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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 1,747 records · Page 97
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Presentation: Steam Condensation Scaled Experiment in the Presence of Non-condensable Gas for Reactor Containment Passive Safety Analysis

This study presents scaled experiments using steam condensation with non-condensable gas (NCG)—helium, simulating hydrogen—as these experiments are pivotal for water-cooled reactor passive containment cooling system (PCCS) design and analysis. Research into PCCSs for small modular reactors (SMRs) is especially important in light of SMR system design; however, studies in the literature reflect limitations due to test geometry and operational condition variations, without considering SMR prototypic design. To address these challenges, a scaled test facility was developed to accurately replicate SMR PCCSs. This facility includes vertical down-flow condensing test sections with 1-, 2-, and 4-in.-diameter condensing tubes, accompanied by annular water cooling. Experiments were conducted using both superheated and saturated steam, with steam mass flow rates varying from 55 to 66 kg/hr., in the presence of helium as the NCG mass flow rate ranges from 1.8 to 22 kg/hr. Test data were collected on (a) the axial temperatures of the annular cooling water; (b) the outer wall temperature of the condensers; and (c) the mass flow rate, temperature, and pressure at the test section inlets and outlets. These primary test data were used in conjunction with a standard data reduction methodology to estimate essential thermal parameters such as heat fluxes, heat transfer coefficients, and condensation rates. The effects of NCGs on steam condensation within the geometry of the scaled test sections were then presented in regard to various testing conditions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1994

Development of pneumatic test techniques for subsonic high-lift and in-ground-effect wind tunnel investigations

Wind tunnel evaluations of two-dimensional high-lift airfoils and of vehicles operating in ground effect near the tunnel floor require special test facilities and procedures. These are needed to avoid errors caused by proximity to the walls and interference from the wall boundary layers. Pneumatic test techniques and facilities were developed for GTRI aerodynamic research tunnels and calibrated to verify that these wall effects had been removed. The modified facilities were then employed to evaluate the aerodynamic characteristics of blown very-high-lift airfoils and of racing hydroplanes operating in ground effect at various levels above the floor. The pneumatic facilities, techniques and calibrations are discussed and typical aerodynamic data recorded both with and without the test-section blowing systems are presented.

Englar, Robert J.↗
DOE OSTI2022

HTS Cable Conductor for Compact Fusion Tokamak Solenoids

Significant progress has been made recently in the US fusion community to develop a strategic plan to enable engineering design and construction of a fusion pilot plant (FPP). Princeton Plasma Physics Laboratory (PPPL) is working on developing high current density HTS conductors for next fusion experiments. Partnering with the US industry, we are evaluating feasibility and affordability of Conductor on Round Core (CORC ® ) developed by Advanced Conductor Technologies (ACT) for the next compact fusion tokamak facility. High current density achieved by a CORC ® cable based on a four-layer model coil recently tested at National High Magnetic Field Laboratory (NHMFL) motivated its consideration for low cost, reduced size fusion magnet application. This is of interest to PPPL because of its scalability to tokamak central solenoid (CS) coils in terms of required flux swings for plasma startup operations. Furthermore, partnering with ACT, we designed and built a two-layer model coil solenoid directly wound with CORC to demonstrate its applicability for compact solenoids such as that used in the national spherical torus experiment (NSTX), NSTX-upgrade (NSTX-U) and the US sustained high-power density test facility (SHPD). The ~160 mm diameter solenoid wound by a two-layer CORC is being tested in early 2022 under electromagnetic cyclic loading at NHMFL in a unique 160 mm bore, 14 T background field magnet facility.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2024

HIGH BURNUP FUEL-COOLANT INTERACTION ANALYSIS SUPPORTING FUEL SAFETY TESTING AT IDAHO NATIONAL LABORATORY

In the near future, experiments on HBu fuel under loss-of-coolant accident (LOCA) and reactivity-initiated accident (RIA) conditions will be performed within the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL). These experiments will be performed using the Transient Water Irradiation System for TREAT (TWIST) experiment vehicle. To support these experiments, analysis of fuel-coolant interaction (FCI) energetics is underway. This paper discusses FCIs in the context of light water reactor (LWR) safety, differentiating between the severe accident focus of commercial reactors and experimental RIA test programs where FCIs have occurred. However, it is highlighted that as the nuclear industry aims for increased burnup limits, the FCI events observed in RIA test programs may become relevant to commercial LWR safety analysis. The paper then presents developments to the UW-FCI computer program to enable simulation of FCIs initiated by solid fuel particles dispersing into the coolant during RIAs.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2001

Development and Implementation of NASA's Lead Center for Rocket Propulsion Testing

With the new millennium, NASA's John C. Stennis Space Center (SSC) continues to develop and refine its role as rocket test service provider for NASA and the Nation. As Lead Center for Rocket Propulsion Testing (LCRPT), significant progress has been made under SSC's leadership to consolidate and streamline NASA's rocket test infrastructure and make this vital capability truly world class. NASA's Rocket Propulsion Test (RPT) capability consists of 32 test positions with a replacement value in excess of $2B. It is dispersed at Marshall Space Flight Center (MSFC), Johnson Space Center (JSC)-White Sands Test Facility (WSTF), Glenn Research Center (GRC)-Plum Brook (PB), and SSC and is sized appropriately to minimize duplication and infrastructure costs. The LCRPT also provides a single integrated point of entry into NASA's rocket test services. The RPT capability is managed through the Rocket Propulsion Test Management Board (RPTMB), chaired by SSC with representatives from each center identified above. The Board is highly active, meeting weekly, and is key to providing responsive test services for ongoing operational and developmental NASA and commercial programs including Shuttle, Evolved Expendable Launch Vehicle, and 2nd and 3rd Generation Reusable Launch Vehicles. The relationship between SSC, the test provider, and the hardware developers, like MSFC, is critical to the implementation of the LCRPT. Much effort has been expended to develop and refine these relationships with SSC customers. These efforts have met with success and will continue to be a high priority to SSC for the future. To data in the exercise of its role, the LCRPT has made 22 test assignments and saved or avoided approximately $51M. The LCRPT directly manages approximately $30M annually in test infrastructure costs including facility maintenance and upgrades, direct test support, and test technology development. This annual budges supports rocket propulsion test programs which have an annual budget in excess of $150M. As the LCRPT continues to develop, customer responsiveness and lower cost test services will be major themes. In that light, SSC is embarking on major test technology development activities ensuring long range goals of safer, more responsive, and more cost effective test services are realized. The LCRPT is also focusing on the testing requirements for advanced propulsion systems. This future planning is key to defining and fielding the ability to test these new technologies in support of the hardware developers.

Dawson, Michael C.↗
DOE OSTI2016

Technical Design Report for the FACET-II Project at SLAC National Accelerator Laboratory

Electrons can “surf” on waves of plasma – a hot gas of charged particles – gaining very high energies in very short distances. This approach, called plasma wakefield acceleration, has the potential to dramatically shrink the size and cost of particle accelerators. Research at the SLAC National Accelerator Laboratory has demonstrated that plasmas can provide 1,000 times the acceleration in a given distance compared with current technologies. Developing revolutionary and more efficient acceleration techniques that allow for an affordable high-energy collider has been the focus of FACET, a National User Facility at SLAC. FACET used part of SLAC’s two-mile-long linear accelerator to generate high-density beams of electrons and their antimatter counterparts, positrons. Research into plasma wakefield acceleration was the primary motivation for constructing FACET. In April 2016, FACET operations came to an end to make way for the second phase of SLAC’s x-ray laser, the LCLS-II, which will use part of the tunnel occupied by FACET. FACET-II is a new test facility to provide the unique capability to develop advanced acceleration and coherent radiation techniques with high-energy electron and positron beams. FACET-II represents a major upgrade over current FACET capabilities and the breadth of the potential research program makes it truly unique.

43 PARTICLE ACCELERATORS↗
NASA NTRS2010

A Framework for Integrated Component and System Analyses of Instabilities

Instabilities associated with fluid handling and operation in liquid rocket propulsion systems and test facilities usually manifest themselves as structural vibrations or some form of structural damage. While the source of the instability is directly related to the performance of a component such as a turbopump, valve or a flow control element, the associated pressure fluctuations as they propagate through the system have the potential to amplify and resonate with natural modes of the structural elements and components of the system. In this paper, the authors have developed an innovative multi-level approach that involves analysis at the component and systems level. The primary source of the unsteadiness is modeled with a high-fidelity hybrid RANS/LES based CFD methodology that has been previously used to study instabilities in feed systems. This high fidelity approach is used to quantify the instability and understand the physics associated with the instability. System response to the driving instability is determined through a transfer matrix approach wherein the incoming and outgoing pressure and velocity fluctuations are related through a transfer (or transmission) matrix. The coefficients of the transfer matrix for each component (i.e. valve, pipe, orifice etc.) are individually derived from the flow physics associated with the component. A demonstration case representing a test loop/test facility comprised of a network of elements is constructed with the transfer matrix approach and the amplification of modes analyzed as the instability propagates through the test loop.

Ahuja, Vineet↗
DOE OSTI2016

Pressure drop measurements across a helical coil steam generator configuration - 15997

The primary heat transport loop heat exchanger within a Very-High Temperature Reactor (VHTR) is most commonly a Helical Coil Steam Generator (HCSG). This design has been regarded as having increased heat transfer within a smaller design in comparison to other heat exchangers. The fluidic heat transfer properties of the HCSG is still under evaluation in tangent with HTRs due to its complex design. Heat transfer between these layers relies heavily on their relative location and fluidic properties. A model was constructed to study the pressure field on a rod s surface as well as across multiple layers of helically facing tubes. Experiments focused on evaluating the impact relative geometry played on the overall pressure drop of the test section with inlet fluid velocity ranging from 0 to 3 m/s. A study to compare the experimental results with the Gaddis-Gnielinski Correlative theoretical values of pressure drop in tube bundles shows that the complex helical geometry results in pressure drops similar to that of fully developed rod bundle although the test facility is comprised of two half rod helical coils. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Gravity-Based Non-Invasive System for Reactor Coolant Inventory Monitoring

This paper summarizes the progress in the use of non-invasive methods for measuring liquid level in small modular nuclear reactors. High precision gravimeters are used to measure changes in liquid level in a test facility environment, and results are compared to simulations. Possible configurations for the meters use in the NuScale SMR are also discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Radiation characterization summary of the NETL beam port 1/5 free-field environment at the 128-inch core centerline adjacent location

The characterization of the neutron, prompt gamma-ray, and delayed gamma-ray radiation fields in the University of Texas at Austin Nuclear Engineering Teaching Laboratory (NETL) TRIGA reactor for the beam port (BP) 1/5 free-field environment at the 128-inch location adjacent to the core centerline has been accomplished. NETL is being explored as an auxiliary neutron test facility for the Sandia National Laboratories radiation effects sciences research and development campaigns. The NETL reactor is a TRIGA Mark-II pulse and steady-state, above-ground pool-type reactor. NETL is intended as a university research reactor typically used to perform irradiation experiments for students and customers, radioisotope production, as well as a training reactor. Initial criticality of the NETL TRIGA reactor was achieved on March 12, 1992, making it one of the newest test reactor facilities in the US. The neutron energy spectra, uncertainties, and covariance matrices are presented as well as a neutron fluence map of the experiment area of the cavity. For an unmoderated condition, the neutron fluence at the center of BP 1/5, at the adjacent core axial centerline, is about 8.2×10 12 n/cm 2 per MJ of reactor energy. About 67% of the neutron fluence is below 1 keV and 22% above 100 keV. The 1-MeV Damage-Equivalent Silicon (DES) fluence is roughly 1.6×10 12 n/cm 2 per MJ of reactor energy.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Investigation of similarity metrics for simulation based scaling analysis

Scaling analysis is a technique for characterizing degree of similarity between processes to occur in a real prototypic system and those to occur in its reduced scale (experimental) analog. In the context of nuclear reactor thermal-hydraulics and safety, scaling analysis is instrumental in assessing the applicability of data obtained from a scaled test facility for calibration of models and validation of simulation codes. Quantification of scale-related distortions is the key element in scaling analysis. In this work, we examine an approach for scaling analysis using computer simulation codes. Similarity analysis using information-theoretic approach is performed by comparing performance of experiment and application to variation of system conditions. Three different information theoretic measures namely mutual information, Kullback-Leibler divergence and overlap coefficient are selected for their potential use as metric for similarity analysis. We test the metrics by comparing them against physics-based scaling ratios, which are well known in fundamental thermal-hydraulic processes. In this exploratory analysis, a case of pressure drop of a steady-state flow through a channel is employed to test the proposed approach. It was found that in the absence of model uncertainties, mutual information serve as an effective measure of the scale effects present in the experiment. However, model uncertainty can significantly impact the quality of similarity analysis. Our analysis indicates that model uncertainties can have significant impact on the information theoretic metrics used for scaling analysis. Furthermore, if the model uncertainties exceeds beyond a critical value, the scale effect present in the experiment cannot be adequately evaluated. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Report on Oak Ridge National Laboratory Testing of Ambient Cure TransPowr E3X®

A key to industry acceptance of a new technology is extensive validation in field trials. The Powerline Conductor Accelerated Test facility (PCAT) at Oak Ridge National Laboratory (ORNL) is specifically designed to evaluate the performance and reliability of novel overhead conductor and conductor instrumentation under field conditions. The facility is set up to capture large amounts of data during testing. General Cable used the ORNL PCAT facility to validate the performance of TransPowr® with Ambient Cure E3X™ Technology—a standard overhead conductor with a high emissivity, low absorptivity ambient cure surface coating. Extensive testing has demonstrated a significant improvement in conductor performance across a wide range of operating temperatures with respect to uncoated conductor of the same type, indicating that E3X Technology can provide a reduction in temperature, a reduction in sag, and an increase in ampacity when applied to the surface of any overhead conductor. This report provides results of that testing.

24 POWER TRANSMISSION AND DISTRIBUTION↗
NASA NTRS1995

Calculated Effects of Nitric Oxide Flow Contamination on Scramjet Performance

The level of nitric oxide contamination in the test gas of the NASA Langley Research Center Arc-Heated Scramjet Test Facility and the effect of the contamination on scramjet test engine performance were investigated analytically. The study was conducted for standard facility conditions corresponding to Mach 6, 7, and 8 flight simulations. The analytically determined levels of nitric oxide produced in the facility are compared with experimentally measured levels. Results of the analysis indicate that nitric oxide levels range from one to three mole percent, which corroborates the measured levels. A three-stream combustor code with finite rate chemistry was used to investigate how nitric oxide affects scramjet performance in terms of combustor pressure rise, heat release, and thrust performance. Results indicate minimal effects on engine performance for the test conditions of this investigation.

Fischer, Karen E.↗
DOE OSTI2016

SMR Re-Scaling and Modeling for Load Following Studies

This study investigates the creation of a new set of scaling parameters for the Oregon State University Multi-Application Small Light Water Reactor (MASLWR) scaled thermal hydraulic test facility which modeled the NuScale power SMR design. As part of a study being undertaken by Idaho National Lab involving nuclear reactor load following characteristics, full power operations need to be simulated, and therefore properly scaled. Presented here is the scaling analysis and plans for RELAP5-3D simulation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

HTTR-GT/H2 test plant - conceptual design study on new intermediate heat exchanger - 18747

Japan Atomic Energy Agency is developing a nuclear cogeneration demonstration test facility named HTTR-GT/H2 that is connected to the existing test reactor HTTR. In the facility, an intermediate heat exchanger (IHX) at maximum service temperature of 900 deg. C is to be newly constructed to supply the high temperature helium flows to both the helium gas turbine system and the hydrogen production system. The conceptual design study on the IHX is conducted considering the future application to the Gas Turbine High Temperature Reactor 300 Cogeneration (GTHTR300C). Several new concepts are proposed and incorporated in the demonstrating IHX design in order to (1) ease integration of the IHX with the gas turbine system to simplify the cogeneration plant layout; (2) allow conventional low-cost steel for construction of the IHX pressure vessel and nozzles; (3) flexibility to accommodate thermal expansion of the bundle and internal structures.; (4) maintaining the structural integrity at elevated service temperature range. This paper describes the conceptual design study results for the IHX and technical issues to be solved in the future design phases. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

OECD-NEA HTTF Benchmark Progress and Updates

General slides discussing High Temperature Test Facility benchmark progress, Prismatic HTGR deployment, modeling and simulation tools for validating systems, and verification and validation issues. Also discuss collaboration and conclusions of RELAP5-3D validation activities based on HTTF with trends in data and international impact to accelerate deployment of prismatic HTGR microreactors by providing an opportunity for designers to assess their codes against experimental data and solutions from other codes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Design and Testing for a New Thermosyphon Irradiation Vehicle

The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) requires most materials and all fuel experiments to be placed in a pressure containment vessel to ensure that internal contaminants such as fission products cannot be released into the primary coolant. It also requires that all experiments be capable of withstanding various accident conditions (e.g., loss of coolant) without generating vapor bubbles on the surface of the experiment in the primary coolant. These requirements are intended to artificially increase experiment temperatures by introducing a barrier between the experimental materials and the HFIR coolant, and by reducing heat loads to the HFIR primary coolant, thus ensuring that no boiling can occur. A proposed design for materials irradiation would remove these limitations by providing the required primary containment with an internal cooling flow. This would allow for experiments to be irradiated without concern for coolant contamination (e.g., from cladding failure of advanced fuel pins) or for specimen heat load. This report describes a new materials irradiation experiment design that uses a thermosyphon cooling system to allow experimental materials direct access to a liquid coolant. The new design also increases the range of conditions that can be tested in HFIR. This design will provide a unique capability to validate the performance of current and advanced fuels and materials. Because of limited supporting data for this kind of irradiation vehicle, a test program was initiated to obtain operating data that can be used to (1) qualify the vehicle for operation in HFIR and (2) validate computer models used to perform design- and safety-basis calculations. This report also describes the test facility and experimental data, and it provides a comparison of the experimental data to computer simulations. A total of 51 tests have been completed: four tests with pure steam, 12 tests with argon, and 35 tests with helium. A total of 10 tests were performed at subatmospheric pressure, and four of these were performed with pure steam. One test was conducted at a high power of 92.7 kW, six tests were HFIR startups, and two tests were HFIR loss of offsite power (LOOP). Pressures up to 10 MPa, vapor temperatures up to 583 K (310°C), and heater temperatures above 600 K (327°C) have been reached in these tests. Two computer programs, RELAP5-3D and TRACE, have been used to simulate the tests. The TRACE code has shown good agreement with the test data and has been used to model a variety of tests. This experimental facility has been very useful in demonstrating the viability of this new type of irradiation facility.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1967

Ceramic research

Load measuring equipment for mechanical testing facility

LOAD TESTING MACHINE↗