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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 1,189 records · Page 66
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Using GTO-Velo to Facilitate Communication and Sharing of Simulation Results in Support of the Geothermal Technologies Office Code Comparison Study

The Geothermal Technologies Office Code Comparison Study (GTO-CCS) aims to support the DOE Geothermal Technologies Office in organizing and executing a model comparison activity. This project is directed at testing, diagnosing differences, and demonstrating modeling capabilities of a worldwide collection of numerical simulators for evaluating geothermal technologies. Teams of researchers are collaborating in this code comparison effort, and it is important to be able to share results in a forum where technical discussions can easily take place without requiring teams to travel to a common location. Pacific Northwest National Laboratory has developed an open-source, flexible framework called Velo that provides a knowledge management infrastructure and tools to support modeling and simulation for a variety of types of projects in a number of scientific domains. GTO-Velo is a customized version of the Velo Framework that is being used as the collaborative tool in support of the GTO-CCS project. Velo is designed around a novel integration of a collaborative Web-based environment and a scalable enterprise Content Management System (CMS). The underlying framework provides a flexible and unstructured data storage system that allows for easy upload of files that can be in any format. Data files are organized in hierarchical folders and each folder and each file has a corresponding wiki page for metadata. The user interacts with Velo through a web browser based wiki technology, providing the benefit of familiarity and ease of use. High-level folders have been defined in GTO-Velo for the benchmark problem descriptions, descriptions of simulator/code capabilities, a project notebook, and folders for participating teams. Each team has a subfolder with write access limited only to the team members, where they can upload their simulation results. The GTO-CCS participants are charged with defining the benchmark problems for the study, and as each GTO-CCS Benchmark problem is defined, the problem creator can provide a description using a template on the metadata page corresponding to the benchmark problem folder. Project documents, references and videos of the weekly online meetings are shared via GTO-Velo. A results comparison tool allows users to plot their uploaded simulation results on the fly, along with those of other teams, to facilitate weekly discussions of the benchmark problem results being generated by the teams. GTO-Velo is an invaluable tool providing the project coordinators and team members with a framework for collaboration among geographically dispersed organizations.

Collaborative Tools↗
DOE OSTI2014

Reducing EnergyPlus Run Time For Code Compliance Tools

Integration of the EnergyPlus ™ simulation engine into performance-based code compliance software raises a concern about simulation run time, which impacts timely feedback of compliance results to the user. EnergyPlus annual simulations for proposed and code baseline building models, and mechanical equipment sizing result in simulation run times beyond acceptable limits. This paper presents a study that compares the results of a shortened simulation time period using 4 weeks of hourly weather data (one per quarter), to an annual simulation using full 52 weeks of hourly weather data. Three representative building types based on DOE Prototype Building Models and three climate zones were used for determining the validity of using a shortened simulation run period. Further sensitivity analysis and run time comparisons were made to evaluate the robustness and run time savings of using this approach. The results of this analysis show that the shortened simulation run period provides compliance index calculations within 1% of those predicted using annual simulation results, and typically saves about 75% of simulation run time.

energyplus, simulation, runtime, run time, comchec↗
DOE OSTI2016

Overview of the U.S. DOE Hydrogen Safety, Codes and Standards Program. Part 4: Hydrogen Sensors; Preprint

Hydrogen sensors are recognized as a critical element in the safety design for any hydrogen system. In this role, sensors can perform several important functions including indication of unintended hydrogen releases, activation of mitigation strategies to preclude the development of dangerous situations, activation of alarm systems and communication to first responders, and to initiate system shutdown. The functionality of hydrogen sensors in this capacity is decoupled from the system being monitored, thereby providing an independent safety component that is not affected by the system itself. The importance of hydrogen sensors has been recognized by DOE and by the Fuel Cell Technologies Office's Safety and Codes Standards (SCS) program in particular, which has for several years supported hydrogen safety sensor research and development. The SCS hydrogen sensor programs are currently led by the National Renewable Energy Laboratory, Los Alamos National Laboratory, and Lawrence Livermore National Laboratory. The current SCS sensor program encompasses the full range of issues related to safety sensors, including development of advance sensor platforms with exemplary performance, development of sensor-related code and standards, outreach to stakeholders on the role sensors play in facilitating deployment, technology evaluation, and support on the proper selection and use of sensors.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2016

Bringing Fenton Hill into the Digital Age: Data Conversion in Support of the Geothermal Technologies Office Code Comparison Study Challenge Problems

The Geothermal Technologies Office Code Comparison Study (GTO-CCS) was established by the U.S. Department of Energy to facilitate collaboration among members of the geothermal modeling community and to evaluate and improve upon the ability of existing codes to simulate thermal, hydrological, mechanical, and chemical processes associated with complex enhanced geothermal systems (EGS). The first stage of the project, which has been completed, involved comparing simulations for seven benchmark problems that were primarily designed using well-prescribed, simplified data sets. In the second stage, the participating teams are tackling two challenge problems based on the EGS research conducted in hot dry rock (HDR) at Fenton Hill, near Los Alamos, New Mexico. The Fenton Hill project, conducted by Los Alamos National Laboratory (LANL) from 1970 to 1995, was the world’s first HDR demonstration project. One of the criteria for selecting this experiment as the basis for the challenge problems was the amount and availability of data for generating model inputs. The Fenton Hill HDR system consisted of two reservoirs – an earlier Phase I reservoir tested from 1974 to 1981 and a deeper Phase II reservoir tested from 1980 to 1995. Detailed accounts of both phases of the HDR project have been presented in a number of books and reports, including a recently published summary of the lessons learned and a final report with a chronological description of the Fenton Hill project, prepared by LANL. Project documents and records have been archived and made public through the National Geothermal Data System (NGDS). Some of the data acquired from Phase II are available in electronic format readable on modern computers. These include the microseismic data from some of the important experiments (e.g. the massive hydraulic fracturing test conducted in 1983) and the injection/production wellhead data from the circulation tests conducted between 1992-1995. However, much of the data collected during the project, while publicly available, currently only exist in the form of tables or graphs within scanned documents. Therefore, in support of the GTO-CCS, the data needed for developing simulation inputs are being compiled and converted to platform independent, open readable formats so that all participating teams will have access to the same electronic data set. In some cases this requires conversion using optical character recognition, digitizing existing images, and generating the appropriate metadata from project documents. The GTO-Velo knowledge management framework, developed by Pacific Northwest National Laboratory (PNNL), was used for the benchmark problem stage of the comparison study and will also be used as the data repository for the challenge problem data sets. It is staggering and impractical to convert all published data for the Fenton Hill site, so the focus is on data that supports simulations for the three topical areas defined by the study for the challenge problems: 1) reservoir creation/stimulation, 2) reactive and passive transport, and 3) thermal recovery. Conversion of these data provide value not only to GTO-CCS participants, but also to members of the geothermal community at large who may be interested in revisiting the Fenton Hill experiment in the future.

Code Comparison Study↗
DOE OSTI2017

How the Geothermal Community Upped the Game for Computer Codes

The Geothermal Technologies Office Code Comparison Study brought 11 research institutions together to collaborate on coupled thermal, hydrologic, geomechanical, and geochemical numerical simulators. These codes have the potential to help facilitate widespread geothermal energy development.

15 GEOTHERMAL ENERGY↗
DOE OSTI2016

Modelling of the Gadolinium Fuel Test IFA-681 using the BISON Code

In this work, application of Idaho National Laboratory’s fuel performance code BISON to modelling of fuel rods from the Halden IFA-681 gadolinium fuel test is presented. First, an overview is given of BISON models, focusing on UO2/UO2-Gd2O3 fuel and Zircaloy cladding. Then, BISON analyses of selected fuel rods from the IFA-681 test are performed. For the first time in a BISON application to integral fuel rod simulations, the analysis is informed by detailed neutronics calculations in order to accurately capture the radial power profile throughout the fuel, which is strongly affected by the complex evolution of absorber Gd isotopes. In particular, radial power profiles calculated at IFE–Halden Reactor Project with the HELIOS code are used. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project. Some slide have been added as an Appendix to present the newly developed PolyPole-1 algorithm for modeling of intra-granular fission gas release.

BISON code↗
DOE OSTI2023

ASME Code Development - Nonmetallics

ASME Code Development - Nonmetallics slides to include behavior models, licensing & code, graphite R&D Program, As-Fab'd Properties, Mechanisms and Analysis, and Irradiation. Discussion on where these fit into Irradiation behavior, Ceramic Composites, Design, and Component Failure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Verification of the shift Monte Carlo code with the C5G7 reactor benchmark

Shift is a new hybrid Monte Carlo/deterministic radiation transport code being developed at Oak Ridge National Laboratory. At its current stage of development, Shift includes a parallel Monte Carlo capability for simulating eigenvalue and fixed-source multigroup transport problems. This paper focuses on recent efforts to verify Shift's Monte Carlo component using the two-dimensional and three-dimensional C5G7 NEA benchmark problems. Comparisons were made between the benchmark eigenvalues and those output by the Shift code. In addition, mesh-based scalar flux tally results generated by Shift were compared to those obtained using MCNP5 on an identical model and tally grid. The Shift-generated eigenvalues were within three standard deviations of the benchmark and MCNP5-1.60 values in all cases. The flux tallies generated by Shift were found to be in very good agreement with those from MCNP. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Mesh generation and energy group condensation studies for the jaguar deterministic transport code

The deterministic transport code Jaguar is introduced, and the modeling process for Jaguar is demonstrated using a two-dimensional assembly model of the Hoogenboom-Martin Performance Benchmark Problem. This single assembly model is being used to test and analyze optimal modeling methodologies and techniques for Jaguar. This paper focuses on spatial mesh generation and energy condensation techniques. In this summary, the models and processes are defined as well as thermal flux solution comparisons with the Monte Carlo code MC21. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Pool scrubbing under jet injection regime: An enhancement of the SPARC90 code

The SPARC90 code was developed to calculate the aerosol pool trapping during vent discharge processes, at low gas velocities. However, there are accident sequences, like SGTR core meltdown sequences, at which particle laden gases reach the aqueous ponds at very high velocities and new particle removal mechanisms become effective right at the inlet. As a result of the shearing off of roll wave water crests, water droplets are entrained in the gas core and sweep out aerosol particles, mainly by inertial impaction and interception. This paper summarizes the update of the SPARC90 code based on state-of-the-art equations for jet hydrodynamics and aerosol removal. Equations for variables like droplets population, size and velocity have been implemented. Based on the anticipated conditions in case of an SGTR severe accident sequence, comparisons of estimates from this new version (SPAR90-Jet) and the original one are set in terms of decontamination factor. Even though further work is still ahead, this work highlights how substantial particle retention at the pool inlet can reach under jet regime and how different aerosol removal mechanisms are with respect to the globule injection regime. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Generation of Bondarenko F factors with MCNP5 for use in scale transport codes

Generally there are three methods of cross-section processing available when using the Scale computer code. They are NITAWL, BONAMI, and CENTRM, with CENTRM being the most common and accurate, but computationally expensive. In order to improve the accuracy of BONAMI (which uses the Bondarenko Method), new Bondarenko/F Factors were to be generated that will smooth out the current F Factors that are being generated using CENTRM. The case discussed here involves using a dedicated Monte Carlo code (MCNP5) to calculate the transport solution from which shielded cross-sections can be produced directly for the transport geometry/mesh. A simple program was created to parse and collect the tallied cross-sections from the MCNP output, which was fed into a modified CLAROL input (a module that replaces or adds data in an AMPX master library), which used the cross-sections obtained from MCNP to calculate new F-factors and update the SCALE library. This approach allows the use of other methods to generate the shielded cross-sections and for easy comparison to existing results. Initial proof-of-principle calculations were carried out for an various cases using various transport solvers, such as NEWT, KENO, and XSDRN, with BONAMI in SCALE. Poor results were obtained using cross-sections generated using infinite homogeneous cases, but good results were obtained by using pin cells in an infinite lattice. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Calculation of natural convection test at Phenix using the NETFLOW++ code

The present paper describes modeling and analyses of a natural convection of the pool-type fast breeder reactor Phenix. The natural convection test was carried out as one of the End of Life Tests of the Phenix. Objective of the present study is to assess the applicability of the NETFLOW++ code which has been verified thus far using various water facilities and validated using the plant data of the loop-type FBR 'Monju' and the loop-type experimental fast reactor 'Joyo'. The Phenix primary heat transport system is modeled based on the benchmark documents available from IAEA. The calculational model consists of only the primary heat transport system with boundary conditions on the secondary-side of IHX. The coolant temperature at the primary pump inlet, the primary coolant temperature at the IHX inlet and outlet, the secondary coolant temperatures and other parameters are calculated by the code where the heat transfer between the hot and cold pools is explicitly taken into account. A model including the secondary and tertiary systems was prepared, and the calculated results also agree well with the measured data in general. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Improvement of three-field based safety analysis code, SPACE, through verification and validation

To verify and validate the state-of-the-art safety analysis code, SPACE, features of three field governing equations were arranged and the comparative analysis with both analytic solution and experimental benchmarks were performed. With the world wide streamline on the safety analysis, SPACE uses two-fluid, three-field governing equations which aim to implement multi-dimensional and multi-scaled analysis. Through the Verification and Validation (V and V), developing code based on the three-field governing equation have not only been confirmed but also improved. V and V activities on SPACE have been conducted as to interfacial drag model in vertical turbulent flow. The verification of the interfacial drag model was based on cell size sensitivity test for confirming feasible calculation range of cell size. Furthermore, through the validation with analytic solution, the modification of the interfacial drag model in vertical turbulent annulus flow was carried out and the nominated model was incorporated. Modified interfacial drag model, Asali's correlation, yields improved result than previous interfacial drag model. The validation with experimental benchmarks had also been performed on the nearly horizontal counter current limitation. To confirm the possibility of the further application of the CCFL model in case with horizontal pipe, validation with nearly horizontal pipe experiment was carried out. V and V activities of SPACE with analytic solution and experimental benchmarks allow us to confirm/identify the tailorability for safety analysis. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Development of one-dimensional computational fluid dynamics code 'GFLOW' for groundwater flow and contaminant transport analysis

Prediction of groundwater movement and contaminant transport in soil is an important problem in many branches of science and engineering. This includes groundwater hydrology, environmental engineering, soil science, agricultural engineering and also nuclear engineering. Specifically, in nuclear engineering it is applicable in the design of spent fuel storage pools and waste management sites in the nuclear power plants. Ground water modeling involves the simulation of flow and contaminant transport by groundwater flow. In the context of contaminated soil and groundwater system, numerical simulations are typically used to demonstrate compliance with regulatory standard. A one-dimensional Computational Fluid Dynamics code GFLOW had been developed based on the Finite Difference Method for simulating groundwater flow and contaminant transport through saturated and unsaturated soil. The code is validated with the analytical model and the benchmarking cases available in the literature. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Overview of development and design of MPACT: Michigan parallel characteristics transport code

MPACT (Michigan Parallel Characteristics Transport Code) is a new reactor analysis tool. It is being developed by students and research staff at the University of Michigan to be used for an advanced pin-resolved transport capability within VERA (Virtual Environment for Reactor Analysis). VERA is the end-user reactor simulation tool being produced by the Consortium for the Advanced Simulation of Light Water Reactors (CASL). The MPACT development project is itself unique for the way it is changing how students do research to achieve the instructional and research goals of an academic institution, while providing immediate value to industry. The MPACT code makes use of modern lean/agile software processes and extensive testing to maintain a level of productivity and quality required by CASL. MPACT's design relies heavily on object-oriented programming concepts and design patterns and is programmed in Fortran 2003. These designs are explained and illustrated as to how they can be readily extended to incorporate new capabilities and research ideas in support of academic research objectives. The transport methods currently implemented in MPACT include the 2-D and 3-D method of characteristics (MOC) and 2-D and 3-D method of collision direction probabilities (CDP). For the cross section resonance treatment, presently the subgroup method and the new embedded self-shielding method (ESSM) are implemented within MPACT. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Recent advances in the Mercury Monte Carlo particle transport code

We review recent physics and computational science advances in the Mercury Monte Carlo particle transport code under development at Lawrence Livermore National Laboratory. We describe recent efforts to enable a nuclear resonance fluorescence capability in the Mercury photon transport. We also describe recent work to implement a probability of extinction capability into Mercury. We review the results of current parallel scaling and threading efforts that enable the code to run on millions of MPI processes. (authors)

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2013

Nuclear data processing for energy release and deposition calculations in the MC21 Monte Carlo code

With the recent emphasis in performing multiphysics calculations using Monte Carlo transport codes such as MC21, the need for accurate estimates of the energy deposition-and the subsequent heating - has increased. However, the availability and quality of data necessary to enable accurate neutron and photon energy deposition calculations can be an issue. A comprehensive method for handling the nuclear data required for energy deposition calculations in MC21 has been developed using the NDEX nuclear data processing system and leveraging the capabilities of NJOY. The method provides a collection of data to the MC21 Monte Carlo code supporting the computation of a wide variety of energy release and deposition tallies while also allowing calculations with different levels of fidelity to be performed. Detailed discussions on the usage of the various components of the energy release data are provided to demonstrate novel methods in borrowing photon production data, correcting for negative energy release quantities, and adjusting Q values when necessary to preserve energy balance. Since energy deposition within a reactor is a result of both neutron and photon interactions with materials, a discussion on the photon energy deposition data processing is also provided. (authors)

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2016

Modeling thermochemistry of fuel and coupling to fuel performance codes

A thermochemical library, Thermochimica, utilizes new algorithms and solvers designed for calculating equilibria of multicomponent and multiphase systems and has been coupled to the nuclear fuel performance code BISON. Thermochimica utilizes a generalized thermochemical database for urania-based fuels that has been under development for a number of years and continues to be expanded. One of the key motivations for use of thermodynamic models is that they can better represent the physics of oxygen migration. The coupling of the BISON code with Thermochimica and examples of its use in representing fuel phenomena are described. In the current implementation in BISON, fluxes are driven by the elemental concentration gradient. However, equilibrium is characterized by the absence of spatial variations of chemical potentials of the system species/components. The driving force to equilibrium (species flux) is proportional to the deviation from equilibrium, i.e., the gradient in chemical potential. It is therefore necessary to integrate chemical potential driven diffusion in the BISON representation of oxygen transport. The diffusion kernels based on thermodynamic models have been developed to simulate multicomponent transport in light water reactor (LWR) materials. The Thermochimica library was used to calculate various thermodynamics properties needed for transport calculations. The library also can calculate properties that are specific to various transport mechanisms for different fuel materials. For example, it can calculate defect site fractions among the sublattices in oxide fuels as a function of temperature, burnup, and stoichiometry. The transport and thermodynamic models are being integrated with other simulation efforts such as calculation of mobility functions that couple fluxes and chemical gradients. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗