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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 271 records · Page 15

An Experimental Investigation of Multilayer Flow in a Slide Die Coating Process

Roll-to-roll coating technology has been actively applied not only for conventional film production but also for emerging energy devices such as organic solar cells, organic light emitting diodes, batteries, smart windows, and fuel cells. The devices are multilayer structures requiring several coating and drying steps when implemented in a serial fashion, which increases capital costs and energy consumption. Simultaneous multilayer (ML) coating technologies present an opportunity for cost reductions and improved process efficiency by reducing equipment footprint and the number of process steps. In this talk, we experimentally investigate ML slide die coating flow for the manufacturing of fuel cell membrane electrode assemblies (MEAs). For ML slide coating, understanding fluid properties such as density, surface tension, and viscosity are necessary for stable flows and limiting intermixing of the miscible layers. Here, we utilize a design-of-experiment methodology - Box-Behnken design (BBD) - to correlate ink formulation parameters to fluid properties. Using the ink properties obtained from the BBD, we performed numerical flow simulations to obtain a proper process conditions to guide experimental coating trials. Finally, experimental verification is performed on a ML slide die and a degree of miscibility was observed for various ink formulation conditions.

28 EE - Advanced Manufacturing Office (EE-5A)↗

Feasibility Study for a Subcritical Assembly at the Oak Ridge National Laboratory

A conceptual design has been completed for a new subcritical assembly at Oak Ridge National Laboratory (ORNL). This subcritical assembly will support the Nuclear Criticality Safety Program (NCSP) training and education program to augment current subcritical assembly capabilities at Sandia National Laboratories (SNL) and at the National Criticality Experiments Research Center (NCERC). The proposed subcritical assembly uses legacy AGN-201M research reactor fuel plates that are available at the Y-12 National Security Complex. This subcritical assembly contains approximately 620 grams of 235 U as UO 2 particulates distributed homogeneously in radiation-stabilized polyethylene. The fuel plates will have a graphite neutron reflector to obtain a core multiplication, $M$, from 10 to 50, corresponding to a k eff of 0.9 to 0.95, respectively. The subcritical assembly will be able to support at least four experiments for the training courses: the addition of fissile material to increase neutron multiplication (mass), a core separation experiment (interaction), the effect of adding moderators to the core (moderation), and the effect of adding neutron absorbers to the core (poison/absorption). The proposed ORNL subcritical assembly will provide additional student bandwidth to support NCSP training courses and to provide for some backup capacity in the event that nuclear facility operations are disrupted at SNL or NCERC due to some unforeseen issue. The subcritical assembly will be designed to be an inherently safe subcritical assembly—subcritical under all normal and abnormal conditions—and it will provide the capability to conduct hands-on training to support NCSP and general nuclear criticality safety (NCS) staff training and qualification goals.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Preliminary NBSR Design-Demonstration Element Thermal-Hydraulics and Structural Analyses

1.1 OBJECTIVE The United State High Performance Research Reactor (USHPRR) program aims to eliminate more than 200kg of High Enriched Uranium (HEU) from commerce annually by converting five U.S. high-performance research reactors and one associated critical assembly to Low Enriched Uranium (LEU) fuel using a high-density alloy of uranium-10 wt% molybdenum (U-10Mo). The National Bureau of Standards Reactor (NBSR) is one of five research reactors selected for this program. The objective of this report is to provide preliminary thermal-hydraulic and mechanical analyses of the hydrodynamic effects in the NBSR Design Demonstration Element (DDE) under conservative approximations for the plate power distribution. This report provides details on the modeling approach and the simulation results obtained, including pressure, flow velocity, temperature, and oxide layer over the design demonstration experiment for the irradiation cycles in the Belgian Reactor (BR)-2.

42 ENGINEERING↗

Design of a Resistively Heated Thermal Hydraulic Simulator for Nuclear Rocket Reactor Cores

A preliminary design study is presented for a non-nuclear test facility which uses ohmic heating to replicate the thermal hydraulic characteristics of solid core nuclear reactor fuel element passages. The basis for this testing capability is a recently commissioned nuclear thermal rocket environments simulator, which uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce high-temperature pressurized hydrogen flows representative of reactor core environments, excepting radiation effects. Initially, the baseline test fixture for this non-nuclear environments simulator was configured for long duration hot hydrogen exposure of small cylindrical material specimens as a low cost means of evaluating material compatibility. It became evident, however, that additional functionality enhancements were needed to permit a critical examination of thermal hydraulic effects in fuel element passages. Thus, a design configuration was conceived whereby a short tubular material specimen, representing a fuel element passage segment, is surrounded by a backside resistive tungsten heater element and mounted within a self-contained module that inserts directly into the baseline test fixture assembly. With this configuration, it becomes possible to create an inward directed radial thermal gradient within the tubular material specimen such that the wall-to-gas heat flux characteristics of a typical fuel element passage are effectively simulated. The results of a preliminary engineering study for this innovative concept are fully summarized, including high-fidelity multi-physics thermal hydraulic simulations and detailed design features.

Litchford, Ron J.↗

Preliminary Neutronics Design and Analysis of the Fast Modular Reactor

General Atomics is developing a new 100-MW(thermal) fast modular reactor (FMR) that provides safe, carbon-free electricity and is capable of incremental capacity additions. The modular design allows it to be factory built and assembled onsite to keep the capital cost low, while the use of dry cooling facilitates siting to complement renewables in nearly any location. The FMR uses high-assay low-enriched uranium-dioxide fuel encapsulated by recognized irradiationresistant silicon carbide composite (SiGA®) cladding that is derisked in the current accident-tolerant fuel program. The FMR fuel assembly is a hexagonal fuel bundle of 120 fuel rods. The total length of the fuel assembly is less than 4 m, with an active fuel length of 1.8 m. The fuel assemblies are configured in an annular core that is located and supported by the reactor internals. The coolant material is helium at a normal operating pressure of 7 MPa. The core is surrounded by zirconium silicide (Zr 3 Si 2 ) and graphite reflector blocks. The fuel, coolant, internals, and reflectors are contained within a reactor pressure vessel. Here, the preliminary nuclear design and analysis established the arrangement of the active core and reflector blocks. The nuclear design analyses of the FMR defined the design parameters, such as fuel enrichments, excess reactivity, fueling scheme, fuel cycle, power distribution, and control rod worth. The preliminary conceptual design determined the three-batch fueling scheme with the allowable total power peaking factor of 1.5. The average discharge burnup is 100 GW days per ton of uranium.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Brayton cycle solarized advanced gas turbine

Described is the development of a Brayton Engine/Generator Set for solar thermal to electrical power conversion, authorized under DOE/NASA Contract DEN3-181. The objective was to design, fabricate, assemble, and test a small, hybrid, 20-kW Brayton-engine-powered generator set. The latter, called a power conversion assembly (PCA), is designed to operate with solar energy obtained from a parobolic dish concentrator, 11 meters in diameter, or with fossil energy supplied by burning fuels in a combustor, or by a combination of both (hybrid model). The CPA consists of the Brayton cycle engine, a solar collector, a belt-driven 20-kW generator, and the necessary control systems for automatic operation in solar-only, fuel-only, and hybrid modes to supply electrical power to a utility grid. The original configuration of the generator set used the GTEC Model GTP36-51 gas turbine engine for the PCA prime mover. However, subsequent development of the GTEC Model AGT101 led to its selection as the powersource for the PCA. Performance characteristics of the latter, thermally coupled to a solar collector for operation in the solar mode, are presented. The PCA was successfully demonstrated in the fuel-only mode at the GTEC Phoenix, Arizona, facilities prior to its shipment to Sandia National Laboratory in Albuquerque, New Mexico, for installation and testing on a test bed concentractor (parabolic dish). Considerations relative to Brayton-engine development using the all-ceramic AGT101 when it becomes available, which would satisfy the DOE heat engine efficiency goal of 35 to 41 percent, are also discussed in the report.

Source record↗

Development of an alkaline fuel cell subsystem

A two task program was initiated to develop advanced fuel cell components which could be assembled into an alkaline power section for the Space Station Prototype (SSP) fuel cell subsystem. The first task was to establish a preliminary SSP power section design to be representative of the 200 cell Space Station power section. The second task was to conduct tooling and fabrication trials and fabrication of selected cell stack components. A lightweight, reliable cell stack design suitable for the SSP regenerative fuel cell power plant was completed. The design meets NASA's preliminary requirements for future multikilowatt Space Station missions. Cell stack component fabrication and tooling trials demonstrated cell components of the SSP stack design of the 1.0 sq ft area can be manufactured using techniques and methods previously evaluated and developed.

Source record↗

Tomographic Image Reconstruction for the Parallel-Slit Ring Collimator Fast Neutron Emission Tomography System

For the past three years, Oak Ridge National Laboratory (ORNL) has been developing a passive fast-neutron emission tomography capability. The goal of this development is the ability to quantify the neutron source strength of individual fuel pins (rods) in spent nuclear fuel assemblies. Such a system could be used to measure the burnup of each fuel pin in a spent fuel assembly in order to take burnup credit when loading dry storage casks or to count individual fuel pins in spent fuel assemblies for safeguards purposes. At present, a laboratory prototype imager is under construction. The purpose of this prototype is to demonstrate imaging capability sufficient to resolve individual fuel pins in spent fuel assemblies. This report documents the development of the iterative reconstruction code used to perform tomographic image reconstruction, the imager response calculation used by the reconstruction code, and the results of reconstructions of simulated tomographic imaging measurements for the prototype imager design.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Guidance Regarding Probable Approaches and Instrumentation to Liquid Fuel Molten Salt Reactor Material Control and Accounting (MC&A) Rev. 1

With the rapid development of advanced reactors and numerous companies requesting pre-application engagements with the Nuclear Regulatory Commission (NRC), there is need for the NRC to best prepare themselves for new challenges presented by advanced technologies. One specific reactor type of interest is the molten salt reactor (MSR) that uses liquid salt as the fuel for the reactor. Liquid salt fuel reactors have specific challenges with regards to performing material control and accounting (MC&A). The challenges result from the fuel form being a continuous fissile material form rather than discrete units like is found with fuel assemblies in traditional light water reactors (LWRs). The goal of this document is to provide the NRC with guidance on the different approaches and technologies that may be proposed by reactor designers to address these problems to prepare staff for regulatory reviews.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Thermal analyses of high-power advanced thermoacoustic radioisotope power system for future space exploration missions

This paper presents the results of 3-D thermal analyses of the heat source assembly of a 440 We advanced radioisotope power system (RPS) for future space exploration missions. This high thermal efficiency RPS employs a heat source assembly of eight Step 2 General Purpose Heat Source (GPHS) modules and four Thermoacoustic Power Converter (TAPC) units. Each unit is thermally coupled to two GPHS modules. Here, the analyses investigated the temperatures of the Iridium (Ir) alloy cladding of the 238 PuO 2 fuel pellets and the surface of the Fine Weave Pierced Fabric (FWPF) aeroshell in the GPHS modules to ensure reliable operation and safety in case of an unlikely reentry. Conductive coupling of the GPHS modules to the heater heads of the TAPC units produces relatively uniform axial and azimuthal distributions of heat flux and temperature. However, for this case the temperature of the Ir alloy cladding is lower than the desired values of 1173 K to maintain sufficient ductility. Adding a 5 mm wide He filled gap and decreasing the thickness of the graphite sleeve to 10 mm increased Ir cladding temperature to 1197–1174 K, which are in the desired range. The maximum surface temperatures of the FWPF graphite aeroshell of 1022–1131 K is well below the NASA specified limit of 1373 K. In addition, the total mass of the heat source assembly decreased by 19.6 kg, ~43% saving compared to original design.

42 ENGINEERING↗

Lunar Surface Access Module Pump-Fed Engine Turbopump Technology

The need for a high specific impulse LOX/LH2 pump-fed lunar lander engine has been established by NASA for the new Exploration architecture. Preliminary studies indicate that a 4 engine cluster in the thrust range of 9,000-lbf each is a likely configuration for the main propulsion of the manned lunar lander vehicle. The main Lunar Surface Access Module engines will likely be responsible for mid-course correction burns, lunar orbit insertion burns, a deorbit burn, and the powered descent to the lunar surface. This multi-task engine philosophy imposes a wide throttling requirement on the engines in the range of 10:1. Marshall Space Flight Center has initiated an internal effort to mature the technologies needed for full scale development of such a LOX/LH2 pump-fed engine. In particular, a fuel turbopump is being designed and fabricated at MSFC to address the issues that a small high speed turbopump of this class will face. These issues include adequate throttling performance of the pump and turbine over a very wide operating range. The small scale of the hardware presents issues including performance scaling, and manufacturing issues like that will challenge the traditional methods we have used to fabricate and assemble larger scale turbopumps. The small high speed turbopump being developed at MSFC will operate at speeds greater than 100,000-rpm. These speeds create issues that include structural dynamics and high cycle fatigue as well as rotordynamic stability. The fuel turbopump development at MSFC will address these issues, and plans are in work for component level testing as well as operation in a test bed engine environment. The fuel turbopump design is nearing completion and described herein.

Thornton, Randall J.↗

IER-297: TEX-HEU Baseline Assemblies Benchmark and Results [Slides]

The pending resolution of comments was accepted by the ICSBEP TRG in December 2021. Subgroup includes Catherine Percher, Dave Heinrichs, Michael Zerkle (NNL), and Jeffrey Favorite (LANL). The full resolution of comments is anticipated by March 31, 2022. Plans to execute TEX-HEU-Hf variant of experiment design are anticipated for FY22.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Subcritical Assembly at ORNL [Slides]

This presentation discusses the US Department of Energy (DOE) Nuclear Criticality Safety Program (NCSP), specifically regarding the training and education courses. Also discussed is the AGN-201M reactor and fuel disk. Additionally, the ORSA conceptual design and design changes are presented. In conclusion, the ORSA design has moved from conceptual to implementable. ORSA hands-on experiments will remain subcritical under normal and credible abnormal conditions. The ORSA location will have minimal security and safety basis impact and is feasible for a diversity of students. The implementation of ORSA will improve upon NCSP T&E’s vision to remain adaptable and responsive.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Cell module and fuel conditioner

Measurements of stack height changes with temperature and cell material characteristics were made. Stack 559 was assembled and components were fabricated for 560, 561, and 562. Stack 425 was transferred from the parallel DOE program and installed in the OS/IES simulation loop for mechanical and electrical testing. Construction and preliminary checkout of the 2 kW test facility was completed and design and procurement of the 8 kW test facility was initiated. The fuel conditioning subsystem design continued to evolve and the state points for the current design were calculated at full and part load conditions. Steam reforming catalyst activity tests were essentially completed and aging tests and CO shift converter tests were initiated.

Hoover, D. Q., Jr.↗

COBRA-SFS Version 6.2.1 User Guide: A Thermal Hydraulic Analysis Code for Spent Fuel Storage and Transportation Casks

This document is an updated and reduced version of the complete documentation of the COBRA-SFS thermal analysis code (COolant Boiling in Rod Arrays – Spent Fuel Storage) (Michener et al. 1995). This manual provides the analyst with input instructions and fundamental guidelines for contrasting models using COBRA-SFS. COBRA-SFS is designed to analyze the dry storage and transportation casks that house spent or used nuclear fuel. The primary focus and application is on square light water reactor fuel assemblies, however the code also has the capability to model hexagonal assemblies and is will suited to any geometry that can be defined as an extruded 2-D map.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Energy efficient engine: Turbine intermediate case and low-pressure turbine component test hardware detailed design report

A four stage, low pressure turbine component has been designed to power the fan and low pressure compressor system in the Energy Efficient Engine. Designs for a turbine intermediate case and an exit guide vane assembly also have been established. The components incorporate numerous technology features to enhance efficiency, durability, and performance retention. These designs reflect a positive step towards improving engine fuel efficiency on a component level. The aerodynamic and thermal/mechanical designs of the intermediate case and low pressure turbine components are presented and described. An overview of the predicted performance of the various component designs is given.

Leach, K.↗

3-D CFD Simulation and Validation of Oxygen-Rich Hydrocarbon Combustion in a Gas-Centered Swirl Coaxial Injector using a Flamelet-Based Approach

Injector design is a critical part of the development of a rocket Thrust Chamber Assembly (TCA). Proper detailed injector design can maximize propulsion efficiency while minimizing the potential for failures in the combustion chamber. Traditional design and analysis methods for hydrocarbon-fuel injector elements are based heavily on empirical data and models developed from heritage hardware tests. Using this limited set of data produces challenges when trying to design a new propulsion system where the operating conditions may greatly differ from heritage applications. Time-accurate, Three-Dimensional (3-D) Computational Fluid Dynamics (CFD) modeling of combusting flows inside of injectors has long been a goal of the fluid analysis group at Marshall Space Flight Center (MSFC) and the larger CFD modeling community. CFD simulation can provide insight into the design and function of an injector that cannot be obtained easily through testing or empirical comparisons to existing hardware. However, the traditional finite-rate chemistry modeling approach utilized to simulate combusting flows for complex fuels, such as Rocket Propellant-2 (RP-2), is prohibitively expensive and time consuming even with a large amount of computational resources. MSFC has been working, in partnership with Streamline Numerics, Inc., to develop a computationally efficient, flamelet-based approach for modeling complex combusting flow applications. In this work, a flamelet modeling approach is used to simulate time-accurate, 3-D, combusting flow inside a single Gas Centered Swirl Coaxial (GCSC) injector using the flow solver, Loci-STREAM. CFD simulations were performed for several different injector geometries. Results of the CFD analysis helped guide the design of the injector from an initial concept to a tested prototype. The results of the CFD analysis are compared to data gathered from several hot-fire, single element injector tests performed in the Air Force Research Lab EC-1 test facility located at Edwards Air Force Base.

Richardson, Brian↗

Integral Experiment Request 554: CED-2 Summary Report

This report summarizes the critical experiment design 2 (CED-2) or final design study for Integral Experiment Request 554 (IER-554). This experiment describes the effects and importance of adding a commercially available neutron absorber material to a known light-water low-enriched uranium assembly. The assembly in question is the Seven Percent Critical Experiments (7uPCX) at Sandia National Laboratories. The neutron absorber plates considered for the experiment are known as Boralcan and are made of boron carbide (B 4 C) particles embedded in 1100 aluminum alloy. The concentrations of B 4 C and 1100 aluminum were changed as well as the thickness and the positions of the plates, and the fuel rod configuration was adapted to ensure that the assembly would be critical in each case studied. A total of 6 critical configurations, each with a neutron absorber plate inserted, are described in this report. No results of high-quality integral experiments involving neutron absorbers made with B 4 C and 1100 aluminum plates are currently publicly available. Sensitivity to the neutron absorber plate material definition, isotopes, and cross sections of specific regions and configurations are also analyzed. As part of the final design of the experiment, a safe plate holder made of 3003 aluminum was designed. The study results indicate that these experiments are achievable with sufficiently low uncertainties and minimal modification to the assembly, and the CED-3 phase can start when funding allows.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗