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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 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
NASA NTRS1974

A new formulation of the conservation equations of fluid dynamics

The computation of time-dependent flows has inspired a new, higher-dimensional formulation of the conservation equations of fluid dynamics in which time is treated as a fourth coordinate. The formulation is derived for a constant-density flow, and then extended to a variable-density flow by introducing a fifth, fictitious coordinate. This new coordinate can also act as a source coordinate, so that external source terms can be included. The analysis is carried out for both incompressible, stratified flow, and compressible equilibrium flow. The results are then extended to non-equilibrium and magnetohydrodynamic flows. Several applications of the new formulation to the computation of time-dependent flows are discussed.

Vinokur, M.↗
NASA NTRS1997

Long-term variability in bright hard X-ray sources: 5+ years of BATSE data

The operation of the Compton Gamma Ray Observatory (CGRO)/burst and transient source experiment (BATSE) continues to provide data for inclusion into a data base for the analysis of long term variability in bright, hard X-ray sources. The all-sky capability of BATSE provides up to 30 flux measurements/day for each source. The long baseline and the various rising and setting occultation flux measurements allow searches for periodic and quasi-periodic signals with periods of between several hours to hundreds of days to be conducted. The preliminary results from an analysis of the hard X-ray variability in 24 of the brightest BATSE sources are presented. Power density spectra are computed for each source and profiles are presented of the hard X-ray orbital modulations in some X-ray binaries, together with amplitude modulations and variations in outburst durations and intensities in recurrent X-ray transients.

Robinson, C. R.↗
DOE OSTI2020

SCALE Code System

The SCALE Code System is a widely used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules, including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Source Term of Accidents Involving Vehicle Crashes and Fuel Fires

This calculation is to be used when developing accident analysis scenarios involving fuel fires at the WSF. The objectives of the calculation are to determine (1) the Source Term from releases of TRU waste involved in impacts, impacts with fuel fire, and fuel fire with no impact; and (2) the plume sensible heat from various sizes of gasoline pool fires.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Source Term of Accidents Involving Aircraft Crashes and Fuel Fires

This calculation is to be used when developing accident analysis scenarios involving aircraft fuel fires at the Waste Storage Facility (WSF). The objectives of the calculation are to determine (1) the Source Term from releases of TRU waste involved in impacts, impacts with fuel fire, and fuel fire with no impact; and (2) the plume sensible heat from various sizes of gasoline pool fires.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Improved Understanding of Airborne Release Fraction and Respirable Fraction from Postulated Free-Fall Spill in DOE Nuclear Facilities (Annual Project Summary Report for NSRD-38)

The two-year experimental project has been completed to update the Airborne Release and Respirable Fraction (ARF and RF) estimates from the free-fall spills of liquid waste simulants. The objectives of the project at the Savannah River National Laboratory (SRNL) are to: (1) establish a technically defensible bounding approach for a reasonably conservative limit to the fall height used in the Ballinger correlation for the ARF, such that spill heights for near-water density liquid solutions above a critical height result in the same value of the ARF; and (2) provide a revised technical basis and data to support the ARF and RF correlation for free-fall liquid spills of heights from 1 to 10 meters. After peer review, these data and associated correlation will be available for consideration in a revision to DOE-HDBK-3010, Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1974

Analysis of acoustic radiation in a jet flow environment

The analysis of sound fields from arbitrary source distributions in terms of Legendre and spherical Hankel functions is well known. The purpose of this paper is to extend this classical method of analysis to environments such as jet flows where flow and flow gradients are inherently present. The wave-equation governing the radiation of sound in such an environment is derived. The steady state flow and flow gradients in the axial and transverse directions appear as coefficients in the terms of the wave-equation. A semi-numerical method is used to solve the wave-equation in terms of modified spherical harmonics yielding the phase velocities and the directivities of an infinite set of modes. The directivity of each mode is obtained in terms of modified Legendre functions by numerical integration. Some results of these directivity and phase-velocity calculations are presented for a limited number of frequency and flow parameters. Both convective and shear refraction are shown to be important.

Mungur, P.↗
DOE OSTI2023

Application of NEAMS Multiphysics Framework for Species Tracking in Molten Salt Reactors

This report from Idaho National Laboratory (INL) summarizes the key modeling and simulation activities conducted under the Department of Energy (DOE) Molten Salt Reactor (MSR) Campaign during the Fiscal Year 2023 (FY23). The focus of the work was to leverage state-of-the-art modeling capabilities from the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) codes to enable novel multiphysics and multiscale modeling and simulation of MSRs. Through collaboration with NEAMS code developers, advanced multiphysics analysis capabilities for MSR systems were demonstrated by coupling depletion, thermal-hydraulics, and thermochemistry into an innovative framework for chemical species transport in MSRs. As a result, the framework can track nuclides throughout their lifetimes in the core, from production (depletion) to advection throughout the salt volume (thermal-hydraulics) and off-gassing or precipitation outside of the salt (thermochemistry). This work supports the near-term deployment of MSRs by integrating the synergistic efforts between the DOE’s MSR Campaign and NEAMS program. The resulting framework will help better connect system design modelers with experimentalists to better understand and predict complex physical behaviors in MSRs. Researchers and MSR developers alike can now leverage these new modeling and simulation capabilities to perform novel analyses with applications including: • MSR dynamics during normal operational transients and accident scenarios • Off-gas system design and performance for fuel cycle and depletion analysis • Corrosion and active chemistry control for reactor component health and lifetime determination • Source term, decay heat and activity determination in accident scenarios • Special nuclear material accountancy and chemical forensic analysis for safeguards • Digital twin development of experiments and experimental reactor demonstrations • Measurement requirements for instrumentation and control design • Uncertainty and sensitivity analysis of missing data to inform future experimental data collection.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Initial development of a generic fluoride salt-cooled reactor model

Fluoride high-temperature reactors (FHRs) are high-temperature, low-pressure reactor concepts that use tri-structural isotropic (TRISO) fuel and molten fluoride salt coolant. These reactors have the potential to provide both electrical power and high-temperature process heat. We used generic FHR parameters for a pebble-bed FHR to develop an initial model with fresh fuel for a generic FHR (gFHR) in MELCOR and SCALE (NEWT and KENO). In this paper, we present the development of our gFHR models, which will serve as the baseline for a sensitivity and uncertainty analysis to quantify the range of possible source terms for FHRs in severe accidents. We present MELCOR results for fuel and coolant temperatures through the core, a nodalization study for the steady-state thermal hydraulic model, and development of reactor physics models in SCALE. As this work progresses, these models will be used to calculate source terms for a loss-of-forced-flow accident and to conduct a sensitivity study on this accident to establish a range of possible source terms. SCALE will provide reactor physics parameters like isotopic inventory, decay heat generation, and temperature coefficients of reactivity. Using the uncertainty quantification tools within SCALE, we will generate distributions for those parameters and will use the uncertainty quantification code RAVEN or DAKOTA to sample those distributions in MELCOR to quantify the impact of reactor physics and thermal hydraulic uncertainties on FHR source terms. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Constellation's best estimate alternate source term methodology overview

Safety analyses for a nuclear power plants need to consider postulated accidents that results in at risk of accidental release of radiation. The regulations require plant specific safety analysis reports to include an evaluation of the requirements of 10CFR50.67. Such a safety analysis report mandates limits such that calculated radiological consequences relative to certain dose locations do not exceed a total effective dose equivalent (TEDE) limits following a postulated release of radioactivity. Calculations are performed to estimate the radiological consequences, in terms of dose, to people and equipment to ensure the estimated doses are within the prescribed limits. Analysis should demonstrate, with reasonable assurance, that these prescribed limits are complied with. Conventional methodologies utilize conservative approaches to address lack of uncertainty quantification in the utilized approaches, methods, and/or inputs. These built-in excess conservatisms often result in compounding effects and hence overly conservative results in the estimated radiological consequences, which leads to inaccurate margin evaluation for operation and accident mitigation. Therefore, evaluating accurate dose consequences is needed for both operational and safety reasons. The research documented in this paper provides an overview of Constellation's Best Estimate Alternate Source Term (BEAST) Methodology. BEAST methodology relies upon the use of realistic yet bounding input distributions for key analysis parameters, which replaces use of conservative deterministic singular inputs that bound overall analysis domain. This approach enables evaluating a more accurate accident analysis response, while enabling a bounding licensing basis envelope via more accurate quantification of uncertainty in the application. Therefore, built-in margin for a given scenario is more accurately evaluated.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1972

The interaction of Sco X-1 with its environment.

Analytic discussion of the ionization produced in a uniform medium by a bremsstrahlung X-ray source. In particular, the ionization of interstellar gas in the vicinity of Sco X-1 is considered, and the analysis is formulated in terms of the specific parameters associated with this source. A constant temperature in the medium surrounding Sco X-1 is assumed, and simple numerical and analytic solutions are obtained for electron density and emitted radiation variation with distance from the X-ray source.

Silk, J.↗
DOE OSTI2022

Application of Constellation's best estimate alternate source term methodology

Safety analyses for a nuclear power plants need to consider postulated accidents that results in a risk of accidental release of radiation. The regulations require plant specific safety analysis reports to include an evaluation of the requirements of 10CFR50.67. Such a safety analysis report mandates limits such that calculated radiological consequences at certain locations do not exceed established limits in 10CFR50.67 following a postulated release of radioactivity. Analyses should demonstrate, with reasonable assurance, that these prescribed limits are complied with. Conventional methodologies utilize conservative approaches to address lack of uncertainty quantification in the utilized approaches, methods, and/or inputs. These built-in excess conservatisms often result in with compounding effects and hence overly conservative results in the estimated radiological consequences, which leads to inaccurate margin evaluation for operation and accident mitigation. Therefore, evaluating accurate dose consequences is needed for both operational and safety reasons. This paper illustrates the use of the best estimate source term (BEAST) methodology to the case of the loss of coolant accident (LOCA) dose analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

WSF B696R Calcs

This calculation is applied to the accident analysis scenario of a limited combustibles fire breaching eight SWBs in B696R R1010. For a fire in a building with no breaches, there is only a ground release. The source term (ST) of the deflagration is determined by the Waste Storage Facilities Documented Safety Analysis (Ref. 1). The doses from the release are calculated by using MACCS2 (Ref. 2). This software take hourly meteorological data for a year as input. There are 8760 hours in a year. MACCS2 calculates a dose for each of the 8760 hours, taking into account the associated source term from Reference 1, the wind speed, the stability class (i.e., A through F), the directional sector of the wind (with associated distance to the site boundary). For each directional sector (i.e., 1 through 16) a MACCS2 run was performed. From the total collection of MACCS2 output files, Microsoft Excel is used to extract the dose that correspond to the meteorological conditions of each hour in the year. Note: although MACCS2 inputs a single year of meteorological data, five years are used (2010 through 2014) by performing multiple runs. This is discussed in Section 2.2. Conclusions: The final 95th percentile dose at the collocated worker used for accident analysis of a limited combustibles fire breaching eight SWBs in B696R R1010 is 18.4 rem.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2023

Integrated Resource and System Plannning Based on Energy Costs and Emission Analysis for a Sustainable Future

Efficient integration of renewable energy sources (RES) - such as wind, solar, geothermal, tidal, and wave - into the electric power grids can help improve energy security and contribute toward a sustainable energy future. To date, however, electric power grids are critically dependent on conventional sources, such as fossil fuels. Given the increasing penetration of RES and dynamic loads, there is a need to thoroughly reanalyze the performance of power grids in terms of cost, efficiency, and energy security. This paper presents a comprehensive analysis in terms of energy costs and emissions estimates of dynamic power grids incorporating various power-generating sources.

distributed energy resources↗
DOE OSTI2025

Development of an Advanced Multiphysics Simulation Capability for Radiant's Microreactor Design

Argonne National Laboratory and Idaho National Laboratory, through a Department of Energy Gateway for Accelerated Innovation in Nuclear Voucher, supported key analysis needs of Radiant related to (i) air jacket thermal fluid performance, (ii) evaluation of decay heat source terms defining air jacket technical requirements, and (iii) assessment of modeling methodologies employed for core analysis. All of these activities center on numerical simulation of various aspects of Kaleidos using the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework, the Cardinal multiphysics application, the OpenMC Monte Carlo code, and the Nek5000 computational fluid dynamics (CFD) code. This project builds upon an earlier Nuclear Energy Advanced Modeling and Simulation (NEAMS) Thermal-Hydraulic (T/H) Center of Excellence (CoE) project focused on initial demonstration of Cardinal multiphysics simulation of High Temperature Gas Reactors (HTGRs) and now focuses on Radiant’s Kaleidos concept.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1999

EGRET Team Papers to Be Published in the Proceedings of the Fifth Compton Symposium

Papers to be included in this symposium are: Detecting the Attenuation of Blazar Gamma-Ray Emission by Extragalactic Background Light with GLAST; Some Aspects of the Radio Emission of EGRET-Detected Blazars; GeV Outburst in Mrk 501; Spectral Modeling of the EGRET 3EG Gamma Ray Sources Near the Galactic Plane; X-ray and Gamma-ray Observations of the COS-B Field 2CG 075+00; Multiwavelength Studies of the Peculiar Gamma-Ray Source 3EG J1835+5918; EGRET/COMPTER Observations of an Unusual, Steep-Spectrum Gamma-Ray Source; Preliminary Results from a New Analysis Method for EGRET Data; and A Systematic Search for Short-term Variability of EGRET Sources.

Source record↗
DOE OSTI2021

Assessment of tritium effluent from Prototype Generation IV Sodium-cooled Fast Reactor

We report a tritium transport model for a pool-type sodium cooled fast reactor was developed for estimation of tritium effluents from the Prototype Generation IV Sodium-cooled Fast Reactor (PGSFR). For this purpose, the transport model developed for Experimental Breeder Reactor II and Fast Flux Test Facility was updated and validated. The tritium production rate from ternary fission and activations in PGSFR were calculated by the Argonne Reactor Computation (ARC) code suite for fast reactor analysis, and due to the potential uncertainties in source terms and variations of cold trap performance and material permeability during reactor operation, the tritium distributions and tritium effluents were estimated for several bounding cases. The upper bounding gaseous and liquid tritium release rates from PGSFR are similar to ~440.7 Ci/year and 14.7 Ci/year, respectively. The gaseous tritium effluent is about few orders of magnitude smaller than the regulatory constraint of the tritium concentration in the air. The peak tritium concentration in the steam-generator feedwater is similar to ~1.1x10 -3 mu Ci/g, which is required to be diluted prior to releasing as tritiated water. The accumulated tritium in the cold trap after five cycle operations is ~ 52,000 Ci. Because of significant conservatism in these upper bounding estimations, it is expected that the tritium effluent at the nominal case with plausible source terms would be about 2 orders of magnitude smaller than that of the upper bounding case.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

(α,n) nuclear data scoping study

Neutrons from the (α,n) reaction are an important component of nondestructive assay techniques to determine enriched uranium and other actinide inventories in a variety of critical points in the nuclear fuel cycle. However, uncertainties in the cross section, total neutron yield and neutron spectrum, and gamma emissions from these reactions, such as 19 F(α,n) and 17,18 O(α,n), introduce large uncertainties in the determination of mass of actinides of interest and can represent several significant quantities in unaccounted material in certain facility processes. Calculations and measurements depend on accurate nuclear data; however, much of the relevant data in use today was measured in the 1980s and earlier and has not been updated. Thus, the current uncertainties in the cross sections and neutron emission spectra are unacceptably large. This report documents the results of a scoping study of (α,n) reaction data that considered the current state of the data and recommends areas of improvement. It also addresses the codes use to calculate the (α,n) neutron and gamma source terms and recommends code improvements to support required analysis.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗