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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.

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At least 1,081 records · Page 60
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Executing BISON-CASL Fuel Performance Cases Using VERA-CS Output

As the Consortium for Advanced Simulation of Light Water Reactors (CASL) moves forward with more complex multiphysics simulations, there is increased focus on incorporating fuel performance analysis methods. The coupled neutronics/thermal-hydraulics capabilities within the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) have become relatively stable and major advances have been made in analysis efforts, including the simulations of twelve cycles of Watts Bar Nuclear Unit 1 (WBN1) operation. However, VERA-CS approaches for treating fuel pin heat transfer have well-known limitations that could be eliminated through better integration with the BISON-CASL fuel performance code. Several approaches are being taken to improve or replace the VERA-CS fuel models and improve integration with BISON-CASL. Tiamat is being developed to replace the fuel models with a direct coupling of BISON-CASL and the neutronics and coolant heat transfer of VERA-CS. Tiamat will execute in an inline mode that allows for in-memory transfer of VERA-CS data to BISON-CASL. This will provide fuel performance results during core simulation. However, fuel performance typically undergoes an independent analysis using a stand-alone fuel performance code with manually specified input defined from an independent core simulator solution or set of assumptions. This milestone covers initial efforts to facilitate using VERA for core simulation and fuel performance to operate in this mode of execution. This approach will be used for several CASL fuel performance challenges, such as RIAs and DNB. These issues require time-dependent full core results to establish the boundary conditions for the high-fidelity BISON-CASL simulations. The approach will also be used to improve fuel temperature models in VERA-CS, guide the development of Tiamat, and simplify the use of BISON-CASL within VERA for industry users.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Fused Global-Local Economic Model Predictive Control for Real-Time Eco-Optimal Control of a Heavy Duty Truck

This paper introduces a fused global-local economic model predictive control (MPC) with varying time steps for maximizing the fuel economy of a heavy-duty truck, where global MPC is performed over a long horizon using cloud-based dynamic programming (DP), and local MPC is performed on-board the vehicle using sequential quadratic programming (SQP). Because of the computational complexity of DP, an in-depth analysis of how the discretization levels, time constant, and horizon length influence the energy savings and computational complexity of online DP has been performed. The fused optimization technique has been evaluated over different route profiles, accounting for traffic, using a medium-fidelity proprietary Volvo model. Results show 4-7% fuel savings, as compared to a baseline model.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2021

Irradiation-enhanced diffusion and diffusion-limited creep in U 3 Si 2

U 3 Si 2 is an advanced fuel candidate due to its relatively high fissile density and attractive thermal properties. Compared to standard UO 2 fuel, there are significant data gaps for the thermophysical and thermomechanical properties of U 3 Si 2 . Point defect concentrations and mobilities under irradiation govern a number of important fuel performance properties, such as creep and fission gas release. In this work, we utilized density functional theory (DFT) data to inform a cluster dynamics framework to predict point defect concentrations in U 3 Si 2 under irradiation. Molecular dynamics (MD) simulations were used to examine the contribution of atomic mixing during ballistic cascades to diffusion, as well as the diffusivity of U and Si at grain boundaries. These atomic scale models for diffusivity were then used to inform a creep model based on bulk (Nabarro-Herring) and grain boundary (Coble) diffusional creep, and climb-limited dislocation creep. The model compares well against available experimental data and has been implemented in the BISON fuel performance code. Finally, a demonstration case using simple power profiles has been carried out, showing that negligible creep occurs due to the low temperatures experienced by U 3 Si 2 in-reactor, a consequence of its high thermal conductivity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Assessment and Improvement of Fission Product Transport Predictions of Particle Fuel in BISON

The U.S. Department of Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program aims to develop predictive capabilities by applying computational methods to the analysis and design of advanced reactor and fuel cycle systems. This program has been providing engineering-scale support for the development of BISON, a high-fidelity and high-resolution fuel performance tool. This study was motivated by the need to incorporate more physics-based models in BISON in order to foster tri-structural isotropic (TRISO) applications. This document details the integration of new modeling capabilities in BISON, including (1) development of pyrolytic carbon (PyC) and silicon carbide (SiC) layer anisotropic thermal and mass transport capabilities, (2) verification of the mass diffusion solution in TRISO modeling, (3) calibration of fission product diffusivity using Advanced Gas Reactor (AGR) experiments, (4) improved fission product release modeling by developing compact diffusion modeling capabilities, and (5) documentation of accelerated failure analysis on the BISON website. Improvements made to the diffusion models and parameters were documented and validated against AGR-1 and -2 experiment data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1982

Optimal short-range trajectories for helicopters

An optimal flight path algorithm using a simplified altitude state model and a priori climb cruise descent flight profile was developed and applied to determine minimum fuel and minimum cost trajectories for a helicopter flying a fixed range trajectory. In addition, a method was developed for obtaining a performance model in simplified form which is based on standard flight manual data and which is applicable to the computation of optimal trajectories. The entire performance optimization algorithm is simple enough that on line trajectory optimization is feasible with a relatively small computer. The helicopter model used is the Silorsky S-61N. The results show that for this vehicle the optimal flight path and optimal cruise altitude can represent a 10% fuel saving on a minimum fuel trajectory. The optimal trajectories show considerable variability because of helicopter weight, ambient winds, and the relative cost trade off between time and fuel. In general, reasonable variations from the optimal velocities and cruise altitudes do not significantly degrade the optimal cost. For fuel optimal trajectories, the optimum cruise altitude varies from the maximum (12,000 ft) to the minimum (0 ft) depending on helicopter weight.

Slater, G. L.↗
DOE OSTI2024

A realistic TRACE PWR LOCA model with application to high-burnup analysis

This paper presents a TRACE model of a postulated large-break loss-of-coolant accident (LBLOCA) event in a four-loop pressurized water reactor (PWR). The input files have been made publicly available and serve as a starting point for researchers to improve upon or apply to their purposes. The model, based on a legacy PWR model, contains numerous improvements and modifications consistent with US Nuclear Regulatory Commission LBLOCA analysis guidelines. The model is applied to analyze high-burnup (>62 GWD/MTU) fuel behavior during LBLOCA, improving on previous work to provide realistic thermal hydraulic predictions for future fuel performance analyses and experiments related to fission fragment, relocation, and dispersal (FFRD). Sensitivity studies are presented to quantify the impact of the vessel-modeling approach, core radial discretization, and other phenomena. In conclusion, the model is intended for research purposes and contains sufficient plant geometric, operational, and safety system details to represent the main physical phenomena pertaining to LBLOCA.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

NUCLEAR ENERGY SYSTEM COST MODELING

The U.S. Department of Energy’s Fuel Cycle Technologies (FCT) Program is preparing to perform an evaluation of the full range of possible Nuclear Energy Systems (NES) in 2013. These include all practical combinations of fuels and transmuters (reactors and sub-critical systems) in single and multi-tier combinations of burners and breeders with no, partial, and full recycle. As part of this evaluation, Levelized Cost of Electricity at Equilibrium (LCAE) ranges for each representative system will be calculated. To facilitate the cost analyses, the 2009 Advanced Fuel Cycle Cost Basis Report is being amended to provide up-to-date cost data for each step in the fuel cycle, and a new analysis tool, NE-COST, has been developed. This paper explains the innovative “Island” approach used by NE-COST to streamline and simplify the economic analysis effort and provides examples of LCAE costs generated. The Island approach treats each transmuter (or target burner) and the associated fuel cycle facilities as a separate analysis module, allowing reuse of modules that appear frequently in the NES options list. For example, a number of options to be screened will include a once-through uranium oxide (UOX) fueled light water reactor (LWR). The UOX LWR may be standalone, or may be the first stage in a multi-stage system. Using the Island approach, the UOX LWR only needs to be modeled once and the module can then be reused on subsequent fuel cycles. NE-COST models the unit operations and life cycle costs associated with each step of the fuel cycle on each island. This includes three front-end options for supplying feedstock to fuel fabrication (mining/enrichment, reprocessing of used fuel from another island, and/or reprocessing of this island’s used fuel), along with the transmuter and back-end storage/disposal. Results of each island are combined based on the fractional energy generated by each islands in an equilibrium system. The cost analyses use the probability distributions of key parameters and employs Monte Carlo sampling to arrive at an island’s cost probability density function (PDF). When comparing two NES to determine delta cost, strongly correlated parameters can be cancelled out so that only the differences in the systems contribute to the relative cost PDFs. For example, one comparative analysis presented in the paper is a single stage LWR-UOX system versus a two-stage LWR-UOX to LWR-MOX system. In this case, the first stage of both systems is the same (but with different fractional energy generation), while the second stage of the UOX to MOX system uses the same type transmuter but the fuel type and feedstock sources are different. In this case, the cost difference between systems is driven by only the fuel cycle differences of the MOX stage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Release 1.0 of MARMOT: A Mesoscale Fuel Performance Code

MARMOT is the mesoscale fuel performance code under development as part of the US DOE Nuclear Energy Advanced Modeling and Simulation Program. With this report, we release a new version of MARMOT, summarize its implemented capabilities, and document the ongoing development of new capabilities planned for the next release. The purpose of MARMOT is to predict the coevolution of microstructure and material properties of nuclear fuel and cladding due to stress, temperature, and irradiation damage. It accomplishes this using the phase field method coupled to solid mechanics and heat conduction. MARMOT is based on the Multiphysics Object-Oriented Simulation Environment (MOOSE), and much of its basic capability in the areas of the phase field method, mechanics, and heat conduction come directly from MOOSE modules.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Fuel utilization effects on system efficiency in solid oxide fuel cell gas turbine hybrid systems

Here, a computational analysis was conducted to optimize the design of a solid oxide fuel cell - gas turbine hybrid power generator, focusing on the impact that fuel utilization within the fuel cell has on system efficiency and installed costs. This is the first ever design-study considering the effect of fuel utilization on performance, as well as on the optimum power split. This hybrid system attained high electric generation efficiencies (>70%) over a wide range of operating conditions (60% < fuel utilization < 90%) while the fuel cell stack size decreased in proportion to decreasing the fuel utilization. A one-dimensional fuel cell model was used to simulate the fuel cell while GateCycle® was used to simulate the performance of the associated recuperated turbine and various subsystems necessary for thermal management. For each test case, the size of the solid oxide fuel cell, gas turbine, and recuperator, as well as the fuel and air flow rates, hot-air bypass set point, and heat exchange effectiveness in the solid oxide fuel cell manifold were varied to obtain 550 MWe output. In addition, anode recycle, turbomachinery efficiency, and various thermal management options were tested. The maximum system efficiency (75.6%) was attained for the single-pass solid oxide fuel cell with highly efficient turbomachinery when the solid oxide fuel cell used 80% of the incoming fuel. Efficiency was essentially flat from 75% fuel utilization through 85% fuel utilization. Employing anode recycle starting at 65% resulted in roughly 1 percentage point efficiency decrease for each percent increase in fuel utilization. For minimized solid oxide fuel cell degradation, a near 50:50 power split case was studied resulting in 68.6% efficiency and the solid oxide fuel cell using 55% of the incoming fuel. Because of shifting half of the power generation to the gas turbine, the size of the fuel cell stack was reduced by 25% as compared to that at maximum efficiency (80% fuel utilization).

42 ENGINEERING↗
DOE OSTI2023

BSM Sensitivity Analysis and Meta-Modeling Next Steps [Slides]

We performed a global sensitivity analysis of the Bioenergy Scenario Model (BSM), using Elementary Effects analysis to identify potentially influential factors, with a focus on sustainable aviation fuel (SAF) production. The process was iterative and informed key model changes and updates over the course of the study in addition to identifying the key model inputs that affect the cumulative production of SAF from the present through 2050. More work will be performed to gain deeper insight into the sensitivity of individual pathways as part of our larger efforts to develop a reduced-form version of the BSM.

09 BIOMASS FUELS↗
DOE OSTI2020

Production Through Simulation: Using Simulation Technologies to Create and Evaluate Nuclear Fuel Fabrication Facility Designs

This project explores how physics-based and discrete-event simulation technologies can be jointly employed to model a suite of nuclear fuel fabrication facility designs with variable radiological environments, optimize their design to lower operational costs subject to a set of design constraints, and evaluate their performance. Initially, three variants of U-20Pu-10Zr metallic nuclear fuel in three specific geometric configurations are modeled and studied across more than 50 physics simulations in nuclear physics software packages SCALE and Monte Carlo N-Particle Transport (MCNP6.2). From these tests, values for effective dose rate and criticality are recorded for each of the nine alloy-geometry configurations. These values are then incorporated into discrete-event models of fuel fabrication in the simulation program ExtendSim Pro 10 as process attributes. This allows one to track the dose to facility personnel as fuel is fabricated, dually enabling us to evaluate the safety of a given facility design in terms of absorbed dose per worker per year and to craft operational guidelines that adhere to federal and local safety regulations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Anisotropic Effects of Loading on Zircaloy Cladding Performance

This presentation provides recent results related to the mechanical behavior of Zircaloy cladding for the Light Water Reactor Fuels Workshop/CRAFT Initiative Update presented by the Advanced Fuels Campaign. This presentation focuses on Zircaloy cladding response to pellet-cladding interactions and how mechanical anisotropy makes the cladding sensitive to the testing mode. Results are compared to existing material models used by BISON, and implications for reactor performance are briefly discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

Modeling of selected ceramic processing parameters employed in the fabrication of 238 PuO 2 fuel pellets

Recent deep space missions utilize the thermal output of the radioisotope plutonium-238 as the fuel in the thermal to electrical power system. Since the application of plutonium in its elemental state has several disadvantages, the fuel employed in these deep space power systems is typically in the oxide form such as plutonium-238 dioxide ( 238 PuO 2 ). As an oxide, the processing of the plutonium dioxide into fuel pellets is performed via ''classical'' ceramic processing unit operations such as sieving of the powder, pressing, sintering, etc. Modeling of these unit operations can be beneficial in the understanding and control of processing parameters with the goal of further enhancing the desired characteristics of the 238 PuO 2 fuel pellets. A finite element model has been used to help identify the time-temperature-stress profile within a pellet during a furnace operation taking into account that 238 PuO 2 itself has a significant thermal output. The results of the modeling efforts will be discussed.

07 ISOTOPE AND RADIATION SOURCES↗
NASA NTRS2004

Reduced Gravity Studies of Soret Transport Effects in Liquid Fuel Combustion

Soret transport, which is mass transport driven by thermal gradients, can be important in practical flames as well as laboratory flames by influencing transport of low molecular weight species (e.g., monatomic and diatomic hydrogen). In addition, gas-phase Soret transport of high molecular weight fuel species that are present in practical liquid fuels (e.g., octane or methanol) can be significant in practical flames (Rosner et al., 2000; Dakhlia et al., 2002) and in high pressure droplet evaporation (Curtis and Farrell, 1992), and it has also been shown that Soret transport effects can be important in determining oxygen diffusion rates in certain classes of microgravity droplet combustion experiments (Aharon and Shaw, 1998). It is thus useful to obtain information on flames under conditions where Soret effects can be clearly observed. This research is concerned with investigating effects of Soret transport on combustion of liquid fuels, in particular liquid fuel droplets. Reduced-gravity is employed to provide an ideal (spherically-symmetrical) experimental model with which to investigate effects of Soret transport on combustion. The research will involve performing reduced-gravity experiments on combustion of liquid fuel droplets in environments where Soret effects significantly influence transport of fuel and oxygen to flame zones. Experiments will also be performed where Soret effects are not expected to be important. Droplets initially in the 0.5 to 1 mm size range will be burned. Data will be obtained on influences of Soret transport on combustion characteristics (e.g., droplet burning rates, droplet lifetimes, gas-phase extinction, and transient flame behaviors) under simplified geometrical conditions that are most amenable to theoretical modeling (i.e., spherical symmetry). The experiments will be compared with existing theoretical models as well as new models that will be developed. Normal gravity experiments will also be performed.

Shaw, Benjamin D.↗
DOE OSTI2011

Technical Manual for the SAM Biomass Power Generation Model

This technical manual provides context for the implementation of the biomass electric power generation performance model in the National Renewable Energy Laboratory's (NREL's) System Advisor Model (SAM). Additionally, the report details the engineering and scientific principles behind the underlying calculations in the model. The framework established in this manual is designed to give users acomplete understanding of behind-the-scenes calculations and the results generated.

09 BIOMASS FUELS↗
DOE OSTI2012

Spent fuel pool analysis using TRACE code

The storage requirements of Spent Fuel Pools have been analyzed with the purpose to increase their rack capacities. In the past, the thermal limits have been mainly evaluated with conservative codes developed for this purpose, although some works can be found in which a best estimate code is used. The use of best estimate codes is interesting as they provide more realistic calculations and they have the capability of analyzing a wide range of transients that could affect the Spent Fuel Pool. Two of the most representative thermal-hydraulic codes are RELAP-5 and TRAC. Nowadays, TRACE code is being developed to make use of the more favorable characteristics of RELAP-5 and TRAC codes. Among the components coded in TRACE that can be used to construct the model, it is interesting to use the VESSEL component, which has the capacity of reproducing three dimensional phenomena. In this work, a thermal-hydraulic model of the Maine Yankee spent fuel pool using the TRACE code is developed. Such model has been used to perform a licensing calculation and the results obtained have been compared with experimental measurements made at the pool, showing a good agreement between the calculations predicted by TRACE and the experimental data. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

GOTHIC Aerosol Source Depletion Studies

Pacific Northwest National Laboratory has continued work to develop an aerosol-laden flow modeling capability with the Generation of Thermal Hydraulic Information in Containment (GOTHIC TM ) computer code to perform simulations for thermal hydraulic conditions and aerosol transport and deposition in spent fuel casks. This report describes our recent work to expand our model that was originally developed in 2019 to allow for thermal characterization, carrier gas flow determination, and tracking of particulate behavior throughout the entire canister volume. This was achieved through conversion of decay heat source in all fuel tubes within the model from heaters to thermal conductors in GOTHIC as well as remeshing the internal volume of the canister. These model improvements allowed for simulations of base case scenarios for comparison of similar efforts by our collaborators with alternative tools at Sandia and Oak Ridge. Several characteristics of the GOTHIC code were elucidated by this effort and were documented for consideration in how the code is used for this effort moving forward. For instance, treatment of particle size distributions, initial spatial distributions, mesh gradients, and minimum volume concentrations were identified as important issues for consideration. Ultimately, the current version of the code is capable of tracking temperatures, flow rates, and particle behavior throughout the canister internal volume. This report presents predicted depletion times and preferred deposition patterns within the canister. Additionally, this report documents code performance and insights for further model development and use for comparisons to other codes and for predicting system behavior in experiments being performed and planned by our collaborators.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Status Report on Activities of the Systems Assessment Task Force, OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs

The Organization for Economic Cooperation and Development /Nuclear Energy Agency (OECD/NEA) Nuclear Science Committee approved the formation of an Expert Group on Accident Tolerant Fuel (ATF) for LWRs (EGATFL) in 2014. Chaired by Kemal Pasamehmetoglu, INL Associate Laboratory Director for Nuclear Science and Technology, the mandate for the EGATFL defines work under three task forces: (1) Systems Assessment, (2) Cladding and Core Materials, and (3) Fuel Concepts. Scope for the Systems Assessment task force (TF1) includes definition of evaluation metrics for ATF, technology readiness level definition, definition of illustrative scenarios for ATF evaluation, and identification of fuel performance and system codes applicable to ATF evaluation. The Cladding and Core Materials (TF2) and Fuel Concepts (TF3) task forces will identify gaps and needs for modeling and experimental demonstration; define key properties of interest; identify the data necessary to perform concept evaluation under normal conditions and illustrative scenarios; identify available infrastructure (internationally) to support experimental needs; and make recommendations on priorities. Where possible, considering proprietary and other export restrictions (e.g., International Traffic in Arms Regulations), the Expert Group will facilitate the sharing of data and lessons learned across the international group membership. The Systems Assessment task force is chaired by Shannon Bragg-Sitton (Idaho National Laboratory [INL], U.S.), the Cladding Task Force is chaired by Marie Moatti (Electricite de France [EdF], France), and the Fuels Task Force is chaired by a Masaki Kurata (Japan Atomic Energy Agency [JAEA], Japan). The original Expert Group mandate was established for June 2014 to June 2016. In April 2016 the Expert Group voted to extend the mandate one additional year to June 2017 in order to complete the task force deliverables; this request was subsequently approved by the Nuclear Science Committee. This report provides an update on the status Systems Assessment Task Force activities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗