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At least 1,369 records · Page 76

Correlating Nanoscale Secondary Ion Mass Spectrometry and Atom Probe Tomography Analysis of Uranium Enrichment in Metallic Nuclear Fuel

Distribution and enrichment of 235U within nuclear fuels are essential to material performance and reactor safety. With the development of metallic U fuels, understanding 235U homogeneity has become increasingly important, however, it is difficult to map with high spatial, and mass resolution to analyze both matrix and secondary phase precipitates. Here, we analyze 235U enrichment in matrix and carbide phases in LEU-Mo via two chemical imaging modalities: nanoscale secondary ion mass spectrometry (nanoSIMS) and atom probe tomography (APT). NanoSIMS and APT provided consistent results, with no statistically significant difference between measured and nominal enrichment (19.75 % 235U).

U-10Mo, atom probe tomography, nanoscale secondary↗

In-situ Electrochemical Testing of Uranium Oxide Alteration under Anoxic Conditions

Prediction of the corrosion behavior of spent uranium oxide (UO 2 ) fuel is needed for developing predictive performance assessment models for a geologic repository. Currently, the Fuel Matrix Dissolution Model (FMDM) is being used for modeling UO 2 chemistry in the Engineered Barrier System (EBS) and includes the effects of oxidants, oxygen and hydrogen peroxide generated from radiolysis of the fuel in contact with water, as well as the role of hydrogen on limiting corrosion. In this report, efforts to validate the FMDM model using in-situ liquid cells in the scanning electron microscopy (SEM) are described. The System for Analysis at the Liquid Vacuum Interface (SALVI), was invented at the Pacific Northwest National Laboratory (PNNL) and is a device that enables the investigation of solid-solution systems within the electron microscope. Initial efforts have concentrated on building the testing system for radioactive materials and exploring the use of the SALVI cell on UO 2 particles.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Mechanistic FCCI model for BISON

In this report, multi-scale simulations to develop a mechanistic model of fuel-cladding chemical interaction (FCCI) conducted under the auspices of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in FY23 are described. The model assumes that cladding wastage is dominated by the intermetallic phase (Fe,Cr)$_{17}$Nd$_2$, and calculates production and diffusion of Nd in the fuel, diffusion into the cladding, and formation and growth of the intermetallic wastage layer. Atomistic simulations are used to calculate diffusivity of Nd in the (Fe,Cr) system. Mesoscale simulations, informed by previous atomistic calculations, are used to determine the effective diffusivity of Nd through the fuel, accounting for empty and sodium-filled porosity within the fuel. The BISON model incorporates this data and calculates the wastage layer thickness as a function of time, assuming parabolic growth kinetics based on the best available data. The results of the new mechanistic BISON model are compared to existing assessment cases for the X447 experiment in EBR-II, and show good agreement with experimental data and the existing BISON empirical model. To improve predictions of kinetics for future work, a phase-field model that includes fuel, wastage, and cladding phases is developed. Preliminary testing indicates that the kinetics may differ significantly from the assumed parabolic growth law, suggesting direction for future improvement of the model that will allow it to be applied to other fuel designs with greater confidence.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Aviation fuels outlook

Options for satisfying the future demand for commercial jet fuels are analyzed. It is concluded that the most effective means to this end are to attract more refiners to the jet fuel market and encourage development of processes to convert oil shale and coal to transportation fuels. Furthermore, changing the U.S. refineries fuel specification would not significantly alter jet fuel availability.

Momenthy, A. M.↗

Fuel optimal control of an experimental multi-mode system

In this paper, the dynamic characteristics associated with the fuel optimal control of a harmonic oscillator are utilized in the development of a near fuel optimal feedback control strategy for spacecraft vibration suppression. In this scheme, single level thrust actuators are governed by recursive computations of the standard deviations of displacement and velocity at the actuator's locations. The algorithm was tested on an experimental structure possessing a significant number of flexible body modes. The structure's response to both single and multiple mode excitation is presented.

Redmond, J.↗

Assessment of coated cladding impact on large-break LOCA with TRACE-DAKOTA

Since 2011, the U.S. Department of Energy has launched new research and development (R&D) programs for developing new accident tolerant fuels (ATFs). These new fuel concepts are expected to bring several advancements in terms of plant safety. However, development, licensing, as well as deployment of ATF require a substantial investment among fuel vendors, operating utilities, and regulatory authorities. For ATF to be economically feasible, credited safety benefits may be necessary to provide an economic incentive for transition of nuclear power plants from Zr/UO2-based fuel technologies to ATFs. As a result, it is essential to evaluate the economic viability of ATF by conducting a comprehensive safety benefits study informed by R&D and demonstration studies. In the present work, based on the best estimate plus uncertainty methodology, sensitivity analysis and uncertainty quantification for key safety quantities under LB-LOCA are performed for the Peach Bottom boiling water reactor (BWR) with GE14 nuclear fuel design adopting the ATF concepts. A coupled TRACE-DAKOTA analysis framework is adopted to perform the uncertainty quantification. Based on the computational results and obtained statistics, the coping time increase, fuel performance improvement as well as the most dominating phenomena during the accident for Cr-coated cladding are investigated in detail. Overall, the results obtained are useful for subsequent plant PRA study and support a move to more modernized performance-based regulatory requirements to realize plant economic benefits.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Plutonium: Advancing our Understanding to Support Sustainable Nuclear Fuel Cycles

With Global energy needs increasing, real energy solutions to meet demands now, are needed. Fossil fuels are not an ideal candidate to meet these needs because of their negative impact on the environment. Renewables such as wind and solar have huge potential, but still need major technological advancements (particularly in the area of battery storage) before they can effectively meet growing world needs. The best option for meeting large energy needs without a large carbon footprint is nuclear energy. Of course, nuclear energy can face a fair amount of opposition and concern. However, through modern engineering and science many of these concerns can now be addressed. Many safety concerns can be met by engineering advancements, but perhaps the biggest area of concern is what to do with the used nuclear fuel after it is removed from the reactor. Currently the United States (and several other countries) utilize an open fuel cycle, meaning fuel is only used once and then discarded. It should be noted that fuel coming out of a reactor has utilized approximately 1% of the total energy that could be produced by the uranium in the fuel rod. The answer here is to close the fuel cycle and recycle the nuclear materials. By reprocessing used nuclear fuel, all the U can be repurposed without requiring disposal. The various fission products can be removed and either discarded (hugely reduced waste volume) or more reasonably, utilized in specialty reactors to make more energy or needed research/medical isotopes. While reprocessing technology is currently advanced enough to meet energy needs, completing research to improve and better understand these techniques is still needed. Better understanding behavior of fission products is one area of important research. Despite it being discovered over 75 years ago, plutonium is still an exciting element to study because of the complex solution chemistry it exhibits. In aqueous solutions Pu can exist simultaneously in multiple oxidation states, including 3+, 4+, and 6+. It also readily forms a variety of metal-ligand complexes depending on solution pH and available ligands. Understanding of the behavior of Pu in solution remains an important area of research today, with relevance to developing sustainable nuclear fuel cycles, minimizing its impact on the environment, and detecting and preventing the spread of nuclear weapons technology.

Lines, Amanda M.↗

Preparation of Diffusion Couples for Irradiation and High Temperature Testing of Representative TRISO PyC/SiC

Understanding the release of fission products from intact tristructural-isotropic (TRISO) coated particle fuel during normal and off-normal service is necessary to ensuring safe and efficient reactor operation. Planar diffusion couples have been fabricated to determine the diffusion kinetics of select fission product species (Ag, Ag+Pd, Eu, and Sr) in pyrocarbon (PyC) and silicon carbide (SiC) materials which are representative of TRISO fuel. Obtaining accurate diffusion coefficients for these systems in representative materials will allow for higher fidelity inputs for fuel performance models. Three different diffusion couple variants have been produced which include a Baseline, PyC, and SiC variant. The Baseline variant serves as an example of representative PyC and SiC layers as the layer properties mimic those found in TRISO fuel from the second Advanced Gas Reactor Fuel Qualification and Development Program. The PyC and SiC variants include variation in the respective layer properties to understand the impact of PyC density and SiC grain size on fission product diffusion. The diffusion couples were designed to allow for exploration of neutron irradiation on diffusion reflecting in reactor behavior as well as high temperature thermal diffusion reflecting diffusive transport during off-normal conditions. The methodology for selecting the diffusion couple design, refinement of the individual variant layer properties, and the complete process for diffusion couple construction are summarized in this report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

U-Mo Monolithic Fuel for Nuclear Research and Test Reactors

The metallic fuel selected to replace the current HEU fuels in the research and test reactors is the LEU-10 weight % Mo alloy in the form of a thin sheet or foil encapsulated in AA6061 aluminum alloy with a zirconium interlayer. In order to effectively lead this pursuit, new developments in processing and fabrication of the fuel elements have been initiated, along with a better understanding of material behavior before and after irradiation as a result of these new developments. This editorial note gives an introduction about research and test reactors, need for HEU to LEU conversion, fuel requirements, high uranium density monolithic fuel development and an overview of the four articles published in the December 2017 issue of JOM under a special topic titled “U-Mo Monolithic Fuel for Nuclear Research and Test Reactors”.

LEU fuel development↗

Solar Thermochemical Fuel Production via a Novel Low Pressure, Magnetically Stabilized, Non-volatile Iron Oxide Looping Process (University of Florida)

The UF Solar Fuels team has developed a modular reactor at the 10kW scale for the thermochemical conversion of H 2 O and CO 2 to H 2 and CO, respectively, using concentrated solar energy to drive high-temperature redox reactions. Various materials, including ferrites, doped and mixed metal ferrites, and ceria, were investigated and reported on. Ceria was chosen as the reactive material, with thermogravimetric and bench-scale analysis concurring with literature review and pointing towards reasonable kinetics and stability at temperatures on the order of 1500°C. A combined radiation, conduction, and species reaction/mass transport model was developed, utilizing Monte Carlo ray tracing, Lattice Boltzmann, and random walk particle tracking for the respective components to direct development and optimization of the reactor. With experimental data used for reaction rates, the model showed a path to efficiencies >20%, which could be market competitive with current PV-hydrolysis hydrogen generation systems. Economic analysis of a larger-scale plant (100kW modular system with cost performed at 500 modules) has been performed, assuming 0.5% and 10% internal rate of return. Without compression costs, H 2 production cost using this is technology at 12 USD/kg H 2 and 17 USD/kg H 2 for realized efficiencies of 20% and 15%, respectively.

14 SOLAR ENERGY↗

The NASA MERIT program - Developing new concepts for accurate flight planning

It is noted that the rising cost of aviation fuel has necessitated the development of a new approach to upper air forecasting for flight planning. It is shown that the spatial resolution of the present weather forecast models used in fully automated computer flight planning is an important accuracy-limiting factor, and it is proposed that man be put back into the system, although not in the way he has been used in the past. A new approach is proposed which uses the application of man-computer interactive display techniques to upper air forecasting to retain the fine scale features of the atmosphere inherent in the present data base in order to provide a more accurate and cost effective flight plan. It is pointed out that, as a result of NASA research, the hardware required for this approach already exists.

Steinberg, R.↗

Development of an In Situ Fission Gas Release Instrument for Fuel Sample Irradiations in the High Flux Isotope Reactor

Experimental measurement of gaseous fission product release with respect to temperature and burnup is a critical aspect of understanding nuclear fuel performance, validating predictive models, and qualifying new fuels. To measure this phenomenon in real-time, Oak Ridge National Laboratory has developed an instrument for measuring in situ fission gas release from small-scale fuel samples irradiated in the High Flux Isotope Reactor (HFIR). The instrument uses a continuous flow of Heover the fuel samples to sweep gaseous fission products from a sealed capsule in the HFIR Be reflector to an instrument station adjacent to the reactor. The instrument station houses two high-purity germanium (HPGe) detectors that measure decay gamma rays from fission products passing through a room temperature dwell chamber placed over the detector crystal. The sealed capsules in the reactor are designed to modulate fuel sample temperatures between 700 and 1,100°C by changing the Ar/He gas mixture surrounding the capsules during irradiation. N-type thermocouples are incorporated into the capsule housing to record real-time fuel temperatures. The capsules are heated primarily by prompt gamma rays emitted from the HFIR core with minimal heat contributions from fission in the fuel samples to minimize temperature gradients in the specimens for separate-effects characterization of the material. This paper describes modeling of time-dependent nuclear heating and fission product formation in fuel samples, thermal characteristics of the in-core capsules, and expected gaseous fission product gamma spectra at the HPGe instrument station.

Mulligan, Padhraic L [ORNL] (ORCID:000000025826540↗

Operation of a 50-kW th chemical looping combustion test facility under autothermal conditions

The United States Department of Energy is developing transformational fossil fuel combustion technologies that result in more efficient power cycles and/or substantial reductions in GHG emissions. Chemical Looping Combustion (CLC) has the potential to achieve a lower cost of electricity than a supercritical pulverized coal power plant with an amine absorption system. Although the concept and research on chemical looping has a long history, CLC has numerous technical issues that must be addressed prior to a commercial debut. This effort is directed toward the following major issues for CLC: (1) demonstrating continuous solids handling, operability, and reactor performance at autothermal conditions and (2) preliminary assessments of oxygen carrier attrition and relative material make-up costs. These are critical issues that should be addressed under realistic conditions prior to directing resources toward larger scale demonstrations. This paper will also describe NETL’s small 50-kWth natural gas circulating CLC test facility located in Morgantown, WV. Despite its small size, this experimental test unit has operated at autothermal conditions for 11 h using a copper-iron-alumina oxygen carrier developed at NETL. Here, this paper will describe the test facility and operating experiences using NETL’s “Gen 2.0” oxygen carrier material. Natural gas conversion to CO2 ranged from 70 to 90% over approximately 75 oxidation-reduction cycles during the autothermal operating period. Estimates for the process specific attrition rate during autothermal operation are also presented. Finally, estimates for the oxygen carrier makeup cost, specific to the NETL process configuration, are several orders of magnitude greater than the NETL target. Future work should focus on reducing the oxygen carrier material cost and improving the attrition resistance of the oxygen carrier material under autothermal test conditions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

GDSA Framework Development and Process Model Integration FY2021

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Spent Fuel & Waste Disposition (SFWD) is conducting research and development (R&D) on geologic disposal of spent nuclear fuel (SNF) and highlevel nuclear waste (HLW). A high priority for SFWST disposal R&D is disposal system modeling (DOE 2012, Table 6; Sevougian et al. 2019). The SFWST Geologic Disposal Safety Assessment (GDSA) work package is charged with developing a disposal system modeling and analysis capability for evaluating generic disposal system performance for nuclear waste in geologic media.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

High-Temperature Desulfurization of Heavy Fuel-Derived Reformate Gas Streams for SOFC Applications

Desulfurization of the hot reformate gas produced by catalytic partial oxidation or autothermal reforming of heavy fuels, such as JP-8 and jet fuels, is required prior to using the gas in a solid oxide fuel cell (SOFC). Development of suitable sorbent materials involves the identification of sorbents with favorable sulfidation equilibria, good kinetics, and high structural stability and regenerability at the SOFC operating temperatures (650 to 800 C). Over the last two decades, a major barrier to the development of regenerable desulfurization sorbents has been the gradual loss of sorbent performance in cyclic sulfidation and regeneration at such high temperatures. Mixed oxide compositions based on ceria were examined in this work as regenerable sorbents in simulated reformate gas mixtures and temperatures greater than 650 C. Regeneration was carried out with dilute oxygen streams. We have shown that under oxidative regeneration conditions, high regeneration space velocities (greater than 80,000 h(sup -1)) can be used to suppress sulfate formation and shorten the total time required for sorbent regeneration. A major finding of this work is that the surface of ceria and lanthanan sorbents can be sulfided and regenerated completely, independent of the underlying bulk sorbent. This is due to reversible adsorption of H2S on the surface of these sorbents even at temperatures as high as 800 C. La-rich cerium oxide formulations are excellent for application to regenerative H2S removal from reformate gas streams at 650 to 800 C. These results create new opportunities for compact sorber/regenerator reactor designs to meet the requirements of solid oxide fuel cell systems at any scale.

Flytzani-Stephanopoulos, Maria↗

Perovskite-Derived Cs 2 SnCl 6 –Silica Composites as Advanced Waste Forms for Chloride Salt Wastes

Advanced materials and processes are required to separate halides and fission products from complex salt waste streams associated with the chemical reprocessing of used nuclear fuels and molten salt reactor technologies for immobilization into chemically durable waste forms. Here, in this work, we explore an innovative concept using metal-halide perovskites as advanced host phases to incorporate Cs and Cl with very high waste loadings. Wet chemistry-synthesized Cs 2 SnCl 6 powders from CsCl salt solutions are successfully encapsulated into a silica matrix to form a composite using low-temperature spark plasma sintering with tunable Cs and Cl loadings up to 31 wt.% and 26 wt.%, respectively. Chemical durability testing of the composite waste forms by semi-dynamic leaching experiments demonstrates that incongruent leaching mechanism dominated. The metal-halide perovskite-silica composite waste forms display exceptional chemical durability with the long-term release rates of Cs and Cl comparable to or outperforming the state-of-the-art waste form materials but with significantly higher waste loadings. The scalable synthesis of the metal-halide perovskite from wet-chemistry processes opens up new opportunities in designing perovskite-glass composite waste forms for salt wastes with very high waste loadings and exceptional chemical durability for the sustainable development of advanced fuel cycles and next-generation reactor technologies.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

GDSA Framework Development and Process Model Integration FY2022

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Spent Fuel & Waste Disposition (SFWD) is conducting research and development (R&D) on geologic disposal of spent nuclear fuel (SNF) and high-level nuclear waste (HLW). A high priority for SFWST disposal R&D is disposal system modeling (Sassani et al. 2021). The SFWST Geologic Disposal Safety Assessment (GDSA) work package is charged with developing a disposal system modeling and analysis capability for evaluating generic disposal system performance for nuclear waste in geologic media. This report describes fiscal year (FY) 2022 advances of the Geologic Disposal Safety Assessment (GDSA) performance assessment (PA) development groups of the SFWST Campaign. The common mission of these groups is to develop a geologic disposal system modeling capability for nuclear waste that can be used to assess probabilistically the performance of generic disposal options and generic sites. The modeling capability under development is called GDSA Framework (pa.sandia.gov). GDSA Framework is a coordinated set of codes and databases designed for probabilistically simulating the release and transport of disposed radionuclides from a repository to the biosphere for post-closure performance assessment. Primary components of GDSA Framework include PFLOTRAN to simulate the major features, events, and processes (FEPs) over time, Dakota to propagate uncertainty and analyze sensitivities, meshing codes to define the domain, and various other software for rendering properties, processing data, and visualizing results.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

GDSA Framework Development and Process Model Integration FY2023

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Spent Fuel & Waste Disposition (SFWD) is conducting research and development (R&D) on geologic disposal of spent nuclear fuel (SNF) and high-level nuclear waste (HLW). A high priority for SFWST disposal R&D is disposal system modeling (Sassani et al. 2021). The SFWST Geologic Disposal Safety Assessment (GDSA) work package is charged with developing a disposal system modeling and analysis capability for evaluating generic disposal system performance for nuclear waste in geologic media.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗