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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 181 records · Page 10
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Code and Solution Verification Assessment of the CTF Thermal Hydraulic Subchannel Code

CTF is a thermal-hydraulics subchannel code jointly developed by Oak Ridge National Laboratory and North Carolina State University. Over the past seven years, the Consortium for Advanced Simulation of Light Water Reactors (CASL) has made a significant investment in developing CTF so it can be used to model light water reactors, including nominal operating conditions, departure from nucleate boiling analysis, and transients ranging from loss of flow to reactivity insertion accidents. In addition to implementing new modeling capabilities and developing the user input and output interface, extensive work has been performed to improve the code’s quality assurance program, resulting in a development process that conforms with NQA-1 requirements. The CASL program follows the Predictive Capability Maturity Model (PCMM) approach for assessing code quality, which emphasizes performing code verification(ensuring the code converges to the correct answer) and solution verification (ensuring the code converges for the intended application). Code and solution verification are used to identify uncertainty errors introduced by numerical approximations in the code and are important for demonstrating that the model is coded without error, which is an important aspect of the Best Estimate plus Uncertainty method. This paper presents a comprehensive overview of the code and solution verification testing that has been performed on CTF. A top-down approach is taken in which the intended CTF applications are presented, followed by the code features required for their modeling. These features are then linked to the applicable code and solution verification tests that demonstrate proper functioning. Past testing efforts are summarized, and new tests are added to help close gaps in the presented test matrix. Rather than performing “one-off” exercises, these tests are added to the automated CTF regression test suite to ensure continual code quality.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

STAT7 v1.2 User Guide: The STAT7 Code for Statistical Propagation of Uncertainties in Steady-State Thermal Hydraulics Analysis of Plate-Fueled Reactors

The STAT7 software was developed to perform steady-state, single-phase thermal hydraulics analysis of plate-fueled reactors based on statistical propagation of uncertainties. Application of the software is for non-power research and test reactors, including conversion to low-enriched uranium fuel of U.S. High-Performance Research Reactors such as Massachusetts Institute of Technology Research Reactor. Since it can be necessary to repeat analysis during fuel reloading, STAT7 accommodates flexibility in analyzing many realistic aspects of reactor fuel management. STAT7 uses a Monte Carlo approach to model uncertainty in common fuel fabrication parameters and other key reactor operating parameters required for thermal hydraulics analyses of research and test reactors. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux, or onset of flow instability. STAT7 supports water properties based on the IAPWS-IF97 functions (The International Association for the Properties of Water and Steam Industrial Formulation 1997 for the Thermodynamic Properties of Water and Steam) in addition to the fit functions. STAT7 predicts axial profiles of fuel, cladding, and coolant temperature along a lateral stripe that runs the full length of the fuel plate from the bottom to the top. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one latera stripe of a fuel element. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis can be performed by analyzing the margin to onset of nucleate boiling and onset of flow instability for each axial node of each stripe of each plate of each fuel element in the core.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

STAT7 v2.0 User Guide

The STAT7 software was developed to perform steady-state, single-phase thermal hydraulics analysis of plate-fueled reactors based on statistical propagation of uncertainties. Application of the software includes non-power research and test reactor analysis, and it has been used for the conversion to low- enriched uranium fuel of U.S. High Performance Research Reactors such as the Massachusetts Institute of Technology Research Reactor. Since it can be necessary to repeat reactor safety analysis, such as during fuel reloading, STAT7 accommodates flexibility in analyzing many practical aspects of reactor fuel management. STAT7 uses a Monte Carlo approach to model uncertainty in common fuel fabrication parameters and other key reactor operating parameters required for reactor thermal hydraulics analysis. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux, or onset of flow instability. STAT7 supports water properties based on the IAPWS-IF97 functions (The International Association for the Properties of Water and Steam Industrial Formulation 1997 for the Thermodynamic Properties of Water and Steam) in addition to fitted functions. STAT7 predicts axial profiles of fuel, cladding, and coolant temperature along a lateral stripe that runs the full length of the fuel plate from the bottom to the top. STAT7 can simultaneously analyze every axial node in each lateral stripe of all fuel plates and coolant channels in every fuel element of an entire reactor core. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. In a single execution, STAT7 can be used to perform full core analysis by analyzing the margin to onset of nucleate boiling and onset of flow instability for each axial node of each stripe of each plate of each fuel element in the core.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Advanced Fuels Campaign Time-at-Temperature Irradiation Campaign and Post-Irradiation Examination Plan

With increasing per capita energy consumption and rapidly growing energy consumption predicted for data centers, the US faces challenges in meeting projected energy demands. The current administration is attempting to accelerate the deployment of new nuclear construction by providing significant investments in advanced nuclear and by modernizing regulatory approaches for the licensing new reactors. However, enabling extended power uprates (EPUs) for the current nuclear reactor fleet would provide a shorter-term solution to better calibrate near-term energy generation capacity to the ever-growing energy demands of the modern era. One pathway to achieve power uprates is to increase the operational window by reassessing fuel safety limits around anticipated transients. Existing light-water reactors (LWRs) utilize a targeted operational window with well-defined operational efficiencies, yet this window is also often bound by conservative safety limits that result in suboptimal operational performance. For example, the operation of the existing fleet is highly constrained by safeguards to operation related to anticipated operational occurrences (AOOs), which are moderate-frequency transients expected with a frequency greater than 0.01 per reactor-year. These AOOs include temperature excursions where the critical heat flux for the system is exceeded, resulting in a departure from nucleate boiling (DNB) at the cladding surface. This DNB event is a thermohydraulic condition that, with current conservative fuel safety limits, results in the fuel rods being classified as “failed,” meaning that the cladding may not be returned to service.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Advanced Fuels Campaign Time-at-Temperature Irradiation Campaign and Post-Irradiation Examination Plan

With increasing per capita energy consumption and rapidly growing energy consumption predicted for data centers, the US faces challenges in meeting projected energy demands. The current administration is attempting to accelerate the deployment of new nuclear construction by providing significant investments in advanced nuclear and by modernizing regulatory approaches for the licensing new reactors. However, enabling extended power uprates (EPUs) for the current nuclear reactor fleet would provide a shorter-term solution to better calibrate near-term energy generation capacity to the ever-growing energy demands of the modern era. One pathway to achieve power uprates is to increase the operational window by reassessing fuel safety limits around anticipated transients. Existing light-water reactors (LWRs) utilize a targeted operational window with well-defined operational efficiencies, yet this window is also often bound by conservative safety limits that result in suboptimal operational performance. For example, the operation of the existing fleet is highly constrained by safeguards to operation related to anticipated operational occurrences (AOOs), which are moderate-frequency transients expected with a frequency greater than 0.01 per reactor-year. These AOOs include temperature excursions where the critical heat flux for the system is exceeded, resulting in a departure from nucleate boiling (DNB) at the cladding surface. This DNB event is a thermohydraulic condition that, with current conservative fuel safety limits, results in the fuel rods being classified as “failed,” meaning that the cladding may not be returned to service.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1975

Measurements of the performance of an electrohydrodynamic heat pipe

The performance of an electrohydrodynamic heat pipe formed by enclosing three straight axial electrodes with a 1.27 cm O.D. internally threaded stainless steel tube was measured with Freon-11 as working fluid. The condenser and evaporator conductances at low power were comparable to those measured in a conventional capillary driven heat pipe using similar working fluids. The evaporator conductance improved at high power probably due to the onset of nucleate boiling. The maximum heat throughput for this pipe was substantially higher than that of a conventional pipe.

Loehrke, R. I.↗
NASA NTRS1990

Testing Bearing Balls For Ignition In Liquid Oxygen

Inexpensive, safe apparatus built to test cooling by, ignition in, and nucleate boiling to film boiling of liquid oxygen flowing around ball bearings. Provides for heating of bearing specimens, direct observations, and measurements of temperatures and temperature distributions by thermocouples and infrared sensors. Used to evaluate suitability of various materials and surface treatments for ball bearings in high-pressure liquid-oxygen turbopumps. Ball rests in fixture while liquid oxygen flows around it. Flow reversed so observer or infrared sensor can view phenomena on trailing or leading side of ball.

Wagner, William R.↗
NASA NTRS1990

Transient analysis of chilldown in a cryogenic transfer line

A numerical model was developed, with the SINDA'85/FLUINT program, for calculating the thermal and hydrodynamic transients that occur during the chilldown of a cryogenic transfer line, using a well documented test case to validate the modeling process. Using this model, a total of ten cases were analyzed to evaluate the effects of variable inlet valve position, inlet pressures, and the use of an internal flow liner to promote nucleate boiling. It was found that an efficient transfer line cooldown can be achieved if the inlet flow is throttled, to reduce the flow rate and quality, and an internal flow liner such as Teflon is used.

Martin, T.↗
NASA NTRS1992

Tank Pressure Control Experiment/thermal Phenomena (TPCE/TP)

The 'Tank Pressure Control Experiment/Thermal Phenomena (TPCE/TP)' is a reflight of the tank pressure control experiment (TPCE), flown on STS-43 in a standard Get-Away Special (GAS) container in August 1991. The TPCE obtained extensive video and digital data of the jet induced mixing process in a partially filled tank in low gravity environments. It also provided limited data on the thermal processes involved. The primary objective of the reflight of TPCE is to investigate experimentally the phenomena of liquid superheating and pool nucleate boiling at very low heat fluxes in a long duration low gravity environment. The findings of this experiment will be of direct relevance to space based subcritical cryogenic fluid system design and operation. Experiment hardware and results from the first TPCE are described in outline and graphic form.

Hasan, M. M.↗
NASA NTRS1993

Microtextured metals for stray-light suppression in the Clementine startracker

Anodized blacks for suppressing stray light in optical systems can now be replaced by microscopically textured metal surfaces. An application of these black surfaces to the Clementine star-tracker navigational system, which will be launched in early 1994 to examine the Moon, en route to intercept an asteroid, is detailed. Rugged black surfaces with Lambertian BRDF less than 10(exp -2) srad(sup -1) are critical for suppressing stray light in the star-tracker optical train. Previously available materials spall under launch vibrations to contaminate mirrors and lenses. Microtextured aluminum is nearly as dark, but much less fragile. It is made by differential ion beam sputtering, which generates light-trapping pores and cones slightly smaller than the wavelength to be absorbed. This leaves a sturdy but light-absorbing surface that can survive challenging conditions without generating debris or contaminants. Both seeded ion beams and plasma immersion (from ECR plasmas) extraction can produce these microscopic textures without fragile interfaces. Process parameters control feature size, spacing, and optical effects (THR, BRDF). Both broad and narrow absorption bands can be engineered with tuning for specific wavelengths and applications. Examples are presented characterized by FTIR in reflection librators (0.95 normal emissivity), heat rejection, and enhanced nucleate boiling.

Johnson, E. A.↗
NASA NTRS1992

Space nuclear power systems 1989; Proceedings of the 6th Symposium, Albuquerque, NM, Jan. 8-12, 1989. Vols. 1 & 2

The present conference discusses such space nuclear power (SNP) issues as current design trends for SDI applications, ultrahigh heat-flux systems with curved surface subcooled nucleate boiling, design and manufacturing alternatives for low cost production of SNPs, a lightweight radioisotope heater for the Galileo mission, compatible materials for uranium fluoride-based gas core SNPs, Johnson noise thermometry for SNPs, and uranium nitride/rhenium compatibility studies for the SP-100 SNP. Also discussed are system issues in antimatter energy conversion, the thermal design of a heat source for a Brayton cycle radioisotope power system, structural and thermal analyses of an isotope heat source, a novel plant protection strategy for transient reactors, and beryllium toxicity.

El-Genk, Mohamed S.↗
NASA NTRS2007

High Specific Power Motors in LN2 and LH2

A switched reluctance motor has been operated in liquid nitrogen (LN2) with a power density as high as that reported for any motor or generator. The high performance stems from the low resistivity of Cu at LN2 temperature and from the geometry of the windings, the combination of which permits steady-state rms current density up to 7000 A/cm2, about 10 times that possible in coils cooled by natural convection at room temperature. The Joule heating in the coils is conducted to the end turns for rejection to the LN2 bath. Minimal heat rejection occurs in the motor slots, preserving that region for conductor. In the end turns, the conductor layers are spaced to form a heat-exchanger-like structure that permits nucleate boiling over a large surface area. Although tests were performed in LN2 for convenience, this motor was designed as a prototype for use with liquid hydrogen (LH2) as the coolant. End-cooled coils would perform even better in LH2 because of further increases in copper electrical and thermal conductivities. Thermal analyses comparing LN2 and LH2 cooling are presented verifying that end-cooled coils in LH2 could be either much longer or could operate at higher current density without thermal runaway than in LN2.

Brown, Gerald V.↗
NASA NTRS1985

Rotary engine cooling system

A rotary engine has a substantially trochoidal-shaped housing cavity in which a rotor planetates. A cooling system for the engine directs coolant along a single series path consisting of series connected groups of passages. Coolant enters near the intake port, passes downwardly and axially through the cooler regions of the engine, then passes upwardly and axially through the hotter regions. By first flowing through the coolest regions, coolant pressure is reduced, thus reducing the saturation temperature of the coolant and thereby enhancing the nucleate boiling heat transfer mechanism which predominates in the high heat flux region of the engine during high power level operation.

Jones, Charles↗
NASA NTRS2007

High Specific Power Motors in LN2 and LH2

A switched reluctance motor has been operated in liquid nitrogen (LN2) with a power density as high as that reported for any motor or generator. The high performance stems from the low resistivity of Cu at LN2 temperature and from the geometry of the windings, the combination of which permits steady-state rms current density up to 7000 A/sq cm, about 10 times that possible in coils cooled by natural convection at room temperature. The Joule heating in the coils is conducted to the end turns for rejection to the LN2 bath. Minimal heat rejection occurs in the motor slots, preserving that region for conductor. In the end turns, the conductor layers are spaced to form a heat-exchanger-like structure that permits nucleate boiling over a large surface area. Although tests were performed in LN2 for convenience, this motor was designed as a prototype for use with liquid hydrogen (LH2) as the coolant. End-cooled coils would perform even better in LH2 because of further increases in copper electrical and thermal conductivities. Thermal analyses comparing LN2 and LH2 cooling are presented verifying that end-cooled coils in LH2 could be either much longer or could operate at higher current density without thermal runaway than in LN2.

Brown, Gerald V.↗
NASA NTRS2010

A Heat Transfer Investigation of Liquid and Two-Phase Methane

A heat transfer investigation was conducted for liquid and two-phase methane. The tests were conducted at the NASA Glenn Research Center Heated Tube Facility (HTF) using resistively heated tube sections to simulate conditions encountered in regeneratively cooled rocket engines. This testing is part of NASA s Propulsion and Cryogenics Advanced Development (PCAD) project. Nontoxic propellants, such as liquid oxygen/liquid methane (LO2/LCH4), offer potential benefits in both performance and safety over equivalently sized hypergolic propulsion systems in spacecraft applications. Regeneratively cooled thrust chambers are one solution for high performance, robust LO2/LCH4 engines, but cooling data on methane is limited. Several test runs were conducted using three different diameter Inconel 600 tubes, with nominal inner diameters of 0.0225-, 0.054-, and 0.075-in. The mass flow rate was varied from 0.005 to 0.07 lbm/sec. As the current focus of the PCAD project is on pressure fed engines for LO2/LCH4, the average test section outlet pressures were targeted to be 200 psia or 500 psia. The heat flux was incrementally increased for each test condition while the test section wall temperatures were monitored. A maximum average heat flux of 6.2 Btu/in.2 sec was achieved and, at times, the temperatures of the test sections reached in excess of 1800 R. The primary objective of the tests was to produce heat transfer correlations for methane in the liquid and two-phase regime. For two-phase flow testing, the critical heat flux values were determined where the fluid transitions from nucleate boiling to film boiling. A secondary goal of the testing was to measure system pressure drops in the two-phase regime.

VanNoord, Jonathan↗