• CONTENTS
• 
  
• Preface
• List of Figures
• List of Tables
• 
  
• Chapter 1 Role for Nuclear Energy
• 1.1 Introduction
• 1.2 Nuclear Power Production
• 1.2.1 Transport of Radioactive Materials
• 1.2.2 Reprocessing
• 1.2.3 Waste Management
• 1.2.4 Nuclear Process Heat for Industry
• 1.3 Radiology and Nuclear Medicine
• 1.4 Security Devices
• 1.5 Food Irradiation
• 1.6 Industrial Uses (IAEA, 2004)
• 1.6.1 Gauging Devices
• 1.6.2 Well Logging
• 1.6.3 Industrial Radiography
• 1.6.4 Sterilization
• 1.7 Environmental Application (IAEA, 2007)
• 1.7.1 Flue Gas Treatment
• 1.7.2 Waste Water Treatment
• 1.7.3 Sewage Treatment
• 1.8 Risk Perception
• Summary
• Bibliography
• Assignments
• 
  
• Chapter 2 Nuclear Reactors
• 2.1 Introduction
• 2.2 Reactor Core
• 2.3 Coolant
• 2.4 Control Rods
• 2.5 Moderator
• 2.6 Other Core Components
• 2.7 Containment
• 2.8 Steam Generator
• 2.9 Turbine Generator
• 2.10 Steam/Water System
• 2.11 Fuel Handling
• 2.12 Spent Fuel Cooling
• 2.13 Emergency Core Cooling
• 2.14 Nuclear Reactor Types
• 2.14.1 Gas Cooled Graphite Moderated
• 2.14.2 Heavy Water Cooled and Moderated
• 2.14.3 Water Cooled and Moderated (PWR)
• 2.14.4 Water Cooled, Graphite Moderated (RBMK)
• 2.14.5 Fast Reactors
• Summary
• Bibliography
• Assignments
• 
  
• Chapter 3 Safety Principles
• 3.1 Introduction
• 3.2 Safety Objectives
• 3.3 Technical Aspects of Safety
• 3.3.1 Siting
• 3.3.2 Design and Construction
• 3.3.3 Commissioning
• 3.3.4 Operation and Maintenance
• 3.3.5 Radioactive Waste Management and Decommissioning
• 3.4 Defence in Depth
• 3.4.1 Multiple Barriers
• 3.4.2 Levels of Defence
• 3.5 Redundancy, Diversity and Independence
• 3.5.1 Redundancy
• 3.5.2 Diversity
• 3.5.3 Independence
• 3.6 Event Analysis
• 3.7 Core Inventory – Radionuclides
• 3.8 Mitigation of Radiological Consequences
• 3.8.1 On-site Emergency Response
• 3.8.2 Off-site Emergency Response
• Summary
• Bibliography
• Assignments
• 
  
• Chapter 4 Safety Approach
• 4.1 Introduction
• 4.2 Deterministic Safety Analysis
• 4.3 Postulated Initiating Events (PIE)
• 4.4 Design Basis Events (DBE)
• 4.4.1 Category-1 Events: Normal Operation and Operational Transients
• 4.4.2 Category-2 Events: Events of Moderate Frequency
• 4.4.3 Category-3 Events: Events of Low Frequency
• 4.4.4 Category-4 Events: Multiple Failures and Rare Events
• 4.4.5 Beyond Design Basis Events (BDBE)
• 4.5 Examples of Safety Analysis
• 4.6 Acceptance Criteria
• 4.7 Risk and Probabilistic Safety Analysis (PSA)
• 4.7.1 Fault Tree Analysis
• 4.7.2 Event Tree Analysis
• 4.7.3 Failure Rates
• Summary
• Bibliography
• Assignments
• 
  
• Chapter 5 Quality Assurance
• 5.1 Introduction
• 5.2 QAP Requirements
• 5.3 Classification of Plant Components
• 5.4 Issues for Consideration
• 5.4.1 Materials
• 5.4.2 Design Aspects
• 5.4.3 Fabrication and Inspection
• 5.4.4 Operation
• 5.5 Maintenance, Surveillance and In-service Inspection (MS&I)
• 5.5.1 Maintenance
• 5.5.2 Surveillance
• 5.5.3 In-service Inspection
• 5.5.4 Quality Assurance
• 5.6 Training and Qualification of Personnel
• 5.7 Plant Ageing
• 5.8 Quality Survey and Audit
• 5.8.1 Quality System Survey
• 5.8.2 Quality System Audit
• Summary
• Bibliography
• Assignments
• 
  
• Chapter 6 Siting of Nuclear Power Plants
• 6.1 Introduction
• 6.2 Basic Requirements
• 6.3 Impact of External Events on the Plant
• 6.3.1 Natural Phenomena
• 6.3.2 Human Induced External Events
• 6.4 Impact of Plant on Site, Environment and Public
• 6.4.1 Radiological Impact Study
• 6.4.2 Population Distribution
• 6.4.3 Environmental Impact Study
• 6.5 Emergency Preparedness
• 6.6 Other Considerations
• 6.7 Quality Assurance
• Summary
• Bibliography
• Assignments
• 
  
• Chapter 7 Engineered Safety Systems
• 7.1 Introduction
• 7.2 Shutdown Systems
• 7.2.1 Design
• 7.2.2 Speed of Control Rod
• 7.2.3 Shutdown System – Reactivity Worth
• 7.2.4 Design Safety
• 7.3 Signals and Safety Logic
• 7.4 Independence of Safety System
• 7.5 Heat Removal Systems
• 7.5.1 Heat Removal Capability
• 7.5.2 Secondary Side Heat Removal
• 7.5.3 Shutdown Cooling System
• 7.5.4 Other Heat Removal Possibilities
• 7.6 Emergency Core Cooling (ECC)
• 7.6.1 Crash Cool-down
• 7.7 Containment and Subsystems
• 7.7.1 Design Pressure and Leak Rate
• 7.7.2 Pressure Control and Heat Removal
• 7.7.3 Containment Isolation
• 7.7.4 Reliability
• 7.7.5 Other Functions of Containment
• 7.8 Plant Monitoring
• Summary
• Bibliography
• Assignments
• 
  
• Chapter 8 Assessment of Radiological Consequences of Incidents
• 8.1 Introduction
• 8.2 Historical Basis of Containment
• 8.3 Quantities of Radioactive Products
• 8.4 Neutron Activation of Structural Materials
• 8.5 Release Rates
• 8.5.1 Transfer and Deposit in Reactor Systems
• 8.5.2 Transfer and Deposit in Buildings
• 8.5.3 Leak Rate to Atmosphere and Filtering
• 8.5.4 Environmental Transport and Deposit
• 8.6 Calculations of Radioactive Release
• 8.7 Source Term Estimation
• Summary
• Bibliography
• Assignments
• 
  
• Chapter 9 Safety Regulation in India
• 9.1 Introduction
• 9.2 Atomic Energy Regulatory Board
• 9.2.1 Mission
• 9.2.2 Powers and Functions
• 9.2.3 Organization
• 9.3 Preparation of Safety Documents
• 9.4 Safety Review of Nuclear Power Projects
• 9.4.1 Siting
• 9.4.2 Design
• 9.4.3 Consent for Construction
• 9.4.4 Commissioning
• 9.4.5 Regulatory Inspections
• 9.5 Project Safety Reviews of New Designs
• 9.5.1 Tarapur-3 & 4
• 9.5.2 PFBR
• 9.5.3 Kudankulam
• 9.6 Regulatory Review of Operating NPPs
• 9.7 Authorization Period Review
• 9.8 Renewal of Authorization
• 9.9 License Renewal
• 9.10 Regulatory Inspection
• 9.11 Licensing of Operating Personnel
• 9.12 Safety Upgradations in Old Plants
• 9.12.1 Rajasthan Unit-2
• 9.12.2 Tarapur-1 & 2
• 9.13 Major Accident Reviews
• 9.13.1 TMI-2 Accident
• 9.13.2 Chernobyl Accident
• 9.13.3 Narora Fire Incident
• 9.13.4 Fukushima Accident
• 9.14 Decommissioning Review
• 9.15 SARCAR
• 9.15.1 Radiation Facilities
• 9.15.2 Medical X-ray Installations
• 9.15.3 Gamma Processing Plants
• 9.15.4 Transport Safety
• 9.15.5 Regulatory Inspection
• 9.15.6 Radioactivity in Foodstuffs
• Summary
• Bibliography
• Assignments
• 
  
• Chapter 10 Safety Practices in Indian NPP
• 10.1 Introduction
• 10.2 Radiological Protection of Workers
• 10.2.1 Dose Limits
• 10.2.2 Radiation Protection Organization
• 10.2.3 ALARA Principle
• 10.2.4 AERB Review
• 10.3 Public Protection
• 10.4 Waste Management
• 10.5 Environmental Surveillance
• 10.6 Emergency Preparedness
• 10.6.1 Laws and Regulations
• 10.6.2 Zoning Concept
• 10.6.3 Emergency Measures
• 10.6.4 Emergency Classification
• 10.7 On-site Emergency Response
• 10.7.1 Declaration Criteria
• 10.7.2 Infrastructure
• 10.8 Off-site Emergency Response
• 10.8.1 Declaration Criteria
• 10.8.2 Infrastructure
• 10.8.3 Roles and Responsibilities
• 10.9 Training and Exercises
• 10.9.1 Training
• 10.9.2 Mock Exercises
• 10.10 Medical Assistance
• Summary
• Bibliography
• Assignments
• 
  
• Chapter 11 Passive Safety
• 11.1 Introduction
• 11.2 Safety Terminologies
• 11.3 Passive Systems Categorization
• 11.4 Passive Shutdown Systems
• 11.4.1 PHWR/CANDU Reactor
• 11.4.2 Sodium Fast Reactor
• 11.5 Passive Decay Heat Removal
• 11.5.1 Core Flooding Tanks
• 11.5.2 Natural Circulation Loops
• 11.5.3 Gravity Drain Tanks
• 11.5.4 Steam Generator Cooling
• 11.6 Passive Containment Cooling
• 11.6.1 Pressure Suppression Pools
• 11.6.2 Heat Removal Systems
• 11.6.3 Passive Spray Systems
• 11.7 Hydrogen Removal
• 11.8 Advanced Reactors
• 11.8.1 AP-600
• 11.8.2 AHWR
• 11.9 Issues in Passive Safety
• 11.9.1 Uncertainties
• 11.9.2 Dynamic Reliability
• 11.9.3 Active vs Passive Comparison
• Summary
• Bibliography
• Assignments
• 
  
• Chapter 12 Public Acceptance of Nuclear Power Plants
• 12.1 Introduction
• 12.2 Social Acceptance
• 12.2.1 Trust-based Path
• 12.2.2 Technology-based Path
• 12.3 Risk-based Approach
• 12.4 Causes of Unacceptance
• 12.4.1 Public Education
• 12.4.2 Media and Politicians
• 12.4.3 Anti-nuclear Groups
• 12.4.4 Neighboring Countries
• 12.4.5 Waste and Fuel Cycle Issues
• 12.4.6 Proliferation
• 12.5 Non-rational Causes
• 12.6 Positive Aspects of Nuclear Energy
• 12.7 Negative Effects of Current Approaches
• 12.8 Proposed Approaches
• 12.8.1 Transparency
• 12.8.2 Target Groups
• 12.8.3 Messaging Content
• 12.8.4 Opinion Polls
• 12.8.5 Risk Communication
• 12.9 International Experiences
• 12.10 Indian Experience
• 12.10.1 Siting
• 12.10.2 Operation
• 12.10.3 Public Hearing
• 12.11 Indian Context Approaches
• Summary
• Bibliography
• Assignments
• 
  
• Annexures
• A History of Events in Nuclear Power Plants and Radiation Facilities
• B Event Categories - PHWR
• C Sequence of Events after a PIE
• D Boiling Heat Transfer
• E Large LOCA Analysis (PHWR-220 MWe)
• F Loss of Flow Accidents in Fast Breeder Reactor
• G Public Outreach Case Studies
• H Radiation Effects and Dose Limits
• I Additional Reading
• 
  
• Index

Dr. G. Vaidyanathan is a retired Outstanding Scientist and Director, Fast Reactor Technology from the Indira Gandhi Centre for Atomic Research, India. His experience of 38 years includes Design, Analysis, Experimentation and Project Management in the area of nuclear energy. He has been involved in the Sodium Cooled Fast Reactors and Pressurised Heavy Water Reactors, which form the main stay of Indian Nuclear Power Programme. He was a visiting professor at SRM University for nearly 8 years, where he taught the subjects on Nuclear Engineering, safety, and Alternative systems of Energy. He was also a guest faculty in Nuclear Engineering at the Indian Institute of Technology, Madras, where he taught post graduate students. He has also taught at Amity University and National Institute of Technologies.