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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 649 records · Page 36
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Demonstration of Load-Follow Operations with Standalone BISON through VERA

As part of the Consortium for Advanced Simulation of Light Water Reactors (CASL), the Virtual Environment for Reactor Applications (VERA) is being developed to provide high-fidelity multiphysics simulations of nuclear reactor cores. VERA contains a number of single physics packages that are coupled together. MPACT is the primary deterministic neutron transport solver, CTF provides thermal hydraulics solutions, and BISON - which is built off of the MOOSE framework - is used for fuel performance simulations. Over the past year, a standalone BISON capability has been developed using a one-way coupling between MPACT/CTF coupling calculations and BISON. This has allowed CASL contributors to gain insight into the fuel performance characteristics while a more tightly coupled methodology between MPACT, CTF, and BISON has been under development. The standalone capability has been used to simulate Watts Bar Unit 1 (WBN1) Cycles 1-3 and to screen for quantities of interest to pellet-clad interaction in Cycles 6-7. The simulation and screening results are presented here. CASL recently partnered with Exelon and the University of Illinois at Urbana Champaign (UIUC) to begin simulating some of Exelon's plants, with particular focus on load-follow operations. In such operations, the power output from the plant will vary based on the demand or anticipated demand at the time. For example, demand is much lower late at night than during the day when people are awake, so it operating the plant at lower power in the evenings could be advantageous. Historically, nuclear power generation has been among the cheapest sources of energy, so most plants in the United States have been operated at full power except for outages and various small scale events. However, in the current economic climate, priorities have changed. A number of questions arise when load-follow operations are considered, particularly with respect to VERA's tools and capabilities. Can the coupled neutronics and thermal hydraulics simulations with MPACT and CTF perform well without substantial convergence issues? Can BISON handle the somewhat rapid power changes present with load-follow? Will clad hoop stresses be alarmingly high as a result of these power changes? This work addresses these questions and demonstrates operations to assess feasibility and robustness before Exelon and UIUC proceed with their simulations using VERA. To demonstrate load-follow capability, two test problems were developed. The first is a single rod that experiences nominal operations during Cycle 1, and then it simulates 18 months of representative load-follow operations in Cycle 2. This single-rod case was used to test the robustness of MPACT, CTF, and BISON on a typical rod. The second problem is a quarter-core model based on WBN1, Cycle 3, with a single month of load-follow operations inserted into the middle of the cycle. This provides better understanding of the stresses to the clad during the numerous power changes. Sections of this article are adapted from the original, full technical report on this capability.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

A Novel Workflow for Coupled Simulation of Hydraulic Stimulation with Simultaneous Injection of Proppant

Enhanced Geothermal Systems (EGS) are located in impermeable host rock. Therefore, to achieve promising circulation, hydraulic stimulation is a key factor. Stimulation involves injecting a combination of fluids and proppant particles into existing fractures under high pressure. The effectiveness of stimulation depends on a variety of different factors, including the pressurization scheme and the optimal choice of carrying fluid and proppant particles to be used. For example, high-density particles can settle out of the injected slurry quickly, whereas lighter particles as well as a dense carrying fluid allow for a long transport into the reservoir. A numerical tool is therefore essential to take all these factors into account and to make robust predictions about the performance of the hydraulic stimulation. We present a novel computational application for coupled numerical simulation of hydraulic stimulation with concurrent injection of proppant in a fractured geothermal reservoir. This application is based on the open-source MOOSE framework that has been enriched with a set of equations to account for the circulation of proppant-laden slurries. It accounts for both the fluid and proppant components of the mixture, as well as particle-driven processes that include gravity settling, particle-particle interactions, and strong density and viscosity contrasts. A particular challenge is the mutual coupling with fracture mechanical processes in a propagating fracture using the FE code Defmod. Aperture and fluid pressure on one side and slurry density and viscosity changes on the other side interact and affect each other. Their basic understanding is therefore essential for the success of the stimulation. We compare our application to common numerical benchmark problems for both slurry flow and hydraulic stimulation.

15 GEOTHERMAL ENERGY↗
DOE OSTI2022

Transient analysis of a micro-reactor using the DireWolf code suite

Transient analyses of heat pipe micro-reactors are necessary to ensure that hypothetical accident scenarios do not comprise reactor safety. Due to its small size and reliance on heat-pipes for cooling, the micro-reactor design introduced in this paper is a tightly coupled system which requires multi-physics tools to accurately model transient events. Idaho National Laboratory's DireWolf code suite based on the MOOSE framework is tailor-built to model heat-pipe reactors. This paper demonstrates DireWolf's ability to simulate the coupled thermal-neutronics transient behavior of a heat-pipe micro-reactor. The transient events presented here include an inadvertent rotation of all control drums simultaneously and a sudden complete rotation of a single control drum. A detailed description of each event is provided along with simulation results, including time dependent power and temperature distributions, and discussion. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Multigroup cross section generation capability in GRIFFIN

GRIFFIN is an advanced reactor multiphysics application built on the object-oriented simulation environment (MOOSE) and is jointly developed by Idaho National Laboratory and Argonne National Laboratory. The cross section application programming interface, originally developed for the PROTEUS code, has been integrated into GRIFFIN to prepare cross sections for thermal reactor applications with heterogeneous geometries. Additional improvements have been made by implementing an on-the-fly slowing down method, a double heterogeneity treatment capability, and updating the procedure to generate the fine multigroup library. The cross section preparation capability in GRIFFIN was verified for graphite-moderated TRISO fuel-based reactor benchmark problems: unit-cell problems of VHTR and EMPIRE micro reactor and HTTR assembly problems. Eigenvalues and multigroup cross sections of GRIFFIN agreed very well with those of the continuous-energy Monte Carlo code Serpent2 within 200 pcm in eigenvalue and 2% in cross sections. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Implementation of hybrid finite element method based transport solver in GRIFFIN

A new transport solver option based on the hybrid FEM (HFEM) was implemented in GRIFFIN, the MOOSE-based reactor analysis code, as an effort to support routine core design calculations for advanced reactor applications. The HFEM formulation with P{sub N} (spherical harmonics expansion), akin to the variational nodal method, is effective for solving a spatially homogenized problem with strong transport effect. The residual and Jacobian evaluations of the HFEM weak form were derived and successfully implemented in GRIFFIN, having the diffusion and the PN options available in the new HFEM based transport solver. The performance was tested with the simplified ABTR benchmark problems. The results indicate that the HFEM-based transport solver is a feasible option for solving problems with spatially homogenized and strong streaming by providing superior accuracy with a proper p-refinement. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Transient convective delayed neutron precursor of {sup 235}U for the molten salt reactor experiment

This work investigates the delayed neutron precursors from the fission of {sup 235}U in molten salt reactors. The six delayed neutron groups predict the spatial distribution of the advective transport for radioactive mass transfer in the molten-salt reactor experiment (MSRE). The Mole code, using the framework of the Multiphysics Object-Oriented Simulation Environment (MOOSE), was used to compute the delayed neutron concentration of the whole system. This approach can be used for analysis of reactivity in circulating conditions. The effect of delayed neutron concentrations on the static or dynamic behavior of the system in MSRE was analyzed. Using this approach, the pattern and validation of delayed neutron for a range of parameters of interest are reported and discussed herein. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

COUPLED MULTIPHYSICS PRIMARY LOOP SIMULATIONS OF THE Mk1-FHR IN THE VIRTUAL TEST BED

The Virtual Test Bed (VTB) repository hosts a wide range of challenge problems to showcase modeling and simulation capabilities to support advanced reactor demonstration. A coupled multiphysics model of the Mk1 pebble-bed fluoride-salt-cooled high-temperature reactor (PB-FHR) is presented here. The analysis leverages NEAMS tools (Griffin), SAM, Pronghorn, and the MOOSE heat conduction module) for core neutronics, thermal hydraulics of the core and the primary loop, and multiscale fuel performance simulations. The analysis was entirely created by coupling standalone simulations of the reactor previously available on the VTB. All input files and documentation developed for this example are available on the VTB website: \url{mooseframework.inl.gov/virtual_test_bed/}}. This model was featured in the NRIC Tech Talk presented in December 2021.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Prediction of The TREAT Facility Power and Temperature of Sirius-2c Nuclear Propulsion Fuel Experiment

This paper presents simulation results of the Sirius-2c full power experiment that was performed at the TREAT facility to support experiment modeling and transient analysis of the NASA-sponsored Sirius series of experiments for nuclear thermal propulsion applications. We presents a novel approach to predict the control rod motion and power of the reactor and the temperature of the specimen during transient experiments using the Griffin and Bison packages of the Multiphysics Object Oriented Simulation Environment (MOOSE). First, the model predicts the transient rod motion that results in the desired reactor power shape. Then, the temperature of the specimen is predicted based on the power deposited in the specimen. The prediction results are in good agreement with the measured reactor power and control rod position and over-predicted the specimen temperature which is mainly related to specimen surface emissivities. The reactor reaches a peak power of 80 MW and a specimen temperature of 2500 K and considering the right specimen emissivities the predicted temperature is in an excellent agreement with the measured values. This model can be utilized in helping and supporting the design and optimization a new experiments that are going to be irradiated in the TREAT facility.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Verification and Validation Activities of Molten Salt Reactors Multiphysics Coupling Schemes at Idaho National Laboratory

This paper presents the latest verification and validation activities in molten salt reactor modeling and simulation performed at Idaho National Laboratory. Multiphysics solutions are obtained by coupling the neutronics code Griffin, the thermal hydraulics code Pronghorn, and the system analysis code SAM, under the MOOSE framework. We present various multiphysics coupling schemes with these codes for molten salt reactor problems and provide verification and validation results. First, we present verification test results of the Griffin-Pronghorn coupled scheme for the CNRS benchmark. Then validation test results are presented for the Griffin-SAM coupled scheme for the pump startup and coast down transients of the Molten Salt Reactor Experiment. Finally, the Griffin-Pronghorn-SAM coupled scheme is demonstrated for the Molten Salt Reactor Experiment reactivity insertion transient using a domain-overlapping coupling algorithm between Pronghorn and SAM. The results of these various coupling schemes demonstrate the ability to capture the effect of fuel flow and the various feedback mechanisms important to MSRs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

High Fidelity Simulations of Air-Cooled Reactor Cavity Cooling System

Nuclear energy is increasingly acknowledged as pivotal in the global shift towards cleaner energy solutions. Advanced nuclear technologies, including High Temperature Gas-cooled Reactors (HTGRs), stand out as appealing options among Generation IV reactors due to their high temperature heat output and potential for cogeneration. HTGR designs incorporate passive safety systems, such as the Reactor Cavity Cooling System (RCCS), which utilize natural principles to manage heat dissipation from the reactor pressure vessel (RPV) during accidents or routine shutdowns. Regulatory bodies require thorough validation of safety systems like the RCCS to ensure they meet specified standards. Consequently, there is a pressing need within the industry for advanced simulation tools capable of assessing these systems’ performance accurately. There is a knowladge gap in the literature concerning high-fidelity data for the RCCS, which motivates the focus of this study. This research focuses on a specific RCCS designed for the Modular High-Temperature Gas Reactor developed by General Atomics (GA- MHTGR). Experimental studies on a scaled version of the air-cooled RCCS of GA-MHTGR were conducted by the University of Wisconsin-Madison. This work contributes to a broader initiative aimed at establishing a numerical benchmark based on the UW-Madison experiments. As first step we performed high fidelity simulations of the experimental facility setup, to analyze flow physics in such systems and validate NekRS and the MOOSE heat transfer and radiation modules.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Directed Energy Deposition Process Modeling, Validation, and Process-Informed Optimization

The directed energy deposition (DED) process, one of the most popular additive manufacturing techniques in use today, involves various complex physical mechanisms that are not yet well understood. In this regard, computational tools show promise for elucidating the manufacturing process and enabling nondestructive performance evaluations of manufactured parts. To better control and optimize the DED process?thereby improving the manufactured product? the present work develops and demonstrates a novel artificial intelligence (AI)-based process control and optimization technique. Specifically, a geometry-free thermo-mechanical model with adaptive subdomain construction is developed to accurately capture the material?s thermo-mechanical response under cyclical reheating and high cooling rates [1]. The model is demonstrated and validated with experimental measurements, in light of various geometries and processing parameters. Moreover, based on this thermo-mechanical model, a physics-informed reduced-order model is developed to enable quick predictions of the temperature field at every time step. Furthermore, an AI-based controller is developed that can adapt to the ever-changing system states by automatically adjusting the manufacturing process parameters. This entire work is based on the open-source Multiphysics Object-Oriented Simulation Environment (MOOSE) [2] and its recent integration with Libtorch (the C++ frontend of PyTorch [3]). The combined development of the adaptive material deposition scheme, thermo-mechanical model, associated reduced-order model, and AI-based process controller carries great potential for improving advanced manufacturing processes.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Calibration of a mesoscale tritium transport model for ceramic breeder materials in TMAP8 using experimental data

Due to the scarcity of the long-term external tritium supply, lithium-containing breeder materials are used in the blanket of fusion reactors to produce tritium faster than the deuterium-tritium fusion reaction consumes it. A cellular breeder is a promising material with higher lithium density and thermal conductivity than conventional ceramic pebbles and enhanced tritium extraction with interconnected pores. Although several experimental and modeling efforts have improved our understanding of tritium transport, the exact mechanisms governing tritium release from breeder materials are still unknown. As a result, current models cannot reliably assess the tritium breeding capabilities of cellular breeder materials. This work presents a tritium transport multiphysics model and calibrates it using experimental data of deuterium absorption and desorption from cellular breeder materials at different temperatures and pressures. This model accounts for ceramic and pore diffusion, trapping and detrapping, and several surface reactions at the pore surface. After calibrating the model and performing sensitivity analysis, we discuss pre-dominant mechanisms governing tritium release from cellular breeder materials. The model is part of the Tritium Migration Analysis Program [TMAP8], a multiscale, multiphysics framework for tritium transport based on the finite element multiphysics framework MOOSE. This study demonstrates some of TMAP8’s capabilities and provides insight into the mechanisms governing tritium transport in cellular ceramic breeder materials.

08 HYDROGEN↗
DOE OSTI2022

Sensitivity Study of Multiscale and Phenomenological Elasto-Viscoplastic Grade 91 Material Models for Component-Scale Response

Many advanced nuclear reactor concepts currently being developed are targeting higher operating temperatures relative to the current fleet of light water nuclear reactors, for efficiency gains and other operational considerations. The design of high temperature structural components with reliable long-term operational performance will depend on material models that accurately capture the inelastic deformation mechanisms active in these environments. In this work, we perform a detailed parameter sensitivity analysis of two unified elasto-viscoplastic Grade 91 material models capable of capturing long term high temperature creep deformation. The first model is a phenomelogical material model from the Nuclear Engineering Material Library (NEML) developed at Argonne National Lab. The NEML model parameters and their uncertainty were fit to a range of Grade 91 experimental data using Bayesian Markov Chain Monte Carlo analysis. The second model is a LAROMance data-driven surrogate material model developed at Los Alamos National Lab. The LAROMance model is fit to a large database of responses produced by a mechanistic crystal plasticity based polycrystal model. Parameters for the LAROMance surrogate material model reflect the pedigree of the Grade 91 microstructure. Both material models have been integrated into the Grizzly code, based on the open-source MOOSE multiphysics simulation framework, to simulate both the progression of aging mechanisms and the effects of that aging on nuclear power plant structures. Grizzly is used analyze a three-dimensional Grade 91 piping system to compare the long-term inelastic response predicted by these two fundamentally different models and assess the sensitivity of the material model input parameters on this quantity of interest.

42 ENGINEERING↗
DOE OSTI2023

Modeling of Prismatic High Temperature Reactors in Pronghorn

Pronghorn is a MOOSE based thermal-hydraulics code developed at Idaho National Laboratory (INL) for advanced nuclear reactor analysis. It has been previously applied to model pebble-bed high temperature reactors (HTRs), liquid-metal cooled reactors, and molten salt reactors, among others. This work leverages the coarse-mesh modeling capabilities in Pronghorn to model the Oregon State University (OSU)'s High Temperature Test Facility (HTTF). The HTTF is a 1:4 height scaled-down facility of General Atomics' Modular High Temperature Gas-cooled Reactor (MHTGR). The facility is primarily built to generate data for code and model validation, and does not precisely replicate MHTGR conditions. Nevertheless, it encompasses the main physics associated with MHTGR transients.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Dynamic Modeling and Characterization of Nuclear-grade Graphite

Idaho National Labs serves as the spearhead for many innovative energy solutions to the world's energy crisis. One such solution is the INL's Microreactor which is designed to deploy to extreme/remote environments where other sources of power are either unavailable or unreliable. In order to best design these energy solutions for their operational environments, it is crucial to understand how the design, components, and materials will respond to the environmental conditions. One key material in these innovative designs is a nuclear-grade graphite known as PCEA. This study examines the behavior of PCEA graphite under dynamic loading, similar to that which may occur in extreme environments. The objective is to characterize the dynamic behavior and produce an accurate, reliable constitutive model suitable for use in simulation tools such as INL's MOOSE. Graphite specimens were tested using a Split Hopkinson Pressure Bar (SHPB) to administer the dynamic compressive load. The SHPB was charged at various pressures to produce a range of strain rates on the material in compression. Data was acquired via strain gauges on the SHPB setup, from which stress, strain, and time data were collected. Analysis revealed the stress-strain behavior of the material as well as insights into the material behavior's relationship to strain rate. Further work must continue to characterize the various other dynamic behaviors of the material which will combine to create a substantially trustworthy constitutive model for this grade of nuclear-grade graphite. Ultimately, this will allow for realistic simulation of the material in reactor designs, allowing for prediction of design weaknesses and leading to improved designs for increased resilience, security, and reliability.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

High Performance Heat Pipe Power Transient Testing at SPHERE Facility

Microreactors are being researched, designed, and built at Idaho National Laboratory (INL). Microreactors are small reactors defined at less than 20MW of power. These reactor concepts are also being looked at throughout industry for various applications. An important aspect of these reactor designs is economic feasibility i.e. lower overnight capital cost. The driving factors for implementing microreactors are quick setup and takedown, minimal operators, and the ability to manufacture them readily and to fit in mid-sized containers for transport. A specific area of research to aid in successful integration of these factors within the designs is passive heat removal of the core’s thermal power. Interest in heat pipes to achieve this passive heat removal has been shown across multiple industry partners. Because of this interest, INL has developed a test facility to facilitate experimental tests for sodium filled heat pipes. INL has developed the Single Primary Heat Extraction and Removal Emulator (SPHERE) facility to run experiments on high performance, sodium filled heat pipes. As mentioned above, heat pipes are passive heat transfer devices. Radially, heat pipes are broken up into an outer wall, a small annular gap, a wick structure, and a centerline gap. They function by utilizing latent heat transfer. Heat pipes are traditionally separated into three regions, an evaporator (heat input), an adiabatic region, and finally a condenser region (heat removal). As heat is being applied to the evaporator, the working fluid undergoes a phase change to a vapor. This phase change causes a differential pressure across the axial length of the pipe driving flow down the center gap of the heat pipe. The vapor flows down past the adiabatic region to the condenser where the heat is removed. This heat removal forces the working fluid to phase change back to a liquid. The wick structure is then utilized to drive the flow back towards the evaporator by capillary forces. This backflow is aided by the annular gap. Because this heat transfer mechanism functions with latent heat transfer, the heat pipe is close to isothermal down the axial length. Heat pipes can operate under a wide range of working fluids. Considerations for these working fluids are primarily driven by operating temperatures amongst other important factors based around overall performance. Sodium filled heat pipes operate from 450°C up to 900°C. This temperature range works well for the current microreactor designs. In conjunction with this experimental capability, INL has developed a modeling software to simulate heat pipe physics within reactor cores. This modeling software is called Sockeye and functions under the established INL Multiphysics Object Oriented Simulation Environment (MOOSE). SPHERE also supports Sockeye development by providing the modeling team with experimental data on an array of setups and operating parameters to support validation efforts. A power transient experiment was performed utilizing the SPHERE facility to continue to aid with Sockeye development. The testing followed a proposed test plan to ramp up and down the temperature of the heat pipe. Sockeye models steady state heat pipe operation with high accuracy, the data provided by the power transient testing aims to assist with the validation efforts and further enhance transient modeling capability of the tool [2].

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Modeling Heat Pipes with Non-Condensable Gases

The report showcases advancements in modeling heat pipes, considering both scenarios with and without non-condensable gases. To achieve this, two distinct modeling approaches are juxtaposed. Firstly, the effective conduction model, implemented in the MOOSE-based code Sockeye, is employed. Secondly, a first-of-a-kind two-phase Euler-Euler Computational Fluid Dynamics (CFD) model is developed using the STAR-CCM+ code. Both models undergo validation against experimental data, acknowledging the inherent uncertainties associated with each modeling assumption. Interestingly, the non-tuned CFD model surpasses the performance of the calibrated conduction model for heat pipes operating under both conditions: with and without non-condensable gases. It's worth noting, however, that the CFD models entail significantly longer runtimes compared to the conduction models. Nevertheless, the insights garnered from the CFD model shed invaluable light on the intricate operational dynamics of heat pipes. Future work involves broadening the validation scope of these models and continuing their development to enhance their utility as robust tools for heat pipe design and operational support.

42 ENGINEERING↗