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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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All sourcesNASA NTRSDOE OSTI
NASA NTRS2010

The NASA Langley Isolator Dynamics Research Lab

The Isolator Dynamics Research Lab (IDRL) is under construction at the NASA Langley Research Center in Hampton, Virginia. A unique test apparatus is being fabricated to support both wall and in-stream measurements for investigating the internal flow of a dual-mode scramjet isolator model. The test section is 24 inches long with a 1-inch by 2-inch cross sectional area and is supplied with unheated, dry air through a Mach 2.5 converging-diverging nozzle. The test section is being fabricated with two sets (glass and metallic) of interchangeable sidewalls to support flow visualization and laser-based measurement techniques as well as static pressure, wall temperature, and high frequency pressure measurements. During 2010, a CFD code validation experiment will be conducted in the lab in support of NASA s Fundamental Aerodynamics Program. This paper describes the mechanical design of the Isolator Dynamics Research Lab test apparatus and presents a summary of the measurement techniques planned for investigating the internal flow field of a scramjet isolator model.

Middleton, Troy F.↗
DOE OSTI2019

Test Plan for the Horizontal Dry Cask Simulator

The thermal performance of commercial spent nuclear fuel dry storage casks is evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission(NRC). Canistered dry storage cask systems rely on ventilation between the inner canister and the overpack to convect heat away from the canister to the surrounding environment for both horizontal and vertical configurations. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. While several testing programs have been previously conducted, these earlier validation studies did not integrate all the physics or components important in a modern, horizontal dry cask system. The purpose of the investigation described in this test plan is to produce data sets that can be used to benchmark the codes and best practices presently used to determine cladding temperatures and induced cooling air flows in modern horizontal dry storage systems. The horizontal dry cask simulator(HDCS) has been designed to generate this benchmark data and add to the existing knowledgebase. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure(MAWP)rating of 2,400 kPa at400 °C. An existing electrically heated but otherwise prototypic boiling water reactor(BWR), Incoloy-clad test assembly will be deployed inside of a representative storage basket and canister. An insulated sheet metal enclosure will be used to mimic the thermal properties of the concrete vault enclosure used in a modern horizontal storage system. Radial and axial temperature profiles along with induced cooling air flow will be measured for a wide range of decay powers and representative(and higher)cask pressures using various backfills of helium, argon, or air. The single assembly geometry with well-controlled boundary conditions simplifies computational requirements while preserving relevant physics. The proposed test apparatus integrates all the underlying thermal-hydraulics important to defining the performance of a modern horizontal storage system. These include combined-mode heat transfer from the electrically-heated assembly to the canister walls and the primarily natural-convective heat transfer from the canister to the cooling air flow passing through the horizontal vault enclosure. The objective of the HDCS is not to reproduce the performance of a commercial dry storage system for any given set of operational parameters. Rather ,the objective is to capture the dominant physics in a well-characterized test apparatus. The close coupling between the thermal response of the canister system and the resulting induced cooling air flow rate is of particular importance. While incorporating the best available information based on thermal-hydraulic scaling arguments as well as previous vertical testing, this test plan is subject to changes due to improved understanding or from as built deviations to designs. As-built conditions and actual procedures will be documented in the final test report.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2018

IV&V Planning & Execution Initiative

Independent verification and validation (IV&V) is a systems engineering discipline which seeks to provide confidence that mission-critical software-dependent systems will operate safely and achieve all mission objectives. In order to optimize application of IV&V resources, it is necessary to identify specific assurance objectives that are most likely to reveal difficulty and therefore to increase confidence. Software assurance literature deals predominantly with the code itself. However, our experience is that the most significant software-related issues are discovered via analysis of other system artifacts such as operations concepts, requirements and design. Consequently, it is desirable to, when assessing risk and defining assurance objectives, consider all facets of the software system architecture. We describe in this presentation progress and lessons learned from an IV&V capability development initiative to enhance the IV&V project planning process using all available architecture information.

Risk↗
DOE OSTI2017

Validation of the Computational Fluid Dynamics Method using the Aboveground Configuration of the Dry Cask Simulator

Dry cask applicants submit their spent nuclear fuel dry storage and transportation cask designs to the U.S. Nuclear Regulatory Commission (NRC) for certification under Title 10 of the Code of Federal Regulations (10 CFR) Part 72, 'Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste'. The NRC staff performs its technical review of these designs in accordance with 10 CFR Part 72 and NUREG-1536, 'Standard Review Plan for Spent Fuel Dry Storage Systems at a General License Facility,' Revision 1, issued July 2010. To ensure that the cask and fuel material temperatures of the dry cask storage system remain within the allowable limits or criteria for normal, off-normal, and accident conditions, the NRC staff performs a thermal review as part of the technical review. Recent applications increasingly have used thermalhydraulic analyses using computational fluid dynamics (CFD) codes (e.g., ANSYS FLUENT) to demonstrate the adequacy of the thermal design. The NRC performs validation studies of the FLUENT CFD code to assist it in making regulatory decisions to ensure adequate protection for storage and transportation casks. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. In efforts to further validate these designs, the NRC in conjunction with the Department of Energy and Sandia National Laboratories (SNL) are conducting an experiment to capture the effects of elevated helium pressures and accurately portray the external convection of aboveground and belowground canister dry cask systems.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Feasibility of a multigroup Boltzmann–Fokker–Planck solution for electron beam dose calculations

Legacy nuclear-reactor Boltzmann solvers start clinical deployment as an alternative to Monte Carlo (MC) codes and Fermi–Eyges semiemprical models in radiation oncology treatment planning. Today’s certified clinical solvers are limited to photon beams. In this paper, ELECTR, a state-of-the-art multigroup electron cross sections generation module in NJOY is presented and validated against Lockwood’s calorimetric measurements, EGS-nrc and GEANT-4 for 1–20 MeV unidirectional electron beams. The nuclear-reactor DRAGON-5 solver is upgraded to access the library and solve the Boltzmann–Fokker–Planck (BFP) equation. A variety of heterogeneous radiotherapy and radiosurgery phantom configurations were used for validation purpose. Case studies include a thorax benchmark, that of a typical breast Intra-Operative Radiotherapy and a high-heterogeneity patient-like benchmark. For all beams, 100% of the water voxels satisfied the American Association of Physicists in Medicine accuracy criterion for a BFP-MC dose error below 2%. At least, 97.0% of adipose, muscle, bone, lung, tumor and breast voxels satisfied the 2% criterion. The average BFP-MC relative error was about 0.56% for all voxels, beams and materials combined. By irradiating homogeneous slabs from Z = 1 (hydrogen) to Z = 99 (einsteinium), we reported performance and defects of the CEPXS mode [US. Sandia National Lab., SAND-89-1685] in ELECTR for the entire periodic table. For all Lockwood’s benchmarks, NJOY-DRAGON dose predictions are within the experimental data precision for 98% of voxels.

42 ENGINEERING↗
DOE OSTI2016

Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices

A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). In conclusion, this paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

IER 489: Critical Unresolved Region Integral Experiment (CURIE) CEDT Phase-2 Final Design

The final design for the CURIE integral benchmark experiment is presented in this report. Measurements will be performed on the Comet vertical assembly machine and include various critical configurations consisting of High Enriched Uranium (HEU) metal plates and Teflon moderator plates surrounded by a thick copper reflector. The various configurations with different Teflon plate thickness planned allow for complimentary measurements with differing nuclear data sensitivity in the Unresolved Resonance Region (URR). Similar to previously performed Zeus experiments, the proposed work will serve to validate vital nuclear data quantities and uncertainties and identify deficiencies in current radiation transport codes. Additionally, it will assess the applicability of optimization approaches to nuclear critical experiment design. The final experiment design is detailed in this report and includes design, optimization, uncertainty and continuous-energy nuclear data sensitivity analysis for the final experimental configurations that will provide the basis for a future benchmark evaluation intended for submission to the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook [1]. This document discusses the method, approach, considerations and final design chosen with conclusions. It also includes cost and schedule. The basis and principles of this document come from the preliminary design report [2]. Section 2 discusses the content and potential impact; section 3 gives an overview of the nuclear data and unresolved resonance region; section 4 gives a summary of the work done during the preliminary design; section 5 discusses the machine learning method and principles used for final design; section 6 summaries the results from final design; section 7 provides conclusions and the range of configurations chosen to carry forward to experiment execution; and section 8 includes a summary of costs.

42 ENGINEERING↗
NASA NTRS2000

Development of an Integrated Rocket Thrust Chamber Assembly Analysis and Design Toolkit (TCAT)

There are many individual computational tools and methods used to analyze and design components of a rocket engine thrust chamber assembly (TCA) 11,21. To analyze and design a thrust chamber assembly these tools are usually used in sequence while communication of information between these tools is usually done manually. Each component of a TCA is usually designed and optimized individually with limited considerations on system optimized design. This approach is often iterative and can be prone to error. Also at present, expert knowledge of each tool is required. The objective of this software development effort is to select, integrate and automate the best tools into an easy-to-use computational package, the Thrust Chamber Analysis Toolkit (TCAT). This tool will provide a seamless process to analyze and design all or individual components of a thrust chamber assembly. This paper describes the current status of the MAT software development. The selected component codes and their capabilities are briefly described. User and module interfaces and integration status are presented. MAT pre- and post-processing added options are presented and overall MAT capabilities are demonstrated with the results of sample cases. Ongoing and planned model validation efforts are also described. Finally, the MAT future plan options are discussed.

Farhangi, S.↗
DOE OSTI2016

Plasma Interactions with Mixed Materials and Impurity Transport

The project brings together three discipline areas at LLNL to develop advanced capability to predict the impact of plasma/material interactions (PMI) on metallic surfaces in magnetic fusion energy (MFE) devices. These areas are (1) modeling transport of wall impurity ions through the edge plasma to the core plasma, (2) construction of a laser blow-off (LBO) system for injecting precise amounts of metallic atoms into a tokamak plasma, and (3) material science analysis of fundamental processes that modify metallic surfaces during plasma bombardment. The focus is on tungsten (W), which is being used for the ITER divertor and in designs of future MFE devices. In area (1), we have worked with the University of California, San Diego (UCSD) on applications of the UEDGE/DUSTT coupled codes to predict the influx of impurity ions from W dust through the edge plasma, including periodic edge-plasma oscillations, and revived a parallel version of UEDGE to speed up these simulations. In addition, the impurity transport model in the 2D UEDGE code has been implemented into the 3D BOUT++ turbulence/transport code to allow fundamental analysis of the impact of strong plasma turbulence on the impurity transport. In area (2), construction and testing of the LBO injection system has been completed. The original plan to install the LBO on the National Spherical Torus Experiment Upgrade (NSTX-U) at Princeton and its use to validate the impurity transport simulations is delayed owing to NSTX-U being offline for substantial magnetic coil repair period. In area (3), an analytic model has been developed to explain the growth of W tendrils (or fuzz) observed for helium-containing plasmas. Molecular dynamics calculations of W sputtering by W and deuterium (D) ions shows that a spatial blending of interatomic potentials is needed to describe the near-surface and deeper regions of the material.

36 MATERIALS SCIENCE↗
NASA NTRS2004

Overview 2003 of NASA Multi-D Stirling Convertor Development and DOE and NASA Stirling Regenerator R and D Efforts

This paper will report on (1) continuation through the 3rd year of a NASA grant for multi-dimensional Stirling CFD code development and validation, (2) continuation through the 3rd and final year of a Department of Energy, Golden Field Office (DOE) regenerator research effort. Results of the NASA multi-D code development effort and the DOE regenerator research efforts will be summarized. Early results and planning for the new regenerator microfabrication contract will also be discussed.

Twe, Roy↗
NASA NTRS1999

LeRC-HT: NASA Lewis Research Center General Multiblock Navier-Stokes Heat Transfer Code Developed

For the last several years, LeRC-HT, a three-dimensional computational fluid dynamics (CFD) computer code for analyzing gas turbine flow and convective heat transfer, has been evolving at the NASA Lewis Research Center. The code is unique in its ability to give a highly detailed representation of the flow field very close to solid surfaces. This is necessary for an accurate representation of fluid heat transfer and viscous shear stresses. The code has been used extensively for both internal cooling passage flows and hot gas path flows--including detailed film cooling calculations, complex tip-clearance gap flows, and heat transfer. In its current form, this code has a multiblock grid capability and has been validated for a number of turbine configurations. The code has been developed and used primarily as a research tool (at least 35 technical papers have been published relative to the code and its application), but it should be useful for detailed design analysis. We now plan to make this code available to selected users for further evaluation.

Heidmann, James D.↗
DOE OSTI2017

Verification of the Two-Fluid Model in the Spectral Element Code Nek-2P for Adiabatic Flow Conditions

The simulation of multiphase flow is a difficult physical, computational, and numerical problem. Sophisticated tools are necessary to provide results encompassing a wide range of physical phenomena. The development of these tools must be supported by rigorous analysis and ongoing efforts of verification and validation to ensure the accuracy of simulations. The Nek-2P multiphase code is being developed at Argonne with the general goal of simulating a wide range of multiphase flows and the primary goal of simulating boiling flows in LWR fuel assemblies. As part of this, there is an ongoing effort to verify the code and validate the underlying physical models. The objective of this work is to present recent results in the verification of Nek-2P. As a starting point, this work is focused on adiabatic flows for the sake of simplicity with the immediate next planned step to expand the effort to include heat and mass transfer. Details of the verification of the implemented mass fraction transport, momentum conservation, and pressure equations in Nek-2P are presented. Results of simulations are compared to analytic solutions when possible, however due to the complexity of multiphase flows, these are seldom available. To overcome this difficulty, an independent model is used to provide solutions for comparison.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1987

Development of N-version software samples for an experiment in software fault tolerance

The report documents the task planning and software development phases of an effort to obtain twenty versions of code independently designed and developed from a common specification. These versions were created for use in future experiments in software fault tolerance, in continuation of the experimental series underway at the Systems Validation Methods Branch (SVMB) at NASA Langley Research Center. The 20 versions were developed under controlled conditions at four U.S. universities, by 20 teams of two researchers each. The versions process raw data from a modified Redundant Strapped Down Inertial Measurement Unit (RSDIMU). The specifications, and over 200 questions submitted by the developers concerning the specifications, are included as appendices to this report. Design documents, and design and code walkthrough reports for each version, were also obtained in this task for use in future studies.

Lauterbach, L.↗
NASA NTRS2004

A History of Full-Scale Aircraft and Rotorcraft Crash Testing and Simulation at NASA Langley Research Center

This paper summarizes 2-1/2 decades of full-scale aircraft and rotorcraft crash testing performed at the Impact Dynamics Research Facility (IDRF) located at NASA Langley Research Center in Hampton, Virginia. The IDRF is a 240-ft.-high steel gantry that was built originally as a lunar landing simulator facility in the early 1960's. It was converted into a full-scale crash test facility for light aircraft and rotorcraft in the early 1970 s. Since the first full-scale crash test was preformed in February 1974, the IDRF has been used to conduct: 41 full-scale crash tests of General Aviation (GA) aircraft including landmark studies to establish baseline crash performance data for metallic and composite GA aircraft; 11 full-scale crash tests of helicopters including crash qualification tests of the Bell and Sikorsky Advanced Composite Airframe Program (ACAP) prototypes; 48 Wire Strike Protection System (WSPS) qualification tests of Army helicopters; 3 vertical drop tests of Boeing 707 transport aircraft fuselage sections; and, 60+ crash tests of the F-111 crew escape module. For some of these tests, nonlinear transient dynamic codes were utilized to simulate the impact response of the airframe. These simulations were performed to evaluate the capabilities of the analytical tools, as well as to validate the models through test-analysis correlation. In September 2003, NASA Langley closed the IDRF facility and plans are underway to demolish it in 2007. Consequently, it is important to document the contributions made to improve the crashworthiness of light aircraft and rotorcraft achieved through full-scale crash testing and simulation at the IDRF.

Jackson, Karen E.↗
DOE OSTI2020

Development of risk mitigation guidance for sensor placement inside mechanically ventilated enclosures – Phase 1

Guidance on Sensor Placement was identified as the top research priority for hydrogen sensors at the 2018 HySafe Research Priority Workshop on hydrogen safety in the category Mitigation, Sensors, Hazard Prevention, and Risk Reduction. This paper discusses the initial steps (Phase 1) to develop such guidance for mechanically ventilated enclosures. This work was initiated as an international collaborative effort to respond to emerging market needs related to the design and deployment equipment for hydrogen infrastructure that is often installed in individual equipment cabinets or ventilated enclosures. The ultimate objective of this effort is to develop guidance for an optimal sensor placement such that, when integrated into a facility design and operation, will allow earlier detection at lower levels of incipient leaks, leading to significant hazard reduction. Reliable and consistent early warning of hydrogen leaks will allow for the risk mitigation by reducing or even eliminating the probability of escalation of small leaks into large and uncontrolled events. To address this issue, a study of a real-world mechanically ventilated enclosure containing GH2 equipment was conducted, where CFD modeling of the hydrogen dispersion (performed by AVT and UQTR, and independently by the JRC) was validated by the NREL Sensor laboratory using a Hydrogen Wide Area Monitor (HyWAM) consisting of a 10-point gas and temperature measurement analyzer. In the release test, helium was used as a hydrogen surrogate. Expansion of indoor releases to other larger facilities (including parking structures, vehicle maintenance facilities and potentially tunnels) and incorporation into QRA tools, such as HyRAM is planned for Phase 2. It is anticipated that results of this work will be used to inform national and international standards such as NFPA 2 Hydrogen Technologies Code, Canadian Hydrogen Installation Code (CHIC) and relevant ISO/TC 197 and CEN documents.

08 HYDROGEN↗
DOE OSTI2017

Gas Gun Model and Comparison to Experimental Performance of Pipe Guns Operating with Light Propellant Gases and Large Cryogenic Pellets

Injection of multiple large (~10 to 30 mm diameter) shattered pellets into ITER plasmas is presently part of the scheme planned to mitigate the deleterious effects of disruptions on the vessel components. To help in the design and optimize performance of the pellet injectors for this application, a model referred to as “the gas gun simulator” has been developed and benchmarked against experimental data. The computer code simulator is a Java program that models the gas-dynamics characteristics of a single-stage gas gun. Following a stepwise approach, the code utilizes a variety of input parameters to incrementally simulate and analyze the dynamics of the gun as the projectile is launched down the barrel. Using input data, the model can calculate gun performance based on physical characteristics, such as propellant-gas and fast-valve properties, barrel geometry, and pellet mass. Although the model is fundamentally generic, the present version is configured to accommodate cryogenic pellets composed of H2, D2, Ne, Ar, and mixtures of them and light propellant gases (H2, D2, and He). The pellets are solidified in situ in pipe guns that consist of stainless steel tubes and fast-acting valves that provide the propellant gas for pellet acceleration (to speeds ~200 to 700 m/s). The pellet speed is the key parameter in determining the response time of a shattered pellet system to a plasma disruption event. The calculated speeds from the code simulations of experiments were typically in excellent agreement with the measured values. With the gas gun simulator validated for many test shots and over a wide range of physical and operating parameters, it is a valuable tool for optimization of the injector design, including the fast valve design (orifice size and volume) for any operating pressure (~40 bar expected for the ITER application) and barrel length for any pellet size (mass, diameter, and length). Key design parameters and proposed values for the pellet injectors for the ITER disruption mitigation systems are discussed.

Combs, Stephen Kirk↗
DOE OSTI2021

Development of phenomena identification and ranking table for Westinghouse lead fast reactor’s safety

The Westinghouse Lead-cooled Fast Reactor (LFR) is a medium-size, passively safe, economic, Gen-IV nuclear reactor. An important effort within the Westinghouse LFR program is the development of the safety analysis methodology, which comprises computer code development, model development, and experimental testing. A key initial task associated with the development of the safety analysis methodology is the identification of processes and phenomena that affect the plant's capability to meet selected safety performance indicators. This is accomplished through the development of a Phenomena Identification and Ranking Table (PIRT) for selected accident scenarios which, for this specific PIRT effort, included selected postulated design basis accidents and hypothetical beyond design basis accidents in LFRs. This paper describes the role of PIRT in the development of the Westinghouse LFR safety analysis methodology and the process used in the PIRT development. Specifically, the Westinghouse LFR PIRT assessed importance of pertinent phenomena and identified gaps in their knowledge-base by evaluating current modeling capabilities and data available for validation. The ultimate goal was to provide guidance on computer code development and validation efforts and to prioritize testing to support LFR design and licensing. The key phenomena and processes that are deemed highly important for the safety performance indicators, but for which the state of knowledge is low, are presented. The testing program and analyses development are planned to address significant phenomena in the PIRT.

Lead fast reactor↗
NASA NTRS2023

X-57 Maxwell Airworthiness Validation Plan

This report is a Final Airworthiness Validation Plan (AVP) and describes how an aircraft like X-57 does (and does not) meet current airworthiness standards. The objective of this report is to create an example certification basis, associated means of compliance (MoC), and method of compliance for a distributed electric propulsion airplane under 14 Code of Federal Regulations (CFR) Part 21, “Certification Procedures for Products and Articles,” and its associated relevant sections of 14 CFR for “Airworthiness Standards” of Part 23, “Normal Category Airplanes,” Part 33 “Aircraft Engines,” and Part 35 “Propellers.” The approach to meet the objective is to use NASA’s X-57 Modification (Mod) IV flight demonstrator as an example and categorize its applicability to the regulations and standards according to the following three conditions: 1. Identify, where applicable, that the MoC and methods of compliance can be associated with existing Standard Specifications and Standard Practices of (ASTM) Committee F39 on Aircraft Systems and ASTM Committee F44 on General Aviation Aircraft; 2. If relevant ASTM standards do not exist, identify means and-or methods of compliance from appropriate Federal Aviation Administration (FAA) Advisory Circulars and other sources to use for the X-57 Mod IV vehicle; or 3. If no relevant certification rule, MoC, or method of compliance exists, highlight this omission and provide recommendations.

Herbert W Schlickenmaier↗