Honors & Awards

1. Top Student Award, Alexandria University, 1981 - 1985
   
2. B.Sc. in Nuclear Engineering with Honors, 1985
3. Argonne National Laboratory Fellowship, 1989
4. George Westinghouse Signature Award of Excellence, 1993
5. Author Recognition Award of the ANS Winter Meeting, 1993
6. Westinghouse Commercial Nuclear Fuel Division, Time Based Management Award for reduced design cycle time, 1998
7. Reactor Physics Division Best Paper Award of the ANS Winter Meeting, 1999
8. Two Westinghouse Core Engineering Creating Value Awards for excellence in teaching and customer responsiveness, 1999
9. Two Westinghouse Core Engineering Creating Value Awards for reduced design cycle time and customer responsiveness, 1999
10. Gold Process Improvement Award for fuel rod design process automation, 1999
11. Silver Product Innovation Award for development of fuel management code, ALPS, 1999
12. Westinghouse Core Engineering Creating Value Award for excellence in teaching, 2000
13. Reactor Physics Division Best Paper Award of the ANS Winter Meeting, 2000
14. Idaho National Engineering and Environmental Laboratory Faculty Fellowship, 2004
15. Idaho National Laboratory Faculty Fellowship, 2005

Education

Massachusetts Institute of Technology Cambridge, MA

D.Sc. degree in Nuclear Engineering, September 1992. Nuclear Reactor Physics thesis under Prof. A.F. Henry on "Simple Quadratic Nodal Model for Hexagonal Geometry”. A few group diffusion nodal code for Savannah-River-Laboratory-type of reactors. GPA: 5.0/5.0 

Alexandria University Egypt

M.Sc. degree in Nuclear Engineering, April 1988. Nuclear Radiation Shielding thesis on "Characteristics of the Produced Gamma-Radiation in A Medium Irradiated with Fast Neutrons”. A Monte Carlo neutron-photon transport code was developed to simulate neutron beam penetration of a steel slab. GPA: 5.0/5.0

Alexandria University Egypt

B.Sc. degree in Nuclear Engineering, June 1985.  Broad curriculum in nuclear and mechanical engineering with emphasis on nuclear reactor physics and engineering.  Thesis on "Decay Heat Removal After LOCA in CANDU”. Top Student Award for five consecutive years.  GPA: 5.0/5.0

Experience 

Massachusetts Institute of Technology Cambridge, MA (January 2007 - Present)

Visiting Professor, Department of Nuclear Science and Engineering.
Responsible for teaching and supervising graduate students’ research in reactor physics.

King Abdul Aziz University (KAU) Saudi Arabia (November 2002 to January 2007)

Associate Professor in the Department of Nuclear Engineering. Developed a Radiotherapy Planning optimization technique which was based on biased multi-objective function optimization. This technique was originally developed for in-core nuclear fuel management for Westinghouse PWRs. A 3-D Cartesian MATLAB script was written to find the optimum radiation beam configuration that satisfied dose objective functions for tumors and sensitive organs alike.

• Coordinator of ABET Response Team which is responsible for addressing EC2000 college-wide concerns in the area of Engineering Design. Responsible for developing, teaching and coordinating a new course, Introduction to Engineering Design, whose objectives include, Self Regulation, Communication, Critical Problem Solving, Working Collaboratively, Modeling and Quality. The course uses student teams, continuous improvement of the learning process and constructivist learning exercises in class.
• Principal Investigator in three university-funded research projects to standardize the implementation of and implement Active Learning in various engineering core courses.
• Coordinator and presenter of "Education Excellence through Active Learning Workshop”. Education professionals received and practiced a 2-day overview of how effective Active Learning can be in enhancing students’ overall learning experience and retention, Jeddah, 2005
• Coordinator and presenter of "An Active Learning Approach to Engineering Education for the 21st Century Workshop,” Al-Qassim University, Al-Qassim, 2006.
• Coordinator and presenter of "An Active Learning Approach to Language Education Workshop,” English Language Center, KAU, 2006.
• Coordinator and presenter of "An Active Learning Approach to Sciences Education Workshop,” Faculty of Sciences, KAU, 2006
• Coordinator and presenter of "An Active Learning Approach to Engineering Education for the 21st Century Workshop,” Department of Civil Engineering, KAU, 2006
• Coordinator and presenter of "Introduction to Engineering Design Short Course,” College of Engineering, Taibah University, 2006

Massachusetts Institute of Technology Cambridge, MA (Summer 2006)

Visiting Research Scientist, Department of Nuclear Science and Engineering.

Elements of the new Global Nuclear Energy Partnership (GNEP) in the USA call for the expansion of domestic use of nuclear power and the minimization of nuclear waste. To achieve both goals in the short term the transmutation of trans-uranic (TRU) elements in Combined Non-Fertile and UO2 (CONFU) Generation-III PWR assemblies was evaluated. These assemblies were composed of a mix of standard UO2 fuel pins and pins made of recycled TRU in an inert matrix and are designed to fit in currently deployed PWRs. In this study, a CONFU-Self-Contained (CONFU-C) assembly was shown to achieve a net TRU destruction in a self-contained TRU multi-recycling system. The system might consist of as little as one currently operating reactor that did not depend on other reactors to supply it with its inventory of recycled TRU. This was considered a major advantage of the new design over its predecessors since it eliminated the need for designating a whole fleet of CONFU reactors to produce recycled TRU for the reactor under consideration. Degradation of fissile content of the multi-recycled TRU was compensated for by drawing from legacy TRU that already comes from standard UO2 spent fuel and the usage of fresh UO2 fuel with different enrichments depending on fuel cooling time after discharge. One recycling strategy which used a 4.5-year in core irradiation, followed by one of three cooling periods (6-, 18-, and 32-year) after discharge and reprocessing was considered. Calculations showed the equilibrium CONFU-C assembly could have a net TRU destruction of approximately 20.6 (for the 6-yr cooling) and 2.7 (for the 18-yr cooling) kg of TRU per TWhe. This represented a net burning rate of ~13% (6-yr cooling) and 3% (18-yr cooling) of the TRU loaded per assembly. However, Fuel Cycle Costs for the equilibrium CONFU-C was shown to be 12.8 (6yr-cooling) and 14.2 (18yr-cooling) mills/KWhe. Due to the relatively long period of the third option (32-yr cooling), a CONFU-C assembly could not be designed to achieve net TRU destruction in a self-contained manner.

To support the achievement of two elements of the GNEP to minimize nuclear waste and develop Advanced Burner Reactors, a new steady-state fast neutron Advanced Burner Test Reactor was conceptually designed. This study presented a new concept for a CANDU-based fast test reactor that was horizontal in orientation, with individual pressure/calandria tubes (PT/CT) running the entire length of the Calandria which was filled with Lead-Bismuth-Eutectic (LBE). This approach for a test reactor would provide more flexibility in refueling, sample removal, and ability to completely reconfigure the core to meet different users’ requirements. Full core neutronic analysis showed a small hexagonal, LBE-cooled, TRU-67Zr fuel with HT-9 cladding and structures, with a core power of 100 MWth produced a fast flux of 1.4x1015 n/cm2 sec averaged over the whole length of six irradiation channels with a total testing volume of more than 77 liters. Loading of TRU from legacy UO2 spent fuel allowed core continuous operation for 180 effective full power days with a net TRU burning rate of 6.4% which was calculated using MCNP-4C/MCODE2. Since high neutron fluence impact on PT/CT might be an issue of concern for this design, oxide dispersion strengthened (ODS) ferritic steel was used as PT/CT material without any impact on core neutronic behavior. ODS ferritic steels have high radiation resistance and superior creep rupture strength at elevated temperature due to the finely distributed Y2O3 particles in the ferritic matrix. A preliminary thermal hydraulic and safey analysis of LBE produced acceptable results.

ABET Baltimore, MD

October 2005

Program Evaluator (PEV) of the Accreditation Board for Engineering and Technology (ABET). Evaluated the Nuclear Engineering and Radiological Sciences (NERS) program of the University of Michigan, Ann Arbor. Evaluation was based on Engineering Criteria (EC-2000).

Idaho National Laboratory Idaho Falls, ID

Summer 2005

Faculty Fellow in the Nuclear and Energy Systems Division. The original concept of the CANDU-based Fast Irradiation Test Reactor (FITR) was further refined. Reactor total power was reduced to 100 MWth while the fast flux value reached 1.5x1015 averaged over the total irradiation volume of about 77 liters with an active length of 135 cm. The new design had a lattice much tighter than the regular CANDU and is cooled by lead-bismuth-eutectic (LBE) or helium The fuel used was U-Pu-10Zr with 15.7% weapons-grade plutonium. In addition to the CANDU attractive features, the new concept had a unique shutdown/control system that enabled the removal of the old control-rod-based system. The new shutdown/control system consists of boric acid injected into the core through un-fuelled pressure tubes without perturbing the axial flux shape resulting in less limiting safety analysis. The removal of the old control rods allowed for employing tighter pressure tube pitch and ultimately the freedom to reconfigure the irradiation channel area for enhanced fast flux values. Preliminary thermal-hydraulic analysis was performed and proved the validity of both coolants although LBE was found to be superior to helium. Basic neutronic safety analysis calculations were also performed to verify the design safe performance in case of accidents.

A Very High Temperature Reactor Self-Sustainable Oasis (VHTR-SSO) for electricity, fresh water, and hydrogen production has been conceptually designed. The SSO uses a 600 MWth VHTR to drive the power cycle and generate net electricity, after subtracting what is needed by High Temperature Electrolysis of Steam (HTES), of 176 MWe. In the current configuration, about 100 MWe is sent to the power grid and the rest is used in a hybrid MSF-RO desalination plant to produce fresh water at a rate of 182,400 m3/d. Waste heat from the VHTR power cycle is used to pre-heat inlet seawater streams for the hybrid Multi-Stage Flash Reverse Osmosis (MSF-RO) plant to increase the water-to-power ratio and consequently increase the water production efficiency. Hydrogen is produced at a rate of 1.185 kg/s via HTES that consumes electricity and heat from the VHTR. Hydrogen is produced at an efficiency of ~50% and is sufficient to fuel approximately 280,000 light vehicles. The VHTR-SSO can be barge-mounted to facilitate its usage in remote costal regions or islands where electricity, energy sources, and fresh water is severely lacking.

Idaho National Engineering and Environmental Laboratory, Idaho Falls

Summer 2004

Faculty Fellow in the Nuclear and Energy Systems Division. Developed a new design for the FTR based on a modified CANDU design. The new design has a lattice tighter than the regular CANDU and fueled with U-10Zr (20% enrichment) to meet a 1015 fast flux target in two irradiation channels. It is cooled by helium or steam and submerged in a pool of Lead-Bismuth-Eutectic as a scattering medium. The on-line refueling feature has been proven very attractive for conducting experiments and loading the irradiation channels without the need to shutdown. MCNP5/MOCUP code package was used to prove the neutronics validity of the new design and compare its performance to various combinations of fuel materials, coolants, and core layouts.

Developed a novel nodal method for Pebble Bed Reactor (PBR) analysis. The new method is based on Coarse Mesh Finite Difference (CMFD) method using Green’s Function to solve the two-node problem in cylindrical geometry. New algorithm has been developed and implemented in INEEL’s PEBBED code that is used for evaluating the PBR option of the US DOE GEN IV initiative.

Westinghouse Electric Corporation, Pittsburgh, PA

June 2000 to November 2002

Principal Engineer in the Core Analysis A Group, Nuclear Fuel Business Unit. Lead Engineer for core analysis tasks for Diablo Canyon, Unit 2. Responsible for leading a team to perform fuel management optimization, plant safety analysis, and generate nuclear design reports, and nuclear plant operational packages. Responsibilities included teaching nuclear reactor physics introductory courses to utility personnel. Lead Engineer for a developmental program to institute the usage of new neutron burnable absorbers (BA), Erbium and Gadolinium, in nuclear reactor fuel in conjunction with the already-successful Integrated Fuel BA (IFBA) product. Responsibilities included leading a team to resolve technical difficulties associated the implementation of the new BA and interfacing with several functional groups to resolve any common problems.

Westinghouse Electric Corporation, Pittsburgh, PA

February 1996 to June 2000

Senior Engineer in the Core Analysis A Group, Commercial Nuclear Fuel Division. Responsible for supporting core analysis tasks for Diablo Canyon, Beaver Valley and D.C. Cook units. Exposure to performing fuel management optimization, and plant safety analysis, generating nuclear design reports, and nuclear plant operational packages which served as strong work experience to help develop more effective solutions to real-life core design problems. Responsibilities included teaching nuclear reactor physics introductory courses to utility personnel. Serve as Lead Engineer to develop and enhance the loading pattern optimization code, ALPS, by introducing heuristics and biased multi-objective function combinatorial optimization in the loading pattern search logic.

Westinghouse Electric Corporation, Pittsburgh, PA

Summer 1994 to January 1996

Senior Engineer B in the BEACON Systems and Methods Group, Commercial Nuclear Fuel Division. Lead engineer to develop an analytical method to accurately capture mixed-oxide fuel spectral interaction with UO2 PWR fuel which eventually helped improve dramatically the accuracy of predicting core behavior loaded with both types of fuel.

Alexandria University, Egypt

Fall 1993 to Summer 1994

Assistant professor in the Department of Nuclear Engineering. Responsible for teaching Nuclear Reactor Theory and Physics courses on the under-graduate as well as the post-graduate levels. Responsible for supervising post-graduate research in the field of Nuclear Reactor Analysis.

Westinghouse Electric Corporation, Pittsburgh, PA

Fall 1992 to Summer 1993

Engineer in the Product and Methods Development Group, Commercial Nuclear Fuel Division. Responsible for nodal methods development for core analysis to support core design activities in the Core Engineering Department. Developed a diffusion nodal method based on the Nodal Expansion Method for hexagonal geometries as a part of the fuel contract for the Czech power plant at the Temelin site. The method employs a unique technique of conformally mapping the hexagon into a rectangle which then can be treated as a regular PWR assembly.

Massachusetts Institute of Technology Cambridge, MA

Fall 1990 to Spring 1992

Research Assistant for a DOE-sponsored project to develop a nodal method for analyzing D2O reactors located in Savannah River Lab. Static and transient codes have been developed for a hex-z geometry in which the face-averaged flux was assumed to take a quadratic shape. A variety of numerical analysis  methods were used to create a fast-converging stable iteration schemes. A WIGL thermal hydraulic feedback model was incorporated to help analyze more realistic transients.

Nuclear Fuel Industries, Ltd. Tokyo, Japan

Summer 1990

Research Associate in the Core Engineering Department. Theoretical model and computer programming were completed for the analysis of Reactivity Initiated Accidents RIA for licensing a 9*9 BWR assembly. Numerical verification of the code developed was done against a point-reactor and 1-D problems.

Massachusetts Institute of Technology Cambridge, MA

Spring 1990

Finished neutronic calculations for the MIT reactor MITR-III. Design will replace the existing reactor, MITR-II, with TRIGA fuel to meet MIT research requirements. TRIGA fuel assemblies were used with D2O as a moderator and a coolant in a multigroup-diffusion code, VENTURE, to get the required flux-peaking values at the MITR-III experimental ports.

Massachusetts Institute of Technology Cambridge, MA

Spring 1989 to Spring 1990

Research Assistant for a DOE-sponsored project on "Safety Characteristics of Operating Light Water Reactors of Western Design”. Presentation and comparison of the safety characteristics of operating light water reactors of Western design in the areas of Reactivity Control, Fuel Design, Pressure Control, Reactor Vessel Materials, Water Level Measurement, Power Sources, Electric Networks, Emergency Water Makeup, Post-Accident Heat Removal, Containment Design, and Severe Accident Features. Chronological listing by country, Family Trees, of all plants of a given standard design have been prepared.

Massachusetts Institute of Technology Cambridge, MA

Fall 1988

Teaching Assistant of "Nuclear Reactor Physics (I)”. The course introduces problems of fission-reactor physics, covering nuclear reactions induced by neutrons, nuclear fission, slowing down of neutrons in infinite media, diffusion theory, the few-group approximation, and point kinetics. Emphasizes the nuclear physical bases of reactor design and their relation to reactor engineering problems.

Alexandria University Egypt

Spring 1987 to Spring 1988

Lecturer of Nuclear and Atomic Physics. Responsibilities included teaching, tutoring, preparing, and grading exams for freshmen Atomic Physics and sophomore Nuclear Physics courses in the Department of Nuclear Engineering.

Alexandria University Egypt

Fall 1984 to Spring 1985

1-D heat transfer code was developed to investigate CANDU’s fuel bundle behavior in case of LOCA. The fail-safe feature of the CANDU reactor was emphasized by controlling the pressure and calandria tubes emissivities.

Publications

(PDF)