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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Impacts of Irradiation Structural Behavior on Thermal Hydraulics Safety Analysis to Support MURR LEU Conversion

The University of Missouri Research Reactor (MURR) located in Columbia, Missouri is one of six U.S. High Performance Research Reactors (USHPRR), including one critical facility, that is actively collaborating with U. S. Department of Energy (DOE) National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Office of Reactor Conversion and Uranium Supply to convert from the use of highly enriched uranium (HEU; ≥20 wt% U-235) to low-enriched uranium (LEU; <20 wt% U-235) fuel. A new type of very high-density LEU fuel based on an alloy of uranium and 10 wt% molybdenum (U-10Mo) is expected to allow the conversion to LEU of MURR, as well as four other USHPRR. MURR has been working with the Reactor Conversion Pillar at Argonne to perform fuel element design and fuel cycle performance analyses, steady-state thermal hydraulics safety analysis, and accident safety analyses in preparation for the conversion of MURR and to support a preliminary safety analysis report for conversion to LEU fuels. Subsequent analyses have also been performed, including transition cycles where all-fresh LEU fuel elements are introduced upon conversion and progressing through reactor operations the core is brought to equilibrium. Thermal hydraulics safety analyses performed as part of the above have employed an assumption on channel gap reduction due to burnup-related phenomena including fuel swelling, irradiation creep, and oxide layer buildup. Recently, a series of structural analyses have been performed on the MURR LEU fuel plates and an element due to significant differences between the plate and element designs of the MURR HEU and LEU fuels. In addition, NUREG-1537 indicates that structural phenomena are to be evaluated. Two separate types of structural analyses were performed for the MURR LEU fuel element: fluid-structure interaction (FSI) and irradiation thermo-mechanical. The FSI analysis evaluated the effects of hydraulic forces on the MURR LEU fuel element to quantify the flow-induced plate deflection, and a minimal impact to the channel gap thickness was predicted under prototypic and bounding conditions. The irradiation thermo-mechanical analysis evaluated the effects of fuel swelling, irradiation creep, and thermal expansion for the MURR LEU plates and the element for prototypic thermal and irradiation conditions based on a high-fidelity approach multiphysics approach. Overall, this thermo-mechanical analysis predicts smaller gap thickness changes in previously limiting regions. Larger changes are predicted in the middle of channels, and for end channels where power density is not typically a maximum. An additional thermo-mechanical analysis was performed for the outermost HEU fuel plate, which showed a similar magnitude of deflection as the outermost LEU plate. Due to substantial differences between the channel gap reductions assumed for the previous safety analyses and those predicted by the irradiation thermo-mechanical analysis, a need to evaluate their impact on the thermal hydraulics safety analyses arose. This report presents the results from the steady-state safety analyses for normal operation as well as the accident analyses for the two most limiting accident scenarios.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Coupled Multiphysics Primary Loop Simulations of the Mk1-FHR in the Virtual Test Bed

To support advanced reactor demonstrations, the Virtual Test Bed (VTB) [1] repository hosts a wide range of challenge problems for showcasing modeling and simulation capabil- ities in support of advanced reactor demonstrations. This document presents a coupled multiphysics model of the Mark 1 pebble-bed fluoride-salt-cooled high-temperature reactor (PB-FHR). The analysis leverages NEAMS tools (Griffin [2], SAM [3], Pronghorn [4], and the MOOSE [5] heat con- duction module) for core neutronics, thermal hydraulics of the core and primary loop, and multiscale fuel performance simulations. The analysis was entirely created by coupling standalone simulations of the reactor that were previously available on the VTB. All input files and documentation de- veloped for this example are available on the VTB website: mooseframework.inl.gov/ virtual_test_bed/ . This model was featured in the National Reactor Innovation Cen- ter Tech Talk presented in December 2021.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Assessment of CTF for Steady-state and Transient Post-CHF Conditions in Support of Time-at-Temperature Modeling Applications

The US nuclear industry is exploring options to improve operational economics and uprate the current fleet of light-water reactors by investigating transitioning to cladding performance–based safety criteria as opposed to the current limit, which requires complete avoidance of critical heat flux (CHF)/dryout. Past experience has shown that not all events leading to a dryout are severe enough to cause fuel performance degradation. Allowing temporary dryout of the fuel—that is, using a time-at-temperature (TaT) strategy—could allow for economic improvements via large power uprates and enhanced operational flexibility for current plants without compromising fuel integrity. To support this effort, the US Department of Energy is executing a comprehensive program that includes generating cladding material data under TaT conditions, developing new mechanistic models, and demonstrating modeling and simulation capabilities for transients of interest. This paper presents work performed to assess the CTF thermal-hydraulics subchannel code. CTF is a package used in the VERA core simulator, which will ultimately be used for TaT analysis. CTF will provide the thermal-hydraulic boundary conditions that will be needed for fuel performance analysis in the BISON code. Quantifying both the accuracy and uncertainty of post-CHF models will therefore be necessary. This paper outlines the strategy for the assessment of TaT and presents the results of using the steady-state and transient dryout experiments of the Boiling Fine-mesh Bundle Tests for CTF validation. The results show that the current model tends to overpredict steady-state critical power. This behavior translates to the transient tests, in which CTF is unable to capture transient dryout behavior. Some discussion of sensitivity analysis work being performed is provided to indicate which models must be further analyzed to properly model transient dryout and its uncertainty.

Salko Jr, Robert [ORNL] (ORCID:0000000253566679)↗
DOE OSTI2017

Two-Dimensional Neutronic and Fuel Cycle Analysis of the LEU-Fueled Transatomic Power Molten Salt Reactor

A recent Third Way report detailing $1.3 billion in private investment in advanced reactor technology includes several liquid-fueled molten salt reactor (MSR) concepts. Interest in these MSR concepts is driven by the enhanced safety, economic, and promising fuel cycle benefits of these advanced reactor concepts. The US Department of Energy (DOE) Office of Nuclear Energy established the Gateway for Accelerated Innovation in Nuclear (GAIN) to provide private companies pursuing innovative nuclear energy technologies with access to the technical support necessary to move toward commercialization. GAIN activities in 2016 included competitive selection of projects for vouchers to fund applicant-specific work at national laboratories. One of these GAIN small business voucher awards was to Transatomic Power Corporation (TAP) to work with Oak Ridge National Laboratory (ORNL) to perform neutronic and fuel cycle analysis and design of TAP's innovative MSR concept. The initial phase of this project is to perform two-dimensional (2D) analysis with ORNL-developed tools to verify TAP-calculated metrics and to develop an understanding of the effects of fuel feed and isotopic removal rates to prepare for the more complex three-dimensional (3D) analysis in coming phases. Design information was based on the TAP white papers wherever possible. This summary discusses the results of this initial phase of the collaboration between ORNL and TAP.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Cooled Gas Turbine and Combined Cycle Analysis for NH 3 -CH 4 Fuel Mixes (Up to 100% NH 3 )

In this study, a cooled gas turbine (GT) analysis was conducted for varying levels of ammonia (NH 3 ) blends with methane. The ultimate goal is to have a gas turbine design that can be used for all the fuel blends (including 100% NH 3 ) without any changes to the system. The technological developments in the cooling system, gas turbine design, and materials that will be required for NH 3 combustion were identified and analyzed in this study to develop an advanced gas turbine design for NH 3 fuels. The study includes a combined cycle performance analysis with the NH 3 fuel blends using the advanced gas turbine design developed in this study. A techno-economic analysis was conducted for analyzing the impact of the NH 3 fuels on the levelized cost of electricity and cost sensitivities to fuel price and capacity factor.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2023

Cooled Gas Turbine and Combined Cycle Analysis for H 2 -CH 4 Fuel Mixes (Up to 100% H 2 )

In this study, a cooled gas turbine analysis was conducted for varying levels of hydrogen (H 2 ) blends with the natural gas. The ultimate goal is to have a gas turbine design that can be used for all the fuel blends (including 100% H 2 ) without any changes to the system. The technological developments in the cooling system, gas turbine design and materials that will be required for H 2 combustion were identified and analyzed in this study to develop an advanced gas turbine design for H 2 fuels. The study includes a combined cycle performance analysis with the H 2 fuel blends using the advanced gas turbine design developed in this study. A techno-economic analysis was conducted for analyzing the impact of the H 2 fuels on the levelized cost of electricity and cost sensitivities to fuel price and capacity factor.

03 NATURAL GAS↗
DOE OSTI2020

A Metallic Fuel Performance Benchmark Problem Based on the IFR-1 Experiment

Metallic nuclear fuels are an active area of research and development for use in advanced reactors. Robust and accurate metallic fuel performance models are necessary for the design, safety analysis, and licensing of such reactors. However, metallic fuel performance models are not as mature as ceramic fuel performance models and require additional development. In order to better gauge the accuracy of existing models, identify high-priority models for development, and quantify the improvements made by future model development, ORNL has developed a benchmark problem based on the IFR-1 experiment. This work gathers all relevant information on the IFR-1 experiment and uses it to develop the benchmark problem. It simulates the problem using the fuel performance code BISON and compare the results to post-irradiation examination data from the experiment. Then a sensitivity analysis is performed on the BISON model to determine the benchmark problem’s sensitivity to uncertainty in the input parameters. The results suggest that BISON’s mechanical models require additional development. Plastic deformation is under-predicted in the cladding and axial swelling is over-predicted in the fuel. These problems may be related. Additionally, there is a bias in the temperature which may be a result of uncertainty in the input parameters rather than an issue with the fuel performance models.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1950

Evaluation of a ducted-fan power plant designed for high output and good cruise fuel economy

Theoretical analysis of performance of a ducted-fan power plant designed both for high-output, high-altitude operation at low supersonic Mach numbers and for good fuel economy at lower fight speeds is presented. Performance of ducted fan is compared with performance (with and without tail-pipe burner) of two hypothetical turbojet engines. At maximum power, the ducted fan has propulsive thrust per unit of frontal area between thrusts obtained by turbojet engines with and without tail-pipe burners. At cruise, the ducted fan obtains lowest thrust specific fuel consumption. For equal maximum thrusts, the ducted fan obtains cruising flight duration and range appreciably greater than turbojet engines.

ENGINES, DUCTED PROPELLER↗
DOE OSTI2020

Temperature Effects on Interdiffusion of Al and U-Mo under Irradiation

A high-energy Xe ion irradiation experiment was conducted to investigate the temperature dependence of interdiffusion in bilayer Al-UMo samples under irradiation. The amount of interdiffusion achieved at a fixed dose with the increase of temperature showed a clear transition at 175 degrees C (with an estimated error in the range of ± 10 degrees C) from temperature-independent to temperature-dependent behavior. The activation energy derived from the curve of interdiffusion quantity vs. irradiation temperature is 0.77 ± 0.16 eV. This information has been utilized to understand the temperature effect on the interdiffusion process that occurred at the interfaces of U-Mo particles and the Al matrix in U-Mo/Al dispersion fuels, whose magnitude significantly impacts the fuel's performance. Although this temperature effect was deemed important, it cannot be examined directly using in-pile irradiation data, as fuel temperatures cannot be measured in reactor irradiation and are highly correlated with fission rate and thermal conductivity evolution. To connect the knowledge accumulated from ion irradiation with in-pile irradiation data, simulation of a full-sized U-Mo/Al dispersion fuel plate irradiated in the FUTURE test in the BR2 reactor was performed with the Dispersion Analysis Research Tool (DART), a dispersion fuel performance code. DART is equipped with an interaction or interdiffusion layer (IL) growth correlation formulated to describe the temperature dependence of ion mixing results. The agreement between calculated and measured fuel meat constituent volume fractions and swelling data demonstrated that the temperature effect on in-pile Al-UMo interdiffusion is well captured with the correlation. In this case, the fitted activation energy is 0.70 eV. Considering the uncertainties associated with the ion irradiation data, the activation energy obtained from in-pile data fitting is in accord with that from ion irradiation results.

36 MATERIALS SCIENCE↗
NASA NTRS1984

Design of Long-Endurance Unmanned Airplanes Incorporating Solar and Fuel Cell Propulsion

Preliminary performance analysis and conceptual design are described for a class of unmanned airplanes possessing multi-day endurance capability. A mixed-mode electric power system incorporates solar cells for daytime energy production and a non-regenerative H2-02 fuel cell to supply energy for night flight. The power system provides energy for all onboard systems, including propulsion., payload, and avionics. Excess solar energy is available during significant portions of the day, and may be used for climbing, maneuvering, or payload functions. By jettisoning fuel cell reactant product (water) during flight, vehicle endurance may be increased under certain conditions. Empirical structure sizing algorithms are combined with low-Reynolds number aerodynamics algorithms to estimate airplane size and geometry to meet prescribed mission requirements. Initial calculations for summertime, high-altitude flight (above 40,000 ft (12 km)) at moderate latitude (31 deg N) indicate that mission endurance of several days may be possible for configurations having wing loadings on the order of 0.9 to 1.3 lb/ft(exp 2). These aircraft tend to be somewhat smaller than solar-powered aircraft previously conceived for multi-month endurance utilizing regenerative fuel cell systems for night flight.

Youngblood, James W.↗
DOE OSTI2016

Assessment of the Analysis Capability for Core-wide PWR Pellet-Clad Interaction (PCI) Screening

The Level 1 Milestone, L1:CASL.P13.03, “Assess the analysis capability for core-wide PWR Pellet-Clad Interaction (PCI) screening and demonstrate detailed 3-D analysis on selected sub-region.”, has been completed by the Fuels, Materials and Chemistry and Physics Integration focus areas of CASL. This report illustrates the application of VERA simulation to PWR core-wide PCI screening, with emphasis on the multi-dimensional analysis capabilities of BISON fuel performance code. The VERA codes utilized in this effort are: the fuel performance code BISON and the VERA-CS sub-element, which combines the neutron transport code MPACT and the thermal-hydraulics code COBRA-TF (CTF). The target reactor for this work is Watts Bar Unit 1, Cycles 6 and 7, selected on the basis that PCI failures had occurred during power start-up in both cycles, despite the moderate power levels, which offered an interesting test of the code’s veracity and a particularly challenging condition for this Milestone’s assessment. The analysis began by performing BISON-IN-VERA, i.e. VERA-CS + BISON, quarter-core depletion calculations, from which limiting assemblies and limiting fuel rods were selected using a four-parameter selection criterion: high power, high temperature, small gap thickness and high cladding hoop stress. The cladding material being Westinghouse Zirlo, the analysis utilized a new material model developed for this cladding by CASL’s FMC team using open literature sources. The VERA-CS + BISON analysis resulted in identifying 51 limiting rods as candidates for more detailed PCI analysis using stand-alone BISON, employing local 2D and 3D geometric models. This local-effects BISON analysis provided detailed PCI failure assessment, considering the impact of power histories on the two potential sources of PCI failure: the classical SCC-PCI and the missing pellet surface geometry MPS-PCI. The 51-rod analyses of Watts Bar Unit 1 cycle 6 and 7 revealed that the classical SCC-PCI stress levels were well below the failure threshold determined in a previous CASL study, which implicated the other mode of failure, namely MPS-PCI. Considering MPS defects to be the root cause for the failure, the local-effects calculations was repeated using MPS defect sizes equal to and larger than Westinghouse quality assurance standards. Initial results of this new analysis showed a significant increase in the cladding stress, indicating that an MPS may have been present in cycle 6 and 7 for the failure to occur. In order to determine the size of MPS needed for cycle 6 and 7 rods to fail, a sensitivity study using 2-D and 3-D models was performed. The results showed cladding hoop stresses, while remaining below the failure level for an MPS of equal size to the acceptance standard, continued to increase as the size of the MPS increased. This report has demonstrated the use of BISON, in consort with VERA’s simulation methodology, for core-wide PCI failure screening capability. Utilizing this demonstration, and emulating the PCI margin analysis methodology developed in the EPRI PCI guidelines, BISON-IN-VERA simulation can be used very effectively to develop core power histories to minimize PCI failure potential.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Circumnuclear Structures in Megamaser Host Galaxies

Using the Hubble Space Telescope , we identify circumnuclear (100–500 pc scale) structures in nine new H{sub 2}O megamaser host galaxies to understand the flow of matter from kpc-scale galactic structures down to the supermassive black holes (SMBHs) at galactic centers. We double the sample analyzed in a similar way by Greene et al. and consider the properties of the combined sample of 18 sources. We find that disk-like structure is virtually ubiquitous when we can resolve <200 pc scales, in support of the notion that non-axisymmetries on these scales are a necessary condition for SMBH fueling. We perform an analysis of the orientation of our identified nuclear regions and compare it with the orientation of megamaser disks and the kpc-scale disks of the hosts. We find marginal evidence that the disk-like nuclear structures show increasing misalignment from the kpc-scale host galaxy disk as the scale of the structure decreases. In turn, we find that the orientation of both the ∼100 pc scale nuclear structures and their host galaxy large-scale disks is consistent with random with respect to the orientation of their respective megamaser disks.

79 ASTRONOMY AND ASTROPHYSICS↗
NASA NTRS1994

Proton exchange membrane fuel cells for space and electric vehicle applications: From basic research to technology development

The proton exchange membrane fuel cell (PEMFC) is one of the most promising electrochemical power sources for space and electric vehicle applications. The wide spectrum of R&D activities on PEMFC's, carried out in our Center from 1988 to date, is as follows (1) Electrode Kinetic and Electrocatalysis of Oxygen Reduction; (2) Optimization of Structures of Electrodes and of Membrane and Electrode Assemblies; (3) Selection and Evaluation of Advanced Proton Conducting Membranes and of Operating Conditions to Attain High Energy Efficiency; (4) Modeling Analysis of Fuel Cell Performance and of Thermal and Water Management; and (5) Engineering Design and Development of Multicell Stacks. The accomplishments on these tasks may be summarized as follows: (1) A microelectrode technique was developed to determine the electrode kinetic parameters for the fuel cell reactions and mass transport parameters for the H2 and O2 reactants in the proton conducting membrane. (2) High energy efficiencies and high power densities were demonstrated in PEMFCs with low platinum loading electrodes (0.4 mg/cm(exp 2) or less), advanced membranes and optimized structures of membrane and electrode assemblies, as well as operating conditions. (3) The modeling analyses revealed methods to minimize mass transport limitations, particularly with air as the cathodic reactant; and for efficient thermal and water management. (4) Work is in progress to develop multi-kilowatt stacks with the electrodes containing low platinum loadings.

Srinivasan, Supramaniam↗
DOE OSTI2023

The STAT7 Code for Statistical Propagation of Uncertainties in Steady State Thermal Hydraulics Analysis of Plate Fueled Reactors

The STAT7 software was developed to perform steady-state thermal hydraulic analyses. Application of the software is for non-power research and test reactors, including conversion to low-enriched uranium fuel of U.S. High-Performance Research Reactors such as MITR-II. Since it can be necessary to repeat analysis during fuel reloading, STAT7 accommodates flexibility in analyzing many realistic aspects of reactor fuel management. STAT7 uses a Monte Carlo approach to model common fabrication parameters and other key reactor analysis uncertainties required for research and test reactor thermal hydraulic analyses. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings on avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one lateral stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis can be performed by analyzing the margin to onset of nucleate boiling and onset of flow instability for each axial node of each stripe of each plate of each element in the core.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Implications of point defect accumulation on UO 2 thermal conductivity and fission gas release under accelerated fuel irradiation

Evaluation of thermal properties is a crucial factor for nuclear fuel performance. During reactor operation, the accumulation of fission products and irradiation-induced lattice defects are responsible for degradation in thermal conductivity. Consequently, it affects fuel temperature and fission gas release (FGR) among other Multiphysics processes important for economics and safety analysis. We analyze the implications of point defects (PD) accumulation described using a rate theory (RT) Model on lattice thermal conductivity of UO 2 . Here, we demonstrate that fission rate-dependent point defect concentrations have the largest impact on in-pile thermal conductivity in the periphery of light water reactor fuels below a temperature threshold governed by the migration barrier of defects. Our analysis provides a mechanistic description of this phenomena which current fuel performance codes treat empirically. The reduction of thermal conductivity in the low -temperature rim region acts as additional thermal resistance and leads to a temperature notably larger than suggested by Lucuta thermal conductivity correlation. These effects are anticipated to have notable impacts when fuels are exposed to accelerated radiation. The impact of such point defect-informed treatment of thermal conductivity on fuel performance is evaluated by a detailed analysis of fission gas behavior and its release. We consider several models capturing different stages of fission gas bubble evolution and fission gas release (FGR). Finally, a new fission rate-dependent correction to the Lucuta correlation is proposed. The results show a significant reduction in thermal conductivity at the fuels’ periphery and an increase in fuel centerline temperature specifically at low burnups. Ultimately a modified LC shows a higher FGR compared to the original LC, while the acceleration process results in a reduction in overall FGR.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Critical and Subcritical Californium Source Driven Noise Analysis Experiments with Fresh PWR Fuels Pins

These experiments were performed in 1983 between August 1 and September 16 and used 35 operational days of critical facility time at the Babcock & Wilcox critical facility in Lynchburg, Va. These measurements were to assess the capability of the Cf source driven noise analysis method to measure the subcriticality of light water reactor fresh fuel configurations. This report documents the experiments of 1983 that were not reported at that time. These measurement with 2.459 wt. % 235 U enriched uranium oxide fresh fuel pins started at the delayed critical configuration of the fully assembled pressurized water reactor (PWR) core configuration with 1511 parts per million (ppm) natural boron (19.8 wt. % 10 B) in the water. This was a cylindrical configuration of 4961 fuel pins. After the initial measurements at delayed criticality, subcritical measurements by the Cf source driven noise analysis method were performed in which the number of detectors and Cf source locations were varied. Additional boron was added to the water moderator and reflector to reduce the neutron multiplication factor, with resulting boron concentrations of 1561, 1613, 1765, 1880, 2104, 2384, 2975, 3606, and 4303 ppm. After the completion of these measurements the increased, the boron concentration was returned to that at delayed critical (1511 ppm) and the reactivity was decreased by reducing the number of fuel pins in steps to 3713, 2553, 1281, 749, and 333 fuel pins and then to that (289) of a s ingle square 17 x 17 pin array, corresponding to a single PW fuel element. For all measurement that started at delayed criticality and then reduced the neutron multiplication factor to various subcritical states, inverse kinetics rod drop measurements at the higher reactivities were also performed but not reported here. At all subcritical states the break frequency noise analysis data was fitted to obtain the prompt neutron decay constant. Some data presented in this report are from notes not in the logbook. The purpose of this report is to document the experimental information for the measurements performed so that a later date someone could perform the required uncertainty and calculational analyses and documentation to use these data for an International Nuclear Criticality Safety Benchmark Program (INCSBEP) or Nuclear Energy Agency (NEA) benchmark. The data from these measurements is available from the Laboratory Records Department of Oak Ridge National Laboratory and the logbook is available from INCSBEP at Idaho National Laboratory.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗