Results for “Boiling curve” — Page 3

NASA NTRS2011

New Technique for Cryogenically Cooling Small Test Articles

Convective heat removal techniques to rapidly cool small test articles to Earth-Moon L2 temperatures of 77 K were accomplished through the use of liquid nitrogen (LN2). By maintaining a selected pressure range on the saturation curve, test articles were cooled below the LN2 boiling point at ambient pressure in less than 30 min. Difficulties in achieving test pressures while maintaining the temperature tolerance necessitated a modification to the original system to include a closed loop conductive cold plate and cryogenic shroud

Rodriquez, Karen M.↗

NASA NTRS2015

Numerical Simulation of Liquid Nitrogen Chilldown of a Vertical Tube

This paper presents the results of a one-dimensional numerical simulation of the transient chilldown of a vertical stainless steel tube with liquid nitrogen. The direction of flow is downward (with gravity) through the tube. Heat transfer correlations for film, transition, and nucleate boiling, as well as critical heat flux, rewetting temperature, and the temperature at the onset of nucleate boiling were used to model the convection to the tube wall. Chilldown curves from the simulations were compared with data from 55 recent liquid nitrogen chilldown experiments. With these new correlations the simulation is able to predict the time to rewetting temperature and time to onset of nucleate boiling to within 25% for mass fluxes ranging from 61.2 to 1150 kg/(sq m s), inlet pressures from 175 to 817 kPa, and subcooled inlet temperatures from 0 to 14 K below the saturation temperature.

GFSSP↗

DOE OSTI2024

Improved Statistics for F-theory Standard Models

Much of the analysis of F-theory-based Standard Models boils down to computing cohomologies of line bundles on matter curves. By varying parameters one can degenerate such matter curves to singular ones, typically with many nodes, where the computation is combinatorial and straightforward. The question remains to relate the (a priori possibly smaller) value on the original curve to the singular one. In this work, we introduce some elementary techniques (pruning trees and removing interior edges) for simplifying the resulting nodal curves to a small collection of terminal ones that can be handled directly. When applied to the QSMs, these techniques yield optimal results in the sense that obtaining more precise answers would require currently unavailable information about the QSM geometries. This provides us with an opportunity to enhance the statistical bounds established in earlier research regarding the absence of vector-like exotics on the quark-doublet curve.

Bies, Martin↗

NASA NTRS1987

Review of nucleation and incipient boiling under pool and forced convection conditions

An overview of liquid-vapor nucleation is given. The result of thermodynamic equilibrium across curved liquid-vapor interfaces is presented. The extension of this to include the interaction with idealizations of surface cavities is made to demonstrate how superheat requirements for nucleation will be affected by surface roughness, flow velocity and buoyancy. Experimental measurements of high liquid superheats and nucleation delay times are presented as examples of homogeneous nucleation. Examples of nucleation and boiling on smooth glass substrates and on metal surfaces with various surface roughnesses are presented.

Merte, Herman, Jr.↗

NASA NTRS1991

Flow boiling with enhancement devices for cold plate coolant channel design

Future space exploration and commercialization will require more efficient heat rejection systems. For the required heat transfer rates, such systems must use advanced heat transfer techniques. Forced two phase flow boiling heat transfer with enhancements falls in this category. However, moderate to high quality two phase systems tend to require higher pressure losses. This report is divided into two major parts: (1) Multidimensional wall temperature measurement and heat transfer enhancement for top heated horizontal channels with flow boiling; and (2) Improved analytical heat transfer data reduction for a single side heated coolant channel. Part 1 summarizes over forty experiments which involve both single phase convection and flow boiling in a horizontal channel heated externally from the top side. Part 2 contains parametric dimensionless curves with parameters such as the coolant channel radius ratio, the Biot number, and the circumferential coordinate.

Boyd, Ronald D.↗

NASA NTRSDate not supplied

Simulations of Cryogenic Line Chilldown with Advanced Sub-Grid Wall Boiling Models

A meso-scale model developed at MIT [2][11] for boiling processes in water was adapted for cryogens and demonstrated for chilldown of propellant transfer lines in both liquid nitrogen and hydrogen. The sub-grid boiling model accurately captures the contributions to heat transfer from the generation of bubble nuclei, growth, and interaction of the bubbles in the microlayer as well as quenching of the boiling surface following bubble departure. It was adapted for cryogenic fluids using thermodynamic scaling concepts taking into account non-dimensional pressures and temperatures that are scaled by the corresponding critical values for the fluid. The boiling model was demonstrated for line chilldown in liquid nitrogen (Darr et al. [6]). The predicted wall temperature at which quenching occurs was close to the test data while the slope of the temperature curve after quenching is initiated, showing a steeper variation than the test data. Simulations were also performed for liquid hydrogen by simulating experiments of Hartwig et al. [7]. Chilldown times in liquid hydrogen are much more rapid due to higher heat transfer in the film boiling regime and accounting for the higher turbulence levels were found to be important. Furthermore, at the much lower fluid temperatures in liquid hydrogen flows, accounting for the variable thermal properties of the solid material is critical and has a dramatic impact on the quench times. The need for additional studies to better understand the evolution of the quench front in liquid hydrogen was noted.

Computational Fluid Dynamics↗

NASA NTRSDate not supplied

Simulations of Cryogenic Line Chilldown with Advanced Sub-Grid Wall Boiling Models

A meso-scale model developed at MIT [2][11] for boiling processes in water was adapted for cryogens and demonstrated for chilldown of propellant transfer lines in both liquid nitrogen and hydrogen. The sub-grid boiling model accurately captures the contributions to heat transfer from the generation of bubble nuclei, growth, and interaction of the bubbles in the microlayer as well as quenching of the boiling surface following bubble departure. It was adapted for cryogenic fluids using thermodynamic scaling concepts taking into account non-dimensional pressures and temperatures that are scaled by the corresponding critical values for the fluid. The boiling model was demonstrated for line chilldown in liquid nitrogen (Darr et al. [6]). The predicted wall temperature at which quenching occurs was close to the test data while the slope of the temperature curve after quenching is initiated, showing a steeper variation than the test data. Simulations were also performed for liquid hydrogen by simulating experiments of Hartwig et al. [7]. Chilldown times in liquid hydrogen are much more rapid due to higher heat transfer in the film boiling regime and accounting for the higher turbulence levels were found to be important. Furthermore, at the much lower fluid temperatures in liquid hydrogen flows, accounting for the variable thermal properties of the solid material is critical and has a dramatic impact on the quench times. The need for additional studies to better understand the evolution of the quench front in liquid hydrogen was noted.

Computational Fluid Dynamics↗

NASA NTRS1993

Vaporization heat transfer of dielectric liquids on enhanced surfaces covered with screen wicks

Experiments were conducted to investigate the vaporization heat transfer characteristics for the dielectric liquid FC-72 on several wicking surfaces which may be used in flat-plate heat pipes. The wicking materials studied included microstructure enhanced surfaces and coarse surfaces covered with screen meshes. Experimental data for q versus deltaT curves and critical heat fluxes were obtained for the two different operating conditions of a heat pipe, evaporation, and shallow pool boiling. When the liquid level was above the heated surface, the height of the liquid level above the surface was varied from 0 to 10 mm. When the liquid level was below the heated surface, the distance from the liquid level to the edge of the surface was adjusted from 0 to 15 mm. Experimental results revealed that for shallow pool boiling when the heated surface was covered with a wire screen mesh, the heat transfer coefficient increased at lower heat fluxes but the critical heat flux (CHF) decreased for all the surfaces tested. In the case of evaporation, both CHF and the heat transfer coefficient increased as the microstructure surfaces were covered with screen meshes.

Gu, C. B.↗

DOE OSTI2025

Development of Large Bore Rabbit Capsules in Support of BWR Cladding Irradiations in HFIR

The High Flux Isotope Reactor (HFIR) is an ideal tool for materials irradiation testing because of its intense steady-state neutron flux. Many programs take advantage of HFIR’s central flux trap for irradiation experiments using capsules, also known as rabbits, to support advanced materials development and reactor design. The facility that makes up the HFIR flux trap has recently undergone a design change that increases the HFIR primary coolant volumetric flow rate by removing restrictions in the system. As a result, the usable cross-sectional area within the facility increased, opening the door to increase the cross-sectional area of the rabbit capsules that fill the facility. This report documents a new large-diameter rabbit housing that increases the usable volume within the rabbit capsule by 22.6%. However, challenges arise with increasing the capsule size, such as establishing a new maximum capsule operating pressure and determining the thermal-hydraulic characteristics. This report addresses those challenges with previously adopted HFIR safety methods. The rupture pressure of the rabbit housings is demonstrated while verifying that capsule swelling during and after rupture will not block coolant flow. Then, a safety factor is applied to ascertain an administrative operating pressure. Additionally, the thermal-hydraulic performance of the HFIR facility filled with large-diameter rabbit capsules is shown to not violate previously determined safety criteria. Next, heat transfer coefficients are determined for use in design calculations. Furthermore, this report gives an example of internal configurations for the new, larger rabbit capsules that use relevant boiling water reactor (BWR) cladding geometry. Finally, this report documents an example thermal design performance for a rabbit capsule containing six gauge-curved tensile tube specimens. The thermal performance gives predicted temperature distributions within the capsule and shows the expected temperature of the passive thermometers for post-irradiation temperature comparisons.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2025

Development of Large Bore Rabbit Capsules in Support of BWR Cladding Irradiations in HFIR

The High Flux Isotope Reactor (HFIR) is an ideal tool for materials irradiation testing because of its intense steady-state neutron flux. Many programs take advantage of HFIR’s central flux trap for irradiation experiments using capsules, also known as rabbits, to support advanced materials development and reactor design. The facility that makes up the HFIR flux trap has recently undergone a design change that increases the HFIR primary coolant volumetric flow rate by removing restrictions in the system. As a result, the usable cross-sectional area within the facility increased, opening the door to increase the cross-sectional area of the rabbit capsules that fill the facility. This report documents a new large-diameter rabbit housing that increases the usable volume within the rabbit capsule by 22.6%. However, challenges arise with increasing the capsule size, such as establishing a new maximum capsule operating pressure and determining the thermal-hydraulic characteristics. This report addresses those challenges with previously adopted HFIR safety methods. The rupture pressure of the rabbit housings is demonstrated while verifying that capsule swelling during and after rupture will not block coolant flow. Then, a safety factor is applied to ascertain an administrative operating pressure. Additionally, the thermal-hydraulic performance of the HFIR facility filled with large-diameter rabbit capsules is shown to not violate previously determined safety criteria. Next, heat transfer coefficients are determined for use in design calculations. Furthermore, this report gives an example of internal configurations for the new, larger rabbit capsules that use relevant boiling water reactor (BWR) cladding geometry. Finally, this report documents an example thermal design performance for a rabbit capsule containing six gauge-curved tensile tube specimens. The thermal performance gives predicted temperature distributions within the capsule and shows the expected temperature of the passive thermometers for post-irradiation temperature comparisons.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

A User’s Guide to the PLTEMP/ANL Code

PLTEMP/ANL V4.4 is a program that obtains a steady-state flow and temperature solution, in geometrical extent, for a nuclear reactor core, or a single fuel assembly, including the effect of manufacturing and modelling uncertainties by the means of hot channel factors. It is based on an evolutionary sequence of codes originally used for calculating plate temperatures, hence “PLTEMP”, developed at Argonne National Laboratory over the past 30 years. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each with its own axial power shape. Depending upon the heat transfer method selected, the temperature solution is effectively 2-dimensional or 3-dimensional. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available to determine safety margins such as onset of nucleate boiling ratio (ONBR), departure from nucleate boiling ratio (DNBR), and onset of flow instability ratio (OFIR). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability for select design and safety parameters is available to greatly reduce the reactor analyst’s time.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Boiling and cavitation caused by transient heat transfer in superfluid helium-4

Superfluid helium-4 (He II) has been widely utilized as a coolant in various scientific and engineering applications due to its superior heat transfer capability. An important parameter required in the design of many He II based cooling systems is the peak heat flux $q^*$, which refers to the threshold heat flux above which boiling spontaneously occurs in He II. Past experimental and numerical studies showed that $q^*$ increases when the heating time $t_h$ is reduced, which leads to an intuitive expectation that very high $q^*$ may be achievable at sufficiently small $t_h$. Knowledge on how $q^*$ actually behaves at small $t_h$ is important for applications such as laser ablation in He II. Furthermore we present a numerical study on the evolution of the thermodynamic state of the He II in front of a planar heater by solving the He II two-fluid equations of motion. For an applied heat flux, we determine the heating time beyond which the He II near the heater transits to the vapor phase. As such, a curve correlating $q^*$ and $t_h$ can be obtained, which nicely reproduces some relevant experimental data. Surprisingly, we find that there exists a critical peak heat flux $q^*_c$, above which boiling occurs nearly instantaneously regardless of $t_h$. We reveal that the boiling in this regime is essentially cavitation caused by the combined effects of the first-sound and the second-sound waves in He II. Based on this physical picture, an analytical model for $q^*_c$ is developed, which reproduces the simulated $q^*_c$ values at various He II bath temperatures and hydrostatic head pressures. This work represents a major progress in our understanding of transient heat transfer in He II.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

DOE OSTI2020

A User’s Guide to the PLTEMP/ANL Code

PLTEMP/ANL V4.3 is a program that obtains a steady-state flow and temperature solution for a nuclear reactor core, or for a single fuel assembly. It is based on an evolutionary sequence of codes originally used for plate temperatures, hence “PLTEMP”, developed at Argonne National Laboratory over several decades. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each with its own axial power shape. Depending upon the heat transfer method selected, the temperature solution is effectively 2-dimensional or 3-dimensional (Appendix XV). It begins with a solution across all coolant channels and fuel plates or tubes within a given fuel assembly, at the entrance to the assembly. The temperature solution is repeated for each axial node along the length of the fuel assembly. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available to determine safety margins such as onset-of-nucleate boiling ratio(ONBR), departure from nucleate boiling ratio (DNBR), and onset of flow instability ratio (OFIR). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability is available (Appendix XII) to greatly reduce the reactor analyst’s time.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

A User’s Guide to the PLTEMP/ANL Code (V.4.3)

PLTEMP/ANL V4.3 is a program that obtains a steady-state flow and temperature solution for a nuclear reactor core, or for a single fuel assembly. It is based on an evolutionary sequence of codes originally used for plate temperatures, hence “PLTEMP”, developed at Argonne National Laboratory over several decades. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each with its own axial power shape. Depending upon the heat transfer method selected, the temperature solution is effectively 2-dimensional or 3-dimensional (Appendix XV). It should be noted that for a 3-dimensional solution the axial power shape is modeled without accounting for axial heat conduction. It begins with a solution across all coolant channels and fuel plates or tubes within a given fuel assembly, at the entrance to the assembly. The temperature solution is repeated for each axial node along the length of the fuel assembly. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available to determine safety margins such as onset-of-nucleate boiling ratio(ONBR), departure from nucleate boiling ratio (DNBR), and onset of flow instability ratio (OFIR). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability is available (Appendix XII) to greatly reduce the reactor analyst’s time.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1976

Cryogenic storage tank thermal analysis

Parametric study discusses relationship between cryogenic boil-off and factors such as tank size, insulation thickness and performance, structural-support heat leaks and use of vapor-cooled shields. Data presented as series of nomographs and curves.

Wright, J. P.↗

DOE OSTI2020

Manufacturing Process Design of a Micro-Scale Liquid-Liquid Extractor and Multi-phase Separator

Advances in process innovation within the chemical process industry are impeded by the cost of producing new innovative components. In this paper, various manufacturing process routes were evaluated for scaling up a micro-scale liquid-liquid extractor and multi-phase separator capable of separating immiscible phases that are used for the extraction of a chemical product into a lower boiling point solvent to make downstream distillation easier. A manufacturing process design methodology was used to evaluate both photochemical machining and binder jetting as a means for producing the micropost arrays within the separator. Cost curves show that for a 0.3 mm deep micropost array, photochemical machining was found to be much cheaper than binder jetting. However, because binder jetting processes enable microposts with over 13 times higher depth-to-width aspect ratios, binder jetting was found to decrease the overall costs of the extractor/separator by more than half for a 500 barrel per day plant, with breakeven as low as 15 barrels per day. These economics were confirmed by building extractor/separator plates and characterizing them dimensionally and functionally. The binder jet plates were found to perform better than photochemical plates leading to the complete separation of the organic phase from water. The utilization of slower sintering cycles combined with surface grinding provided acceptable dimensional tolerances for the binder jet plates. Using pressurized air during depowdering improved the economics for the binder jetting process.

42 ENGINEERING↗

DOE OSTI2023

Sensitivity Analysis of Transient Critical Heat Flux by RIA Under High-Pressure Flow Boiling Conditions in TRTL

A reactivity-initiated accident (RIA) is a design-basis accident under which critical heat flux (CHF) is likely to be exceeded. The operational margin for RIAs is currently determined using steady-state CHF lookup tables, which provide conservative estimates relative to transient CHF phenomena. The Transient Reactor Test Loop (TRTL) facility at Oregon State University is capable of performing out-of-pile rapid heating experiments representative of a RIA at conditions representative of a pressurized water reactor (PWR). Here, to further our understanding of and ability to predict transient CHF under PWR conditions, we performed a sensitivity analysis on a RELAP5-3D model of the TRTL facility coupled to the RAVEN code framework to define a proposed experimental test matrix to be performed at the TRTL facility. We then implemented a flow boiling CHF correlation into RELAP5-3D and performed a secondary sensitivity analysis inspecting the impact of the built-in RELAP5-3D CHF and heat transfer multipliers on both the prediction of CHF and key safety parameters, such as peak cladding temperature and heat flux. The results show that the multiplier with the highest influence toward the prediction of CHF occurrence and the safety parameters is the transient CHF multiplier. Operational performance envelopes have been developed for each of the test matrix cases and will be used for validation once the experiments are performed. The TRTL facility is currently performing shakedown testing to verify system performance prior to proceeding with the experimental campaign. Restart testing results include pump curve restart testing, pressure tests, and heater rod thermocouple transients.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Improved evaluation of safeguards parameters from spent fuel measurements with the Differential Die-Away (DDA) instrument

The Differential Die-Away (DDA) technique is a highly sensitive non-destructive assay method for characterizing and detecting the presence of fissile material within an item of interest. DDA utilizes a series of pulses from a neutron generator (NG) to actively interrogate an item of interest. The die-away time of the neutron population induced by this active interrogation and the integral of the total differential die-away signal can be used to characterize items such as nuclear waste drums and spent nuclear fuel assemblies. In this work, Los Alamos National Laboratory (LANL) conceptualized, designed, and fabricated a DDA instrument that was deployed for field test measurements at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab) in Oskarshamn, Sweden. The instrument performed multiple static measurements at fixed locations and dynamic axial scans of 15 pressurized water reactor (PWR) and 10 boiling water reactor (BWR) spent fuel assemblies, collecting both passive and active measurement data. The static assays of the assemblies measured the differential die-away signal, die-away time, and total passive neutron emission rate to create calibration curves for the evaluation of assembly multiplication, burnup, initial enrichment, effective fissile mass, and total elemental plutonium mass. Each calibration curve was optimized by minimizing the relative root mean square error (RRMSE) of assembly assay results compared to declared assembly parameters. The same quantities were also measured with the axial scans, and the resulting data were applied in two ways: (1) in the creation of calibration curves to improve evaluation of the same safeguards parameters as static assays, and (2) for comparison to simulation. In most cases, across both PWR and BWR assemblies, axial scan data improved the estimation of the above parameters, quantified by decreasing the calibration curve RRMSE. These axial scan results demonstrate the ability of the DDA instrument and analysis method to characterize spent PWR and BWR fuel as well as, or better than, a static assay of the same assembly. Furthermore, the DDA instrument’s unique ability to obtain both active and passive data in a single, axial scan of an entire spent fuel assembly represents a more efficient and accurate way of assaying spent fuel for verification purposes. These results represent a significant advancement for characterizing spent nuclear fuel compared to current technologies.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗