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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 883 records · Page 49
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Atomic and Molecular Physics Effects on Ion Flow into Hot Surfaces

Research during this project at the University of Wisconsin Fusion Technology Institute (UW FTI) on ion and neutral flow through an arbitrary, monotonic potential difference created by nearly transparent electrodes accomplished the following: (1) developed and implemented an integral equation approach for atomic physics effects in helium plasmas; (2) extended the analysis to coupled integral equations that treat atomic and molecular deuterium ions and neutrals; (3) implemented the key deuterium and helium atomic and molecular cross sections; (4) added negative ion production and related cross sections; and (5) benchmarked the code against experimental results. The analysis and codes treat the species D 0 , D 2 0 , D + , D 2 + , D 3 + , D - and, separately at present, He 0 and He + . Extensions enhanced the analysis and related computer codes to include He ++ ions plus planar and cylindrical geometries.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2013

Development of an object-oriented ORIGEN for advanced nuclear fuel modeling applications

The ORIGEN package serves as the core depletion and decay calculation module within the SCALE code system. A recent major re-factor to the ORIGEN code architecture as part of an overall modernization of the SCALE code system has both greatly enhanced its maintainability as well as afforded several new capabilities useful for incorporating depletion analysis into other code frameworks. This paper will present an overview of the improved ORIGEN code architecture (including the methods and data structures introduced) as well as current and potential future applications utilizing the new ORIGEN framework. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Summary of improvements in modeling for fidelity and performance in MAMBA during FY18

This milestone report summarizes the FMC work on improvements to the MAMBA modeling methodology. The work was performed in close collaboration with the PHI as well as THM focus area and it was undertaken to improve both the physical fidelity and computational performance of MAMBA. We first describe the path towards identifying and then resolving issues in the MAMBA code and the underlying physics models that were still remaining at the beginning of FY18, after the code re-write in the previous year. These were primarily related to boiling and species transport, especially boron, within the CRUD. The code problems originated in the MAMBA3D code, from which the development of the new re-written code started, but they were only identified during testing of the new MAMBA code. We purposefully discuss a number of alternative explanations and solutions to the main issue identified, since the two-phase character of fluid transport in the CRUD and the related mass transport are still sources of model uncertainty that we continue to investigate. The hands-on code development referenced in this report was done under PHI and this milestone mainly provided support in the form of analysis, literature review, code review and feedback to the code developers.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2012

High conversion Th-U{sup 233} fuel assembly for current generation of PWRs

This paper presents a preliminary design of a high conversion Th-U{sup 233} fuel assembly applicable for current generation of Pressurized Water Reactor (PWRs). The considered fuel assembly has a typical 17 x 17 PWR lattice. However in order to increase the conversion of Th{sup 232} to U{sup 233}, the assembly was subdivided into the two regions called seed and blanket. The central seed region has a higher than blanket U{sup 233} content and acts as a neutron source for the peripheral blanket region. The latest acts as a U{sup 233} breeder. While the seed fuel pins have a standard dimensions the blanket fuel radius was increased in order to reduce the moderation and to facilitate the resonance neutron absorption in blanket Th{sup 232}. The U{sup 233} content in the seed and blanket regions was optimized to achieve maximal initial to discharged fissile inventory ratio (FIR) taking into account the target fuel cycle length of 12 months with 3-batch reloading scheme. In this study the neutronic calculations were performed on the fuel assembly level using Helios deterministic lattice transport code. The fuel cycle length and the core k{sub eff} were estimated by applying the Non Linear Reactivity Model. The applicability of the HELIOS code for the analysis of the Th-based high conversion designs was confirmed with the help of continuous-energy Monte-Carlo code SERPENT. The results of optimization studies show that for the heterogeneous seed and blanket (SB) fuel assembly the FIR of about 0.95 can be achieved. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Identification of RESNET HERS Index Values Corresponding to Minimal Compliance with the IECC Performance Path

Pacific Northwest National Laboratory analyzed the relationship between the Residential Energy Services Network (RESNET) Home Energy Rating System (HERS) Index and the traditional simulation-based Performance Path used in the International Energy Conservation Code (IECC). The analysis evaluates, for a single-family residence with various characteristics, the ranges of HERS Index values that would imply compliance with the 2012 IECC Performance Path. Several building characteristics considered likely to result in quantifiable differences in the outcomes of the two approaches, or otherwise believed to be of interest to code developers and policy makers, are considered in the analysis.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2023

Thermomechanical analysis and modeling of involute-shaped fuel plates using the Cheverton–Kelley experiments for the High Flux Isotope Reactor

Three research reactors with involute-shaped fuel plates are pursuing conversion from highly enriched uranium to low-enriched uranium fuel. Various core design and safety evaluation studies are essential to assess the feasibility of the conversion. The use of 3D computational multiphysics codes is being explored in these analyses and therefore they must undergo a thorough evaluation and quality assurance process due to their potential impact on nuclear safety. Here, the Cheverton and Kelley physical tests performed in the late 1960s to investigate the deflections of HFIR’s outer plate under uniform pressure and temperature fields are simulated by employing commercially available computational codes, with the goals to (1) verify and validate the models and numerical solvers implemented in the codes for thermomechanical analysis of involute reactor plates and (2) to develop a benchmark computational test to evaluate future versions of existing software or newly developed computational codes. The results of the simulations showed good agreement with each other as well as against the Cheverton–Kelley experimental data. Some minor deviations were observed for a few multiphysics cases and their potential origins and impact on the analysis results is investigated in the paper. The validated models increase the confidence in using multiphysics codes to evaluate existing or new LEU designs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Three-dimensional analysis using polygons in particle and heavy ion transport code system (PHITS) - 25504

Recently a function to treat tetrahedral-mesh geometry was introduced in Particle and Heavy Ion Transport code System (PHITS). Owing to this function, three-dimensional analysis using polygons became possible. The work flow of the three-dimensional analysis using a particular polygon object was demonstrated, converting the data into tetrahedral mesh with the tetrahedral-mesh generator TetGen, computing the radiation transport with PHITS, and finally analyzing with three-dimensional data analysis application ParaView. (author)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2017

Flow-induced vibration analysis of a helical coil steam generator experiment using large eddy simulation

This paper describes a numerical study of flow-induced vibration in a helical coil steam generator experiment conducted at Argonne National Laboratory in the 1980s. In the experiment, a half-scale sector model of a steam generator helical coil tube bank was subjected to still and flowing air and water, and the vibrational characteristics were recorded. The research detailed in this document utilizes the multi-physics simulation toolkit SHARP developed at Argonne National Laboratory, in cooperation with Lawrence Livermore National Laboratory, to simulate the experiment. SHARP uses the spectral element code Nek5000 for fluid dynamics analysis and the finite element code DIABLO for structural analysis. The flow around the coil tubes is modeled in Nek5000 by using a large eddy simulation turbulence model. Transient pressure data on the tube surfaces is sampled and transferred to DIABLO for the structural simulation. The structural response is simulated in DIABLO via an implicit time-marching algorithm and a combination of continuum elements and structural shells. Tube vibration data (acceleration and frequency) are sampled and compared with the experimental data. Currently, only one-way coupling is used, which means that pressure loads from the fluid simulation are transferred to the structural simulation but the resulting structural displacements are not fed back to the fluid simulation

CFD↗
DOE OSTI2025

Quantifying and Zoning Urban Heat Island Effects Using Unsupervised Machine Learning

This work explores the Urban Heat Island (UHI) effects in Maricopa County, Arizona, employing a simulation-based approach that combines large-scale building energy modeling with advanced spatial analysis. Utilizing the Automatic Building Energy Modeling (AutoBEM) software suite, we simulated the energy consumption for approximately 1.35 million buildings based on the Model America version 1.0 (MAv1) dataset. Our methodology incorporated spatial analysis at multiple scales, including individual buildings, clusters of zones determined by K-means clustering, and geographical level evaluation based on Zip codes. The results revealed significant variations in energy consumption and heat emissions across different building types and urban zones. High-emission hotspots identified through clustering pointed to areas most contributing to the UHI effects. Zip code-based area analysis further contextualized these findings, offering an urban context-based perspective on emission distribution and informing potential urban energy policies for mitigating UHI effects.

Chowdhury, Shovan [ORNL]↗
DOE OSTI2016

Analysis of Time Dependent Data and PRA

In the past decades, several numerical simulation codes have been employed to simulate accident dynamics (e.g., RELAP5-3D, MELCOR, MAAP). In order to evaluate the impact of uncertainties into accident dynamics, several stochastic methodologies have been coupled with these codes. These stochastic methods range from classical Monte-Carlo and Latin Hypercube sampling to stochastic polynomial methods. Similar approaches have been introduced into the risk and safety community where stochastic methods (such as RAVEN, ADAPT, MCDET, ADS) have been coupled with safety analysis codes in order to evaluate the safety impact of timing and sequencing of events. These approaches are usually called Dynamic PRA or simulation-based PRA methods. These uncertainties and safety methods usually generate a large number of simulation runs (database storage may be on the order of gigabytes or higher). The scope of this paper is to present a broad overview of methods and algorithms that can be used to analyze and extract information from large data sets containing time dependent data. In this context, 'extracting information' means constructing input-output correlations, finding commonalities, and identifying outliers. Some of the algorithms presented here have been developed or are under development within the RAVEN statistical framework. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Preliminary Analysis of FFTF Loss-of-Flow Tests with SAS4A/SASSYS-1

The Fast Flux Test Facility (FFTF) was the most recent Sodium-cooled Fast Reactor (SFR) built in the United States. The Passive Safety Test (PST) program was carried out on FFTF between 1984 and 1986 to demonstrate the capability of passive reactor shutdown and decay heat removal in response to protected and unprotected transients. In order to recover the information and knowledge that has been gained from FFTF, a research project was initiated focusing on the simulation of Unprotected Loss of Flow Without Scram (LOFWOS) tests similar to the knowledge preservation activities conducted for EBR-II by Argonne National Laboratory in earlier years. These knowledge preservation activities are intended to support the validation of computational tools and methods in the area of SFR passive safety. Comparisons with experimental data and other safety codes provide opportunities to improve SFR design and safety analysis codes and methods.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

PetIGA: A framework for high-performance isogeometric analysis

We present PetIGA, a code framework to approximate the solution of partial differential equations using isogeometric analysis. PetIGA can be used to assemble matrices and vectors which come from a Galerkin weak form, discretized with Non-Uniform Rational B-spline basis functions. We base our framework on PETSc, a high-performance library for the scalable solution of partial differential equations, which simplifies the development of large-scale scientific codes, provides a rich environment for prototyping, and separates parallelism from algorithm choice. We describe the implementation of PetIGA, and exemplify its use by solving a model nonlinear problem. To illustrate the robustness and flexibility of PetIGA, we solve some challenging nonlinear partial differential equations that include problems in both solid and fluid mechanics. Lastly, we show strong scaling results on up to 4096 cores, which confirm the suitability of PetIGA for large scale simulations.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2011

Summary of Results for Converging Shock Wave Problems

The purpose of this note is to provide a brief overview of an attempt to simulate a Guderley-like imploding shock problem in xRAGE. Prior to this study, the following have been completed with respect to the Guderley problem itself: A canonical definition for use with ASC codes has been proposed (LA-UR-11-02765); 1D simulations with spherical symmetry on uniform grids have been performed, with associated spatial convergence analysis. The xRAGE code showed near 1st order convergence for this problem in the initial studies (LA-UR-10-06435); A version of the Guderley problem as been added to the xRAGE regression test suite. More recently, a large number of simulations have been performed for the Guderley problem (and a similar problem) using xRAGE’s (v. 1009.02) AMR capability.

Mathematics & Computing(97)↗
DOE OSTI2025

Fluoride-Cooled High-Temperature Pebble-Bed Reactor Reference Plant Model Updates

This work presents the latest improvements to, and investigations performed with, the Fluoride-Cooled High-Temperature Pebble-Bed Reactor reference plant models for the United States Nuclear Regulatory Commission. These models, developed with the Comprehensive Reactor Analysis Bundle, or BlueCRAB, serve as the foundation for the future development of detailed design evaluation models based on license applications. BlueCRAB is the code suite proposed for non-light-water reactor systems safety analyses, and it incorporates various simulation tools developed by the Nuclear Energy Advance Modeling and Simulation program, including the Griffin code for reactor physics, the Pronghorn and SAM codes for core thermal fluids, the BISON code for solid conduction and fuel performance, and the SAM code for system analysis. The primary objective of this work is to assess the level of readiness of BlueCRAB for modeling fluoride-cooled high-temperature pebble-bed reactors. We first developed numerical models in BlueCRAB that include the key physics for this technology, ensuring an adequate level of fidelity for modeling the core performance during accident scenarios. This was followed by simulation of transient scenarios, two loss-of-forced-cooling events (one protected and one unprotected), and two control rod withdrawal events (one delayed and one prompt supercritical reactivity insertion). The analysis includes comparisons between the 2-D thermal fluid porous media models in Pronghorn and SAM, comparisons between coupled Pronghorn-Griffin and coupled SAM-Griffin models for two loss-of-forced cooling events and one control rod withdrawal event, and comparisons between SAM single-solve and domain-overlapping approaches for multi-scale thermal fluid coupling. In addition, we performed comparisons between 3-D, 2-D, and 0-D neutronic models for the two control rod withdrawal scenarios with Pronghorn-Griffin. The results show that the BlueCRAB models led to physically intuitive solutions for the scenarios examined. The changes in the various scalar and vector fields such as the neutron flux, power, temperatures, densities, pressures, and velocities are all within the expected ranges, and their distributions can be explained from the system response of the transients and the geometric and material variations. Several comparisons suggest that the porous media models in Pronghorn and SAM can lead to similar solutions, even though they are based on different methodologies. The simulations demonstrate that there are differences between the various levels of fidelity, and it is advisable to have flexible tools that can cover the breadth and depth of needs that may arise in future technical evaluations. We believe that BlueCRAB’s capabilities represent a significant asset for confirmatory analyses aimed at resolving important safety questions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

RAVEN Beta Release

This documents the release of the Risk Analysis Virtual Environment (RAVEN) code. A description of the RAVEN code is provided, and discussion of the release process for the M2LW-16IN0704045 milestone. The RAVEN code is a generic software framework to perform parametric and probabilistic analysis based on the response of complex system codes. RAVEN is capable of investigating the system response as well as the input space using Monte Carlo, Grid, or Latin Hyper Cube sampling schemes, but its strength is focused toward system feature discovery, such as limit surfaces, separating regions of the input space leading to system failure, using dynamic supervised learning techniques. RAVEN has now increased in maturity enough for the Beta 1.0 release.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

First use of three-dimensional equilibrium, stability, and transport calculations for interpretation of ELM-triggering with magnetic perturbations in NSTX

The application of non-axisymmetric magnetic perturbations has been demonstrated to destabilize edge-localized modes (ELMs) in the National Spherical Torus Experiment. A model 3D equilibrium has been calculated for these experiments using the VMEC code, which assumes nested flux surfaces and therefore that resonant perturbations are shielded. First, a free-boundary equilibrium is calculated using the NSTX coil set, with pressure and current profiles matched to a standard 2D reconstruction, but with up-down symmetry enforced. A new equilibrium is then calculated with the n = 3 field applied at a level consistent with experiment. This equilibrium is then used as the basis of further calculations using codes developed for analysis of stellarator plasmas. The neoclassical transport due to the 3D fields is calculated to be small compared with the experimental transport rates. Initial stability analysis has been performed, and indicates a modest degradation in ballooning stability with 3D fields applied. A new 3D equilibrium is also calculated using the SIESTA code, which allows for the formation of islands and stochastic regions. A comparison of the field structure between the SIESTA calculation and the assumption of fully penetrated vacuum perturbation indicates smaller island sizes and very small stochastic transport in the SIESTA case.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2012

Uncertainties propagation in the framework of a Rod Ejection Accident modeling based on a multi-physics approach

The control of uncertainties in the field of reactor physics and their propagation in best-estimate modeling are a major issue in safety analysis. In this framework, the CEA develops a methodology to perform multi-physics simulations including uncertainties analysis. The present paper aims to present and apply this methodology for the analysis of an accidental situation such as REA (Rod Ejection Accident). This accident is characterized by a strong interaction between the different areas of the reactor physics (neutronic, fuel thermal and thermal hydraulic). The modeling is performed with CRONOS2 code. The uncertainties analysis has been conducted with the URANIE platform developed by the CEA: For each identified response from the modeling (output) and considering a set of key parameters with their uncertainties (input), a surrogate model in the form of a neural network has been produced. The set of neural networks is then used to carry out a sensitivity analysis which consists on a global variance analysis with the determination of the Sobol indices for all responses. The sensitivity indices are obtained for the input parameters by an approach based on the use of polynomial chaos. The present exercise helped to develop a methodological flow scheme, to consolidate the use of URANIE tool in the framework of parallel calculations. Finally, the use of polynomial chaos allowed computing high order sensitivity indices and thus highlighting and classifying the influence of identified uncertainties on each response of the analysis (single and interaction effects). (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

PCI margin analysis methodology using the FALCON fuel performance code

ANATECH Corp., a wholly owned subsidiary of Structural Integrity Associates, Inc. has been at the forefront of PCI (pellet-cladding interaction) margin analyses as principal author of the EPRI PCI Guidelines documents and through the performance of numerous PCI margin evaluations for light-water reactors. In this paper, ANATECH first provides a discussion of specific features of the methodology developed over a number of years of application including a) selection of limiting assemblies for the detailed FALCON full length fuel rod (i.e., R-Z model) and PCI (i.e., R-θ model) evaluations, b) development of FALCON input characteristics including high fidelity rod-by-rod power histories, and c) application of failure criteria and remedies to the core designer in the case where the criteria are exceeded. Second, ANATECH provides guidance based on evaluations performed to-date. Analysis results are presented to show critical aspects of the evaluations and how they are used to provide insights to inform and improve commercial plant operation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗