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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 685 records · Page 38
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

A New Analytic-Adaptive Model for EGS Assessment, Development and Management Support

To increase understanding of the energy extraction capacity of Enhanced Geothermal System(s) (EGS), a numerical model development and application project is completed. The general objective of the project is to develop and apply a new, data-coupled Thermal-Hydrological-Mechanical-Chemical (T-H-M-C) model in which the four internal components can be freely selected from existing simulation software without merging and cross-combining a diverse set of computational codes. Eight tasks are completed during the project period. The results are reported in five publications, an MS thesis, twelve quarterly, and two annual reports to DOE. Two US patents have also been issued during the project period, with one patent application originated prior to the start of the project. The “Multiphase Physical Transport Modeling Method and Modeling System” (U.S. Patent 8,396,693 B2, 2013), a key element in the GHE sub-model solution, is successfully used for EGS studies. The “Geothermal Energy Extraction System and Method" invention (U.S. Patent 8,430,166 B2, 2013) originates from the time of project performance, describing a new fluid flow control solution. The new, coupled T-H-M-C numerical model will help analyzing and designing new, efficient EGS systems.

15 GEOTHERMAL ENERGY↗
DOE OSTI2014

A New Analytic-Adaptive Model for EGS Assessment, Development and Management Support

To increase understanding of the energy extraction capacity of Enhanced Geothermal System(s) (EGS), a numerical model development and application project is completed. The general objective of the project is to develop and apply a new, data-coupled Thermal-Hydrological-Mechanical-Chemical (T-H-M-C) model in which the four internal components can be freely selected from existing simulation software without merging and cross-combining a diverse set of computational codes. Eight tasks are completed during the project period. The results are reported in five publications, an MS thesis, twelve quarterly, and two annual reports to DOE. Two US patents have also been issued during the project period, with one patent application originated prior to the start of the project. The “Multiphase Physical Transport Modeling Method and Modeling System” (U.S. Patent 8,396,693 B2, 2013), a key element in the GHE sub-model solution, is successfully used for EGS studies. The “Geothermal Energy Extraction System and Method" invention (U.S. Patent 8,430,166 B2, 2013) originates from the time of project performance, describing a new fluid flow control solution. The new, coupled T-H-M-C numerical model will help analyzing and designing new, efficient EGS systems.

15 GEOTHERMAL ENERGY↗
DOE OSTI2014

Coupled Neutronics Thermal-Hydraulic Solution of a Full-Core PWR Using VERA-CS

The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing a core simulator called VERA-CS to model operating PWR reactors with high resolution. This paper describes how the development of VERA-CS is being driven by a set of progression benchmark problems that specify the delivery of useful capability in discrete steps. As part of this development, this paper will describe the current capability of VERA-CS to perform a multiphysics simulation of an operating PWR at Hot Full Power (HFP) conditions using a set of existing computer codes coupled together in a novel method. Results for several single-assembly cases are shown that demonstrate coupling for different boron concentrations and power levels. Finally, high-resolution results are shown for a full-core PWR reactor modeled in quarter-symmetry.

Core Simulator↗
DOE OSTI2014

OECD NEA Benchmark Database of Spent Nuclear Fuel Isotopic Compositions for World Reactor Designs

Experimental data on the isotopic concentrations in irradiated nuclear fuel represent one of the primary methods for validating computational methods and nuclear data used for reactor and spent fuel depletion simulations that support nuclear fuel cycle safety and safeguards programs. Measurement data have previously not been available to users in a centralized or searchable format, and the majority of accessible information has been, for the most part, limited to light-water-reactor designs. This paper describes a recent initiative to compile spent fuel benchmark data for additional reactor designs used throughout the world that can be used to validate computer model simulations that support nuclear energy and nuclear safeguards missions. Experimental benchmark data have been expanded to include VVER-440, VVER-1000, RBMK, graphite moderated MAGNOX, gas cooled AGR, and several heavy-water moderated CANDU reactor designs. Additional experimental data for pressurized light water and boiling water reactor fuels has also been compiled for modern assembly designs and more extensive isotopic measurements. These data are being compiled and uploaded to a recently revised structured and searchable database, SFCOMPO, to provide the nuclear analysis community with a centrally-accessible resource of spent fuel compositions that can be used to benchmark computer codes, models, and nuclear data. The current version of SFCOMPO contains data for eight reactor designs, 20 fuel assembly designs, more than 550 spent fuel samples, and measured isotopic data for about 80 nuclides.

Gauld, Ian C↗
DOE OSTI2014

Modernization Enhancements in SCALE 6.2

SCALE is a widely used suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, industry, and research institutions around the world have used SCALE for nuclear safety analysis and design. However, the underlying architecture of SCALE is based on a 40-year old design with dozens of independent functional modules and control programs, primarily implemented in the Fortran programming language, with extensive use of customized intermediate files to control the logical flow of the analysis. Data are passed between individual computational codes using custom binary files that are read from and written to the hard disk. The SCALE modernization plan provides a progression towards SCALE 7, which will provide an object-oriented parallel-enabled software infrastructure with state-of-the-art methods implemented as reusable components. This paper provides a brief overview of the goals of SCALE modernization and details some modernized features available with SCALE 6.2.

modern↗
DOE OSTI2013

Synchrotron Radiation Workshop (SRW)

"Synchrotron Radiation Workshop" (SRW) is a physical optics computer code for calculation of detailed characteristics of Synchrotron Radiation (SR) generated by relativistic electrons in magnetic fields of arbitrary configuration and for simulation of the radiation wavefront propagation through optical systems of beamlines. Frequency-domain near-field methods are used for the SR calculation, and the Fourier-optics based approach is generally used for the wavefront propagation simulation. The code enables both fully- and partially-coherent radiation propagation simulations in steady-state and in frequency-/time-dependent regimes. With these features, the code has already proven its utility for a large number of applications in infrared, UV, soft and hard X-ray spectral range, in such important areas as analysis of spectral performances of new synchrotron radiation sources, optimization of user beamlines, development of new optical elements, source and beamline diagnostics, and even complete simulation of SR based experiments. Besides the SR applications, the code can be efficiently used for various simulations involving conventional lasers and other sources. SRW versions interfaced to Python and to IGOR Pro (WaveMetrics), as well as cross-platform library with C API, are available.

Chubar, O.↗
DOE OSTI2013

LLNL-Earth3D

Earth3D is a computer code designed to allow fast calculation of seismic rays and travel times through a 3D model of the Earth. LLNL is using this for earthquake location and global tomography efforts and such codes are of great interest to the Earth Science community.

Source record↗
DOE OSTI2013

Nonlinear Analysis for Event Forewarning (NLAfEW)

The NLAfEW computer code analyses noisy, experimental data to forewarn of adverse events. The functionality of the analysis is a follows: It removes artifacts from the data, converts the continuous data value to discrete values, constructs time-delay embedding vectors, comparents the unique nodes and links in one graph, and determines event forewarning on the basis of several successive occurrences of one (or more) of the dissimilarity measures above a threshold.

Hively, Lee Mizener↗
DOE OSTI2015

VERAIn

CASL's modeling and simulation technology, the Virtual Environment for Reactor Applications (VERA), incorporates coupled physics and science-based models, state-of-the-art numerical methods, modern computational science, integrated uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs), single-effect experiments, and integral tests. The computational simulation component of VERA is the VERA Core Simulator (VERA-CS). The core simulator is the specific collection of multi-physics computer codes used to model and deplete a LWR core over multiple cycles. The core simulator has a single common input file that drives all of the different physics codes. The parser code, VERAIn, converts VERA Input into an XML file that is used as input to different VERA codes.

Simunovic, Srdjan↗
DOE OSTI2016

Recovery Act. Development and Validation of an Advanced Stimulation Prediction Model for Enhanced Geothermal System

The research project aims to develop and validate an advanced computer model that can be used in the planning and design of stimulation techniques to create engineered reservoirs for Enhanced Geothermal Systems. The specific objectives of the proposal are to: 1) Develop a true three-dimensional hydro-thermal fracturing simulator that is particularly suited for EGS reservoir creation. 2) Perform laboratory scale model tests of hydraulic fracturing and proppant flow/transport using a polyaxial loading device, and use the laboratory results to test and validate the 3D simulator. 3) Perform discrete element/particulate modeling of proppant transport in hydraulic fractures, and use the results to improve understand of proppant flow and transport. 4) Test and validate the 3D hydro-thermal fracturing simulator against case histories of EGS energy production. 5) Develop a plan to commercialize the 3D fracturing and proppant flow/transport simulator. The project is expected to yield several specific results and benefits. Major technical products from the proposal include: 1) A true-3D hydro-thermal fracturing computer code that is particularly suited to EGS, 2) Documented results of scale model tests on hydro-thermal fracturing and fracture propping in an analogue crystalline rock, 3) Documented procedures and results of discrete element/particulate modeling of flow and transport of proppants for EGS applications, and 4) Database of monitoring data, with focus of Acoustic Emissions (AE) from lab scale modeling and field case histories of EGS reservoir creation.

15 GEOTHERMAL ENERGY↗
DOE OSTI2015

Neutron Source Facility Training Simulator Based on EPICS

A plant operator training simulator is developed for training the plant operators as well as for design verification of plant control system (PCS) and plant protection system (PPS) for the Kharkov Institute of Physics and Technology Neutron Source Facility. The simulator provides the operator interface for the whole plant including the sub-critical assembly coolant loop, target coolant loop, secondary coolant loop, and other facility systems. The operator interface is implemented based on Experimental Physics and Industrial Control System (EPICS), which is a comprehensive software development platform for distributed control systems. Since its development at Argonne National Laboratory, it has been widely adopted in the experimental physics community, e.g. for control of accelerator facilities. This work is the first implementation for a nuclear facility. The main parts of the operator interface are the plant control panel and plant protection panel. The development involved implementation of process variable database, sequence logic, and graphical user interface (GUI) for the PCS and PPS utilizing EPICS and related software tools, e.g. sequencer for sequence logic, and control system studio (CSS-BOY) for graphical use interface. For functional verification of the PCS and PPS, a plant model is interfaced, which is a physics-based model of the facility coolant loops implemented as a numerical computer code. The training simulator is tested and demonstrated its effectiveness in various plant operation sequences, e.g. start-up, shut-down, maintenance, and refueling. It was also tested for verification of the plant protection system under various trip conditions.

Park, Young Soo↗
DOE OSTI2016

ITER Port Interspace Pressure Calculations

The ITER Vacuum Vessel (VV) is equipped with 54 access ports. Each of these ports has an opening in the bioshield that communicates with a dedicated port cell. During Tokamak operation, the bioshield opening must be closed with a concrete plug to shield the radiation coming from the plasma. This port plug separates the port cell into a Port Interspace (between VV closure lid and Port Plug) on the inner side and the Port Cell on the outer side. This paper presents calculations of pressures and temperatures in the ITER (Ref. 1) Port Interspace after a double-ended guillotine break (DEGB) of a pipe of the Tokamak Cooling Water System (TCWS) with high temperature water. It is assumed that this DEGB occurs during the worst possible conditions, which are during water baking operation, with water at a temperature of 523 K (250 C) and at a pressure of 4.4 MPa. These conditions are more severe than during normal Tokamak operation, with the water at 398 K (125 C) and 2 MPa. Two computer codes are employed in these calculations: RELAP5-3D Version 4.2.1 (Ref. 2) to calculate the blowdown releases from the pipe break, and MELCOR, Version 1.8.6 (Ref. 3) to calculate the pressures and temperatures in the Port Interspace. A sensitivity study has been performed to optimize some flow areas.

Interspace ITER Vacuum vessel↗
DOE OSTI2015

Target Design Optimization of KIPT Neutron Source Facility

Argonne National Laboratory (ANL) of the United States developed and designed a neutron source facility for Kharkov Institute of Physics and Technology (KIPT) of Ukraine. The facility was constructed at Kharkov, Ukraine and its commissioning process has been started. The facility has an electron accelerator driving a subcritical assembly. The electron beam power is 100 kW using 100 MeV electrons. The subcritical assembly has WWR-M2 fuel assemblies with U-235 enrichment of 19.7 wt%. The facility will be utilized to perform basic and applied nuclear research, to produce medical isotopes, and to train young nuclear specialists. Solid target design with stacked disks is selected and each target disk is cooled by water from both side. Tungsten or natural uranium is the target material. This paper presents the target design optimization to maximize the neutron yield and the neutron flux level of the facility while satisfying the thermal-hydraulic design criteria. Monte Carlo computer code MCNPX is utilized for the neutron yield analyses with ENDF/B-VII.0 nuclear data libraries, as a function of target thickness using a simplified target model without cladding or coolant channels. It’s found that the neutron yield saturates at target thickness 60 ~ 70 mm for both tungsten and uranium materials. For the final target design with cladding and coolant channels, the total uranium thickness is 56.5 mm, and the total tungsten thickness is only 33.0 mm. The tungsten material has a large absorption cross section for thermal neutrons. If the tungsten thickness is increased, the neutron yield gain is offset by the neutron absorption reaction. The paper presents the design optimization analyses for both target materials.

electron accelerator↗
DOE OSTI2016

Revised Extended Grid Library

The Revised Eolus Grid Library (REGL) is a mesh-tracking library that was developed for use with the MCNP6TM computer code so that (radiation) particles can track on an unstructured mesh. The unstructured mesh is a finite element representation of any geometric solid model created with a state-of-the-art CAE/CAD tool. The mesh-tracking library is written using modern Fortran and programming standards; the library is Fortran 2003 compliant. The library was created with a defined application programmer interface (API) so that it could easily integrate with other particle tracking/transport codes. The library does not handle parallel processing via the message passing interface (mpi), but has been used successfully where the host code handles the mpi calls. The library is thread-safe and supports the OpenMP paradigm. As a library, all features are available through the API and overall a tight coupling between it and the host code is required. Features of the library are summarized with the following list: Can accommodate first and second order 4, 5, and 6-sided polyhedra; any combination of element types may appear in a single geometry model; parts may not contain tetrahedra mixed with other element types; pentahedra and hexahedra can be together in the same part; robust handling of overlaps and gaps; tracks element-to-element to produce path length results at the element level; finds element numbers for a given mesh location; finds intersection points on element faces for the particle tracks; produce a data file for post processing results analysis; reads Abaqus .inp input (ASCII) files to obtain information for the global mesh-model; supports parallel input processing via mpi; and support parallel particle transport by both mpi and OpenMP.

Martz, Roger L.↗
DOE OSTI2017

Weapon Physicist Declassifies Rescued Nuclear Test Films

The U.S. conducted 210 atmospheric nuclear tests between 1945 and 1962, with multiple cameras capturing each event at around 2,400 frames per second. But in the decades since, around 10,000 of these films sat idle, scattered across the country in high-security vaults. Not only were they gathering dust, the film material itself was slowly decomposing, bringing the data they contained to the brink of being lost forever. For the past five years, Lawrence Livermore National Laboratory (LLNL) weapon physicist Greg Spriggs and a crack team of film experts, archivists and software developers have been on a mission to hunt down, scan, reanalyze and declassify these decomposing films. The goals are to preserve the films’ content before it’s lost forever, and provide better data to the post-testing-era scientists who use computer codes to help certify that the aging U.S. nuclear deterrent remains safe, secure and effective.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2017

Finite Element Solver for Fission Dynamics

FELIX is a physics computer code used to model fission fragment mass distributions in a fully quantum-mechanical, microscopic framework that only relies on our current knowledge of nuclear forces [1]. It is an implementation of the time-dependent generator coordinate method (TDGCM), which simulates the dynamics of a collective quantum wave-packet assuming the motion is adiabatic. In typical applications of the TDFCM, the nuclear collective wavepacket is obtained as a superposition of wavefunctions obtained by solving the Hartree-Fock-Bogoliubov equations of nuclear density functional theory (DFT). The program calculates at each time step the coefficients of that superposition.

Regnier, D.↗
DOE OSTI2017

Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT)

The restart of the Transient REActor Test (TREAT) facility provides a unique opportunity to engage the fast reactor fuels community to reinitiate in-pile experimental safety studies. Historically, the TREAT facility played a critical role in characterizing the behavior of both metal and oxide fast reactor fuels under off-normal conditions, irradiating hundreds of fuel pins to support fast reactor fuel development programs. The resulting test data has provided validation for a multitude of fuel performance and severe accident analysis computer codes. This paper will provide a review of the historical database of TREAT experiments including experiment design, instrumentation, test objectives, and salient findings. Additionally, the paper will provide an introduction to the current and future experiment plans of the U.S. transient testing program at TREAT.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Containment Sodium Chemistry Models in MELCOR.

To meet regulatory needs for sodium fast reactors’ future development, including licensing requirements, Sandia National Laboratories is modernizing MELCOR, a severe accident analysis computer code developed for the U.S. Nuclear Regulatory Commission (NRC). Specifically, Sandia is modernizing MELCOR to include the capability to model sodium reactors. However, Sandia’s modernization effort primarily focuses on the containment response aspects of the sodium reactor accidents. Sandia began modernizing MELCOR in 2013 to allow a sodium coolant, rather than water, for conventional light water reactors. In the past three years, Sandia has been implementing the sodium chemistry containment models in CONTAIN-LMR, a legacy NRC code, into MELCOR. These chemistry models include spray fire, pool fire and atmosphere chemistry models. Only the first two chemistry models have been implemented though it is intended to implement all these models into MELCOR. A new package called “NAC” has been created to manage the sodium chemistry model more efficiently. In 2017 Sandia began validating the implemented models in MELCOR by simulating available experiments. The CONTAIN-LMR sodium models include sodium atmosphere chemistry and sodium-concrete interaction models. This paper presents sodium property models, the implemented models, implementation issues, and a path towards validation against existing experimental data.

Louie, David↗