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- Herausgeber: John Wiley & Sons
- Kategorie: Wissenschaft und neue Technologien
- Sprache: Englisch
Developing sufficient energy resources to replace coal, oil and gas is a globally critical necessity. Alternatives to fossil fuels such as wind, solar, or geothermal energies are desirable, but the usable quantities are limited and each has inherent deterrents. The only virtually unlimited energy source is nuclear energy, where safety of infrastructure systems is the paramount concern.
Infrastructure Systems for Nuclear Energy addresses the analysis and design of infrastructures associated with nuclear energy. It provides an overview of the current and future nuclear power industry and the infrastructure systems from the perspectives of regulators, operators, practicing engineers and research academics. This book also provides details on investigations of containment structures, nuclear waste storage facilities and the applications of commercial/academic computer software.
Specific environments that challenge the behavior of nuclear power plants infrastructure systems such as earthquake, blast, high temperature, irradiation effects, soil-structure interaction effect, etc., are also discussed.
Key features:
- Includes contributions from global experts representing academia and industry
- Provides an overview of the nuclear power industry and nuclear infrastructure systems
- Presents the state-of-the-art as well as the future direction for nuclear civil infrastructure systems
Infrastructure Systems for Nuclear Energy is a comprehensive, up-to-date reference for researchers and practitioners working in this field and for graduate studies in civil and mechanical engineering.
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Table of Contents
Title Page
Copyright
List of Contributors
Preface
Acronyms
Chapter 1: Introduction
1.1 International Workshop on Infrastructure Systems for Nuclear Energy
1.2 Overview of Nuclear Power Plants
1.3 Infrastructure for Nuclear Power Industry
1.4 Containment Structures
1.5 Nuclear Waste Storage Facilities
Part I: Infrastructure for Nuclear Power Industry
Chapter 2: Current Status and Future Role of Nuclear Power
2.1 Introduction
2.2 Installed Nuclear Power Capacity in 2011
2.3 Discussion
2.4 Conclusions
2.5 Further Reading
References
Chapter 3: Seismic Probabilistic Risk Assessment for Nuclear Power Plants
3.1 Introduction
3.2 Conventional SPRA Methodologies
3.3 The Methodology of Huang et al.
3.4 Summary and Conclusions
References
Chapter 4: Seismic Abatement Method for Nuclear Power Plants and Seismic-Isolation Systems for Structural Elements
4.1 Main Principles of the Method
4.2 Theorem and Proof
4.3 Finite Element Construction
4.4 Pros and Cons of the Method
4.5 Application of the Method to Seismic Isolation Design of Whole Building
4.6 Seismic Isolation Devices to Protect Various Elements and Units
4.7 Applications
4.8 Conclusions
References
Chapter 5: Framework for Design of Next-Generation Base-Isolated Nuclear Structures
5.1 Introduction
5.2 Development of Seismic Isolation Systems
5.3 Seismic Isolation of New Nuclear Power Plant Structures
5.4 Performance-Based Design and Evaluation Framework
5.5 Conclusions
References
Chapter 6: Development of Nuclear Energy in Taiwan
6.1 Introduction
6.2 Brief Illustration of Nuclear Power Plants
6.3 Safety of Nuclear Power Generation
6.4 Nuclear Safety Enhancement
6.5 Radioactive Waste Management
6.6 Conclusions
Chapter 7: Regulatory Challenges on Safety of Nuclear Power Plants in Taiwan
7.1 Introduction
7.2 Challenge I: New Evidence of Active Faults Near Plants
7.3 Challenge II: Aging Management
7.4 Challenge III: Risk-Informed In-Service Inspection (RI-ISI)
7.5 Challenge IV—Chinshan Independent Spent Fuel Storage Installation (ISFSI) Program
7.6 Challenge V: Post-Fukushima Safety Reassessment of NPPs
7.7 Concluding Remarks
References
Chapter 8: Concrete Properties, Safety, and Sustainability of Nuclear Power Plant Infrastructures: New Tools and Themes for Future Research
8.1 Introduction
8.2 Tools for Design and Analysis: Advanced Damage Modeling
8.3 Application to Reinforced Concrete Structures
8.4 Aging Monitoring
8.5 Perspectives and Conclusions
References
Chapter 9: Small Modular Reactors: Infrastructure and Other Systems
9.1 Introduction
9.2 Advantages of SMRs
9.3 Regulatory and Technical Issues
9.4 Design Features of iPWRs
9.5 Conclusions
Part II: Containment Structures
Chapter 10: Seismic Design of Reinforced Concrete Structures in Japan: NPP Facilities and High-Rise Buildings
10.1 Introduction
10.2 Safety Review System of Facilities in Japan
10.3 Design Earthquake Motion for Structures
10.4 Modeling of Structures for a Response Analsyis
10.5 Design Criteria of Structures
10.6 Concluding Remarks
References
Chapter 11: Nonlinear Modeling of 3D Structural Reinforced Concrete and Seismic Performance Assessment
11.1 Introduction
11.2 Construction of a Non-Orthogonal Cracking Model for Three Dimensions and Six Directions
11.3 Path-Dependent Variables Defining the Non-Orthogonal Crack Group and its Setting
11.4 Verification at the Element Level (Uniform Field)
11.5 Verification at the Member Level (Uniform Stress Field)
11.6 Conclusions
References
Chapter 12: Shear Ductility and Energy Dissipation of Reinforced Concrete Walls
12.1 Introduction
12.2 Shear Theory
12.3 Softened Membrane Model (SMM)
12.4 Conversion of Biaxial Strains to Uniaxial Strains
12.5 Constitutive Model of Concrete in CSMM
12.6 Constitutive Model of Mild Steel Bars in CSMM
12.7 Hysteretic Loops
12.8 Cyclic Shear Ductility and Energy Dissipation
12.9 Framed Shear Walls Under Cyclic Loading
12.10 Earthquake Application
12.11 Conclusions
References
Chapter 13: Behavior of Reinforced Concrete Elements Subjected to Tri-Directional Shear Stresses
13.1 Introduction
13.2 Previous Research Studies on Structures Subjected to a 3D State of Stress
13.3 Modeling of RC Elements under a 3D State of Stress
13.4 The Universal Panel Tester
13.5 Installation of Out-of-Plane Hydraulic Cylinders
13.6 Application of Out-of-Plane Shear in the Universal Panel Tester
13.7 Test Program
13.8 Behavior of Test Panels Under Tri-Directional Shear Loads
13.9 Interaction Surface of Bi-Directional Shear Stresses
13.10 Summary and Conclusions
Acknowledgments
References
Chapter 14: Pre-Stressed Concrete Containment Structural Design in China
14.1 Introduction
14.2 Design Improvements of Pre-Stressed Concrete Containment in Chashma Nuclear Power Plant
14.3 Performance Analysis and Experiment Investigation of Third-Generation Containments
14.4 Applications of Other Containment Structures in Domestic Nuclear Power Plants
14.5 Conceptual Design of Spherical Pre-Stressed Concrete Containment Structures
14.6 Conclusions on Pre-Stressed Concrete Containments
References
Chapter 15: Steel Plate Concrete Walls for Containment Structures in Korea: In-Plane Shear Behavior
15.1 Introduction
15.2 Fundamentals
15.3 In-Plane Shear Behavior Models
15.4 Experimental Programs
15.5 Conclusions
References
Chapter 16: Lessons Learned from Kashiwazaki-Kariwa NPP after Niigataken Chuetsu-Oki Earthquake (2007) in View of SSI Effect
16.1 Introduction
16.2 Outline of the Earthquake, KK-NPP, and Observed Events
16.3 Simulation Analyses of Observation Records
16.4 Parametric Study on Floor Response
16.5 Conclusions
References
Chapter 17: Blast, Shock, and Impact Hazards to Nuclear Structures
17.1 Introduction
17.2 Hazard Environments and Loads
17.3 Experimental Observations
17.4 Computational and Experimental Analysis
17.5 Design and Construction
17.6 Summary
References
Chapter 18: History of Shear Design Provisions in the ASME/ACI Code for Concrete Reactor Vessels and Containments
18.1 Introduction
18.2 Background of ASME/ACI Code
18.3 Tangential Shear Design Provisions
18.4 Peripheral Shear Design Provisions
18.5 Radial Shear Design Provisions
18.6 Summary
References
Chapter 19: US NRC Requirements for Containment Structure Design
19.1 Introduction
19.2 Seismic Analysis for Containment Structures
19.3 Design of Containment Structure
19.4 Conclusions
Disclaimer
References
Part III: Computer Software for Containment Structures
Chapter 20: FE Program SCS for Analyzing Wall-Type Concrete Structures
20.1 Introduction
20.2 Material Scale
20.3 Element Scale
20.4 Structure Scale
20.5 Validation
20.6 Conclusions
References
Chapter 21: Modeling and Analysis of Nuclear Power Plant Structures Using ANATECH-ANACAP Software System
21.1 Introduction
21.2 Concrete Constitutive Formulation in ANACAP-U
21.3 Example Applications
References
Chapter 22: SASSI FE Program for Seismic Response Analysis of Nuclear Containment Structures
22.1 Introduction
22.2 Methodology
22.3 Summary
Acknowledgments
References
Chapter 23: FE Program LS-DYNA for Analysis of NPP Structures Including Seismic Soil–Structure Interaction
23.1 Introduction
23.2 Relevant Strengths of LS-DYNA
23.3 Analysis Framework
23.4 Perfectly Matched Layer (PML)
23.5 Effective Seismic Input (ESI)
23.6 Numerical Results
23.7 Conclusions
References
Chapter 24: FE Program ATENA for Safety Assessment of NPP Containments
24.1 Introduction
24.2 Material Model for Concrete
24.3 Validation
24.4 Nonlinear Analysis of Containment Structures
24.5 Conclusions
References
Part IV: Nuclear Waste Storage Facilities
Chapter 25: Properties of Concrete Required in Nuclear Power Plants
25.1 Introduction
25.2 Chemical Attack, Freezing-and-Thawing Cycling
25.3 Permeability and Diffusivity
25.4 Radiation-Shielding Capability and Irradiation Effects
25.5 Volume Changes and Creep
25.6 Thermal and Fire Exposure
25.7 Concrete for Waste-Disposal Structures
25.8 Conclusions
Acknowledgments
References
Chapter 26: Concrete under High Temperature
26.1 Introduction
26.2 The Coupling Among Hygro-Thermo-Mechanical Loading
26.3 Modeling Coupling
26.4 Acceleration of Basic Creep of Concrete by Temperature
26.5 Experimental Data
26.6 High Temperature Test Data
26.7 Concrete Strength Data
26.8 Remarks on Temperature Concrete Data
26.9 Thermo-Elastoplastic Concrete Model
26.10 Loss of Bounded Material Response
26.11 Conclusions
References
Chapter 27: Irradiation Effects on Concrete Structures
27.1 Introduction
27.2 Background
27.3 Microstructures
27.4 Interaction Between Radiation and Materials
27.5 Mechanism of Concrete Deterioration
27.6 Gamma Ray Irradiation Tests
27.7 Conclusions
Acknowledgments
References
Chapter 28: Activities in Support of Continuing the Service of Nuclear Power Plant Safety Related Concrete Structures
28.1 Introduction
28.2 Concrete Structures
28.3 In-Service Inspection and Testing Requirments
28.4 Renewal of Operating Licenses
28.5 Operating Experience and Material Performance
28.6 Management of Aging
28.7 Potential Research Topics
28.8 Summary
References
Chapter 29: Spent Nuclear Fuel Final Disposal in Taiwan
29.1 Introduction
29.2 Disposal Program
29.3 Operation Organization and Work Delegation
29.4 Nuclear Backend Fund
29.5 2009 Progress Report
29.6 Conclusions
References
Chapter 30: Safety Features of Dry Storage System at Chinshan Nuclear Power Plant
30.1 Introduction
30.2 Major Components and Operation Sequence
30.3 Major Safety Features
30.4 Conclusions
References
Chapter 31: Seismic Consequence Modeling for the Yucca Mountain Repository Project
31.1 Introduction
31.2 Description of the Repository
31.3 The Pre-Closure Safety Case
31.4 The Post-Closure Safety Case
31.5 Summary
References
Index
This edition first published 2014
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Library of Congress Cataloging-in-Publication Data
Infrastructure systems for nuclear energy / editors, Thomas T. Hsu, Chiun-lin Wu, Jui-Liang Lin.
pages cm
Summary: “Infrastructure Systems for Nuclear Energy summarizes this progress with an up-to-date reference to guide the future research and design of infrastructure systems for nuclear energy”— Provided by publisher.
Includes bibliographical references and index.
ISBN 978-1-119-97585-4 (hardback)
1. Nuclear power plants— Earthquake effects. 2. Reinforced concrete construction— Design and construction. 3. Nuclear power plants— Design and construction. 4. Nuclear power plants— Asia— Case studies. 5. Nuclear facilities— Earthquake effects. 6. Nuclear facilities— Design and construction. 7. Earthquake resistant design. I. Hsu, Thomas T. C. (Thomas Tseng Chuang), 1933-
TK9152.163I54 2013
621.48— dc23
ISBN 9781119975854
List of Contributors
Preface
Creating sufficient energy resources to replace traditional, carbon-polluting coal and oil is a crucial question for the health of Planet Earth and the survival of human beings. We are at the cross-roads of deciding whether the unlimited potential of nuclear energy should be utilized. The 2011 Fukushima accident re-opened in the public's mind the question: “Are nuclear power plants safe?” This book intends to sort out the facts from the fallacies on nuclear infrastructures.
This book, Infrastructure Systems for Nuclear Energy, addresses both the design and analysis of infrastructures associated with nuclear energy. It first leads readers through an overview of the current/future nuclear power industry and its infrastructure systems. To achieve a comprehensive overview, we invited contributions from the world's top authorities in key areas of expertise, such as regulations, operation, engineer practices, teaching and research. Only through open sharing of scientific studies, not emotional conjecture, can real consensus based on knowledge be built.
This book then provides a thorough investigation of containment structures, nuclear waste storage facilities, and the associated applications of selected commercial/academic computer softwares. Further chapters are devoted to the behavior of infrastructure systems when they are challenged by specific environment issues faced by nuclear power plants, such as the impact of high temperatures, irradiation effects, blasts, earthquakes, soil–structure interaction effects, and so on.
The authors of the book's 31 chapters come from all over the world and are leaders at the cutting edges of their professions. This book, therefore, not only represents the state-of-the-art of nuclear infrastructure systems, but also serves as an authoritative guide for teaching, research, and training of engineers and scientists of the future. This book will serve readers from diverse disciplines and backgrounds. For clarity and convenience, we provide a list of acronyms and their fuller definitions.
This book is the result of institutional cooperation between the University of Houston THSRL (Thomas T. C. Hsu Structural Research Laboratory) in the USA and NCREE (National Center for Research on Earthquake Engineering) of Taiwan. This cross-continent collaboration gave impetus to the signing of an Agreement of Scholarly Exchange and Collaboration in 2001 that was renewed in 2006 and 2011.
The editors would like to thank Professors Shyh-Jiann Hwang, Liang-Jenq Leu, Kuo-Chun Chang, and Chien-Chz Hsu of the National Taiwan University and NCREE for their strong support in terms of space, facilities, manpower, and expenses. The editors also wish to express their appreciation to Ms. Hui-Chu Ni of DnE Information Service Net, whose talent and diligence made possible the completion of this book.
Thomas T. C. HsuChiun-Lin WuJui-Liang LinJuly 4, 2013
Acronyms
Chapters
10 CFR
Title 10 of the Code of Federal Regulations
19
3D
Three Dimension
16
A
ABWR
Advanced Boiling Water Reactor (light-water cooled)
2
ACI
American Concrete Institute
18, 19, 28
AEC
Atomic Energy Council, Taiwan
7, 29, 30
AFNR
Advanced Fast Neutron spectrum Reactor (sodium cooled)
2
AGR
Advanced Gas-cooled Reactor (carbon dioxide cooled)
2
AIJ
Architectural Institute of Japan
15
ALWR
Advanced Light-Water Reactor
2
AMP
Aging Managing Programs
2
ANS
American Nuclear Society
28
AOS
Add-On-Shield
30
ASCE
American Society of Civil Engineers
5
ASME
American Society of Mechanical Engineers
7, 18, 19, 26, 28
ASP
Aging Surveillance Programs
2
B
BARC, India
Bhabha Atomic Research Centre in India
24
BDBE
Beyond Design-Basis Earthquake
5
Breeders-FBRs
See FBR
2
BWR
Boiling Water Reactor
1, 2, 10, 16, 17, 28
BWRVIP
Boiling Water Reactor Vessel and Internals Project
7
C
C
Celsius
19
CAMUS
French Seismic Research Program
8
CANDU
CANada Deuterium Uranium (reactor) (heavy water moderated)
2
CCF
Common-Cause Failure
2
CDF
Core Damage Frequency
2
CEA
French Atomic Agency; Also, French Alternative Energies and Atomic Energy Commission
8
CFR
Code of Federal Regulations
1, 18, 28, 31
CO2
Carbon Dioxide
2
CONMOD
CONcrete containment MODeling and management
28
CPR
Chinese (design) Pressurized (light-water cooled) Reactor
2
CRWMS M&O
Civilian Radioactive Waste Management Systems, Management and Operating contractor
31
D
DBA
Design-Basis Accident
17
DBE
Design-Basis Earthquake
5, 30
DBGM
Design-Basis Ground Motions
31
DID
Defense-In-Depth
2
DM
Direct Method
22
DNBM
Department of Nuclear Backend Management
29
DOD
Department of Defense
17
DOE
Department of Energy
1, 9, 17, 31
E
E/B
External Building
10
EACD
Earthquake Analysis of Concrete Dams, a computer program developed at the University of California, Berkeley
23
EBS
Engineered Barrier System
31
ECCS
Emergency Core Cooling Systems
1, 2
EPR
European (evolutionary) Pressurized (light-water cooled) Reactor, a third-generation PWR
1, 2, 22
EPRI
Electric Power Research Institute
7, 21, 28
EPU
Extended Power Up-rate
2
ESBWR
Economic Simplified Boiling Water Reactor (light-water cooled)
2
ESI
Effective Seismic Input
23
ETH Zurich
Eidgenössische Technische Hochschule Zürich, Swiss Federal Institute of Technology, Zurich
24
EUR
European Utility Requirements
22
EW
East–West
16
F
F
Fahrenheit
19
FBR
Fast (neutron spectrum) Breeder Reactor
2
FE
Finite Element
22
FEM
Finite Element Method
16
FEP
Features, Events, and Processes
29
FSAR
Final Safety Analysis Report
30
FTS
Free Thermal Strain
26
G
GALL
Generic Aging Lessons Learned
28
GDC
General Design Criteria, or General Design Criterion
1, 18, 19
GE
General Electric Company
2
GEH
GE-Hitachi nuclear energy co.
2
GEL
Green Energy and Environment Research Laboratories
29
GGE
Greenhouse Gas Emission
2
GWh
GigaWatt-hour (10
9
W-hour)
2
H
HCLPF
High-Confidence-Low-Probability of Failure
3, 5
HDB
High-Damping rubber Bearing
5
HE
High-Explosive
17
HF
High-Frequency
19
HLW
High-Level Waste
31
I
I/C
Inner Concrete
10
IAEA
International Atomic Energy Agency
1, 2, 28
IED
Improvised Explosive Device
17
ILRT
Integrated Leakage Rate Test
1
INER
Institute of Nuclear Energy Research, Taiwan
29, 30
IPP
Independent Power Producers
6
iPWR
integrated PWR
9
ISFSI
Independent Spent Fuel Storage Installation
30
ISG
Interim Staff Guidance used by the Spent Fuel Project Office of US Nuclear Regulatory Commission
19, 22
ITRI
Industrial Technology Research Institute
29
ITS
Important To Safety
31
ITWI
Important To Waste Isolation
31
J
JEAG
Japan Electric Association Guidelines
5, 15
JMA
Japan Meteorological Agency
16
JNES
Japan Nuclear Energy Safety Organization
5
JRC
Joint Research Centre, European Commission
8
K
KEPCO
Kansai Electric Power COmpany
10
KEPRI
Korea Electric Power Research Institute
15
KK
Kashiwazaki-Kariwa NPP
16
KM
Knowledge Management
2
KSSC
Korean Society of Steel Construction
15
kWh
kiloWatt-hour (10
3
W-hour)
2
L
LCS
Large-Core Solution
22
LDB
Low-Damping rubber Bearing
5
LITS
Load-Induced Thermal Strain
26
LLNL
Lawrence Livermore National Laboratory
23
LLRT
Local Leakage Rate Test
1
LOCA
Loss-Of-Coolant Accident
1, 2
LR
License Renewal
2
LRB
Lead-Rubber Bearing
5
LSTC
Livermore Software Technology Corporation
23
LTO
Long-Term Operation
2
LWR
Light-Water Reactor
27
M
MAFE
Mean Annual Frequency of Exceedance
3
MCMC
Markov Chain Monte Carlo
8
METI
Ministry of Economy, Trade and Industry, Japan
10
MLIT
Ministry of Land, Infrastructure, Transport and tourism, Japan
10
MSM
Modified Subtraction Method
22
MUR
Measurement Uncertainty Recapture
6
MUR PU
Measurement Uncertainty Recapture Power Up-rate
2
MWd
MegaWatt-day
2
MWe
MegaWatt-electrical
2
MWh
MegaWatt-hour (10
6
W-hour)
2
N
NCO
Niigataken Chuestu-Oki earthquake
16
NEA
Nuclear Energy Agency
2, 28
NEUP
Nuclear Energy University Program
1
NGNP
Next Generation Nuclear Plant
2
NI
Nuclear Island
22
NISA
Nuclear and Industrial Safety Agency, Japan
10
NPP
Nuclear Power Plant
2, 3, 6, 16, 24, 27, 28, 30
NRC
Nuclear Regulatory Commission, See US NRC
1, 2, 9, 17, 18, 19, 31
NS
North–South
16
NSC
Nuclear Safety Commission of Japan
10
NSSS
Nuclear Steam Supply System
6
NUPEC
Nuclear Power Engineering Cooperation, Japan
17, 21
O
O&M
Operation and Maintenance
2
OBE
Operating-Basis Earthquake
19
OECD
Organization of Economic Cooperation and Development (Europe)
2
OFR
Office of Federal Regulations
28
OPs
Operational Practices
2
ORNL
Oak Ridge National Laboratory
26
P
PCRV
Pre-stressed Concrete Reactor Vessel
26
PGA
Peak Ground Acceleration
3, 8, 31
PGV
Peak Ground Velocity
31
PHWR
Pressurized Heavy-Water Reactor (CANDU)
2
P-I
Pressure-Impulse
17
PLiM
Plant Life Management program
2
PLM
Plant Life Management
27
PML
Perfectly Matched Layer
23
pp fibers
Polypropylene fibers. The recommended volume content of 0.1% pp fibers is approximately a dosage of 0.9 kg/m3 monofilament
25
PRA
Probabilistic Risk Assessment
2
PRM
Pontiroli, Rouquand, Mazars damage model
8
PSHA
Probabilistic Seismic Hazard Analysis
3, 5
psig
pounds per square inch, gauge
19
PSR
Periodic Safety Review
2
PT
Penetrant Test
30
PU
Power Up-rate
2
PVP
Pressure Vessels and Piping
26
PWh
PetaWatt-hour (10
15
W-hour)
2
PWR
Pressurized Water Reactor
1, 2, 9, 10, 17, 28
Q
QA
Quality Assured
2
R
R.G.
Regulatory Guide
28
R/C
Reinforced Concrete
10
RB
Reactor Building
30
RBMK
Reaktor Bolshoy Moschnosti Kanalij (Russian high power channel-type reactor)
2
RCCV
Reinforced Concrete Containment Vessel
10, 21
RGs
Regulatory Guides
19
RH
Relative Humidity
8
RILEM
International Union of Laboratories and Experts in Construction Materials, Systems, and Structures
28
RIP
Reactor Internal Pump
2
RPV
Reactor Pressure Vessel
2
RT
Radiographic Test
30
S
S/G
Steam Generators
10
SA
Spectral Acceleration
31
SAR
Safety Analysis Report
30
SASSI
System for Analysis of Soil–Structure Interaction, a computer program
22
SDC
Seismic Design Category
5
SFP
Spent Fuel Pool
1, 30
SG
Steam Generator
9
SKM
SKin Method
22
SM
Subtraction Method
22
SMiRT 18
18th International Conference on Structural Mechanics in Reactor Technology
19
SMR
Small Modular Reactor
9
SNF
Spent Nuclear Fuel
31
SNL
Sandia National Laboratories
17, 31
SPRA
Seismic Probabilistic Risk Assessment
3
SPU
Stretch Power Up-rate
2
SRV
Steam Relief Valve
2
SSC
Structures, Systems, and Components (in NPPs)
1, 28
SSCs
see SSC
2, 19
SSE
Safe-Shutdown Earthquake
19
SSHAC
Senior Seismic Hazard Analysis Committee
3
SSI
Soil–Structure Interaction
16, 19, 22, 30
SSSI
Structure–Soil–Structure Interaction
19
T
TEPCO
Tokyo Electric Power C0mpany
10
TEV
Transport and Emplacement Vehicle
31
TFR
Transfer Cask
30
TG/DTA
Thermogravimetry and Differential Thermal Analysis
27
TMSL
Tokyo Mean Sea Level
10
TPC
Taiwan Power Company
7, 29, 30
TPRI
Taiwan Power Research Institute
29
TSC
Transportable Storage Canister
30
TSPA
Total System Performance Assessment
31
TTC
Transitional Thermal Creep
26
TWh
TeraWatt-hour (10
12
W-hour)
2
U
UD
Up–Down
16
UHRS
Uniform Hazard Response Spectrum
5
UO2
Uranium Dioxide
2
US DOE
United States Department Of Energy
28
US NRC
United States Nuclear Regulatory Commission
5, 21, 28
UT
Ultrasonic Test
30
V
VCC
Vertical Concrete Cask
30
VHTR
Very High Temperature Reactor
2
VVER
Vodo-Vodyanoi-Energetichesky-Reactor; See WWER
2
W
WANO
World Association of Nuclear Operators
6
Wh
Watt-hour
2
WSCP
Water and Soil Conservation Plan
30
WWER
Water-cooled, Water-moderated light water Reactor (Russian design of a PWR); See VVER
2
Chapter 1
Introduction
Thomas T. C. Hsu1, Chiun-Lin Wu2, and Jui-Liang Lin2
1Moores Professor, Department of Civil and Envionmental Engineering, University of Houston, Houston, TX, USA
2Associate Research Fellow and Research Fellow, respectively, National Center for Research on Earthquake Engineering, Taipei, Taiwan
1.1 International Workshop on Infrastructure Systems for Nuclear Energy
The International Workshop on Infrastructure Systems for Nuclear Energy was organized by the joint effort of the University of Houston, USA, and the National Center for Research on Earthquake Engineering (NCREE), Taiwan, and was held on December 15–17, 2010, in Taipei. The workshop called upon top international experts in research, industry, and governmental regulatory agencies to share knowledge, discuss current events/data, evaluate the newest technologies and practices in nuclear energy infrastructure systems, and formulate the best strategies for design of nuclear infrastructure—specifically focused on safety issues. Speakers came from the United States, Japan, Korea, Italy, France, Switzerland, Netherland, Russia, the Czech Republic, China, and of course, Taiwan (http://iwisne.ncree.org.tw). Thirty papers covering the most important current topics in nuclear infrastructure are included in this book, Infrastructure Systems for Nuclear Energy.
It is our sincere hope that our work will guide future studies on nuclear infrastructure and will encourage timely international workshops where open discussions can lead the way towards a safer and energy-sufficient future. In fact, a previous international workshop on nuclear containment structures was held in 1991 at the University of Houston, USA. It was organized by Professors Thomas T. C. Hsu and S. T. Mau, and was sponsored by the National Science Foundation. That program produced the book Concrete Shear in Earthquake (1992, Elsevier Applied Science), which has, indeed, guided many researchers and practitioners over the past 20 years.
Safety has been, and will continue to be, a major factor in nuclear energy, realistically and politically. The worst of the three major reactor meltdown accidents in history was the 1986 “Chernobyl Disaster” in Russia, in which case the absence of containment structures to protect the nuclear reactors accounted for the unmitigated spilling of radioactivity and the resulting disastrous impacts. The 1979 “Three Mile Island Incident” in the USA was an operational error and the radiation leak was sufficiently protected by the containment structures such that, to date, no known death had resulted from that accident. The 2011 “Fukushima Disaster” in Japan demonstrates the importance of anticipating and preparing against possible multiple threats (earthquake and tsunami) of unprecedented magnitude.
According to the September, 2013 report by the International Atomic Energy Agency (IAEA) (http://www.world-nuclear.org/info/reactors.html), there are 432 nuclear power reactors operable around the world in some 30 countries, predominantly in Europe, followed by the Americas and Asia. Additionally, 68 nuclear power reactors are under construction, mostly in Asia and Russia (Table 1.1). Nuclear energy generates between 2–75% of the total electrical power of these countries (Table 1.2) (http://www.world-nuclear.org/info/nshare.html).
Table 1.1 Number of nuclear reactors around the world (as of September 2013)
Table 1.2 Share of nuclear power, 2002–2012 (as of May 2013) (http://www.world-nuclear.org/info/nshare.html)
In order to further develop nuclear energy, the engineers dealing with infrastructure systems must assure the world that they can solve two basic problems:
To build safe and economical containment structures to protect nuclear reactor vessels from leaking radioactive pollutants, particularly in earthquake regions.
To safely and economically store nuclear wastes, both in nuclear power plants and in designated underground sites for long-term storage of radioactive nuclear wastes.
The future of nuclear energy depends strongly on creating safe and affordable infrastructure systems that can safely resist earthquakes, as well as other natural and artificial hazards.
The 2010 workshop was made possible by the strong support of Taiwan's National Center for Research on Earthquake Engineering (NCREE), and all the sponsors and co-sponsors (http://iwisne.ncree.org.tw). Chief among the academic institutions are the National Taiwan University, the University of Houston, the University of Tokyo, the Politecnico di Milano, and the Seoul National University. The strongest governmental and industrial supports included the National Science Council of Taiwan, the National Applied Research Laboratories, the Atomic Energy Council of Taiwan, the Institute of Nuclear Energy Research, the Fuel Cycle and Materials Administration, the Taiwan Power Company, the Sinotech Engineering Consultants, Ltd., and the Taiwan Chapter of International Academy of Engineering.
1.2 Overview of Nuclear Power Plants
At present, electricity is mostly generated by thermal power plants. In such an installation, coal or oil is burned in a furnace to heat water into steam in order to spin a turbine. A steam turbine drives an electrical generator to produce electricity, which is transported long distances to the users by high-voltage copper wires. A nuclear power plant (Figure 1.1) is essentially the same as a thermal power plant using fossil fuels, except that the source of energy to heat water comes from nuclear fission of radioactive materials, such as uranium and plutonium, in a nuclear reactor vessel.
Figure 1.1 A schematic drawing of the nuclear power plant (Source: US NRC)
In a thermal power plant, the burning of fossil fuels emits large amounts of carbon dioxide (CO2) which causes global warming. Efforts to reduce these undesirable outputs are urgent and widespread. In a nuclear power plant, the nuclear fission of uranium in the reactor vessel does not produce CO2, but does release various types of harmful radiations. In order to prevent the radiations from contaminating the atmosphere during a rare nuclear meltdown accident, the reactor vessel is housed in a large and strong containment structure, which serves as a barrier to insulate the radioactivity from the environment. From this point of view, the containment structure is one of the most important infrastructure systems ensuring the safety of a nuclear power plant.
In a nuclear reactor vessel, the spent fuel assemblies (each consisting of 60 to 80 fuel rods) are replaced by new ones over a 3–6-year period. Because the temperature and radioactivity of spent fuel assemblies are still very high, their storage becomes an important issue in a nuclear power plant. After leaving the reactor vessel, the spent fuel assemblies should be immerged in a spent fuel pool full of water. Once the decaying heat and radiation of the spent fuel assemblies reach an acceptable level, these assemblies are moved from the spent fuel pool into a dry cask for longer-term storage. This book, therefore, emphasizes two kinds of infrastructure systems in a nuclear power plant, namely, the nuclear containment structures and the nuclear waste storage facilities. The latter consists of the spent fuel pool, the dry cask storage and the final storage at, possibly, the Yucca Mountain Repository in the USA, or the Onkalo spent nuclear fuel repository currently under construction in Finland.
Depending on the progress of the nuclear reactor vessels since the 1950s, nuclear power plants are divided into five generations (I, II, III, III+, and IV) as shown in Figure 1.2 (OECD, 2012).1 At present, Generation III+ plants are being constructed and Generation IV plants are on the drawing board and not expected to be available for commercial construction before 2030.
Figure 1.2 Five generations of nuclear power plants (Reproduced with permission from OECD/Nuclear Energy Agency (2012) Nuclear Energy Today Second Edition http://www.oecd-nea.org/pub/nuclearenergytoday/6885-nuclear-energy-today.pdf)
1.3 Infrastructure for Nuclear Power Industry
The term “infrastructure for nuclear energy” has different meanings for different people. To the planners and managers of nuclear power plants, “nuclear infrastructure” or “nuclear industry infrastructure” is considered to have four components:
technological infrastructure
regulatory infrastructure
computer technology infrastructure
human infrastructure.
1.3.1 Technological Infrastructure
Two essential industries are required to support a new or existing nuclear plant: (i) suppliers and fabricators of nuclear fuel and safety related components; and (ii) suppliers of balance-of-plant equipment, construction materials, electronics, instrumentation, and so on.
In the USA, the fuel-cycle industry has undergone significant changes. Future fluctuations in uranium prices, the deployment of new enrichment technologies, significant consolidation of fuel-cycle supply companies, and the possible recycling of spent fuel could all affect the supply chain for nuclear fuel. New reactor deployment must demonstrate the existence of an industrial infrastructure to supply the proposed fuel.
The waste-management component of fuel-cycle infrastructure should address the on-site storage of spent nuclear fuel; but the current lack of a national system for high-level waste disposal is a problem common to all new technologies. The Nuclear Waste Policy Act (first passed in 1983) provides for the development of the Yucca Mountain site in Nevada as a mined, geologic repository for high-level waste and the development of a transportation system linking US nuclear power plants, an interim storage facility, and the permanent repository (NEI, 2002c).
The recycling of spent fuel to recover fissile material and long-lived heavy elements would reduce the heat generation and volume of final waste products. At the time that most existing US nuclear plants were built, the industry—encouraged by the federal government—planned to recycle used nuclear fuel by recovering plutonium. In 1979, however, the United States deferred the reprocessing of all commercial used nuclear fuel because of concerns about the possible proliferation of nuclear weapons, and the industry was forced to adopt a once-through, single-use fuel cycle. As a result, reprocessing and recycling are currently not prevalent in the United States.2
1.3.2 Regulatory Infrastructure
The US Nuclear Regulatory Commission (US-NRC) is perhaps the best-known government agency for regulating the construction and operation of nuclear reactors, nuclear containment structures, and waste-management facilities. This agency also regulates nuclear fuel-cycle operations, including the mining, milling, conversion, enrichment, and fabrication of fuels.
NRC regulations limit commercial reactor licenses to an initial 40 years (based on economic and antitrust considerations) but also allow licenses to be renewed (NRC, 2002). The license renewal process involves confirmation that structures, systems, and components (SSC) can continue to perform safely beyond the original term of the operating license. License renewal is important for maintaining a significant portion of existing energy production. The first plant reached the end of its original license period in 2006; approximately 10% reached 40 years by the end of 2010; and more than 40% will reach 40 years by 2015.
A key aspect of regulatory development for near-term deployment is the efficient implementation of 10 CFR Part 52. Created in 1989, this regulation established three new licensing processes for future plants: early site permitting (i.e., NRC approval of a site before a decision has been made to build a plant); design certification (i.e., NRC approval of a standard design); and combined licenses (i.e., a combined construction permit and operating license).
1.3.3 Computer Technology Infrastructure
The most dramatic contribution to changes in design tools in the last 30 years has been computer technology. With today's computational speeds, optimization and simplification of new plant design and construction seems to be limited only by the imagination.
The contribution of computer technology advancement to the design of nuclear power plants is well demonstrated by the five chapters in Part Three of this book. The computer softwares for concrete containment structures are now applicable to three-dimensional analysis and are striving to use realistic nonlinear constitutive relationships.
1.3.4 Human Infrastructure
One of the most important aspects of nuclear infrastructure is the development of human resources. The nuclear industry faces the dual challenge of an aging workforce and a growing gap between its employment needs and the number of graduating students. Replacement of the aging workforce is essential for both existing plants and new facilities.
Several initiatives to strengthen the human resources have been made by the DOE, NEI, and the American Nuclear Society (DOE, 2001). In 2002, DOE announced the establishment of a new program, Innovations in Nuclear Infrastructure and Education, that offers millions of dollars in awards to university consortia to encourage investment in programs on research reactors and nuclear engineering and in strategic partnerships with national laboratories and industry. This was followed in 2009 by the establishment of the Nuclear Energy University Program (NEUP), which awarded in 2013 a total of 61 research grants, totaling 42 million dollars (https://inlportal.inl.gov/portal/server.pt/community/neup_home/600/FY13_R&D_awards).
Other public and private efforts to bolster the educational infrastructure for the nuclear industry are underway throughout the world. As a result, many universities and national laboratories associated with nuclear science and engineering now offer scholarships and fellowship programs.
Eight chapters dealing with “nuclear infrastructure” and written by leaders of the nuclear industry are assembled in Part One of this book from Chapter 2 to Chapter 9.
1.4 Containment Structures3
To the engineers who design the physical facilities of nuclear power plants, “nuclear infrastructure” could be defined as the whole physical plant except the nuclear reactor vessel and the piping system. In a nuclear power plant, the nuclear reactor vessel and the piping systems are the operating systems that generate the electricity. All the other structures that support the operating system are called the infrastructure systems. The two most important components of the infrastructure systems are the containment structures, in Part Two (10 chapters) and Part Three (5 chapters) of this book, and nuclear waste storage facilities, in Part Four (7 chapters). Since these two components are the focus of this book, the title of the book has been chosen to be Infrastructure Systems for Nuclear Energy.
A containment structure, in its most common usage, is a steel or reinforced concrete structure enclosing a nuclear reactor. It is designed to contain the escape of radiation up to a maximum pressure in the range of 410 to 1400 kPa (60 to 200 psi). The containment is the fourth and final barrier to radioactive release (part of a nuclear reactor's defense in depth strategy). The first barrier is the ceramic casing for fuel pellets, the second is the metal cladding tubes made of stainless steel or zirconium alloy, and the third is the reactor vessel and coolant system (Figure 1.3).
Figure 1.3 Schematic drawing of the containment building
In early designs, the containment boundary was typically an airtight steel structure enclosing the reactor vessel and normally sealed off from the outside atmosphere. The steel is either free-standing or attached to the concrete missile shield. In the United States, the design and thickness of the containment and the missile shield are governed by federal regulations (10 CFR 50.55a), and must be strong enough to withstand the impact of a fully loaded passenger airliner without rupture.
While the containment plays a critical role in the most severe nuclear reactor accidents, it is only designed to contain or condense steam in the short term (for large break accidents) and long-term heat removal still must be provided by other systems. In the Three Mile Island accident, the containment pressure limit was maintained, but due to insufficient cooling, radioactive gas was intentionally let from containment by operators to prevent over-pressurization. This, combined with further failures, caused the release of minimal amounts of radioactive gas to the atmosphere during the accident.
Information is still being gathered on the failures at Fukushima. While the plant had operated safely since 1971, an earthquake and tsunami well beyond the design basis resulted in the failure of all safety systems, including AC power, backup generators, and backup batteries. This resulted in partial or complete meltdown of fuel rods, damage to fuel storage pools and buildings, significant release of radioactive debris to surrounding area, air, and sea, and resorting to the use of fire engines, concrete mixing trucks, and helicopters to deliver cooling water to spent fuel pools and containments.
If the outward pressure from steam in a limiting accident is the dominant force, containments tend to be a spherical design. If the weight of structure is the dominant force, designs tend to be a can shape. Modern designs tend toward a combination of both.
Containment structures for nuclear power reactors vary in size, shape, materials used, and in their suppression systems. Suppression systems are critical to safety analysis and greatly affect the size of containment. Suppression refers to condensing the steam after a major break has released it from the cooling system. Because decay heat does not go away quickly, there must be some long term method of suppression. For this purpose, containments are categorized as either “large-dry,” “sub-atmospheric,” or “ice-condenser.”
1.4.1 The Pressurized Water Reactors
For a pressurized water reactor (PWR), the containment also encloses the steam generators and the pressurizer, and is the entire reactor building. The missile shield around it is typically a tall cylindrical or domed building. PWR containments are typically large (up to 10 times larger than a BWR) because the design strategy of the containment is to provide adequate volume for the steam/air mixture that results from a loss-of-coolant accident to expand into, thus limiting the ultimate pressure (driving force for leakage) reached in the containment building.
Early containment designs—including those of Siemens, Westinghouse, and Combustion Engineering—had a mostly can-like shape built with reinforced concrete. As concrete has very good compression strength compared to tensile, this is a logical design for the building materials since the extremely heavy top part of the containment exerts a large downward compressive force that counteracts some tensile stress if internal pressure suddenly goes up. As reactor designs have evolved, many nearly spherical containment designs for PWRs have also been constructed. This is the most logical design for containing a large internal pressure. Most current PWR designs involve some combination of the two, with a cylindrical lower part and a half-spherical top. Figure 1.4 shows some examples of the containments with different shapes.
Figure 1.4 Examples of the containments with different shapes. Note that the large concrete structures of hyperbolic shape in (a) and (d) are cooling towers. (a) Three Mile Island was an early PWR design with ‘can’ type containments (Source: http://ma.mbe.doe.gov/me70/history/photos.htm); (b) A more detailed image for the ‘can’ type containment in France (Source: http://en.wikipedia.org/wiki/File:Centrale_nucleaireBrennilis.jpg); (c)The twin PWR reactor containments in U.S. (Source: http://en.wikipedia.org/wiki/File:Donald_Cook_Nuclear_Power_Plant_1993.jpg); (d) A nearly completely spherical containment design in Germany (Courtesy of Rainer Lippert)
Newer designs, notably the AP1000 and the European Pressurized Reactor, call for both a steel and concrete containment (http://www.ap1000.westinghousenuclear.com/ exploreap1000.html). The outer concrete shield building provides missile protection and the inner steel structure resists internal pressures to prevent leakage. The outer shield building of AP1000 is made of steel–concrete composite with a thick concrete wall sandwiched between two steel plates at the surface. In order to cool the inner steel structure during leakage, the shield building has a large ring tank on the conical roof to provide water by gravity, as well as providing air inlets just below the roof and an air outlet at the roof center for air circulation.
1.4.2 The Boiling Water Reactors
In a boiling water reactor (BWR), the containment strategy is different. A BWR's containment consists of a drywell where the reactor and associated cooling equipment is located. The drywell is much smaller than a PWR containment and plays a larger role. During the theoretical leakage design basis accident, the reactor coolant flashes to steam in the drywell, pressurizing it rapidly. Vent pipes or tubes from the drywell direct the steam to below the water level maintained in the wetwell (also known as a torus or suppression pool), thus condensing the steam and limiting the pressure ultimately reached. Both the drywell and the wetwell are enclosed by a secondary containment building, maintained at a slight sub-atmospheric or negative pressure during normal operation and refueling operations.
The GE containment designs are referred to by the names Mark I (http://nuclearstreet.com/nuclear-power-plants/w/nuclear_power_plants/mark-i-containment.aspx), Mark II (http://nuclearstreet.com/nuclear-power-plants/w/nuclear_power_plants/mark-ii-containment.aspx), and Mark III (http://nuclearstreet.com/nuclear-power-plants/w/nuclear_power_plants/mark-iii-containment.aspx). The Mark I is the oldest, characterized by a drywell containment which resembles an inverted light bulb above the wetwell, which is a steel torus containing water. The Mark II was used with late BWR-4 and BWR-5 reactors. It is called an “over-under” configuration with the drywell forming a truncated cone on a concrete slab. Below is a cylindrical suppression chamber made of concrete rather than just sheet metal. Both use a lightweight steel or concrete “secondary containment” over the top floor, which is kept at a slight negative pressure so that air can be filtered. The top level is a large open space with an overhead crane suspended between the two long walls for moving heavy fuel caskets from the ground floor, and removing/replacing hardware from the reactor and reactor well. The reactor well can be flooded and is straddled by pools separated by gates on either side for storing reactor hardware, normally placed above the fuel rods, and for fuel storage. The Mark III uses a concrete dome, somewhat like PWRs, and has a separate building for storing used fuel rods on a different floor level. All three types of containment use the large body of water in the suppression pools to quench steam released from the reactor system during transients.
The Mark I containment was used in those reactors at the Fukushima I Nuclear Power Plant which were involved in the Fukushima I nuclear accident. The site suffered from a combination of two beyond design-basis events: first a powerful earthquake, which may have damaged reactor plumbing and structures, and then a 15-meter tsunami which destroyed fuel tanks, generators, and wiring systems, causing backup generators to fail. Eventually, the battery-powered pumps also failed. Insufficient cooling water led to partial or possible complete meltdown of fuel rods that were completely uncovered by water. This led to releases of significant amounts of radioactive material to the air and sea, and to hydrogen explosions. Since the thin secondary containments were not designed to withstand hydrogen explosions, the blow-out destroyed roofs and walls, causing the destruction of all equipment on the refueling floor, the cranes, as well as the refueling platform itself. Unit 3 suffered a particularly spectacular explosion, which created a plume of debris over 300 m high. This resulted in the collapse of the north end of the top floor, and the buckling of concrete columns on the west side, as shown by aerial photographs. Although the secondary containments were fitted with modified hardened vent systems to vent hydrogen into the exhaust stacks, they may have not been effective without power. Unit 2 had a large panel purposely removed to vent gases, but suffered an explosion in the lower suppression area. Even before the Fukushima incident, the Mark I containment had been criticized as being more likely to fail during a blackout.
From a distance, the BWR design looks very different from PWR design because a square building is usually used for containment. Also, because the turbines and reactor are designed as one loop only, the steam going through the turbines is also slightly radioactive, and the turbine building should be shielded as well. This leads to two buildings of rectangular shape, with the taller one housing the reactor and the shorter but longer one housing the turbine hall and supporting structures (Figure 1.5).
Figure 1.5 (a) A representative one-unit German BWR showing containment around both the turbine and reactor buildings (Courtesy of Hinnerk); (b) A typical two-unit BWR in the USA (Courtesy of Doc Searls); (c) Modern plants have tended towards a design that is not completely cylindrical or spherical, like this painted containment at the Clinton Nuclear Generating Station (Courtesy of Daniel Schwen)
1.4.3 Design and Testing Requirements
Title 10 of the Code of Federal Regulations, Part 50, Appendix A, General Design Criteria (GDC 54–57), or some other design basis, provides the basic design criteria for isolation of pipelines penetrating the containment wall. Each major pipe penetrating the containment, such as the steam lines, has generally two isolation valves, one on the inside and one on the outside of the containment. For large, high-pressure lines, maintenance considerations and providing space for relief valves cause the designers to install the isolation valves close to the wall where the lines exit the containment. In the event of a leak in the high-pressure piping that carries the reactor coolant, these valves could close rapidly to prevent radioactivity from escaping the containment. The containment isolation valves may also close on a variety of other signals, such as high pressure in the containment experienced during a high-energy line break (e.g., main steam or feed water lines). The containment building serves to contain the steam/resultant pressure, but there is typically no radiological consequences associated with such a break at a pressurized water reactor (PWR). Valves on lines for standby systems penetrating the containment are normally closed.
During normal operation, the containment is air-tight and access is only through marine style airlocks. High air temperature and radiation from the reactor core limit the time, measured in minutes, that workers can spend inside the containment while the plant is operating at full power. In the event of a worst-case emergency, called a “design basis accident” in the Nuclear Regulatory Commission (NRC) regulations, the containment is designed to seal off and contain a meltdown. Redundant cooling systems are installed to prevent a meltdown, but as a matter of policy, at least one is required for a containment building. For design purposes, the reactor vessel's piping is assumed to be breached, causing a “LOCA” (loss-of-coolant accident) where the water in the reactor vessel is released inside the containment and flashes into steam. The resulting pressure increase inside the containment, which is designed to withstand the pressure, triggers containment sprays (“dousing sprays”) to turn on in order to condense the steam and thus reduce the pressure. An emergency shutdown by inserting control rods into the reactor core (SCRAM in BWRs or “reactor trip” in PWRs) initiates very shortly after the break occurs. The safety systems close non-essential lines into the air-tight containment by shutting the isolation valves. Emergency Core Cooling Systems are quickly turned on to cool the fuel and prevent it from melting. The exact sequence of events depends on the reactor design.
Containment buildings in the USA are subjected to mandatory testing of the containment and containment isolation provisions under 10 CFR Part 50, Appendix J. Integrated Leakage Rate Tests (ILRTs or Type A tests) are performed on a 15-year basis. Local Leakage Rate Tests (LLRTs or Type B or Type C tests) are performed much more frequently both to identify the possible leakage in an accident and to locate and fix leakage paths. LLRTs are performed on containment isolation valves, hatches, and other appurtenances penetrating the containment. A nuclear plant is required by its operating license to prove containment integrity prior to restarting the reactor after each maintenance shutdown. The requirement can be met with satisfactory local or integrated test results (or a combination of both when an ILRT is performed).
Ten chapters dealing with the design and analysis of containment structures are given in Part Two of this book, from Chapter 10 to Chapter 19. Five chapters dealing with the computer software for containment structures are given in Part Three, from Chapter 20 to Chapter 24.
1.5 Nuclear Waste Storage Facilities
1.5.1 Spent Fuel Pools4
Spent fuel pools (SFP) are storage pools for spent fuel from nuclear reactors. They are typically 12 m (40 ft) or more in depth, with the bottom 4.3 m (14 ft) equipped with storage racks designed to hold fuel assemblies removed from the reactor (http://en.wikipedia.org/wiki/File:Carso_Fuel_pool.jpg). A reactor's pool is specially designed for the reactor in which the fuel was used, and is situated at the reactor site. In many countries, the fuel assemblies, after being in the reactor for 3 to 6 years, are stored underwater for 10 to 20 years before being sent for reprocessing or dry cask storage. The water cools the fuel and shields its radiation.
While only about 2.4 m (8 ft) of water is needed to keep radiation levels below acceptable levels, the extra depth provides a safety margin and allows fuel assemblies to be manipulated without special shielding to protect the operators.
The NRC estimates that many of the nuclear power plants in the United States will have no more room in their spent fuel pools by 2015, most likely requiring the use of temporary storage of some kind.
1.5.2 Operation
About a quarter to a third of the total fuel load of a reactor is removed from the core every 12 to 18 months and replaced with fresh fuel. Spent fuel rods generate intense heat and dangerous radiation that must be contained. Fuel is moved from the reactor and manipulated in the pool generally by automated handling systems, although some manual systems are still in use. The fuel bundles fresh from the core normally are segregated for several months for initial cooling before being sorted and moved to other parts of the pool to wait for final disposal. Metal racks keep the fuel in safe positions to avoid the possibility of a “criticality”—a nuclear chain reaction occurring. Water quality is tightly controlled to prevent the fuel or its cladding from degrading. The maximum temperature of the spent fuel bundles decreases significantly between 2 and 4 years, and less from 4 to 6 years. The fuel pool water is continuously cooled to remove the heat produced by the spent fuel assemblies. Pumps circulate water from the spent fuel pool to heat exchangers, then back to the spent fuel pool. The water temperature under normal operating conditions is held below 50 °C (120 °F). Radiolysis, the dissociation of molecules by radiation, is of particular concern in wet storage, as water molecules may be split by residual radiation, producing hydrogen gas that may accumulate and increase the risk of explosions. For this reason the air in the room of the pools, as well as the water, must permanently be monitored and treated.
1.5.3 Dry Cask Storage5
Dry cask storage is a method of storing high-level radioactive waste, such as spent nuclear fuel that has already been cooled in the spent fuel pool for at least one year (Figure 1.6). The fuel is surrounded by inert gas inside a large container. These casks are typically steel cylinders that are sealed either by welding or bolting. Ideally, the steel cylinder provides leak-tight containment of the spent fuel. Each cylinder is surrounded by additional steel, concrete, or other material to provide radiation shielding to workers and the public. Some of the cask designs can be used for both storage and transportation.
Figure 1.6 (a) and (b) Examples of the dry cask storage area (Source: US NRC, http://www.connyankee.com/html/fuel_storage.html. (b) reproduced from Shih-Jen Wang, Jyh-Jun Chen, Chen-Sheng Chien, Jyh-Tong Teng, A study of heat capacity temperature limit of BWR, Nuclear Engineering and Design, 243 (2012)); (c) and (d) the detail of the spent fuel storage cask (Source: US NRC)
There are various designs for dry storage cask systems. With some designs, the steel cylinder containing the fuel stands vertically in a concrete or steel vault, which provides the radiation shielding. The casks are placed on a concrete pad at a dry cask storage site (Figures 1.6a and b). Other designs orient the cylinders horizontally (Figure 1.6c). During the 2000s, dry cask storage design was used in the United States, Canada, Germany, Switzerland, Belgium, Sweden, the United Kingdom, Japan, and Lithuania.
In the late 1970s and early 1980s, the need for alternative storage in the United States began to grow when pools at many nuclear power plants began to fill up with stored spent fuel. As the proposed national permanent storage facilities (such as the Yucca Mountain nuclear waste repository) were not in operation, utility companies began looking at other options to store spent fuel. Dry cask storage was one of the most practical options for temporary storage. The 2008 NRC guideline calls for fuels to be submerged in a storage pool for at least five years before being moved to dry casks. It describes the dry casks used in the USA as “designed to resist floods, tornadoes, projectiles, temperature extremes, and other unusual scenarios.”
Spent fuel is currently stored in dry cask systems at a growing number of nuclear power plant sites. The NRC estimates that many of the nuclear power plants in the United States will be out of room in their spent fuel pools by 2015, most likely requiring the use of temporary storage of some kind. Yucca Mountain Nuclear Waste Repository was planned to open in 2017, but remains embroiled in controversy
Seven chapters dealing with nuclear waste storage facilities are given in Part Four of this book, from Chapter 25 to Chapter 31.
1 OECD/Nuclear Energy Agency (2012) Nuclear Energy Today Second Edition, http://www.oecd-nea.org/pub/nuclearenergytoday/6885-nuclear-energy-today.pdf
2 Hastings, P. S., Licensing and building new nuclear infrastructure. The Bridge, Expanding Frontiers of Engineering, 32(4), Winter 2002 issue, NAE publication. http://www.nae.edu/Publications/Bridge/ExpandingFrontiersofEngineering7308.aspx
3 Wikipedia, http://en.wikipedia.org/wiki/Containment_building
4 Wikipedia, http://en.wikipedia.org/wiki/Spent_fuel_pool
5 Wikipedia, http://en.wikipedia.org/wiki/Dry_cask_storage
