Research

All university-level research programs have certain common goals: Discovering new knowledge, Integrating that knowledge with what we already know to achieve a deeper understanding of the world around us, Applying knowledge to make the world a better place, and Teaching that knowledge to others so that they can continue the process.  Within that framework each individual faculty member has specific long- and short-term goals related to the development of their field.  My research has been focused on the application of the principles of heat, mass, and momentum transfer to nuclear reactor systems.   In the long-term, my goal is to become a nationally and internationally recognized expert in multiphase flow processes, specifically in the improvement of nuclear reactor performance and safety and energy efficiency.  I plan to accomplish this goal by:

• Improving our scientific understanding of key phenomena important to industrial applications such as turbulence, interfacial behavior, and so on. 
• Applying fundamental principles of physics to develop, evaluate, and refine models for multiphase flow systems
• Validating those models using high-quality experimental measurements and innovative measurement techniques
• Mentoring PhD candidates to produce qualified, creative faculty to continue developing the field.

      My general research interests have been focused in a few key areas.  One has been creating new methods for evaluating and optimizing multiphase flow models.  From this area has grown an interest in the development of new instrumentation and the continuing improvement of existing types of instrumentation for measuring important parameters in two-phase flows.  Finally, I have also been involved in applied research, solving key problems related to industrial applications of multiphase flows. 

     Throughout this process, my efforts have led to numerous publications and citations and significant research funding.  I have published a total of 56 journal publications.  This number does not include two book chapters, one two-phase flow in large diameter pipes and one on small modular reactors, nor does it include peer-reviewed conference proceedings.  Several have been published in a journals with impact factor of 4 or higher, with the highest impact factor being 6.2. Perhaps more important than publishing is making sure that people are reading and using the work. Based on data obtained from my Google Scholar profile, I have an h-index of 20. 

I have obtained $665,000 in funding over the last 5 years.  Additional details can be found in my Curriculum Vitae, in Appendix A.  I am also continuing to submit research proposals.  Based on my publication and funding record, according to the Dean of CEC, I “have demonstrated outstanding scholarship in thermal hydraulics.”

      Also important is the development of a national and international reputation for excellence.  I have already begun to develop such a reputation.  I have developed a collaboration with Dr. Xuizhong Shen, a researcher at the Kyoto University Research Reactor Institute in Japan.  I have been approached by the Institute of Nuclear Safety Systems, a subsidiary of the Kansai Electric Power Company in Japan, to perform funded research on reactor safety analysis codes and model development.  Finally, my work has been cited by researchers in a wide range of fields and by researchers across the globe.  The fields range from nuclear applications, to the propagation of gas bubbles in volcanic magma, to the development of models for gas-oil flows in oil wells. Recently the Korean Atomic Energy Research Institute (KAERI) began including a correlation I developed in their nuclear reactor safety analysis code SPACE. This work has also led to a Young Member Achievement Award from the Thermal Hydraulics Division of the Atomic Energy Society of Japan, citing my “extensive and original research contributions to the development of the interfacial area transport equation.”

      The data resulting from this research has been used by organizations ranging from the U.S. Nuclear Regulatory Commission, Bettis Atomic Power Laboratory, and Chevron Energy Technology in order to validate computer codes.  At the NRC this includes validating TRACE, an industry-standard nuclear reactor safety analysis code, and the development of TRACE-T, a beta-version code which includes implementation of detailed bubble coalescence and breakup models.  Bettis Atomic Power Laboratory has used the experimental data to validate the multiphase flow models in commercial CFD code CFX, produced by ANSYS.  Chevron Energy Technology is using data I produced to develop design improvements in their oil processing systems, saving money and improving resource utilization. Detailed void fraction measurements at various points along the suction piping for ECCS pumps were used by Westinghouse Nuclear to justify the safety of the systems to the NRC in response to Generic Letter 2008-1.  The results were used by NRC to create training materials for nuclear plant operators.  INSS implemented the resulting model in their custom version of RELAP5.  The results of the steam condensation study will be used by the Consortium members in their licensing applications to the US NRC.

     Last, but not least, the mentoring of PhD candidates is an important part of our research mission.  In 2018 my first PhD student graduated.  He served as the Director of Engineering at Phase Change Energy Solutions, Inc., a premier engineering design firm developing thermal management solutions across several industries from electronics cooling to food storage and transportation.  He is now a research scientist in the Systems Integration Laboratory at INL.  Seven other students obtained a PhD since. Two other students are working at INL, two have positions in the Navy Nuclear Laboratory, one is working in fuel design for Framatome, and two are planning to enter academia. The Dean of CEC wrote in his recommendation for tenure that this was “truly exceptional.”

Description of Research

Multiphase Flow Experiments and Modeling

      The first step in the scientific method is observation – the collection of experimental data, and a key part of creating new knowledge.  Experimental data is also intimately involved in testing hypotheses (models) that are developed to explain those observations.  To that end, I have performed a great deal of experimental work over the course of my career. A significant portion of that experimental work has been the collection of an extensive database of bubble behavior relevant in a wide range of systems. These include:

• Large diameter tubes such as oil wells and vertical risers, chemical processing systems, and vertical risers in natural circulation Boiling Water Reactors.
• Tube bundles such as nuclear reactor cores and steam generators and chemical process cooling systems
• Rectangular channels such as those found in many nuclear research reactors and nuclear reactors for Naval applications

      I have established the Thermal Hydraulics Experiment, Modeling, and Engineering Simulation (THEMES) Laboratory at Missouri S&T.  Infrastructure necessary for experimental research was recently completed, and I am establishing test facilities for multiphase flow research.  The highlights of the THEMES laboratory include

• A 50 hp air compressor capable of delivering up to 207 acfm of compressed air at a pressure of 150 psi and a 30 hp centrifugal pump which produces 90 ft of head at a flow rate of 1000 gpm;
• Modular test facility design to reduce construction times;
• Instrumentation including various flow meters, electrical conductivity void probes and electrical impedance void meters, etc.;
• Facilities for testing phase change materials including a Transient Hot Bridge (THB) and high-precision, temperature controlled environments allowing thermal property measurements at temperatures ranging from -5ºC to 200ºC;
• Access to the Materials Research Center (MRC) at Missouri S&T for advanced materials characterization;
• An existing multiphase flow test facility for rectangular channels.

      In the spirit of enhancing the ability of researchers to provide high-quality data for validation of models, I have also been part of the development of advanced two-phase flow instrumentation.  As a graduate student I developed a computer-controlled electronic positioning system capable of positioning void probes inside of a flow channel to within 0.1 mm.  Previously, void probes had been positioned by hand using micrometer scales.  I then combined this with a multiple-void-probe system and a high-capacity data acquisition system capable of collecting up to 2.5 million samples per second.  As a result I was able to perform complex experiments in 15-20 minutes that previously required 3-4 hours.

      I have been working to develop more robust methods for constructing probes.  Typical failure rates for probe construction are greater than 50%, leading to significant losses in both resources and time.  Using our in-house electronics facility, I have been working to improve both the construction and characterization process.

Some key improvements include:

• Use of high-strength dielectric epoxy to coat sensors, with roughened sensor surface to improve adhesion and reduce film draining and beading;
• Soldering the wires to sensors rather than crimping, to reduce breakage and create a more stable electrical connection;
• Use of a camera-equipped microscope, allowing more accurate measurement of sensor positions through image analysis;
• Storage of images and improving the ability to confirm, review, and repeat data analysis (as compared to handwritten notes).

There are also some major concerns regarding the data processing methods used for these types of probes.  Specifically, most data processing schemes use the bubble chord length measured by the probe to classify bubbles as Group 1 (small spherical and distorted-spherical bubbles) or Group 2 (Taylor cap and slug or churn-turbulent bubbles).  There is a concern that this process incorrectly categorizes many Group 2 bubbles near the size limit as Group 1, since the chord length is often significantly smaller than the bubble diameter.  As a result of this and other concerns I have implemented a number of improvements to the data processing software, such as:

• Implemented a moving comparison algorithm rather than a threshold to reduce incorrect grouping of rapidly-following trailing bubbles into one large bubble;
• Implement a trust-region method to calculate the diameter of all bubbles previously categorized as Group 1 using a solid-sphere approximation and the interfacial velocity of the front and rear interface to reduce bubble categorization error.

The results from the electrical conductivity probes were also compared with measurements performed using optical void probes.  Optical void probes were provided by the mREAL laboratory at Missouri S&T, led by Prof. Muthanna Al-Dahhan. Previous studies have confirmed that the total void fraction and interfacial area measurements made using these two sensors agree well, however the bubble group categorization has not been confirmed. These improvements were implemented and the two data processing methods were compared for a small number of test cases. 

      The second through fourth steps of the scientific method are developing hypotheses, testing hypotheses, and revising hypotheses.  The development, optimization and evaluation of two-phase flow models falls under these steps. As part of this work I have developed new flow regime maps and drift-flux models in large diameter tubes. The publication that resulted from this work, published in 2010, became one of the 10 most-cited papers in the journal Progress in Nuclear Energy from 2010 to 2014.  I have also spent time evaluating the prediction uncertainty of the two-phase flow models used in TRACE and RELAP, industry-standard safety analysis codes.  As part of this work I developed a revised drift-flux model for the prediction of interfacial drag.  The revised model was able to significantly improve the prediction of high void fraction cases in the vertical riser of advanced boiling water reactors.  I also developed a new interfacial area concentration correlation by deriving the Sauter mean diameter of both small, spherical bubbles and large Taylor bubbles from the steady state two group interfacial area transport equation.  After benchmarking with experimental data, the prediction of interfacial area concentration was significantly improved over current industry-standard approaches.

      The centerpiece of this portion of my research over the past few years has been the development of a modular, one-dimensional two-phase flow analysis code using MATLAB.  Based on the two-fluid model used in RELAP and TRACE, I implemented a full two-bubble-group approach with void transport and interfacial area transport.  At this time the code is limited to vertical flows without heat transfer, but it is a powerful tool for evaluating and comparing models.  All of the constitutive models within the code are modular, allowing me to evaluate the sensitivity of the system to changes in the various parameters that are key to accurately predicting multiphase flows.

      In the spirit of that effort, I have developed an objective optimization technique for two-phase flow models that uses this code as a key component.  Using principles from Pareto optimization and implementing a modified form of the Gauss-Newton algorithm, I was able to make some key revisions to the two-group bubble coalescence and breakup models for large diameter tubes.  The resulting model was able to reduce the interfacial area concentration prediction error from 52% to 33%.    I am continuing to use this approach to evaluate the sensitivity of the code to key models. At this time I am focusing on the drift-flux type correlations which are used to calculate the interfacial drag forces, thereby determining the phase concentrations, flow rates, and coolant inventory in nuclear reactor systems.  In the near future I will also be evaluating various interfacial area concentration correlation schemes and the addition of void covariance effects in the two-fluid model, a phenomenon which has been neglected until now.

Enhancing Passive Safety in Nuclear Reactor Systems

      My interest in enhancing passive safety in nuclear reactor systems is driven by the rise of small modular reactor (SMR) systems.  These reactors are much smaller, and therefore have much smaller thermal loads.  This lends them to various passive safety systems that are impractical in larger designs.  One of the most direct applications is the use of Phase Change Materials (PCMs) in SMRs.  PCMs are materials that are designed to freeze and melt at a specific temperature, and have a high heat of fusion.  This allows them to absorb large amounts of thermal energy at a relatively constant temperature.  In the long term, I plan to develop research on the application of high-temperature PCMs to enhance passive safety in nuclear reactor systems.  I hope to design PCM systems that can be incorporated into the emergency core cooling systems of modern reactor systems such as the suppression pool and reactor containment.  Incorporating PCMs with phase change temperatures in the range of 80 – 90ºC in the suppression pool and containment structure has the potential to absorb more heat, reducing the containment pressure and temperature.  This allows smaller containment and reduced construction costs.

      I began this work by investigating room-temperature PCMs.  While this is a first step in a larger research program, it also has important potential impacts.  The use of PCMs can reduce energy consumption in heating and cooling by 30% to 50%, reducing energy costs and greenhouse gas emissions. My research team developed a novel eutectic PCM using Methyl Palmitate and Lauric Acid, both naturally occurring fatty acids.  The resulting PCM has a melting temperature of 25.5ºC and a heat of fusion of 205.2 kJ/kg.  Properties drifted by only 1% during 3000 melt/freeze cycles, which represents about 80 years of daily thermal loading and unloading.  This PCM may also have applications to space travel: NASA’s ORION mission uses and PCM based heat exchanger to reduce the required radiator size for thermal management.  There are two major concerns associated with this type of PCM.  First is leakage of the liquid phase, which causes loss of material and therefore degrades performance.  Second is the low thermal conductivity of most organic PCMS, resulting in large thermal gradients that impede the ability of the materials to maintain the internal environment at a constant temperature.  To remedy these concerns, two modifications to the PCM were made.  First a gelling agent was added.  This gelling agent results in a solid-gel phase change and a form-stable PCM that will not leak.  Second, graphene nanoplatelets were added to the mixture.  These changes had no effect on the melting temperature, but reduced the energy storage capacity to about 180 J/g and increased the thermal conductivity by 100%.  Further, the addition of nanoparticles reduced the supercooling typical of organic PCMs by providing nucleation sites for freezing to begin. This reduces the difference between the melting and freezing temperatures, improving the temperature management capability of the PCM.

      Recent research on this subject has focused on the ability of these materials to resist radiation damage during use.  Two different PCMs were exposed to radiation using the Missouri S&T Reactor (MSTR) and the Missouri University Research Reactor (MURR).  The melting temperature and latent heat of the samples were measured before and after irradiation in order to evaluate the potential lifetime of the materials in a radiation environment.  The measurements found no significant change in these properties at radiation doses up to 2800 Gy, representing just under one year in a nuclear reactor containment or the approximate duration of a manned journey to Mars.

Steam Condensation for Reactor Safety Applications

      I was also a co-PI on a project for the Small Modular Reactor Research and Education Consortium (SMRrec).  This project involves the design and construction of a test facility to investigate scaling effects on condensation heat transfer in SMR passive cooling systems.  The Passive Containment Cooling System (PCCS) is one of the most important passive safety systems used in small modular reactors (SMRs). The containment vessel (CV) forms an integral part of the PCCS system.  At the time an accident is initiated, steam is released from the Reactor Pressure Vessel (RPV) into the CV.  This steam condenses on the CV walls.  This leads to condensation heat transfer from the RPV steam to the containment wall. The condensate is returned to the reactor core through drain lines.  It is well documented that the presence of even a small quantity of non-condensable gases (NCGs) greatly influences the condensation process. Research in the COndensation Rate for Passive Safety (CORPS) test facility aims to study the characteristics of heat transfer of a PCCS in the presence of non-condensable gases. Specific objectives for this research are:

• Review and evaluate existing data and models for condensation heat transfer for application to Westinghouse SMR (W-SMR) containment condensation
• Perform experiment and CFD simulations to evaluate the scalability to predict condensation heat transfer with and without NCGs.
• Evaluate and validate the effectiveness of the CFD simulations in scaling of condensation phenomena for different diameter pipes

The experiment will be compared with CFD predictions generated using STAR-CCM+.  It was observed that the software predicts the general trends of temperature distribution at various axial and radial locations.  However errors in the prediction of heat transfer coefficient were significant, as they ranged from 68% near the inlet to 38%. This can be attributed to the heat flux calculation method adopted by the software. Specifically the nucleation site density for condensation is a user-input parameter that has significant effects on the calculated heat transfer rate, however little to no guidance or framework exists for modeling or selecting this parameter.  This shortcoming limits the ability of STAR-CCM+ for predictive design calculations, where the nucleation site density is not known a priori.

Graduate Student Supervision

Ph.D. Degrees Completed:

Date	Student	Thesis/Dissertation Title	Current Position
May 2023	Sungje Hong	Validation of CFD Models for Interfacial Area Transport in Large Diameter Channels	Engineer Framatome
May 2023	Alexander Swearingen	Sensitivity of Numerical Modeling of Multiphase Hydrodynamics to Constitutive Models	Research Scientist Idaho National Laboratory
December 2022	Monica Gehrig	Using Computational Methods to optimize High Heat Flux Component Thermal Performance in Magnetic Confinement Fusion Reactor Research	DOE Postdoctoral Fellowship Fusion Energy Sciences Oak Ridge National Laboratory
August 2021	Ryan Steere	Radiation Effects on Phase Change Material Performance	Research Scientist Bettis Atomic Power Laboratory
August 2021	Palash Bhowmik	Condensation Heat Transfer Rates in Passive Safety Systems	Research Scientist Idaho National Laboratory
August 2019	Hayder Al-Naseri	Bubble Dynamic Properties in Low Height Bubble and Slurry Bubble Column with Internals for Fischer-Tropsch Synthesis
May 2019	Hiralkumar Patel	Experimental Investigation of Liquid Contact in the Developing Post-Dryout CHF Flow Boiling Regime Using Surface Mounted Thermistors	Research Scientist Naval Nuclear Laboratory
August 2018	Chandller Mills	Measurement of Interfacial Area Concentration
May 2018	Rami Saeed	Advances in Phase Change Materials for Thermal Energy Storage	Research Scientist Idaho National Laboratory.

M.S. Degrees Completed:

Date	Student	Thesis/Dissertation Title	Current Position
December 2023	Emin Ozdem	Verification of ATHLET Predictions for SB-LOCA experiments in PUMA	Safety Engineer
August 2021	William McCauley	Non-Thesis	Reactor Operator Idaho National Laboratory
May 2021	Jacob Ford	Non-Thesis
May 2021	Murat Tuter	Severe Accident Analysis Using ATHLET-CD	Safety Engineer
May 2017	Varun Kalra	CFD Validation and Scaling of Condensation Heat Transfer
May 2017	Raymond Fanning	W-SMR Passive Safety Natural Convection Heat Exchanger
May 2016	Rami Saeed	Thermal Characterization of Phase Change Materials for Thermal Energy Storage	Research Scientist Idaho National Laboratory.