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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 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
NASA NTRS2002

Momentum and Heat Flux Measurements in the Exhaust of VASIMR Using Helium Propellant

Electromagnetic thrusters typically use electric and magnetic fields to accelerate and exhaust plasma through interactions with the charged particles in the plasma. The energy required to create the plasma, i.e. ionization energy, is potential energy between the electron and ion. This potential energy is typically lost since it is not recovered as the plasma is exhausted and is known as frozen flow loss. If the frozen flow energy is a small fraction of the total plasma energy, this frozen flow loss may be negligible. However, if the frozen flow energy is a major fraction of the total plasma energy, this loss can severely reduce the energy efficiency of the thruster. Recovery and utilization of this frozen flow energy can improve the energy efficiency of a thruster during low specific impulse operating regimes when the ionization energy is a large fraction of the total plasma energy. This paper quantifies the recovery of the frozen flow energy, i.e. recombination energy, via the process of surface recombination for helium. To accomplish this task the momentum flux and heat flux of the plasma flow were measured and compared to calculated values from electrostatic probe data. This information was used to deduce the contribution of recombination energy to the total heat flux on a flat plate as well as to characterize the plasma conditions. Helium propellant was investigated initially due to its high ionization potential and hence available recombination energy.

Chavers, D. Gregory↗
DOE OSTI2017

TRACE Analysis of Loss of Flow for the APR1400

The APR1400 is a pressurized water reactor (PWR). The APR1400 nuclear steam supply system consists of a reactor pressure vessel (RPV), which contains the reactor core, and two primary reactor coolant system loops connected in parallel to the RPV. Each coolant loop consists of a hot leg (HL), a u-tube steam generator (SG), two reactor coolant pumps (RCPs) and two cold legs (CLs). A pressurizer is connected to the hot leg on one of the coolant loops. During normal operation, the RCPs circulate high pressure water through the reactor core to remove the heat. The heated water exits the RPV and passes through the hot legs into the steam generator tubes. Heat is transferred through the walls of the steam generator tubes into a lower pressure secondary coolant system. Cooled primary system water from the steam generators is returned through the RCPs and cold legs to the RPV. The RCPs operate at constant speed; the primary coolant system is designed to provide a prescribed rate of coolant flow through the reactor core. The APR1400 SG secondary coolant system pressure is controlled through modulation of valves in the steam lines to maintain a prescribed turbine inlet pressure. The SG secondary coolant inventory is controlled by using a feedwater control system to maintain a prescribed liquid level in the steam generator downcomer. This water inventory level is sufficient to cover the outer surfaces of the steam generator tubes while preventing overfill that would cause liquid to pass from the steam generators through the steam lines into the turbine system. In the current work, the Nuclear Regulatory Commission's reactor systems analysis code, TRACE?2 was used to perform an analysis of a loss of flow anticipated operational occurrence (AOO). The loss of flow is initiated by a loss of off-site power (LOOP), resulting in a trip of the RCPs. The regulatory purpose of the current work is to confirm safety analyses documented in the APR1400 Design Control Document (DCD)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1987

An experimental study on the effects of tip clearance on flow field and losses in an axial flow compressor rotor

Detailed measurement of the flow field in the tip region of a compressor rotor was carried out using a Laser Doppler Velocimeter (LDV) and a Kiel probe at two different tip clearance heights. At both clearance sizes, the relative stagnation pressure and the axial and tangential components of relative velocities were measured upstream, inside the passage and downstream of the rotor, up to about 20 percent of the blade span from the annulus wall. The velocities, outlet angles, losses, momentum thickness, and force defect thickness are compared for the two clearances. A detailed interpretation of the effect of tip clearance on the flow field is given. There are substantial differences in flow field, on momentum thickness, and performance as the clearance is varied. The losses increase linearly within the passage and their values increase in direct proportion to tip clearance height. No discernable vortex (discrete) is observed downstream of the rotor.

Lakshminarayana, B.↗
DOE OSTI2012

Behavior of an heterogeneous annular FBR core during an unprotected loss of flow accident: Analysis of the primary phase with SAS-SFR

In the framework of a substantial improvement on FBR core safety connected to the development of a new Gen IV reactor type, heterogeneous core with innovative features are being carefully analyzed in France since 2009. At EDF R and D, the main goal is to understand whether a strong reduction of the Na-void worth - possibly attempting a negative value - allows a significant improvement of the core behavior during an unprotected loss of flow accident. Also, the physical behavior of such a core is of interest, before and beyond the (possible) onset of Na boiling. Hence, a cutting-edge heterogeneous design, featuring an annular shape, a Na-plena with a B{sub 4}C plate and a stepwise modulation of fissile core heights, was developed at EDF by means of the SDDS methodology, with a total Na-void worth of -1 $. The behavior of such a core during the primary phase of a severe accident, initiated by an unprotected loss of flow, is analyzed by means of the SAS-SFR code. This study is carried-out at KIT and EDF, in the framework of a scientific collaboration on innovative FBR severe accident analyses. The results show that the reduction of the Na-void worth is very effective, but is not sufficient alone to avoid Na-boiling and, hence, to prevent the core from entering into the primary phase of a severe accident. Nevertheless, the grace time up to boiling onset is greatly enhanced in comparison to a more traditional homogeneous core design, and only an extremely low fraction of the fuel (<0.1%) enters into melting at the end of this phase. A sensitivity analysis shows that, due to the inherent neutronic characteristics of such a core, the gagging scheme plays a major role on the core behavior: indeed, an improved 4-zones gagging scheme, associated with an enhanced control rod drive line expansion feed-back effect, finally prevents the core from entering into sodium boiling. This major conclusion highlights both the progress already accomplished and the need for more detailed future analyses, particularly concerning: the neutronic burn-up scheme, the modeling of the diagrid effect and the control rod drive line expansion feed-backs, as well as the primary/secondary systems thermal-hydraulics behavior. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1974

Pressure-loss and flow coefficients inside a chordwise-finned, impingement, convection, and film air-cooled turbine vane

Total-pressure-loss coefficients, flow discharge coefficients, and friction factors were determined experimentally for the various area and geometry changes and flow passages within an air-cooled turbine vane. The results are compared with those of others obtained on similar configurations, both actual and large models, of vane passages. The supply and exit air pressures were controlled and varied. The investigation was conducted with essentially ambient-temperature air and without external flow of air over the vane.

Hippensteele, S. A.↗
DOE OSTI2012

Multi-scale approach to the modeling of fission gas discharge during hypothetical loss-of-flow accident in gen-IV sodium fast reactor

The required technological and safety standards for future Gen IV Reactors can only be achieved if advanced simulation capabilities become available, which combine high performance computing with the necessary level of modeling detail and high accuracy of predictions. The purpose of this paper is to present new results of multi-scale three-dimensional (3D) simulations of the inter-related phenomena, which occur as a result of fuel element heat-up and cladding failure, including the injection of a jet of gaseous fission products into a partially blocked Sodium Fast Reactor (SFR) coolant channel, and gas/molten sodium transport along the coolant channels. The computational approach to the analysis of the overall accident scenario is based on using two different inter-communicating computational multiphase fluid dynamics (CMFD) codes: a CFD code, PHASTA, and a RANS code, NPHASE-CMFD. Using the geometry and time history of cladding failure and the gas injection rate, direct numerical simulations (DNS), combined with the Level Set method, of two-phase turbulent flow have been performed by the PHASTA code. The model allows one to track the evolution of gas/liquid interfaces at a centimeter scale. The simulated phenomena include the formation and breakup of the jet of fission products injected into the liquid sodium coolant. The PHASTA outflow has been averaged over time to obtain mean phasic velocities and volumetric concentrations, as well as the liquid turbulent kinetic energy and turbulence dissipation rate, all of which have served as the input to the core-scale simulations using the NPHASE-CMFD code. A sliding window time averaging has been used to capture mean flow parameters for transient cases. The results presented in the paper include testing and validation of the proposed models, as well the predictions of fission-gas/liquid-sodium transport along a multi-rod fuel assembly of SFR during a partial loss-of-flow accident. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1991

Analysis of losses in supersonic mixing and reacting flows

A method for analyzing flow losses and thrust potential in supersonic combustors is presented. This method relies on a complete and consistent one-dimensional representation of a three-dimensional flow-field. Numerical results for flush wall fuel injection into a Mach 3 flow are examined and comparisons are made with experimental measurements of fuel concentration. Mixing results for a swept injection ramp, a straight (unswept) injection ramp, and a thirty degree downstream-directed flush wall jet in the same combustor duct are analyzed. The flow loss/thrust potential of the flush wall jet and the swept ramp are investigated (based on reacting solutions) using computed combustor effectiveness. The wall jet displays slightly higher thrust potential than the swept ramp at the end of the combustor.

Riggins, D. W.↗
DOE OSTI2025

Loss of Flow Conditions in a Modern Pool-type SFR and In-Pile Experiment Design

The metallic fuel safety performance under unprotected design-basis transients is a key consideration for the deployment of advanced sodium fast reactors (SFRs). Reliable data are needed to validate advanced safety codes, reduce uncertainty in cladding failure thresholds, and strengthen confidence in licensing approaches. To address this need, this report develops blueprints for a conceptual sodium loss-of-flow (LOF) experiment in the Mk-IIIR loop at the Transient Reactor Test Facility (TREAT), providing the technical foundation for future integral testing.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Coupled 3D-neutronics / thermal-hydraulics analysis of an unprotected loss-of-flow accident for a 3600 MWth SFR core

The core behaviour of a large (3600 MWth) sodium-cooled fast reactor (SFR) is investigated in this paper with the use of a coupled TRACE/PARCS model. The SFR neutron spectrum is characterized by several performance advantages, but also leads to one dominating neutronics drawback - a positive sodium void reactivity. This implies a positive reactivity effect when sodium coolant is removed from the core. In order to evaluate such feedback in terms of the dynamics, a representative unprotected loss-of-flow (ULOF) transient, i.e. flow run-down without SCRAM in which sodium boiling occurs, is analyzed. Although analysis of a single transient cannot allow general conclusions to be drawn, it does allow better understanding of the underlying physics and can lead to proposals for improving the core response during such an accident. The starting point of this study is the reference core design considered in the framework of the Collaborative Project on the European Sodium Fast Reactor (CP-ESFR). To reduce the void effect, the core has been modified by introducing an upper sodium plenum (along with a boron layer) and by reducing the core height-to-diameter ratio. For the ULOF considered, a sharp increase in core power results in melting of the fuel in the case of the reference core. In the modified core, a large dryout leads to melting of the clad. It seems that, for the hypothetical event considered, fuel failure cannot be avoided with just improvement of the neutronics design; therefore, thermal-hydraulics optimization has been considered. An innovative assembly design is proposed to prevent sodium vapour blocking the fuel channel. This results in preventing a downward propagation of the sodium boiling to the core center, thus limiting it to the upper region. Such a void map introduces a negative coolant density reactivity feedback, which dominates the total reactivity change. As a result, the power level and the fuel temperature are effectively reduced, and a large dryout is prevented. As a next step, the classical sodium plenum is replaced by a fission gas plenum (with lower sodium fraction), thus improving flow stability. Stable boiling at a steady power level is achieved in this final configuration. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Preliminary Analysis of FFTF Loss-of-Flow Tests with SAS4A/SASSYS-1

The Fast Flux Test Facility (FFTF) was the most recent Sodium-cooled Fast Reactor (SFR) built in the United States. The Passive Safety Test (PST) program was carried out on FFTF between 1984 and 1986 to demonstrate the capability of passive reactor shutdown and decay heat removal in response to protected and unprotected transients. In order to recover the information and knowledge that has been gained from FFTF, a research project was initiated focusing on the simulation of Unprotected Loss of Flow Without Scram (LOFWOS) tests similar to the knowledge preservation activities conducted for EBR-II by Argonne National Laboratory in earlier years. These knowledge preservation activities are intended to support the validation of computational tools and methods in the area of SFR passive safety. Comparisons with experimental data and other safety codes provide opportunities to improve SFR design and safety analysis codes and methods.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS2015

Understanding and Mitigating Tip Leakage and Endwall Losses in High Pressure Ratio Cores

Reducing endwall and tip secondary flow losses will be a key enabler for the next generation of commercial and military air transport and will be an improvement on the state-of-the-art in turbine loss reduction strategies. The objective of this research is three-fold: 1) To improve understanding of endwall secondary flow and tip clearance losses 2) To develop novel technologies to mitigate these losses and test them in low-speed cascade and rig environments 3) To validate predictive tools To accomplish these objectives, Pratt & Whitney (P&W) has teamed with Pennsylvania State University (PSU) to experimentally test new features designed by P&W. P&W will create new rim-cavity features to reduce secondary flow loss and improve purge flow cooling effectiveness and new blade tip features to manage leakage flows and reduce tip leakage secondary flow loss. P&W is currently developing technologies in these two areas that expect to be assimilated in the N+2/N+3 generation of commercial engines.

Turbomachinery↗
DOE OSTI2019

Uncertainty Quantification on SAM Simulations of EBR-II Loss-of-Flow Tests

The System Analysis Module (SAM) is a modern system analysis tool being developed at Argonne National Laboratory for advanced non-LWR safety analysis. It aims to provide fast-running, whole-plant transient analyses capability with improved-fidelity for SFR, LFR, and MSR/FHR. In prior work, SAM has performed well in modeling the EBR-II benchmark based on the SHRT-17 and SHRT-45R protected and unprotected loss-of-flow experiments. However, underlying uncertainties in the modeling parameters need to be quantified to gain confidence in any best-estimate simulations. In this work, a Python coupling framework between Dakota and SAM has been developed to investigate these uncertainties, alongside code enhancements in SAM for user-flexibility in perturbing physical closure models and fluid equations of state. 25 significant input parameters covering boundary conditions, geometry, physical closure models, and equations of state were identified with their uncertainties defined from literature. Two SAM transient models for SHRT-17 and SHRT-45R were perturbed on these input parameters using Latin Hypercube Sampling and the uncertainties in predicted maximum fuel, cladding, and coolant temperatures were quantified as the output responses. Spearman rank correlation of input parameters to the predicted temperatures indicated that the most sensitive parameters include the initial core power, pipe wall drag, fuel & coolant conductivity, and peak channel flow area. Sensitivity of input parameters was found to differ between the two transients and core channel modeling strategies. The overall perturbations did not result in large uncertainties in the predicted temperatures, despite the efforts made to obtain accurate yet conservative input uncertainties.

Mui, Travis↗
NASA NTRS1995

Off-design computer code for calculating the aerodynamic performance of axial-flow fans and compressors

An off-design axial-flow compressor code is presented and is available from COSMIC for predicting the aerodynamic performance maps of fans and compressors. Steady axisymmetric flow is assumed and the aerodynamic solution reduces to solving the two-dimensional flow field in the meridional plane. A streamline curvature method is used for calculating this flow-field outside the blade rows. This code allows for bleed flows and the first five stators can be reset for each rotational speed, capabilities which are necessary for large multistage compressors. The accuracy of the off-design performance predictions depend upon the validity of the flow loss and deviation correlation models. These empirical correlations for the flow loss and deviation are used to model the real flow effects and the off-design code will compute through small reverse flow regions. The input to this off-design code is fully described and a user's example case for a two-stage fan is included with complete input and output data sets. Also, a comparison of the off-design code predictions with experimental data is included which generally shows good agreement.

Schmidt, James F.↗
NASA NTRS1994

Fiber Optic Switch For Broadband Emission Spectroscopy

Many high-temperature processes comprise large-scale phenomena. Studying spatial and temporal correlations of physical processes between several locations within characteristic scales provides desired information on macroscopic physical processes. Achieved with emission spectroscopy by use of multiple optical fibers. Simultaneous coupling of light from these fibers into single available spectrometer and/or monochromator not accomplished without added expense of two-dimensional array and increased complexity of calibration. Quasi-simultaneous coupling, while maintaining optimum alignment and maximum throughput of broadband emission, achieved by use of fiber optic multiscanner. Instrument used successfully in study of frozen-flow losses internal to flow of plasma inside nozzle of arc jet. Instrument includes two hollow disks of different sizes and stepping motor.

De Groot, Wim↗