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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 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Requirements Description of DASSH-F

This report reviews the modeling and simulation capabilities of Argonne National Laboratory’s DASSH code that is used in present reactor analysis activities. These capabilities will be used to establish the set of verification tasks necessary to verify DASSH for use on commercial projects. A similar approach was taken for the PERSENT, REBUS and DIF3D software packages. The DASSH program is a thermal analysis code designed to rapidly allow a reactor design engineer to obtain flow rates requirements that satisfy peak temperature constraints in the domain. DASSH is a follow-on development to the SE2-ANL software and SUPERENERGY-2 software that it is based upon. DASSH was designed to account for both neutron and gamma heating and is inherently connected to the GAMSOR part of the ARC suite of fast reactor analysis software. SE2-ANL is a developed piece of software from the 1980s while DASSH is a modern implementation with notable improvements in geometry handling. The most important upgrade of DASSH relative to SE2-ANL is that it can analyze multiple time points in a single run where SE2-ANL can only treat a single time point. This allows the user to understand the impact of and search the flow distribution for the entire operational period of a reactor design considering pressure drop, peak coolant and fuel temperatures, and thermal striping. DASSH has three input paths that have to be verified. The first input path builds the geometry and power distribution based upon the DIF3D model but ignores the gamma heating aspects of the problem. The second input path also builds the geometry from the DIF3D model but it takes the neutron and gamma heating distributions from GAMSOR. The third input path is to take the geometry and power distribution directly from user input (i.e. not coupled to DIF3D or GAMSOR). DASSH also has many built in correlations for material properties along with a user defined specification of the fuel, structure, and coolant properties. There are correlations for flow split, mixing, pressure drop, and heat transfer coefficients (subchannel rather than a direct methodology). In total, verification of DASSH will require an extensive testing to cover all possible user features of the software.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

A High-Fidelity Model of the Peach Bottom 2 Turbine-Trip Benchmark Using VERA

This work presents a high-fidelity simulation of the Peach Bottom turbine trip (PBTT) benchmark using the Virtual Environment for Reactor Applications (VERA), a multiphysics reactor modeling tool developed by the U.S. Department of Energy’s Consortium for Advanced Simulation of Light Water Reactors energy innovation hub. The PBTT benchmark, based on a 1977 transient event at the end of cycle 2 in a General Electric Type-4 boiling water reactor (BWR), is a critical test case for validating core physics models with thermal feedback during rapid reactivity events. VERA was employed to perform end-to-end, pin-resolved simulations from conditions at the beginning of cycle 1 through the turbine-trip transient, incorporating detailed neutron transport, fuel depletion, and subchannel thermal hydraulics. The simulation reproduced key benchmark observables with high accuracy: the peak power excursion occurred at 0.75 s, matching the scram time and closely aligning with the benchmark average of 0.742 s; the simulated maximum power spike was approximately 7600 MW, which is within 3% of the benchmark average of 7400 MW; and void-collapse dynamics were consistent with benchmark expectations. Reactivity predictions during cycles 1 and 2 remained within 1500 pcm and 400 pcm of criticality, respectively. These results confirm VERA’s ability to model complex coupled neutronic and thermal hydraulic behavior in a BWR turbine-trip transient, which will support its use in future studies of modeling dryout, fuel performance, and uncertainty quantification for transients of this type.

BWR↗
NASA NTRS1979

Studies of oceanic tectonics based on GEOS-3 satellite altimetry

Using statistical analysis, geoidal admittance (the relationship between the ocean geoid and seafloor topography) obtained from GEOS-3 altimetry was compared to various model admittances. Analysis of several altimetry tracks in the Pacific Ocean demonstrated a low coherence between altimetry and seafloor topography except where the track crosses active or recent tectonic features. However, global statistical studies using the much larger data base of all available gravimetry showed a positive correlation of oceanic gravity with topography. The oceanic lithosphere was modeled by simultaneously inverting surface wave dispersion, topography, and gravity data. Efforts to incorporate geoid data into the inversion showed that the base of the subchannel can be better resolved with geoid rather than gravity data. Thermomechanical models of seafloor spreading taking into account differing plate velocities, heat source distributions, and rock rheologies were discussed.

Poehls, K. A.↗
NASA NTRS1984

Adaptive mobile access protocol (AMAP) for the message service of a land mobile satellite experiment (MSAT-X)

This paper describes a feasibility study of the adaptive mobile access protocol (AMAP) for MSATA-X, a proposed experimental mobile satellite communication network. The mobiles are dispersed over a wide geographical area and the channel data rate is limited due to the size and cost limitations of mobile antennas. AMAP is a reservation based multiple-access scheme. The available bandwidth is divided into subchannels, which are divided into reservation and message channels. The ALOHA multiple-access scheme is employed in the reservation channels, while the message channels are demand assigned. AMAP adaptively reallocates the reservation and message channels to optimize system performance. It has been shown that if messages are generated at a rate of one message per hour, AMAP can support approximately 2000 active users per 2400 bit/s channel with an average delay of 1.4 s.

Li, V. O. K.↗
NASA NTRS1984

Performance analysis of an adaptive multiple access scheme for the message service of a land mobile satellite experiment (MSAT-X)

This paper describes an Adaptive Mobile Access Protocol (AMAP) for the message service of MSAT-X., a proposed experimental mobile satellite communication network. Message lengths generated by the mobiles are assumed to be uniformly distributed. The mobiles are dispersed over a wide geographical area and the channel data rate is limited. AMAP is a reservation based multiple access scheme. The available bandwidth is divided into subchannels, which are divided into reservation and message channels. The ALOHA multiple access scheme is employed in the reservation channels, while the message channels are demand assigned. AMAP adaptively reallocates the reservation and message channels to optimize the total average message delay.

Yan, T.-Y.↗
NASA NTRS1994

A new presentation of complex voltage data for Goldstone radar astronomy

A presentation method similar to a vector field display or a data-based grid has been used to visualize complex voltage test data taken over multiple subchannels. Unlike simple plots of all data points on the complex plane, the position of the data in the time series is part of the presentation, providing additional information to aid in fault isolation during readiness testing. The 'phase-magnitude' presentation, as it has come to be called, was designed for the demands of real-time data acquisition and processing with remote monitoring over low bandwidth communication links.

Howard, S. D.↗
NASA NTRS2000

Heat Transfer in a Complex Trailing Edge Passage for a High Pressure Turbine Blade - Part 1: Experimental Measurements: Experimental Measurements - Part 1

A combined experimental and computational study has been performed to investigate the detailed heat transfer coefficient distributions within a complex blade trailing edge passage. The experimental measurements are made using a steady liquid crystal thermography technique applied to one major side of the passage. The geometry of the trailing edge passage is that of a two-pass serpentine circuit with a sharp 180-degree turning region at the tip. The upflow channel is split by interrupted ribs into two major subchannels, one of which is turbulated. This channel has an average aspect ratio of roughly 14:1. The spanwise extent of the channel geometry includes both area convergence from root to tip, as well as taper towards the trailing edge apex. The average section Reynolds numbers tested in this upflow channel range from 55,000 to 98,000. The tip section contains a turning vane near the extreme comer. The downflow channel has an aspect ratio of about 5:1, and also includes convergence and taper. Turbulators of varying sizes are included in this channel also. Both detailed heat transfer and pressure distribution measurements are presented. The pressure measurements are incorporated into a flow network model illustrating the major loss contributors.

Bunker, Ronald S.↗
DOE OSTI2016

VERA-CS Modeling and Simulation of PWR Main Steam Line Break Core Response to DNB

The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time step of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.

CTF↗
DOE OSTI2017

Review of CTF’s Fuel Rod Modeling Using FRAPCON-4.0’s Centerline Temperature Predictions

Coolant Boiling in Rod Arrays–Two Fluid (COBRA-TF), or CTF1 [1], is a nuclear thermal hydraulic subchannel code used throughout academia and industry. CTF’s fuel rod modeling is originally developed for VIPRE code [2]. Its methodology is based on GAPCON [3] and FRAP [4] fuel performance codes, and material properties are included from MATPRO handbook [5]. This work focuses on review of CTF’s fuel rod modeling to address shortcomings in CTF’s temperaturepredictions. CTF is compared to FRAPCON which is U.S. NRC’s steady-state fuel performance code for light-water reactor fuel rods. FRAPCON calculates the changes in fuel rod variables and temperatures including the eects of cladding hoop strain, cladding oxidation, hydriding, fuel irradiation swelling, densification, fission gas release and rod internal gas pressure. It uses fuel, clad and gap material properties from MATPRO. Additionally, it has its own models for fission gas release, cladding corrosion and cladding hydrogen pickup. It allows finite dierence or finite element approaches for its mechanical model.In this study, FRAPCON-4.0 [6] is used as a reference fuel performance code. In comparison, Halden Reactor Data for IFA432 Rod 1 and Rod 3. CTF simulations are performed in two ways; informing CTF with gap conductance value from FRAPCON, and using CTF’s dynamic gap conductance model.First case is chosen to show temperature is predicted correctly with CTF’s models for thermal and cladding conductivities once gap conductance is provided. Latter is to review CTF’s dynamic gap conductance model. These Halden test cases are selected to be representative of cases with and without any physical contact between fuel-pellet and clad while reviewing functionality of CTF’s dynamic gap conductance model. Improving the CTF’s dynamic gap conductance model will allowprediction of fuel and cladding thermo-mechanical behavior under irradiation, and better temperature feedbacks from CTF in transient calculations.

Toptan, Aysenur↗
DOE OSTI2017

Implementation of a pressure correction term in the CTF sub-channel code

CTF [1] is a modernized and improved version of the legacy subchannel code, COBRA-TF [2], that is being jointly developed and maintained by Oak Ridge National Laboratory (ORNL) and North Carolina State University. The code was adopted by ORNL for use in the Consortium for Advanced Simulation of Light Water Reactors (CASL) in 2012 for aiding in addressing CASL challenge problems. Since that time, activities related to CTF have included implementing software quality assurance measures, implementing new features and models, performing validation and verification testing, establishing and supporting a CTF User Group, and developing the code for use in coupled applications [3, 4, 5]. CTF is applicable to single- and two-phase flows in light water reactor geometries at normal and accident operating conditions. The two-phase flow model consists of the six-equation two-phase flow model [6] augmented by equations for the droplet field and also includes appropriate source terms. This work focuses on the numerical stability of the six-equation two-phase flow model and is part of a recent eort to improve CTF capabilities for simulating flows in boiling nuclear reactors (BWRs). As it is of common knowledge among the thermodynamic community, the six-equation two-phase flow model is not strongly hyperbolic, meaning some of its eigenvalues are either not real or not distinct [7]. In a Cauchy sense and for an initial-value problem, strong hyperbolicity was found to be equivalent to well-posedness which ensures uniqueness of the solution in the space-time used. This concept is also of importance when designing high-order numerical methods and also when implementing boundary conditions that rely on real eigenvalues such as Riemann solvers. A tremendous amount of research is available in the literature on how to hyperbolize the six-equation two-phase flow model by adding ad hoc terms to the system of equations often referred to as virtual mass or interfacial pressure correction term. From a mathematical prospective, the main idea driving this approach is to add terms to the phasic momentum equation containing first-order partial derivatives of the unknowns to modify the Jacobian matrix and thus canceling the imaginary part of the eigenvalues.For this work it was chosen to implement a interfacial pressure correction (IPC) term on the same model as what was done for the CATHARE system code [8], that allows to recover a two-pressure formulation. In the following, the six-equation two-phase flow model is recalled along with the IPC term and its influence on the eigenvalues. Then, numerical results and convergence studies are presented for a water-faucet problem [9] at a pressure of 10 bar. Finally, conclusions and future work are addressed.

Delchini, Marco↗
DOE OSTI2018

Mechanistic Modeling of Departure from Nucleate Boiling under Transient Scenarios

The critical heat flux (CHF) corresponding to the departure from nucleate boiling (DNB) is one of the major limiting factors in the design and operation of pressurized water reactors (PWRs). Various predictive tools have been proposed for steady-state conditions. Empirical correlations and look-up tables yield relatively good agreement with specific experimental datasets and are widely used in subchannel codes for PWR transient simulations. However, experimental studies have revealed that during fast transients, the CHF values can become significantly higher than those in steady-state or slow transient scenarios, causing this modeling approach to result in overly conservative DNB prediction. This paper presents a mechanistic transient CHF model. Based on prior work, two DNB triggering mechanisms prevail in this model—the hydrodynamic thinning process and the thermal thinning process—both of which rely on the generally admitted liquid sublayer dryout theory. These mechanisms evaluate the depletion of the liquid sublayer underneath vapor slugs flowing over the channel. This model is further validated against three sets of power transient experiments at different operating conditions. While it clearly outperforms steady-state approaches and generally agrees closely with measurements, it still remarkably under-estimates CHF for very fast transients at low pressure. Future investigations will address this limitation.

Zhao, X. G.↗
DOE OSTI2017

Validation of CTF Droplet Entrainment and Annular/Mist Closure Models using Riso Steam/Water Experiments

This document summarizes the work to validate the droplet entrainment and de-entrainment models, as well as two-phase closure models in the CTF code [1] by comparison with experimental data obtained at Risø National Laboratory [2]. CTF is the modernized version of the subchannel code, COBRA-TF [3], being jointly developed by Oak Ridge National Laboratory and North Carolina State University for applications in the Consortium for Advanced Simulation of Light Water Reactors (CASL). The Risø data included a series of more than 250 steam/water experiments that were performed in both tube and annulus geometries over a range of various pressures and outlet qualities. Experimental conditions wereset so that the majority of cases were in the annular/mist flow regime. Measurements included liquid film flow rate, droplet flow rate, film thickness, and two-phase pressure drop. CTF was used to model 180 of the tubular geometry cases, matching experimental geometry, outlet pressure, and outlet flow quality to experimental values. CTF results were compared to the experimental data at the outlet of the test section in terms of vapor flow fractions, and entrained liquid flow fractions, and pressure drop per unit length.

Wysocki, Aaron↗
DOE OSTI2018

CTF_

CTF is a thermal-hydraulic simulation tool designed for modeling the reactor vessel region of light water reactors. The code is capable of modeling both coolant and solid regions and includes a nuclear fuel rod model. Coolant is represented with a 9-equation (two phases plus droplets and non-condensable gas) model with closure relationships being provided for normal and accident conditions. Meshing is done using the subchannel approach, which models the coolant region as a collection of parallel 1-D flow paths that communicate laterally through "gaps". CTF also includes a collection of supporting software tools that include: Xml2ctf - A utility for creating native CTF input decks from a simplified input file; pyCTF - A collection of Python utilities used for post-processing CTF results, manipulating CTF input models, and testing differences between output files; CTFFuel - A separate front-end to the CTF fuel rod solver package; Multistate - A separate driver for CTF that communicates through the code coupling interface; and veraAPI - A c++ interface to the code.

Salko, Robert K↗
DOE OSTI2018

Development of Preliminary VERA-CS Crud Induced Localized Corrosion Modeling Capability

The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing multiphysics core-simulator software for light water reactors (LWRs) known as VERA-CS in order to improve the state of the art in modeling and simulation of challenge problems that are limiting to the nuclear industry. One of these challenge problems includes fuel rod crud deposition, which can lead to crud-induced power shift (CIPS) and crud-induced localized corrosion (CILC). This paper documents work that was performed to develop a preliminary CILC-modeling capability in VERA-CS in support of the crud challenge problem. The CILC capabilities were developed by coupling VERA-CS to the CASL-developed Cicada code, which provides 1D and 3D clad conduction and oxide growth modeling tools, as well as coupling to the CASL-developed MAMBA code, which is used for modeling clad crud deposition. An approach called rod thermal-hydraulic reconstruction (ROTHCON) was developed and integrated into VERA-CS. This allows the modeler to capture spacer-grid turbulence and heat transfer effects in the CTF subchannel code so that the spatial resolution of crud and oxide rod surface growth could be better resolved. After implementing these capabilities, several assessments were performed to ensure that the capabilities function as expected, and a pin-resolved quarter-core simulation was run as a demonstration.

Salko Jr, Robert↗
DOE OSTI2019

Results of the Phenix Dissymmetric Test Benchmark Exercise

In the context of the European H2020 project SESAME, a work package was dedicated to the validation of code coupling for the thermal-hydraulics of pool type reactors. To this end, CEA proposed a benchmark exercise of the so-called dissymmetric test performed at the end of life of the PHENIX reactor. The PHENIX reactor was an experimental reactor operated by CEA in France from 1974 until 2009. The knowledge and data acquired over the lifetime of this reactor are of paramount importance for the development of future sodium-cooled generation IV reactors. At the beginning of this test, the reactor is operated at nominal power (350 MWth) with 3 primary pumps and 4 intermediate heat exchangers. During this test, a pump trip occurs at the secondary side of one of the two loops (one loop provides coolant to two intermediate heat exchangers). The chain reaction is then stopped while the pumps remain operating at nominal speed. Dissymmetric flow patterns are then observed in the cold pool as hot sodium leaves the intermediate heat exchangers that are no longer cooled. The failed secondary pump is then restarted while the rotational speed of the other is reduced. Eventually, a stable homogeneous state is reached. This transient is challenging for both modelers and codes since this exercise tackles an industrial scale reactor featuring complex flows and interactions between different scales. The physical phenomena observed include thermal stratification, flow mixing and buoyancy effects. In order to accurately model these phenomena for a reasonable computational costs, the partners used System Thermal Hydraulics codes coupled with Computational Fluid Dynamics codes and one even included a subchannel code. In this paper, the PHENIX reactor and its behavior during the dissymmetric test are described. Then, a comparison of the different modeling options selected by the partners is offered. Finally, the results are compared to the experimental data and discussed and modelling recommendations are proposed.

code coupling↗
DOE OSTI2019

NOBLE METAL MASS TRANSPORT MODEL FOR MOLTEN SALT REACTOR ANALYSIS IN VERA-CS

Molten Salt Reactor designs require robust multiphysics modelling capabilities to further the development of this next generation technology. Due to the unique liquid fuel design, traditional reactor physics analysis codes do not capture all of the important physics of these reactors. This work builds upon a general species transport model implemented into the multiphysics core simulator suite VERA-CS, by adding source and sink models for non-soluble fission products, specifically the noble metals. Noble metal fission products are born in the fuel-salt, but do not form stable fluorides, and instead plate out on various surfaces in the reactor fuel loop. A concentration gradient diffusion-convection based mass transfer model is implemented into the general species transport model within CTF-the subchannel thermal hydraulics code in VERA-CS-and verified. Two noble metal decay chains are analyzed, and their steady state bulk and surface concentrations are calculated for a simple flow loop in CTF that is representative of the Oak Ridge National Laboratory Molten Salt Reactor Experiment (MSRE). Non-soluble species 99Mo to 99Tc captures wall plate out of decaying species, and non-soluble 135Te to soluble 135I and gaseous 135Xe captures wall loss of decaying species. Results are compared with MSRE data and the possible multiphysics effects of this phenomena are identified. Future work to enhance the robustness of the physical model and couple it with neutronics to evaluate these effects of Noble Metal mass transport is discussed.

Walker, Samuel↗
DOE OSTI2022

Machine Learning-based Prediction of Departure from Nucleate Boiling Power for the PSBT Benchmark

Machine Learning (ML) has seen an exponential growth in its applications due to its advanced data driven prediction capabilities. The study presents a data-driven approach as a preliminary attempt to predict the power at which departure from nucleate boiling (DNB) occurs in pressurized water reactors (PWRs) by constructing an advanced ML algorithm that takes outlet pressure, inlet temperature and inlet mass flux as the input features. DNB is a critical heat flux (CHF) phenomenon seen in PWRs. The experimental data from the PWR subchannel and bundle tests (PSBT) benchmark is first used to train an artificial neural network (ANN) to predict the DNB power, which produces a root mean square error (RMSE) of 6.89 kW/m when tested on a blind subset of the PSBT data. Since the PSBT dataset is relatively small to train an accurate ANN, a data augmentation methodology based on generative adversarial networks (GANs) is used to expand the training dataset. By assuming that the real data follows a certain distribution, GANs try to learn that underlying distribution to generate similar synthetic data to augment the database and to improve the predictive capabilities of the ANN. The data generated from GANs are validated using 1-nearest neighbor and kernel maximum mean discrepancy. To further ensure data from GAN is similar to PSBT, the data is tested and filtered out using the sub-channel thermal-hydraulic code CTF. The results indicate that with the addition of 120 data points from GAN the RMSE reduces to 4.84 kW/m showing promising results for future developments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗