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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 1,423 records · Page 79
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Influence of construction and demolition waste management on the environmental impact of buildings

Highlights: Black-Right-Pointing-Pointer Environmental impacts of different demolition practices. Black-Right-Pointing-Pointer 'Top-down' approach to the Life Cycle Analysis methodology. Black-Right-Pointing-Pointer Results based on real buildings measurements and demolition contractor activities. Black-Right-Pointing-Pointer Not every type of selective demolition brings about environmental benefits. - Abstract: The purpose of this study is to quantify comparable environmental impacts within a Life Cycle Analysis (LCA) perspective, for buildings in which the first (Materials) and last (End of Life) life cycle stages are adjusted to several waste/material management options. Unlike most LCAs, the approach is 'top-down' rather than 'bottom-up', which usually involves large amounts of data and the use of specific software applications. This approach is considered appropriate for a limited but expedient LCA designed to compare the environmental impacts of different life cycle options. Present results, based on real buildings measurements and demolition contractor activities, show that shallow, superficial, selective demolition may not result in reduced environmental impacts. Calculations actually show an increase (generally less than 5%) in most impact categories for the Materials and End of Life stages because of extra transportation needs. However, core material separation in demolition operations and its recycling and/or reuse does bring environmental benefits. A reduction of around 77% has been estimated in the climate change impact category, 57% in acidification potential and 81% in the summer smog impact (for the life cycle stages referred).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Utilizing the deuterium-tritium fusion resonance to diagnose thermal runaway in igniting plasmas

For high-efficiency inertial confinement fusion implosions, it is predicted that a burning hot spot will successfully encompass all surrounding fuel and then transition into a thermal runaway where the internal energy increase from fusion occurs on a timescale faster than the expansion of the fuel is able to quench the fusion chain reaction after ignition occurs. Observation of this dynamic phase transition would indicate distinct burn properties and indicate an implosion's robustness. A technique for diagnosing the presence of thermal runaway from measurements of nuclear reaction history is presented. The technique is based on taking the logarithmic derivative of the nuclear reaction history, called the 𝛼 curve, and allowing a mathematical decoupling of the mass, volume, and thermal reactivity in the fusion reaction rate equation. During thermal runaway, where the thermal temperature dominates the burn dynamics, a maximum in the 𝛼 curve is found where there is a maximum in the first derivative of the thermal fusion reactivity, an effect to the deuterium-tritium (DT) fusion cross-section resonance. This provides a distinct signature related to the fundamental nature of the DT fusion nuclear resonance and signifies the transition into the fusion thermal instability. Impacts of charged particle transport on the effect are also assessed and the analytical formulas are compared and found to be in agreement with radiation hydrodynamic codes.

high-energy-density plasmas↗
DOE OSTI2014

TH-A-12A-01: Medical Physicist's Role in Digital Information Security: Threats, Vulnerabilities and Best Practices

I. Information Security Background (Speaker = Kevin McDonald) Evolution of Medical Devices Living and Working in a Hostile Environment Attack Motivations Attack Vectors Simple Safety Strategies Medical Device Security in the News Medical Devices and Vendors Summary II. Keeping Radiation Oncology IT Systems Secure (Speaker = Bruce Curran) Hardware Security Double-lock Requirements “Foreign” computer systems Portable Device Encryption Patient Data Storage System Requirements Network Configuration Isolating Critical Devices Isolating Clinical Networks Remote Access Considerations Software Applications / Configuration Passwords / Screen Savers Restricted Services / access Software Configuration Restriction Use of DNS to restrict accesse. Patches / Upgrades Awareness Intrusion Prevention Intrusion Detection Threat Risk Analysis Conclusion Learning Objectives: Understanding how Hospital IT Requirements affect Radiation Oncology IT Systems. Illustrating sample practices for hardware, network, and software security. Discussing implementation of good IT security practices in radiation oncology. Understand overall risk and threats scenario in a networked environment.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2011

Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC): Subcontinuum-scale Verification and Validation Strategy

The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. Achieving the objective of modeling the performance of a disposal scenario requires describing processes involved in waste form degradation and radionuclide release at the subcontinuum scale, beginning with mechanistic descriptions of chemical reactions and chemical kinetics at the atomic scale, and upscaling into effective, validated constitutive models for input to high-fidelity continuum scale codes for coupled multiphysics simulations of release and transport. Verification and validation (V&V) is required throughout the system to establish evidence-based metrics for the level of confidence in M&S codes and capabilities, including at the subcontiunuum scale and the constitutive models they inform or generate. This Report outlines the nature of the V&V challenge at the subcontinuum scale, an approach to incorporate V&V concepts into subcontinuum scale modeling and simulation (M&S), and a plan to incrementally incorporate effective V&V into subcontinuum scale M&S destined for use in the NEAMS Waste IPSC work flow to meet requirements of quantitative confidence in the constitutive models informed by subcontinuum scale phenomena.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Assessing the Effect of Parametric Uncertainty on the FRAPCON-3.4 Thermal Conductivity Model

In the absence of a well-accepted, mathematical description of how units of material respond to one or more external actions, scientists must often rely upon empirically-derived models to make predictions. These models are often employed to model very complex phenomena, with an unfortunate side-effect that intuition, as to the values of particular model parameters, may not serve to guide their use. Given our increased reliance on modeling and simulation to make predictions in the absence of experimental data, it befits the scientific and engineering communities to explore and report upon uncertainty quantification techniques applied to previously adopted (or at least well-accepted) empirical models that derive from or pertain to substantial experimental data sets. This report represents a collection of three methodologies aimed at assessing the predictive capability of the empirical thermal conductivity model adopted by the nuclear fuel performance code, FRAPCON-3.4. Each of these methodologies considers the effect of uncertain parameters – a plausible reality in the context of empirically-derived models – on the ability of the model to predict uranium dioxide conductivity data from open literature sources. The results lead the authors to question the predictive capability of the FRAPCON model for predicting the thermal conductivities associated with irradiated fuel samples. The report concludes with a preliminary examination of Idaho National Laboratory’s (INL) nuclear fuel performance code, BISON, developed under INL’s Multiphysics Object Oriented Simulation Environment (MOOSE).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Design of the Energy Partitioning Energy Coupling (EPEC) NIF Experiment

Using scaled experiments performed at the National Ignition Facility (NIF), the Energy Partitioning Energy Coupling (EPEC) experimental platform will provide benchmark data from scaled high energy density (HED) environments to validate codes for prompt yield determination following a near surface explosion. A set of numerical codes has been used to design the experiment and predict the experimental results. Preliminary NIF shots assessing performance of the EPEC energy source (a halfraum) have been carried out. Modeling of the EPEC experiments is discussed, with comparisons to results from the source shots.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2013

A 3D Orthotropic Elastic Continuum Damage Material Model

A three dimensional orthotropic elastic constitutive model with continuum damage is implemented for polymer matrix composite lamina. Damage evolves based on a quadratic homogeneous function of thermodynamic forces in the orthotropic planes. A small strain formulation is used to assess damage. In order to account for large deformations, a Kirchhoff material formulation is implemented and coded for numerical simulation in Sandia’s Sierra Finite Element code suite. The theoretical formulation is described in detail. An example of material parameter determination is given and an example is presented.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Cavern leaching at SPR associated with 2017 oil sales and exchanges

In 2017 small-scale drawdowns at the Strategic Petroleum Reserve (SPR) supported oil sales for the Bipartisan Budget Act of 2015 and the 21st Century Cures Act of 2015 as well as exchanges supporting relief efforts for hurricane Harvey. These drawdowns may affect cavern stability and available drawdowns, thus is important to assess the leaching effects on the cavern shape. Cavern shape estimates from the SANSMIC solution mining code suggest the shapes of 28 caverns were altered in 2017 to varying degrees depending on the total volume of water injected, the initial cavern shape and the distance between the hanging string depth and the oil-brine interface depth. A flaring of the cavern floor occurred in 13 caverns, a geomechanically unfavorable outcome that may require operational changes to preserve cavern integrity. Of the three caverns with post-sale sonars, SANSMIC predictions compared favorably to two but underpredicted the third.

04 OIL SHALES AND TAR SANDS↗
DOE OSTI2024

Tools for Water Ingress Testing

The Safety Storage and Engineering Team, as part of the Production Support Services division (PSS-2), is tasked with ensuring the safety of containers used for handling and storage of nuclear materials. As part of this work, water ingress tests are conducted to evaluate the water-tightness of containers intended for in-glovebox use. In collaboration, the statistics group of the Computer and Computational Sciences Division (CCS-6) provided support in developing a statistically defensible approach for determining appropriate sample sizes for water ingress testing. Water ingress testing involves multiple measurements on multiple containers. Our approach uses a simple random effects model to analyze a pilot data set, implementing prediction limits to evaluate the efficacy of collecting additional data. Although this study capitalizes on available data, our approach can be used with estimates of the ratio of between and within variability and average values, often available from past testing or expert knowledge. An interactive Shiny tool was developed as a final user-friendly product for future testing. The Shiny interface is an open-source package providing a framework for building web applications. Raw data exploration and prediction interval-based sample size assessments can quickly be conducted by the engineering team without needing to interact with the underlying code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

FAST-1.1: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients (Developed under NQA-1-2017)

Fuel Analysis under Steady-state and Transients (FAST) is the U.S. Nuclear Regulatory Commission (NRC)’s computer code that calculates the steady-state and transient response of nuclear reactor fuel rods during long-term in-reactor burnup, anticipated operational occurrences (AOOs), design basis accidents (DBAs), and dry storage conditions. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include: * heat conduction through the fuel and other materials * heat transfer from the cladding-to-coolant * cladding elastic and plastic deformation, including creep * fuel-cladding mechanical interaction * fission gas release from the fuel * rod internal pressure and void volume * cladding oxidation The code contains necessary material and coolant properties, as well as clad-to-coolant heat transfer correlations, for normal operation through postulated accidents and AOOs for today’s U.S.-based light water reactor (LWR) fuel designs. FAST-1.1 also contains preliminary materials and models for new LWR fuel concepts, such as accident tolerant fuel (ATF), and non-LWR fuel concepts such as metallic fuels for sodium fast reactors (SFRs). FAST has been developed for use on Windows and Linux operating systems. This document describes FAST-1.1 and is one of a series of documents on the code; the other documents detail the material properties used by FAST as well as its integral assessment to experiments and commercial data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

FAST-1.2: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients

Fuel Analysis under Steady-state and Transients (FAST) is the U.S. Nuclear Regulatory Commission (NRC)’s computer code that calculates the steady-state and transient response of nuclear reactor fuel rods during long-term in-reactor burnup, anticipated operational occurrences (AOOs), design basis accidents (DBAs), and dry storage conditions. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include: • heat conduction through the fuel and other materials • heat transfer from the cladding-to-coolant • cladding elastic and plastic deformation, including creep • fuel-cladding mechanical interaction • fission gas release from the fuel • rod internal pressure and void volume • cladding oxidation The code contains necessary material and coolant properties, as well as clad-to-coolant heat-transfer correlations, for normal operation through postulated accidents and AOOs for today’s U.S.-based light water reactor (LWR) fuel designs. FAST-1.2 also contains preliminary materials and models for new LWR fuel concepts, such as accident tolerant fuel (ATF), and non-LWR fuel concepts such as metallic fuels for sodium fast reactors (SFRs). FAST has been developed for use on Windows and Linux operating systems. This document describes FAST-1.2 and is one of a series of documents on the code; the other documents detail the material properties used by FAST as well as its integral assessment to experiments and commercial data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Development and Assessment of CTFFuel

This milestone report documents an approach that seeks to develop a standalone fuel solver to simulate steady-state and transient thermo-mechanical responses of light-water reactor fuel rods and assess its temperature predictions against analytical solutions, and a well-known and validated fuel performance code. This manuscript focuses on five major topics: (1) development of the CTFFuel, (2) discussion of the newly implemented fuel deformation models in dynamic gap conductance model of CTFFuel, (3) enabling the dynamic gap conductance model in VERA-CS, (4) verification of the solid conduction solution and comparison of the CTFFuel with respect to the workhorse code, CTF, and (5) benchmark of the thermal conductivity degradation using the burnup dependent fuel thermal conductivity models and benchmark of the fuel centerline temperature using the dynamic gap conductance model with Halden measurements and FRAPCON-4.0 simulation results. CTFFuel temperature predictions are compared to predictions from FRAPCON-4.0 for UO 2 , UO 2 +Gd 2 O 3 and MOX fuels through reactor’s life. The main motivation is to understand discrepancies between CTFFuel and FRAPCON-4.0’s temperature predictions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

FAST-1.2.2: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients

Fuel Analysis under Steady-state and Transients (FAST) is the U.S. Nuclear Regulatory Commission (NRC)’s computer code that calculates the steady-state and transient response of nuclear reactor fuel rods during long-term in-reactor burnup, anticipated operational occurrences (AOOs), design basis accidents (DBAs), and dry storage conditions. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include heat conduction through the fuel and other materials, heat transfer from the cladding-to-coolant, cladding elastic and plastic deformation (including creep), fuel-cladding mechanical interaction, fission gas release from the fuel, rod internal pressure, void volume, and cladding oxidation. The code contains necessary material and coolant properties, as well as clad-to-coolant heat transfer correlations, for normal operation through postulated accidents and AOOs for today’s U.S.-based light water reactor (LWR) fuel designs. FAST-1.2.2 also contains preliminary materials and models for new LWR fuel concepts, such as accident tolerant fuel (ATF), and non-LWR fuel concepts such as metallic fuels for sodium fast reactors (SFRs). FAST has been developed for use on Windows and Linux operating systems. This document describes FAST-1.2.2 and is one of a series of documents on the code; the other documents detail the material properties used by FAST as well as its integral assessment to experiments and commercial data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

NON-LTE INVERSIONS OF THE Mg ii h and k AND UV TRIPLET LINES

The Mg ii h and k lines are powerful diagnostics for studying the solar chromosphere. They have become particularly popular with the launch of the Interface Region Imaging Spectrograph ( IRIS ) satellite, and a number of studies that include these lines have lead to great progress in understanding chromospheric heating, in many cases thanks to the support from 3D MHD simulations. In this study, we utilize another approach to analyze observations: non-LTE inversions of the Mg ii h and k and UV triplet lines including the effects of partial redistribution. Our inversion code attempts to construct a model atmosphere that is compatible with the observed spectra. We have assessed the capabilities and limitations of the inversions using the FALC atmosphere and a snapshot from a 3D radiation-MHD simulation. We find that Mg ii h and k allow reconstructing a model atmosphere from the middle photosphere to the transition region. We have also explored the capabilities of a multi-line/multi-atom setup, including the Mg ii h and k, the Ca ii 854.2 nm, and the Fe i 630.25 lines to recover the full stratification of physical parameters, including the magnetic field vector, from the photosphere to the chromosphere. Finally, we present the first inversions of observed IRIS spectra from quiet-Sun, plage, and sunspot, with very promising results.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2024

Systematic and objective evaluation of Earth system models: PCMDI Metrics Package (PMP) version 3

Systematic, routine, and comprehensive evaluation of Earth system models (ESMs) facilitates benchmarking improvement across model generations and identifying the strengths and weaknesses of different model configurations. By gauging the consistency between models and observations, this endeavor is becoming increasingly necessary to objectively synthesize the thousands of simulations contributed to the Coupled Model Intercomparison Project (CMIP) to date. The Program for Climate Model Diagnosis and Intercomparison (PCMDI) Metrics Package (PMP) is an open-source Python software package that provides quick-look objective comparisons of ESMs with one another and with observations. The comparisons include metrics of large- to global-scale climatologies, tropical inter-annual and intra-seasonal variability modes such as the El Niño–Southern Oscillation (ENSO) and Madden–Julian Oscillation (MJO), extratropical modes of variability, regional monsoons, cloud radiative feedbacks, and high-frequency characteristics of simulated precipitation, including its extremes. The PMP comparison results are produced using all model simulations contributed to CMIP6 and earlier CMIP phases. An important objective of the PMP is to document the performance of ESMs participating in the recent phases of CMIP, together with providing version-controlled information for all datasets, software packages, and analysis codes being used in the evaluation process. Among other purposes, this also enables modeling groups to assess performance changes during the ESM development cycle in the context of the error distribution of the multi-model ensemble. Quantitative model evaluation provided by the PMP can assist modelers in their development priorities. In this paper, we provide an overview of the PMP, including its latest capabilities, and discuss its future direction.

54 ENVIRONMENTAL SCIENCES↗
NASA NTRS1985

Fuel and oxidizer turbine loss analysis

The turbine losses for the fuel and oxidizer turbines at the FPL condition were assessed by a quasi-3D loss analysis method. This loss analysis method uses two flow codes - MERIDL and TSONIC - to calculate the flow velocities along the blade surfaces and endwalls. The velocities are then used as input to the boundary layer code - BLAYER - to calculate the friction losses due to incidence, secondary flow, and tip clearance. The loss analysis for the fuel turbine indicated an overall two-stage efficiency of about 90%. The largest loss was due to rotor tip clearance. The loss analysis for the oxidizer turbine is nearly completed. Results for the first stage of the two-stage design indicates an efficiency of about 80%, with high losses due to rotor incidence and blade and endwall friction.

Haas, J. E.↗
NASA NTRS1987

A critical assessment of wind tunnel results for the NACA 0012 airfoil

A large body of experimental results, obtained in more than 40 wind tunnels on a single, well-known two-dimensional configuration, has been critically examined and correlated. An assessment of some of the possible sources of error has been made for each facility, and data which are suspect have been identified. It was found that no single experiment provided a complete set of reliable data, although one investigation stands out as superior in many respects. However, from the aggregate of data the representative properties of the NACA 0012 airfoil can be identified with reasonable confidence over wide ranges of Mach number, Reynolds number, and angles of attack. This synthesized information can now be used to assess and validate existing and future wind tunnel results and to evaluate advanced Computational Fluid Dynamics codes.

Mccroskey, W. J.↗
NASA NTRS1988

A critical assessment of wind tunnel results for the NACA 0012 airfoil

A large body of experimental results, obtained in more than 40 wind tunnels on a single, well known two-dimensional configuration, was critically examined and correlated. An assessment of some of the possible sources of error was made for each facility, and data which are suspect were identified. It was found that no single experiment provided a complete set of reliable data, although an investigation stands out as superior in many respects. However, from the aggregate of data the representative properties of the NACA 0012 airfoil can be identified with reasonable confidence over wide range of Mach numbers, Reynolds number, and angles of attack. This synthesized information can now be used to assess and validate existing and future wind tunnel results and to evaluate advanced Computational Fluid Dynamics codes.

Mccroskey, W. J.↗