Nuclear Reactor Safety – Principles and Concepts

495

Author: Vaidyanathan G

ISBN: 9789380381565

Copy Right Year: 2017

Pages:  372

Binding: Soft Cover

Publisher:  Yes Dee Publishing

SKU: 9789380381565 Category:

Description

Nuclear Reactor Safety is an introductory book on nuclear reactors and safety. The book begins with introduction to the applications of nuclear energy in power production, industry, medicine and food preservation, followed by the different types of reactors. Then it deals with the different safety principles, safety approaches and quality assurance and engineered safety of Nuclear reactors. Safety regulation and practices in the Indian NPPs have been detailed out for a good understanding. Finally it introduces the reader to the passive safety approaches being utilized and contemplated for new designs besides sharing some aspects of public acceptance of nuclear energy. Basic reactor physics and analysis of some exceptional events in nuclear reactors are given in the annexures.

Additional information

Weight .5 kg
Dimensions 23 × 16 × 2 cm

Table of Content

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-2Events: Events of Moderate Frequency
4.4.3 Category-3Events: 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 External Events due to 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 the Outside Atmosphere and Filtering Provisions
8.5.4 Environmental Transport and Deposit Conditions
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 (540 MWe each)
9.5.2 PFBR(500MWe)
9.5.3 Kudankulam (2 × 1000 MWe)
9.6 Regulatory Review of Operating NPPs
9.7 Normal Review During Authorization Period
9.8 Safety Review for Renewal of Authorization
9.9 License Renewal
9.10 Regulatory Inspection During Operation
9.11 Licensing of Operating Personnel
9.12 Safety Upgradations in Old Plants
9.12.1 RajasthanUnit-2
9.12.2 Tarapur-1&-2
9.13 Safety Review Following Major Accidents
9.13.1 TMI-2Accident
9.13.2 Chernobyl Accident
9.13.3 Fire Incident in Narora Atomic PowerStation (NAPS)
9.13.4 Fukushima Accident
9.14 Safety Review for Decommissioning
9.15 Safety Review Committee for Applications of Radiation (SARCAR)
9.15.1 Radiation Facilities
9.15.2 Medical X-ray Installations
9.15.3 Gamma Radiation Processing Plants
9.15.4 Safety in Transport of Radioactive Materials
9.15.5 Regulatory Inspection of Radiation Facilities
9.15.6 Radioactivity in Foodstuffs
Summary
Bibliography
Assignments
Chapter 10 Safety Practices in Indian NPP
10.1 Introduction
10.2 Radiological Protection of the Workers
10.2.1 Dose Limits
10.2.2 Organization for Radiation Protection
10.2.3 Steps for ALARA Exposures
10.2.4 Radiation Protection Review by AERB
10.3 Radiological Protection to Public
10.4 Radioactive Waste Management
10.5 Environmental Radiological Surveilance
10.6 Emergency Preparedness
10.6.1 National Laws, Regulations and Requirements
10.6.2 Zoning Concept and Emergency Planning
10.6.3 Emergency Measures
10.6.4 Classification of Emergencies
10.7 Features of On-site Emergency Preparedness Response (EPR) Plan
10.7.1 Criteria for Declaration of Emergency
10.7.2 Infrastructure for On-site Emergency Response
10.8 Features of Off-site EPR Plan
10.8.1 Criteria for Declaration of Emergency
10.8.2 Infrastructure for Off-site Emergency Response
10.8.3 Roles and Responsibilities for Off-site Emergency Response
10.9 Training and Exercise
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 Some Safety Terminologies
11.3 Categorisation of Passive Systems
11.4 Passive Reactor Shutdown Systems
11.4.1 PHWR/CANDU Reactor
11.4.2 Sodium Fast Reactor
11.5 Passive Decay Heat Removal in LWR/PHWR Reactors
11.5.1 Pre-pressurized Core Flooding Tanks
11.5.2 Elevated Tank Natural Circulation Loops
11.5.3 Elevated Gravity Drain Tanks
11.5.4 Passively Cooled Steam Generator Natural Circulation
11.6 Passive Safety Systems for Containment Cooling and Pressure Suppression
11.6.1 Containment Pressure Suppression Pools
11.6.2 Containment Passive Heat Removal/Pressure Suppression Systems
11.6.3 Passive Containment Spray Systems
11.7 Hydrogen Removal
11.8 New Passive Featured Reactors
11.8.1 AP-600
11.8.2 Advanced Heavy Water Reactor (AHWR)
11.9 Issues Related to Passive Safety
11.9.1 Uncertainties
11.9.2 Dynamic Reliability
11.9.3 Comparative Assessment between Active and Passive Systems
Summary
Bibliography
Assignments
Chapter 12 Public Acceptance of Nuclear Power Plants 
12.1 Introduction
12.2 Social Acceptance
12.2.1 The Trust-based Path
12.2.2 The Technology-based Path
12.3 Risk based Approach
12.4 Rational Root Causes of Unacceptance
12.4.1 Public Education
12.4.2 Attitude of Media and Politicians
12.4.3 Antinuclear Groups
12.4.4 Neighboring Countries
12.4.5 Nuclear Waste and Fuel Cycle Back End
12.4.6 Proliferation
12.5 Non-rational Root Causes
12.6 Positive Points of Nuclear Energy
12.6.1 Cost Competitive
12.6.2 Independence of Supply
12.6.3 Climate Change
12.6.4 Nuclear Safety
12.6.5 Future Energy Needs
12.6.6 Positive Local Impacts
12.7 Negative Effects of Some Present Approaches
12.8 Proposed Approaches
12.8.1 Transparency
12.8.2 Target Groups
12.8.3 Content of the Messages
12.8.4 Opinion Polls
12.8.5 Risk Communication
12.9 Experiences of Korea, Czekeslovakia and Finland
12.10 The Indian Experience
12.10.1 NPP Siting
12.10.2 NPP Operation
12.10.3 Public Hearing
12.11 Approaches in the Indian Context
Summary
Bibliography
Assignments

• Annexure A − History of Events in Nuclear Power Plants and Radiation
• Facilities
• Annexure B –Event Categories-PHWR
• Annexure C –Sequence of Events after a PIE
• Annexure D − Boiling Heat Transfer
• Annexure E − Large LOCA Analysis of Indian Pressurized Heavy Water
• Reactor-220 MWe
• Annexure F − Loss of Flow Accidents in a Fast Breeder Reactor
• Annexure G − Public Outreach Case Studies
• Annexure H –Radiation Effects and Dose Limits
• Annexure I − Additional Reading
• Index

About The Author

G Vaidyanathan is a visiting professor in Nuclear Engineering at the SRM university, Chennai, India. He is also a guest faculty in Nuclear Engineering at the Indian Institute of Technology, Madras. He retired as Outstanding Scientist and Director, Fast Reactor Technology from the Indira Gandhi Centre for Atomic Research, Kalpakkam, after an experience of 38 years in Design, Analysis, Experimentation and Project Management. 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.

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