The effects of orientation angle, subcooling, heat flux, mass flux, and pressure on bubble growth and detachment in subcooled flow boiling

Thesis (S.M. and S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012.Cataloged from PDF version of thesis.Includes bibliographical references (p. 119-122).The effects of orientation angle, subcooling, heat flux, mass flux, and pressure on bubble growth and detachment in subcooled flow boiling were studied using a high-speed video camera in conjunction with a two-phase flow loop that can accommodate a wide range of flow conditions. Specifically, orientation angles of 0' (downward-facing horizontal), 30°, 45°, 60°, and 90° (vertical); mass flux values of 250, 300, 350, and 400 kg/m²s, with corresponding Froude numbers in the range of 0.42 to 1.06; pressures of 101 (atmospheric), 202, and 505 kPa; two values of subcooling (10°C to 20°C); and two heat fluxes (0.05 to 0.10 MW/m²) were explored. The combination of the test section design, high-speed video camera, and LED lighting results in high accuracy (order of 20 microns) in the determination of bubble departure diameter. The data indicate that bubble departure diameter increases with increasing heat flux, decreasing mass flux, decreasing levels of subcooling, and decreasing pressure. Also, bubble departure diameter increases with decreasing orientation angle, i.e. the largest bubbles are found to detach from a downward-facing horizontal surface. The mechanistic bubble departure model of Klausner et al. and its recent modification by Yun et al. were found to correctly predict all the observed parametric trends, but with large average errors and standard deviation: 35.7+/-24.3% for Klausner's and 16.6±11.6% for Yun's. Since the cube of the bubble departure diameter is used in subcooled flow boiling heat transfer models, such large errors are clearly unacceptable, and underscore the need for more accurate bubble departure diameter models to be used in CFD.by Rosemary M. Sugrue.S.M.and S.B

An experimental study of bubble departure diameter in subcooled flow boiling including the effects of orientation angle, subcooling, mass flux, heat flux, and pressure

The effects of orientation angle, subcooling, heat flux, mass flux, and pressure on bubble departure diameter in the isolated bubble regime of subcooled flow boiling were studied by high-speed video in a two-phase flow loop that can accommodate a wide range of flow conditions. Specifically, the following ranges were explored: orientation angles of 0° (downward-facing horizontal), 30°, 45°, 60°, 90° (vertical), and 180° (upward-facing horizontal); mass flux values of 250, 300, 350, and 400 kg/m2 s, corresponding to Froude numbers between 0.42 and 1.06; pressures of 101 (atmospheric), 202, and 505 kPa; two values of the subcooling degrees (10 and 20 °C); and two heat fluxes (0.05 and 0.10 MW/m2). The combination of the test section design, high-speed video camera and LED lighting results in high accuracy (order of 20 μm) in the determination of the bubble departure diameter. The data indicate that the bubble departure diameter increases with increasing heat flux, decreasing mass flux, decreasing subcooling, and decreasing pressure. Also, the bubble departure diameter increases with decreasing orientation angle, i.e. the largest bubbles are found to detach from a downward-facing horizontal surface. The mechanistic bubble departure diameter model of Klausner et al. and its recent modification by Yun et al. were found to correctly predict all the observed parametric trends, but with large average errors and standard deviation: 65.5 ± 75.8% for Klausner's and 37.9 ± 51.2% for Yun's. Since the cube of the bubble departure diameter is used in subcooled flow boiling heat transfer models, such large errors are clearly unacceptable, and underscore the need for more accurate bubble departure diameter models.Douglas C. Spreng FundNuclear Energy Institut

A robust momentum closure approach for multiphase computational fluid dynamics applications

Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2017.This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.Cataloged from student-submitted PDF version of thesis.Includes bibliographical references (pages [183]-190).Multiphase computational fluid dynamics (M-CFD) modeling approaches allow for the prediction of critical three-dimensional thermal-hydraulics phenomena in nuclear reactor applications. The advancement and consistent adoption of such tools could transform the industry's approach to the design of reliable systems, and the efficient operation of systems existing, which in the past have been dependent upon correlation-based sub-channel analysis codes. The success of these M-CFD methods in simulating two-phase flow and boiling heat transfer depends on their demonstrated accuracy and robustness, which signals a dual need for the comprehensive analysis of existing data and a reevaluation of the underlying physics. By virtue of the Eulerian-Eulerian two-fluid approach, additional terms manifest in the M-CFD phase momentum equations, which represent information that has been lost, and require closure through prescription of interfacial force models. These momentum "closures" are vital to M-CFD prediction of mean flow profiles, including void fraction and phase velocity distributions, and require high-resolution, robust models to perform accurately throughout a diverse array of flow conditions. While an overwhelming number of models have been developed with a wide range of varying performance, no consensus exists about how to assemble these models successfully in a CFD framework, suggesting that their predictive power is still limited. The lift force, responsible for lateral void fraction redistribution, is particularly refractory to the development of a consistent modeling strategy for these closures. Historically, CFD approaches have been forced to inconsistently leverage existing models derived for single bubbles in laminar flow, which disregard the complex dynamics and interactions of bubbles with turbulence and bubble wakes. Current understanding of the lift force in turbulent flow has been limited to qualitative evidences that small spherical bubbles experience a positive lift, resulting in a wall-peaked void fraction distribution, while larger deformed bubbles experience a negative lift and corresponding center-peaked profile. The present work brings forward a new physical interpretation of the lift force in turbulent bubbly flow through a synthesis of information from DNS studies, fine and coarse scale experiments, and analytical investigations. To overcome the limitations of previous models, a simple dimensionless quantity, the Wobble number, a number which systematically describes the unsteady behavior of bubbles in turbulent flow conditions, is proposed. Introducing this dependency into the lift formulation allows for precise identification of lift inversion, which alone exceeds capabilities of existing models. Additionally, the model is extended to account for group behavior with the introduction of a swarm-like function based on void fraction. Its formulation is built on the conceptual understandings of drift phenomena, bubble interaction probability, and the maximum packing factor for dispersed bubbly flow. These two key mechanisms are assembled into a lift model for turbulent bubbly flow, which is implemented in CFD and validated on several experimental databases spanning an extensive range of flow conditions. Error analyses demonstrate the new formulation's robustness and predictive abilities, allowing for a more comprehensive representation of the two-phase phenomena particularly significant in nuclear reactor applications; moreover, it avoids the introduction of case-specific adjustments to unphysical coefficients and tunable parameters which are characteristic and typical limitations of previous models, indicating another valuable improvement. Finally, the new model's performance in a prototypical rod bundle is evaluated and a qualitative assessment of its applicability in a nuclear reactor geometry context is demonstrated.by Rosemary M. Sugrue.Ph. D

Development and evaluation of a calibrator material for nucleic acid-based assays for diagnosing Aspergillosis

Twelve laboratories evaluated candidate material for an Aspergillus DNA calibrator. The DNA material was quantified using limiting-dilution analysis; the mean concentration was determined to be 1.73-1010 units/ml. The calibrator can be used to standardize aspergillosis diagnostic assays which detect and/or quantify nucleic acid

Conceptual Design of Molten Salt Loop Experiment for MIT Research Reactor

Molten salt is a promising coolant candidate for Advanced High Temperature Reactor (AHTR) Gen-IV designs. The low neutron absorption, high thermal capacity, chemical inertness, and high boiling point at low pressure of molten salt coolants could potentially lead to enhanced safety and lower cost of AHTR designs as compared with conventional Light Water Reactors. Improved economics are expected to be a result of the higher possible operating temperatures (700oC), improving thermal efficiency, availability of process heat for industrial applications, and reduced containment costs. Improved safety margins arise from the use of highly robust TRISO particles fuel in either pebble or graphite block form, greater thermal inertia, low pressure and high boiling point of molten salts relative to water cooled reactor designs. Currently, one of the main challenges associated with further advancement of AHTR design is predicting reactor core materials’ interactions with molten salt coolant over long time scales in a radiation environment. In the Fall of 2010, the Nuclear Engineering Design Project Course (22.033/33) undertook the challenge to design a molten salt test loop to be installed in the MIT Research Reactor (MITR) that would recreate anticipated AHTR operating conditions and fill the knowledge gap in understanding of materials behavior in such environment. In addition to simulating neutronic, thermal and chemical conditions similar to those of AHTR, the test loop must also meet the safety and operating requirements of the MITR. During the course, a preliminary design was developed that features an annular in core molten salt flow channel to maximize the volume available for testing materials’ samples and maintaining the salt temperature at 700oC and flow velocity at 6 m/s, while avoiding boiling at the outside surface of the loop, as prescribed by MITR safety requirements. A number of additional requirements were addressed by the students including reactivity insertion, power peaking, tritium production, shielding, and others. The design tasks were subdivided into four key areas of neutronics, thermal hydraulics, chemistry and materials, and instrumentation and control. The molten salt chosen was LiF-BeF[subscript 2] (FLiBe) with lithium enriched in [superscript 7]Li isotope up to 99.995% because this salt is the leading coolant candidate for AHTR. Hastelloy-N was chosen as the material in contact with the molten salt due to its high resistance to corrosion, good material properties at high temperature and extensive use in previous experiments. The presence of corrosion products, free fluorine and production of tritium in the molten salt were found to be important phenomena challenging the loop design. Therefore, various methods for the salt chemistry control and tritium release were evaluated and resulted in a design of multi-component system for monitoring the salt conditions, maintaining redox potential and removing the impurities and tritium from the salt. Another challenge was managing the loop operation given the relatively high freezing point of the salt at about 460oC. Procedures were developed for start-up, steady-state, shutdown and transient operation of the loop. The thermal hydraulic analyses indicate that 1.8 kW of strap heating along the loop outside the core section and a 1.5 to 2 kW pump were required, depending on final design choices. In addition, preliminary cost estimates of constructing the loop experiment at MITR were performed. The main constraints on the choice of the loop’s individual components and diagnostics were: 1) the ability to function at the designed operating temperature, pressure, and flow rates; 2) the ability to function in a nuclear radiation environment; and 3) the necessity to meet MITR safety requirements. A database of vendors for the loop’s components, instrumentation, and diagnostics was compiled. To support further work on the molten salt test loop an electronic library of references was compiled as well. Finally, a number of potential accident scenarios were examined and their effects on the safety and operation of MITR were evaluated and found to represent no danger to the public or interfere with normal operations. Minor leakages of either the reactor water or the molten salt coolant inside the loop were found to be self-sealing with little to no effects on the safety and operation of MITR. A complete failure of the loop’s heating and pumping systems was found to lead to FLiBe’s cooling and freezing inside the loop, with the freezing time ranging from several minutes to ~1 hour depending on the choice of the loop thermal insulation material