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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 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

STM/S Grid LDOS Data and Analysis Code for Deciphering Majorana Zero Modes in Topological Superconductor

This dataset provides raw millikelvin scanning tunneling microscopy/spectroscopy (STM/S) grid spectroscopy data and Python analysis scripts supporting the manuscript “Deciphering Majorana Zero Modes in Topological Superconductor FeTe0.55Se0.45 with Machine-Learning-Assisted Spectral Deconvolution.” The dataset includes a raw grid spectroscopy file acquired on FeTe0.55Se0.45 at 40 mK under magnetic field, together with Python/Jupytext analysis scripts used for STM/S data processing, visualization, spectral deconvolution, Lorentzian peak fitting, feature extraction, machine-learning-assisted clustering, and figure generation. These files support the analysis of vortex-core local density of states and the identification of zero-bias-peak-related spectral components from complex in-gap states. The dataset is intended to provide a citable archival record of the data and analysis code associated with the published manuscript and to support transparency and reproducibility of the reported STM/S and machine-learning workflow.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2017

Frequency spectrum analysis code for nuclear noise based on improved wavelet algorithm - 337

The core of the narrow flow channel generally faced the problem of local bubble boiling with the temperature rise of the nuclear fuel element, it is necessary and meaningful to answer and deal with the problem on the technical level. Reactor noise analysis begins with the study of the physical characteristics of the core, after nearly 30 years of research and development it has become a powerful tool in the auxiliary monitoring of the power reactor operating. By using this noise frequency spectrum analysis technology, the reactor dynamic parameter such as prompt neutron decay constant can be measured. Also, through such system, the frequency characteristics of reactor can be real-time monitored, which is helpful for safety operation and fault diagnosis. Spectrum analysis technology in the Fourier analysis era is now barely available, the reason is that the local spectral characteristics of this technology are not obvious, but this defect can be greatly improved by wavelet analysis technique, however, the traditional wavelet analysis will introduce the problem of spectral confusion. To solve the problem, based on the improved wavelet algorithm, frequency spectrum analysis code for nuclear noise is developed,from the perspective of fine spectral analysis, the local boiling accident of the coolant can be prejudged and avoid. Based on the analysis and test of experimental data,The results show that the calculation ability of the code is in accordance with the expected, it can be applied in the operation of research reactor. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

COBRA-SFS Version 6.2.1 User Guide: A Thermal Hydraulic Analysis Code for Spent Fuel Storage and Transportation Casks

This document is an updated and reduced version of the complete documentation of the COBRA-SFS thermal analysis code (COolant Boiling in Rod Arrays – Spent Fuel Storage) (Michener et al. 1995). This manual provides the analyst with input instructions and fundamental guidelines for contrasting models using COBRA-SFS. COBRA-SFS is designed to analyze the dry storage and transportation casks that house spent or used nuclear fuel. The primary focus and application is on square light water reactor fuel assemblies, however the code also has the capability to model hexagonal assemblies and is will suited to any geometry that can be defined as an extruded 2-D map.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Updated Economic Model for Estimation of GDP Losses in the MACCS Offsite Consequence Analysis Code RDEIM Model Report for MACCS v4.2

This report updates the Regional Disruption Economic Impact Model (RDEIM) GDP-based model described in Bixler et al. (2020) used in the MACCS accident consequence analysis code. MACCS is the U.S. Nuclear Regulatory Commission (NRC) used to perform probabilistic health and economic consequence assessments for atmospheric releases of radionuclides. It is also used by international organizations, both reactor owners and regulators. It is intended and most commonly used for hypothetical accidents that could potentially occur in the future rather than to evaluate past accidents or to provide emergency response during an ongoing accident. It is designed to support probabilistic risk and consequence analyses and is used by the NRC, U.S. nuclear licensees, the Department of Energy, and international vendors, licensees, and regulators. The update of the RDEIM model in version 4.2 expresses the national recovery calculation explicitly, rather than implicitly as in the previous version. The calculation of the total national GDP losses remains unchanged. However, anticipated gains from recovery are now allocated across all the GDP loss types – direct, indirect, and induced – whereas in version 4.1, all recovery gains were accounted for in the indirect loss type. To achieve this, we’ve introduced new methodology to streamline and simplify the calculation of all types of losses and recovery. In addition, RDEIM includes other kinds of losses, including tangible wealth. This includes loss of tangible assets (e.g., depreciation) and accident expenditures (e.g., decontamination). This document describes the updated RDEIM economic model and provides examples of loss and recovery calculation, results analysis, and presentation. Changes to the tangible cost calculation and accident expenditures are described in section 2.2. The updates to the RDEIM input-output (I-O) model are not expected to affect the final benchmark results Bixler et al. (2020), as the RDEIM calculation for the total national GDP losses remains unchanged. The reader is referred to the MACCS revision history for other cost modelling changes since version 4.0 that may affect the benchmark. RDEIM has its roots in a code developed by Sandia National Laboratories for the Department of Homeland Security to estimate short-term losses from natural and manmade accidents, called the Regional Economic Accounting analysis tool (REAcct). This model was adapted and modified for MACCS. It is based on I-O theory, which is widely used in economic modeling. It accounts for direct losses to a disrupted region affected by an accident, indirect losses to the national economy due to disruption of the supply chain, and induced losses from reduced spending by displaced workers. RDEIM differs from REAcct in in its treatment and estimation of indirect loss multipliers, elimination of double-counting associated with inter-industry trade in the affected area, and that it is intended to be used for extended periods that can occur from a major nuclear reactor accident, such as the one that occurred at the Fukushima Daiichi site in Japan. Most input-output models do not account for economic adaptation and recovery, and in this regard RDEIM differs from its parent, REAcct, because it allows for a user-definable national recovery period. Implementation of a recovery period was one of several recommendations made by an independent peer review panel to ensure that RDEIM is state-of-practice. For this and several other reasons, RDEIM differs from REAcct.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2024

Investigation of Numerical Methods for Performance Improvement of MOOSE-based System Analysis Codes

The main objective of this study is to investigate the feasibility of implementation of staggered-grid finite volume method (SG-FVM) using the MOOSE framework to support the development of the advanced system analysis code SAM. This study successfully demonstrated the integration of staggered-grid finite volume method in the application level using the MOOSE framework, although not in the framework level which should be investigated in the future. Several important properties of the implemented SG-FVM, e.g., high-order spatial accuracy and monotonicity preserving, have been demonstrated with selected numerical test cases. The superior performance in execution time was also evident based on a selected one- dimensional flow problem in a loop configuration.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2025

Systems and methods for binary code analysis

Human-readable (HR) code may be derived from a binary. The HR code may be configured to have statistical properties suitable for machine-learned (ML) translation. The HR code may comprise source code, intermediate code, assembly code, or the like. A machine-learned translator may be configured to translate the HR code into labels comprising semantic information pertaining to respective functions of the binary, such as a function name, role, or the like. Execution of the binary may be blocked in response to translating the HR code to a label associated with malware, such as cryptocurrency mining malware or the like. Conversely, the binary may be permitted to proceed to execution in response to determining that the translation is free from labels indicative of malware.

Anderson, Matthew W.↗
DOE OSTI2020

Razorback – A reactor transient analysis code for large rapid reactivity additions in a natural circulation research reactor

Research reactors play an important role in higher education, scientific research, and medical radioisotope production around the world. It is thus important to ensure the safety of facility workers and the public. This work presents a new reactor transient analysis code referred to as Razorback. The code has been developed for the evaluation of large rapid reactivity addition in research reactors. Its initial focus is the Annular Core Research Reactor (ACRR) at Sandia National Laboratories. Results have been validated against ACRR pulse operations. Razorback models the reactor kinetics, fuel element heat transfer, fuel element thermal expansion, and natural circulation coolant channel thermal-hydraulic response. Simulation results for ACRR pulse operations are shown to agree very well with operational data obtained from the ACRR. Finally, Razorback is expected to be a valuable tool for ACRR pulse performance prediction and ACRR reactor safety analyses.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Analysis of stimulated Raman backscatter and stimulated Brillouin backscatter in experiments performed on SG-III prototype facility with a spectral analysis code

Experiments about the observations of stimulated Raman backscatter (SRS) and stimulated Brillouin backscatter (SBS) in Hohlraum were performed on Shenguang-III (SG-III) prototype facility for the first time in 2011. In this paper, relevant experimental results are analyzed for the first time with a one-dimension spectral analysis code, which is developed to study the coexistent process of SRS and SBS in Hohlraum plasma condition. Spectral features of the backscattered light are discussed with different plasma parameters. In the case of empty Hohlraum experiments, simulation results indicate that SBS, which grows fast at the energy deposition region near the Hohlraum wall, is the dominant instability process. The time resolved spectra of SRS and SBS are numerically obtained, which agree with the experimental observations. For the gas-filled Hohlraum experiments, simulation results show that SBS grows fastest in Au plasma and amplifies convectively in C{sub 5}H{sub 12} gas, whereas SRS mainly grows in the high density region of the C{sub 5}H{sub 12} gas. Gain spectra and the spectra of backscattered light are simulated along the ray path, which clearly show the location where the intensity of scattered light with a certain wavelength increases. This work is helpful to comprehend the observed spectral features of SRS and SBS. The experiments and relevant analysis provide references for the ignition target design in future.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Dynamic event tree analysis with the SAS4A/SASSYS-1 safety analysis code

The consequences of a transient in an advanced sodium-cooled fast reactor are difficult to capture with the traditional approach to probabilistic risk assessment (PRA). Numerous safety-relevant systems are passive and may have operational states that cannot be represented by binary success or failure. In addition, the specific order and timing of events may be crucial which necessitates the use of dynamic PRA tools such as ADAPT. The modifications to the SAS4A/SASSYS-1 sodium-cooled fast reactor safety analysis code for linking it to ADAPT to perform a dynamic PRA are described. A test case is used to demonstrate the linking process and to illustrate the type of insights that may be gained with this process. Finally, newly-developed dynamic importance measures are used to assess the significance of reactor parameters/constituents on calculated consequences of initiating events.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Modification of the SAS4A Safety Analysis Code for Integration with the ADAPT Discrete Dynamic Event Tree Framework.

It is difficult to assess the consequences of a transient in a sodium-cooled fast reactor (SFR) using traditional probabilistic risk assessment (PRA) methods, as numerous safety-related sys- tems have passive characteristics. Often there is significant dependence on the value of con- tinuous stochastic parameters rather than binary success/failure determinations. One form of dynamic PRA uses a system simulator to represent the progression of a transient, tracking events through time in a discrete dynamic event tree (DDET). In order to function in a DDET environment, a simulator must have characteristics that make it amenable to changing physical parameters midway through the analysis. The SAS4A SFR system analysis code did not have these characteristics as received. This report describes the code modifications made to allow dynamic operation as well as the linking to a Sandia DDET driver code. A test case is briefly described to demonstrate the utility of the changes.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2024

Development and Integration of a Stochastic Clad Damage Propagation Model into PRONGHORN-SC Subchannel Analysis Code

The failure of fuel pins in nuclear reactors is intrinsically stochastic. Typically, a combination of variation in manufacturing that affects the material characteristics and the fuel assembly dimensions, variation in operating conditions, such as local power, coolant flow rate, and irradiation induced changes in material properties lead to a large uncertainty in failure margin of the fuel pins. Failure, therefore, may occur in exceptional pins with adverse combinations of these variations. Upon a metal fuel pin (U-Pu-Zr/HT9) failure, depressurization of the fuel pin takes place by release of fission gas, liquid sodium bond, and potentially solid fuel particles or molten/eutectic fuel droplets through the hole in cladding. The effect of a fission gas jet on neighbor fuel pins and possible propagation of a clad damage during normal operation was studied experimentally in 1970s and it was found that the post-failure fission gas jet insulates the jet impingement area of the target fuel pin surface and could increase the target pin’s surface temperature by as much as 100 – 200 K during the failed pin depressurization. It was concluded that the effect should not lead to fuel pin failure propagation during normal operation. In accident scenarios of sodium and lead fast reactors such as Unprotected Loss-Of-Flow (ULOF) or Unprotected Transient Over Power (UTOP), the fuel pins can be subjected to higher clad temperatures and fuel pin pressures or fuel clad mechanical/chemical interaction where thermal creep margin becomes significantly lower compared to the normal operation conditions. Therefore, possible stochastic failure and the post-failure fission gas/fuel jet impingement could be critical in order to predict fuel pin failure propagation. Pin depressurization due to fission gas release may degrade the heat transfer by formation of a gas blanket on a neighboring pin surface, which is a local phenomenon, and by causing coolant flow deceleration and starvation, which could affect a surrounding region as well. Furthermore, the potential presence of solid fuel particles or molten fuel at the time of clad failure could boost post-failure jet induced degradation even further. The present study models the U-Pu-Zr/HT9 metal fuel pin failure and stochastic clad damage propagation by biased sampling based on a Cumulative Damage Fraction (CDF) type clad failure criterion and the normal distribution of fuel failure probability density as a function of logarithm of Cumulative Damage Fraction. In addition, the effect of post-failure fission gas jet on heat transfer degradation is modeled for the target pins. This model is called stochastic Clad Damage Propagation (CDAP). The CDAP model is now fully integrated into developmental version of PRONGHORN-SC subchannel analysis code, allowing for modeling local failures and its propagation potential. Section 2 describes the components of the CDAP models. Section 3 describes the model implementation to PRONGHORN-SC and input specifications. Section 4 describes the CDAP model validation coupled to PRONGHORN-SC.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Demonstration of object location, classification, and characterization by developed deep learning dust ablation trail analysis code package using plasma jets

Based on deep learning, a Dust Ablation Trail Analysis (DATA) code package is developed to detect dust ablation trails in tokamaks, which is intended to analyze a large amount data of tokamak dusts. To validate and benchmark the DATA code package, 2440 plasma jet images are exploited for the training and test of the deep learning DATA code package, since plasma jets resemble the shape and size of dust ablation clouds in tokamaks. After being trained by 1920 plasma jet images, the DATA code package is able to locate 100% plasma jets, classify plasma jets with the accuracy of >99.9%, and output image skeleton information for classified plasma jets. The DATA code package trained by the plasma jet images is also used to analyze the dust ablation trails captured in the Experimental Advanced Superconducting (EAST) tokamak with the satisfactory performance, further verifying its applicability in the fusion dust ablation investigation. Based on its excellent performance presented here, it is demonstrated that our DATA code package is able to automatically identify and analyze dust ablation trails in tokamaks, which can be used for further detailed investigations, such as the three-dimensional reconstruction of dusts and their ablation trails.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

RAZORBACK - A Research Reactor Transient Analysis Code Version 1.0 - Volume 1: User's Manual

Razorback is a research reactor transient analysis computer code designed to simulate the operation of a research reactor (such as Sandia National Laboratories Annular Core Research Reactor (ACRR)). The code provides a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, the equation of motion for fuel element thermal expansion, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This input manual describes how an input file is composed, and facilitates an understanding of the various code input parameters. The makeup of the various code output files is also described. This manual also provides instructions for the installation and setup of the code, and how to report bugs and/or errors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

RAZORBACK - A Research Reactor Transient Analysis Code (Version 1.0 - Volume 2: Software Design and Development Report)

RAZORBACK is a research reactor transient analysis computer code designed to simulate the operation of a research reactor (such as Sandia National Laboratories' Annular Core Research Reactor (ACRR)). The code provides a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, the equation of motion for fuel element thermal expansion, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This Software Design and Development Report describes how the code is designed to implement the functional requirements established in the Software Requirements Specification document for RAZORBACK. The description presents the physical model equations used in the design, and how these equations are set up to be solved numerically. This report also provides an evaluation of how the software design meets each of the functional requirements.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Evaluation of Design & Analysis Code, CACTUS, for Predicting Crossflow Hydrokinetic Turbine Performance

CACTUS, developed by Sandia National Laboratories, is an open-source code for the design and analysis of wind and hydrokinetic turbines. While it has undergone extensive validation for both vertical axis and horizontal axis wind turbines, and it has been demonstrated to accurately predict the performance of horizontal (axial-flow) hydrokinetic turbines, its ability to predict the performance of crossflow hydrokinetic turbines has yet to be tested. The present study addresses this problem by comparing the predicted performance curves derived from CACTUS simulations of the U.S. Department of Energy’s 1:6 scale reference model crossflow turbine to those derived by experimental measurements in a tow tank using the same model turbine at the University of New Hampshire. It shows that CACTUS cannot accurately predict the performance of this crossflow turbine, raising concerns on its application to crossflow hydrokinetic turbines generally. The lack of quality data on NACA 0021 foil aerodynamic (hydrodynamic) characteristics over the wide range of angles of attack (AoA) and Reynolds numbers is identified as the main cause for poor model prediction. A comparison of several different NACA 0021 foil data sources, derived using both physical and numerical modeling experiments, indicates significant discrepancies at the high AoA experienced by foils on crossflow turbines. Users of CACTUS for crossflow hydrokinetic turbines are, therefore, advised to limit its application to higher tip speed ratios (lower AoA), and to carefully verify the reliability and accuracy of their foil data. Accurate empirical data on the aerodynamic characteristics of the foil is the greatest limitation to predicting performance for crossflow turbines with semi-empirical models like CACTUS. Future improvements of CACTUS for crossflow turbine performance prediction will require the development of accurate foil aerodynamic characteristic data sets within the appropriate ranges of Reynolds numbers and AoA.

13 HYDRO ENERGY↗
DOE OSTI2017

Status of the SAS4A/SASSYS-1 Safety Analysis Code

SAS4A/SASSYS-1 is a simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced reactors. This report summarizes the current status of SAS4A/SASSYS-1, including the implementation of software quality assurance procedures for configuration management, new licenses, current users, and a gap assessment of the test suite. The existing copyright has been extended for an additional five years, which was essential so that Argonne can continue developing, distributing, and supporting the software for both domestic use and international collaborations. SAS4A/SASSYS-1 Version 5.2 was completed in March 2017 and released to users in May. The importance of SAS4A/SASSYS-1 maintenance reflects its relevance to a number of U.S. Department of Energy programs as well as domestic and international collaborations. External collaborations have also produced improvements in the SAS4A/SASSYS-1 code, and these are noted in this report. SAS4A/SASSYS-1 has a growing user base that continues to strengthen the promotion of advanced non-LWR reactor concepts. Additional users will help solidify DOE’s leadership role in fast reactor safety both domestically and in international collaborations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

COBRA-SFS thermal-hydraulic analysis code for spent fuel storage and transportation casks: Models and methods

COBRA-SFS, a thermal-hydraulics code developed for steady-state and transient analysis of multi-assembly spent-fuel storage and transportation systems, has been incorporated into the Used Nuclear Fuel-Storage, Transportation and Disposal Analysis Resource and Data System tool as a module devoted to spent fuel package thermal analysis. This paper summarizes the basic formulation of the equations and models used in the COBRA-SFS code, showing that COBRA-SFS fully captures the important physical behavior governing the thermal performance of spent fuel storage systems, with internal and external natural convection flow patterns, and heat transfer by convection, conduction, and thermal radiation. Of particular significance is the capability for detailed thermal radiation modeling within the fuel rod array.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

The SAS4A/SASSYS-1 Version 5.8 Safety Analysis Code System

SAS4A/SASSYS-1 is a software simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced nuclear reactors. Detailed, mechanistic models of steady-state and transient thermal, hydraulic, kinetic, and mechanical phenomena are employed to describe the response of the reactor core, the reactor primary and secondary coolant loops, the reactor control and protection systems, and the balance-of-plant to accidents caused by changes in coolant flow, loss of heat rejection, or reactivity insertion. The consequences of single and double-fault accidents can be modeled, including fuel and coolant heating, fuel and cladding mechanical behavior, core reactivity feedbacks, coolant loop performance including natural circulation, and decay heat removal. Analyses are typically terminated upon demonstration of reactor and plant shutdown to permanently coolable conditions, or upon violation of design basis margins. The objective of the analysis is to quantify accident consequences as measured by the transient behavior of system performance parameters, such as fuel and cladding temperatures, reactivity, and cladding strain. Originally developed for analysis of sodium cooled reactors with oxide fuel clad by stainless steel, the models were subsequently extended and specialized to metallic fuel clad with advanced alloys and to several other coolant options, including lead, LBE, and water.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗