Accelerated Materials Evaluation for Nuclear Applications Utilizing Irradiation and Integrated Modeling: Current and Advanced Nuclear Fuels
Sponsored by: TMS Structural Materials Division, TMS: Nuclear Materials Committee
Program Organizers: Assel Aitkaliyeva, University of Florida; Peter Hosemann, University of California - Berkeley; Samuel Briggs, Oregon State University; David Frazer, Los Alamos National Laboratory
Tuesday 2:00 PM
February 25, 2020
Room: Theater A-8
Location: San Diego Convention Ctr
Session Chair: Assel Aitkaliyeva, University of Florida
2:00 PM
Alpha Self-Irradiation of Archive and Irradiated Fast Reactor Fuels: Thierry Wiss1; Oliver Dieste1; Emanuele De Bona1; Dragos Staicu1; 1European Commission - Jrc
Mixed Uranium-Plutonium Oxides are suitable fuels for fast reactors and can be used for the transmutation of the radiotoxic minor actinides but also burning of excess plutonium. The high alpha-activity of the minor actinides lead to the formation of large amount of defects, which will appear even before irradiation in the reactor. In this work we are reporting on experimental observations of the alpha-damage effects on different minor actinide containing fuel samples. The microstructure of these samples has been investigated by Electron Transmission Microscopy to assess the formation of extended defects. Lattice parameter and strain in damaged samples have been measured as a function of the cumulated alpha dose and the corresponding concentration of point defects estimated. Differential Scanning Calorimetry measurements were also performed on several samples and the stored energy released during anneals attributed to the different type of defects.
2:20 PM
Comparison of Radial Microstructural Changes in Fast Reactor MOX Fuels Across Varying Burnup Profiles: Riley Parrish1; Assel Aitkaliyeva2; 1Sandia National Laboratory; 2University of Florida
This work will show results from the examination of three fast reactor mixed-oxide (MOX) fuel pellets irradiated to approximately 3%, 14%, and 21% fissions per initial metal atom (FIMA) in the Fast Flux Test Facility. Results from scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS) analysis will demonstrate the evolution of microstructural features along the radial thermal gradient of the fuel pellet. Across all burnups, the morphology and distribution of solid fission products show distinct differences at each of the radial positions. At 14% and 21% FIMA, a Pd-rich metallic precipitate begins to form alongside the five metal ε-phase particles and insoluble perovskite oxide. The fuel-cladding gap has accumulated metallic species on the cladding surface along with a dense layer rich in volatile fission products consistent with the Joint Oxide Gain.
2:40 PM
Synthesis of Intermetallic UZr2+x and Its Phase Transformation: Tiankai Yao1; Michael Benson1; Jason Harp1; Lingfeng He1; Jian Gan1; 1Idaho National Laboratory
UZr2+x (where x=0, 0.3, 0.6) is an intermetallic compound in the U-Zr system. Its significance grows with the rekindled research interest on U-Zr based metallic nuclear fuels for next generation fast reactors. As part of the Thermal Energy Transport under Irradiation (TETI) EFRC, INL team revisited the synthesis of UZr2+x by arc melting alloys and studying the phase transformation of these intermetallic compounds. The alloys were heat treated by different combinations of annealing and quenching. Theoretically, UZr2+x crystalizes in hexagonal structure (delta phase) at room temperature and is a cubic structure (gamma phase) at high temperature. However, a transformation from delta to gamma phase was triggered by ion irradiation even significantly below the phase transformation temperature. Our results suggest that the UZr2+x phase performs as gamma phase with isotopic properties due to the body centered cubic crystal structure regardless of irradiation temperatures inside a reactor.
3:00 PM
Testing of Nuclear Fuels and Materials in the Advanced Fuels Campaign: Geoffrey Beausoleil1; Christopher Petrie2; Walt Williams3; 1Idaho National Laboratory; 2Oakridge National Laboratory; 3Purdue University
The Department of Energy (DOE) Advanced Fuels Campaign (AFC) has initiated an accelerated fuel irradiation and qualification pathway for developing advanced reactors fuel for lead test assembly (LTA) testing. This process includes the performance of a Phenomena Identification Ranking Table (PIRT) analysis for a fuel form and an evaluation of the existing modelling and experimental capabilities to support investigation of the highest priority phenomena. The campaign can then utilize the separate effect testing platforms MiniFuel at ORNL and DISECT from NSUF as well as the semi-integral testing platform FAST at INL. These irradiation tools are designed with a focus on high throughput high-volume experiments that can accelerate the time to reach deep burnup targets. These tools can be performed in parallel to supporting fuel performance modelling efforts to support the insertion of a LTA in an accelerated time frame (i.e., a reduction of a 20 year process to 7 years).
3:20 PM
In-situ Neutron Characterization of Advanced Nuclear Fuels - The Road to a New Neutron Irradiation Testing Capability: Edward Obbard1; Claudia Gasparrini1; Patrick Burr1; Kyle Johnson2; Denise Lopes2; Clara Anghel3; Simon Middleburgh3; Daniel Gregg4; Klaus Dieter Liss4; Grant Griffiths4; Nicholas Scales4; Gordon Thorogood4; Greg Lumpkin4; 1UNSW Sydney; 2Kungliga Tekniska Hogskolan (KTH); 3Westinghouse Electric Sweden AB; 4ANSTO
Neutron diffraction offers a method to characterise encapsulated nuclear materials, reducing the need to handle contaminating samples. Precise thermal expansion measurements and phase characterisation were performed from room temperature to 1873K, for U3Si2 and UN/U3Si2 accident tolerant fuel composites. U3Si2 shows a negative, linear temperature dependence of the instantaneous thermal expansion described by α(T) = 2.10E-05 – 7.25E-09 x T (1/K). Precise thermal expansion for UN was obtained over this temperature range. Post annealing characterisation highlighted the reaction between vanadium canning material and UN-U2.8Si2 composite, accompanied by vanadium grain boundary transport and formation of V5Si3 and V3Si. Results are interpreted in their direct context, and also regarding wider aims to enable testing of advanced fuels in research reactors that may not have extensive fuel PIE facilities but nonetheless do have world-leading small angle, wide angle and neutron tomography capability for this type of non-contact experimentation.
3:40 PM Break
3:55 PM
Changes in the Starting Microstructures of U-Mo Fuels due to the Effects of Neutron Irradiation: Dennis Keiser1; Brandon Miller1; Jan-Fong Jue1; Adam Robinson1; Kelley Verner1; 1Idaho National Laboratory
The Materials Management and Minimization Reactor Conversion Fuel Qualification Program is developing low enriched U-Mo fuels to replace the use of fuels containing highly enriched uranium. A monolithic and dispersion fuel is being developed. As part of this development, U-Mo fuel plates have been irradiated in the Advanced Test Reactor under a variety of conditions and then characterized using scanning electron microscopy transmission electron microscopy, and other techniques. Based on this characterization, changes that occur in the starting microstructures of both fuel types, due to the effects of irradiation, have been identified. This presentation will describe the microstructural evolution that occurs and how data that have been generated can be employed to better model the behavior of U-Mo fuels during irradiation. Examples of data about U-Mo alloys that have been produced include those related to: amorphization, recrystallization/polygonization, fission product phase development, and changes in fission gas bubble size/distribution/morphology.
4:15 PM
In-situ Observation of Radiation-induced Phase Transformation in U-Mo: Bei Ye1; Weiying Chen1; Yinbin Miao1; Abdellatif Yacout1; Yipeng Gao2; 1Argonne National Laboratory; 2Idaho National Laboratory
The metastable γ-U phase in U-Mo fuels is prone to phase decomposition to become α-U phase and enriched U2Mo, primarily at cell or grain boundaries. This mixture reverses back to the γ-U phase under neutron irradiation. It was speculated that recrystallization in U-Mo fuels may be related to the lattice stresses accumulated during the reordering of lattice atoms from one crystal structure to another, based on previous ion irradiation experiments. Since the recrystallization process accelerates fission gas swelling at high burnup, it is of great importance to investigate whether the α-phase reversal promotes the initiation of recrystallization. In this study, in-situ irradiation was performed at the IVEM-Tandem facility to directly observe the microstructural evolution of decomposed γ-U region under radiation. Results from this study will not only help deepen the fundamental understanding of the defect evolution process during α-phase reversal but also provide critical information for theoretic modeling.
4:35 PM
Impact of Ionization Effects and Defect Trapping on Microstructure Evolution in Light Ion Irradiated Uranium Dioxide: Marat Khafizov1; Yuzhou Wang1; M Riyad1; Janne Pakarinen2; Lingfeng He3; Anter El-Azab4; David Hurley3; 1Ohio State University; 2Belgian Nuclear Research Center (SCK•CEN); 3Idaho National Laboratory; 4Purdue University
This study considers impact of defects produced by light ion irradiation on properties of polycrystalline uranium dioxide (UO2). UO2 samples were irradiated with 2.6 MeV hydrogen at 300°C and 3.9 MeV helium ions at 200°C to a comparable dose of 0.05 dpa. These conditions were chosen to limit mobility of point defects and effectively capture early stages of damage evolution. Lattice expansion was determined from XRD and modulated thermoreflectance method was used to measure thermal conductivity. These characterization revealed notably larger lattice expansion and thermal conductivity reduction in He samples. We analyze our results using standard point defect models for lattice expansion and thermal conductivity reduction that were informed by atomic level simulation reported in the literature. Our analysis suggests that the difference in the observed evolution of microstructure under two different irradiation conditions can be attributed to ionization effect and trapping of implanted ions.
4:55 PM
Diffusion Analysis of Metallic Fission Products in Tristructural-isotropic Coated Fuel Using Representative Diffusion Couples: Rachel Seibert1; Tyler Gerczak1; 1Oak Ridge National Laboratory
Tristructural-isotropic (TRISO) coated particle fuel is a well-established fuel considered for use in a variety of reactor concepts, most notably the high-temperature gas-cooled reactor. Release of select metallic fission products from TRISO has the potential to limit the fuel’s operational lifetime and creates potential concerns for maintenance workers. Planar diffusion couple systems consisting of representative pyrocarbon (PyC) and silicon carbide (SiC) are investigated to conduct diffusion analysis relevant to this release problem, with the intent to provide data for fuel performance models to improve fuel reliability and efficiency. The diffusion couple systems were fabricated using a fluidized-bed chemical vapor deposition technique, and were ion implanted into the PyC with either palladium and silver, silver, europium, or strontium. The influence of temperature, grain structure, and neutron irradiation on the PyC/SiC interface has been examined through electron microscopy and depth profiling techniques and will be discussed here.
5:15 PM
Microstructural and Micro-chemical Characterization of Safety Tested TRISO UCO Fuel Kernels Irradiated in the Advanced Test Reactor
: Zhenyu Fu1; Lingfeng He2; Xiang Liu2; Isabella van Rooyen2; Yong Yang1; 1University of Florida; 2Idaho National Laboratory
In this study, the AGR-2-222-RS36 and AGR-2-222-RS19 TRISO fuel particles were irradiated to a burnup of 12.55% FIMA in ATR, followed by safety testing at 1600ºC for 300 hours. Transmission electron microscopy (TEM) lamellas were prepared at the fuel kernel center, half center, recoil zone, and the fuel kernel-buffer layer interface. The Talos F200X scanning/TEM (STEM) was used to characterize the microstructural and micro-chemical evolutions in those two irradiated fuel kernels. Preliminary data analysis gives the following findings: (1) Rod-shaped precipitates rich in fission product Pd are scattered from the fuel kernel center to the edge; (2) Sub-micro and nano-sized precipitates rich in fission product Pd, Rh, and Ag are located in the buffer layer; (3) Amorphous carbon is found to diffuse into the fuel kernel up to a distance of 50 m from the kernel outer surface; and (4) Fission gas bubbles were mainly observed in the Mo-rich phase.