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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 1,189 records · Page 66
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

An Efficient Tool-Set for Modeling Containment Ventilation in Very Complex Facilities - 17215

The applicability of a multi-zone ventilation analysis program, CONTAM, to complex radioactive waste material handling or processing plants is evaluated. Validation results are based on simulations from three-dimensional multi-physics computational fluid dynamics. Insight into the application limits of CONTAM and conditions under which the use of higher fidelity tools are required are also discussed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

A Dissipation-Based Method for Improving the Accuracy of Computational Fluid Dynamics Simulations of High Level Non-Newtonian Waste - 17351

Analysis of shear rate in dissipative scales of turbulence was performed using a quasi-direct numerical simulation (Q-DNS) of flow through a pipe involving Bingham fluid. Dependency of deformation rate on dissipative scales of turbulence was studied and a modification algorithm was developed and implemented in Reynolds-averaged Navier-Stokes (RANS) modeling of the same flow using the standard k-e model. It was shown that the modification was effective in increasing accuracy of the RANS simulations when the velocity profile was modified to match experimental data. In both the Q-DNS and RANS simulations, the Herschel-Bulkley (HB) rheology equation was used to relate shear stress of the non-Newtonian fluid to shear deformation. Additional analysis of dependency of shear on the turbulent dissipation rate (TDR) was performed using various flow conditions in RANS simulations with a modified viscosity model (alpha model). In the past, the alpha model has been shown to produce accurate simulations for the same flow conditions. Results of the analysis shows that a qualitative agreement exists between shear dependency in the RANS-alpha and Q-DNS-HB simulations when the ratio of velocity to length scale is used. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Development of Transient Coupled Chemical-Thermal-Fluids Simulation for Unsealed Aluminum-Clad Spent Nuclear Fuel Storage Canister - 19273

This study examines a preliminary three-dimensional (3D) multi-physics computational fluid dynamics (CFD) model of the unsealed canisters containing aluminum-clad spent nuclear fuel from INL's ATR in the irradiated fuel storage facility. This 3D CFD model considers fully resolved fuel plates in an unsealed canister with small mass exchange with surrounding air as time evolves in the simulation and contains a network of chemical reactions to consider due to several of the side reactions occurring in the presence of water vapor, nitrogen, oxygen and carbon dioxide rather than in helium backfilled DOE standard canisters. Prior to the CFD modeling, the full set of chemical reactions under these conditions was analyzed, reduced, and refit to a more manageable number of equations for the computational efficiency of the 3D CFD simulation. To deal with the long time-scale of interest for tracking the possible buildup of trace species such as nitric acid, NOx, and hydrogen, a segregated time stepping algorithm is used, wherein the velocity field is only updated periodically, and assumed to be in a pseudo steady state. This study first examines the theory of the model, workflow to establish radiolysis reaction network, and initial simulations of the evolutions of thermal fields, hydrogen gas and nitric acid concentrations within unsealed canisters in conditions relevant to the INL INTEC CPP-603 facility over a long period of time. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Melter Modeling over a Range of Scales to Support Vitrification at the WTP - 19664

Computational modeling is a powerful complement to laboratory experiments and testing of systems at various scales. When properly applied and validated, computational fluid dynamics (CFD) can provide information not readily obtainable by physical testing, especially in cases where the operating environment does not permit visual observation or the insertion of typical data acquisition instrumentation as in the case of waste glass vitrification. A series of CFD and heat transfer models over a range of scales are being developed to support melter operations at the Hanford Waste Treatment and Immobilization Plant. This paper provides a description of the models and the resulting information they provide that can be used to improve melter throughput, as well as aid in evaluating and resolving operating plant issues or upsets. This work is important for the WTP operation by providing a potential for model-predictive control of the melter operation. Increases in the rate of glass production and avoiding idling can significantly decrease the duration and cost of the vitrification campaign, potentially saving billions of dollars. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Computationally Aided Design of a Small-Scale Waste-Glass Melter - 20452

To provide mission support to the WTP for the vitrification of legacy nuclear tank waste at the Hanford site, a reduced-scale vitrification pilot system is being designed to process simulated and actual radioactive tank wastes. The waste stored in underground tanks will be separated into HLW and LAW fractions, where the majority is LAW by mass (approximately 90%) and HLW by activity (greater than 95%). The first tank waste to be processed at the WTP will be LAW. Because Hanford tank waste and the resultant melter feed compositions are known to vary widely, pilot-scale operations are essential to identify potential problems, provide needed data, confirm assumptions, and determine impacts to full-scale melters and off-gas systems from these different feeds. In addition, the reduced-scale melter system must be capable of quickly providing results to operations without the typically high cost of full-scale radioactive operations and to provide an engineering platform in close proximity to local operations staff. The objectives of this computational modeling effort are to produce process conditions like those encountered in the WTP LAW melters and to minimize the volume of radioactive waste necessary for the evaluations. To this end, melter design features are being explored to increase production without significantly increasing surface area or melter-glass volumes or compromising data quality. In support of these objectives, a set of computational fluid dynamic models was developed to provide insight into the flow patterns within these small melters and evaluate design features to increase production without significantly increasing glass inventory. Computed velocities were evaluated at the interface between the cold cap and the molten-glass pool and within the bulk-glass pool. Since melt rate correlates with the velocity underneath the cold cap, this information can be used to select design features that produce the highest gains. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Development of CFD Flow Analysis for Nuclear Facility Ventilation Systems

Understanding and modeling of particulate transport has importance related to safety in ventilated environments that contain hot cells, gloveboxes, and fume hoods. In particular, good modeling of the airflow in these spaces is required to understand areas of stagnation where particles drop out of the free stream and accumulate. This accumulation of particulate can cause either contamination and/or higher than acceptable local dose levels. It is known that multiple sites across the DOE complex take credit for hot cells, gloveboxes, and/or hoods in their safety basis for providing a defense-in-depth benefit for both onsite and offsite releases. By providing confinement of radioactive materials, such features serve to reduce direct doses to facility workers and mitigate the consequences to the environment due to an uncontrolled release. Each of these features has access points that interface with the personnel space. Understanding how air flow behaves at these access points is of great interest to those performing hazard analysis. Computational Fluid Dynamics (CFD) can be used to analyze and model the flows in hot cells and glove boxes as a way to confirm operation. While CFD capability continues to advance, there are still important modeling assumptions that are left to the analyst's discretion. These are most notably the choice of the turbulence models, mesh structure, and wall boundary condition assumptions used in the model. In this work, we apply CFD in modeling airflow through an occupiable work area of the Alpha Gamma Hot Cell Facility (AGHCF) at Argonne National Laboratory (ANL). This modeling is then compared to field test data to both inform and verify the modeling assumptions. Based on the results of field tests, modeling assumptions and boundary conditions are refined and the process is repeated until the results are found to be reliable with a high level of confidence.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Evaluation of the shear rate effect on the interfacial forces acting on a single bubble

Owing to the outstanding characteristics of mass and heat transfer efficiency, two-phase bubbly flows have wide applications in the nuclear engineering fields. However, the physics behind the bubbly flows phenomenon can be very complex due to the various parameters and phenomena involved. Thanks to the increasing computing power and advanced algorithms, computational fluid dynamics (CFD) approach has emerged as an effective tool for exploring the bubbly flows. In typical laminar channel or pipe flow, the flow field develops a parabolic velocity profile due to the wall friction. Assuming the shear rate across the bubble diameter is constant, the bubble will encounter various shear rates depending on its transverse position. Different shear rates may lead to the acceleration of transverse migration speed and influence the bubble rising velocity. In this summary, we evaluate the effect of shear rate on the interfacial forces (lift and drag) acting on a single bubble with the shear rate ranging from 3.8 to 8.3 s{sup -1} and the Eotvos number ranging from 2.58 to 4.45.

42 ENGINEERING↗
DOE OSTI2017

CFD Simulation of Bubble Growing Under a Downward Facing Surface

The heat removing of downward facing surface is an important issue in many areas, including nuclear safety, solar storage system and electronic cooling devices. Therefore, the characteristic and heat removal capability of downward facing boiling were investigated by many experiments in recent years. Among these studies, the thickness of bubble is considered as an important parameter. The increasing of bubble thickness will increase the thermal resistance of cooling process and reduce the cooling ability. In order to focus on the dynamics of bubble thickness, Das et al. injected air into sodium dodecyl sulphate (SDS) solution through a nozzle under the horizontal downward facing surface. The bubble will be formed by the air injection and kept under the horizontal plate. The diameter of nozzle, horizontal plate and SDS solution tank are 2 mm, 20 mm and 30 mm, respectively. The growing processing and the stable thickness of the bubble were recorded in the experiment. Basing on the conditions of Das's experiment, the appropriate computational fluid dynamics (CFD) simulation was conducted. The growing process and the stable thickness of bubble under difference contact angle are predicted by CFD in this study.

42 ENGINEERING↗
DOE OSTI2017

Time-Resolved PIV/PTV Measurements on Interior Subchannels of a Wire-Wrapped 61-pin Hexagonal Fuel Bundle

Liquid metal fast reactor fuel bundles using sodium as a coolant typically utilize a tightly packed triangular lattice of fuel pins enclosed in a hexagonal duct. Often, the fuel pins are helically wrapped with a wire spacer. The primary function of the wire spacers is to maintain a gap between adjacent fuel pins. They also enhance sub-channel mixing to increase convective heat transfer and mitigate vortex-induced vibration. Texas A and M University has conducted isothermal flow experiments in a wire-wrapped 61-pin hexagonal fuel bundle to support the research on advanced nuclear fuel development sponsored by the Department of Energy (DOE). These tests produced high spatial and temporal resolution measurements of the velocity field and pressure at different locations in the fuel bundle. The experimental facility employs matched-index-of-refraction (MIR) techniques and laser diagnostic velocity measurement techniques such as particle image velocimetry (PIV), particle tracking velocimetry (PTV), and laser Doppler velocimetry (LDV). The result of this experimental activity will be an experimental database of pressure and velocity field measurements that will be suitable for validating computational fluid dynamics (CFD) methods, such as Reynolds-averaged Navier-Stokes (RANS) and highly resolved large eddy simulation (LES). PIV results on the set of edge subchannels near the hexagonal duct have been presented in another paper. This paper presents PIV results on two sets of interior subchannels, defined as Interior- 1 and Center-2, at a Reynolds number (Re) of 19,000 in the turbulent regime.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

The Flow Visualization for the Separation of Air-water Bubbly Two-phase Flow Through a Vertical T-junction

T-junction is a typical structure in many industrial systems such as power plants, chemical process, and hydrocarbon production industries. The separation of two-phase flow passing through T-junction is a common flow phenomenon. There are lots of investigations regarding two-phase flow separation in branching tees and wyes tees. The characteristic of the two-phase flow between the two outlets are always non-uniform, and this kind of fluid status may contribute to the unpredictable damage of downstream facilities. For the nuclear power plants, severe accidents analysis is important especially for Light-Water Reactors (LWR) when Loss of Coolant Accident (LOCA) occurs. The small break of the pipe may cause the coolant become two-phase flow due to the pressure drop. Under certain conditions, the behavior of T-junction is similar to a gas-liquid separator, as a result of the gas enters the side arm and the liquid flows straight along the run arm. The partial separation of gas-liquid mixtures can reduce the load on the main separator, probably leading to smaller units. The study aims to establish complete visual observations for the benchmark of Computational Fluid Dynamics (CFD) model. As a consequence, a vertical T-junction (D = 26 mm) experimental loop was set up for the purpose of elaborating the complex two-phase fluid phenomena. In this experiment, variables were manipulated under highly controlled conditions and a photographic method was proposed to analyze the images of phase separation. Furthermore, the images obtained by camera have been discussed to receive valuable information about the two phase flow behavior around the T-junction.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Flow Characteristics of Intermediate Heat Exchanger with CFD Methodology on Nuclear Safety Application

The intermediate heat exchanger (IHX) plays an important role in the field of nuclear power plant, and its safety margin could strongly affected by the flow distribution of secondary flow inlet. As a result, the size of secondary inlet is crucial in thermal design. However, the IHX contains thousands of heat exchanger tubes, and it takes larger time to simulate each tube accurately. Other than that, there are more delicate devices in a nuclear power plant, so it would take too much time to simulate actual model. Although system code could solve problems effectively, it could not detect local deformation. On nuclear safety application, how to save time and receive specific results in an efficient way is necessary. Therefore, the porous media model and the heat exchanger model from computational fluid dynamics (CFD) were adopted in this study. The methodology not only can save much time than actual model but also receive more specific results than system code. To ensure the accuracy, validation of the methodology is also presented in this study by modeling RHR heat exchanger of L nuclear power plant. The results of the simulation were compared to experimental values from L nuclear power plant and receive small errors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

A non-linear k-{epsilon} turbulent model sensitivity analysis for flow across a PWR spacer grid with mixing vanes

In this work we studied the sensitivity of different computational fluid dynamics (CFD) numerical solutions on turbulence model, domain discretization mesh type, and inlet boundary conditions. The case under examination is a turbulent one-phase flow (water) through a scaled nuclear pressurized water reactor (PWR) fuel assembly. The main feature of this geometrical configuration is the presence of a spacer grid with mixing vanes. The design of a spacer grid with mixing vanes plays a key role for the core heat transfer efficiency, increasing the coolant heat transfer coefficient. The main reason for this is the introduction of the top mixing vanes, which consist of an ordered repetitive succession of wings that generate secondary swirling flows. The increase of heat transfer efficiency also increases the critical heat flux (CHF), making it possible to increase the design and operational limits, far from boiling crises (departure of nucleate boiling in case of a PWR). The main focus of this work was to analyze the simulation capabilities of Reynolds-averaged Navier-Stokes (RANS) models (a cubic non-linear k-epsilon model formulation) for this complex flow geometry. The CFD code used was CD-Adapco STAR-CCM+.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Validation of the Computational Fluid Dynamics Method using the Aboveground Configuration of the Dry Cask Simulator

Dry cask applicants submit their spent nuclear fuel dry storage and transportation cask designs to the U.S. Nuclear Regulatory Commission (NRC) for certification under Title 10 of the Code of Federal Regulations (10 CFR) Part 72, 'Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste'. The NRC staff performs its technical review of these designs in accordance with 10 CFR Part 72 and NUREG-1536, 'Standard Review Plan for Spent Fuel Dry Storage Systems at a General License Facility,' Revision 1, issued July 2010. To ensure that the cask and fuel material temperatures of the dry cask storage system remain within the allowable limits or criteria for normal, off-normal, and accident conditions, the NRC staff performs a thermal review as part of the technical review. Recent applications increasingly have used thermalhydraulic analyses using computational fluid dynamics (CFD) codes (e.g., ANSYS FLUENT) to demonstrate the adequacy of the thermal design. The NRC performs validation studies of the FLUENT CFD code to assist it in making regulatory decisions to ensure adequate protection for storage and transportation casks. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. In efforts to further validate these designs, the NRC in conjunction with the Department of Energy and Sandia National Laboratories (SNL) are conducting an experiment to capture the effects of elevated helium pressures and accurately portray the external convection of aboveground and belowground canister dry cask systems.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Thermal Evaluations for Hypothetically Drain-Down Spent Fuel Storage Facility at SRS

The L-Area Spent Nuclear Fuel (SNF) underwater storage facility at Savannah River Site (SRS), referred to as L-Basin SNF facility, has an array of Vertical Tube Storage (VTS) racks and a group of High Flux Irradiation Reactor (HFIR) fuel cores. The underwater facility is integrally designed as a wet SNF storage system with a chemically-controlled water pool. The SNF storage racks are seismically qualified. The L-Basin pool holds about 3.4 million gallons of water with concrete wall thickness of about 5 ft and water depths of 17 ft to 30 ft. The facility has mainly two different elevation levels of 30 ft and 17 ft below the ground level. At the level 30 ft below the ground, a maximum of 110 racks of VTS and 5 racks of Dry Cave can be stored. At the level of 17 ft below the ground, 120 HFIR cores can be stored, considering that each rack contains two HFIR cores. In case of a hypothetical accident, all cooling water in the storage pool may be drained away, resulting in dry SNF storage condition. The primary objective of the work is to estimate the maximum fuel temperature in a conservative manner when the L-Basin facility is completely dry due to the hypothetical drain-down of the pool water. A three-dimensional Computational Fluid Dynamics (CFD) modeling approach was taken for the thermal evaluations of the water-drained SNF storage facility. The modeling calculations were performed, assuming that all of the storage racks in L-Basin are fully loaded with SNF assemblies containing decay heat sources. In the case of full loading, the L-Basin facility stores 14,600 SNF assemblies in VTS racks and 120 cores in HFIR racks. The thermal analysis results of the L-Basin model yielded maximum temperatures of VTS and HFIR fuel assemblies for different fuel loading scenarios and boundary conditions of the SNF storage racks.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Numerical Simulation of Twin Jets using LES with the Nek5000 Code

The twin jet activity, a joint effort between Texas A and M University, University of Tennessee - Knoxville, and Argonne National Laboratory, is being used to analyze the performance of various computational fluid dynamics (CFD) code packages. The twin jet water facility (TJWF) is based off a semi-classical problem of combining planar turbulent jets in a large free-shear environment or enclosure. The problem is similar to physics seen in advanced reactor designs such as in the large upper plena of a sodium fast reactor. Utilizing the experimental data sets created using laser Doppler velocimetry (LDV) and particle image velocimetry (PIV), the Nek5000 code, among others, is qualitatively compared to the experimental work. This effort is a starting point for determining the appropriate simulation techniques needed for resolving the twin jet physics.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1990

Mechanical design problems associated with turbopump fluid film bearings

Most high speed cryogenic turbopumps for liquid propulsion rocket engines currently use ball or roller contact bearings for rotor support. The operating speeds, loads, clearances, and environments of these pumps combine to make bearing wear a limiting factor on turbopump life. An example is the high pressure oxygen turbopump (HPOTP) used in the Space Shuttle Main Engine (SSME). Although the HPOTP design life is 27,000 seconds at 30,000 rpms, or approximately 50 missions, bearings must currently be replaced after 2 missions. One solution to the bearing wear problem in the HPOTP, as well as in future turbopump designs, is the utilization of fluid film bearings in lieu of continuous contact bearings. Hydrostatic, hydrodynamic, and damping seal bearings are all replacement candidates for contact bearings in rocket engine high speed turbomachinery. These three types of fluid film bearings have different operating characteristics, but they share a common set of mechanical design opportunities and difficulties. Results of research to define some of the mechanical design issues are given. Problems considered include transient strat/stop rub, non-operational rotor support, bearing wear inspection and measurement, and bearing fluid supply route. Emphasis is given to the HPOTP preburner pump (PBP) bearing, but the results are pertinent to high-speed cryogenic turbomachinery in general.

Evces, Charles R.↗
NASA NTRS1996

Investigation of Gas Effects on Cryocooler Resonance Characteristics

Cryocooler thermal and vibrational performance is determined, fundamentally, by the dynamic interactions between the mechanical system and the working fluid. This paper explores the effect of working-fluid characteristics on the mechanical response of the cooler. Experimental data collected from two coolers characterized under the Jet Propulsion Laboratory's extensive program of cryocooler testing and characterization show that a classical single-degree-of-freedom spring-mass-damper model does not capture the full frequency dependence of the mechanical response. The data from the two modes of cooler operation (slosh and head-to-head) are used to motivate the explanation that working-fluid characteristics dominate at high frequencies, and mechanical system characteristics dominate at lower frequencies. Operating temperature is shown to be a significant factor in determining resonance behavior. Finally, the discussion provides a framework within which resonate parameters and cooler characteristics can be extracted from the experimental data.

cryocooler vibration working-fluid characteristics↗
NASA NTRS1992

Connector Mechanism Has Smaller Stroke

System for connecting electrical and/or fluid lines includes mechanism reducing length of stroke necessary to make or break connections. Feature enables connection and disconnection in confined space, and compensates for misalignment between connectors. Connector in active member moves upward at twice the speed of downward stroke of passive member. Stroke amplified within connector system. Applications include connections between modular electronic units, coupled vehicles, and hydraulic systems.

Milam, M. Bruce↗