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All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

A machine learning approach to thermal conductivity modeling: A case study on irradiated uranium-molybdenum nuclear fuels

A deep neural network was developed for the purpose of predicting thermal conductivity with a case study performed on irradiated nuclear fuel. Traditional thermal conductivity modeling approaches rely on existing theoretical frameworks that describe known, relevant phenomena that govern the microstructural evolution processes during neutron irradiation (such as recrystallization, and pore size, distribution and morphology). Current empirical modeling approaches, however, do not represent all irradiation test data well. We develop a machine learning approach to thermal conductivity modeling that does not require a priori knowledge of a specific material microstructure and system of interest. Our approach allows re-searchers to probe dependency of thermal conductivity on a variety of reactor operating and material conditions. The purpose of building such a model is to allow for improved predictive capabilities linking structure-property-processing-performance relationships in the system of interest (here, irradiated nuclear fuel), which could lead to improved experimental test planning and characterization. The uranium-molybdenum system is the fuel system studied in this work, and historic irradiation test data is leveraged for model development. Our model achieved a mean absolute percent error of approximately 10%. Results indicate our model generalizes well to never before seen data, and thus use of deep learning methods for material property predictions from limited, historic irradiation data is a viable approach. This work challenges the current paradigm in materials science, where material property models are based on limited experimental data fitted to low-dimensionality phenomenological models, or by simulations in which fundamental equations are explicitly solved. The work presented here aims to demonstrate the potential and limitations of machine learning in the field of materials science and material property modeling.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Nuclear Technology Research and Development (NTRD) Technical Monthly June FY18

The Nuclear Technology Research and Development (NTRD) Technical Monthly Report for June FY18 covers the following areas: Advanced Fuels Campaign, Material Recovery and Waste Forms Development, MPACT Campaign, Fuel Cycle Options Campaign, Joint Fuel Cycle Study Activities, and, AFCI-HQ Program Support.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Electrolytic Renewable Fuel Production Optimal Operation Investigation - H2@Scale: Cooperative Research and Development Final Report, CRADA Number CRD-19-00818

This work explored the optimal design and operation of electrolytic hydrogen production from renewable power. While there are many financial incentives for renewable hydrogen, this work investigated the effects of Low Carbon Fuel Standards (LCFS) on the hydrogen breakeven cost. To this end, this project developed optimized operational strategies for electrolytic hydrogen facilities in the California. Four specific model projects were used as the basis for the analysis. Three model projects will be located in California and one in Texas or the Midwest to be determined during the project. All projects include interconnection to the natural gas grid as a method of transport for the product fuel and the hydrogen cases will also consider alternative modes of transport. The operational optimization will maximize project return through fuel production and grid services based on future scenarios for the value of each. The effort will rely on modeling tools developed by NREL and the UCI for grid modeling as well as the integrated resource planning tool (IRP), RESOLVE, which is the official IRP tool used by the California Public Utility Commission (CPUC). The development of this project can help to inform future hydrogen deployment, policy and regulation makers, research and development decisions, and private and public investment.

08 HYDROGEN↗
NASA NTRS2010

Concepts and Challenges for Environmentally Friendly En Route Operations

A flight trajectory optimization algorithm with fuel and contrails models, which develops alternative flight paths, provides policy makers the necessary data to make tradeoffs between persistent contrails mitigation and aircraft fuel consumption. This study develops an algorithm that calculates wind-optimal trajectories for cruising aircraft while avoiding the regions of airspace prone to persistent contrails formation. The optimal trajectory is derived using Singular Perturbation Method. The regions of airspace favorable to persistent contrails formation are modeled as high-risk areas that aircraft should avoid and are adjustable. The tradeoffs between persistent contrails formation and additional travel time are investigated for wind-optimal trajectories and various contrails-avoidance trajectories at 10 different cruising altitudes for flights departing from Chicago and San Diego to New York. The additional travel times required for avoiding 100% persistent contrails formation at various flight altitudes ranged from approximately 0% to 4.3% for flights from Chicago to New York. For flights between San Diego and New York, additional traveling times vary between 1.3% and 5% depending on the cruise altitude and the percentage of contrail avoidance. Talk will present the results of aircraft fuel consumptions that are proportional to the travel time for cruising aircraft.

Sridhar, Banavar↗
DOE OSTI2024

Pebble Tanker Model for Nuclear Criticality Safety Needs

This report documents a study performed to investigate the requirements for criticality safety benchmark experiments for high-assay, low-enriched uranium (HALEU) fuel in transportation applications. In this work, an exploratory application model, the “Pebble Tanker,” was developed to represent TRISO fuel in a transportation scenario for an analysis of the validation basis in industrial quantities. An aspect of the criticality validation process involves assessing the “similarity” between application and experimental benchmark systems through an integral index parameter evaluation. Here, this includes propagating nuclear data uncertainties and calculating a correlation coefficient (hereinafter referred to as “c k ”) to evaluate the similarity of benchmark experiments compared with the application Pebble Tanker model. Finding sufficient critical benchmark experiments allows for the evaluation of bias and bias uncertainty, thus determining the upper subcritical limit (USL) of the transportation package. A target k eff of ~0.94 was used in this work to establish appropriate modeling conditions, reflecting a reasonable estimate for a USL. Two container models were investigated: one with the Hermes-type pebble and one with the Pebble Bed Modular Reactor (PBMR)–type pebble. The models were simplified, considering only fuel, containment structure, and either water or air. This allows a focus on the underlying physics of applications involving TRISO fuel pebbles using the Pebble Tanker model. A crucial consideration is the transport package's ability to safely hold pebbles while flooded, maintaining subcritical conditions. Tools available in the SCALE 6.3.1 suite—the CSAS6-Shift, TSUNAMI-3D-Shift, and TSUNAMI-IP sequences—were employed for neutronics and sensitivity and uncertainty (S/U) analysis of the Pebble Tanker. Findings demonstrated sufficient available critical experiment benchmarks to perform a validation of the Pebble Tanker in the most reactive state, i.e., when the Tanker is flooded.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1973

Mass flow controller for gaseous propellants

Gaseous propellants exhibit large variations in pressure and temperature and hence in fuel delivery. All-mechanical, mass flow controller which compensates for these variations has been developed to maintain constant fuel rate of gas. Further work is necessary to ease inlet pressure limitation.

Source record↗
DOE OSTI2017

Investing in Alternative Fuel Infrastructure: Insights for California from Stakeholder Interviews: Preprint

Increased interest in the use of alternative transportation fuels, such as natural gas, hydrogen, and electricity, is being driven by heightened concern about the climate impacts of gasoline and diesel emissions and our dependence on finite oil resources. A key barrier to widespread adoption of low- and zero-emission passenger vehicles is the availability of refueling infrastructure. Recalling the 'chicken and egg' conundrum, limited adoption of alternative fuel vehicles increases the perceived risk of investments in refueling infrastructure, while lack of refueling infrastructure inhibits vehicle adoption. In this paper, we present the results of a study of the perceived risks and barriers to investment in alternative fuels infrastructure, based on interviews with industry experts and stakeholders. We cover barriers to infrastructure development for three alternative fuels for passenger vehicles: compressed natural gas, hydrogen, and electricity. As an early-mover in zero emission passenger vehicles, California provides the early market experience necessary to map the alternative fuel infrastructure business space. Results and insights identified in this study can be used to inform investment decisions, formulate incentive programs, and guide deployment plans for alternative fueling infrastructure in the U.S. and elsewhere.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
NASA NTRS1992

Fuel consumption in optimal control

A method has been developed for comparing three optimal control strategies based on fuel consumption. A general cost function minimization procedure was developed by applying two theorems associated with convex sets. Three cost functions associated with control saturation, pseudofuel, and absolute fuel are introduced and minimized. The first two cost functions led to the bang-bang and continuous control strategies, and the minimization of absolute fuel led to an impulsive strategy. The three control strategies were implemented on two elementary systems and a comparison of fuel consumption was made. The impulse control strategy consumes significantly less fuel than the continuous and bang-bang control strategies. This comparison suggests a potential for fuel savings in higher-order systems using impulsive control strategies. However, since exact solutions to fuel-optimal control for large-order systems are difficult if not impossible to achieve, the alternative is to develop near-optimal control strategies.

Redmond, Jim↗
DOE OSTI2021

Evaluation of pebble scanning strategies for fuel qualification by simple simulated radiography

Reactor designs using tristructural-isotropic (TRISO) particles in pebble fuel forms have been developed in multiple programs and are now being designed and deployed by several companies around the world. Fuel pebbles for these designs commonly contain an inner fueled zone packed with TRISO particles and a fuel-free zone on the surface of the pebble to protect the particles within. Therefore, qualification of pebbles for use in a reactor commonly requires determination that there are no TRISO particles within a given minimum distance to the pebble surface. This determination may be made non-destructively by pebble radiography, in which a series of x-ray projections of the pebble are acquired. In this work, a method to create and image representative digital pebbles was developed to allow for rapid testing of various pebble radiography strategies to evaluate their effectiveness in detecting escapee particles. Using this method, imaging strategies in two and three dimensions were tested with fixed and adaptive angular distributions. Here it was found that an adaptive two-dimensional pebble scanning strategy was the most effective for fuel pebble qualification.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

The electric current effect on electrochemical deconsolidation of spherical fuel elements - 18447

For High-Temperature Gas-cooled Reactor in China, fuel particles are bonded into spherical fuel elements by a carbonaceous matrix. In conjunction with the development of a system to determine fuel failure fractions through {gamma}-analysis of individual and undamaged fuel particles, an automated electrochemical deconsolidation apparatus was developed in this study to separate the particles from the carbonaceous matrix by disintegrating the matrix into fine graphite powder, based on the anodic oxidation of graphite in an electrolyte containing nitric acid. A series of tests were conducted under a range of deconsolidation conditions to develop procedures and refine the equipment. The easy-handle un-irradiated spherical fuel elements were used as samples for initial tests. It is found that the morphology, size, and element content of deconsolidated graphite matrix and free particles were notably affected by electric current intensity. During the electrolysis, the nitric acid electrolyte gradually infiltrated into the micropores of graphite matrix, and the electrochemical reaction would take place around micropore structure as well as matrix surface. With much higher surface reaction capability due to the large specific area of micropore structure, the expansion stress originated from inner surface of micropores widened the c-spacing of the layer lattice of carbonaceous matrix, which was thought to be one of the main driving force of deconsolidation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Modeling ammonia-fueled co-flow dual-channel protonic-ceramic fuel cells

This paper reports the model development for a dual-channel protonic-ceramic fuel cell (PCFC) operating on ammonia fuel. The model considers the coupled interactions of several physical and chemical processes, including three-dimensional heat conduction within the bipolar plates and the membrane-electrode assembly (MEA), one-dimensional flow within the fuel and air channels, detailed heterogeneous catalytic reactions within the porous composite anode structure, Butler–Volmer representation of the charge-transfer chemistry, and Nernst–Planck transport of three charged defects (protons, oxygen vacancies, and small polarons) within the dense electrolyte membrane. The membrane-electrode assembly is composed of a Ni-BCZYYb (BaCe 0.7 Zr 0.1 Y 0.1 Yb 0.1 O 3-δ ) anode, a BCZYYb electrolyte membrane, and a BCFZY (BaCo 0.4 Fe 0.4 Zr 0.1 Y 0.1 O 3-δ ) cathode. Chemical and physical parameters for the MEA model are established using previously published button-cell data. One aspect of the study is to investigate the partial ammonia decomposition upstream of the fuel cell. Such fuel cracking increases the H 2 content of the fuel entering the PCFC, which may have benefits. However, endothermic ammonia pyrolysis within the composite anode structure assists with thermal control of the cell. The dual-channel model can be considered as the unit cell of a full fuel-cell stack.

25 ENERGY STORAGE↗
DOE OSTI2021

Modeling Environmental Effects on Ventilated Spent Fuel Storage Systems

This report describes newly developed external environment wind effects models of spent nuclear fuel (SNF) dry storage systems. The primary purpose of these wind effects models is to better understand particle deposition on SNF canisters in the context of chloride-induced stress corrosion cracking. The goal of this effort is to further the understanding and improve the wind effects model of the Nuclear Horizontal Modular Storage (NUHOMS ® ) Advanced Horizontal Storage Module–High Seismic (AHSM-HS) storage system in support of the planned Canister Deposition Field Demonstration (CDFD) project (Durbin et al. 2021, Suffield et al. 2021). The steps in this wind effects investigation begin with a validation case of comparing experimental data with a STAR-CCM+ computational fluid dynamics (CFD) model of the Belowground Vertical Dry Cask Simulator (BVDCS) with the external environment explicitly modeled. Next, a test case is performed with the AHSM-HS models comparing solar-loading strategies for a standalone and wind effects model in STAR-CCM+ and a standalone model in ANSYS Mechanical Parametric Design Language (APDL). The final portion of this study compares results of the validation exercise with applications of wind effects models for two horizontal storage systems, a site specific NUHOMS ® horizontal storage module and a NUHOMS ® AHSM-HS.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

The characterization of microstructure and chemistry of transmutation fuels

The goal of the fast reactor development programs is fabrication and certification of a nuclear fuel system that performs all functions of a conventional fuel while destroying recycled actinides. Metal fuels, such as U-Pu-Zr fuels, are considered to be candidates for this application. The physical and mechanical properties, as well as behavior laws and correlations of metal alloy fuel, all of which define the behavior and performance of metal alloy fuels has been investigated in detail. The results indicated that the prospect for fueling a future fast spectrum reactor using metal fuels is excellent. There are several challenges associated with the fuel implementation in reactor environment. One of the limiting factors in lifetime and safety of thermal and fast reactor systems is fuel swelling and resulting fission product transport to cladding. Understanding diffusion kinetics and phases formed between metal fuel and cladding upon exposure to high temperatures typical in reactor environment is critical for ensuring integrity, safety, and performance of the material, advancement of this type of fuel in the future, and securing long-term success of the nuclear fleet. The first step to understanding fuel-cladding interaction is investigation of an as-cast fuel microstructure. Despite the importance of these fuels, detailed description of microstructures in these materials is extremely limited. The majority of published results are based on characterization of as-cast microstructures of these fuels in scanning electron microscopes (SEM). In this work, we present the results from microstructural characterization of metallic Pu-based fuels, which is essential in the validation of current and future atomistic and meso-scale modeling efforts. A firm understanding of microstructural evolution in reactor materials is imperative and will lead to deeper understanding of in-reactor degradation behavior, creating a new generation of radiation-resistant materials and fuels with the potential for significant increases in burn-up.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Development of R&D GREET 2023 Rev1 to Estimate Greenhouse Gas Emissions of Sustainable Aviation Fuels for 40B Provision of the Inflation Reduction Act

The federal Interagency Working Group on sustainable aviation fuels (SAF) tasked Argonne National Laboratory with developing a modified version of the Greenhouse Gases, Regulated Emissions, and Energy Use in Technologies (GREET) model based on R&D GREET 2023. The goal of the new GREET version is to simulate the life-cycle greenhouse gas (GHG) emissions associated with seven sustainable aviation fuel (SAF) pathways for consideration under the 40B Provision of the Inflation Reduction Act – Sustainable aviation fuel credit. The Provision includes a new GHG-based tax credit to incentivize SAF production and reduce the costs of these fuels.

02 PETROLEUM↗
DOE OSTI2026

Results of the AGR-5/6/7 UCO TRISO fuel irradiation test in the Advanced Test Reactor

AGR-5/6/7 was the last in a series of irradiation experiments sponsored by the U.S. Department of Energy in support of the development and qualification of TRISO coated particle fuel for use in a high-temperature gas-cooled reactor. This experiment was conducted to verify the performance of the reference-design TRISO-coated low-enriched UCO fuel for modular high-temperature gas-cooled reactor normal operating conditions and to explore fuel performance at temperatures substantially beyond those typical of normal operation. A total of 194 UCO fuel compacts in five capsules were irradiated in the Advanced Test Reactor for 360.9 effective full-power days, achieving final burnup ranging from 5.66% to 15.26% fissions per initial heavy metal atom and fast neutron fluence ranging from 1.62 × 10 25 to 5.55 × 10 25 n/m 2 (E > 0.18 MeV). Calculated time-averaged fuel temperatures ranged from 467 °C to 1432 °C, with a peak fuel temperature of 1536 °C. During the first five irradiation cycles (∼180 effective full-power days), 85m Kr fission gas release-rate-to-birth-rate ratios were 10 −7 –10 −6 , indicating no particle failures. Near the end of the sixth cycle, a substantial number of in-pile particle failures occurred in Capsule 1. The fission gas release from this capsule impacted the readings from the other four capsules and led to unreliable fission gas release measurements for all capsules during the last four cycles. A preliminary post-irradiation examination of Capsule 1 fuel and internal components revealed the in-pile particle failures resulted from operational issues with the capsule, not subpar performance of the fuel particles.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Volatility Studies of Lanthanide Hexafluoroacetylacetone Chelates

The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy (DOE) Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors (LWRs). Some of the leading candidates to replace traditional zirconium-based cladding are alumina-forming ferritic alloys (e.g., FeCrAl) and silicon carbide (SiC) composites. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative geometry in the High Flux Isotope Reactor (HFIR) under relevant LWR temperatures and in some cases under prototypic heat flux. These designs allow for post-irradiation examination (PIE) of cladding which closely resembles expected commercially viable geometries and microstructures. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost for moving advanced cladding materials closer to commercialization. This work describes the capsule designs that have been developed at ORNL, some initial results, and plans for future irradiations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Mechanical Reliability and Life Prediction of Coated Metallic Interconnects within Solid Oxide Fuel Cells

Metallic cell interconnects (IC) made of ferritic stainless steels, i.e., iron-based alloys, have been increasingly favored in the recent development of planar solid oxide fuel cells (SOFCs) because of their advantages in excellent imperviousness, low electrical resistance, ease in fabrication, and cost effectiveness [1]. Typical SOFC operating conditions inevitably lead to the formation of oxide scales on the surface of ferritic stainless steel, which could cause delamination, buckling, and spallation resulting from the mismatch of the coefficient of thermal expansion and eventually reduce the lifetime of the interconnect components. Various protective coating techniques have been applied to alleviate these drawbacks [2]. In the present work, a fracture-mechanics-based quantitative modeling framework has been established to predict the mechanical reliability and lifetime of the spinel-coated, surface-modified specimens under an isothermal cooling cycle. Analytical solutions have been formulated to evaluate the scale/substrate interfacial strength and determine the critical oxide thickness in terms of a variety of design factors, such as coating thickness, material properties, and uncertainties. The findings then are correlated with the experimentally measured oxide scale growth kinetics to quantify the predicted lifetime of the metallic interconnects.

Solid oxide fuel cell (SOFC), Ferritic stainless s↗
DOE OSTI2020

Benchmark Modeling and Simulation of the FFTF LOFWOS Test #13 Using SAM

The Fast Flux Test Facility (FFTF) was a 400 MW thermal powered, oxide-fueled, liquid sodium cooled test reactor, built to assist development and testing of advanced fuels and materials for fast breeder reactors. In July 1986, a series of unprotected Loss of Flow Without Scram (LOFWOS) transients were performed in FFTF as part of the Passive Safety Testing (PST) program. The LOFWOS Test #13, which was initiated at 50% power and 100% flow with the pump pony motors left off, has been chosen as a benchmark case by IAEA to support collaborative efforts within international partnerships on the validation of simulation tools and models in the area of sodium fast reactor passive safety in an IAEA Coordinated Research Project (CRP), launched in October 2018. The System Analysis Module (SAM) is an advanced and modern system analysis tool under development at Argonne National Laboratory for advanced non-LWR safety analysis. It utilizes the object-oriented application framework MOOSE to leverage the modern software environment and advanced numerical methods. The capabilities of SAM are being extended to enable the transient modeling, analysis, and design of various advanced nuclear reactor systems. To participate the IAEA CRP and enhance the SAM validation base for advanced reactor transient safety analysis, benchmark simulations of the FFTF LOFWOS Test #13 are performed using the SAM code. In this first phase of the validation effort, the thermal-hydraulic behavior of the reactor system is the focus and the reactor kinetics is not considered in the SAM FFTF model. Instead, the results of Argonne’s neutronics calculations are directly used, including the power shape of the active core region and the power history during the transient. The simulation results of FFTF at steady state agreed well with the measured data from the test. During the transient, reasonably good agreement were also obtained. Future work to improve the model will focus on introducing the reactivity predictions into the model, as well as better understanding or resolving the current discrepancies with the measured data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗