Education

PhD, Nuclear Engineering, Massachusetts Institute of Technology, 2000.
B.S., Nuclear Engineering, Polytechnic of Milan, 1996.

Teaching

Engineering of Nuclear Reactors (22.312), Thermal Hydraulics in Power Technology (22.313J), Thermal-Fluids Engineering II (2.006); MIT-INPO Reactor Technology Course for Utility Executives (Co-Director)

Honors and Awards

• ASME Heat Transfer Division Best Paper, 2008.
• Best Paper Award at the 1st ASME Micro/Nanoscale Heat Transfer International Conference, Tainan, Taiwan, January 6-9, 2008
• Junior Bose Award for Excellence in Teaching, School of Engineering, 2007
• PAI Outstanding Teaching Award (presented by the student chapter of the American Nuclear Society), 2006 and 2007
• The Ruth and Joel Joel Spira Award for Distinguished Teaching, 2006
• Graduate Teaching Award, 2005
• Norman C. Rasmussen Career Development Chair, 2004
• ANS Mark Mills Award, 2001
• Best Technical Paper Award, Thermal-Hydraulic Track at the 8th International Conference on Nuclear Engineering (ICONE-8), 2000
• Manson Benedict Fellowship for Outstanding Academic Achievements, 2000
• Martin Society Fellow, 1999
• Alpha Nu Sigma Honor Society Member, 1999
• Best NED Teaching Assistant Award, 1998 and 1999

Publications (pdf)

 

Research Interests

Multi-phase flow and heat transfer; advanced reactor design; reactor thermal-hydraulic, neutronic and structural analysis.

Nanofluids for Nuclear Applications

By seeding the nuclear reactor coolant with nanoparticles it is possible to enhance the rate at which energy is removed from the nuclear fuel under normal and accident conditions, thus improving the reactor's economic and safety performance. The resulting particle-fluid system is called a 'nanofluid'. I have been studying the heat transfer and colloidal behavior of nanofluids, including boiling and quenching, resistance to nuclear irradiation, long-term stability, etc. This research program is in collaboration with the MIT Nuclear Reactor Laboratory, and has enjoyed the support of AREVA, the Nuclear Regulatory Commission, the National Science Foundation, the Department of Energy, the Idaho National Laboratory, and the King Abdulaziz City of Science and Technology, ABB and the Electric Power Research Institute (EPRI).

The area of nanofluid boiling has been the source of exciting results. The experiments in my lab indicate that significant critical heat flux (CHF) enhancement is possible with nanofluids at modest concentrations, and this enhancement was proven to occur for the first time also in a flow system. This work also cast light on the mechanism of CHF enhancement in nanofluids. Briefly, buildup of a porous layer of nanoparticles on the heater surface was observed during nucleate boiling, and it was shown that this layer significantly improves the surface wettability, which in turn increases greatly the CHF.

This video shows the occurrence of CHF, the transition from nucleate boiling to film boiling, on an electrically-heated wire submerged in water.

Another noteworthy piece of work in the nanofluid boiling area relates to quenching heat transfer. Quenching is the rapid cooling of a very hot solid object by exposure to a much cooler liquid. For example, following a loss of coolant accident in a light water reactor, the overheated nuclear fuel is quenched by cold water from the emergency core cooling system. Using metallic spheres and rodlets heated to ~1000°C and then plunged in cold nanofluids, it was shown that a layer of nanoparticles on the sphere and rodlet surface can prematurely destabilize film boiling, and greatly accelerate the quenching process.

Quenching of a steel rodlet in clean water proceeds through the development of a coherent and slow-moving boiling front (top), while quenching of the same rodlet in a nanofluid occurs via a much-faster global collapse of the vapor film (bottom) . Note the different time scales in the two sequences.

Among the possible nuclear applications of nanofluids that I have been exploring, one has proven particularly promising so far, i.e., mitigation of postulated severe accidents during which the core melts and relocates to the bottom of the reactor vessel. In such accidents it is desirable to retain the molten fuel within the vessel by removing the decay heat through the vessel wall. This process is limited by the occurrence of CHF on the vessel outer surface. Analyses indicate that using a nanofluid vs water as the coolant can improve the in-vessel retention capabilities of nuclear reactors by as much as 30-40%.

In 2007 I launched the International Nanofluid Property Benchmark Exercise (INPBE), to help eliminating the many inconsistencies in the nanofluid thermal conductivity database. This initiative has received an overwhelming response: 34 organizations from the US, UK, Italy, Germany, Belgium, France, Switzerland, Poland, Japan, South Korea, India, China, Singapore and Puerto Rico independently measured the thermal conductivity of identical control nanofluid samples, and reported the data to MIT. The results have shown that the effective medium theory can predict nanofluid thermal conductivity, drawing the controversy on this topic closer to an end. The effort was completed in early 2009 and a summary paper with >60 authors has been published in the J. Applied Physics.

Fundamentals of Boiling

Revolutionary advances in the analysis of boiling phenomena are within reach through the systematic use of multi-phase Computational Fluid Dynamics (CFD), specifically Interface Tracking Methods (ITM). With such methods the geometry of the vapor-liquid interface is not assumed (e.g., bullet-shaped bubbles), but actually calculated from ‘first principles’. Moreover, CFD can resolve (through Direct Numerical Simulation or Large Eddies Simulation) the velocity and temperature gradients near the interface, so prediction of the exchange of momentum and heat at the interface requires no empirical closure relations. The key issue is the availability of high-quality experimental data to validate the CFD codes. My lab has state-of-the art capabilities, i.e., infra-red thermometry, high-speed imaging and optical probe, that have been used to obtain detailed and fundamental time- and space-resolved data on pool boiling heat transfer, specifically bubble departure diameter and frequency, growth and wait times, nucleation site density and near-wall void fraction. If successful, the application of ITMs to the complex two-phase phenomena encountered in the core and other major components of nuclear reactors may result in significantly improved accuracy in the predictions of the reactor behavior. Elimination of unnecessary conservatism in the analysis could lead to power uprates and/or safety margin enhancement.

Schematic of the boiling facility with high-speed infra-red thermometry (top). The IR camera records the 2D time-dependent temperature distribution on an indium-tin-oxide (ITO) heater, laid over a transparent sapphire substrate. The technique allows for the direct measurement of nucleation sites and bubble departure frequency (video below).

Double-tip optical probe used to measure the near-wall void fraction during boiling.

Recent experiments have revealed that nano-engineered surfaces can enhance boiling heat transfer dramatically. However, the role of nano-scale structures in boiling heat transfer is not at all understood, since according to the traditional theory of boiling, it is the liquid-vapor-solid interactions at the micro-scale that dominate bubble nucleation and boiling heat transfer. Using custom-fabricated nano-patterned surfaces, studied with the high-speed infrared imaging technique and molecular dynamic simulations, we are probing the fundamental mechanisms of boiling heat transfer at the nanoscale.

High-Temperature Molten Salts for Nuclear and Solar Power Applications

Since the primary costs in nuclear or solar power systems are in the plant hardware, it is imperative that the heat-to-electricity conversion occurs with high efficiency in these systems. This requirement favors the use of a high-temperature heat transfer medium, and in the case of solar power, also a high-temperature storage medium. Molten salts are promising heat transfer and storage media, due to their high thermal capacity and relative chemical inertness at high temperatures, up to 1000°C. At such high temperatures, thermal radiation transport within the molten salts, which are semi-transparent media, becomes very significant, and must be accounted for in the thermal analysis of the system. Therefore, in the process of developing molten salt media for advanced high-temperature nuclear and solar energy applications, we are measuring the attenuation of light (from IR to UV) within various molten salts, e.g., chloride salts like NaCl-KCl and carbonate salts like Li2CO3-Na2CO3-K2CO3.

Looking down at a bath of molten (and transparent) KCl-NaCl salt at ~700°C (high-temperature cell deployed in my lab).

Development and Modeling of Advanced Nuclear Fuels

Increasing the core power density is an attractive way to improve the economics of current and future nuclear plants. We have developed an innovative fuel assembly design that affords a 20% power density increase in the Boiling Water Reactor (BWR) core while maintaining the safety margins of lower power reactors. The new assembly uses an optimized large square lattice with smaller fuel pins and control rods than traditional BWR assemblies, hence it was named the Large-Assembly Small-Pins concept, or LASP.

Schematic of the LASP concept (right) as compared to 4 traditional 9×9 BWR fuel assemblies (left)

A new code, the Fuel Engineering and Structural analysis Tool (FEAST), was developed to predict the irradiation performance of metal and oxide fuel in sodium-cooled fast reactors. FEAST comprises several modules working in coupled form to describe key irradiation phenomena, such as fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze the fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios.

Predictions of the FEAST code for fission gas release (top) and peak clad strain (bottom) in the X425 metal-fuel test conducted in the EBR-II reactor.