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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 OSTI2016

The Effect of Support Rod on Film Boiling Heat Transfer for Hot Metal Sphere in Subcooled Water

In fuel-coolant interaction (FCI), the heat transfer of corium has been conventionally evaluated by the correlations with a sphere developed in boiling region for subcooled film. However, these correlations do not give significant consideration on the support rod for the sphere from their developments, although Nishio et al. pointed out that the heat transfer and the vapor film collapsing mode would depend on H/D ratio. In this study, we performed experimental tests with the SS316 sphere (10 mm) and found the noticeable increase on the heat transfer in subcooled film boiling region with the deeper immersion depth and larger diameter of support rod. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Sensitivity study on the fin effect of thermocouple mounted on the heated surface under film boiling condition

In the reactivity-initiated accident (RIA) scenario, right after the power burst with the occurrence of departure from nucleate boiling (DNB), the cladding temperature quickly reaches its maximum and is maintained in a steady film boiling regime. Then, under limited heat transfer performance of film boiling, the local cooling effect by the thermocouple wire may not be neglected. A computational model is developed to estimate the local cooling effect of the thermocouple wire mounted on the cladding surface under film boiling condition. The model predicts the temperature profile along the cladding surface as physically expected and is verified against the benchmark problem. Sensitivity studies reveal the significant effects of the thickness of vapor film layer and the diameter of thermocouple on the temperature drop near the thermocouple, for they increase the heat transfer area to the liquid water region. The impact of film boiling heat transfer coefficient, however, is not significant, as the primary heat transfer mechanism near the thermocouple is the heat conduction to the wire. Future study will include the contribution of radiative heat transfer between the cladding surface and the vapor film layer, as well as a transient power history which represents an actual RIA scenario.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Effects of liquid subcooling and initial rod temperature on the minimum film boiling temperature

This paper presents an experimental study on the minimum film boiling temperature (T{sub min}) and its sensitivity to liquid subcooling and initial rod temperature. T{sub min} is measured through a quenching experiment using a vertical stainless steel rod that is quenched in a distilled water bath. The cylindrical stainless steel rod has a flat circular end with diameter of 9.5 mm similar to that of a fuel rod and length of 25 cm. Physically, T{sub min} is used to define the boundary between film boiling and transition boiling, beyond which temperature liquid would not be able to come into physical contact with the solid surface. Results show that the quenching curves are affected by varying the liquid subcooling but are not significantly altered by the initial rod temperature. Additionally, the experiments include visualization of the boiling regimes. The visualization gives insight into the film flow behavior that is found to depend strongly on the liquid subcooling. It also provides an understanding of the phenomenon for quench front axial propagation (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Computational Study of Droplet Dynamics in Dispersed Flow Film Boiling Regime in PWR Geometries

Dispersed flow film boiling (DFFB) regime is exhibited during low-flooding rate conditions following the postulated Loss-of-Coolant Accident (LOCA) in nuclear reactor flow channels. LOCA is a type of design-basis accident scenario, which is crucial for reactor design and licensing. Accurate models to predict the heat transfer in DFFB regime is critical for the reactor to maintain safety limits. The nominal operating linear heat rate is also set based on the maximum linear heat rate determined for a LOCA transient. It is understood that the behavior of entrained droplets in DFFB regime greatly affects the heat transfer by evaporating and desuperheating the surrounding steam. Furthermore, interaction with flow structures also aids in increasing the evaporation rate of entrained droplets. In order to quantify this enhancement in heat transfer, the size of droplets downstream of flow structures (Spacer Grids and Mixing Vanes) is of crucial importance. Thermal hydraulic analysis codes like COBRA-TF, CATHARE-3 and adopt a three-field equation model that allows them to model the droplet field separately. However, to ensure the accuracy of these field equations, various correlations and physical models are required. This paper proposes a method for characterizing the mean Sauter Diameter of droplets across spacer-grids and mixing vanes in PWR sub-channels using PHASTA - a high fidelity flow solver with multiphase capabilities using the level-set method. The PHASTA code is a finite element based computational fluid dynamics code designed for solving compressible and incompressible turbulent flows. It has been demonstrated to be massively scalable (up to 786,432 cores), which is crucial for simulating large realistic domains. The code incorporates the level-set method for interface tracking of the gas-liquid interface.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Study of the film boiling heat transfer and two-phase flow interface behavior using image processing

Here, in this study, based on a small-scale quench test facility, the two-phase flow interface behavior during quench transients is visualized and analyzed utilizing an image processing framework. The high-fidelity experimental results obtained for two-phase flow in the current framework can support various studies both in the time domain and in the frequency domain. In particular, visualization of the data obtained from different heating surfaces under different test conditions are used to perform a full-scale transient 2-D vapor film reconstruction. The liquid-vapor interface variations in various heat transfer regimes as well as at the initial film breakup point can be directly obtained through the processed data. Moreover, the temporal variation of the interfacial wave frequency approaching quench is investigated in detail. Based on the high-resolution data obtained for the liquid-vapor interface, the detailed phase velocity and temperature profiles are obtained through theoretic analysis, based on which the film boiling heat transfer coefficient (HTC) can be determined. In addition, an improved film boiling HTC model is developed considering the effects of wall superheat, liquid subcooling temperature, vapor film thickness as well as fluid properties. The model is found to predict film boiling HTC well within 15% error.

42 ENGINEERING↗
DOE OSTI2025

Insight into pseudo-film-boiling in supercritical water under simplified interfacial conditions

The focus of this paper is the physics of supercritical water film boiling through the simulation of natural convection of supercritical water in a rectangular cavity. It has been widely believed that above a fluid's critical point, only one phase exists. However, recent research has indicated the existence of distinct gas-like and liquid-like regions separated by the Widom line, which is the locus of the maxima of the specific heat. Along the Widom line, density decreases sixfold, viscosity drops by a factor of two, while specific heat spikes by an order of magnitude. These variations, specifically in density and viscosity, produce a thick pseudo-interface and flow dynamics behavior akin to film-boiling, denoted here as pseudo-two-phase flow dynamics. When the cavity is heated from the bottom, a pseudo-film of gas-like fluid is formed at the heated wall, and the flow complexity evolves, characterized by the formation of multiple vortices. Results demonstrate the ability of modern conjugate heat transfer computational fluid dynamics to capture the pseudo two-phase regime, with the formation of a pseudo-film at the heated wall. Further, the boiling curve is studied, showing an increase in heat flux with an increase in the temperature differential between the wall and the pseudo-critical temperature. Furthermore, results also show a previously uncharacterized instability, where the convection cells in the cavity oscillate and can flip directions under certain conditions.

Computational fluid dynamics↗
DOE OSTI2016

Multiphysics modeling of two-phase film boiling within porous corrosion deposits

Porous corrosion deposits on nuclear fuel cladding, known as CRUD, can cause multiple operational problems in light water reactors (LWRs). CRUD can cause accelerated corrosion of the fuel cladding, increase radiation fields and hence greater exposure risk to plant workers once activated, and induce a downward axial power shift causing an imbalance in core power distribution. In order to facilitate a better understanding of CRUD's effects, such as localized high cladding surface temperatures related to accelerated corrosion rates, we describe an improved, fully-coupled, multiphysics model to simulate heat transfer, chemical reactions and transport, and two-phase fluid flow within these deposits. Our new model features a reformed assumption of 2D, two-phase film boiling within the CRUD, correcting earlier models' assumptions of single-phase coolant flow with wick boiling under high heat fluxes. This model helps to better explain observed experimental values of the effective CRUD thermal conductivity. Finally, we propose a more complete set of boiling regimes, or a more detailed mechanism, to explain recent CRUD deposition experiments by suggesting the new concept of double dryout specifically in thick porous media with boiling chimneys. - Highlights: • A two-phase model of CRUD's effects on fuel cladding is developed and improved. • This model eliminates the formerly erroneous assumption of wick boiling. • Higher fuel cladding temperatures are predicted when accounting for two-phase flow. • Double-peaks in thermal conductivity vs. heat flux in experiments are explained. • A “double dryout” mechanism in CRUD is proposed based on the model and experiments.

42 ENGINEERING↗
DOE OSTI2017

Multi-Functional Zirconium-Silicide Coatings on Zirconium-alloy for Improved Accident Tolerance

Recently, development of protective coatings on zirconium-alloy (Zr-alloy) fuel cladding in light water reactors has been extensively investigated to enhance safety margins in accident scenarios. One accident scenario is degradation of the reactor core under a coolant-limited condition, where a large amount of Zr-alloy claddings would react with high temperature steam, leading to oxidation, hydrogen production, and loss of intrinsic mechanical properties of the fuel cladding. Although, total replacement of Zr-alloy with SiC-SiCf composites or FeCrAl is being considered for the long-term, application of oxidation resistant coatings on zirconium-alloy is being investigated as a near-term solution to mitigate oxidation of Zr-alloy and increase the accident coping time. Coatings, such as chromium, Ti{sub 2}AlC, SiC, and Cr{sub 3}C{sub 2}- NiCr, deposited by various deposition techniques have been investigated. In addition, recent studies on modified heating surface for heat transfer enhancement provides basis on the optimal design of coated Zr-alloy claddings so that the new cladding can be achieve accident tolerance and heat transfer efficiency for both accident and normal operation conditions. Therefore, boiling heat transfer from the coated surface to water environment must be evaluated in parallel. In heat transfer performance, boiling heat transfer is characterized by nucleate boiling, critical heat flux (CHF), transition boiling, minimum film boiling temperature (T{sub min}), and film boiling. Zirconium-silicide (ZrSi{sub 2}) is a potentially attractive protective coating material for cladding applications. Although this material has not been studied extensively, it appears to be a promising candidate because of the high melting points, and the low neutron penalty, and the generally good oxidation resistance to transition metal silicides. It has been reported that oxidation of ZrSi{sub 2} powder in a flowing air environment at elevated temperatures produced ZrO{sub 2}, SiO{sub 2}, and ZrSiO{sub 4}, pointing to the possibility of achieving highly protective surface layers of SiO{sub 2} and ZrSiO{sub 4} that would protect the material from further oxidation. In this study, magnetron sputter deposition of ZrSi{sub 2} on zirconium-alloy flats and cylindrical rods was performed. The coated samples were oxidized in high temperature air (700 deg. C, 1000 deg. C and 1200 deg. C) to evaluate their oxidation resistance. The coated cylindrical rods were quenched in saturated water to observe transition boiling, T{sub min}, film boiling regime, while the coated flats were tested in pool boiling facility to investigate nucleation boiling and CHF.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Single-Bubble Flow Boiling Phenomena - Interface Tracking Simulation

Boiling, as one of the most efficient heat transfer mechanisms, is widely used in various engineering systems. Better understanding and modeling of this process remains a major challenge in multiphase flow research. The distribution of vapor in boiling system affects the heat transfer rate and may cause unfavorable conditions, such as heater burn-out. A number of flow regimes have been observed experimentally. In general, there are three regimes under boiling conditions: nucleate boiling, transition boiling, and film boiling. The nucleate boiling regime is categorized into partial boiling regime and fully developed nucleate boiling, according to behavior of bubble dynamics and heat transfer characteristics. In the first region, as the wall temperature increases, the fraction of the wall surface subject to nucleate boiling increases until bubble formation occupies the entire heated surface. Heat transfer in this region is a complex mixture of single-phase forced convection and nucleate boiling. As bubble density increases, bubbles turn to coalesce and form insulating vapor patches, which impedes liquid flow back to the surface. The heat transfer rate under this condition reduces dramatically. This point is called Critical Heat Flux (CHF), which leads to boiling crisis. The behavior of boiling crisis is dependent on local fluid conditions due to bubble growth and departure. Experiments are the ultimate tool for boiling model development and verification, but they still lack some capabilities needed for better understanding of the local mechanism of this phenomena. It's difficult to capture many important details at the pressure/temperature conditions of interest using experimental techniques, such as full three dimensional measurements of velocity and temperature distribution and estimates of heat flux partitioning during the boiling process. The interface tracking simulation (ITS) is one of the promising approaches to describe heat transfer of boiling phenomena and their underlying mechanisms. The advances of high performance computing (HPC) in recent years made it possible to apply DNS to a wide variety of adiabatic flows. Some work has been done to implement evaporation / condensation models as well. Lee and Nydahl performed a numerical simulation of bubble growth and departure assuming that the bubble has the shape of a truncated sphere and a micro-layer exists during bubble growth. A 2-D cylindrical bubbles simulation has been done by Son with a simplified model to predict the heat flux in a thin liquid micro-layer. A direct simulation of 2D film boiling is conducted by Juric and Tryggvason. The film boiling, below critical pressure with a modified Level Set Method, is studied by Son and Dhir. Ose and Kunugi have developed a boiling and condensation model on subcooled boiling phenomena. Fuchs performed numerical simulations of bubble growth and departure considering wall conduction effects. Fully transient heat and fluid flow is modeled with a free surface of a rising bubble and a periodic calculation of the whole cycling of a growing, detaching, and rising bubble. Mukherjee and Kandikar have done a numerical simulation of an evaporating meniscus on a moving heated surface with level set method. The heat transfer associated with the advancing and receding contact angle has been studied. To develop the capability of physics-based boiling process, in the presented work the energy and momentum equations are coupled to simulate single-bubble growth scenarios. For a single bubble simulation in infinite liquid, the analytical solution exists and can be compared with the numerical results for verification purposes. The presented capabilities will be utilized in the large scale DNS of boiling flows in the future. The finite-element based code, PHASTA is used for the simulation. It is a parallel, hierarchic, higher-order accurate, adaptive, stabilized, transient analysis flow solver, which has been shown to be an effective tool for a wide range of single and two-phase flow problems. It has been shown to be highly scalable on top supercomputers ({approx}85% scaling on up to 3 million mesh partitions using a 92 billion element mesh . PHASTA has been shown to reliably predict the details of adiabatic bubbly flows as well as single-phase flows with heat transfer. To develop the capability of physics-based boiling process, in the presented work the energy and momentum equations are coupled to simulate a single-bubble growth scenario. For a single bubble growth in superheated water, the analytical solution exists and can be compared with the numerical results for verification purposes. The second-order accurate spatial and time discretization is used for the flow solver in the presented work. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Characterization of SiC-SiC accident tolerant fuel cladding after stress application

In this work, SiC-SiC tubes with representative LWR (water cooled reactor) cladding dimensions were subjected to cycles of out-of-pile thermal stress followed by quenching. Samples were held at elevated temperatures and dropped into a water bath to replicate the high thermal stresses generated during reflooding. The samples were then characterized to measure changes in strength, permeability, and structure. Results were evaluated both for quenching into room temperature and boiling water. The results show that SiC-SiC composite tubes quenched from 1000 Celsius degrees in air into a boiling water bath exhibited no increase in He leak rate through the cylinder wall up to a pressure difference of 34.5 kPa compared to the as-fabricated state. SiC-SiC composite tubes quenched from 300-400 C. degrees into room temperature water showed an increase in He leak rate compared to the as-fabricated state. We attribute this to the more severe thermal gradient generated through the wall thickness due to higher heat flux characteristic of nucleate boiling compared to film boiling. A persistent vapor film forms around the samples quenched into boiling water, whereas no vapor film forms for the samples quenched into room temperature water. Another important result is that not loss in strength was noted compared to unquenched samples and that the stresses generated by the quenching were not enough to cause fiber fracture.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Evaluation for Effect of TRACE Spacer Grid Model on RBHT Reflood Tests

Usually, the spacer grid decreases the flow area and then enhances the convective heat transfer due to the flow acceleration and the turbulence increases. Their phenomena is shown in both egg-crate style spacer grids and mixing vane grids. Also, the mixing vane swirls the flow and enhances the convective heat transfer. In this study, the effect of spacer grid model of TRACE was performed on RBHT (Rod Bundle Heat Transfer) reflood tests. The RBHT tests were designed to conduct several separate-effects tests such as single phase steam cooling, low flow boiling and dispersed flow film boiling heat transfer, which is very important during a large break loss of coolant accident (LOCA). This facility has a heated length of 3.66 m (12.0 ft), which is typical of a pressurized water reactor (PWR) fuel assembly. The test section consists of a 7 x 7 rod bundle with 45 heated electrical rods, and seven mixing vane grids. The electrical rods have a diameter of 9.5 mm and are arranged with a 12.6 mm pitch with 45 heater rods and four unheated rods. Tests 1096, 1108, 1170, 1196, 1285 and 1383, which are in the TRACE assessment manual , were selected to evaluate the effect of the spacer grid model of TRACE. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Experimental investigation on the phenomena around the onset nucleate boiling during the impacting of a droplet on the hot surface

Onset of nucleate boiling of a droplet when impacted onto hot surface was investigated. Three kinds of surfaces, normal stainless steel (NSS), stainless steel with TiO{sub 2} coating (UVN), and stainless steel with TiO{sub 2} coating and radiated by ultraviolet ray were employed to examine the effect of wettability. The droplet size was 2.4 mm diameter, and dropped under different We number. The image is generated by high speed camera with the frame speed of 1000 fps. The boiling conditions are identified as natural convection, nucleate boiling, critical heat flux, transition, and film boiling. In the present report, the discussion will be focused on the beginning of nucleate boiling on the droplet. Nucleate boiling occurs when bubbles are generated. These bubbles are probably caused by nucleation on the impurities within the liquid rather than at nucleation sites on the heated surface because the bubbles appear to be in the bulk of the liquid instead of at the liquid-solid interface. In addition, the smaller the contact angle, the fastest the boiling.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2023

Tree-Based Ensemble Learning Models for Wall Temperature Predictions in Post-Critical Heat Flux Flow Regimes at Subcooled and Low-Quality Conditions

Accurately predicting post-critical heat flux (CHF) heat transfer is an important but challenging task in water-cooled reactor design and safety analysis. Although numerous heat transfer correlations have been developed to predict post-CHF heat transfer, these correlations are only applicable to relatively narrow ranges of flow conditions due to the complex physical nature of the post-CHF heat transfer regimes. In this paper, a large quantity of experimental data is collected and summarized from the literature for steady-state subcooled and low-quality film boiling regimes with water as the working fluid in vertical tubular test sections. In addition, a low-quality water film boiling (LWFB) database is consolidated with a total of 22,813 experimental data points, which cover a wide flow range of the system pressure from 0.1 to 9.0 MPa, mass flux from 25 to 2750 kg/m 2 s, and inlet subcooling from 1 to 70 °C. Two machine learning (ML) models, based on random forest (RF) and gradient boosted decision tree (GBDT), are trained and validated to predict wall temperatures in post-CHF flow regimes. The trained ML models demonstrate significantly improved accuracies compared to conventional empirical correlations. To further evaluate the performance of these two ML models from a statistical perspective, three criteria are investigated and three metrics are calculated to quantitatively assess the accuracy of these two ML models. For the full LWFB database, the root-mean-square errors between the measured and predicted wall temperatures by the GBDT and RF models are 5.7% and 6.2%, respectively, confirming the accuracy of the two ML models.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Characterization of Fuel-to-Coolant Heat Transfer During Reactivity-Initiated Accidents Using Tightly Coupled Thermal Hydraulics and Fuel Thermomechanics

The reactivity-initiated accident (RIA) is a complex scenario with several tightly interacting physical phenomena. Accurately predicting fuel behavior during these transients is difficult due to limitations in the modeling of fuel-to-coolant heat transfer. Common approaches to simulate RIAs involve standalone calculations using either a fuel performance code or a thermal-hydraulic code. The complex interdependencies of thermal-hydraulic and fuel mechanical behavior suggest that a tight coupling between these codes may provide more accurate predictions of fuel-to-coolant heat transfer and cladding mechanical response. Here, RELAP5-3D and BISON are coupled in this paper to simulate RIAs, and a sensitivity analysis is performed to rank key thermal properties and two-phase heat transfer parameters relevant for fuel-to-coolant heat transfer and cladding failure mechanisms in UO 2 –Zircaloy-4 systems. Gas gap conductance, film boiling heat transfer uncertainty, pulse width, fuel-specific heat capacity, and cladding-specific heat capacity were identified as important parameters. Variations in figures of merit resulting from changes to pulse width and the material thermal properties indicate that time-dependent heat transfer rates are significant for safety-relevant mechanical parameters due to the time dependence of cladding ductility and pellet-cladding mechanical interaction loading. The results suggest that the thermal-hydraulic factors have a nonnegligible influence on the thermomechanical solution and vice versa. Tight coupling of both sets of physics is recommended to improve prediction of fuel behavior during RIAs. Highlights include the following: 1. The RELAP5-3D thermal-hydraulic code and the BISON fuel performance code are tightly coupled for simulation of RIA transients with energy depositions at the Zircaloy-4 cladding failure threshold. 2. Departure from nucleate boiling occurred for all simulated cases. Due to the ductility of fresh fuel, substantial ballooning occurred in most cases. 3. Gas gap conductance, fuel-specific heat capacity, cladding-specific heat capacity, transient pulse width, and film boiling heat transfer were the dominant thermal factors impacting the safety figures of merit at energy depositions.

Critical Heat Flux (CHF)↗
DOE OSTI2016

TRACE/PARCS Analysis of ATWS with Instability for a MELLLA+BWR/5

A TRACE/PARCS model has been developed to analyze anticipated transient without SCRAM (ATWS) events for a boiling water reactor (BWR) operating in the maximum extended load line limit analysis-plus (MELLLA+) expanded operating domain. The MELLLA+ domain expands allowable operation in the power/flow map of a BWR to low flow rates at high power conditions. Such operation exacerbates the likelihood of large amplitude power/flow oscillations during certain ATWS scenarios. The analysis shows that large amplitude power/flow oscillations, both core-wide and out-of-phase, arise following the establishment of natural circulation flow in the reactor pressure vessel (RPV) after the trip of the recirculation pumps and an increase in core inlet subcooling. The analysis also indicates a mechanism by which the fuel may experience heat-up that could result in localized fuel damage. TRACE predicts the heat-up to occur when the cladding surface temperature exceeds the minimum stable film boiling temperature after periodic cycles of dryout and rewet; and the fuel becomes “locked” into a film boiling regime. Further, the analysis demonstrates the effectiveness of the simulated manual operator actions to suppress the instability.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Assessment on DNB performance for high pressure subcooled boiling with modified CASL boiling model

Understanding and predicting the Departure from Nucleation Boiling phenomena in the nuclear power industry has been a grand challenge due to its large impact on the reactor safety and operational economics. Over the past several years, CASL THM team has extensively investigated and demonstrated a state of art Multiphase CFD framework to provide a high-fidelity DNB simulation capability based on the two-fluid Eulerian model in conjunction with a wall heat flux partitioning approach. The overarching goal of this task is to develop and demonstrate a robust CFD-based DNB model which is capable of predicting an entire boiling history up to the DNB condition in which the wall boiling mode is transitioning from the nucleate boiling mode to the film boiling mode. Three different flow channel configurations (pipe flow, 5x5 fuel bundle with and without mixing vane tests) were studied and thorough closure sensitivity studies were performed to evaluate the wall boiling and momentum closure effect on the DNB performance. Predicated DNB values were compared to the measurements from high pressure subcooled flow boiling tests resulting in a consistent under-predictive trend for all three validation campaigns in FY16-18. With extensive closure sensitivity study, a modified CASL boiling closure model (i.e. Gen1-FY19) is introduced with the following key changes: 1) improved wall boiling closure for high pressure application, 2) newly calibrated lift and wall lubrication force models with modified interfacial length scale model, and 3) updated liquid phase condensation model calibrated for the subcooled boiling application.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Hydrodynamic instabilities and heat transfer characteristics in the duct flow of a fluid in the supercritical thermodynamic regime

The behavior of fluids at supercritical thermodynamic conditions is inherently complex due to large variations in thermodynamic and transport properties. Recent numerical and experimental investigations illustrate ongoing interest for these fluids, especially supercritical CO 2 and supercritical water, for a variety of applications. For example, supercritical water reactors (SCWR) operate in this extreme condition of high-pressure and temperature, resulting in highly dynamic flow fields and unexpected heat transfer regimes. The potential heat transfer benefits in this regime are directly associated with the extreme variations in thermodynamic and transport properties, which occur at, and above, the critical point. This work characterizes the hydrodynamic instabilities that arise for fluids at supercritical thermodynamic conditions when buoyancy forces are significant. Two specific configurations are considered, a natural convection cavity flow, and a mixed convection, heated, horizontal channel flow. Natural convection flow in a cavity is a classical configuration with expected behavior below the critical point. This configuration aids in characterizing the effect of the variable properties in the supercritical thermodynamic regime. Further, limited studies in the existing literature have been conducted for low-Reynolds and intermediate-Rayleigh numbers, mixed-convection channel flows for supercritical water, which is the focus of the channel flow configuration. To investigate the thermally driven hydrodynamic instabilities in this regime, a high-order fully-implicit numerical method is used. Such strong variations in thermophysical properties (in particular, density) are difficult to simulate and an altogether compressible framework is needed. Therefore, the compressible Navier-Stokes equations are solved without any additional assumptions. The fully implicit, high-order in space and time, reconstructed discontinuous Galerkin method as implemented within the multi-physics code called ALE3D (Arbitrary Lagrangian and Eulerian in 2D and 3D), developed at Lawrence Livermore National Laboratory (LLNL), is used. This fully implicit, L-stable method accurately captures the compressible nature of the ow in the limit of very low Mach number. It has been widely accepted that above the critical point, only one phase is observed. However, recent research has indicated the existence of the distinct gas-like and liquid-like regions separated by the Widom line, the locus of the maxima of the specific heat. Along the Widom line, density decreases 6-fold, viscosity drops by a factor of 2, while specific heat spikes by an order of magnitude. These variations, specifically in density and viscosity, produce a thick pseudo-interface and flow dynamics behavior akin to film boiling. A pseudo-film at the heated wall of the cavity and the horizontal channel is observed where buoyancy forces induce mixing through the specific configurations. Further the local Rayleigh and Richardson numbers provide maps of the flow field and the buoyancy forces driving the microscopic mixing. In the first chapter, I describe a background of supercritical fluid and the various applications. The second chapter focuses on the mathematical model and numerical method used for simulations, where a description of the equation of state for supercritical water is described. The third chapter focuses on the natural convection cavity with a heated bottom wall. In this cavity a gas-like and a liquid-like flow within the supercritical thermodynamic regime are observed. The fourth chapter focuses on a forced convection, horizontal channel, distinguishing between the gas-like, liquid-like, and mixed flow regimes. Mixed convection flow, with the addition of gravitational forces in the horizontal channel show the influence of variable properties on the hydrodynamic development, heat transfer, and rising instabilities. The last chapter of this research focuses on characterizing the unstable hydrodynamics through time-averaging processes and analysis of the movement of energy through the developing plumes.

42 ENGINEERING↗
DOE OSTI2016

Clad to coolant heat transfer during RIA in PWR conditions

The power pulse during an insertion of reactivity accident (RIA) leads to a transient heating of the fuel. Heat transfer toward the coolant through the clad and associated temperature transient of the clad is of safety significance since it may largely impact the risk of clad failure. For sufficiently large enthalpy deposited within the fuel, clad temperature increases rapidly and this can lead to coolant boiling. This paper concerns the critical review of the present understanding and of the models for the clad to coolant heat transfer in RIA for PWR conditions. The results of past experiments are reviewed and the reliability of current models dedicated to these conditions is analyzed. It appears that the use of classical models for wall to fluid heat transfer does not allow predicting the clad temperature transient as measured in RIA-related experiments. The deviation is all the more large when film boiling conditions are reached. To reproduce the experimental data, some correlations of heat transfer can be fitted but the reliability of such model out of the range of conditions of the experimental data is low. It is mainly due to a rather low level of understanding of the heat transfer in very transient conditions (i.e. when the time scale of variation of the main heat transfer parameters, like a wall temperature, is less than the typical internal time scale of the heat transfer process itself). It results in a very large scattering in the numerical simulation of clad temperature when several fuel codes are considered. To improve the reliability of these calculations, it is believed that the study of transient boiling heat transfer thanks to devoted experimental facilities is required.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗