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E-grāmata: Risk and Safety Analysis of Nuclear Systems

(University of Michigan, Ann Arbor, MI), (University of Washington)
  • Formāts: PDF+DRM
  • Izdošanas datums: 17-Aug-2011
  • Izdevniecība: John Wiley & Sons Inc
  • Valoda: eng
  • ISBN-13: 9781118043448
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Risk and Safety Analysis of Nuclear SystemsPalielināt
  • Formāts: PDF+DRM
  • Izdošanas datums: 17-Aug-2011
  • Izdevniecība: John Wiley & Sons Inc
  • Valoda: eng
  • ISBN-13: 9781118043448
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"The book has been developed in conjunction with NERS 462, a course offered every year to seniors and graduate students in the University of Michigan NERS program. The first half of the book covers the principles of risk analysis, the techniques used to develop and update a reliability data base, the reliability of multi-component systems, Markov methods used to analyze the unavailability of systems with repairs, fault trees and event trees used in probabilistic risk assessments (PRAs), and failure modes of systems. All of this material is general enough that it could be used in non-nuclear applications, although there is an emphasis placed on the analysis of nuclear systems. The second half of the book covers the safety analysis of nuclear energy systems, an analysis of major accidents and incidents that occurred in commercial nuclear plants, applications of PRA techniques to the safety analysis of nuclear power plants (focusing on a major PRA study for five nuclear power plants), practical PRA examples, and emerging techniques in the structure of dynamic event trees and fault trees that can provide a more realistic representation of complex sequences of events. The book concludes with a discussion on passive safety features of advanced nuclear energysystems under development and approaches taken for risk-informed regulations for nuclear plants"--

"The first half of the book covers the principles of risk analysis, the techniques used to develop and update a reliability data base, the reliability of multi-component systems, Markov methods used to analyze the unavailability of systems with repairs, fault trees and event trees used in probabilistic risk assessments (PRAs), and failure modes of systems. All of this material is general enough that it could be used in non-nuclear applications, although there is an emphasis placed on the analysis of nuclear systems. The second half of the book covers the safety analysis of nuclear energy systems, an analysis of major accidents and incidents that occurred in commercial nuclear plants, applications of PRA techniques to the safety analysis of nuclear power plants (focusing on a major PRA study for five nuclear power plants), practical PRA examples, and emerging techniques in the structure of dynamic event trees and fault trees that can provide a more realistic representation of complex sequences of events"--

Provided by publisher.

 

The book has been developed in conjunction with NERS 462, a course offered every year to seniors and graduate students in the University of Michigan NERS program.

The first half of the book covers the principles of risk analysis, the techniques used to develop and update a reliability data base, the reliability of multi-component systems, Markov methods used to analyze the unavailability of systems with repairs, fault trees and event trees used in probabilistic risk assessments (PRAs), and failure modes of systems. All of this material is general enough that it could be used in non-nuclear applications, although there is an emphasis placed on the analysis of nuclear systems.

The second half of the book covers the safety analysis of nuclear energy systems, an analysis of major accidents and incidents that occurred in commercial nuclear plants, applications of PRA techniques to the safety analysis of nuclear power plants (focusing on a major PRA study for five nuclear power plants), practical PRA examples, and emerging techniques in the structure of dynamic event trees and fault trees that can provide a more realistic representation of complex sequences of events. The book concludes with a discussion on passive safety features of advanced nuclear energy systems under development and approaches taken for risk-informed regulations for nuclear plants.

Preface xii
Permissions and Copyrights xiv
List of Tables
xvi
List of Figures
xviii
1 Risk and Safety of Engineered Systems
1(14)
1.1 Risk and Its Perception and Acceptance
1(5)
1.2 Overview of Risk and Safety Analysis
6(2)
1.3 Two Historical Reactor Accidents
8(1)
1.4 Definition of Risk
9(1)
1.5 Reliability, Availability, Maintainability, and Safety
10(2)
1.6 Organization of the Book
12(3)
References
13(2)
2 Probabilities of Events
15(44)
2.1 Events
15(2)
2.2 Event Tree Analysis and Minimal Cut Sets
17(2)
2.3 Probabilities
19(6)
2.3.1 Interpretations of Probability
19(1)
2.3.2 Axiomatic Approach to Probabilities
20(1)
2.3.3 Intersection of Events
21(1)
2.3.4 Union of Events
22(3)
2.3.5 Decomposition Rule for Probabilities
25(1)
2.4 Time-Independent Versus Time-Dependent Probabilities
25(1)
2.5 Time-Independent Probabilities
26(5)
2.5.1 Introduction
26(1)
2.5.2 Time-Independent Probability Distributions
27(4)
2.6 Normal Distribution
31(4)
2.7 Reliability Functions
35(6)
2.8 Time-Dependent Probability Distributions
41(9)
2.8.1 Erlangian and Exponential Distributions
42(1)
2.8.2 Gamma Distribution
43(1)
2.8.3 Lognormal Distribution
44(2)
2.8.4 Weibull Distribution
46(1)
2.8.5 Generalized "Bathtub" Distribution
47(1)
2.8.6 Selection of a Time-Dependent Probability Distribution
48(2)
2.9 Extreme-Value Probability Distributions
50(2)
2.10 Probability Models for Failure Analyses
52(7)
References
53(1)
Exercises
53(6)
3 Reliability Data
59(26)
3.1 Estimation Theory
59(6)
3.1.1 Moment Estimators
60(1)
3.1.2 Maximum Likelihood Estimators
61(3)
3.1.3 Maximum Entropy Estimators
64(1)
3.1.4 Comparison of Estimators
65(1)
3.2 Bayesian Updating of Data
65(5)
3.2.1 Bayes Equation
65(2)
3.2.2 Applications of the Bayes Equation
67(3)
3.3 Central Limit Theorem and Hypothesis Testing
70(4)
3.3.1 Interpretation of the Central Limit Theorem
71(1)
3.3.2 Hypothesis Testing with the Central Limit Theorem
72(2)
3.4 Reliability Quantification
74(11)
3.4.1 Central Limit Theorem for Reliability Quantification
74(2)
3.4.2 Engineering Approach for Reliability Quantification
76(1)
3.4.3 χ2-Distribution for Reliability Quantification
77(1)
3.4.4 Three-Way Comparison and Concluding Remarks
78(2)
References
80(1)
Exercises
80(5)
4 Reliability of Multiple-Component Systems
85(24)
4.1 Series and Active-Parallel Systems
86(7)
4.1.1 Systems with Independent Components
86(2)
4.1.2 Systems with Redundant Components
88(2)
4.1.3 Fail-to-Safety and Fail-to-Danger Systems
90(3)
4.2 Systems with Standby Components
93(3)
4.3 Decomposition Analysis
96(4)
4.4 Signal Flow Graph Analysis
100(1)
4.5 Cut Set Analysis
101(8)
References
104(1)
Exercises
104(5)
5 Availability and Reliability of Systems with Repair
109(32)
5.1 Introduction
109(2)
5.2 Markov Method
111(7)
5.2.1 Markov Governing Equations
111(2)
5.2.2 Solution of Markov Governing Equations
113(3)
5.2.3 An Elementary Example
116(2)
5.3 Availability Analyses
118(10)
5.3.1 Rules for Constructing Transition Rate Matrices
118(1)
5.3.2 Availability Transition Rate Matrices
119(4)
5.3.3 Time-Dependent Availability Examples
123(4)
5.3.4 Steady-State Availability
127(1)
5.4 Reliability Analyses
128(5)
5.4.1 Reliability Transition Rate Matrices
129(1)
5.4.2 Time-Dependent Reliability Examples
130(1)
5.4.3 Mean Time to Failure
130(3)
5.5 Additional Capabilities of Markov Models
133(8)
5.5.1 Imperfect Switching Between System States
134(2)
5.5.2 Systems with Nonconstant Hazard Rates
136(1)
References
137(1)
Exercises
137(4)
6 Probabilistic Risk Assessment
141(38)
6.1 Failure Modes
142(1)
6.2 Classification of Failure Events
143(7)
6.2.1 Primary, Secondary, and Command Failures
143(1)
6.2.2 Common Cause Failures
144(4)
6.2.3 Human Errors
148(2)
6.3 Failure Data
150(2)
6.3.1 Hardware Failures
150(1)
6.3.2 Human Errors
150(2)
6.4 Combination of Failures and Consequences
152(4)
6.4.1 Inductive Methods
152(2)
6.4.2 Event Tree Analysis
154(2)
6.5 Fault Tree Analysis
156(9)
6.5.1 Introduction
156(1)
6.5.2 Fault Tree Construction
157(1)
6.5.3 Qualitative Fault Tree Analysis
157(3)
6.5.4 Quantitative Fault Tree Analysis
160(5)
6.5.5 Common Cause Failures and Fault Tree Analysis
165(1)
6.6 Master Logic Diagram
165(3)
6.7 Uncertainty and Importance Analysis
168(11)
6.7.1 Types of Uncertainty in PRAs
168(1)
6.7.2 Stochastic Uncertainty Analysis
169(1)
6.7.3 Sensitivity and Importance Analysis
170(2)
References
172(1)
Exercises
172(7)
7 Computer Programs for Probabilistic Risk Assessment
179(18)
7.1 Fault Tree Methodology of the SAPHIRE Code
179(4)
7.1.1 Gate Conversion and Tree Restructuring
180(1)
7.1.2 Simplification of the Tree
180(2)
7.1.3 Fault Tree Expansion and Reduction
182(1)
7.2 Fault and Event Tree Evaluation with the SAPHIRE Code
183(2)
7.3 Other Features of the SAPHIRE Code
185(1)
7.4 Other PRA Codes
185(2)
7.5 Binary Decision Diagram Algorithm
187(10)
7.5.1 Basic Formulation of the BDD Algorithm
187(2)
7.5.2 Generalization of the BDD Formulation
189(4)
7.5.3 Zero-Suppressed BDD Algorithm and the FTREX Code
193(1)
References
194(1)
Exercises
195(2)
8 Nuclear Power Plant Safety Analysis
197(62)
8.1 Engineered Safety Features of Nuclear Power Plants
197(18)
8.1.1 Pressurized Water Reactor
198(12)
8.1.2 Boiling Water Reactor
210(5)
8.2 Accident Classification and General Design Goals
215(5)
8.2.1 Plant Operating States
217(1)
8.2.2 Accident Classification in 10 CFR 50
217(2)
8.2.3 General Design Criteria and Safety Goals
219(1)
8.3 Design Basis Accident: Large-Break LOCA
220(11)
8.3.1 Typical Sequence of a Cold-Leg LBLOCA in PWR
221(4)
8.3.2 ECCS Specifications
225(2)
8.3.3 Code Scaling, Applicability, and Uncertainty Evaluation
227(4)
8.4 Severe (Class 9) Accidents
231(2)
8.5 Anticipated Transients Without Scram
233(8)
8.5.1 History and Background of the ATWS Issue
233(2)
8.5.2 Resolution of the ATWS Issues
235(2)
8.5.3 Power Coefficients of Reactivity in LWRs
237(4)
8.6 Radiological Source and Atmospheric Dispersion
241(9)
8.6.1 Radiological Source Term
242(1)
8.6.2 Atmospheric Dispersion of Radioactive Plume
243(4)
8.6.3 Simple Models for Dose Rate Calculation
247(3)
8.7 Biological Effects of Radiation Exposure
250(9)
References
252(2)
Exercises
254(5)
9 Major Nuclear Power Plant Accidents and Incidents
259(44)
9.1 Three Mile Island Unit 2 Accident
260(3)
9.1.1 Sequence of the Accident---March 1979
260(1)
9.1.2 Implications and Follow-Up of the Accident
260(3)
9.2 PWR In-Vessel Accident Progression
263(9)
9.2.1 Core Uncovery and Heatup
265(1)
9.2.2 Cladding Oxidation
266(2)
9.2.3 Clad Melting and Fuel Liquefaction
268(2)
9.2.4 Molten Core Slumping and Relocation
270(1)
9.2.5 Vessel Breach
271(1)
9.3 Chernobyl Accident
272(5)
9.3.1 Cause and Nature of the Accident---April 1986
272(2)
9.3.2 Sequence of the Accident
274(1)
9.3.3 Estimate of Energy Release in the Accident
275(1)
9.3.4 Accident Consequences
275(1)
9.3.5 Comparison of the TMI and Chernobyl Accidents
276(1)
9.4 Fukushima Station Accident
277(2)
9.4.1 Sequence of the Accident---March 2011
277(1)
9.4.2 March 2011 Perspectives on the Fukushima SBO Event
278(1)
9.5 Salem Anticipated Transient Without Scram
279(4)
9.5.1 Chronology and Cause of the Salem Incident
279(2)
9.5.2 Implications and Follow-Up of the Salem ATWS Event
281(2)
9.6 LaSalle Transient Event
283(8)
9.6.1 LaSalle Nuclear-Coupled Density-Wave Oscillations
283(4)
9.6.2 Simple Model for Nuclear-Coupled Density-Wave Oscillations
287(2)
9.6.3 Implications and Follow-Up of the LaSalle Incident
289(2)
9.7 Davis-Besse Potential LOCA Event
291(12)
9.7.1 Background and Chronology of the Incident
291(2)
9.7.2 NRC Decision to Grant DB Shutdown Delay
293(2)
9.7.3 Causes for the Davis-Besse Incident and Follow-Up
295(2)
References
297(3)
Exercises
300(3)
10 PRA Studies of Nuclear Power Plants
303(46)
10.1 WASH-1400 Reactor Safety Study
304(7)
10.2 Assessment of Severe Accident Risks: NUREG-1150
311(29)
10.2.1 Background and Scope of the NUREG-1150 Study
311(2)
10.2.2 Overview of NUREG-1150 Methodology
313(2)
10.2.3 Accident Frequency Analysis
315(5)
10.2.4 Accident Progression Analysis
320(4)
10.2.5 Radionuclide Transport Analysis
324(3)
10.2.6 Offsite Consequence Analysis
327(3)
10.2.7 Uncertainty Analysis
330(1)
10.2.8 Risk Integration
331(6)
10.2.9 Additional Perspectives and Comments on NUREG-1150
337(3)
10.3 Simplified PRA in the Structure of NUREG-1150
340(9)
10.3.1 Description of the Simplified PRA Model
340(4)
10.3.2 Parametric Studies and Comments on the Simplified PRA Model
344(1)
References
345(2)
Exercises
347(2)
11 Passive Safety and Advanced Nuclear Energy Systems
349(52)
11.1 Passive Safety Demonstration Tests at EBR-II
349(15)
11.1.1 EBR-II Primary System and Simplified Model
350(7)
11.1.2 Unprotected Loss-of-Flow and Loss-of-Heat-Sink Tests
357(4)
11.1.3 Simplified Fuel Channel Analysis
361(1)
11.1.4 Implications of EBR-II Passive Safety Demonstration Tests
362(2)
11.2 Safety Characteristics of Generation III+ Plants
364(18)
11.2.1 AP1000 Design Features
364(2)
11.2.2 Small-Break LOCA Analysis for AP1000
366(5)
11.2.3 Economic Simplified Boiling Water Reactor
371(4)
11.2.4 Reliability Quantification of SBWR Passive Safety Containment
375(7)
11.3 Generation IV Nuclear Power Plants
382(19)
11.3.1 Sodium-Cooled Fast Reactor
383(4)
11.3.2 Hypothetical Core Disruptive Accidents for Fast Reactors
387(6)
11.3.3 VHTR and Phenomena Identification and Ranking Table
393(3)
References
396(3)
Exercises
399(2)
12 Risk-Informed Regulations and Reliability-Centered Maintenance
401(16)
12.1 Risk Measures for Nuclear Plant Regulations
402(4)
12.1.1 Principles of Risk-Informed Regulations and Licensing
402(3)
12.1.2 Uncertainties in Risk-Informed Decision Making
405(1)
12.1.3 Other Initiatives in Risk-Informed Regulations
406(1)
12.2 Reliability-Centered Maintenance
406(11)
12.2.1 Optimization Strategy for Preventive Maintenance
407(2)
12.2.2 Reliability-Centered Maintenance Framework
409(1)
12.2.3 Cost-Benefit Considerations
410(3)
References
413(2)
Exercises
415(2)
13 Dynamic Event Tree Analysis
417(26)
13.1 Basic Features of Dynamic Event Tree Analysis
418(3)
13.2 Continuous Event Tree Formulation
421(5)
13.2.1 Derivation of the Stochastic Balance Equation
421(2)
13.2.2 Integral Form of the Stochastic Balance Equation
423(2)
13.2.3 Numerical Solution of the Stochastic Balance Equation
425(1)
13.3 Cell-to-Cell Mapping for Parameter Estimation
426(8)
13.3.1 Derivation of the Bayesian Recursive Relationship
427(3)
13.3.2 CCM Technique for Dynamic Event Tree Construction
430(4)
13.4 Diagnosis of Component Degradations
434(9)
13.4.1 Bayesian Framework for Component Diagnostics
434(3)
13.4.2 Implementation of the Probabilistic Diagnostic Algorithm
437(4)
References
441(1)
Exercises
442(1)
Appendix A Reactor Radiological Sources
443(6)
A.1 Fission Product Inventory and Decay Heat
443(3)
A.2 Health Effects of Radiation Exposure
446(3)
References
448(1)
Appendix B Some Special Mathematical Functions
449(4)
B.1 Gamma Function
449(2)
B.2 Error Function
451(2)
References
451(2)
Appendix C Some Failure Rate Data
453(4)
Appendix D Linear Kalman Filter Algorithm
457(5)
References
461(1)
Answers to Selected Exercises 462(5)
Index 467
JOHN C. LEE, PhD, has been Professor of Nuclear Engineering at the University of Michigan since 1974, following five years of employment at Westinghouse Electric Corporation and General Electric Company. He has written for approximately 180 publications on broad areas of nuclear reactor physics and engineering, including nuclear systems analysis and diagnostics. Dr. Lee is a Fellow of the American Nuclear Society.

NORMAN J. McCORMICK, PhD, is an emeritus professor of mechanical engineering at the University of Washington who retired in 2003. From 1966 until the early 1990s, he was a professor of nuclear engineering. Dr. McCormick is the author of the book Reliability and Risk Analysis Methods and Nuclear Power Applications (upon which part of NERS 462 is based) and has authored approximately 150 journal articles. He is a Fellow of the American Nuclear Society.
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