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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 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Mercury Speciation During Vitrification of LAW

n this work, laboratory and engineering-scale tests were conducted that mimic the reactive environment of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) low activity waste (LAW) melter and off-gas system in order to assess the speciation of mercury at select points in the processing system. The experimental protocols, test equipment, feed materials used, and rationale for testing are detailed in the Test Plan for this work. Particular attention was paid to the amount and speciation of mercury in submerged bed scrubber (SBS) solutions, which are intended to be recycled back to the melter feed at the WTP. Results are presented from detailed mercury analysis of these solutions using Environmental Protection Agency (EPA) Method 1630 and modifications to this method for dimethyl mercury and methyl mercury as well as EPA Method 1631 and modifications to this method for total, dissolved, elemental, suspended, and ionic mercury. Testing was conducted in two separate sets of experiments: (1) One set of tests (crucible-scale furnace tests) that involved heating small batches of mercury-spiked melter feed in crucibles to determine the amount of mercury retained in the glass and the species of mercury in the exhaust gases; (2) A second set of tests (DM10 melter tests) was planned that involved creating LAW melter plenum gas compositions using the DM10 melter system, injecting mercury into the off-gas stream, and passing that stream though a reactor that simulates various plenum gas conditions. Operational issues led to the need to use one of the VSL DM100 melters in place of the planned DM10 melter for these tests. The speciation of mercury after exposure to those conditions was monitored. In both sets of tests, the exhaust gases were run through a scrubber that was intended to mimic the LAW SBS in order to determine how much of the mercury exiting the melter would be retained in the primary off-gas system fluids and in what form. After passing through the SBS, the exhaust stream was analyzed to determine particulate, ionic, elemental, and total mercury passing downstream of the SBS. In tests employing gases derived from the DM100 melter that were spiked with elemental mercury, the processing system provided gas temperatures and residence times that are representative of the WTP LAW vitrification system in order to assess the effect of those conditions on mercury speciation. The Decontamination Factor (DF) across the system in the tests with mercury-spiked DM100 melter exhaust was determined using the analytical data from EPA Method 30B (Fluegas Adsorption Mercury Speciation (FAMS TM )) exhaust samples and the amount of mercury detected in the SBS solutions. Mercury species used in melter feed crucible scale tests were divalent (chloride and iodide), monovalent (chloride and fluoride), and elemental mercury. Individual tests included only a single form of mercury in the feed. The results for total mercury mass balances in the crucible tests are also presented.

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

Optimization of Hydrogen Production from Pyrolysis of Biomass Waste [Poster]

We developed an CFD aided ML-based tool for rapid assessment and optimization of different compositions of biomass in a fluidized bed reactor. First, we use CFD to simulate fluidized bed reactors with known inlet biomass mixtures and obtain corresponding syngas yields. A lumped kinetic mechanisms represents the conversion of cellulose, hemicellulose, and lignin, as well as subsequent cracking of tars into non-condensable gases (H 2 , CO, CO 2 , CH 4 ). We use Bayesian analysis/optimization to obtain the ideal operational and mass flow conditions for hydrogen production.

08 HYDROGEN↗
DOE OSTI2023

Microstructural characterization of the oxidation of nuclear graphite under chronic and accident conditions via XCT and synchrotron XCT

Graphite is a candidate material to moderate fast neutrons and for structural components in the US next-generation graphite-moderated reactors. A graphite core is conceived as a large formation of interconnected bricks that primarily serves as the moderator of fast neutrons, holds instrumentation, fuel elements, control rods, and is a receptacle for molten salts for Molten Salt Reactors (MSRs) designs. During the operation of a graphite-moderated power plant, graphite components might be subjected to chronic oxidation during normal operating conditions or aggressive oxidation as a result of accidental ingress of air that reacts vigorously with the graphite core. Using synchrotron, x-ray computed tomography (XCT), this research systematically characterized microstructural changes that accompanied these two oxidation scenarios. Chronic oxidation was studied by characterizing IG-110, PCEA, and NBG-18 specimens that were gradually oxidized in air at a low temperature (520°C). The accidental ingress of air into the graphite reactor was simulated by subjecting the grades mentioned above to high-temperature oxidation at approximately 750°C. This research is the first, in situ, systematic characterization of nuclear graphite microstructural evolution that can be associated with the two possible oxidation scenarios and provide insight into related repercussions. The results show that the microstructure and pore connectivity influence the rate of oxidation and evolution of the microstructure under the two oxidation regimes. These results are crucial to understanding which grades are more resilient to each type of oxidation and documenting the damage created in the graphite.

Arregui Mena, Jose'↗
DOE OSTI2026

Onset of Fluidization in MP-PIC Simulations using MFIX-Exa

Fluidized bed reactors are used across a variety of industries, including for energy processes like pyrolysis that result in low-cost energy products. Design and scale-up of fluidized beds is de-risked by modeling and simulation, utilizing tools like NETL’s MFIX-Exa High-Performance Computing (HPC) code for reacting multiphase flow. This report summarizes an investigation into the breadth of problems to which MFIX-Exa may be applied, specifically with regard to low fluid velocities and the onset of fluidization. A simple fluidization study is conducted both experimentally and numerically for particles of interest, then reactor simulations are compared to cold flow experiments for uniform distributor plates. Approaches for modeling bubble caps are also presented.

discrete particle method↗
DOE OSTI2022

CTF: A modernized, production-level, thermal hydraulic solver for the solution of industry-relevant challenge problems in pressurized water reactors

CTF is a thermal hydraulic (T/H) subchannel tool that has been extensively developed over the past ten years as part of the Consortium for Advanced Simulation of Light Water Reactors (CASL) program. The code was selected early in the CASL program for support of high-impact challenge problems that were found to be relevant to the nuclear industry and its currently operating fleet of pressurized water reactors (PWRs), including issues such as departure from nucleate boiling (DNB), crud-induced power shifts (CIPSs), and reactivity-insertion accidents (RIAs). By incorporating CTF into the multiphysics Virtual Environment for Reactor Application (VERA) core simulator software developed by CASL, CTF has become the primary means of providing fluid and fuel thermal feedback, as well as T/H figure-of-merits (FOMs) in large-scale reactor simulations. With the goal of solving industry challenge problems, CASL placed great emphasis on developing high-quality, high-performance, validated software tools that offer higher fidelity than what is currently possible with current industry methods. In support of this effort, CTF was developed from a research tool into an nuclear quality assurance (NQA-1)–compliant, production-level software tool that is capable of addressing the stated challenge problems and goals of CASL. This work presents a review of the major technological achievements that were realized in developing CTF over the past decade of the CASL program and presents an overview of the code solution approach and closure models.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Impact of Variable Gas Mixtures on Bubble Size Distribution and Mass Transfer in Gas Fermentation Reactors

Gas fermentation has emerged as a promising new technology for the generation of fuels and chemicals from mixtures of greenhouse and energy rich gas streams (CO2/CH4/H2/CO) via microbial bioreaction. Example pathways include biomethanation (CO2/H2 to CH4), biogas upgrading, CO fermentation and wet-waste conversion. Effective Gas-liquid mass-transfer is an important physical phenomenon that determines the design and scale-up of these systems. There is currently a knowledge-gap regarding bubble-size distributions when using a mixture of gases with vastly different properties, which can have a significant impact on overall mass-transfer. For example, hydrogen bubbles are more buoyant compared to other relatively heavier gases (CO2/CH4/CO), resulting in a large distribution of residence times and bubble sizes. This work therefore develops a deeper understanding of bubble dynamics and interphase mass transfer in such heterogenous gas mixtures through well-resolved computational models. We use a detailed multiphase computational-fluid-dynamics (CFD) model to study the impact of gas-mixtures on overall mass-transfer in bubble column and air-lift reactors. The CFD tool previously developed by the authors (1) for simulating aerobic fermentation reactors at scale is used in this study. The Reynolds-averaged mass, momentum, energy, and species transport equations are solved for interpenetrating gas and liquid phase in this model. We use a population balance-based bubble-size-distribution model that is validated against small-scale experiments in our solver. Results pertaining to multiple simulations of gas-fermentation reactors are presented where gas mixtures with varying compositions of CO2/CH4/CO/H2 are imposed at the sparger boundaries. The spatio-temporal variations in bubble-size distribution and mass transfer coefficient is analyzed for varying superficial velocities and gas-compositions for varying sizes of bubble-column and airlift reactors.

BIOMASS FUELS,MATHEMATICS AND COMPUTING↗
DOE OSTI2021

Predicting thermodynamic and thermophysical properties of molten chloride salts from ab-initio and classical molecular dynamics simulations

Molten salt reactors (MSRs) are among the advanced concepts pursued under the generation IV nuclear energy technology umbrella. However, the basic concept is not new and was first developed as part of the effort to power aircrafts with nuclear energy in the 1950’s. Later in the 1960’s, Oak Ridge National Laboratory (ORNL) built and operated the Molten-Salt Reactor Experiment (MSRE). This reactor used a fluoride salt with uranium as fuel. Fluorides salts are still highly relevant and proposed in several designs. In addition, chloride salts are being considered for MSRs operating in the fast neutron spectrum. This report focuses on chloride salts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Conceptual Design of Passive Neutron Albedo Reactivity and Active Neutron Albedo Reactivity Instruments for Various Arrangements of Pebble Bed Reactor Pebbles

MCNP6 simulations were performed to assess the anticipated capability of the Passive Neutron Albedo Reactivity technique for nuclear safeguards measurements of irradiated pebble/pebbles. Two physical setups were examined. One with a single pebble of three different burnups and one with 27 pebbles in a cube. For all cases the sensitivity to removing all the fissile material was examined. For both of the physical setups, 4 assay cases were examined: (a) singles count rates for which the only source neutrons were the inherent neutrons in the fuel, (b) singles count rates again except this time two relatively weak AmLi sources were placed above and below the top central pebble and the combined effect of both neutron source terms were examined, (c) doubles count rates for which the only source neutrons were the inherent neutrons in the fuel, and (d) doubles count rates again except this time two relatively weak AmLi sources were placed above and below the top central pebble and the combined effect of both neutron source terms were examined. Finally, a few parameters such as Cd near the 3 He tubes were perturbed to see if the results could be improved.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Incorporation of Thermal Hydraulic Models for Thermal Power Dispatch into a PWR Power Plant Simulator

This report describes the development, modeling, and results of a generic pressurized water reactor power plant simulator that incorporates coupled electrical and thermal power dispatch to an industrial process located approximately one kilometer from the nuclear power plant. The simulator is a commercial PWR simulator that has been modified to include thermal power dispatch as described in past milestone reports [ , ]. The commercial PWR simulator is a generic simulator available from GSE SYSTEMS® (Sykesville, MD, USA) that is built using RELAP5-HDTM Real-Time Solution and in-house software developed by GSE Systems. This generic PWR (GPWR) simulator performs real-time simulation of the complete power plant from the reactor neutronics to the electricity generation and distribution. All primary, secondary, and auxiliary systems are modeled including all control logic in order to provide the most accurate representation of actual nuclear power plant (NPP) operation, and the simulator results have been rigorously verified by an actual NPP operating at approximately 1 GWe. This report is a continuation of worked performed in previous years, and supplemental information from previous reports is included in the appendix for reference.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

VERA BWR progression problems

During the first phase of the Consortium for Advanced Simulation of Light Water Reactors (CASL) program, the Virtual Environment for Reactor Applications (VERA) was developed with a focus on capabilities for high-fidelity, multiphysics simulation of pressurized water reactors (PWRs). During this development effort, a set of progression problems was created ranging from smaller pin cell calculations to larger 3D full-core calculations. These progression problems helped to guide the development of the software and served as benchmarks against which to test VERA. Since 2019, efforts have been made to extend the capabilities of VERA to model boiling water reactors (BWRs). Because BWR simulations come with many unique challenges, a set of BWR progression problems was developed to aid in this new effort. The BWR progression problems range from 2D lattice calculations to 3D mini-core problems, and reference neutronic solutions were computed using continuous-energy Monte Carlo codes. MPACT, one of the neutronics code in VERA, was benchmarked using the BWR progression problems. The code is capable of computing solutions to all problems. The eigenvalues computed by MPACT agree well with the Monte Carlo reference solutions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Simulations of neutron noise in the research reactor AKR-2: comparison between a discrete ordinates and a diffusion-based method

A diffusion-based and a discrete ordinates method are used to simulate a neutron noise experiment in the research reactor AKR-2 at the Technical University in Dresden, Germany. The AKR-2 reactor provides an interesting case for the comparison between the two methods because it is characterized by large heterogeneities and regions with low macroscopic neutron cross-sections. For the calculations, the same spatial discretization and the same set of two-energy macroscopic neutron cross-sections with isotropic scattering are used. Significant discrepancies between the diffusion-based and discrete ordinates methods are found in regions of the systems where the diffusion approximation is expected to be inaccurate in reproducing characteristics of the static neutron flux and neutron noise. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Transient Modeling and Simulation of a Generic Stable Salt Reactor

A SAM system-level model of a generic stable salt reactor has been developed to investigate thermal-hydraulic behavior and safety performance under steady and transient conditions. The model integrates information generated from a reactor physics analysis using PROTEUS and PERSENT, and a computation fluid dynamics (CFD) analysis using STAR-CCM+. A loose, iterative coupling scheme between PROTEUS and SAM is implemented to calculate the equilibrium power and temperature distributions in the steady-state critical core condition. The converged steady-state model is then used in PERSENT to calculate the four reactivity feedback temperature coefficients (Doppler, fuel density, coolant density, and core radial expansion) and kinetic parameters that are needed in SAM to model the temperature feedback effects in transient simulations. Within the fully enclosed liquid fuel pins, natural convection is the dominant heat transfer mechanism. The STAR-CCM+ model of the fuel pin considers conjugate heat transfer from the liquid fuel salt to the pin cladding and external reactor coolant. The CFD results of the axial and radial temperature profiles are used to empirically determine an effective fuel salt thermal conductivity in the SAM fuel pin model so that the temperatures predicted by the SAM model match as closely as possible the CFD results. In the central region of the fuel pin, the effective thermal conductivity is as high as similar to 60 times the physical fuel salt thermal conductivity. The whole-plant SAM model is then used to simulate an unprotected station blackout transient. The results of this simulation showed that the large negative fuel axial expansion reactivity feedback reduces fission power to similar to 2.4% nominal power. The core is cooled by natural circulation, which removes heat in the core to the emergency heat removal system, and ultimately, to the ambient. However, peak fuel salt and cladding temperatures can potentially reach as high as 1500 K, albeit briefly, if the shutdown mechanism fails to operate.

stable salt reactor; transient simulations; system↗
DOE OSTI2022

Control rod modeling in liquid metal-cooled fast reactors

Control rod modeling in Liquid Metal-cooled Fast Reactors (LMFRs) is important for an accurate simulation, especially in depletion calculations. Recently, control rod search and cusping models have been added to the LUPINE multiphysics fast reactor simulator. LUPINE stands for the 'LMFR Utility for Physics Informed Nuclear Engineering' and is currently being developed at North Carolina State University. LUPINE models the coupled multiphysics effects in LMFRs, including neutronics, thermal hydraulics, thermal expansion, and depletion. The control rod search has been implemented using a Newton-secant search in an inexact-Newton iteration and the cusping model uses a polynomial technique to correct for control rod cusping. The control rod cusping and search models were demonstrated by modeling the Advanced Burner Reactor (ABR) MET-1000 Sodium-cooled Fast Reactor (SFR) and a long-life Lead-cooled Fast Reactor (LFR) based on a Westinghouse Electric Company, LLC (WEC) design. A differential control rod worth curve was calculated for both reactor models to demonstrate the control rod cusping model. The SFR and LFR models were used to demonstrate the importance of modeling control rod movement during depletion calculations and the adverse effect of control rods on cycle length is demonstrated. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Development of Whole System Digital Twins for Advanced Reactors: Leveraging Graph Neural Networks and SAM Simulations

Here, in this work, we introduce a novel method to develop whole system digital twins (DTs) for advanced nuclear reactors. This method treats a complex reactor system as a heterogeneous graph: with the system components as different types of graph nodes and their physical interconnections as edges. Based on the heterogeneous graph, a graph neural network combining graph convolution and temporal node attention is developed as the DT, facilitating a comprehensive understanding of the system's dynamic behavior. By utilizing the System Analysis Module (SAM) code for simulating various operational transients, we develop a graph-based database that trains the DT. This DT is characterized by two primary functions: It can infer the entire system's status using sparse node information, and it can predict the progress of transients based on current and historical system information. Our approach is validated through case studies on the Experimental Breeder Reactor II (EBR-II) system and a generic Fluoride-salt-cooled High-temperature Reactor (gFHR), demonstrating the DT's accuracy in forecasting operational transients. The DT's rapid computation capabilities enhance its potential for supporting advanced reactor operations, offering benefits in intelligent simulation, autonomous control, and anomaly detection, paving the way for improved safety analysis and intelligent component health management for advanced reactor systems and reducing their operations and maintenance cost.

EBR-II↗
DOE OSTI2021

Development and Performance Simulations of the MPACT-AGREE Code Coupling Interface for MAGNOX Reactors

MPACT is a whole-core 3D neutron transport code jointly developed by Oak Ridge National Laboratory and the University of Michigan to perform high-fidelity light water reactor (LWR) analysis. Its capability was recently extended for gas-cooled reactor analysis by coupling to the thermo-fluids code AGREE for thermal feedback. This paper presents the MPACT-AGREE code coupling for the simulation of MAGNOX-type gas-cooled graphite-moderated reactors. A coupling interface has been developed, along with methods to accurately simulate MAGNOX type reactors. The coupling mechanics were tested using two problems derived from the Calder Hall reactor. Simulations of these models demonstrated that the codes have been successfully coupled and can provide reasonable results in a tractable simulation time.

Gentry, Cole↗