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At least 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
NASA NTRS1997

Heat Transfer Computations of Internal Duct Flows With Combined Hydraulic and Thermal Developing Length

This study investigated the Navier-Stokes computations of the surface heat transfer coefficients of a transition duct flow. A transition duct from an axisymmetric cross section to a non-axisymmetric cross section, is usually used to connect the turbine exit to the nozzle. As the gas turbine inlet temperature increases, the transition duct is subjected to the high temperature at the gas turbine exit. The transition duct flow has combined development of hydraulic and thermal entry length. The design of the transition duct required accurate surface heat transfer coefficients. The Navier-Stokes computational method could be used to predict the surface heat transfer coefficients of a transition duct flow. The Proteus three-dimensional Navier-Stokes numerical computational code was used in this study. The code was first studied for the computations of the turbulent developing flow properties within a circular duct and a square duct. The code was then used to compute the turbulent flow properties of a transition duct flow. The computational results of the surface pressure, the skin friction factor, and the surface heat transfer coefficient were described and compared with their values obtained from theoretical analyses or experiments. The comparison showed that the Navier-Stokes computation could predict approximately the surface heat transfer coefficients of a transition duct flow.

Wang, C. R.↗
NASA NTRS1993

Thermal and hydraulic considerations regarding the fate of water discharged by the outflow channels to the Martian northern plains

The identification of possible shorelines in the Martian northern plains suggests that the water discharged by the circum-Chryse outflow channels may have led to the formation of transient seas, or possibly even an ocean, covering as much as one-third of the planet. Speculations regarding the possible fate of this water have included local ponding and reinfiltration into the crust; freezing, sublimation, and eventual cold-trapping at higher latitudes; or the in situ survival of this now frozen water to the present day -- perhaps aided by burial beneath a protective cover of eolian sediment or lavas. Although neither cold-trapping at higher latitudes nor the subsequent freezing and burial of flood waters can be ruled out, thermal and hydraulic considerations effectively eliminate the possibility that any significant reassimilation of this water by local infiltration has occurred given climatic conditions resembling those of today. The arguments against the local infiltration of flood water into the northern plains are two-fold. First, given the climatic and geothermal conditions that are thought to have prevailed on Mars during the Late Hesperian (the period of peak outflow channel activity in the northern plains), the thickness of the cryosphere in Chryse Planitia is likely to have exceeded 1 km. A necessary precondition for the widespread occurrence of groundwater is that the thermodynamic sink represented by the cryosphere must already be saturated with ice. For this reason, the ice-saturated cryosphere acts as an impermeable barrier that effectively precludes the local resupply of subpermafrost groundwater by the infiltration of water discharged to the surface by catastraphic floods. Note that the problem of local infiltration is not significantly improved even if the cryosphere were initially dry, for as water attempts to infiltrate the cold, dry crust, it will quickly freeze, creating a seal that prevents any further infiltration from the ponded water above. The second argument against the local infiltration of flood water in the northern plains is based on hydraulic considerations. Repeated impacts have likely brecciated the Martian crust down to a depth of roughly 10 km. Given a value of permeability no greater than that inferred for the top 10 km of the Earth's crust (approximately 10(exp -2) darcies), a timescale as much as a billion years or more for the Martian groundwater system to achieve hydrostatic equilibrium, and the approximately 2-4 km elevation difference between the outflow channel source regions and the northern plains, the water confined beneath the frozen crust of the northern plains should have been under a significant hydraulic head. Thus, the existence of a hydraulic pathway between the ponded flood waters above the northern plains and the confined aquifer lying beneath it would not have led to the infiltration of flood water back into the crust, but rather the additional expulsion of groundwater onto the surface.

Clifford, S. M.↗
NASA NTRS1999

Overview of Loop Heat Pipe Operation

Loop heat pipes (LHP's) are two-phase heat transfer devices that utilize the evaporation and condensation of a working fluid to transfer heat, and the capillary forces developed in the porous wicks to circulate the fluid. The LHP was first developed in the former Soviet Union in the early 1980s, about the same time that the capillary pumped loop (CPL) was developed in the United States. The LHP is known for its high pumping capability and robust operation mainly due to the use of fine-pored metal wicks and an integral evaporator/hydro-accumulator design. The LHP technology is rapidly gaining acceptance in aerospace community. It is the baseline design for thermal control of several spacecraft, including NASA's GLAS and Chemistry, ESA's ATLID, CNES' STENTOR, RKA's OBZOR, and several commercial satellites. Numerous LHP papers have been published since the mid-1980's. Most papers presented test results and discussions on certain specific aspects of the LHP operation. LHP's and CPL's show many similarities in their operating principles and performance characteristics. However, they also display significant differences in many aspects of their operation. Some of the LHP behaviors may seem strange or mysterious, even to experienced CPL practitioners. The main purpose of this paper is to present a systematic description of the operating principles and thermal-hydraulic behaviors of LHP'S. LHP operating principles will be given first, followed by a description of the thermal-hydraulics involved in LHP operation. Operating characteristics and important parameters affecting the LHP operation will then be described in detail. Peculiar behaviors of the LHP, including temperature hysteresis and temperature overshoot during start-up, will be explained. For simplicity, most discussions will focus upon LHP's with a single evaporator and a single condenser, but devices with multiple evaporators and condensers will also be discussed. Similarities and differences between LHP's and CPL's will be addressed throughout the paper whenever appropriate.

Ku, Jentung↗
NASA NTRS2006

Preliminary Analysis of a Water Shield for a Surface Power Reactor

A water based shielding system is being investigated for use on initial lunar surface power systems. The use of water may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. A simple 1-D thermal model indicates the necessity of natural convection to maintain acceptable temperatures and pressures in the water shield. CFD analysis is done to quantify the natural convection in the shield, and predicts sufficient natural convection to transfer heat through the shield with small temperature gradients. A test program will he designed to experimentally verify the thermal hydraulic performance of the shield, and to anchor the CFD models to experimental results.

Pearson, J. Boise↗
NASA NTRS1988

Reduced gravity boiling and condensing experiments simulated with the COBRA/TRAC computer code

A series of reduced-gravity two-phase flow experiments has been conducted with a boiler/condenser apparatus in the NASA KC-135 aircraft in order to obtain basic thermal-hydraulic data applicable to analytical design tools. Several test points from the KC-135 tests were selected for simulation by means of the COBRA/TRAC two-fluid, three-field thermal-hydraulic computer code; the points were chosen for a 25-90 percent void-fraction range. The possible causes for the lack of agreement noted between simulations and experiments are explored, with attention to the physical characteristics of two-phase flow in one-G and near-zero-G conditions.

Cuta, Judith M.↗
NASA NTRS2011

Updates on HRF Payloads Operations in Columbus ATCS

The NASA developed Human Research Facility 1 (HRF1) and Human Research Facility (HRF2) experiment racks have been operating in the European Space Agency (ESA) Columbus module of the International Space Station (ISS) since Summer 2008. The two racks are of the same design. Since the start of operations, unexpected pressure spikes were observed in the Columbus module's thermal-hydraulic system during the racks activation sequence. The root cause of these spikes was identified in the activation command sequence in the Rack Interface Controller (RIC), which controls the flow of thermal-hydraulic system fluid through the rack. A new Common RIC Software (CRS) release fixed the bug and was uploaded on both racks in late 2009. This paper gives a short introduction to the topic, describes the Columbus module countermeasures to mitigate the spikes, describes the ground validation test of the new software, and describes the flight checks performed before and after the final upload. Finally, the new on-orbit test designed to further simplify the racks hydraulic management is presented.

DePalo, Savino↗
NASA NTRS1988

Reduced-gravity two-phase flow experiments in the NASA KC-135

An adequate understanding is sought of flow and heat transfer behavior in reduced and zero gravity conditions. Microgravity thermal-hydraulic analysis capabilities were developed for application to space nuclear power systems. A series of reduced gravity two phase flow experiments using the NASA KC-135 were performed. The objective was to supply basic thermal hydraulic information that could be used in development of analytical tools for design of space power systems. The experiments are described. Two main conclusions were drawn. First, the tests demonstrate that the KC-135 is a suitable test environment for obtaining two phase flow and heat transfer data in reduced gravity conditions. Second, the behavior of two phase flow in low gravity is sufficiently different from that obtained in 1 g to warrant intensive investigation of the phenomenon if adequate analytical tools are to be developed for microgravity conditions.

Cuta, Judith M.↗
NASA NTRSDate not supplied

Comparison of Convective Heat Transfer Correlations and Their Application to NTP Reactors

This study analyzes various Nusselt number and friction factor correlations and applies them to a Small Nuclear Rocket Engine model with a Sinusoidal power distribution profile to understand their effects on the temperatures and pressures inside the reactor. A nodal thermal hydraulic solver was used to determine the fluid and channel surface temperatures while also incorporating variable fluid properties and channel roughness. The results showed that the considered friction factors could essentially be used interchangeably given that their difference affected the pressure by less than 1%. However, large variations in the tube surface temperature were obtained for the different Nusselt numbers while the fluid temperature distribution was forced to remain the same. Supersonic flow conditions presented by Maynard Taylor are investigated to serve as a baseline for how experimental errors can lead to uncertainties in the reported empirical correlations. Detailed experimental investigation is necessary to determine the Nusselt number correlation that will provide the best prediction of the thermal hydraulic performance inside the reactor fluid flow channels.

Nuclear Thermal Propulsion↗
NASA NTRSDate not supplied

Comparison of Convective Heat Transfer Correlations and Their Application to Nuclear Thermal Propulsion Reactors

This study analyzes various Nusselt number and friction factor correlations and applies them to a Small Nuclear Rocket Engine model with a Sinusoidal power distribution profile to understand their effects on the temperatures and pressures inside the reactor. A nodal thermal hydraulic solver was used to determine the fluid and channel surface temperatures while also incorporating variable fluid properties and channel roughness. The results showed that the considered friction factors could essentially be used interchangeably given that their difference affected the pressure by less than 1%. However, large variations in the tube surface temperature were obtained for the different Nusselt numbers while the fluid temperature distribution was forced to remain the same. Supersonic flow conditions presented by Maynard Taylor are investigated to serve as a baseline for how experimental errors can lead to uncertainties in the reported empirical correlations. Detailed experimental investigation is necessary to determine the Nusselt number correlation that will provide the best prediction of the thermal hydraulic performance inside the reactor fluid flow channels.

Experiment↗
NASA NTRS1998

Nuclear modules for space electric propulsion

Analysis of interplanetary cargo and piloted missions requires calculations of the performances and masses of subsystems to be integrated in a final design. In a preliminary and scoping stage the designer needs to evaluate options iteratively by using fast computer simulations. The Oak Ridge National Laboratory (ORNL) has been involved in the development of models and calculational procedures for the analysis (neutronic and thermal hydraulic) of power sources for nuclear electric propulsion. The nuclear modules will be integrated into the whole simulation of the nuclear electric propulsion system. The vehicles use either a Brayton direct-conversion cycle, using the heated helium from a NERVA-type reactor, or a potassium Rankine cycle, with the working fluid heated on the secondary side of a heat exchanger and lithium on the primary side coming from a fast reactor. Given a set of input conditions, the codes calculate composition. dimensions, volumes, and masses of the core, reflector, control system, pressure vessel, neutron and gamma shields, as well as the thermal hydraulic conditions of the coolant, clad and fuel. Input conditions are power, core life, pressure and temperature of the coolant at the inlet of the core, either the temperature of the coolant at the outlet of the core or the coolant mass flow and the fluences and integrated doses at the cargo area. Using state-of-the-art neutron cross sections and transport codes, a database was created for the neutronic performance of both reactor designs. The free parameters of the models are the moderator/fuel mass ratio for the NERVA reactor and the enrichment and the pitch of the lattice for the fast reactor. Reactivity and energy balance equations are simultaneously solved to find the reactor design. Thermalhydraulic conditions are calculated by solving the one-dimensional versions of the equations of conservation of mass, energy, and momentum with compressible flow.

Difilippo, F. C.↗
NASA NTRS2005

Hardware Progress Made in the Early Flight Fission Test Facilities (EFF-TF) To Support Near-Term Space Fission Systems

The EFF-TF provides a facility to experimentally evaluate thermal hydraulic issues through the use of highly effective non-nuclear testing. These techniques provide a rapid, more cost effective method of evaluating designs and support development risk mitigation when concerns are associated with non-nuclear aspects of space nuclear systems. For many systems, electrical resistance thermal simulators can be used to closely mimic the heat deposition of the fission process, providing axial and radial profiles. A number of experimental and design programs were underway in 2004. Initial evaluation of the SAFE-100a (19 module stainless steel/sodium heat pipe reactor with integral gas neat exchanger) was performed with tests up to 17.5 kW of input power at core temperatures of 1000 K. A stainless steel sodium SAFE-100 heat pipe module was placed through repeated freeze/thaw cyclic testing accumulating over 200 restarts to a temperature of 1000 K. Additionally, the design of a 37-fuel pin stainless steel pumped sodium/potassium (NaK) loop was finalized and components procured. Ongoing testing at the EFF-TF is geared towards facilitating both research and development necessary to field a near term space nuclear system. Efforts are coordinated with DOE laboratories, industry, universities, and other NASA centers. This paper describes some of the 2004 efforts.

VanDyke, Melissa↗
NASA NTRS2007

High Fidelity Thermal Simulators for Non-Nuclear Testing: Analysis and Initial Results

Non-nuclear testing can be a valuable tool in the development of a space nuclear power system, providing system characterization data and allowing one to work through various fabrication, assembly and integration issues without the cost and time associated with a full ground nuclear test. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Testing with non-optimized heater elements allows one to assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. High fidelity thermal simulators that match both the static and the dynamic fuel pin performance that would be observed in an operating, fueled nuclear reactor can vastly increase the value of non-nuclear test results. With optimized simulators, the integration of thermal hydraulic hardware tests with simulated neutronie response provides a bridge between electrically heated testing and fueled nuclear testing, providing a better assessment of system integration issues, characterization of integrated system response times and response characteristics, and assessment of potential design improvements' at a relatively small fiscal investment. Initial conceptual thermal simulator designs are determined by simple one-dimensional analysis at a single axial location and at steady state conditions; feasible concepts are then input into a detailed three-dimensional model for comparison to expected fuel pin performance. Static and dynamic fuel pin performance for a proposed reactor design is determined using SINDA/FLUINT thermal analysis software, and comparison is made between the expected nuclear performance and the performance of conceptual thermal simulator designs. Through a series of iterative analyses, a conceptual high fidelity design can developed. Test results presented in this paper correspond to a "first cut" simulator design for a potential liquid metal (NaK) cooled reactor design that could be applied for Lunar surface power. Proposed refinements to this simulator design are also presented.

Bragg-Sitton, Shannon M.↗
NASA NTRS2007

High Fidelity, Fuel-Like Thermal Simulators for Non-Nuclear Testing: Analysis and Initial Test Results

Non-nuclear testing can be a valuable tool in the development of a space nuclear power system, providing system characterization data and allowing one to work through various fabrication, assembly and integration issues without the cost and time associated with a full ground nuclear test. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Testing with non-optimized heater elements allows one to assess thermal, heat transfer. and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. High fidelity thermal simulators that match both the static and the dynamic fuel pin performance that would be observed in an operating, fueled nuclear reactor can vastly increase the value of non-nuclear test results. With optimized simulators, the integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics and assess potential design improvements at relatively small fiscal investment. Initial conceptual thermal simulator designs are determined by simple one-dimensional analysis at a single axial location and at steady state conditions; feasible concepts are then input into a detailed three-dimensional model for comparison to expected fuel pin performance. Static and dynamic fuel pin performance for a proposed reactor design is determined using SINDA/FLUINT thermal analysis software, and comparison is made between the expected nuclear performance and the performance of conceptual thermal simulator designs. Through a series of iterative analyses, a conceptual high fidelity design is developed: this is followed by engineering design, fabrication, and testing to validate the overall design process. Test results presented in this paper correspond to a "first cut" simulator design for a potential liquid metal (NaK) cooled reactor design that could be applied for Lunar surface power. Proposed refinements to this simulator design are also presented.

Bragg-Sitton, Shannon M.↗
NASA NTRS1998

A Report from the Thermal Science Research Center (TSRC)

A vertical flow loop was designed and assembled to determine the local (circumferential and axial) and mean wall temperature distributions for single-phase and two-phase (subcooled and saturated) downward flow in both uniformly-heated and single-side heated vertical channels. Freon-11 was used as the working fluid in order to directly relate and compare the results with a previous experimental campaign which employed this same working fluid. For a given steady-state experiment, the following parameters were held constant: (1) exit pressure, (2) inlet temperature, and (3) mass velocity. For a given configuration of the 2.2 m long cylindrical channel test section, which had a 1.2 m long heated section, the applied heat rate was varied from zero through successive quasi-steady states to a level which corresponded to localized film boiling in the test section. The measurements showed that the boiling curve changes significantly at higher mass velocities with respect to both the circumferential and axial directions. The slope of the boiling curve changes in a non-monotonic fashion with respect to the circumferential directions. The slope of the boiling curve changes in a non-monotonic fashion with respect to the circumferential direction. The measurements point to the existence of a dry-out phenomenon occurring at multiple levels of the applied heat for the single-side heated channel. In comparing the heat transfer for horizontal channel flow with a vertically downward flow, the results show that significantly lower heat transfer occurs in the horizontal flow. However, this trend reverses as both the Reynolds number and the applied heat rate increase. Both the Liu-Winterton and Shah correlations were compared with the experimental data. The Shah correlation predicted the uniformly heated tube data better. When a thermal hydraulic diameter approach was used for the single-side heated case, the data at upstream locations for Z/L less than 0.5 was bounded above by the Liu-Winterton correlation and below by the Shah correlation. At Z/L = 0.5, the Shah correlation bounded the data; and for Z/L greater than 0.5, both correlations overpredicted the data with the Shah correlation being closest to the data. The present results indicate that additional correlational development is needed. In addressing some of the advanced space thermal management objectives concerning accommodating high heat fluxes in non-uniformly heated systems, a large battery of experiments 88 have been completed where local two-dimensional wall temperature variations were measured for both single-phase and two-phase flow in a single-side heated circular tube. As noted above, the results show significant axial and circumferential variations. Accurately accounting for such variations can result in optimized future advanced space, enhanced (high heat flux) thermal management systems.

Boyd, Ronald D.↗
NASA NTRS1992

An improved heat transfer configuration for a solid-core nuclear thermal rocket engine

Interrupted flow, impingement cooling, and axial power distribution are employed to enhance the heat-transfer configuration of a solid-core nuclear thermal rocket engine. Impingement cooling is introduced to increase the local heat-transfer coefficients between the reactor material and the coolants. Increased fuel loading is used at the inlet end of the reactor to enhance heat-transfer capability where the temperature differences are the greatest. A thermal-hydraulics computer program for an unfueled NERVA reactor core is employed to analyze the proposed configuration with attention given to uniform fuel loading, number of channels through the impingement wafers, fuel-element length, mass-flow rate, and wafer gap. The impingement wafer concept (IWC) is shown to have heat-transfer characteristics that are better than those of the NERVA-derived reactor at 2500 K. The IWC concept is argued to be an effective heat-transfer configuration for solid-core nuclear thermal rocket engines.

Clark, John S.↗