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At least 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Fuel injection assembly for a turbomachine

Fuel injection assemblies and combustors are provided. A fuel injection assembly includes a fuel injector having a first annular wall that about an axial centerline and extends from a primary inlet disposed at a first end to a primary outlet disposed at a second end. A second annular wall surrounds the first annular wall. A fuel plenum is defined between the first annular wall and the second annular wall. A fuel duct extends from a fuel outlet defined in the second annular wall to a fuel inlet. wherein the fuel duct is in fluid communication with the fuel plenum. The fuel duct includes a polygonal segment and a cylindrical inlet segment. The polygonal segment extends from the fuel outlet to the cylindrical inlet segment.

Griffis, Clay Thomas↗
DOE OSTI2012

Parametric analysis of thermal stratification during the Monju turbine trip test

CFD-based simulation techniques are evaluated using a simplified symmetric Monju model to study multi-dimensional mixing and heat transfer in the upper plenum during a turbine trip test. When the test starts and core outlet temperatures drop due to reactor shutdown, the cooler sodium is trapped near the bottom of the vessel and the hotter (less dense) primary sodium at the higher elevations stays largely stagnant for an extended period of time inhibiting natural circulation. However, the secondary flow through a set of holes on the inner barrel bypasses the thermally stratified region as a shorter path to the intermediate heat exchanger and improves the natural circulation flow rate to cool the core. The calculations with strict adherence to benchmark specifications predict a much shorter duration for thermal stratification in the upper plenum than the experimental data indicates. In this paper, the results of a parametric analysis are presented to address this discrepancy. Specifically, the role of the holes on the inner barrel is reassessed in terms of their ability to provide larger by-pass flow. Assuming inner barrel holes with rounded edge produces results more consistent with the experiments. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Progress in developing laser based post irradiation examination infrastructure

To be able to understand the performance of reactor fuels and materials, irradiated materials must be characterized effectively and efficiently in high rad environments. This characterization work is performed remotely and in environments generally hostile to instrumentation. Laser based characterization techniques provide the ability to operate remotely and robustly in a hot-cell environment. Laser based instrumentation also can provide high spatial resolution suitable for scanning and imaging large areas. The INL is currently developing three laser based Post Irradiation Examination (PIE) stations for the Hot Fuel Examination Facility (HFEF) at the INL. These laser based systems will characterize irradiated materials and fuels. The characterization systems are the following: -) laser shock, -) laser based ultrasonic C-scan system and -) gas assay, sample and recharge system (GASR, up-grade to an existing system). The laser shock technique will characterize material properties and failure loads/mechanisms in various materials such as Light Water Reactor (LWR) fuel, plate fuel, and next generation fuel forms in high radiation areas. The laser shock-technique induces large amplitude shock waves to mechanically characterize interfaces such as the fuel-clad bond. As part of the laser shock system, a laser-based ultrasonic C-scan system will be used to detect and characterize de bonding caused by the application of the laser shock as well as performing classical non-destructive evaluation testing and imaging functions such as microstructure characterization, flaw detection and dimensional metrology. A spectroscopy capability can be added later as the salient components to perform Light Induced Breakdown Spectroscopy will be in place. Additional fiber ports in the feed-through will also enable future spectroscopic techniques to be efficiently installed and operated. The purpose of the GASR is to measure the pressure/volume of the plenum of an irradiated fuel element/capsule and obtain fission gas samples for analysis. The study of pressure and volume in the plenum of an irradiated fuel element/capsule and the analysis of fission gases released is important to understanding the performance of reactor fuels and materials. The main function of the laser in this application is to puncture the fuel rod or plate to allow the fission gas to escape and if necessary to weld the hole close. This system may also be used to measure the pressure/volume of other components (such as control blades) and obtain gas samples from these components for analysis. The GASR upgrade will match the functionality of the current system and provide these important functions on fuels made from current and future engineered materials. The GASR station will also have the inherent capability to perform cutting welding and joining functions within a hot-cell. The extra fiber ports built into the hot cell feed-through will also enable the efficient installation of future laser processing work benches. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Development of Inspection Tools for the AY-102 Double-Shell Tank at the Hanford DOE Site - 16383

Florida International University (FIU) has been developing inspection tools to assist in identifying the location of the leak in the primary tank of AY-102. This effort has led to the development of a magnetic miniature rover and a pneumatic pipe crawler. The rover will travel through the refractory cooling channels and the crawler will travel through the air supply lines to the central plenum underneath the primary tank. Both tools will provide visual feedback and potentially identify the location of the leak. The magnetic miniature tool is a remote controlled rover with four wheels driven by independent micro DC motors. The rover will have to navigate through over 40 feet of channel that vary in width from 1.5 to 3.0 inches with four 90 deg. turns to reach the center of the tank. To avoid debris in the channel, the device will travel upside down magnetically attached to the bottom of the primary tank. The pneumatic pipe crawler is a worm type robot with a modular design, composed of interchangeable cylindrical modules connected with flexible links. The design is an evolution of previous peristaltic crawlers developed at FIU, and uses pneumatic actuators to emulate the contractions of the peristaltic movement. The crawler must travel approximately 100 feet through three and four inch piping and traverse through vertical risers, elbows and reducers to reach the central plenum. Prototypes for both the rover and the crawler have been designed and manufactured. Initial bench scale testing has validated the design concepts, demonstrating their ability to navigate through complex geometries. Future work will include full scale testing and evaluation of the systems' reliability. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Engineering Scale Testing of Robotic Inspection Tools for Double Shell Tanks at Hanford - 18349

Waste found in the annulus of the AY-102 double-shell tank (DST) at Hanford has prompted the need for the development of inspection tools that are able to access confined spaces around and underneath the tanks. During the last year, Florida International University (FIU) has been developing inspection tools that are capable of accessing the tank floors through the air supply line, the air refractory channels and the drain lines. This effort has led to the development of a pneumatic pipe crawler and a miniature rover. In this paper, updates to the inspection tool designs and the full scale testing of both systems are presented. The tools provide visual feedback for the operator, carrying with them onboard cameras and light sources. The images are streamed back live via a tether, which also provides power to the devices. To increase the sensing capabilities of both tools, temperature, humidity and radiation sensor modules are being integrated onto the developed systems, and can be interchanged depending on the inspection objectives. The miniature rover measures 7 cm in length and 3.8 cm in width, and travels upside down on the bottom of the tank to avoid debris accumulated in the channels. The video is also being processed real-time to provide control of the unit semi-autonomously. The pipe crawler measures approximately 76 cm in length and 6.4 cm in diameter, and travels using pneumatic actuators that emulate the peristaltic movement contractions typical of worm type robots. A full-scale sectional mock-up of the DST foundation, refractory system and annulus is currently being developed as a test-bed for the testing of the inspection tools. The intention of the test-bed is to demonstrate the viability of using both inspection tools in configurations similar to the actual tank. The test-bed includes a 2.43 m wide section of both tank liners, refractory channels and concrete foundation with drain slots. The test-bed is designed to be modular so that various plate thicknesses can be easily utilized and different refractory configurations can be incorporated. Currently, the refractory channels are of the tree type structure with turns that are approximately 30 degrees. Both inspection tools were demonstrated in the mock-up facility. The miniature rover ran successfully through the drain slots from the annulus to central plenum and had no issues making the 90-degree turns. The peristaltic crawler was able to navigate through the 7.62 cm air supply line from the annulus to the central plenum. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

CFD Predictions of Severe Accident Natural Circulation Flows in a Combustion Engineering PWR

Computational fluid dynamics (CFD) is used to predict natural circulation flows between a reactor vessel upper plenum and a steam generator during specific severe accidents in a pressurized-water reactor (PWR). The predictions are based on a model for a simplified Combustion Engineering (CE) design. This work is part of a larger study to investigate the potential for severe accident induced steam generator (SG) tube failures in PWRs being conducted by the U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research. The CFD model builds upon the lessons learned from previous studies of a generic Westinghouse PWR under similar conditions. The CE SG is significantly larger than the Westinghouse design and geometric differences are expected to reduce the mixing and increase the potential thermal challenge to the SG tubes. The CFD predictions are utilized to improve the basis for simplified flow and mixing models applied in one-dimensional severe accident codes. Parameters are determined to support system code modeling of hot leg and SG inlet plenum mixing, hot leg and SG tube bundle flow rates, and tube inlet temperatures. This updated modeling uses boundary conditions obtained from system analysis code predictions of severe accident behavior in a CE PWR. Model refinements include an improved tube bundle representation and improvements to the thermal boundary conditions. Results indicate less mixing and higher tube inlet temperatures for the CE design compared to previous Westinghouse studies. (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Spectral Analysis of LDV Measurements inside of the TAMU Air-Cooled Reactor Cavity Cooling System - 18688

Spectral analysis was applied to the LDV measurements inside of the TAMU air-cooled RCCS facility at two different locations to both evaluate the feasibility of spectral methods for this experiment and gain a first insight into the flow behavior using a Fourier representation. Application of the wavelet transform showed that there are patterns in the flow signified by repeating frequencies and flow structures. Although the sampling rate of 7.92 Hz at the data point centered at the bottom center of riser 4 was lower than the higher frequency sampling rate of 57.92 Hz at the top data point of the upper plenum, the wavelet transforms of the higher frequency data point showed that the dominant frequencies were bounded within 4 Hz. Application of wavelet analysis is limited by the size of the interrogation window but may be modified according to the frequency range of interest by adjusting the size of the time width for the undilated Morlet mother wavelet. Spectral analysis can be applied to future experiments to understand uncharacterized flow behavior such as flow reversal inside of the upper plenum and for benchmarking future CFD simulations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Sensitivity study of PWR upper vessel flow and phenomena using RELAP5-3D

During normal pressurized water reactor (PWR) operation, water enters the reactor pressure vessel (RPV) from the cold leg and exits through the hot leg. While most of this water flows through the downcomer and reactor core, a number of other flow paths exist, including the core baffle/bypass and the upper plenum spray. The upper plenum sprays normally account for a small fraction (less than 10%) of the total inlet flow, and provide cooling to the upper vessel dome and instrumentation/control rod channels. In the event of a loss-of-coolant accident (LOCA), especially one with core blockage, this alternative flow path may play a key role in maintaining core coolability. Because of the possible safety implications of the upper vessel flow, it is important to ensure the models used to simulate this region of the vessel are accurate. In this study, the model was altered to better reflect the internal flow path of the upper head and this was compared to the reference model. The simulation of interest is a hot leg double-ended guillotine (HL DEG) break followed later by a full blockage of the core and core bypass. In this scenario, the coolant can only reach the core through the upper spray bypass, so its impact on core coolability can be clearly seen. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Mixed effects associated with air ingress in graphite channels

The next generation of nuclear power plants, such as Modular High Temperature Gas-Cooled Reactors (MHTGR), have passive safety design features to overcome transients during off-normal operation but it is unproven whether their safety systems will perform as per design requirements. The design and construction of a facility, to investigate the passive safety response of the complete reactor system will require significant resources. Therefore, a practical approach is to improve the understanding of this multi-physics or mixed effect problem in laboratory scale set-ups and computational models. The HTGR related separate effects experiments at academic institutions and other laboratories have been developed to provide validation data and complement computational efforts to identify critical issues related to passive safety design. There are several separate effect experimental studies currently under investigation or recently conducted by several academic institutions in U.S and throughout the world. The High Temperature Test Facility (HTTF) has been constructed at Oregon State University to conduct integral thermal-hydraulics experiments for the development of HTGRs. For Depressurized Conduction Cooling (DCC) experiments in HTTF, a mixture of nitrogen and helium gas is used to study internal natural convection up to 650?C. Similarly many other experimental studies have emphasized on HTGR core and plenum-to-plenum thermal-hydraulic behavior but with inert gases such as nitrogen or non-oxidizing solid substitutes for graphite. There is limited experimental data of mixed oxidation effects upon air ingress in nuclear grade graphite leading changes to thermal properties, heat generation due to chemical reactions and natural convection of air-helium mixture inside core at high temperatures (above 900?C). This work describes a unique experimental set-up developed by authors to conduct these mixed effect experiments. The current work is equally relevant to the designs well known by other names or geometries such as Very High Temperature Reactors (VHTRs) and Pebble Bed Modular Reactors (PBMRs). A common feature of HTGRs with passive safety design is that under certain accident conditions the decay heat is dissipated passively from the core by radiation/conduction/convection heat transport to the surrounding environment. Passive decay heat removal is one of the prime passive safety features of HTGRs and the heat conduction through the internal core region, graphite matrix, to the periphery is one of the major mechanisms of passive safety in HTGRs. The conduction and radiation heat transfer provide long-term heat removal but, during a Depressurized Loss of Forced Coolant (DLOFC) accident, the anticipated break in the coolant system will allow air ingress into the hot core region and can lead to graphite oxidation, which will negatively impact the core thermal behavior. In addition, air will also impact natural circulation of Helium and thus diminish the heat removal rate, which can further increase the graphite temperature. Therefore, both air ingress can eventually give rise to thermal instabilities and then to possible situations for breaching the safety envelope of TRISO fuel. This research summary is organized to present the background on oxidation studies, description of experimental setup and discussion of results in the following sections. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Development of TDLAS for Water Vapor Concentration Measurement during VHTR Steam Ingress

Very-high-temperature reactors (VHTR) utilize a graphite-moderated, helium-cooled design that provides higher efficiencies and a low-carbon source of process heat (>700 deg. C) for many industrial applications. Safety concerns have led to extensive accident condition investigations. Steam generator heat-exchanger tube rupture accidents (SGTR) can lead to increased reactor core steam density, which causes neutron slowing-down and chemical corrosion of the graphite fuel elements and structural columns. It is important to be able to computationally model the propagation of steam throughout the lower plenum and nuclear reactor core, but experimental data is needed to validate such models. The primary objective of this project is to instrument a scaled down experiment of a VHTR designed by Ohio State University (OSU), with tunable diode laser absorption spectroscopy (TDLAS) diagnostics to measure steam concentration changes and map how water vapor propagates through the test facility's lower plenum and core. The basic principle of absorption spectroscopy is based on a molecular species ability to absorb photons of specific energies. A collimated laser beam of known energy is passed through a test section containing a sample gas. When a transition of the absorbing gas species resonates with the laser frequency, a certain amount of energy will be absorbed based on the rovibrational characteristics of the molecule. The intensity of the transmitted laser light can be measured by a detector after exiting the test section. The Beer-Lambert Law of absorbance is used to recover this information relating beam attenuation to the system dimensions, gas composition, pressure, and temperature. The line-strength is a function of temperature alone and relates the fraction of molecules in a given rovibrational state through the partition function. The partition function describes the statistical properties of a system in thermodynamic equilibrium where the states obey a Boltzmann distribution. Thus, the number density of molecules in a particular state can be determined from the total internal partition function at a particular temperature. The lineshape function of a given absorption transition is derived from the physical interaction between the absorbing species and is dependent on temperature and pressure. Random thermal motion of absorbing molecules result in broadening of the linewidth, called Doppler broadening, which follows a Gaussian distribution. Molecular collisions result in broadening and shifting of the linewidth, called pressure broadening, which follows a Lorentzian distribution. The overall lineshape results in a Voigt profile, which is a convolution of the Gaussian and Lorentzian distributions. The Voigt profile does not have a simple analytical form and is most often calculated numerically or approximated by complex functions. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Developing a Nuclear Quality Assurance Compliant Design Methodology for Neutronic Analysis of Xe-100 Design

The primary objective of this work is to develop a design methodology compliant with nuclear quality assurance standards for the Xe-100 neutronic design verification studies. To achieve this, a Monte Carlo model of the Xe-100 reactor was constructed using the exclusion principle, transformation technique, and universe-based level specification following Idaho National Laboratory (INL) NQA level-1 compliant standards and an NQA-1 compliant version of MCNP6. The model encompasses the entire reactor core structures, including the upper plenum, core region, and lower plenum sections, along with all sub-components. The active core section was represented using the spectral regions, each comprising a particular fuel composition and temperature averaged over the considered zone, calculated by X-energy using Very Superior Old Programs (VSOP). Additionally, a component-wise temperature map was implemented into the model, not only for the core region but also for the structural components. Temperature-dependent cross-section libraries, along with thermal scattering law libraries, generated using INL NQA-1 compliant version of NJOY21, were utilized for each isotope in the burnt fuel and the structural materials. Furthermore, the volume of each modeled component was estimated using a stochastic approach with the ray tracing method in MCNP and criticality calculations were performed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

RELAP5-3D in the OECD-NEA HTGR Thermal Hydraulics Benchmark

The OECD-NEA HTGR Thermal Hydraulics benchmark uses data from the High Temperature Test Facility for a series of code-to-code and code-to-data exercises aimed at improving the state of knowledge on existing thermal hydraulics modeling and simulation tools for prismatic HTGR applications. This benchmark includes problems representing hot gas mixing in the lower plenum, the depressurized conduction cooldown accident and the pressurized conduction cooldown accident. This presentation discusses the benchmark and RELAP5-3D's role in that benchmark, including as a tool for predicting behavior in the core and as a tool for providing boundary conditions to computational fluid dynamics analysis in the lower plenum.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Summary Report Of The FY25 Computational Fluid Dynamics Verification And Validation Exercises In The Advanced Reactor Technologies - Gas-cooled Reactor Program

Verification and Validation (V&V) of numerical tools is critical for ensuring reasonable predictions during design, safety analysis, and licensing. Recent work in the Advanced Reactor Technologies - Gas-cooled Reactor (ART-GCR) program has focused on V&V of common Computational Fluid Dynamics (CFD) tools that are used within the Untied States. This report presents an update on these CFD V&V activities. These Generation IV Forum (GIF) Very High Temperature Reactor (VHTR) Computational Methods, Validation, and Benchmarking (CMVB) is an international organization dedicated to the verification and validation of High Temperature Gas-Cooled Reactor (HTGR) simulation tools. Participation in the CMVB provides additional value to the V&V activities, as it allows for access to a wider range of data, and provides valuable benchmarking exercises. Three HTGR phenomena are targeted: Reactor Cavity Cooling System (RCCS) performance, core bypass flow, and lower plenum mixing. Simulations of the University of Wisconsin-Madison (UW-Madison) RCCS facilities are performed with Reynolds Averaged Navier-Stokes (RANS) in StarCCM+. Results are compared for both forced and natural convection conditions, with both exhibiting good agreement with experimental measurements. The Idaho National Laboratory (INL) matched index of refraction (MIR) and Korean Atomic Energy Research Institute (KAERI) bypass flow expeirments are used to validation CFD predictions of bypass flow. Simulations are performed with RANS in StarCCM+ and with Large Eddy Simulation (LES) in NekRS. Finally, preliminary simulations of the Institute of Nuclear and New Energy Technology (INET) lower plenum mixing facilities are presented. Initial work has developed models with LES, RANS, and porous media models. These preliminary models are presented and compared to each other to gauge differences in predictions with each of the three methods.

and Benchmarking (CMVB)↗
DOE OSTI2023

Machine learning from RANS and LES to inform coarse grid simulations

Nuclear system thermal hydraulic analysis has historically relied on computationally inexpensive 1D codes. However, such tools are unable to capture multiscale multidimensional effects in large nuclear reactor enclosures. On the other hand, simulations with higher fidelity can be too expensive for such purposes. One of the ways to reduce computational cost is to perform simulations on a coarse grid, which, unfortunately, introduces large discretization errors. In this paper, two high-to-low data-driven approaches are investigated: (1) a coarse grid turbulence model to predict eddy viscosity and (2) correction of errors in coarse grid velocity fields. The approaches aim to reduce grid- and turbulence model-induced errors in coarse grid Reynolds-averaged Navier–Stokes (RANS) simulations. Two sources of high-fidelity data, RANS and large eddy simulations (LES), are explored. To extract the eddy viscosity from the LES data, an inverse optimization problem is solved. However, the LES eddy viscosity is shown to be comparable to the RANS eddy viscosity in terms of error reduction. Therefore, the directly available RANS eddy viscosity was used to develop a coarse grid data-driven turbulence model. Additionally, error correction in velocity is used to reduce the remaining uncertainties and bring the results closer to reality. In conclusion, the performance of the frameworks is demonstrated for a scaled upper plenum of a gas-cooled reactor facility.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Summary Report of the FY24 DOE Contributions to the GIF VHTR CMVB

The Generation-IV Forum (GIF) Very-High-Temperature Reactor-Computational Methods Validation and Benchmark (VHTR-CMVB) initiative, involving organizations from Korea Atomic Energy Research Institute (KAERI) (South Korea), Institute of Nuclear and New Energy Technology of Tsinghua University (INET) (China), U.S. Department of Energy (DOE) (U.S.), Joint Research Centre (JRC) (Europe), and Japan Atomic Energy Agency (JAEA) (Japan), is dedicated to the verification and validation of tools for High-Temperature Gas-Cooled Reactors (HTGRs) analysis, using data shared by Computational Methods Validation and Benchmark (CMVB) signatories. For FY24, the US DOE CMVB has committed to several critical activities. Under WP1, led by the US, the integration of the High Temperature Gas Cooled Reactor - Pebble-Bed Module (HTR-PM) Phenomena Identification and Ranking Table (PIRT) into the comparative PIRT is progressing, with a new draft of the comparison tables issued earlier this year and currently being utilized by INET for their contribution. Neutronic validation efforts under WP3 include the preparation of the burnup analysis benchmark, preliminary calculations, and the development of reference models and results. In WP2, a validation exercise for hot gas mixing in the lower plenum of HTR-PM is in progress, using experimental data from INET (China) to validate modeling approaches. A model of the experimental facility has been developed using StarCCM+, with initial calculations slated for presentation at the GIF CMVB meeting this fall. Another WP2 activity focuses on validating numerical models for air-cooled Reactor Cavity Cooling System (RCCS) with experimental data from the Wisconsin Madison RCCS facility. A high-fidelity model, developed using NEK-RS, is currently being validated with available data from a low power forced convection test. These efforts are aimed at enhancing and confirming the accuracy of HTGR analysis tools, ensuring their alignment with experimental data and regulatory requirements.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Mass flow rate effect on a rotating detonation combustor with an axial air injection

Changing the flow rate of reactants being injected into a rotating detonation combustor (RDC) results in interesting behavior of the system. Prior studies have found that an increase in mass flow rate gradually increases the detonation wave speed before splitting the wave into multiple fronts. Here, the focus of this study is in understanding the physics of such behavior through a combination of experiments and numerical simulations. For this purpose, the axial air inlet-based RDC geometry is used. In the experiments, the wave velocity increased similar to prior studies. Corresponding full-scale simulations show that increase in mass flow rate by increasing pressure of the feed plenums shortens the recovery time of the injectors. This causes a more uniform fuel–air mixture to form prior to the arrival of the detonation wave. As a result, a more ideal detonation is observed, which leads to an increased wave velocity. Details of the detonation structure and variation with mass flow rate are analyzed. The presence and variation of the different deflagration zones in such non-premixed RDCs are also discussed.

42 ENGINEERING↗
DOE OSTI2019

Multiphysics Simulation of Moisture-Graphite Oxidation in MHTGR

A small amount of moisture on the ppm level could present in the primary helium coolant under normal operations of High-Temperature Gas-cooled Reactors (HTGRs). One potential safety issue related to HTGRs is the chronic moisture-graphite oxidation, which could affect the integrity of the fuel blocks and reflectors in the reactor core, and the support columns in the hot plenum. However, it is infeasible to perform chronic moisture-graphite oxidation tests under the prototypic high-pressure, high-temperature conditions over a long period of time that is comparable to the full service time, i.e., 36 months for Modular High Temperature Gas-cooled Reactor (MHTGR). As an alternate, accelerated moisture-graphite oxidation tests at the atmospheric pressure and high moisture concentrations have been performed in the literature. Through these accelerated tests, global reaction rate equations have been proposed, which makes it possible to study the moisture-graphite oxidation numerically. This paper is aimed to establish a multiphysics model that can evaluate the moisture-graphite oxidation under MHTGR normal operation conditions. COMSOL Multiphysics was applied to couple the fluid flow, heat and mass transfer, chemical reaction and material structural changes. To reduce the calculation time, the prototypic three-dimensional (3-D) structures were simplified into a two-dimensional (2-D) simulation domain. The performance of four nuclear grades of graphite (IG-110, 2114, PCEA, and NBG-17) over a 36-month service period was then investigated under the most likely prototypic conditions of MHTGR. The simulation results indicate that most oxidation occurs in the three or four bottom fuel blocks due to their higher temperatures. In general, however, only a thin layer of 1 mm into the graphite will be considerably oxidized even for those three or four bottom fuel blocks.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

An open source fast fluid dynamics model for data center thermal management

Although computational fluid dynamics (CFD) has been widely adopted to improve data center thermal management, the high computational demand limits its applications, such as multivariate optimal design and operation. Fast fluid dynamics (FFD), which has been applied for fast airflow simulation, shows great potential. However, few research applied FFD for optimal design and operation of data center thermal management. This research improves the FFD model for data centers and conducts a comprehensive evaluation and demonstration. First, the FFD model is improved by solving the advection and diffusion equations together using an upwind scheme instead of a semi-Lagrangian advection solver in the conventional FFD model. Second, new features for data centers are added, such as a pressure correction method to simulate plenum airflow and dynamic boundary conditions for IT racks. The new FFD model is first validated with two indoor environment cases and the results show that the new FFD model has slightly better overall prediction accuracy and faster speed compared to the conventional FFD model. It is also observed that both FFD models achieve acceptable accuracy, except for a few localized disparities with experimental data, which might be due to simplified handling of turbulence viscosity near the boundaries. Furthermore, validation with a real data center shows that the FFD model achieves a similar level of accuracy as CFD when compared to the experimental measurements with some level of uncertainties. It is then demonstrated for data center optimal design and operation, which saves 53.4–58.8% of annual energy while still meeting the thermal requirements. In conclusion, with a much faster speed and comparable accuracy compared to CFD, the FFD model parallelized on a graphics processing unit is promising for practical model-based data center early design and operation.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗