Nuclear Systems Volume II
Elements of Thermal Hydraulic Design
- Available for pre-order. Item will ship after November 24, 2021
Nuclear Systems Volume II: Elements of Thermal Hydraulic Design, Second Edition provides advanced coverage of a wide variety of thermal fluid systems and technologies in nuclear power plants, including discussions of the latest reactor designs and their thermal/fluid technologies. Beyond the thermal hydraulic design and analysis of the core of a nuclear reactor, the book covers other components of the nuclear power plant, such as the pressurizer, containment, and the entire primary coolant system.
Placing more emphasis on the appropriate models for small scale resolution of the velocity and temperature fields through computational fluid mechanics, the book shows how this enhances the accuracy of predicted operating conditions in nuclear plants. It introduces considerations of the laws of scaling and uncertainty analysis, along with a wider coverage of the phenomena encountered during accidents.
The book serves as a textbook for advanced undergraduate and graduate students taking courses in nuclear engineering, studying thermal/hydraulic systems in nuclear power plants.
Table of Contents
1. Formulation of the Reactor Thermal Hydraulic Design Problem 1.1 INTRODUCTION 1.2 POWER REACTOR HYDRAULIC CONFIGURATIONS 1.3 BOUNDARY CONDITIONS FOR THE HYDRAULIC PROBLEM 1.4 PROBLEMS TREATED IN THIS BOOK 1.5 FLOW IN SINGLE CHANNELS 1.5.1 UNHEATED CHANNEL 1.5.2 HEATED CHANNEL 1.6 FLOW IN MULTIPLE, HEATED CHANNELS CONNECTED ONLY AT PLENA 1.7 FLOW IN INTERCONNECTED, MULTIPLE HEATED CHANNELS 1.8 APPROACHES FOR REACTOR ANALYSIS 1.8.1 BWR AND LMR CORE ANALYSIS 1.8.2 PWR CORE ANALYSIS 1.9 LUMPED AND DISTRIBUTED PARAMETER SOLUTION APPROACHES PROBLEMS ACRONYMS 2. Scaling of Two-Phase Flows in Complex Nuclear Reactor Systems 2.1 INTRODUCTION 2.1.1 MOTIVATION FOR SCALING ACTIVITY 2.1.2 LIMITATIONS TO THE APPLICATION OF SCALING 2.2 SCOPE OF THIS CHAPTER 2.3 DIMENSIONAL ANALYSIS AND THE BUCKINGHAM PI THEOREM 2.3.1 MOTIVATION FOR USE OF THIS ANALYSIS AND THEOREM 2.3.2 BUCKINGHAM PI THEOREM METHODOLOGY 2.3.3 LIMITATIONS 2.4 LINEAR SCALING 2.4.1 DEFINITION 2.4.2 DEVELOPMENT 2.4.3 LIMITATIONS 2.5 VOLUME (POWER TO VOLUME) SCALING 2.5.1 DEFINITION 2.5.2 DEVELOPMENT 2.5.3 TEST FACILITIES 2.6 ZUBER SCALING CONTRIBUTIONS 2.6.1 ZUBER’S PERSPECTIVE 2.6.2 HIERARCHICAL TWO-TIERED SCALING (H2TS) 2.7 ISHII SCALING 2.7.1 BACKGROUND 2.7.2 THREE-LEVEL SCALING 2.7.3 ADVANTAGES OF THREE-LEVEL SCALING 2.7.4 TEST FACILITIES 2.7.5. ILLUSTRATIVE EXAMPLES 2.8 MODIFIED LINEAR SCALING 2.8.1 GOALS 2.8.2 COMPARISON TO OTHER SCALING APPROACHES 2.8.3 LIMITATIONS 2.9 FRACTIONAL SCALING ANALYSIS (FSA) 2.9.1 THE FSA APPROACH 2.9.2 QUANTITATIVE PHENOMENA RANKING 2.10 DYNAMICAL SYSTEM SCALING 2.10.1 DYNAMICAL SYSTEM SCALING METHODOLOGY FUNDAMENTALS 2.10.2 THE PROCESS METRIC 2.10.3 SIMILARITY CRITERIA PROBLEMS ACRONYMS DEFINITIONS REFERENCES 3 Single, Heated Channel Transient Analysis 3.1 SIMPLIFICATION OF TRANSIENT ANALYSIS 3.2 SOLUTION OF TRANSIENTS WITH APPROXIMATIONS TO THE MOMENTUM EQUATION 3.2.1 SECTIONALIZED, COMPRESSIBLE FLUID (SC) MODEL 3.2.2 MOMENTUM INTEGRAL MODEL (MI) - INCOMPRESSIBLE BUT THERMALLY EXPANDABLE FLUID 3.2.3 SINGLE MASS VELOCITY (SV) MODEL 3.2.4 THE CHANNEL INTEGRAL (CI) MODEL 3.3 SOLUTION OF TRANSIENTS BY THE METHOD OF CHARACTERISTICS (MOC) 3.3.1 BASICS OF THE METHOD 3.3.2 APPLICATIONS TO SINGLE-PHASE TRANSIENTS 3.3.3 APPLICATIONS TO TWO-PHASE TRANSIENTS PROBLEMS ACRONYMS REFERENCES 4 Multiple Heated Channels Connected Only at Plena 4.1 INTRODUCTION 4.2 GOVERNING ONE-DIMENSIONAL, STEADY STATE FLOW EQUATIONS 4.2.1 CONTINUITY EQUATION 4.2.2 MOMENTUM EQUATION 4.2.3 ENERGY EQUATION 4.3 STATE EQUATION 4.4 APPLICABLE BOUNDARY CONDITIONS 4.4.1 CHANNEL BOUNDARY CONDITIONS 4.4.2 PLENA HEAT TRANSFER BOUNDARY CONDITIONS 4.5 THE GENERAL SOLUTION PROCEDURE 4.6 CHANNEL HYDRAULIC CHARACTERISTICS 4.6.1 THE FRICTION-DOMINATED REGIME 4.6.2 THE GRAVITY-DOMINATED REGIME 4.7 COUPLED CONSERVATION EQUATION: SINGLE-PHASE, NONDIMENSIONAL SOLUTION PROCEDURE 4.7.1 DERIVATION OF A SINGLE, COUPLED MOMENTUM–ENERGY EQUATION 4.7.2 NONDIMENSIONAL EQUATIONS 4.7.3 ONSET OF MIXED CONVECTION (UPFLOW) 4.7.4 ADIABATIC CHANNEL FLOW REVERSAL 4.7.5 STABILITY OF COOLED UPFLOW 4.7.6 STABILITY OF HEATED DOWNFLOW 4.7.7 PREFERENCE FOR UPFLOW 4.7.8 LIMITS OF THE SOLUTION PROCEDURE OF SECTION 4.7 4.8 DECOUPLED CONSERVATION EQUATION: ANALYTICAL SOLUTION PROCEDURE FOR HIGH FLOW RATE CASES 4.8.1 PRESCRIBED CHANNEL PRESSURE DROP CONDITION: SOLUTION PROCEDURE 4.8.2 PRESCRIBED TOTAL FLOW CONDITION: SOLUTION PROCEDURE PROBLEMS REFERENCES 5 Analysis of Interacting Channels by the Porous Media Approach 5.1 INTRODUCTION 5.2 APPROACHES TO OBTAINING THE RELEVANT EQUATIONS 5.3 FUNDAMENTAL RELATIONS 5.3.1 POROSITY DEFINITIONS 5.3.2 THEOREMS 5.4 DERIVATION OF THE VOLUME-AVERAGED MASS CONSERVATION EQUATION 5.4.1 SOME USEFUL DEFINITIONS OF AVERAGES 5.4.2 DERIVATION OF THE MASS CONSERVATION EQUATION: METHOD OF INTEGRATION OVER A CONTROL VOLUME 5.4.3 DERIVATION OF THE MASS CONSERVATION EQUATION: APPLICATION OF CONSERVATION PRINCIPLES TO A VOLUME CONTAINING DISTRIBUTED SOLIDS 5.5 DERIVATION OF THE VOLUMETRIC AVERAGED LINEAR MOMENTUM EQUATION 5.6 DERIVATION OF THE VOLUMETRIC AVERAGED EQUATIONS OF ENERGY CONSERVATION 5.6.1 ENERGY EQUATION IN TERMS OF INTERNAL ENERGY 5.6.2 ENERGY EQUATION IN TERMS OF ENTHALPY 5.7 CONSTITUTIVE RELATIONS 5.8 CONCLUSION PROBLEMS REFERENCES 6 Interacting Channels - Subchannel Analysis 6.1 INTRODUCTION 6.2 CONTROL VOLUME SELECTION 6.3 DEFINITIONS OF TERMS IN THE SUBCHANNEL APPROACH 6.3.1 GEOMETRY 6.3.2 MASS FLOW RATES 6.3.3 AXIAL MASS FLOW RATE 6.3.4 TRANSVERSE MASS FLOW RATE PER UNIT LENGTH 6.3.5 MOMENTUM AND ENERGY TRANSFER RATES 6.4 DERIVATION OF THE SUBCHANNEL CONSERVATION EQUATIONS: METHOD OF SPECIALIZATION OF THE POROUS MEDIA EQUATIONS 6.4.1 GEOMETRIC RELATIONS 6.4.2 CONTINUITY EQUATION 6.4.3 ENERGY EQUATION 6.4.4 AXIAL LINEAR MOMENTUM EQUATION 6.4.5 TRANSVERSE LINEAR MOMENTUM EQUATION 6.5 APPROXIMATIONS INHERENT IN THE SUBCHANNEL APPROACH 6.6 COMMONLY USED FORMS OF THE SUBCHANNEL CONSERVATION EQUATIONS 6.6.1 DEFINITIONS 6.6.2 THE COBRA CONTINUITY EQUATION 6.6.3 THE COBRA ENERGY EQUATION 6.6.4 THE COBRA AXIAL MOMENTUM EQUATION 6.6.5 THE COBRA TRANSVERSE MOMENTUM EQUATION 6.7 CONSTITUTIVE EQUATIONS 6.7.1 SURFACE HEAT TRANSFER COEFFICIENTS (PARAMETER 1) AND AXIAL FRICTION AND DRAG (PARAMETER 4) 6.7.2 ENTHALPY (PARAMETER 3) AND AXIAL VELOCITY (PARAMETER 6) TRANSPORTED BY PRESSURE-DRIVEN CROSS-FLOW 6.7.3 TRANSVERSE FRICTION AND FORM DRAG COEFFICIENT (PARAMETER 7) 6.7.4 TRANSVERSE CONTROL VOLUME ASPECT RATIO (PARAMETER 8) 6.7.5 EFFECTIVE CROSS-FLOW RATE FOR MOLECULAR AND TURBULENT MOMENTUM AND ENERGY TRANSPORT (PARAMETERS 2 AND 5) 6.8 BEYOND THE FUNDAMENTALS OF SUBCHANNEL ANALYSIS METHODOLOGY OF SECTIONS 6.1 TO 6.7 6.9 APPLICATION OF THE SUBCHANNEL APPROACH TO CORE ANALYSIS 6.9.1 THE MULTISTAGE AND ONE-STAGE METHODS FOR CORE THERMAL HYDRAULIC SUBCHANNEL ANALYSIS 6.9.2 MULTIPHYSICS SIMULATION OF CORE PERFORMANCE PROBLEMS ACRONYMS REFERENCES 7 Flow Loops 7.1 INTRODUCTION 7.2 LOOP FLOW EQUATIONS 7.3 STEADY STATE, SINGLE-PHASE, NATURAL CIRCULATION 7.3.1 DEPENDENCE ON ELEVATIONS OF THERMAL CENTERS 7.3.2 FRICTION FACTORS IN NATURAL CONVECTION 7.4 STEADY STATE, TWO-PHASE, NATURAL CIRCULATION 7.5 LOOP TRANSIENTS 7.5.1 SINGLE-PHASE LOOP TRANSIENTS 7.5.2 TWO-PHASE LOOP TRANSIENTS 7.5.3 DETAILED PUMP REPRESENTATION PROBLEMS ACRONYMS REFERENCES 8 Steady State and Transient Analysis of Centrifugal Pumps 8.1 INTRODUCTION 8.2 CENTRIFUGAL PUMP PERFORMANCE 8.2.1 STEADY STATE OPERATION OF CENTRIFUGAL PUMPS 8.2.2 PUMP CHARACTERISTIC CURVE VERSUS SYSTEM CURVE 8.2.3 PUMP EFFICIENCY, BRAKE AND HYDRAULIC HORSEPOWER 8.2.4 PREVENTION OF PUMP CAVITATION – NPSH 8.2.5 REQUIRED VERSUS AVAILABLE NPSH 8.2.6 NPSH OF ECCS PUMPS FOLLOWING LOCA 8.2.7 PUMP SIMILARITY RULES 8.3 TRANSIENT ANALYSIS OF REACTOR COOLANT PUMPS 8.3.1 IMPELLER SPEED FOLLOWING LOSS OF POWER TO OPERATING PUMP 8.3.2 LOOP FLOW TRANSIENT 8.3.3 SIMPLIFICATIONS OF LOOP MOMENTUM EQUATION 8.3.4 NON DIMENSIONALIZATION OF IMPELLER ANGULAR MOMENTUM EQUATION 8.3.5 SOLUTION OF FLOW DECAY FOLLOWING PUMP TRIP 8.3.6 FLOW RATE FOLLOWING PUMP STARTUP 8.3.7 PUMP MATHEMATICAL MODEL FOR PLANT EVENTS 9 Transient Analysis of PWR Pressurizer 9.1 INTRODUCTION 9.2 PRESSURIZER DESCRIPTIONS 9.2.1 PRESSURIZER SURGE LINE 9.3 PRESSURIZER FUNCTIONS 9.3.1 PRESSURIZER HEATERS 9.3.2 PRESSURIZER SAFETY AND RELIEF VALVES 9.3.3 PRESSURIZER SPRAY 9.3.4 CHEMICAL AND VOLUME CONTROL SYSTEM 9.3.5 PRESSURIZER CONTROL SYSTEM 9.3.6 PRESSURIZER RESPONSE TO TRANSIENTS 9.4 FORMULATION FOR TRANSIENT ANALYSIS 9.4.1 MODELING APPROACH 9.4.2 PROCESSES CROSSING CONTROL SURFACE 9.4.3 APPLICATION OF CONSERVATION EQUATIONS – CONTINUITY 9.4.4 APPLICATION OF CONSERVATION EQUATIONS – ENERGY 9.4.5 CLOSURE BY CONSTITUTIVE EQUATION – VOLUME CONSTRAINT 9.4.6 SOLUTION OF THE SET OF EQUATIONS 9.4.7 INTEGRATION OF THE STATE VARIABLES 9.5 EVALUATION OF CONSTITUTIVE EQUATIONS 9.5.1 WALL HEAT TRANSFER 9.5.2 CONDENSATION IN PRESSURIZER 9.5.3 MAIN SPRAY FLOW RATE 9.5.4 FLOW THROUGH SAFETY AND RELIEF VALVES 9.5.5 SURGE FLOW RATE 9.5.6 PRESSURIZER HEATER 9.5.7 PRESSURIZER WATER LEVEL 9.5.8 EXCHANGES AT THE BULK INTERFACE 9.6 CLASSIFICATION OF RCS BREAK SIZES 9.6.1 TOTAL LOSS OF FEEDWATER AND ONCE-THROUGH CORE COOLING 9.6.2 THE THREE MILE ISLAND ACCIDENT PROBLEMS ACRONYMS & ABBREVIATIONS REFERENCES 10 Transient Thermal Hydraulic Analysis of Containment 10.1 INTRODUCTION 10.2 TYPES OF CONTAINMENT BUILDINGS 10.3 DESIGN BASIS ACCIDENT (DBA) 10.3.1 LOCA EVALUATION 10.3.2 MSLB EVALUATION 10.4 CONTAINMENT DESIGN LIMITS 10.4.1 CONTAINMENT PRESSURE 10.4.2 CONTAINMENT TEMPERATURE 10.5 MIXTURE OF NON-REACTIVE IDEAL GASES 10.6 CONTAINMENT RESPONSE TO THERMAL LOADS 10.6.1 FORCING FUNCTIONS –FLOW RATES OF MASS AND ENERGY 10.6.2 CONSERVATIONS OF MASS AND ENERGY FOR CONTAINMENT 10.6.3 ALTERNATIVE SOLUTION OF CONTAINMENT EQUATIONS 10.7 PARTITION OF BREAK FLOW 10.8 PHASE CHANGE, POOL – ATMOSPHERE INTERACTION 10.8.1 PROCESSES AT THE VAPOR – LIQUID INTERFACE 10.9 HEAT CONDUCTORS HEAT TRANSFER 10.9.1 CONDENSATION HEAT TRANSFER COEFFICIENT – HEAT CONDUCTORS 10.10 SIMPLE RELATION BETWEEN LOCA ENERGY, PPEAK AND VC 10.11 EQUIPMENT QUALIFICATION 10.12 EFFECT OF DEBRIS ON LONG TERM COOLING 10.12.1 DEBRIS DEFINITION 10.12.2 ECCS FUNCTION 10.12.3 DEBRIS EFFECTS 10.12.4 DEBRIS EFFECTS AT SUMP STRAINER 10.13 CONTAINMENT ANALYSIS COMPUTER CODES PROBLEMS ACRONYMS REFERENCES 11 Analysis of Steam Generators and Condensers 11.1 INTRODUCTION 11.1.1 TYPES OF PWR STEAM GENERATORS 11.1.2 FLOW PATH IN VERTICAL UTSG AND OTSG 11.1.3 DEGREE OF SUBCOOLING AND DEGREE OF SUPERHEAT 11.2 STEAM GENERATOR CONTROL SYSTEM 11.3 STEAM GENERATOR TUBE INTEGRITY 11.3.1 TUBE FAILURE MECHANISMS 11.3.2 VERTICAL VERSUS HORIZONTAL SG 11.3.3 STEAM GENERATOR TUBE RUPTURE EVENT 11.4 ANALYSIS OF PWR OTSG 11.4.1 ONSET OF NUCLEATE BOILING AND SATURATION 11.4.2 HEAT EXCHANGER ANALYSIS – SG ECONOMIZER REGION 11.4.3 TEMPERATURE PROFILE – SG EVAPORATOR REGION 11.5 THERMAL DESIGN OF PWR UTSG 11.6 PWR UTSG DESIGN OPTIMIZATION 11.6.1 UTSG COST COMPONENTS 11.7 SG DRYOUT AND ESTIMATION OF TIME TO UNCOVER CORE 11.8 TRANSIENT ANALYSIS OF UTSG 11.8.1 SG TRANSIENT MODEL 11.8.2 TUBE TEMPERATURE DISTRIBUTION 11.9 ANALYSIS OF POWER PLANT CONDENSER PROBLEMS ACRONYMS & ABBREVIATIONS REFERENCES 12 Fundamentals of Reactor Transient Simulation 12.1 INTRODUCTION 12.2 LUMPED PWR MODEL 12.2.1 ALLOCATION OF CV 12.2.2 PROCESSES CROSSING CONTROL SURFACE 12.2.3 BALANCING EQUATIONS AND UNKNOWNS 12.2.4 FORMULATION OF PROCESSES IN THE RCS AND PZR 12.3 DATA DESCRIPTION AND PREPARATION 12.3.1 RCS FLOW RATE 12.3.2 RCS TEMPERATURE DISTRIBUTION 12.3.3 SG SECONDARY SIDE 12.3.4 HEAT TRANSFER COEFFICIENT 12.4 PWR DETAILED NODALIZATION 12.5 APPROACHES IN FORMULATING VARIOUS THERMAL HYDRAULIC MODELS 12.5.1 THREE-EQUATION MODEL 12.5.2 FOUR-EQUATION MODEL 12.5.3 FIVE-EQUATION MODEL 12.6 TRANSPORT EQUATIONS – SINGLE PHASE FLOW 12.7 TWO-FLUID MODEL 12.7.1 SIX-EQUATION MODEL – SINGLE PRESSURE 12.7.2 SEVEN-EQUATION MODEL – TWO PRESSURE 12.7.3 CONSTITUTIVE RELATIONS 12.8 SOLUTION METHOD PROBLEMS ACRONYMS REFERENCES 13 Treatment of Uncertainties in Reactor Thermal Analysis 13.1 OVERVIEW 13.2 SCOPE 13.3 STATISTICAL FUNDAMENTALS - ESTIMATION OF DISTRIBUTION PROPERTIES 13.3.1 ESTIMATING THE MEAN AND STANDARD DEVIATION OF DISTRIBUTIONS 13.3.2 THE NORMAL DISTRIBUTION 13.3.3 CONFIDENCE LEVEL 13.3.4 ESTIMATING THE POPULATION MEAN 13.3.5 ESTIMATING THE POPULATION STANDARD DEVIATION 13.4 FUNDAMENTALS OF DETERMINISTIC APPROACHES 13.4.1 DETERMINISTIC APPROACHES – FORWARD SENSITIVITY ANALYSIS 13.4.2 DETERMINISTIC APPROACHES – ADJOINT SENSITIVITY ANALYSIS 13.4.3 DETERMINISTIC APPROACHES – RELATIONSHIP TO SENSITIVITY ANALYSIS 13.5 RELEVANT FUNDAMENTALS – STATISTICAL-BASED APPROACHES 13.5.1 MONTE CARLO 13.5.2 ORDER STATISTICS USING WILKS’ FORMULA 13.5.3 CSAU (CODE SCALING, APPLICABILITY AND UNCERTAINTY) 13.5.4 METHOD OF EXTRAPOLATION OF OUTPUT UNCERTAINTIES (CIAU) 13.6 HOT SPOTS AND SUBFACTORS 13.7 COMBINATIONAL METHODS: SINGLE HOT SPOT IN CORE 13.7.1 DETERMINISTIC METHOD FORMULATIONS 13.7.2 STATISTICAL METHOD FORMULATIONS 13.8 EXTENSION TO MORE THAN ONE HOT SPOT 13.9 OVERALL CORE RELIABILITY 13.9.1 METHODS THAT DO NOT DISTINGUISH BETWEEN THE CHARACTER OF VARIABLES 13.9.2 METHODS THAT DO DISTINGUISH BETWEEN THE CHARACTER OF VARIABLES 13.10 CONCLUSION PROBLEMS ACRONYMS REFERENCES Appendix A - Selected Nomenclature Appendix B - Physical and Mathematical Constants Appendix C - Unit Systems Appendix D - Mathematical Tables Appendix E - Thermodynamic Properties Appendix F - Thermophysical Properties of Some Substances Appendix G - Dimensionless Groups of Fluid Mechanics and Heat Transfer Appendix H - Multiplying Prefixes Appendix I - List of Elements Appendix J - Square and Hexagonal Rod Array Dimensions Appendix K - Parameters for Typical BWR-5 and PWR Reactors Appendix L - Discretization of Lumped Parameter Conservation Equations L.1 INTRODUCTION L.2 EULERIAN DESCRIPTION AND SIGN CONVENTION L.3 DISCRETIZATION L.3.1 CONSERVATION OF MASS EQUATION L.3.2 CONSERVATION OF ENERGY EQUATION – INTERNAL ENERGY L.3.3 CONSERVATION OF ENERGY EQUATION – ENTHALPY L.3.4 CONSERVATION OF LINEAR MOMENTUM EQUATION L.3.4.1 Averaging Rate of Change of CV Momentum Term L.3.4.2 Averaging Gravity Term L.3.4.3 Averaging Differential Static Pressure Term L.3.4.4 Averaging Frictional Pressure Loss Term L.3.4.5 Averaging Momentum Flux Term L.3.4.6 Discretized Momentum Equation L.3.5 SECOND LAW FORMULATION REFERENCES Appendix M - Proof of Local Volume-Averaging Theorems of Chapter 5
Neil E. Todreas is a Professor Emeritus in the Departments of Nuclear Science and Engineering and Mechanical Engineering at the Massachusetts Institute of Technology. He held the Korea Electric Power Corporation (KEPCO) chair in nuclear engineering from 1992 until his retirement to part- time activities in 2006. He served an 8-year period from 1981 to 1989 as the Nuclear Engineering Department Head. Since 1975, he has been a codirector of the MIT Nuclear Power Reactor Safety summer course, which presents current issues of reactor safety significant to an international group of nuclear engineering professionals. His area of technical expertise includes thermal and hydraulic aspects of nuclear reactor engineering and safety analysis. He started his career at Naval Reactors working on the Shippingport reactor and surface nuclear vessels after earning a BEng and an MS in Mechanical Engineering at Cornell University. Following his ScD in Nuclear Engineering at MIT, he worked for the Atomic Energy Commission (AEC) on organic-cooled/heavy water-moderated and sodium-cooled reactors until he returned as a faculty member to MIT in 1970. He has an extensive record of service for government (Department of Energy (DOE), U.S. Nuclear Regulatory Commission (USNRC) and national laboratories), utility industry review committees including INPO, and international scientific review groups. He has authored more than 250 publications and a reference book on safety features of light water reactors. He is a member of the U.S. National Academy of Engineering and a fellow of the American Nuclear Society (ANS) and the American Society of Mechanical Engineers (ASME). He has received the American Nuclear Society Thermal Hydraulics Technical Achievement Award, its Arthur Holly Compton Award in Education, and its jointly conferred (with the Nuclear Energy Institute) Henry DeWolf Smyth Nuclear Statesman Award.
Mujid S. Kazimi was a Professor in the Departments of Nuclear Science and Engineering and Mechanical Engineering at the Massachusetts Institute of Technology. He was the first director of the Center for Advanced Nuclear Energy Systems (CANES) and the Tokyo Electric Power Company (TEPCO) chair professor in Nuclear Engineering both at MIT. Prior to joining the MIT faculty in 1976, Dr. Kazimi worked for a brief period at the Advanced Reactors Division of Westinghouse Electric Corporation and at Brookhaven National Laboratory. At MIT, he was Head of the Department of Nuclear Engineering from 1989 to 1997 and Chair of the Safety Committee of the MIT Research Reactor from 1998 to 2009. He served from 1990 to 2015 as the codirector of the MIT Nuclear Power Reactor Safety summer course. He was active in the development of innovative designs of fuel and other components of nuclear power plants and in analysis of the nuclear fuel cycle options for sustainable nuclear energy. He co-chaired the 3-year MIT interdisciplinary study on the Future of the Nuclear Fuel Cycle, published in 2011. He served on scientific advisory committees at the U.S. National Academy of Engineering and several other national agencies and laboratories in the United States, Japan, Spain, Switzerland, Kuwait, the United Arab Emirates and the International Atomic Energy Agency. He authored more than 200 articles and papers that have been published in journals and presented at international conferences. Dr. Kazimi earned a BEng at Alexandria University in Egypt and an MS and PhD at MIT, all in Nuclear Engineering. He was a fellow of the American Nuclear Society and the American Association for the Advancement of Science. Among his honors is the Technical Achievement Award in Thermal Hydraulics by the American Nuclear Society and membership in the U.S. National Academy of Engineering.
Mahmoud Massoud is an adjunct Professor at the Department of Mechanical Engineering, University of Maryland, College Park, where he has taught undergraduate and graduate level engineering courses since 1988. Dr. Massoud has been active in the Nuclear Industry as a Principal Engineer at the Exelon Corporation and as a Consulting Engineer since 1989. He has authored a textbook in engineering and over 20 journal and conference papers. Dr. Massoud earned his B.S. and M.S. in Mechanical Engineering with emphasis in Internal Combustion Engines from Tehran University. Subsequently, he earned his M.S. and Nuclear Engineer’s degree from MIT’s Mechanical and Nuclear Engineering Departments, respectively. He earned his Ph.D. in Nuclear Engineering from the University of Maryland College Park.