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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 1,603 records · Page 89
All sourcesNASA NTRSDOE OSTI
NASA NTRS2001

Scramjet Tests in a Shock Tunnel at Flight Mach 7, 10, and 15 Conditions

Tests of the Hyper-X scramjet engine flowpath have been conducted in the HYPULSE shock tunnel at conditions duplicating the stagnation enthalpy at flight Mach 7, 10, and 15. For the tests at Mach 7 and 10 HYPULSE was operated as a reflected-shock tunnel; at the Mach 15 condition, HYPULSE was operated as a shock-expansion tunnel. The test conditions matched the stagnation enthalpy of a scramjet engine on an aerospace vehicle accelerating through the atmosphere along a 1000 psf dynamic pressure trajectory. Test parameter variation included fuel equivalence ratios from lean (0.8) to rich (1.5+); fuel composition from pure hydrogen to mixtures of 2% and 5% silane in hydrogen by volume; and inflow pressure and Mach number made by changing the scramjet model mounting angle in the HYPULSE test chamber. Data sources were wall pressures and heat flux distributions and schlieren and fuel plume imaging in the combustor/nozzle sections. Data are presented for calibration of the facility nozzles and the scramjet engine model. Comparisons of pressure distributions and flowpath streamtube performance estimates are made for the three Mach numbers tested.

Rogers, R. C.↗
DOE OSTI2023

Nuclear fuel irradiation testbed for nuclear security applications

The nuclear security community has long been interested in the identification and quantification of nuclear material signatures to understand a material’s provenance, use, and ultimate application. New forensics signatures and methods intended for non-traditional or advanced nuclear fuel applications require fuel irradiation experiments to demonstrate viability and validity. Integral fuel irradiations have historically required significant costs and long timelines to design, irradiate, and characterize. This paper describes how a recently developed nuclear fuel irradiation testbed can be used to provide a low cost, rapid turnaround, modular test environment for irradiation and evaluation of nuclear fuel specimens for nuclear security applications. The irradiation testbed houses six small ‘MiniFuel’ samples within hermetically sealed capsules inside targets that can be removed in between each ∼25-day operating cycle of the High Flux Isotope Reactor (HFIR). As many as nine targets can be irradiated using a single irradiation position (reflector region) in HFIR, allowing for varying irradiation temperatures and burnups. A suite of hot cell capabilities have been established to perform post-irradiation examination for measuring performance (e.g., fuel swelling, fission gas release) and facilitating experiment disassembly for subsequent property measurements, microstructural analysis, or chemical assay. This new testbed allows fuel irradiations to be conducted on an accelerated timeframe to enable rapid proof of concept testing and to provide reference material for nuclear fuel security applications. Recent applications using this testbed include the testing of isotopic taggants in UO 2 fuel (intentional forensics), testing of U-10Mo fuel for down-conversion of highly enriched uranium–fueled reactors, and the production of irradiated UO 2 fuel material for signature analysis of its isotopic composition (plutonium, fission gases, etc.).

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2019

OVERVIEW OF ALUMINUM CLADDING OXIDE PREDICTION MODELS FOR EUROPEAN HIGH POWER RESEARCH REACTORS

UMo/Al dispersion fuel clad with aluminum alloy is a primary candidate fuel form that is being developed for the European high power research reactors (EUHPRR). In BR2 tests, water oxidation of cladding grew thick due to high powers. Excessive cladding oxidation is a cause for serious fuel performance degradation in a test while fuel performance itself may be sound. It is necessary for a fuel designer to have a reliable model to predict cladding oxidation for fuel licensing activities. Using oxide data from BR2 tests, this study reviews the prominent oxide prediction models and cladding-to-coolant heat transfer correlations. Because the oxide growth models were based upon different coolant flow conditions from those of the EUHPRR, the effect of the heat transfer correlations at the cladding-to-coolant interface on the oxide thickness prediction were also examined. The best combination of the oxide prediction model and the heat transfer correlation applicable for the EUHPRR is recommended. Since the existing oxide models were all developed for aluminum alloy types different from those currently considered for the EUHPRR, the effect of cladding alloy type is also discussed.

Research reactor fuel, aluminum oxidation, in-pile↗
DOE OSTI2016

Fabrication of rodlets with coated and lined Mo-alloy cladding for testing and irradiation

Molybdenum alloy fuel cladding with an outer protective layer has been proposed to achieve enhanced tolerance to severe loss of coolant accidents for light water reactors. The accident tolerant fuel cladding will need to meet stringent fuel reliability requirements and significantly delay steam reaction during a severe accident to provide plant operators additional coping time to recover from the accident. A laboratory measurable target for this feasibility study with coated Mo cladding is to demonstrate its capability to survive in steam and maintain fuel rods in coolable geometry at 1200-1500 Celsius degrees for at least 24 hours. Preliminary studies indicate such a capability may provide additional coying of 4-19 hours. A feasibility study undertaken since 2012 has demonstrated feasibility of fabricating long, thin-wall Mo and Mo-ODS tubes, and forming a metallurgically bonded protective coating of FeCrAl or Zr-alloy via physical vapor deposition to provide protection of the Mo cladding in 1200 C. degrees for 24 hours. The diametral ductility of the Mo and Mo-ODS thin wall tubes can be increased to 5-15% pre-irradiation to meet fuel operational requirements. Fabrication of the outer protective layer using thermo-mechanical processing, such as hot isostatic pressing and hot coextrusion, has been under evaluation to achieve high quality lined Mo cladding from commercial scale production. Welding of lined Mo tubes to end-caps are being developed using a combination of solid state welding of laser or electron beam seal welding. Rodlets of lined Mo tubes are being fabricated for further testing to validate its reliability under normal operation and capability of accident tolerance to the 1200- 1500 C. degrees range. Test reactor irradiation is also being planned. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1996

A performance comparison of two small rocket nozzles

An experimental study was conducted on two small rockets (110 N thrust class) to directly compare a standard conical nozzle with a bell nozzle optimized for maximum thrust using the Rao method. In large rockets, with throat Reynolds numbers of greater than 1 x 10(exp 5), bell nozzles outperform conical nozzles. In rockets with throat Reynolds numbers below 1 x 10(exp 5), however, test results have been ambiguous. An experimental program was conducted to test two small nozzles at two different fuel film cooling percentages and three different chamber pressures. Test results showed that for the throat Reynolds number range from 2 x 10(exp 4) to 4 x 10(exp 4), the bell nozzle outperformed the conical nozzle. Thrust coefficients for the bell nozzle were approximately 4 to 12 percent higher than those obtained with the conical nozzle. As expected, testing showed that lowering the fuel film cooling increased performance for both nozzle types.

Arrington, Lynn A.↗
DOE OSTI2024

Sister Rod Destructive Examinations (FY22) Appendix E: Mechanical Testing

As a part of the DOE NE High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 GWd/MTU) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called sister rods or sibling rods, are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin Zirc-4, ZIRLO, and M5. The DEs are being conducted to obtain a baseline of the HBU rod’s condition before dry storage and are focused on understanding overall SNF rod strength and durability. Composite fuel and defueled cladding will be tested to derive material properties. Although the data generated can be used for multiple purposes, one primary goal for obtaining the post-irradiation examination data and the associated measured mechanical properties is to support SNF dry storage licensing and relicensing activities by (1) addressing identified knowledge gaps and (2) enhancing the technical basis for post-storage transportation, handling, and subsequent disposition.This appendix documents the status of the ORNL Phase 1 DE activities related to the mechanical testing of selected sister rods in Phase 1 of the sister rod test program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Irradiation performance of a U-7Mo in Al-Si matrix dispersion full-size fuel plate assembly

The Korea Atomic Energy Research Institute (KAERI) is leading the Ki-Jang Research Reactor (KJRR) project with the intent to develop a new reactor for medical isotope production and other nuclear research purposes. The KJRR core is designed to use high density fuel system where uranium alloyed with 7 wt% molybdenum (U-7Mo) particles are dispersed in a matrix of aluminum alloyed with 5 wt% silicon (Al-5Si) and clad in aluminum alloy 6061 (Al-6061) to form fuel plates. KAERI developed a fabrication facility to construct KJRR fuel assemblies and partnered with the Idaho National Laboratory (INL) to irradiate a full-size fuel assembly, with 21 total fuel plates, in the Advanced Test Reactor (ATR). Irradiation testing and subsequent Post Irradiation Exam (PIE) campaigns were performed successfully over a multi-year project. Monte Carlo neutronic calculations, coupling with a depletion code, were performed based on ATR’s as-run power history which showed that the highest power plate (plate 20) reached 83.1 % end-of-life (EOL) local burnup based on initial 235 U content. Additionally, finite element thermal modeling was performed based as-run power history which showed a beginning-of-life (BOL) peak local heat flux of 184 W/cm 2 . No anomalous fuel performance was observed during the irradiation and target test conditions were achieved. PIE showed favorable performance of the fuel assembly regarding all important phenomena. This paper describes the KJRR fuel assembly irradiation conditions and PIE data to support the conclusion that it performed well, without evidence of unexpected or problematic fuel performance, within an irradiation test designed to bound the KJRR design environment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Flowsheet Development for the University of Tokyo YAYOI (UTY) Fuel

The Savannah River National Laboratory (SRNL) was requested by H-Canyon Engineering to determine the flowsheet parameters needed to dissolve and store the YAYOI Material Test Reactor (MTR) uranium fuel safely and efficiently. In response to this request, a literature review of existing SRNL MTR dissolution flowsheets and general fuel dissolution literature in the nuclear fuel processing industry was performed to evaluate chemical dissolution parameters required to dissolve the YAYOI fuel and the tin-plated carbon steel product cans (PC). Based on past dissolutions of similar fuel in the H-Canyon and open literature reviews on chemical dissolution of spent nuclear fuel (SNF), the following conclusions and flowsheet recommendations were made.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1982

Low and medium heating value coal gas catalytic combustor characterization

Catalytic combustion with both low and medium heating value coal gases obtained from an operating gasifier was demonstrated. A practical operating range for efficient operation was determined, and also to identify potential problem areas were identified for consideration during stationary gas turbine engine design. The test rig consists of fuel injectors, a fuel-air premixing section, a catalytic reactor with thermocouple instrumentation and a single point, water cooled sample probe. The test rig included inlet and outlet transition pieces and was designed for installation into an existing test loop.

Schwab, J. A.↗
NASA NTRS1984

Fuel property effects on USAF gas turbine engine combustors and afterburners

Since the early 1970s, the cost and availability of aircraft fuel have changed drastically. These problems prompted a program to evaluate the effects of broadened specification fuels on current and future aircraft engine combustors employed by the USAF. Phase 1 of this program was to test a set of fuels having a broad range of chemical and physical properties in a select group of gas turbine engine combustors currently in use by the USAF. The fuels ranged from JP4 to Diesel Fuel number two (DF2) with hydrogen content ranging from 14.5 percent down to 12 percent by weight, density ranging from 752 kg/sq m to 837 kg/sq m, and viscosity ranging from 0.830 sq mm/s to 3.245 sq mm/s. In addition, there was a broad range of aromatic content and physical properties attained by using Gulf Mineral Seal Oil, Xylene Bottoms, and 2040 Solvent as blending agents in JP4, JP5, JP8, and DF2. The objective of Phase 2 was to develop simple correlations and models of fuel effects on combustor performance and durability. The major variables of concern were fuel chemical and physical properties, combustor design factors, and combustor operating conditions.

Reeves, C. M.↗
DOE OSTI2016

Instrumentation Development for Transient Testing in TREAT

Development of nuclear fuels requires experimentation and modeling for behavior during transient power and cooling conditions. The Transient Reactor Test facility (TREAT) will provide unique capability in the U.S. to perform controlled experimentation of such conditions in a nuclear environment. To support experiment objectives and provide validation data for modeling and simulation tools, experiment instrumentation plays a critical role in monitoring test specimen behavior (temperature, dimensional changes, relocation, microstructural evolution, etc.) and specimen boundary conditions (thermal, mechanical, nuclear) at various length and time scales with the ultimate goal of reducing uncertainties in experiment conditions. This summary presents an overview of current instrument development objectives and progress for TREAT experiments with emphasis on near-term activities to support Reactivity Initiated Accident (RIA) experiments in LWR systems. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Fuel element design and analysis for potential LEU conversion of the Advanced Test Reactor

This study provides a summary of the status of the design process for a low-enrichment replacement for the high-enrichment fuel that has historically been used in the Advanced Test Reactor (ATR) at Idaho National Laboratory. Under current long-term Department of Energy policy, the ATR will be converted to operate using low-enrichment fuel by 2030. To this end, an engineering evaluation and design process has been underway for more than ten years. Initially performed as a series of scoping studies for different design options, the last five years have seen a transition to a formal Safety-in-Design approach to the design and safety evaluation process. This paper focuses on such efforts in progress since 2012 to identify pre-conceptual and conceptual design candidates for ATR conversion. The Enhanced LEU Fuel (ELF) design is selected as the most promising conceptual design. Analyses and trade-off studies performed that led to that selection are described. Analysis of certain aspects of core performance for an ELF-fueled core relative to the current high-enrichment fuel show that only small and manageable changes to core operations will be necessary to support conversion of ATR to low-enrichment fuel.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Data Challenges in Multi-Sensor Data Science System for Monitoring a Solvent Extraction Process

Idaho National Laboratory (INL) is maintaining and gaining knowledge into the nuclear fuel cycle by building a test bed to allow researchers the opportunity to study nuclear fuel processing operations. This includes studying solvent extraction processes that use centrifugal contactors. As part of INL’s mission, the goal of this project is to develop a system that utilizes non-traditional measurement sources such as vibration, acoustics, current, light, flow, and temperature in conjunction with data-based, machine learning techniques that will allow for signal discovery. This multisensory data can support the development of safeguards by design, provide operator process awareness, and discover process anomalies. This poster will highlight some of the data collection and analytics challenges for the multi-sensor system as well as the mitigation strategies to build a robust system. Additionally, some preliminary data from the first testing campaign will be shown to help illustrate the data needs of the system.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
NASA NTRS1980

Atomizing characteristics of swirl can combustor modules with swirl blast fuel injectors

Cold flow atomization tests of several different designs of swirl can combustor modules were conducted in a 7.6 cm diameter duct at airflow rates (per unit area) of 7.3 to 25.7 g/sq cm sec and water flow rates of 6.3 to 18.9 g/sec. The effect of air and water flow rates on the mean drop size of water sprays produced with the swirl blast fuel injectors were determined. Also, from these data it was possible to determine the effect of design modifications on the atomizing performance of various fuel injector and air swirler configurations. The trend in atomizing performance, as based on the mean drop size, was then compared with the trends in the production of nitrogen oxides obtained in combustion studies with the same swirl can combustors.

Ingebo, R. D.↗
DOE OSTI2020

Uranium nitride tristructural-isotropic fuel particle

A summary of activities around concept development, design, manufacturing, and early irradiation testing of a novel coated fuel particle, uranium nitride tristructural-isotropic fuel is provided. This fuel particle offers significantly higher uranium density over historic manifestations of coated fuel particles and may be more optimal for a range of advanced reactor applications. After reviewing the design process, steps involving production of this fuel form are discussed. Low burnup irradiation data on this fuel particle are now available, indicating that performance metrics have been met to date based on the original design expectations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2010

Computational Analysis of Dynamic SPK(S8)-JP8 Fueled Combustor-Sector Performance

Civil and military flight tests using blends of synthetic and biomass fueling with jet fuel up to 50:50 are currently considered as "drop-in" fuels. They are fully compatible with aircraft performance, emissions and fueling systems, yet the design and operations of such fueling systems and combustors must be capable of running fuels from a range of feedstock sources. This paper provides Smart Combustor or Fuel Flexible Combustor designers with computational tools, preliminary performance, emissions and particulates combustor sector data. The baseline fuel is kerosene-JP-8+100 (military) or Jet A (civil). Results for synthetic paraffinic kerosene (SPK) fuel blends show little change with respect to baseline performance, yet do show lower emissions. The evolution of a validated combustor design procedure is fundamental to the development of dynamic fueling of combustor systems for gas turbine engines that comply with multiple feedstock sources satisfying both new and legacy systems.

Ryder, R.↗