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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2011

Pre-test CFD Calculations for a Bypass Flow Standard Problem

The bypass flow in a prismatic high temperature gas-cooled reactor (HTGR) is the flow that occurs between adjacent graphite blocks. Gaps exist between blocks due to variances in their manufacture and installation and because of the expansion and shrinkage of the blocks from heating and irradiation. Although the temperature of fuel compacts and graphite is sensitive to the presence of bypass flow, there is great uncertainty in the level and effects of the bypass flow. The Next Generation Nuclear Plant (NGNP) program at the Idaho National Laboratory has undertaken to produce experimental data of isothermal bypass flow between three adjacent graphite blocks. These data are intended to provide validation for computational fluid dynamic (CFD) analyses of the bypass flow. Such validation data sets are called Standard Problems in the nuclear safety analysis field. Details of the experimental apparatus as well as several pre-test calculations of the bypass flow are provided. Pre-test calculations are useful in examining the nature of the flow and to see if there are any problems associated with the flow and its measurement. The apparatus is designed to be able to provide three different gap widths in the vertical direction (the direction of the normal coolant flow) and two gap widths in the horizontal direction. It is expected that the vertical bypass flow will range from laminar to transitional to turbulent flow for the different gap widths that will be available.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Experimental and Analytic Study on the Core Bypass Flow in a Very High Temperature Reactor

Core bypass flow has been one of key issues in the very high temperature reactor (VHTR) design for securing core thermal margins and achieving target temperatures at the core exit. The bypass flow in a prismatic VHTR core occurs through the control element holes and the radial and axial gaps between the graphite blocks for manufacturing and refueling tolerances. These gaps vary with the core life cycles because of the irradiation swelling/shrinkage characteristic of the graphite blocks such as fuel and reflector blocks, which are main components of a core's structure. Thus, the core bypass flow occurs in a complicated multidimensional way. The accurate prediction of this bypass flow and counter-measures to minimize it are thus of major importance in assuring core thermal margins and securing higher core efficiency. Even with this importance, there has not been much effort in quantifying and accurately modeling the effect of the core bypass flow. The main objectives of this project were to generate experimental data for validating the software to be used to calculate the bypass flow in a prismatic VHTR core, validate thermofluid analysis tools and their model improvements, and identify and assess measures for reducing the bypass flow. To achieve these objectives, tasks were defined to (1) design and construct experiments to generate validation data for software analysis tools, (2) determine the experimental conditions and define the measurement requirements and techniques, (3) generate and analyze the experimental data, (4) validate and improve the thermofluid analysis tools, and (5) identify measures to control the bypass flow and assess its performance in the experiment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Investigation on the Core Bypass Flow in a Very High Temperature Reactor

Uncertainties associated with the core bypass flow are some of the key issues that directly influence the coolant mass flow distribution and magnitude, and thus the operational core temperature profiles, in the very high-temperature reactor (VHTR). Designers will attempt to configure the core geometry so the core cooling flow rate magnitude and distribution conform to the design values. The objective of this project is to study the bypass flow both experimentally and computationally. Researchers will develop experimental data using state-of-the-art particle image velocimetry in a small test facility. The team will attempt to obtain full field temperature distribution using racks of thermocouples. The experimental data are intended to benchmark computational fluid dynamics (CFD) codes by providing detailed information. These experimental data are urgently needed for validation of the CFD codes. The following are the project tasks: (1) Construct a small-scale bench-top experiment to resemble the bypass flow between the graphite blocks, varying parameters to address their impact on bypass flow. Wall roughness of the graphite block walls, spacing between the blocks, and temperature of the blocks are some of the parameters to be tested. (2) Perform CFD to evaluate pre- and post-test calculations and turbulence models, including sensitivity studies to achieve high accuracy. (3) Develop the state-of-the art large eddy simulation (LES) using appropriate subgrid modeling. (4) Develop models to be used in systems thermal hydraulics codes to account and estimate the bypass flows. These computer programs include, among others, RELAP3D, MELCOR, GAMMA, and GAS-NET. Actual core bypass flow rate may vary considerably from the design value. Although the uncertainty of the bypass flow rate is not known, some sources have stated that the bypass flow rates in the Fort St. Vrain reactor were between 8 and 25 percent of the total reactor mass flow rate. If bypass flow rates are on the high side, the quantity of cooling flow through the core may be considerably less than the nominal design value, causing some regions of the core to operate at temperatures in excess of the design values. These effects are postulated to lead to localized hot regions in the core that must be considered when evaluating the VHTR operational and accident scenarios.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Bypass flow computations on the LOFA transient in a VHTR

Bypass flow in the prismatic gas-cooled very high temperature reactor (VHTR) is not intentionally designed to occur, but is present in the gaps between graphite blocks. Previous studies of the bypass flow in the core indicated that the cooling provided by flow in the bypass gaps had a significant effect on temperature and flow distributions for normal operating conditions. However, the flow and heat transports in the core are changed significantly after a Loss of Flow Accident (LOFA). This study aims to study the effect and role of the bypass flow after a LOFA in terms of the temperature and flow distributions and for the heat transport out of the core by natural convection of the coolant for a 1/12 symmetric section of the active core which is composed of images and mirror images of two sub-region models. The two sub-region models, 9 x 1/12 and 15 x 1/12 symmetric sectors of the active core, are employed as the CFD flow models using computational grid systems of 70.2 million and 117 million nodes, respectively. It is concluded that the effect of bypass flow is significant for the initial conditions and the beginning of LOFA, but the bypass flow has little effect after a long period of time in the transient computation of natural circulation.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2011

Bypass Flow Study

Fluid dynamics experiments are being conducted in the matched-index-of-refraction flow system at the Idaho National Laboratory to develop benchmark databases. The data will be used to assess computational fluid dynamics solutions of the momentum equations, scalar mixing, and turbulence models for the flow ratios between coolant channels and bypass gaps in the interstitial regions of typical prismatic standard fuel element or upper reflector block geometries of typical modular high-temperature gas-cooled reactors in the limiting case of negligible buoyancy and constant fluid properties. The experiments use particle image velocimetry to measure the velocity fields that will populate the bypass flow study database. The matched-index-of-refraction modular high-temperature gas-cooled reactor bypass flow experiment measures flow velocity fields in the coolant channels and interstitial gaps between typical prismatic block standard fuel elements (SFEs) or upper reflector blocks.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Active bypass flow control for a seal in a gas turbine engine

An active bypass flow control system for controlling bypass compressed air based upon leakage flow of compressed air flowing past an outer balance seal between a stator and rotor of a first stage of a gas turbine in a gas turbine engine is disclosed. The active bypass flow control system is an adjustable system in which one or more metering devices may be used to control the flow of bypass compressed air as the flow of compressed air past the outer balance seal changes over time as the outer balance seal between the rim cavity and the cooling cavity wears. In at least one embodiment, the metering device may include a valve formed from one or more pins movable between open and closed positions in which the one pin at least partially bisects the bypass channel to regulate flow.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2017

Active bypass flow control for a seal in a gas turbine engine

An active bypass flow control system for controlling bypass compressed air based upon leakage flow of compressed air flowing past an outer balance seal between a stator and rotor of a first stage of a gas turbine in a gas turbine engine is disclosed. The active bypass flow control system is an adjustable system in which one or more metering devices may be used to control the flow of bypass compressed air as the flow of compressed air past the outer balance seal changes over time as the outer balance seal between the rim cavity and the cooling cavity wears In at least one embodiment, the metering device may include an annular ring having at least one metering orifice extending therethrough, whereby alignment of the metering orifice with the outlet may be adjustable to change a cross-sectional area of an opening of aligned portions of the outlet and the metering orifice.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2016

Effects of bypass flow and power change and deviation on performance of thermal mixing structure of HTR-PM - 18424

The temperature of bypass flow not flowing through the reactor core is obviously lower than that of the main coolant helium in a High Temperature Gas-cooled Reactor (HTGR). In addition, change and deviation of reactor power because of power regulation will introduce higher radial temperature difference of coolant helium from reactor to steam generator. In order to investigate the effects of bypass flow and power change and deviation on radial temperature difference of the helium flow entering the steam generator, model experiments and simulation calculations are proposed to evaluate the mixing performance of the thermal mixing structure at HTR-PM reactor core outlet. The experiments on Test Facility Hot Gas Mixing (TF-HGM) are carried out to observe the influences of bypass flow and reactor power deviation on mixing performance. In addition, some simulation calculations are conducted for the main mixing structure of the TF-HGM facility and the thermal mixing structure of HTR-PM. According to the results of experiments and simulation calculations, the maximum radial temperature difference at the outlet of the mixing structure fulfils the requirement by steam generator, less than 30 deg. C ({+-}15 deg. C), under all considered conditions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Active magnetic regenerative liquefier using process gas pre-cooling from bypass flow of heat transfer fluid

A process for liquefying a process gas that includes: introducing a heat transfer fluid into an active magnetic regenerative refrigerator apparatus that includes a low magnetic or demagnetized field section; continuously diverting a bypass portion of the heat transfer fluid from a cold side of the low magnetic or demagnetized field section into a bypass flow heat exchanger at a first cold inlet temperature; and continuously introducing the process gas into the bypass flow heat exchanger at a first hot inlet temperature and discharging the process gas or liquid from the bypass flow heat exchanger at a first cold exit temperature; wherein the temperature difference between bypass heat transfer first cold inlet temperature and the process gas first cold exit temperature is 1 to 5 K.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2012

Bypass Flow Computations using a One-Twelfth Symmetric Sector For Normal Operation in a 350 MWth VHTR

Significant uncertainty exists about the effects of bypass flow in a prismatic gas-cooled very high temperature reactor (VHTR). Bypass flow is the flow in the gaps between prismatic graphite blocks in the core. The gaps are present because of variations in their construction, imperfect installation and expansion and shrinkage from thermal heating and neutron fluence. Calculations are performed using computational fluid dynamics (CFD) for flow of the helium coolant in the gap and coolant channels along with conjugate heat generation and heat transfer in the fuel compacts and graphite. A commercial CFD code is used for all of the computations. A one-twelfth sector of a standard hexagonal block column is used for the CFD model because of its symmetry. Various scenarios are computed by varying the gap width from zero to 5 mm, varying the total heat generation rate to examine average and peak radial generation rates and variation of the graphite block geometry to account for the effects of shrinkage caused by irradiation. The calculations are for a 350 MWth prismatic reactor. It is shown that the effect of increasing gap width, while maintaining the same total mass flow rate, causes increased maximum fuel temperature while providing significant cooling to the near-gap region. The maximum outlet coolant temperature variation is increased by the presence of gap flow and also by an increase in total heat generation with a gap present. The effect of block shrinkage is actually to decrease maximum fuel temperature compared to a similar reference case.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Bypass Flow Resistance in Prismatic Gas-Cooled Nuclear Reactors

Available computational fluid dynamics (CFD) predictions of pressure distributions in the vertical bypass flow between blocks in a prismatic gas-cooled reactor (GCR) have been analyzed to deduce apparent friction factors and loss coefficients for systems and network codes. We performed calculations for vertical gap spacings "s" of 2, 6 and 10 mm, horizontal gaps between the blocks of two mm and two flow rates, giving a range of gap Reynolds numbers Re Dh of about 40 to 5300. Laminar predictions of the fully-developed friction factor f fd were about three to ten per cent lower than the classical infinitely-wide channel In the entry region, the local apparent friction factor was slightly higher than the classic idealized case but the hydraulic entry length L hy was approximately the same. The per cent reduction in flow resistance was greater than the per cent increase in flow area at the vertical corners of the blocks. The standard k-ϵ model was employed for flows expected to be turbulent. Its predictions of f fd and flow resistance were significantly higher than direct numerical simulations for the classic case; the value of L hy was about thirty gap spacings. Initial quantitative information for entry coefficients and loss coefficients for the expansion-contraction junctions between blocks is also presented. Our study demonstrates how CFD predictions can be employed to provide integral quantities needed in systems and network codes.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Validation of RANS Models for Predicting Bypass Flow in Prismatic VHTRs

In prismatic Very-High Temperature Reactors (VHTRs), gaps occur between neighboring fuel assemblies referred to as bypass gaps. This work aims to measure the flow distribution between the bypass gaps between neighboring assemblies and the coolant channels within the assemblies. Computational fluid dynamics (CFD) was done using Reynolds-Averaged Navier Stokes (RANS) modeling in STAR-CCM+ to assess the ability of RANS models to accurately capture this phenomena. The RANS results were compared to experimental worked performed by Seoul National University (SNU) and Korea Atomic Energy Research Institute (KAERI) for validation. The CFD results showed good comparison with the experimental data available from SNU and flow split was characterized for multiple configurations, inlet mass flow rates, and bypass gap sizes.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Simulations of Bypass Flow in Prismatic VHTRs

In prismatic Very-High Temperature Reactors (VHTRs), gaps occur between neighboring fuel assemblies referred to as bypass gaps. This work aims to measure the flow distribution between the bypass gaps between neighboring assemblies and the coolant channels within the assemblies. Computational fluid dynamics (CFD) was done using Reynolds-Averaged Navier Stokes (RANS) modeling in STAR-CCM+ to assess the ability of RANS models to accurately capture this phenomena. The RANS results were compared to experimental worked performed by Seoul National University (SNU) and Korea Atomic Energy Research Institute (KAERI) for validation. The CFD results showed good comparison with the experimental data available from SNU and flow split was characterized for multiple configurations, inlet mass flow rates, and bypass gap sizes.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Bypass Flow Model Implementation for VERA BWR

The Virtual Environment for Reactor Applications (VERA) has been developed by the Consortium for Advanced Simulation of LWRs (CASL) over the past 10 years to address difficult problems facing commercial light-water reactors (LWRs). VERA comprises several high-fidelity physics packages, each of which models a different aspect of the reactor. These individual pieces are then linked to provide a fully coupled multiphysics solution for various LWR analyses.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Exhaust bypass flow control for exhaust heat recovery

An exhaust system for an engine comprises an exhaust heat recovery apparatus configured to receive exhaust gas from the engine and comprises a first flow passage in fluid communication with the exhaust gas and a second flow passage in fluid communication with the exhaust gas. A heat exchanger/energy recovery unit is disposed in the second flow passage and has a working fluid circulating therethrough for exchange of heat from the exhaust gas to the working fluid. A control valve is disposed downstream of the first and the second flow passages in a low temperature region of the exhaust heat recovery apparatus to direct exhaust gas through the first flow passage or the second flow passage.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗