Nuclear Power Safety in Saudi Arabia

150 Pages·2016·4.07 MB·English

UC Irvine
UC Irvine Electronic Theses and Dissertations
Title
Nuclear Power Safety in Saudi Arabia
Permalink
https://escholarship.org/uc/item/41f529ch
Author
Khubrani, Fahad Ali
Publication Date
2016

Peer reviewed|Thesis/dissertation
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University of California
UNIVERSITY OF CALIFORNIA,
IRVINE

Nuclear Power Plant Safety in Saudi Arabia

THESIS

submitted in partial satisfaction of the requirements
for the degree of

MASTER OF SCIENCE

in Electrical Engineering

by

Fahad Ali Khubrani

Thesis Committee:
Professor Nader Bagherzadeh, Chair
Professor Najmedin Meshkati
Professor Jean-Luc Gaudiot

2016
© 2016 Fahad Ali Khubrani
DEDICATION

To

my parents and my wife

ii
TABLE OF CONTENTS

Page

LIST OF FIGURES iv

LIST OF TABLES vii

ACKNOWLEDGMENTS viii

ABSTRACT OF THE THESIS ix

INTRODUCTION 1

CHAPTER 1: Lessons Learned from Severe Nuclear Accidents 5
Three Mile Island 6
Chernobyl 9
Fukushima 12
Onagawa Nuclear Power Station 15
Concluding Remark 17

CHAPTER 2: Nuclear Power Reactor Technology 18
Generation III + Reactors 18
SMRs 27
Concluding Remark 64

CHAPTER 3: Nuclear Safety Systems and Programs 66
Nuclear Safety Perspective 66
I&C Systems 82
Important Nuclear Plant Assets 97
Concluding Remark 100

CHAPTER 4: Safety Culture 103
Safety Culture Model 103
Characteristics of Safety Culture 105
Safety Culture Development 119
Influence of the Regulatory Body on Safety Culture 123
Safety Culture in Pre-Operational Phase 124
Concluding Remark 135

CHAPTER 5: Conclusion 137

REFERENCES 138

iii
LIST OF FIGURES

Page

Figure 1.0 Examples of Events at Nuclear Facilities 4

Figure 1.1 TMI-2 (PWR) Schematic 6

Figure 1.2 Damaged of the Chernobyl-4 Reactor 9

Figure 1.3 Damage to the Fukushima reactor buildings 12

Figure 1.4 Onagawa Nuclear Power Station Location 15

Figure 2.1 AP1000 Passive Safety Systems 18

Figure 2.2 AP1000 Passive Containment Cooling System 19

Figure 2.3 APR 1400 Schematic 21

Figure 2.4 APR 1400 Divisional & Quadrant Layout Arrangement 23

Figure 2.5 EPR Safety Features 24

Figure 2.6 EPR Meltdown Core Compartment 25

Figure 2.7 Reactor System Configuration of CAREM-25 28

Figure 2.8 CAREM-25 Reactor Pressure Vessel 29

Figure 2.9 Reactor System Configuration of ACP-100 32

Figure 2.10 ACP-100 Reactor Building Underground Location 33

Figure 2.11 Layout of AHWR 35

Figure 2.12 Reactor System Configuration of IRIS 39

Figure 2.13 Reactor System Configuration of SMART 42

Figure 2.14 SMART Engineered Safety Systems 45

Figure 2.15 Reactor System Configuration of Westinghouse SMR 47

Figure 2.16 Westinghouse SMR Diverse Decay Heat Removal Methods 50
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Figure 2.17 NuScale Plant Schematic 52

Figure 2.18 NuScale Emergency Core Cooling System 54

Figure 2.19 Single SC-HTGR Reactor Module 56

Figure 2.20 Reactor Configuration of HTR-PM 60

Figure 2.21 TRISO Fuel Element 61

Figure 2.22 HTR-PM Twin Unit Power Conversion Components 64

Figure 3.1 Defense in Depth Protection Layers 67

Figure 3.2 PRA Model 68

Figure 3.3 IRIDM Process 73

Figure 3.4 Risk-Informed Deliberation 74

Figure 3.5 Seismic Probabilistic Risk Assessment 75

Figure 3.6 Spent Fuel Pool 79

Figure 3.7 Single Dry Cask 81

Figure 3.8 NUHMS Canister and Casks 82

Figure 3.9 HI-STORM UMAX Technology 83

Figure 3.10 Cyber Actors 92

Figure 3.11 Cyber-Attacks Examples 93

Figure 3.12 Computer Security Architecture 94

Figure 3.13 I&C Lifecycle Development Model 97

Figure 3.14 I&C life cycle processes and V&V Activities 98

Figure 3.15 NPP Control Operating System Architecture 100

Figure 4.1 Schein’s Three Level Culture Model 104

Figure 4.2 Nuclear Pre-operational Project Phases 125

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Figure 4.3 Integrated Management Systems 128

Figure 4.4 Safety Culture Model 131

vi
LIST OF TABLES

Page

Table 1.0a Typical Activity Content of a (BWR) Core 2

Table 1.0b INES Event Scale 3

Table 2.1 Characteristics of APR 1400 22

Table 2.2 Characteristics of CREM-25 30

Table 2.3 Characteristics of ACP-100 34

Table 2.4 Characteristics of AHWR 36

Table 2.5 Characteristics of SMRT 44

Table 2.6 Characteristics of Westinghouse SMR 48

Table 2.7 Characteristics of NuScale 54

Table 2.8 Characteristics of SC-HTGR 57

Table 2.9 Characteristics of HTR-PM 62

Table 3.1 Characteristics of NuScale 89

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ACKNOWLEDGMENTS

I would like to express the deepest appreciation to my committee chair, Professor Nader
Bagherzadeh, for his great attitude and infinite support. He continually and convincingly
conveyed a spirit of adventure with respect to research in the nuclear field. Without his
guidance and persistent assistance, this thesis would not have been possible.

I would like to thank my committee member, Professor Jean-Luc Gaudiot for his support
and acceptance to be a member of my thesis committee.

I would like to thank my distinguished committee member, Professor Najmedin Meshkati
from the University of Southern California, USC, who has spared no efforts to guide me,
share his literature work, and keep me posted on the latest nuclear safety matters.

Special thanks to Mr. Ted Quinn, a past ANS president, who shared with me his work in I&C
systems and referred me to the MIT Nuclear Plant Safety short program. His motivation for
learning about nuclear safety has the lasting effect.

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