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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 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Methods and Tools to allow molecular flow simulations to be coupled to higher level continuum descriptions of flows in porous/fractured media and aerosol/dust dynamics

The purpose of this project was to develop methods and tools that will aid in safety evaluation of nuclear fuels and licensing of nuclear reactors relating to accidents.The objectives were to develop more detailed and faster computations of fission product transport and aerosol evolution as they generally relate to nuclear fuel and/or nuclear reactor accidents. The two tasks in the project related to molecular transport in nuclear fuel and aerosol transport in reactor vessel and containment. For both the tasks, explorations of coupling of Direct Simulation Monte Carlo with Navier-Stokes solvers or the Sectional method were not successful. However, Mesh free methods for the Direct Simulation Monte Carlo method were successfully explored.These explorations permit applications to porous and fractured media, and arbitrary geometries.The computations were carried out in Mathematica and are fully parallelized. The project has resulted in new computational tools (algorithms and programs) that will improve the fidelity of computations to actual physics, chemistry and transport of fission products in the nuclear fuel and aerosol in reactor primary and secondary containments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

High-Fidelity CFD Modeling of Cryogenic Hydrogen Isotope Extrusion for Fusion Reactor Pellet Fueling

This study investigates the extrusion processes of deuterium and protium using ANSYS-Polyflow. The geometries and computational fluid dynamics (CFD) settings closely replicate the experimental setups and data acquired from the extruder experiments at Oak Ridge National Laboratory (ORNL) for validation purposes. We explore the impacts of (1) slip versus non-slip boundary conditions and (2) the use of constant, temperature-, and shear rate–dependent viscosities, concluding that the implementation of non-slip wall boundary conditions combined with shear rate–dependent viscosity produced more accurate predictions. The simulations achieved excellent agreement with the experimental data, with relative differences of only 5% for deuterium, and 3% to 6% for protium. This is the first time that experimental extrusion data at ORNL have been accurately predicted through high-fidelity CFD modeling. In conclusion, the advancements offer valuable insights and a foundational modeling tool for optimizing pellet injectors for ITER and other future reactor-scale devices.

ANSYS-Polyflow↗
NASA NTRS2013

Deleterious Thermal Effects due to Randomized Flow Paths in Pebble Bed, and Particle Bed Style Reactors

Reactor fuel rod surface area that is perpendicular to coolant flow direction (+S) i.e. perpendicular to the P creates areas of coolant stagnation leading to increased coolant temperatures resulting in localized changes in fluid properties. Changes in coolant fluid properties caused by minor increases in temperature lead to localized reductions in coolant mass flow rates leading to localized thermal instabilities. Reductions in coolant mass flow rates result in further increases in local temperatures exacerbating changes to coolant fluid properties leading to localized thermal runaway. Unchecked localized thermal runaway leads to localized fuel melting. Reactor designs with randomized flow paths are vulnerable to localized thermal instabilities, localized thermal runaway, and localized fuel melting.

Moran, Robert P.↗
NASA NTRS2013

Deleterious Thermal Effects Due To Randomized Flow Paths in Pebble Bed, and Particle Bed Style Reactors

A review of literature associated with Pebble Bed and Particle Bed reactor core research has revealed a systemic problem inherent to reactor core concepts which utilize randomized rather than structured coolant channel flow paths. For both the Pebble Bed and Particle Bed Reactor designs; case studies reveal that for indeterminate reasons, regions within the core would suffer from excessive heating leading to thermal runaway and localized fuel melting. A thermal Computational Fluid Dynamics model was utilized to verify that In both the Pebble Bed and Particle Bed Reactor concepts randomized coolant channel pathways combined with localized high temperature regions would work together to resist the flow of coolant diverting it away from where it is needed the most to cooler less resistive pathways where it is needed the least. In other words given the choice via randomized coolant pathways the reactor coolant will take the path of least resistance, and hot zones offer the highest resistance. Having identified the relationship between randomized coolant channel pathways and localized fuel melting it is now safe to assume that other reactor concepts that utilize randomized coolant pathways such as the foam core reactor are also susceptible to this phenomenon.

Moran, Robert P.↗
DOE OSTI2021

A simulation-based parametric study of CLOU chemical looping reactor performance

Chemical looping with oxygen uncoupling (CLOU) is a variant on chemical looping combustion in which the oxygen carrier releases gaseous O 2 in the fuel reactor, making it well-suited for solid fuels, since the released gaseous O 2 readily reacts with solid char. This work presents several computational fluid dynamic (CFD) simulations of copper-based CLOU in a dual fluidized bed system, each with different operating conditions. The modeling predicted that coal particle sizes as large as 1000 μm did not significantly affect performance. Increased oxygen carrier copper loading resulted in an excess of gaseous oxygen in the product gas stream. Decreasing the oxygen carrier bed mass as well as reducing the air reactor fluidizing velocity did not supply enough oxygen to the fuel reactor to complete combustion of the coal. This generated a failure state in which the temperature continued to decrease in the fuel reactor from the lack of combustion, which in turn reduced the O 2 equilibrium partial pressure, further lowering the amount of combustion possible. Sufficient O 2 can be maintained in the fuel reactor by ensuring a high enough air reactor velocity and a large enough supply of oxygen carrier inventory to handle the chosen coal feed rate.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Preliminary design of control rods in the single-fluid double-zone thorium molten salt reactor (SD-TMSR)

Recent studies on Molten Salt Reactors (MSRs) showed that the excess reactivity at the beginning of the operation is large for many fueling strategies and must be compensated by a reactivity control system. The current work introduces a reliable safety system based on control rods in addition to the online feed system reactivity control in the Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR). Three different initial fissile loadings are considered: 233 U, reactor-grade Pu, and transuranic (TRU) elements as a startup fuel. We applied six different absorbing materials to investigate the main operational and safety parameters in the SD-TMSR: natural B 4 C, enriched B 4 C with 90% 10 B, HfB 2 , HfH 1.62 , Eu 2 O 3 , and Gd 2 O 3 . The present work focuses on control rod design, integral and differential control rod worth, shutdown margin, and shadowing effects at steady-state. We employed the SERPENT-2 Monte-Carlo code to calculate the reactivity worth and analyze the performance of the reactivity control system. We showed that 233 U and reactor-grade Pu startup cores maintain adequate shutdown margin with all considered absorbers. Finally, this report proposes a design of control rod clusters that compensate the excess reactivity of the SD-TMSR loaded with different initial fissile material.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Benchmark of Spacer Grid Models used in Nuclear Power Plants

The fuel bundles of both PWR and BWR reactors require spacers to maintain the integrity and spacing of the slender fuel rods. These spacers can provide benefits such as increased mixing and turbulence, which increases heat distribution, leading to better Departure from Nucleate Boiling (DNB) conditions in PWRs. They also enhance the critical power ratios by increasing the thickness of the liquid layer in BWRs. However, the spacers also cause a pressure drop which must be engineered into the performance and operation. The flow within a rod bundle involves a number of complicated factors, such as cross flow, (de)entrainment of liquid layer, bubble or droplet break up/join. This requires specialized 3 dimensional computational fluid dynamics for the complete flow, however, most of the newly developed models have been accomplished by industry, so they are proprietary and in addition require extensive programming and cost to implement. In the public domain, a number of 1 dimensional equations have been developed that could benefit in initial design and decision making, if the equations are valid in the new and different applications the models are used. In 2002, an OECD/NRC BFBT benchmark workshop determined the variables used in the development of models, to predict flow within nuclear fuel bundles, had not been updated with technological advances to replace general theory and characteristics. If those properties could be identified and modeled, they could be applied to a wider range of conditions. This study utilizes NUPEC BFBT tests' measured data, on an average spacer over a range of pressure and flows, which was determined to be the most detailed measured data available. The focus is to describe the mathematical models of spacer grids and validate them with experimental results. This study consists of compiling and comparing available and established single phase equations and models computations. While these models may not have the desired degree of accuracy for final decisions, they may be accurate enough for the initial phases of design. No conclusion on a universal equation that fits all boundary conditions accurately or detailed effect of properties was determined. However, by using the BFBT experimental data, boundary conditions where the spacer grid models can be useful were identified. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Method advancements for fuel assessments during seismic and LOCA events

The evaluation of fuel assembly structural responses of light-water reactors to seismic and LOCA excitations has received increased attention in the last five years. Real seismic events such as those that affected the Fukushima Daiichi and North Anna plants in 2011 have caused the industry to more closely examine our preparedness for seismic events. Furthermore, the review of design certifications for new plants has raised questions about the methods used to evaluate these events and is leading to revised guidance from regulators (e.g. US NRC Information Notice 2012-09). As a result, the industry is seeing fuel structural design margins based on current methods begin to erode for seismic and LOCA events. This paper will review points for consideration as the industry responds to these changes. To begin, it is worthwhile for the industry to understand the history and the critical thought behind the current methods such that improved criteria and methods can be formulated. This review identifies three known areas of weakness in the current framework for this methodology. First, a broader understanding of spacer grid behavior is needed. Spacer grid behavior in response to impact loads is of paramount importance, but the industry needs to re-frame our characterization of this behavior in order to focus on the actual critical safety function of this component. Secondly, despite guidance from the Standard Review Plan, the effects of irradiation, on spacer grid characteristics, in particular, must be considered. Thirdly, the area of fluid-structure-interaction may be the greatest frontier to conquer in this methodology and may provide our best possibility to recover eroded margins. However, fluid-structure-interactions must be cautiously evaluated and applied with conservatism. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Characterization of Safeguards Implementation Schedules for Advanced Reactors

Advanced reactor designs differ from the traditional light-water reactor designs and operational requirements in nearly every facet, including the heat transfer fluids, operating temperatures and pressures, and fuel cycle requirements. To account for these differences, significant effort will be required to ensure appropriate safeguards implementation for US-based reactors intended for global deployment. This is the second in a series of three reports that examine facets of this issue. The first paper (ORNL/TM-2021/2142) examines potential construction schedules for advanced reactors to create a baseline understanding of the timeframes involved. This second report characterizes International Atomic Energy Agency (IAEA) safeguards implementation schedules for potential advanced reactors. Finally, the third report (ORNL/TM-2021/2208) examines existing gaps for safeguards implementation at advanced reactors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Fluoride-Cooled High-Temperature Pebble-Bed Reactor Reference Plant Model Updates

This work presents the latest improvements to, and investigations performed with, the Fluoride-Cooled High-Temperature Pebble-Bed Reactor reference plant models for the United States Nuclear Regulatory Commission. These models, developed with the Comprehensive Reactor Analysis Bundle, or BlueCRAB, serve as the foundation for the future development of detailed design evaluation models based on license applications. BlueCRAB is the code suite proposed for non-light-water reactor systems safety analyses, and it incorporates various simulation tools developed by the Nuclear Energy Advance Modeling and Simulation program, including the Griffin code for reactor physics, the Pronghorn and SAM codes for core thermal fluids, the BISON code for solid conduction and fuel performance, and the SAM code for system analysis. The primary objective of this work is to assess the level of readiness of BlueCRAB for modeling fluoride-cooled high-temperature pebble-bed reactors. We first developed numerical models in BlueCRAB that include the key physics for this technology, ensuring an adequate level of fidelity for modeling the core performance during accident scenarios. This was followed by simulation of transient scenarios, two loss-of-forced-cooling events (one protected and one unprotected), and two control rod withdrawal events (one delayed and one prompt supercritical reactivity insertion). The analysis includes comparisons between the 2-D thermal fluid porous media models in Pronghorn and SAM, comparisons between coupled Pronghorn-Griffin and coupled SAM-Griffin models for two loss-of-forced cooling events and one control rod withdrawal event, and comparisons between SAM single-solve and domain-overlapping approaches for multi-scale thermal fluid coupling. In addition, we performed comparisons between 3-D, 2-D, and 0-D neutronic models for the two control rod withdrawal scenarios with Pronghorn-Griffin. The results show that the BlueCRAB models led to physically intuitive solutions for the scenarios examined. The changes in the various scalar and vector fields such as the neutron flux, power, temperatures, densities, pressures, and velocities are all within the expected ranges, and their distributions can be explained from the system response of the transients and the geometric and material variations. Several comparisons suggest that the porous media models in Pronghorn and SAM can lead to similar solutions, even though they are based on different methodologies. The simulations demonstrate that there are differences between the various levels of fidelity, and it is advisable to have flexible tools that can cover the breadth and depth of needs that may arise in future technical evaluations. We believe that BlueCRAB’s capabilities represent a significant asset for confirmatory analyses aimed at resolving important safety questions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Computational Analysis of Different Sparging Systems and their Influence in the Fluid-Dynamic Behavior of Bubble Column Reactors

Bubble column bioreactors are being actively considered for gas fermentation applications, specifically for CO2 utilization, and sugars to fuels conversion. Their main advantages include good mass transfer without any moving parts and low-cost of operation and maintenance. However, the design and scale-up of such reactors is challenging specifically for carbon capture applications where a mixture of gases (e.g. CO2/CO/H2) with variable solubilities is used. The overall performance of scaled-up bioreactors (e.g., mass transfer rate) is largely affected by gas holdup, bubble size distribution (BSD), and multiphase hydrodynamics. We investigate the effect of gas sparger designs on the performance of these large-scale bioreactors using computational fluid dynamics simulations in this work, so as to improve CO2 conversion at scale. The gas distribution systems in bubble column reactors not only determines operational regime, but also affects the evolution of the BSD, which in turn influences interfacial mass transfer and ultimately the efficiency of the gas-liquid exchange process. In addition to the BSD, uniformity in gas sparging affects gas holdup and bubble residence time which constitute important metrics of performance in gas-liquid systems. In this work, we use computational models to simulate high fidelity representations of different sparger designs and their effect on the operation of a bubble column reactor. Four different types of spargers have been selected for the computational study (Fig. 1): ladder, multi-ring, single-ring and toroidal. Their effect on superficial velocity, gas holdup mixing efficiency, and BSD will be evaluated in this work. The model uses a multiphase Eulerian framework similar to [1] and include a composition of mixtures of H2/CO/CO2 gases, common in fermentation applications.

BIOMASS FUELS,MATHEMATICS AND COMPUTING↗
DOE OSTI2022

Safeguards for the Lithium Fluoride Thorium Reactor: A Preliminary Nuclear Material Control and Accounting Assessment

This report was prepared by Oak Ridge National Laboratory (ORNL) for Flibe Energy Inc. (FEI), a US-based advanced reactor company founded in 2011 and headquartered in Huntsville, Alabama. FEI is developing the lithium-fluoride thorium reactor (LFTR), which is a modern two-fluid molten salt reactor (MSR) design operating on a thorium/ 233 U fuel cycle. FEI intends for LFTR to become a self-sustaining clean energy source that can create or breed its own fuel from thorium. Each LFTR is intended to breed enough fissile material to compensate for the amount it consumes. Consequently it would not require fissile replenishment during its operational lifetime. This self-sustaining nature would eliminate the need for uranium enrichment infrastructure to support LFTRs after the first generation.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Aerosol Particle Deposition on a Spent Nuclear Fuel Assembly Spacer Grid

The flow and particle deposition patterns on surfaces in an idealized spacer grid for a 17x17 pressurized water reactor (PWR) assembly in a spent fuel canister are modeled using computational fluid dynamics (CFD) with laminar flow. The effects of gravitational settling, non-Stokesian flow, and particle slip are first rigorously analyzed. From the analysis, non-Stokesian effects and slip may be neglected for the particle sizes and conditions expected in a canister. For particles that do not settle out, a swirling flow pattern at the corners of a spacer grid channel directs particles to the leeward side of the flow vanes where much of the deposition occurs. Particle deposition increases with increasing particle diameter. Deposition also increases with decreasing flow velocity as this provides more time for particles to settle and deposit on the leeward side of the flow vanes. The fraction of particles that are transmitted through a spacer grid is determined as a function of inlet gas velocity and particle diameter by running the CFD calculation for each set of conditions and for each particle diameter. Curve fits of the transmission curve as a function of particle diameter for a specified spacer grid and flow velocity are applied to a lognormal particle mass density function for the inlet particles. The resulting mass density function and aerosol mass fraction that passes through the spacer grid can be determined analytically without resorting to numerical iteration. A sample calculation of the analytical solution is demonstrated for a lognormal particle mass density function.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1992

Nuclear thermal rockets - Key to moon-Mars exploration

The Space Exploration Initiative (SEI) calls for lunar and Martian exploration missions for which solid-core nuclear thermal rockets (NTRs), in virtue of their single-stage, fully-reusable nature, are ideally suited. NTRs promise double the specific impulse of chemical propulsion. A lunar mission employing a reusable NTR is currently being conducted by NASA. The NTR would be assembled in LEO in such a way that it remained 'radioactively cold' during earth-to-orbit deployment by a heavy-lift chemical booster, and therefore presented no radioactive hazard. Also under consideration is a particle-bed reactor in which the hydrogen propulsive fluid directly cools coated-particle fuel spheres.

Borowski, Stanley K.↗
DOE OSTI2016

Study of Melt Formation and Behavior of a Two-Component Debris Bed in LIVE Facility

The main objective of the LIVE program at Karlsruhe Institute of Technology (KIT) is to study the melt pool behavior in the reactor pressure vessel (RPV). The study is carried out in large-scale 2D and 3D facilities and in supporting separate-effects tests in order to provide a reasonable estimate of the remaining uncertainty. Within the LIVE experimental program several tests have been performed with water and with non-eutectic melts (mixture of KNO{sub 3} and NaNO{sub 3}) as simulant fluids to study the heat flux distribution in the conditions when the melt pool is covered by water from the top. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Application of computational fluid dynamics for the simulation of cryogenic molecular sieve bed absorber of hydrogen isotopes recovery system for Indian LLCB-TBM

One of the most challenging tasks in the design of the fuel cycle system lies in the effective design of Tritium Extraction System (TES) which involves proper extraction and purification of tritium in the fuel cycle of the fusion reactor. Indian Lead Lithium cooled Ceramic Breeder Test Blanket Module (LLCB-TBM) would extract hydrogen isotopes through Cryogenic Molecular Sieve Bed (CMSB) adsorber system. A prototype Hydrogen Isotopes Recovery System (HIRS) is being developed to validate the concepts for tritium extraction by adsorption mass transfer mechanism. In this study, a design model has been developed and analyzed to simulate the adsorption mass transfer kinetics in a fixed bed adsorption column. The simulation leads primarily to effective design of HIRS, which is a state-of-the-art technology. The paper describes the process simulation approach and the results of Computational Fluid Dynamics (CFD) analysis. The effects of different operating conditions are studied to investigate their influence on the hydrogen isotopes adsorption capacity. The results of the present simulation study would be used to understand the best optimized transport phenomenon before realizing the TES as a system for LLCB-TBM. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗