Search NASA⌕ Search

SEARCH · Search NASA

Results for “analysis codes”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 667 records · Page 37
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

WAVEMOTH-FAST SPHERICAL HARMONIC TRANSFORMS BY BUTTERFLY MATRIX COMPRESSION

We present Wavemoth, an experimental open source code for computing scalar spherical harmonic transforms (SHTs). Such transforms are ubiquitous in astronomical data analysis. Our code performs substantially better than existing publicly available codes owing to improvements on two fronts. First, the computational core is made more efficient by using small amounts of pre-computed data, as well as paying attention to CPU instruction pipelining and cache usage. Second, Wavemoth makes use of a fast and numerically stable algorithm based on compressing a set of linear operators in a pre-computation step. The resulting SHT scales as O(L{sup 2}log{sup 2} L) for the resolution range of practical interest, where L denotes the spherical harmonic truncation degree. For low- and medium-range resolutions, Wavemoth tends to be twice as fast as libpsht, which is the current state-of-the-art implementation for the HEALPix grid. At the resolution of the Planck experiment, L {approx} 4000, Wavemoth is between three and six times faster than libpsht, depending on the computer architecture and the required precision. Because of the experimental nature of the project, only spherical harmonic synthesis is currently supported, although adding support for spherical harmonic analysis should be trivial.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2017

Fast Reactor Core Analysis Using AZNHEX Neutron Diffusion Code

The AZTLAN platform project emerged as an initiative to rank Mexico at a competitive level in software aspects for the analysis of nuclear reactors, in an effort to achieve autonomy in this area. The AZTLAN platform is conformed by a thermo-hydraulic code (AZTHECA), a neutron transport code (AZTRAN) and two neutron diffusion codes (AZKIND, AZNHEX). AZNHEX is a 3D diffusion code that solves numerically the time dependent neutron diffusion equations in hexagonal z-geometry using the RTN0 nodal finite element method. With this code, it is possible to simulate and analyze light water reactors, such as the Russian VVER, and variety of fast reactors. In this work, a sodium-cooled fast reactor core in steady state, with control rods fully withdrawn, is simulated. The comparison and validation of the values obtained with AZNHEX was done with the results obtained with MCNP6 Monte Carlo code.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Analysis of a nuclear data experiment: MANTRA-3

To improve the accuracy on the nuclear data, the uncertainties should be kept as low as possible. The major portion of these uncertainties comes from the input data and the assumptions made in the simulation models. For a high profile project like MANTRA, the pre and post irradiation analysis is very important. The current paper discusses the pre and post irradiation analysis of MANTRA-3 that could affect the end results using MCNP5 (a stochastic code) and ECCO-ERANOS (a deterministic code). This analysis was very helpful for understanding the possible uncertainties in MANTRA-3 and for calculating the final results. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

tonalli : an asexual genetic code to characterize APOGEE-2 stellar spectra. I. Validation with synthetic and solar spectra

ABSTRACT We present tonalli, a spectroscopic analysis python code that efficiently predicts effective temperature, stellar surface gravity, metallicity, $\alpha$-element abundance, and rotational and radial velocities for stars with effective temperatures between 3200 and 6250 K, observed with the Apache Point Observatory Galactic Evolution Experiment 2 (APOGEE-2). tonalli implements an asexual genetic algorithm to optimize the finding of the best comparison between a target spectrum and the continuum-normalized synthetic spectra library from the Model Atmospheres with a Radiative and Convective Scheme (MARCS), which is interpolated in each generation. Using simulated observed spectra and the APOGEE-2 solar spectrum of Vesta, we study the performance, limitations, accuracy, and precision of our tool. Finally, a Monte Carlo realization was implemented to estimate the uncertainties of each derived stellar parameter.

Adame, Lucía (ORCID:0000000263286099)↗
DOE OSTI2018

Verification and Validation of the PLTEMP/ANL Code for Thermal-Hydraulic Analysis of Experimental and Test Reactors

This report documents the verification and validation (V&V) of the PLTEMP/ANL Version 4.3 code. A brief description of the code and its history are given in Section 1.1. Section 1.2 provides an overview of PLTEMP/ANL verification and validation work done to date. Section 1.3 lists some publications by other authors on PLTEMP/ANL verification and validation. Section 1.4 describes a UNIX script developed to routinely verify the code after a code modification. The V&V work documented in this report was performed over several years, as code development progressed. In each verification/validation testing discussion, the version of PLTEMP/ANL which was initially used for the V&V work is referenced, but all of the testing presented in this report is applicable to current release version 4.3 of the code.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Verification and Validation of the PLTEMP/ANL Code for Thermal-Hydraulic Analysis of Experimental and Test Reactors, Volume 2

This is Volume 2 of a two-volume document that collects the verification and validation (V&V) works done for the PLTEMP/ANL code during the years of its development and improvement. Volume 1 describes the V&V of sixteen capabilities of the PLTEMP/ANL code that were identified by research reactor analysts as frequently used in their thermal-hydraulic analysis. Volume 2 describes the V&V of developments and improvements since the release of PLTEMP/ANL Version 4.3. Each chapter of the document focuses on verifying or validating a specific part of the software that calculates a particular phenomenon, e.g., channel flow calculation, coolant property calculation, heat transfer calculation, and flow instability calculation. Software verification is performed by comparing the code with a hand calculation, Microsoft spreadsheet calculation, Mathematica calculation, or MATLAB calculation. The software validation is achieved by comparing the code with experimental data or a widely tested code like the RELAP5 code. In addition, some PLTEMP/ANL V&V works that are available in the open literature are simply referenced in Volume 1 of the document. PLTEMP/ANL has been used in safety analysis reports of several US and foreign research reactors licensed and converted from highly enriched uranium fuel to low-enriched uranium fuel. A list of such reactors is given in Volume 1 of the document.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Verification and Validation of the PLTEMP/ANL Code for Thermal-Hydraulic Analysis of Experimental and Test Reactors, Volume 1

This is Volume 1 of a two-volume document that collects the verification and validation (V&V) works done for the PLTEMP/ANL code during the years of its development and improvement. Volume 1 describes the V&V of sixteen capabilities of the PLTEMP/ANL code that were identified by research reactor analysts as frequently used in their thermal-hydraulic analysis. Volume 2 describes the V&V of developments and improvements since the release of PLTEMP/ANL Version 4.3. Each chapter of the document focuses on verifying or validating a specific part of the software that calculates a specific phenomenon, e.g., channel flow calculation, coolant property calculation, heat transfer calculation, and flow instability calculation. Software verification is performed by comparing the code with hand calculation, Microsoft spreadsheet calculation, Mathematica calculation, or MATLAB calculation. The software validation is done by comparing the code with experimental data or a widely tested code like the RELAP5 code. In addition, some PLTEMP/ANL V&V works that are available in the open literature are simply cited in Volume 1 of the document. PLTEMP/ANL has been used in the safety analysis reports of several US and foreign research reactors licensed and converted from highly enriched uranium fuel to low-enriched uranium fuel. A list of such reactors is given in Volume 1 of the document.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

An integrated coupling model for solving multiscale fluid-fluid coupling problems in SAM code

In this study, an integrated coupling method has been developed for solving multiscale fluid-fluid coupling problems in plant-scale safety analysis models in SAM (System Analysis Module) code. In this method, a higher-fidelity multi-dimensional (3D) flow module is used for reactor components of complex flow features (e.g., reactor core) and a lumped parameter one-dimensional (1D) flow module for plant-scale flow loops (e.g., primary loop pipe network), respectively. In this method, the 3D fluid equation/domain and 1D fluid equation/domain are tightly coupled at the residual level and solved simultaneously using the Newton’s method to overcome the convergence issues typically seen in existing approaches like separate domain approach, where the 3D fluid equation and 1D fluid equation are solved separately. Extensive and successful code verifications and demonstrations have been performed for this newly developed method. This new modeling approach significantly simplify the work flow in developing high-fidelity plant-scale safety analysis model, e.g. for pool-type reactors and pebble-bed reactors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Characteristics of Axial Power Shape and Flow Field at Low Flow Condition

The representative event which experiences the low flow condition is SLB (Steam Line Break) event. During the SLB event with LOOP (Loss of Offsite Power), the reactor core is tripped and all RCPs (Reactor Coolant Pump) are stopped. The biggest worth control rod also stuck out. As a result, the flow rate and pressure in reactor core decrease and the peaking factor is relatively high compared to a nominal operating condition. In this paper, we study the tendency of the core axial power distribution and flow field with respect to the core average power at the SLB accident condition. Generally, the condition has been analyzed using the neutronics code with one dimensional flow result of the assemblies. However, if the core average power increases, the coolant density in the hot channel decreases and mass flux increases. The assumption that the mass flux is constant at the one dimensional analysis is incorrect. Hence, it is needed to couple the neutronics code and the subchannel analysis code in order to give feedback to neutronics code with the 3-D flow information. In this study, we coupled the nuclear design code and the subchannel analysis code and showed the characteristics of the axial power distribution and the flow field at the low flow condition. Since there is no experimental data at the condition, the characteristics of the power distribution and flow field were analyzed to understand the physical phenomena. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

What, why and when to go virtual: An international analysis of early adopters of virtual building energy codes inspections

To meet greenhouse gas reduction targets, several countries are pursuing more ambitious policies in their buildings and construction sectors, such as introducing zero net energy/carbon building codes. Countries often report not having enough qualified staff for performing building energy code inspections and many are exploring faster, easier, and more reliable methods to check the compliance of buildings with their codes. Building inspections are a critical element for ensuring code compliance and they have traditionally been performed in person. However, in-person inspections can be labor and travel intensive, costly, and prone to human error. In this paper, the authors explore how virtual inspections, particularly in light of the recent COVID-19 pandemic, have impacted processes for building code compliance checks in jurisdictions and communities around the world. Here, the authors collected data on four key parameters (time and financial savings, scope of inspections, changing practices and technological innovation, and benefits to consumers) from six jurisdictions and communities in five countries (Australia, Canada, Singapore, United Arab Emirates, and the United States) to analyze the impacts of virtual inspections on code compliance checks. The analysis found the greatest value from virtual inspections in geographically dispersed regions and for cities experiencing rapid building construction. The study also explored emerging technologies that are being piloted for virtual inspections. Although many of these technologies hold promise, more resources and capacity are needed to make them viable for use in building energy code inspections.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2018

Snakes on a Spaceship—An Overview of Python in Heliophysics

Computational analysis has become ubiquitous within the heliophysics community. However, community standards for peer review of codes and analysis have lagged behind these developments. This absence has contributed to the reproducibility crisis, where inadequate analysis descriptions and loss of scientific data have made scientific studies difficult or impossible to replicate. The heliophysics community has responded to this challenge by expressing a desire for a more open, collaborative set of analysis tools. Here, this article summarizes the current state of these efforts and presents an overview of many of the existing Python heliophysics tools. It also outlines the challenges facing community members who are working toward the goal of an open, collaborative, Python heliophysics toolkit and presents guidelines that can ease the transition from individualistic data analysis practices to an accountable, communalistic environment.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2014

An Overview of the XGAM Code and Related Software for Gamma-ray Analysis

The XGAM spectrum-fitting code and associated software were developed specifically to analyze the complex gamma-ray spectra that can result from neutron-induced reactions. The XGAM code is designed to fit a spectrum over the entire available gamma-ray energy range as a single entity, in contrast to the more traditional piecewise approaches. This global-fit philosophy enforces background continuity as well as consistency between local and global behavior throughout the spectrum, and in a natural way. This report presents XGAM and the suite of programs built around it with an emphasis on how they fit into an overall analysis methodology for complex gamma-ray data. An application to the analysis of time-dependent delayed gamma-ray yields from 235 U fission is shown in order to showcase the codes and how they interact.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Experimentally Validated Computational Modeling of Creep and Creep-Cracking for Nuclear Concrete Structures

In a Nuclear Power Plant, one of the most important components is the concrete nuclear reactor cavity, which serves both a structural and protective function as the biological radiation shield. Given that creep has been identified as a major knowledge gap in the assessment of nuclear structures (NUREG/CR-7153), this work helps to further the understanding of creep behavior of massive concrete containment structures for decades to enable safe and long-term operation of these facilities. This project has developed a robust, experimentally validated model to predict creep in nuclear concrete structures for up 60 years using short-term creep data thereby enabling a longer service life of critical facilities and early detection of structural failure. The work presented in this report is a pairing of computational and experimental methods. For the first time, the time temperature superposition (TTS) principle was successfully used to generate a uniaxial creep compliance master curve to predict mortar creep response for up to 22,500 days (nearly 60 years) at a reference temperature of 20°C. These data were used as input into finite element analysis (FEA) codes that use highly realistic random, 3D concrete microstructures from reconstructed coarse limestone aggregates. Finite element analysis performed provides the ability to quickly upscale mortar viscoelastic behavior to long-term concrete creep/relaxation data. A master creep compliance curve, constructed from the TTS principle, spanning 27 years, was used to validate two and a half decades of simulated concrete creep. Concurrently, three different simulated wall specimens were designed to mimic the behavior of post-tensioned concrete nuclear containment facility vessel walls over time as a result of concrete creep. The specimens were designed with different thicknesses, transverse and longitudinal reinforcement ratios, and level of post-tensioning stress. Each specimen contained various instrumentation to measure internal concrete temperature, concrete strain, and post-tensioning strain hourly for over 3 years. The concrete creep model developed in this project, based on the FEA concrete simulations, was applied to simulate the structural-scale experiments of prestressed concrete walls conducted in this project using the Grizzly code. These models can represent the effects of reinforcing and prestressing. Although there are some discrepancies with the experimental data, the model can predict the overall trends of the creep response in these experiments. One of these experimental models was also applied to an extended time to demonstrate how the findings from this study can be used to predict the behavior of actual structures of interest that have been in service for extended periods of time.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Managing genomic variant calling workflows with Swift/T

Bioinformatics research is frequently performed using complex workflows with multiple steps, fans, merges, and conditionals. This complexity makes management of the workflow difficult on a computer cluster, especially when running in parallel on large batches of data: hundreds or thousands of samples at a time. Scientific workflow management systems could help with that. Many are now being proposed, but is there yet the “best” workflow management system for bioinformatics? Such a system would need to satisfy numerous, sometimes conflicting requirements: from ease of use, to seamless deployment at peta- and exa-scale, and portability to the cloud. We evaluated Swift/T as a candidate for such role by implementing a primary genomic variant calling workflow in the Swift/T language, focusing on workflow management, performance and scalability issues that arise from production-grade big data genomic analyses. In the process we introduced novel features into the language, which are now part of its open repository. Additionally, we formalized a set of design criteria for quality, robust, maintainable workflows that must function at-scale in a production setting, such as a large genomic sequencing facility or a major hospital system. The use of Swift/T conveys two key advantages. (1) It operates transparently in multiple cluster scheduling environments (PBS Torque, SLURM, Cray aprun environment, etc.), thus a single workflow is trivially portable across numerous clusters. (2) The leaf functions of Swift/T permit developers to easily swap executables in and out of the workflow, which makes it easy to maintain and to request resources optimal for each stage of the pipeline. While Swift/T’s data-level parallelism eliminates the need to code parallel analysis of multiple samples, it does make debugging more difficult, as is common for implicitly parallel code. Nonetheless, the language gives users a powerful and portable way to scale up analyses in many computing architectures.

Science & Technology - Other Topics↗
DOE OSTI2012

Modern Nuclear Data Evaluation with the TALYS Code System

This paper presents a general overview of nuclear data evaluation and its applications as developed at NRG, Petten. Based on concepts such as robustness, reproducibility and automation, modern calculation tools are exploited to produce original nuclear data libraries that meet the current demands on quality and completeness. This requires a system which comprises differential measurements, theory development, nuclear model codes, resonance analysis, evaluation, ENDF formatting, data processing and integral validation in one integrated approach. Software, built around the TALYS code, will be presented in which all these essential nuclear data components are seamlessly integrated. Besides the quality of the basic data and its extensive format testing, a second goal lies in the diversity of processing for different type of users. The implications of this scheme are unprecedented. The most important are: 1. Complete ENDF-6 nuclear data files, in the form of the TENDL library, including covariance matrices, for many isotopes, particles, energies, reaction channels and derived quantities. All isotopic data files are mutually consistent and are supposed to rival those of the major world libraries. 2. More exact uncertainty propagation from basic nuclear physics to applied (reactor) calculations based on a Monte Carlo approach: 'Total' Monte Carlo (TMC), using random nuclear data libraries. 3. Automatic optimization in the form of systematic feedback from integral measurements back to the basic data. This method of work also opens a new way of approaching the analysis of nuclear applications, with consequences in both applied nuclear physics and safety of nuclear installations, and several examples are given here. This applied experience and feedback is integrated in a final step to improve the quality of the nuclear data, to change the users vision and finally to orchestrate their integration into simulation codes.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2012

Radial profile measurement of electron temperature in edge stochastic magnetic field layer of LHD using intensity ratio of extreme ultraviolet line emissions

Vertical profile of neon line emissions in 30-650 A wavelength range has been observed in horizontally elongated plasma cross section of Large Helical Device (LHD). Intensity ratio between the neon line emissions is studied to measure the radial profile of electron temperature in the edge stochastic magnetic field layer of LHD. The edge temperature profile successfully obtained from the line ratio of NeVIII 2s-3p to 2p-3s transitions is compared with the simulation based on three-dimensional edge transport code. The result shows a reasonably good agreement with the edge temperature profile analyzed from atomic data and analysis structure code. The electron temperature at last closed flux surface measured from the intensity ratio is also in good agreement with that measured from Thomson scattering.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

Integrated FHR technology development: Tritium management, materials testing, salt chemistry control, thermal hydraulics and neutronics, associated benchmarking and commercial basis

This report summarizes the results of a 3-year Integrated Research Project (IRP) sponsored by the U.S. Department of Energy to address major technical challenges in the development of the Fluoride-salt-cooled High-temperature Reactor (FHR). The IRP universities included the Massachusetts Institute of Technology (MIT), the University of California at Berkeley (UCB), the University of Wisconsin (UW), and the University of New Mexico (UNM). The FHR is a new reactor concept less than 20 years old that combines (1) a clean fluoride salt coolant, (2) the graphite-matrix coated-particle fuel originally developed for High- Temperature Gas-Cooled Reactors and (3) passive decay heat removal systems from sodium fast reactors. The base-line design is a pebble-bed FHR that delivers heat between 600 to 700°C with a 7 Li 2 BeF 4 (flibe) salt coolant to the power cycle or industrial users. The Executive Summary describes what was accomplished and summarizes major conclusions. The main report contains stand-alone sections in each major area of work. Highlights are summarized below. Tritium. The coolant salt produces tritium necessitating understanding and control of tritium. Experimental work included measuring hydrogen uptake on carbon, tritium uptake during irradiation of graphite in 700°C salt and the permeability of tritium through metals with and without coatings. An FHR systems model was developed to predict tritium behavior and enable design of tritium control systems. The leading candidate for tritium removal from the flibe salt is carbon bed outside the reactor core. Thermal hydraulics and Neutronics. A 10-meter high FHR integral effects test facility, using Dowtherm A®, has been built and used to simulate the integral response of a representative FHR to improve understanding of FHR thermal hydraulics and validate thermal hydraulics codes used for the FHR. Neutronic analysis included code-to-code benchmarking, multiphysics modeling, uncertainty analysis for nuclear data uncertainties and shielding analysis. Twisted-tube heat exchangers were examined that may provide significant cost reductions because of their improved heat-transfer performance with liquid salts. The impacts on safety and design caused by uncertainties in physical properties of the salt were assessed with recommendations on where added physical property measurements are needed. Work began on understanding radiative heat transfer that becomes important in salt systems at higher temperatures. Materials. Static corrosion tests were conducted on graphite and materials in 700°C salt -- both in the UW laboratory and in the MIT reactor. In addition to static corrosion testing, a circulating salt corrosion loop was built and is operating. Methods to measure salt properties including redox were developed. The presence of carbon increases corrosion. With good redox control, the evidence indicates stainless steel can be used as a material of construction. Benchmarking. An additional series of workshops where held to enable benchmarking and comparison of results between different investigators. Commercialization. The FHR delivers higher average-temperature (~650°C) heat to the power cycle or industry than other reactors. This enables power cycles with lower-cost heat storage for variable electricity to the grid while the reactor operates at base-load -- providing a major competitive advantage by maximizing revenue from electricity sales. In October, 2016, a new startup company named Kairos Power was formed, and opened its first office in Jack London Square, Oakland, California, in January 2017. Kairos Power exists largely due to the work that was accomplished by this IRP and the previous DOE-funded FHR IRP that performed research on FHR technology from 2012 to 2015. In September, 2018, as this IRP is concluding, Kairos Power was staffed at 60 full time employees, working to develop technology derived from this IRP. Many of these employees worked with the IRP before joining Kairos Power 1 (https://kairospower.com). Other students and researchers from this IRP have joined other startup companies, including the Terrapower Molten Chloride Fast Reactor and the Terrestrial Energy development efforts.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Sensitivity analysis of small modular reactor mPower with ATHLET

Small modular reactor (SMR) possesses the characteristics of less radioactive source term, simpler structure, modular design and features like passive safety as well as low risk of proliferation. Developed and designed by Babcock and Wilcox, the mPower reactor is one of pronouncing small modular reactors that adopts the integrated design of the pressurized water reactor, steam generator and a built-in internal control rod drive mechanism. In this paper, the code ATHLET (Analysis of fossil-hydraulics of Leaks and Transients) is used to model the mPower reactor. At the same time, the sensitivity analysis of the core structure and geometric parameters of the pipeline loop, as well as the steady-state simulation and post-exothermic simulation are carried out depending on the given scenarios. In addition, the sensitivity analysis of pressure, temperature and mass flow in the secondary loop is performed. The results of this study will mitigate some of the frequent misunderstanding in the field and will help to better understand the design of SMR and provide basic research data for further research. However, there are limitations in the current research. Considering that the mPower is still in the conceptual design, the established model is only based on the limited public data while detailed parameters are still in myth, which leads to certain level of uncertainties in the calculation. Besides, the research is mainly on the influence of sensitivity analysis under steady state conditions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗