Search NASA⌕ Search

SEARCH · Search NASA

Results for “FLUID FUEL REACTORS”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Distributed Temperature Sensing within a 19-pin Wire-Wrapped Fuel Bundle Model

It has always been difficult to measure local temperature and flow velocity within fuel bundle geometries because of their complex and compact layouts. High-resolution mapping of the flow field is especially challenging. High-resolution data is typically obtained using optical instruments but optical access to the interior regions of a bundle is severely limited. Velocity fields within fuel bundle models have been mapped with particle image velocimetry (PIV), an optical technique that employs lasers and cameras. Optical access to the bundle interior is obtained using transparent materials and matching the refractive index of the physical structures with that of the fluid. Testing is necessarily isothermal because refractive indices vary with temperature. However, the ultimate aim of such studies is understanding fluid dynamics and heat transfer within real reactors and so flow field mapping under nonisothermal conditions is of particular interest. Temperature fields can be mapped at high-resolution using laser-induced fluorescence (LIF). However, as an optical technique it is unsuitable for nonisothermal liquid flows. This is due to the same sort of optical distortion that restricts PIV. It is also unsuitable for opaque fluids and setups with restricted optical access. Fiber optic distributed temperature sensing (DTS) offers a high-resolution alternative that does not rely upon imaging. A single sensor can map temperature fields at resolutions far exceeding that achievable with conventional sensors such as thermocouples. Though fiber optic sensors cannot match the spatial resolution and bandwidth of LIF, they are suitable for nonisothermal and opaque flows. And though fibers are physical probes that perturb the flow, they can be strung across complex geometries to provide data in regions lacking the optical access necessary for techniques like LIF. Distributed temperature sensing has the potential to provide fuel bundle temperature data of a type that cannot be obtained with optical methods or conventional sensors. This paper reports on preliminary results of efforts to use fiber optic distributed temperature sensors in a fuel bundle model. We are conducting a feasibility study exploring implementation issues such as strain control in the mounting configuration and signal noise associated with flow-induced vibration. We describe here the setup and present preliminary data from our first phase testing with stagnant flow. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

V and V of the HTR Code Package (HCP) as a comprehensive HTR steady state and transient safety analysis framework - 18664

The HTR code package (HCP) allows for the simulation of several safety-related aspects of a High Temperature Reactor core in a highly integrated manner. HCP currently couples the thermo-fluid dynamics and time dependent neutronics code MGT, the spectrum code TRISHA, the burn up code TNT, the fuel management code SHUFLE and the fission product release code STACY. During its development, state-of-the-art programming techniques and standards are applied. The cross sections in HCP are generated by a 0D- or 1D solver featuring an innovative approach to treat the double heterogeneity of pebble fuel. Also major improvements have been made to optimize the nuclear data library based on ENDF/B-VII. Two advanced fuel management models are provided by the code module SHUFLE that go far beyond the capabilities of existing system codes like VSOP. The source term analysis code module STACY is coupled to HCP providing release rate calculations with a high spatial resolution making use of the nuclide densities provided by TNT. An outstanding new feature of HCP is the possibility to simulate long-term operation scenarios based on OTTO or MEDUL fuel shuffling schemes as well as selected transients in one integrated code package. Even alternating steady state/transient/steady state simulations are possible. So with HCP different fuel strategies and their influence on various kinds of accidents can be examined with one consistent reactor model. This paper provides an overview of the development status of the HCP and reports about selected benchmark results. It is demonstrated that the new system code HCP is capable to replace existing stand-alone codes like VSOP, TINTE/MGT, FRESCO or PANAMA while introducing new features, which so far to our knowledge were not available in the field of HTR safety research. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Fuel Cycle Modeling and Simulation of the Molten Salt Breeder Reactor

The purpose of this work is to perform a neutronic and fuel cycle analysis of the single-fluid Oak Ridge National Laboratory (ORNL) Molten Salt Breeder Reactor (MSBR) as described in ORNL-4541 [1]. Expected design requirements and system outputs were published in this referenced report; the work described herein is focused on building a model of the core, determining the best strategy for modeling the reactor core, and ultimately implementing this modeling/analysis strategy for future research in liquid fueled reactors. Similar analyses for the MSBR using unit cell and full-core models performed using MCNP [2] showed that multiregion models can accurately represent the spectral properties of the multizone MSBR [3,4].

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Advanced Pellet Cladding Interaction Modeling Using the US DOE CASL Fuel Performance Code: Peregrine

The US DOE’s Consortium for Advanced Simulation of LWRs (CASL) program has undertaken an effort to enhance and develop modeling and simulation tools for a virtual reactor application, including high fidelity neutronics, fluid flow/thermal hydraulics, and fuel and material behavior. The fuel performance analysis efforts aim to provide 3-dimensional capabilities for single and multiple rods to assess safety margins and the impact of plant operation and fuel rod design on the fuel thermomechanical- chemical behavior, including Pellet-Cladding Interaction (PCI) failures and CRUD-Induced Localized Corrosion (CILC) failures in PWRs. [1-3] The CASL fuel performance code, Peregrine, is an engineering scale code that is built upon the MOOSE/ELK/FOX computational FEM framework, which is also common to the fuel modeling framework, BISON [4,5]. Peregrine uses both 2-D and 3-D geometric fuel rod representations and contains a materials properties and fuel behavior model library for the UO2 and Zircaloy system common to PWR fuel derived from both open literature sources and the FALCON code [6]. The primary purpose of Peregrine is to accurately calculate the thermal, mechanical, and chemical processes active throughout a single fuel rod during operation in a reactor, for both steady state and off-normal conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Advanced Pellet-Cladding Interaction Modeling using the US DOE CASL Fuel Performance Code: Peregrine

The US DOE’s Consortium for Advanced Simulation of LWRs (CASL) program has undertaken an effort to enhance and develop modeling and simulation tools for a virtual reactor application, including high fidelity neutronics, fluid flow/thermal hydraulics, and fuel and material behavior. The fuel performance analysis efforts aim to provide 3-dimensional capabilities for single and multiple rods to assess safety margins and the impact of plant operation and fuel rod design on the fuel thermo-mechanical-chemical behavior, including Pellet-Cladding Interaction (PCI) failures and CRUD-Induced Localized Corrosion (CILC) failures in PWRs. [1-3] The CASL fuel performance code, Peregrine, is an engineering scale code that is built upon the MOOSE/ELK/FOX computational FEM framework, which is also common to the fuel modeling framework, BISON [4,5]. Peregrine uses both 2-D and 3-D geometric fuel rod representations and contains a materials properties and fuel behavior model library for the UO2 and Zircaloy system common to PWR fuel derived from both open literature sources and the FALCON code [6]. The primary purpose of Peregrine is to accurately calculate the thermal, mechanical, and chemical processes active throughout a single fuel rod during operation in a reactor, for both steady state and off-normal conditions.

CASL, Consortium for Advanced Simulation of LWRs, ↗
DOE OSTI2020

Strategies for thorium fuel cycle transition in the SD-TMSR

Liquid-fueled Molten Salt Reactor (MSR) systems represent advances in safety, economics, and sustainability. The MSR has been designed to operate with a Th/ 233 U fuel cycle with 233 U used as startup fissile material. Since 233 U does not exist in nature, we must examine other available fissile materials to start up these reactor concepts. This work investigates the fuel cycle and neutronics performance of the Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR) with different fissile material loadings at startup: High Assay Low Enriched Uranium (HALEU) (19.79%), Pu mixed with HALEU (19.79%), reactor-grade Pu (a mixture of Pu isotopes chemically extracted from Pressurized Water Reactor (PWR) spent nuclear fuel (SNF) with 33 GW d/tHM burnup), transuranic elements (TRU) from Light Water Reactor (LWR) SNF, and 233 U. The MSR burnup routine provided by SERPENT-2 is used to simulate the online reprocessing and refueling in the SD-TMSR. The effective multiplication factor, fuel salt composition evolution, and net production of 233 U are studied in the present work. Additionally, the neutron spectrum shift during the reactor operation is calculated. The results show that the continuous flow of reactorgrade Pu helps transition to the thorium fuel cycle within a relatively short time (≈ 4.5 years) compared to 26 years for 233 U startup fuel. Finally, using TRU as the initial fuel materials offers the possibility of operating the SD-TMSR for an extended period of time (≈ 40 years) without any external feed of 233 U.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Research reactor plate fuel multi-physics hotspot model for the HFIR

Ongoing development of a multi-physics hotspot model for nuclear research reactor plate fuel is presented with a focus on the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The background of plate fueled research reactors in general and the HFIR in particular are discussed along with the need to convert from high enriched uranium (HEU) to low enriched uranium (LEU) fuel, which drives the need to improve the analysis capability of fuel thermal-fluid performance. HFIR HEU fuel plate fabrication is described to explain the origin of fuel segregations and cladding non-bonds that currently contribute to the hotspot flux peaking factors. Previous efforts at modeling hotspots caused by these defects are reviewed and areas for improvement associated with high resolution multi-physics simulation are identified. Modes of heat transfer through non-bonds, including thermal contact conductance, are examined. Conjugate heat transfer over hotspots is also explored to high resolution up to and through the wall boundary. The challenges associated with multi-physics verification and validation (VV), particularly the requirements beyond VV 20, are presented. An approach to Validation within the limits of existing data for conduction heat transfer through the fuel plate, and convection into the cooling fluid is set forth. Preliminary simulations are presented assessing hot spot modeling approaches for contact conductance, and fluid turbulence modeling within COMSOL. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1991

A nuclear driven metallic vapor MHD coupled with MPD thrusters

Nuclear energy as a source of power for space missions, represents an enabling technology for advanced and ambitious space applications. Nuclear fuel in a gaseous or liquid form has been configured as a promising and practical candidate in this regard. The present study investigates and performs a feasibility analysis of an innovative concept for space power generation and propulsion. The system embodies a conceptual nuclear reactor with an MHD generator and coupled to MPD thrusters. The reactor utilizes liquid uranium in droplet form as fuel and superheated metallic vapor as the working fluid. This ultrahigh temperature vapor core reactor brings forward varied and challenging technical issues, and it has been addressed to in this paper. A parametric study of the conceived system has been performed in a qualitative and quantitative manner. Preliminary results show enough promise for further indepth analysis of this novel system.

Anghaie, Samim↗
DOE OSTI2018

Magnetostrictive Cold Spray Sensor Feasibility Assessment

Ultrasound sensors and inspection systems are frequently used to generate acoustic waves in metal structures capable of detecting and characterizing cracks, pits, erosion, inclusions, weld anomalies, and other material and structural features. One significant problem with piezoelectric transducers is the difficulty to achieve good coupling between the transducer and the surface being examined. This is particularly true in harsh conditions with high temperatures, cyclical hot and cold temperatures, highly radioactive fields found near nuclear reactors or spent nuclear fuel, caustic or corrosive fluids, and other extreme environmental conditions, as well as in long-term monitoring applications where repair or replacement of the sensor is difficult or expensive. Typically, coupling between the surface and the transducer is achieved with water, gel, grease, viscous shear coupling material, or pressure, which might not be possible or appropriate for long-term applications in which the impedance -matching materials wear away, evaporate, or simply stop functioning due to changes in surface conditions. Fluid couplings can evaporate or drain away from the transducer-substrate interface; glue- based couplings may foul or fail and are notoriously unreliable at high temperatures and in radioactive environments. This work explores the behavior of a magnetostrictive cold-spray patch that is metallurgically bonded to a stainless steel inspection target surface, and compares it to the performance of a standard adhesively-bonded ferrous-cobalt magnetostrictive strip solution. Cold-spray is a coating process where 10 –100 micron diameter powdered metal is accelerated to Mach 2 to Mach 3 (2–3 × speed of sound) and impacted on the surface to be coated. Each powder particle forms a kinetic bond with the substrate or other coating particles to produce a metallurgically bonded layer. If the powder is nickel or cobalt with high magnetostrictive coefficients, this surface can serve as the base of a magnetostrictive sensor suitable for crack or pitting-damage inspection and monitoring that is not subject to temporal or environmental degradation.

47 OTHER INSTRUMENTATION↗
DOE OSTI2016

The coupled neutronics and thermo-fluid dynamics design characteristics of the XE-100 200 MW{sub TH} reactor - 18552

The Xe-100 is a high temperature helium-cooled, graphite moderated, pebble bed reactor with a 10 times through, multi-pass (MEDUL=MEhrfacDUrchLauf) fueling scheme. The design, characterized as a Generation IV reactor, is based on proven technology that has been demonstrated a number of times both experimentally and as commercial reactors. This proven foundation ensures high levels of design and technical readiness. In the overview provided of the design it is shown how the excess reactivity is limited by continuous refueling, whilst the selected core power density will ensure passive heat removal. In this way the intrinsic safety design features of a typical GEN IV design are guaranteed. It is shown that no upset event will lead to a significant amount of radiological release. The coupled core neutronic and thermo-fluid dynamics design for the Xe-100 is performed with the VSOP-A system of codes. For the equilibrium core, VSOP results demonstrate that the spherical fuel powers (maximum 1.9 kW) and operational temperatures (<900 deg. C) fall well within the envelope of the design criteria. Adequate reactivity control and long-term, cold shutdown are provided by two separate, diversely-actuated systems, while the overall negative reactivity temperature coefficient is illustrated over the total operational range. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

An Approach to Assessing the Added Mass for a Plate in a Channel

The Global Threat Reduction initiative (GTRI) was established by the National Nuclear Security Administration (NNSA) in 2004. One of the highlights of the program is to convert research and test reactors at home and abroad from Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU) fuels. By switching directly from HEU to LEU, an LEU reactor produces fewer neutrons than an HEU reactor, and in some cases, a direct switch from HEU to LEU results in the reactor being unable to achieve criticality. Under certain circumstances, criticality is achieved and the original neutron fluxes are restored with LEU fuel by adding more fuel elements to the core. Not all reactors, exemplified by the Advanced Test Reactor (ATR) at the Idaho National Lab (INL), have the luxury of simply adding more fuel to the core. In such cases, entirely new designs must be created and tested. The ATR utilizes plate type fuel. During the design and testing of the ATR fuel, flow induced vibrations were identified as potential source for experiment and fuel failure. Flow induced vibrations are problematic to solve because both the frequency of excitation and the modal frequencies of the object need to be determined. For the frequency of the object the frequency is often determined, either theoretically or experimentally without any influence of fluid. When a fluid is introduced, the fundamental frequency of an object can be greatly reduced allowing for resonance between the excitation frequency and the fundamental frequency of the plate. The consequence of such resonance is often failure of the component. Current methods for determining the frequency of a body immersed in a fluid requires the use of multi-physics software in order to determine (if at all possible) the frequency of the object. The work herein provides an alternative means of calculating the frequency of a plate by combining plate frequency theory to determine the added mass on the plate when in the presence of narrow channels. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Fluid structural interaction simulation of the MP-1 plate performance irradiated in the advanced test reactor

Due to the high neutron fluxes they generate and increased heat transfer performance, plate type fuels are used in the U.S. high-performance research reactors. During irradiation, a significant amount of fission energy (i.e., ~200 MeV per atom) is released by the U-235 chain reaction that is carried away by the coolant. Through thermal hydraulic analysis, the coolant’s heat transfer capability is investigated to ensure that the desired fuel temperature can be maintained. At high temperatures, the plate undergoes elastic/plastic deformation, creep, and swelling as a result of both the temperature gradients and the fission gas production within the fuel. These effects are studied via fuel performance analysis. When the plate deformation is small enough that the coolant flow remains relatively unchanged, conducting these two types of analyses independently will suffice. But at high fuel burnups, the swollen plates may encroach into the coolant channels that separate the fuel plates from each other and cause these channels to narrow. Large reductions in channel gap size imperil cooling performance, causing fuel temperatures to rise. Further, if the plate deformation is asymmetric, the fuel centerline will shift toward one side of the channel, causing an uneven reduction in coolability. In addition, a boehmite (oxide) layer will, over time, grow on the plate surface, further obstructing the heat transfer process. To precisely predict fuel temperatures/deformation, a complete coupled analysis that considers coolant flow, heat transfer, oxide growth, elastic/plastic deformation, creep, and swelling is needed; however, this type of analysis method is not available in the literature. To fill the gap, this research developed a fluid structure interaction (FSI) approach to the fuel plate analysis, then successfully applied it to the Mini-Plate (MP)-1 experiment, which was irradiated in the Advanced Test Reactor (ATR) for both one and two cycles. The complete analysis coupled STAR-CCM+, a computational fluid dynamic (CFD) software for calculating flow, with Abaqus, a finite element analysis code for calculating plate deformation. Improvements in the results were found when comparing the fully coupled analysis to the independently conducted analyses but they were not significant due to the miniature size of the plates and the relatively short irradiation time. In the future, the fully coupled approach presented herein will be applied to full-size fuel plates with longer irradiation cycles once additional experiments become available.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Computational Investigation of Fuel Dispersal Phenomena during Large-Break Loss of Coolant Accident in Light-Water Reactors

In the event of cladding rupture, which could occur in light water reactor fuel assemblies during a loss-of-coolant accident (LOCA), fuel particles, along with fission gases, can be expelled into the reactor core from the fractured fuel rod. This expulsion of fragmented fuel particles, referred as fuel dispersal, is the subject of investigation to evaluate the safety implications of increasing fuel burnup in light water reactors, with a specific focus on fuel fragmentation, relocation, and dispersal. Particle trajectories and the resulting mass distribution of the settled particles within the reactor pressure vessel can pose a long-term cooling challenge for the reactor core. The fuel dispersal phenomenon is significantly influenced by the ejection characteristics of the fuel fragments, as well as the size and shape of the cladding rupture and fuel rod depressurization history during LOCA transients. In this study, the transport of fuel particles within a scaled 5 × 5 lattice of a pressurized water reactor rod bundle geometry is modeled through a two-fluid Eulerian framework that treats the gas and solid phases as interpenetrating continua. The required boundary conditions are evaluated from the fuel performance code BISON in a postulated large-break LOCA scenario. The modeling framework considers solid fuel particles as granular matter, interacting with the gaseous dry steam phase and fission gases through the governing interfacial momentum and energy exchange between the gas and solid phases. The simulation results provide the volume fraction of the solids settled on the bottom surface of the fuel bundle, quantifying the deposition within the bundle geometry.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Pressure drop measurements across a helical coil steam generator configuration - 15997

The primary heat transport loop heat exchanger within a Very-High Temperature Reactor (VHTR) is most commonly a Helical Coil Steam Generator (HCSG). This design has been regarded as having increased heat transfer within a smaller design in comparison to other heat exchangers. The fluidic heat transfer properties of the HCSG is still under evaluation in tangent with HTRs due to its complex design. Heat transfer between these layers relies heavily on their relative location and fluidic properties. A model was constructed to study the pressure field on a rod s surface as well as across multiple layers of helically facing tubes. Experiments focused on evaluating the impact relative geometry played on the overall pressure drop of the test section with inlet fluid velocity ranging from 0 to 3 m/s. A study to compare the experimental results with the Gaddis-Gnielinski Correlative theoretical values of pressure drop in tube bundles shows that the complex helical geometry results in pressure drops similar to that of fully developed rod bundle although the test facility is comprised of two half rod helical coils. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Real-Time Elemental and Isotopic Measurements of Molten Salt Systems through Laser-Induced Breakdown Spectroscopy

Molten salt reactors are an emerging advanced nuclear reactor concept in which the fuel is dissolved into the working fluid in the form of a high-temperature molten salt. The complex, mobile, and corrosive nature of this fluid presents a fundamental challenge for analytical measurements. Tracking species throughout the reactor is important for ensuring proper operation. This article presents laser-induced breakdown spectroscopy (LIBS) used to monitor the elemental composition of a molten salt and corresponding hydrogen isotopic shifts in real-time. A NaNO 3 –KNO 3 eutectic salt was saturated with protium and deuterium gases, then the effluent aerosol stream formed using an argon sparging vessel was monitored with LIBS. This modular LIBS system permitted several spectrometers to be used simultaneously to capture high-resolution isotopic shifts and provide broadband elemental coverage. Further, the results exhibit how LIBS can be used to understand salt–gas chemical and physical interactions such as diffusion. Furthermore, LIBS’ broad elemental coverage can provide greater insight into the chemical reactions within the salt vessel such as the formation of water vapor by monitoring hydrogen and oxygen signatures simultaneously. Ultimately, this study demonstrates the analytical possibilities of LIBS for real-time monitoring of isotopes and elemental composition in molten salt systems.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2014

Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear Research Reactors

Turbulent flow coupled with heat transfer is investigated for a High Flux Isotope Reactor (HFIR) fuel plate. The Reynolds Averaged Navier-Stokes Models are used for fluid dynamics and the transfer of heat from a thermal nuclear fuel plate using the Multi-physics code COMSOL. Simulation outcomes are compared with experimental data from the Advanced Neutron Source Reactor Thermal Hydraulic Test Loop. The computational results for the High Flux Isotope Reactor core system provide a more physically accurate simulation of this system by modeling the turbulent flow field in conjunction with the diffusion of thermal energy within the solid and fluid phases of the model domain. Recommendations are made regarding Nusselt number correlations and material properties for future thermal hydraulic modeling efforts

COMSOL LEU HFIR↗
DOE OSTI2016

Graphite Dust Deposition on HTGR Steam Generator: Effect of Particle-wall Impact - 18520

Generated by abrasion of fuel spheres, graphite dust is an important issue for the operation and maintenance of high-temperature gas-cooled reactor (HTGR), since the transport of fission product is closely coupled with graphite dusts. AVR experiments on dust specific radioactivity indicate that graphite dust deposition on steam generator (SG) should be of special interest. Most previous simulations on graphite dust deposition assume a hit-and-stick nature for particles, which may overestimate the deposition rate. In this work we uses a critical velocity model to account for particle-wall interaction. Combined with computational fluid dynamics simulation and discrete phase method, the deposition of graphite dust in the upper 22 layers of SG tubes was studied with particle-wall interaction considered. The result shows that the deposition rate is highly overestimated if hit-and-stick is assumed. For particles larger than 5 {mu}m, the difference can be as large as 3 times. Particle deposition rate reaches maximum at about 3 {mu}m. As for the spatial distribution, few particles deposit on the first layer of tubes due to relative low Stokes number while more depositions occur after entering bundle zone, which indicates that the unsteady wake among tubes could be a major mechanism for the deposition. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Large Eddy Simulation of Convective Heat Transfer in a Random Pebble Bed Using the Spectral Element Method

The development of fluoride-cooled high-temperature reactors has drastically increased the demand for an in-depth understanding of the heat transfer (HT) in packed beds cooled by liquid salts. The complex flow fields and space-dependent porosity found in a pebble bed require a detailed understanding to ensure the proper cooling of the reactor core during normal and accident conditions. As detailed experimental data are complicated to obtain for these configurations, high-fidelity simulation such as large eddy simulation and direct numerical simulation (DNS) can be employed to create a high-resolution heat transfer numerical database that can assist in addressing industrial-driven issues associated with the heat transfer behavior of fluoride-cooled high-temperature reactors. In this paper, we performed a series of large eddy simulation using computational fluid dynamics (CFD) code NekRS to investigate the heat transfer for a bed of 1741 pebbles. Further, the characteristics of the flow, such as average, rms, and time series of velocity and temperature, have been analyzed. Porous media averages have also been performed. The simulation results show a good agreement between non-conjugate heat transfer and conjugate heat transfer. The generated data will be used to benchmark heat transfer modeling methods and local maxima/minima of heat transfer parameters. It will also be used for supporting convective heat transfer quantification for Kairos Power and benchmarking lower fidelity models.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗