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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 973 records · Page 54
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Insights Gained from Forensic Analysis with MELCOR of the Fukushima-Daiichi Accidents.

Since the accidents at Fukushima-Daiichi, Sandia National Laboratories has been modeling these accident scenarios using the severe accident analysis code, MELCOR. MELCOR is a widely used computer code developed at Sandia National Laboratories since ~1982 for the U.S. Nuclear Regulatory Commission. Insights from the modeling of these accidents is being used to better inform future code development and potentially improved accident management. To date, our necessity to better capture in-vessel thermal-hydraulic and ex-vessel melt coolability and concrete interactions has led to the implementation of new models. The most recent analyses, presented in this paper, have been in support of the of the Organization for Economic Cooperation and Development Nuclear Energy Agency’s (OECD/NEA) Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF) Project. The goal of this project is to accurately capture the source term from all three releases and then model the atmospheric dispersion. In order to do this, a forensic approach is being used in which available plant data and release timings is being used to inform the modeled MELCOR accident scenario. For example, containment failures, core slumping events and lower head failure timings are all enforced parameters in these analyses. This approach is fundamentally different from a blind code assessment analysis often used in standard problem exercises. The timings of these events are informed by representative spikes or decreases in plant data. The combination of improvements to the MELCOR source code resulting from analysis previous accident analysis and this forensic approach has allowed Sandia to generate representative and plausible source terms for all three accidents at Fukushima Daiichi out to three weeks after the accident to capture both early and late releases. In particular, using the source terms developed by MELCOR, the MACCS software code, which models atmospheric dispersion and deposition, we are able to reasonably capture the deposition of radionuclides to the northwest of the reactor site.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Clean Energy to Communities: Gap Region Outreach Project

The Clean Energy to Communities (C2C) program provides expertise and tools to local and regional governments, Tribes, community-based organizations, municipal utilities, and rural electric cooperatives to help them achieve their clean energy goals. The U.S. Department of Energy (DOE) launched C2C in 2022, following engagement with more than 150 stakeholders representing 95 communities in 40 states and 6 Tribes. The program offers three levels of support: expert match, peer-learning cohorts, and in-depth partnerships. In its first year, C2C worked with over 150 communities. To ensure broad participation, the National Renewable Energy Laboratory (NREL) examined regional representation. This analysis revealed that four regions are under-represented in program participation: Gulf South, Appalachia, Midwest, and Northern Plains. To help address these gaps, NREL led an effort to identify barriers faced by potential C2C participants from these regions and develop strategies to facilitate participation. This document summarizes findings from this effort and provides recommendations for program changes.

C2C↗
DOE OSTI2024

Iron-Chromium-Aluminum Accident Tolerant Fuel Concept Source Term Accident Sequence Analysis - High Burnup Fuel Source Term Accident Sequence Analysis Supplement

To extend NUREG-1465 and high burnup fuel source term (SAND2023-01313) recommendations, representative radiological releases to containment – patterned after NUREG-1465 – have been evaluated for LWRs utilizing iron-chromium-aluminum (FeCrAl) alloys in place of zirconium-based alloys in major core structures (cladding and fuel canisters) and high burnup fuel with enrichments of 8% and 10% for PWRs and BWRs, respectively. Representative radionuclide releases are generated for this accident tolerant fuel concept by applying non-parametric bootstrap methods to MELCOR simulation results. Accident scenarios considered in this analysis include principle contributors to historical core damage frequency estimates for a range of nuclear reactor technologies representative of the operating U.S.A. fleet of nuclear reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Cr-coated Accident Tolerant Fuel Concept Source Term Accident Sequence Analysis - High Burnup Fuel Source Term Accident Sequence Analysis Supplement

To extend NUREG-1465 and high burnup fuel source term (SAND2023-01313) recommendations, representative radiological releases to containment – patterned after NUREG-1465 – have been evaluated for LWRs utilizing the chromium-coating on major zircaloy structures (cladding and fuel canisters) and high burnup fuel with enrichments of 8% and 10% for PWRs and BWRs, respectively. Representative radionuclide releases are generated for this accident tolerant fuel concept by applying non-parametric bootstrap methods to MELCOR simulation results. Accident scenarios considered in this analysis include principle contributors to historical core damage frequency estimates for a range of nuclear reactor technologies representative of the operating U.S.A. fleet of nuclear reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Beyond Design Basis Event Pilot Evaluations: Results and Recommendations for Improvements to Enhance Nuclear Safety at DOE Nuclear Facilities (DOE, 2013. A Report to the Secretary of Energy)

In the six months after the March 2011 Fukushima Daiichi nuclear power plant accident in Japan, the U.S. Department of Energy (DOE) took several actions to review the safety of its nuclear facilities and identify situations where near-term improvements could be made. These actions and recommendations were addressed in an August 2011 report to the Secretary of Energy, Review of Requirements and Capabilities for Analyzing and Responding to Beyond Design Basis Events. Based on recommendations in the August 2011 report, DOE embarked on a project to develop and refine guidance that supports improvements in DOE’s processes for analyzing and mitigating beyond design basis events (BDBEs), i.e., events such as earthquakes that are more severe than the events that formed the basis of the design for DOE’s nuclear facilities. The results of this BDBE project and recommendations for further DOE actions are provided in this follow-on report. The main activity of the BDBE project was the pilot application of guidance for evaluation of BDBE analysis and mitigation features at four DOE nuclear facilities representing a range of DOE sites, nuclear facility types/activities, and responsible program offices. The pilot evaluations focused on: (1) BDBE evaluations as documented in the facility Documented Safety Analysis (DSA); (2) potential BDBE vulnerabilities and margins to failure for facility safety features as obtained from general area and specific system walkdowns and design document reviews; and, (3) preparations made in facility and site emergency management programs to respond to severe accidents. The BDBE project also evaluated whether draft BDBE guidance on safety analysis and emergency management could be used to improve the analysis of, and preparations for, mitigating severe accidents and BDBEs. The pilot evaluation team paid close attention to related actions being pursued by the U.S. Nuclear Regulatory Commission, the U.S. commercial nuclear industry, and the international nuclear safety community to help refine the guidance developed for the pilot evaluations.

61 RADIATION PROTECTION AND DOSIMETRY↗
NASA NTRS1980

Computational models for the nonlinear analysis of reinforced concrete plates

A finite element computational model for the nonlinear analysis of reinforced concrete solid, stiffened and cellular plates is briefly outlined. Typically, Mindlin elements are used to model the plates whereas eccentric Timoshenko elements are adopted to represent the beams. The layering technique, common in the analysis of reinforced concrete flexural systems, is incorporated in the model. The proposed model provides an inexpensive and reasonably accurate approach which can be extended for use with voided plates.

Hinton, E.↗
NASA NTRS1995

Program For Analyzing Failures In A Complex System

Configurable Real-Time Analysis System (CRANS) computer program provides means for representing logically interconnected items (e.g., modules of equipment in complicated system) in matrix format. Capabilities provided helpful in real-time environment (e.g., operational environment of such system), in which need for quick evaluation of results of changes or failures. Program facilitates rapid logic-tree analysis. Written in C language.

Mccluney, R. Kevin↗
NASA NTRS1975

Numerical solution of quadratic matrix equations for free vibration analysis of structures

This paper is concerned with the efficient and accurate solution of the eigenvalue problem represented by quadratic matrix equations. Such matrix forms are obtained in connection with the free vibration analysis of structures, discretized by finite 'dynamic' elements, resulting in frequency-dependent stiffness and inertia matrices. The paper presents a new numerical solution procedure of the quadratic matrix equations, based on a combined Sturm sequence and inverse iteration technique enabling economical and accurate determination of a few required eigenvalues and associated vectors. An alternative procedure based on a simultaneous iteration procedure is also described when only the first few modes are the usual requirement. The employment of finite dynamic elements in conjunction with the presently developed eigenvalue routines results in a most significant economy in the dynamic analysis of structures.

Gupta, K. K.↗
DOE OSTI2016

Integrated Agent-Based and Production Cost Modeling Framework for Renewable Energy Studies

The agent-based framework for renewable energy studies (ARES) is an integrated approach that adds an agent-based model of industry actors to PLEXOS and combines the strengths of the two to overcome their individual shortcomings. It can examine existing and novel wholesale electricity markets under high penetrations of renewables. ARES is demonstrated by studying how increasing levels of wind will impact the operations and the exercise of market power of generation companies that exploit an economic withholding strategy. The analysis is carried out on a test system that represents the Electric Reliability Council of Texas energy-only market in the year 2020. The results more realistically reproduce the operations of an energy market under different and increasing penetrations of wind, and ARES can be extended to address pressing issues in current and future wholesale electricity markets.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2016

Effect of soil property uncertainties on permafrost thaw projections: a calibration-constrained analysis: Modeling Archive

This dataset contains an ensemble of thermal-hydro soil parameters including porosity, thermal conductivity, thermal conductivity shape parameters, and residual saturation of peat and mineral soil. The ensemble was generated using a Null-Space Monte Carlo analysis of parameter uncertainty based on a calibration to soil temperatures collected at the Barrow Environmental Observatory site by the NGEE team. The micro-topography of ice wedge polygons present at the site is included in the analysis using three 1D column models to represent polygon center, rim and trough features. The Arctic Terrestrial Simulator (ATS) was used in the calibration to model multiphase thermal and hydrological processes in the subsurface. This dataset includes one zipped file, one pdf file, and one csv file.This Modeling Archive is in support of an NGEE Arctic publication. The citation for that publication can be found within the references section.The Next-Generation Ecosystem Experiments: Arctic (NGEE Arctic), was a research effort to reduce uncertainty in Earth System Models by developing a predictive understanding of carbon-rich Arctic ecosystems and feedbacks to climate. NGEE Arctic was supported by the Department of Energy's Office of Biological and Environmental Research.The NGEE Arctic project had two field research sites: 1) located within the Arctic polygonal tundra coastal region on the Barrow Environmental Observatory (BEO) and the North Slope near Utqiagvik (Barrow), Alaska and 2) multiple areas on the discontinuous permafrost region of the Seward Peninsula north of Nome, Alaska.Through observations, experiments, and synthesis with existing datasets, NGEE Arctic provided an enhanced knowledge base for multi-scale modeling and contributed to improved process representation at global pan-Arctic scales within the Department of Energy's Earth system Model (the Energy Exascale Earth System Model, or E3SM), and specifically within the E3SM Land Model component (ELM).

54 ENVIRONMENTAL SCIENCES↗
NASA NTRS1979

The alteration of profile analysis to accommodate testing functions

The development of a methodology was studied for testing differences among several pilot functions, where the data points represent averages at various frequencies. Topics discussed include: basic assumptions, hypothesis, profile analysis, alteration of profile analysis to accommodate testing functions, test and procedures, and power of tests.

Myers, R. H.↗
DOE OSTI2020

Sample problems for Section III, Division 5 design by inelastic analysis of Grade 91 components

This report works through two sample design analyses of representative Grade 91 components using the ASME Boiler & Pressure Vessel Code Section III, Division 5, Appendix HBB-T rules for the strain and deformation and creep-fatigue design criteria using a design by inelastic analysis approach. The component geometries and loadings were selected to span a wide variety of potential applications. The sample components are an endplate geometry, which is fairly representative of typical pressure vessel construction, and a core block, representing a critical component in a heat pipe microreactor, which has a complicated geometry and complex thermal stress history. These sample problems serve several functions. First, they are complete worked examples of how to apply the current design by inelastic analysis rules. Second, they serve as test cases to evaluate the current rules and provide feedback on how they could be optimized to deal with complicated component geometries or to be better automated in modern finite element analysis software. Finally, the analysis here uses an inelastic constitutive model for Grade 91 developed by Argonne National Laboratory and proposed for inclusion in the ASME Boiler & Pressure Vessel Code Section III, Division 5 rules as a reference constitutive model. Using this constitutive model in the sample problem analyses tests the model in the context of the ASME Code rules. The final conclusions of this report are that the material model and the current Code rules are satisfactory, but that the Code rules could be optimized to simplify automation and reduce over conservatism. Future work should address the specific optimization topics discussed in this report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

A new approach for modeling and analysis of molten salt reactors using SCALE

The Office of Fuel Cycle Technologies (FCT) of the DOE Office of Nuclear Energy is performing an evaluation and screening of potential fuel cycle options to provide information that can support future research and development decisions based on the more promising fuel cycle options. [1] A comprehensive set of fuel cycle options are put into evaluation groups based on physics and fuel cycle characteristics. Representative options for each group are then evaluated to provide the quantitative information needed to support the valuation of criteria and metrics used for the study. Included in this set of representative options are Molten Salt Reactors (MSRs), the analysis of which requires several capabilities that are not adequately supported by the current version of SCALE or other neutronics depletion software packages (e.g., continuous online feed and removal of materials). A new analysis approach was developed for MSR analysis using SCALE by taking user-specified MSR parameters and performing a series of SCALE/TRITON calculations to determine the resulting equilibrium operating conditions. This paper provides a detailed description of the new analysis approach, including the modeling equations and radiation transport models used. Results for an MSR fuel cycle option of interest are also provided to demonstrate the application to a relevant problem. The current implementation is through a utility code that uses the two-dimensional (2D) TRITON depletion sequence in SCALE 6.1 but could be readily adapted to three-dimensional (3D) TRITON depletion sequences or other versions of SCALE. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Poster — Thur Eve — 03: Application of the non-negative matrix factorization technique to [{sup 11}C]-DTBZ dynamic PET data for the early detection of Parkinson's disease

In this work, a class of non-negative matrix factorization (NMF) technique known as alternating non-negative least squares, combined with the projected gradient method, is used to analyze twenty-five [{sup 11}C]-DTBZ dynamic PET/CT brain data. For each subject, a two-factor model is assumed and two factors representing the striatum (factor 1) and the non-striatum (factor 2) tissues are extracted using the proposed NMF technique and commercially available factor analysis software “Pixies”. The extracted factor 1 and 2 curves represent the binding site of the radiotracer and describe the uptake and clearance of the radiotracer by soft tissues in the brain, respectively. The proposed NMF technique uses prior information about the dynamic data to obtain sample time-activity curves representing the striatum and the non-striatum tissues. These curves are then used for “warm” starting the optimization. Factor solutions from the two methods are compared graphically and quantitatively. In healthy subjects, radiotracer uptake by factors 1 and 2 are approximately 35–40% and 60–65%, respectively. The solutions are also used to develop a factor-based metric for the detection of early, untreated Parkinson's disease. The metric stratifies healthy subjects from suspected Parkinson's patients (based on the graphical method). The analysis shows that both techniques produce comparable results with similar computational time. The “semi-automatic” approach used by the NMF technique allows clinicians to manually set a starting condition for “warm” starting the optimization in order to facilitate control and efficient interaction with the data.

60 APPLIED LIFE SCIENCES↗
NASA NTRS1991

Multi-resolution analysis for ENO schemes

Given an function, u(x), which is represented by its cell-averages in cells which are formed by some unstructured grid, we show how to decompose the function into various scales of variation. This is done by considering a set of nested grids in which the given grid is the finest, and identifying in each locality the coarsest grid in the set from which u(x) can be recovered to a prescribed accuracy. This multi-resolution analysis was applied to essentially non-oscillatory (ENO) schemes in order to advance the solution by one time-step. This is accomplished by decomposing the numerical solution at the beginning of each time-step into levels of resolution, and performing the computation in each locality at the appropriate coarser grid. An efficient algorithm for implementing this program in the 1-D case is presented; this algorithm can be extended to the multi-dimensional case with Cartesian grids.

Harten, Ami↗
DOE OSTI2021

Material Interactions in Severe Accidents – Benchmarking the MELCOR V2.2 Eutectics Model for a BWR-3 Mark-I Station Blackout: Part II – Uncertainty Analysis

Single case comparisons between severe accident simulations can provide detailed insights into severe accident model behavior, however, they cannot offer insights into model uncertainty, sensitivity to uncertain parameters, or underlying model biases.Here in this analysis, the single case benchmark comparison of the MELCOR material interaction models for a station blackout (SBO) scenario of a boiling water reactor (BWR) using representative Fukushima Daiichi Unit 1 boundary conditions is expanded to include an uncertainty analysis. As part of this uncertainty analysis, 1200 simulations are performed for each material interaction model (2400 total), with random sampling of 14 uncertain MELCOR input parameters. Input parameters are selected for their impact on models representing core degradation processes. These include candling, fuel rod failure, debris quenching and dryout. The analysis performed here is not a traditional “best-estimate” uncertainty analysis that uses best-estimate parameters or identifies best-estimate figure of merit distributions. Instead, it is an exploratory uncertainty analysis that identifies and interrogates underlying model form biases of the two material interaction models (eutectics and interactive materials models). Uniform distributions are applied to all uncertain parameters to ensure coverage of the model parameter uncertainty space. Key findings from this study include underlying model form biases exhibited by material interaction models, and notable differences in accident progression outcomes between the material interaction models. This uncertainty study extends and confirms the conclusions from the first part of this study, which compared the impact of material interaction modeling on simulation of a short-term station blackout scenario with representative Fukushima Daiichi Unit I boundary conditions. In particular, this study confirms that the eutectics model generally exhibits accelerated degradation and failure of fuel components, the core plate, and the lower head. The eutectics model also has a tendency to exhibit a greater degree of core degradation, greater debris mass formation, and larger debris mass ejection. Finally, the eutectics model exhibits higher maximum temperatures for fuel, cladding, particulate debris, oxidic molten pool, and metallic molten pool components than the interactive materials model; interactive materials model simulations exhibit a soft “limitation” on maximum temperatures that is related to the temperature at which material relocation occurs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Enabling liquid vapor analysis using synchrotron VUV single photon ionization mass spectrometry with a microfluidic interface

Vacuum ultraviolet (VUV) single photon ionization mass spectrometry (SPI-MS) is a vacuum-based technique typically used for analysis of gas phase and solid samples, but not for liquids, due to the challenge in proposing volatile liquids in vacuum. In this work we introduce the first demonstration of in situ liquid analysis by integrating the System for Analysis at the Liquid Vacuum Interface (SALVI) microfluidic reactor into VUV SPI-MS. Four representative volatile organic compound (VOC) solutions were used to illustrate feasibility of liquid analysis. Our results reflect accurate mass identification of the VOC molecules and reliable determination of appearance energy (AE) that is consistent with ionization energy (IE) for gaseous species in literature as reported. This work validates that the vacuum-compatible SALVI microfluidic interface can be utilized at the synchrotron beamline and enable in situ research of gas-phase molecules evaporating off the surface of a liquid, which holds importance in the study of condensed matter chemistry.

47 OTHER INSTRUMENTATION↗
NASA NTRS1993

Multi-resolution analysis for ENO schemes

Given a function u(x) which is represented by its cell-averages in cells which are formed by some unstructured grid, we show how to decompose the function into various scales of variation. This is done by considering a set of nested grids in which the given grid is the finest, and identifying in each locality the coarsest grid in the set from which u(x) can be recovered to a prescribed accuracy. We apply this multi-resolution analysis to Essentially Non-oscillatory Schemes (ENO) schemes in order to reduce the number of numerical flux computations which is needed in order to advance the solution by one time-step. This is accomplished by decomposing the numerical solution at the beginning of each time-step into levels of resolution, and performing the computation in each locality at the appropriate coarser grid. We present an efficient algorithm for implementing this program in the one-dimensional case; this algorithm can be extended to the multi-dimensional case with cartesian grids.

Harten, Ami↗