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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 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
NASA NTRS1991

On the application of ENO scheme with subcell resolution to conservation laws with stiff source terms

Two approaches are used to extend the essentially non-oscillatory (ENO) schemes to treat conservation laws with stiff source terms. One approach is the application of the Strang time-splitting method. Here the basic ENO scheme and the Harten modification using subcell resolution (SR), ENO/SR scheme, are extended this way. The other approach is a direct method and a modification of the ENO/SR. Here the technique of ENO reconstruction with subcell resolution is used to locate the discontinuity within a cell and the time evolution is then accomplished by solving the differential equation along characteristics locally and advancing in the characteristic direction. This scheme is denoted ENO/SRCD (subcell resolution - characteristic direction). All the schemes are tested on the equation of LeVeque and Yee (NASA-TM-100075, 1988) modeling reacting flow problems. Numerical results show that these schemes handle this intriguing model problem very well, especially with ENO/SRCD which produces perfect resolution at the discontinuity.

Chang, Shih-Hung↗
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↗
NASA NTRS1992

On the numerical treatment of nonlinear source terms in reaction-convection equations

The objectives of this paper are to investigate how various numerical treatments of the nonlinear source term in a model reaction-convection equation can affect the stability of steady-state numerical solutions and to show under what conditions the conventional linearized analysis breaks down. The underlying goal is to provide part of the basic building blocks toward the ultimate goal of constructing suitable numerical schemes for hypersonic reacting flows, combustions and certain turbulence models in compressible Navier-Stokes computations. It can be shown that nonlinear analysis uncovers much of the nonlinear phenomena which linearized analysis is not capable of predicting in a model reaction-convection equation.

Lafon, A.↗
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 NTRS2001

An Exact Form of Lilley's Equation with a Velocity Quadrupole/Temperature Dipole Source Term

There have been several attempts to introduce approximations into the exact form of Lilley's equation in order to express the source term as the sum of a quadrupole whose strength is quadratic in the fluctuating velocities and a dipole whose strength is proportional to the temperature fluctuations. The purpose of this note is to show that it is possible to choose the dependent (i.e., the pressure) variable so that this type of result can be derived directly from the Euler equations without introducing any additional approximations.

Goldstein, Marvin E.↗
DOE OSTI2020

Mechanistic source term development for liquid fueled MSRs - Model Development Update

This report provides results from a critical literature review on the development of a mechanistic source term model (MST) for liquid fueled molten salt reactors (MSRs). It is submitted in fulfillment of milestone M4AT-20AN040601075 in the Advanced Reactor Technologies Molten Salt Reactors Campaign work package AT-20AN04060107.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS2000

Implementation of New Turbulence Spectra in the Lighthill Analogy Source Terms

The industry-standard MGB approach to predicting the noise generated by a given aerodynamic flow field requires that the turbulence velocity correlation be specified so that the source terms in the Lighthill acoustic analogy may be computed. The velocity correlation traditionally used in MGB Computations is inconsistent with a number of basic qualitative properties of turbulent flows. In the present investigation the effect on noise prediction of using two alternative velocity correlations is examined.

Woodruff, S. L.↗
DOE OSTI2022

Source Term Analysis of Xenon (STAX): An effort focused on differentiating man-made isotope production from nuclear explosions via stack monitoring

An overview of the hardware and software developed for the Source Term Analysis of Xenon (STAX) project is presented which includes the data collection from two stack monitoring systems installed at medical isotope production facilities, infrastructure to transfer data to a central repository, and methods for sharing data from the repository with users. STAX is an experiment to collect radioxenon emission data from industrial nuclear facilities with the goal of developing a better understanding of the global radioxenon background and the effect industrial radioxenon releases have on nuclear explosion monitoring. The final goal of this work is to utilize collected data along with atmospheric transport modeling to calculate the contribution of a peak or set of peaks detected by the International Monitoring System (IMS) to provide desired discriminating information to the International Data Centre (IDC) and National Data Centers (NDCs). Types of data received from the STAX equipment are shown and collected data was used for a case study to predict radioxenon concentrations at two IMS stations closest to the Institute for RadioElements (IRE) in Belgium. The initial evaluation of results indicate that the data is very valuable to the nuclear explosion monitoring community.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

Transition of SRT Source Term Software to Modern Version Control System to Enable More Efficient Verification, Validation, and User Engagement

The Simplified Radionuclide Transport (SRT) code assists in reactor design and licensing by providing a detailed analysis of radionuclide transport under specific transient scenarios. This analysis is essential for ensuring that reactors meet safety and regulatory standards before they are authorized to operate. Mechanistic source term analysis, which involves a realistic evaluation of radionuclide behavior from the source to the environment, is a key component of this process. It provides a comprehensive understanding of potential release scenarios, supporting the development of robust safety measures.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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 OSTI2021

Unclassified Source Term and Radionuclide Data for the Groundwater Flow and Contaminant Transport Model for Corrective Action Units 101 and 102: Central and Western Pahute Mesa, Nye County, Nevada, Revision 1

This report documents the evaluation of information and data available on the unclassified source term and radionuclide (RN) contamination for Central and Western Pahute Mesa (PM): Corrective Action Units (CAUs) 101 and 102. The Central and Western PM CAUs constitute two areas of the Nevada National Security Site (NNSS) used for underground nuclear testing (Figure 1-1). Eighty-five underground nuclear tests were conducted at the PM CAUs between 1965 and 1992 (NNSA/NFO, 2015b). Underground nuclear tests deposit radioactive elements in the subsurface, resulting in groundwater contamination in the vicinity of the underground test locations. At the PM CAUs, the testing resulted in more than 2.68 × 10 7 curies (Ci) of radioactivity (decay-corrected to September 30, 2012) remaining underground (Finnegan et al., 2016). Of the 85 underground nuclear tests conducted at the PM CAUs, three were shallow tests detonated within 200 meters (m) of the surface and referred to as the “near-surface tests”; the remaining 82 were detonated deeper than 200 m below ground surface (bgs) and are referred to as the “deep underground tests” in this document.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1989

LU-SGS implicit algorithm for three-dimensional incompressible Navier-Stokes equations with source term

A numerical method is developed for solving the incompressible Navier-Stokes equations using the concept of pseudocompressibility. A lower-upper symmetric-Gauss-Seidel implicit scheme is developed for three-dimensional incompressible viscous flow computations. The present algorithm offers additional advantages when solving the flow equations with source terms. Complete vectorizability of the algorithm on oblique planes of sweep in three-dimensions is accomplished in a new flow solver, INS3D-LU code. Spatial differencing is a second-order accurate semi-discrete finite-volume method augmented by a third-order accurate numerical dissipation model which is based on spectral-radii. Comparison of numerical solutions for a curved duct with experimental data shows good agreement. The method is applied to calculate the inducer flow of the Space Shuttle Main Engine turbopump.

Yoon, Seokkwan↗
DOE OSTI2023

Advanced Modeling and Simulation to Characterize Advanced Boiling Water Reactor Source Terms to Support a Regulatory Approval Pathway for Right-Sized Emergency Planning Zone. Advanced Boiling Water Reactor (A-BWR) Project Phase 2

The United States Department of Energy (US DOE) is currently supporting the development of various small modular reactor (SMR), microreactor and advanced reactor designs. These reactors have improved safety features as compared to conventional Large Light Water Reactors (LLWRs), which include features that improve potential reduction of radiological source terms in the event of design and beyond-design basis accidents. Specifically, some reactors feature a relatively smaller containment volume with respect to the available surface area for fission product deposition; these include Integrated Pressurized Water Reactors (iPWRs), such as the NuScale SMR and SMR-type Boiling Water Reactors (BWRs) such as General Electric-Hitachi’s BWRX-300 design.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Development of a BlueCRAB/MELCOR Framework for Supporting Realistic Mechanistic Source Term Calculations in Microreactors

Efforts are currently underway to deploy microreactor modeling and simulation tools to better support vendors and regulatory authorities in submitting and reviewing licensing applications. In particular, the Nuclear Regulatory Commission is expected to rely on the Comprehensive Reactor Analysis Bundle (BlueCRAB) multiphysics toolset in performing design- and beyond-design-basis accident analyses. In addition, the Nuclear Regulatory Commission has been using the MELCOR code to estimate mechanistic source terms during accidents. As MELCOR relies on isotopic inventory and reactor temperature/power evolution profiles during accident conditions—all of which can theoretically be obtained from BlueCRAB—the ultimate goal of this activity is to establish a common BlueCRAB-MELCOR framework. However, prior to the present research, BlueCRAB had never been used to calculate such quantities of interest at the full-core level. While there are many Monte Carlo (MC) codes capable of computing such quantities of interest, they are unable to readily account for multiphysics feedback. BlueCRAB allows for the coupling of different physics codes together to perform multiphysics-informed calculations. Therefore, the purpose of this fiscal year 2023 work is to investigate the feasibility and challenges of performing such calculations within BlueCRAB so as to generate the data that MELCOR relies on. To demonstrate the methodology, the proposed workflow was applied to a prototypical heat pipe-cooled microreactor model. To predict isotopic concentrations (taking into account the ac- cumulation of fission products during operation), the necessary microscopic cross sections were generated via OpenMC and tabulated with respect to temperature and burnup. Next, a recently developed capability in Griffin (the reactor physics application in BlueCRAB) was used to convert the OpenMC output format into the ISOXML format used by Griffin. A multiphysics microscopic depletion calculation that involved performing a coupled full-core, heterogeneous neutron trans- port and thermal calculation at each depletion step was conducted to deplete the core to end of life (EOL) conditions so as to provide both isotopics and the initial condition for the transient calculation. Following a brief null-transient to verify that the initial condition had been properly restarted and was indeed in thermal equilibrium, a heat pipe failure transient was simulated. Thus, the entire workflow of using BlueCRAB to generate MELCOR inputs, from cross-section generation to producing isotopic inventory and power/temperature evolution profiles during transients, is demonstrated. This report also details the identified gaps in the workflow and how they were (for the most part) addressed. Future work should focus on directly including MEL- COR into the workflow by performing a MELCOR calculation using the BlueCRAB-generated input data. In addition, the heat pipe reactor design should be improved so as to reflect more prototypical burnup characteristics at EOL.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRSDate not supplied

LES Simulation of Cooling Airflow of High-Pressure Turbine Using the Source Term Approach

In this study, the Open National Combustion code (OpenNCC) is applied to simulate the airflow inside the high-pressure turbine (HPT) of the energy efficient engine (EEE). The main objective of this study is to validate the proposed methodology of a three-dimensional unsteady calculation (LES) of the HPT with a moving mesh capability to capture a relative motion at the stator and rotor interface. In our previous study, the adiabatic wall assumption was made at all the solid surfaces of HPT (i.e., the cooling airflows were not included.) To relax this assumption, we implement the surface source team approach, in which we impose the source term at a specific area of a cooling airflow exit at the solid surfaces by specifying an injection angle, temperature, turbulent intensity, and mass flowrate of each cooling airflow. The beneficial feature of this approach is that it does not require a mesh for each hole. For the validation of the model, the data for EEE model is obtained from the General Electric (GE) test campaign, and the full scale warm-air rig test condition representing a test point close to the integrated core/low spool design condition is considered.

CFD Hot-streaks↗
NASA NTRSDate not supplied

Numerical Investigation of the Effect of Cooling Airflow inside High-Pressure Turbine using the Source Term Approach

In this study, the Open National Combustion code (OpenNCC) is applied to simulate the airflow inside the high-pressure turbine (HPT) of the energy efficient engine (EEE). The main objective of this study is to compare the characteristics of the HPT performance obtained by using three different approaches (Case 1: the realistic combustion products are considered at the HPT inlet, but the cooling airflows are not included, Case 2: the working fluid is assumed to be a single gas (i.e., air), and the cooling airflows are considered, Case 3: the cooling airflow and the combustion products at the HPT inlet are included). To mimic the cooling airflows at the HPT surface, we use the surface source team approach, in which we impose the source term at a specific area of a cooling airflow hole at the solid surfaces by specifying an injection angle, temperature, turbulent intensity, and mass flowrate of each cooling airflow. For the validation of the model, the data for the EEE model was obtained from the General Electric (GE) test campaign (RDG10), and the full scale warm-air rig test condition representing a test point close to the integrated core/low spool design condition was considered. This validation test was done using the stationary uniform inflow condition (i.e., the reported total temperature and total pressure). Secondly, we considered the Sea-Level Take-off (SLTO) engine condition and separately performed the combustor simulation to obtain a more realistic time-averaged, spatially-nonuniform HPT inflow condition. This work provides a better understanding the effect of the cooling airflow model with different assumptions on the HPT performance. It is found that using different working gas have a minor effect on the performance metric, however the presence of the cooling airflows significantly affects the HPT performance such as the turbine efficiency.

CFD Hot-streaks↗
DOE OSTI2023

Application of MELCOR for Simulating Molten Salt Reactor Accident Source Terms

Molten Salt Reactor (MSR) systems can be divided into two basic categories: liquid-fueled MSRs in which the fuel is dissolved in the salt, and solid-fueled systems such as the Fluoride-salt-cooled High-temperature Reactor (FHR). The molten salt provides an impediment to fission product release as actinides and many fission products are soluble in molten salt. Nonetheless, under accident conditions, some radionuclides may escape the salt by vaporization and aerosol formation, which may lead to release into the environment. Here, we present recent enhancements to MELCOR to represent the transport of radionuclides in the salt and releases from the salt. Some soluble but volatile radionuclides may vaporize and subsequently condense to aerosol. Insoluble fission products can deposit on structures. Thermochimica, an open-source Gibbs Energy Minimization (GEM) code, has been integrated into MELCOR. With the appropriate thermochemical database, Thermochimica provides the solubility and vapor pressure of species as a function of temperature, pressure, and composition, which are needed to characterize the vaporization rate and the state of the salt with fission products. Since thermochemical databases are still under active development for molten salt systems, thermodynamic data for fission product solubility and vapor pressure may be user specified. This enables preliminary assessments of fission product transport in molten salt systems. In this paper, we discuss modeling of soluble and insoluble fission product releases in a MSR with Thermochimica incorporated into MELCOR. Separate-effects experiments performed as part of the Molten Salt Reactor Experiment in which radioactive aerosol was released are discussed as needed for determining the source term.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗