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DOE OSTI2022

Multimodal analysis and characterization of the boehmite layer formed on AA6061 before and after alkaline etching

Low-enriched uranium (LEU) alloyed with 10% Mo (U-10Mo) is being considered as a promising alternative to oxide-based dispersion fuel with high-enriched uranium for use in research reactors. The configuration of this proposed LEU monolithic LEU fuel plate consists of a U-10Mo plate-type fuel foil with a 25 µm zirconium interlayer barrier clad with an aluminum alloy (AA6061). In certain research reactors, the clad AA6061 is coated with a boehmite layer to prevent corrosion. The boehmite layer has a high-pH passivation range, which makes it resistant to oxidation. Boehmite is usually formed on the AA6061 surface by autoclave processing. Before the boehmite layer is added, the surface of the AA6061 is cleaned using techniques such as polishing and wet etching. In this study, we use multimodal analysis to examine how pretreatment of AA6061 using polishing followed by alkaline etching affects the chemical composition of the boehmite layer. X-ray photoelectron microscopy (XPS), transmission electron microscopy (TEM), and x-ray diffraction (XRD) were used to study the chemical changes in the boehmite layer caused by alkaline etching pretreatment. XPS provides quantitative analysis for the Al:O ratio as well as oxidation states present on the surface, which suggests slight oxidation of the boehmite surface after alkaline etching of the AA6061 surface. We further explored this suggested oxidation of the boehmite surface using high-resolution transmission electron microscopy with selected area electron diffraction (SAED) and grazing incidence x-ray diffraction (GI-XRD), which suggested only a small amount of aluminum oxide at the surface. The multimodal analysis and imaging yielded new insights for optimizing boehmite growth on AA6061 for research reactors.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Thermal conductivity measurement of the interaction layer between UMo and Al produced by high-energy heavy ion irradiation

In this work, we report the first direct thermal conductivity measurement results for Al-UMo interaction layer (IL), which is typically observed in UMo/Al dispersion fuel plates under irradiation. The investigated IL was formed by irradiating Al coated UMo substrate using 80 MeV iodine ions at 180 degrees C up to 3.03 x 10 17 ions/cm 2 fluence. Microstructural characterization indicated that the induced IL is amorphous with an approximately (U 0.8 ,Mo 0.2 )Al 5.3 stoichiometry, which is similar to that formed under in-pile irradiation. Focused ion beam (FIB) was used to prepare nine specimens of various lengths from the IL that could be suspended across a microfabricated device for thermal conductivity measurement. The measured thermal conductivity values of the IL were significantly lower than the values for both the original UMo fuel and the Al. The successful measurement of the Al-UMo IL provides valuable information for the development and qualification of UMo/Al dispersion fuels for research and test reactor conversion applications, and further demonstrates the promising capabilities of utilizing the suspended bridge method in nuclear fuel research.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

Mapping the Microhardness of Al Matrix in U-7Mo/Al Dispersion Fuels at Medium and High Burn-up by in situ Nanoindentation

The mechanical degradation of the Al matrix in U-7Mo dispersion fuels was detected after irradiation in previous studies, but more details of the degradation mechanism are not clear yet to clarify the root cause. In this study, microhardness mapping in fresh and irradiated U-7Mo dispersion fuels coated with ZrN, with burn-up of 3.35 × 10 21 and 6.28 × 10 21 f/cm 3 , are measured to reveal the hardening of the Al matrix as a function of distance to fuel particles. Micro cubes were pulled by plasma focused ion beam scanning electron microscopy and tested by in situ nanoindenter with Berkovitch tip. Size effects are measured in medium and high burn-up samples. The higher burn-up specimen exhibits a more pronounced size effect when indent load is lower than 20 mN. Size effect correction models are also calculated by plots fitting, which predicts the true microhardness without size effect. Results shows that the hardness of the Al matrix near the ZrN coating has the highest value, then decreases moving away from the coating and become stable when the distance reaches ∼10 µm. The hardness starts increasing again getting closer to the next fuel particle, which makes the hardness distribution between two fuel particles to be “U” shape.

36 - MATERIALS SCIENCE↗

DOE OSTI2023

Irradiation performance of a U-7Mo in Al-Si matrix dispersion full-size fuel plate assembly

The Korea Atomic Energy Research Institute (KAERI) is leading the Ki-Jang Research Reactor (KJRR) project with the intent to develop a new reactor for medical isotope production and other nuclear research purposes. The KJRR core is designed to use high density fuel system where uranium alloyed with 7 wt% molybdenum (U-7Mo) particles are dispersed in a matrix of aluminum alloyed with 5 wt% silicon (Al-5Si) and clad in aluminum alloy 6061 (Al-6061) to form fuel plates. KAERI developed a fabrication facility to construct KJRR fuel assemblies and partnered with the Idaho National Laboratory (INL) to irradiate a full-size fuel assembly, with 21 total fuel plates, in the Advanced Test Reactor (ATR). Irradiation testing and subsequent Post Irradiation Exam (PIE) campaigns were performed successfully over a multi-year project. Monte Carlo neutronic calculations, coupling with a depletion code, were performed based on ATR’s as-run power history which showed that the highest power plate (plate 20) reached 83.1 % end-of-life (EOL) local burnup based on initial 235 U content. Additionally, finite element thermal modeling was performed based as-run power history which showed a beginning-of-life (BOL) peak local heat flux of 184 W/cm 2 . No anomalous fuel performance was observed during the irradiation and target test conditions were achieved. PIE showed favorable performance of the fuel assembly regarding all important phenomena. This paper describes the KJRR fuel assembly irradiation conditions and PIE data to support the conclusion that it performed well, without evidence of unexpected or problematic fuel performance, within an irradiation test designed to bound the KJRR design environment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Reactivity, Formation, and Solubility of Polyoxometalates Probed by Calorimetry

Room temperature calorimetry methods were developed to describe the energy landscapes of six polyoxometalates (POMs), Li–U 24 , Li–U 28 , K–U 28 , Li/K–U 60 , Mo 132 , and Mo 154 , in terms of three components: enthalpy of dissolution (ΔH diss ), enthalpy of formation of aqueous POMs (ΔH f,(aq) ), and enthalpy of formation of POM crystals (ΔH f,(c) ). ΔH diss is controlled by a combination of cation solvation enthalpy and the favorability of cation interactions with binding sites on the POM. In the case of the four uranyl peroxide POMs studied, clusters with hydroxide bridges have lower ΔH f,(aq) and are more stable than those containing only peroxide bridges. Here, in general for POMs, the combination of calorimetric results and synthetic observations suggest that spherical topologies may be more stable than wheel-like clusters, and ΔH f,(aq) can be accurately estimated using only ΔH f,(c) values owing to the dominance of the clusters in determining the energetics of POM crystals.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2020

AFIP6-MkII and RERTR-12 Porosity Data Collection and Analysis for Modeling and Simulation

Gathering data for the improvement of nuclear fuel modeling and simulation efforts is the primary driver for this work. Mechanistic models allow for a better understanding of the material on a micro- and macrostructural level while saving time and money over traditional experiment efforts. Historically, summarized data and correlations are the inputs for empirical material models and model validation. When improving these models for nuclear fuels with experimental results, there is a lack of reliable data readily available. Experiments - RERTR-12 and AFIP6-MkII - were conducted to understand the irradiation behavior of metallic U-10Mo monolithic fuels for use in extreme reactor environments such as research reactors like the Advanced Test Reactor (ATR) or the High Flux Isotope Reactor (HFIR). Microstructural characteristics of fission gas pores (FGP) in each experiment are collected using an automated image analysis technique developed at the University of Florida and presented here. A series of statistical tests are performed to explore the reliability of the results, as well as understand where the data is lacking and what future data collection is necessary to provide sufficient information to assist modeling efforts. The focus is on the porosity, pore size, and eccentricity of FGPs formed during irradiation in three AFIP6-MkII samples and one RERTR-12 sample. From the analysis, it is clear there are substantial impacts of fission density on the pore structure, but there also exist also underlying connections between each sample and the behavior observed in the pores. Further analyses of the pre- and post-irradiation microstructure are needed to improve the understanding of these connections. An early method for microstructural data analysis is presented within and is currently being expanded to include other microstructure data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

β- and γ-Counting for Pre-detonation Nuclear Forensics on Eu-155

Post-detonation nuclear forensics was performed at Los Alamos National Laboratory (LANL) on 155Eu, a fission product on the wing of the fission product production curve whose yield is sensitive to fission fuel and neutron energy. With a half-life of 4.753 years, 155Eu provides a longer-lived option for these measurements than other fission products with similar mass numbers. The Chemistry Division Group - Nuclear and Radiochemistry, at LANL routinely measures a suite of fission products from 235U fissions in thermal neutron flux experiments known as thermal calibration exercises, using a mixture of gas proportional β-decay counting and γ-spectrometry on HPGe detectors. The fission products of interest are reported relative to a high-yield reference fission product from the same sample to create a running average ratio specific to neutron energy and fuel type; Equation 1 below shows the ratio-of-ratios R-value measured fission products are reported in: R i x = [A( i X)/A( 99 Mo)] unknown [A( i X)/A( 99 Mo)] 235 U n,th where A denotes activity, i X is the nuclide of interest, and 235 U n,th denotes irradiations of 235 U with thermal neutrons. Measurements of unknown fission spectra are ratioed to a running-average of thermal calibration results, and the resulting R-value can be referred to as against a library of irradiation conditions.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2022

Alternate Methods for Cleaning Zirconium Plate

The United States High Performance Research Reactor Conversion Program identified the need to transition from highly enriched uranium fuel to lowly enriched uranium fuel for high power research and civil reactors. One of the fuel configurations selected for these reactors was a uranium (U) foil alloyed with 10% molybdenum (Mo). The U-10Mo fuel foil was then covered with a zirconium (Zr) interlayer and pressed into an aluminum cladding. Idaho National Laboratory prescribed a need to remove surface oxidation on the Zr prior to bonding it with U-10Mo fuel foil. To determine the best method of oxidation removal, a variety of polishing methods, including hand polishing with a diamond paste or Scotch-Brite™ pads, mechanically wet-polishing with abrasive belts, and chemical etching with hydrofluoric/nitric acid mixtures, were investigated on pre-polished Zircaloy 702 sheets purchased directly from the manufacturer. No change in surface roughness was detected using any of these methods relative to the as-received material. In fact, scanning electron microscopy detected grit from hand polishing with Amplex Grade 30 waste soluble diamond paste and Scotch-Brite™ pads. Visible staining was found, and fluorine detected via X-ray photoelectron spectroscopy on the flash etching samples. These contaminants were not seen on the as-received material. In addition to the need to remove the oxide layer on the Zr prior to co-rolling, there are concerns that the Neolube debonding agent on the co-rolling can may contaminate the Zr surface after co-rolling. The experimental results found that zirconium in contact with Neolube coated steel plates was easily cleaning with ethanol wiping after heat treated and no carbide inclusions remained on the surface. Overall, all cleaning methods tested did not significantly change the surface roughness of the plates or reduce the oxide layer present on the zirconium. Carbon was found on all the samples, before and after cleaning, except for the flash etch samples which detected fluorine instead. The insignificant changes seen in the cleaned plates relative to the as received plates question the need for cleaning prior to rolling. Further investigation in the requirements for cleaning at this step is recommended. Current methods for cleaning the surface after hot rolling with alcohol appear sufficient. After hot rolling, if Neolube does peel off the can and onto the zirconium coated foil surface, a dry abrasive or diamond paste polish is not recommended due to the chance of abrasive being embedded into the surface. However, the ability of an etchant to lift the lubricant off the zirconium surface indicates that a spot etch method could work to clean specific spots where Neolube is present, if the need arises.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Model Validation and Uncertainty Quantification on the KRUSTY Microreactor Design Using GRIFFIN Neutron Transport Code [Poster]

Argonne National Laboratory (ANL) and INL have developed a GRIFFIN steady state neutronics model for the multiphysics simulations of the Kilopower Reactor Using Sterling TechnologY (KRUSTY) microreactor in the Multiphysics Object Oriented Simulation Environment (MOOSE). The reliability of such deterministic neutronics models can be validated by comparing with computations from Monte Carlo codes (e.g. MCNP, SERPENT, OpenMC, Shift, etc). Furthermore, potential modeling/design improvements can be identified by incorporating uncertainty quantification (UQ), which can be performed by MOOSE’s Stochastic Tools Module (STM). KRUSTY is a prototype for a 5-kW thermal nuclear-powered space reactor. Its primary components consist of nuclear fuel, heat pipes, a control rod, a reflector, and the shielding. The fuel consists of 3 stacked U-7.65Mo cylinders with a hole in the center for the control rod. 8 liquid sodium heat pipes transfer fission energy from the solid fuel block to the Sterling power conversion system where the energy is extracted, and the cooled sodium flows back to the core via capillary action . The movable Boron Carbide control rod regulates the neutron population during startup or when a reactor temperature boost is needed . The beryllium oxide reflector is in 3 places in the reactor; it surrounds the core axially, it lies beneath the core on a platen, and it is present in the shim. The axial and lower reflectors rest on an adjustable stainless-steel platen that moves upward to cover the fuel and help the reactor reach criticality. Lastly, radial stainless steel surrounds the core offering protection from radiation exposure .

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Fuel Design and Fabrication: Research Reactor Fuel

The objective of this chapter is to inform the reader on the essential aspects of design and fabrication of nuclear fuel systems utilized in research and test reactors. Fuel attributes of interest include mechanical, physical, and thermal properties as well as overall manufacturability. The chapter focuses primarily on plate-type fuels utilized in the highest power research reactors where design and function are critical to operational safety and successfully meeting each reactor’s unique mission.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Recent Progress in U-10Mo Mechanical and Thermophysical Property Characterization

Results are presented on the mechanical and thermophysical property characterization of U-10Mo as part of the MP-1 qualification campaign. The thermal diffusivity, specific heat and thermal conductivity of as fabricated U-10Mo fuel foils and plates were measured using the thermal conductivity microscope (TCM), differential scanning calorimeter (DSC) and laser flash analyzer (LFA) all of which located in a shielded glove box – thermal property cell (TPC). A novel inverse method based on finite element analysis was developed to evaluate the thermal diffusivity of composite/layered materials via the LFA technique such as the 5-lyered U-10Mo monolithic fuel mini-plates.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Recent Progress in U-10Mo Mechanical and Thermophysical Property Characterization -- Poster

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1993

Partitioning of Nb, Mo, Ba, Ce, Pb, Th and U between immiscible carbonate and silicate liquids: Evaluating the effects of P2O5,F, and carbonate composition

Previously we have reported carbonate liq./silicate liq. partition coefficients (D) for a standard suite of trace elements (Nb, Mo, Ba, Ce, Pb, Th, and U) and Ra and Pa as well. In brief, we have found that immiscible liquid partitioning is a strong function of temperature. As the critical temperature of the carbonate-silicate solvus is approached, all partition coefficients approach unity. Additionally, for the overwhelming majority of the partitioning elements, InD is a linear function of 'ionic field strength,' z/r, where z is the charge of the partitioned cation and r is its ionic radius.

Jones, J. H.↗

DOE OSTI2026

The porosity surrounding carbides and second phase stringers in monolithic U-10Mo fuel plate after irradiation

Post-irradiation microstructure characterization plays an important role in qualifying the low-enriched uranium (LEU) monolithic U-10 wt%Mo plate-type fuel for United States high-performance research reactors (USHPRRs) program. Inhomogeneous features resulting from manufacturing and irradiation processes, including carbides, second phase stringers, and extensive void spaces caused by the combining of small porosities, may increase the risk of heat concentration in local regions of the fuel plate over the operating conditions. In this study, characteristics of carbides, stringers, and porosity after multiple levels of irradiation at varying fission densities were studied by electron microscopes to decipher the morphology of pores and the porosity evolution in U-10 wt%Mo. For carbides, the result shows that porosities start forming on UMo grain boundaries, then on UMo/carbides interfaces as the burn-up going higher. However, the porosities surrounding carbides grow larger than the ones on UMo grain boundaries. The porosities around the uranium carbides could interconnect to form larger void space. The study revealed that the void spaces larger than 5 µm were found around uranium carbides after high burnup, while no evidence was observed to support the similar voids formed near second phase stringers even though the size of the stringers (> 50 µm) was much larger than uranium carbides (< 20 µm). The evolution of porosities suggests that the formation of second phase stringers may not create more significant porosities compared to regular uranium carbides regions during fuel operating conditions.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Impact of chemical ordering on thermodynamic properties of point defects and Xe substitutional in U-10Mo

The accurate knowledge of defect energetics is critical to understanding the aging and irradiation behavior of U-10Mo nuclear fuel, which is selected as the fuel type for conversion of the United States High-Performance Research Reactors (HPRRs). Furthermore, using hybrid molecular dynamics and Monte Carlo (MDMC) simulation, we studied the impact of chemical ordering on the formation energies of vacancies, interstitials, and the solution energy of the Xe substitutionals. Instead of forming a random solid solution (RSS), substantial short-range-order (SRO) develops in U-10Mo, particularly at low temperatures. Mo atoms are found to repel each other and prefer U-rich local atomic environments within the 1st nearest neighbor (1NN) cutoff. Compared to the case of a RSS, the state with equilibrated Mo ordering shifts the distributions of vacancy and interstitial formation energies due to the dependence of defect energies on the local atomic environment, without a clear effect on Xe solution energy. In the operation temperature range (100–250 °C) of U-Mo fuels, neglecting SRO can lead to an inaccurate estimate of thermal equilibrium point defect concentrations by over an order of magnitude and incorrectly predict the preference among different types of dumbbells, highlighting the critical importance of accounting for the impact of chemical ordering for accurate atomistic calculations of defect properties.

Molecular dynamics and Monte Carlo (MDMC)↗

DOE OSTI2022

SCALE modeling of foil irradiations at WSU’s TRIGA with sensitivity/uncertainty analysis

We report the Washington State University TRIGA reactor was modeled in the SCALE system’s KENO neutron transport code. The model includes 119 fuel rods, a central boral control rod, 4 control blades, 5 irradiation ports, and 20 graphite reflector blocks within a rectangular aluminum grid box. The model accurately simulated the irradiation of depleted uranium and highly enriched uranium foils and predicted reaction rate ratios for 238 U(n,2n) 237 U / 238 U(n,γ) 239 U and 99 Mo / fission, which compared favorably to empirical values previously measured at Los Alamos. Finally, a sensitivity/uncertainty analysis was performed using the continuous-energy TSUNAMI-3D module of SCALE.

237Np↗

DOE OSTI2020

A Potts Model parameter study of particle size, Monte Carlo temperature, and “Particle-Assisted Abnormal Grain Growth”

A Potts Model was proposed to account for temperature and particle heterogeneity-dependent Zener Pinning. This was accomplished by relating the random fluctuations of grain boundary position allowed by the Potts Model switching probability at finite simulation temperature to experimentally observed grain growth stagnation behavior. We assume that these fluctuations arise from random fluctuations in the thermodynamic and kinetic properties of the grain boundaries and/or interfaces as they change with temperature. As an application of this model, the grain growth kinetics of U-10 wt. % Mo nuclear fuels were simulated with the input of microstructural images during different heat treatment processes. Simulated average grain growth behavior is in good agreement with experiments.

Frazier, William E.↗

DOE OSTI2022

Reexamination of 2.5-Ga “whiff” of oxygen interval points to anoxic ocean before GOE

Transient appearances of oxygen have been inferred before the Great Oxygenation Event (GOE) [~2.3 billion years (Ga) ago] based on redox-sensitive elements such as Mo and S—most prominently from the ~2.5-Ga Mount McRae Shale in Western Australia. We present new spatially resolved data including synchrotron-based x-ray spectroscopy and secondary ion mass spectrometry to characterize the petrogenesis of the Mount McRae Shale. Sediments were primarily composed of organic matter and volcanic ash (a potential source of Mo), with U-Pb ages revealing extremely low sedimentation rates. Catagenesis created bedding-parallel microfractures, which subsequently acted as fluid pathways for metasomatic alteration and recent oxidative weathering. Our collective observations suggest that the bulk chemical datasets pointing toward a “whiff” of oxygen developed during postdepositional events. Nonzero Δ 33 S in trace-metal–poor, early diagenetic pyrite and the unusually enriched organic carbon at low sedimentation rates instead suggest that environmental oxygen levels were negligible ~150 million years before the GOE.

59 BASIC BIOLOGICAL SCIENCES↗