Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “FUEL PELLET”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Reactor fuel pellets with thermally-conductive inserts, and related reactor fuel pellet arrangements

Fuel pellets and fuel pellet arrangements include thermally-conductive inserts within a fuel. The inserts have at least one portion of a thermally-conductive material, such as radially-extending fins. The inserts are configured to dissipate heat during use of the fuel pellets, while minimizing the amount of the total volume of the fuel pellet that is occupied by non-fissile material. The inclusion of heat-dissipating inserts enables the fuel pellets to exhibit improved thermal performance over the lifetime of the fuel, including a relatively low peak temperature and relatively low integrated average temperatures, while the minimal volume of the inserts avoids significantly decreasing the percent of enrichment achievable.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Uncovering Uranium Isotopic Heterogeneity of Fuel Pellets from the Fifth Collaborative Materials Exercise of the Nuclear Forensics International Technical Working Group

In 2017, the Nuclear Forensics International Technical Working Group (ITWG) organized their fifth 37 Collaborative Materials Exercise (CMX-5). The exercise samples were two uranium dioxide fuel pellets 38 manufactured from the same starting materials by different processes to have similar bulk isotopic 39 composition, but different spatial uranium isotopic distributions. Sets of identical materials were sent to 40 all participating laboratories, who then utilized their existing nuclear forensic capabilities to 41 independently analyse fuel pellets and identify similarities and differences of the materials’ 42 characteristics. The analytical methods used to probe the fuel pellets included ex situ, such as sectioning 43 or breaking up the pellets and analyzing dissolved pieces using inductively coupled plasma mass 44 spectrometry (ICP-MS), analyzing particles collected from intact or fragmented pellets by secondary ion 45 mass spectrometry (SIMS), as well as in situ methods, such as laser ablation coupled with ICP-MS, 46 autoradiography and nanoSIMS. In this paper we present the results of these independent analyses and 47 compare the capabilities of those nuclear forensic analytical methods to uncover details of the isotopic 48 heterogeneity of uranium fuel pellets.

Nuclear Forensic Analysis of Uranium Fuel Pellets,↗
DOE OSTI2018

Grain Size and Phase Purity Characterization of U 3 Si 2 Pellet Fuel

Characterization of U 3 Si 2 fresh fuel pellets is important for quality assurance and validation of the finished product. Grain size measurement methods, phase identification methods using scanning electron microscopes equipped with energy dispersive spectroscopy and x-ray diffraction, and phase quantification methods via image analysis have been developed and implemented on U 3 Si 2 pellet samples. A wide variety of samples have been characterized including representative pellets from an initial irradiation experiment, and samples produced using optimized methods to enhance phase purity from an extended fabrication effort. The average grain size for initial pellets was between 16 and 18 µm. The typical average grain size for pellets from the extended fabrication was between 20 and 30 µm with some samples exhibiting irregular grain growth. Pellets from the latter half of extended fabrication had a bimodal grain size distribution consisting of coarsened grains (>80 µm) surrounded by the typical (20-30 µm) grain structure around the surface. Phases identified in initial uranium silicide pellets included: U 3 Si 2 as the main phase composing about 80 vol. %, Si rich phases (USi and U 5 Si 4 ) composing about 13 vol. %, and UO 2 composing about 5 vol. %. Initial batches from the extended U 3 Si 2 pellet fabrication had similar phases and phase quantities. The latter half of the extended fabrication pellet batches did not contain Si rich phases, and had between 1-5% UO 2 : achieving U 3 Si 2 phase purity between 95 vol. % and 98 vol. % U 3 Si 2 . The amount of UO 2 in sintered U 3 Si 2 pellets is correlated to the length of time between U 3 Si 2 powder fabrication and pellet formation. These measurements provide information necessary to optimize fabrication efforts and a baseline for future work on this fuel compound.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Qualitative comparison of bremsstrahlung X-rays and 800 MeV protons for tomography of urania fuel pellets

We present an assessment of x-rays and proton tomography as tools for studying the time dependence of the development of damage in fuel rods. We also show data taken with existing facilities at Los Alamos National Laboratory that support this assessment. Data on surrogate fuel rods have been taken using the 800 MeV proton radiography (pRad) facility at the Los Alamos Neutron Science Center (LANSCE), and with a 450 keV bremsstrahlung X-ray tomography facility. The proton radiography pRad facility at LANSCE can provide good position resolution (<70 {mu}m has been demonstrate, 20 {mu}m seems feasible with minor changes) for tomography on activated fuel rods. Bremsstrahlung x-rays may be able to provide better than 100 {mu}m resolution but further development of sources, collimation, and detectors is necessary for x-rays to deal with the background radiation for tomography of activated fuel rods.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

The Feasibility and the Benefits of the Advanced Nuclear Fuel Pellet Designs with Radially Varying Fuel Zoning and Burnable Poison Concentration (Final Summary Report)

As part of the work supported by the US Department of Energy (DOE) Office of Nuclear Energy (NE) Gateway for Accelerated Innovation in Nuclear (GAIN) FY 2021 Voucher, Exelon Generation, now Constellation, and Oak Ridge National Laboratory (ORNL) entered into a cooperative research and development agreement (CRADA) to evaluate and assess the feasibility and impact of various conceptual advanced nuclear fuel pellet designs (ANFPDs). The objective of this project was to perform modeling and simulation and analyses using the advanced modeling and simulation capabilities of VERA/BISON, developed by DOE, to determine the viability and benefits of numerous advanced nuclear fuel pellet design concepts in terms of fuel cycle costs, operational safety, and margin improvement. Whereas the detailed coupled neutronic and thermal hydraulic analyses performed using VERA provided in-depth knowledge in terms of fuel cycle performance, the detailed VERA results were used in subsequent fuel performance analyses using BISON. These subsequent analyses focused on several key fuel performance criteria, such as peak fuel centerline temperature (FCT), fission gas release (FGR), gap closure, plenum pressure, and cladding hoop stress. These key fuel performance criteria were analyzed for some of the conceptual fuel designs and were compared with the results obtained for UO 2 fuel. The results provided detailed information to enable better understanding of the performance of the fuel types analyzed. Understanding the advantage of loading these conceptual fuel designs into the core is important not only to Constellation but also to the entire light-water reactor (LWR) fleet in the United States.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Fracture toughness measurement of ceramics with fuel pellet geometry

The crack initiation of nuclear fuel pellets, driven by extreme temperature gradients, is a critical factor impacting reactor performance. Although this fact is widely acknowledged, there is limited fracture toughness data for specimens with representative geometry and fabricated using relevant processing methods. Here, to address this, a new chevron-notched short-rod specimen configuration with diameter and length both equal to 10 mm, representative of a conventional fuel pellet and tested in tension, was developed in an attempt to validly measure Mode I fracture toughness, K Ic . Test calibration was carried out by testing polycrystalline ZrO 2 , SiC, and Si 3 N 4 ceramics fabricated with conventional fuel pellet geometry and known fracture toughnesses. The results suggest that the method described in this paper produces valid K Ic measurements using conventional ceramic fuel pellets.

Ceramics↗
DOE OSTI2019

Microscale Isotopic Variation in Uranium Fuel Pellets with Implications for Nuclear Forensics

Until recently, the analysis and identification of nuclear fuel pellets in the context of a nuclear forensics investigation have been mainly focused on macroscopic characteristics, such as fuel pellet dimensions, uranium enrichment, and other reactor-specific features. Here, we report microscale isotopic heterogeneity observed in different fuel pellet fragments that were characterized in situ by nanoscale secondary ion mass spectrometry (NanoSIMS). The materials analyzed include fuel fragments obtained as part of the Collaborative Materials Exercise (CMX-4) organized by the Nuclear Forensics International Technical Working Group (ITWG), as well as a fuel pellet fragment from a commercial power reactor. Although the commercial fuel pellet showed a homogeneous 235 U/ 238 U ratio across the sample (within analytical error), NanoSIMS imaging of the CMX-4 fuel pellet fragments showed distinct microscale variations in the uranium isotopic composition. The average 235 U enrichments were 2.2 and 2.9% for the two samples; however, the measured 235 U/ 238 U ratios varied between 0.0081 and 0.035 (0.79–3.3 atom % 235 U) and between 0.0090 and 0.045 (0.89–4.3 atom % 235 U). The measurement of 236 U in one of the CMX-4 samples suggested the use of at least three uranium oxide powders of different isotopic compositions (“source terms”) in the production of the pellets. These variations were not detected using the conventional bulk, macroscopic techniques applied to these materials. Furthermore, our study highlights the importance of characterizing samples on the microscale for heterogeneities that would otherwise be overlooked and demonstrates the potential use of NanoSIMS in guiding further nuclear forensic analysis.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

Comparative chemical and structural analyses of two uranium dioxide fuel pellets

Here, a comparative investigation of the trace element, U and Pb isotopic compositions, and structural attributes of two uranium dioxide fuel pellets was conducted for nuclear forensic applications. One fuel pellet consists of natural UO 2 (0.71% 235 U), while the other contains enriched UO 2 (3.98% 235 U). Numerous impurities, including transition metals, lanthanides, and main group elements were detected in both fuel pellets with total trace element contents that vary between 9.77 and 11.37 μg/g. The relative abundances of the transition metal impurities in the natural UO 2 fuel pellet are linked to the source U ore and provide important insights into its provenance. X-ray fluorescence spectroscopy was utilized to investigate and compare the macro-(cm)-scale distribution of impurities within each of the fuel pellets. In addition, structural properties of the fuel pellets were examined using powder X-ray diffraction (PXRD) and Raman and infrared spectroscopy. Differences in the intensity of the T2 g U-O stretching vibrational mode were observed in the Raman spectra of the fuel pellets. Lastly, Pb isotope ratios were investigated to potentially differentiate between the natural and enriched UO 2 samples and are linked to the process history of these materials.

36 MATERIALS SCIENCE↗
DOE OSTI2012

LIBS Spectral Data for a Mixed Actinide Fuel Pellet Containing Uranium, Plutonium, Neptunium and Americium (V.1)

Laser-induced breakdown spectroscopy (LIBS) was used to analyze a mixed actinide fuel pellet containing 75% UO 2 /20% PuO 2 /3% AmO 2 /2% NpO 2 . The preliminary data shown here is the first report of LIBS analysis of a mixed actinide fuel pellet, to the authors knowledge. The LIBS spectral data was acquired in a plutonium facility at Los Alamos National Laboratory where the sample was contained within a glove box. The initial installation of the glove box was not intended for complete ultraviolet (UV), visible (VIS) and near infrared (NIR) transmission, therefore the LIBS spectrum is truncated in the UV and NIR regions due to the optical transmission of the window port and filters that were installed. The optical collection of the emission from the LIBS plasma will be optimized in the future. However, the preliminary LIBS data acquired is worth reporting due to the uniqueness of the sample and spectral data. The analysis of several actinides in the presence of each other is an important feature of this analysis since traditional methods must chemically separate uranium, plutonium, neptunium, and americium prior to analysis. Due to the historic nature of the sample fuel pellet analyzed, the provided sample composition of 75% UO 2 /20% PuO 2 /3% AmO 2 /2% NpO 2 cannot be confirm without further analytical processing. Uranium, plutonium, and americium emission lines were abundant and easily assigned while neptunium was more difficult to identify. There may be several reasons for this observation, other than knowing the exact sample composition of the fuel pellet. First, the atomic emission wavelength resources for neptunium are limited and such techniques as hollow cathode discharge lamp have different dynamics than the plasma used in LIBS which results in different emission spectra. Secondly, due to the complex sample of four actinide elements, which all have very dense electronic energy levels, there may be reactions and interactions occurring within the plasma, such as collisional energy transfer, that might be a factor in the reduction in neptunium emission lines. Neptunium has to be analyzed alone using LIBS to further understand the dynamics that may be occurring in the plasma of the mixed actinide fuel pellet sample. The LIBS data suggests that the emission spectrum for the mixed actinide fuel pellet is not simply the sum of the emission spectra of the pure samples but is dependent on the species present in the plasma and the interactions and reactions that occur within the plasma. Finally, many of the neptunium lines are in the near infrared region which is drastically reduced in intensity by the current optical setup and possibly the sensitivity of the emission detector in the spectral region. Once the optics are replaced and the optical collection system is modified and optimized, the probability of observing emission lines for neptunium might be increased significantly. The mixed actinide fuel pellet was analyzed under the experimental conditions listed in Table 1. The LIBS spectra of the fuel pellet are shown in Figures 1-49. The spectra are labeled with the observed wavelength and atomic species (both neutral (I) and ionic (II)). Table 2 is a complete list of the observed and literature based emission wavelengths. The literature wavelengths have references including NIST Atomic Spectra Database (NIST), B.A. Palmer et al. 'An Atlas of Uranium Emission Intensities in a Hollow Cathode Discharge' taken at the Kitt Peak National Observatory (KPNO), R.L. Kurucz 1995 Atomic Line Data from the Smithsonian Astrophysical Observatory (SAO), J. Blaise et al. 'The Atomic Spectrum of Plutonium' from Argonne National Laboratory (BFG), and M. Fred and F.S. Tomkins, 'Preliminary Term Analysis of Am I and Am II Spectra' (FT). The dash (-) shown under Ionic State indicates that the ionic state of the transition was not available. In the spectra, the dash (-) is replaced with a question mark (?). Peaks that are not assigned are most likely real features and not noise but cannot be confidently assigned to a transition without further investigation. Several peaks have multiple assignments due to limited resolution of the spectrometer used (20,000, $\lambda$/$\Delta$$\lambda$) and without the availability, at this point in time, of pure PuO 2 , AmO 2 , and NpO 2 to confirm the identity of the peaks. A different spectrometer was used in the plutonium facility to collect the mixed actinide fuel pellet data (Echelle 3000) than the DUO 2 , ThO 2 and uranium ore previously reported (Echelle 4000) which accounts for the slight shift in the observed wavelength of the uranium emission lines.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Machining induced fissures in relation to microstructure of sintered U3Si2 fuel pellets

As uranium silicide (U3Si2) becomes a more desired fuel replacement for uranium dioxide in light water reactors, it is import to understand the fabrication process. Moreover, it is important to understand how the as-sintered microstructure of the fuel pellet affects the final product; uranium silicide fuel pellets are currently fabricated using a powder metallurgy process. Pellet quality and integrity is a concern as any unwanted defects may cause adverse irradiation behavior. Thus, pellets need to remain relatively free of defects (surface chipping, cracking, undesired microstructure, etc.) in order to be placed into a fuel assembly. Recently, it was observed that uranium silicide fuel pellets developed surface fissures on the machined surface after a centerless grinding process. Initial observations indicate a relation between microstructure and machining induced fissures; in essence, fissures on the machined surface are only observed in pellets containing a non-uniform texture (fine grains surrounding coarse grains). Experiments have been focused on determining the cause(s) of the non-uniform texture. Once the cause was determined and grain size was reproduced, samples with a coarse-grained microstructure, along with samples containing a fine-grained microstructure were centerless ground to confirm the relation between surface fissures and microstructure. Samples were also subjected to microscopic analysis, grain size analysis, and x-ray diffraction (XRD).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Enhanced steam oxidation resistance of uranium nitride nuclear fuel pellets

Here, the steam oxidation resistance of UN and UN-(20 vol%)ZrN fuel pellets is evaluated to enhance understanding of steam corrosion mechanisms in advanced nuclear fuel materials. In situ neutron diffraction shows the modified UN fuel pellets form a (U 0.77 ,Zr 0.23 )N solid-solution and the sole crystalline oxidation product detected in bulk is (U 0.77 ,Zr 0.23 )O 2 . U 2 N 3 is not detected in significant quantities during the steam oxidation of UN or (U 0.77 ,Zr 0.23 )N and stable lattice parameters show that hydriding does not take place. Steam oxidation rates, obtained via sequential Rietveld refinement show how (U 0.77 ,Zr 0.23 )N has a higher activation energy (79 ± 1 kJmol -1 vs. 50 ± 5 kJmol -1 ), higher onset temperature (430 °C vs. 400 °C) and slower reaction rates for steam oxidation up to 616 °C, than pure UN. Throughout, both UN and (U 0.77 ,Zr 0.23 )N exhibit linear (non-protective) oxidation kinetics, signifying that degradation of the fuel pellets is caused by the evolution of gaseous products at the interface followed by oxide scale spallation. This quantitative and mechanistic understanding of material degradation enables better defined operating regimes and points towards (U,Zr)N solid solutions as a promising strategy for the design of advanced nuclear fuel materials with enhanced steam corrosion resistance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Natural isotopic compositions of titanium, iron, and nickel observed in commercial fuel pellets – Promising candidate elements for stable isotope tagging

Stable isotope taggants would constitute unique identifiers for nuclear fuel cycle materials, resulting in expedited timelines and high confidence provenance assessments for nuclear forensics investigations. However, reliably identifying and interpreting stable isotope taggants in nuclear materials recovered from outside of regulatory control will largely be predicated on the assumption that the taggant element intrinsic to the untagged nuclear material exhibits natural isotopic ratios. Here, we present high-precision Ti, Fe, and Ni isotope compositions in 13 commercial low-enriched uranium (LEU) fuel pellets to assess the suitability of these transition metals as stable isotope taggants. Our investigations reveal limited isotope variations among the fuel pellets in all three elements, which are consistent with small mass-dependent isotope fractionations, comparable to variations previously reported for natural samples. In practice, isotopically tagged nuclear materials are expected to fall along isotopic mixing lines, since intrinsic background levels of taggant elements dilute the taggant towards natural isotope compositions. Furthermore, the observation that Ti, Fe, and Ni isotope compositions in a suite of LEU fuel pellets are close to or indistinguishable from estimates for the Bulk Silicate Earth demonstrates that a two end-member mixing assumption would be valid for these transition metals, indicating that all three are promising candidate elements for stable isotope tagging. Finally, we present mass balance calculations to quantify isotopic perturbations expected from admixing isotopically anomalous Ti, Fe, and Ni taggants to assess the interplay between elemental and taggant concentrations and find favorable compromises for facilitating successful taggant identification with current analytical methods.

Intentional forensics↗
DOE OSTI2019

Thermal Transport and Fracture Behavior of Sintered Fuel Pellets: Experimental Validation of NEAMS Tool MARMOT

The project targets at experimental validation of NEAMS tool MARMOT for predicting thermal transport and fuel fracture. The key components of the project include: (1) sintering UO 2 samples with well-controlled microstructures (grain size, pore, fission bubble, porosity and pore distribution either in intra- or inter-granular pores); (2) thermal conductivity measurements of these wellcharacterized polycrystalline samples; (3) indentation testing of sintered UO 2 with various grain size, porosity and stoichiometry to obtain fracture and crack propagation mechanisms (inter or intragranular fracture); and (4) validation and uncertainty quantification of the MARMOT thermal transport and fracture models. The high quality experimental data will be used to validate the predictions of MARMOT for thermal transport and fracture, using average data and local microstructure information. In this work, UO 2 fuels with controlled microstructure (grain size, pore size/structure and fuel chemistry) were successfully prepared by high energy ball milling of the hyper-stoichiometric UO 2+x powders purchased commercially, and consolidated by spark plasma sintering. Dense nano-sized UO 2 fuel pellets with controlled stoichiometry were sintered by SPS. The effects of different SPS sintering parameters on the pore size and distribution were investigated, and the correlation among the SPS parameter – microstructure control – stoichiometry/defect structure was established. To further control the pore structure and fuel chemistry, systematic thermal annealing of the sintered fuels with the controlled pore structure was performed in a reduced. The controlled microstructure and fuel chemistry enable the investigation of the separate effects of pore structure and fuel chemistry on thermal transport behavior. The mechanical properties of the sintered UO 2 fuels were tested by nanoindentation and microindentation testing at different temperatures. The fracture behavior of the sintered fuels as functions of grain structure and temperature was investigated. Nanocrystalline UO 2 display higher hardness than microcrystalline counterpart, consistent with the Hall-Petch strengthening mechanism. Greater Young’s modulus and fracture toughness are also identified for the nanocrystalline UO 2 , and hardness and Young’s modulus decrease with temperature, suggesting better ductility of oxide fuels at greater temperature at small length scale. Hyper-stoichiometric UO 2 specimen displays higher hardness and fracture toughness than stoichiometric UO 2 , due to the impediment of the crack propagation by the oxygen interstitial atoms. These results are useful in understanding the mechanical properties of the high burn-up structure (HBS) formed in nuclear fuels during reactor operation, and also provide critical experimental data as the input for the development and validation of the MARMOT fracture model of nuclear fuels. The thermal properties of the sintered UO 2 fuel pellets with controlled grain structure, fuel chemistry and pore structures are measured by laser flash and the impact of the microstructure characteristics on the thermal transport properties is elucidated. Through a close collaboration with the MARMOT team at Idaho National Laboratory, the thermal properties of the sintered oxide fuel pellets with controlled microstructures are measured experimentally, and modeled using the developed MARMOT thermal transport models for a joint effort of demonstrating high spatial resolution measurements at the microstructural level on as-fabricated nuclear fuels and materials and obtain critical experimental data for model validation and properties predictions. In this work, we also conducted sensitivity analysis and uncertainty quantification (UQ) on a mesoscale simulation that has been applied to understand the effective thermal conductivity of UO 2 reactor fuel using the MARMOT tool. We also conduced UQ on mechanistic macroscale models that have been developed to predict the effective thermal conductivity. The most sensitive parameter is the thermal conductivity of bulk UO 2 for the thermal conductivity models. In the future, the predicted simulation distributions need to be compared to experimental data for validation of the models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Development of Innovative Accident Tolerant High Thermal Conductivity UO 2 -Diamond Composite Fuel Pellets

The University of Florida (UF) evaluated a composite fuel consisting of UO 2 powder mixed with diamond micro particles as a candidate as an accident-tolerant fuel (ATF). The research group had previous extensive experience researching with diamond micro particles as an addition to reactor coolant for improved plant thermal performance. The purpose of this research work was to utilize diamond micro particles to develop UO 2 -Diamond composite fuel pellets with significantly enhanced thermal properties, beyond that already being measured in the previous UF research projects of UO 2 – SiC and UO 2 – Carbon Nanotube fuel pins. UF is proving with the current research results that the addition of diamond micro particles to UO 2 may greatly enhanced the thermal conductivity of the UO 2 pellets producing an accident-tolerant fuel. The Beginning of life benefits have been proven and fuel samples are being irradiated in the ATR reactor to confirm that the thermal conductivity improvements are still present under irradiation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Design of a Continuous Pellet Fueling System for Wendelstein 7-X

A continuous pellet fueling system (CPFS) is currently being designed at the Oak Ridge National Laboratory (ORNL) for the long pulse operation of the Wendelstein 7-X (W7-X) stellarator. The purpose of the CPFS is to provide deep continuous fueling for feedback-controlled high-density operation and mitigation of predicted hollow density profiles. As described here, the system will provide the capability to inject cylindrical pellets of solid hydrogen or deuterium into the plasma core, with flexibility to vary the pellet size, velocity, and injection frequency. Pellets are nominally of 3 mm in diameter and have a length between 1 and 4 mm. The heart of the CPFS is a vertically oriented, twin-screw extruder, cooled by three Gifford-McMahon cryocoolers in parallel, designed to form a continuous filament of hydrogen or deuterium. The filament width, which determines the pellet length, can be adjusted by means of a variable nozzle driven by a linear actuator at the base of the extruder. Coupled to the nozzle is a solenoid-operated gas gun and cutter assembly. A pneumatic propellant valve pulses a burst of ~60 bar helium to accelerate the cut pellet into W7-X. Three gaps in the guide tubes provide pumping locations to remove the helium propellant before it reaches the plasma. The maximum velocity of the pellet is limited by its ability to survive navigating the guide tube trajectory intact. A microwave cavity located within the guide tube provides the capability to measure the pellet size and velocity. The mechanical and thermal designs of the W7-X extruder, adjustable nozzle, and gun and cutter assembly design are described. A guide tube design and experimental pellet survivability test results are presented.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Molybdenum and Tungsten isotope compositions of UO 2 fuel pellets: Implications for isotopically enriched taggants

The addition of isotopically enriched taggants to material at the front end of the nuclear fuel cycle could be a powerful tool used to assist law enforcement authorities should material outside of regulatory control be found. Two potential candidates for this purpose are molybdenum (Mo) and tungsten (W) as both elements have five or more stable isotopes and are trace elements contained within nuclear fuel. So there is a concern that Mo and W could undergo isotope fractionation during processes like uranium enrichment and to date, it is unknown if nuclear fuels have natural Mo and W isotope compositions. If Mo and W isotopic variability is present in nuclear fuels, this would hinder the use of these elements as isotopic taggants because it would be difficult to discern the original taggant isotope composition with high confidence. Therefore, a set of 16 low enriched uranium (LEU) fuel pellets from US commercial producers was analyzed using multi collector-inductively coupled plasma mass spectrometry (MC-ICPMS) to determine Mo and W isotope compositions (i.e., 94 Mo/ 92 Mo, 95 Mo/ 92 Mo, 96 Mo/ 92 Mo, 97 Mo/ 92 Mo, 98 Mo/ 92 Mo, 183 W/ 182 W, 184 W/ 182 W, and 186 W/ 182 W). Relative to terrestrial standards, LEU fuel pellets have variable Mo and W isotope compositions, thereby complicating the use of these elements as isotopically enriched taggants. As such, this work demonstrates that the isotope composition of any potential taggant must be well characterized in the base nuclear fuel prior to any taggant addition. Furthermore, these results suggest that Mo and W are not ideal candidates for isotopically enriched taggants.

58 GEOSCIENCES↗
DOE OSTI2016

Bison Modeling of SiC/SiC Cladding Including Fuel-Pellet Interaction

SiC/SiC composites are considered leading candidates for accident tolerant fuel cladding in light water reactors. However, when SiC-based materials are exposed to neutron irradiation, they experience significant changes in dimensions and physical properties. In the extreme nuclear reactor environment, SiC-based fuel cladding will see neutron damage, significant heat flux, as well as a corrosive environment. In this report, we evaluate the effect of non-uniform changes in the dimensions caused by neutron irradiation with variable temperatures, along with the closing of the fuel – clad gap, on the stress development in the clad over the course of time. To ensure reliable and safe operation of accident tolerant fuel cladding concepts such as SiC-based materials, it is important to assess its thermo-mechanical performance under in-reactor conditions of irradiation and realistic temperature distributions. In this work, we have used the BISON fuel performance modeling code to perform the thermo-mechanical analysis of SiC/SiC cladding. A constitutive model is constructed and solved numerically to predict the stress distribution and variation in the cladding under normal operating conditions. The dependence of dimensions and physical properties variation with irradiation and temperature has been incorporated. These initial results from the analyses provides insights about the stress distribution and variation with time in the cladding as well as the interaction of fuel pellet with the cladding under different conditions of initial fuel pellet-cladding gap and steady state power levels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗