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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 577 records · Page 32
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

chimes_calculator

The Chebyshev Interaction Model for Efficient Simulation (ChIMES) is a machine-learned interatomic potential targeting chemistry in condensed phase systems. ChIMES models are able to approach quantum-accuracy through a systematically improvable explicitly many-bodied basis comprised of linear combinations of Chebyshev polynomials. Though originally developed as to enable description of organic molecular materials, ChIMES has successfully been applied to systems spanning ambient water to molten carbon, and leveraged as correction for density functional based tight binding simulations. The ChIMES calculator comprises a flexible toolset for evaluating ChIMES interactions (e.g. in simulations, single point calculations, etc). Users have the option of directly imbedding the ChIMES calculator within their codes, or evaluating interactions through the beginner-friendly serial interface, each of which have Python, C++, C, and Fortran API's.

Goldman, Nir↗
DOE OSTI2021

LCOC (Levelized Cost of Light-Duty EV Charging Calculator)

This code was developed to calculate the levelized cost of light-duty electric vehicle (EV) charging (LCOC) for the paper "Levelized Cost of Charging Electric Vehicles in the United States" by Borlaug et al.(https://doi.org/10.1016/j.joule.2020.05.013). The calculation takes inputs from a variety of publicly available data sources (e.g., utility tariffs from the Utility Rate Database; EV charging station locations from the Alternative Fuels Data Center; energy price forecasts from the Energy Information Administration) and produces the current LCOC, considering when, where, and how light-duty EVs are recharged. In addition, the lifetime fuel cost savings (LFCS) calculation estimates the discounted fuel cost savings for a new battery EV (BEV) or plug-in hybrid EV (PHEV) over a 15-year time horizon using state-level gasoline price and LCOC projections.

Borlaug, Brennan↗
DOE OSTI2021

SCC (Step characteristics calculator) [SWR-21-80]

The SCC software calculates the step response characteristics based on the data collected from the plant. The tools available to characterize the step response needs a plant model. Our tool calculates the characteristics based on the data captured from the plant. This prevents the user from building a plant model and directly calculates the response.

Prabakar, Kumaraguru↗
DOE OSTI2024

Plastics Environmental Risk Calculator

SF-24-074 The Plastics Environmental Risk Calculator (PERC) was developed in Microsoft Excel and estimates the environmental distribution and lifetime of new biobased and conventional plastics from commonly measured properties of the plastics. A Random Forest regression model embedded in the calculator calculates plastic degradation rates and lifetimes as a proxy for environmental risk. Model default assumptions may be overwritten by the user.

Beckman, Kevin [Argonne National Laboratory (ANL),↗
DOE OSTI2025

Calculation Of Neutrino Flux

The CONFLUX software framework, Calculation Of Neutrino FLUX, is built with the goal to simplify and standardize the calculation. CONFLUX packages three methods to calculate neutrinos generated from reactor neutrinos or individual beta decays with common nuclear data and beta theories for direct cross-method comparison. The software prepacked the latest nuclear database, including ENDF.B/VIII, JEFF 3.3, and ENSDF, as well as methods to process the uncertainties. It also allows customized nuclear data and beta theories and user generated time dependent reactor models for convenient adjustment of fission products, theoretical corrections.

Zhang, Xianyi [Lawrence Livermore National Laborat↗
DOE OSTI2022

Space-dependent calculation of the multiplicity moments for shells

In earlier work, we extended the methodology of multiplicity counting in nuclear safeguards, by elaborating the one-speed stochastic transport theory of the calculation of the so-called multiplicity moments, i.e. the factorial moments of the number of neutrons emitted from a fissile item, following a source event from an internal neutron source (spontaneous fission and (α, n) reactions). Calculations were made for spheres and cylinders of various shapes. In all our work so far, the material of the items was homogeneous, and the distribution of the internal source was assumed to be uniformly distributed within the item, with the neutron emission assumed to be isotropic. In the present work the calculations are extended to the case of a point source inside either a solid sphere or in a spherical shell. This necessitates the extension of the theory to non-homogeneous items and non-uniform and non-isotropic sources. This work describes the extension of the theory and provides some quantitative results. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Accelerated coupled Monte Carlo-Thermal hydraulic calculations using a hybrid GTF-diffusion-based prediction block: first results

Accurate predictions of spatial power and temperature distributions require the coupling of a neutron transport solver with a thermal-hydraulic (TH) feedback. Nowadays, Monte Carlo (MC) codes are widely coupled to TH solvers, typically via a Picard iteration (PI) method, due to the higher fidelity that such frameworks can produce. To speed up a PI, a prediction step can produce an improved initial guess for a source distribution and feed it to the MC code. Recent work investigated a prediction step that uses generalized transfer functions (GTFs) to predict the macroscopic cross sections' variations following a perturbation in TH properties, such as coolant density. The previous method also relied on first order perturbation (FOP) theory to predict perturbed power profiles, rather than using an expensive MC iterate. The implemented FOP method relied on generating a fission matrix from which the forward and adjoint Eigenmodes were extracted and later used to for power calculations. The generation of the fission matrix can introduce a significant computational overhead, therefore undermining the performance of the proposed hybrid technique when applied to high-dimensional problems, e.g., full core calculations. This work attempts to improve the GTF-FOP prediction step by replacing the FOP solver with a nodal diffusion solver, thus eliminating the need to calculate a fission matrix. The GTF-diffusion step was tested for various moderator density perturbations. In each case, the predicted power distribution showed good agreement with the reference case. The latter is attributed to the generally good prediction of most spatially distributed macroscopic cross sections, except the transport cross section, which will become the focus of future work. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Impact of Increased Latent Generations on Sensitivity Calculations with SCALE

Analyses of cross section sensitivity data from systems with fissile material allow analysts to associate an importance for each material, nuclide, reaction, and neutron energy by simulating real world criticality scenarios. Although criticality safety validation efforts can be guided by the cross-section sensitivity and uncertainty data generated for a particular system, these calculations can often be computationally expensive and sometimes cumbersome without proper guidance. The TSUNAMI suite within the SCALE code package has several methods for generating sensitivity data, including multigroup and continuous energy (CE) capabilities. The release of SCALE 6.3 has three different CE methods for generating cross section sensitivity data: (1) the Iterated Fission Probability (IFP) method with the KENO Monte Carlo transport solver, (2) the IFP method with the Shift Monte Carlo transport solver, and (3) the Contributon-Linked eigenvalue sensitivity/Uncertainty estimation via Tracklength importance CHaracterization (CLUTCH) method with the KENO Monte Carlo transport solver. Although the CLUTCH method has additional parameters for generating sensitivity data files relative to the IFP method, all three methods use latent generations, which are the generations between an event (i.e., fission) and the assessment of importance based on the asymptotic population of progeny neutrons. Increasing the number of latent generations in a calculation leads to increased discrimination of the sensitivity coefficients but at the cost of the increased uncertainty associated with those generated values. Analysts must balance the accuracy of the sensitivity calculations and its uncertainty with the associated computational cost involved in generating the values. This paper discusses the impact of adjusting the latent generation parameter for a range of sensitivity values and how these changes compare with the direct perturbation values obtained from a change of ±0.5% Δ k in both benchmark and safety application models. Two benchmarks from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the MPC-32 dual purpose canister for spent nuclear fuel are used for analysis.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Analysis of SCALE Criticality and Sensitivity Calculations for Reflected HEU Cylinders

The SCALE code package offers multiple nuclear data libraries and sensitivity and uncertainty (S/U) methods supporting and derived from Monte Carlo (MC) transport. The CSAS and TSUNAMI-3D sequences use KENO MC, utilizing either continuous-energy (CE) cross sections or multigroup (MG) cross section libraries. TSUNAMI-3D has two CE calculational methods: the iterated fission probability (IFP) method, and the Contribution-Linked eigenvalue sensitivity/Uncertainty estimation via Tracklength importance CHaracterization (CLUTCH) method. Previous work has shown poor agreement between CLUTCH and confirmatory direct perturbation calculations in specific applications (e.g., fissionable and polyethylene reflectors). The HEU-MET-FAST-084 (HMF-084) International Criticality Safety Benchmark Evaluation Project evaluation consists of 27 cylindrical highly enriched uranium metal cores with 14 unique reflector materials of 0.5 and 1 in. thicknesses. Included in this list of reflector materials are natural uranium and polyethylene. This work utilized SCALE 6.2.4 models of the HMF-084 evaluation, with additional non-physical configurations to test both the MG bias and CLUTCH functionality across a variety of reflector material thicknesses. The evaluation’s use of concentric cylinders allowed for examination of several MG self-shielding methods: infinite homogenous, cylindrical, and spherical. The results indicate that the use of polyethylene reflectors with CLUTCH is not fundamentally impossible but sensitive to geometry. The poor performance of CLUTCH with fissionable reflectors was reaffirmed. The 2 in. and greater polyethylene-reflected calculations demonstrate the necessity of using the 302-group library for fast systems. The nickel MG bias was substantial, as discussed in a companion paper, as were cobalt and iron.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Comparative study of decay heat calculations with FLUKA and MCNP/CINDER2008

In designing a high-power tungsten target, decay heat driven temperature rise in the spallation volume is a safety concern during maintenance and in loss of coolant accidents. As tungsten hydrates and becomes volatile in steam at above 800 ◦C, it is important to keep the target temperature below this threshold when active cooling is unavailable. Decay heat in a target is calculated with particle transport simulation codes combined with transmutation codes. The calculated decay heat usually differs depending on the nuclear cross sections and the decay particle transport models built in the code architecture. In this paper, we calculated decay heat of a water-cooled tungsten target using popular particle transport codes, FLUKA and MCNP6® paired with CINDER2008 and compared the results. The target-moderator-reflector (TMR) system is modeled with a water-cooled solid tungsten target, water premoderators, liquid hydrogen cold moderators and beryllium reflectors. Water-cooled stainless-steel shielding is modeled around the TMR system. The tungsten volume is clad with a thin layer of erosion/corrosion resistant material. This study provides information about the uncertainty range in decay heat prediction of high-power spallation targets for hazard analysis.

Lee, Yong Joong↗
DOE OSTI2021

ambrosia: An R package for calculating and analyzing food demand that is responsive to changing incomes and prices

The ambrosia R package was developed to calculate food demand for staples and non-staple commodities that is responsive to changing levels of incomes and prices. ambrosia implements the framework to quantify food demand as established by Edmonds et al. (Edmonds et al. 2017) and allows the user to explore and estimate different variables related to the food demand system. Currently ambrosia provides 3 main functions: (1) calculation of food demand for any given set of parameters including income levels and prices (2) estimation of parameters within a given a dataset. Note: ambrosia is used to calculate parameters for the food demand model implemented in the Global Change Analysis Model(GCAM) (3) exploration and preparation of raw data before starting a parameter estimation

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2020

Harnessing the power of ab initio calculations, distributed computing and machine learning to efficiently locate extreme molecules for use in carbon-based solar cells (Final Technical Report)

The use of high-throughput virtual screening (HTVS) tools is a powerful tool to expedite the materials discovery of commercially relevant materials. In previous years, our group has developed a molecular discovery platform to generate libraries in order to obtain suitable candidates for different applications, starting from the Harvard Clean Energy Project. This platform is suitable to test in-silico on traditional supercomputing clusters and shared resources, for example, in the IBM World Community Grid. In this project, we used the molecular discovery platform to create and screen a library of candidates of organic photovoltaic (OPVs) molecules. Based on a set of candidates created with combinations of molecular moieties, we were able to filter, by conformation stability, the energy of electronic orbitals and approximated power conversion efficiencies (PCE). To improve the predictions of orbital energies calculated and the PCEs, we used Gaussian Process regression and two sets of molecules. These sets correspond to electronic structure calculations of a higher level of theory and experimental PCE values, respectively. Finally, we selected a subset of the best candidates (molecules with a PCE higher than 10%) to understand its absorbance properties with TD-DFT. This project has demonstrated the capabilities of our molecular discovery platform for HTVS. Finally, machine learning can help us to introduce more complex effects included in bulk conditions and computational intensive calculations on models.

14 SOLAR ENERGY↗
DOE OSTI2020

Subcritical Multiplication Ex-Core Calculations with Shift

This report details the development and testing of Virtual Environment for Reactor Applications (VERA) as applied to full core subcritical multiplication ex-core calculations with Shift. The limiting factor in being able to run these detector calculations with Monte Carlo is the amount of computational memory used by a fully-loaded full core problem. The main development features added to VERA in this effort were enabling source biasing and separable weight windows and integrating multithreading in Shift. With these capabilities, a subcritical multiplication detector response calculation on a full core Watts Bar Nuclear Plant Unit 1 (WBN1) ex-core model was successfully run on a moderately sized computing cluster.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

New Polarizability Functions for Calculating Atomic Hydrogen S State Spectral Lines in Exact Agreement with NIST Measurements

This paper presents a new approach to calculating atomic hydrogen energy levels which is simple, fast and accurate. One advantage of this new approach is the ability to explicitly incorporate the polarizability resulting from a scattering event between a hydrogen atom and a photon. The polarizability effect is small but significant, producing exact S state spectral energies. The general calculational approach is applicable to both atomic hydrogen and multi-electron energy levels. Although motivated by the multi-electron case for chemistry applications, the focus of this paper will be calculations that are specific to atomic hydrogen which include the polarizability.

74 ATOMIC AND MOLECULAR PHYSICS↗
DOE OSTI2021

Geochemical data package for performance assessment calculations related to the Savannah River Site

The Savannah River Site (SRS) disposes of low-level radioactive waste (LLW) and stabilizes high-level radioactive waste (HLW) tanks in the subsurface environment. Calculations used to establish the radiological limits of these facilities are referred to as Performance Assessments (PAs), Special Analyses (SAs), and Composite Analyses (CAs). The purpose of this document is to revise the existing geochemical data package used for these calculations (Kaplan, 2016). This work builds on earlier compilations of geochemical data, referred to a geochemical data packages (Kaplan, 2007; Kaplan, 2010; Kaplan, 2016; McDowell-Boyer et al., 2000). This work is being conducted as part of the on-going maintenance program of the SRS PA programs that periodically updates calculations and data packages when new information becomes available.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Validation of the Public Radiation Exposure Calculation for the Incident at the National Institute of Standards and Technology Center for Neutron Research on February 3, 2021

The Department of Energy/National Nuclear Security Administration Consequence Management Program was contacted by the Health Physics Chief of the National Institute of Standards and Technology Center for Neutron Research (NCNR) to review public radiation exposure calculations for an event that occurred on its Gaithersburg, Maryland, campus on February 3, 2021. Subject matter experts from the Nuclear Emergency Support Team (NEST) assets, specifically the Consequence Management Home Team (CMHT) and the National Atmospheric Release Advisory Center (NARAC), were selected to provide support. CMHT used three separate modeling codes to validate the results the scientists at NCNR calculated using the HotSpot model. The analyses were performed using NARAC’s in-house Lagrangian dispersion codes known as LODI and Aeolus, as well as the Turbo FRMAC software from Sandia National Laboratories. The team used parameters provided by the NCNR scientists regarding the site, applicable observable meteorological data, and environmental survey and sampling data to estimate public exposure. Each model estimated public dose at much less than 0.5 mrem. CMHT concurs with the NCNR public radiation exposure calculations which state that members of the public at the 400-meter boundary would have received a radiological dose of less than 0.5 mrem.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2021

Assessment of Tools for Molten Salt Reactor Dose Rate Calculations

This report discusses a preliminary assessment of the capabilities of current state-of-the-art stochastic codes Shift and MCNP6 to calculate the ex-core radiation dose rates for a simplified Molten Salt Reactor (MSR) model. The Monte Carlo code Shift has been under significant development in recent years at ORNL as part of the CASL program and is now supported by NEAMS. Originally, Shift was developed for LWR ex-core calculations but with dose rate and shielding calculations specifically requested by the NEAMS program’s MSR industry partners, the MSR Application Drivers team was tasked with assessing Shift for non-LWR applications. This was the first application of the Shift code for non-LWRs and the findings can be considered preliminary due to the activities occurring only over a 5-month period. Attractive features of Shift include massive parallelization and advanced automated variance reduction techniques such as CADIS and FW-CADIS.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Feedstock Carbon Intensity Calculator (FD-CIC) Users’ Manual and Technical Documentation

A transparent and easy-to-use tool for feedstock-specific, farm-level CI calculation of feedstocks is helpful. With the ARPA-E support, we have developed a tool–the Feedstock Carbon Intensity Calculator (FD-CIC). The first version of FD-CIC was released with the GREET® model in 2020 (Wang et al., 2020) so that corn feedstock producers could use this publicly available tool (https://greet.es.anl.gov/tool_fd_cic) to quantify corn grain CIs with farm-level input data and management practices. In the 2021 version, we expand the tool’s capabilities by including additional feedstocks such as soybeans, sorghum, and rice. Similar to corn, it calculates the farm-level CI for these feedstocks by allowing user-defined farm-level farming inputs and incorporating the GHG emission intensities of these inputs from GREET (in particular, GREET1, the fuel cycle model of GREET).

09 BIOMASS FUELS↗