Index Manuals The United States of America Seventh National Report for the Convention on Nuclear Safety (2016)
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NUREG-1650
Revision 6
The United States of
America Seventh National
Report for the Convention
on Nuclear Safety
Manuscript Completed: August 2016
Date Published: October 2016
Prepared by
U.S. Nuclear Regulatory Commission (NRC)
Institute of Nuclear Power Operations (INPO)
Office of Nuclear Reactor Regulation
ABSTRACT
The U.S. Nuclear Regulatory Commission has prepared Revision 6 to NUREG-1650, “The
United States of America Seventh National Report for the Convention on Nuclear Safety,” for
submission for peer review at the seventh review meeting of the Convention on Nuclear Safety,
to be convened at the International Atomic Energy Agency in Vienna, Austria, in March 2017.
This report addresses the safety of land-based commercial nuclear power plants in the United
States. It demonstrates how the U.S. Government achieves and maintains a high level of
nuclear safety worldwide by enhancing national measures and international cooperation, and by
meeting the obligations of all the articles established by the Convention. These articles address
the safety of existing nuclear installations, the legislative and regulatory framework,
the regulatory body, responsibility of the licensee, the priority given to safety, financial and
human resources, human factors, quality assurance, assessment and verification of safety,
radiation protection, emergency preparedness, siting, design and construction, and operation.
This report also addresses the principles of the Vienna Declaration adopted by the Contracting
Parties in February 2015.
Similar to the U.S. National Report issued in 2013, this revised document includes a section
developed by the Institute of Nuclear Power Operations describing work that the U.S. nuclear
industry has done to ensure safety. The primary responsibility for the safety of a nuclear
installation rests with the license holder; therefore, Part 3 explains how the nuclear industry
maintains and improves nuclear safety.
iii
CONTENTS
ABSTRACT
iii
EXECUTIVE SUMMARY
xi
ACKNOWLEDGMENTS
xiii
ABBREVIATIONS
xv
PART 1
INTRODUCTION
3
Purpose and Structure of this Report
3
Changes to the Sixth U.S. National Report
4
SECTION 1. SUMMARY
13
1.1
The U.S. Policy Toward Nuclear Activities
13
1.1.1 Regulatory Body Organizational Values
13
1.1.2 Regulatory Body Challenges
14
1.2
National Nuclear Programs
15
1.2.1 Reactor Oversight Process
15
1.2.2 License Renewal
16
1.2.3 Power Uprates
17
1.2.4 New Reactor Licensing
18
1.3
Safety and Regulatory Issues, and Regulatory Accomplishments
19
1.3.1 Safety and Regulatory Issues Discussed in the Sixth U.S. National
Report
19
1.3.2 Current Safety and Regulatory Issues
40
1.3.3 Major Regulatory Accomplishments
50
1.4
International Peer Reviews and Missions
57
1.4.1 Convention on Nuclear Safety
57
1.4.2 Integrated Regulatory Review Service
60
1.4.3 Operational Safety Review Team
60
PART 2
ARTICLE 6. EXISTING NUCLEAR INSTALLATIONS
63
6.1
Introduction
63
6.2
Nuclear Installations in the United States
64
6.3
Regulatory Processes and Programs
65
6.3.1 Reactor Licensing
65
6.3.2 Reactor Oversight Process
66
6.3.3 Industry Trends Program
67
6.3.4 Accident Sequence Precursor Program
68
6.3.5 Operating Experience Program
70
6.3.6 Generic Issues Program
70
6.3.7 Rulemaking
71
6.3.8 Fire Regulation Program
72
6.3.9 Decommissioning
73
6.3.10 Reactor Safety Research Program
74
6.3.11 Special Programs for Public Participation
75
6.4
Fukushima Lessons Learned
77
6.5
Vienna Declaration on Nuclear Safety
77
ARTICLE 7. LEGISLATIVE AND REGULATORY FRAMEWORK
81
7.1
Legislative and Regulatory Framework
81
7.2
Provisions of the Legislative and Regulatory Framework
82
v
7.2.1 National Safety Requirements and Regulations
82
7.2.2 Licensing of Nuclear Installations
83
7.2.3 Inspection and Assessment
84
7.2.4 Enforcement
84
ARTICLE 8. REGULATORY BODY
87
8.1
The Regulatory Body
87
8.1.1 Mandate
87
8.1.2 Authority and Responsibilities
87
8.1.2.1 Scope of Authority
87
8.1.2.2 The NRC as an Independent Regulatory Agency
88
8.1.3 Structure of the Regulatory Body
88
8.1.3.1 The Commission
88
8.1.3.2 Component Offices of the Commission
89
8.1.3.3 Offices of the Executive Director for Operations
90
8.1.3.4 Advisory Committees
92
8.1.3.5 Atomic Safety and Licensing Board Panel
93
8.1.3.6 Office of the Inspector General
93
8.1.4 Position of the NRC in the Governmental Structure
93
8.1.4.1 Executive Branch
93
8.1.4.2 The States (i.e., of the United States)
95
8.1.4.3 Congress
96
8.1.5 International Responsibilities and Activities
96
8.1.5.1 International Standards
100
8.1.5.2 Integrated Regulatory Review Service Mission
101
8.1.5.3 Operational Safety Assessment Review Team
102
8.1.6 Financial and Human Resources
103
8.1.6.1 Financial Resources
103
8.1.6.2 Human Resources
103
8.1.7 Openness and Transparency
107
8.2
Separation of Functions of the Regulatory Body from Those of Bodies
Promoting Nuclear Energy
111
8.3
Ethic Rules Applying to NRC Employees and Former Employees
112
ARTICLE 9. RESPONSIBILITY OF THE LICENSE HOLDER
113
9.1
Introduction
113
9.2
The Licensees Prime Responsibility for Safety
113
9.3
NRC Enforcement Program
114
9.4
Openness and Transparency
117
ARTICLE 10. PRIORITY TO SAFETY
119
10.1 Background
119
10.2 Probabilistic Risk Assessment Policy
120
10.3 Applications of Probabilistic Risk Assessment
120
10.3.1 Risk-Informed Special Treatment
121
10.3.2 Risk-Informed Inservice Inspection
122
10.3.3 Risk-Informed Technical Specification Changes
123
10.3.4 Development of Standards
124
10.3.5 Level 3 Probabilistic Risk Assessment Project
124
10.4 Safety Culture
125
10.4.1 Safety Culture Policy Statement
125
10.4.2 NRC Monitoring of Licensee Safety Culture
127
10.4.2.1 Background
127
vi
10.4.2.2 Enhanced Reactor Oversight Process
127
10.4.3 The NRC Safety Culture
128
10.5 Managing the Safety and Security Interface
130
ARTICLE 11. FINANCIAL AND HUMAN RESOURCES
133
11.1 Financial Resources
133
11.1.1 Financial Qualifications for Construction and Operations
134
11.1.1.1 Construction Permit Reviews
134
11.1.1.2 Operating License Reviews
134
11.1.1.3 Combined License Application Reviews
135
11.1.1.4 Postoperating License Nontransfer Reviews
135
11.1.1.5 Reviews of License Transfers
135
11.1.2 Financial Qualifications Program for Decommissioning
136
11.1.3 Financial Protection Program for Liability Claims Arising from
Incidents
137
11.1.4 Insurance Program for Onsite Property Damages Arising from
Incidents
138
11.2 Regulatory Requirements for Qualifying, Training, and Retraining Personnel ..138
11.2.1 Governing Documents and Process
138
11.2.2 Experience
140
ARTICLE 12. HUMAN FACTORS
141
12.1 Goals and Mission of the Program
141
12.2 Program Elements
141
12.3 Significant Regulatory Activities
142
12.3.1 Human Factors Engineering
142
12.3.2 Emergency Operating Procedures and Plant Procedures
143
12.3.3 Shift Staffing
144
12.3.4 Fitness for Duty
145
12.3.5 Human Factors Information System
146
12.3.6 Support to Event Investigations and For-Cause Inspections and
Training
146
12.4
Fukushima Lessons Learned
148
ARTICLE 13. QUALITY ASSURANCE
149
13.1 Background
149
13.2 Regulatory Policy and Requirements
149
13.2.1 Appendix A to 10 CFR Part 50
150
13.2.2 Appendix B to 10 CFR Part 50
150
13.2.3 Approaches for Adopting More Widely Accepted International
Quality Standards
151
13.3 Quality Assurance Regulatory Guidance
151
13.3.1 Guidance for Staff Reviews for Licensing
151
13.3.2 Guidance for Design and Construction Activities
151
13.3.3 Guidance for Operational Activities
151
13.4 Quality Assurance Programs
152
13.5 Quality Assurance Audits Performed by Licensees
152
13.5.1 Audits of Vendors and Suppliers
153
13.6
Vendor Inspection Program
153
ARTICLE 14. ASSESSMENT AND VERIFICATION OF SAFETY
155
14.1 Ensuring Safety Assessments throughout Plant Life
155
14.1.1 Assessments of Safety
156
14.1.2 Maintaining the Licensing Basis
156
vii
14.1.2.1 Governing Documents and Process
156
14.1.3 Power Uprates
158
14.1.3.1 Governing Documents and Process
158
14.1.3.2 Experience
160
14.1.4 License Renewal
161
14.1.4.1 Governing Documents and Process
161
14.1.4.2 Experience
163
14.1.4.3 Operating Beyond 60 Years
163
14.1.5 The United States and Periodic Safety Reviews
165
14.1.5.1 The NRC’s Robust and Ongoing Regulatory Process
and the Current Licensing Basis
166
14.1.5.2 The Backfitting Process: Timely Imposition of New
Requirements
167
14.1.5.3 The NRC’s Extensive Experience with Broad-Based
Evaluations
168
14.1.5.4 License Renewal Confirms Safety of Plants
169
14.1.5.5 Risk-Informed Regulation and the Reactor Oversight
Process
170
14.1.5.6 Licensee Responsibilities for Safety: Regulations and
Initiatives Beyond Regulations
171
14.1.5.7 Summary
171
14.2 Verification by Analysis, Surveillance, Testing, and Inspection
173
14.3 Fukushima Lessons Learned
173
14.4 Vienna Declaration on Nuclear Safety
174
ARTICLE 15. RADIATION PROTECTION
177
15.1 Authorities and Principles
177
15.2 Regulatory Framework
179
15.3 Regulations
179
15.4 Radiation Protection Activities
181
15.4.1 Control of Radiation Exposure of Occupational Workers
181
15.4.2 Control of Radiation Exposure of Members of the Public
182
15.5 Fukushima Lessons Learned
183
ARTICLE 16. EMERGENCY PREPAREDNESS
185
16.1 Background
185
16.2 Offsite Emergency Planning and Preparedness
186
16.3 Emergency Classification System and Emergency Action Levels
187
16.4 Recommendations for Protective Action in Severe Accidents
189
16.5 Inspection Practices - Reactor Oversight Process for Emergency
Preparedness
190
16.6 Responding to an Emergency
192
16.6.1 Federal Response
192
16.6.2 Licensee, State, and Local Response
193
16.6.3 The NRC’s Response
194
16.6.4 Aspects of Security that Support Response
194
16.7 Communications with Neighboring States and International Arrangements
195
16.8 Communications with the Public
196
16.9 Fukushima Lessons Learned
197
ARTICLE 17. SITING
201
17.1 Background
201
17.2 Safety Elements of Siting
202
viii
17.2.1 Background
202
17.2.2 Assessments of Seismic and Geological Aspects of Siting
203
17.2.3 Assessments of Seismic and Geological Aspects of Siting
204
17.2.4 Assessments of Radiological Consequences from Postulated
Accidents
204
17.3 Environmental Protection Elements of Siting
206
17.3.1 Governing Documents and Process
206
17.3.2 Other Considerations for Environmental Reviews
207
17.4 Re-evaluation of Site-Related Factors
209
17.5 Consultation with other Contracting Parties to be Affected by the Installation ..209
17.6 Vienna Declaration on Nuclear Safety
209
ARTICLE 18. DESIGN AND CONSTRUCTION
211
18.1 Defense-in-Depth Philosophy
211
18.1.1 Governing Documents and Process
211
18.1.2 Experience
214
18.2 Technologies Proven by Experience or Qualified by Testing or Analysis
215
18.3 Design for Reliable, Stable, and Easily Manageable Operation
215
18.3.1 Governing Documents and Process
215
18.3.2 Experience
216
18.3.2.1 Human Factors Engineering
216
18.3.2.2 Digital Instrumentation and Controls
216
18.3.2.3 Cyber Security
218
18.4 New Reactor Construction Experience Program
219
18.5 Fukushima Lessons Learned
220
18.6 Vienna Declaration on Nuclear Safety
221
ARTICLE 19. OPERATION
223
19.1 Initial Authorization to Operate
224
19.2 Definition and Revision of Operational Limits and Conditions
226
19.3 Approved Procedures
226
19.4 Procedures for Responding to Anticipated Operational Occurrences and
Accidents
227
19.5 Availability of Engineering and Technical Support
227
19.6 Incident Reporting
228
19.7 Programs to Collect and Analyze Operating Experience
229
19.8 Radioactive Waste
230
19.9 Vienna Declaration on Nuclear Safety
231
PART 3
CONVENTION ON NUCLEAR SAFETY REPORT: THE ROLE OF THE
INSTITUTE OF NUCLEAR POWER OPERATIONS IN SUPPORTING THE UNITED
STATES COMMERCIAL NUCLEAR POWER INDUSTRY’S FOCUS ON
NUCLEAR SAFETY
235
Executive Summary
237
Organization and Governance
239
INPO’s Role within the Federal Regulatory Framework
243
Responsibilities of INPO and Its Members
245
Principles of Sharing (Openness and Transparency)
246
Priority to Safety (Safety Culture)
247
Operations, Activities and Actions
248
Evaluation Programs
248
Plant Evaluations
249
ix
Corporate Evaluations
251
Other Review Visits
252
Training and Accreditation Programs
255
Training and Qualification Guidelines
257
Courses and Seminars
258
Analysis and Information Exchange Programs
258
Events Analysis Program
259
Development of Documents and Products
259
Workshops and Meetings
261
Nuclear Network® System
261
Performance Data Collection and Trending
261
Equipment Performance Data
262
Operating Experience for New Plant Construction
262
Other Analysis Activities
262
Comprehensive Performance Monitoring Program
262
Special Focus Program
263
Assistance Programs
264
New Plant Development
264
Relationship with World Association of Nuclear Operators
265
Industry Response to the Accident in Fukushima
266
Conclusion
275
APPENDIX A NRC STRATEGIC PLAN 2014-2018
A-1
APPENDIX B NRC MAJOR MANAGEMENT CHALLENGES FOR THE FUTURE
B-1
APPENDIX C REFERENCES
C-1
APPENDIX D U.S. COMMERCIAL NUCLEAR POWER REACTORS
D-1
x
EXECUTIVE SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) has prepared Revision 6 to NUREG-1650,
“The United States of America Seventh National Report for the Convention on Nuclear Safety,”
for submission for peer review at the seventh review meeting of the Convention on Nuclear
Safety, to be convened at the International Atomic Energy Agency in Vienna, Austria, in March
2017. This report addresses the safety of land-based commercial nuclear power plants in the
United States. It demonstrates how the U.S. Government achieves and maintains a high-level of
nuclear safety worldwide by enhancing national measures and international cooperation and by
meeting the obligations of all the articles established by the Convention. These articles address
the safety of existing nuclear installations, the legislative and regulatory framework,
the regulatory body, responsibility of the licensee, the priority given to safety, financial and
human resources, human factors, quality assurance, assessment and verification of safety,
radiation protection, emergency preparedness, siting, design and construction, and operation.
This report addresses the issues identified through the peer review conducted during the sixth
review meeting in April 2014 and discusses challenges and issues that have arisen since that
time. The sixth review meeting identified the following six U.S. challenges:
(1)
Fukushima-related activities
(2)
risk-informed fire protection regulations requiring extensive resources for the transition to
National Fire Protection Association 805
(3)
ensuring continuity during the oversight transition from plant construction to operation
(4)
nuclear industry strategy
(5)
reporting status of the periodic safety review gap being developed
(6)
status of the NRC’s work on the issuance of possible license renewals to operate
beyond 60 years (i.e., subsequent license renewal)
This report discusses the status of safety issues raised in the sixth U.S. National Report,
including implementation of Fukushima lessons, nondestructive examinations, concrete
degradation, cumulative effects of regulations, economic consequences, counterfeit and
fraudulent items, and construction inspection. The report also addresses the following safety
and regulatory issues that have necessitated significant attention since 2013:
(1)
baffle-former bolts
(2)
digital instrumentation and control systems
(3)
open phase conditions in electric power systems
(4)
risk-informing regulations and processes
(5)
spent fuel pool neutron-absorbing materials
(6)
staff readiness to transition plants from construction to operations
(7)
staff readiness to transition plants from operation to decommissioning
(8)
subsequent license renewal
(9)
implementation of the principles of the Vienna Declaration on Nuclear Safety
(10)
Project Aim
The Institute of Nuclear Power Operations has also provided input to this report. The primary
responsibility for the safety of a nuclear installation rests with the license holder; therefore,
Part 3 explains how the nuclear industry maintains and improves nuclear safety, including
challenges and the implementation of lessons learned from the Fukushima accident.
xi
ACKNOWLEDGMENTS
Contributors to this report include the following technical and regulatory experts from the U.S.
Nuclear Regulatory Commission:
Akstulewicz, Frank
Flanders, Scott
Koenick, Stephen
Algama, Don
Foggie, Kirk
Kohen, Marshall
Allwein, Russell
Folk, Kevin
Kolb, Tim
Anderson, James
Fong, CJ
Kozak, Tom
Anderson, Joseph
Frahm, Ron
Kratchman, Jessica
Armstrong, Kenneth
Franovich, Rani
Kugler, Andrew
Arndt, Steven
Freeman, Eric
Kulesa, Gloria
Bales, Michelle
Frumkin, Dan
Kuntz, Robert
Barczy, Theresa
Gallucci, Ray
Kuritzky, Alan
Basu, Sudhamay
Gardocki, Stan
Lane, John
Bauer, Laurel
Garry, Steve
Lappert, Glenna
Beltz, Terry
Gaslevic, Jim
Laur, Steve
Benowitz, Howard
Gavrilas, Mirela
Lee, Eric
Billings, Sally
Gendelman, Adam
Lee, Richard
Billoch, Araceli
Gennardo, David
Lehman, Bryce
Biro, Mihaela
Gibson, Lauren
Lewis, Doris
Bloom, Steve
Goetz, Robert
Lian, Jocelyn
Bowman, Gregory
Gordon, Dennis
Lising, Jason
Boyce, Tom
Grant, Jeffrey
Lois, Kosmas
Buford, Angela
Green, Kimberly
Lukes, Robert
Burke, John
Hardies, Bob
Lupold, Tim
Campbell, Steve
Hart, Michelle
Lyons, Sara
Campbell, Tison
Hayes, Michelle
Lyons-Burke, Kathy
Carte, Norbet
Held, Wesley
Mahlahla, LaTonya
Cauffman, Chris
Henderson, Karen
Martin, Kamishan
Chen, Yen-Ming
Hill, Brittan
Mathew, Roy
Clark, Brooke
Hiser, Allen
McCartin, Timothy
Clark, Jefferson
Hollcraft, Zachary
McIntyre, David
Clark, Sheldon
Hopkins, Jon
McKirgan, John
Coyne, Kevin
Huffert, Antony
Merzke, Daniel
Cruz, Holly
Huffman, Bill
Miller, Barry
Cruz, Zahira
Imboden, Stacey
Miller, Chris
Culp, Lisa-Anne
Iyengar, Raj
Miller, Keith
Cumblidge, Stephen
Jackson, Chris
Milligan, Patricia
Cushing, Jack
Jackson, Deborah
Mizuno, Geary
Decker, David
Jarriel, Lisa
Monninger, John
Dhir, Neha
Johnson, Don
Montecalvo, Michael
Dickson, Elijah
Kahler, Robert
Mossman, Timothy
Dinsmore, Steve
Karwoski, Kenneth
Moyer, Carol
Dougherty, Jay
Keefe, Molly
Muller, David
Doyle, Dan
Keim, Andrea
Murray, Charles
Ennis, Rick
Kichline, Michelle
Nakoski, John
Ferkile, Andrea
Kim, Grace
Nourbakhsh, Hossein
Ferrell, Kimberly
Klein, Alex
Oesterle, Eric
xiii
Olmstead, Joan
Rosenberg, Stacey
Temple, Jeffrey
Orders, William
Ross-Lee, Mary-Jane
Thompson, Catherine
Orr, Mark
Rossi, Mathew
Trocine, Leigh
Quinones, Lauren
Rothschild, Trip
Vanden Berghe, John
Palmrose, Donald
Ruffin, Steve
Vechioli, Lucieann
Pantalo, Charity
Sanders, Serita
Watson, Bruce
Pasquale, Dan
Schofer, Fred
Waugh, Andrew
Patel, Ami
Schrader, Eric
Weber, Carl
Pedersen, Renee
Scott, Cathy
Weerakkody, Sunil
Pedersen, Roger
Seshagiri Rao, Tammara
Westreich, Barry
Peduzzi, Francis
Shoop, Undine
Whited, Jeffrey
Peralta, Juan
Sieracki, Diane
Whitman, Jennifer
Portillo, Veronica
Sigmon, Rebecca
Wilkins, Lynnea
Prescott, Paul
Smith, Shawn
Williams, Donna
Prescott, Peter
Smith, Will
Williams, Kevin
Purnell, Blake
Syndor, Russell
Wong, Albert
Rautzen, William
Tadesse, Rebecca
Wong, Emma
Reed, Timothy
Tallarico, Alison
Wood, Kent
Rini, Brett
Taylor, Robert
Wray, John
Rivera, Alison
Thomas, George
Yip, Brian
Robinson, Jay
Thompson, Catherine
Yoder, Matt
Rodriguez, Veronica
Truong, Tung
Rosales-Cooper, Cindy
Telson, Ross
Contributors to this report include the following experts from the Institute of Nuclear Power
Operations:
Aitken, Steve
Hembree, Dave
Masters, Glen
Berko, Dave
Johnson, Steve
Nichols, Steve
Cox, Tammy
Klodnicki, Mike
Place, Jeff
Crane, Randy
Koves, Ken
Simon, Shawn
Crawley, Carmen
Love, Tammy
Spinnato, Roger
Gambrill, Bob
Martinez, Ramon
Weaver, Claire
xiv
ABBREVIATIONS
ABWR
advanced boiling-water reactor
ADAMS
Agencywide Documents Access and Management System (NRC)
ALARA
as low as reasonably achievable
ANS
American Nuclear Society
ANSI
American National Standards Institute
AP
advanced passive
APR
advanced power reactor
ASME
American Society of Mechanical Engineers
BSAF
Benchmark Study of the Accident at the Fukushima Dai-ichi
BRIIE
baseline risk index for initiating events
BWR
boiling-water reactor
CEO
chief executive officer
CFR
Code of Federal Regulations
CFSI
Counterfeit, Fraudulent, Suspect Items
CNS
Convention on Nuclear Safety
ConE
construction experience program
DG
draft regulatory guide
DHS
U.S. Department of Homeland Security
DOE
U.S. Department of Energy
EPA
U.S. Environmental Protection Agency
EPRI
Electric Power Research Institute
ERDA
Energy Research and Development Administration
ESBWR
economic simplified boiling-water reactor
FEMA
Federal Emergency Management Agency
FLEX
diverse and flexible coping strategies
FR
Federal Register
FY
fiscal year
GL
generic letter
IAEA
International Atomic Energy Agency
ICES
INPO Consolidated Event System
ICRP
International Commission on Radiological Protection
IEEE
Institute of Electrical and Electronics Engineers
IER
INPO event reports
IN
information notice
INPO
Institute of Nuclear Power Operations
IP
inspection procedure
IPSR
INPO performance summary report
IRRS
Integrated Regulatory Review Service
ISAP
integrated safety assessment program
ISG
interim staff guidance
xv
ITAAC
inspection(s), test(s), analysis (analyses), and acceptance criterion/criteria
LOCA
loss of coolant accident
LOOP
loss of offsite power
MD
management directive
MWe
megawatt electric
MWt
megawatt thermal
NANTeL
National Academy for Nuclear Training e-Learning
NATF
North American Transmission Forum
NEA
Nuclear Energy Agency
NEI
Nuclear Energy Institute
NEIL
Nuclear Electric Insurance Limited
NEPA
National Environmental Policy Act
NIMS
National Incident Management System
NRC
U.S. Nuclear Regulatory Commission
NTTF
Near-Term Task Force
OMB
Office of Management and Budget
OSART
Operational Safety Assessment Review Team
PML
performance monitoring leaders
POC
performance objectives and criteria
PRA
probabilistic risk assessment
PWR
pressurized-water reactor
RFI
request for information
RG
regulatory guide
RIS
regulatory issue summary
SAMGs
severe accident management guidelines
SAREF
SAfety REsearch Opportunities post-Fukushima
SAT
systems approach to training
SBO
station blackout
SEP
Systematic Evaluation Program
SFP
spent fuel pool
SOER
Significant Operating Experience Report
SSC
structure, system, and component
TI
temporary instruction
U.S.
United States
US APWR
U.S. Advanced Pressurized-Water Reactor
US EPR
U.S. Evolutionary Power Reactor
WANO
World Association of Nuclear Operators
xvi
PART 1
INTRODUCTION
This section describes the purpose and structure of the “United States of America Seventh
National Report for the Convention on Nuclear Safety,” and provides a summary of changes in
the seventh U.S. National Report.
Purpose and Structure of This Report
The United States of America is submitting this updated report for peer review to the seventh
review meeting of the Contracting Parties to the Convention on Nuclear Safety (hereafter
referred to as the Convention, or CNS). The scope of this report considers only the safety of
land-based commercial nuclear power plants, consistent with the definition of nuclear
installations provided in Article 2 and the scope of Article 3 of the Convention.
This report demonstrates how the U.S. Government meets the following objectives described in
Article 1 of the Convention:
(i)
to achieve and maintain a high level of nuclear safety worldwide through the
enhancement of national measures and international cooperation including,
where appropriate, safety-related technical cooperation
(ii)
to establish and maintain effective defenses in nuclear installations against potential
radiological hazards to protect individuals, society, and the environment from harmful
effects of ionizing radiation from such installations
(iii)
to prevent accidents with radiological consequences and to mitigate such consequences
should they occur
Technical and regulatory experts from the U.S. Nuclear Regulatory Commission (hereafter
referred to as the NRC, Commission,1 agency, or staff) updated the seventh U.S. National
Report, principally using agency information that is publicly available. This updated report
follows the format of the sixth U.S. National Report published in 2013 and is designed to be a
standalone document. Therefore, this report duplicates some of the information presented in the
2013 report. To facilitate peer review, Table 1 includes a summary of the main changes to the
report. This table in Part 1 is followed by a high level summary of the report, consistent with the
guidance of the Convention.
Part 2 discusses the Convention’s Articles 6 through 19. Chapters are numbered according to
the article of the Convention under consideration. Each chapter begins with the text of the
article, followed by an overview of the material covered and a discussion of how the
United States meets the obligations described in the article. Articles 6 through 9 summarize the
existing nuclear installations and the legislative and regulatory system governing their safety
and discuss the adequacy and effectiveness of that system. Articles 10 through 16 address
general safety considerations and summarize major safety-related features. Articles 17 through
19 address the safety of installations.
1 Commission may also refer to the Chairman and Commissioners who head the NRC.
3
Similar to the 2013 report, Part 3 of this document includes a contribution by the Institute of
Nuclear Power Operations (INPO) describing work that the U.S. nuclear industry has done to
ensure safety. INPO is a nongovernmental corporation founded in 1979 by the U.S. nuclear
industry to collectively promote the highest levels of safety and reliability at U.S. nuclear plants.
The primary responsibility for the safety of a nuclear installation rests with the license holder;
therefore, Part 3 explains how the nuclear industry maintains and improves nuclear safety.
The report concludes with a series of appendices that discuss the NRC’s main challenges, the
agency’s Strategic Plan, references, and a list of nuclear plants in the United States.
This report does not explicitly discuss Articles 1 through 5 because the general text of the
report, and indeed the very existence of the report, fulfills the requirements of these articles. In
accordance with Article 1, the report illustrates how the U.S. Government meets the objectives
of the Convention. The report discusses the safety of nuclear installations according to the
definition in Article 2 and the scope of Article 3. It addresses implementing measures (such as
national laws, legislation, regulations, and administrative means) according to Article 4.
Submission of this report fulfills the obligation under Article 5 on reporting. In addition, the
information in this report is available in more detail on the NRC’s public Web site.
Changes to the Sixth U.S. National Report
To facilitate peer review of this report, Table 1 summarizes the changes to the sixth U.S.
National Report. A revision bar along the left margin of the page identifies changes from the
sixth report.
Table 1 Summary of Changes to the Sixth U.S. National Report
Report Section
Change
Abstract
Updated to add discussion about the sixth CNS
and the Vienna Declaration on Nuclear Safety
Executive Summary
Updated to add discussion about the sixth CNS
and the Vienna Declaration on Nuclear Safety
PART 1
Introduction
Updated to add discussion about the sixth CNS
Purpose and Structure of This Report
Updated to add discussion about the sixth CNS
Summary of Changes to the Sixth U.S. National
Updated table
Report
Section 1
SUMMARY
Updated to add discussion about the sixth CNS
1.1
The U.S. Policy toward Nuclear
No changes
Activities
1.1.1
Regulatory Body Organizational Values
No changes
1.1.2
Regulatory Body Challenges
Updated to add discussion on most recent NRC
Strategic Plan and Inspector General report
1.2
National Nuclear Programs
Editorial changes only
1.2.1
Reactor Oversight Process
Updated to add discussion about the
self-assessment
1.2.2
License Renewal
Updated to add discussion about units entering
the 41st year of operation
1.2.3
Power Uprate Program
No changes
4
Report Section
Change
1.2.4
New Reactor Licensing
Summarized. Updated to add discussion about
applications under review.
1.3
Safety and Regulatory Issues, and
Editorial changes only
Regulatory Accomplishments
1.3.1
Safety and Regulatory Issues
Completely updated to add current status
Discussed in the Sixth U.S. National
Report
1.3.2
Current Safety and Regulatory Issues
Completely updated. Discusses 7 new topics.
1.3.3
Major Regulatory Accomplishments
Completely updated. Discusses 6 new topics.
1.4
International Peer Reviews and
Editorial changes only
Missions
1.4.1
Convention on Nuclear Safety
Updated to include results from sixth CNS report
and Rapporteurs’ findings. A new subsection on
the Vienna Declaration on Nuclear Safety has
been added.
1.4.2
Integrated Regulatory Review Service
Updated to include reference to mission results
1.4.3
Operational Safety Review Team
Updated to include reference to mission results
PART 2
Article 6
EXISTING NUCLEAR INSTALLATIONS
Updated to state that the article addresses the
Vienna Declaration on Nuclear Safety
6.1
Introduction
Updated to add performance goals
6.2
Nuclear Installations in the United
Updated to include status of plants in operation
States
and shutdowns
6.3
Regulatory Processes and Programs
No change
6.3.1
Reactor Licensing
Updated to include information on combined
licenses issued
6.3.2
Reactor Oversight Process
Updated to discuss current plant performance
status
6.3.3
Industry Trends Program
Updated to discuss 2014 trends results
6.3.4
Accident Sequence Precursor Program
Updated to include a discussion about the
accident sequence precursor program status
report issued in 2015
6.3.5
Operating Experience Program
Editorial changes only
6.3.6
Generic Issues Program
Updated to reflect changes to streamline the
program
6.3.7
Rulemaking
No changes
6.3.8
Fire Protection Regulation Program
Updated to discuss challenges and updates on
the transition to risk-informed rule
6.3.9
Decommissioning
Updated to discuss new decommissioning
rulemaking
6.3.10
Reactor Safety Research Program
No changes
6.3.11
Public Participation
Updated to refine discussion on rulemaking
process
6.4
Lessons Learned from Fukushima
Editorial changes only
6.5
Vienna Declaration on Nuclear Safety
New section
Article 7
LEGISLATIVE AND REGULATORY
Updated to state that no changes to U.S.
FRAMEWORK
legislative framework were made post-
Fukushima
7.1
Legislative and Regulatory Framework
Updated to add discussion on the Convention on
the Physical Protection on Nuclear Material
7.2
Provisions of the Legislative and
No changes
Regulatory Framework
5
Report Section
Change
7.2.1
National Safety Requirements and
Updated to refine discussion on rulemaking
Regulations
process
7.2.2
Licensing of Nuclear Installations
Editorial changes only
7.2.3
Inspection and Assessment
Editorial changes only
7.2.4
Enforcement
Editorial changes only
7.3
Lessons Learned from Fukushima
Section deleted. No changes to U.S. legislative
framework post-Fukushima.
Article 8
REGULATORY BODY
Updated to state that regulatory actions post-
Fukushima are discussed in Section 1
8.1
The Regulatory Body
No changes
8.1.1
Mandate
No changes
8.1.2
Authority and Responsibilities
No changes
8.1.2.1
Scope of Authority
No changes
8.1.2.2
The NRC as an Independent
Editorial changes only
Regulatory Agency
8.1.3
Structure of the Regulatory Body
No changes
8.1.3.1
The Commission
Editorial changes only
8.1.3.2
Component Offices of the
Editorial changes only
Commission
8.1.3.3
Offices of the Executive Director for
Updated to reflect organizational changes
Operations
8.1.3.4
Advisory Committees
No changes
8.1.3.5
Atomic Safety and Licensing Board
No changes
Panel
8.1.3.6
Office of the Inspector General
Editorial changes only
8.1.4
Position of the NRC in the
No changes
Governmental Structure
8.1.4.1
Executive Branch
Updated to discuss role of the National Security
Council
8.1.4.2
The States (i.e., of the United States)
Editorial changes only
8.1.4.3
Congress
No changes
8.1.5
International Responsibilities and
Updated throughout
Activities
8.1.5.1
International Standards
Updated to include committee representation and
efforts to harmonize with NRC guidance
8.1.5.2
Integrated Regulatory Review Service
Updated to add discussion on findings and
Mission
followup mission results
8.1.5.3
Operational Safety Assessment
Updated to add discussion on findings and
Review Teams
upcoming mission
8.1.6
Financial and Human Resources
No changes
8.1.6.1
Financial Resources
Updated to add funds for fiscal year 2015
8.1.6.2
Human Resources
Updated to discuss survey findings and
knowledge management initiatives
8.1.7
Openness and Transparency
Updated to include most recent numbers
associated to public outreach activities
8.2
Separation of Functions of the
Updated to refine discussion on means by which
Regulatory Body from Those of Bodies
effective separation and independence is
Promoting Nuclear Energy
ensured
8.3
Fukushima Lessons Learned
Deleted section. Regulatory actions post-
Fukushima are discussed in Section 1.
8.3
Ethics Rules Applying to NRC
New section. Replaces the original 8.3 section.
Employees and Former Employees
6
Report Section
Change
Article 9
RESPONSIBILITY OF THE LICENSE
Updated to state that overall responsibility of the
HOLDER
license holder did not change post-Fukushima
9.1
Introduction
No changes
9.2
The Licensee’s Primary Responsibility
Updated to refine discussion on means to ensure
for Safety
licensee has resources for managing an accident
9.3
The NRC Enforcement Program
Updated to discuss recent enforcement actions
9.4
Openness and Transparency
Editorial changes only
9.5
Fukushima Lessons Learned
Deleted section. Overall responsibility of the
license holder did not change post-Fukushima.
Article 10
PRIORITY TO SAFETY
Updated to state that no changes to policies on
priority to safety have been made as a result of
Fukushima.
10.1
Background
Updated throughout. New references.
10.2
Probabilistic Risk Assessment Policy
No changes
10.3
Applications of Probabilistic Risk
Editorial changes only
Assessment
10.3.1
Risk-Informed Special Treatment
Updated to discuss update on pilot application
10.3.2
Risk-Informed Inservice Inspection
Updated to provide new references
10.3.3
Risk-Informed Technical Specification
Updated initiatives
Changes
10.3.4
Development of Standards
Updated to discuss updates to standard
10.3.5
Level 3 Probabilistic Risk Assessment
New section
Project
10.4
Safety Culture
Restructured to include the policy statement
discussion in Section 10.4.1
10.4.1
Safety Culture Policy Statement
Renamed. Updated to discuss new safety culture
traits.
10.4.2
The NRC Monitoring of Licensee Safety
Renumbered
Culture
10.4.2.1
Background
Renumbered. No wording changes.
10.4.2.2
Enhanced Reactor Oversight Process
Renumbered. Updated to discuss issuance of
revised procedures.
10.4.3
The NRC Safety Culture
Renumbered. Updated throughout. Discusses
safety culture components.
10.5
Managing the Safety and Security
Updated to add new references
Interface
10.6
Fukushima Lessons Learned
Section deleted. No changes to policies on
priority to safety as a result of Fukushima.
Article 11
FINANCIAL AND HUMAN
Updated to state that no changes in financial
RESOURCES
resources measures have taken place post-
Fukushima
11.1
Financial Resources
Updated to discuss financial qualification
regulations
11.1.1
Financial Qualifications for Construction
Minor editorial change in the section name
and Operations
11.1.1.1
Construction Permit Reviews
Editorial changes only
11.1.1.2
Operating License Reviews
Editorial changes only
11.1.1.3
Combined License Application Reviews
Updated to refine discussion on application
requirements
11.1.1.4
Postoperating License Nontransfer
Updated to discuss issuance of interim staff
Reviews
guidance
7
Report Section
Change
11.1.1.5
Reviews of License Transfers
Updated to discuss financial qualification
regulations
11.1.2
Financial Qualifications Program for
Updated to discuss decommissioning funding
Decommissioning
regulations
11.1.3
Financial Protection Program for
Minor editorial change in the section name.
Liability Claims Arising from Incidents
Updated the retrospective premium pool
requirements
11.1.4
Insurance Program for Onsite Property
Editorial changes only
Damages Arising from Accidents
11.2
Regulatory Requirements for Qualifying,
No changes
Training, and Retraining Personnel
11.2.1
Governing Documents and Process
Updated references
11.2.2
Experience
No changes
11.3
Fukushima Lessons Learned
Section deleted. No changes to licensee financial
requirements have taken place post-Fukushima.
Article 12
HUMAN FACTORS
Editorial changes only
12.1
Goals and Mission of the Program
No changes
12.2
Program Elements
No changes
12.3
Significant Regulatory Activities
Editorial changes only
12.3.1
Human Factors Engineering
Approved extended power uprates and editorial
changes. Minor change in section title.
12.3.2
Emergency Operating Procedures and
Updated status on Fukushima activities and
Plant Procedures
information on mitigating strategies rule.
12.3.3
Shift Staffing
Updated to include discussion on staffing plan
validation
12.3.4
Fitness for Duty
Updated reference on fatigue and fitness-for-duty
12.3.5
Human Factors Information System
Update on database
12.3.6
Support to Event Investigations and For-
Updated to add discussion on recent inspections
Cause Inspections and Training
12.4
Fukushima Lessons Learned
Update on status of Fukushima activities
Article 13
QUALITY ASSURANCE
Updated to state that regulatory actions post-
Fukushima are discussed in Section 1
13.1
Background
No changes
13.2
Regulatory Policy and Requirements
Editorial changes only
13.2.1
Appendix A to 10 CFR Part 50
No changes
13.2.2
Appendix B to 10 CFR Part 50
No changes
13.2.3
Approaches for Adopting More Widely
No changes
Accepted International Quality
Standards
13.3
Quality Assurance Regulatory Guidance
No changes
13.3.1
Guidance for Staff Reviews for
Reference update
Licensing
13.3.2
Guidance for Design and Construction
No changes
Activities
13.3.3
Guidance for Operational Activities
Reference update
13.4
Quality Assurance Programs
Reference update
13.5
Quality Assurance Audits Performed by
No changes
Licensees
13.5.1
Audits of Vendors and Suppliers
Editorial changes only
13.6
Fukushima Lessons Learned
Section deleted. Fukushima actions are
discussed in Section 1.
13.6
Vendor Inspection Program
New section. Replaces the original 13.6 section.
8
Report Section
Change
Article 14
ASSESSMENT AND VERIFICATION
Editorial changes only
OF SAFETY
14.1
Ensuring Safety Assessments
Editorial changes only
throughout Plant Life
14.1.1
Assessment of Safety
Updated to include information on plant licensing
basis.
14.1.2
Maintaining the Licensing Basis
Editorial changes only
14.1.2.1
Governing Documents and Process
No changes
14.1.1.2
Regulatory Framework for the Restart of
No changes
Browns Ferry, Unit 1
14.1.3
Power Uprates
No changes
14.1.3.1
Governing Documents and Process
Editorial changes only
14.1.3.2
Experience
Updated information on power uprates to date,
including details on Peach Bottom and Monticello
14.1.4
License Renewal
No changes
14.1.4.1
Governing Documents and Process
Updated discussion of the continued storage
rule, and added new references and information
on new guidance documents
14.1.4.2
Experience
Updated discussion about renewed license to
date
14.1.4.3
Operating Beyond 60 Years
Completely updated
14.1.5
The United States and Periodic Safety
Editorial changes only
Reviews
14.1.5.1
The NRC’s Robust and Ongoing
Editorial changes only
Regulatory Process and the Current
Licensing Basis
14.1.5.2
The Backfitting Process: Timely
Updated information on protections similar to the
Imposition of New Requirements
backfitting rule
14.1.5.3
The NRC’s Extensive Experience with
Editorial changes only
Broad-Based Evaluations
14.1.5.4
License Renewal Confirms Safety of
No changes
Plants
14.1.5.5
Risk-Informed Regulation and the
Updated to describe a risk informed approach
Reactor Oversight Process
14.1.5.6
Licensee Responsibilities for Safety:
Editorial changes only
Regulations and Initiatives Beyond
Regulations
14.1.5.7
The NRC’s Regulatory Process
Updated to discuss actions to address some of
Compared with International Safety
the Integrated Regulatory Review Service
Reviews
findings
14.2
Verification by Analysis, Surveillance,
Editorial changes only
Testing, and Inspection
14.3
Fukushima Lessons Learned
Update on Fukushima activities
14.4
Vienna Declaration on Nuclear Safety
New section
Article 15
RADIATION PROTECTION
No changes
15.1
Authorities and Principles
Updated and new references provided, including
information on a proposed rulemaking
15.2
Regulatory Framework
Editorial changes only
15.3
Regulations
Updated status of regulatory activities and
references
15.4
Radiation Protection Activities
No changes
9
Report Section
Change
15.4.1
Control of Radiation Exposure of
Updated collective doses
Occupational Workers
15.4.2
Control of Radiation Exposure of
Deleted outdated information on ground water
Members of the Public
contamination
15.5
Fukushima Lessons Learned
Update on Fukushima activities
Article 16
EMERGENCY PREPAREDNESS
No changes
16.1
Background
No changes
16.2
Offsite Emergency Planning and
Editorial changes only
Preparedness
16.3
Emergency Classification System and
Updated on emergency action level guidelines.
Emergency Action Levels
16.4
Recommendations for Protective Action
No changes
in Severe Accidents
16.5
Inspection Practices - Reactor Oversight
No changes
Process for Emergency Preparedness
16.6
Responding to an Emergency
Editorial changes only
16.6.1
Federal Response
Update on information on Department of
Homeland Security and National Response
Framework
16.6.2
Licensee, State, and Local Response
No changes
16.6.3
The NRC’s Response
No changes
16.6.4
Aspects of Security that Support
Rewrite of actions taken after September 2011
Response
16.7
Communications with Neighboring
Updated information on agreements and added
States and International Arrangements
discussion on observation of a U.S. exercise
16.8
Communications with the Public
No changes
16.9
Fukushima Lessons Learned
Updated status of Fukushima activities
Article 17
SITING
Updated to state that no changes on siting
regulatory activities have taken place post-
Fukushima. Addresses CNS consultancy
meeting templates for Articles 17 and 18.
17.1
Background
Deleted short paragraph with duplicate
information
17.2
Safety Elements of Siting
No changes
17.2.1
Background
No changes
17.2.2
Assessments of Nonseismic Aspects of
Updated references
Siting
17.2.3
Assessments of Seismic and Geological
No changes
Aspects of Siting
17.2.4
Assessments of Radiological
Added updated references and editorial changes
Consequences from Postulated
Accidents
17.3
Environmental Protection Elements of
Editorial changes only
Siting
17.3.1
Governing Documents and Process
Added updated references
17.3.2
Other Considerations for Siting Reviews
Added discussion on rulemaking and other
relevant regulatory developments. Minor change
in section title.
17.4
Reevaluation of Site-Related Factors
Editorial changes only
17.5
Consultation with other Contracting
Editorial changes only
Parties To Be Affected by the
Installation
10
Report Section
Change
17.6
Fukushima Lessons Learned
Section deleted. No changes on siting regulatory
activities have taken place post-Fukushima.
17.6
Vienna Declaration on Nuclear Safety
New section. Replaces the original 17.6.
Article 18
DESIGN AND CONSTRUCTION
Addresses CNS consultancy meeting templates
for Articles 17 and 18
18.1
Defense-in-Depth Philosophy
No changes
18.1.1
Governing Documents and Process
Updated references throughout the section
18.1.2
Experience
Updates on status of Watts Bar licensing
activities
18.1.2.1
Regulatory Framework for the
No changes
Reactivation of Watts Bar, Unit 2
18.1.2.2
Design Certifications
No changes
18.2
Technologies Proven by Experience or
No changes
Qualified by Testing or Analysis
18.3
Design for Reliable, Stable, and Easily
No changes
Manageable Operation
18.3.1
Governing Documents and Process
No changes
18.3.2
Experience
References to new reactors, small modular
designs, and Multinational Design Evaluation
Program working groups updated
18.3.2.1
Human Factors Engineering
Cyber security rule update
18.3.2.2
Digital Instrumentation and Controls
Discussion of communication of national
operating experience events and information
exchanges with international counterparts
18.3.2.3
Cyber Security
Updated to discuss status of Fukushima activities
18.4
New Reactor Construction Experience
Discussion of communication of national
Program
operating experience events and information
exchanges with international counterparts
18.5
Fukushima Lessons Learned
Updated to discuss status of Fukushima activities
18.6
Vienna Declaration on Nuclear Safety
New section
Article 19
OPERATION
Updated to state that regulatory actions
post-Fukushima are discussed in Section 1
19.1
Initial Authorization to Operate
Update on new reactor licensing activities
19.2
Definition and Revision of Operational
Editorial changes only
Limits and Conditions
19.3
Approved Procedures
Editorial changes only
19.4
Procedures for Responding to
Updated to add discussion on mitigating
Anticipated Operational Occurrences
strategies rulemaking
and Accidents
19.5
Availability of Engineering and
No changes
Technical Support
19.6
Incident Reporting
New references provided
19.7
Programs To Collect and Analyze
Updated to add examples of operating
Operating Experience
experience communications
19.8
Radioactive Waste
Updated to add discussion about the U.S. Court
of Appeals order on Yucca Mountain
19.9
Fukushima Lessons Learned
Section deleted. Fukushima actions are
discussed in Section 1.
19.9
Vienna Declaration on Nuclear Safety
New section. Replaces the original 19.9
11
PART 3
Convention on Nuclear Safety Report: The Role of
Updated
the Institute of Nuclear Power Operations in
Supporting the U.S. Commercial Nuclear Power
Industry’s Focus on Nuclear Safety
APPENDICES AND ANNEXES
APPENDIX A NRC STRATEGIC PLAN
Updated to add most recent and updated
Strategic Plan
APPENDIX B NRC MAJOR MANAGEMENT
Updated to add most recent report from the
CHALLENGES FOR THE FUTURE
Inspector General
APPENDIX C U.S. SUPPORT OF THE
Deleted. Incorporated into Fukushima writeup
INTERNATIONAL ATOMIC ENERGY AGENCY
under Section 1.3.1.
ACTION PLAN ON NUCLEAR SAFETY
APPENDIX C REFERENCES
Renumbered. Updated.
APPENDIX D U.S. COMMERCIAL NUCLEAR
Renumbered. Updated. Power uprates and new
POWER REACTORS
licenses issued.
ANNEX 2 U.S. NUCLEAR ELECTRIC INDUSTRY
Deleted
PERFORMANCE INDICATOR GRAPHS
12
SECTION 1. SUMMARY
The Summary in the National Report should highlight the Contracting Party’s continued
efforts in achieving the Convention’s objectives. It should serve as a major information
source by summarizing updated information on matters that have developed since the
previous National Report, focusing discussion on significant changes in national laws,
regulations, administrative arrangements, and practices related to nuclear safety, and
demonstrating followup from one Review Meeting to the next.
This section provides a high level summary of U.S. policy toward safety; the regulatory body’s
organizational values, including transparency; and its challenges. It summarizes the national
nuclear programs; provides an update on important safety and regulatory issues identified in the
previous National Report; and addresses those safety and regulatory issues and regulatory
accomplishments that have arisen since the last National Report was issued (see
NUREG-1650, “The United States of America Sixth National Report for the Convention on
Nuclear Safety,” Revision 5, issued in August 2013). Lastly, this section summarizes the results
of international peer reviews and missions.
1.1 The U.S. Policy toward Nuclear Activities
The Energy Reorganization Act of 1974 created the U.S. NRC as an independent agency of the
Federal Government. The agency’s mission is to license and regulate the Nation’s civilian use of
byproduct, source, and special nuclear materials to ensure adequate protection of public health
and safety, promote the common defense and security, and protect the environment. In
addition, the agency’s export licensing and domestic safeguards programs are integral to the
U.S. Government’s commitment to nuclear nonproliferation. The NRC’s safety and security
responsibilities stem from the Atomic Energy Act of 1954, as amended. The agency
accomplishes its mission by licensing and overseeing nuclear reactor operations and other
activities that apply to the possession of nuclear materials and wastes, ensuring that nuclear
materials and facilities are safeguarded from theft and radiological sabotage, issuing rules and
standards, inspecting nuclear facilities, and enforcing regulations.
1.1.1 Regulatory Body Organizational Values
In conducting its work, the NRC adheres to seven organizational values to guide its actions:
integrity, service, openness, commitment, cooperation, excellence, and respect. The NRC’s
Principles of Good Regulation help carry out NRC regulatory activities. These principles focus
on ensuring safety and security while appropriately balancing the interests of stakeholders,
including the public and licensees. These principles are independence, efficiency, clarity,
reliability, and openness. The NRC’s final decisions are based on objective, technical
assessments of all information, and are documented with reasons explicitly stated. As a learning
organization, the NRC establishes ways to evaluate and continually upgrade its regulatory
capabilities. Its regulations are coherent, logical, practical, and based on the best available
knowledge from research and operational experience.
The NRC also views nuclear regulation as a service to the public and, as such, it must be
transacted openly. The NRC is committed to being a trusted, independent, transparent, and
effective regulator. The NRC’s Open Government Plan, first published April 7, 2010, is a
reflection of the agency’s long history of, and commitment to, openness with the public and
13
transparency in the regulatory process. The agency’s goal to ensure openness explicitly
recognizes that the public must be informed about, and have a reasonable opportunity
to participate meaningfully in, the regulatory process. Except for proprietary information,
security-related information, predecisional information, and information supplied by foreign
governments that is deemed to be sensitive, the NRC makes the documentation that it uses in
its decisionmaking process available in the agency’s Public Document Room in Rockville, MD,
and on the agency’s public Web site at http://www.nrc.gov. Over the past several years, the
NRC also has embraced social media as an important new tool for reaching a wider public
audience. As a result, a significant amount of information about nuclear activities and the
national policy regarding them is available to everyone.
1.1.2 Regulatory Body Challenges
The NRC identified major challenges for the future in NUREG-1614, Volume 6, “Strategic Plan:
Fiscal Years 2014-2018,” dated August 2014. External factors may cause changes to the
regulatory environment. To adapt to these changes, the NRC must use its resources efficiently,
revise the regulatory framework as appropriate to disposition existing or emerging issues, and
provide adequate infrastructure to maintain staff competence and readiness. Some current and
expected future challenges include:
x
continued implementation of enhancements to nuclear safety based on insights arising
from operating experience reviews and lessons learned from the 2011 nuclear accident
at the Fukushima Dai-ichi nuclear facility in Japan
x
continual learning and adaptation of the regulatory framework, as necessary, to address
knowledge of and response to the specific hazards, uncertainties, and risks associated
with each nuclear site
x
continued readiness to review applications involving new technologies such as small
modular reactors, medical isotope production facilities, and rapidly evolving digital
instrumentation and control systems
x
changes in the demographics, experience, and knowledge of the workforce
x
continued awareness of and support to the development of nuclear safety and security
regulations around the world
x
changing economic conditions in the energy market affecting current and planned
applications to construct and operate new nuclear facilities or licensee decisions to
decommission existing ones
x
globalization of nuclear technology and the nuclear supply chain, driving the need for
increased international engagement on the safe and secure use of radioactive material
and the need for new oversight approaches, including ensuring that foreign components
used in U.S. nuclear facilities are in compliance with NRC requirements
x
continuous monitoring of the threat environment to ensure the security of nuclear
facilities and radioactive materials
14
As stated in the Strategic Plan, the following key external factors could affect the agency’s
ability to achieve its strategic goals:
x
market pressures on operating plants and license applications
x
a significant operating incident (domestic or international)
x
globalization of the nuclear technology and the nuclear supply chain
x
a significant terrorist incident
x
legislative and executive branch initiatives
x
international nuclear standards developments
x
international treaties and conventions
x
lost, misplaced, intercepted, or delayed information
By law, the Inspector General of each Federal agency (as discussed in Article 8) identifies the
agency’s most serious management and performance challenges and assesses progress in
addressing them. The NRC’s Inspector General’s annual assessment of the major management
challenges confronting the agency appear on the NRC’s public Web site. These challenges
represent what the Inspector General considers to be inherent and continuing program
challenges relative to maintaining effective and efficient oversight and internal controls. As a
result, it is likely they will continue to be challenges from year to year. Challenges do not
necessarily equate to problems. The 2015 assessment report described the main challenges in
the following areas of NRC’s work, discussed in more detail in Appendix B to this report.
x
regulation of nuclear reactor safety programs
x
regulation of nuclear materials and radioactive waste programs
x
management of security over internal infrastructure (personnel, physical, and cyber
security) and nuclear security
x
management of information technology and information management
x
management of financial programs
x
management of administrative functions
1.2 National Nuclear Programs
The NRC has several programs and processes to protect public health and safety and the
environment and to meet the obligations of the Convention on Nuclear Safety (CNS). Key
programs in the reactor arena comprise a well-established regulatory process, which includes:
(1) reactor oversight, (2) license renewal, (3) power uprates, and (4) new reactor licensing.
1.2.1 Reactor Oversight Process
The regulatory framework for NRC’s Reactor Oversight Process consist of three strategic
performance areas: reactor safety, radiation safety, and safeguards. Within each strategic
performance area are cornerstones that reflect the essential safety aspects of facility operation.
These seven cornerstones include: initiating events, mitigating systems, barrier integrity,
emergency preparedness, public radiation safety, and physical security. Satisfactory licensee
15
performance in the cornerstones provide reasonable assurance of safe facility operation and
that the NRC’s safety mission is being accomplished. Each cornerstone contains performance
indicators to ensure that their objectives are being met.
The results of the annual self-assessments and other independent or focused evaluations have
stated that the risk-informed performance-based Reactor Oversight Process has remained
transparent and showed that the Reactor Oversight Process has effectively and openly
supported the agency’s mission and strategic goals of safety and security.
Inspection reports, including the results of emergency exercise evaluations, are on the NRC
public Web site at http://www.nrc.gov/NRR/OVERSIGHT/ASSESS/listofrpts_body.html. Article 6
of this report discusses the Reactor Oversight Process in detail.
1.2.2 License Renewal
The NRC’s review of license renewal applications focuses on maintaining plant safety and
specifically considers the effects of aging on important structures, systems, and components.
The review of a renewal application proceeds along two paths—one to review safety issues and
the other to assess potential environmental impacts. Applicants must demonstrate that they
have identified and can manage the effects of aging and can continue to maintain an acceptable
level of safety throughout the period of extended operation. Applicants must also address the
environmental impacts from extended operation. The Commission has seen sustained, strong
interest in license renewal, which allows plants to operate up to 20 years beyond their current
operating licenses. The Atomic Energy Act established the original 40-year term, a timeframe
based on economic and antitrust considerations, rather than the technical limitations of the
nuclear facility.
The decision to seek license renewal is voluntary and rests entirely with nuclear power plant
owners. The decision typically is based on the plant’s economic viability and whether it can
continue to meet the Commission’s requirements. Currently, more than three quarters of the
plants in the United States have had their operating licenses renewed. Based on statements
from industry representatives, the Commission expects nearly all sites to apply for license
renewal. As reported in the sixth U.S. National Report, 18 units entered their 41st year of
operation (the period of extended operation) between 2011 and 2013.2 By the end of 2016,
20 additional units will have entered the period of extended operation as listed below. In addition
to these plants, Indian Point Nuclear Generating Station, Units 2 and 3, are operating beyond 40
years under the timely renewal provision.3
2 In October 2015, Pilgrim Unit 1, which entered the period of extended operation in 2012, announced that the
plant will cease operations by mid-2019. In April 2016, Entergy announced that Pilgrim will be permanently
shutdown on May 2019.
3 Visit http://www.nrc.gov/info-finder/reactors/ip/ip-timely-renewal.html for more information on Indian Point’s timely
renewal.
16
Year 2014
Year 2015
Year 2016
x Calvert Cliffs Nuclear Power,
x Millstone Power
x Beaver Valley Power Station,
Unit 1
Station, Unit 2
Unit 1
x Oconee Nuclear Station, Unit 3
x St. Lucie Nuclear Power Plant,
x Arkansas Nuclear One, Unit 1
Unit 1
x Edwin I. Hatch Nuclear Plant,
x Browns Ferry Nuclear Plant,
Unit 1
Unit 3
x Peach Bottom Nuclear Plant,
x Calvert Cliffs Nuclear Power
Unit 3
Plant, Unit 2
x Donald C. Cook Atomic Power
x Salem Nuclear Generating
Station, Unit 1
Station, Unit 1
x Browns Ferry Nuclear Plant,
x Brunswick Steam Electric Plant,
Unit 2
Unit 1
x Brunswick Steam Electric
Plant, Unit 2
x James A. FitzPatrick Nuclear
Power Plant4
x Three Mile Island Nuclear
Station, Unit 1
x Cooper Nuclear Station
x Duane Arnold Energy Center
x Prairie Island Nuclear
Generating Plant, Unit 2
Section 1.3.2 of this report provides a discussion on subsequent license renewal (i.e., renewal
beyond 60 years). Section 6.2 of this report provides additional discussion on the Pilgrim
Nuclear Power Station, Unit 1, and James A. FitzPatrick Nuclear Power Plant license status.
Article 14 of this report discusses the license renewal process in detail, including a discussion of
the update to the Generic Environmental Impact Statement for license renewal.
1.2.3 Power Uprates
Under its licensing program, the NRC carefully reviews requests to raise the maximum thermal
power level at which a plant may be operated. In reviewing these power uprate requests, the
NRC’s review focuses on safety. The agency closely monitors operating experience to identify
safety issues that may affect the implementation of power uprates.
Power uprates can be classified as:
(1) measurement uncertainty recapture power uprates,
(2) stretch power uprates, and (3) extended power uprates. Measurement uncertainty recapture
power uprates are less than a 2 percent increase in power and are achieved by implementing
higher precision feedwater flow measurement devices to more accurately calculate reactor
power. Stretch power uprates have increased power up to 7 percent and are generally within
the original design capacity of the plant. Stretch power uprates usually involve changes to
instrumentation setpoints and generally do not involve major plant modifications. Extended
4 FitzPatrick entered the period of extended operation in October 2014. However, in March 2016, Entergy certified
that FitzPatrick would cease operations in January 2017.
17
power uprates usually increase power more than 7 percent and require significant modifications
to major balance-of-plant equipment. The NRC has approved extended power uprates of up to
20 percent.
Article 14 of this report discusses the power uprate process in detail.
1.2.4 New Reactor Licensing
The NRC’s new reactor program focuses on licensing reviews for small and large light-water
reactors and advanced nonlight-water reactors; oversight and construction inspection activities;
preapplication and readiness reviews for current and future reactor licensing; and infrastructure
development to support oversight and reactor licensing. The NRC is in the process of
completing ongoing licensing reviews; supporting construction activities associated with four
new reactor units in the United States licensed under Title 10 of the Code of Federal
Regulations (10 CFR) Part 52, “Licenses, Certifications, and Approvals for Nuclear Power
Plants”; and is making enhancements to increase the efficiency and predictability of small
light-water and advanced reactor reviews. The NRC’s new reactor program is also actively
engaged in several international cooperative activities to promote enhanced safety in new
reactor designs, strengthen reactor siting reviews, and improve the effectiveness and efficiency
of inspections and the collection and sharing of construction experience.
The NRC staff is interacting with vendors and utilities on new reactor applications and licensing
activities. The NRC staff is actively reviewing four combined license applications for a total of
seven new nuclear plants, two design certification applications, and one design certification
renewal application. All combined license applicants are using the licensing process specified
in 10 CFR Part 52. This licensing process resolves all safety and environmental issues, as well
as emergency preparedness and security issues, before a new nuclear power plant is
constructed.
In addition to working on domestic issues for new reactor construction, the NRC has been a
leader in cooperating with other national nuclear regulatory authorities to address reactor
licensing activities. The NRC is a founding member of, and fully participates in, the Multinational
Design Evaluation Program, a unique international forum with members from the regulatory
authorities of Canada, China, Finland, France, Hungary, India, Japan, the Republic of Korea,
the Russian Federation, South Africa, Sweden, the United Arab Emirates, the United Kingdom,
and the United States. The Nuclear Energy Agency (NEA) from the Organisation for Economic
Co-operation and Development performs the technical secretariat duties for the Multinational
Design Evaluation Program.
The activities of the Multinational Design Evaluation Program include:
(1) cooperation on
specific safety design reviews of Westinghouse Electric Company’s Advanced Passive
(AP) 1000, Korea Electric Power Corporation and Korea Hydro and Nuclear Power Co., Ltd.’s
Advanced Power Reactor 1400 (APR1400), General Electric Nuclear Energy’s Advanced
Boiling-Water Reactor, and AREVA Nuclear Power’s U.S. Evolutionary Power Reactor
(US EPR), and (2) exploration of opportunities to harmonize and converge on regulatory
practices in the areas of safety goals, safety classification, digital instrumentation and controls,
mechanical codes and standards, and vendor inspection cooperation.
18
The Multinational Design Evaluation Program interacts with various representatives from the
industry, including vendors and operators, standards development organizations, and the World
Nuclear Association.
Articles 17 and 18 of this report discuss new reactor licensing in more detail. Sections 1.3.3, 6.2,
and 18.1.2 of this report discuss licensing of Watts Bar, Unit 2, in more detail.
1.3 Safety and Regulatory Issues, and Regulatory Accomplishments
This section provides an update on important safety and regulatory issues identified in the sixth
U.S. National Report and addresses those safety and regulatory issues and regulatory
accomplishments that have necessitated significant attention since the last National Report was
issued.
1.3.1 Safety and Regulatory Issues Discussed in the Sixth U.S. National Report
In the sixth U.S. National Report, the NRC staff reported to be working with the safety and
regulatory issues listed in this section. An update on the following items is provided:
x
concrete structural issues
x
construction inspection program lessons learned
x
counterfeit, fraudulent, and suspect items
x
cumulative effects of regulation
x
evaluation of economic consequences
x
Fukushima lessons learned
x
nondestructive evaluations
x
steam generator integrity
Concrete Structural Issues
Since 2009, several significant conditions adverse to quality have occurred or were discovered
in safety-related concrete structures of operating reactors in the United States. These conditions
involve the following:
x
shield building laminar cracking at Davis Besse
x
alkali-silica reaction concrete degradation at Seabrook
Each of the above issues was or is being addressed by the respective licensee under its
Corrective Action Program. A brief description of each of these issues is provided below:
Shield Building Laminar Cracking at Davis-Besse
The shield building at the Davis-Besse nuclear power plant is a reinforced-concrete structure
that surrounds the freestanding steel containment vessel, and has nominal wall thickness of
30 inches with vertical and horizontal rebar grids on both the inside and outside face. The
functions of the shield building are to provide:
(1) biological shielding, (2) environmental
protection of the containment vessel, and (3) control release of annulus atmosphere during
accidents.
19
During the October 2011 mid-cycle outage, while cutting a construction opening to replace the
reactor vessel closure head, laminar cracking was identified in the shield building cylindrical
wall. The licensee performed impulse response mapping, core bores, and in-depth calculations
to determine the operability of the shield building. The cracks were very tight (i.e., hairline
cracks) and the rebar and concrete were generally found to be in good condition. The licensee
determined that the reason for the concrete laminar cracking was a combination of the design
specification for construction of the shield building, which did not require application of an
exterior sealant from moisture, and environmental conditions associated with the blizzard of
1978.
The NRC issued “Confirmatory Action Letter - Davis-Besse Nuclear Power Station,” on
December 2, 2011, detailing planned actions that the licensee had to take to provide continued
long-term confidence of the shield building’s ability to maintain its safety functions. The licensee
performed structural evaluations to capture bounding conditions and took corrective actions
including (1) applying a sealant to the surface of the shield building to address the cause of the
cracking and prevent new cracks and (2) development of a shield building monitoring program.
In the summer of 2013, while performing additional testing according to the shield building
monitoring program, the licensee identified new crack indications. The licensee determined that
some of these cracks were previously missed because of limitations of the boroscope used for
earlier examinations of core bores and others were considered evidence of possible growth of
the existing cracks. The licensee performed more testing and analysis that confirmed crack
growth and issued an apparent cause evaluation that attributed the crack growth to ice wedging.
The licensee modified the shield building monitoring program in response to these new findings.
The NRC has completed an inspection of the licensee’s apparent cause evaluation for the crack
growth. The NRC staff determined that the shield building laminar cracking condition remained
bounded by the licensee’s structural evaluation. The NRC has concluded that the licensee
provided reasonable assurance that, with the current condition, the shield building will perform
its safety function, including withstanding earthquakes and tornadoes. The NRC staff continues
to follow the licensee’s corrective actions related to this issue under the baseline inspection
program.
Alkali-Silica Reaction Concrete Degradation at Seabrook Station
Alkali-silica reaction is a slow chemical process that can cause degradation over time in
hardened concrete. For this reaction to occur, it is necessary for the concrete to contain reactive
aggregate, high alkali content in the cement, and adequate moisture to form a gel. The gel
expands by absorbing water initially, resulting in a network of microcracks in the concrete.
Depending on its progression and severity, the alkali-silica reaction can reduce or affect
mechanical properties of concrete (i.e., compressive, tensile, shear, and bond strengths, elastic
modulus, and the Poisson’s ratio) used in design to different extents, and could also affect
empirical code relationships between concrete mechanical properties assumed in the American
Concrete Institute design and construction codes. Alkali-silica reaction expansion could also
lead to structural displacement or deformation and discrete macrocracking not considered in the
concrete design, and could affect structural performance over time.
In August 2010, during an assessment for the license renewal application by the Seabrook
Station, the licensee identified concrete degradation due to alkali-silica reaction in below-grade
walls of several safety-related structures with ground water intrusion. Seabrook is the first U.S.
20
commercial nuclear power plant where this type of degradation has been identified. The
licensee’s root cause analysis determined that, along with other causal factors, the alkali-silica
reaction developed in Seabrook’s concrete primarily because the concrete mix used a
susceptible aggregate that was slow-reacting. The potential reactivity of this aggregate was
undetected by the testing specified by the applicable American Society for Testing and Materials
construction standards at the time of construction in the late 1970s. Since that time, the role of
slow-reacting aggregate in alkali-silica reaction has been identified in the construction industry
and improved standard tests are now available to better identify slow reactive aggregates before
use.
Seabrook engineers have continued detailed testing, walkdowns, crack monitoring, and
evaluations to address and manage the issue comprehensively in the short- and long-term. On
May 16, 2012, the NRC staff issued a letter to the licensee to confirm commitments to address
this issue. The letter focuses on assuring operability of the affected structures pending review of
a formal root cause analysis and short- and long-term monitoring actions while plant-specific
alkali-silica reaction research and development continues. The research and development
results will be used, in part, to address long-term effects on structural performance and
management of the issue, and to provide a technical basis for resolution of the operability
determination and for identifying corrective actions, if required.
The NRC staff reviewing Seabrook’s license renewal application is focusing on the discovery of
this concrete degradation because the aging effects of alkali-silica reaction on the affected
structures may be different in character or magnitude after the term of the current operating
license. The licensee needs to demonstrate that the aging effects during the period of extended
operation will be adequately managed. The NRC is currently evaluating the proposed plant-
specific, first-of-a-kind, alkali-silica reaction aging management program.
The NRC staff’s plant oversight reviews are focused on ensuring that the alkali-silica reaction
issue at Seabrook is comprehensively addressed and managed such that there is reasonable
assurance that the affected structures will continue to perform their intended safety functions
through the expected service life. The staff has performed detailed inspections to verify and
assess the adequacy of the licensee’s interim operability basis and actions and commitments to
address the impact on reinforced concrete structures at Seabrook. The NRC, through followup
inspections, verified the adequacy of planned actions related to the alkali-silica reaction
structures monitoring program, and the large-scale testing to reconcile this issue with the design
and licensing basis. The large-scale testing has been completed and the licensee is currently
compiling and reviewing the results of the testing program in preparation of a license
amendment request submittal. An initial determination of the test program is that out-of-plane
(through-thickness) expansion is also an important parameter for monitoring the progression of
alkali-silica reaction. The licensee is in the process of installing extensometers at select plant
locations to measure future out-of-plane expansion.
The NRC also has engaged external stakeholders and members of the public through public
meetings and written communications under the reactor oversight and license renewal
processes. On November 18, 2011, the NRC issued Information Notice (IN) 2011-20, “Concrete
Degradation by Alkali-Silica Reaction,” to inform licensees of the occurrence of alkali-silica
reaction-induced concrete degradation of safety-related structures at Seabrook.
21
The NRC’s oversight review of this issue determined that there are no immediate safety
concerns based on existing safety margins, the slow nature of the degradation, and ongoing
monitoring. This review has included an evaluation of the licensee’s prompt operability
determinations for various structures affected by alkali-silica reaction. These operability
determinations address the alkali-silica reaction impacts on material properties due to
microcracking, as well as the impacts due to macrocracking and building deformation due to
alkali-silica reaction expansion. The NRC’s oversight includes ongoing assessment of the
continued acceptability of the operability determinations. The most recent results of the NRC
staff’s review are documented in “Seabrook Station, Unit No. 1 - Integrated Inspection Report
05000443/2015004 and Independent Spent Fuel Storage Installation Report No.
07200063/2015001,” dated February 12, 2016. The NRC continues its oversight of the
alkali-silica reaction issue and has formed a multioffice, multidiscipline working group to guide
the agency’s ongoing approach to respond to this safety issue.
Construction Inspection Program Lessons Learned
To provide regulatory oversight of the construction of four AP1000 units, NRC implemented a
construction inspection program, including development of governing documents and
procedures. The NRC’s Region II Office in Atlanta, Georgia, has the primary responsibility for
implementing the construction inspection program. Region II has as many as five resident
construction inspectors at the construction sites during the preoperational phase of construction
to oversee the day-to-day activities of the licensee and its contractors, and supplements this
inspection staff with additional personnel from Region II, other regional offices, and
headquarters technical staff, as needed, to ensure that the as-built facility conforms to the
conditions of the license.
The NRC conducts vendor inspections to ensure that products and services furnished to U.S.
reactors meet established regulatory requirements for quality and other safety factors.
The NRC began to fully implement its construction inspection program with the issuance of the
licenses for Vogtle, Units 3 and 4, in February 2012. The program was expanded to include
V.C. Summer, Units 2 and 3, when their licenses were issued in March 2012.
The NRC evaluated inspection results to identify lessons learned that could be used as
feedback to improve the construction inspection program, to focus future inspection activities,
and to inform licensees of needed improvement areas. Over the course of the first several years
of the full implementation of the program, the following lessons learned were identified:
x
Design and configuration control —Licensees must align with designers, suppliers, and
constructors to achieve effective design control, configuration and change management
and comply with 10 CFR Part 52 when making changes to the certified design.
x
Supplier oversight —Licensees must effectively oversee all contractors, subcontractors,
and vendors to ensure that they are aware of and meeting regulatory and inspections,
tests, analyses, and acceptance criteria requirements.
x
Digital instrumentation and control —Licensees must focus on digital instrumentation
and control systems to ensure compliance with licensing commitments and address
design verification and validation issues early on.
22
x
Corrective Actions —Implementing an effective corrective action program at construction
sites presents unique challenges to licensees that have a large number of activities
occurring across many organizations (licensee, contractors, vendors, etc.).
x
Licensee’s ultimate responsibility —The licensee holds the ultimate responsibility to
meet its obligations under its license and must demonstrate that it is a competent and
capable operator. Engineering, procurement, and construction contracts and their
implementation must preserve these principles.
Counterfeit, Fraudulent, and Suspect Items
The integrity of the supply chain is a fundamental element of an effective quality assurance
program for the NRC’s licensed facilities and their associated suppliers. Although there is no
evidence of significant counterfeit activity impacting U.S. nuclear facilities, the NRC emphasizes
the importance of robust quality assurance programs in protecting against counterfeit,
fraudulent, and suspect items.
Over the past 3 decades, the NRC has published multiple documents to inform stakeholders of
counterfeit or misrepresented products and services. On March 21, 1989, the NRC issued
Generic Letter (GL) 89-02, “Actions to Improve the Detection of Counterfeit and Fraudulently
Marketed Products,” to inform licensees of effective program elements for detecting counterfeit
or fraudulently marketed products and for assuring the quality of vendor-supplied products.
More recently, the NRC has published several other documents related to counterfeit,
fraudulent, and suspect items:
x
IN 2008-04, “Counterfeit Parts Supplied to Nuclear Power Plants,” dated April 7, 2008,
informs addressees of the potential for counterfeit parts to enter their supply chains.
x
IN 2012-22, “Counterfeit, Fraudulent, Suspect Items (CFSI) Training Offerings,” dated
January 25, 2013, provides a list of training resources that can be used for educating
personnel involved in NRC-regulated activities on current trends in CFSI and techniques
to prevent the use of CFSI.
x
IN 2013-15, “Willful Misconduct/Record Falsification and Nuclear Safety Culture,” dated
August 23, 2013, describes a vendor’s criminal actions to destroy serial numbers in an
attempt to conceal a component’s origin before it was installed in a U.S. nuclear plant.
x
Regulatory Issue Summary (RIS) 2015-08, “Oversight of Counterfeit, Fraudulent, and
Suspect Items in the Nuclear Industry,” dated June 24, 2015, heightens awareness of
the existing NRC regulations and how they apply to counterfeit, fraudulent, and suspect
items within the scope of the NRC’s regulatory jurisdiction.
x
Regulatory Guide (RG) 5.71, “Cyber Security Programs for Nuclear Facilities,” dated
January 2010, addressed the procurement of digital assets including supply chain
security.
23
Licensees and industry organizations have also focused on the challenges presented by
counterfeit, fraudulent, and suspect items. In a joint effort with the Nuclear Energy Institute, the
Electric Power Research Institute (EPRI) developed EPRI-1019163, Revision 1, “Plant Support
Engineering: Counterfeit and Fraudulent Items,” in July 2014. The guidance document is
intended for use by licensees to aid in preventing the introduction of counterfeit, fraudulent, and
suspect items into nuclear facilities.
The NRC has also been involved in other U.S. Government activities related to addressing
counterfeit, fraudulent, and suspect items. The NRC participates in the U.S. Department of
Homeland Security’s National Intellectual Property Rights Coordination Center, which leverages
the combined resources of partner agencies to better combat intellectual property theft and to
dismantle the criminal organizations that seek to profit from the manufacturing, importation and
sale of counterfeit items.
To consolidate information and references for its activities related to counterfeit, fraudulent, and
suspect items, the NRC developed an extensive public Web site, which can be accessed at
http://www.nrc.gov/about-nrc/cfsi.html. The information shared on the Web site includes
historical agency documents, presentations from various public meetings, and ongoing NRC
activities to address counterfeit, fraudulent, and suspect items.
Cumulative Effects of Regulation
In 2009, the NRC began addressing ways to mitigate the cumulative effects of regulation in
response to Commission direction. Since then, the staff prepared four Commission
papers: SECY-11-0032, “Consideration of the Cumulative Effects of Regulation in the
Rulemaking Process,” dated March 2, 2011, SECY-12-0137, “Implementation of the Cumulative
Effects of Regulation Process Changes,” dated October 5, 2012, COMSECY-14-0014,
“Cumulative Effects of Regulation and Risk Prioritization Initiative: Update on Recent Activities
and Recommendations for Path Forward,” dated April 9, 2014, and SECY-15-0050, “Cumulative
Effects of Regulation Process Enhancements and Risk Prioritization Initiative,” dated
April 1, 2015.
The Commission approved the rulemaking enhancements proposed by the staff in
SECY-11-0032, which included providing increased stakeholder interactions, publishing
supporting guidance concurrent with rules, requesting specific comment on cumulative effects of
regulation process improvements in proposed rules, and developing informed implementation
timeframes.
In response to SECY-12-0137, the Commission directed the following:
x
Any expansion of the consideration of the cumulative effects of regulation should be
considered in the broader context of actions directed from
COMGEA-12-0001/COMWDM-12-0002, “Proposed Initiative To Improve Nuclear Safety
and Regulatory Efficiency.”
x
The staff should continue to develop and implement outreach tools that will allow NRC to
consider more completely the overall impacts of multiple rules, orders, generic
communications, advisories, and other regulatory actions on licensees and their ability to
focus effectively on items of greatest safety import.
24
x
The staff should engage industry to seek volunteer facilities to perform “case studies” to
review the accuracy of cost and schedule estimates used in the NRC’s regulatory
analysis.
In COMSECY-14-0014, the staff provided a summary of the results of case studies to review the
accuracy of cost and schedule estimates used in the agency’s regulatory analyses. In addition,
the staff proposed to merge the cumulative effects of regulation and risk prioritization initiative
deliverables, which was approved by the Commission. In SECY-15-0050, the staff provided
options to the Commission for incorporating risk insights into the decisionmaking process to
prioritize regulatory activities for operating reactors. In response to SECY-15-0050, the
Commission continued to support ongoing cumulative effects of regulation process
enhancements, including consideration of risk insights in regulatory decisionmaking through
existing agency processes.
In addition, the NRC is applying the process enhancements principles, to the extent practicable,
to the post-Fukushima regulatory actions. For instance, the NRC has engaged in significant
public interaction (through public meetings, comment periods, etc.) during the development of
these actions, and is providing implementation guidance when necessary.
Evaluation of Economic Consequences
The NRC’s regulatory framework accounts for the offsite economic consequences associated
with unintended releases of radionuclides with subsequent land contamination. Specifically,
offsite property damage is considered during the evaluation of cost-justified substantial safety
enhancements (i.e., backfit analysis), as well as in regulatory and environmental analyses. The
NRC uses similar guidance documents to conduct the cost-benefit determinations of these
analyses. In performing these economic analyses, the NRC has traditionally considered two
categories of property, onsite and offsite. Generally, onsite property is owned or controlled by
the license- or certificate-holder and located within the boundaries of the licensed facility,
whereas offsite property is located outside of the site boundaries, and is not owned or controlled
by the license- or certificate-holder. However, in these economic analyses, the distinction
between offsite and onsite property goes beyond the location or ownership of the property.
Onsite property costs include replacement power, decontamination costs, and costs associated
with refurbishment or decommissioning. Offsite property costs include both the direct costs
associated with property damage (e.g., diminution of property values) and indirect costs
(e.g., tourism, manufacturing, and agriculture disruption).
In response to SECY-12-0110, “Consideration of Economic Consequences Within the U.S.
Nuclear Regulatory Commission’s Regulatory Framework,” the Commission directed the NRC
staff to enhance the currency and consistency of the existing framework through updates to
guidance documents integral to performing cost-benefit analyses. The Commission also found
that economic consequences should not be treated as equivalent in regulatory character to
matters of adequate protection of public health and safety. Therefore, the Commission does not
plan to discuss economic consequences in future CNS reports.
Fukushima Lessons Learned
After the accident at Fukushima, the NRC took prompt action to ensure that there were no
immediate safety concerns at U.S. facilities and to verify nuclear power plant operators’
preparedness to respond to and mitigate the consequences of beyond-design-basis events.
25
These actions included issuance of IN 2011-05, “Tohoku-Taiheiyou-Oki Earthquake Effects on
Japanese Nuclear Power Plants,” dated March 18, 2011, to provide information about the
Fukushima accident to U.S. licensees. The NRC also issued two inspection procedures (IPs),
called temporary instructions (TIs), to NRC inspection staff to evaluate specific aspects of
licensee preparedness to respond to an event like that which occurred at the Fukushima facility
(TI 2515/183, “Follow up to the Fukushima Daiichi Nuclear Station Fuel Damage Event,” dated
March 23, 2011, and TI 2515/184, “Availability and Readiness Inspection of Severe Accident
Management Guidelines (SAMGs),” dated April 29, 2011). Finally, the NRC issued
Bulletin 2011-01, “Mitigating Strategies,” dated May 11, 2011, to request information from U.S.
licensees regarding their preparations for dealing with such an event.
On March 23, 2011, the Commission approved formation of the Near-Term Task Force (NTTF),
comprised of senior NRC staff and management, to systematically and methodically review the
NRC’s processes and regulations in light of the Fukushima accident. The Commission tasked
the NTTF with determining whether the NRC should make additional improvements to its
regulatory system, and to make policy recommendations to the Commission. The NTTF issued
its report, titled “Recommendations for Enhancing Reactor Safety in the 21st Century: The
Near-Term Task Force Review of the Insights from the Fukushima Daiichi Accident,” on
July 12, 2011. The NTTF concluded that continued operation of U.S. nuclear plants and ongoing
NRC licensing activities posed no imminent risk to public health and safety. The NTTF also
concluded that enhancements to safety and emergency preparedness were warranted, and
made 12 overarching recommendations for Commission consideration. In addition, the NTTF
concluded that nuclear power plant facilities and designs that were under NRC review, or
licensed, since 2011, also needed to demonstrate their ability to cope with hazards that are
beyond the design basis. The staff proposed a prioritization of the NTTF’s recommendations.
The Commission approved the staff’s proposal in SRM-SECY-11-0137, “Prioritization of
Recommended Actions to be Taken in Response to Fukushima Lessons Learned,” dated
December 15, 2011, and stated its support for staff action on the near-term recommendations.
The NRC formed the Japan Lessons-Learned Project Directorate (currently called the Japan
Lessons Learned Division) to perform a longer-term review of the March 11, 2011, Japanese
earthquake and tsunami and lead the implementation of the associated safety enhancements.
This organization reports to a Steering Committee of senior NRC officials, which is chaired by
the Deputy Executive Director for Reactor and Preparedness Programs and it is comprised of
Office Directors from many of the NRC program offices and regional offices. As a first step in its
assessment of lessons-learned from the accident, the staff considered whether any of the NTTF
recommendations identified an imminent hazard to public health and safety. The staff ultimately
agreed with the NTTF’s conclusion that the accident did not reveal any imminent risk to public
health and safety. The Japan Lessons-Learned Project Directorate then prioritized the twelve
NTTF recommendations by tiers and expanded upon the task force recommendations to include
proposals from the international community, the U.S. Congress, the NRC’s Advisory Committee
on Reactor Safeguards, and other stakeholders.
26
Tier 1 Recommendations
The first tier consists of actions that the NRC determined should be started without unnecessary
delay and for which sufficient resource flexibility, including availability of critical skill sets, exists.
The Tier 1 recommendations consist of the following:
x
seismic and flood hazard reevaluations (Recommendation 2.1)
x
seismic and flood walkdowns (Recommendation 2.3)
x
station blackout (SBO) regulatory actions (Recommendation 4.1)
x
mitigating strategies for beyond-design-basis events (Recommendation 4.2)
x
reliable hardened vents for Mark I and Mark II containments (Recommendation 5.1)
x
spent fuel pool (SFP) instrumentation (Recommendation 7.1)
x
strengthening and integration of emergency operating procedures, SAMGs, and
extensive damage mitigation guidelines (Recommendation 8)
x
emergency preparedness regulatory actions (staffing and communications)
(Recommendation 9.3)
Tier 2 Recommendations
The second tier recommendations are actions that originally could not be initiated because of a
need for further technical assessment and alignment, dependence on Tier 1 issues, or lack of
availability of critical skill sets. These were recommendations that the staff determined did not
require long-term study and could be initiated when sufficient technical information and
applicable resources become available. The Tier 2 recommendations include the following:
x
SFP makeup capability (Recommendations 7.2, 7.3, 7.4, and 7.5)
x
emergency preparedness actions (Recommendation 9.3)
x
reevaluation of external hazards other than seismic and flooding (e.g., tornados,
hurricanes, and drought) (additional issue)
As a result of further assessment after these items were prioritized as Tier 2 recommendations,
the NRC determined that SFP makeup capability and the emergency preparedness actions
would be more efficiently and effectively addressed as part of the Tier 1 activity for mitigating
strategies for beyond-design-basis events. Therefore, these two Tier 2 activities have been
consolidated into Tier 1.
In SECY-16-0074, “Assessment of Fukushima Tier 2 Recommendation Related to Evaluation of
Natural Hazards Other Than Seismic And Flooding,” dated June 2, 2016, the staff concluded
that other than seismic and flooding, only those natural hazards associated with high winds and
27
snow loads warranted further assessments and stakeholder interactions to address the
recommendation. The staff intends to complete its assessment by the end of 2016.
Tier 3 Recommendations
The third tier consists of actions that required further staff study to support regulatory action,
relied on the result of an associated short-term action to inform the long-term action, depended
on the availability of critical skill sets, or related to potential revisions to the regulatory
framework that balances defense-in-depth and risk considerations (Recommendation 1). The
following recommendations are included in Tier 3:
x
periodic confirmation of seismic and flooding hazards (dependent on
Recommendation 2.1) (Recommendation 2.2)
x
potential enhancements to the capability to prevent or mitigate seismically-induced fires
and floods (long-term evaluation) (Recommendation 3)
x
reliable hardened vents for other containment designs (long-term evaluation)
(Recommendation 5.2)
x
hydrogen control and mitigation inside containment or in other buildings (long-term
evaluation) (Recommendation 6)
x
emergency preparedness enhancements for prolonged SBO and multiunit events
(dependent on availability of critical skill sets) (Recommendation 9.1/9.2)
x
Emergency Response Data System capability (related to long-term evaluation under
Recommendation 10) (Recommendation 9.3)
x
additional emergency preparedness topics for prolonged SBO and multiunit events
(long-term evaluation) (Recommendation 10)
x
emergency preparedness topics for decisionmaking, radiation monitoring, and public
education (long-term evaluation) (Recommendation 11)
x
Reactor Oversight Process modifications to reflect the recommended defense-in-depth
framework (dependent on Recommendation 1) (Recommendation 12.1)
x
staff training on severe accidents and resident inspector training on SAMGs (dependent
on Recommendation 8) (Recommendation 12.2)
x
basis of emergency planning zone size (additional issue)
x
prestaging of potassium iodide beyond 10 miles (additional issue)
x
expedited transfer of spent fuel to dry cask storage (additional issue)
The evaluation of the need to expedite transfer of spent fuel to dry cask storage was addressed
by the staff in COMSECY-13-0030, “Staff Evaluation and Recommendation for Japan
28
Lessons-Learned Tier 3 Issue on Expedited Transfer of Spent Fuel,” dated November 12, 2013.
The Commission agreed with the staff’s assessment that expedited transfer was not warranted.
In addition, a number of Tier 3 recommendations are being addressed by the Tier 1 “Mitigation
of Beyond-Design Basis Events” rulemaking, which can be found in www.regulations.gov
(Docket ID: NRC-2014-0240). In SECY-15-0137, “Proposed Plans for Resolving Open
Fukushima Tier 2 and 3 Recommendations,” dated October 29, 2015, the staff proposed
resolution plans, including necessary resource requirements, for those Tier 2 and 3
recommendations that had not been previously closed or addressed by other, higher-priority
recommendations. Staff Requirements Memorandum (SRM)-SECY-15-0137, dated February 8,
2016, approved the staff’s resolution plans for these Tier 2 and 3 recommendations. The
Commission approved closing a subset of these recommendations and directed the staff to
continue issuing status updates on those recommendations that remain open.
Subsequently, in SECY-16-0041, “Closure of Fukushima Tier 3 Recommendations Related to
Containment Vents, Hydrogen Control, And Enhanced Instrumentation,” dated March 31, 2016,
the staff informed the Commission of the final assessment and closure of Fukushima-related
Tier 3 recommendations regarding evaluations of reliable vents for containment types other than
BWR Mark I and Mark II containments, hydrogen control and mitigation, and reactor and
containment instrumentation enhancements.
Post-Fukushima Safety Enhancements
Using existing regulatory processes (e.g., orders, rulemaking, and requests for information
(RFI)), the Japan Lessons-Learned Division provides project management, technical review,
and oversight of implementation of Fukushima lessons learned.
On March 12, 2012, the NRC issued the first regulatory requirements, in the form of orders, for
the operating reactors based on lessons learned from the accident. These orders require safety
enhancements of operating reactors, construction permit holders, and combined license
holders. Specifically, they require nuclear power plants to implement safety enhancements
related to (1) mitigation strategies to respond to beyond-design-basis external events,
(2) ensuring severe-accident-capable reliable hardened containment vents for boiling-water
reactors (BWRs) with Mark I and II containment designs, and (3) enhancing SFP
instrumentation. All licensees will be in compliance with the SFP instrumentation order, and the
majority of the licensees will be in compliance with the mitigating strategies order, by the end of
2016.
The NRC also issued an RFI on March 12, 2012, requiring each reactor licensee to reevaluate
the seismic and flooding hazards at its site using present-day guidance, methods, and
information; conduct walkdowns of its facilities to ensure protection against the hazards in its
current design-basis; and assess its emergency communications systems and staffing levels.
The walkdowns have been completed. The NRC has received the reevaluated hazard reports
for seismic and flooding for most licensees and has begun to issue assessments of those
reports. As appropriate, licensees are performing additional evaluations to determine how the
reevaluated hazards would affect the plant. They are also assessing whether the mitigating
strategies developed in response to the order described above would remain available under
the conditions of the reevaluated hazards. If not, then under the proposed related rulemaking,
they would be required to enhance those strategies, develop alternate mitigating strategies, or,
possibly, make design basis changes under the NRC’s backfit provisions.
29
Information about the NRC’s activities associated with the Fukushima lessons learned are
specified within the individual Articles of this report and in NUREG-1650, “The United States of
America National Report for the 2012 Convention on Nuclear Safety Extraordinary Meeting,”
Revision 4, issued in July 2012. Specifically, per paragraph 23 of the Final Summary Report for
the 2nd CNS Extraordinary Meeting, the Contracting Parties agreed that the National Reports
should cover:
(a)
The results of reassessments of external events, of periodic safety assessments, and of
any peer reviews, and any followup actions taken or planned, including upgrading
measures.
The NRC is undertaking near-term regulatory activities to reevaluate and
enhance, as necessary, the protection of structures, systems, and components
(SSCs) against seismic and flooding events for all operating reactors in the U.S.
These activities are based on NTTF Recommendations 2.1 and 2.3, as modified
by subsequent NRC senior management and Commission direction. These
activities include requesting that licensees reevaluate the seismic and flooding
hazards at their sites using updated methods and perform “walkdowns” to identify
plant-specific vulnerabilities. Additional details and actions for other external
events are discussed in Section 18.5.1 of this report.
(b)
Actions taken or planned to cope with natural hazards more severe than those
considered in the design basis.
The NRC is evaluating topics related to external events that exceed a facility’s
design basis. Activities associated to these topics include the following:
x
evaluation of external hazards other than seismic and flooding
x
updating of natural hazard information and addressing any new and
significant information
x
development of mitigating strategies for beyond-design-basis external
events
Additional details are discussed in Section 1.3.3 of this report.
(c)
For new nuclear power plants, improved safety features and additional improvements, if
any, to address external hazards and to prevent accidents and, should an accident
occur, to mitigate its effects and avoid offsite contamination.
In response to the Fukushima accident, the NRC issued an order requiring all
licensees to have mitigating strategies that would preserve core cooling, SFP
cooling, and containment. The NRC issued the Mitigation of
Beyond-Design-Basis Events proposed rulemaking that would make this
requirement generally applicable and incorporate lessons learned from
implementation of the order. Furthermore, the NRC used its regulatory processes
to request that licensees reevaluate the seismic and flooding hazards at their
sites using present-day regulatory guidance and methodologies. Information from
these evaluations will be used to determine whether additional regulatory actions
30
are necessary (e.g., updating the design-basis and SSCs important to safety) to
protect against the updated hazards. Additional details are discussed in Section
1.3.3 of this report.
(d)
Upgrading of accident management measures for extreme natural events, including, for
example, measures to ensure core cooling and SFP cooling, the provision of alternate
water sources for the reactor and for the SFP, the availability of the electrical power
supply, measures to ensure containment integrity, and filtration strategies and hydrogen
management for the containment; the development of probabilistic safety assessments
to identify additional accident management measures should be considered as a
possible future activity.
The NTTF recommendations for upgrading accident management measures
were discussed earlier in this section. The development of probabilistic safety
assessments to identify additional accident management measures are
discussed in Section 10.3.5 of this report.
(e)
Measures taken or planned to ensure the effective independence of the regulatory body
from undue influence, including, where appropriate, and information on the hosting of
Integrated Regulatory Review Service (IRRS) missions.
As noted in Section 8.3 of this report, the U.S. Congress created the NRC as an
independent agency in 1974. As a result of the Fukushima nuclear accident,
there have been no changes in the U.S. legislative framework that governs the
NRC and the regulations of the U.S. nuclear industry. Regarding the International
Atomic Energy Agency’s (IAEA) peer review missions, the United States hosted
an IRRS mission in 2010 and the followup mission in 2014. Additional details are
discussed in Section 8.1.5.2 of this report.
(f)
Enhancements of emergency preparedness and response measures, including, for
example, for multiunit sites, approaches and methods of source term estimation and
initiatives in the field of remediation. The enhancements should include defining the
additional responsibilities up to appropriate levels of the national government and the
development of procedures and joint actions of various agencies and improvements in
international cooperation.
The Fukushima accident highlighted the need to ensure that sufficient staff and
resources are available to respond to a multiunit event and a prolonged SBO. As
such, the NRC undertook actions to enhance emergency preparedness with
respect to communications and staffing. Additional information can be found in
Section 16.9 of this report.
(g)
Information on how IAEA safety standards are taken into account.
The NRC actively participates in the development of the IAEA’s safety standards.
Where appropriate, the NRC also references the safety standards in NRC
regulations and regulatory guidance. Additional information can be found in
Section 8.1.5.1 of this report.
31
(h)
Information on activities undertaken to enhance openness and transparency for all
stakeholders.
Openness is the second of six “Principles of Good Regulation” that the NRC first
established in 1977. These principles guide all of the agency’s activities.
Openness is also one of seven organizational values, adopted in 1995, to which
the agency adheres in all its work. After the Fukushima event, the NRC updated
its crisis communication plan with lessons learned and added staff to the Office
of Public Affairs’ technical briefer list to support public and media outreach efforts
in future emergency response events. Additional information can be found in
Sections 8.1.7 and 8.3 of this report.
The NRC staff has also continued to support and participate in NEA’s post-Fukushima activities,
notably through its Committee on Nuclear Reactor Activities and Committee on the Safety of
Nuclear Installations. The NRC has been involved in a number of working groups, including
those associated with defense-in-depth and defining an effective nuclear regulator, along with
research activities. The NRC staff considers the NEA guidance and research results for
harmonization with its actions. The NRC will continue to support the NEA’s long-term
post-Fukushima research activities and will consider the results for harmonization with its
planned actions.
The NRC has long been involved with the international community in collaborative efforts
related to safety analysis and assessment. Since the Fukushima accident, international
cooperation has been strengthened further through safety related projects coordinated by NEA,
International Atomic Energy Commission, and other regional organizations. The Benchmark
Study of the Accident at the Fukushima Dai-ichi Nuclear Power Station (BSAF) project under
the auspices of NEA is an example of such cooperative effort. The U.S. Department of Energy
(DOE), the NRC, EPRI, and others participate in this benchmark phased study. The main
objectives of the project are to benchmark severe accident analysis tools (codes) against the
Fukushima Dai-ichi accident, and to help Japan prepare for decommissioning of Fukushima
Dai-ichi, Units 1, 2, and 3, by using the codes to predict core debris locations to better inform
approaches for access and retrieval of radioactive material. Phase 1 of the project focused on
thermal-hydraulics of the reactor systems and the containment, and an estimation of the
distribution of degraded core materials and their composition for the first 6 days of the accident.
Phase 1 was completed and the final report was issued in March 2015. Phase 2 started in April
2015 with the goal of analyzing fission product transport and distribution on-site and off-site for
the first 21 days of the accident. It is estimated that Phase 2 will continue through mid-2018.
Another Fukushima-related safety project under the auspices of NEA is SAfety REsearch
Opportunities post-Fukushima (SAREF). The project has two objectives:
(1) to address safety
research gaps and advance safety knowledge base, and (2) to support Japan in achieving safe
and prompt decommissioning. A report summarizing the safety research gaps is in preparation.
NEA is also sponsoring a study of fast-running software tools used to model radionuclide
releases during nuclear accidents in the frame of the Fukushima Dai-ichi accident. The objective
of this activity is to benchmark software tools used to estimate consequences of accidents at
nuclear facilities. A report summarizing the results of these studies will be presented for
approval during the NEA’s Committee on Nuclear Regulatory Activities (CNRA) meeting in
June 2016.
32
The NRC’s post-Fukushima activities are also consistent with the IAEA’s report titled “Action
Plan on Nuclear Safety,” which was adopted by the General Conference in September 2011.
The report identified specific initiatives to address lessons learned from the nuclear accident
and to enhance multilateral communication. The examples described below are some actions
taken by the United States in support of the IAEA’s Action Plan.
x
Safety Assessments in Light of the Accident at Tokyo Electric Power Company’s
Fukushima Dai-ichi Nuclear Power Station: The United States immediately undertook a
comprehensive assessment of its operating nuclear power plants as discussed in this
section and Section 1.3.3 of this report.
x
IAEA Peer Reviews: The United States has strongly supported the IAEA’s suite of peer
review services since their inception. The NRC regularly provides technical experts to
participate in IRRS and Operational Safety Review Team (OSART) missions around the
world, often at a senior leadership level. Additional information can be found in
Section 1.4 of this report.
x
Emergency Preparedness and Response: The United States has undertaken
significant activities to assess and strengthen, where appropriate, its emergency
preparedness and response programs following the Fukushima accident. The United
States has also worked closely with Canada and Mexico to enhance North American
cooperation in this area. Additional information can be found in Sections 16.9 and 16.7
of this report.
x
National Regulatory Bodies: The NRC has devoted significant resources to address
the findings and recommendations from the U.S. IRRS mission. A thorough assessment
of the U.S. nuclear safety regulatory infrastructure and current regulations was a key
component in the NRC’s assessment methodology. Additional information can be found
in Section 8.1.5.2 of this report.
x
Operating Organizations: The licensee has primary responsibility for safety in the
United States. The U.S. Government continues to work closely with INPO to ensure
clear communication with each of our licensees and has directed each licensee to
implement recommendations from nuclear safety assessments. The United States also
continues to host OSART missions on a regular basis. Additional information can be
found in this section and in Sections 1.3.3 and 8.1.5.3, Article 9, and Part 3 of this report.
x
IAEA Safety Standards: Through its representation on the IAEA Commission on Safety
Standards and all of the IAEA Safety Standards Committees, the United States is
actively participating in IAEA’s efforts to review the effectiveness of the international
safety standards and to recommend revisions as appropriate. The United States also
takes the IAEA safety standards into account in the development of new or revised
regulations and regulatory guidance. Additional information can be found in
Section 8.1.5.1 of this report.
x
International Legal Framework: The United States conducts international activities
related to statutory mandates, international treaties and conventions, international
organizations, bilateral relations, and research. For example, in this report, the United
States has endeavored to address, in detail, all the areas specified in the revised CNS
33
guidance documents, Rapporteur’s findings, and the Vienna Declaration on Nuclear
Safety. The United States has encouraged other contracting parties to do likewise
through its bilateral and multilateral activities. Additional information can be found in
Sections 1.4.1 and 8.1.5 of this report.
x
Member States Planning to Embark on a Nuclear Power Program: The NRC
coordinates the International Regulatory Development Partnership collaboratively with
Advanced Systems Technology and Management, Inc. The International Regulatory
Development Partnership assists countries with emerging nuclear power programs in
developing organizational and programmatic resources for regulatory oversight.
Additional information can be found in Section 8.1.5 of this report.
x
Capacity Building: The United States continues to work to ensure the availability of
ample resources necessary to ensure a high level of nuclear safety and safe,
responsible, and sustainable use of nuclear technologies. Additional information can be
found in Sections 1.3.2, 8.1.6.2 and 11.2 of this report.
x
Protection of People and the Environment from Ionizing Radiation: The United States,
through its regulatory framework, regulations, and radiation protection programs,
continues to ensure that radiation exposure to the workers caused by a nuclear
installation is kept as low as reasonably achievable, and that the potential for exposures
of individuals to radiation doses that exceed the prescribed national dose limits are
minimized. Additional information can be found in Section 15 of this report.
x
Communication and Information Dissemination: The United States places a high
priority on effective and transparent communication with the public in the event of an
emergency. Additional information can be found in Section 16.8 of this report.
Information about the NRC’s openness and transparency policies and practices can be
found in Section 8.1.7 of this report.
x
Research and Development: The United States continues to play a lead role in
international nuclear safety research. International research is an efficient mechanism
for leveraging limited resources and for promoting collaborative work that encourages
the use of diverse approaches and viewpoints while discouraging duplication. Additional
information can be found in Section 6.3.10 of this report.
The NRC’s post-Fukushima activities are consistent with the IAEA Fukushima Report.
Furthermore, the NRC activities address, as appropriate, the observations and challenges
identified by the Special Rapporteur on Fukushima, which are documented in the 6th CNS
President’s report. Specifically, the Special Rapporteur identified the following challenges:
x
Minimize Gaps Between Contracting Parties’ Safety Improvements: Through its
representation on the IAEA standards committees, the United States is actively
participating in IAEA’s efforts to review the effectiveness of the international safety
standards and to recommend revisions as appropriate. The IAEA’s safety standards are
used as reference documents to inform the development of requirements and guidance
in the NRC’s reactor, radiation protection, and waste management programs. The NRC
also provides senior expert assistance to the IAEA to support studies designed to
34
advance the safety standards program and minimize gaps in safety improvements.
Additional information can be found in Section 8.1.5.1 of this report.
x
Achieve Harmonized Emergency Plans and Response Measures: After the Fukushima
event, the NRC undertook actions to enhance emergency preparedness for licensees
with respect to communications and staffing. For example, the NRC requested that
licensees evaluate their current communications systems and the equipment that would
be used during an emergency event. The NRC also improved its communication
strategy and conducted an extensive analysis of the emergency planning zone.
Additional details can be found in Section 16.9 of this report.
x
Make Better Use of Operating and Regulatory Experience, and International Peer
Review Services: The United States supports IAEA peer review services, such as
IRRS and OSART missions. The NRC also has a robust operating experience program,
which is considered essential in the implementation of the agency’s mission. Additional
information can be found in Sections 1.4 and 6.3.5 of this report.
x
Improve Regulators’ Independence, Safety Culture, Transparency and Openness: The
United States places a high priority on effective and transparent communication with the
public in the event of an emergency. After the Fukushima event, the NRC updated its
crisis communication plan to support public and media outreach efforts in future
emergency response eventbas. The actions that the NRC has taken are broadly
supportive of the priority given to safety and a strong safety culture. For example, the
NRC has ensured that licensees have adequate staffing to respond to emergencies and
clearly defined roles and responsibilities for those responders. The NRC also developed
a Safety Culture Policy Statement, which applies to all of its regulated entities. The NRC
has updated all appropriate guidance and inspection documents to address the Safety
Culture concepts in the Policy Statement. Additional information can be found in this
section and Sections 1.3.3, 8.1.7, 8.3, and 10.4 of this report.
x
Engage All Countries to Commit and Participate in International Cooperation: The
United States has undertaken significant activities to assess and strengthen, where
appropriate, international cooperation. The United States has worked closely with
Canada and Mexico to enhance North American cooperation in this area. The United
States also continues to lead by example in the implementation of conventions and
treaties, such as the CNS. The United States has encouraged other contracting parties
to do likewise through its bilateral and multilateral activities. Additional information can
be found in Sections 1.4.1, 8.1.5, 16.9, and 16.7 of this report.
Nondestructive Evaluations
There have been several recent issues of operating experience where nondestructive
evaluation has yielded results that have caused the NRC to take action. In some cases,
nondestructive evaluation results have identified the occurrence of degradation at rates that are
different than had been anticipated causing the NRC to review inspection requirements and
regulatory practices. In other cases, the failure of nondestructive evaluation to find significant
degradation has led to updated approaches to improve the performance of the evaluation.
Several examples of the implications of these nondestructive evaluation issues are provided
below.
35
Failed Inspections of Diablo Canyon, Unit 2, Weld Overlays
In 2008, Diablo Canyon, Unit 2, installed Alloy 52 full structural weld overlays on six pressurizer
welds. The inspection vendor performed an acceptance examination using conventional manual
ultrasonic search units to detect possible welding defects in the overlays. The acceptance
review found no unacceptable indications.
In 2013, the licensee performed an inservice inspection using a manual phased array ultrasonic
search unit. This examination found several embedded laminar flaws missed by the 2008
acceptance examinations. One of the flaws was 16 inches long and a second was essentially
360 degrees around the overlay. It was determined that the IP and search units used in the
2008 acceptance examinations were able to detect the laminar flaws, but the procedure had
been implemented poorly. The original ultrasonic search units were difficult to hold
perpendicular to the pipe at high scanning speeds and procedures failed to address this
limitation. The phased-array search units were contoured to fit the pipe surface and were easier
to use. Additionally, the phased-array search units provide a more persistent image of the flaws,
allowing the inspector more time to detect and identify the flaws.
After a structural analysis was performed by the licensee and the analysis was reviewed by the
NRC staff, it was determined that the laminar flaws did not challenge the structural integrity of
the weld overlays. No repairs were required.
The licensee has implemented several changes to the weld overlay inspections. The licensee is
disallowing the use of the 2008 IP in future inspections. The licensee will only employ phased
array probes for subsequent examinations of pressurizer weld overlays. The IP used in the 2008
examination was revised by EPRI to provide additional guidance for the maximum scan speed.
Indications in the Belgian Pressure Vessel Forgings
In June 2012, a new ultrasonic inspection was performed to examine the pressure vessel for
possible underclad cracks at the Doel, Unit 3, plant in Belgium. Although no underclad cracks
were detected, the inspection detected nearly laminar indications in the lower and upper shells
of the vessel. As this inspection was focused near the inside diameter of the vessel, a followup
inspection was performed in July 2012, with an array of ultrasonic transducers to inspect the full
volume of the vessel forging rings. This followup inspection found more than 8,000 nearly
laminar indications in the Doel 3 pressure vessel forging rings, with a typical size of roughly
10 millimeters in diameter. In September 2012, the pressure vessel of the Belgian Tihange, Unit
2, which was made using forgings from the same manufacturing facility, also was inspected.
The inspection of the Tihange, Unit 2, forging rings found over 2,000 similar indications in the
lower and upper shells. The indications in Doel, Unit 3, and Tihange, Unit 2, were determined to
be hydrogen flakes introduced during the vessel manufacture.
Fracture mechanics calculations show that the laminar orientation of the flaws makes them
relatively benign to the toughness of the pressure vessel. The Belgian regulator, the Federal
Agency for Nuclear Control, reviewed the licensee fracture evaluation and causal analysis and
determined that Doel, Unit 3, and Tihange, Unit 2, would be permitted to restart. The Federal
Agency for Nuclear Control has placed a number of conditions on the restart of both reactors.
Among these conditions, the licensee was required to perform irradiation studies on some
surrogate materials. In 2014, the results of the irradiation studies became available and
36
indicated the surrogate materials experienced more embrittlement than expected. As a result,
Doel, Unit 3, and Tihange, Unit 2, entered into early refueling outages and remained shut down
due to pending investigation of anomalous embrittlement. The licensee performed additional
irradiation studies and determined the anomalous embrittlement was a function of the surrogate
material, and was not representative of the actual reactor pressure vessel forging material. As a
result, the Federal Agency for Nuclear Control permitted Doel 3 and Tihange 2 to resume
operations.
The NRC staff reviewed fabrication information and determined that several reactor pressure
vessels in the United States contain ring forgings that had been produced in the same
fabrication shop as the rings in the Doel, Unit 3, and Tihange, Unit 2, vessels. The NRC also
determined that other vessels in the United States contained ring forgings that had been
produced in other fabrication shops. The staff notified industry about the possibility of hydrogen
flaking in reactor pressure vessel ring forgings. In response, the industry retrieved the original
fabrication records of ultrasonic examinations of vessel ring forgings and documented the ability
of the construction-era ultrasonic examination techniques to detect indications similar in nature
to hydrogen flakes, and documented the requirement for recording such indications. The
industry also performed bounding structural integrity assessments to ensure that reactor
pressure vessel integrity would be maintained under accident conditions, even in the presence
of hydrogen flaking.
On September 22, 2013, the NRC issued IN 2013-19, “Quasi-Laminar Indications in Reactor
Pressure Vessel Forgings,” to inform industry of the quasi laminar indications observed in the
Belgian reactor pressure vessel forgings. Additionally, the NRC hosted a public meeting with
industry and stakeholders on March 5, 2013, to discuss these indications. The industry
presented plans to investigate the type of ultrasonic examination techniques used during
construction and to perform a probabilistic fracture mechanics evaluation of the structural
integrity effect on U.S. reactors of potentially undiscovered quasi laminar indications.
Subsequently, in October 2013, EPRI published its findings in EPRI-3002000647, “Materials
Reliability Program (MRP): Evaluation of the Reactor Vessel Beltline Shell Forgings of
Operating U.S. PWRs [Pressurized Water Reactors] for Quasi Laminar Indications (MRP 367).”
The objectives of the report were twofold:
(1) to evaluate whether reactor pressure vessel
forgings in U.S. plants were likely to have indications similar to those found in Doel 3 and
Tihange 2, and (2) to evaluate the structural significance of indications if they did exist in a
reactor pressure vessel. The report concluded that the ultrasonic techniques used during
construction of U.S. vessels were capable of detecting quasi laminar indications, and the
reporting requirements would have caused the indications to be recorded if they were present.
The report included a probabilistic fracture mechanics analysis of a set of conditions based on
data from Doel 3 and Tihange 2. The industry concluded that even if quasi laminar indications
were present in a U.S. reactor vessel forging, the incremental increase in the vessel failure
probability under pressurized thermal shock loading is negligible.
The NRC staff’s evaluation consisted of reviewing the analyses performed by the Belgian
licensee as well as the evaluation performed by the industry. Specifically, the NRC staff
reviewed evaluations of nondestructive examination records performed by the U.S. industry to
examine the likelihood of the presence of the quasi laminar indications in U.S. reactor pressure
vessels. Structural evaluations were also performed to determine the risk significance even if
the quasi laminar indications were present. This was followed by a risk-informed evaluation
documented in “Technical Assessment of Potential Quasi-Laminar Indications in Reactor
37
Pressure Vessel Forgings,” dated September 8, 2015. The staff’s current understanding is that
the identified hydrogen flaking is structurally insignificant. Accordingly, there are no current
plans to require additional ultrasonic examinations to look for hydrogen flaking.
In the longer term, the staff and industry have agreed to approach the American Society of
Mechanical Engineers (ASME) to ensure that lessons learned from the discovery of hydrogen
flakes in an operating reactor pressure vessel are appropriately incorporated into applicable
codes and standards.
Identification of Cracking in a Bottom Mounted Instrument Nozzle at a French Plant
During an inspection of a bottom mounted instrument nozzle at a plant in France, ultrasonic
nondestructive evaluation identified cracks in the nozzle material adjacent to the J-groove weld
that attaches the nozzle to the bottom head. The French operational inspection agency, the
Nuclear Safety Authority, met with NRC counterparts and described the findings and the French
regulatory response to require ultrasonic inspection of bottom mounted nozzles at all reactors.
During the information exchange meeting, the NRC staff provided information related to similar
findings of cracking of a bottom mounted instrument nozzle at South Texas Project Nuclear
Plant in 2003. Current requirements to inspect the bottom mounted instrumentation nozzles at
U.S. plants are contained in the ASME Boiler and Pressure Vessel Code (BPV Code). The BPV
Code currently only requires visual examination of the outer surface of the reactor vessel where
the bottom mounted nozzles exit the lower head. Because of the nondestructive evaluation
findings at the French plant, the NRC staff approached ASME to initiate BPV Code changes to
require volumetric inservice inspection of bottom mounted instrument nozzles for all nuclear
plants having material susceptible to primary water stress-corrosion cracking. The ASME
Code committee considered developing a code case, but found only a limited number of
licensees would consider using the code case. Therefore, the committee elected not to develop
a code case. Licensees who want to perform volumetric inspections in lieu of visual inspections
can do so using the NRC’s relief request process.
Steam Generator Integrity
Steam generators in PWRs contain components that form part of the reactor coolant pressure
boundary (e.g., tubing and the channel head). Managing steam generator tube degradation has
been a significant area of focus by industry since the first operating reactors were brought into
service. The industry has moved to different heat treatments and alloys in successive
generations of steam generators, in addition to improved control of secondary water chemistry,
in an effort to decrease the susceptibility of steam generator tubing to various corrosion
mechanisms. Operating conditions and maintenance items can potentially affect the useful
lifetime of a steam generator and may affect the integrity of the steam generator tubing. Several
examples of steam generator integrity issues with implications for the NRC staff are provided
below.
Tube-To-Tube Wear
Wear attributed to tube-to-tube contact has been detected in both once-through and
recirculating steam generators.
38
Replacement Once-Through Steam Generators
Wear indications, attributed to tube-to-tube contact, at Three Mile Island, Unit 1, were first
reported in fall 2011, after one cycle of operation with the replacement steam generators. After
the Three Mile Island findings were shared with other plants with once-through steam
generators, subsequent re-analysis of prior eddy current inspection data by these plants
indicated tube-to-tube wear was present at some of the other units. The re-analyses indicated
that the tube-to-tube wear at these plants is shallow and slow growing. Subsequent to these
findings, the NRC issued IN 2012-07, “Tube-to-Tube Contact Resulting in Wear in
Once-Through Steam Generators,” dated July 17, 2012, to provide licensees with lessons
learned from the discovery of these indications. Licensees were expected to review the
information for applicability and consider actions, as appropriate, to avoid similar problems.
These findings highlight the importance of performing comprehensive inspections of new and
replacement equipment to ensure they perform as expected. The cause of the tube-to-tube
contact was determined to be a result of a combination of factors including: nonconservatism
in the margin to tube buckling in the design, the outside temperature of the steam generator
shell is cooler than the value used in the design, the tube preload is less tensile than the value
used in the design analysis, and the lateral loads or accelerations are sufficient to cause the
observed wear. Since this degradation is readily managed through the licensees’ Steam
Generator Tube Integrity Programs, no additional regulatory action was deemed necessary.
Additional information can be found in the NRC’s Agencywide Documents Access and
Management System (ADAMS) under Accession No. ML13178A358.
Replacement Recirculating Steam Generators
San Onofre Nuclear Generating Station replaced the Unit 2 steam generators in 2010 and the
Unit 3 steam generators in 2011. On January 31, 2012, San Onofre, Unit 3, was operating at
100 percent rated thermal power when a primary-to-secondary leak was detected. Although the
leak rate was initially small, it increased enough in a short period of time that the plant was shut
down. Unit 3 was in its first cycle of operation with replacement steam generators.
At the time of the leak in Unit 3, Unit 2 already was shut down for maintenance and refueling,
having just completed its first cycle of operation with replacement steam generators. Tube wear
was detected at a number of locations in both units. The wear was attributed to the tubes
interacting with tube support plates, antivibration bars, retainer bars, and other tubes. The wear
attributed to the retainer bars and tube-to-tube contact was not expected. All tubes in Unit 2 had
adequate integrity. At Unit 3, there were eight tubes that did not have adequate integrity
because of tube-to-tube wear.
A root cause evaluation report that the plant owner prepared stated that the U-bend portion of
some of the tubes experienced fluid elastic instability in the inplane direction which caused the
tubes to wear against each other. The wear in the tubes near the retainer bars was a result of
the design of the smaller diameter retainer bars, which was insufficient to prevent excessive
flow-induced vibration of the retainer bar.
Ultimately, the plant did not restart. By letter dated June 12, 2013, Southern California Edison
notified the NRC of its decision to permanently cease operations at San Onofre Nuclear
Generating Station, Units 2 and 3. Additional information on the decision to cease operations at
San Onofre is discussed in Section 6.2 of this report.
39
The NRC staff assessed the lessons learned from the San Onofre event. The results of that
effort were documented on March 6, 2015. Additional information can be found under ADAMS
Accession No. ML15062A125.
Low Alloy Steel Channel Head Corrosion Operating Experience
In response to international operating experience on corrosion of the low-alloy steel steam
generator channel head beneath the channel head cladding (in the vicinity of the channel head
drain line), some U.S. plants have performed inspections of their steam generator channel
heads. On October 3, 2013, the NRC staff issued IN 2013-20, “Steam Generator Channel Head
and Tubesheet Degradation,” to address this issue. As of February 2016, no corrosion near the
channel head drain line has been identified in the U.S. steam generators; however, some minor
corrosion of the low alloy steel channel head at a different location has been observed at a
couple U.S. facilities. The U.S. industry’s response has been effective at addressing this issue;
therefore, no further regulatory action has been deemed necessary.
1.3.2 Current Safety and Regulatory Issues
The NRC and its licensees are evaluating and resolving the following potential safety and
regulatory issues:
x
baffle-former bolts
x
digital instrumentation and control systems
x
open phase conditions in electric power system
x
risk-informing regulations and processes
x
SFP neutron-absorbing materials
x
staff readiness to transition plants from construction to operations
x
staff readiness to transition plants from operation to decommissioning
x
subsequent license renewal
x
Project Aim
Baffle-Former Bolts
The core baffle is a portion of the reactor vessel internals in a Westinghouse PWR. The core
baffle is located within the core barrel and functions to direct the coolant flow through the core
and provide some lateral support to the fuel assemblies. Vertical baffle plates are bolted to the
edges of horizontal former plates, which are attached to the inside surface of the core barrel.
There are typically eight levels of former plates located at various elevations within the core
barrel. The bolts that secure the baffle plates to the former plates are referred to as
baffle-former bolts. To cool the baffle structure, some water flowing through the reactor vessel is
directed between the core barrel and the baffle plates in either a downward direction
(i.e., downflow configuration), or an upward direction (i.e., upflow configuration).
Degradation of the baffle-former bolts was first noted in the late 1980s in PWR facilities outside
the United States. The NRC communicated operating experience on baffle-former bolt
degradation to U.S. licensees in IN 98-11, “Cracking of Reactor Vessel Internal Baffle Former
Bolts in Foreign Plants,” dated March 25, 1998. On January 9, 2012, EPRI issued MRP-227-A,
“Power Reactor Internals Inspection and Evaluation Guidelines,” which includes an inspection of
the baffle-former bolts during the timeframe when bolt degradation is most likely to appear, as
40
demonstrated by operating experience. The NRC endorsed MRP-227-A in 2012. The guidelines
of MRP-227-A provide for the development of an aging management program for PWR reactor
vessel internals that meets the NRC requirements for issuance of a renewed operating license.
The degradation of the baffle-former bolts is attributed to irradiation assisted stress corrosion
cracking. Baffle-former bolts are subjected to significant stresses and irradiation over years of
plant operation. PWRs with a downflow configuration place additional stress on the baffle-former
bolts due to the pressure differential across the vertical baffle plates. At this time, significant
degradation of the baffle-former bolts has only been observed in Westinghouse four-loop PWR
reactors with the downflow configuration and bolts made of type 347 stainless steel. Seven
reactors in the U.S. match this description.
In spring 2016, two U.S. nuclear power plants identified a large number of type 347 stainless
steel baffle-former bolts with indications of degradation during the performance of ultrasonic
inspections following MRP-227-A guidelines. In general, both units replaced potentially
degraded bolts with type 316 stainless steel bolts in an improved design. The NRC has
inspected repair activities at these facilities.
The NRC performed a risk-informed evaluation of the safety impact that degradation of
baffle-former bolts could present to operating reactors. The NRC concluded that this issue did
not pose an immediate shutdown of any facilities. Analyses will be performed to determine the
material condition of the baffle-former bolts that were removed from the two units and provide
additional insights on the degradation mechanism. The NRC is considering a future generic
communication to address the issue, which will be informed by the results of these ongoing
analyses. The remaining susceptible reactors have indicated that they will accelerate schedules
for performing the MRP-227-A inspection of the baffle-former bolts. The U.S. nuclear industry
has formed a working group to consider potential changes to the MRP-227-A inspection regime.
The NRC is monitoring the industry response to this issue and will take appropriate regulatory
actions.
Digital Instrumentation and Control Systems
The NRC maintains a robust regulatory program for ensuring the safety and security of nuclear
facilities protected and operated with analog and digital instrumentation and control systems.
Using its current regulatory infrastructure, the NRC staff continues to review and approve
license amendments for specific digital instrumentation and control systems, and evaluate new
reactor applications that fully incorporate highly integrated digital technologies. However, the
efficiency and predictability of the processes can be improved. The staff is developing an action
plan to implement an integrated strategy to modernize the NRC’s digital instrumentation and
control regulatory infrastructure and provide for consistent, predictable, and efficient
implementation of digital technology.
A Steering Committee, comprised of NRC senior managers, is providing oversight in the
development of the integrated action plan. The Steering Committee provides oversight for
updating and implementing the plan in order to ensure a sound strategy to modernize the NRC’s
digital instrumentation and control regulatory infrastructure. The plan is being developed through
engagement of Institute of Electrical and Electronics Engineers (IEEE) standards setting
committees, digital instrumentation and control vendors, licensees, and other external
stakeholders in order to reach a common understanding of the digital instrumentation and
control regulatory challenges, priorities and potential solutions. The plan is intended to
41
encompass all NRC digital instrumentation and control activities and regulatory challenges.
Examples include incorporation of IEEE standards into NRC regulations, updates to NRC’s
policy on common-cause failure, and guidance for applicants and licensees to implement digital
instrumentation and control systems. The plan will ensure that any new or revised requirements:
(1) are performance-based (rather than prescriptive), (2) are technology neutral, (3) apply in the
same manner to operating and new reactors, and (4) do not pose an unnecessary impediment
to advancement in nuclear applications of digital technology. Also, the integrated action plan will
ensure that the modernized regulatory infrastructure will improve the predictability and
consistency of the agency’s regulatory process for licensing and oversight for digital
instrumentation and control systems.
In May 2016, the NRC staff submitted the initial integrated action plan to the Commission for
review and approval.
Open Phase Conditions in Electric Power Systems
Operating experience has identified design vulnerabilities associated with open phase
conditions in offsite power systems at operating nuclear plants. These events involved offsite
power supply circuits that were rendered inoperable by open-circuited phase conditions. This
condition can degrade the performance capabilities of both offsite and onsite power systems.
The operating events indicated that the design of the electric power systems to minimize the
probability of losing electric power from any of the remaining power supplies as a result of, or
coincident with, the loss of power from the transmission network were inadequate because it did
not take into account the possibility of the loss of a single phase between the transmission
network and the onsite power distribution system.
The January 30, 2012, operating event at Byron Station, Unit 2, revealed a significant design
vulnerability, which resulted in the loss of safety functions for electric power systems. At Byron
Station, Unit 2, both offsite and onsite electric power systems were unable to perform their
intended safety functions to provide electric power to the engineered safety feature buses with
sufficient capacity and capability to permit functioning of SSCs important to safety. The staff
determined that a design-basis event concurrent with an undetected open phase condition
would likely have resulted in the plant exceeding criteria specified in 10 CFR 50.46,
“Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power
Reactors,” and the accident analyses assumptions. Based on the Byron Station operating event,
the staff issued IN 2012-03, “Design Vulnerability in Electric Power System,” dated March 1,
2012.
A review of other operating experience identified similar design vulnerabilities associated with
single phase open circuit conditions at South Texas , Unit 2, Beaver Valley Power Station,
Unit 1, and a single event that affected Nine Mile Point, Unit 1, and the neighboring James A.
FitzPatrick power plant.5 These events involved offsite power circuits that were rendered
inoperable due to an open circuit in one phase. In each instance (except in the South Texas,
5
South Texas Project, Unit 2, Licensee Event Report (LER) 2001-001 (ADAMS Accession No. ML011010017);
Beaver Valley Power Station, Unit 1, LER 2007-002-00 (ADAMS Accession No. ML080280592); Nine Mile Point
Nuclear Station, Unit 1, LER 2005-04 (ADAMS Accession No. ML060620519); and James A. FitzPatrick LER
2005-006 (ADAMS Accession No. ML060610079).
42
Unit 2, event), the condition went undetected for several weeks because offsite power was not
aligned to the engineered safety feature buses during normal operation and the surveillance
procedures, which recorded phase-to-phase voltage, did not identify the loss of the single
phase. At South Texas, Unit 2, offsite power was normally aligned to the engineered safety
feature and nonsafety plant buses, and the reactor was manually shut down by the operator
when the three circulating water pumps were tripped by the open phase condition. In addition,
operating experience has identified three similar international events with one or two open
phase conditions at reactors located in Canada, Sweden, and United Kingdom.
In the domestic and international events discussed above, the protective relaying schemes did
not detect the open phase conditions due to inadequate detection schemes. The open phase
event at Byron Station6, Unit 2, resulted in degraded and unbalanced voltage conditions on
redundant engineered safety features buses, which led to the tripping of equipment required for
normal plant operations and safe shutdown. The inability of the protection scheme to detect an
open phase condition and automatically transfer power from the affected electric power system
allowed the degraded offsite power system to remain connected to engineered safety features
buses, and prevented other onsite AC sources (e.g., emergency diesel generators) from starting
and powering these buses. As a result, certain equipment required for safe operations remained
powered by the degraded AC source and were put in jeopardy to either rely on internal safety
features to lockout and protect the vulnerable components or risk damage from overheating.
Furthermore, equipment relied on for safe shutdown was also at risk of being unavailable for a
period of time outside the plant’s accident analysis, even after restoration of an operable power
source.
On July 27, 2012, the staff issued Bulletin 2012-01, “Design Vulnerability in Electric Power
System,” to confirm that licensees comply with 10 CFR 50.55a(h)(2), 10 CFR 50.55a(h)(3),
General Design Criterion 17, “Electric Power Systems,” or applicable principal design criteria
specified in the updated final safety analysis report. The NRC staff has reviewed the information
that the licensees provided and the details of this review are documented in Bulletin 2012-01,
“Design Vulnerability in Electric Power System: Summary Report,” dated February 26, 2013.
The staff concluded that this design vulnerability exists at all operating plants, except for
Seabrook Nuclear Power Plant because of plant-specific switchyard features.
The staff developed Branch Technical Position 8-9, “Open Phase Conditions in Electric Power
System,” in July 2015, to provide guidance to staff on reviewing any licensing actions to address
this design vulnerability. The licensees have implemented interim corrective actions and
compensatory measures to address the operability of the electric power system until plant
modifications are completed by December 31, 2018. Plant modifications can include installing a
system or component that will detect and automatically isolate a single-phase open circuit
condition and/or a high impedance ground fault condition on the electrical systems at applicable
nuclear power plant sites.
6 Byron Station, Units 1 and 2, LER 12-001-00 (ADAMS Accession No. ML12090A492); and Byron Station, Units 1
and 2, LER 12-001-01 (ADAMS Accession No. ML12272A358).
43
Risk-Informing Regulations and Processes
The NRC is advancing the use of risk-information in regulatory decisionmaking and processes,
while continuing to emphasize defense-in-depth and safety margins. Several related initiatives,
such as the NRC’s Risk Informed Steering Committee, and risk-informed improvements to
standard technical specifications serve as key examples that are shaping the direction that the
agency will take in regard to risk-informed decisionmaking.
NRC’s Risk Informed Steering Committee
The NRC's Risk Informed Steering Committee is an NRC senior management committee that
provides strategic direction to the NRC staff to advance the use of risk-informed decisionmaking
in licensing, oversight, rulemaking, and other regulatory areas, consistent with the
Commission’s “Policy Statement on Use of PRA Methods in Nuclear Activities,” dated
August 16, 1995. The Committee is chaired by the Director of the Office of Nuclear Reactor
Regulation, with membership consisting of Deputy Office Directors from the Offices of New
Reactors, Nuclear Regulatory Research, Nuclear Materials Safety and Safeguards, Nuclear
Security and Incident Response, and Nuclear Reactor Regulation, as well as the Region I
Administrator.
The nuclear industry has its own Risk Informed Steering Committee, which is a counterpart to
the NRC’s Committee, with its membership comprising licensee chief nuclear officers and other
senior level executives, as well as representation from the Nuclear Energy Institute. The NRC
Risk Informed Steering Committee has held several public meetings with the industry’s
committee. The NRC and industry each agreed to form two working groups that focused on
guidance in two selected areas related to probabilistic risk assessment (PRA) technical
adequacy and dealing with uncertainties in risk-informed decisionmaking. These working groups
formed problem statements and action plans that have been provided to the applicable line
organizations for consideration and continued work.
The NRC Risk Informed Steering Committee is also providing direction to the NRC staff
concerning efforts to provide credit for mitigating strategies put in place in response to the
Commission Orders after the events at the Fukushima Dai-ichi nuclear power plant. The NRC’s
Risk Informed Steering Committee will continue to hold public meetings with the industry to
discuss current and upcoming risk-informed initiatives of interest.
Risk-Informed Improvements to Standard Technical Specifications
In 1992, the NRC issued the improved Standard Technical Specifications to clarify the content
and form of requirements necessary to ensure safe operation of nuclear power plants in
accordance with 10 CFR 50.36, “Technical Specifications.” As the Standard Technical
Specifications mature, areas for improvement have been identified. One process used to
efficiently initiate changes to the Standard Technical Specifications involves the
industry-sponsored Technical Specifications Task Force submitting a proposed change,
commonly known as a “Traveler,” to the NRC for review, approval, and subsequent
incorporation into the next revision of the Standard Technical Specifications.
The NRC reviews the proposed change, with the end product being a model application, a
model safety evaluation, and a review plan that licensees may use in subsequent license
amendment requests. Licensees applying to incorporate these proposed changes into their
44
Technical Specifications must provide a plant-specific justification acceptable to the NRC staff in
their amendment request. The NRC staff is currently reviewing several risk-informed license
amendment requests in accordance with Traveler TSTF-425, Revision 3, “Relocate Surveillance
Frequencies to Licensee Control - RITSTF Initiative 5b,” and Traveler TSTF-505, Revision 1,
“Provide Risk-Informed Extended Completion Times (CTs) - RITSTF Initiative 4b.”
TSTF-425/RITSTF Initiative 5b provides a risk-informed methodology to identify, assess,
implement, and monitor proposed changes to frequencies of technical specification surveillance
requirements. TSTF-505/RITSTF Initiative 4b allows licensees to modify selected required
actions to permit extending completion times, provided risk is assessed and managed within an
acceptable configuration risk management program. These initiatives are intended to maintain
and improve safety through incorporation of risk assessment and management techniques in
the Technical Specifications, while reducing unnecessary burden. The staff continues to work
on the risk-informed technical specifications initiatives to add a risk-informed component to the
Standard Technical Specifications.
Spent Fuel Pool Neutron-Absorbing Materials
The NRC requires that power reactor license holders maintain SFP subcriticality in accordance
with 10 CFR 50.68, “Criticality Accident Requirements,” General Design Criterion 62,
“Prevention of Criticality in Fuel Storage and Handling,” and other equivalent regulatory criteria.
The NRC has a similar requirement included in the technical specifications for nonpower
reactors.
Neutron-absorbing materials have been used in SFPs for more than 30 years to allow for
increases in SFP storage capacity while maintaining safety margins against inadvertent
criticality. Plates or sheets of these materials are used in SFP racks and are comprised of a
compound, alloy, or a composite material that serves as a matrix to contain a neutron absorber
nuclide, primarily boron-10. Several types have been deployed and include the following:
x
boron carbide (B4C) in a silicone polymer
x
B4C in a phenol formaldehyde resin matrix
x
B4C in an aluminum matrix with aluminum cladding
x
natural boron in a stainless steel matrix
x
B4C in an aluminum metal matrix composite
The NRC is performing confirmatory research on SFP neutron-absorbing materials used in the
U.S. commercial nuclear power industry, the surveillance methodologies used by the industry,
and the surveillance intervals of the neutron-absorbing materials used in the SFPs.
Furthermore, the NRC has issued three technical letter reports7 discussing some of the
methods that license holders use to monitor the degradation of neutron-absorbing materials, the
uncertainties in the methodologies employed to monitor the performance, and the degradation
mechanisms.
7
“Boraflex, RACKLIFE and BADGER: Description and Uncertainties,” dated September 30, 2012, “Initial
Assessment of Uncertainties Associated with BADGER Methodology,” dated September 30, 2012, and
“Monitoring Degradation of Phenolic Resin-Based Neutron Absorbers in Spent Nuclear Fuel Pools,” dated
June 5, 2013.
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