing data archive from a neutron irradiation campaign of nuclear structural alloys fabricated by powder metallurgy with hot isostatic pressing (PM-HIP). Data in this archive include uniaxial tensile testing data, fractography of selected tensile bars, and nanoindentation. The irradiation campaign was designed to facilitate a direct comparison of PM-HIP to conventional casting or forging. Five common nuclear structural alloys were included in the campaign: 316L stainless steel, SA508 pressure vessel steel, Grade 91 ferritic steel, and Ni-base alloys 625 and 690. Irradiations were carried out in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to target doses of 1 and 3 displacements per atom (dpa) at target temperatures of 300ºC and 400ºC through the DOE Office of Nuclear Energy, Nuclear Science User Facilities (NSUF) project 15-8242.

Wharry, Janelle↗
DOE OSTI2024

Length scale effects of micro- and meso-scale tensile tests of unirradiated and irradiated Zircaloy-4 cladding

Zircaloy-4 is an essential material for cladding structures within fission-based reactors. To explore the changes in properties measured on differing length scales, FIB-machined micro-scale tensile tests were performed on both irradiated and control groups of Zircaloy-4. This was correlated with tensile testing on femtosecond laser-machined meso-scale specimens. Pronounced size effects were found when varying specimen geometry. Increases in tensile geometry size were associated with a reduction in measured yield stress for both irradiated and unirradiated samples. Here, meso-scale testing found strength and strain values similar to that of bulk-scale testing.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Post-irradiation examination of UN-Mo-W fuels for space nuclear propulsion

The National Aeronautics and Space Administration's return to space nuclear propulsion stems from the need for a more efficient method of space travel. Nuclear thermal propulsion systems have been shown to be two times more efficient than chemical propulsion. NASA's Sirius program was created to fabricate and test fuels for space nuclear propulsion, specifically to determine their performance under prototypical startup conditions. The Sirius project featured 4 test capsules, Sirius-1 featured uranium nitride fuel dispersed in a matrix of tungsten and rhenium, while Sirius-2A, -2B, and -3 featured uranium nitride-molybdenum-tungsten fuel (UN-Mo-W). This study discusses the Sirius-2A and -2B irradiation experiments at the Idaho National Laboratory, specifically their performance under irradiation at the Transient Reactor Test Facility. It was found that the fuel samples overall did not exhibit significant cracking, though the Sirius-2A fuel did have one large crack on the surface of the fuel. There was minimal hydrogen absorption in the samples, though it is unknown if the absorption occurred during irradiation or during fabrication. Mechanical testing indicated that the UN fuel demonstrated ceramic behavior as expected, and the Mo/W matrix demonstrated linear elastic behavior to failure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Effects of elevated temperature exposure on the residual stress state and microstructure of PVD Cr coatings on SiC investigated via in situ X-ray diffraction and transmission electron microscopy

Pure Cr coatings are being considered as corrosion-resistant layers for SiC-SiC f composite fuel cladding in light water reactors. The residual stress state and interdiffusion behavior of Cr coatings deposited on SiC substrates using bipolar high-power impulse magnetron sputtering (B-HiPIMS) and direct current impulse magnetron sputtering (DCMS) were studied using in situ X-ray diffraction experiments at 300 °C, 700 °C, and 1200 °C. After elevated temperature exposure, evolution of the coating microstructure and coating/substrate interdiffusion was characterized by transmission electron microscopy. The compressive residual stress of the dense B-HiPIMS coating at room temperature was gradually relieved at 300 °C, while the stress transitioned to a tensile state at 700 °C. Grain growth of the Cr coating and a Cr–SiC interdiffusion layer consisting of newly formed Cr–C and Cr–Si phases were observed in the B-HiPIMS coating after exposure at 700 °C. The phases in the interdiffusion layer were identified to be Cr 7 C 3 , Cr 3 Si, and Cr 5 Si 3 C x experimentally and by thermodynamic predictions. For the DCMS coating, however, changes in residual stress and microstructure at the elevated temperatures were limited by the inherent porosity and inhomogeneous structure of the as-deposited coating. At 1200 °C, both coating types interacted rapidly with the underlying SiC substrate in the early stages of the experiment. The results from this study suggest that the characteristics of the Cr coatings do not significantly change at the reactor operating temperature of 300 °C, while changes in microstructure and stress state are expected in off-normal conditions of 700 °C and 1200 °C. Finally the results from this study suggest the B-HiPIMS coating is best-suited as a protective coating for SiC fuel cladding.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Small Scale Creep Testing of 14YWT via In-situ Transmission Electron Microscopy Irradiation and Nanoindentation

The next generation of nuclear materials must withstand harsh operating conditions such as high temperatures and irradiation doses. Nanostructured ferritic alloys like oxide dispersion strengthened steels, especially 14YWT, have shown promise as a structural material to withstand these harsh operating conditions. For application, understanding the irradiation enhanced creep of the structural components is fundamental to evaluating the service life in a reactor. Challenges with evaluating irradiation enhanced creep are related to the limited space in nuclear reactors and the expense of conducting post irradiation experiments on irradiated cladding. Ion irradiations are considered to expedite irradiation testing, but such experiments have restricted depth of penetration into the material, limiting the ability to characterize changes to material properties after irradiation. Small scale mechanical testing can be used with ion beam irradiations as a potential avenue to evaluate the irradiation enhanced creep of materials. In this study, in-situ transmission electron microscopy nanopillar creep studies on 14YWT were performed and simultaneously ion beam irradiated with 2.8 MeV Au4+ ions. It was observed that the ion beam irradiation did increase the measured strain rate of the materials. In addition, ex-situ nanoindentation creep studies were performed over a range of temperatures on control 14YWT, and it was observed that there was a change in the deformation mechanism between 873 K and 1073 K that agrees well with macro-scale mechanical testing. Furthermore, these results show promise for applying these techniques to nuclear materials in the future.

14 YWT↗
DOE OSTI2023

The effects of radiation-induced grain subdivision and dislocations on the fracture properties of uranium dioxide

Microcantilever bending tests were applied on uranium dioxide samples irradiated by 84 MeV Xe ions. Depending on the initial grain sizes, different levels of grain subdivision and dislocation development were observed near the irradiated surface. These radiation damage features were shown to degrade the fracture properties of both samples. The fundamental aspects of the radiation-microstructure-property relationships were discussed in this paper, taking a holistic consideration of the microcantilever bending data, fractography and microstructures. The separate effects of the irradiation-induced grain subdivision and dislocations on the fracture initiation and propagation of uranium dioxide were discussed. In conclusion, it was observed that radiation-induced dislocations and grain subdivision, without the presence of Xe bubbles co-located with the defects, cause reductions in the Young's modulus, fracture strain, and fracture stress, and a transition to intergranular fracture.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Post-irradiation characterization of a high burnup mixed oxide fuel rod with minor actinides

The Advanced Fuels Campaign performed a series of irradiation tests of minor actinide-bearing mixed oxide fuel (MA-MOX), the so-called AFC-2C&D experiments, to investigate the transmutation of long-lived transuranic actinide isotopes contained in spent nuclear fuel via fast reactor technology at burnups exceeding 10 % fission of initial metallic atoms. This manuscript reports the test results derived from one of the five MA-MOX rodlets taken to higher burnup in the AFC-2D irradiation. This includes both non-destructive investigations, such as gamma and neutron spectrometry, and destructive investigations, such as fission gas release, ceramography, and chemical burnup analysis. In addition, the microstructure of the fuel was investigated using advanced electron microscopy techniques including electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). Here, it was observed with EBSD that the pellet had subdivision of the grains and the TEM observed migration of cladding material into the 5 metal precipitates in the fuel which could have been from the higher than desired oxygen/metal ratio. The TEM also showed an enrichment of Cr in fuel clad chemical interaction (FCCI) layer.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Acoustic mapping by picosecond ultrasonics for elastic property measurement: Experimental demonstration on a TRISO fuel compact

Picosecond ultrasonics has been demonstrated on a tristructural isotropic (TRISO) fuel compact to measure the elastic properties of each compact layer. This technique utilizes an ultrashort pump laser pulse to excite vibrations in a gold transducer film covering the surface of each component and a second probe laser pulse to record the resulting acoustic strain induced change in optical reflectance. From the damping of this film vibration, the acoustic reflection coefficient, which couples the elastic properties of the transducer film and the sample, can be obtained, enabling a calculation of the sample’s acoustic velocity and elastic modulus. Overall, results obtained from this method are consistent with known values of elastic moduli, namely that the SiC coating is the stiffest component of the compact while the carbonaceous matrix is the most compliant. Nanoindentation was conducted as a benchmark technique on the same sample and shows satisfactory agreement with the results of picosecond ultrasonics. Compared to other methods like nanoindentation, picosecond ultrasonics is multimodal with a capability of measuring several key properties simultaneously and has potentials to be coupled into optical fibers for remote sensing. Thus, these demonstration measurements reveal the methodology to be a promising candidate for in-situ and high-throughput optical characterizations of nuclear materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Post-irradiation examination of legacy high burnup fuel to support safety testing

Safety/transient testing to evaluate performance under off-normal conditions is an essential pillar for both the development of Accident Tolerant Fuels (ATF) and the optimization of fuel operation economics beyond current discharge burnups. Among other factors, the successful interpretation of the transient testing results relies upon the knowledge of the initial conditions of the test, including the characteristics of the fuel system under scrutiny. When testing pre-irradiated material, the assumptions that the fuel and the cladding still have the same properties as in the pre-irradiation stage is obviously wrong and could affect the results of the test. This is particularly true the more burnup accumulates in the fuel rod and irradiation progresses. The knowledge of the initial microstructure of both fuel and cladding allows a clearer interpretation of the subsequent transient testing results, provides validation of the physical phenomena underlying the model predictions and eliminates the uncertainties related to the limited knowledge of the sample status before the test. One example is the phenomenon of fine fragmentation that occurs in Light Water reactor (LWR) fuel. During a Loss of Coolant Accident (LOCA) or Reactivity Initiated Transient (RIA) the fuel can severely fragment. During LOCA, high burnup fuel tend to finely fragment, which has raised safety concerns due to the increased likelihood of dispersal of such small particles once the cladding has burst and due to the increased fission gas release. Therefore, efforts have been devoted to the assessment of a pulverization threshold that could determine the conditions under which fine fragmentation is predominant. However, the lack of information regarding the initial conditions of the fuel, and the connections between those conditions and the pre-transient irradiation history, have hindered the development of a fully mechanistic fragmentation and pulverization criterion. The empirical relationships rely on conservative estimations, due to the lack of information on critical material properties and characteristics. More generally, experimental evidence of the irradiation-induced modifications at microstructural scale are necessary to determine the behavior of the material at the macroscopic scale, with the latter being the one of technological interest. Significant progress has been made in the last two decades in the developments of analytical materials science techniques that can be applied to highly radioactive materials, such as high burnup fuels. The availability of new techniques and the improvement of existing ones has enabled investigations previously not possible that can deepen the understanding of the fuel characteristics and properties at high burnup. The better knowledge of material behavior and irradiation-induced phenomena could help the prediction of its performance. In this context, the scope of the present work is to apply a wide portfolio of advanced characterization techniques to determine properties that are relevant for safety and performance. The results are interpreted in the context of engineering scale post-irradiation examinations and available information on the irradiation conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Micro-cantilever beam experiments and modeling in porous polycrystalline UO 2

Understanding the impact of microstructure on the thermo-mechanical behavior of oxide nuclear fuels is vital to predicting their performance through multiscale models. Evaluating the mechanical properties at the sub-grain length scale is key to developing these multiscale models. In this work, 3D finite element (FE) models were constructed to simulate the micrometer-scale bending of micro-cantilever beams fabricated using porous polycrystalline uranium dioxide (UO 2 ) and tested at room temperature. Here, the results showed that the porosity and elastic anisotropy of individual grains can play a significant role in determining the effective mechanical properties of the material deduced from the tests. Specifically, the porosity had a non-negligible effect, given that the pore size was of the same order of magnitude as the dimensions of the micro-beams. Correlations between load-deflection data, pore location, and elastic properties (effective Young's modulus) were investigated using UO 2 micro-beam FE models, where pore clusters were included and placed at different locations along the length of the beam. Results indicated that the presence of pore clusters near the substrate, i.e., the clamp of the micro-cantilever beam, has the strongest effect on the load-deflection behavior, with the porosity leading to a reduction of stiffness that is the largest for any location of the pore clusters. Furthermore, it was also found that pore clusters located towards the middle of the span and close to the end of the beam have a comparatively small effect on the load-deflection behavior. Therefore, it is concluded that accurate estimates of Young's modulus can be obtained from micro-cantilever experiments after accounting for porosity on the one third of the beam length close to the clamp. This, in turn, provides an avenue to improve microscale experiments and their analysis in porous, anisotropic elastic materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗