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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

Abstract The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important potential comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The irradiations were the beginning tests to qualify U-10wt%Zr as a driver fuel for FFTF. The FFTF core, with a 91.4 cm tall fuel column and a chopped cosine neutron flux profile, operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in previous EBR-II experiments that had a 32-cm height core. The MFF-3 and MFF-5 qualification assemblies operated in FFTF to >10 at% burnup, and performed very well with no cladding breaches. The MFF-3 assembly operated to 13.8 at% burnup with a peak inner cladding temperature of 643°C, and the MFF-5 assembly operated to 10.1 at% burnup with a peak inner cladding temperature of 651°C. Because of the very high operating temperatures for both the fuel and the cladding, data from the MFF assemblies are most comparable to the data obtained from the EBR-II X447 experiment, which experienced two pin breaches. The X447 breaches were strongly influenced by a large amount of fuel/cladding chemical interaction (FCCI). The MFF pins benefitted from different axial locations of high burnup and peak cladding temperature, which helped to reduce interdiffusion between rare earth fission products and stainless steel cladding. Post-irradiation examination evidence illustrates this advantage. Comparing other performance data of the long MFF pins to prior EBR-II test data, the MFF fuel inside the cladding grew less axially, and the gas release data did not reveal a definitive difference.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Determination of Scale Bar for AGHCF Metallography Data

The DOE Nuclear Energy Advanced Reactor Technologies (ART) Fast Reactor Program (FRP) has supported the development of several databases containing information on the safety performance of fast reactors, components, and fuels. This growing collection of legacy experimental data, operating data, and analysis is available online to registered users. Metallography data represents one of the most critical types of post-irradiation examination (PIE) data being collected, organized, and archived in several ART Fast Reactor Databases (https://frdb.ne.anl.gov), including the Metallic Fuels Irradiation & Physics Database (FIPD), Out-of-Pile Transient Database (OPTD), and TREAT (the Transient Reactor Test Facility) Experimental Relational Database (TREXR). These databases contain three principal sets of metallography data. The first set comprises metallography data from Experimental Breeder Reactor-II (EBR-II) and Fast Flux Test Facility (FFTF) irradiated fuel pins examined in the Hot Fuel Examination Facility (HFEF). The second set consists of metallography data from EBR-II irradiated fuel pins examined in the Alpha-Gamma Hot Cell Facility (AGHCF). The third set includes metallography data from transient-tested fuel pins (including both out-of-pile furnace tests and TREAT tests) examined in AGHCF. Since both the second and third sets were generated in AGHCF, they are governed by identical specifications. The metallography data in the databases consist of digital images scanned from either positive or negative photographic films. To analyze the microstructure of a fuel pin, a series of preparatory steps are required, including sectioning, epoxy mounting, mechanical grinding and polishing, and etching. Following sample preparation, specimens are transferred for metallographic examination. The AGHCF and HFEF metallography data were generated using optical microscopes manufactured by Leitz and Bausch and Lomb (B&L). Images were recorded on Polaroid film at preset magnifications. Magnification verification for the Leitz and B&L metallographs was conducted every two months prior to 1989 and at least every six months from 1989 through the conclusion of the IFR program. Magnifications determined from imaging of microslide standards were compared to the instrument settings for magnifications ranging from 50× to 500×. If the magnifications determined from standards deviated from the instrument settings, adjustments were made to the bellows extension until agreement was achieved. The specifications for AGHCF and HFEF legacy metallography data have been established based on available hard-copy and digital records, most of which have been incorporated into the data repositories associated with FIPD, OPTD, and TREXR. Detailed specifications including hard-copy records, digitized records, cutting diagrams and sectioning schemes, high-magnification photographs, photomosaics (composites), information tags, scale bars, and magnification verification procedures can be found in a separate report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Specifications of Legacy U(Pu)Zr Metallography Data

The DOE Nuclear Energy Advanced Reactor Technologies (ART) Program has supported the creation of several databases with information describing the safety performance of fast reactors, components, and fuels. This growing collection of legacy experimental data, operating data, and analysis is available online to registered users. Metallography data is one of the most important types of PIE data being collected, organized and stored in several ART Fast Reactor Databases (https://frdb.ne.anl.gov), including the Fuels Irradiation & Physics Database (FIPD), Out-of-Pile Transient Database (OPTD), and TREAT (the Transient Reactor Test Facility) Experimental Relational Database (TREXR). These databases contain three main sets of metallography data. The first set is the metallography data from Experimental Breeder Reactor-II (EBR-II) and Fast Flux Test Facility (FFTF) irradiated fuel pins measured in the Hot Fuel Examination Facility (HFEF); the second set is the metallography data from EBR-II irradiated fuel pins measured in the Alpha-Gamma Hot Cell Facility (AGHCF); the third set is the metallography data from transient tested fuel pins (the transients tests include the out-of-pile tests and TREAT tests) measured in AGHCF. Since both the second and third sets of data were measured in AGHCF, they are governed by the same specification. The metallography data in the databases are digital images scanned from either positive or negative photos. The quality of the images, including their resolution and contrast, relies on the preserved quality of the pictures and the scanning conditions. To analyze the microstructure of a fuel pin, a series of preparatory steps must be undertaken. These include sectioning, epoxy mounting, mechanical grinding and polishing, and often etching. The resulting samples were then transferred to a secondary hot cell, or glovebox (depending the strength of radiation field) for microscopy examination. The specifications provided herein focus on the metallography examinations. The sample grinding, polishing and etching processes are also discussed. The procedures involving sectioning and epoxy mounting are out of the scope of the current specification; details can be found in the corresponding operation manuals. If more data are collected and added to the ART Fast Reactor Databases, this specification will be updated to accommodate the additional data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Development of PNNL’s Plutonium Metallography Capability

The goal of this project was to develop plutonium metallography capabilities and establish workflows for characterization at PNNL’s Radiological Processing Facility (RPL). Developing and expanding on this capability opens more opportunities for PNNL to better support science through collaborations with other labs, plants, and sites withing the Department of Energy Complex and through programs within the National Nuclear Security Administration. This project established metallography equipment in an air glovebox, polishing protocols, as well as radiological and facility controls and procedures so that gram quantities of plutonium metal could be analyzed on multiple characterization tools. Data was collected and analyzed on two different delta phase plutonium - Gallium samples and a calciothermically reduced alpha plutonium metal using a combination of optical and electron microscopy, powder X-ray diffraction, and atom probe tomography. This report highlights the metallography capabilities, some preliminary data collections, and plans at PNNL to support plutonium material science.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Sectioning, Metallography and Burnup Determination of MFF Fuel Pins

This test plan describes the requirements for sample preparation only. Metallography is also covered. Burnup and density exams will be performed in the Idaho National Laboratory (INL) Analytical Laboratory and planning will be done using the required paperwork of that facility; the retained fission gas work will be covered by a separate test plan.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Metallography Box Improvements

The information included in this poster is designed to provide information on a visualization system needed in a hot-cell with a high radiation dose level. The system is necessary to enable real-time remote monitoring of microscope function. The current configuration for viewing function of in-cell equipment is through a single window. This makes for difficult viewing of the microscope workstations which are located away from the window. A properly designed camera system enhances troubleshooting capability in-cell. The poster will aid in the exposition of the design process and implementation of the system described above.

42 ENGINEERING↗
DOE OSTI2014

Metallography and microstructure interpretation of some archaeological tin bronze vessels from Iran

Archaeological excavations in western Iran have recently revealed a significant Luristan Bronzes collection from Sangtarashan archaeological site. The site and its bronze collection are dated to Iron Age II/III of western Iran (10th–7th century BC) according to archaeological research. Alloy composition, microstructure and manufacturing technique of some sheet metal vessels are determined to reveal metallurgical processes in western Iran in the first millennium BC. Experimental analyses were carried out using Scanning Electron Microscopy–Energy Dispersive X-ray Spectroscopy and Optical Microscopy/Metallography methods. The results allowed reconstructing the manufacturing process of bronze vessels in Luristan. It proved that the samples have been manufactured with a binary copper–tin alloy with a variable tin content that may relates to the application of an uncontrolled procedure to make bronze alloy (e.g. co-smelting or cementation). The presence of elongated copper sulphide inclusions showed probable use of copper sulphide ores for metal production and smelting. Based on metallographic studies, a cycle of cold working and annealing was used to shape the bronze vessels. - Highlights: • Sangtarashan vessels are made by variable Cu-Sn alloys with some impurities. • Various compositions occurred due to applying uncontrolled smelting methods. • The microstructure represents thermo-mechanical process to shape bronze vessels. • In one case, the annealing didn’t remove the eutectoid remaining from casting. • The characteristics of the bronzes are similar to other Iron Age Luristan Bronzes.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Characterization of Pu-238 Heat Source Granule Containment

The Milliwatt Radioisotopic Themoelectric Generator (RTG) provides power for permissive-action links. Essentially these are nuclear batteries that convert thermal energy to electrical energy using a doped silicon-germanium thermopile. The thermal energy is provided by a heat source made of 238 Pu, in the form of 238 PuO 2 granules. The granules are contained by 3 layers of encapsulation. A thin T-111 liner surrounds the 238 PuO 2 granules and protects the second layer (strength member) from exposure to the fuel granules. An outer layer of Hastalloy-C protects the T-111 from oxygen embrittlement. The T-111 strength member is considered the critical component in this 238 PuO 2 containment system. Any compromise in the strength member seen during destructive testing required by the RTG surveillance program is characterized. The T-111 strength member is characterized through Scanning Electron Microscopy (SEM), and Metallography. SEM is used in the Secondary Electron mode to reveal possible grain boundary deformation and/or cracking in the region of the strength member weld. Deformation and cracking uncovered by SEM are further characterized by Metallography. Metallography sections are mounted and polished, observed using optical microscopy, then documented in the form of microphotographs. SEM mat further be used to examine polished Metallography mounts to characterize elements using the SEM mode of Energy Dispersive X-ray spectroscopy (EDS).

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2017

Microstructures of chemical vapor deposited high-purity tungsten achieved by two different precursors

Highlights: • CVD tungsten used as electron emitters of thermionic fuel elements was investigated. • Microstructures were characterized with metallography analysis. • The preferred orientation was determined with XRD and EBSD. • The surface morphology was observed with SEM and WLI. • Tungsten via thermolysis of WCl{sub 6} is better for thermionic fuel elements. - Abstract: Chemical vapor deposited (CVD) high-purity tungsten can be manufactured as electron emitters in thermionic fuel elements. Exposed surface planes of the emitter emit electrons and affect the efficiency and load capacity of the thermionic fuel element. In order to determine the electron emission planes of the emitter, the metallurgic structures, the preferred orientation and the as-deposited surface morphology of CVD tungsten have been investigated. In this work, high-purity (over 99.99%) tungsten was achieved by two different precursors: hydrogen reduction of tungsten hexafluoride (WF{sub 6}) and thermolysis of tungsten hexachloride (WCl{sub 6}). The microstructures of CVD tungsten were characterized by metallography analysis, X-ray diffraction (XRD), electron backscattered diffraction (EBSD), and white light interference (WLI). Tungsten fabricated by hydrogen reduction of WF{sub 6} and thermolysis of WCl{sub 6} has <100>-preferred columnar microstructures. However, the latter exhibits larger columnar grains and preferred <100> less. The preferred orientation is caused by the higher growth rate of the <100> orientation and competition between the <100> orientations of different grains. The as-deposited surface consists of tiny pyramids, with <100> axis and {111} side faces. In contrast to tungsten produced by hydrogen reduction of WF{sub 6}, the pyramid on tungsten synthesized via thermolysis of WCl{sub 6} contains small {110} facets on the four edges. Therefore, tungsten made via thermolysis of WCl{sub 6} is a better candidate for thermionic fuel elements, because the higher work function of {110} planes enables better output efficiency.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Comparison of Zirconium Redistribution in BISON EBR-II Models Using FIPD and IMIS Databases with Experimental Post Irradiation Examination

Metallic fuels have seen increased interest for future sodium fast reactors due to their material properties: high thermal conductivities and advantageous neutronic properties allow for greater fission densities. One drawback to typical metallic fuels is zirconium redistribution, which impacts this advantageous material and its neutronic properties. Unfortunately, the processes behind zirconium migration behavior are understood using first principles, so before these fuels are implemented in future fast reactors, characterization and fuel qualification regimes must be completed. These activities can be supported through the use of robust modeling using the most accurate empirical models currently available to fuel researchers around the world. The tool that allows researchers to model this complex coupled thermo-mechanical behavior and nuclear properties is BISON. Additionally, BISON model parameters need to be compared against PIE measurements. The current work utilizes two fuel pins from EBR-II experiment X441 to optimize various model parameters, including porosity correction factor, thermal conductivity, phase transition temperature, and diffusion coefficient multipliers, before implementing the final model for seven fuel pins with differing characteristics. To properly evaluate the BISON simulations, the results are compared to PIE metallography data for each fuel pin, to ensure the zirconium redistribution is properly reflected in the simulation results. Six out of seven analyzed fuel pins demonstrate good agreement between the metallography images and BISON results, showing alignment of the Zr-rich, Zr-depleted, and moderately Zr-enriched zones at various axial heights along the fuel pins. Further work is needed to refine the model parameters for general pin use.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Summary of FUTURIX-FTA Non-destructive Examination

A summary of results from the non-destructive examination of the FUTURIX-FTA irradiation are presented. This irradiation was conducted under true fast spectrum conditions in the Phenix reactor. Baseline non-destructive examinations of these tests have been completed at this time. Results from visual examination, neutron radiography, gamma spectrometry, and dimensional inspection have been collected. The data collected from this irradiation will be compared to sister irradiations performed in pseudo-fast spectrum conditions at the INL Advanced Test Reactor helping to confirm fuel performance behavior observed in Advanced Test Reactor irradiations is representative of fast reactor conditions. All indications from the non-destructive examinations performed on the FUTURIX-FTA elements indicate a successful irradiation with fuel performance in line with expectations. The intact state of the nitride fuel is especially notable as these pellets were fragile after fabrication, but they appear to have survived the irradiation and extensive shipping well. These rodlets will undergo further examination to include gamma tomography to investigate the radial fission product distribution, fission gas release analysis, optical metallography, microhardness testing, and chemical burnup analysis. If possible, cross sections from optical metallography may also be examined by more advanced techniques such as scanning electron microscopy and transmission electron microscopy of specific sites through the use of focused ion beam lift outs. The results of all these examinations will be compared with the results from the AFC-1B, AFC-I1D, AFC-IF, AFC-1G, AFC-1H and AFC-LAE irradiations. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Multimodal Defect Imaging of Pure Tungsten Components Fabricated via Electron Beam Powder Bed Fusion

The utilization of additive manufacturing (AM) techniques for refractory materials in high-temperature environments has significantly expanded because of the ability to fabricate geometrically complex components. Electron beam powder bed fusion (EB-PBF), which provides lower residual stress, a cleaner vacuum environment, and better efficiency for high melting point, is one of the best-suited AM methods to produce advanced refractory components. However, the property variation attributed to the heterogeneous microstructure and process-induced defects has hindered the widespread adoption of EB-PBF-produced material like tungsten. While numerous in-situ monitoring and defect detection methods have been demonstrated for EB-PBF, a workflow that compares and evaluates process-induced abnormalities from different imaging perspectives is still limited. This study examines a feature-embedded tungsten component manufactured via the EB-PBF process to demonstrate the defect detection capabilities of a multimodal defect imaging workflow. The predefined and process-induced defects are evaluated by harnessing various imaging techniques, including in-situ electron imaging, layerwise near-infrared (NIR) imaging, post-build high-energy x-ray computed tomography (CT), and conventional destructive metallography. The results highlight the strengths and limitations of distinctive defect imaging techniques concerning specific defect types, sizes, and conditions. It was found that electron imaging can provide more abnormal detection capabilities while maintaining a higher measuring accuracy, against the conventional metallography in this case study, compared with NIR and CT imaging techniques.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Austenite formation and cementite dissolution during intercritical annealing of a medium-manganese steel from a martensitic condition

In the present study, in situ high-energy X-ray diffraction (HEXRD), dilatometry, and metallography were conducted for the purpose of understanding microstructure evolution during intercritical annealing of a Fe-0.19C-4.39Mn steel from a martensitic initial condition. Three different simulations for austenite growth and solute partitioning during isothermal holding were conducted using the DICTRA™ module of Thermo-Calc®. The microstructures after intercritical annealing exhibit film-like retained austenite; however, in situ HEXRD indicates that retained austenite in the initial martensitic microstructure had decomposed upon heating to the isothermal holding temperature, suggesting that film-like austenite may be generated during intercritical annealing via nucleation and growth of new austenite and without necessarily preserving initial retained austenite films in the martensitic microstructure. Metallography and in situ HEXRD also indicate that cementite had formed upon heating but did not readily dissolve during an 1800 s isothermal hold although considerable growth of austenite had occurred during the isothermal hold, suggesting that austenite nucleation and initial growth during intercritical annealing does not solely occur due to carbon partitioning. A simulation incorporating cementite suggests that the increase in austenite fraction during intercritical annealing initially occurs in association with Mn partitioning followed by austenite growth induced by gradual cementite dissolution.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Non-destructive Quantitative Phase Analysis and Microstructural Characterization of Zirconium Coated U-10Mo Fuel Foils via Neutron Diffraction

This report uses neutron diffraction to investigate the crystal phase composition of uranium-molybdenum alloy foils (U-10Mo) for the CONVERT MP-1 Reactor Conversion Project, and determines the effect on alpha-uranium contamination following the deposition of a Zr metal diffusion layer by various methods: plasma spray deposition of Zr powders at LANL and hot co-rolling with Zr foils at BWXT. In summary, there is minimal decomposition of the gamma phase U-10Mo foil to alpha phase contamination following both plasma spraying and hot co-rolling. The average unit cell volume, i.e. lattice spacing, of the Zr layer can be mathematically extracted from the diffraction data; co-rolled Zr matches well with literature values of bulk Zr, while plasma sprayed Zr shows a slight increase in the lattice spacing, indicative of interstitial oxygen in the lattice. Neutron diffraction is a beneficial alternative to conventional methods of phase composition, i.e. x ray diffraction (XRD) and destructive metallography. XRD has minimal penetration depth in high atomic number materials, particularly uranium, and can only probe the first few microns of the fuel plate; neutrons pass completely through the foil, allowing for bulk analysis of the foil composition and no issues with addition of cladding layers, as in the final, aluminum-clad reactor fuel plates. Destructive metallography requires skilled technicians, cutting of the foil into small sections, hazardous etching conditions, long polishing and microscopy times, etc.; the neutron diffraction system has an automated sample loader and can fit larger foils, so there is minimal analysis preparation; the total spectrum acquisition time is ~ 1 hour per sample. The neutron diffraction results are limited by spectra refinement/calculation times and the availability of the neutron beam source. In the case of LANSCE at Los Alamos, the beam operates ~50% of the year. Following the lessons learned from these preliminary results, optimizations to the process and analysis can be made, and neutron diffraction can become a viable and efficient technique for gamma/alpha phase composition determination for nuclear fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Examination of Legacy Metallic Fuel Pins (U-10Zr) Tested in FFTF

The MFF series of metallic fuel (U-10wt%Zr) tests performed in the Fast Fuel Test Facility (FFTF) were the beginning tests to qualify the fuel as a driver fuel for FFTF. They all performed very well, to relatively high burnup and with no pin breaches. Tests MFF-3 and MFF-5 were chosen to be destructively examined because they were run at high peak cladding temperatures, 643 and 649°C respectively, and to modest to high burnup (138 and 101 MWd/kgM respectively). They were the only sodium fast reactor (SFR) metallic tests, with long (91.4 cm) fuel columns and clad in HT9, to be operated at these high temperatures. Extensive operating condition reconstruction was performed to accompany the examination, providing detailed operating conditions axially along the pins. In addition, detailed ORIGEN calculations provided fuel burnup along the pin length. These calculations and detailed cross-section metallography allowed a PhD dissertation to be performed where the fuel/cladding chemical interaction (FCCI) was modeled using Fickian and Soret Effect driven diffusion of rare earth fission products to the cladding surface and interact with the cladding. The examination of the MFF-3 and MFF-5 pins included neutron radiography, axial profilometry including pin bow and length measurements, precision gamma scanning, plenum gas puncturing to measure fission gas release, detailed chemical/isotopic analyses of fuel samples to confirm burnup calculations, and metallography of pin cross-sections, including micro-hardness measurements of fuel and cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗